ML19270G517
| ML19270G517 | |
| Person / Time | |
|---|---|
| Site: | 05000502 |
| Issue date: | 05/31/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19270G518 | List: |
| References | |
| NUDOCS 7906120147 | |
| Download: ML19270G517 (3) | |
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WNSCOnSin Elecinc row couem 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 May 31, 1979 Mr. Harold R. Denten Director Office of fluclear Reactor Regulation U. S. IlUCLEAR REGULATORY C0ff11SSION Washington, D. C. 20355
Dear Mr. Denton:
DOCKET (10. STil 50-502 AMErlDituiT 18 TO THE PRELIf1IfiARY SAFETY AttALYSIS REPORT WISC014SI!! UTILITIES PROJECT Wisconsin Electric Power Company, Wisconsin Power and Light Company and Wisconsin Public Service Corporation submit herewith three signed originals w and sixty copies of Amendment 18 to the Preliminary Safety Analysis Report filed in the above docket.
Very truly yours, n
Sol Burstein Execu'.ive Vice President Enclosures 2344 052 l
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BEFORE TiiE UNITED STATES NUCLEAR REGULATORY C0lti!SSION In the Hatter of HISCONSIll ELECTRIC POWER C0t!PANY Docket No.
WISCONSIN POWER AND LIGHT COMPANY, and STN 50-502 WISCONSIN PUBLIC SERVICE CORPORATION AMEND
- TENT NO. 13 PRELIMINARY SAFETY AMLYSIS REPCRT Wisconsin Electric Power Company, on its own behalf and on behalf of Wisconsin Power and Light Company and Wisconsin Public Service Corporation (all hereinafter collectively referred to as " Applicants"), hereby amends the Preliminary Safety Analysis Report (PSAR), filed as part of the Application for Licenses in this docket. This amendment consists of Applicants' responses to the Nuclear Regulatory Commission's requests for additional information forwarded by Mr. Olan D. Parr's letters dated February 6 cnd 16,1979 and i March 20,1979. These responses were previously submitted to the NRC for early review in a question-answer formt with Applicants' letter s dated April 2,1979 and May 10,' 1979. ,.
7 The changes mentioned herein are contained on the replacement pages enclosed herewith and made part hereof, which pages are to be inserted in the PSAR in accordance with the accompanying instructions.
Dated May 31, 1979.
Respectfully submitted.
WISCONSIN ELECTRIC POWER C0f tPANY J
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STATE OF WISCO:4SI!1 HILWAUKEE COUiiTY 58' SQL BURSTElil, being first duly sworn, on oath says that he has read the foregoing statenent and knows the contents thoroof, that the sano is true to the best of his knowledge and belief, and that this verification is made by affiant for the reason that Wisconsin Electric Power Company is a corporation, and he is an officer of such corporation, to-wit, Executive Vice President of such corporation, and is duly authorized to rake this verification for and on its behalf.
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Subscribed and sworn to before me this 31st day of May,1979.
Lu mn 9P h uA,--m s _
itatary Public, State of Wisconsin 2344 54 ;
My Cornission expires N# 6, / 97C .
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WUP Amendment 18 PSAR 6/79 INSERTION INSTRUCTIONS Correction pages to the Preliminary Safety Analysis Report (Volumes 1-9) are identified by " Amendment 18, 6/798 and are printed on green stock.
Vertical bars with the numeral 18 have been placed in the outside margins of revised text pages and tables to show the location of any technical changes originating with this amencknent. Bars showing changes originating with previous amendments have also been kept. No change bars are used on figures or new questions and responses.
Entries berein beginning with T or F designate tables or figures, respectively.
All other entries are page numbers.
Remove Old Insert New _ Location VOLUME 1 WE letter to Before WE letter Harold R. Denton to Harold R.
for Amendment 16/ blank Denton for Amend-ment 17 USNRC Attachsent for Amendment 18/ blank Sol Burstein Affidavit / blank MEP-1/MEP-2 -
After Vol. 1 Title Page MEP-1/MEP-2 Before Vol. 1 Title Page EP.1-1/ blank EP.1-1/ blank After Chap. 1 T1.3-1 (9 of 13) T1.3-1 (9 of 13) Tab EP.3-1/EP.3-2 EP.3-1/EP.3-2 After Chap. 3 3-xv/3-xv1 3-xv/3-xvi Tab 3.1-49/3.1-50 3.1-49/3.1-50 3.11-2a thru 3.11-7/ blank 3.11-3 thr,2 3.11-8 F3.6-1 F3.6-1 VOLUME :
EP.5-1/ blank EP.5-1/ blank After Chap. 5 5.2-33/5.2-34 5.2-33 thru 5.2-34a/ blank Tab 5.2-51/5.2-52 5.2-51/5.2-52 5.4-5/5.4-6 5.4-5/5.4-6 VOLUME 3 EP.6-1/EP6-2 EP.6-1/EP.6-2 After Chap. 6 6-vii/6--iiii 6-vii/6-viii Tab 6.5-4a through 6.5-11/ blank 6.5-5 thru 6.5-12 F6.5.2-1 F6.5.2-1 VOLUME 4 EP.8-1/ blank EP.8-1/ blank After Chap. 8 8-1/8-11 8-1/8-11 Tab 8.2-1/8.2-2 8.2-1/8.2-2 8.3-2a thru 8.3-4a/ blank 8.3-3 thru 8.3-4a/ blank 8.3-21 thru 8.3-24b ".3-21 thru 8.3-24volank 2344 055 I-1 b(1O bl
WUP Amendment 18 PSAR 6/79 Remove Old Insert New Location VOLUME 4 (CONT)
EP.9-1/EP.9-2 EP.9-1/EP.9-2 Atter Chap. 9 9.4-12a thru 9.4-14 9.4-13 thru 9.4-14a/ blank Tab F9.4.5-1 F9.4.5-1 VOLINE 6 EP.13-1/ blank EP.13-1/ blank After Chap. 13 13-vii/ blank 13-vii/ blank Tab 13.1-15 thru 13.1-18 13.1-15 thru 13.1-18 F13.1-3 F13.1-3 EP.15-1/EP.15-2 EP.15-1/EP.15-2 After Chap. 15 15-vil thru 15-vilia/olank 15-vii thru 15-villa / blank Tab T15.4.1-la T 15.4 .1-1A T15.4.1-2a T15.4.1-2A T15.4.1-3a T15.4.1-3A T15.4.1-4 a T15. 4 .1-4 A T15.4.1-Sa T15.4.1-5A VOLUME 7 EP.16-1/ blank EP.16-1/ blank After Chap. 16 16.4-19/16.4-20 16.4-19 thru 16.4-20a/ blank Tab EP.17-1/ blank EP.17-1/ blank Af ter Chap. 17 17.1-33/ blank 17.1-33/ blank Tab EP.A-1/ blank EP.A-1/ blank After App. A Tab A-vii-A-viii A-vil thru A-ix/blan%
A.1-11/A.1-12 A.1-11/A.1.12 A.1-29/A.1-30 A.1-29/A.1-30 A.1-33/A.1-34 A.1-33 thru A.1-37/ blank EP.B-1/ blank EP.B-1/ blank After App. B Tab B.3-1/B.3-2 B.3-1/D.3-2 B.3-13 thru B.3-18 B.3-13 thru B.3 *8 B.3-47/B.3-48 B.3-47/B.3-48 VOLUME 8 EP.Q-I thru EP.Q-3/ blank EP.Q-1 thru EP.Q-3/ blank e.2ter Acceptance Review Questions and Responses Tab AEC-lii thru AEC-x AEC-lii thru AEC-xi/ blank After AEC Questions and Responses Tab 0042.4 8-1/Q042. 48-2 After Tab 040, Q042.49-1/Q042.49-2 Q042.47-2 Q110.18-1/ blank thru After Tab 110, Q110.21-1/ blank Q110.17-1/ blank VOLUME 9 Q121.10-1/ blank thru After Tab 120, Q121.12-1/ blank Q121.9-1/ blank Q130.25-1/Q130.25-2 After Tab 130, Q130.26-1 thru Q130.26-6 Q130.24-1/ blank Q130.27-1/ blank thru Q130.3C-1/ blank 0130.34-1/ blank
,_, 2344 056
WUP Amendment 18 PSAR 6/79 Remove Old Insert New Location VOLUME 9 (CONT)
Q214.37-1/ blank After Tab 210, FQ214.37-1 and -2 Q214.36-5/ blank Q214.38-1/ blank thru Q214.43-1/ blank Q221.50-1/Q221.50-2 After Tab 220, Q221.49-4 Q222.6-1/ blank After Tab 220, Q222.7-1/Q222.7-2 Q222.3-1/ blank Q222.8-1/Q222.8-2 Q240.2-1/Q240.2-2 After Tab 240, TQ240.2-1 thru TQ240.2-3/ blank Q240.1-1/ blank FQ240.2-1 and -2 Q241.3-1/ blank thru After Tab 240, Q241.5-1/ blank Q241.2-1/ blank Q312.9-1/ blank thru Atter Tab 312, Q312.11-1/ blank Q312.2-1/ blank Q331.28-1/ blank thru After Tab 330, Q331.29-2 Q331.27-1/ blank Q411.17-1/ blank After Tab 410, Q411.18-1/ blank Q411.16-1/ blank Q412.5-1/ blank After Tab 410, Q412.4-1/ blank Q430.4-1/ blank After Tab 430, Q430.3-1/ blank Q432.10-1/ blank After Tab 430, Q432.1-1/ blank 2344 057 I-3
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT MASTER LIST OF EFFECTIVE PACJS The Lists of Effective Pages for the Preliminary Safety Analysis Report are compiled for each chapter and appendix in Volumes 1 through 7 and the NRC Questions and Responses in Volumes 8 and 9.
The Master List of Effective Pages presents the dates of issue for each amendment, and the revision number of the General Table of Contents (found at the front of each volume) and each List of Effective Pages (found immediately af ter the respective tab).
Issue Date Issue Date Issue Date Original 5/23/74 Amendment 8 6/16/75 Amendment 17 2/79 Amendment 0 8/9/74 Amendment 9 9/8/75 Amendment 18 6/79 Amendment 1 9/27/74 Amendment 10 10/6/75 Amendment 2 1/3/75 Amendment 11 11/24/75 Amendment 3 1/20/75 Amendment 12 2/16/76 Amendment 4 2/10/75 Amendment 13 7/15/76 Amendment 5 3/ 17/75 Amendment 14 5/26/78 Amendment 6 5/2/75 Amendment 15 9/22/78 Amendment 7 6/6/75 Amendment 16 11/78 General Table of Contents Chapter 8 i thru v -
EP.8-1 18 vi 14 Chapter 9 Lists of Effective Pages EP.9-1 and EP.9-2 18 Chapter 1 Chapter 10 EP.1-1 18 EP.10-1 17 Chapter 2 Chapter 11 EP.2-1 16 EP.11-1 17 Chapter 3 Chapter 12 EP.3-1 and EP.3-2 18 EP.12-1 16 Chapter 4 Chapter 13 EP . 4 -1 16 EP.13-1 18 Chapter 5 Chapter 14 EP.5-1 18 EP.14-1 17 Chapter 6 Chapter 15 EP.6-1 and EP.6-2 18 EP.15-1 18 Chapter 7 EP.15-2 17 EP.7-1 16 Chapter 16 EP.16-1 18 2344 058 MEP-1
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT MASTER LIST OF EFFECTIVE PAGES (CONT'D)
Revision Page Ntumber Chapter 17 EP.17-1 18 Appendir A EP.A-1 18 Appendix B EP.B-1 18 Ouestions and kesponses EP.Q-1 thru EP.Q-3 18 2344 059 O
O MEP-2
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 1 Page, Table (T) , or Revision Figure (F) Number 1-i and 1-ii 7 1-iii and 1-iv 10 1-v -
1-v1 10 1.1-1 through 1.1-2 -
1.2-1 through 1.2-12 -
1.3-1 -
1.4-1 through 1.4-5 -
1.5-1 through 1.5-27 10 1.6-1 . -
T.1.1-1 (4 pages) 9 T.1.3-1 (page 1) 15 (page 2) 12 (page 3) 11 (Page 4) la (pages 5) 15 (Page 6) -
(page 7) 15 (Page 8) -
(page 9) 18 (page 10) -
(page 11) 15 (pages 12 and 13) -
T.1.6-1 (12 pages) -
F.1.1-1 -
F.1.2-1 15 F.1.2-2 -
F.1.2-3A through F.1.2-3I -
F .1. 2 -3J 17 F.1.2-3K -
F.1.2-4 and F.1.2-5 -
F .1. 5 - 1 -
F.1.5-2 and F.1.5-3 10 EP.1-1
WUP Amen &nent 18 PSAR 6/79 TABLE 1.3-1 (COffr*D)
DESIGN COMPARISON Chapter Otapter Title Significant Significant Number System / Component References Similarities Differences 9.0 containment Air Section 9.4.7.1 Beaver Valley Unit 1 tbne (Cont'd) Filtration North Anna Units 1 & 2 Containment Purge Air Section 9.4.7.2 Beaver Valley Unit 1 North Anna Units 1 & 2 and Beaver System North Anna mits 1 & 2 Valley Unit 1 do not incorporate debris screens into the purge supply and exhaust lines or the 18 capability to purge during Mode 3.
Containment Air Section 9.4.7.3 Beaver Valley Unit 1 terth Anna Units 1 & 2 and Recirculation System North Anna Units 1 & 2 Deaver Valley Unit 1 use 3-50 percent capacity f an coolers; fan coolers are NNS. North Anna Units 1 & 2 use service water and Beaver Valley Unit 1 uses river water as back-up cooling medium.
CRDM Cooling Section 9.4.7.4 Beaver Valley Unit 1 lbrth Anna Units 1 & 2 and North Anna Units 1 & 2 Beaver Valley Unit 1 use cooling coils and <xxnponent cooling water.
tbrth Anna Units 1 & 2 have 6 fans (3 standbys) and Beaver Valley uses 3 fans.
Diesel Generator Building Section 9.4.8 Beaver Valley Unit 1 tbrth Anna Units 1 & 2 and Heating and Ventilation North Anna Units 1 & 2 Beaver Valley Unit 1 do not have recirculation capability.
Service Water Pumphouse Section 9.4.9 Beaver Valley Unit 1 lbrth Anna Unita 1 & 2 and Ventilation North Anna Units 1 & 2 Deaver Valley Unit I have fan on N exhaust duct.
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wup Amendment 18 PSAR 6/79 PRELIMIIERY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 3 Page, Table (T) , or Revision Page, Table (T) , or Revision Piqure Ws Number Fiqure (F) Ramber 3-1 through 3-xiv 8 3.7-9 through 3.7-12a 1 3-xv and 3-xvi 18 3.7-12b 6 3-xvii and 3-xviii 8 3.7-13 -
3-xix 16 3.7-14 2 3 xx 13 3.7-15 -
3-xxi and 3-xxii 8 3.7-16 through 3.7-18b 4 3 xxiii 15 3.7-19 through 3.7-20c 13 3.7-21 through 3.7-25 -
3.1- 1 -
3.7-26 13 3.1-2 and 3.1-3 7 3.7-27 through 3.7-30 -
3.1-4 through 3.1-8 -
3.7-31 14 3.1-9 14 3.7-32 -
3.1-10 through 3.1-15 -
3.7-33 and 3.7-34 6 3.1-16 2 3.7-34a 2 3.1-16a 3 3.7-35 through 3.7-37 -
3.1-17 through 3.1-41 -
3.1-42 17 3.8-1 through 3.8-2a 15 3.1-43 through 3.1-48 -
3.8-3 through 3.8-5 -
3.1-49 18 3.8-6 and 3.8-6a 1 3.1-50 through 3.1-52 -
3.8-7 -
3.1-53 6 3.8-8 4 3.1-54 and 3.1-55 17 3.8-9 10 3.1-56 2 3.8-10 11 3.1-57 through 3.1-61 -
3.8-11 through 3.8-16 10 3.8-17 through 3.8-19 -
3.2-1 through 3.2-2a 2 3.8-20 16 3.2-3 and 3.2-4 -
3.8-21 and 3.8-22 -
3.8-23 4 3.3-1 -
3.8-24 -
3.3-2 and 3.3-2a 1 3.8-25 through 3.8-26a 4 3.3-3 and 3.3-4 2 3.8-27 and 3.8-28 6 3.3-4a 13 3.8-29 -
3.3-5 2 3.8-30 16 3.8-31 11 3.4-1 2 3.8-32 through 3.8-35 10 3.4-2 9 3.8-36 and 3.8-37 11 3.8-38 through 3.8-40 10 3.5-1 and 3.5-2 6 3.8-41 -
3.5-2a 7 .8-42 13 3.5-3 and 3.5-4 -
3.8-03 -
3.5-5 through 3.5-7 6 3.8-44 1 3.8-44a and 3.8-45 10 3.6-1 7 3.8-46 through 3.8-49 6 3.6-2 and 3.6-3 6 3.8-50 10 3.6-4 and 3.6-4a 7 3.8-50a 6 3.6-5 through 3.6-6f 6 3.8 51 -
3.6-7 through 3.6-8b 2 3.8-52 10
- 3. 6-9 1 3.8-53 8 3.6-10 2 3.8-54 7 3.8-54a 15 3.7-1 13 3.8-55 9 3.7-2 15 3.8-56 8 3.7-2a 3 3.8-56a 10 3.7-3 through 3.7-8 -
3.8-57 -
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONT'D)
Chapter 3 Page, Table (T) , or Revision Page, Table (T) , or Revision Piqure (F) Number Fiqure (F) Number 3.8-58 through 3.8-60 15 F.3.7.1-4 through F.3.7.1-14 13 3.8-61 -
F.3.7.2-1 -
3.8-62 and 3.8-63 15 F.3.7.3-1 through F.3.7.3-5 -
3.8-64 4 F.3.8.1-1 through F.3.8.1-4 17 F.3.8.1-5 8 3.9-1 13 F.3.8.1-6 through F.3.8.1-8 17 3.9-2 through 3.9-4a 4 F.3.8.1-9 4 3.9-5 through 3.9-16 - F.3.8.1-10 through F.3.8.1-13 -
3.9-17 1 F.3.8.1-14 5 3.9 ,8 through 3.9-25 -
F.3.8.1-15 4 F.3.8.1-16 through F.3.8.1-21 -
3.10-1 14 F.3.8.1-22 17 3.10-2 2 F.3.8.1-23 -
3.10-3 14 F.3.8.1-24 4 3.10-4 2 F.3.8.1-25 -
F.3.8.1-25a and F.3.8.1-25b 17 3.11-1 and 3.11-2 16 F.3.8.1-26 -
3.11-3 through 3.11-9 18 F.3.8.4-1 15 F.3.8.4-2 through F.3.8.4-7 15 T.3.1.2-1 -
F.3.8.4-8 (3 pages) 15 T.3.2.5-1 (pages 1 and 2) 8 F.3.8.4-9 15 (page 3) 10 F.3.8.4-9A 15 (page 4) 9 F.3.8.4-10 15 (pages 5 and 6) 2 F.3.8.4-10A 15 (page 7) 4 F.3.8.4-11 and F.3.8.4-12 15 (page 8) 15 F. 3. 8. 4 -13 (deleted) 15 (pages 9 and 10) 4 F.3.8.4-14 and F.3.8.4-15 8 (pages 11 through 13) 6 F.3.8.4-16 (deleted) 8 (pages 14 through 17) 4 F.3.8.4-17 2 (page 18) 2 F.3.8.5-1 -
T.3.3.1-1 through T.3.3.1-7 -
T.3.5.1-1 (2 pages) 7 T.3.5.2-1 7 T.3.5.3-1 through T.3.5.3-6 -
T.3.5.4-1 6 974 pf7 T.3.6-1 4 L J S Ij t UVJ T.3.6-2 (page 1) 4 (page 2) 1 T.3.6-3 (2 pages) 4 T.3.7-1 through T.3.7-6 -
T.3.8.1-1 and T.3.8.1-2 -
T.3.8.1-3 (page 1) 2 (pages 2 through 4) -
T.3.8.1-4 and T.3.8.1-5 -
T.3.8.1-6 2 T.3.8.1-7 -
T.3.9-1 (2 pages) -
T.3.9-2 (2 pages) 6 T.3.9-3 through T.3.9-6 -
T.3.9-7 (2 pages) -
T.3.11 I and T.3.11-2 16 F.3.6-1 18 F.3.6-2 2 F.3.7.1-1 through P.3.7.1-2 13 F.3.7.1-3 -
WUP Amendment 18 PSAR 6/79 CHAPTER 3 TABLE OF CONTLNTS (CONT 'D)
Section Title Page 3.5.2.5 Design and Installation Criteria, 3.9-13 Pressure-Relief Devices 3.9.2.6 Stress Levels for Category I Components 3.9-15 3.9.2.7 Field Run Piping System 3.9-15 3.9.3 Components Not Covered by ASME Code 3.9-15 3.9.3.1 Stone & Webster Scope of Supply 3.9-15 3.9.3.2 Westinghouse Scope of Supply 3.9-16 3.9.3.2.1 Faulted Conditions 3.9-17 3.9.3.2.2 Reactor Internals Response under 3.9-17 Blowdown and Seismic Excitation 3.9.3.2.3 Acceptance Criteria 3.9-18 3.9.3.2.4 Methods of Analysis 3.9-20 3.9.3.2.5 Blowdown Forces Due to Cold and 3.9-20 Hot Leg Break 3.9.3.2.6 Methods of Blowdown Analysis 3.9-24 (Mechanical) 3.9.3.2.7 Control Rod Drive Mechanisms 3.9-24 3.10 SEISMIC DESIGN OF SEISMIC CATEGORY I 3.10-1 INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10.1 Seismic Design Criteria 3.10 -1 3.10.2 Seismic Analyses, Testing Procedures, 3.10-1 and Restraint Measures 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.11.1 Equipment Identification 3.11-2 3.11.2 Qualification Tests and Analyses 3.11-3 3.11.3 Qualification Test Results (FSAR) 3.11-7 3-xv 2344 064
WUP Amendment 18 PSAR 6/79 CIIAPTER 3 TABLE OF CONTENTS (CONT'D)
Section Title Page 3.11.4 Loss of Ventilation 3.11-7 2344 065 0
O 3-xvi
WUP Amendment 18 PSAR 6/79 residual, steady-state, and transient stresses, and (3) size of flaws.
Discussion All ferritic materials in the reactor containment liner are toughness tested in accordance with the provisions of ASME Boiler and Pressure Vessel Code, Secton III, Division 2, CC-2520 "Special Materials Testing" with the following exception: l18
- 1) Notch toughness testing is required for materials with a nominal thickness of 5/8 in. and greater, rather than greater than 5/8 in. as stated in the code.
A minimum service temperature of 500F is specified -for f erritic containment liner plate and penetration materials. The normal 18 operating temperature of the containment is 900F.
The above requirements ensure that ferritic containmant liner plate and penetration materials do not exhibit brittle fracture characteristics under conditions of minimum service temperature, maximum defect size, and yield stress levels.
Ferritic materials thinner than 5/8 in. are impact tested by full l18 or subsized Charpy V - notch impact tests. The acceptance criteria for subsized specimens shall be as shown in Table 16 of SA-20, ASME Boiler and Pressure Vessel Code,Section II - Part A.
Weld procedure qualification demonstrates that the toughness of the weld metal and toughness of the heat affected zones meet the code requirements.
Reference Title Section Materials, Quality Control, and 3.8.2.7 Special Construction Techniques (Steel Containment System) 3.1.2.52 Capability for Containment Leakage Rate Testing (Criterion 52)
Criterion The reactor containment and other equipment which may be sub-jected to containment test conditions shall be designed so that periodic. integrated leakage rate testing can be conducted at containment design pressure.
2344 066 3.1-49
WUP PSAR Discussion The containment structure and related equipment, which are subjected to the containment structural test conditions as described in Section 6.2.1.4, are designed so that the periodic integrated leakage rate testing can be conducted at che containment design pressure. Containment leakage rate testing is described in Section 9.5.9.
References Title Section Testing and Inspection (Containment 6.2.1.4 Functional Design)
Containment Leakage Monitoring System 9.5.9 3.1.2.53 Provisions for Containment Testing and Inspection (Criterion 53)
Criterion The reactor containment shall be designed to permit (1) appro-priate periodic inspection of all huportant areas, such as penetrations, (2) an appropriate surveillance prog::am, and (3) periodic testing at containment design pressui of the leaktightness of penetrations which have resilient seals and expansion bellows.
Discussion The reactor containment is designed to permit testing all penetrations and liner seam welds for leaktightness.
The design of the containment includes the placement of leak chase channel systems over at least those liner seam welds inaccessible after construction and over penetration-to-liner welds. These channels are capable of being conveniently and periodically pressurized to containment structure design pressure so as to permit te.c.ing the leaktightness of the welds covered by the channels.
Electrical penetratian assemblies which employ resilient seals or expansion bellows may be periodic: ally tested as a unit for leaktightness following installation in the containment.
2344 067 0
3.1-50
WUP Amendment 18 PSAR 6/79
- 6. Containment heat removal f an cooler valves - up to 5 minutes after the DBA,
- 7. Power, control, instrumentation cables, and electrical penetrations for equipment under Items 1 through 6 above -
times up to 6 months after the DBA, consistent with required operating times.
In addition to the above, mechanical equipment (e .g . ,
accumulators and piping) of ESF inside the containment structure would be required to function following the DBA.
Systems outside the containment which contain equipment that would be exposed to the recirculated containment sump water or that would handle post DBA containment atmosphere or air from the contiguous areas are listed below:
- 1. Containment Heat Removal Systems (Section 6.2.2)
- 2. Emergency Core Cooling System (Section 6.3)
- 3. Containment Air Purification and Cleanup System (Section 6.2.3)
- 4. Contain..ent Isolation System (Section 6. 2.0)
- 5. Reactor Plant Ventilation System Emergency Filtration (Secton 6.5.1)
- 6. Combustible Gas Control System (Section 6.2.5)
- 7. Instrumentation and cables associated with Items 1 through 6 3.11.1 Qualification Tests and Analyses Application of IEEE Std 323-1974, IEEE Standard for Oualifying Class IE Equipment for Nuclear Power Generating Stations 4 Stone and Webster Scope of Supply l7 All Class IE equipment will be environmentally qualified in accordance with IEET 323r1974ca) to provide assurance that it is capable of perfornfing its safety function, within the specified g time, when subjected to any of the extremes of the specified environmental envelope. The following information will be provided to the NRC when available, but prior to the OL application date:
3,,,_3 2344 068
WUP Amendment 18 PSAR 6/79 (a) Identification of Class IE equipment including:
(i) Manufacturer, (ii) Manufacturer's type, and (iii) Manufacturer's model number.
(b) Applicant's equipment procurement specification requirements, including:
(i) System safety function requirements, (ii) Environmental envelope, and (iii) Time required to fulfill safety function when subjected to extremes of environmental envelope.
(c) Equipment manufacturer e s test plan, (d) Equipment manufacturer's test set-up, (e) Equipment manufacturer e s test procedures, (f) Equipment manufacturer's acceptability goals and requiremente as approved by the applicant.
The above information will be provided for at least one item in 18 each of the following groups of Class IE equipment as soon as it is available and prior to the OL application:
(a) Switchgear (b) Motor Control Centers (c) Valve Operators (in containment)
(d) Motors (e) Logic Equipment (f) Cables (g) Diesel Generator Control Equipment (h) Sensors (i) Limit Switches (j) Heaters (k) Fans (1) Control Boards g
(m) Instrument Racks and Panels 4 (n) Connectors (o) Penetrations (p) Splices Timely completion of the project schedule may require the design, procurement, and test process to continue while the NRC is reviewing this material.
A summary of the test results will be included with the OL application. In addition, complete, detailed qualification sinformation and test results will be available for an NRC audit.
'The qualification and test results will demonstrate that all Class IE equipment is qualified to the program outlined above.
3.11-4
WUP Amendment 18 PSAR 6/79 Electrical Penetrations Electrical penetrations will be designed, tested, and documented in accordance with IEEE 317-1976(2), and Regulatory Guide l16 1.63(3).
Containment Isolation Valve Actuators The valves inside the containment are designed to operate under the environmental conditions tabulated in Section 3.11. The containment isolation valves are air-operated, f ail-closed valves with solenoid pilot valves. The solenoid valves are high temperature coils with watertight housings. l7 Qualification Tests for Cables Cables in the containment which may be required to function during and after a DBA are qualified in accordance with IEEE 383-1974(*) for the DBA environment of temperature, pressure, 4 humidity, chemical spray, and radiation.
Cable insulation and jacket material are selected to operate in the environments of normal operation or that of the post-accident periods, as required.
Cables inside the containment are designed to withstand the normal radiation dosage and a superimposed DBA radiation dosage, as well as the post-accident environment.
Westinghouse Scope of Supply Class IE equipment will be qualified in accordance with IEEE 323-1974, or related standards as defined by the generic 16 resolution being pursued by Westinghouse and the NRC.
Motors All motors for safety-related items located inside the containment are rated for 1.0 or 1.15 service factor load. The l4 insulation systems are Class B or better. All motors are given at least .tdu) standard NEMA, MGI Routine Tests and Class IE motors are certified to start the specified load with 70 percent terminal voltage.
The containment air fan cooler motors are continuous duty Class I motors used in the containment heat removal system and as such are designed for normal and post-accident environment.
Continuous duty motors inside the containment will be qualified in accordance with IEEE 334-1975(5) and Regulatory Guide 1.40(*). y continuous duty motors outside the containment will be qualified in accordance with IEEE 334-1975C5).
2344 070 3.11-5
WUP Amendment 18 PSAR 6/79 Electric Valve Operators Stone & Webster supplied electric valve operators in the containment, which may be required to fimction during and after a DBA, are tested in accordance with IEEE 382-1972(7) as modified by Regulatory Guide 1.73(*). Electric valve operators outside 7 the containment will be qua3ified in accordance with IEEE 382-1972(7).
Westinghouse supplied electric valve operators will be qualified in accordance with the generic resolution being pursued by 1gl Westinghouse and the NRC.
Motor-operated valve insulation material is selected to operate in the environments of normal operation and that of the post-accident period, as required.
Motor-operated valves inside the containment are designed to withstand normal radiation dosage and a superimposed DBA radiation dosage, as well as the post-accident environment.
Mechanical Equipment All the mechanical equipment (e.g., pumps, valves, tanks, heat exchangers, piping, etc.) of the systems that will recirculate containment sump water or that will be located in the containment and be required to function af ter a DBA will be designed and fabricated of corrosion resistant materials that will remain functionally unaffected by radiation. Qualification tests for these materials will not be performed.
The procurement specifications for each component of the systems define the maximum radiation level at the location where the equipment is installed and also define a radiation dose equal to the 40-year normal accumulated dose in addition to the superimposeu DBA dose. The procurement specifications also define other environmental denign parameters (e.g., temperature and humidity) for the specified equipment. Sufficient design margins are incorporated into the procurement specifications so that the equipment is capable of withstanding the most severe environmental conditions without loss of their safety f unction.
Mechanical seals are specified for pumps which recirculate the containment sump water. Leading seal manufacturers have carried out tests on seals operating under combined high temperature and pressure conditions in a boric acid environment. The seals were previously irradiated to a radiation dose similar to that which would be experienced during and after a DBA. The performance tests indicate that the seals operate without any sign of instability or malfunction such as seal face popping, abnormal power consumption, or hangup.
2344 071 0 3.11-6
WUP Amendment 18 PSAR 6/79 Air-Conditioning and Ventilation Equipment The air-conditioning and ventilation equipment required to service safety-related systems and the control room is specified and procured for the environmental conditions predicted under the normal and post-accident conditions at the installed locations.
Qualification testing of equipment to the applicable environmental conditions is required except for those components whose material properties are not degraded by the environment of pressure, temperature, humidity, and radiation, and which can withstand the analyzed temperature effects such as thermal expansion. The procurement specification and qualification test procedures have identical requirements to those discussed under mechanical equipment.
The control room environmental design bases are described in Section 9.4.
3.11.3 Qualification Test Results (FSAR)
The results of the qualification tests for each type of equipment will be provided in the FSAR.
3.11.4 Loss of Ventilation The design basis and analysis of the air-conditioning and ventilation systems, as described in Sections 6.4.1 and 9.4.1, show that the systems contain sufficient redundancy so that a single failure does not affect the systems' capabilities to provide environmental control, during both normal operation and abnormal conditions, including the DBA. Therefore, safety-related control and electrical equipment in the control building, diesel generator building, and the auxiliary building (outside of the ECCS equipment areas) need not be qualified to extreme environmental conditions.
The essential environmental equipment for the main contcol room, instrument and relay room, and emergency switchgear rooms will be Seismic Category I designed and tornado missile protected. Pover to operate the equipment and instruments will be provided from two separate, independent and redundant Class IE power supplies.
All instrumentation and actuation systems will conform to the l7 requirements of IEEE 279-1971(*).
In the unlikely event of a total failure of the redundant air-conditioning and ventilation systems, the duration of operation of the safety-related controls and electrical equipment located in the affected areas would limit the operation of the s af ety-related systems. The worst case environment requires an eventual shutdown of the reactor because of the effect on the safety-related systems. The conditions of the worst case environment will be presented in the FSAR.
2344 072 3.11-7
WUP Amendment 18 PSAR 6/79 The qualification testing of the control, air-conditioning and ventilation, and electrical equipment under extreme environmental conditions is described in Section 3.11.2 and also in the following sections:
Habitability Systems Functions Design 6.4.1 Control Building Ventilation 9.4.1 References
- 1. IEEE - 323, Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations, 1974.
- 2. IEEE-317, Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations, 1976.
- 3. Regulatory Guide 1.63, Electric Penetration Assemblies in Containment Structures for Water Cooled Nuclear Power Plants.
- 4. IEEE-383, Standard for Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations, 1974.
- 5. IEEE-334, Type Tests of Continuous Duty Class I Motors 18 Installed Inside the Containment of Nuclear Power Generating Station, 1975.
- 6. Regulatory Guide 1.40, Qualification Tests of Continuous Duty Motors Installed Inside the Containment of Water Cooled Nuclear Power Plants.
- 7. IEEE-382, Trial Use Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations, 1972.
- 8. Regulatory Guide 1.73, Qualification Tests of Electric Valve Operators Installed inside the Containment of Nuclear Power Plants.
- 9. IEEE-279, Criteria for Protection Systems for Nuclear Power Generating Stations, 1971.
2344 073 O
3.11-8
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AMENDMENT 18 T J
WUP Amendment 18 PSAR 6/79 PRELInINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 5 Page, Table (T) , or Revision Page, Table (T) , or Revision Piqure (F) Number Fiqure (F) Number 5-i -
T.5.2-6 -
5-11 8 T.S.2-7 (3 pages) 17 5-111 17 T.5.2-8 (3 pages) -
5-iv through 5-vi 8 T.5.2-9 2 5-vii through 5-viiia 17 T.5.2-10 and T.5.2-11 -
5-ix and 5-x 14 T.5.4-t and T.5.4-2 -
5-xi through 5-xiii 17 T.S.5-1 and T.S.5-2 -
5-xiv 8 T.5.5-3 17 5.1-1 through 5.1-6 -
T.5.5-4 (deleted) 17 T.S.5-5 (2 pages) 17 T.5.5-6 through T.S.5-16 -
5.2-1 through 5.2-4a -
5.2-5 through 5.2-24 -
F.5.1-1 2 5.2-25 through 5.2-26c 4 F.5.1-2 through F.5.1-7 -
5.2-27 -
F.5.2-1 through F.5.2-5 -
5.2-28 and 5.2-28a 4 F.5.3-1 -
5.2-29 and 5.2-30 -
F.5.5-1 and F.5.5-2 -
5.2-31 and 5.2-32 17 F.5.5-3 and F.5.5-3A 17 5.2-33 6 F.5.5-4 through F.5.5-8 -
5.2-34 and 5.2-34a 18 F.5.5-9 8 5.2-35 16 5.2-36 and 5.2-36a 7 5.2-37 through 5.1-40c 17 5.2-41 -
5.2-42 and 5.2-42a 2 5.2-43 -
5.2-44 through 5.2-46a 2 5.2-47 through 5.2-51 -
5.2-52 18 5.3-1 -
5.4-1 through 5.4-4 -
5.4-5 18 5.4-6 and 5.4-7 -
5.5-1 through 5.5-7 -
5.5-8 through 5.5-13 17 5.5-14 through 5.5-28 -
5.5-21 5.5-22 through 5.5-25 5.5-26 13 1
2}ff [
5.5-26a 7 5.5-27 14 5.5-28 -
5.5-28a 14 5.5-29 through 5.5-43 -
5.6-1 t.irough 5.6-9 -
T.S.1-1 (2 pages) 17 T.5.2-1 (deleted) 6 P.S.2-2 (3 pages) -
. 5.2-3 T.S.2-4 4 T.S.2-5 (2 pages) -
WUP Amendment 6 PSAR 1/3/75 The chemical and volume control system provides a means for adding chemicals to the RCS which control the pH of the coolant during initial startup and subsequent operation, scavenge oxygen frcm the coolant during startup, and control the oxygen level of the coolant due to radiolysis during all power operations subsequent to startup. The oxygen content and pH limits for power operations are shown in Table 5.2-9.
The pH control chemical employed is lithium hydroxide. This chemical is chosen for its compatibility with the materials and water chemistry of borated water / stainless steel / zirconium /
Inconel systems. In addition, lithium is produced in solution from the neutron irradiation of the dissolved boron in the coolant. The lithium hydroxide is introduced into the RCS via the charging flow. The solution is prepared in the laboratory and poured into the chemical mixing tank. Primary grade water is then used to flush the solution to the sucticn manifold of the charging pumps. The concentration of lithium 7 hydroxide in the RCS is maintained in the range specified for pH control. If the concentration exceeds this range, the cation bed demineralizer is employed in the letdown line in series operation with the mixed bed demineralizer. Since the amount of lithium to be removed is small and its buildup can be readily calculated and determined by analysis, the flow through the cation bed demineralizer is not required to be full letdown flow.
During reactor startup from the cold condition, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution is introduced into the RCS in the same manner as described above for the pH control agent.
Dissolved hydrogen is employed to control and scavenge oxygen produced due to radiolysis of water in the core region.
Sufficient partial pressure of hydrogen is maintained in the volume control tank such that the specified equilibrium concentration of hydrogen is maintained in the reactor coolant.
A self-contained pressure control valve maintains a minimum pressure in the vapor space of the volume control tanM. This can be adjusted to provide the correct equilibritLa hydrogen concentration.
Components with stainless steel sensitized in the manner expected during component f abrication and installation (eg . , cladding) will operate satisfactorily under normal plant chemistry conditions in pressurized water reactor systems because chlorides, fluorides, and particularly oxygen are controlled to very low levels. The pH range given in Table 5.2-9 is the expected range at 256C. At operating temperature in the reactor coolant system, the actual pH range is much smaller since the ionization of boric acid decreases with increases in temperature. 2 Strong base alkali (Li70H) is added to the reactor coolant to provide a basic pH environment relative to pure water at operating temperature. As the reactor coolant temperature 5.2-33
WUP Amendment 18 PSAR 6/79 increases, the acidic nature of boric acid decreases, and the 2 alkaline nature of Li2OH predominates. These conditions ensure that reactor coolant system corrosion rates are minimal.
5.2.4 Fracture Toughness 5.2.4.1 Compliance with Cocle Requirements Assurance of adequate fracture toughness of ferritic materials in the RCPB (ASME Section III, Class 1 components) is provided by compliance with the new requirements for fracture toughness testing included in NB 2300 to Section III of the ASME Boiler and Pressure Vessel Code.
All primary system pressure boundary materials, which are within 18 the scope of 10CFR50, Appendix G, will be shown to satisfy the applicable requirements in the FSAR.
Class 2 and 3 components will comply with ASME Section III fra ture toughness test methods and acceptance criteria where 2
required. Class 2 vessels will comply with Summer 1972 Code Addendum.
5.2.4.2 Acceptable Fracture Energy Levels The initial upper shelf fracture energy levels for the materials of the reactor vessel beltline (including welds) shall be determined by Charpy V-notch tests. The specimens shall be oriented as required by NB 2300 of Section III of the ASME Boiler and Pressure Vessel Code.
Appendix G to 10CFR50 specifies additional fracture toughness requirements for ferritic materials of pressure retaining 2
components of the RCPB that are not required by Paragraph NB-2300 of Section III of the ASME Boiler and Pressure Vessel Code.
18l 5.2.4.3 operating Limitations during Startup and Shutdown The operating curves, including pressure-temperature limitations, are calculated in accordance with 10CFR50, Appendix G and ASME 7
Code Section III, Appendix G requirements. Actual material property test data will be used. The methods outlined in Appendix G to Section III will be employed for the shell regions in the analysis of protection against nonductile failure. The initial operating curves are calculated assuming a period of reactor operation such that the beltline material will be limiting (Fig. 5.2-1 and 5.2-2.). Beltline material properties degrade with radiation exposure, and this degradation is measured in terms of the adjusted reference nil-ductility temperature which includes a reference temperature ( ARTNDT) . Predicted A RT NDT values are derived by using Fig. 5.2-3 and the maximum fluence from Fig. 5.2-4 at 1/4T (thickness) and 3/4T locations 2344 078 5.2-34
WUP Amendment 18 PSAR 6/79 (tips of the Code reference flaw when the flaw is assumed at I.D.
and O.D. locations, respectively).
2344 079 1
5.2-34a
WUP PSAR
- 2. A man-way is provided in the pressurizer top head to allow access for internal inspection of the pressurizer.
- 3. The insulation covering all component and piping welds and adjacent base metal is designed for ease of removal and replacement in areas where external inspection is planned.
- 4. Removable plugs in the concrete floor above the reactor coolant pumps permit removal of the pump motor. This allows for internal inspection of the pumps.
- 5. The reactor coolant loop compartments are designed to allow personnel entry during refueling operations. This permits direct inspection access to the external portion of the piping and components.
5.2.8.2 Equipment for Inservice Inspections Various equipment is available to perform reactor vessel and nozzle inservice inspections. The specific inspection tool used depends upon the contractor selected to perform the inservice inspection. All contractors under consideration use devices capable of ultrasonically inspecting all reactor vessel shell welds and nozzle safe-end welds from inside the vessel.
5.2.8.3 Recording and Comparing Data The Applicants will compile appropriate examination records of the preoperational examination for the purpose of comparison with the examination results of subsequent inservice inspections. The system used to record data will be developed at the FSAR stage of licensing and will incorporate experience gained at the Applicants' Point Beach Nuclear Plant.
5.2.8.4 Reactor Vessel Acceptance Standards The acceptance standards used to establish acceptability of the reactor vessel are in accordance with Article 1GB-3000 and Appendix A of ASME XI.
5.2.8.5 Coordination of Inspection Equipment with Access Provisions To ensure access provisions in the design for inservice inspection, the Applicants identify pressure vessels, piping, pumps, and valves which require inspection. An inservice inspection coordinator is designated by the architect-engineer to review the Applicants' information and assure compliance with governing inspection requirements.
The Applicants' inservice inspection program is designed to utilize proven equipment and techniques. It is not anticipated s.2-s, 2344 080
WUP Amendment 18 PSAR 6/79 that significant developmental or experimental programs will be required.
5.2.9 References
- 1. Cooper, K.; Starek, R. M.; and Miselis, V. Overpressure Protection for Westinghouse Pressurized Water Reactors.
WCAP-7769, Revision 1, June 1972.
- 2. Nay, J. A. Process Instrumentation for Westinghouse Nuclear Steam Supply System. WCAP-7671, April 1971.
- 3. Hazelton, W. S., et. al. Basis for Heatup and Cooldown Limit Curves. WCAP-7924, July 1972.
- 4. Shabbits, W. O. Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel Plate. WCAP-7623, December 1970.
- 5. Szyslowski J. J. and Salvatori, R. Determination of Design Pipe Breaks for the Westinghouse Reactor Coolant System.
WCAP-7503, February 1972.
- 6. Enrietto, J. F. Control of Delta Ferrite Austenitic Stainless 2
Steels Weldments. WCAP-8324, May 1974.
18 2344 081
- n >
O 5.2-52
WUP Amendment 18 PSAR 6/79 they are acceptable. The gr~na stresses are low and thus have a negligible effect on the strta intensity in the vessel.
5.4.3.4 Thermal Stresses Due to Loss-of-Coolant Accident Eracture mechanics evaluation of the reactor vessel due to thermal stresses following a LOCA are discussed in Section 5.2.
5.4.3.5 Heatup and Cooldown Heatup and cooldown requirements for the reactor vessel are discussed in Section 5.2.
5.4.3.6 Irradiation Surveillance Program The program will conform with ASTM E,-185-73
- Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" and 10CFR50, Appendix H. Six surveillance capsules will be used in the surveillance program to monitor changes in the RTNDT of the beltline region material, predicted to be most limiting based on original fracture toughness properties and estimated changes considering copper and phosphorous contents. Beltline region material sufficient for at least two additional capsules will be retained. Each capsule will contain the following type and number of specimens:
No. of No. of No. of 1/2 T Material Cha:7pys Tensiles Compact Tensions Limiting Base Material
- 15 3 4 Limiting Base Material ** 15 3 4 Weld Metal *** 15 3 4 15 - -
Weld Heat.g:ffect . Zone
- Specimens oriented in the major rolling or working direction.
- Specimens oriented normal to the major rolling or working direction.
- Weld metal to be selected per ASTM E185-73.
The reactor vessel materials surveillance capsule withdrawal schedule will be in compliance with 10CPR50, Appendix H and ASTM 18 E-185. The withdrawal schedule will be discussed in the FSAR.
2344 082 5.4-5
WUP PSAR Archive material from beltline region plates or forgings and a core region weldment will be retained.
The capsules can be removed for testing when the vessel head is removed, and the vessel is prepared for the refueling operations.
A guide basket is welded to the neutron pad assembly. Samples are contained in a stainless steel sheathed capsule which is fastened to the basket lid. Sample recovery is accomplished by remote handling equipment. If necessary, capsules can be replaced when the internals are removed.
5.4.3.7 Capability for Annealing the Reactor Vessel There are no special design features which would prohibit the in situ annealing of the vessel. If the unlikely need for an annealing operation was required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature greater than 6500F for a period up to 168 hr would be applied. Various modes of heating may be used depending on the annealing temperature.
The reactor vessel materials surveillance program is adequate to accanmodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the annealing treatment.
5.4.4 Tests and Inspections The reactor vessel quality assurance program is given in Table 5.4-2 and Chapter 17. All nondestructive testing, inspection, and acceptance standards are in accordance with requirements of Section III of the ASME Code.
5.4.4.1 Ult _rasonic Examinations
- 1. All forged materials are inspected in accordance with Section III of the ASME Co4e, except that rings and hollow forgings are additionally examined using angle beam techniques in the 40-65 desji.2e range.
- 2. During fabrication, angle beam inspection of 100 percent of plate material is performed in afiition to the design code straight beam ultrasonic test.
- 3. All full penetration welds and heat affected zones in the primary pressure boundary welds are inspected upon completion of welding and intermediate heat treatment.
- 4. The reactor vessel is examined af ter hydro-testing to provide a base line map for use as a reference document in relation to later inservice inspections.
2344 083 O
5.4-6
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 6 Page, Table (T) , or Revision Page, Table (T) , or Revision Fiqure (F) Number Fiqure (F) Number 6-i 8 6.3-4 and 6.3-5 2 6-11 17 6.3-6 -
6-111 through 6-v1 15 6.3-7 through 6.3-10 2 6-vii 18 6.3-11 through 6.3-13 -
6-viii 8 6.3-14 14 d-1x throuch 6-xa 17 6.3-14a 1 6-xi and 6-xii 8 6.3-15 -
6-xiii and 6-xiv 17 6.3-16 14 6-xv 11 6.3-17 through 6.3-18a 2 6.3-19 and 6.3-20 -
6.1-1 -
6.3-21 1 6.1-2 10 6.3-22 and 6.0-22a 14 6.1-3 14 6.3-23 14 6.1-4 -
6.3-24 and 6.3-24a 9 6.3-25 -
6.2-1 -
6.3-26 through 6.3-28 16 6.2-2 6 6.3-29 1 6.2-2a 10 6.3-30 7 6.2-3 8 6.3-30a 11 6.2-4 17 6.3-31 and 6.3-32 2 6.2-4a through 6.2-4c 3 6.3-32a 11 6.2-4d 8 6.3-33 11 6.2-4e through 6.2-49 17 6.3-34 -
6.2-5 and 6.2-6 17 6.2-ba through 6.2-69 8 6.4-1 14 6.2-7 4 6.4-2 2 6.2-8 through 6.2-10 -
6.4-3 1 6.2-11 4 6.4-4 and 6.4-4a 7 6.2- 12 and 6.2-13 -
6.4-5 7 6.2-14 through 6.2-14c 1 6.4-6 14 0.2-14d and 6.2-14e 8 6.4-7 15 6.2-15 throagh 6.2-16d 17 6.4-8 7 6.2-17 through 6.2-22 8 6.4-9 17 6.2-23 through 6.2-27 -
6.4-10 14 6.2-28 9 6.4-11 through 6.4-13 7 6.2-28a through 6.2-28c 8 6.4-14 and 6.4- 15 14 6.2-28d through 6.2-32b 17 6.4-16 7 6.2-33 -
6.4-17 14 6.2-34 8 6.2-35 through 6.2-36a 17 6.5-1 and 6.5-2 -
6.2-37 through 6.2-52b 15 6.5-3 and 6.5-4 16 6.2-52c through 6.2-52e 17 6.5-5 through 6.5-12 18 6.2-52t and 6.2-52g 15 6.2-53 -
T.6-1 (page 1) 15 6.2-54 5 (pages 2 and 3) 11 6.2-55 through 6.2-56a 17 (page 4) 10 6.2-57 and 6.2-58 -
(page 5) 6 6.2-59 5 (page 6) 11 6.2-60 6 (pages 7 and 8) 6 6.2-60a 5 (page 9) 10 6.2-61 3 (pages 10 through 12b) 6 6.2-62 and 6.2-62a 6 (pages 13 and 14) 4 6.2-63 1 (pages 15 and 15) 11 (pages 17 and 1P) -
6.3-1 and 6.5-2 2 T.6-2 (page 1) 2 6.3-3 -
(page 2) 11 tv 6-i 2344 084
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CO W D)
Chapter 6 Page, Table (T) , or Revision Page, Table (T) , or Revision Fiqure (F) Number Fiqure (F) Number T.6-2 (page 3) 2 F.6.2.1-16 through T.6.2.1-1 through F.6.2.1-22 -
T.6.2.1-3 -
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WUP Amendment 18 PSAR 6/79 CHAPTER 6 TABLE OF CONTENTS (CONT ep)
Section Title Page 6.5 EMERGENCY FILTRATION SYSTEMS 6.5-1 6.5.1 Reactor Plant Ventilation System Emergency Filtration 6.5-1 6.5.1.1 Design Bases 6.5-1 6.5.1.2 System Description 6.5-2 6.5.1.3 Design Evaluation 6.5-4 6.5.1.4 Test and Inspection 6.5-4 6.5.1.5 Instrumentation Requirements 6.5-5 6.5.2 Fuel Building Ventilation System Emergency Filtration 6.5-6 6.5.2.1 Design Bases 6.5-6 6.5.2.2 System Description 6.5-7 6.5.2.3 Design Evaluation 6.5-9 6.5.2.4 Test and Inspection Requirements 6.5-10 6.5.2.5 Instrumentation Applications 6.5-11 2344 086 6-vii
WUP Amendment 8 PSAR 6/16/75 CHAPTER 6 LIST OF TABLES Table Title 6-1 Containment Design Evaluation Parameters 6-2 Typical Materials Employed for Components of Con-tainment Spray, Containment Structure Ventilation, Combustible Gas Control, and Containment Isolation Systems 6.2.1-1 Summary of Heat Transfer Correlations Used to Calculate Steam Generator Heat Flow in the SATAN Code 6.2.1-2 Core Stored Energy for Generic 17x17 Fuel 6.2.1-3 Energy Balance Table, Double-Ended Pump Suction Break 6.2.1-4 Available Energy in Steam Generator 6.2.1-5 Minimum ECCS Hydraulic Chara cteris tics for Post Reflood (One Intact Loop) at 183 Seconds g 6.2.1-6 Minimum ECCS Hydraulic Characteristics for Post Reflood (Broken Loop) at 183 Seconds 6.2.1-7 Accident Chronology Pump Suction DER, Minimum ESF 6.2.1-8 Subcompartment Design Pressure Differentials 6.2.1-9 Mass and Energy Release Rates 150 In.2 Cold Leg LDR-Reactor Vessel Cavity 6.2.1-10 Summary of Reactor Cavity Subcompartment Vent Loss Coefficients 6.2.1-11 Mass and Fnergy Release Rates Pump Suction LDR (4.88 f t2) - Steam Generator Cubicle 6.2.1-12 Node Length to Area Ratio's Steam Generator Com-partment with RELAP4 6.2*.1-13 K-Factors and Vent Areas and Vent Flow Models for Steam Generator Analysis 6.2.1-14 Mass and Energy Release Rates Pressurizer Surge Line Break-Pressurizer Cubicle O
2344 087 6-viii
WUP Amendment 18 PSAR 6/79
- 2. HEPA filters -
Tested individually by the appropriate Filter Test Facility listed in the current USDOE Environmental Health and Safety Bulletin for Filter Unit Inspection and Testing Service. 14
- 3. Adsorber impregnated carbon - Each batch, original or replacement, conforms with Table 2 of Regulatory Guide 1.52.
Inplace Tests Following Initial Installation
- 1. Emergency filtration collected ductwork is balanced and system flow is measured to ensure that the trains and associated ductwork, dacpers, and valves meet the design criteria.
- 3. HEPA filters -
Inplace DOP testing in accordance with 14 ANSI N510-1975 Section 10. The test is to confirm a penetration of less than 0.05 pereent at rated flow.
- 4. Adsorber - The leak test is performed in accordance with ANSI N510-1975 Section 12 using gaseous halogenated hydrocarbon refrigerant with an upstream concentration no greater than 20 ppm. The allowed bypass leakage is 14 0.05 percent. Following the completion of the test, air flow is continued until ef fluent refrigerant gas is less than 0.01 ppm.
Periodic tests and inspections of the RPVS energency filtration are performed in accordance with Technical Specifications (Chapter 16) .
6.5.1.5 Instrumentation Applications
- 1. Differential pressure switches are provided across the filters in each train to detect clogged filters and actuate an alarm in the control room on high differential pressure.
- 2. Flow switches downstream of the fan in each train actuate an alarm in the control room if the fan motor receives a start signal and no flow is realized, indicating a failure of the train or of the isolation damper.
2344 088 6.5-5 0?
WUP Amendment 18 PSAR 6/79 6.5.2 Fuel Building Ventilation System Emergency Filtration 6.5.2.1 Design Bases y The fuel building ventilation system (FBVS) incorporates two emergency carbon adsorbers/HEPA filtration trains designed to:
- 1. Draw sufficient air from the fuel building to maintain a g negative pressure of 0.25 in. water gauge in the structure during fuel handling operations.
- 2. Filter fuel building exhaust flow through carbon 14 adsorbers and HEPA filters to remove radioiodine and particulates, respectively.
- 3. Provide a backup to the containment structure ventilation system for purging the containment structure 18l (except during fuel handling operations) .
The FBVS emergency tiltration trains are used to mitigate the consequences of a fuel handling accident and therefore are designed for a single active failure in the short term (Section 3.1.1) . Operation of one of the FBVS emergency 18l filtration trains is required during fuel handling operations.
It is desirable that the fuel building supply air handling units be stopped to reduce the exhaust rate from the building in the event of a fuel handling accident, but failure of the operator to do so dces not result in a conservatively calculated dose to the public in excess of the guideline values established by 10CFR100.
No credit is taken for a reduction in fuel building exhaust rate or for an elevated release in the analysis of the fuel handling accident (Section 15.4.5) .
The only source of radioiodine in the fuel building is the spent fuel stored in the spent fuel pool. Since the fuel is stored in Seismic Category I racks which prevent damage to the fuel during an earthquake, the only credible mechanisms by which significant quantities of radioiodine can be released are due to a damaged fuel assembly or the drop of a heavy object onto stored spent fuel assemblies.
One of the FBVS emergency filtration trains is required to be in operation:
- 1. Whenever spent fuel is being handled (ref ueling, defueling, and during removal of spent fuel from the spent fuel pool for offsite shipment) .
- 2. Whenever work in the fuel building requires handling of heavy objects over stored spent fuel.
2344 089 6.5-6
WUP Amendment 18 PSAR 6/79 operation of the FBVS emergency filtration trains is initiated manually from either the fuel building or from the control room.
Normal fuel building ventilation exhaust is through a HEPA filter to minimize the release of airborne particulates (Section 9.4.5) .
The fuel building vent monitor detects high radioactivity in the fuel building vent and actuates an alarm in the control room. l18 Operation of one of the FBVS emergency filtration trains is then manually initiated to reduce releases.
The FBVS emergency filtration trains can be used to filter containment purge air in the event that both containment air filtration filters are inoperable (Section 9.4.7.2) . This function is not safety-related. The FBVS emergency filtration trains are normally isolated from the containment purge exhaust duct by a normally closed valve under administrative control. l18 The fuel building ventilation system includes unit coolers to accommodate equipment heat loads in the fuel building. The unit coolers in safety-related equipment cubicles are seismic category I safety-related and are supplied from the safety-related reactor plant component cooling water system (Section 9.2.2) . Under no conditions is it necessary to supply outside air to the fuel building for equipment cooling.
The fuel building is a seismically designed, safety-related structure with gasketed doors and hatches and sealed piping penetrations. The scroll door to the railroad car bay is of a special design to limit inleakage and is required to be closed l18 during fuel handling operations. Building inleakage is conservatively estimated to be 3,000 cfm for the 0.25 in. water l18 gauge negative pressure maintained within the structure.
The extent of compliance of the FbVS emergency filtration trains with Regulatory Guide 1.52 is discussed in Appendix A.
6.5.2.2 System Description A diagram of fuel building ventilation system emergency filtration is shown on Fig. 6.5.2-1.
The fuel building is normally maintained at a slightly negative pressure by supplying air at a rate of 8,000 cfm while exhausting air through a HEPA filter at a rate of 10,000 cfm (Section 9.4.5) . During fuel handling operations (as defined in Section 6.5.2.1) , exhaust flow is diverted through one of the 11,000 cfm FBVS emergency filtration trains while maintaining 18 normal supply. Eight thousand cfm are drawn from the spent fuel pool surface area, while 3,000 cfm are drawn from general areas in the fuel building. Frel building negative pressure is maintained at a minimum of 0.25 in. water gauge relative to 18 outside atmosphere during fuel handling operations. In the event of a fuel handling accident, the supply fan is manually stopped 2344 090
.Np 6.5-7
WUP Amendment 18 PSAR 6/79 to reduce the fuel building exhaust rate. At times other than during fuel handling operations, the operator manually initiates l
operation of one of the FBVS emergency filtration trains on a high radioactivity alarm in the fuel building vent.
To meet the single failure criteria for an active failure in the short term, there are two 100 percent capacity FBVS emergency filtration trains. One train is in operation and the other is on automatic standby. If the start switch for a given train is in the "ON" position and flow is lost (indicating a loss of power, a motor operator failure, or a fan motor failure), the redundant train automatically starts, power is cut off to the failed train, and an alarm is actuated in the control room. The automatic controls are designed for a single failure in that no single failure of the control circuitry can stop operation of both trains. The controls are energized by the vital bus (Section 8.3.2) .
Prior to actuation of the FBVS emergency filtration train, the normal exhaust from the HEPA filters (Section 9.4.5) is 18 terminated manually by closing the two air-operated isolation valves downstream of the normally operating HEPA filters and the fuel building normal ventilation exhaust fan is manually stopped.
Actuation of the above ensures that an open flowpath exists to the FBVS emergency filtration trains and that all filtered air is 18l exhausted from the fuel building to maintain the negative pressure within the building. The air operated isolation valve and the motor-operated dampers are Seismic Category I and safety related.
Prior to operating one of the FBVS emergency filtration trains during refueling, manually-operated dampers in the FBVS distribution and collection ductwork are aligned to provide an air sweep across the spent fuel pool as described in 18l Section 9.4.5. Administrative control is exercised over the dampers in the collection ductwork such that during fuel handling operations an open flowpath to the FBVS emergency filtration trains is ensured. The manually operated isolation damper just upstream of 18l handling operations.the emergency filtration trains is open during fuel The manually operated isolation damper just upstream of the FBVS emergency filtration trains is closed only if the trains are to be used to filter the containnent purge air. In order to align the system for containment purge:
- 1. The manually operated isolation damper upstream of the emergency filtration trains is closed.
- 2. The normally closed valve in containment purge line 18 connection to the FBVS is opened. The containment purge discharge damper to the normal fuel building vent is
, 6.5-8 2344 Qgl
WUP Amendment 18 PSAR 6/79 remote manually closed which automatically opens the motor-operated containment air filter bypass damper in 18 containment.
The FBVS emergency filtrEtion trains are Seismic Category I and safety related. The trains are located in the saftty-related fuel building which affords the trains protection from externally generated missiles and extreme natural phenomena. Protection from internally generated missiles and biological shielding for plant personnel is provided by the concrete FBVS emergency filtration train cubicles.
The FBVS emergency filtration trains incorporate electric heaters to limit the relative humidity of incoming air to 70 percent.
Charcoal bed depth is a minimum of 2 inches. Units of this type have demonstrated iodine removal efficiencies of 99.9 percent for both elemental and organic forms of iodine. Iodine removal efficiency is conservatively assumed to be 95 percent for accident analysis.
The carbon adsorbers employed in the FBVS emergency filtration l14 trains are of gasketless design to facilitate removal of charcoal and to reduce the potential for leakage. Fire protection for the g adsorber is provided by manually actuated water spray systems (Section 9.5.1) .
Low-flow air bleed cooling systems are not provided for the FBVS emergency filtration trains. Analysis indicates that the maximum temperature for the carbon adsorbers is less than the maximum 16 temperature specified in Section 4.9 of ANSI N509 Nuclear Power Plant Air Cleaning Units and Components (1976).
6.5.2.3 Design Evaluation The FBVS emergency filtration trains are adequately sized and designed to: 1) maintain a negative pressure of 0.25 in. water gauge in the fuel building during fuel handling operations and 18
- 2) provide the capability for processing fuel building exhaust air through carbon adsorbers and HEPA filters to minimize radioiodine and particulates in the fuel building vent.
Operation of one of the two available trains is required during l18 fuel handling operations. The FBVS emergency filtration trains are designed to meet a single active failure in the short term (Section 3.1.1) , which satisfies the requirements of the fuel handling accident analysis (Section 15.4.5) . A 95 percent removal efficiency for both organic and elemental forms of iodine is ensured by the design of the trains (humidity control and carbon bed depth) and by the program of inspections and testa described in Section 6.5.2.4. Protection of the trains is ensured by their location within a shielded cubicle in the Seismic Category I, reinforced concrete, safety-related fuel building.
ro*
a
. 2344 092 6.5-9
WUP Amendment 18 PSAR 6/79 Low-flow air bleed cooling systems are not provided since analysis indicates that the maximum carbon adsorber temperature is less than that. allowed in Section 4.9 of ANSI N509, Nuclear Power Plant Air Cleaning Units and Components (1976).
Radioactive iodine inventories are based upon collecting 100 percent of the radioiodines released in a puff using the assumptions of Safety Guide 25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling 16 and Pressurized Water Reactors. This inventory is collected on the surface of the carbon adsorbers, which are immediately isolated. The energy due to beta decay is deposited on the surface of the carbon adsorber while the gamma energy is uniformly deposited in the carbon adsorber. No credit is taken for heat transfer to the surroundings nor for absorption in the filter's structural materials. Parameters used in the analysis are listed in Tables 6.5-1 and 15.4.5-1. The carbon adsorber surface temperature as a tunction of time is given in Table 6.5-3.
During fuel handling operations, the building will be maintained at a negative pressure of 0.25 in. water gauge relative to outside atmosphere, and will be monitored by redundant ditterential pressure indicators. When a fuel building emergency filter is operating and low negative pressure alarm is received (e.g. , less than 0.25 in. water gauge vacuum), the normal ventilation supply isolation dampers (Fig. 9. 4. 5-1) are manually closed until the cause of excessive air leakage is determined to ensure that no open flowpaths are available which could bypass 18 the emergency filtration trains in the event of a fuel handling accident. The normal ventilation supply isolation dampers are Seismic Category I.
A recirculation line with a motor-operated damper is provided at each filtration fan discharge to prevent excessively low pressures in the building if the supply air system is not operating . If a high negative pressure signal (e.g., greater than 0.75 in. water gauge vacuum) is received when the emergency filtration system is operating, the recirculation damper for the operating train will open, and a portion of the discharge air will flow through the recirculation ductwork into the general areas of the building. This will prevent excessive negative pressure in the building and will provide a greater filtration rate for the building air.
6.5.2.4 Test and Inspection Requirements The following tests and inspections are performed on the FBVS emergency filtratioa trains:
2344 093 $
6.5-10
WUP Amendment 18 PSAR 6/79 Pre-installation Tests ,
- 1. Filter train pressure test - Trains are evacuated to the test pressure and are observed for 2 hr. No more than 1 in. of water change should be observed.
- 2. HEPA filters -
Tested individually by the appropriate y Filter Test Facility listed in the current USDOE Environmental Health and Safety Bulletin for Filter Unit Inspection and Testing Service.
- 3. Adsorber (impregnated carbon) - Each batch, original or replacement, conforms with Table 2 of Guide 1.52. Regulatory l14 Inplace Tests Following Initial Installation
- 1. Emergency filtration collection ductwork is balanced and system flow is measured to ensure that the trains and associated ductwork, dampers, and valves meet the design criteria.
- 3. HEPA filters -
Inplace DOP testing in accordance with ANSI N510-1975 Section 10. The test is to confirm a penetration of less than 0.05 percent at rated flow. g
- 4. Adsorber - The leak test is performed in accordance with ANSI N510-1975 Section 12 using gaseous halogenated hydrocarbon refrigerant with an upstream concentration no greater than 20 ppm. The allowed bypass leakage is 0.05 percent. Following the completion of the test, air flow is continued until effluent refrigerant gas is less than 0.01 ppm.
Periodic Tests and Inspections Periodic tests and inspections of the FBVS emergency filtration are performed in accordance with Technical Specifications (Chapter 16) .
6.5.2.5 Instrumentation Applications
- 1. Differential pressure switches are provided acrosc the filters in each train to detect clogged filters and actuate an alarm in the control room on high differential pressure.
- 2. Flow switches downstream of the fan in each train actuate an alarm in the control room if power is supplied to the train and flow is lost, indicating a 6.5-11
WUP Amendment 18 PSAR 6/79 fas. lure of the train or of the isolation dampers. Lou flow starts the redundant train.
- 3. Differential pressure switches initiate high and low alarms to monitor fuel building negative pressure.
18 Moreover, they initiate recirculation of the eergency filtration units to prevent excessive negative pl.essures in the building if the supply air system is not operating.
2344 095 0
O 6.5-12
a
/
d FUEL BUILDING uA -
A H
FUEL s , FR0W FBVS SUPPLY I Bull 0 LNG VENT
- - AIR HANDLING UNITS i
\
FIG. 9.4.5-1 r *-~-----
i j p. a TO
^ FROM FBVS I
'FIC.9.4.5-1
' u r'
1 r-
/ NCRMAL HEPA l
-- T FC FC' N FilfERS pig,g,4,5_; [
SC 3 >l4 NNS_ 50 3 I 4Ah '
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FSN I
I 3 -
l i
[J.___._________
i t
r-------T ~~~'------] '
Ah
~
I V TO
[_gFIG.9.4.5-1 NOTES:
- 1. VALVES AND DAMPERS SHOWN AllGNED FOR REFUELING.
- 2. SYSTEM IS SAFETY CLASS 3 (SC-3) FICEPT AS SHOWN.
2344 096 "s
f
6 F _ _ -
3\
FROM f 2 ( G5NE AL AREAS l 1
NN S *+SC-3 8,000 CFM TO FBYS '
MORMAL HEPA _
FILTERS i
t_;
n -
SC-3 SPENT FUEL l
_____ _ _ _ . 4 POOL
~
E vv l
~ <
_s F.C. SC-3 >4 NNS vv; l
_ _ _ _ _ _ _ _ . g FROM CONTAINMENT
, I r, , x PURGE AIR I E vv l-u FIG. 9.4.7-2 J vv N vv
> vv F.C.
/
_ _ _ . _ _ _ a I I
hb t F IG. 6.5.2-1 FUEL BUILDING EMERGENCY FILTRATION WISCONSIN UTILITIES PROJECT ,
PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT 18 )
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 8 Page, Table (T) , or Revision Fiqure iF) Number 8-i 18 B-11 17 8-111 8 8-iv 16 8.1-1 7 8.1-2 through 8.1-6 -
8.1-7 through 8.1-9 4 8.2-1 18 8.2-2 2 8.3-1 -
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8.3-11 2 8.3-12 and 8.3-12a 6 8.3-13 through 8.3-14a 6 u.3-15 and 8.3-16 16 8.3-17 through 8.3-21 17 8.3-22 through 8.3-24a 18 8.3-25 1 8.3-26 through 8.3-28 15 8.3-29 and 8.3-30 1 8.3-31 16 8.3-32 7 8.3-33 through 8.3-34a 2 8.3-35 6 8.3-36 -
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':::3.;_, ; 2344 078 F.8.3.1-2 -
F.8.3.1-3 and F.8.3.1-4 7 F.8.3.1-3 and F.8.3.1-6 16 F.8.3.2 ' -
WUP hmendment 18 PSAR 6/79 CHAPTER 8 ELECTRIC POWER TABLE OF CONTENTS Section Title Page
8.1 INTRODUCTION
8.1-1 8.1.1 Utility Grid 8.1-1 8.1.2 Interconnection to Other Grids 8.1-1 8.1.3 Transmission System at Site 8.1-1 8.1.4 Onsite A-C System 8.1-1 8.1.5 Onsite D-C Systems 8.1-3 8.1.6 Identification of Safety-Related Systems 8.1-4 8.1.6.1 System Functions 8.1-4 8.1.6.2 Power Supply Sources 8.1-5 8.1.7 Identification of Safety Criteria 8.1-6 8.2 OFFSITE POWER SYSTEMS 8.2-1 8.2.1 Description 6.2-1 8.2.2 Analyses 8.2-2 8.3 ONSITE POWER SYSTEMS 8.3-1 8.3.1 A-C Power Systems 8.3-1 8.3.1.1 Description 8.3-1 8.3.1.1.1 Plant Distributton Network 8.3-1 8.3.1.1.2 Onsite Standby Power Supply 8.3-5 8.3.1.1.3 120 V A-C Vital Bus System 8.3-10 8.3.1.1.4 Equipment Criteria 8.3-11 8.3.1.2 Analysis 8.3-14 8.3.1.2.1 Compliance Analysis 8.3-14 8.3.1.2.2 Hostile Environments 8.3-17 2344 0?9-8-i
WUP Amendment 18 PSAR 6/79
_ CHAPTER 8 TABLE OF CONTENTS (CONT *D)
Section Title Page 8.3.1.2.2.1 Equipment Identification (Refer to 8.3-17 Section 3.11.1) 8.3.1.2.2.2 Qualification Tests (Refer to 8.3-18 Section 3.11.2) 8.3.1.3 Conformance with Appropriate Quality 8.3-20 Assurance Standards 8.3.1.4 Independence 01 Redundant Systems 8.3-20 8.3.1.4.1 Principal Criterion 8.3-20 8.3.1.4.2 Administrative Responsibility for 8.3-20 Compliance 8.3.1.4.3 Equipment Consideration 8.3-20 8.3.1.4.4 Cable Considerations 8.3-22 8.3.1.5 Physical Identification of Safety-Related Equipment 8.3-29 O 8.3.2 D-C Power Systems 8.3-32 8.3.2.1 Description 8.3-32 8.3.2.1.1 General 8.3-32 8.3.2.1.2 Safety-Related Systems 8.3-32 8.3.2.1.3 Nonsafety-Related System 8.3-32 8.3.2.1.4 Plant D-C System 8.3-33 8.3.2.1.5 Testing 8.3-34 8.3.2.2 Analysis 8.3-35
, 2344 100 g 8-ii
WUP Amendment 18 PSAR 6/79 8.2 OFFSITE POWER SYSTEMS 8.2.1 Description Electric energy generated at 22 kV is stepped up to 345 kV by the main transformers, and delivered through 345 kV, 25,000 MVA, 2,000 ampere circuit breakers to the 345 kV switching station located at the plant site. This station uses a breaker arrangement as shown on Fig. 8. b2. Each generating unit can be connected to one or both of the two 345 kV bus sections. All 345 kV and 69 kV circuit breakers will be furnished with two independent trip coils. Two batteries, each with its own charger, will be utilized in the switchyard. One trip coil will be operated from the primary protective relaying and by operation of the control switch using the d-c supply from one battery. The 2 other trip coil will be operated by backup relaying with its d-c supply from the other battery. Two 480 V full capacity a-c supplies will be installed to the switchyard to serve the switchyard auxiliary equipment such as battery chargers, air compressors, lighting, heating, ventilating, etc. Normally, each supply will serve one half of the auxiliaries. If one supply is lost, all of the auxiliary load can be supplied from the other 480 V supply. At least four transmission lines will tie into the interconnected transmission network of the Wisconsin Utilities.
This network also includes extensive underlying lower voltage systems operating at 138, 115, and 69 kV. The Wisconsin Utilities in turn are interconnected with CcInmonwealth Edison Con.pany, Northern States Power Company, and Upper Peninsular Power Company. These interconnections consist of four 345 kV tie lines and three lower voltage tie lines.
The two 345 to 69 kV auxiliary transformers are connected to 345 kV bus sections which are not adjacent. These transformers supply the 69 kV ring bus as shown on Fig. 8. b2. A combustion turbine generator may be installed and connected to the 69 kV switchyard. This unit would be used as a backup supply to plant auxiliary loads for system peaking and system standby reserve. l18 Alarms, controls, and indications would be provided so that the combustion turbine generator could be started, synchronized, loaded, and tripped locally and from the control room.
The onsite distribution system for each unit is supplied by two l18 independent 69 kV underground lines connected to nonadjacent sections of the 69 kV ring bus as shown on Fig. 8. b2 and 8.2-1.
2344 101 8.2-1
WUP Amentdment 2 PSAR 1/3/75 8.2.2 Analyses The Wisconsin Utilities generation and transmission system is designed to withstand severe contingencies without resulting in an uncontrolled widespread tripping of lines and/or generators.
Such contingencies include the outage of a double circuit transmission line, an entire substation, or an entire power plant. This means that sequential loss of two units at this plant would not adversely affect the transmission system, and the power to replace the lost generation could be supplied by the Wisconsin Utilities internal reserve and its interconnected transmission system. The 345 kV transmission lines serving the plant would continue to be energized from the transmission system. System design for stability and circuit isolation will prevent the sudden loss of one unit from causing the second unit to trip. Results of a complete steady state and transient 2 stability analysis are provided in the Site Addendum to the PSAR for each site. The transmission lines are designed to withstand 1/2 in. of ice with a 4 psf wind load at 00F; or a 100 mph wind on bare conductors, whichever is greater. This exceeds the most severe ice and wind loading expected.
The switchyards and transmission lines from the plant are designed to meet the requirements of General Design Criteria 17 and Regulatory Guide 1.32.
2344 1 0'2 O
8.2-2
WUP Amendment 18 PSAR 6/79
- 1. System phase fault on the 4.16 kV bus
- 2. Feeder phase fault with a stuck breaker 6
For these conditions, the onsite source will not be connected to the emergency bus. The two independent emergency 4,160 V buses constitute separate power supplies to the safety-related loads.
Certain safety-related loads may receive power from either emergency 4,160 V bus H2O or emergency 4,160 V bus H40. A system of interlocks is provided so that the redundant emergency systems are never tied together.
The emergency 4,160 V buses consist of indoor metal-clad switchgear located within the control building which is a Seismic Category I, tornado protected structure. These buses are physically and electrically segregated so that any single failure which might affect one bus will not jeopardize proper operation of the other bus as shown on Fig. 8.3.1-5. The structure contains an automatic fire protection system as described in l16 Section 9.5.1. There is no possibility of accident generated missiles in the area. Redundant air-conditioning is available to provide temperature control of the rooms in which the switchgear is located. This ventilation system is discussed in Section 9.4.
Bus feeder circuit breaker control switches and bus synchronizing switches for the normal and emergency buses are located in the control room. In addition, controls required to maintain the unit in a hot shutdown condition are provided at locations outgide the control room for the contingency that the control room 'is not accessible as discussed in Section 7.4.1. Emergency bus circuit breakers have the capability of being manually operated at the switchgear.
The manual controls for the emergency bus circuit breakers located at the switchgear consist of mechanical trip-close lever or push buttons. These controls are located inside the breaker cubicle on the breaker itself, and the cubicle door would have to be opened to gain access to them.
The breakers could not be inadvertently tripped because access to 2 the emergency switchgear area is limited and under administrative control and the switchgear cubicle doors are closed at all times.
It would take a deliberate conscious effort to trip the breaker using these controls. The trip condition of the breaker is alarmed in the control room. This applies to the diesel generator breakers, also.
Instrumentation is provided in the control room to indicate 4,160 V bus loads and voltage and to alarm any abnormality.
In general, motors 300 hp and up are operated at 4,160 V or 6,900 V and motors up to 300 hp are fed from 480 V switchoesr or 8.3-3 2344 l
WUP Amendment 18 PSAR 6/79 motor control centers. All safety-related motors are designed for direct across-the-line starting. 480 V Systems Power for 480 V nonsafety-related auxiliaries will be supplied from five nonsafety-related single ended unit substations con-sisting of dry type transformers and associated metal-clad switchgear. The unit substations are supplied from the 4,160 V nonemergency buses as shown on Fig. 8.3.1-1. Power for safety-related auxiliaries is supplied from four, single ended unit substations consisting of dry type transformers and associated metal-clad switchgear. Each emergency 4,160 V bus supplies two safety-related unit substations sized to meet safety-related load requirements. In no case will unit substations fed from different 4,160 V emergency buses be connected together. Power for motors, approximately 50 hp and smaller, and other small power requirements is, in general, fed from motor control centers (MCC) supplied from the normal or emergency 480 V unit substations. The motor control centers are self-supporting metal-clad structures with combination magnetic, reversing or nonreversing motor starters and molded case air circuit breakers. Motor starters have built-in 480-120 V transformers for control circuit power. Emergency unit substations and emergency MCC are located within They are physically separated 16l (Fig. Seismic Category 8.3.1-5 and I8.3.1-6) structures. so that any single failure which might affect one bus will not jeopardize proper operation of the other bus. Tests and Inspections Preoperational tests and inspections are performed to demonstrate that components are correct and properly mounted; all connections are correct and continuous; components are operational and metering; and protective devices are properly calibrated and adjusted. Following satisfactory checkout of all components of a system, an initial system test is performed. The initial system tests are operational tests conducted to demonstrate that the equipment operates within design limits and that the system is operational and will meet its performance specifications. These tests demonstrate that the safety-related loads can operate on the preferred power supply; the loss of preferred power supply can be detected; the standby power supply can be started and can accept design load within the design basis time; and that the standby power supply is independent of the preferred power supply. 234'4 104 8.3-4
WUP Amendment 18 PSAR 6/79 Periodic tests are directed at detecting the deterioration of the system toward an unacceptable condition and will demonstrate that components which are not exercised during normal operation are operable. The 4,160 V and 480 V drawout circuit breakers and associated devices may be tested while individual equipment is not in service. The circuit breakers may be placed in the " testa position and tested functionally. Protective relays and trip g devices are tested under a simulated overload or fault condition and their calibration is verified. The breaker opening and closing may also be exercised. Availability of breaker control power is indicated by breaker indicating lights. The inservice periodic testing requirements of the safety-related loads are defined in the individual system discussions in Chapters 5, 6, 7, 9, and 10. In general, these requirements are 2344 105 8.3-4a
NUP Amendment 17 PSAR 2/79 8.3.1.4.3 Equipment Consideration Design features of the major components of the Class IE system to ensure conformance with IEEE-308-1971 are described below. This portion of the discussion excludes the criteria and basis for the installation of electrical cable for the systems. The safety-related portions 'f o the a-c station service system are divided into two load groups; the safety-related actions of each load group are independent of the safety actions provided by its redundant counterpart. Two Class IE a-c power system, each consisting of a diesel generator, a 4,160 V switchgear, 480 V unit substations, and motor control centers are furnished to supply power to the safety-related loads. The redundant components of the Class IE power systems are located in separate rooms or are separated by barriers (Fig. 8.3.1-5 and 8.3.1-6) . l16 These areas are protected from the maximum probable flood as discussed in Section 3.4.4. One centrifugal charging pump and one reactor plant component cooling water pump may be connected to either 4,160 V emergency bus manually with the use of a key interlock system. A manual transfer switch (Fig. 8.3.1-1) , equipped with a key interlock, is provided for each pump breaker in each 4,160 V emergency switchgear. A key is required to operate the transfer switch and also to permit either the train A or train B breaker to be racked into the operating position. The key for the charging pump equipment is not interchangeable with the key for the reactor plant component cooling pump equipment. This design prevents connecting the redundant emergency 4,160 V buses together and satisfies the independence requirements of Regulatory Guide 1.6. This equipment is not subject to common mode failure through failure of the ventilation system. The two diesels have independent ventilation systems fed from the 480 V emergency motor control center located in the adjacent emergency switchgear room. The ventilation system in the switchgear room is not subject to a single failure which could degrade the environment beyond the point to which the equipment is qualified as discussed in Section 9.4. The emergency switchgear and diesel generators are 1qcated in fire protected areas. The equipment is not subject to failure due to operation of the fire protection system since the fire protection system discharge nozzles do not directly impinge on the equipment. The fire protection system is further discussed and analyzed in Section 9.5.1. 2344 106 8.3-21
WUP Amendment 18 PSAR 6/79 8.3.1.4.4 Cable Considerations I The criteria and basis for the installation of safety-related electrical cables for the Class IE cable system are described below. Cable Splicing Cable splicing will not be permitted in trays. If splicing becomes necessary it will be done in an enclosed metal box. Cable Derating and Cable Tray Fill Cables are derated to compensate for ambient temperatures and for adjacent power cables. Power cables are sized and derated on the basis of Power Cable Ampacity, published by the Insulated Power Cable Engineers Association (IPCEA Publication P-4 6-4 2 6) . Six-thousand nine hundred volt power, 4,160 V power, and large 480 V power cables, when installed in cable trays, are arranged in a single layer with maintained spacing. Control cables 60 not occupy the same t ray as 6,900 V cables, 4,160 V cables, and 480 V large power cables. Small power cables, approximately No. 4 AWG and smaller, are sized in accordance with IPCEA Publication No. P-54-440. These derated cables, and cables for intermittent duty (e.g. , valve operators) or 120 V control cables, are not restricted to one layer and may occupy the same tray. Instrumentation, communication, and low voltage control cables may occupy the same raceway and normally are separated from small power cables and 120 V control cables. Cable tray fill is limited to 50 percent of the available cross-sectional area of the cable tray, for the trays in which maintained spacing is not required. Safety-related cable trays are not loaded above their side rails. Cable Routing in Congested Areas and Areas of Hostile Environment The safety-related cables for redundant systems have isolation and/or separation to assure that no single credible event will prevent operation of the required number of redundant ESF, surveillance devices, or protection system devices. or protection system devices. Redundant safety-related calbes are run in separate raceways. Separation distances between these redundant raceways are described in subsequent paragraphs of this section. Cable trays for redundant safety-related systems are not routed through an area where combustible material is present, unless it is unavoidable. Where such routing is unavoidable, only one system of redundant safety-related cables is allowed in the area. 16l Cables in the containment which are required to function during and after a DBA are type tested for the DBA environment of temperature, pressure, humidity, chemical spray, and radiation (Section 3.11) . Cable insulation and jacket materials are 8.3-22
,. 2344 107
wuP Amendment 18 PSAR 6/79 selected to operated in the environments of normal operation or that of the post-accident period as required. 93 Electrical Penetrations There are 82 penetrations. Four 20-inch penetrations are for the reactor coolant pump motor leads. Thirty-six 12-inch penetrations are for nonsafety-related power, control, and instrumentation cables. Twenty-one 12-inch penetrations are for safety-related power, control, and instrumentation cables associated with Train A and Channels I and II. The remaining twenty-one 12-inch penetrations are for safety-related power, control, and instrumentation cables associated with Train B and Channels III and IV. Penetra^. ion areas, both inside and outside the containment, are reserved for electrical cables and their supports. No piping having a potential for damaging electrical cables is installed in penetration areas. Redundant penetrations do not occupy the same group. The redundant penetrations are separated from each other 16 as well as from the nonsafety penet. rations by fire barriers. The anticipated physical arrangement and assignments of circuits are shown on Fig. 8.3.1-3. Electrical penetrations will conform to the provisions of IEEE 317-1976 (Ref . 4) . A plan of the penetration area is shown on Fig. 8.3.1-4. In lieu of Type B testing, each electrical penetration is provided with a permanently installed leakage surveillance system which pressurizes the penetration test chamber to a pressure not less than Pa and monitors the penetration for leakage (10CER50 Appendix J, paragraphs IIIB.1(c) and IIIB.3 (b)) . If the pressure maintained is less than P , Type B testing will a be performed every other refueling shutdown or at an interval not greater than 3 years (10CFR50 Appendix J, paragraph IIID.2) . The electrical penetations are designed to withstand a single l18 failure of their overcurrent protection as required by IEEE 317-1976 and Regulatory Guide 1.63. The fault current resulting from an overcurrent protection failure will not cause 16 the electrical penetrations to fail mechanically; thus, there will be no loss of containment integrity. I Where it is not possible to provide electrical penetrations l18 capable of withstanding severe overcurrents, the normal 16 overcurrent protection schemes will be modified to prevent severe overcurrents through the use of one of the following fast acting l18 backup systems. e
. 2344 108 8.3-23
WUP Amendment 18 PSAR 6/79 6,900 Volt Penetrations The 6,900 V penetrations are protected by air circuit breakers located in the medium voltage switchgear buses J10, J20, and J40. These circuit breakers are controlled by relays which detect both overload and fault conditions. Breaker failure relays are provided which will trip the appropriate 6,900 V bus supply breaker in the event a feeder breaker fails to trip. The ability of the penetration to withstand fault current is a significant factor in choosing the proper time delay for the breaker failure relays. 480 Volt Penetrations The 480 V penetrations which are connected to Load Center Unit Substations (LCUS) are protected by air circuit breakers. These breakers are controlled by static trip units which detect both overload and fault conditions. Breaker failure relays similar to those used on the 6,900 V system are provided to trip the LCUS 16 supply breaker should the appropriate feeder breaker fail to trip. For the cases in which 480 V penetrations connected to motor control centers (MCC) cannot withstand the failure of a molded case circuit breaker, the following is done for:
- 1. Class IE MCC
- a. The load center breaker feeding the MCC is coordinated to trip if the molded case circuit breaker fails to trip (the penetratico is sized to withstand the overlaod condition until the backup device trips) ; or
- b. Fuses are connected in series with the molded case 18l circuit breaker at the MCC load terminals.
- 2. Non Class IE MCC 16 Puses are connected in series with the breaker at the MCC load terminals. The fuses are coordinated with the molded case breakers to prevent unnecessary fuse blowing while maintaining the integrity of the penetrations for breaker failure.
Where fuses are used as backup overcurrent protection, they will 18 be included in the periodic test program described in Section 8.3.1.1.1. Sharing of Cable Trays In general, nonsafety-related cables do not share the same cable trays and safety-related cables. In some instances, however, it may be necessary that cables for nonsafety-related circuits be 8.3-24 2344 109
WUP Amendment 18 PSAR 6/79 run in the same tray with safety-related cables. Nonsafety-related cables have the same quality and installation design criteria as the safety-related cables and therefore do not compromise the protective function cabling. Where a nonsafety-related cable is routed in a cable tray with cables of one redundant system, that cable may not be installed in a tray containing safety-related cables of a mutually redundant system and it will not be installed in a tray carrying nonsafety-related cables without first going through an isolation device. A i nonsafety-related raceway system is furnished. Nonsafety-related cables will be installed in this raceway system or as described above. Fire Detection and Protection Fire detection and protection systems, either automatically or manually initiated, are provided in those areas required to preserve the intecrity of circuits for redundant safety-related services. The areas and systems are as described in Section 9.5.1. Fire protection equipment is provid6d in the following areas: Emergency Switchgear Room Cable Tray Spreading Room Reactor Building Cable Vault and Tunnel Diesel Generator Rooms The fire hazard to safety-related cables is reduced by the following provisions. Control and instrument cables installed in the cable tray or in areas common to safety-related cables have an overall flame retardant jacket and flame retardant, nonwicking fillers. Power cables installed in areas common to safety-related cables have either a flame retardant jacket or are installed in ducts or conduit. Cable and Cable Tray Marking Emergency power system components such as cables, trays, and raceways have a system of unique colors to identify safety-related systems. These unique color markers are readily visible to the operators or maintenance craftsmen so that the s af ety-related cable, trays, or raceways can be recognized. Cables which are in safety-related systems are identified by permanent cable identification markers at each end of the cables. These markers contain the alphanumeric identification of the cable and its color code, thereby indicating the train or channel involved. Additional markers may be attached to the cable at intervals so that routing can be verified during construction. 8.3-24a
WUP Amendment 18 PSAR 6/79 PRELIMDEARY SAFETY ANALYSIS REPORT
];.IST OF EFFECTIVE PAGES Chapter 9 Page, Table (T) , or Piqure IF)
Revision Page, Table (T) , or Revision NJmber Fiqure (F) Number 9-1 and 9-11 15 9.4-29 - 9-111 through 9-vi - 9.4-30 14 9-vii 15 9-vili - 9.5-1 through 9.5-26 14 9-ix and 9-x 17 9.5-27 through 9.5-30 16 9-xi and 9-xii 14 9.5-31 through 9.5-38 2 9-xiia 16 9.5-39 and 9.5-40 6 9-xiib 14 9.5-41 4 9-xiii 15 9.5-42 6 9-xiv through 9-xvi 8 9.5-42a 3 9-xvii 15 9.5-43 2 9-xviii 8 9.5-44 and 9.5-44a 3 9-xix 14 9.5-45 3 9-xx 16 9.5-46 through 9.5-48a 15 9.5-49 through 9.5-51 2 9.1-1 through 9.1-28 15 9.2-1 through 9.2-18 15 T.9.1-1 15 9.2-19 through 9.2-22 - T.9.2.1-1 15 9.2-23 and 9.2-24 5 T.9.2.1-2 - 9.2-25 2 T.9.2.1-3 1 9.2-26 and 9.2-26a 14 T.9.2.2.1-1 (2 pages) 15 9.2-27 6 T.9.2.2.1-2 (page 1) 1 9.2-28 through 9.2-28d 9 (page 2) - 9.2-29 1 T.9.2.5-1 4 9.2-30 2 T.9.3.2-1 16 T.9.3.4-1 - 9.3-1 3 T.9.3.4-2 (7 pages) - 9.3-2 - T.9.3.6-1 through T.9.3.6-7 - 9.3-3 2 T.9.5.4-1 1 9.3-4 and 9.3-4a 3 T.9.5.8-1 15 9.3-5 1 T.9.5.8-2 15 9.3-6 and 9.3-6a 3 9.3-7 and 9.3-8 - F.9.1.1-1 and F.9.1.1-2 - 9.3-9 16 F.9.1.2-1 15 9.3-10 through 9.3-12 - F.9.1.3-1 15 9.3-13 and 9.3-14 15 F.9.1.3-2 and F.9.1.3-3 (deleted) 15 9.3-15 through 9.3-16a 1 F.9.1.4-1 through F.9.1.4-3 - 9.3-17 through 9.3-46 - F. 9.1. 4 -4 9 9.3-47 2 F.9.1. 4-5 and F.9.1.4-6 - 9.3-48 through 9.3-51 - F.9.2.1-1 through F.9.2.1-3 15 9.3-52 9 F.9.2.2.1-1 through 9.3-53 and 9.3-54 15 F.9.2.2.1-3 15 F.9.2.2.2-1 15 9.4-1 through 9.4-6a 17 F.9.2.2.2-2 - 9.4-7 2 F.9.2.2.2-3 (deleted) 15 9,4-8 and 9.4-9 - F.9.2.2.3-1 - 9.4-10 3 F.9.2.2.5-1 - 9.4-11 4 F.9.2.4-1 2 9.4-12 3 F.9.2.5-1 9 9.4-13 through 9.4-14a 18 F.9.2.5-2 2 9.4-15 through 9.4-16a 2 F.9.2.5-3 through F.9.2.5-7 4 9.4-17 0 F.9.2.5-8 6 9.4-18 - F.9.2.6-1 0 9.4-19 through 9.4-26a 17 F.9.3.1-1 and F.9.3.1-2 - 9.4-27 2 F.9.3.2-1 and F.9.3.2-2 15 9.4-28 and 9.4-28a 6 F.9.3.3-1 and F.9.3.3-2 - EP.9-1 2344 111
IfUP Aasendument 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF trrz.sT VE P P * (COfff*D) Chapter 9 Page, Table (T) , or Revision Piqure (F) Number F.9.3.3-3 2 F.9.3.3-4 15 F.9.3.4-1 4 F.9.3.4-2 through F.9.3.4-9 - F.9.3.6-1 15 F.9.3.6-2 and F.9.3.6-3 (deleted) 9 F.9.4.1-1 17 F.9.4.1-2 - F.9.4.1-3 (deleted) 14 F.9.4.2-1 and F.9.4.2-2 13 F.9.4.4-6 13 F.9.4.5-1 18 F.9.4.6-1 13 F.9.4.7-1 17 F.9.4.7-2 17 F.9.4.7-3 7 F.9.4.7-4 3 F.9.4.8-1 2 F.9.4.9-1 - F.9.4.10-1 (deleted) 14 F.9.5.1-1 15 F.9.5.1-2 14 P.9.5.1-3 - F.9.5.1-4 15 F.9.5.1-5 16 F.9.5.4-1 - 97//7 LJ 119 I IL F.9.5.5-1 6 F.9.5.6-1 6 F.9.5.7-1 2 P.9.5.8-1 2 F.9.5.9-1 - O EP.9-2
WUP Amendment 18 PSAR 6/79 Hot water heating maintains the fuel building at 700F, coincident with an outside temperature of -220F. l18 The fuel building emergency filtration (Section 6.5.2) exhausts the pool surface area following a loss of power. Unit coolers are located in selected areas throughout the fuel building to remove equipment heat loads and to minimize the quantity of outside air required for ventilation. Exhaust air, prior to release to the atmosphere, passes through a high efficiency particulate air (HEPA) filter, which removes l18 airborne radioactive particulates. The HEPA filter has an efficiency of approximately 99.97 percent when filtering particulates 0.3 microns or larger. The fuel building emergency filtration system is described in Section 6.5.2. 9.4.5.2 System Description The fuel building ventilation system is shown on Figure 9.4.5-1. The supply portion of the fuel building ventilation system consists of two 100 percent capacity air handling units. Each air handling unit includes a roughing filter, glycol / water heating coil, chilled water cooling coil, and centrifugal fan. The air handling units are each rated at approximately 8,000 cfm, and deliver outside air via ductwork to areas of the building. Approximately 6,000 cfm are utilized for the spent fuel pool surface air flow requirements during and immediately following refueling. An exhaust rate of approximately 7,000 ctm is maintained from the pool surface areas during normal operation to l18 ensure removal of contaminated air. The supply air for the spent fuel pool surface is distributed via strip diffusers located just above the pool surface. The diffusers extend the entire length of one side of the pool. The exhaust air from the pool surface is drawn into registers attached to ductwork at the opposite side of the pool. The air supply for the fuel pool can be manually diverted by closing a damper in the fuel pool supply ductwork and opening a damper in the bypass ductwork supplying air to- other areas of the building. The exhaust portion of the fuel building normal ventilation consists of two 100 percent capacity air handling units. Each air handling unit is rated at approximately 10,000 cim, and includes a roughing filter, HEPA filter, and centrifugal fan. Unit coolers recirculate and cool the surrounding air. Each unit cooler consists of a chilled water cooling coil and a fan. 2344 1;3 9.4-13
WUP Amendment 18 PSAR 6/79 During fuel handling operations, the normal exhaust bypasses the HEPA exhaust filters and is manually diverted through the fuel h building emergency filters (Section 6.5.2) by stopping the HEPA 18l exhaust fan and starting the emergency filter fan, and opening and closing the necessary dampers. 9.4.5.3 Safety Evaluation The fuel building ventilation is designed to maintain a controlled environment for personnel and equipment and to minimize the release of radioactive airborne material to the atmosphere. It is not required to operate during an accident and is not safety related. Dur%ng fuel handling operations, the fuel H building emergency filtration ventilation system (Section 6.5.2) is in operation in order to maintain a minimum negative pressure of 0.25 in. water gauge. 9.4.5.4 Test and Inspection Requirements The HEPA filters are tested for efficiency utilizing the DOP method as described in Section 6.2.3.1. After installation, the filters are checked for leakage. The fuel building ventilation system is inspected after initial installation, and each supply and exhaust fan is tested for air balance. All portions of the system are tested for the total air quantity being delivered or exhausted. The fuel building ventilation system is continually in operation. Routine surveillance and preventive maintenance eliminate the need for periodic testing. 9.4.5.5 Instrumentation Applications The fuel building ventilation fans are locally controlled and are interlocked to ensure operation of an exhaust fan if a supply fan is in operation. Only one supply fan and one exhaust fan are in operation at any time. A flow switch located in each fan discharge duct automatically starts the other fan and annunciates on the auxiliary building control panel on low discharge flow. The fuel building vent monitor (Section 11.4) provides control room indication and annunciation of fuel building ventilation g exhaust radioactivity levels. In the event of high radioactivity alarm when fuel is not being handled, the fuel building emergency ventilation is manually placed in operation. 9.4.6 Service Building Heating, Ventilation, and Air Conditioning (HVAC) 9.4.6.1 Design Bases The service building HVAC system is not safety related. The system is designed to provide a controlled ents_ronment for 9.4-14
WUP Amendment 18 PSAR 6/79 personnel and equipment and to minimize the potential release of radioactive material to the environment. The required flow rates for the various portions of the service building HVAC are based on conservative estimates of the following: 2344 115 9.4-14a
4 d
,__________.7_
rA , l I T , l l l l i s FS I l l I I i1 l l 1 ;g (EACH) 8000 CFM L-_ _ _, I I l NORMAL VENTILATION l
}- - l A l SUPPLY (NNS)
I l- >lJ t fr--------H-FC FC ll l OUTSIDE AIR il! t l S C-3 +*. NNS O CLYCOL HEATING C0ll 2000 CFM C CHILLED WATER C0ll CONTINUQUS SUPPLY TO AREAS OF FUEL BLOG. h i r 6000 CFM TN SUPPLY TO ARE (WHEN N T
/
N FS TO FUEL
,t ___l [--
-----t ;
(EACH) 10,000 CFM BUILDING - N/ ' l l NORMAL VENTILATION I " 8} 6.5.2-1 F------h-I u--__, FS l
- s Z M [w$ @
HEPA FILTERS 2344 116 'k THIS SYSTEM IS SAFETY CLASS 3(SC-3) EXCEPT AS OTHERWISE NOTED. 9
\
j .
i 1 I I UNIT COOLER UNIT HEATER (TYPICAL) (TYPICAL)
.- 7000 CFM (NORMAL) 8000 CFM (REFUELING)
EXHAUST FROM 6000 CFM FUEL POOL SUPPLY TO FUEL POOL (WHEN REFUELING) V> ' _ [ h b r 4 - SC-3 yy5-> > u m>> >>mn>> FUEL POOL 1 f LING) NNS SC-3
-E Z' v ,, ,,
IO FUEL BUILDING 7000 CFfA 3000 CFM VENT SYSTEM E1HAUST FROM AREAS CONTINUOUS EXHAUST 7 EMERGENCY FILTRAfl0N OF FUEL BUIL0lNC FROM AREAS OF (WHEN NOT REFUEllNC) FUEL BUILDING 2)kk i TRAINS FIG.6.5.2-1 FIG. 9.4.5-1 FUEL BUILDING NORMAL VENTILATION WISCONSIN UTILITIES PROJECT PREllMINARY SAFETY ANALYSIS REPORT AMEN 0 MENT 18
.Es
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 13 Page, Table (T) , or Revision Fiqure (F) Number 13-1 13 13-ii - 13-111 8 13-iv 17 13-v 14 13-vi - 13-vii 18 13.1-1 through 13.1-4 13 13.1-5 through 13.1-12 - 13.1-13 and 13.1-14 0 13.1-15 and 13.1-16 18 13.1-17 - 13.1-18 18 13.1-19 and 13.1-20 - 13.1-21 through 13.1-23 13 13.2-1 13 13.2-2 and 13.2-3 - 13.2-4 through 13.2-7 17 13.3-1 and 13.3-2 - 13.3-3 and 13.3-4 16 13.3-5 through 13.3-8 - 13.4-1 and 13.4-2 - 13.5-1 and 13.5-2 - 13.6-1 - 13.7-1 14 T.13.2-1 (2 pages) 10 F.13.1-1 13 F.13.1-2 0 F.13.1-3 18 F.13.1-4 - 2344 118 EP.13-1
WUP Amendment 18 PSAR 6/79 CHAPTER 13 LIST OF FIGURES Figure Title 13.1-1 Wisconsin Electric Power Company Corporate Organization 13.1-2 Wisconsin Electric Power Company Nuclear Project Office Organizatien Chart 13.1-3 WUP Plant Organization Chart for One Unit 13.1-4 Point Beach Nuclear Plant Organization Chart 2344 II9 13-vii
WUP Amendment 18 PSAR 6/79
- a. Responsible for the preparation of technical documents for, and coordination and follow with suppliers of, field change kits Gesigned to improve operational and safety features of Navy Nuclear Power Plants.
- b. Responsible for the preparation of specifications and Technical Ordering Documents for static power supplies to replace existing rotating machinery.
- 2. 16 years -- Delco Electronics Division - General Motors Corporation.
- a. Lead man of team that perform 7d accuracy analysis and prepared test reports for electro-mechanical servomechanism systems.
- b. Lead man of team that assembled and tested first production unit of inertial guidance system.
- c. Responsible for design, analysis, and testing of complex servomechanisms used in inertial guidance systems.
B. Educational Background
- 1. Nuclear power training classes
- 2. Analog and digital computer training classes
- 3. M.S. in Electrical Engineering, 1962.
- 4. B.S. in Electrical Engineering, 1956.
13.1.2 Operating Organization 44 120 13.1.2.1 Plant Organization Fig. 13.1-3 presents the organization chart for a one-unit l g i station and indicates the normal plant staff. This organization will be in effect prior to issuance of the Unit 1 operating i license. This organization has been developed by utilizing the experience obtained in the previous staffing of the Yankee-Rowe, Connecticut Yankee, and Point Beach Nuclear Plants. It reflects due consideration to the requirements of startup testing and operation of a two-unit nuclear plant. Fig. 13.1-4 describes the organization of the Point Beach Nuclear Plant which has been in successful operation since October 1970. A comparison between the two figures indicates that the new plant will have an l18 increased staff. The specific staffing requirements in regard to licensed operators and operating crews is defined in the Technical Specifications, Chapter 16. 13.1-15
WUP Amendment 18 PSAR 6/79 Haven plant will be staffed with key personnel who meet or 18l The exceed the qualification requirements of ANSI N18.1 - 1977, lf
" Selection and Training of Nuclear Power Plant Personnel."
13.1.2.2 Personnel Functions, Responsibilities and Authorities 13.1.2.2-1 General Superintendent The General superintendent is responsible for the safe and effi-cient operation of the plant. He is the senior member of company management at the plant site and is totally responsible for all aspects of plant operation. He has the authority to take whatever action is necessary to insure that all aspects of plant activities are conducted within the limits of normal management practices and all legal requirements as defined by the appropriate federal and state licenses and permits. He reports directly to the Executive Vice President in charge of power generation. 13.1.2.2.2 Operations Superintendent The Operations Superintendent reports directly to the General Superintendent and has the responsibility and authority for insuring the safe and efficient operation of the operating equipment in the plant including the nuclear units, steam plant, and all supporting systems in accordance with the applicable plant licenses, operating procedures, emergency operating procedures, technical specifications, and safety rules. He also serves as a member of the Duty and Call Superintendent Staff and a r. ember of the General Superintendent *s Supervisory Staff. He must be qualified for cold Senior Reactor Operator Licensing. 13.1.2.2.3 Assistant to the General Superintendent The Assistant to the General Superintendent reports directly to the General Superintendent and is responsible for assisting the General Superintendent in the administrative aspects of plant operation that relate to the technical areas, such as:
- 1. Overall Technical Specification surveillance program.
- 2. Progress and Technical Reports.
- 3. Coordination of all technical records and documentation.
- 4. Overall schedule planning.
- 5. Coordination and publication of significant operating events and abnormal occurrences.
6 *. Onsite Quality Assurance coordination. O 2344 121 13.1-16
WDP PSAR
- 7. Recording secretary for General Superintendent *s Super-visory Staff.
- 8. Plant training programs.
This position will normally be filled by a licensed SRO Super-visor who will also serve as a Duty and Call Superintendent and as a member of the General Superintendent *s Supervisory Staff. 13.1.2.2.4 Maintenance Superintendent The Maintenance Superintendent is responsible to the General Superintendent for all electrical and mechanical maintenance activities throughout the plant, with the exception of instrumentation and control activities. He is assisted by two general foremen, one of whom specializes in electrical and the other mechanical maintenance. In addition, he will have technical assistants assigned as necessary. He is a member of the General Superintendent *s Supervisory Staff and may be a Duty and Call Superintendent if so qualified. 13.1.2.2.5 Instrumentation and Control Engineer The Instrumentation and Control Engineer reports directly to the General Superintendent and is responsible for all instrumentation and control activities throughout the plant. He is a member of the General Superintendent *s Supervisory Staff and he may hold an SRO license and be a Duty and Call Superintendent, if so qualified. 13.1.2.2.6 Reactor Engineer The Reactor Engineer reports directly to the General Superin-tendent, and is responsible for all Reactor Engineering Activities throughout the plant. Such activities include:
- 1. Maintaining special nuclear material accountability records and procedures,
- 2. Preparation of tests and procedures involved in the receipt, inspection, movement, storage, and shipment of special nuclear materials including fuel assemblies,
- 3. Preparation of plant performance test procedures, including such items as core loading, physics tests, seconda::y plant performance tests, etc., and
- 4. Computer programming.
He is a member of the General Superintendent *s Supervisory Staff and, .if qualified, he may serve as a Duty and Call Superintendent. 13.1-17
WUP Amendment 18 PSAR 6/79 13.1.2.2.7 Radiochemical Engineer The Radiochemical Engineer, as head of the Chemistry and Health Physics Group, directs and has overall responsibility for all chemistry, radio-chemistry, radiation protection, and personnel safety aspects of the plant. His responsibilities include such areas as:
- 1. Site environmental monitoring programs.
- 2. Receipt and shipment of radioactive materials and wastes other than nuclear fuel.
- 3. Nonradioactive wastes treatment and disposal.
- 4. Preparation and implementation of plant radioactive and nonradioactive emergency plans, including training and drills.
- 5. All chemistry and radiation procedures and equipment.
- 6. Maintenance of proper chemistry and radiochemistry con-trol of all plant systems.
- 7. Development and maintenance of records of all radio-active and nonradioactive releases from the plant and personnel exposure records.
He is a member of the General Superintendent's Supervisory Staff and reports directly to the General Superintendent. 13.1.2.2.8 Health Physicist The Health Physicist reports directly to the Radiochemical Engineer but has the option of direct communication with, and reporting to, the General Superintendent on matters of radiation protection. He exercises supervisory control over all radiological and industrial safety aspects of plant operation. He has the authority and responsibility to stop any work or activity where he believes employee health and safety is being He is a member of the General Superintendent *s jeopardized. Supervisory Staff. 13.1.2.2.9 Office Supervisor The Office Supervisor reports directly to the General Superin-tendent and is responsible for the general administrative 18l functions of the plant. He is a member of the General Superintendent *s Supervisory Staff. O 13.1-18 23 /j ej l}}
GENERAL ASSISTANT TO (1 ) (1) GENERAL SECRETARY SU PERI NTEN DENT (2-5) TRAINING TECHNICAL COO R DIN ATOR ASSISTANTS I l I i i OPERATIO N S I M AI NTEN ANCE INST. S CONTROL CHEMISTRY S REACTOR ENG. ADMINISTR ATION SECURITY (1) l (1) ! (t) HEALTH PHYSICS (9) l (g ) l (g) l (g) SUPER N Et DENT RADIOCMCAL SECURITY SUPERINTENDENT CONTROL ENGINEER ENGINEER REACTOR ENGINEER OFFICE SUPER. SRO/CL SUPER. (1) (2) (t) (1) (2) (2) TECHNICAL TECHNICAL TECHNICAL
^S3'$^" HEALTH TECHNICAL CLERK CONTRACT SRO/CL ASSISTANTS ASSISTANT PHYSICIST AGSISTANTS TYPIST GUARDS (1) (2) (1) (3) (2)
QU A LITY ASSURANCE GENERAL i S C FOREMAN TECHNICAL (j) STOCKMAN ASSISTANT FOREMAN ASSISTANTS HP FOREMAN (6) (8) (3) (3) SU V SOR ECH ANIC- RADIOCHEMICAL (2) I S C TECHNICIAN SRO/CL ELECTRICIAN TECHNICIAN RA DI ATION CONTROL (5) (6) (2) OPERATING SUPERVISOR REPAI R M AN IEC REPAIRMAN (3) SRO/CL R A DI ATIO N TOTAL NORM ALLY CO" L gELPER ASSIGNED PERSONNEL (5) (4) :97-100
.) CONTROL REPAIRMA N -a OPERATOR 1. SRO-SENIOR REACTOR OPERATOR,10 CFR 55 SRO/HL HEL PE R
- 2. PO- RE ACTOR OPERATOR,10 CFR 55 z.:. (10) (3) 3. CL-COLD LICENSED FI G. 13.1 - 3 AUXILI ARY UTILITY M AN 4 H L -HOT LICENSED V/UP PLANT ORGANIZATION CHART
~ OPERATORS 'O 5. HEALTH PHYSICIST HAS OPTIONAL FOR ONE UNIT REPORTING ROUTE TO GENER AL
- 5) "^ ' AL WISCONSIN UTILITIES PROJECT EQUIPMENT H A PHYS SM RS PRELIMINARY SAFETY ANALYSIS REPORT OPERATORS 6. AUXILI ARY OPER ATORS AND EQUIPMENT CPERATORS PROVIDE SCME RADIATION PROTECTION FUNCTIONS AMENDMENT 18
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIS? OF EFFECTIVE PAGES Chapter 15 Page, Table (T) , or Revision Page, Table (T) , or Revision Fiqure (P) Number Fiqure (F) Number 15-1 through 15-111 8 15.4-26 6 15-iv 15 15.4-27 15 15-v and 15-vi 10 15.4-28 and 15.4-28a 7 15-vii through 15-viiia 18 15.4-29 - 15-ix through 15 xii 10 15.4-30 and 15.4-30a 1 15-xiii through 15-xvii j 17 15.4-30b 7 15.4-31 through 15.4-32a 7 15.1-1 through 15.1-6 - 15.4-33 through 15.4-35 - 15.1-7 and 15.1-8 4 15.4-36 15 15.1-9 through 15.1-14 - 15.4-37 1 15.1-15 and 15.1-16 1 T.15.1-1 through T.15.1-3 - 15.2-1 - T.15.1-4 (deleted) 4 15.2-2 and 15.2-2a 1 T.15.1-5 - 15.2-3 through 15.2-13 - T.15.1-6 (page 1) - 15.2-14 and 15.2-14a 7 (pages 2 and 3) 1 15.2-15 7 T.15.2-1 (7 pages) - 15.2-16 - T.15.2-2 and T.15.2-3 - 15.2-17 9 T.15.2-4 (2 pages) 1 15.2-18 and 15.2-18a 3 T.15.3-1 1 15.2-19 - T.15.3.1-1 and T.15.3.1-2 10 15.2-20 and 15.2-20a 1 T.15.3.4-1 - 15.2-21 through 15.2-24 - T.15.3.5-1 - 15.2-25 through 15.2-26a 3 T.15.4-1 17 15.2-27 through 15.2-42 - T.15.4.1-1 17 15.2-43 13 T15.4.1-1A 18 15.2-44 15 T15.4.1-2 17 15.2-44a 13 T15.4.1-2A 18 15.2-44b 5 T15.4.1-3 17 15.2-45 and 15.2-46 - T15.4.1-3A 18 15.2-47 13 T 15. 4.1-4 17 T15. 4 .1-4 A 18 15.3-1 through 15.3-2d 10 T15.4.1-5 17 15.3-3 - T15.4.1-5A 18 15.3-4 and 15.3-5 7 T.15.4.1-6 through T.15.4.1-8 10 15.3-6 3 T.15.4.1-9 and T.15.4.1-10 17 15.3-7 15 T.15.4.2-1 - 15.3-8 and 15.3-8a 10 T.15.4.2-2 7 15.3-9 and 15.3-10 7 T.15.4.2-3 3 T.15.4.3-1 - 15.4-1 and 15.4-2 10 T.15.4.4-1 and T.15.4.4-2 - 15.4-3 through 15.4-4 j 17 T.15.4.5-1 - 15.4-5 through 15.4-6a 4 T.15.4.5-2 4 15.4-7 - T.15.4.6-1 and T.15.4.6-2 - 15.4-8 15 15.4-9 and 15.4-10 7 F.15.1-1 through F.15.1-6 - 15.4-10a 15 F.15.2-1 through F.15.2-45 - 15.4-11 through 15.4-16 7 F.15.3.1-1 through F. 15.3.1-10b 10 15.4-17 - F.15.3.1-11a and F.15.3.1-11b 10 15.4-18 15 F.15.3.1-12a and F.15.3.1-12b 10 15.4 -18a 7 F.15.3.3-1 through F.15.3.3-5 - 15.4-19 - F.15.3.4-1 through F.15.3.4-6 - 15.4-20 and 15.4-20a 1 F.15.3.5-1 and F.15.3.5-2 9 15.4-21 and 15.4-22 15 (deleted) 15.4-23 and 15.4-24 7 F.15.4.1-1a through F.15.4.1-1d 17 15.4-25 15 F.15.4.1-2a through F.15.4.1-2d 17 EP.15-1
)
WUP Amendment 17 PSAR 2/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONT'D) Chapter 15 Page, Table (T) , or Revision Fiqure (F) Number F.15.4.1-3a through F.15.4.1-3d 17 F.15.4.1-4a through F.15.4.1-4d 17 F.15.4.1-Sa through F.15.4.1-5d 17 F.15.4.1-6a through F.15.4.1-6d 17 F.15.4.1-7a through F.15.4.1-7d 17 P.15.4.1-Ba through F.15.4.1-8c 17 F.15.4.1-9a through F.15.4.1-9d 17 F.15.4.1-10a through 17 F.15.4.1-10h F.15.4.1-11a through 17 F.15.4.1-11d F.15.4.1-12a through 17 F.15.4.1-12d F.15.4.1-13a through 17 F.15.4.1-13d F.15.4.1-14a through 17 F.15.4.1-14d F.15.4.1-15a and F.15.4.1-15b 17 F.15.4.1-16a and F.15.4.1-16b 17 F .15. 4 .1- 17 17 F.15.4.2-1 - througn F.15.8i .2-4 F.15.4.2-5 7 F.15.4.2-6 through 9 F.15.4.2-9 (deleted) P.15.4.3-1 - P.15.4.3-2 through 9 F.15.4.3-5 (deleted) F.15.4.4-1 through F.15.4.4-7 - F.15.4.4-8 through F.15.4.4-11 9 (deleted) F.15.4.5-1 and F.15.4.5-2 9 (deleted) F.15.4.6-1 through F.15.4.6-4 - F.15.4.6-5 through F.15.4.6-8 (deleted) 9 2344 126 O EP.15-2
WUP Amendment 18 PSAR 6/79 CHAPTER 15 LIST OF TABLES Table Title 15.1-1 Huclear Steam Supply System Power Ratings 15.1-2 Trip Setpoints and Time Delays to Trip Assumed in Accident Analyses 15.1-3 Determination of Maximum Overpower Trip Point-Power Range Neutron Flux Channel - Based on Nominal Setpoint Considering Inherent Instrumentation Errors 15.1-4 Average Gap Activities 15.1-5 Core Temperature Distribution 15.1-6 Summary of Initial Conditions and Computer Codes Used 15.2-1 Time Sequence of Events for Applicable Condition II Events 15.2-2 Minimum Calculated DNBR for Cases of Rod Cluster Control Assembly Misalignment and Dropped Rod Cluster Control Assembly 15.2-3 Natural Circulation Flow 15.2-4 Equipment Assumed to Function after Condition II Events 15.3-1 Equipment Assumed to Function after Condition III Events 15.3.1-1 Time Sequence of Events 15.3.1-2 Small Break 15.3.4-1 Time Sequence of Events for Condition III Events 15.3.5-1 Maximum Radioisotope Inventory Released from Process
, Gas Charcoal Bed Adsorber and Associated Piping 15.4-1 Equipment Assumed to Function After Condition IV Events 15.4.1-1 Large Break (Typical 3-Loop Containment) 15.4.1-1A Large Break (HUP Containment) 15.4.1-2 Containment Data (Typical 3-Icop Containment) 15-vii
WUP Amendment 18 PSAR 6/79 CHAPTER 15 LIST OF TABLES (CONT'D) Table Title 15.4.1-2A Containment Data (WUP Plant) 15.4.1'-3 Reflood Mass and Energy Releases for Limiting Break DECLG (CD = 0.4) (Typical 3-Loop Containment) 15.4.1-3A Reflood Mass and Energy Releases for Limiting Break DECLG (CD = 0.4) (WUP Containment) 15.4.1-4 Broken Loop Accumulator Flow Rate to Containment for Limiting Case (Typical 3-Loop Containment) 15.4.1-4A Broken Loop Flow Rate to Containment for Limiting Case (WUP Containment) 15.4.1-5 Large Break - Time Sequence of Events (Typical 3-Loop Containment) 15.4.1-5A Large Break Time Sequence of Events (WUP Containment) 15.4.1-6 Radioactivity of the Sump Water 0 Min after a IDCA 15.4.1-7 Radioactivity of the Sump Water 20 Min after a LOCA 15.4.1-8 Radioactivity of the Sump Water 60 Min after a LOCA 15.4.1-9 Fuel Release Rate Multiples (R/Ro) Used to Calculate Reactor Coolant Iodine Concentration Spike after Shutdown 15.4.1-10 Analysis of Containment Purge Concurrent with LOCA Maximum Calculated Coolant Concentrations and Releases 15.4.2-1 Core Parameters Used in Steam Break Analysis 15.4.2-2 Time Sequence of Events for Postulated Feedline Rupture 15.4.2-3 Radioactivity Released from Main Steam Line Rupture Accident 15.4.3-1 Radioactivity Released from a Steam Generator Tube Rupture Accident 15.4.4-1 Summary of Results for Locked Rotor Transients 15-viii
}}/;4 l}g
WUP Amendment 18 PSAR 6/79 CHAPTER 15 LIST OF TABLES (CONT *D) Table Title 15.4.4-2 Radioactivity Released from a Reactor Coolant Pump Locked Rotor Accident 15.4.5-1 Fuel Handling Accident Radioactivity Released to the Fuel Building Atmosphere 15.4.5-2 Fuel Handling Accident Radioactivity Released to the Env.tronment 15.4.6-1 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident 15.4.6-2 Radioactivity Released from Rod Ejection Accident 2344 129 15-viiia
WUP Amendment 18 PSAR 6/79 TABLE 15.4.1-1A LARGE BREAK (WUP CONTAINMENT) l18 DECLG Results ,1C_o = 0. 4) Peak clad temperature (F) 2151 Peak clad location (ft) 7.5 Local Zr/H 2O reaction (%) 6.1 local Zr/HgO location (ft) 7.5 Total Zr/H,0 reaction (%) <0.3 Hot rod burst time (sec) 31.6 Hot rod burst location (ft) 6.0 Calculation License core power plus 2 percent instrument error (MWt) 2830.5 Peak linear power 102 percent (kW/ft) 11.17 Peaking factor (at license rating) 2.05 Accumulator water volume (fts) 1,000 Fuel Region and Cycle Analyzed Cycle Region Unit I 1 All Unit II 1 All 2344 130 1 of 1
WUP Amendment 18 PSAR 6/79 TABLE 15.4.1-2A CONTAINMENT DATA (WUP PLANT) l18 Net Free Volume 2.52 x 106 ft8 Initial Conditions Pressure 14.7 psia Temperature 90 F RWST temperature 45 F Service water temperature 33 F Outside temperature -22 F Spray System Number of pumps operating 2 Maximum spray flow per pump 3,450 gpm Actuation time 69 sec Safeguards Fan Coolers Number of fan coolers operating 4 Fastest post accident initiation of fan coolers 27 sec Structural Heat Sinks Thickness (ft) Area (ftz) 1.0 Concrete 160,000 2.75 Concrete 0.03125 Steel 12,700 10.0 Concrete 0.03125 Steel 21,100 4.5 Concrete 0.03125 Steel 40,500 4.5 Concrete 0.04167 Steel 29,000 2.5 Concrete 0.0065774 153,396 0.024171 125,351 0.05778 56,643 0.086312 11,821 0.116545 26,757 0.2561 1,032 2344 131 1 of 1
WUP Amendment 18 PSAR 6/79 TABLE 15.4.1-3A REFLOOD MASS AND ENERGY RELEASESC13 FOR LIMITING BREAK (DECLG (CD = 0.4)) (WUP CONTAINMENT) l18 Time Mass Flow Energy Flow _(sec) (1bm/sec) (Btu /sec) 37.139 0.0 0.0 37.814 0.000551 0.602 38.164 0.599 7.829x102 41.877 33.61 4.3935x10* 51.118 213.81 1.3265x105 67.218 255.28 1.3794x105 87.718 272.38 1.3548x105 111.418 283.33 1.3144x105 137.918 292.30 1.2684x105 200.418 309.66 1.1699x105 289.118 338.73 1.0829x105 488.618 353.50 0.9834x105 NOTE: (8) Accumulator nitrogen was released between 45.7 and 65.7 sec at a mass flow rate of 192.26 lbm/sec. 2344 132 1 of 1
WUP Amendment 18 PSAR 6/79 TABLE 15.4.1-4A BROKEN LOOP ACCUMULATOR FLOW RATE TO CONTAINMENT FOR LIMITING CASE (WUP CONTAINMENT) l18 Time Mass Rate (13 (sec) _(1bm/sec) 0.0 4600.1 2.0 3777.3 4.0 3289.4 6.0 2953.5 8.0 2702.3 10.0 2502.6 13.0 2267.5 16.0 2085.6 20.0 1905.2 24.0 1769.7 24.46 0.0 NOTE: O )Enthalpy of accumulator water is 59.6 Btu /lbm. 2344 133 1 of 1
WUP Amendment 18 PSAR 6/79 TABLE 15.4.1-5A LARGE BREAK TIME SEQUENCE OF EVENTS (WUP CONTAINMENT) l18 DECLG (C n =0.4) (secF Start 0.0 Reactor trip signal 0.497 Safety injection signal 1.29 Accumulator injection 15.5 End of bypass 24.79 End of blowdown 28.96 Pump injection 26.29 Bottom of core recovery 37.189 Accumulator empty 45.62 2344 134 1 of 1
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Unacter 16 Page, Table (T) , or Revision Fiqure (F) Number 16-i 8 16-11 - 16-111 through 16-vii 8 16-viii 14 16-ix 8 16-x 7 16.1-1 - 16.1-2 2 16.1-3 through 16.1-6 - 16.2-1 through 16.2-5 - 16.2-6 and 16.2-7 18 16.2-8 through 16.2-10 - 16.3-1 through 16.3-52 - 16.4-1 and 16.4-2 3 16.4-3 through 16.4-5 - 16.4-6 14 16.4-7 and 16.4-8 17 16.4-9 through 16.4-16 - 16.4-17 and 16.4-18 9 16.4-19 2 16.4-20 and 16.4-20a 18 16.4-21 6 16.5-1 through 16.5-4 - 16.6-1 through 16.6-9 - 16.6-10 and 16.6-11 14
}f T.16.3-1 -
T.16.3-2 (2 pages) - T.16.3-3 and T.16.3-4 T.16.4-1 (page 1) 3 (pages 2 and 3) 6 T.16.4-2 and T.16.i!-3 T.16.4-4 (pages 1 through 4) - (Page 5) 3 T.16.4-5 T.16.4-6 (2 pages) - T.16.6-1 T.16.6-2 (3 pages) - F.16.2-1 through F.16.2-3 - F.16.3-1 through F.16.3-11 - EP.16-1
WUP Amendment 2 PSAR 1/2/75 measurements and identification of individual radionuclides by gamma analysis such that at least 90 percent of the gross beta and gamma radioactivity in a sample is accounted for by the analysis. Ventilation vents shall be sampled quarterly for tritium vapor. (2) Prior "to containment purging, the containment atmosphere shall be analyzed for tritium. (3) Radiation Monitoring Systems
- a. Area Radiation Monitoring System The area radiation monitoring system shall be checked and calibrated at intervals specified in Table 16.4-1.
- b. Process and Effluent Radiation Monitoring System 2
The process and effluent radiation monitoring system shall be checked and calibrated at intervals specified in Table 16.4-1.
- c. Airborne Radioactivity Monitoring System The airborne radioactivity monitoring system shall be checked and calibrated at intervals specified in Table 16.4-1.
References 2344 136 Title Section
- 1. Air Conditioning, Heating, Cooling, and 9.4 Ventilation Systems
- 2. Main Condenser and Main Condenser 10.4 Evaluation System
- 3. Radioactive Liquid Waste System 11.2
- 4. Radioactive Gaseous Waste System 11.3
- 5. Process and Effluent Radiation 11.4 Monitoring Systems
- 6. Effluent Release Limits 16.3.9
- 7. Area Radiation Monitoring System 12.1.4 2
- 8. Airborne Radioactivity Monitoring System 12.2.4 16.4-19
WUP Amendment 18 PSAR 6/79 16.4.11 Emergency Power System Tests Applicability Applies to periodic testing and surveillance requirements of the emergency power system. Obiective To verify that the emergency power system will respond promptly and properly when required. Specification The following tests and surveillance shall be performed as stated:
- 1. Diesel Generators
- a. A manually initiated test consisting of a start of each diesel generator, followed by manual synchronization with other power sources and assumption of load for at least 30 minutes, shall be conducted monthly.
- b. An operational test shall be conducted during reactor shutdown for major fuel reloading. The operational test shall verify proper operation of the automatic start, load shed, and load sequencing functions for the following simulated conditions:
(1) Loss of offsito power, (2) Safety injection signal, (3) Loss of offsite power together with a safety injection signal. In addition, after the diesel generator has carried its load for a minimum of 5 min, automatic load shedding and load sequencing functions shall be retested by manually interrupting the power to the emergency bus.
- c. Each diesel generator shall be given a thorough inspection at least annually following the manufacturer's recommendations for this class of standby service.
- 2. Station Batteries 61 a. The specific gravity, electrolyte level, voltage of Il each cell in each safety-related battery, and the 16.4-20
WUP Amendment 18 PSAR 6/79 safety-related d-c bus voltage shall be measured and I recorded monthly.
- b. Every three months the specific gravity of each safety-related battery cell, the temperature reading of every fifth cell, the voltage reading of each cell y and total battery terminal voltage, the float voltage, the battery load with the battery on float charge, the level of electrolyte of each cell, and the amount of water added to any cell shall be measured and recorded.
2344 138 16.4-20a
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY AIULLYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 17 Page, Table (T), or Revision Figure (F) Number 17-i and 17-11 14 17-iil - 17-iv 7 17-v 13 17-vi 7 17-vii 13 17.1 13 17.1-1 14 17.1-2 2 17.1-3 and 17.1-4 3 17.1-5 13 17.1-6 and 17.1-6a 14 17.1-7 14 17.1-8 2 17.1-9 through 17.1-12a 14 17.1-13 through 17.1-15 - 17.1-16 14 17.1-17 through 17.1-27 - 17.1-28 13 17.1-29 - 17.1-30 16 17.1-31 13 17.1-32 - 17.1-33 18 . 17.2-1 - T.17.1.3-1 (4 pages) 2 T 1".1.3-2 (8 pages) - F.17.1.1-1 through F.17.1.1-3 13 F.17.1.1-4 14 F.17.1.1-5 .(17 pages) - F.17.1.1-6 (3 pages) - 2344 139 EP.17-1
WUP Amendment 18 PSAR 6/79 meaningful and are eeffectively complying with the corporate policy and 10C8150 App. B criteria.
- 2. Audits by the WE-QA organization to provide a comprehensive independent verification and evaluation of quality procedures and activities to assure that they rare meaningful and are effectively complying with the QA program.
- 3. External audits conducted by WE and its designated agent, sub-contractors, and vendors performing activities in the early stages of design and procurement. These audits include verification and evaluation of their QA program, procedures and activities to assure compliance with all aspects of the QA program and procurement requirements.
- 4. Audits performed by the prime contractors, subcontractors, and vendors to verify and evaluate their own and suppliers QA program, procedures and activities to assure compliance with the QA program and procurement activities.
17.1.2 Quality Assurance during Design and Construction Stone & Webster The contents of this section can be found in the Stone & Webster l2 Topical Report, SHSQAP 1-74A " Standard Nuclear Quality Assurance lM Program" Revision C dated December 1, 1978. QA Category I items covered by this program are listed in Section 3.2.5 of this PSAR. l6 17.1.3 Westinghouse Nuclear Systems Divisions Quality Assurance Plan The contents of this section can be found in WCAP-8370, " Quality 12 Assurance Plan," Revision 8-A dated September 1977 with the 118 addition of Table 17.1.3-1, Application of Quality Systems Requirements and Table 17.1.3-2, NSSS Functional Responsibilities. 2 Reference to Table 3.2-1 on page 17.1-19 in WCAP-8370 is Table 3.2.5-1 of this PSAR which identifies the safety-related structures, systems, and components controlled by this program. 2344 140 17.1-33
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Appendix A Page, Table (T) , or Revision Figure (F) Number A-i and A-ii - A-iii through A-vi 14 A-vii through A-ix 18 A-1 - A.1-1 1 A.1-2 through A.1-4 14 A.1-5 - A.1-6 14 A.1-7 and A.1-8 16 A.1-9 14 A.1-10 16 A.1-11 and A.1-12 18 A.1-13 through A.1-14a 17 A.1-15 through A.1-22 14 A.1-23 through A.1-26 16 A.1-27 and A.1-28 14 A.1-29 and A.1-30 18 A.1-31 through A.1-33 14 A.1-34 through A.1-37 18 A.2-1 and A.2-2 16 T.A-1 (2 pages) - T.A-2 (2 pages) - T.A-3 (2 pages) - T.A-4 (2 pages) -
'k lff EP.A-1
WUP Amendment 18 PSAR 6/79 TABLE OF CONTENTS (CONT *D) Section Title Page of Nuclear Power Plants (Regulatory Guide 1.73) A.1-29 A.1-1.74 Quality Assurance Terms and Definitions (Regulatory Guide 1.74) A.1-29 A.1-1.75 Physical Independence of Electric System (Regulatory Guide 1.75) A.1-30 A.1-1.76 Design Basis for Nuclear Power Plants (Regulatory Guide 1.76) A.1-30 A.1-1.77 Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors (Regulatory Guide 1.77) A.1-30 A.1-1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant during a Postulated Hazardous Chemical Release (Regulatory Guide 1.78) A.1-30 A.1-1.79 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors (Regulatory Guide 1.79) A .1 -31 A.1-1.80 Preoperational Testing of Instrument Air Systems (Regulatory Guide 1.80) A.1-32 A.1-1.81 Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants (Regulatory Guide 1.81) A.1-33 A.1-1.82 Sumps for Emergency Core Cooling and Containment Spray Systems (Regulatory Guide 1.82) A.1-33 A.1-1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes (Regulatory Guide 1.83) A.1-33 A.1-1.84 Code Case Acceptability ASME Section III Design and Fabrication (Regulatory Guide 1.84) A.1-33 A.1-1.85 Code Case Acceptability ASME Section III Materials (Regulatory Guide 1.85) A.1-34 A-vii
WUP Amendment 18 PSAR 6/79 TABLE OF CONTENTS (CONT 8D) Section Title Page A.1-1.86 Termination of Operating Licenses for Nuclear Power Plants (Regulatory Guide 1.86) A.1-34 A.1-1.87 Construction Criteria for Class 1 Components in Elevated Temperature Reactors (Supplement to ASME Section III Code cases 1592, 1594, 1595, and 1596) (Regulatory Guide 1.87) A.1-34 A.1-1.88 Collection, Storage, and Maintenance of Nuclear Power Plants Quality Assurance Records A.1-34 A.1-1.94 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel during the Construction Phase of Nuclear Power Plants (Regulatory Guide 1.94) A.1-34 A.1-1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluat-ing Compliance with 10 CFR Part 50, Appendix I A.1-34 A.1-1.111 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors A.1-35 A.1-1.112 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-cooled Power Reactors A.1-35 A.1-1.113 Estimating Aquatic Disper.sion of Effluent from Accidental x3 Routine Reactor Releases for the Purpose of Implementing Appendix I A.1-35 A.1-1.116 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems (Regulatory Guide 1.116) A.1-35 A-viii
WUP Amendment 18 PSAR 6/79 TABLE OF CONTENTS (CONT *D) Section Title Page A.1-1.123 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Regulatory Guide 1.123) A.1-35 A.1-1.142 Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments) (Regulatory Guide 1.142) A.1-35 A.2 OTHER DIVISION REGULATORY GUIDES A.2-1 A.2-3.2 Efficiency Testing of Air-Cleaning Systems Containing Devices for Removal of Particles (Regulatory Guide 3.2) A.2-1 A.2-4.1 Measuring and Reporting of Radio-activity in the Environs of Nuclear Power Plants (Regulatory Guide 4.1) A.2-1 A.2-8.1 Radiation Symbol (Regulatory Guide 8.1) A.2-1 A.2-8.5 Immediate Evacuation Signal (Regulatory Guide 8.5) A.2-1 A.2-8.8 Information Relevant to Maintaining . Occupational Radiation Exposure As Low As Practicable (Nuclear Reactors) (Regulatory Guide 8.8) A.2-1 2344 144 A-ix
WUP Amendment 18 PSAR 6/79 Discussion Position C.3: The systems listed comply with NRC Position C.3. These are the most critical systems in a nuclear plant and must be carefully protected from contamination, especially stainless steel. These same systems are the only ones the NRC has singled out in Regulatory Guide 1.44, regarding prevention of stainless steel sensitization. For other safety-related systems, it is adequate to use water defined by ANSI-N45.2.1-73, except that the flush water is matched as close as practicable to that intended for normal system operation. For example, l17
" demineralized water" of ANSI is used for systems that operate with demineralized /
deionized / condensate water. It is not necessary to flush such systems with water containing 0.15 ppm chlorides when the 1.0 ppm maximum chlorides required by ANSI would be adequate to prevent contamination. Position C.4: Contamination levels in expendable products are based upon safe practices and industrial availability. Contaminant levels are controlled such that subsequent removal by standard cleaning methods will result in the achievement of final acceptable levels which are not detrimental to the materials. A.1-1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Regulatory Guide 1.38) Quality assurance requirements for pack aging, shipping, receiving, storage, and handling comply to meet the regulatory position of Regulatory Guide 1.38, Rev. 2, issued May 1977, with 16 the following exceptions and modifications: l18 Position C.1.c Contamination levels in expendable products g and C.2.c: and controlled and surfaces are subsequently cleaned to remove contaminants to final acceptance levels. 2344 145 A.1-11
WUP Amendment 18 PSAR 6/79 Discussion Position C.1.c See discussion C.4 of Section A.1-1.37. N and C.2.c: A.1-1.39 Housekeeping Requirements for Water-Cooled Nuclear Power Plants (Regulatory Guide 1.39) g Quality assurance requirements defined in Chapter 17 for housekeeping commit to meet the regulatory position of Regulatory Guide 1.39, Rev. 2, issued September 1977. A.1-1.40 Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (Regulatory Guide 1.40) Continuous-duty Class I motors, installed inside the containment comply with Regulatory Guide 1.40, issued March 1973. Methods of compliance are discussed in Sections 3.11 and 8.3.1.2. A.1-1.41 Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments (Regulatory Guide 1.41) Onsite electric power systems are tested in accordance with Regulatory Guide 1.41, issued March 1973, as part of the initial lh preoperational testing program and also af ter major modifications or repairs. 2344 146 O A.1-12
WUP Amendment 16 PSAR 11/78 welds is controlled and does not present any significant problems. In addition, shop welds of limited accessibility are repetitive due to multiple production of similar components, and such welding is closely supervised. For field application, the type of qualification should be con-sidered on a case-by-case basis due to the great variety of cir-cumstances encountered. A.1-1.72 Spray Pond Plastic Piping (Regulatory Guide 1.72) The design, fabrication, and testing of fiber glass-reinforced thermosetting plastic piping, if used in the spray ponds, complies with the requirements specified in Regulatory Guide 1.72, issued December 1973. A.1-1.73 Oualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plant (Regulatory Guide 1.73) g Stone & Webster Scope of Supply Electric valve operators will be qualified in accordance with the regulatory guide. With regard to aging refer to PSAR statement on IEEE Std-323-1974 in Section 3.11.2, Amendment 4. Westinghouse Scope of Supply. Westinghouse complies wium the requirements of IEEE-382-72 and thereby with Regulatory Guide 1.73 positions with the exception that stem mounted switches are not environmentally tested along with the valve motor operator. A.1-1.74 Ouality Assurance Terms and Definitions (Regulatory Guide 1.74) Stone & Webster Scope of Supply The Quality Assurance Program described in Chapter 17.1.2 commits to meet the regulatory position of Regulatory Guide 1.74 issued 16 February 1974. 14 Westinghouse' Scope of Supply The Quality Assurance Program described in Chapter 17.1.3 commits to meet the regulatory position of Regulatory Guide 1.74, issued IO February 1974. gy4 2344 147 A.1-29
WUP Amendment 18 PSAR 6/79 14l A.1-1.75 Physical Inde>=ndence of Electric Systems gl (Regulatory Guide 1.75) 14l Physical independence of electric system complies with Regulatory Guide 1.75 issued January 1975. Methods of compliance are Il discussed in Section 7.1.2.12. A.1-1.76 Design Basis for Nuclear Power Plants (Regulatory Guide 1.76) The design basis tornado characteristics used for the design of structures, systems, and componente important to safety will at least meet the values of the parameters specified in Table I for the site region as given in Figure 1 of Regulatory Guide 1.76 dated April 1974. If a given site is characterized by less conservative values for the parameters than the regional values in Table I, less con-servative values may be used if justified by a comprehensive 14 analysis. The design basis tornado characteristics are discussed in Sections 2.3.1 and 3.3.2. A.1-1.77 Assumptions Used for Evaluating a Control Rod Eiection Accident for Pressurized Water Reactors (Regulatory Guide 1.77) The results of the Westinghouse analysis show compliance with the regulatory position given in Section C of Regulatory Guide 1.77. In addition, Westinghouse complies with the intent of the assumptions given in Appendix A of that regulatory guide. A.1-1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant During a Postulated 18l Hazardous Chemical Release (Regulatory Guide 1.78) 14 l The assumptions used for identifying chemicals, potentially 18l hazardous to the control room al- in accordance with the provisions of Regulatory Guide 1.78 issued June 1974. Evalu ation of control room habitability is conducted on a case-by-case basis for each specific potential hazard identified. Self-contained breathing apparatus and protective clothing will be provided, if 14 necessary, in the event that applicable toxicity limits may be exceeded to ensure that the control room remains habitable. Potentially hazardous chemicals are discussed in section 2.2.2 of the Site Addendum and Section 9.5.8. 2344 148 g A.1-30
WUP Amendment 14 PSAR 5/26/78 The plant instrument air systems do have an interface with components that are part of safety-related systems. These air controlled components are individually tested to verify that upon I loss of their nonsafety-related air supply, they will respond by assuming their designed fail safe position. Testing to determine plant response to a complete loss of instrument air will not be performed as a single test. A.1-1.81 Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants (Regulatory Guide 1.81) 4 The degree to which emergency and shutdown electric systems will be shared will comply, in all respects, with the provisions of Regulatory Guide 1.81, Rev. 1, issued January 1975. l9 Methods of compliance of shared systems are discussed in Section 8.3.1.2. 4 A.1-1.82 Sumps for Emergency Core Cooling and Containment spray Systems (Regulatory Guide 1.82) The containment recirculation sumps comply in all respects to the provisions of Regulatory Guide 1.82, issued June 1974. The 5 containment recirculation sumps are discussed in Section 6.2.2.2. A.1-1.83 Inservice Inspection of Pressurized Water Reactor 4 Steam Generator Tubes (Regulatory Guide 1.83) Design of the steam generator support and restraint syst9m will permit inservice inspection access to the steam generator Lanway 9 openings. The program for stear. generator tubing inservice inspection will 4 comply with the recommendations of Regulatory Guide 1.83, Rev. 1, issued July 1975, with the following exception: l14 Inservice inspection may be performed during any plant 4 shutdown for maintenance and repair, except that inspection intervals shall not be greater than 24 months nor less than 6 114 months. Details of the program will be provided in the FSAR. 4 A.1-1.84 Code Case Acceptability ASME Section III Design and Fabrication (Regulatory Guide 1.84) The Wisconsin Utilities Project will comply with the intent of Regulatory Guide 1.84, Rev. 11, issued November 1977. Because this guide is revised approximately every 4 months, later j4 revisions of the guide need to be evaluated. Where the guide limits the applicability of a particular Code Case by referencing an additional Regulatory Guide, the Wisconsin Utilities Project will comply to the extent specified by the referenced guide. 2344 149 A.1-33
WUP Amendment 18 PSAR 6/79 A .1-1. 8 5 Code Case Acceptability ASME Section III Materials 4 (Regulatory Guide 1.85) The Wisconsin Utilities Project will comply with the intent of Regulatory Guide 1.85, Rev. 11, issued November 1977. Because this guide is revised approximately every 4 months, later g revisions of the guide will need to be evaluated. Where the guide limits the applicability of a particular Code Case by referencing an additional Regulatory Guide, the Nisconsin Utilities Project will comply to the extent specified by the referenced guide. A.1-1.86 Termination of Operating Licenses for Nuclear 4 Power Plants (Regulatory Guide 1.86) The Wisconsin Utilities plants will be retired based on guidance 9 in Regulatory Guide 1.86, issued June 1974, or by other options acceptable to the NRC at the time of retirement. A.1-1.87 Construction Criteria for Class 1 Components in Elevated Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1594, 1595, and 1596) (Regulatory Guide 1.87) 4 Regulatory Guide 1.87 is not applicable because a pressurized water reactor is utilized. A.1-1.88 Collection, Storage, and Maintenance of Nuclear Power Plants Quality Assurance Records The Quality Assurance Program described in Chapter 17 commits to 16 meet the regulatory position of Regulatory Guide 1.88, Rev. 2, issued October 1976. A.1-1.94 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel during the Construction Phase of Nuclear Power Plants (Regulatory Guide 1.94) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.94, Rev. 1, dated April 1976, subject to the alternatives listed in 18 SWSQAP 1-74A, Appendix VIII. A.1-1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I The dose calculations from radioactive releases were performed in accordance with the methodology and equations of Regulatory Guide 1.109, Rev. 1, issued October 1977. Estimated doses are discussed in Chapter 11 of the Site Addendum. 2344 150 A.1-34
WUP Amendment 18 PSAR 6/79 A.1-1.111 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors The annual average and grazing season average CHI /Q and D/Q estimates were calculated using methodology consistent with Regulatory Guide 1.111, Rev. 1, issued July 1977. Calculation techniques are discussed in Section 2.3.5 of the Site Addendum. A.1-1.112 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors Radioactive source terms are calculated in accordance with Regulatory Guide 1.112, Rev. 0-R, issued May 1977. Liquid and gaseous releases are discussed in Chapter 11. A.1-1.113 Estimating Aquatic Dispersion of Effluent from Acci-dental and Routine Reactor Releases for the Purpose of Implementing Appendix I The initial dilution of the discharges is calculated using models consistent with Regulatory Guide 1.113, Rev. 1, issued April 1977. Dilution estimates are discussed in Section 11.2 of 18 the Site Addendum. A.1-1.116 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems (Regulatory Guide 1.116) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.116, Rev. 0-R, dated June 1976. A.1-1.123 Ouality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants (Regulatory Guide 1.123) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.123, Rev. 1, dated July 1977. A.1-1.142 Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments) (Regulatory Guide 1.142) The procedures and requirements for the design, fabrication, erection, and testing of safety-related concrete structures (other than the reactor containment) comply with Regulatory Guide 1.142, Rev. O, issued April 1978, with the following exceptions: 2344 l5l A.1-35
WUP Amendment 18 PSAR 6/79
- 1. Regulatory Position C.3 - ACI-349, and not the special ductility requirements of Appendix A of ACI-318-71, will be used for design.
All safety-related structures are designed for elastic behavior when subjected to earthquakes much greater in magnitude than those implicit in ACI-318. Where the design earthquake response is required to be in the elastic range, Appendix A inherently does not apply.
- 2. Regulatory Position C.8 - Pipes which contain liquid, gas, or vapor embedded in concrete will be pressure tested under the applicable piping standard.
Whenever a piping system has an applicable testing standard, that standard should be used rather than the test requirement established by ACI-349. Different piping systems have different pressure and time limits for testing. Thus, setting a separate standard for piping embedded in concrete may lead to two sets of tests for the same piping system. However, if a piping system does not have a testing standard, then the pressure testing requirements for embedded pipes set forth in Section 6.3 of ACI-318-71 will be called for in M addition to the requirements of Section 6.3 of ACI-349-76.
- 3. Regulatory Position C.9b, C.9d, and C.9c - The load factors given in ACI-349 will be used.
A. The load factor applied to Pa will be 1.25, as stipulated in load combination (6) of ACI-349-76, in place of the 1.5 factor referred to in Paragraph C.9.b of Regulatory Guide 1.142. The 1.25 factor for Pa is satisfactory for these structures. The purpose of these structures is to withstand the pressure loads, not to act as a leakage barrier to confine the pressure load, as is the case for a reactor containment structure. Furthermore, a load factor of 1.0, similar to that used for the safe shutdown earthquake and tornado, would be reasonable to use. It seems unrealistic to require the same criteria for these safety-related structures as for the leaktight barrier of the reactor containment. B. The load factor applied to Eo will be 1.7, as stipulated in load combinations (2) and (10) of ACI-349-76, in place of the 1.9 factor referred to in paragraph C.9.d of Regulatory Guide 1.142. The use of a 1.9 load factor is excessively conservative. A.1-36
WUP Amendment 18 PSAR 6/79 Use of a load factor of 1.9 for OBE would result in OBE governing the design of some structural elements. In these cases, the OBE ultimate load based on amplifications associated with OBE damping and the 1.9 load factor, is greater than the SSE ultimate load based on amplifications associated with SSE damping and appropriate SSE load factor. For the OBE to govern the design of numerous structural elements would seem to be contradictory to the basic definition of the SSE. The SSE is, by definition, the earthquake which produces the maximum ground motion for which safety-related structures are designed to remain functional. To force the OBE to govern the design of these structures by using an excessively high load factor would appear contradictory to the concept of 18 establishing a safe shutdown earthquake. C. The load factor applied to both Pa and Eo will be 1.15, as defined in load combination (7) of ACI-349-76, in place of the 1.25 factor referred to in Paragraph C.9.c of Regulatory Guide 1.142. The basis for this load factor is defined in paragraphs A and B above. A combined meeting of members of both the ASME Section III, Division 2, and the ACI-349 Code Committees was held in Toronto, Ontario in April 1978 to discuss the difference in load factors used for the loading combinations in the two codes. The joint meeting concluded that the load factors used in each of the codes were appropriate when the consequences of the failure of each type of structure was considered. 2344 153 A.1-37
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Appendix B Page, Table (T) , or Revision Figure (F) Number B-i 14 B.1-1 through B.1-3 14 B.2-1 through B.2-24 14 B.3-1 18 B.3-2 15 B.3-3 through B.3- 8 14 B.3-9 through B.3-12 17 B.3-13 13 B.3-14 through B.3-161 18 B.3-17 18 B.3-18 14 B.3-19 15 B.3-20 17 B.3-21 14 B.3-22 15 B.3-23 14 B.3-24 and B.3-24a 16 B.3-25 and B.3-26 17 B.3-27 14 B.3-28 17 B.3-29 through B.3-39 14 B.3-40 and B.3-41 17 B.3-42 15 B.3-43 through B.3-47 17 B 3-48 18 2344 154 EP.B-1
WUP Amendment 18 PSAR 6/79 QUALIFICATION REVIEW MATTERS A. CHANGES FROM THE BASE PLANT DESIGN A.1 Revision to QA Program Refer to Sections 17.1.2 and 17.1.3 for commitments to the 18 later versions of standard QA programs. 2344 155-B.3-1
WUP Amendment 15 PSAR 9/22/78 A.2 Changes in Fuel Design
- a. See response D.31 and subsections 4.2.1.3.1 and 4.4.1.1.
15 b. See subsections 4.2.1.1.1, 4.2.1.2.1, and 4.2.1.3.1.
- c. See subsection 4.2.1.3.1.
2344 ISip 9 O B.3-2
WUP Amendment 16 PSAR 11/78 l16 D.19 Protection Against Low-Tra-iectory Turbine Missiles (Regulatory Guide 1.115) Turbine missiles are discussed in Section 10.2.3 of the PSAR, including an analysis for high-trajectory missiles. The design of the plant complies with the guidelines of l15 Regulatory Guide 1.115, Revision 1, issued July 1977. D.20 Tornado Design Classification (Regulatory Guide 1.117) The design of the plant complies with the guidelines of Regulatory Guide 1.117, Revision 0, issued June 1976. D.21 Periodic Testing of Electric Power and Protection Systems _(Regulatory Guide 1.118) Stone & Webster Scope of Supply The periodic testing of electric power and protection systems complies with the requirements of Regulatory Guide 1.118, Revision 1, issued November 1977 with the following clarifications regarding sensor-response time testing. The response time of electrical protective relays and motion sensors will be verified by including them in the response time test of the protective system. The response time of passive sensors, i.e., potential transformers and current transformers, which are used ar interfaces between the electrical system and protective relays, does not degrade without changes in the steady state performance of the device. Pressure and level sensors will be included in the response time test of the protection system to verify that the response time is adequate with regard to achieving the intended safety function. The response time testing is only required to ensure the t the response time is less than or equal to the response time used in the accident analysis. Radiation, temperature, and some types of in-line electrical and fluid sensors will not be quantitatively tested for response time after installation. Undetected response time degradation of certain types of sensors is not known to occur provided they are properly installed and calibrated. Verification of proper installation and calibration will be used as an indirect means of assuring proper response times. Equipment suppliers will be required to verify the sensor response time adequacy. These response times will be combined with system response time, which is measured independently of the sensor. B.3-13
WUP Amendment 18 PSAR 6/79 Bench testing, in lieu of in situ testing, will-not be required, since removal of a sensor for the purpose of response time testing would introduce new problems due to occupational radiation doses and potential maintenance / operator error. The overall sensor reliability would be decreased as a result of failures introduced by testing. Repeated testing to detect response time degradation in sensors of proven historical performance would only compromise the safety of plant operation. Westinghouse scope of Supply The periodic testing of electric power and protection systems complies with the requirements of Regulatory Guide 1.118, Revision 1, issued November 1977, with the following clarifications:
- 1. The recommendations will ba considered to be discretionary in accordance with the Applicant's interpretation of the intent of IEEE 338-1971.
- 2. It is believed that testing design features without change proposed in the PSAR, supplemented by vesting procedure and surveillance requirements >f the Technical Specifications, including those for response time testing, will be acceptable for the Haven Plant.
The online testing design features and procedures are addressed in the response to Regulatory Guide 1.22 in Appendix A, and are described in Section 7.1.
- 3. The neutron detectors are considered exempt from the requirements of response time testing.
- 4. Equipment performing control functions, but actuated from protection system sensors is not part of the safety system and will not be tested for time response.
- 5. Readings on channels which monitor the same variable will be adjusted for spatial, environmental and other factors prior to comparing the values of the variables.
- 6. Testing will not be tied to accident conditions, but only to the range of the parameter that is varied.
- 7. The " expected environmental and mechanical installation" will not be duplicated for the testing of sensors which must be removed to accomplish respon.ie time testing unless it can be shown that the duplication is practical and that the duplicated factors significantly influence the sensor time
. response.
B.3-14 2344 lh8
WUP Amendment 18 PSAR 6/79
- 8. Temporary test instrumentation including the use of temporary jumper wires, removal of fuses and other equipment not hard-wire into the protection system will be used where applicable.
D.22 Design Limits and Loading Combinations for Class Linear-Type Component Supports (Regulatory Guide 1.124) The Applicant will use the ASME Code Section III, Sub-section NF as the design basis for reactor coolant system equipment supports (see Section 5.5.14) . The design limits and loading combinations for Class 1 linear-type supports comply with Regulatory Guide 1.124, Rev. 1, dated January 1978, with the following exceptions and additions, as noted in the letter from S. B. Jacobs of Stone & Webster Engineering Corporation to the Secretary of the Commission, US NRC dated June 8, 1978.
- 1. The following paragraph should be added to Regulatory Position C.3:
C.3.c The bending stress limit Fb resulting from tension and bending in structural membera as specified in Appendix XVII-2214 of Section III, Div. 1, should be the smaller value of 0.66 S or 0.55 S u for compact sections, 0.75 S y or 0.63 S for doubly symmetrical members with bending about the minor axis, and 0.6 Sy or 0.5 Su for box-type flexural members and 18 miscellaneous members.
- 2. The second paragraph under Regulatory Position C.4 should be replaced with the following:
However, all increases (i .e . , those allowed by NF-3231.1 (a) XVII-2110 (a), and F-1370 (a)) shall always be limited by XVII-2110 (b) of Section III. The critical buckling strengths defined by XVII-2110 (b) or sectie III should be calculated using material properties at temperature. The increase allowable permitted for tensile stress in bolts shall not exceed the lesser of 0.70 Su or 0.70 Sy at temperature. The increased allowable permitted for shear stress in bolts shall not exceed 0.42 Su at temperature.
- 3. Paragraph C.S.a should be revised as follows:
- a. "The stress limits of XVII-2000 of Section III, and Regulatory Position 3 of this guide, should not be exceeded for component supports designed by the linear elastic analysis method. These 4
B.3-15 2344 159
WUP Amendment 18 PSAR 6/79 stress limits may be increased according to the provisions of NF-3231.1 (a'i af Section III and Regulatory Position 4 of t:lis guide when ef fects resulting from constaaint of free-end displacement and anchor motion are added to the loading combination."
- 4. Regulatory Position C.8 should read as follows:
Supports for the " active" components that are required only during an emergency or faulted plant condition and that are subjected to loading combinations described in Regulatory Positions C.6 and C.7 should be designed within the design limita described in Regulatory Position C.5 or other justifiable design limits. These limits should be defined by the Design Specification and stated in the PSAR, such that the function of the supported systems will be maintained when they are subjected to the loading combinations described in Regulatory Positions C.6 and C.7. Discussion 18
- 1. The paragraph added to Regulatory Position C.3 is necessary because of an apparent oversight in applying the 5/6 factor to bending stress allowables. -
- 2. The third sentence in the second paragraph of Regulatory Position C.4 prohibits the application of the increased allowables presently permitted by NF-3231.1 (a) and F-1370 (a) to Service Limits A or B for bolted connections. The danger of applying the increase presently allowed by Subsection NF has been pointed out at Subsection NF Committee meetings. The SSW position asserts that maximum safe increased allowables are achieved by limiting bolting tensile stess to the lesser of 0.7 S u or Sy at temperature and bolting shear stress to 0.42 Q at temperature. The 0.7 Su lbnit is well recognized in Section III of the Code. The average shear strength of bolting material is about 0.62 S u according to test data, with a standard deviation of 0.033. Results indicate that the ratio of shear strength to tensile strength is independent of the bolt grade. Curves showing this appear on page 50 of Guide to Design for Loited and Riveted Joints, by J.W. Fisher. Test data are given in a paper by J.J. Wallaert and J.W. Fisher, Shear Strength of High-Strength Bolts, Journal of the Structural Division, ASCE, Volume 91, ST3, June 1965.
- 3. Loads developed by anchor motions are also deformation limited and as such are considered to be grouped in a the same category as loads from restraint of free-end B.3-16 2344 160
WUP Amendment 18 PSAR 6/79 displacement. The resulting stresses are essentially of the secondary type.
- 4. Regulatory Position C.8 is revised as shown because this section implies that the lower stress limits associated with the Design Levels A and B Service Limits must be used for any component support that serves a safety-related function during an Emergency or Faulted plant condition. This would seem to imply that 3 main coolant pump support, which constitutes a passive element in the main coolant loop, would have to be designed to meet the Design Level A and B Service Tdmits during an Emergency or Faulted (LOCA) 18 plant condition. This would require that a snubber providing restraint on an RHR line would have to be designed to the Design Level A and B Service Limits during an Emergency or Faulted plant condition. If this is the intent, it is a severe departure from current practice. Only active components, such as valves whose operation is required for safe shutdown during an Emergency or Faulted condition have been required to meet design stress limits for these plant conditions. Level C and D Service TAmits have been considered adequate to assure pressure boundary integrity under the more severe operating conditions.
D.23 Design Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports (Regulatory Guide 1.130) 15 The Applicant will use the ASME Code, Section III, Sub-section NF as the design basis for reactor coolant system equipment supports (seo Section 5.5.24) . The service limits and loading combinations for Class 1 plate and shell type component supports comply with Regulatory Guide 1.130, Rev. 1, dated October 1978 with the following exceptions:
- 1. Paragraph C.3, Regulatory Position should be revised as follows:
Service limits for component supports designed by 18 linear elastic analysis should always be limited by the critical buckling strength. The critical buckling strength should be calculated using material at temperature properties. Conservative factors of safety for flat plates and for shells should be maintained for each design and service limit.
- 2. Regulatory Position C.7 should read as follows:
Supports for the " active" components that are required only during an emergency or faulted plant condition B.3-16a 2544 1b1
WDP Amendment 18 PSAR 6/79 and that are subjected to loading combinations described in Regulatory Positions C.5 and C.6 should be designed within the design limits described in Regulatory Position C.4 or other justifiable design limits. These limits should be defined by the Design Specification such that the function of the supported system will be maintained when they are subjected to the loading combinations described in Regulatory Positions C.5 and C.6. Discussion
- 1. The margin of 2 for flat plates and 3 for shells, now required by paragraph C.3 of Regulatory Guide 1.130 Rev. 1, may be unnecessarily conservative because of the effects of boundary conditions. The margins were arbitrarily selected by the NRC on the basis of the precedent set by Subsection NB and NE of the ALC code for shells (F.S. = 3) , and by the AISC column factors of safety in the case of plates (P.S. varies from 1.67 to 1.92). The NRC has authorized a study of generic bucxling criteria by a group of consultants. Initial reports indicate they are proposing establishhment of 18 critical buckling stresses based on lower bound values of test data from the aerospace industry (thin shells) without specific tolerance requirements of construction. Prudent safety factors would still have to be chosen. A Task Group under the W.G. on Containment, Section III of ASME Code is investigating this problem; considerable work needs to be done in order to establish safe and yet reasonable generic buckling limits for plates and shells.
- 2. Regulatory Position C.7 is revised as shown because this section implies that the lower stress limits associated with the Design Levels A and B Service Limits must be used for any component support that serves a safety-related function during an Emergency or Faulted plant condition. This would seem to imply that a main coolant pump support, which constitutes a passive element in the main coolent loop, would have to be designed to meet the Design Level A and B Service Limits during an Emergency or Faulted (LOCA) plant condition. This would also require that a snubber providing restraint on an RER line would have to be designed to the Design Level A and B Service Limits during an Emergency or Faulted plant condition.
If this is the intent, it is a severe departure from current practice. Only active components, such as valves, . .whose operation is required for saf e shutdown during an Emergency or Faulted condition have been required to meet design stress limits for these plant conditions. Level C and D Service Limits have been B.3-16b M4f l
WUP Amendment 18 PSAR 6/79 considered adequate to assure pressure boundary N integrity under the more severe operating conditions. D.24 Fuel-Oil Systems for Standby Diesel Generators (Regulatory Guide 1.137) The design of the fuel oil systens for standby diesel generators complies with the regulatory positions of 17 Regulatory Guide 1.137 issued Januarf 1978. D.25 Information Relevant to Ensuring that Occupational Radia-tion Exposures at Nuclear Power Stations Will Be As Low An Is Reasonably Achievable (Regulatory Guide 8.8) The application of Regulatory Guide 8.8, Revision 2, issued March 1977 is discussed in Appendix A, Item A.2-8.8. D.26 Guidelines for Fire Protection for Nuclear Power Plants (BTP APCSB 9.5-1) Guidelines for fire protection have been revised and are addressed in Section 9.5. The guidance in Draft Regulatory Guide 1.120 has been used in developing fire protection programs and system design as alternatives to Branch Technical Position APCSB 9.5-1, issued August 1976. D.27 Design, Testing, and Maintenance Criteria for Normal Venti-lation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Reactor Plants (BTP ETSB 11-2) BTP ETSB 11-2 has been issued as P0gulatory Guide 1.140. The normal ventilation exhaust system and filtration units comply with the guidance of Regulatary Guide 1.140 with the following exceptions. Containment Air Filtration c.2.f - Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510-1975 ac 15 recommended in this paragraph. However, ductwork tests are performed using acceptable methods of the Associated Air Balance Council. c.3.c - For HEPA and adsorber mountings the requirement of ANSI N509-1976, Section 5.6.3, will be complied with except for the tolerance requirements. The tolerances for HEPA and adsorber mounting frames will be sufficient to pass the bank leak tests of paragraphs 5.c and 5.d of the guide. c.3.f - Exception is taken to Section 5.10.3.5 of ANSI N509-1976, ductwork, as a structure, will have a resonant frequency above 25 Hz. A considerable portion of the sheet B.3-16c b
WUP Amendment 18 PSAR 6/79 steel bounded by the four corners of the duct and its reinforcing angles contributed little to the stiffness and hence the resonant frequency of the structure. These areas in general vibrate as membranes and normally exhibit several natural frequencies below 25 Hz. Resonant disturbing frequencies do not result in excessively large deflections or stresses in membranes and therefore no attempt will be made to obtain membrane fundamental frequencies above 25 Hz. c.3.h - Exception is taken to Section S.2.2.4 of ANSI N509-1976 which calls for a means of compaction to uniform density. Where uniform compaction can be demonstrated, ompacting means shall not be required. c.3.i - 1) System resistances will be determined in accordance with Section 5.7.1 of PNSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan inlet and outlet losses including a conservative safety factor will be estimated utilizing information available at time of specification. The purchased systems Mill be balanced and adjusted during installation to ensure design flow. 15 2) Exception is taken to Section 5.7.2 of ANSI N509-1976. Copies of fan ratings or test reports are not necessary when certified fan performance curves are furnished.
- 3) Exception is taken to Section 5.7.3 of ANSI N509-1976. Balancing techniques specified may not be followed. Maximum permissible vibration velocity level method may not be complied with.
Dynamic balancing of the fan is required with the maximum double displacement not exceeding established industry values for various rotational speeds. Fan balancing will be performed again after installation.
- 4) Exception is taken to Section 5.7.5 of ANSI N509-1976. Where AMCA certification ratings are submitted, documentation will not be furnished in accordance with Section 5.7.5.
c.3.k - The air flow distribution will be within 120 percent of the average air flow as tested in accordance with ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution can not be achieved. c'.3.1 - Exception is taken to the provisions in Section 5.9 of ANSI N509-1976: B.3-16d 2 b
WUP Amendment 18 PSAR 6/79
- 1) Dampers will not be designed to the specifications of ANSI B31.1.
- 2) Butterfly valves will not be used.
- 3) Class B leakage rates shall be determined for one damper of each type instead of every damper for the most severe conditions anticipated for each damper type. In-place system leakage testing will ensure the absence of excessive leakage.
- 4) Minimum diameter of damper shaft length 24 inches and under shall be 1/2 inch, and 3/4 inch for shafts between 25 and 48 inches.
Note: Items (1) and (2) do not apply to containment penetrations. Service Building Air Filtration c.2.a - The use of prefilterc and downstream HEPA will be assessed on a case by case basis. c.2.f - Housing leak tests are performed in accordance with g the provisions of Section 6 of ANSI N510-1975 as recommended in this paragraph. However, ductwork tests are performed using acceptable methods of the Associated Air Balance council. c.3.c - For HEPA and adsorber mountings the requirement of ANSI N509-1976 Section 5.6.3 will be complied with except for the tolerance requirements. The tolerances for HEPA and adsorber mounting frames will be sufficient to pass the bank leak tests of paragraphs 5.c and 5.d of the Guide. c.3.h - Exception is taken to Section 5.2.2.3 of ANSI N509-1976 which calls for a means of compaction to uniform density. Where uniform compaction can be demonstrated, compacting means shall not be required. c.3.1 - 1) System resistances will be determined in accordance with Section 5.7.1 of ANSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan inlet and outlet losses including a conservative safety factor will be estimated utilizing information available at time of specification. The purchased systems will be balanced and adjusted during installation to ensure design flow.
- 2) Exception is taken to Section 5.7.2 of ANSI N509-B.3-16e 2344 1"^5
WUP Amendment 18 PSAR 6/79 1976. Copies of fan ratings or test reports are not necessary when certified fan performance curves are furnished.
- 3) Exception is taken to Section 5.7.3 of ANSI N509-1976. Balancing techniques specified may not be followed. Maximum permissible vibration velocity level method may not be complied with.
Dynamic balancing of the fan is required with the maximum double displacement not exceeding established industry values for various rotational speeds. Fan balancing will be performed again after installation.
- 4) Exception is taken to Section 5.7.5 of ANSI N509-1976. Where AMCA certification ratings are submitted, documentation will not be furnished in accordance with Section 5.7.5.
c.3.k - The air flow distribution will be within 120 percent of the average air flow as tested in accordance with ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution cannot be achieved. 15 c.3.1 - Exception is taken to the provisions in Section 5.9 of ANSI H509-1976 of designing to ANSI B31.1:
- 1) Dampers will not be designed to the specifications of ANSI B31.1.
- 2) Butterfly valves will not be used.
- 3) Class B leakage rates shall be determined for one damper of each type instead of every damper for the most severe conditions anticipated for each damper type. In place system leakage testing will ensure the absence of excessive leakage.
- 4) Minimum diameter of damper shaft length 24 inches and under shall be 1/2 inch, and 3/4 inch for shafts between 25 and 48 inches.
Auxiliary Building Filtration c.2.b - Filter component layouts will normally consist of a maximum of 3 HEPA filters high and 10 HEPA filters wide. If individual filter components with capacities greater than 1,000 cfm are used, the system flow rate may be greater than 30,000 cfm. c.3.f - Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510-1975 as recommended in this paragraph. However, ductwork tests are 2344 166 6 B.3-16f
WUP Amendment 18 PSAR 6/79 performed using acceptable methods of the Associated Air Balance Council. c.3.c - For HEPA and adsorber mountings the requirements of ANSI N509-1976 Section 5.6.3 will be complied with except for the tolerance requirements. The tolerances of HEPA and adsorber mounting frames will be sufficient to pass the bank leak tests of Paragraphs 5.c and 5.d of the Guide. c.3.h - Exception is taken to Section 5.2.2.4 of ANSI H509-1976 which calls for a means of compaction to uniform density. Where uniform compaction can be demonstrated, compacting means shall not be required. c.3.1 - 1) System resistances will be determined in accordance with Section 5.7.1 of ANSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan inlet and outlet losses including a conservative safety factor will be estimated utilizing information available at time of specification. The purchased systems will be balanced and adjusted during installation to ensure design flow.
- 2) Exception is taken to Section 5.7.2 of ANSI N509- 15 1976. Copies of fan ratings or test reports are not necessary when certified fan performance curves are furnished.
- 3) Exception is taken to Section 5.7.3 of ANSI N509-1976. Balancing techniques specified may not be followed. Maximum pennissible vibration velocity level method may not be complied with.
Dynamic balancing of the fan is required with the maximum double displacement not exceeding established industry values for various rotational speeds. Fan balancing will be performed again after installation.
- 4) Exception is taken to Section 5.7.5 of ANSI N509-1976. Where AMCA certification ratings are submitted, documentation will not be furnished in accordance with Section 5.7.5.
c.3.k - The air flow distribution will be within 120 percent of the average air flow as tested in accordance with ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution cannot be achieved. c.3.1 - Exception is taken to the provisions in Section 5.9 of ANSI H509-1976 of designing to ANSI B31.1: 2344 167 B.3-16g
WUP Amendment 18 PSAR 6/79
- 1) Dampers will not be designed to the specifications of ANSI B31.1.
- 2) Butterfly valves will nut be used.
- 3) Class B leakage rates shall be determined for one damper of each type instead of every damper for the most severe conditions anticipated for each damper type. In place system leakage testing will ensure the absence of excessive leakage.
- 4) Minimum diameter of damoer shaft length 24 inches and under shall be 1/2 inch and 3/4 inch for shafts between 25 and 48 inches.
Process Vent Filter Assemblies c.2.f - Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510-1975 as recommended in this paragraph. However, ductwork tests are performed using acceptable methods of the Associated Air Balance Council. H c.3.c - For HEPA and adsorber mountings the requirement of ANSI H509-1976 Section 5.5.3 will be complied with except for the tolerance requirements. The tolerances for HEPA and adsorber mounting grames will be sufficient to pass the bank leak tests of Paragraphs 5.c and 5.d of the Guide. c.3.g - The dwell time for a minimum 2 inches of charcoal adsorber unit will be 0.25 sec. For bed depths greater than 2 inches, where the dwell time is less than 0.25 sec per 2 inch depth of total bed depth, experimental verification of filter efficiency will be provided. c.3.h - Exception is taken to Section 5.2.2.4 of ANSI N509-1976 which calls for a means of compaction to uniform density. Where uniform compaction can be demonstrated, compacting means shall not be required. c.3.1 - 1) System resistances eill be determined in accordance with Sectica 5.7.1 of ANSI N509-1976 except tht fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan inlet and outlet losses including a conservative safety factor will be estimated utilizing information available at time of specification. The purchased systems will be balanced and adjusted during installation to ensure design flow.
- 2) Exception is taken to Section 4.7.2 of ANSI N509-h B.3-16h bb
WUP Amendment 18 PSAR 6/79 1976. Copies of fan ratings or test reports are not necessary when certified fan performance curves are furnished.
- 3) Exception is taken to Section 5.7.3 of ANSI N509-1976. Balancing techniques specified may not be followed. Maximum permissible vibration velocity level method may not be complied with.
Dynamic balancing of the fan is required with the maximum double displacement not exceeding established industry values for various rotational speeds. Fan balancing will be performed again after installation.
- 4) Exception is taken to Section 5.7.5 of ANSI N509-1976. Where AMCA certification ratings are submitted, documentation will not be furnished in accordance with Section 5.7.5.
c.3.k - The air flow distribution will be within 120 percent of the average air flow as tested in accordance with ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution cannot be achieved. c.3.1 - Exception is taken to the provisions in Section 5.9 15 of ANSI N509-1976 of designing to ANSI B31.1:
- 1) Dampers will not be designed to the specifications of ANSI B31.1.
- 2) Butterfly valves will not be used.
- 3) Class B leakage rates shall be determined for one damper of each type instead of avery damper for the most severe conditions antic ra;ed for each damper type. In place system leakage testing will ensure the absence of excessive leakage.
- 4) Minimum diameter of damper shaft length 24 inches and under shall be 1/2 inch and 3/4 inch for shafts between 25 and 48 inches.
Iodine Removal Unit c.2.f - Housing Ir E, tests are performed in accordance with the provisions of Section 6 of ANSI N510-1975 as recommended h r Of paragraph. However, ductwork tests are performed ur As; 1...,ptable methods of the Associated Air Balance Counctl. c.3.c - For HEPA and adsorber mountings the requirement of ANSI N509-1976 Section 5.6.3 vill be complied with except for the tolerance requirements. The tolerances for HEPA B.3-161 2344 169
WUP Amendment 18 PSAR 6/79 and adsorber mounting frames will be sufficient to pass the bank leak test of Paragraphs 5.c and 5.d of the Guide. c.3.g - The dwell time for a minimum 2 inches of charcoal adsorber unit will be 0.25 sec. For bed depths greater than 2 inches, where the dwell time is less than 0.25 sec per 2 inch depth of total bed depth, experimental verification of filter efficiency will be provided. c.3.h - Exception is taken to Section 5.2.2.4 of ANSI N509-1976 which calls for a means of compaction to uniform density. Where uniform compaction can be demonstrated, compacting means shall not be required. c.3.k - The air flow distribution will be within 20 percent of the average air flow as tested in accordance with ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution cannot be achieved. c.3.1 - Exception is taken to the provisions in Section 5.9 of ANSI N509-1976 of designing to ANSI B31.1:
- 1) Dampers will not be designed to the specifications of g ANSI B31.1.
- 2) Butterfly valves will not be used.
- 3) Class B leakage rates shall be determined for ona damper of each type instead of every damper for the most severe conditions anticipated for each damper type. In place system leakage testing will ensure the absence of excessive leakage.
- 4) Minimum diameter of damper shaft length 24 inches and under shall be 1/2 inch, and 3/4 inch for shafts between 25 and 48 inches.
Fuel Building Normal Ventilation c.2.a - The use of prefilters will be assessed on a case by case basis. c.2.f - Housing leak tests are performed in accordance with the provisions of Section 6 of ANSI N510 *?75 as recommended in this paragraph. However, ductwork tests are performed using acceptable methods of the Associated Air Balance Council. c.3.c - For HEPA mountings the requirement of ANSI N509-1976 Section 5.6.3 will be complied with except for the tolerance requiremen- The tolerances for HEPA and adsorber mounting frame; will be sufficient to pass the bank leak tests of Paragraphs 5.c and 5.d of the Guide. B.3-16j 2344 i/0
NUP Amendment 18 PSAR 6/79 c.3.1 - 1) System resistances will be determined in accordance with Section 5.7.1 of ANSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan inlet and outlet losses including a conservative safety factor will be estimated utilizing information available at time of specification. The purchased systems will be balanced and adjusted during installation to ensure design flow.
- 2) Exception is taken to Section 5.7.2 of ANSI N509-1976. Copies of fan ratings or test reports are not necessary when certified fan performance curves are furnished.
- 3) Exception is taken to Section 5.7.3 of ANSI N509-1976. Balancing techniques specified may not be followed. Maximum permissible vibration velocity level method may not be complied with.
Dynamic level method may not be complied with. Dynamic balancing of the fan is required with 15 the maximum double displacement not exceeding established industry values for various rotational speeds. Fan balancing will be performed again after installation.
- 4) Exception is taken to Section 5.7.5 of ANSI N509-1976. Where ACMA certification ratings are submitted, documentation will not be furnished in accordance with Section 5.7.5.
c.3.k - The air flow distribution will be within 120 percent of the average air flow as tested in accordance
- 11th ANSI N510-1975. Turning vanes will be provided only where a uniform air distribution cannot be achieved.
c.3.1 - Exception is taken to the provisions in Section 5.9 of ANSI N509-1976 of designing to ANSI B31.1: a Dampers will not be designed to the specifications of ANSI B31.1.
- 2) Butterfly valves will not be used.
- 3) Class B leakage rates shall be determined for one damper of each type instead of every damper for the most severe conditions anticipated for each damper type. In place system leakage testing will ensure the absence of excessive leakage.
2344 171 B.3-16k
WUP Amendment 18 PSAR 6/79
- 4) Minimum diameter of damper shaft length 24 inches and 15 under shall be 1/2 inch, and 3/4 inch for shafts between 25 and 48 inches.
D.28 Design Guidance for Solid Radioactive Water Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants (BTP ETSB 11-3) The solidification system that will be utilized conforms to the design guidance presented in ETSB 11-3. A description of the solidification system is provided in Section 11.5. 2344 172 O O B.3-161
WUP Amendment 18 PSAR 6/79 D.29 Use of ACI 349 ACI-349, " Standard Code Requirements for Nuclear Safety Related Concrete Structures - June 1976" is the appropriate industry standard used for establishing the structural design criteria discussed in Sections 3.8.3, 3.8.4, and 3.8.5 of this PSAR. Based on the above, the Applicant proposes the continued use of ACI-349 as the principal code for design of all nuclear safety-related concrete structures (other than the containment shell, mat and dome) . Refer to the response to Question 130.29 and the compliance statement for Regulatory Guide 1.142 discussed in g Item A.1-1.142 of Appendix A for further information on the acceptability of ACI-349. 2344 173 d B.3-17
WUP Amendment 14 PSAR 5/26/78 D.30 Fuel Handling Accident The calculated dose resulting from a fuel handling accident inside containment is well within the guidelines of 10CFR100. A conservative analysis of this accident utilizing the assumptions of Regulatory Guide 1.25 with no credit taken for filtration through carbon adsorbers, containment mixing, or containment isolation, results in a calculated site boundary thyroid dose of 78 rem in the worst sector. A more realistic analysis with credit taken for the non-nuclear safety grade carbon adsorbers in the containment air filtration system, which normally is in operation during containment purge operations, results in a calculated thyroid dose of 7.8 rem in the worst sector. Further reduction in the calculated dose would be realized due to the automatic isolation of the containment purge air system upon receipt of a high radiation signal from the redundant containment purge air vent monitors. The containment purge air system is described in Section 9.4.7.2. The containment purge air vent monitors are described in Section 11.4. Following isolation, the radioactivity in the containment atmosphere continues to be reduced by operation of the containment air filtration system in the recirculation mode or allowing the iodines to decay. If additional purification is required prior to release, the containment purge can be manually aligned to the fuel building emergency filters which are safety related. 2344 174 O B.3-18
WUP Amendment 17 PSAR 2/79 The releases resulting from a steam generator are essentially the same as for the Model D4 and results in no change in consequences. Based on the above discussion it can be concluded that the 17 Model F steam generator does not significantly change the analysis presented in the PSAR and the conclusion regarding compliance with the acceptance criteria and safety margins remain valid with the substitution of Model F steam generator. 2344 175 B.3-47
WUP Amendment 18 PSAR 6/79 TABLE E.10-1 _ COMPARISON OF MODEL D4 AND MODEL F PARAMETERS D4 F Tube I.D. (in .) 0.664 0.608 Number of tubes 4,578 5,626 Primary volume (fta) 938 967 Primary pressure drop (psi) 34.0 34.9 Tube bundle height 336 343 Steam generator heat transfer 48,300 55,000 U surface area (ftz) Secondary water mass inventory (1bm)
- Full load 103,500 109,000
- No load 163,000 166,000 Steam generator level setpoints (percent of narrow range span)
- Span (in.) 122 128
- Hi-Hi level 85% 78.1%
- Lo-Lo level 37% 15.6%
E.11 Meteorological Computations The text of the Site Addendum, Section 2.3.4, tables of CHI /Q values for accident conditions, and the resulting
- Chapter 15 tables on DBA doses have neen updated. The reason for the change is the utilization of a more advanced g calculation model now used by the Staff, based on recent wind tracer tests for the determination of 5-percent CHI /Q calculations.
23A4 176 lll D.3-48
WUp Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES NRC Ouestions Page, Table (T) , or Revision Page, Table (T) , or Revision Piqure (F) Number Fiqure (F) Number ARQ i through ARQ iii 15 QO20.13-1 6 Q01.1-1 through QO1.4-1 1 QO20.14-1 4 002.1-1 through QO2.24-1 1 Q020.15-1 through Q020.19-1 2 QO3.1-1 4 QO20,20-1 and QO20.21-1 15 QO3.1-2 1 Q020.22-1 through QO20.25-1 2 QO3.2-1 6 0020.26-1 4 QO3.3-1 and QO3.4-1 1 F.020.26-1 and F.020.26-2 4 Q11.1-1 1 Q020.27-1 through QO20.27-3 5 Q11.2-1 and Q11.2-2 1 QO20.28-1 through QO20.43-1 2 Q11.3-1 through Q11.8-1 1 Q020.44-1 through QO20.48-1 6 Q11.9-1 4 9020.49-1 15 Q13.1-1 through Q13.13-1 1 Q020.50-1 through QO20.54-2 6 Q21.1-1 and Q21.2-1 1 Q020.55-1 and Q020.56-1 6 Q21.3-1 5 0020.57-1 and 9020.57-2 16 Q21.4-1 and Q21.5-1 1 Q21.6-1 13 0042.1-1 and Q042.1-2 17 Q21.7-1 through Q21.14-1 1 F.042.1-1 17 022.1-1 through Q22.3-1 1 Q042.2-1 3 Q22.4-1 5 0042.3-1 2 Q22.5-1 and Q22.6-1 1 Q042.4-1 and Q042.5-1 3 Q31.1-1 and Q31.2-1 1 Q042.6-1 10 Q31.3-1 .- Q042.7-1 2 Q31.4-1 15 0042.8-1 4 Q31.5-1 1 Q042.9-1 and Q042.10-1 2 Q31.6-1 0 0042.11-1 5 Q31.7-1 4 Q042.12-1 and Q042.12-2 8 Q31.8-1 13 Q042.13-1 10 Q31.9-1 and Q31.10-1 1 Q042.14-1 8 Q31.11-1 6 Q042.14-2 10 Q31.12-1 1 Q042.15-1 and 0042.16-1 8 Q31.13-1 2 0042.17-1 2 Q31.14-1 and Q31.15-1 1 0042.18-1 and 0042.19-1 8 Q31.16-1 3 Q042.20-1 2 Q33.1-1 through Q33.9-1 0 Q042.21-1 4 Q41.1-1 1 Q042.22-1 7 Q42.1-1 through Q42.5-1 1 Q042.23-1 2 0042.24-1 5 AEC-i through AEC-lii 16 Q042.25-1 4 AEC-iv through AEC-xi 18 0042.26-1 2 0042.27-1 and Q042.28-1 11 QO3.2-1 6 Q042.29-1 and Q042.30-1 2 Q042.31-1 through Q042.31-10 11 0010.1-1 through QO10.5-2 2 F.042.31-1 and F.042.31-2 11 0010.6-1 and QO10.6-2 8 0042.32-1 through Q042.39-1 6 QO10.7-1 7 0042.40-1 through 0042.42-1 8 0010.8-1 through Q010.8-5 6 0042.43-1 through 0042.46-1 6 Q010.9-1 6 Q042.47-1 and Q042.47-2 17 Q010.10-1 and QO10.11-1 16 Q042.48-1 and 0042.48-2 18 Q042.49-1 and Q042.49-2 18 Q020.1-1 2 QO20.2-1 through QO20.7-1 2 Q110.1-1 4 0020.8-1 4 Q110.2-1 and Q110.3-1 2 QO20.9-1 3 Q110.4-1 and Q110.5-1 4 Q020.10-1 and Q620.10-2 2 Q110.6-1 2 0020.11-1 4 Q110.7-1 4 QO20.12-1 2 Q110.8-1 }}4f l(( 2 EP.Q-1
ygp Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES f?yf 'U) HRC Ouestions Page, Table (T) , or Revision Page, Table 07 , or Revision Fiqure (F) Number Fiqure (F) Number Q110.9-1 13 Q214.10-1 through Q214.12-1 4 Q110.10-1 6 F.214.12-1 and F.214.12-2 7 Q110.11-1 7 Q214.13-1 and Q214.14-1 4 Q110.12-1 through Q110.16-1 6 Q214.15-1 and Q214.16-1 7 Q110.17-1 17 Q214.17-1 and Q214.18-1 3 Q110.18-1 through Q110.21-1 18 F.214.18-1 7 Q214.19-1 7 Q120.1-1 2 Q214.20-1 4 Q120.2-1 and Q120.3-1 4 Q214.21-1 3 Q120.4-1 through Q120.4-7 2 Q214.22-1 through Q214.25-1 7 Q120.5-1 2 Q214.26-1 and Q214.27-1 8 Q120.6-1 through Q120.11-1 6 Q214.28-1 9 Q120.12-1 3 0214.28-2 14 Q120.13-1 through Q120.15-1 2 Q214.29-1 through Q214.33-1 7 Q120.16-1 5 Q214.34-1 11 Q120.17-1 and Q120.17-2 2 Q214.35-1 through Q214.35-3 16 Q121.1-1 and Q121.1-2 6 0214.36-1 through Q214.36-5 16 Q121.2-1 and Q121.5-2 6 Q214.37-1 18 Q121.5-3 9 F.214.37-1 and F.214.37-2 18 Q121.5-4 through Q121.5-8 6 Q214.38-1 through Q214.43-1 18 Q121.6-1 through' Q121.9-1 16 Q121.10-1 through Q121.12-1 18 Q221.1-1 7 Q122.18-1 7 Q221.2-1 4 Q122.18-3 and Q122.18-4 6 Q221.3-1 and Q221.3-2 4 Q221.4-1 12 Q130.1-1 through Q130.5-1 2 Q221.4-2 4 Q130.6-1 6 Q221.5-1 7 Q130.7-1 4 Q221.6-1 and Q221.7-1 4 Q130.7-2 2 Q221.8-1 3 F.130.)-? - Q221.9-1 14 P.130.7-2 tirough F.130.7-4 2 Q221.10-1 3 Q130.8-1 9 Q221.11-1 13 Q130.9-1 4 Q221.12-1 2 Q130.10-1 7 Q221.13-1 and Q221.14-1 4 Q130.11-1 and C130.12-1 2 Q221.15-1 7 Q130.13-1 thrcugh Q130.13-3 4 Q221.16-1 2 Q130.14-1 2 0221.17-1 4 Q130.15-1 through Q130.24-1 6 Q221.18-1 through Q221.20-1 2 Q130.25-1 and Q130.25-2 18 Q221.21-1 7 Q130.26-1 through 0130.26-6 18 Q221.22-1 3 Q130.27-1 through Q130.30-1 18 Q221.23-1 through Q221.30-1 2 Q130.34-1 18 Q221.31-1 6
*
- Q221.32-1 and Q221.32-2 7 Q211.1-1 through Q211.8-1 2 Q221.33-1 through Q221.36-1 6 Q211.9-1 4 9221.37-1 7 Q211.10-1 and Q211.11-1 2 Q221.38-1 and Q221.39-1 13 Q211.12-1 and Q211.13-1 6 Q221.40-1 9 Q214.1-1 and Q214.2-1 2 Q221.41-1 11 Q214.3-1 3 T221.41-1 and T221.41-2 10 0214.4-1 5 (deleted)
Q214.5-1 13 T221.41-3 through T221.41-5 7 Q214.6-1 3 Q221.42-1 7 Q214.7-1 2 0221.43-1 and Q221.44-1 9 Q214.8-1 3 Q221.45-1 6 Q214.9-1 through Q214.9-4 5 Q221.46-1 13 F.214.9-1 through F.214.9-4 5 0221.47-1 through Q221.49-4 17 EP.Q-2 f}ff
WUP Amendment 18 PSAR 6/79 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONT *D) NRC Ouestions Page, Table (T) , or Revision Page, Table (T) , or Revision Piqure (F) Number Figure (F) Number Q221.50-1 and Q221.50-2 18 Q421.1-1 3 Q222.1-1 through Q222.3-1 16 Q421.2-1 10 Q221.6-1 18 Q421.3-1 3 Q222.7-1 and Q222.7-2 18 Q422.1-1 and 0422.1-6 5 Q222.8-1 and Q222.8-2 18 F.422.1-1 through F.422.1-3 5 Q240.1-1 16 Q422.2-1 2 Q240.2-1 and Q240.2-2 18 T.240.2-1 through T.240.2-3 18 Q430.1-1 3 F.240.2-1 and F.240.2-2 18 Q430.2-1 and Q430.2-2 3 Q241.1-1 and Q241.2-1 10 Q430.3-1 17 Q241.3-1 through 0241.5-1 18 Q430.4-1 18 Q242.1-1 and Q242.2-1 2 Q431.1-1 3 Q432.1-1 3 0312.1-1 and Q312.2-1 16 0432.10-1 18 Q312.9-1 18 Q312.10-1 and Q312.10-2 18 Q312.11-1 18 Q331.1-1 7 Q331.2-1 through Q331.6-1 2 Q331.7-1 16 Q331.8-1 through Q331.12-1 2 Q331.13-1 6 Q331.14-1 15 Q331.15-1 and Q331.16-1 6 Q331.17-1 and Q331.17-2 - Q331.18-1 through Q331.21-2 6 Q331.22-1 through Q331.27-1 16 Q 331.28 -1 18 Q331.29-1 and Q331.29-2 18 Q411.1-1 14 Q411.2-1 Q411.3-1 through Q411.5-1 13 2 h}df l[ Q411.6-1 13 Q411.7-1 through Q411.9-1 14 Q441.10-1 13 F.411.10-1 13 Q411.11-1 13 Q411.12-1 2 Q411.13-1 6 Q411.14-1 through Q411.16-1 16 Q411.17-1 and Q411.18-1 18 Q412.1-1 and Q412.1-2 3 Q412.2-1 3 Q412.3-1 through Q412.3-6 6 Q412.4-1 6 Q412.5-1 18 Q413.1-1 and Q413.2-1 3 Q413.3-1 and Q413.4-1 2 Q413.5-1 3 Q413.6-1 through Q413.6-3 4 , 3 Q413.7-1 and Q413.8-1 ' 3 Q413.9-1 2 Q413.10-1 9 Q413.11-1 and Q413.12-1 17 EP.Q-3
WUP Amendment 16 PSAR 11/78 Question No. Subiect Paqe 042.6 (6.2.1.3) Containment Initial conditions 0042.6-1 042.7 (6.2.1.3) Tagami Equation Q042.7-1 - 042.8 (6 . 2.1. 3) Component Cooling Water Temperature Q042.8-1 042.9 (6 . 2 .1. 3) Containment Spray and Fan Coolers Q042.9-1 042.10 (6 . 2 .1. 3) Containment Pressure and Temperature during LOCA Q042.10-1 042.11 (6.2.1.3) Reactor Coolant System Leakage Q042.11-1 042.12 (6.2.1.3) Containment Pressure Response Analysis Q042.12-1 042.13 ( 6 . 2.1. 3) External Design Pressure of the containment Structure Q042.13-1 042.14 (6 . 2.1. 3) Subcompartment Analysis Q042.14-1 042.15 (6 . 2 .1. 3) Reactor Coolant System Pipe Break Q042.15-1 042.16 (6.2.1.3) Subcampartments Not located near a Rupture Q042.16-1 042.17 (6.2.1.3) Volumes and Vent Areas for Sub-compartment Pressore Response Analysis Q042.17-1 042.18 (6 . 2 .1. 3) Surge Line Rupture Q042.18-1 042.19 (b.2.1.3) Pressure Response between the Floors Q042.19-1 042.20 (6.2.1.1) Sand Plugs Q042.20-1 042.21 (6 . 2.1. 3) CUPAT 0042.21-1 042.22 (6.2.2.2) Debris Size in RHR Q042.22-1 042.23 (6.2.2.3) Trapped Reactor Coolant Blowdown Liquid and Spray Water Q042.23-1 042.24 (6.2.2.3) Sump and Spray Systems Q042.24-1 042.25 (t. 2.2.4) Containment Air Recirculation Fan Cooler System Q042.25-1 042.26 (6.2.4) Containment Vent and Purge Valves Q042.26-1 042.27 (6.2.4) Purging of the Containment Q042.27-1 042.28 (6.2.4) LOCA Q042.28-1 042.29 (6.2.5) Mixing of Hydrogen within Con-tainment Subcompartments Q042.29-1 042.30 (6.2.5.3) Curve of Hydrogen Concentration Q042.30-1 042.31 (6.2.1) Postulated Main Steam Line Breaks Q042.31-1 042.32 (6.2.1) ECCS Spillage Q042.32-1 042.33 (6.2.1) Reflood and Post-Reflood Phases of LOCA Q042.33-1 042.34 (6.2.1) Gap Conductance for Concrete Heat Sinks Q042.34-1 042.35 (6.2.1) Heat Sinks Q042.35-1 042.36 (6.2.1) Heat Sinks Equipment and Components Q042.36-1 042.37 (6.2.1) Design Pressure and Differential Pressure QL42.37-1 042.38 (6. 2.1) Fan Cooler Heat Removal Rate Q0,2.38-1 042.39 (6 . 2.1) Containment Spray System Q042.39-1 042.40 (6.2.1) CUPAT Q042.40-1 042.41 (6. 2.1) Inertial Effects on Subcompartments Q042.41-1 042.42 (6. 2.1) L/A Valves for Flow Paths in sub-compartments Q042.42-1 042.43(6.2.4.1) Hydrogen Recombiner Suction Lines Q042.43-1
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2344 180
WUP Amendment 18 PSAR 6/79 Question No. Subiect Page k 042.44 (6.2.5) Monitoring of Hydrogen Concentra-tion Q042.44-1 042.45 (6.2.5) Design Criteria for Hydrogen Analyzers Q042.45-1 042.46 (6.2.5) Hydrogen Production and Accumulation Q042.46-1 042.47 (RSP ,B . 3) Containment Leak Testing Program Q042.47-1 042.48 -(6.2) Containment System Leak Rate Testing Q042.48-1 042.49 (6.2) Steam Generator Water Seal Q042.49-1 110.1 (3.6.2.1) Piping Systems: WCAP 8082 Q110.1-1 110.2 (3.6.4.2) Restrained Piping Systems Q110.2-1 110.3 (3 . 9 .1. 2) Testing Procedures for Mechanical Componente Q110.3-1 110.4 (3 . 9 .1. 3) Pre-operational Test Program Prototype Plant Q.110. 4 -1 110.5 (3 . 9 .1. 5, Faulted Condition Loading
- 3. 9.1. 6) Combinations Q110.5-1 110.6 (3.10) Electrical and Mechanical Equipment Seismic Qrilification Q110.6-1 110.7 (5.2.1) Elastic System Analysis and Component Inelastic System Analysis Q110.7-1 110.8 (16.4) Inservice Test Programs for Pumps and Valves Q110.8-1 110.9 Loose-Parts Monitoring System Q110.9-1 110.10 Main Coolant Loop (s) Q110.10-1 110.11 (3. 9 .1. 6) Dynamic Loading and Loading Combinations Q110.11-1 110.12 (T3. 9-2) ASME Class 2 and 3 Pressure Vessels Q110.12-1 110.13 (T3.9-3, Emergency Loading 3.9.2.3) Q110.13-1 110.14 (3.9.2.2) Design Conditions I and II Q110.14-1 110.15 (3.10) Seismic Listing Methods Q110.15-1 110.16 (10.3.3) Main Steam Stop and Check Valves Q110.16-1 110.17 (App. B) Seismic Qualification of NSSS Supplied Class IE Equipment Q110.17-1 110.18 (App. B) Regulatory Guide 1.121 Q110.18-1 (RSP) 110.19 (App. B) Regulatory Guide 1.124 Q110.19-1 110.20 (App B) Regulatory Guide 1.130 Q110.20-1 (RSP) 110.21 (App. B) Service Limits Q110.21-1 (RSP) 120.1 ~ (u .2.3, Materials with Yield Strength 6.2, 6.33 Greater Than 90,000 psi Q's20.1-1 120.2 (5. 4 . 3, Regulatory Position 2 5.4, 5.5, A.1-150) Q120.2-1 120.3 (10.3, Class 2 and 3 Camponents 10.4', A.1-150) Q120.3-1 120.4 (4.2.2) Regulatory Guide 1.31 Q120.4-1 120.5 (4.2.2) Welder Qualification Q120.5-1
' ^" -*"
2344 lgl
WUP Amendment 18 PSAR 6/79 Question No. Subiect Page 120.6 (5.2.4) Class 1 Components Q120.6-1 120.7 (5.0, 6.0, Fracture Toughness Test Methods 10.0) Q120.7-1 120.8 (5.2.3.4) pH Variance Q120.8-1 120.9 (5.2.6) Flywheel Inservice Inspection Program Q120.9-1 120.10 (5.2.7.4) RCPB Leak Detection Q120.10-1 120.11 (5.4.2) Regulatory Position C.1b (1) Q120.11-1 120.12 (5.5.2.4) Steam Generator Tubing Q120.12-1 120.13 (6.0) Pressure-Retaining Ferritic Materials Q120.13-1 120.14 (6.0) ESF Construction Materials Q120.14-1 120.15 (6.0) Control of pH of ESF Coolants Q120.15-1 120.16 (6.0) Storage of ESF Coolants Q120.16-1 120.17 (6.0) High Energy Turbine Missiles Q120.17-1 121.1 (5.2.8) High Energy Fluid System Piping Q121.1-1 121.2 10CFR50.55a Q121.2-1 121.3 (RSP, Appendix G, 10CFR50 5.2.4) Q121.3-1 121.4 (RSP, Tensil Strength and Toughness Data 5.4.2) Q121.4-1 121.5 (RSP, Turbine Integrity 10.2.3) Q121.5-1 121.6 (5.0) Primary Component Supports Q121.6-1 121.7 (5.2) SA 533 Class 2 Steel Q 121.7-1 121.8 (RSP , B. 3) Regulatory Guide 1.99 P-T Limits Q 121. 8-1 121.9 (5.2.4.2) 10CFR50 Appendix GSH Q121.9-1 121.10 (5.2) Reactor Coolant Pressure Boundary Q121.10-1 Materials 121.11 (5.2.4) Appendices G and H of 10CFR50 Q121.11-1 (5.4.3) (RSP) 121.12 (5.2) Information on Fracture Toughness Q121.12-1 Requirements 122.18 AVT Treatment (5.5.2.3.4) Q122.18-1 130.1 (3.3.2.2) Real Time Analysis Q130.1-1 130.2 (3. 3. 2. 3) Structural Steel Framing Q130.2-1 130.3 (3.5.4) Penetration of Missiles Q130.3-1 130.4 (3.5) Ductility Ratios Q130.4-1 130.5 (3.7.1.3) Damping Valves Q130.5-1 130.6 (3.7.1. 6) Soil Structure Interaction Q130.6-1 130.7 (3.7.2.3) Degrees-of-Freedom Q130.7-1 130.8 (3. 7 . 2 .14 ) Damping Q130.8-1 130.9 (3.7.3.4) Model Responses Q130.9-1 130.10 Peak Shock Recorders (3.7.4.2.3) Q130.10-1 130.11 (3. 8.1.1) Welding Q130.11-1 130.12 (3. 8 .1. 2) Section Errata Q130.12-1 130.13 (3 . 8.1. 4) Containment Structure Analysis Q130.13-1 130.14 (3.8.3.3) Loadina Combinations Q130.14-1 130.15 (3.4) Water Head Q130.15-1 130.16 , Scabbing of Concrete Missile Barrier Q130.16-1 AEC-v 2344 182
WUP Amendment 18 PSAR 6/79 Question No. Subiect Page h 130.17 (3.5.4) Ductility Ratios Q130.17-1 130.18 (3.7.1.6, Embedment and Soil Conditions Q130.18 -1 3.7.2.5) 130.19 Adequate Number of Masses Q130.19-1 130.20 Triaxial Rear Shock Recorder Q130.20-1 (3.7.4.2.3) 130.21 (3. 8 .1. 3) Ambient Temperature Q130.21-1 130.22-(3.8.1.5) Tangential Shear Stress Q130.22-1 130.23 (3.8.3) Local Combination Q130.23-1 130.24 (3. 8 .5. 5) Buoyancy Q130.24-1 130.25 (3.5.4) Tornado Missile Design Criteria Q130.25-1 (RSP) 130.26 (3.5.4) Ductility of Structural Elements Q130.26-1 (RSP) 130.27 (3.7 .1. 6) Analysis of Soil-Structure Inter- Q130.27-1 (3.7.2.5) (RSP) action 130.28 (3.7.2) Failure of Non-Category 1 Structures Q130.28-1 130.29 (3.8.3) ACI-349-76 Q130.29-1 (3.8. 4) (RSP) 130.30 (3.8.4) Subcompartment Pressurization Loads Q130.30-1 130.34 (3 .7.1. 6) Analysis of Soil-Structure Inter- Q130.34-1 (3.7.2.5) (RSP) action 211.1 (3.2) Control Rod Drive Mechanism Housing Q211.1-1 211.2 (3.2) Reactor Coolant Pump Bolting Q211.2-1 211.3 (3.2) Auxiliary Systems of Diesel Genera-tors Q211.3-1 211.4 (3.2) Fuel Transfer Tube Q211.4-1 211.5 (5.1.2) Branch Lines to the Sampling System Q211.5-1 211.6 (5.1. 2) RHR System Q211.6-1 211.7 (5.1.2) Loop Seal Drains Q211.7-1 211.8 (6.2.2) Fuel Pool Cleaning and Cleanup Sys-tem Line Q211.8-1 211.9 (9.2.2.1) Cooling Lines Q211.9-1 211.10 (9.3.3) Nitrogen Gas Supply Q211.10-1 211.11 (9.4.8) Safety Classification of System Components Q211.11-1 211.12 (5. 2.1. 3 ) Quality Group A Components Q211.12-1 211.13 (5.2.1.4) Unacceptable Code Cases Q211.13-1 214.1 (4.4) Effects of 17 x 17 Geometry on DNB Calculations Q214.1-1 214.2 (5.2.2) Overpressure Protection Calculations Q214.2-1 214.3 (5.2.2) Pressurizer Safety Valves Q214.3-1 214.4 (6.3.4) Preoperational Testing of ECCS Q214.4-1 214.5 ~ (6.3) LOCA Analysis Q214.5-1 214.6 (15.2) Flow Coast Down Calculations Q214.6-1 214.7 (15.2.5) Initial Reactor Coolant Flow Rate Q214.7-1 214.8 (15.2.7) Primary Safety Valve Discharge Rates Q214.8-1 214.9 (15.3.6) Single RCCA Withdrawal Q214.9-1 214.10 (15.4.2.1) Location and Worth of Stuck Control Rod Q214.10-1 h 214.11 (15.4.2.1)DNBR Versus Time Curves Q214.11-1 214.12 (15.4.2.2) Steam Flow Rates Q214.12-1 AEC-vi 2344 183
WUP Amendment 18 PSAR 6/79 Question No. Subiect Page 214.13 (15.4.2.2) Trip Signal Generation Q214.13-1 214.14 (15.4.2.2) Steam Generator Level Q214.14-1 214.15 (15.4.2.2) Feedline Break Analysis Q214.15-1 214.16 (15.4.2.2) Auxiliary Feedwater Q214.16-1 214.17 (15.4.2.2) Pressurizer Pressure Q214.17-1 214.18 (15.4.2.2) Pressurizer Safety Valves Q214.18-1 214.19 (15.4.2.2) Auxiliary Flow Rate Heat Removal Capacity Q214.19-1 214.20 ( 15. 4. 3) Tube Rupture Accident Q214.20-1 214.21 (15.2.12) Pressurizer Heaters and Backup Heaters Q214.21-1 214.22 (1.5) Confirmatory Flow Model Tests Q214.22-1 214.23 (1.5) DNB Verification Test Q214.23-1 214.24 (1.5) Verification Tests Measuring Rod Drop Time Q214.24-1 214.25 (5.2.2) Feedline Rupture Analysis Q214.25 214.26 (5.2.2) Overpressure Protection Calculations Q214.26-1 214.27 (5.5.2) Preheater Box in Steam Generator Q214.27-1 218.28 (5.5.7.3) Inlet Isolation Valves Q214.28-1 214.29 (6.3) Field Adjusted and Stem Locked Valves Q214.29-1 214.30 (15.1) Accident Analysis Q214.30-1 214.31 (4.4) HYDNA Q214.31-1 214.32 (15.1.9) Computer Codes Q214.32-1 214.33 (15.2.4) Boron Dilution Anslysis Q214.33-1 214.34 ( 15. 4. 2) Steamline Break and Feedline Break Analysis Q214.34-1 214.35 (B . 3) Overpressure Protection Q214.35-1 214.36 (5.5.7) Residual Heat Removal System and AFW Supply to Ium Q214.36-1 214.37 (5.2.2) IEEE 279 Q214.37-1 (RSP) 214.38 (5.2.2) Seismic Qualification - Over- Q214.38-1 (RSP) pressurization Events 214.39 (5. 5.7) Equipment Required for Cold Shut- Q214.39-1 (RSP) down 214.40 (6.3) Remote Indication of Manual Valves Q214.40-1 (RSP) 214.41 (15.1. 5) Operator Action Consequences Q214.41-1 (RSP) 214.42 (5. 4. 6) Environmental Qualification for Q214.42-1 (5.4.7) (6.3) ECCS Pumps (RPS) 214.43 (3.5.1) Internal Missile Design Criteria Q214.43-1 221.1 (3.1.1) Single Failure Criteria Q221.1-1 221.2 (3.11.2) Safety Related Equipment Qualifi-cation Q221.2-1 221.3 (7.1) Standard Format Q221.3-1 221.4 (7.1. 2) Design Criteria for Instruments and
- - Control Systems Q221.4-1 221.% (7.5) Reactor Coolant Pump Coastdown Q221.5-1 221.6 (7.9) IEEE-Std-279-1971 Q221.6-1 AEC-vii
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WUP Amendment 18 PSAR 6/79 Ouestion No. Subiect Page k 221.7 (7. 2) Anticipatory Trips Q221.7-1 221.8 (16 . 4 .1. 4 ) Calibration Q221.8-1 221.9 (7.3) Injection to Recirculation Switch-over Q221.9-1 221.10 (7.3.2.2) GDC 37 Q221.10-1 221.11 (7. 3. 2 .1) Failure Mode and Effects Analysis Q221.11-1 221.12 (7. 3) Inadvertent Disabling of a Compo-nent Q221.12-1 221.13 (7.3) Redundant Safety Systems Q221.13-1 221.14 (7. 5) Recorders during a Seismic Event Q221.14-1 221.15 (7.5) Safety-Related Display Instru-mentation Q221.15-1 221.16 (7.6.2) Valves between RCS and RHR Systems Q221.16-1 221.17 (7.6) Auxiliary Feedwater System Q221.17-1 221.18 (8.1) Regulatory Guide Compliance Q221.18-1 221.19 (8.2) Offsite Power System Q221.19-1 221.20 (8.2) Gas Turbine Generators Q221.20-1 221.21 (8.2) Grid Stability Analysis Q221.21-1 221.22 (8.2) Interruption of Power to ESF Q221.22-1 221.23 (8 . 3.1.1) Emergency Bus Circuit Breakers Q221.23-1 221.24 ( 8 . 3.1. 2) Four kV Safety Bus Q221.24-1 221.25 (8. 3.1.1) Diesel Generators Q221.25-1 221.26 (8.3) Diesel Generator Qualification Q221.26-1 221.27 (8 . 3.1.1) ESF and Supporting Auxiliary System Q221.27-1 221.28 (8.3) Relay Trip Setpoint Drift Q221.28-1 221.29 (8.3) Thermal Overload Protection Q221.29-1 221.30 (8 . 3.1. 2) Cables in the Containment Q221.30-1 221.31 ( 8. 3. 2.1) D-C System Q221.31-1 221.32 (7. 6.2) Values Used between High and Low Pressure Systems Q221.32-1 221.33 (8. 3.1.1) Diesel Generator Protection Devices Q221.33-1 221.34 ( 8 . 3.1.1) Temperature Monitoring Q221.34-1 221.35 (8.3) Thermal Overload Protection Q221.35-1 221.36 ( 8 . 3. 2.1) Testing Requirements for D-C Systems Q221.36-1 221.37 Single RCCA Withdrawal at Full Power Q221.37-1 221.38 (3.1.1) Single Failure Provisions for Electrical Systems Q221.38-1 221.39 (3.1.1.2) Electrical Equipment Classification Q 221.39-1 221.40 (7.2) Anticipatory Trips Q221.40-1 221.41 (7.3) Injection Mode to Recirculation Mode Switchover Q221.41-1 221.42 (7.3.2.2) ECCS Testing Q221.42-1 221.43 (7.5) Post-Accident Display Instru-mentation Q221.43-1 221.44 (3.11) Safety-Related Display Information Q 221. 44-1 221.45 (10.3) Auxiliary Feedwater System Q221.45-1 221.46 (15.3.4) Underfrequency of Reactor Coolant 221.47 (App. B) Pump Breakers Seismic Qualification of Recorders Q221.46-1 Q221.47-1 l AEC-viii 0
WUP Amendment 18 PSAR 6/79 Question No. Sub-lect Page 221.48 (App. B) Conformance with Regulatory Guide 1.105 Q221.48-1 221.49 (App. B) Submittal of Docuinentation Required by IEEE 323-1974 Q221. 49-1 221.50 (App. B) Environmental Qualification of Electrical Equipment Q221.50-1 222.1 (RSP , B.3) Regulatory Guide 1.108 Q222.1-1 22 2.2 (A.1, Regulatory Guide 1.63
- 8. 3.1. 2) Q222.2-1 222.3 (B.3) Onsite Power Supply Load Shedding Q222.3-1 222.6 (8.3) (RSP) Testing of Onsite Power Sources Q222.6-1 222.7 (8.3) (RSP) Design of Penetration Overload Protection Q222.7-1 222.8 (8.2) Use of Combustion Turbine Generator Q222.8-1 240.1 (RSP ,B . 3) Loose Part Monitoring Q240.1-1 240.2 (6.2.1) Main Steam Line Break Analysis (Dry) Q240.2-1 241.1 (4.2) WCAP-8185 Q241.1-1 241.2 WCAP-8185 Q241.2-1 241.3 (4.2.1.3.1) Analysis Using Approved WCAP 8720 Q241.3-1 (RSP) 241.4 (4 . 2.1. 3.1) Effects of Rod Bow Q241.4-1 (RSP) 241.5 (4.2) (RSP) Analysis of Asymmetric Loads Q241.5-1 242.1 Limiting Power Distributions (4.3.2.2.6) Q242.1-1 242.2 (4. 3. 2. 7) Reactivity of Fuel Storage Facilities Q242.2-1 312.1 (3.11) Qualifications of Class IE Equipment Q312.1-1 312. 2 (RSP, Fuel Handling Accident 15.7.4, B.3) Q312.2-1 312.9 (3.11) Beta Radiation Qualification Requirements Q312.9-1 312.10 (App. B- Mitigation of Fuel Handling D.30) Accident Q312.*0-1 312.11 (16.3.8) Fuel Building Negative Pressure Criteria Q312.11-1 331.1 (12.1. 3) Dose Rate at Site Boundary Q331.1-1 331.2 (12.1.3) Reactor Core as a Source Term Q331.2-1 331.3 (12.1.3) T7rbine Building as a Zone III Region Q331.3-1 331.4 (12.1. 4) Check Intervals for Plant Area Monitors Q331.4-1 331.5 (12.2.4) Air Monitoring Q331.5-1 331.6 (12.2.4.3) Monitoring of Gaseous Iodine Concentrations Q331.6-1 331.7 (T12.2-12) Man-Rem Estimate Q331. 7- 1 331.8 (T 12.2-2) Concentration Calculations Q331.8-1 331.9 (F12.1-1 Layout Discrepancies Q331.9-1 through 12.1-12) 331.10 (12.1-2) Layout Drawing of Service Building Q331.10-1 331.11 (12.1-2) Control Panels in Waste Disposal Building Q331.11-1
% . AEC-ix
WUP Amendment 18 PSAR 6/79 Question No. Subiect Page 331.12 (12.1-3) Fuel Pool Area Location Q331.12-1 331.13 (12.1.2) Qualifications of Personnel Q331.13-1 331.14 Gamma Dose Resulting from Spent (12.1. 2. 3) Fuel Q331.14-1 331.15 ( 12.1. 4) Calibration Intervals Q331.15-1 331.16 Carbon Adsorber Cartridges (12. 2. 4. 2) Q331.16-1 331.17 Air Monitors Q331.17-1 331.18 (12. 2. 4) Plant Airborne Radioactivity Monitors Q331.18-1 331 19 (F12.1-6) Zone VI Radiation Zone Q331.19-1 331.20 Waste Baler (F 12.1-12) Q331.20-1 331.27 ' Service Building Controlled Area (F 12.1 -2 3) Arrangements Q331.21-1 331.22 (A.2) ALARA Program i331.22-1 331.23 (A.2) ALARA Responsibility Q331.23-1 331.24 (A.2) Radioactive Monitoring Systems Q331.24-1 331.25 (A.2) ALARA Dose-Reducing Techniques Q331.25-1 331.26 (A.2) Radiation Protection Personnel and Design and Review Q331.26-1 331.27 (A.2) Decommissioning Features Q331.27-1 331.28 (12.1.2) Personnel Protection during Fuel Transfer Q331.28-1 331.29 (12.1. 2) Cobalt Reduction and Control Q331.29-1 411.1 (17.1.1) QA Program Q411.1-1 411.2 (17.1.1.1) Superintendent of Quality Assurance Q411.2-1 411.3 (17.1.1.1) QA Personnel Q411.3-1 411.4 ( 17.1.1. 2) QA Program Q411.4-1 411.5 Nonconformance Reports (17.1.1.15) Q411.5-1 411.6 QA Program (17.1.1.18) Q411.6-1 411.7 (17.1. 2) Upgrading of QA Q411.7-1 411.8 (17.1. 3) Upgrading of QA Q411.8-1 411.9 (17.1.1) Management Establishing QA Q411.9-1 411.10(17.1.1) Statement of QA Policy Q411.10-1 411.11(17.1.1) Transmittal of QA Policy Q411.11-1 411.12(17.1.1) Resolution of QA Disputes Q411.12-1 411.13(17.1.'1.1) Training of QA Superintendent Q411.13-1 411.14(17.1.1) Quality Assurance Q411.14-1 411.15(9.5.1.1) Fire Protection Organization Q411.15-1 411.16(9.5.1.3) Fire Protection Quality Assurance Q411.16-1 411.17 (App. A) Commitment to QA Regulatory Guides Q411.17-i (17.1) (RSP) 411.18 (App. A) Regulatory Guide 1.38 Q411.18-1 (RSP) 412.1 (13.1.1. 4 ) Delegation of Responsibilities Q412.1-1 412.2 (13.1.2.1) Proposed Organization for One Unit Q412.2-1 412.3 ( 13.1. 2. 3) Shift Crew Composition Q412.3-1 412.4 (13.1.3) Personnel Q412.4-1 412.5 (13.1.2.3) Update of Fig. 13.1-3 Q412.5-1
'" ~*
2344 187
WUP Amendment 18 PSAR 6/79 Ouestion No. Subiect Page 413.1 (14.1) Personnel in Initial Test Program Q413.1-1 413.2 (14.1) Design Organizations Q413.2-1 413.3 (14.1) General Construction Related Prerequisites Q413.3-1 413.4 (14.1) Initial Test Programs Q413.4-1 413.5 (14.1) Abnormal Occurrence Reports Q413.5-1 413.6 (14.1) Scheduling of Test Programs Q413.6-1 413.7 (14.1) Trial-Use of Plant Operating and Emergency Procedures Q413.7-1 413.8 (14.1) Augmenting of Plant Staff during Initial Test Program Q413.8-1 413.9 (14.1) Unique Safety-Related Systems Q413.9-1 413.10 (14.1) Regulatory Guide Compliance Q413.10-1 413.11 (14.1) Initial Test Program Review for the FSAR A413.11-1 413.12 (14.1) Operating and Plant Staff Input to Test Program Q413.12-1 421.1 ( 13 . 7.1. 4) Employee Screening Program Q421.1-1 421.2 (13.7) Design Protection against Industrial Sabotage Q421.2-2 421.3 (RSP) Regulatory Guide 1.17 Q421.3-1 422.1 (13. 3) Koshkonong Emergency Plan Q422.1-1 422.2 (13. 3. 6) Hospitals Q422.2-1 430.1 (13.2) Number of Licensed Operators Q430.1-1 430.2 (13. 2) Operating Supervisor Q430.2-1 43G.3 (13. 2) Fire Brigade Training Program Q430.3-1 430.4 (13.2) Fire Protection Training Program Q430.4-1 431.1 (13. 2) Training Program Evaluation Q431.1-1 432.1 (13.5) Preliminary Schedule for Preparation of Procedures Q432.1-1 432.10 Offaite Fire Fighting Support - (13.3.4.2) Security Restrictions Q432.10-1 2344 188 AEC-xi
WUP Amendment 18 PSAR 6/79 OUESTION 042.48 (6.2) Closed systems outside containment (e.g. , the hydrogen recombiner system, the emergency core cooling system and the containment spray system) will constitute one of the redundant containment isolation barriers, and because of their post-accident function, will become extensions of the containment boundary followirq a LOCA. Since these systems, which will contain contaminated water or gas, include valves and pumps, they may become potential leakage paths for contaminated water or gaa outside of c( utainment. Therefore, discuss the test method ; :) that will be us d to gaantify the leak rates and propose a :est frequency comensurate with the method (s) of testing employed. Also, specify the leakage limit for each of these systems. RESPOJ9E Those closed systems outside containment (e.g., the combustible gas control system, the emergency core cooling system, and the containment spray system), which constitute one of the redundant containment. isolation barriers and become extensions of the boundary following a LOCA, will be tested by a method and frequency to provide assurance that the total potential leakage will be within acceptable limits. After a LOCA, the combustible gas control system (Section 6.2.5) is placed into operation. It then contains containment atmosphere and is considered part of the containment boundary. During Type A testing (Section 6.2.1.4) , both trains of the combustible gas control system will be exposed to the containment atmosphere, and any system leakage will become part of the overall integrated containment structure leakage test. Containment isolation valves for each train will be Type C tested (Section 6.2.1.4) . Since only a single train is required post-LOCA, and after containment pressure has been reduced, inclusion of both trains in the Type A containment structure leakage test plus the addition of the isolation valve leakage test (Type C) provides assurance that any system leakage would be within acceptable limits. If excessive leakage is detected following a LOCA, the leaking train can be isolated without impairing the ability of the combustible gas control system to limit the hydrogen buildup inside containment. The emergency core cooling system (ECCS) and the containment spray system (CSS) are waterfilled and operational following a LOCA. During the recirculation mode, contaminated water is drawn from the containment sump and recirculated through these systems. The maximum probable leakage for these systems is estimated at approximately 4.2 liters /hr (Tables 6.2.2-3 and 6.3-6). In analyzing the radiological consequences of the IOCA (Section 15.4.1.2) , doses were calculated based on a leak rate of 10 liters /hr for the portions of the ECCS and CSS systems outside Q042.48-1
WUP Amendment 18 PSAR 6/79 containment. This leak rate is greater than twice the maximum probable. In conjunction with the Type C tests on the system isolation valves (Section 6.2.1.4) , a visual inspection for leakage will be conducted with the system operational but isolated from the containment. For the low-head safety injection system (Figure 6.3-3) , one train will be isolated (the other train will remain operational) and receive suction from the refueling water storage tank (RWST) . The residual heat removal (RHR) purnp is started with water from the discharge of the RHR heat exchangers recirculated to the pump's suction. With the system at pressure, the system will be visually inspected for any leakage. The second train will be similarly inspected. The high head safety injection system (Figure 6.3-1) will also be visually inspected. Suction for one charging pump will be aligned to the RHR pump discharge. Valve alignments will be made such that the high head portion of the ECCS system outside containment can be pressurized and visually inspected for leakage. The remaining two charging pumps will be sequentially tested for seal leakage. Any component found leaking during this test will be repaired. The containment spray system will also be tested during the Type C test period. Each train will be isolated from containment and tested sequentially. Valve alignments will be as in the normal monthly pump flow test. The containment spray pump will be started. With the system at pressure, piping and components will be visually inspected for leakage. As for the ECCS, any leaking component will be repaired. These systems are in operation or are flow tested on a regular basis, and any component found leaking would be routinely fixed as part of the normal maintenance program. Components which may have a small amount of leakage as part of their normal operation (e.g . , pump seals) will be maintained to keep this leakage as low as is practicable. Total integrated leakage will b limited to 10 liters /hr. If excessive leakage is detected fol) . wing a LOCA, the leaking train can be isolated without impairing the effectiveness of the ECCS and PSS systems. Volatile species evolved from this leakage will be processed by the auxiliary building emergency filtration equipment (Section 6.5.1) to minimize releases to the environment. 2344 I90 0 Q042.48-2
WUP Amendment 18 PSAR 6/79 OUES' TION 042.49(6.2) PSAR Table 6.2.4-2 indicates that penetrations associated with the secondary system (main steam, main feedwater, auxiliary feedwater, etc.) are to be exempt from local leak testing. To permit an evaluation of the acceptability of this, provide justification that the steam generator tube bundles do not constitute a potential containment atmosphere leak path following a LOCA. In this regard, a water seal, established by reflooding the steam generators following a LOCA. may be shown to exist that will preclude containment atmosphere leakage. Therefore, discuss how a water seal can be established and maintained, considering single failures of active components. In addition, provide system drawings showing the routing and elevations of the steam generator piping.
RESPONSE
Those portions of the secondary system that penetrate the containment in a PWR satisfy the requirements of SRP 6.2.4, and are considered to fomm a closed system inside the containment. The valving arrangements outside the containment and the closed system inside the containment are part of the containment isolation system as considered in GDC 54 and 57. They must be designed to Seismic Category I requirements, be protected against high energy line breaks, and be able to close within specified times. Leak testing of PWR secondary system isolation valves is not required by 10CFRSO, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Piping associated with the steam generators, such as main steam, main feedwater, auxiliary feedwater, blowdown, and sampling system piping fall within the scope of GDC 57. These lines are typically provided with isolation valves which are capable of remote manual operation for the purpose of isolation. These valves also receive automatic isolation signals for other safety requirements. (Although these signals may be generated at the same time as containment isolation signals, they are not considered as such.) For large primary side breaks when an orderly shutdown cannot be executed, a plant trip will occur and the emergency core cooling system will be started. The reactor coolant system pressure will decrease at a rate dependent on break size. Sometime after the incident, conitairiment atmosphere will enter the tube bundle of the steam generator through the pipe break. During this initial period, approximately 1/2 to 6 hr after the incident, the steam generator shell side is pressurized, thereby preventing containment atmosphere leakage to the secondary system. After the short term, the barrier is maintained by a head of water above the tube bundle. This is the reason that secondary side Q042.49-1 2344 lhl
WUP Amendment 18 PSAR 6/79 isolation criteria (including isolation valve closure times and valve actuation signal logic) are determined considering limiting secondary side postulated accidents and not the LOCA. Redundant motor-driven pumps and a turbine-driven pump ensure adequate delivery of auxiliary feedwater to the steam generator. The main steam and feedwater lines for the Model F generat9r are routed such that draining of a steam generator is precluded. Due to the present stage of engineering, routing of the blowdown and sample lines is not available. However, these lines will be protected from missiles, pipe rupture, or jet impingement effects as a result of a break in the primary system. These criteria ensure that post-LOCA seal water in the broken loop steam generator will not be drained. 10CFR50, Appendix J, requires Type C leak testing of containment isolation valves that:
- 1. Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation, such as purge and ventilation vacuum relief, and instrument valves;
- 2. Are required to close automatically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation;
- 3. Are required to operate batermittently under post-accident conditions; and
- 4. Are in main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power reactors.
The secondary system isolation valves do not perform these functions and, therefore, are not included in Type C leak testing. 2344 i92 O Q042.49-2
WUP Amendment 18 PSAR 6/79 QUESTION 110.18 (RSP) (APP. B) The exception taken to position C.2.a. (2) and C.2.a. (4) of Regulatory Guide 1.121 in Appendix B, Section D.6 of the PSAR is unacceptable. The requirement for a 300 percent margin against burst failure based on normal operating pressure differential must be satisfied for all types of defects. This margin of safety may be demonstrated either analytically or experimentally. Test data submitted by Westinghouse for certain types of through wall defects has indicated that additional margin remained in the tube beyond the point where bulging occurs. A lower margin of safety may be applicable to this test data, provided it is shown that the remaining strength beyond bulging to gross rupture provides an equivalent margin of safety as required in Regulatory Guide 1.121. Therefore, provide additional information that substantiates the equivalency of the Westinghouse 200 percen t margin, based on Westinghouse performed tests, to the 300 percent margin required by the Regulatory Guide which utilizes a somewhat less conservative definition of tube failure. This equivalency must be justified for all types of tube defects.
RESPONSE
Westinghouse has documented its positions on Regulatory Guide 1.121 by corporate letter to the NRC, and has identified as the major exception the margin of 3 against tube failure for normal operation. Westinghouse defines tube failure as plastic deformation of a crack to the extent that the sides of the crack open to a nonparallel, elliptical configuration. The tubing can sustain added internal pressure beyond those values before reaching a condition of gross failure. Westinghouse has interpreted this to apply as an operating limit for the plant and considers that it introduces a conflict to the established conditions for plant operation as identified in the plant technical specifications. A factor of 3 is quite often used in ASME Code design guidelines. These ASME Code practices apply to the design of hardware and the analyses done on these designs. Conditions which occur during operation of the equipment, and which may affect the equipment so that design values no longer apply, are nnt directly addressed by the initial ASME Code requirements. That is one reason why plant technical specifications have been generated to establish safe limits of operation for power station equipment. The ASME Code is not directly applicable to the operational criteria of steam generator tubir.g. Our tubing design and tubing in the design condition have margins in excess of 3. However, we do not believe that this margin should be utilized as a limiting condition for normal operation. 2344 l93 Q110.18-1
WUP Amendment 18 PSAR 6/79 QUESTION 110.19 (APP. B) The information presented in Appendix B, Section D.22 is acceptable for ferritic bolts only. However, Section D.22 should address Regulatory Guide 1.124, Revision 1, dated January 1978 rather than the November 1976 issue. Therefore, revise Section D.22 to include an assessment of Revision 1 of Regulatory Guide 1.124.
RESPONSE
An assessment of Regulatory Guide 1.124, Rev. 1, dated January 1978 is provided in Item D.22 of Appendix B. 2344 194 Q110 19-1
WUP Amendment 18 PSAR 6/19 OUESTION 110.20 (RSP) ( APP. B) Exceptions No. 1, 2, 8, 9, 10, 11, and 12 in Appendix B, Section D.23 are not completely acceptable. Conformance to the rules of ASME Section III only provides assurance of the structural integrity of a component and does not necessarily guarantee the operability of an active component. Therefore, merely requiring deformation limits of a support structure in the Design Specifications may not assure that the supported active component will remain operable under all loading conditions. The staff position on deformation is that as a minimum, the support for an active component, i.e., a component required for safe shutdown of the plant or to mitigate the consequences of an accident, should not deform to the extent that the operability of the active component is impaired. Further, the fact that the words "most adverse" and " normal function" may not be well defined or may be ambigious is not, in the staff's opinion, acceptable justification for deleting entire sections of Regulatory Guide 1.130. Revision 1 of Regulatory Guide 1.130 dated October 1978 may satisfy some of the concerns expressed in Section D.23 of the PSAR. Therefore, revise Section D.23 by either committing to the positions in Regulatory Guide 1.130, Revision 1 which are associated with the above noted exceptions, or by providing sufficient design criteria to assure the staff that the above staf f position on deformation of supports will be satisfied.
RESPONSE
An assessment of Regulatory Guide 1.130, Rev. 1, dated October 1978 is provided in Item D.23 of Appendix B. 2344 195 Q110.20-1
WUP Amendment 18 PSAR 6/79 QUESTION 110.21 (RSP) (APP. B) Exceptions No. 5 and 7 in Appendix B, Section D.23 of the PSAR are unacceptable. As implied in Position C.3 of Regulatory Guide 1.130, Revision 1, the Staff Position is that for normal and upset plant condition loads, the Level A and B service limits are 50% of the critical buckling strength for flat plates and 33% of the critical buckling strength for shells. For loads associated with the f aulted plant condition, the Level D service limits should not exceed the limits specified in ASME Appendix F, of the critical buckling strengths for shells, unless justified by acceptable analysis and/or testing. Revise Exceptions 5 and 7 to include a commitment to the above position.
RESPONSE
Exceptions No. 5 and 7 in Appendix B, Item D.23 of the PSAR are based upon Rev. O of Regulatory Guide 1.130, dated July 1977. An assessment of Rev. 1 of that regulatory guide is provided in Item D.23 of Appendix B. 2344 i96 Q110.21-1
WUP Amendment 18 PSAR 6/79 QUESTION 121.10 1 21 Table 5.2-7 of the PSAR lists Reactor Coolant Pressure Boundary Materials for Class 1 Primary Components. List the specific location (s) for any ferritic material of pres sure-retaining components of the reactor coolant pressure boundary, other than: (1) Carbon and low alloy ferritic steel pipe, forgings, castings, and pipe with specified minimum yield strengths not over 50,000 psi, (2) Welds and weld heat affected zones in the materials specified in (1) above, (3) Materials for bolting and other types of fasteners with specified minimum yield strengths not over 130,000 psi.
RESPONSE
As indicated in Teble 5.2-7, SA 533 Grade A C2.ss 2, and SA 508 Class 2a materials, having specified minimum yield strengths greater than 50,000 psi, may be used in the fabrication of the Haven steam generators, pressurizer, and reactor vessel. Specifically, SA 533 Grade A Class 2 materials may be used in the following pressure-retaining applications: Steam generator shells, Steam generator heads, Pressurizer shell, Pressurizer heads, Reactor vessel shells (other than core region), Reactor vessel head. SA 508 Class 2a materials may be used in the following pressure-retaining applications: Steam generator nozzles, Steam generator manways, Steam generator tube sheets, Pressurizer nozzles, "'a a Pressurizer manways. 3NN )9[ Q121.10-1
WUP Amendment 18 PSAR 6/79 OUESTION 121.11 [RSP) (5.2.4) (5.4.3) The response to Question 121.9 is not adequate. Your response stated that "the requirements of Appendices G and H of 10CFR50 are met as discussed throughout Section 5.2.4 and 5.4.3" of the WUP PSAR and that "there are no areas of noncompliance" with these requirements. The information contained in Sections 5.2.4 and 5.4.3 is not adequate enough to determine if all of the requirements of these appendices will be met. It is our position that you make a commitment to fully comply with all of the requirements of Appendices G and H of 10CFR50.
RESPONSE
The WUP project will fully comply with the requirements of 10CFR50, Appendices G and H, as they exist on January 1, 1978. The text in Sections 5.2.4.1, 5.2.4.2, and 5.4.3.6 has been revised to clarify this commitment. 2344 198 Q121.11-1
WUP Amendment 18 PSAR 6/79 OUESTION 121.12 (5.2) In response to Question 121.7, you reference Topical Report WCAP-9292 ( Dynamic Fracture Toughness of ASME SA508 Class 2a and SA533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals" - March 1978). Review of this topical report has not yet been completed by the Staff. It should be noted, that if this topical report is found to be acceptable, it will only satisfy the generic requirements of Appendix G of 10CFR Part 50. This is, the adequacy of the subject materials to be describe d in the KIR curve of Appendix G of the ASME Code will have been demonstrated; however, the specific materials to be used in the Haven Nuclear Plant must meet all the fracture toughness requirements of 10CFR Part 50, Appendix G. Therefore, provide a commitment that this information will be supplied in the Haven FSAR.
RESPONSE
The response to Question 121.10 identifies the pressure retaining applications in which SA533 Grade A, Class 2 and SA508 Class 2a materials may be used. If SA533 Grade A, Class 2 or SA508 Class 2a is used in pressure retaining applications in the Haven steam generators or pressurizer, compliance with the applicable ASME Code
- fracture toughness requirements will ensure compliance with the 10CFR50, Appendix G fracture toughness requirements. Also, if SA 533 Grade A, Class 2 is used in pressure retaining Code *,
fracture toughness requirements and the 10 CFR50, Appendix G fracture toughness requirements will be satisfied. 2344 l?9
- In accordance with the requirements of 10CFR50.55a, the appli-cable ASME Code for the Haven pressure vessel which is part of the reactor coolant pressure boundary (i.e., reactor vessel, steam generators, and pressurizer) will be no earlier than the Summer 1972 Addendum of the 1971 Edition of the ASME Code Section III.
Q121.12-1
WUP Amendment 18 PSAR 6/79 QUESTION 130.25 (3.5.4) For the design of Seismic Category I concrete structures against tornado missiles, it is the staff's position that the wall and roof thicknesses should be in accordance with those indicated in for the applicable tornado region. (Tornado regions are described in Regulatory Guide 1.76.) Therefore, describe the extent to which your design will meet this position. The staff notes, that the minimum wall and roof thickness requirements are based on the missile spectrum specified in Table 3.5.2-1 of the WUP PSAR. On other applications, the staff has accepted an alternate missile spectrum which results in decreased wall thicknesses, but possibly increased reinforcing steel requirements .
RESPONSE
The design of Seismic Category I concrete structure against tornado missiles is based on the test data and design methods in Topical Report SWECO-7703, Missile-Barrier Interaction. This report was provided to the NRC on September 23, 1977. Appendix B of this report presents a design method for barrier thickness to prevent s cabbing. Section B.3.1 shows that this method is conservative when compared with Calspan-Bechtel tests and Sandia-EPRI tests. This topical report is used for the Haven tornado missile protection design rather than Attachment 1. Section B.4.1 shows that 20.6 in. of 4,000 psi concrete or 22.7 in. of 3,000 psi concrete will stop the most critical missile, the 743 lb, 12-in. diameter, 15-foot long schedule 40 steel pipe at a speed of 210 fps, without causing scabbing of the walls. For roof design for tornado missiles, the pipe is assumed to be traveling at 0.70 times the design horizontal speed, or 147 fps. A thickness of 16.5 in. of 4,000 psi concrete or 18.2 in. of 3,000 psi concrete is sufficient to prevent scabbing. Therefore, the minimum of 24 in. of 3,000 psi concrete provided for walls and roofs is conservative. This conservatism is increased by the fact that the strength of concrete during the life of the plant is higher than the specified strength at 28 days. 2344 200 Q130.25-1
WUP" Amendment 18 PSAR 6/79 ATTACHMENT 1 STRUCTURAL ENGINEERING BRANCH POSITION REQUIREMENTS FOR 'IORNADO MISSILE PROTECTION As an interim measure, the minimum concrete wall and roof thicknesses for tornado missile protection will be as follows: Concrete Strength ~ Wall Thickness Roof Thickness (psi) (in.) (in.) 3,000 27 24 Region I 4,000 24 21 5,000 21 18 3,000 24 21 Region II 4,000 21 18 5,000 19 16 3,000 21 18 Region III 4,000 18 16 5,000 16 14 NOTE: These thicknesses are for protection against local effects only. Designers must establish independently the thickness requirements for overall structural response. Reinforcing steel should satisfy the provisions of Appendix C, ACI 349 (that is, 0.2 percent minimum, EWEF) . 2344 201 O Q130.25-2
WUP Amendment 18 PSAR 6/79 OUESTION 130.26 (3.5.4) The Staff has developed criteria with respect to ductility of reinforced concrete and steel structural elements subject to impactive or impulsive loads as shown in Attachment 2. Your commitment to meet these criteria should be provided in addition to what is presented in Section 3.5.4 of the WUP PSAR.
RESPONSE
In the evaluation of overall response of structural elements subject to impactive loads, such as impacts due to missiles, the following are limits for maximum structural response: 1.0 REINFORCED CONCRETE MFJ4BERS , 1.1 For beams, slabs, and walls where flexure controls design, the permissible ductility ratio (p) under bnpactive and impulsive loads will be taken as y= 0.05 for p p' 2 0.005 p p' p= 10 for p p* s 0.005 where p and p* are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code. 1.2 For beam-columns, walls, and slabs carrying axial compression loads and subject to impulsive or impactive loads producing flexure, the permissible ductility ratio in flexure will be as follows: (a) When compression controls the design, as defined by an interaction diagram, the permissible ductility ratio will be 1.3. (b) When the compression loads do not exceed 0.1f'c Ag or one third of that which would produce balanced conditions, whichever is smaller, the permissible ductility ratio can be as given in Section 1.1. (c) The permissible ductility ratio will vary linearly from 1.3 to that given in Section 1.1 for conditions between those specified in (a) and (b). 1.3 For structural elements resisting axial compressive impulsive or impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3. 1.4 For shear carried by concrete only: p= 1.0 2344 202 Q130.26-1
WUP Amendment 18 PSAR 6/79 For shear carried by concrete and stirrups or bent bars: p= 1.3 For shear carried entirely by stirrups: p = 3.0 1.5 For members specifically designed for membrane tension: p = 0 . 5 cu, cy where: cu = uniform ultimate strain of rebar cy = strain at yield of rebar The tension mechanism for two-way slabs and beams anchored at their ends is used in barrier design when:
& 1. There is a second barrier between the primary barrier acting in tension and the missile protected area. The second barrier is designed to stop scabbing particles from the primary barrier.
- 2. The primary barrier is not required to carry other loads.
- 3. The missile geometry is such that it cannot slip between the rebar pattern.
- 4. The rebar is continuous in the barrier and fully developed in the barrier support.
- 5. The barrier supports are capable of providing the inplane force necessary for the tension mechanism.
2.0 STRUCTURAL STEEL MEMBERS 2.1 For flexure compression and shear: p = 10.0 2344 203 O Q130.26-2
WUP Amendment 18 PSAR 6/79 2.2 For columns with slenderness ratio (1/r) equal or less than 20: p = 1.3 where: 1 = effective length of the member r = the least radius of gyration For columns with slenderness ratio greater than 20 p = 1.0 2.3 For members subjected to tension: p = 0.5 cu cy where: cu = uniform ultimate strain of the material Ry = strain at yield of material 23$4 204
- r. i .
Q130.26-3
WUP Amendment 18 PSAR 6/79 ATTACHMENT 2 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTS SUBJECTED TO IMPACTIVE OR IMPULSIVE LOADS INTRODUCTION In the evaluation of overall response of reinforced concrete structural elements (e .g . , missile barriers, columns, slabs, etc.) subjected to impactive or hapulsive loads, such as impacts due to missiles, assumption on non-linear response (i . e . , ductility ratios greater than unity) of the structural elements is generally acceptable provided that the safety functions of the structural elements and those of safety-related systems and components supported or protected by the elements are maintained. The following summarizes specific SEB interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements subjected to bnpactive and impulsive loads. SPECIFIC POSITIONS 1.0 REINFORCED CONCRETE MEMBERS, 1.1 For beams, slabs, and walls where flexure controls design, the permissible ductility ratio (p) under impactive and impulsive loads should be taken as: p = 0.05 for p p' 2 0.005 p p' u = 10 for p p* s 0.005 where p and p* are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code. 1.2 If use of a ductility ratio greater than 10 (i .e . , p > 100) is required to demonstrate design adequacy of structural elements against impactive or impulsive loads (e .g. , missile impact) , such a usage should be identified in the plant SAR. Information justifying the use of this relatively high ductility value shall be provided for SEB staf f review. 1.3 For beam-columns, walls, and slabs carrying axial compression loads and subject to impulsive or impactive loads producing flexure, the permissible ductility ratio in flexure should be as follows: (a) When compression controls the design, as defined by an interaction diagram, the permissible ductility ratio shall be 1.3. O Q130.26-4 2344 205
WUP Amendment 18 PSAR 6/79 ATTACHMENr 2 (Cont) (b) When the compression load does not exceed 0.1fc Ag or one-third of that which would produce balanced conditions, whichever is smaller, the permissible ductility ratio can be as given in Section 1.1. (c) The permissible ductility ratio shall vary linearly from 1.3 to that given in Section 1.1 for conditions between those specified in (a) and (b). (See Fig. 1.) 1.4 For structural elements resisting axial compressive impulsive or impactive loads only, without flexure, Me permissible axial ductility ratio shall be 1.3. 1.5 For shear carried by concrete only p= 1.0 For shear carried by concrete and stirrups or bent bars p= 1.3 For shear carried entirely by stirrups p = 3.0 2.0 STRUCTURAL STEEL MEMBERS 2.1 For flexure compression and shear p = 10.0 2.2 For columns with slenderness ratio (1/r) equal or less than 20 p = 1.3 where: 1 = effective length of the member r = the least radius of gyration For columns with slenderness ratio greater than 20 p = 1.0 2344 206 Q130.26-5
WUP Amendment 18 PSAR 6/79 ATTACHMENT 2 (Cont) 2.3 For members subjected to tension p = 0.5 cu CY where: cu = uniform ultimate strain of the material cy w strain at yield of material 2344 207 e O O Q130.26-6
WUP Amendment 18 PSAR 6/79 OUESTION 130.27 (RSP) (3.7.1. 6) (3.7.2.5) It is noted that the soil-structure interaction for Seismic Category I structures founded on soil or soil backfill has been is the staffes evaluated by the finite element method. It position that the conventional method of lumped mass model, coupled by soil spring and dashpot, also be evaluated for these structures and that the floor response spectra used in the design, envelopes the response spectra obtained by these two methods. Therefore, provide a commitment that your final design of these structures will incorporate this position.
RESPONSE
Refer to Question 130.34 which supersedes this question. 2344 208 Q130.27-1
WUP Amendment 18 PSAR 6/79 QUESTION 130.28 (3.7.2) In your description of ihe Seismic Systems Analysis, you do not address the effects of the iailure of Noncategory I structures on Category I structures. Indicate that all Noncategory I structures are designed such that: (1) The collapse of a Noncategory I structure will not strike a Category I structure, or (2) Collapse of a Noncategory I structure will not impair the integrity of the impacted Category I structure, or (3) Noncategory I structures are designed and analyzed to Category I standards.
RESPONSE
Section A.1-1.29 (Appendix A) indicates conformance, as noted, with Regulatory Guide 1.29. This commitment indicates that all non-Category I structures will be designed such that: (1) The collapse of a non-Category I structure will not strike a Category I structure, or (2) Collapse of a non-Category I structure will not impair the integrity of the impacted Category I structure, or (3) Non-Category I structures are designed and analyzed such that they will withstand the effects of a Safe Shutdown Earthquake. 2344 209 Q130.28-1
WUP Amendment 18 PSAR 6/79 QUESTION 130.29 (3 . 8 . 3) (3.8.4) (RSP) In response to Qualification Review Item D.29, you indicate that ACI-349-76 is the appropriate industry standard to use in the design and construction of concrete structures other than the concrete containment. Through Regulatory Guide 1.142, the staff has endorsed the use of ACI-349-76 with a number of exceptions. Therefore, it is the staff's position, that ACI-349-76 may be used if the regulatory position delineated in Regulatory Guide 1.142 is complied with.
RESPONSE
The Applicants' compliance with Regulatory Guide 1.142 is discussed in Item A.1-1.142 of Appendix A. 2344 210 Q130.29-1
WUP Amendment 18 PSAR 6/79 QUESTION 130.30 (3.8.4) In your response to Qualification Review item D.41, you indicate that asymmetric loads on components within containment subcompartments are under investigation. Indicate if these loads are considered in the structural design of subcompartments. Specifically, describe the analytical and design techniques utilized to determine the effects of such loads on the shield wall surrounding the reactor vessel and indicate how these pressurization loads are combined with other coincident loads, including the seismic loads.
RESPONSE
As indicated in the response to Qualification Review item D.41, asymmetric pressure loads will not be developed further until detailed engineering of the plant begins after the receipt of a construction permit. The results of the work will be presented in the FSAR. The analysis and design of the shield wall surrounding the reactor vessel, and the analysis and design of other subcompartment walls will include the effects of asymmetric pressure loads. The effects of asymmetric pressure differences across the walls, and the component support reactions on the walls resulting from the asymmetric pressures will be considered. The dynamic nature of the loads, and the interaction of the radial walls and the shield wall will be considered in the analysis. The pressurization loads are combined with other coincident loads in accordance with Section 3.8.3.3 of the PSAR. 2344 2II Q130.30-1
WUP Amendment 18 PSAR 6/79 QUESTION 130.34 (RSP) (3.7.1.6) (3.7.2.5) In lieu of a response to Question 130.27, provide the following information. It is noted that the soil-structure interaction ior Seismic Category I structures founded on soil or soil backfill has been evaluated by the finite element method. It is the staff *s position, that the methods for implementing the soil-structure interaction analysis should include both the half-space lumped spring and mass representation and the finite element approaches. Seismic Category I structures, systems, and components should be C' signed to responses obtained by any one of the following methods: (1) Envelope of results of the two methods, or (2) Results of one method with conservative design consideration of impact from use of the other method, or (3) Combination of (1) and (2) with provision of adequate conservatism in design. Therefore, we request you to compare the responses obtained by the half-space (lumped parameter) approach at a few typical locations to those responses obtained by the finite element approach at these same locations. We request that you perform this limited comparison for all Seismic Category I structures founded on soil or soil backfill at the Haven Nuclear Plant. The three components of earthquake motion should be considered in the analyses. Floor response spectra should be provided at the basemat, an intermediate elevation and an upper elevation. For the lumped parameter representation, the variations of soil properties should be considered. If the results of the lumped parameter approach are different from the response obtained by the finite element approach, assess the safety significance of the differences in response.
RESPONSE
The soil-structure interaction for each soil-founded Category I structure will be determineet by the finite lement and continuum methods; a comparison of the results will be presented in the FSAR. Amplified response spectra will be obtained at the base mat, an intermediate floor, and at the roof of the structures involved. A comparison of the results of the two analyses will be made to assure that there is no aigniticant difference between the two design bases. 2344 212 Q130.34-1
WUP Amendment 18 PSAR 6/79 OUESTION 214.37 (RSP) (5.2.2) The response to Question 214.35, concerning equipment to mitigate low-temperature overpressurization events, indicates that exception may be taken to the system design requirements of IEEE 279 which are specified in our position. The response does not provide information identifying the exception (s) , nor does it provide justification for the exception (s) . Therefore, we require that any exception to IEEE 279 be identified and justified.
RESPONSE
The low temperature RCS overpressure mitigation circuits are shown on Fig. Q214.37-1 and Q214.37-2. The circuits are redundant. A temperature interlock is required to prevent inadvertent operation of a relief valve due to failure of a temperature transmitter. This permissive provides a cross connection between the redundant circuits. The criteria presented in IEEE-279-1971 are not required for the interlocks of the RCS pressure control associated with this system because the interlocks do not perform a protection function. These interlocks will meet the isolation criteria of IEEE-279-1971 that address control and protection system interaction (Section 4.7) . 2344 213 Q214.37-1
TRAIN A l TRAIN 8 RCS RCS RCS RCS RCS RCS RCS RCS TEMR TEMP TEMP TEMR TEMP TEMR LOOP 1 LOOP 2 LOOP 3 PRES' PRES. LOOP 1 OP 2 LOOP 3 I i
<r 1r i , ,,
I u , , 1 , 1, I I I i l 1 I l PROTECTION RACK l l l PROTECTION RACK I PROTECTION SYSTEM DV O h ' O v O'S
E l
I
\/ O h
__ ______g______, ,__ _,,_,_____ CONTROL SYSTEM 1r i r 1r 1r 1 r 1 r 1r 1r CONTROL RACK A CONTROL RACK B PERMISSIVE FOR PORV P PERMISSIVE FOR PORV A l 1r 1r 1 it AUXILI ARY AUXILI A RY RACK RACK N u , r 1 r A TO POWER OPER ATED TO POWER OPERATED A RELIEF VALVE A RELIEF VALVE B N FI G. Q 214. 37-1
~
" LOW TEMPERATURE RCG OVERPRESSURE MITIG ATION CIRCUIT WISCOllSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT 19
RCS TEMP. RCS PRES. RCS PRES. RCS TEMR TT TT TT TT TT TT LOOP 1 LOOP 2 LOOP 3 LOOP 1 LOOP 2 LOOP 3 i r 1r 1 r i r 1 r 1 r AU CTION EER AUCTIONEER LO LO 1 r 1 r 1 r 1 r P': f ( T) - - P': f ( T ) TB TB FUNCTION FUNCTION
~ -
GE NERATOR GENERATOR , xs P 1 r PB PB PB PB LOW TEMP PERMISSIVE A LOW TEMP. PERMISSIVE r b 1 r i t i t i r b I FIG. 0214.37-2 m LOW TEMPERATURE RCS OVERPRESSURE MITIG ATION CIRCUIT h WISCONSIN UTILITIES PROJECT TO POWER OPER ATED TO POWER OPER ATED PRELIMIN ARY SAFETY AN ALYSIS REPORT RELIEF VALVE A RELIEF VALVE B AMENDMENT 18
WUP Amendment 18 PSAR 6/79 QUESTION 214.38 (RSP) (S.2.2) In the response to Question 214.35, you propose a probabilistic evaluation of equipment to mitigate low-temperature over-pressurization events in lieu of meeting OBE criteria. Such a probabilistic assessment to satisfy our seismic criteria is unacceptable. We require that you comply with the seismic qualification provision of the position.
RESPONSE
The response to Question 214.35 proposes an evaluation to determine if an OBE will cause an overpressure event. This fluid systems evaluation will be performed to analyze the potential for overpressure transients following an OBE. The basis of the evaluation will assume the plant air system as inoperable since it is not seismically qualified. The results of the evaluation will be supplied at a later date. 2344 216 Q214.38-1
WUP Amendment 18 PSAR 6/79 QUESTION 214.39 (RSP) (5.5.7) The response to Question 214.36, concerning the capability to achieve and maintain cold shutdown, did not address the following criteria: (1) The plant must be capable of achieving RHR operational conditions in about 36 hours (assuming the worst single failure and loss of offsite power). (2) Test results specifically justifying the adequacy of boron mixing under natural circulation (assuming the worst single failure and loss of offsite power) must be provided. Discuss how you wil?. satisfy these criteria.
RESPONSE
(1) As explained in the response to Question 214.36, the safe shutdown design basis is hot standby (as it is for all Westinghouse pressurized water reactors) . Extended operation at hot standby is achieved by the service water-auxiliary feedwater cross connections, while manual actions are taken or repairs are made. The response to part (6) of Question 214.36 provided an abbreviated procedure for cooling down using natural circulation. It can reasonably be expected to reach RHR operational conditions within 36 hr from commencement of the cooldown. Thus, the capability of achieving RHR operational conditions in about 36 hr (assuming the worst single failure and loss of offsite power) does inherently exist in a manner consistent with the recommended implementation for Class 2 plants described in BTP RSBS-1, Rev. 1. The actual utilization of this cooldown capability would depend upon the circumstances and plant conditions at the time of such an event. (2) The NSSS and BOP designs permit a natural circulation boron mixing test to be conducted if required. Such a test will be performed during the first cycle of operation when sufficient decay heat is present, if a similar test has not already been performed on any other typical Westinghouse pressurized water reactor. 2344 2l7. Q214.39-1
WUP Amendment 18 PSAR 6/79 OUESTION 214.40 (RSP) (6.3) Inadvertent mispositioning of some emergency core cooling manual valves could compromise system performance. We require control room position indication for any manually operated emergency core cooling valve which, if kept in an incorrect position, would compromise system performance. Discuss your plans to satisfy this requirement.
RESPONSE
The Applicants will preclude mispositioning of those manual emergency core cooling valves which, if kept in an incorrect position, would compromise system performance through a proven system of administrative controls. This system consists of physically locking open the sensitive manual vaives and performing, on a monthly basis, a valve line up check and verification to assure that these manual valves remain locked open. This method of administrative controls has been successfully used at the Applicants' Point Beach Nuclear Plant. It should be further recognized that these manual valves are generally provided and used to facilitate system and component maintenance. When a component, such as a safety injection pump, is isolated for maintenance, plant procedures and practice require that the system be successfully tested following the maintenance. This provides further assurance that the system valve line up, including the manual valves, has been returned to a position which does not compromise system performance. Because of the provision discussed above, the Applicants have no plans to provide manual ECCS valve position indication in the control room. 2344 2l8 Q214.40-1
WUP Amendment 18 PSAR 6/79 QUESTION 214.41 (RSP) (15.1. 5) Following a steam line break and subsequent safety system response, reactor cooldown, and pressurizer refill, operator action may be necessary to prevent compromising reactor vessel integrity because of potential operation in an unacceptable pressure / temperature regime. We require that any manual actions (without credit before 10 minutes) be identified and reflected in Discuss your plans analyses predicting acceptable consequences. to satisfy this requirement.
RESPONSE
Generic vessel integrity analysis for a large steam line break has been performed for a three loop reactor vessel with material properties similar to those expected for the Haven plant. This analysis takes no credit for operator actions before 10 minutes. The results of this analysis show that reactor vessel integrity will be assured for the design life of the vessel. 2344 2J9 Q214.41-1
WUP Amendment 18 PSAR 6/79 ODESTION 214.42 (RSP) (5.4.6) (5.4.7) (6.3) We require assurance that the emergency core cooling pumps will perform their safety functions for an extended period of time following a LOCA under environmental conditions that prevail. Discuss your plans to satisfy this requirement.
RESPONSE
The charging pumps (high-head saf ety injection) and residual heat removal pumps (low-head safety injection) are located outside containment, and are not exposed to the severe environmental conditions inside containment that follow a loss of coolant accident. The equipment specifications for the charging pumps and residual heat removal pumps require them to be " capable of performing their long-term cooling function for 1 year". The same types of pumps have been used extensively in other operating plants. Their function during recurrent normal operations in such plants as Zion, D. C. Cook, Trojan, North Anna, and Farley has successfully demonstrated their performance capability. Periodic inspections have further confirmed their long-term operability. Nevertheless, design provisions are included that would allow maintenance on either pump if necessary during long-term operation. 2344 220 9214.42-1
WUP Amendment 18 PSAR 6/79 QUESTION _ 214.43(3.5.1) Safety-1 elated systems should be protected against loss of function due to internal missile impact. Pressurized components and rotating machinery are potential internal missile sources. These include the vessel seal ring and valve stems. Discuss your plans to consider valve stems, gravity missiles, and secondary missiles in the Haven design.
RESPONSE
As discussed in Section 3.5, saf ety-related systems are protected against loss of function due to internal missile impact. Pressurized components such as valve stems are not considered as credible sources of missiles, as discussed in Section 3.5. Notwithstanding this fact, protective barriers have been provided in the pressurizer area to preclude damage from postulated valve missiles due to the postulated failure of valve body-to-bonnet bolts. Refer to Appendix A, Item A.1-1.29, on Seismic Design Classification for a response to the concern for gravity missiles. The reactor coolant pump flywheel is not considered a source of missiles for the reasons discussed in Section 5.2.6. Whenever a primary missile is postulated, the need for a barrier to protect a Category I item will be evaluated. If a barrier is required and secondary missiles could be generated which would damage the functioning of a Category I item, then the barrier will be designed to prevent secondary missiles. 2344 221 Q214.43-1
WUP Amendment 18 PSAR 6/79 QUESTION 221.50 (APP. B) The response to Question 221.49 is not acceptable to the Staff. Exceptions have been taken to environmental qualification information requested during our qualification review and our first round questions. It is the Staff's position that sufficient information should be provided, prior to granting a construction permit, to assure that all safety-related equipment will be adequately qualified to perform its safety function. (1) As a minimum, we require that descriptions of the programs to be followed to environmentally qualify at least one item in each of the following groups of BCP and NSSS Class IE equipment be provided: (a) Switchgear (b) Motor Control Centers (c) Valve Operators (in containment) (d) Motors (e) Logic Equipment (f)' Cables (g) Diesel Generator Control Equipment (h) Sensors (i) Limit Switches (j) Heaters (k) Fans (1) Control Boards (m) Instrument Racks and Panels (n) Connectors (-o ) Penetrations (p) Splices (2) This qualification program should include: (a) Identification of equipment including:
- 1. Manufacturer, if available
- 11. Manufacturer's type number, if available iii. Manufacturer's model number, if available (b) Equipment design specification requirements ,
including:
- i. The system safety function requirements ii. An environmental envelope which includes all extreme parameters, both maximum and minimum values, expected to occur during plant shut-down, normal operation, abnormal operation and any design basis event.
iii. Time required to f ulfill its saf ety function when subjected to any of the extremes of the environmental envelope specified above. Q221.50-1 344
WUP Amendme2t 18 PSAR 6/79 O (c) Test plan, (d) Test set-up, (e) Test procedures, (f) Acceptability goals and requirements.
RESPONSE
Section 3.11.2 has been revised to include the program to environmentally qualify Class IE equipment in accordance with IEEE 323-1974. 2344 223 O O Q221.50-2
WUP Amendment 18 PSAR 6/79 QUESTION 222.6 (RSP) (8.3) Your response to qualification review item D-35, position 3 is inadequate in its present form. Revise your technical specification to include requirements for tests as set forth in our position 3, to demonstrate the full functional operability and independence or the onsite power sources once per 18 months during shutdown.
RESPONSE
Refer to revised Section 16.4.11, Amendment 18, for the' periodic testing requirements of the onsite power sources. 2344 224 Q222.6-1
WUP Amendment 18 PSAR 6/79 QUESTION 222.7 (RSP) (8.3) Your design of penetration overload protection is inadequate and, therefore, your response to qualification review item D-11 is incomplete. We require that the following requirements of IEEE-279 should be satisfied with regard to the protection of the electrical penetrations: (1) All source and feeder breaker overload and short circuit protection systems are qualified for the service environment including seismic. The seismic qualification for non-Class IE circuit breaker protection systems should as a minimum assure that the protection systems remain operable during an operating basis earthquake. (2) The circuit breaker protection system trip set points must have the capability for test and calibration. Provisions for test under simulated fault conditions should be provided. (3) No single failure shall cause excessive currents in the penetration conductors which will . degrade the penetration seals. (4) Signals for tripping source and feeder breakers shall be independent, physically separated, and powered from separated sources. Modify your design to include the above requirements.
RESPONSE
The design of the penetration overload protection is adequate to prevent a loss of containment integrity due to an overload-caused penetration seal failure. (1) All Class IE source and feeder breaker overload and short circuit protection systems are qualified for the service environment, including seismic. The non-Class IE breakers and their overcurrent trip systems are not separately qualified for seismic vibrations for the following reasons. a) The normal failure mode of a stored energy air circuit breaker during a seismic event is failure to remain closed. One of the most important aspects of seismic qualification of Class IE breakers is that they must not fail to the trip position during a seismic avent, thereby interrupting power to safety-related loads. b) The normal failure mode of relays during a seismic event is contact bounce, which wotGd result in a breaker trip not failure to trip. 2344 225 Q222.7-1
WUP Amendment 18 PSAR 6/79 c) It is expected that all penetrations will be able to withstand the fault cur: rent available at the penetration with the possible emeption of the 6.9 kV reactor coolant pump penetrations. 'Itese penetrations are protected by stored energy air circuit breakers which are controlled by relays, both the breakers and the relays would normally fail to the trip state if they failed at all during a seismic avent. d) It is not considered credible that a piece of equipment, which must withstand the forces generated when starting a reactor coolant pump, would fail and cause an overcurrent (during a seismic event) which simultaneously causes overcurrent relays to resist the magnetic forces acting to close their contacts, or causes stored energy air circuit breakers to fail to trip, while the power supply to the bus feeding the affected equipment is maintained. (2) The circuit breaker protection system trip set points have the capability for test and calibration. Provisions for test under simulated fault conditions are provided (Sections 8.3.1.1.1 and 8.3.1.1.4). (3) No single failure will result in overcurrents of sufficient magnitude and duration to cause mechanical failure of the penetration seals. This is discussed above and in Section 8.3.1.4.4. (4) All signals for tripping source and feeder breakers are independent and physically separated. The control power for the Class IE source and feeder breakers is from different Class IE batteries. The control power for non-Class IE source and feeder breakers is from separate circuits fed by the non-Class IE station battery. All control power circuits for these circuit breakers are monitored in the control room. 2344 226 O Q222.7-2
WUP Amendment 18 PSAR 6/79 QUESTION _ 222.8 (8.2L Section 8.2 of the WUP PSAR and the Haven Site Addendum discuss the possible use of a combustion turbine generator. You state that, "The combustion turbine generator is not a safety-related source of electric power. It is considered to be an offsite power source." You also state that, "this unit would be used as a backup supply to plant auxiliary loads, for black plant startup, system peaking, and standby reserve." Please respond to the following questions concerning the combustion turbine generator: (1) Section 16.3.6.5 of the WUP PSAR (Technical Specifications) states that the reactor shall not be made critical unless several conditions are met. One of these conditions is that two or more 345 kV transmission lines are in service. Explain how the combustion turbine generator could be utilized for a black plant startup when your proposed technical specifications prohibit this type of startup. (2) If the combustion turbine generator is going to be used as an "offsite power source" for the safety-related buses, . provide the following additional information for the unit: (a) The intended modes of operation, including the circumstances under which it will provide "offsite power"; (b) If known, the rating and specifications of the unit; and (c) If known, the testing which will be performed to assure its reliability. (3) Discuss the effect or provide assurance that the combustion turbine generator when operating and connected to the 69 kV bus will not effect the stability or reliability of the preferred source of offsite power.
RESPONSE
Sections 8.2 of the PSAR and the Haven Site Addendum have been revised to remove references to the combustion turbine generator as an offsite power source for the safety-related buses and to remove its use for black plant startup. The combustion turbine generator would continue to be used for system peaking and standby' reserve. When the combustion turbine generator is used, it will be synchronized to the system across its own circuit breaker. The generator protective relays, including overcurrent, generator differential, and reverse power, trip the generator breaker. As backup protection for both the generator and the 69 kV bus, bus overcurrent and differential relays trip the 69 kV breakers on both sides of the combustion turbine generator connection point. There are always a minimum of two circuit 9222.8-i 2344 227
WUP Amendment 18 PSAR 6/79 breakers lines tobetween the the combustion turbine generator and the 69 kV plant or the transformers feeding the bus. This arrangement prevents combustion turbine either primary or backup protection for the source. generator from affecting the preferred power 2344 228 ~ G O Q222.8-2
WUP Amendment 18 PSAR 6/79 OUESTION 240.2 (6.2.1) The methods used to generate mass and energy release data following a main steam line break are described in Westinghouse Topical Report WCAP-8860, " Mass and Energy Releases Following a Main Steam Line Break." This method assumes liquid entrainment for large breaks and dry steam for small breaks. WCAP-8860 has not been approved by the NRC. The mass and energy release data in WCAP-8860 is designed to be applicable for plants with Model D and Model 51 steam generators and the amount of entrainment predicted may not be applicable for Haven which utilizes Type F steam generators. Provide analyses of double ended main steam line breaks as a function of power level assuming no liquid entrainment from the break. Provide mass and energy release data as a function of time and also containment pressure and temperature as a function of time. Failure of an Engineered Safety Features train should be assumed since this is identified as the most severe single failure in Table 6.2.1-28 of the PSAR.
RESPONSE
Westinghouse will subnit to the NRC by July 30, 1979, mass and energy release data assuming entrainment which will be applicable to plants with Model F steam generators. The data will be , submitted as a supplement to WCAP-8860. It is anticipated that review of WCAP-8860 and the Model F supplement will be completed prior to the beginning of the environmental qualification program for WUP. Therefore, the environmental qualification program for equipment inside the containment will be based on the main steam line break analysis that includes the effect of liquid entrainment. This analysis is presented in Section 6.2.1.3.2.2. For information purposes, the requested analysis of double-ended breaks assuming no liquid entrainment is presented below. A spectrum of main steam line double-ended rupture accidents covering different break areas and reactor operating levels is analyzed. (Split ruptures are not included here since the mass and energy releases used for split ruptures in the analysis presented in Section 6.2.1.3.2.2 do not include liquid entrainment.) Mass and energy release data with no liquid entrainment for the Model'F steam generators for the limiting cases are Westinghouse Proprietary, and will be presented in a separate letter from S. Burstein, Executive Vice President, Wisconsin Electric Power Company, to H. Denton, Director, Nuclear Regulatory Commission. These mass and energy release rates are calculated with the Westinghouse MARVEL computer code. A description of the MARVEL analytical model can be found in WCAP-7909, MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System, October 1972. Additional calculations were performed with the mass and energy release data and pertinent balance-of plant parameters in order to obtain WUP mass and energy release rates. Q240.2-1 2344 229
WUP Amendment 18 PSAR 4/79 The method of balance-of plant parameters are given in Section 6.2.1.3.2.2. WUP mass and energy release rates for the limiting containment and pressure cases are presented in Tables Q240.2-2 and Q240.2-3, respectively. The contribution from each source is listed separately. The results of the analysis are presented in Table Q240.2-1 A failure of one emergency bus to energize, causing the loss of one engineered safety feature (ESF) train is assumed for this analysis. This was shown to be the most severe single failure for both contaimment pressure and temperature in the liquid entrainment analysis presented in Section 6.2.1.3.2.2. The initial containment pressure is assumed to be 14.7 psia for this spectrum of breaks. The case giving the highest peak containment pressure is repeated with an initial pressure of 16.2 psia. This was shown to produce the limiting case for peak containment pressure in the liquid entrainment analysis. The highest containment pressure results from a full DER at 70-percent power with an initial containment pressure of 16.2 psia. The highest containment temperature results from a full DER at 102-percent power with an initial containment pressure of 14.7 psia. The containment pressure and temperature transients for both limiting cases are shown on Fig. Q240.2-1 and Q240.2-2, respectively. 2344 230 i O Q240.2-2
WUP Amendment 18 PSAR 6/79 TABLE O240.2-1 CONTAINMEPTT PEAK PRESSURE AND TEMPERAWRE FOLLOWING A MAIN STEAM LINE DOUBLE-ENDED RUPTURE INSIDE CONTAINME?rf (No Moisture Entrainment, Failure of One ESP Train) Time of Time of Time of Time of Time Fan Time Power Break Initial Peak Peak Peak
- Peak Feedwater Steam Line Coolers Sprays level Area (8) Pressure Pressure Presaure Testp . Temp. Isolationaa3 Isolationta3 Startta3 Startta3 (5) fftal fpsia) (psic) (sect f*F) (sec) (sec) (sec) (sec) (sec) 102 Full 14.7 40.93 227 359.9 86 8.0 8.0 8.6 86.5 102 0.7 14.7 34.29 429 342.9 207 12.0 12.0 14.7 209 70 Full 14.7 41.07 287 357.3 85 8.0 8.0 8.5 85.5 70 Full 16.2 43.05 287 349.5 77 8.0 8.0 8.1 77.5 70 0.6 14.7 34.73 614 337.3 261 12.0 12.0 15.9 263 30 Full 14.7 40.94 429 352.9 82 8.5 8. 5 . 8.5 82.5 30 0.5 14.7 38.43 1,059 329.2 409 13.5 13.5 17.9 414 0 Full 14.7 41.04 659 347.4 92 8.0 8.0 8.4 93 0 0.2 14.7 27.93 3,789 310.1 1,919 13.0 146.5 53.8 1,929 NOTES:
tal For full DER, the forward flow area is 1.4 f ta and the reverse flow area is 4.87 fta. IN.) Subsequent to steam piping blowdown, the effective reverse flow area is 2.8 fta. U For smaller DER, the value listed is the flow area to each side of the break. 4 ta) Isolation signals are generated by the NSSS protection system for all cases except 4 the 0.2 fta DER at zero power. The steam line isolation signal for this case is generated by containment pressure. u ta) The times listed are when the system becomes effective for heat removal. 1 of 1
WUP Amendment 18 PSAR 6/79 TABLE Q21.0.2-2 MASS Em ENERGY PELEASE RA*ES _Nhb._ NbMS Full DTR at 70f.-Tower, Failure of One ECF Train, Initial Conteinment Pressure of 16.2 psia Limitirg Case for Containment Pressure (No Ffoistee Entrainment) Forward Flav frors Peverse Flow fron Reverse Flcra frera Rtmtured Log Steam Generstor Turbine Plant Piping _ Intact Stea:n Generators Total Time Ibss Rate Laergy aste Mass Rate Energy Este has Rate Energy hate Mass Rate Energy Rate (See) Qbm/see) Q d Etu/r.ecl (1te/see} QO6 Btu /sec{ (ib.1/see) Qa6_Btufay] Qbefsyj {106Btjsec} E: (1) This table contains proprietary data and has been forvarded to the Staff by S. Burstein, Executive Vice President, Esconsin Electric Power Company to R. Denton, Director, Office of Ecclear Regult. tion, Nuclear Regulatory Commission, letter of June 1979 N U b b N U N 1 of 1 O O O
VCP Amen &sent 18 PSAR 6/79 TABI2 Q2160.?-3 MASS AND PRERGY PEII.ASE RAITS
- ~
MAIN STEAM LIFE BREAK 1 Full DER at 10$ Power, Failure of One ESF Train, Initial Containment Pressure of 11.7 psia 6 Limiting Case for Containment Tenperature (No Moisture Entrainment) Forward Flow from Reverse Flow from Beverse Flow from _Ruytured Log Steam Generator Turbine Plant Piping Intact Steam Generators Total Time Mass Rate Enprgy Bate Mass Rate Er: Mass Rate Engrgy Rate M,as Rate Enfrgy Rate (See) Qtm/sec) QOuB,tu]sec{ _ (1_bm/sec) QO_grgyRateB,tu[ nee), Qtunfecc) (10,_B,tu/sec), (1,befrec{ ,(10" Btu /s_ec{ S* U (1) This table contains proprietary data and has been forwarded to the Staff by S. Burstein, Executive Vice President, Wisconsin Electric Power Company to H. Denton, Director, Office of Nuclear Regulation, Nuclear Regulatory Ccanaission, letter of June 1779 N U A A N U U 1 of 1
HOURS m_, ,,,,J0-',,,,,,,,40-',,,,,,,,J0-*, , , ,,,,il 0-', ,,,,,,,J0, o A i - m FULL DER AT 70% POWER FAILURE OF ONE ESF TRAIN
. /\\
\
y e m Pl= 16.2 PSI A
\ -
a FULL DER AT 102% POWER w FAILURE OF ONE ESF TRAIN w m Pl= 14.7 PSI A E // H o N
//
Z E m
~
5- - f o O h 1 0- "10' ' ' ' ' " 1 0' ' ' ' ' ' " 1 0' ' ' ' ' ' ' " 1 0' ' ' ' ' ' " 1 0' TIME AFTER ACCIDENT. SEC 2344 234 FIG. Q 240.2-1 CONTAINMENT PRESSURE STEAMLINE BREAK WlSCONSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT 18
HOURS 8 .,,,,,1 1 0-', , , ,,,,,i1 0-', , , , , , , , i10-*, , , , , , , , 10-*, ,
,,,,i10,
- FULL DER AT 102 % POWER FAILURE OF ONE ESF TRAIN
@ Pi=14.7 PSIA s O
f} o_ L ,- O__ . _ . i t E D o H o - N @ / FULL DER AT 70 % POWER
>[
to FAILURE OF ONE ESF TRAIN
- a. @ PI=16.2 PSI A o
8 10-# ' ' ' " 1 0" ' ' ' ' ' ' " 1 0' ' ' ' ' ' " 1 0* ' ' ' ' ' ' " 1 0' ' ' ' ' " ' 1 0' TIME AFTER ACCIDENT, SEC 2344 235 FIG. Q240.2- 2 CONTAINMENT TEMPERATURE STEAMLINE BREAK WISCONSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT 18
WUP Amendment 18 PSAR 6/79 QUESTION 241.3 (RSP) (4.2.1.3.1) In your reply to item A.2. (c) of our Qualification Review letter, your steady-state performance evaluation, calculations are performed using the improved Westinghouse analytical model, PAD-3.3. Our approval of this model, which will be the subject of separate correspondence, incorporates sane restrictions for its application. Therefore, provide a commitment that the steady-state performance evaluation for the Haven FSAR will be performed using the approved version of WCAP-8720 that will incorporate these restrictions.
RESPONSE
The steady state performance evaluation for the WUP FSAR will be performed using the approved version of WCAP-8720, and will incorporate the restrictions. WCAP-8720 is referenced in the WUP PSAR Section 4.2.4. 2344 236 Q241.3-1
WUP Amendment 18 PSAR 6/79 QUESTION 241.4 (RSP) (4.2.1.3.1) In your reply to item A. 2. (a) and D.31 of our Qualification Review letter, concerning fuel rod bowing, WCAP-8346 and WCAP-8692 are listed as references. In our letter from J. Stolz (NRC) to T. Anderson (W) dated June 19, 1978, we state that WCAP-8692 is no longer adequate for referencing and needs to be revised. WCAP-8346 had been previously withdrawn. Two interim procedures are available for treating the effects of rod bow on critical heat flux: (1) an acceptable model as described in an NRC memorandum from D.F. Ross and D.G. Eisenhut to D.B. Vassallo and K.R. Goller, dated February 15, 1977, (a copy of which will be forwarded to you under separate cover) and (2) procedures that can be used by an applicant as given in a letter from J. Stolz (NRC) to T. M. Anderson (W) , dated June 19, 1978. The effects of rod bow should be evaluated for the Haven FSAR using one of these methods or a method that might be developed by Westinghouse and approved by NRC.
RESPONSE
The DNB analyses in the PSAR of the Unit 1 17x17 core were performed such that generic DNBR margins described in the NRC's " Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors (Revision 1) February 16, 1977" are available for offsetting rod bow penalties. The appropriate rod bow penalty and any operating restriction in the technical specifications, if required, will be addressed prior to the issuance of the Operating License of this core. 2344 237 9241.4-1
WUP Amendment 18 PSAR 6/79 OUESTION 241.5 (RSP) (4.2) The analysis of asymmetric LOCA loads has been previously discussed in Qualification Review item D.41. Loads from seismic and LOCA events are being reviewed by NRQ as generic tasks (A-2 and B-6, NUREG-0371) . Since no plant structures would be affected by the outcome of this analysis, the results of an analysis that show that the fuel assemblies. can withstand this phenomenon and that coolable geometry is maintained should be provided in the FSAR. For the CP review, we only require that Wisconsin Electric Power commit to address this issue and provide the analysis in the Haven FSAR.
RESPONSE
Analytical results, with the overall safety factor that includes fuel integrity for the limiting LOCA transient, will be provided at the FSAR stage. 2344 238 Q241.5-1
WUP Amendment 18 PSAR 6/79 OUESTION 312.9(3.11) You have indicated that the engineered safety feature (ESF) equipment will be qualified to a post-LOCA gamma dose of 1.0 x 10e rads. Since the post-accident containment environment is expected to have a significant beta radiation component, we believe that all ESF components insidE the containment should be qualified to beta as well as gamma radiettion resulting from the postulated design basis accident.
RESPONSE
While it is true that the post-LOCA beta radiation component in the containment is significant, sufficient shielding is provided by equipment enclosures, conduit, and cable jackets to reduce the beta dose to safety-related components to a negligible amount. As shown in Table 3.11-2, the maximum calculated post-LOCA integrated gamma dose for 0-6 months is 3.2 x 107 rads. Since the mechanism by which the gamma dose is transmitted to the equipment is essentially the same as for beta radiation and the specified value of 1 x 10s rads is conservative, qualification of balance of plant equipment to gamma dose alone is sufficient. As discussed in Section 3.11, protective coatings (paint used in containment) which could be affected by a beta dose are qualified to a total integrated dose of 1.0 to 9.0 x 10S rads using a gamma source. For Westinghouse Electric Company supplied equipment, appropriate gamma and beta doses are specified in Supplement 1 to WCAP-8587 for safety-related electrical equipment. The beta doses specified are applicable only to exposed organic materials. 2344 239 Q312.9-1
WUP Amendment 18 PSAR 6/79 QUESTION 312.10 (APP. B) (D.30) Your response to Question 312.2 does not meet the objective of our position. While the calculated dose resulting from a fuel handling accident inside containment is well within the guidelines of 10CFR100, our objective in analyzing the consequences of such an accident is to assure that the plant design incorporates the capability for prompt mitigation of any radioactivity release. Therefore, it is our position that your design either: (1) Not allow purging during fuel handling inside containment; or (2) Mitigate the accidental release of activity during purging through appropriate sensing and valve response time; er (3) Mitigate the accidental release of activity by processing any releases through appropriately designed filters. Provide the necessary information, analyses, and drawings to show conformance with this position.
RESPONSE
In developing the response to NRC Question 312.2, an extremely conservative approach was utilized in calculating the dose resulting from a fuel handling accident inside containment. All of the gaseous activity from the ruptured fuel assembly was assumed to be instantaneously released to the environment in a puff. The resultant dose (Response to Item D-30, Appendix B) was well within the guidelines of 10CFR100. In actuality, several factors exist which would further mitigate this calculated dose. Realistically, the activity released would mix with the air inside the containment. The containment air recirculation coolers (Section 9.4.7. 3.2) are located above the operating floor (el 26 ft-4 1/2 in., Fig. 3.8.1-4) and take suction above the fuel handling area. Each unit has a normal flow rate of approximately 100,000 etm and provides an upward velocity with subsequent mixing of any release. The containment recirculation fan coolers and associated ring ductwork, which distribute conditioned air to locations above the operating floor, ara Seismic Category I. Nonseismic ductwork evenly distributes a portion of the conditioned air throughout the remainder of the containment structure. No outlets of this system are located in the immediate vicinity of the purge exhaust system inlets. During fuel handling, the normal air flow patterns are from the lower containment structure to the operating floor Q312.10-1 2344 240
WUP Amendment 18 PSAR 6/79 (Fig. 9. 4. 7-3) . Hence, any releases from the fuel handling area must be transported by the containment air recirculation system, and distributed by it throughout the containment structures. Therefore, there are no direct release paths to the environment. The containment purge supply system distributes fresh air above the operating floor, further promoting mixing of this volume. Suction inlets for the purge exhaust system are located next to the containment air filtration units, which are on the containment mat (el (-) 49 f t-7 in. , Fig. 3.8.1-1) , three levels below the operating floor. The containment penetration for the purge exhaust is located two levels below the operating floor at the (-) 10 ft-0 in. elevation (Fig. 3.8.1-2). Since no outlets for the containment air recirculation system are located near the exhaust system inlets, short-circuiting of the mixing mechanism does not occur. Any gaseous release from the fuel handling area would have to follow a torturous path prior to entering the e xhaust system (Section 9.4.7.2). Substantial mixing would also be obtained from the purge supplies and containment recirculation fan coolers
- air flow patterns above the operating floor.
The diluted gaseous release would be sensed by the Seismic Category I, Claps IE redundant radiation monitors located in the purge effluent line* (Qections 9.4.7.2 and 11.4.2.1.8) . High radiation signals from these monitors result in isolation of the containment purge supply and exhaust lines within 4 sec (Table 6.2.4-1 and Section 9.4.7.2). These containment isolation valves and their operators are Seismic Category I. Although greater containment mixing exists, it is assumed that the gaseous release mixes with only 1 percent of the con tainment air, resulting in the early detection of the release and termination of the purge with substantially all of the released activity retained inside the containment. Utilizing the parameters of 1 percent containment mixing , 20,000 cfm purge, and a 4-second valve closure time which includes instrument delays, the resultant dose is estimated to be one-twentieth of the dose calculated for the fuel handling accident inside containment in response to Item D-30, Appendix B. An identical reduction factor is estimated for the site-specific dose. The use of increased mixing inside containment and credit for lodine retention by the on-line containment air filtration carbon adsorber units would further reduce the calculated release and resulting dose. 1 e zm i t. c4) q Q312.10-2 .
WUP Amendment 18 PSAR 6/79 OUESTION 312.11 (16.3.8) We note that the minimum negative pressure to be maintained in the tuel building during fuel handling operations is not included in the Technical Specifications. It is our position that the fuel handling area be maintained at a negative pressure of 20.25 in. of water gauge, relative to the outside atmosphere, during fuel handling opeutions to preclude exfiltration of any untreated activity in the event of a fuel handling accident. Specify the minimum negative pressure which you will maintain in the fuel building during fuel handling operations.
RESPONSE
As stated in the revised ' Sections 6.5.2 and 9.4.5, the fuel building will be maintained at a minimum negative pressure of 0.25 in. of water gauge relative to outside atmosphere during fuel handling operations. 2';44 242 Q312.11-1
WUP Amendment 18 PSAR 6/79 QUESTION 331.28(12.1.2) Describe precautions taken to prevent inadvertent personnel access during fuel transfer to the very high radiation areas in the vicinity of the fuel transfer tube. If there is sufficient permanent shielding to assure acceptable levels in adjacent, potentially occupied areas, provide diagrams of that shielding.
RESPONSE
Personnel, both inside and outside containment, will be excluded from immediate proximity to the fuel transfer tube which will be enclosed in separately shielded cubicles. These areas will be locked and administratively controlled during refueling. Areas adjacent to the fuel transfer tube cubicles will either be administrative 1y controlled with locked access, or sufficiently shielded to reduce the dose rates to acceptable levels. At this time, the detailed design of the shielding has not been finalized, but will be provided in the FSAR. 2344 243 Q331.28-1
WUP Amendment 18 PSAR 6/79 OUESTION 331.29(12.1.2) Provide information concerning action taken to maintain occupational radiation exposure as low as is reasonably achievable by minimizing and controlling the buildup, transport, and deposition of activated corrosion products in reactor coolant and auxiliary systems. Include as a minimum, information on the following steps taken to minimize co-58 and Co-60, including: (1) The use of reduced nickel in primary coolant system alloys. (2) Low cobalt impurity specifications in primary coolant system alloys. (3) The minimization of high cobalt, hard facing wear materials in the primary coolant system. (4) The use of high flow rate /high temperature filtration. (5) The selection of valves and packing materials to minimize crud buildup and maintenance. (6) Provisions of decontamination of reactor coolant components and systems.
RESPONSE
The following considerations are given to minimize the buildup of Co-58 and Co-60 in the Reactor Coolant System: (1) The major source of nickel in the primary system alloys is in the steam generator tubing. Although Westinghouse maintains a continuing effort to evaluate alternate steam generator materials, Inconel 600 still appears to be one of the best. Examination of operating plants has shown that the released corrosion products from Inconel 600 have less nickel than the base metal. Also, Westinghouse utilizes the process of bright annealing, which reduces the corrosion tendency and minimizes the crud retention properties of steam generator tubes. (2) Materials with low cobalt content are specified for primary coolant system alloys. However, the availability, durability, and economics of component materials must always be considered during material procurement. (3) High cobalt, hard facing wear materials such as stellite, which come in contact with the primary coo' ant, are used only where substitute materials cannot meet performance requirements. 2344 244 Q331.29-1
WUP Amendment 18 PSAR 6/79 (4) The provision of high flow rate /high temperature filtration is not a standard practice in Westinghouse plants, and is not currently planned. (5) The selection of Westinghouse supplied valves is made in accordance with their functional requirements for various systems. Further detail regarding valves and their design description is given in Section 5.5. (6) The radiation fields associated with major equipment in the plants are frequently reduced by decontaminating the component. Reactor coolant pump impellers are routinely decontaminated when major maintenance is required. Decontaminating agents have been used for reactor coolant pumps consisting of an alkaline permanganate soaking followed by an oxalic acid-ammonium citrate treatment. The above considerations to minimize the buildup of radio-cobalt are discussed in further detail in WCAP-8872, Nesign, Inspection, Operation, and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures As Low As Reasonably Achievable", April 1977. 23dk 2 O Q331.29-2
WUP Amendment 18 PSAR 6/79 QUESTION 411.17 (RSP) (APP. A) (17.1) The response to Question 411.14 is incomplete. It is the Staff's position that the PSAR for the Haven Nuclear Plant must include a commitment to: (1) Comply with the requirements of ANSI N45.2.5 (Draft 3, Rev. 1, January 1974) or meet the regulatory position of Regulatory Guide 1.94 (April 1975 to April 1976); (2) Comply with the requirements of ANSI N45.2.8 (Draft 3, Rev. 3, April 1974) or meet the regulatory position of Regulatory Guide 1.116 (June 1976 or May 1977); (3) Comply with the requirements of ANSI N4542.12 (Draf t 3, Rev. 4, February 1974) ; and - (4) Comply with the requirements of ANSI N45.2.13 (Draf t 2, Rev. 4, April 1974) or meet the regulatory position of Regulatory Guide 1.123 (October 1976 or July 1977) , Please provide such commitments or propose alternatives for the Staff's evaluation. We note, too, that page 17.1-12a still refers to ANSI N45.2.13-1974. This reference should be changed to agree with item 4 above.
RESPONSE
(1) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.94, Rev. 1, dated April 1976, subject to the alternatives listed in SWSQAP 1-74A, Appendix VII. (2) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.116, Rev. O-R, dated June 1976. (3) The quality assurance program described in Chapter 17 commits to comply with the requirements of ANSI N45.2.12, Draft 3, Rev. 4, February 1974, Requirements for Auditing of Quality Assurance Programs for Nuclear Poder Plants, subject to the alternatives listed in SWSQAP 1-74A, Appendix VII. (4) The quality assurance program described in Chapter 17 commits to meet the regulatory position of Regulatory Guide 1.123, Rev. 1, dated July 1977, subject to the alternatives listed in SWSQAP 1-74A, Appendix VII. The reference to ANSI N45.2.13-1974 on page 17.1-12a will be deleted. The reference shall be changed to agree with response item (4) above. 2344 246 Q411.17-1
WUP Amendment 18 PSAR 6/79 QUESTION 411.18 (RSP) (APP. A) The proposed enceptionAnodification relating to position C.1.b of Regulatory G5ide 1.38 (page A.1-11 of the PSAR) is unacceptable. It is the staff's position that rerating of hoisting equipment should be done in accordance with the Regulatory Guide unless adequate justification (e .g. , an appropriate discussion and analysis of design margin) is provided. Please delete the proposed exception / modification or provide such justification.
RESPONSE
Appendix A, Item A.1-1.38 has been updated to delete the exception to position C.1.b. Rerating of' hoisting equignent will be performed in accordance with the Regulatory Guide. 2344 247 Q411.18-1
HUP Amendment 18 PSAR 6/79 QUESTION 412.5 (13 .1. 2 . 3) You indicate in your answer to Question 41.1 that the plant staff numbers shown on Figure 13.1-3 are for two unit operation. Therefore, since the Haven site is a single unit, please update Figure 13.1-3, and other areas if needed, to reflect that f act.
RESPONSE
Figure 13.1-3 and Section 13.1.2 have been revised to reflect the plant staff numbers for single-unit operation. 2344 248 Q412.5-1
WUP Amendment 18 PSAR 6/79 QUESTION 430.4 (13.2) Your response to Question 430.3 is not complete. The commitment to conduct an initial fire protection training program does not meet the requirements of Standard Review Plan Section 13.2. To meet these requirements, provide the following: (1) A commitment to complete the initial fire protection training prior to receipt of fuel at the site; (2) The provisions that have been made for the indoctrination of construction personnel; and (3) A detailed description of the training program for the individual (s) responsible for the formulation and assurance of the fire protection program. A commitment to meet the requirements of Section A.1 of Appendix A to Branch Technical Position ASB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976," may be made in lieu of suhaitting a program description.
RESPONSE
(1) Initial fire protection training will be completed prior to receipt of fuel at the site. (2) Construction personnel will be indoctrinated in accordance with the Constructor's fire protection program. The Applicants' System Fire Protection Officer will audit the Constructor a s implementation of his program. (3) The Applicants will meet the requirements of Section A.1 of Appendix A to Branch Technical Position ASB 9.5-1. 2344 249 Q430.4-1
WUP Amendment 18 PSAR 6/79 OUESTION 432.10 (13 . 3. 4 . 2 ) In your discussion of Fire Emergency on page 13.3-4, your plans would appear to exclude offsite fire fighting support from restricted areas. Please clarify your intent in this regard.
RESPONSE
It is Applicants
- intent to request fire fighting support from the local fire department for fire situations within the restricted area in the event of need. To accournodate this support, the local fire department is instructed in the operational precautions and special hazards associated with fighting fires on nuclear power plant sites (Section 9.5.1.2) .
At least one fire drill per year will involve the local fire department (Section 13.2.1. 6) . It is Applicants' intent to rely entirely upon the local fire department to handle fire situations within the exclusion area but outside of the restricted area. 2344 250 Q432.10-1}}