ML20150C149
| ML20150C149 | |
| Person / Time | |
|---|---|
| Site: | 05000502 |
| Issue date: | 10/30/1978 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20062C813 | List: |
| References | |
| NUDOCS 7811140171 | |
| Download: ML20150C149 (3) | |
Text
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O Wisconsin Electnc vowcecome 231 WEST MICHIGAN, MILWAUKEE, WISCONSIN 53201 October 30, 1978 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington,.D.~ C. 20555
Dear Mr. Denton:
DOCKET N0. STN 50-502 AMENDMENT 16 TO THE PRELIMINARY S?FETY ANALYSIS REPORT WISC0tlSIN UTILITIES PROJECT Wisconsin Electric Power Company, Wisconsin Power and Light Company, and Wisconsin Public Service Corporation submit herewith three signed originals and sixty (60) copies of Amendment 16 to the Preliminary Safety Analysis Report filed in the above docket.
Very truly yours, 6
Sol Burstein Executive Vice President Enclosures .
78///%5/ y 50 402- K
$6 '(oD 7 ; 1 > .
. _ . .._ - . . . _ _ . . . _ . - . . . . _ _ _ _ .. ~ ..._.. _._. . _. . . . . . _ . . . . . . _ . . _ . - . _ . , . _ . _ . _ _ . . . _ _ . _
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of WISCONSIN ELECTRIC POWER COMPANY, Docket No.
WISCONSIN POWER AND LI.GHT COMPANY, and STN 50-502 WISCONSIN PUBLIC SERVICE CORPORATION AMENDMENT N0. 16 PRELIMINARY SAFETY ANALYSIS REPORT 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 first round requests for additional information. These requests were provided to Applicants as enclosures to Mr. Olan D. Parr's letters dated July 18 and August 3,1978. Applicants have also included several additional responsas to Mr. Roger S. Boyd's letter dated April 5,1978, concerning the Nuclear Regulatory Commission's qualification review of the PSAR.
This amendment also contains a revision to the General Information Section as filed in this Application. This revision amends the application to request a construction permit and all necessary licenses required in connection with the ownership, use, and operation of one nuclear power plant. The revision to the General Information Section also includes a revised Exhibit D concerning the total estimated cost of the proposed facility and updates the listing of principal officers and directors for each of the Applicants.
The changes mentioned herein are contained on the replacement pages attached hereto and made part hereof, which pages are to be inserted in the PSAR and General Information Section .n accordance with the accompanying instructions.
Dated October 30, 1978.
Respectfully submitted, WISCONSIN ELECTRIC POWER COMPANY r - -.
s - .. , . _ , . . . . ,
STATE OF WISCONSIN) ss.
MILWAUKEE COUNTY )
S0L BURSTEIN, being first duly sworn, on oath says that he has read the foregoing statement and knows the contents thereof, that the same 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 make this verification for and on its behalf.
s Subscribed and sworn to before me this 30th day of October 1978.
oM A N kM/et/ % ~
Notary Public, State of Wisconsin My Commission expiresAA< 6,,NBo y
~i . t
l o O WISCONSIN Electnc powra couesur 231 WEST MICHIGAN, MILWAUKEE, WISCONSIN 53201 October 30, 1978 ;
1 l
1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555
Dear Mr. Denton:
DOCKET N0. STN 50-502 AMENDMENT 16 TO THE PRELIMINARY SAFETY ANALYSIS REPORT WISCONSIN UTILITIES PROJECT Wisconsin Electric Power Company, Wisconsin Power and Light Company, o
and Wisconsin Public Service Corporation submit herewith three signed originals and sixty (60) copies of Amendment 16 to the Preliminary Safety Analysis Reoort filed in the above docket.
Very truly yours, Sol Burstein Exect/tive Vice President Enclosures O 7 m 99/ya Jb.fot-K
'BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of WISCONSIN ELECTRIC POWER COMPANY, Docket No.
WISCONSIN POWER AND LIGHT COMPANY, and STN 50-502 WISCONSIN PUBLIC SERVICE CORPORATION AMENDMENT N0. 16 PRELIMINARY SAFETY ANALYSIS REPORT 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 Rtport (PSAR), filed as part of the Application for Licenses in this docket. This amendment consists of Applicants' responses to the Nuclear Regulatory Commission's first round requests for additional information. These requests were provided to Applicants as enclosures to Mr. Olan D. Parr's letters dated July 18 and August 3,1978. Applicants have also included several O additional responses the Nuclear Regulatory to Mr. Rogerqualification Commission's S. Boyd'sreview letterofdated the PSAR. April 5,1978, concerning This amendment also contains a revision to the General Information Section as filed in this Application. This revision amends the application to request a construction permit and all necessary licenses required in connection with the -
, ownership, use, and operation of one nuclear power plant. The revision to the General Information Section also includes a revised Exhibit D concerning the total estimated cost of the proposed facility and updates the listing of principal officers and directors for each of the Applicants.
The changes mentioned herein are contained on the replacement pages attached hereto and made part hereof, which pages are to be inserted in the PSAR l and General-Information Section in accordance with the accompanying instructions.
Dated Octsber 30, 1978.
Respectfully submitted, j . WISCONSIN ELECTRIC POWER COMPANY I
o
i t
STATE OF WISCONSIN) ss.
MILWAUKEE COUNTY )
I
! SQL BURSTEIN, being first duly sworn, on oath says that he has read the foregoing statement and .knows the contents thereof, that the same 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 make this verification for and on
- its behalf.
l l
i O - - - - B Subscribed and sworn to before me this 30th day of October 1978.
d'&ft>& ?)'WClo d=s -
Notary Puf/lic, State of Wisconsin
~
My Commission expiresAd< 6, /v8o o
O
WUP Amendment 16 PSAR 11/78 INSTRUCTIONS FOR MAKING PAGE CHANGES Amendment 16 material is dated November 1978 and printed on green stock.
The following list of material furnished as Amendment 16 serves as a checklist for entering new pages. Enter the revised pages as listed, discarding superseded material.
Remove Old Page Insert New.Page (Front /Back) (Front /Back)
Volume 1 Volume 1 Before WE letter to Harold R. Denton for Amendment 15, insert:
WE letter to Harold R.
Denton for Amendment 16/
blank USNRC Attachment - Amend-O ment 16/ blank Sol Burstein Affadavit/ blank
_ Before General Table of Contents, insert:
MEP.-1/2 iii/iv iii/iv EP.1-1/2 EP.1-1/ blank EP.2-1/ blank EP.2-1/ blank EP.3-1 thru EP.3-5 EP.3-1/2 3-xix/xx 3-xix/xx 3.8-19/20 3.8-19/20 3.8-29/30 3.8-29/30 3.11-1/2 3.11-1/2 3.11-2a/ blank 3.11-2a/ blank 3.11-4a/4b 3.11-4a/4b 3.11-7/ blank 3.11-7/ blank Behind T3.9-7 (2of 2) , insert: l
- T3.11-1 l
- T3.11-2 l
l EP.3A-1/ blank EP.3A-1/ blank L
l 1.
WUP Amendment 16 PSAR 11/78 Remove Old Page Insert New Page 1 (Front /Back) (Front /Back l l
Volume 2 Volume 2 iii/iv iii/iv l EP.4-1/2 EP.4-1/ blank j EP.5-1/2 EP.5-1/ blank 5.2-35/36 5.2-35/36 Volume 3 Volume 3 iii/iv iii/iv EP.6-1 thru 6.1-5 EP.6-1/2 6-vii/viii 6-vii/viii 6-ix/x 6-ix/x 6.3-25/26 6.3-25/26 6.3-27/28 6.3-27/28 6.5-3/4 6.5-3/4 6.5-4a/ blank 6.5-9/10 6.5-9/10 6.5-11/ blank Behind T6.3-9, insert: O T6.5-1 thru 6.5-3 EP.7-1 thru EP.7-3 EP.7-1/ blank Volume 4 Volume 4 iii/iv iii/iv EP.8-1/2 EP.8-1/ blank 8-iii/iv 8-iii/iv 8.3-3/4 8.3-3/4 8.3-15/16 8.3-15/16 8.3-19/20 8.3-19/20 8.3-21/22 8.3-21/22 8.3-23/24 8.3-23/24 8.3-24a/24b 8.3-31/32 8.3-31/32 Behind F8.3.1-4, insert:
F8.3.1-5 and 8.3.1-6 EP.9-1 thru EP.9-4 EP.9-1/2 9-xiia /b 9-xiia /b 9-xix/xx 9-xix/xx 9.3-9/10 9.3-9/10 9.5-27/ blank 9.5-27/28 2.
WUP Amendment 16 PSAR 11/78 Remove Old Page Insert'New Page (Front /Back) (Front /Back 9.5-28/29/30/ blank 9.5-29/30/ blank T9.3.2-1 T9.3.2-1 Behind F9.5.1-4, insert:
- F9.5.1-5 Volume 5 volume 5 iii/iv iii/iv EP.10-1/2 EP.10-1/ blank 10-vii/ blank 10-vii/ blank 10.2-7/ blank 10.2-7/ blank F10.4.8-1 F10.4.8-1 EP.11-1/2 EP.11-1/ blank T11.1-1A (4 pages) T11.1-1A (4 pages)
T11.1-1B (2 pages) T11.1-1B (4 ' pages)
() Volume 6 Volume 6 iii/iv iii/iv EP.12-1/2 EP.12-1/ blank 12-i/ii 12-i/ii 12-v/vi 12-v/vi 12.1-5/6 12.1-5/6 12.1-7/8 12.1-7/8
- 12.1-8a/8b 12.1-17/ blank 12.1-17/ blank 12.3-1/2 12.3-1/2
- 12.3-2a/ blank F12.1-1 F12.1-1 F12.1-6 F12.1-6 F12.1-7 F12.1-7 F12.1-8 F12.1-8 F12.1-9 F12.1-9 F12.1-10 F12.1-10 F12.1-11 F12.1-11
- F12.1-11A F12.1-16 F12.1-16 F12.1-17 F12.1-17 F12.1-18 F12.1-18 F12.1-19 F12.1-19 3.
WUP Amendment 16 PSAR 11/78 Remove Old Page Insert New Page (Front /Back) (Front /Back F12.1-20 F12.1-20 F12.1-22 F12.1-22 EP.13-1/2 EP.13-1/ blank 13.3-3/4 13.3-3/4' EP.14-1/ blank EP.14-1/ blank EP.15-1 thru EP.15-3 EP.15-1/2 Volume 7 Volume 7 iii/iv iii/iv EP.16-1/2 EP.16-1/ blank EP.17-1/2 _?.17-1/ blank 17.1-12a/ blank 17.1-12a/ blank 17.1-29/30 17.1-29/30 17.1-33/ blank 17.1-33/ blank EP.A-1/ blank EP.A-1/ blank A.1-7/8 A.1-7/8 A.1-9/10 A.1-9/10 A.1-11/12 A.1-11/12 A.1-23/24 A.1-23/24 A.1-25/26 A.1-25/26 A.1-29/30 A.1-29/30 A.1-33/34 A.1-33/34 A.2-1/2 A.2-1/2 A.2-3/ blank -
EP.B-1/ blank EP.B-1/ blank B.3-11/12 B.3-11/12 B.3-13/14 B.3-13/14 B.3-23/24 B.3-23/24 B.3-24a/ blank Volume 8 Volume 8 iii/iv iii/iv EP.Q-1 thru EP .Q-6 EP.Q-1 thru EP.Q-3 AEC-i thru AEC-ix AEC-i thru AEC-ix QO10.10-1/ blank QO10.11-1/ blank 0020.57-1/2 Q042.47-1/ blank 4.
l WUP Amendment 16 PSAR 11/78 Remove Old Page Insert New Page (Front /Back) (Front /Back Volume 9 Volume 9 iii/iv iii/iv l
Q121.6-1/ blank Q121.7-1/ blank Q121.8-1/ blank Q121.9-1/ blank Q214.35-1 thru 214.35-3 Q214.36-1 thru 214.36-5
- Behind Q221.46-1, insert:
Q222.1-1/ blank Q222.2-1/ blank Q222.3-1/ blank Behind TAB 240, insert:
Q240.1-1/ blank Behind Q242.1-1/ blank, l
l insert:
I - TAB 312 Accident Analysis Branch Q312.1-1/ blank l
Q312.2-1/ blank Q331.7-1/ blank Q331.7-1/ blank Q331.22-1/ blank i
1 Q331.23-1/ blank l
Q331.24-1/ blank Q331.25-1/ blank Q331.26-1/ blank Q331.27-1/ blank Q411.14-1/ blank Q411.15-1/ blank Q411.16-1/ blank l .
O 5.
WUP PSAR Amendirent 16 HAVEN 11/78 PRELIMINARY SAFETY ANALYSIS REPORT k
MASTER LIS"' OF EFFECTIVE PAGES l
The Lists of Effective Pages for the PSAR are compiled for each chapter and appendix in Volumes 1 throu3h 7 and the NRC Questions and Responses in Volumes 8 and 9. j The Master List of Ef f ecti'/e Pages presents the dates of issue for each amendment, and the revision numoer of the General Table of Contents (found at the '
front of each volume) , and each list of eff ective pages (found immed4.at *1y af ter the respective tab) .
Iseue Date Issue Date
- l Original S/23/74 Amendment 8 . 6/75 Amendment 8/9/74 Amendment 9 9/8/75 Amendment 1 9/27/74 Amendment 10 10/6/75 '
Amendment 2 9/22/78 Amendment 11 11/24/75 Amendment 3 11/78 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 Amenament 6 5/2/75 Amendment 15 9/22/78 Amendment 7 6/6/75 Amendment 16 11/78 Chapter 8 General Table of Contents EP.8-1 16 1/ii -
( iii iv 16
@ pter 9 !
i
[ v -
EP.9-1/2 16 i vi 14 Cnapter 10
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i Lists of Effective Pages l EP.10-1 16 !
Chapter 1 )
Chapter 11 l EP.1-1 16 !
EP.11-1 lb l Chapter 2 i c'napter 12 l EP.2-1 16 EP.12-1 1b Chapter 3 Che
~pt.er 13
, EP.3-1/2 16 EP.13-1 lb Chapter 4 l
Chapt er 14 i EP.4-1 16 l EP.14-1 16 i Chapter 5 i Chapter 15 EP.5-1 16 EP.1s-1/2 16 l Chapter 6 i
Clmpter 16 i EP.6-1/2 16 EP.16-1 16 Chapter 7
, Chapter 17 EP.7-1/ blank 16 EP.17-1 lb MEP-1 I
WUP PSAR Anendment 16
.9AVEN 11/78 PRELIMINARY SAFETY ANALYSIS REPORT MASTER LIST OF EFFEC"PIVE PAGES (CONT)
Appendix A EP.A-1 16 Appendix B EP.B-1 16 Qu.ations and Responses EP.Q-1 thru 3 16 O
W MEP-2
WUP PSAR O
V GENERAL TABLE OF CONTENTS (CONT'D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRA' S 7 INSTRUMENTA JLS III 7.1 Itr 7.2 7.3 /EATURES SYSTEMS QUIRED FOR SAFE SHUTDOWN
, SAFE'1"I-hr: LATED AND POWER GENERATION DISPLAY INSTRUMENTATION f 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY AND POWER GENERATION 7.7 CONTROL SYSTEMS l 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ON61TE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES I
l 9.4 AIR CONDITIONItG , HEATIIG, COOLING, l AND V'NTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS A
U iii
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i l
WUP l PSAR l l
GENERAL TABLE OF CONTENTS LCONT'D) l Chapter / '
Section Tit 3e Volume 10 STEAM AND POWER COINERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 .RALIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS iv
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WUP Amendment 16 PSAR 11/78
, PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES -
Chapter 1 y
Page, Table . (T) or Revision Piqure (F) Number 1-1 through 1-11 7 1-iil through 1-iv 10
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1-vi 10 1.1-1 through 1.1-2 -
1.2-1 through 1.2-12 -
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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) 4 -
(pages 5 through 9) 15 '
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(page 11) 15 T.1.3-1 (pages.12 and 13) -
T.1.6-1 (12 pages) -
F.1.1-1 -
F.1.2-1 8 F.1.2-2 -
?.1.2-3A through F.1.2-3K -
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l F.1.5-1 -
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2.0-1 -
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2.4-3 15 2.5-1 15 2.5-2 -
T.2-1 (page 1) b)' (page 2) 15 7
(pages 3 and 4) 15 l
1 EP.2-1
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WUP Amendment 16 PSAR 11/78
\
PRELIMINARY SAFETY ANALYSIS REPORT
(/
LIST OF EFFECTIVE PAGES Chapter 3 Page, Table (T) or Revision Piqure (F) Number Page, Table (T) or Revision 3.7-26 13 Fiqure (F) Number 3.7-27 through 3.7-30 -
3-1 through 3-xviii 8 3.7-31 14 3-xix 16 3.7-32 -
3-xx 13 3.7-33 through 3.7-34 6 3-xxi through 3-xxii 8 3.7-35 through 3.7-37 -
3-xxiii 15 15 3.8-1 through 3.8-2a 3.1-1 -
3.8-3 through 3.8-5 3.1-2 through 3.1-3 7 3.8-6 through 3.8-6a 1 3.1-4 through 3.1-8 -
3.8-7 14 3.8-8 4 3.1-9 3.1-10 through 3.1-15 -
3.8-9 10 3.1-16 2 3.8-10 11 3.1-16a 3 3.8-11 through 3.8-16 10 3.1-17 through 3.1-52 - 3.8-17 through 3.8-19 3.1-53 through 3.1-55 16 3.0-20 16 3.1-56 2 3.8-21 through 3.8-22 -
3.1~57 through 3.1-61 -
3.8-23 4 3.8-24 -
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("'% 3.2-3 through 3.2-4 - 3.8-27 through 3.d-28 6
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3.8-29 3.8-30 16 3.3-2 through 3.3-2a 1 3.8-31 11 .
3.3-3 through 3.3-4 2 3.8-32 through 3.8-35 10 3.3-4a 13 3.8-36 through ?.d-37 11 3.3-5 2 3.8-38 through 3.8-40 10 3.8-41 -
3.4-1 2 3.8-42 13 3.4-2 9 3.8-43 -
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3.8-50 10 3.5-5 through 3.5-7 6 3.8-50a 6 3.8-51 -
3.6-1 7 3.8-52 10 3.6-2 through 3.6-3 6 3.8-53 8 3.6-4 through 3.6-4a 7 3.8-54 7 3.6-5 through 3.6-6f 6 3.8-54a 15 3.6-7 through 3.6-8b 2 3.8-55 9 3.6-9 1 J.8-56 8 3.6-10 2 3.8-56a 10 3.o-57 -
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3.7-14 2 3.9-1 13 3.7-15 3.9-2 through 3.9-4a 4
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3.7-19 through 3.7-20c 13 3.9-17 1 3.7-21 through 3.7-25 - 3.9-18 through 3.9-25 -
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NUP Amendment 16 PSAR 11/76 PRELIMINARY SAFEli ANALYSIS REPORT LIST OF EFFEL*PIVE PAGES (CONT 'D)
Chapter 3 Page, Table (T) or Revision Page, Table (T) or Revision Piqure 3 Number Figure (F) Nu:nber 3.10-1 14 F.3.8.1-15 4 3.10-2 2 F.3.8.1-16 through F.3.8.1-21 -
3.10-3 14 F.3.8.1-22 1o 3.10-4 2 F.3.8.1-23 -
F.3.8.1-24 4 3.11-1 through 3.11-2a 16 F.3.8.1-25 -
3.11-3 12 F.3.8.1-25a through F.3.8.1-25b 16 3.11-4 13 F.3.8.1-26 -
3.11-4a 16 F. 3. 8. 4 -1 (deleted) 15 3.11-4b 13 F. 3. 8.4 -2 throug h F. 3.8.4 -7 15 3.11-5 througn 3.11-6 7 F . 3. 8 . 4 -8 (3 pages) 15 3.11-7 16 F.3.8.4-9 15 F.3.8.4-9A 15 T.3.1.2-1 -
F. 3. 8. 4 - 10 15 T.3.2.5-1 (paqes 1 and 2) 8 F . 3 . 8 . 4 -10 A 15 (page 3) 10 F.3.8.4 -11 through F.3.8.4-12 15 (page 4) 9 F. 3. 8. 4 -13 3 t (pages 5 and 6) 2 F . 3. 8 . 4 -14 through F.3.8.4-15 8 (page 7) 4 F. 3. 8. 4 -16 (deleted) 8 (page 8) 6 F. 3. 8. 4 -17 2 l (pages 9 and 10) 4 F.3.8.5-1 -
(pages 11 through 13) 6 (pages 14 through 17) 4 T.3.2.5-1 (page 18) 2 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 -
l T.3.5.4-1 6 j T.3.6-1 4 i 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 -
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T.3.9-7 (2 pages) -
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F. 3. 7.1 -4 through F.3.7.1-14 13 F.3.7.2-1 -
F.3.7.3-1 through F.3.7.3-5 -
F.3.8.1-1 through F.3.8.1-8 8 F.3.8.1-9 4 F.3.8.1-10 through F.3.8.1-13 -
F.3.8.1-?4 5 EP.3-2
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WUP Amendment 16 PSAR 11/78-4 CHAPTER 3 (f ,
LIST OF TABLES (CONT 'D)
Table Title 3.9-4 Stress . Criteria for Safety-Related ASME Class 2 and Class 3 Tanks, Westinghouse Scope of Supply 3.9-5 Stress. Criteria for ASME Code Class 2 and Class 3 Inactive Pumps, Westinghouse Scope of Supply 3.9-6 . Stress Criteria for ASME III Class 2 and 3 Active Pumps, i hestinghouse Scope.of Supply 3.9-7 Stress Criteria for Safety-Related ASME Class 2 and 3 Valves,. Load Combinations 'and Allowable Stress, Westinghouse scope of Supply 3.11-1 Calculated Gamma Dose Rates . Vs. Time 3.11-2 Calculated Intergrated Gamma Doses for Various Times O
3-xix
., . . . - - ,,,r- * . . . _ . . . .c, ..., . - .. . . . + . . , . . ......,...,,...c.,
. ~ , ,,,..,,e,.n-.. .e+,..-m.- , . , ,,y.,,. , .m .tr en..y . , , . , - - , , -I
WUP Rmendment 13 PSAR 7/15/76 CHAPTER 3 LIST OF FIGURES Figure Title 3.6-1 Example of Application of Pipe Break Criteria - Main Steam Piping 3.6-2 Typical Restraint Load Amplification Factors 3.7.1-1 Design Response Spectrum Horizontal SSE .20g 3.7.1-2 Design Response Spectrum Vertical SSE .20g 3.7.1-3 Design Response Spectrum Horizontal OBE .06g 3.7.1-4 Design Response Spectrum Vertical OBE .06g 3.7.1-5 Time History Accelerogram Horizontal .20g 3.7.1-6 Time History Accelerogram Vertical .20g 3.7.1-7 Response Spectrum Correlation Horizontal SSE .20g 3.7.1-8 Response Spectrum Correlation Horizontal SSE .20g 3.7.1-9 Response Spectrum Correlation Horizontal SSE .20g 3.7.1-10 Response Spectrum Correlation Horizontal SSE .20g 3.7.1-11 Response Spectrum Correlation Vertical SSE .20g 3.7.1-12 Response Spectrum Correlation Vertical SSE .20g 3.7.1-13 Response Spectrum Correlation Vertical SSE .20g 3.7.1-14 Response Spectrum Correlation Vertical SSE .20g 3.7.2-1 Dynamic Model of Contal: .> ent Structure 3.7.3-1 Representation of Fami!y of Peak Response Curves Within Broadened Resonant Peak 3.7.3-2 Hypothetical Vs. Actual Response of Multiple Modes Within Broadened Resonant Peak 3.7.3-3 Justification of Static Load FactoI 3.7.3-4 Model F?ams G
3-xx
WUP PSAR Complete Identification of Name As Date of the Document and of the Used Issue Sponsor Organization Liner Plates & Pads ASME III Division 2 Accesses: ASME III Division 1 Equipment Hatch Fr.
- Pers. Access Inck L E.
Penetrations ASME III Divis1on 1 Piping NC-Electrical NE Fuel Transfer Tube Sleeve ASME III Division 1, Subsection NE 3.8.1.3 Loads and Loading Combinations 3.8.1.3.1 Containment Structure Shell and Mat The containment structure design at each site is based upon the load criteria given in ACI/ASME (ACI-359) unless otherwise noted herein. Table 3.8.1-3 tabulates the structural load combinations O and, in addition to the ACI/ASME incorporates jet impingement and missile impact loads in the abnormal load combinations, where applicabits.
(ACI-359) load criteria, 3.8.1.3.2 Steel Liner and Penetrations 3.8.1.3.2.1 Liner Plate Load combinations for the design of the liner plate, anchors, and embedments are shown in Table 3.8.1-4. The loads in Table 3.8.1-5 apply to the design of insert and overlay plates and their associated anchors.
Table 3.8.1-4 is based on ASME III, Division 2, Table CC-3200-1.
Fatigue effects arc considered, where necessary, using the methods of AMSE III, Division 1.
3.8.1.3.2.2 Penetrations
- 1. Piping System Penetrations l Table 3.7-3 for Seismic Category I piping applies to unsleeved pipe penetrations and pipe and fluted head of sleeved penetrations.
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3.8-19 L , _ _ _ . _ , . ,_ ... _ _ .. _ . _ , - .. _ _ _ , . _ ..- .
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l WUP Rmendment 16 '
PSAR 11/78 Sleeves are considered part of the containment boundary as shown in Fig. NE-1132-1 of ASME III, Division 1. A~ such, they are designed for the load combinations of Table 3.8.1-6 which is based on Article NE-3000.
- 2. Mechanical System Penetration The fuel transfer tube enclosure is part of the containment boundary. The expansion bellows are designed for the most severe combinations of seismic deflections plus containment pressurization and temperatures. The bellows are connected by a cylinder which is welded to the liner. The cylinder is designed for the loads listed in Table 3.8.1-6.
- 3. Elect.,ical Service Penetrations Where electrical penetrations form a part of the containment boundary, they are designed for the load combinations in Table 3.8.1-6. These loads conform to the requirements of ASME III, Division 1, Article NE-3000.
16 The electrical penetrations are designed for the containment design pressure. This pressure includes a safety margin over the maximum pressure as required by footnote 1 to Article NE-3000.
3.8.1.3.2.3 Access Openings Load combinations for access openings are listed in Table 3.8.1-6. As required by AMSE III, Division 2, para. CC-3831.1, these combinations comply with the requirements of ASME III, Division 1, Article NE-3000. The attached liner is designed for the same loading combinations as the liner plate, Table 3.8.1 4 3.8.1.4 Design and Analysis Procedures 3.8.1.4.1 Containment Structure The design, analysis, and construction of th e containment structure is similar to other plants designed and constructed by Stone & Webster, as listed in Section 1.4.
Tne containment structure is analyzed and designed for the factored loading conditions given in Section 3.8.1.3. The containment structure consists of a hemispherical dome, a cylindrical shell, and a mat supported by an elastic subgrade.
Discontinuities exist at the junction of the mat and cylindrical shell, and at the junction of the shell and dome. The length of the cylindrical shell is great enough to eliminate the influence of one discontinuity on the other.
O 3.8-20
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l WUP l PSAR j
- 3. Embedments Allowables for the corner transition area embedment are given in Table 3.8.1-7.
- 4. Insert and Overlay Plates, Brackets, and Attachments )
The strain limits of Table 3.8.1-7 are used as the criteria in the analysis of the insert plates. For the Design I load case of Table 3.8.1-5, the normal category limits of Table 3.8.1-7 are used, and for.the Design II load case, the abnormal category limits of Table 3,8.1-7 apply.
The allowables used for the Design I load case for overlay plates, brackets, and attachments are per AISC, 1969,
" Specification for the Design, Fabrication, and Erection of Structural Steel for Building". The allowables used with the Design II load case are 1.5 times those used with the Design I load case. These strength considerations are based on paragraph cc-3750 of ASME III, Division 2.
S. Anchors Containment liner anchor allowables are given in Table 3.8.1-1, which is based on ASME III, Division 2, Table cc-3700-2. These i allowables are applied also to the anchors which are used to secure insert plates to the containment wall.
3.8.1.5.2.2 Penetrations
- 1. Pipe Penetrations Allowables for sleeved pipe penetrations are listed in Table 3.8.1-6. As noted in the table, these apply to the sleeve and are derived from the requirements of Article NE-3000, ASME III, Division 1. The fluted head and the pipe are designed to the criteria and by the analytical procedures for piping described in Section 3.7.3.6. The applicable table for loading combinations and stress limits for Seismic Category I piping systems is Table 3.7-3. This includes the dynamic system loads associated with the DBA, except where rupture of the pipe line attached to the penetration is postulated.
Unsleeved penetrations are designed for the criteria listed in Table 3.7-3.
The design criteria for pipe rupture for the sleeve of sleeved
. penetration assemblies are listed in Table 3.8.1-2. The table is based on Appendix F of Section III, Division 1, and includes O
3.8-29 4 w ey,.g.--a , w y.-g ,.wi.- ,s ,+.ar% e r *-wv 'vN **w-es' 't'v er:tt *'m *'W-'=**k-ee'ye'*-~W' * + + + "t*-1'-wv* 1Nnw'*' *w + 9s
WUP Amendment 16 PSAR 11/78 the 85 percent Guide 1.57 and reduction factor recommended by Regulatory paragraph NE-3131.2 for metal containment f components.
- 2. Mechanical System Penetrations The stress allowables listed in Table 3.8.1-6 apply to the cylindrical section of the fuel transfer tube enclosure for the area where it forms a portion of the containment boundary.
- 3. Electrical Service Penetrations Allowables for electrical service penetrations are given in Table 3.8.1-6 which is based on Article NE-3000, ASME III, 16 l Division 1. Regulatory Guide 1.63 references IEEE Std. 317-1976 as being an acceptable criteria.
The intent of both the Regulatory Guide and the IEEE Standard 16l will be met. Accordingly, the criteria listed in Table 3.8.1-6 reflect the requirements of Article NE-3000, particalarly NE-3300 and NE-3700.
3.8.1.5.2.3 Access Openings Allowables for access openings are given in Table 3.8.1-6 and are based on Article NE-3000, ASME III, Division 1.
3.8.1.6 Materials, Quality Control, and Special Construction Techniques For applicable codes, standards, and specifications, see Section 3.8.1.2.
3.8.1.6.1 Concrete Materials and workmanship used in the containment, and for other Seismic Category I structural concrete, conforms to the..following codes and specifications:
ACI 214 " Recommended Practice for Evaluation of Compression Test Results of Field Concrete" ACI 301 " Specifications for Structural Concrete for Buildings" and all specifications of the American Society for Testing and Materials referred to in Section 1.5 and declared to be a part of ACI 301 as if fully set forth herein ACI 304 " Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete" ACI 305 " Recommended Practice for Hot Weather Concreting" h l
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3.8-30 l
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WUP Amendment 16 PSAR 11/78 3.11 ENVIRONMEIRAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT
( Components of safety-related systems and their associated instrumentation and electrical equipment located inside the containment structure and elsewhere, which are required to function during and subsequent to an accident are designed to operate under normal, test, and post-accident environmental conditions that may occur at their installation location. The locations of major components of safety-related systems are given in Table 3.2.5-1, and the , environmental conditions for these locations (except for those locations inside the containment which are given below) are listed in Table 7.5-4. I The most severe environmental conditions to be imposed upon the equipment which would operate inside the containment during normal operation and subsequent to a DBA are as follows:
During Normal First 2 Hours After First 2 Hours Condition Operation Following a DBA Following a DBA 12 Temperature 90 F 275 F (tentative)
- 225 F (tentative)
Pressure 1 psig 45 psig 22.5 psig (tentative)
() Humidity Spray 50%
None 100%
The equipment is 100%
The equipment is assumed to be assumed to be con-continuously tinuously sprayed sprayed with a with a solution of solution of 1.5% 1.5% boric acid boric acid, with with NaOH added to NaOH added to result in a pH of result in a pH 8.5, measured at of 10.5, meas- 25 C ured at 25 C M
Radiation 1-50 mr/hr 2.0 x 106 rads 1.0 x 108 rads access areas plus applicable includes maximum 1-50 r/hr 40 yr normal 40 yr normal dose cubicles dose Stone and Webster Scope of Supply The integrated doses for the duration of the accident including, the first 2 hr, were calculated by assuming a dose point at the center of the base of a hemispherical source which has the same volume as the containment. The source term assumed that j ,_s 100 percent of the core noble gases, 50 percent of the core l t )
l \~/
- The temperature used for qualification will be based on a 12 revised analysis after satisfactory generic resolution of Westinghouse steam line break analysis methods.
3.11-1
WUP Amendment 16 )
PSAR 11/78 l iodines, and 1 percent of the remainder of the core activity were instantaneously released into the containment volume. In considering the gamma dose, no credit was taken for either internal shielding or activity reduction by sprays or other mechanisms except radioactive decay. By contrast, the shielding effect of equipment enclosures and electrical cable jackets to beta radiation is considered sufficient to reduce the beta dose to a negligible fraction of the total dose.
The calculated gamma dose rates vs time, as well as calculated integrated gamma doses for rarious time periods, are presented in is Table 3.11-1.
Class IE equipment and cab. ) be qualified will use a direct exposure gamma radiation environment.
The protective coatings (paint) used within the containment are selected on the basis of a satisfactory demonstration of resistance to gamma radiation at a rate of from 1.0 to 6.4 x 107 rads /hr to a total integrated dose of from 1.0 to 9.0 x 10' rads. A spent fuel assembly is used as the gamma source.
The qualification may be by analysis, test, or a combination of both. The procedures for testing Class 1E equipment are detailed in their applicable IEEE Standards and/or IEEE 323.
Environmental conditions applying specifically to certain equipment items are indicated in Section 3.11.2.
3.11.1 Equipment Identification Equipment required to function inside the containment during and subsequent to a DBA is listed below. Also listed are the periods of time for which the equipment would be required to operate following the DBA.
- 1. Containment isolation valves and other valves in engineered safety features (ESF) (with operators) -
operation completed in the first 5 minutes after the DBA,
- 3. Pressurizer pressure and level channels - up to 30 minutes after the DBA,
- 4. Accumulator pressure channels - up to 5 minutes after the DBA,
- 5. Containment heat removal fans - up to 2 months after the DBA, 3.11-2
- . - . . . . . . . . . - ~ ,.
WUP Amendment 16 PSAR 11/78
- 6. . Containment' heat removal- fan' 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 . ,
accumualtors 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. Containment Isolation System (Section 6.2.4)
- 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 l
3.11-2a
L WUP Amendment 16 PSAR 11/78
() determined qualified life is less than full plant life, ongoing qualification will be used.
primary means of ongoing qualification .will be The operating history of similar equipment installed in similar or more severe environments and a periodic
!2 testing program designed to monitor those equipment parameters which may be degraded by aging.- If the monitored parameters show a possible reduction in the reliability of the equipment, corrective action will be taken. This corrective action will be maintenance, modification, or replacement.
- c. The results of the qualification program will be U
included in the FSAR.
Electrical Penetrations Electrical penetrations will be designed, tested, and documented in accordance with IEEE 317-1976 (Ref . 2) , and Regulatory lg Guide 1.63 (Ref . 6) .
Containment Isolation Valve Actuators The valves inside the containment are designed to operate under the environmental conditions tabulated in Section 3.11. The
/~T containment isolation valves are air-operated, fail-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 (Ref. 7) 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 pe "ods, 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 IEES-323-1974, or related standards as defined by the generic g resolution being pursued by Westinghouse and the NRC.
O v
3.11-4a i
F F =wT-
- g m y. s, mew 9 v..g. ry w, ,3 , , , , . __9,,. _ ,,, . _ ,
WUP Amendment 13 PSAR 7/15/76 Motors h
All motors for safety-related items located inside the 4l containment are rated for 1.0 or 1.15 service f actor load. The insulation systems are Class B or better. All motors are given at least the 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 -1974 (Ref. 3) and Regulatory I ; Guide 1.40 March 1973 (Ref. 3) . Continuous duty motors outside i
the containment will be qualified in accordance with l IEEE-334-1974 (Ref . 3) .
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llh 3.11-4b
WUP Amendment 16 PSAR 11/78 References I
- 1. Deleted N l
- 2. IEEE-317, " Electrical Penetration Assemblies in Containment i Structures for Nuclear Power Generating Stations," 1976. l 16
- 3. IEEE-334, " Type Tests of Continuous Duty Class I Motors Installed Inside the Containment of Nuclear Power Station," 1975. Generhting l16
- 4. IEEE-382, " Trial Use Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations," 1972. i 1
5.
IEEE-279, " Criteria for Protection Systems for Nuclear Power Generating Stations," 1971.
- 6. Regulatory Guide 1.63, " Electric Penetration Assemblies in
, Containment Structures for Water. Cooled Nuclear Power Plants."
- 7. IEEE-383, " Standard for Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations," 1974.
, 8. Regulatory Guide 1.40, " Qualification Tests of Continuous
\_- Duty Motors Installed Inside the Containment of Water Cooled Nuclear Power Plants."
- 9. Regulatory Guide 1.73, " Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants."
- 10. IEEE Std-323 for Qualifying Class IE Equipment for Nuclear Power Generating Stations.
3.11-7
. ~ . . - . - . . - . , - . , , . - . , . , , . , - . . . , . . . - _ . . - . - - . . . , . . - ~ . - . . - , , - . . . . ..... _ ..- - . - - - - . -
. . . . . _ . . .- - . . . . _ . = - . . _ __ . ..
I WUP Amendment 16 PSAR 11/78 l l
1 TABLE 3.11-1 CALCULATED GAMMA DOSE RATES VS TIME Time- Gamma Dose Rate-
[hr) (rads /hr) 0.0 2.1 x 106-2.0 7.3 x-105 16 8.0 3.0 x 10s 24.0 1.3 x.105-96.0 4.2 x 10*
200.0- 2.7 x 104 300.0 1.9 x 104 400.0 1.4 x 10*
500.0 1.1 x 104 600.0 -B.9 x 103 720.0 7.0 x 103
-4830.0 9.8 x 102 O
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, . - . - . . . . . . . . - - . , . . . . . . . - - . . _ . . - . - - . . . . - . - . . . . . . - . . . . . . . . . - _ . , . . . . . . . ~ . - , . . . . . . - . . . . . ~ , . , . - . .
WUP Amendment 16 PSAR 11/78 (s) TABLE 3.11-2 CALCULATED INTEGRATED GAMMA DOSES FOR VARIOUS TIMES Gamma Dose Time Interval (rads) 16 0-2 hr 2.0 x 106 0-8 hr 4.8 x 106 0-24 hr 7.8 x 106 0-96 hr 1.3 x 107 0-30 day 2,3 x 107 0-6 mo 3.2 x 107 hi
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3A-lii through 3A-iv 8 3A-1 through 3A-2 1 3A-3 through 3A-9 -
3A-10 through 3A-11 4 3A-12 through 3A-24 -
3A-25 2 3A-26 through 3A-26a 1 3A-27 through 3A-29 4 T.3A.3-1 through T.3A.3-2 -
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NUP PSAR
( GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION g 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY y,) AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONIIG , HEATItG, COOLIIE, AND VENTILATION SYSTEMS 9.5 OTHER AUXII:IARY SYSTEMS x_J iii
WUP PSAR I l
GENERAL TABLE OF CONTENTS (CONT'D) l l
Chapter / i Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR l
10.3 MAIN STEAM SUPPLY SYSTEM l
10.4 OTHER FEATURES OF STEAM AND POWER I l
CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS O
iv
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WUP Amendment 16 PSAR 11/78
.( PRELIMINARY SAFETY ANALYSIS REPORT A
LIST OF EFFECTIVE PAGES Chapter 4 Page, Table (T) or Revision Page, Table (T) or Revision' ,
Fiqure iF) Number Fiqure (F) Number 4-1 15 F.4.3-36 10 4-11 through 4-ix 10 F.4.3-37 through F.4.3-45 -
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WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSL' REPORT h..J LIST OF EFFECTIVE PAGES Chapter 5 i
Page, Table (T) or Revision Page, Table (T) or Revision Piqure (F) Number Figure (F) Number 5-i -
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WUP Amendment 16 PSAR 11/78 I For a selected time of operation, this shift is assigned a sufficient magnitude so that no unirradiated ferritic materials in other components of the RCS will be limiting in the analysis.
Changes in fracture toughness of the beltline region plates or forgings, weldments, and associated heat affected zones due to radiation damage will be monitored by a surveillance program as g discussed in Section 5.4.3.6. The evalua*ich or the radiation damage in this surveillance program is based on pre-irradiation and post-irradiation testing of Charpy V-notch and tensile specimens and post-irradiation testing of 1/2T compact tension specimens carried out during the lifetime of the reactor vessel.
Specimens are irradiated in capsules located near the core midheight and removable from the vessel at specified intervals.
The results of the radiation surveillance program will be used to verify that the ART ui predicted using Figs. 5.2-3 and 5.2-4 is appropriate and to make any changes necessary to correct Fig.
5.2-3 if ART ei determined from the surveillance program is dif ferent from the predicted 3RT ui
- Temperature lindts for preservice hydrotests, and inservice leak and hydrotests will be calculated using methods in accordance with 10CFR50 Appendix G. 7 5.2.4.4 Compliance with Reactor Vessel Material Surveillance
(-)x
(_ Program Requirements See Section 5.4.
5.2.4.5 Reactor Vessel Annealing See Section 5.4.
5.2.5 Austenitic Stainless Steel The unstabilized austenitic stainless steel material specifications used for the (1) RCPB, (2) systems requireC for reactor shutdown, and (3) systems required for emergency core cooling are listed in Tables 5.2-7 and 5.2-8.
The unstabilized austenitic stainless steel material for the reactor vessel internals which are required for energency core cooling for any mode of normal operation or under postulated accident conditions, and for core structural load bearing members are listed in Table 5.2-10.
\s-5.2-35
WUP Amendment 7 PSAR 6/6/75 All of the above tabulated materials are procured in accordance with the material specification regrirements and include the special requirements of the ASME Code,Section III, plus Addenda and Code Cases as are applicable to meet Appendix B of 10CFR50 in the Federal Register, Vol. 35, No. 125.
5.2.5.1 Cleaning and Contamination Protection Procedures It is required that all austenitic stainless steel materials used in the fabrication, installation, and testing of nuclear steam supply components and systems be handled, protected, stored, and cleaned according to recognized and accepted methods and techniques. The rules covering these controls are stipulated in the Westinghouse Electric Corporation process specifications.
These process specifications supplement the equipment specification and purchase order requirements of every individual austenitic stainless steel component or system which Westinghouse procures for a nuclear steam supply system, regardless of the ASME Code Classification. They are also given to the architect-engineer and to the owner of the power plant for use within their scope of supply and activity to assure compliance with the ANSI N-45 specifications.
To assure that manuf acturers and installers adhere to the rules in these specifications, surveillance of opera tions by Westinghouse personnel is conducted either in residence at the manufacturer's plant and the installer's const ruction site or, when residency is not practical, during periodic engineering and quality assurance visitations and aud2.ts at these locations.
The discovery of any deviation from these rules, whether it be during the "act" or as the result of a subsequent " material-reaction" requires corrective measures to eliminate the condition or replacement of the material and/or component.
The process specifications (PS) which establish these rules and l which are in compliance with The American National Standards Institute N-45 Committee specifications are as follows:
! PS Number l
82560HM Requirements for Pressure Sensitive Tapes for Use on Austenitic Stainless Steels.
83336KA Requirements for Thermal InsuT.ation Used on Austenitic Stainless Steel Piping and Equipment.
l 83860LA Requirements for Marking of Reitetor Plant Components ano i Piping.
84350HA Site Receiving Inspection and Storage Requirements for Systems, Material, and Equipment.
5.2-36
~. . . . _ . - . _ . - _ . _ _ ._._ _ _ _
l WUF
- PSAR GENERAL TABLE OF CONTENTS (CONT'D)
Chapter /
Section Title volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY AND POWER GENERATION ,
7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM I
8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV j 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES l
i 9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS O :
iii
WUP PSAR
{
l GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 10 STEAM AND POWER COhVERSION SYSTEM V 10,1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS
's 3 . 2 TRAINING PROGRAM 13.3 EMERGENCY PLANS O
iv
WUP Amendment 16 PSAR 11/78
(
( PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 6 Page, Table (T) Revision Page, Table (10 or Revision <
or Fiqure (F) Number Fiqure (F) Number 6-1 8 6.3-16 14 6-11 throuch 6-vi 15 6.3-17 through 6.3-1Ba 2 6-vii 16 6.3-19 through 6.3-20 -
6-viii 8 6.3-21 1 6-ix 15 6.3-22 through 6.3-22a 14 6-x 16 6.3-23 14 6-xi through 6-xii 8 6.3-24 through 6.3-24a 9 6-xiii through 6-xv 11 6.3-25 through 6.3-28 16 6.1-2 10 6.3-29 1 6.1-3 14 6.3-30 7 6.1-4 -
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(page 2) 11 EP.6-1
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (COtrT 8D)
Chapter 6 Page, Table (T) or Revision Page, Table (10 or Revision Fiqure (F) Number Fiqure (F) Number (page 3) 2 F.6.2.1-22 -
T.6.2.1-1 through F.6.2.1-23 through T.6.2.1-3 -
F.6.2.1-25 8 T.6.2.1-4 through F.6.2.1-26 1 T.6.2.1-6 8 F.6.2.1-27 through T.6.2.1-7 11 F.6.2.1-3 T.6.2.1-B (3 pagas) 8 F.6.2.1-31 through T.6.2.1-9 through F.6.2.1-35 4 T.6.2.1-19 8 F.6.2.1-36 11 T.6.2.1-20 (2 pages) 8 F.6.2.1-37 10 T.6.2.1-21 through F.6.2.1-38 6 T.6.2.1-23 8 F.6.2.1-39 through T.6.2.1-24 through F.6.2.1-51 8 T.6.2.1-25 10 F.6.2.1-52 through T.6.2.1-26 9 F.6.2.1-53 10 T.6.2.2-1 (2 pages) 15 F.6.2.1-54 8 T.6.2.2-2 (3 pages) 15 F.6.2.2-1 15 T.6.2.2-3 5 F.6.2.2-2 through T.6.2.3-1 (2 pages) 15 F.6.2.2-3 11 T.6.2.3-2 through F. 6 . 2 . 2 -4 4 T.6.2.3-4 15 F.6.2.2-5 through T.6.2.3-5 through F.6.2.2-6 11 T. 6 . 2. 3 -9 (deleted) 11 F.6.2.2-7 through T.6.2.3-10 15 F. 6. 2. 2 -8 4 T.6.2.4-1 (page 1) F.6.2.2-9 11 through (page 2) -
F. 6 . 2. 2 -10 15 (page 3) 6 F.6.2.4-1 through (page 4) -
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WUP Amendment 16 PSAR 11/78 CHAPTER 6 TABLE OF CONTENTS (CONT'D)
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 Lvaluation 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 6.5-5 6.5.2.1 Design Bases 6.5-5 6.5.2.2 System Description 6.5-7 6.5.2.3 Design Evaluation 6.5-9 6.5.2.4 Tests and Inspections Requirements 6.5-10 6.5.2.5 Instrumentation Applications 6.5-11 O
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 ct eristics for Post Reflood (One Intact Loop) at 183 Seconds 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 Energy Release Rates Pump Suction LDR (4.88 f tz) - 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 6-viii i
A,
WUP Amendment 15 ,
PSAR 9/22/78 CHAPTER 6 LIST OF TABLES (CONT ' D)
Table Title 6.2.1-15 Mass and Energy Release Rates Pressurizer Spray Line Break - Pressurizer Cubicle 6.2.1-16 K-Factors, Vent Areas, and Vent Flow Models for Pressurizer Cubicle Analysis
~
6.2.1-17 Length to Area Ratios Pressurizer Cubicle Analysis with RELAP 4 6.2.1-18 Short Term Steam Line DER-Mass and Energy Release Rates 6.2.1-19 Short Term Feedwater Line DER - Mass and Energy Release Rates 6.2.1-20 K-Factors, Vent Areas, and Vent Flow Models for Operating Floor Differential Pressure Analysis 6.2.1-21 Length to Area Ratio for Analysis of Operating Floor Differential Pressure with RELAP4 6.2.1-22 Mass and Energy Release Rates-Steam Line* Break 6.2.1-23 Feedwater Line Break - Mass and Energy Release Rates 6.2.1-24 Peak Containment Conditions Main Steam Line Break 6.2.1-25 Accident Chronology Main Steam Line Break 6.2.1-26 Length-to-Area Ratios Reactor Cavity Analysis with RELAP4 6.2.2-1 Containment Spray System Component Data 6.2.2-2 Failure Analysis for Containment Heat Removal Systems 6.2.2-3 Containment Spray System Leakage Outside Contain-ment Structure 6.2.3-1 Nomenclature Used for Equations in Section 6.2.3 6.2.3-2 Iodine Removal Coefficients 6.2.3-3 Parameters for Calculating Minimum and Maximum O Containment Spray pH During Caustic Addition 6.2.3-4 Parameters for Ultimate Sump pH Calculations 6-ix l
WUP Amendment 16 PSAR 11/78 CHAPTER 6 LIST OF TABLES (CONT 'D)
Table Title 6.2.3-5 through 6.2.3-9 Deleted 6.2.3-10 Containment Volumes Covered by Spray 6.2.4-1 Piping Penetrations through the Reactor Containment 6.2.5-1 Parameters Used in Calculating Hydrogen Sources 6.3-1 Emergency Core Cooling System Component Parameters 6.3-2 Materials Employed for Emergency Core Cooling System Components 6.3-3 Sequence of Change-Over Operation from Injection to Recirculation, Emergency Core Cooling System 6.3-4 Normal Operating Status of Emergency Core Cooling System Components for Core Cooling 6.3-5 Emergency Core Cooling System Shared Functions Evaluation 6.3-6 Maximum Potential Recirculation . Loop Leakage External 6.3-7 Emergency Core Cooling Relief Valve Data 6.3-8 Single Active Failure Analysis, Emergency Core Cooling System, Loss-of-Coolant Accident 6.3-9 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis, Long Term Phase 6.5-1 Parameters Used in Evaluating Low-Flow Air-Bleed Cooling f or RPVS and FBVS Emergency Filtration Trains 6.5-2 Iodine Inventory of Reactor Plant Ventilation System Carbon Absorber (t=240 hr) 6.5-3 Maximum Carbon Absorber Surf ace Temperature Following a LOCA or FIIA O
6-x l
i WUP PSAR
[\- safety features, the core remains in place and intact with its essential heat transfer geometry preserved.
6.3.3.4 Fuel Rod Perforations Discussions of peak clad temperature and metal-water reactions are given in Sections 15.3.1 and 15.4.1. Analyses of the radiological consequences of a fission product release due to a rupture of a pipe in the RCS are given in Section 15.4.1.
6.3.3.5 to 6.3.3.10 These sections apply to BWRs only and do not relate to this plant.
6.3.3.11 Effects of ECCS Operation on the Core The effects of ECCS operation on the reactor core are discussed in Sections 5.2.1, 15.2.13, 15.3, and 15.4.
6.3.3.12 Use of Dual Function Components The ECCS contains components that have no other operating function as well as components that are shared with other' systems. Components in each category are as follows:
( 1. Components of the ECCS which perform no other function are:
- a. One accumulator for each loop which discharges borated water into a cold leg of the reactor coolant loop piping,
- b. Two boron injection recirculation pumps, one of which continuously circulates the 12 weight percent boric acid solution through the boron injection tank.
- c. One boron injection tank,
- d. One boron injection surge tank,
- e. Associated piping, valves, and instrumentation.
- 2. Components which also have a normal operating function are as follows:
- a. The RRR pumps and the RHR heat exchangers - These components are normally used during the latter stages of normal reactor cooldown and when the reactor is held at cold shutdown for core decay
() heat removal. However, during all other plant 6.3-25
i l
WUP Amendment 16 PSAR 11/78 I operating conditions, they are aligned to perform the low head injection function.
- b. The charging pumps - One of these pumps is normally aligned for charging service. As a part of the chemical and volume control system, the normal operation of these pumps is discussed in Section 9.3.4.
- c. The RWST - This tank is used to fill the refueling cavity for refueling operations. During all other plant operating conditions, it is aligned to the suction of the RHR pumps. The charging pumps suctions are automatically aligned to the RWST upon ,
receipt of the SIS signal.
An evaluation of all components required for operation of the ECCS demonstrates that either:
- 1. The component is not shared with other systems; or
- 2. If the component is shared with other systems, it is aligned during normal plant operation to perform its accident function, or if not aligned to its accident function, two valves in para 11cl are provided to align the system for injection, and two valves in series are g provided to isolate portions of the system not utilized W for injection. These valves are automatically actuated by an SIS signal.
Table 6.3-9 indicates the alignment of components during normal operation, and the realignment required to perform the accident function.
6.3.3.13 Dependence on Other Systems Operation of the charging pumps depends on the following:
- 1. Reactor plant ventilation system (Section 9.4.2) init coolers to maintain a suitable temperature in the charging pump cubicles, 2l 2. Reactor plant component cooling water system (Section gg 9 . 2 . 2 .1) to provide cooling water to the pump seal and oil coolers,
- 3. Emergency buses for power to the pumps,
- 4. The RWST to provide a source of injection water (injection phase only) ,
6.3-26
WUP Amendment 16 PSAR 11/78
- 6. The service water system (9.2.1) to provide cooling water g to the reactor plant ventilation system unit coolers described in Item 1.
Operation of the RHR pumps depends upon the following:
- 1. Reactor plant ventilation system unit coolers to maintain a suitable temperature in the RHR pump cubicles,
- 2. Reactor plant component cooling water system to provide cooling water to the pump seal cooler, g
- 3. Emergency buses for power to the pumps,
- 4. The RWST to provide a source of injection water (injection phase only) ,
- 5. The service water system (9.2.1) to provide cooling water to the reactor plant ventilation system unit coolers E described in item 2.
6.3.3.14 Limiting Conditions for Maintenance During Operation The detailed Technical Specifications (Chapter 16) of the FSAR O will establish limiting conditions governing the maintenance of ECCS components during plant operation with the reactor critical.
The design philosophy with respect to active components in the ECCS is to provide redundant equipment. Since there is redundancy, maintenance is possible during operation for a limited period of time to be defined in the Technical Specifications. Routine servicing and maintenance of equipment of this type is normally scheduled for periods of refueling and maintenance outages.
Since the design basis as stated in Section 6.2 is to accept a single active failure during the injection phase or a single active or passive failure in the recirculation phase and yet provide the minimum required engineered safety features equipment, two 100 percent capacity trains of this equipment are installed. (100 percent capacity is defined as a system which is capable of performing its design bases with its stated design criteria . )
6.3.3.15 Lag Times To provide protection for large area ruptures in the RCS, the ECCS is required to respond to rapidly reflood the core following the depressurization and core voiding that is characteristic of large area ruptures. The accumulators act to perform the rapid g reflooding function. with no dependence on the normal or emergency 6.3-27
l WUP Amendment 16 I PSAR 11/78 power sources, and also with no dependence on the receipt of an actuation signal.
Operation of this system with two of the three available accumulators delivering to the reactor vessel (one accumulator spilling through the break) meets the criteria state in Section 6.3.1.
The function of the pumps is to complete the refill of the vessel and ultimately return the core to a subcooled state. The flow from one charging pump and one RHR pump is sufficient to complete the refill of the core and meets the criteria stated in Section 6.3.1.
Initial response of the injection systems is automatic, with appropriate allowances for delays in actuation of circuitry and active nomponents. The active portions of the injection systems are autontatically actuated by an SIS signal. In addition, manual actuation of the entire ECCS and individual components can be accomplished from the control room. In analysis of system performance, delays in reaching the programmed trip points and in actuation of components are conservatively established on the basis that only emergency onsite power is available.
The starting sequence of the charging pumps, the RHR pumps, and the related emergency power equipment is designed so that delivery of the full rated flow is reached within 25 sec after the process parameters reach the set points for the injection signal.
6.3.3.16 Thermal Shock Considerations Thermal shock considerations are discussed in Section 5.2.
6.3.3.17 Limits on System Parmmeter The reactor is not operated at power with the isolation valves in the accumulator lines closed except for brief periods to test the seating effectiveness of the injection line check valve. This is done on one accumulator at a time by depressurizing the pipe between the check valve and the accumulator and measuring water flow into the test line. This test is routinely performed when the reactor is being returned to power after an outage and the RCS pressure is above the accumulator pressure. If leakage through a check valve is excescive, the isolation valve is closed and an orderly shutdown initiated to repair the check valve. The performance of the check valves for this type of service has been carefully studied and it is concluded that it is highly unlikely that the accumulator isolation valves would have to be closed because of leakage. Valves open signals are transmitted to the valves upon an SIS signal.
6.3-28
WUP Amendment 16 PSAR 11/78
() 4. The motor-operated isolation dampers open in the RPVS emergency filtration train that receives the start signal.
Actuation of the above ensures that a slightly negative pressure is maintained in the auxiliary building and that all building exhaust passes through the emergency filtration trains. The RPVS emergency filtration trains are safety-related Seismic Category I !
and powered from the emergency buses. The design flow rate for I each unit is 10,000 cfm. !
The trains are located in the safety-related auxiliary building which protects the trains from externally generated missiles and extreme natural phenomena. Protection from internally generated missiles and biological shielding for plant personnel is provided by the concrete RPVS emergency filtration train cubicles.
Valves used to isolate the normal auxiliary building exhaust, the normally operating filters and the waste disposal building exhaust are missile protected, . Seismic Category I and safety-related. The RPVS emergency filtration trains incorporate electric heaters to limit the relative humidity of incoming air to 70 percent. Carbon bed depth is a minimum of 2 in. Units of this type have demonstrated iodine-removal efficiencies of 99.9 percent for both elemental and organic forms of iodine. Iodine O removal efficiency is conservatively assumed to be 95 percent for accident analysis.
The carbon adsorbers employedintheRPVSemergencyfiltrationlM trains are of gasketless design to facilitate removal of charcoal and to reduce the potential for leakage. Fire protection for the l M adsorber is provided by manually actuated water spray systems i
(Section 9.5.1) .
, Low-flow air bleed cooling systems are not provided for the RPVS emergency filtration trains. Analyses indicate that the maximum g temperature for the carbon adsorbers is less than the maximum
- temperature specified in Section 4.9 of ANSI N509 Nuclear Power Plant Air Cleaning Units and Components (1976). [
Portions of the collection ductwork used for emergency filtration are used for normal ventilation, and are isolated from the normal ventilation flow path by Seismic Category I, safety-related backdraft dampers whenever the emergency filtration trains are in k operation. Emergency filtration collection ductwork to the following areas is Seismic Category I and safety-related:
- 1. Residual heat removal heat exchanger cubicles,
- 2. Charging pump cubicles,
() 3. Residual heat removal / containment spray pump cubicles, 6.5-3
WUP Amendment 16 PSAR 11/78
- 4. Safeguard valve operating areas, h
- 6. RPVS emergency filtration train cubicles.
6.5.1.3 Design Evaluation The RPVS emergency filtration trains are adequately sized and designed to maintain a slightly negative pressure in the auxiliary building. Provisions are made to process auxiliary-or HEPA/ carbon adsorbers to glminimizebuilding exhaust air radioactive through
~
iodineHEPA and particulates in the building exhaust. Operation of one of the filtration trains is required only in the event of a recirculation piping failure in the auxiliary building following a LOCA in containment'. Post-LOCA operation of one of the RPVS emergency filtration trains is ensured by actuation with an SIS signal as described in Section 7.3.1. A 95 percent removal efficiency for both organic and gl elemental forms of iodine is ensured by the design of the trains I (humidity control and carbon bed depth) and by the program of 1 inspections and tests described in Section 6.5.1.4. Protection of the trains is ensured by their location within a shielded cubicle in the Seismic Category I auxiliary building.
Low-flow air bleed cooling systems are not provided since analysis indicates that 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 on collecting 100 percent of the radioiodines from containment leakage and 100 percent of the 16 radioiodines released to the auxiliary building ventilation .
I system due to ESF leakage following a LOCA. This inventory is collected on the surface of the carbon adsorbers which are then ,
isolated at a time which maximizes the ultimate temperature rise. ;
The energy due to beta decay is deposited on the surface of adsorber while that due to gamma energy is uniformly deposited in the carbon. ho credit is taken for energy absorption in 1 surrounding materials nor for conduction or convection losses.
Parameters used in the analysis are in Tables 6.5-1 and 6.5-2. The carbon adsorber surface temperature as a function of time is in Table 6.5-3.
6.5.1.4 Test and Inspection f
The following tests and inspections are performed on the RPVS emergency filtration trains:
Pre-installation Tests
- 1. Filter train pressure - Trains are evacuated to the test pressure and are observed for 2 hr. No more than 1 in.
16l of water change should be observed.
f 6.5-4
WUP Amendment 16 PSAR 11/78
- 2. HEPA filters -
Testod individually by the appropriate Filter Test Facilit3 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.
In-place Tests Following Initial Installation
- 1. Emergency filtration collected ductwork is balanced and system flow is measared to ensure that the trains and associated ductwork, dampers, and valves meet the design criteria.
14 adsorbers is tested for uniformity.
- 3. HEPA filters - In-place DOP testing in accordance with ANSI N510-1975 Section 10. The test is to coniirm a g penetration of less than 0.05 percent at rated flow.
I O
l r
O 6.5-4a
WUP Amendment 16 PSAR 11/78 A
() 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 lu trains are of gasketless design to facilitate removal of charcoal and to reduce the potential for leakage. For protection for the adsorber is provided by manually actuated water sprays systems M (Section 9.5.1) .
Low-flow air bleed cooling systems are not provided for the FBVS emergency filtration' trains. Analysis indicates that the maximum 6
temperature for the carbon adsorbers is less than the maximum 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 slightly negative pressure in the fuel building, and 2) to provide the capability for processing fuel building exhaust air through carbon adsorbers and HEPA M filters to minimize radioiodine and particulates in the fuel building vent. Operation of one of the trains is
/) administratively required during fuel handling operations. The
\m) 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 tests 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.
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 4 Accident in the Etel Handling and Storage facility for Boiling 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 s uniformly deposited in the carbon adsorber. No credit is taken for heat transfer to the surroundings nor for absorption in the O- filter's structural materials. Parameters used in the analysis 6.5-9
WUP Amendment 16 PSAR 11/78 are in Tables 6.5-1 and 15.4.5-1. The carbon adsorber surface 4 temperature as a function of time is in Table 6.5-3.
6.5.2.4 Tests and Inspections Requirements The following tests and inspections are performed on the FBVS emergency filtration trains:
Pre-installation 'lests
- 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.
1 g 2. HEPA filters -
Tested individually by the appropriate l Filter Test Facility listed in the current USDOE i 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 Regulatory Guide Ml 1.52.
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.
g 2. Air flow distribution to the HEPA filters and carbon adsorbers is tested for uniformity.
- 3. HEPA filters -
Inplace DOP testing in accordance with y ANSI N510-1975 Section 10. The test is to confirm a penetration of less than 0.05 percent 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 M no greater than 20 ppm. The allowed bypass leakage is 0.05 percent. Following the completion of the test, air flow is continued until ef fluent ref rigerant gas is less than 0.01 ppm.
Periodic Test and Inspections Periodic tests and inspection of the FBVS emergency filtration are performed in accordance with Technical Specifications (Chapter 16).
O 6.5-10
WUP Amendment 16 PSAR 11/78 6.5.2.5 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 power is supplied to the train and flow is lost, indicating a ,
failure of the train or of the isolation dampers. Low '
flow starts the redundant train.
O
( l w 1 1
6.5-11 1
. . . - . . . ~ , . , . ,. . - , . ,. _ . . - . . . . , . . . . . . , . ..1
. . ~ .. -. . - _ .. .~ - - . - -
WUP Amendment 16 PSAR 11/78 TABLE 6.5-1 PARAMETERS USED IN EVALUATING LOW-FLOW AIR-BLEED COOLING FCE RPVS AND FBVS EMERGENCY FILTRATION TRAINS Maximum Allowable Temperature Specified in Section 4.9 ANSI N509 Nuclear Power Plant Air Cleaning Units and Components 300 F Max 1. mum Building Temperature Following LOCA 120 F Following FHA 105 F g Filter Characteristics Surface Area 246 ft2 Depth 2-inch, 4-inch ;
Rated Flow Rate 10,000 cfm i Heater AT 12 F ,
Carbon Adsorber Characteristics Density 30 lb/fta Heat Capacity 0.2 Btu /lb F Thermal Conductivity O.12 Btu /ft2/ hr F bV .
I J
1 0
1 of 1
WUP Artendment 16 PSAR 11/78 O
v TABLE 6.5-2 IODINE INVENTORY OF REACTOR PLANT VENTIIATION SYSTEM CARBON ADSORBER (t = 240 hr1 C1 16 I 131 5,250 I 13 '> 0 1 133 10.27 I 134 0 I 135 4.44 x 10-7 O
1 0
1 of 1
WUP Amendment 16 PSAR 11/78-TABLE 6.5-3 MAXIMUM CARBON ADSORBER SURFACE TEMPRATURE FOLLOWING A LOCA OR FHA RPVS Carbon Adsorber FBVS Carbon Adsorber Following a LOCA(1) Folleving a FPH Carbon Adsorber Bed Depth (in)-
2_in. 4 in. 2 in. 4 in.
16 i
t(1) T(a) T(a) T(2) Tca)
-(sec) (F) (F) (F) (F) 0 132 132 117 117 102 132.06 132.06 117.03 117.03 los 132.23 132.21 117.10 117.09 10* 133.28 132.82 117.43 117.27 105 143.2 137.76 118.77 117.90 106 205.6 168.9 123.31 121.16 107 248.6 190.3 127.7 123.37 10s 248.6 190.3 127.7 123.37 O
I l
I NOTES:
(*) Assumes collection of iodines from containment leakage and from ESF leakage released to the auxiliary building atmosphere (2) t = time after filter isolation (sec)
T = temperature of carbon adsorber surface (F) 1 of 1
__ _ . . . . _ . . _ _ _ . . ~ . , . _ _ _ . . . . . . _ _ . . _ . . . _ . - _ . _ . . _ . . _ . - - . . _ . - _ _ _ . . _ _ . . . . - . _ . - . _ . . _ , ~ . _ _ _ _ , - -
WUP Azrsndment 16 PSAR 11/78 O PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECI'IVE PAGES Chapter 7 Page, Table (T) or Revision Page, Table (T) or Revision Piqure (F) Number Figure (F) Numtw r 7-i 8 7.6-1 through 7.6-2 -
7-11 through 7-iii 9 7.6-3 through 7.6-5 14 7-iv through 7-v 8 7-vi 14 7.7-1 through 7.7-16 -
7-vii through 7-viii 8 7.7-17 through 7.7-18b 14 7-ix 14 7.7-19 through 7.7-22 -
7-x through 7-xi 8 7.7-23 5 7.7-24 -
7.1-1 through 7.1-2 -
7.1-3 through 7.1-4a 4 T.7.2-1 (2 pages) 9 7.1-5 through 7.1-6a 4 T.7.2-2 (page 1) -
7.1-7 -
T.7.2-2 (page 2) 5 7.1-8 5 T.7.2-3 (2 pages) -
7.1-9 4 T.7.2-4 (6 pages) -
7.1-10 -
T.7.3-1 through T.7.3-2 -
7.1-11 through 7.1-13 4 T.7.3-3 (2 pages) -
7.1-14 through 7.1-17 11 T.7.3-4 through T.7.3-5 -
T.7.5-1 (2 pages) -
7.2-1 through 7.2-7 -
T.7.5-2 (page 1) 11 7.2-8 9 T.7.5-2 (page 2) -
7.2-8a 13 T.7.5-3 (7 pages) -
_, ) 7.2-9 thvough 7.2-10 7 T.7.5-4 4 7.2-11 -
T.7.7-1 (pages) -
7.2-12 through 7.2-13 7 l 7.2-14 through 7.2-16a 9 F.7.2-1 (page 1) 3 7.2-17 through 7.2-21 -
F.7.2-1 (page 2) 7 7.2-22 4 F.7.2-1 (page 3) -
7.2-23 through 7.2-27 -
F.7.2-1 (page 4) 3 i 7.2-28 through 7.2-28a 4 F.7.2-1 (page 5) 9 7.2-29 through 7.2-32 -
F.7.2-1 (page 6) 7.2-33 13 through (page 13) -
7.2-34 11 F.7.2-1 (page 14) 7 7.2-34a 7 F.7.2-1 (page 15) 3 7.2-35 -
F.7.2-2 through F.7.2-4 -
7.2-36 9 F.7.2-5 7 F.7.6-1 14 7.3-1 through 7.3-6 -
F.7.6-2 -
7.3-7 13 F.7.6-3 (deleted) 11 7.3-8 through 7.3-8a 3 F.7.7-1 through F.7.7-10 -
7.3-9 -
7.3-10 through 7.3-10a 4 7.3-11 through 7.3-12 -
7.3-13 3 :
7.3-14 4 1 7.3-14a 11 l 7.3-14b 4 7.3-15 5 7.3-16 -
7.3-17 13 ,
7.4-1 through 7.4-3 3 l
/'"\ 7.4-4 through 7.4-6 14 l U 7.5-1 4 7.5-2 through 7.5-2a 9 7.5-3 through 7.5-4a 4 7.5-5 through 7.5-6 -
I WUP PSAR GENERAL TABLE OF CONTENTS (CONT'D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS l
6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY
[
AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONI!E, HEATING, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS iii
WUP PSAR GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENEPATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM ;
12 RADIATION PROTECTION VI l
12.1 SHIELDING l 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERSTIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS O
iv 1
l
WUP Amendment 16 PSAR 11/78 O
PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 8 Page, Table (T) Revision Ficure (F1 Number 8-i -
8-ii through 8-iii 8 8-iv 16 8.1-1 7 S.1-2 through 8.1-6 -
8.1-7 through 8.1-9 4 8.2-1 through 8.2-2 2 8.3-1 -
8.3-2 7 8.3-2a 6 8.3-3 through 8.3-4 16 8.3-5 -
8.3-6 9 8.3-7 7 8.3-8 6 8.3-9 through 8.3-10 -
8.3-11 2
['
8.3-12 through 8.3-T2a 6 8.3-13 through 8.3-14a 6 8.3-15 through 8.3-16 16 8.3-17 -
8.3-18 15 8.3-19 16 8.3-20 through 8.3-21 7 8.3-22 16 8.3-22a 2 8.3-23 through 8.3-24b 16 8.3-25 1 8.3-26 through 8.3-28 15 8.3-29 through 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 -
T.8.3.1-1 -
8.3.1-2 7 T.8.3.1-3 (page 1) through (page 5) 5 T.8.3.1-3 (page 6) 10 T.8.3.1-3 (page 7) through (page 8) 5 T.8.3.1-3 (page 9) 10 F.8.1-1 -
F.8.1-2 7 F.8.2-1 4 F.8.3.1-1 7 A F.B.3.1-2 -
! A F.8.3.1-3 through F.8.3.1-4 7 b P.8.3.1-5 through F.8.3.1-6 16 F.B.3.2-1 -
WUP Amendment 8 PSAR 6/16/75
() CHAPTER 8 LIST OF TABLES Table Title 8.3.1-d Major Electrical Equipment 8.3.1-2 Onsite A-C Power System Loads 8.3.1-3 Emergency Diesel Generator Loading O
l O
8-iii
WUP Amendment 16 PSAR 11/78 CHAPTER 8 LIST OF FIGURES Figure Title 8.1-1 Main 345 kV Transmission Network for WUMS - January 1981 8.1-2 Offsite Power Connections and Reserve Station Service Transformers Connections 8.2-1 Power Connections Between Switchyard and Onsite Standby Power System 8.3.1-1 Electrical One Line Diagram 8.3.1-2 One Line Diagram Vital Bus and 125 V D-C Connections 8.3.1-3 Local Assignment of Electrical Penetrations 8.3.1-4 Plan View Electrical Penetrations 8.3.1-5 Physical Separation of Onsite Dis-tribution System 8.3.1-6 Physical Separation of Onsite Dis-tribution System 8.3.2-1 One Line Diagram Nonsafety-Related 125 V D-C System and 120 V A-C Instrument Bus O
8-iv
WUP Amendment 16 PSAR 11/78
() Certain safety-related loads may receive power from either eme/gency 4,160 V bus H2O or emergency 4,160 V bus H40. A system of Anterlocks is provided so that the redundant emergency systems ar never tied together.
The emergency 4,160 V buses consist of indo< tal-clad switchgear located within the control building whict aismic Category I, tornado protected structure. Thes, are physically and electrically segregated so that any si _uure which might affect one bus will not jeopardize prop tration of the other bus as shown on Fig. 8,3.1-5. Tht ructure l16 contains an automatic fire protection system as described in 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 outside 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.
2 The breakers could not be inadvertently tripped because access to the emergency switchgear are 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 switchgear or 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 O from five nonsafety-related single ended unit substations con-sisting of dry type transformers and associated metal-clad 8.3-3
WUP Amendment 16 PSAR 11/78 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 sup? lies two safety-related unit substations sized to meet nafety-related load requirements. In no case will unit substations fed from different 4,160 V emerrency 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 's) supplied from the normal or emergency 480 V unit substations. Tne motor control centers are self-supporting metal-clad structures with ccmbination magnetic, reversing or nonreversing motor starters and molded case air circuit breakers.
Motcr st..rters have built-in 480-120 V transformers for control circuit power.
Emergency unit substations and emergency MCC's are located within S They are physically separated i
l M (eismic F' CategoryandI structures.
'.3.1-5 8 . 3.1 -6 ) so that any single failure which mig iect one bun will not ieopardize proper operation of the ot) a.
Tes_ 3 Inspections Preoperational tests and inspections are perforned to demonstrate that components are correct, and properly mounted; all connections are corre ct , and continuous; components are operational, and metering; and protective devicos are properly calibrated and adjusted.
l 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 Ibnits and 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 dete cted; 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.
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 " test" position and tested functionally. Protective relays are tested 8.3-4 ,
I
)
WUP Amendment 16 PSAR 11/78
() The test program consists of factory and onsite tests, as defined in IEEE-387-1972 (Ref . 2) , with clarification of factory test No. 3 as noted below.
Factory tests consist of the following:
- 1. Break-in run for a length of time (determined by the diesel generator manufacturer) adequate to pass through the initial failure period of the diesel generator,
- 2. Starting test,
- 3. Load acceptance test - test load blocks and sequence as limited by diesel generator manufacturer's capability,
- load test, load test,
- 6. .1jection test,
- 7. Electrical test, i
- 8. Functional test.
O The onsite test program is in accordance with Regulatory Guide 1.41 "Preoperational Testing of Redundant Onsite Electric Power Systems To Verify Proper Load Group Assignments."
Onsite tests consist of the following acceptance test series:
- 1. Starting test,
- 2. Load acceptance test, l
- 3. Rated load test, 16 l 4 .. Design load test, l
l
- 5. Load rejection test,
- 6. Electrical test,
- 7. Functional test,
- 8. Multiple start test.
The periodic test program is run in accordance with Regulatory Guide 1.108 " Periodic Testing of Diesel Generator Units Used as
( Onsite Electric Power Systems at Nuclear Power Plants." Periodic tests are run as follows:
- 1. Avai? ability test - Immonth intervals, 8.3-15
WUP Amendment 16 PSAR 11/78 4 2.. Operational test performed during each unit refueling h shutdown.
Only one diesel jenerator is allowed to be tested at any given time. During the test, all required safety features equipment is available on the redta3 ant emergency bus.
Each of the diesel generators are located in a separate Seismic Category I and tornado protected room. Each diesel generator unit has its own independent auxiliary systems.
The onsite fuel oil storage capacity provides for at least 7 days operation of minimum ESF and shutdown loads for each unit.
The Applicants conclude that the onsite standby power system satisfies GDC 17 and 18, IEEE-308-1971 (Ref . 3) , and Regulatory 4
Guides 1.6, 1.9, 1.41, and 1.108 is acceptable.
Diesel Generator Auxiliary Systems The diesel generator fuel oil storage and transfer system for each diesel consist of: (1) an underground diesel generator fuel oil storage tank, containing sufficient oil for 7 days of continuous operation following a DBA, (2) two full capacity electric motor driven diesel generator fuel oil transfer pumps, and (3) a 500 gal fuel oil day tank.
To detect the inadvertent presence of water in the fuel oil tanks, the Applicants will: (1) install near the bottom of each tank a float actuated level switch which will be responsive to the difference in density of water and oil, (2) install moisture sensors in each underground fuel oil storage tank discharge lines, and (3) sample the fuel oil in each storage tank on a monthly basis to check for the presence of water.
Each diesel generator has two air starting systems which can be cross-connected; each is sized for a five start capability without outside power. A separate self-contained closed water-to-water cooling system with two direct diesel driven jacket water circulation pumps, and electric immersion heater, and an a-c motor driven circulating water pump are used to maintain engine jacket water temperatures for reliable rapid startups.
The diesel generator lubrication system of each diesel generator includes two independent lubricating oil pumps direct driven from the diesel generator, and an a-c motor driven lubricating oil pump. The a-c motor driven lubricating oil pump supplies warmed lubricating oil to the diesel engine sump and other necessary components when the diesel is not running for reliable fast startups.
O 8.3-16
l WUP Amendment 16 l PSAR 11/78 l
() 6. Combustible Gas Control System (Section 6.2.5) .
- 7. Instrumentation and cables associated with Items 1 through 6 The procurement specification for each component of the systems in Items 1 through 6 defines the maximum expected radiation level ,
at the place where the equipment is installed and the expected '
life-time dose under which the equipment is required to function.
These procurement specifications also define other environmental design parameters for the specified equipment (e.g. , temperature and humidity) . Sufficient design margins are incorporated into the procurement specifications su that the equipment is capable of withstanding the most severe environmental conditions without loss of its safety function. . Whenever a procurement specification calls for an equipment or materials qualification test, documentation is required from the manufacturer to establish the satisfactory completion of that test.
Materials and equipment required to- operate inside the containment during and after a DBA are specified to include considerations of pressure, temperature, humidity, chemistry, and radiation levels. The procurement specifications define a radiation dose equal to the 40 year normal accumulated dose plus the DBA dose.
Loss of Ventilation (Ref er to Section 3.11.4)
To ensure that loss of the air conditioning and/or ventilation system will not adversely affect the operability of safety-related control and electrical equipment located throughout the plant, the environmental system for these areas will meet the single failure criterion (Section 9.4) . ,
8.3.1.2.2.2 Qualification Tests (Refer to Section 3.11.2)
Electrical Penetrations Electrical penetrations will be designed, tested, and documented in accordance with IEEE-317-1976 (Ref. 4) and Regulatory l g Guide 1.63.
Containment Isolation Valve Actuators (Ref er to Section 3.11.1)
The valve actuators are designed to operate at 60 psia pressure, 280 F temperature, and integrated radioactivity exposure or 1.3 x 106 rad with 100 percent humidity. The containment isolation valves are air-operated, fail-closed valves with solenoid pilot valves. The srlenoid valves are nigh temperature coils with watertight housings.
8.3-19
-. _ . _ _ . _ . . , . _ . _ _ _ . . . . _ . _ . . _ . ~ , , _ _ . _ _ . _ . . - _ . . _ . _ _ _ . _ _
WUP Amendment 7 PSAR 6/6/75 Qualification Tests for Cables h Cables in the containment which may be required to function during and af ter a DBA are qualified in accordance with IEEE-383-1974 for the DBA environment of temperature, pressure, humidity, chemical spray, and radiation.
Cable insulation and jacket material is selected to operate in the environments of normal operation or that of the post-accident period, 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.
Power cables have either an overall flame retardant jacket or are installed in ducts and conduits.
Control and instrument cables will be single or multiple conductor with an overall flame retardant jacket. Fillers are flame retardant and nonwicking.
Motors Continuous duty Class I motors located inside the containment, are designed, tested, and documented in accordance with IEEE-334-1971 (Ref. 5) and Regulato ry Guide 1.40.
All motors ror safety-related equipment located inside the containment have been conservatively rated and will have a service factor of 1.0 or 1.15. The insulation systems are Class B or better. All motors are given at least the standard NEMA, MGI houtine Tests and Class IE motors are certified to start the specified load with 70 percent rated nameplate voltage.
Electric Valve Operators Electrically-operated valves in the containment which may be required to function during and after a DBA are type tested for the post-accident environment of temperature, pressure, humidity, chemical spray, and radiation.
Electrically-operated valve insulation material is selected to operate in the environment during the post-accident period, as required.
Electrically-operated valves inside the containment are designed to withstand the normal radiation dosage and a superimpo sed DBA radiation dosage, as well as the post-accident environment.
Stone & Webster scope valve operators will be type tested in accordance with IEEE-382-1972 (Ref . 6) and Regulatory Guide 1.73 and Westinghouse scope valve operators will be qualified in 8.3-20 J
WUP Amendment 7 l PSAR 6/6/75 l [
\
accordance with the generic resolution- being pursued by Westinghouse and the NRC staff.
l Qualifica' tion Test Results '(PSAR) (Refer to Section 3.11.2) l The results of the qualification tests for each type of equipment will be provided in the FSAR.
8.3.1.3 Conformance With Appropriate Ouality Assurance Standards
.The safety-related portions of the onsite a-c standby. power system are classified as QA Category I. The QA procedures .used during equipment- design, fabrication, shipment, field storage, field installation, system and component checkout, and the records pertaining to each of these during the construction and preoperational test phases of each unit, are described in Chapter 17.
This QA Program, as discussed in Chapter 17, is in conformance with IEEE-336-1971 (Ref. 7) .
8 . 3 .1. 4 Independence of Redundant Systems 8.3.1.4.1 Principal Criterion The principal design criterion that establishes the minimum requirements for preserving the independence of redundant Class IE electrical systems through physical arrangement and separation and for assuring the minimum required equipment availability during any design basis event (Class IE electric system and design basis events are as defined in IE".E-3 08-1971) is as follows:
Class IE electrical equipment is physically separated from its redundant counterpart or mechanically protected as required to prevent the occurrence of common failure modes.
8.3.1.4.2 Administrative Responsibility for Compliance The administrative responsibility and control provided to assure compliance with the criteria that establish the minimum requirements for preserving the independence of redundant Class IE electrical systems during design and construction is presented in Chapter 17, particularly the QA procedures described in Section 17.1.1.3.
O 8.3-21
WUP Rmendment 16 PSAR 11/78 8.3.1.4.3 Equipment Consideration Design features of the major components of the Class IE system to ensure conformance with IEEE 'a08-1971 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 of the a-c station service system are I 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 16l Separated by barriers (Figs . 8.3.1-5 and 8.3.1-6). These areas )
are protected from the maximum probable flood as discussed in i 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 '
satisfies the independence requirements of Regulatory Guide 1.b.
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 located 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.
8.3-22
WUP Amendment 16 PSAR 11/78 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 do not occupy the same tray 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 re g'lired. 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. Redundant safety-related cables are run in separate raceways. Separation distances between these redundant raceways is described in O 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.
l16 Cables in the containment which are required to f unction during l
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 selected to operate in the environments of normal operation or that of the post-accident period as recuired.
16 Electrical Penetrations There are eighty-two penetrations. Four 20-in. penetrations are for the reactor coolant pump motor leads. Thirty-six 12-in.
penetrations are for nonsafety-related power, control, and instrumentation cables. Twenty-one 12-in. pene trations are for safety-related power, control, and instrumentation cables associated with Train A and Channels I and II. The remaining 16 twenty-one 12-in. penetrations are for safety-related power, control, and instrumentation cables associated with Train B and Channels III and IV. ]
Penetration areas, both inside and outside the containment, are O reserved for electrical cables and their supports.
having a potential for damaging electrical cables is installed in No piping 8.3-23
WUP Amendment 16 PSAR 11/78 penetration areas. Redundant penetrations do not occupy the same group. The redundant penetrations are separated from each other as well as from the nonsafety penetrations by fire barriers. The anticipated physical arrangement and assignments of circuits is 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 chamhcr to a pressure not less than Pa and monitors the penetration for leakage (10CFR50 Appendix J, paragraphs IIIB.1 (c) and IIIB.3 (b) ) .
If the pressure maintained is less than Pa Type B testing will be performed every other refueling shutdown or at an interval not greater than 3 years (10CFR50 Appendix J, paragraph IIID.2) .
g Wherever practical, the electrical penetrations are designed to withstand a single f ailure of their overcurrent protection as required by IEEE 317-1976 and Regulatory Guide 1.63. The fault current resulting f rom an overcurrent protection failure will not cause the electrical penetrations to fail mechanically, thus there will be no loss of containment integrity.
Where it is not practical to provide electrical penetrations i capable of withstanding severe overcurrents, the normal overcurrent protection schemes will be modified to prevent severe overcurrents through the use of fast acting, backup systems.
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 abilit" of the penetration to withstand fault current is a significant f actor 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 v ich 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 supply breaker should the appropriate feeder breaker fail to trip. l 8.3-24
WUP Amendment 16
?SAR 11/78 For the cases in which 480 V penetrations connected to motor control centers (MCC 's) cannot withstand the failure of a molded case circuit breaker, the following is done for:
- a. The load center breaker feeding the MCC is coordinated to trip if the molded case circuit breaker fails to trip (the penetration is sized to withstand the overload condition until the backup 16 device trips ); or
- b. Fuses are connected in series with the molded case circuit break at the MCC load terminals.
- 2. Non Class IE MCC's Fuses 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.
Sharing of Cable Trays In general, nonsaf ety-related cables do not share the same cable Ox trays with safety-related cables. In some instances, however, it may be necessary that cables for nonsafety-related circuits be 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 3 l 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 integrity of circuits for redundant safety-related services. The areas and systems are as describe d in Section 9.5.1. Fire protection equipment is provided in the following areas:
Emergency Switchgear Room Cable Tray Spreading Room 8.3-24a l . . .
l WUP Amendment 16 PSAR 11/78 Reactor Building Cable 5?ault and Tunnel !
Diesel Generators Rooms The fire hazard to safety-related cables is reduced by the following provisions. Control and instrument cables installed in i the cable tray or in areas common to safety-related cables have i 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 naintenance craftsmen so "he safety-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 er channel involved. Additional markers may be attached to the cable at intervals so that routing can be verified during construction.
O O
8.3-24b
WUP Amendment 16 PSAR 11/78
() The following example describes the method of identifying an item of mechanical equipment.
Example: 1SWP*P13A "1" stands'for the unit number "SWP" stands for service water system
"*" indicates safety-related equipment "P" stands for pump (motor)
"13A" signifies the unique identification number T
The "A" after 13 indicates that one or more other identical pump motors exist in parallel service (1SWP*P13B etc.). It does not refer to a safety channel or train.
Groups of safety-related equipment within the same load group, train or channel, located together (i.e., in cabinets, racks, control boards, etc.) are marked as a group with the appropriate ,
color code.
This identification scheme of color coding and the alphanumerical equipment identification system makes it evident to the operator or maintenance craftsman, without the necessity for consulting any reference material, whether the equipment, cabling, etc., is safety-related and, if safety-related, which train or channel is involved. -
References
- 1. American National Standard Institute, " Application Guide for A-C High Voltage Circuit Breakers," ANSI C37.010, 1964.
- 2. IEEE-387, " Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations," 1972.
- 3. IEEE-308, " Standard Criteria for Class IE Electric Systems for Nuclear Power Generating Stations," 1971.
- 4. IEEE-317, " Electrical Penetrations Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations,"
1976. lg
- 5. IEEE-334, " Trial Use Guide for Type Tests of Continuous Duty, Class I Motors Installed Inside the Containment- of Nuclear Power Generating Stations," 1975. lg
- 6. IEEE-382, " Trial Use Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations," 1972.
O 8.3-31
WUP Amendment 7 PSAR 6/6/75
- 7. IEEE-336, " Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations," 1971.
7l 8.3.2 D-C Power Systems 8.3.2.1 Description 8.3.2.1.1 General Each unit is designed with five separate 125 V d-c systems. Each of the four safety-related systems consists of a power supply, distribution system and load groups arranged to provide power to the safety-related 6-c loads. The fifth d-c system provides power to the nonsafety-related d-c loads.
Sufficient physical separation and electrical isolation are provided to prevent common failure modes in the redundant safety-related systems.
8.3.2.1.2 Safety-Related Systems Each of the four safety-related systems consists of a battery, battery charger and main distribution bus. The f ollowing major safety-related loads are fed from these systems:
- 1. Engineered safety features actuation system, O
- 2. 120 V vital bus inverters,
- 3. Control power for the emergency a-c buses,
The d-c systems will be operable for at least 2 hr following a loss of all a-c power.
8.3.2.1.3 Nonsafety-Related System The nonsafety-related system also consists of a battery, battery l charger and main distribution bus. It supplies power only to the i nonsafety-related loads such as:
- 1. Turbine a>1xiliary d-c motors.
I 1
llh l l
l 8.3-32 l
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g ll l ll (RED E WHITE) TRAYS 480V NORM LOAD CTR AND TRAIN A U P T O EL. 26'-4" gI '
4lll4 l (OR ANGE) TRAYS 5 TO WATER SERVICE l l
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NOTES:
t DIESEL GENER ATOR LE ADS IN RIGID STEEL CONDUlf ;
EMBEDDED IN FLOOR SL AB TO RESPECTIVE SWITCHGE AR ;
6 2 COLOR CODESt SEE SECT B.S .I.4.4 CH I M CH III CH N RED WHITE BLUE YELLOW TRAIN A TRAIN B ORANGE PURPLE f f RED L ORANGE CABLES MAY BE RUN IN SAME ARE A EL E CTRIC AL BLUE 6 PURPLE CABLES MAY BE RUN IN SAME AREA PEN ETR ATIONS CH I & II AN D WHITE CABLES ARE SEPARATED FROM RED L ORANGE BY [
TR AIN A APPROPRIATE FIRE BARRIERS OR PHYSICAL SEPARATION !
p, , YELLOW CABLES ARE SEPAR ATED FROM BLUE D PURPLE BY
- - APPROPRIATE FIRE BARRIERS OR PHYSICAL SEPARATION ELECTRICAL
-i PENETRATIONS
- % CH. III & II AND
- - TRAIN B C _._._
i.. NON-SAFETY EL ECT RICAL P E N E TR ATIO NS
, F I G. 8.3.1- 5 i PHYSICAL SEPAR ATION OF
. ONSITE DISTRIBUTION SYSTEM WISCONSIN UTILITIES PROJECT PRELIMIN ARY S AFETY ANALYSIS REPORT I t
AMENDMENT 16 l 1
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F I G. 8.3.1 - 6 PHYSICAL SEPAR ATION OF ONSITE DISTRIBUTION SYSTEM WISCONSIN UTILITIES PROJECT }
PRELIMIN ARY SAFETY AN ALYSIS REPORT I AMENDMENT 16
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 9 Page, Table (T) or Revision Page, Table (T) or Revision Piqure (F) Number Fiqure (F) Number 9-1 through 9-11 15 9.3-10 through 9.3-12 -
9-111 through 9 vi -
9.3-13 through 9.3-14 15 9-vii 15 9.3-15 through 9.3-16a 1 9-viii through 9-ix -
9.3-17 through 9.3-46 -
9-x 149 9.3-47 2 x-lia 16 9.3-48 through 9.3-51 -
9-xiib 14 9.3-52 9 9-xii 15 9.3-53 through 9.3-54 7 9-xiv through 9-xvi 8 ;
9-xvii 15 9.4-1 13 9-xviii 8 9.4-2 15 9-xix 14 9.4-3 through 9.4-6a 14 9-xx 16 9.4-7 2
~
9.4-8 through 9.4-9 9.1-1 through 9.1-3 -
9.4-10 3 9.1-4 through 9.1-4a 2 9.4-11 4 9.1-5 through 9.1-6 5 9.4-12 through 9.4-12a 3 9.1-7 -
9.4-13 through 9.4-14 -
9.1-8 1 9.4-15 through 9.4-16a 2 e 9.1-9 through 9.1-11 -
9.4-17 0 9.1-12 through 9.1-12a
('d 2 -
9.4-18 9.1-13 through 9.1-15 -
9.4-19 2 9.1-16 2 9.4-20 -
9.1-17 through 9.1-23 -
9.4-21 11 9.1-24 through 9.1-24b 4 9.4-22 3 9.1-25 -
9.4.23 15 9.4-24 7 9.2-1 -
9.4.25 through 9.4.26 15 9.2-2 through 9.2-4 9 9.4.26a 7 '
9.2-5 4 9.4-27 2 i 9.2-6 -
9.4-28 through 9.4-28a 6 i 9.2-7 through 9.2-Ba 4 9.4-29 -
l 9.2-9 9 9.4-30 14 1 9.2-10 through 9.2-10a 2 9.2-11 through 9.2-12 -
9.5-1 through 9.5-26 14 9.2-13 through 9.2-14 2 9.5-27 through 30 16 9.2-15 through 9.2-16 -
9.5-31 through 9.5-3fl 2 9.2-17 2 9.5-39 through 9.5-40 6 9.2-18 through 9.2-18a -
9.5-41 4 9.2-19 through 9.2-22 -
9.5-42 6 9.2-23 through 9.2-24 5 9.5-42a 3 9.2-25 2 9.5-43 2 9.2-26 through 9.2-26a 14 9.5-44 through 9.5-44a 3 9.2-27 6 9.5-45 3 9.2-28 through 9.2-28d 9 9.5-46 through 9.5-48a 15 9.2-29 1 9.5-49 through 9.5-51 2 9.2-30 2 T.9.1-1 6 9.3-1 3 T.9.2.1-1 2 9.3-2 -
T.9.2.1-2 -
/ 9.3-3 2 T.9.2.1-3 1 9.3-4 through 9.3-4a 3 T.9.2.2.1-1 (2 pages) -
9.3-5 1 T.9.2.2.1-2 (page 1) 1 9.3-6 through 9.3-6a 3 (page 2) -
9.3-7 through 9.3-8 -
T.9.2.5-1 4 9.3-9 16 T.9.3.2-1 16 EP.9-1
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIFF OF EFFECTIVE PAGES (CONF'D)
Chapter 9 Page, Table (T) or Revision Page, Table (T) or Revision Fiqure (F) Number Fiqure (F) Number T.9.3.4-1 -
F. 9 . 5.1 -4 15 T.9.3.4-2 (7 pages) -
F.9.5.1-5 16 T.9.3.6-1 through T.9.3.6-7 -
P.9.5.4-1 -
T.9.5.4-1 1 F.9.5.5-1 6 T.9.5.8-? 15 F.9.5.6-1 6 T.9.5.8-2 15 F.9.5.7-1 2 F.9.5.8-1 2 F.9.1.1-1 through F.9.1.1-2 -
F.9.5.9-1 -
F.9.1.2-1 15 F.9.1.3-1 15 F.9.1.3-2 through F.9.1. 3-3 (deleted) 15 F.9.1.4-1 through F.9.1.4-3 -
F . 9.1. 4 -4 9 F.9.1.4-5 through F.9.1.4-6 -
F.9.2.1-1 through F.9.2.1-3 15 F.9.2.2.1-1 through F.9.2.2.1-3 15 F.9.2.2.2-1 -
F.9.2.2.2-2 15 F.9.2.2.2-3 (deleted) 15 F.9.2.2.3-1 -
F.9.2.2.5-1 -
F.9.2.4-1 2 F.9.2.5-1 9
( F.9.2.5-2 2 i F.9.2.5-3 through F.9.2.5-7 4 i F.9.2.5-8 6 l F.9.2.6-1 0 F.9.3.1-1 through F.9.3.1-2 -
P.9.3.2-1 through F.9.3.2-2 15 F.9.3.3-1 through F.9.3.3-2 -
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 P.9.3.6-2 through F.9.3.6-3 (deleted) 9 l F.9.4.1-1 15 F.9.4.1-2 -
l F.9.4.1-3 (deleted) 14 F.9.4.2-1 through F.9.4.2-2 13 F. 9. 4. 4 -1 13 F.9.4.5-1 -
F.9.4.6-1 13 F.9.4.7-1 13 F . 9. 4 . 7 -2 3 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 P.9.5.1-1 15 F.9.5.1-2 14 F.9.5.1-3 -
EP. 9-2
I 1
l i
WUP Amendment 16 l PSAR 11/78 l 1
() CHAPTER 9 AUXILIARY SYSTEMS TABLE OF CONTENTS (CONT'D)
Section Title Page l
9.5.1.8 Fire Protection Organization 9.5-27 1 9.5.2 Communications Systems 9.5-28 9.5.2.1 Design Bases 9.5-28 9.5.2.2 System Description 9.5-31 9.5.2.3 Safety Evaluation 9.5-33 9.5.2.4 Test and Inspection Requirements 9.5-33 9.5.3 Unit Lighting System 9.5-33 9.5.3.1 Design Bases 9.5-33 :
9.5.3.2 System Description 9.5-34 9.5.3.3 Design Evaluation 9.5-35 9.5.3.4 Test and Inspection Requirements 9.5-35 9.5.3.5 Failure Analysis' 9.5-35 9.5.3.6 Instrumentation Applications 9.5-36 9.5.4 Diesel Generator Fuel Oil Storage and Trans-fer System 9.5-36 .
9.5.4.1 Design Bases 9.5-36 9.5.4.2 System Description 9.5-36 ,
9.5.4.3 Safety Evaluation 9.5-37 9.5.4.4 Test and Inspection Requirements 9.2-38 9.5.4.5 Instrumentation Applications 9.5-38 9.5.5 Diesel Generator Cooling Water System 9.5-39 9.5.5.1 Design Bases 9.5-39
() 9.5.5.2 System Description 9.5-40 9-xiia
WUP Amendment 14 PSAR 5/26/78 NPM 9 AUXILIARY SYSTEMS TABLE OF CONTENTS (CONT ' D)
Section Title Page 9.5.5.3 Safety Evaluation 9.5-40 9.5.5.4 Test and Inspection Requirements 9.5-40 9.5.5.5 Instrumentation Applications 9.5-40 9.5.6 Diesel Generator Starting System 9.5-41 9.5.6.1 Design Bases 9.5-41 9.5.6.2 System Description 9.5-41 9.5.6.3 Safety Evaluation 9.5-42 9.5.6.4 Test and Inspection Requirements 9.5-42 9.5.6.5 Instrumentation Applications 9.5-42 O
O 9-xiib
WUP Amendment 14 PSAR 5/26/78 LIST OF FIGURES (CONT 8 D)
Figure Title 9.3.4-6 Chemical and Volume Control System - Process Flow Diagram j 9.3.4-7 Chemical and Volume Control System - Process Flow Diagram 1
9.3.4-8 Chemical and Volume Control System - Process Flow Diagram 1 9.3.4-9 Chemical and Volume Control System - Process Flow I Diagram l 9.3.6-1 Boron Recovery System 9.3.6-2 Deleted 9.3.6-3 Deleted 9.4.1-1 Control Building Air Conditioning, Heating, and Emergency Ventilation
\' 9.4.1-2 Control Building Chilled Water System 9.4.1-3 Deleted 9.4.2-1 Reactor Plant Ventilation 9.4.2-2 Waste Disposal Building Ventilation 9.4.4-1 Turbine Building Ventilation System 9.4.5-1 Fuel Building Normal Ventilation 9.4.6-1 Service Building HVAC System 9.4.7-1 Containment Air Filtration System 9.4.7-2 Containment Purge Air System 9.4.7-3 Containment Air Circulation 9.4.7-4 CRDM Cooling 9.4.8-1 Diesel Generator Building Heating and Ventilation
-Systems 9.4.9-1 Service Water Pump House Ventilation 9-xix
., . - . _ ~ - . . . . _ , . . . . _ . . . , _ . - . . . , . - . . . ~ . ~ . . . _ . . . . _ _ _ _ . - _ _ _ . . . _ . _ . _ _ _ _ . _ _ . . _ _ . ~ . _ . _ . . . _ - - . _
WUP Amendment 16 PSAR 11/78 LIST OF FIGURES (CONT'D)
Fiqure Title 9.4.10-1 Deleted 9.5.1-1 Fire Protection Arrangement - Yard Plan 9.5.1-2 Fire Protection Arrangement - Safety-Related Areas Piping 9.5.1-3 Fire Prote'ction - Foam 9.5.1-4 Carbon Dioxide Fire Protection System 9.5.1-5 Offsite Management Fire Protection Organization 9.5.4-1 Diesel Generator Fuel Oil System 9.5.5-1 Typical Diesel Engine Cooling Water System 9.5.6-1 Typical Diesel Engine Air Starting System 9.5.7-1 Typical Diesel Engine Lubrication System 9.5.8-1 Reactor Plant Gas Supply System 9.5.9-1 Containment Leakage Monitoring System O
9-xx
WUP '
Amendment 16 PSAR 11/78
/)
\/
(Section 10.4.8) . The steam generatorblowdownsamples,whichl4 check on primary-to-secondary leakage, are taken at the reactor plant sample sink. The steam generator blowdown sample radiation monitor is discussed in Section 11.4.2.1.
Provisions for local sampling for the safety injection accumulators are made. These samples are low temperature and do not require sample coolers. Provisions for local sampling for the two auxiliary boilers are provided. These samples are high temperature and are cooled by turbine plant component cooling water.
Except for the pressurizer relief tank gas space sample sampling lines coming from within the containment structure contain high temperature samples. The sample coole for the high temperature samples reduce the temperature to 140 F or less for safe handling. Sample flows leaving each sample cooler are manually throttled and can be directed to a purge line or to the sampling sink. After sufficient purging, a pressurized sample is obtained by isolating the sample in a sample capsule which is removed from the sampling line for analysis.
An air-operated sample valve is installed in each individual sampling line within the containment structure. These air-operated valves are also installed in certain sampling lines (Fig. 9.3.2-2) .
(~ which do not penetrate the containment structure
( ,}/ Each of these valves can be remotely operated from a sample control panel in the sample room in the auxiliary building.
Steam generator blowdown sampling lines are connected to the steam generator blowdown lines outside the containment structure and downstream of the steam generator blowdown line containment isolation valves.
In addition to the above facilities for periodic sampling, there are facilities in the sampling systems for continuous radiation monitoring of the steam generator blowdown samples and continuous oxygen, pH, hydrazine, conductivity and/or sodium monitoring of the condensate and feedwater systems (Section 10.4.7) .
Local instrumentation at the sample sinks is provided to permit manual control of sampling operations and to ensure that the samples are at suitable temperatures and pressures before diverting flow from the purge line to the sample sinks. Each purge line contains a flow indicator to indicate purge flow rate.
9.3.2.3 Safety Evaluation Except for the sampling line containment isolation valves, which close on a CIA signal, the sampling systems are not required to function during an emergency condition, nor are they required to function to prevent an emergency condition . However, as a 7 s}
(
- minimum, the sampling systems are designed as described in Section 9.3.2.1 to maintain the pressure boundary integrity of 9.3-9
WUp PSAR the systems to which they are connected up to the last isolation valve.
If a reactor coolant sampling line becomes inoperable due to a malfunction, there is at least one alternate path which can be used to obtain a similar sample.
Radiological evaluations for normal operation of the sampling system are provided in Chapter 12, and Sections 9.3.6, 10.4.8, 11.2, 11.3, and 11.4.
9.3.2.4 Test and Inspection Requirements Sampling system components are used regularly during power operation, cooldown, and shutdown. Routine maintenance is performed to ensure the availability of the sampling systems.
The continuous radiation monitors and automatic analyzers are periodically tested, calibrated, and checked to ensure proper instrument response and operation of alarm and warning functions.
The sampling system containment isolation valves are tested as stated in Section 6.2.4.
9.3.2.5 Instrumentation Application Sampling system instrumentation is located adjacent to the sample sinks in appropriate sample rooms.
Temperature and pressure indicators are provided at all sample stations in the sample rooms.
Pressure indicators, upstream and downstream from the high pressure throttling valves on high pressure sampling lines, are provided.
Flow indicators are furnished on purge lines to indicate the sample flow rates.
Where necessary, relief valves are furnished to protect equipment from overpressure.
The steam generator blowdown sample radiation monitor automatically isolates the steam generator blowdown system (Section 10.4.8) from the makeup and blowdown system (Section 10.4.12).
All continuous automatic analyzers and radiation monitors indicate and alarm in the appropriate sample rooms. A common alarm is provided in the control room for the continuous analyzers in the turbine plant sampling system and for the steam generator blowdown analyzers. The steam generator blowdown radiation monitor is alarmed and indicated separately in the control room as part of process radiation monitoring system h (Section 11.4) .
9.3-10
WUP Amendment 16 PSAR -
11/78 f~\
d alarm and annunciate in the control room are provided for these 34 areas.
9.5.1.8 Fire Protection Organization 1he fire protection organization is shown schematically on Fig. 9.S.1-5. The solid line indicates the fire protection i organization line of authority and the dashed line indicates the overall administrative lines of authority.
President, Wisconsin Electric The President, Wisconsin Electric, has both the overall administrative and fire protection responsibilities for the nuclear plant. He has charged the Executive Vice President (Nuclear) with the administration of the nuclear plant and the Senior Vice President with fire protection responsibilities.
Executive Vice President The Executive Vice President directs the activities of management personnel in the Nuclear Projects Office and the nuclear plant to ensure that the health and saf ety of the '6 public are provided for in the design and operation of all systems at the nuclear plant.
O Senior Vice President The Senior Vice President directs the activities of management personnel in the Quality Assurance and Technical Services Department who ensure that fire protection programs are implemented and adhered to; including programs at the nuclear facilities as required by Federal, State ,and insurance organizations.
Fire Protection Officer The System Fire Protection Officer is responsible for directing the formulation and implementation of fire l protection programs. He coordinates the activities of the l Nuclear Projects Office, the nuclear plant and the Insurance and Claims Department, as they pertain to the fire protection analysis, design,and administration.
l l
Once established, all programs are audited periodically by the Fire Protection Officer. The Fire Protection Officer reports to the 9irector, Quality Assurance and Technical Services on all matters pertaining to these programs.
l l e's l
i l
9.5-27
WUP Amendment 16 PSAR 11/78 Superintendent, Insurance and Claims The Superintendent, Insurance and Claims, coordinates the requirements of the insurance organizations (Nuclear Mutual Limited and NEI,-PIA) with the Fire Protection Officer as they pertain to the nuclear plant.
g Administrator, Nuclear Proiects Office The Administrator, Nuclear Projects Office, directs fire protection engineering analysis, design,and licensing for the nuclear plant. He utilizes fire protection design consultants as required to provide these services. The Administrator and Fire Protection Officer jointly review and approve tt~ analysis and design of ire protection systems utilized at the nuclear plant.
9.5.2 Communications Systems 9.5.2.1 Design Bases 14 The communication systems are designed to provided reliable communications between all essential areas of the station and to essentia) locations remote from the plant during normal operations or under emergency conditions. This capability is provided by diverse communication system types.
O 9.5-28
N WUP Amendment 16 PSAR 11/78 O ,
1 l
1 l
l THESE PAGES INTENTIONALLY LEFT BLANK l
l )
9.5-29/30
s WUP Amendment 16 PSAR 11/78 TABLE 9.3.2-1 REP?JTE SAMPLES Typical Typical System Operating System Operating Pressure (psia)(*D Temperature (F) 4 8 3 The following samples are collected at the reactor plant sample sink located in the auxiliary building:
- 1. Pressurizer vapor space sample 2,235 653
- 2. Pressurizer liquid space sample 2,235 653
- 3. Two reactor coolant hot leg samples 2,235 618 4 Reactor coolant filter inlet sample 250 115
- 5. Letdown heat exchanger outlet sample 237 115
- 6. Boron thermal regeneration demineralizers outlet samples 230 50-140
- 7. Volume control tank gas space sample 20-75 115
- 8. Two residual heat exchanger outlet samples 450 400 6
- 9. Charging pump discharge 2,485 115
- 10. Pressurizer relief tank gas space sample 3 120
- 11. Degasifier liquid effluent sample 170 140
- 12. Degasifier condenser process gas effluent sample 2 215
- 13. Process gas receiver sample 75 120 14 Reactor plant component cooling water system sample 150 105 lg
- 15. Three steam generator blowdown samples 949 540 l
%e following samples are collected at the turbine plant sample sink located in the turbine building:
l
- 1. Three individual min steam lines samples 915-1,095 535-558
- 2. Steam generator feedwater pump suction manifold sample 200 395
- 3. Fe9 water heater No. 1 outlet manifold sample 1,200 435
- 4. Condensate pump discharge manifold sample 490 90 l S. Four condenser hotwell samples (as pumped) 25 90 Note:
417 Refer to specific system descriptive sections for more exact inf ormation.
Normal operating data herein are estimates provided for the purposs of review only.
1 of 1
- O O O PRESIDENT l
q l l
, TREASURER SENIOR VICE PRESIDENT EXECUTIVE VICE PRESIDENT l
1 l
., l l DIRECTOR OF QUALITY ASSURANCE
' l __
AND TECHNICAL SERVICES l i l
l I l
1 I I
SUPERINTENDENT OF INSURANCE SYSTEM FIRE PROTECTION PROJECTS ADMINISTRATOR
- AND CLAIMS (NEL-PIA AND NML) 0FFICER l NUCLEAR PROJECT OFFICE 4
l
- r r
(
FIRE INSUR ANCE CONSULTANTS, FIRE PROTECTION l DESIGNERS AND INSPECTORS NUCLEAR POWER PLANT DESIGN CONSULTANT l
FIG. 9. 5.1 - 5
- - - ADMINISTRATIVE ORGANIZATION FIRE PROTECTION ORGAN 12 ATION OFFSITE MANAGEMENT FIRE PROTECTION ORGANIZATION i
WISCONSIN UTILITIES ' PROJECT
, PRELIMINARY SAFETY ANALYSIS REPORT A M END M ENT 16
- a -- - r _ _- - . . , , ,-e
l l
WUP PSAR GENERAL TABLE OF CONTENTS (CONTeD)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTDiS I
6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
i 7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-REIATED AND POWER GENERATION DISPIAY INSTRUMENTATION O 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTE!!
8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONING , HEATI!G, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS x
iii
WUP PSAR GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 10 STEAM AND POWER COlWERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 R A D I O A C T I V E W A S T E M A N A G E M E h"r V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIEIDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS iv
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 10 Page, Table (T) or Revision Page, Table (T) or Revision Fiqure (F) Number Fiqure (F) Humber 10-i through 10-11 -
F.10.4.3-1 -
10-111 14 F.10.4.4-1 5 10-iv 8 F.10. 4 . 5-1 3 10-v 14 F.10.4.7-1 -
10-vi 8 F.10.4.7-2 5 10-711 16 F.10.4.7-3 -
F.10.4.8-1 16 10.1-1 -
F.10.4.8-2 (deleted) 9 F.10.4.9-1 through F.10.4.9-2 15 10.2-1 through 10.2-3 -
F .10 . 4 . 9-3 15 10.2-4 through 10.2-4c 6 F.10 . 4.10 -1 8 10.2-4d 15 F.10.4.11-1 through F.10.4.11-3 -
10.2-5 -
10.2-6 2 10.2-7 16 10.3-1 12 10.3-2 through 10.3-2a 1 10.3-3 6 10.3-4 1 10.3-5 -
\- ' 10.4-1 through 10.4-2 -
10.4-3 15 10.4-4 5 10.4-5 through 10.4-7 -
10.4-8 through 10.4-Ba 1 10.4-9 5 10.4-10 6 10.4-10a 2 10.4-11 5 10.4-12 through 10.4-13 -
10.4-14 through 10.4-14a 15 10.4-15 -
10.4-16 through 10.4-20 14 10.4-21 15 10.4-22 3 10.4-23 through 10.4-24 7 10.4-25 through 10.4-26a 9 10.4-27 -
10.4-28 5 10.4-29 through 10.4-30 15 T.10.1-1 -
T.10.2-1 (2 pages) 6 T.10.2-2 (2 pages) 6 T.10.4-8-1 through T.10.4.8-2 14 T.10.4.10-1 -
T.10.4.11-1 -
F.10.1- 1 through F.10.1-3 -
F.10.2-1 through F.10.2-6 2 F.10.2-7 through F.10.2-9 6
,C) P.10.3-1 5 F.10.3-2
(]
F.10.3-3 through F.10.3-4 5 1
F.10.4.2-1 15 EP.10-1 1
1
, -- - ~ - - . . -
UUP Amendment 16 PSAR- 11/78 LIST OF FIGURES (CONT'D) ,
Figure Title 9
10.4.7-3 Feedwater Heater Drain 10.4.8-1 Steam Generator Blowdown System 10.4.8-2' Deleted 10.4.9-1 Auxiliary Steam System 10.4.9-2 Auxiliary Condensate System ,
10.4.9-3 Auxiliary Boiler Steam and Condensate and Feedwater System 10.4.10-1 Auxiliary Feedwater System 10.4.11-1 Turbine Plant Component Cooling Water System i 10.4.11-2 Turbine Plant Component Cooling Water System 10.4.11-3 Turbine Plant Component Cooling Water System O ;
i l
l J
l i
i 10-vii
WUP Amendment 16 !
PSAR 11/78 O valve is slowly reopened while the other valves close, maintaining constant load on the unit. Limit switches and control valve position instrumentation provide indications that '
the stop and control valves have been reopened. The program then continues with similar testing of the remaining turbine valves. 2 The LP turbine valves are also periodically tested. Each LP stop and control valve is closed and reopened in a predetermined sequence in a manner suitable for checking the proper operation of these valves. The turbine valves are tested at regular intervals as required by the technical specifications.
References '
- 1. " Turbine Missile Analysis," Allis--Chalmers Power Systems, Inc., Engineering Report ER503, July 1975, Amended December '
1, 1975. 16
- 2. " Probability of Turbine Missiles," Allis-Chalmers Power Systems, Inc., Engineering Report ER504, October 1975.
- 3. " Quality Control of Large Castings and Forgings for Steam ,
Turbine Generators," 1971 American Power Conference, '
Proceedings, Vol. 33. ;
" Forgings O 4.
881,000 lb Acceptance From Weight,"
Gigantic Ingot with 140" Diameter and Criteria,"
Part 2 Paper
" Operational Presented at Stresses the and 1972 2 International Forge Masters Meeting, Cherry Hill, New Jersey.
- 5. "New Experimental Technique to Determine Residual Stresses in Large Turbine-Generator Components," Paper presented at the 1974 American Power Conference Chicago, Illinois.
- 6. "HP Rotor Strength and Stress Data," Allis-Chalmers Power Systems, Inc., Rev. A., December 12, 1974.
- 7. Bush, S.H., " Probability of Damage to Nuclear Components Due 6 to Turbine Failures," Nuclear Safety, Vol. 14, No. 3, May-June 1973. I a
i O
10.2-7
..,....-..----,--.-..-,..-.-.-.----.~..-.-----.------N
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FIG.9.3.2-2 2
l Fl M v d
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46 h SC2 V ..SC2h-hNS STEAM CfNERATOR INSIDE MAIN DUTSIDEMAIN i
$ VALVE
- STE 2 V E VE HOUSE INSIDE CONTAINMENT +4- DUTSIDE CONT AINMENT 10 CONTAINMENT SUMP FlG.9.3.3-4 NOTE:
THIS SYSTEM 13 NON-NUCLEAR SAFETY CLASS (NNS) EXCEPT AS NOTED.
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' TO CONDENSATE SYSTEM Q ,
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FIG.10. 4. 8 - 1 l STEAM GENERATOR BLOWDOWN SYSTEM !
WISCONSIN UTILITIES PROJECT r PRELIMINARY SAFETY ANALYSIS REPORT }
AMENDMENT 16 l l
l
WUP Amendment 16 (
PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 11 Page, Table (T) or Revision Page, Table (10 or Revision Figure (F) Number Fiqure (F) Numter 11-1 15 (deleted) 11-i1 through 11-x 14 T.11.6.3-1 (2 pages) 10 T.11.6.4-1 15 11.1-1 through 11.1-10 14 F.11.1-1 -
11.2-1 through 11.2-2 14 F.11.2-1 through F.11.2-2 14 11.2-3 through 11.2-5 15 F.11.2-3 5 11.2-b through 11.2-10 14 F.11.2-4 14 F.11.2-5 through F.11.2-9 5 ,
11.3-1 through 11.3-2a 14 F.11.2-10 (deleted) 14 ,
11.3-3 through 11.3-6 5 P.11.2-11 15 11.3-7 through 11.3-8 14 F.11.2-12 through F.11.2-13 9 (deleted) 11.4-1 through 11.4-3 14 F.11.3-1 14 11.4-4 through 11.4-5 -
F.11.3-2 through F.11.3-6 5 11.4-6 through 11.4-8 14 F.11.3-7 through F.11.3-8 14 11.4-9 -
F.11.5-1 through F.11.5-3 14 P.11.5-4 through F.11.5-7 5 11.5-1 through 11.5-7 14 F.11.6.1-1 through F.11.6.1-3 -
F.11.6.2-1 through F.11.6.2-8 14 11.6-1 through 11.6-3 -
(deleted) 11.6-4 through 11.6-6 14 F.11.6.2-9 through F. 11.6.2-20 9 ,
, (deleted)
T.11.1-1A (4 pages) 16 T.11.1-1B (page 1) (page 2) 14 T.11.1-1B (page 3) (page 4) 16 T.11.1-2A (2 pages) 14 i T.11.1-28 14 T.11.1-3 through T.11.1-4 -
T.11.1-5 (3 pages) 14 r T.11.1-6 (3 pages) 14 T.11.1-7 14 T.11.2-1 (2 pages) 14 T.11.2-2 (2 pages) 14 T.11.2-3 (2 pages) 14 T.11.2-4 through T.11.2-24 14 (deleted)
".11.2-25 through T.11.2-26 -
T.11.2-27 5 T.11.2-28 through T.11.2-29 14 (deleted)
T.11.2-30 (page 1) 2 (page 2) 1 T.11.3.3-1 (2 pages) 5 T.11.3.6-1 (3 pages) 14 T.11.3.6-2 (2 pages) 14 T.11.3.6-3 through T.11.3.6-7 14 T.11.3.9-1 (deleted) 14 T.11. 4 - 1 (2 pages) 14 T.11.5-1 14 .
T.11.5-2 (2 pages) 14 T.11.6.1-1 through T.11.6.1-3 -
T.11.6.1-4 (2 pages) -
T.11.6.1-5 -
'T.11.6.1-6 (2 pages) -
T.11.6.2-1 through T.11.6.2-4 14-EP.11-1
l WUP Am ndment 16 '
PSAR 11/78 i
() TABLE 11.1-1A REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS FOR DESIGN CASE i
Isotope uCi/ce uCi/cm Kr-83M 3.13E-01 4.44E-01 Kr-85M 1.15E+00 1.63E+00 g Kr-85 1. 97E-02 2.79E-02 Kr-87 8.93E-01 1.27E+00 Kr-88 2.31E+00 3.28E-00 Kr-89 7.73E-02 1.10E-01 Xe-131M 6.38E-03 9.05E-03 Xe-133M 3.60E-01 5.11E-01 Xe-133 1.52E+01 2.16E+01 Xe-135M 8.31E-01 1.18E+00 Xe-135 1.92E+00 2.72E+00 Xe-137 1.25E-01 1.77E-01 Xe-138 5.12E-01 7.26E-01 :
Total g Noble 2.37E+01 3.36E+01 Gases Br-83 5.98E-02 8.48E-02 -
Br-84 3.06E-02 4.34E-02 '
Br-85 4.11E-03 5.83E-03 Br-87 2.49E-03 3.53E-03 I-129 2.82E-08 4.00E-08 I-131 1.52E+00 2.16E+00 I-132 6.14E-01 8.71E-01 I-133 2.56E+00 3.63E+00 g I-134 4.14E-01 5.87E-01 I-135 1.47E+00 2408E+00 I-136 4.59E-03 6.51E+03 '
i Total Halogens 6A68E+00 9.47E-00 g '
Cr-51 6.79E-04 9.63E-04 Mn-54 5.64E-04 8.00E-04 Mn-56 2.14E-04 3.03E-04 l16 Fe-59 7.86E-04 1.11E-03 g Co-58 1.86E-02 2.64E-02 Co-60 5.50E-04 7.80E-04 l16 Total g Corrosion 2.14E-02 3.03E-02 Products 1 of 4
WUP Am:ndment 16 PSAR 11/78 TABLE 11.1-1A (CONT
- D)
REACTOR COOIANT EQUILIBRIUM CONCENTRATIONS FOR DESIGN CASE Isotope uCi/cc uCi/qm Se-81 2.56E-04 3.63E-04 Se-83 4.28E-04 6.07E-04 Se-84 1.51E-04 2.14E-04 Rb-88 2.30E+00 3.26E+00 Rb-89 7.35E-02 1.04E-01 Rb-90 1.65E-02 2.34E-02 Rb-91 7.84E-02 1.11E-01 Rb-92 6.35E-03 9.01E-03 Sr-89 2.45E-03 3.47E-03 N Sr-90 1.01E-04 1.43E-04 Sr-91 1.28E-03 1.82E-03 Sr-92 5.31E-04 7.53E-04 Sr-93 3.32E-05 4.71E-05 Sr-94 7.47E-06 1.06E-05 Y-90 1.27E-04 1.80E-04 Y-91M 7.10E-04 1.01E-03 Y-91 3.94E-04 5.59E-04 r~s Y-92 4.88E-04 6.92E-04 s Y-93 2.18E-04 3.09E-04 Y-94 1.13E-05 1.60E-05 Y-95 8.19E-06 1.16E-05 Zr-95 4.07E-04 5.77F-04 Zr-97 2.61E-04 3.70E-04 Nb-95M 8.18E-06 1.16E-05 Nb-95 4.22E-04 5.98E-04 Nb-97M 2.50E-04 3.55E-04 Nb-97 2.80E-04 3.97E-04 Mo-99 2.25E+00 3.19E+00 l16 Mo-101 1.19E-02 1.69E-02 Mo-102 7.68E-03 1.09E-02 Mo-105 2.92E-04 4.14 E-04 Tc-99M 1.25E+00 1.77E+00 l 16 Tc-101 2.29E-02 3.25E-02 Tc-102 9.13E-03 1.29E-02 Tc-105 1.24E-03 1.76E-03 Ru-103 1.96E-04 2.78E-04 Ru-105 2.07E-05 2.94E-05 Ru-106 1.86E-05 2.64E-05 W Ru-107 1.18E-07 1.67E-07 Rh-103M 1.96E-04 2.78E-04 Rh-105M 2.07E-05 2.94E-05 Rh-106 1.86E-05 2.64E-05 Rh-107 6.35E-07 9.01E-07 l16 Sn-127 8.91E-04 1.26E-03
({]) in i
2 of 4
WUP Amendmant 16 PSAR 11/78 TABLE 11.1-1A (CONT *D)
REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS FOR DESIGN CASE g Isotope uCi/cc uCi/qm Sn-128 1.54E-03 2.18E-03 Sn-130 2.32E-04 3.29E-04 Sb-127 5.24E-03 7.43E-03 l16 Sb-128 3.98E-05 5.64E-05 Sb-129 1.37E-02 1.94E-02 Sb-130 1.82E-03 2.58E-03 Sb-131 4.85E-03 6.88E-03 g Sb-132 5.72E-04 8.11E-04 Sb-133 1.21E-03 1.72E-03 Te-127M 1.21E-03 1.72E-03 Te-127 1.07E-03 1.52E-03 l 16 Te-129M 2.24E-02 3.18E-02 Te-129 2.34E-02 3.32E-02 Te-131M 1.41E-02 2.00E-02 Te-131 7.52E-03 1.07E-02 Te-132 1.59E-01 2.2SE-01 g Te-133M 2.23E-02 3.16E-02 Te-133 O Te-134 Cs-134 3.71E-03 2.27E-02 2.14E-01 5.26E-03 3.22E-02 3.03E-01 Cs-136 1.14E-02 1.62E-02 Cs-137 9.54E-01 1.35E+00 l16 Cs-138 6.96E-01 9.87E-01 Cs-139 6.38E-02 9.05E-02 Cs-140 6.87E-03 9.74E-03 Cs-142 3.48E-03 4.94E-03 Ba-139 5.50E-02 7.80E-02 Ba-140 2.53E-03 3.59E-03 Ba-141 8.65E-05 1.23E-04 Ba ",42 3.45E-03 4.89E-03 La-140 7.85E-04 1.11E-03 La-141 1.85E-04 2.62E-04
'La-142 5.51E-07 7.81E-07 La-143 1.41E-05 2.00E-05 g Ce-141 3.98E-04 5.64E-04 Ce-143 3.14E-04 4.45E-04 Ce-144 2.82E-04 4.00E-04 Ce-145 1.51E-06 2.14E-06 Ce-146 5.12E-06 7.26E-06 Pr-143 3.85E-04 5.46E-04 Pr-144 2.82E-04 4.00E-04 Pr-145 1.08E-04 1.53E-04 Pr-146 1.37E-05 1.94E-05 O Nd-147 1.40E-04 1.99E-04 3 of 4
/
WUP Amendment 16 PSAR 11/78 TABLE 11.1-1A (CONT'D)
REACTOR COOLANT EQUILIBRIUM CONCENTRATIONS FOR DESIGN CASE ,
Isotope uCi/qm N uCi/cc Nd-149 1.32E-05 1.87E-05 Nd-151 7.99E-07 1.13E-06 Bn-147 3.94E-05 5.59E-05 Pm-149 6.09E-05 8.64E-05 Pm-151 2.15E-05 3. 05E-0 5 Sm-151 8.19E-09 1.16E-08 Sm-153 8.21E-06 1.16E-05 l5 Total Remainder 8.36E+00 1.19E+01 l4 O
O 4 of 4
WUP Amendment 14 PSAR 5/26/78
() TABLE 11.1-TB PRIMARY COOLANT AND SECONDARY SIDE RADIOACTIVITIES FOR NORMAL EXPECTED CASE Reactor Coolant Steam Generator Steam Generator Isotope .(uCi/qm) Liquid (uCi/qm) Steam (uCi/qm)
Noble Gases Kr-83M 2. 2E-02 0.0 7.2E-09 Kr-85M 8 . 8 E-0 2 0.0 2.9E-08 Kr-85 1. 7E-0 3 0.0 5.6E-10 Kr-87 6.8E-02 0.0 2.1E-08 Kr-88 1. 9E-01 0.0 6.1E-08 Kr-89 6.8E-03 0.0 2.2E-09 Xe-131M 4.4E-03 0.0 S %E-09 Xe-133M 3.4E-02 0.0 1.1E-08 Xe-133 1.4E+00 0.0 4.4E-07 Xe-135M 1.7E-02 0.0 5.5E-09 Xe-135 2.0E-01 0.0 6.4E-08 Xe-137 1.2E-02 0.0 4.0E-09 Xe-138 5.8E-02 0.0 1.9E-08 Halogens N Br-83 5.9E-03 2.0E-07 2. 0E-09 Br-84 3.5E-03 3.8E-08 3.8E-10 l Br-85 4.1E-04 4.7E-10 4.7E-12 I-130 2. 2E -03 1.5E-07 1.5E-09 I-131 2.3E-01 2.1E-05 2.1E-07 I-132 1.2E-01 5.6E-06 5.6E-08 I-133 3.7E-01 2.8E-05 2.8E-07 I-134 6.1E-02 1.0E-06 1.0E-08 I-135 2.1E-01 1.2E-05 1.2E-07 CS, RB RB-86 7. 3E-0 5 6.8E-09 6.8E-12 RB-88 2.7E-01, 1.7E-06 1. 7E-0 9 Cs-134 2.1E-02 2.0E-06 2.0E-09 Cs-136 1.1E-02 1.0E-06 1.0E-09 ,
Cs-137 1.5E-02 1.4E-06 1.4E-09 Water Activation Products N-16 4.0E+01 1.7E-06 1.7E-06 O
1 of 4
WUP Amendment 14 PSAR 5/26/78 TABLE 11.1-1B (CONT 8D)
PRIMARY COOLANT AND SECONDARY SIDE RADIOACTIVITIES FOR NORMAL EXPECTED CASE Reactor Coolant Steam Generator Steam Generator Isotope (uCi/qm) Liquid (uCi/qm) (uCi/qm)
Tritium H-3 1.0E+00 1.0E-03 1. 0E-0 3 ,
Other Nuclides Cr-51 1.6E-03 1.4E-07 1.4E-10 Mn-54 2.6E-04 3.2E-08 3.2E-11 Fe-55 1. 3E-03 1.3E-07 1.3E-10 Fe-59 8.3E-04 j 9.5E-08 9.5E-11 Co-5B 1.3E-02 (#gg 1.3E-06 1.3E-09 Co-60 1.7E-0 3 1.4E-07 1.4E-10 Sr-89 2.9E-04 3.2E-08 3.2E-12 Sr-90 8.3E-06 6.3E-10 6.3E-13 Sr-91 6.9E-04 3.9E-08 3.9E-11 Y-90 1.1E-06 1.3E-10 1.3E-13 g Y-91M 2.3E-11 O Y-91 Y-93 4.7E-04 5.3E-05 3.6E-05 2.3E-08 4.7E-09 1.9E-09 4.7E-12 1.9E-12 Zr-95 5.0E-05 6.3E-09 6.3E-12 Nb-95 4 . 2E-0 5 6.3E-09 6.3E-12 Mo-99 7.4E-02 6.7E-06 6.7E-09 Tc-99M 5.4E-02 6.1E-06 6.1E-09 Ru-103 3.8E-05 3.2E-09 3.2E-12 Ru-106 8.3E-06 6.3E-10 6.3E-13 Rh-103M 5.9E-05 4.5E-09 4.5E-12 Rh-106 1.4E-05 9.3E-10 9.3E-13 Te-125M 2. 4E-05 1.6E-09 1.6E-12 Te-127H 2.3E-04 1.6E-08 1.6E-11 Te-127 9.0E-04 5.9E-08 5.9E-11 Te-129M 1.2E-03 9.5E-08 9.5E-11 Te-129 2.1E-03 1.4E-07 1.4E-10 Te-131M 2. 3E-0 3 1.7E-07 1.7E-10 Te-131 1. 5E-03 4.6E-08 4.6E-11 Te-132 2.4E-02 1.7E-06 1.7E-09 Ba-137M 2.2E-02 2.1E-06 2.1E-09 Ba-140 1.9E-04 1.6E-08 1.6E-11 La-140 1.4E-04 1.2E-08 1.2E-11 Ce-141 5.8E-05 6.3E-09 6.3E-12 Ce-143 3. 7E-0 5 1.7E-09 1.7E-12 Ce-144 2.7E-05 3.2E-09 3.2E-12 Pr-143 4.2E-05 3.2E-09 3.2E-12 O' Pr-144 Np-239 4.4E-05 1.1E-0 3 4.6E-09 1.0E-07 4 6E-12 1.0E-10 2 of 4
WUP Am3ndment 16 PSAR 11/78 TABLE 11.1-1B (CONT 'D)
PRIMARY COOLANT AND SECONDARY SIDE RADIOACTIVITIES FOR NORMAL EXPECTED CASE Reactor Coolant Steam Generator Isotope (Ci) Liquid (Ci)
Noble Gases KR-83M 3.5E+00 0.0 KR-85M 1.4E+01 0.0 KR-85 2.GE-01 0.0 KR-87 1.1E+01 0;0 KR-88 3.0E+01 0.0 KR-99 1.1E+00 0.0 XE-131M 6.9E-01 0.0 XE-133M 5.3E+00 0.0 XE-133 2.1E+02 0.0 XE-135M 2.7E+00 0.0 XE-135 3.1E+04 0.0 l XE-137 1.9E+00 0.0 l XE-138 9.0E+00 0.0 Halogens 16 !
BR-83 9.2E-01 2.5E-05
-BR-84 5.4E-01 426E-06 l BR-85 6.4E-02 5.6E-08 i
! I-130 3.4E-01 1.8E-05 I-131 3.6E+01 2.5E-03 I-132 1.9E+01 6.7E-04 I-133 5.8E+01 3.4E-03 l I-134 9.6E+00 1a2E-04 I-135 3.3E+01 1.5E-03 CS, RB l RB-86 1.1E-02 8.1E-07 RB-88 4.2E+01 2.1E-04 CS-134 3.3E+00 2.4E-04 CS-137 2.42+00 1.7E-04 Water Activation Products N-16 6.2E+03 2.0E-04 Tritium H-3 1.6E+02 1.2E-01 3 of 4
WUP Amendmsnt 16 ,
PSAR 11/78 TABLE 11.1-1B (CONT *D)
PRIMARY COOLANT AND SECONDARY SIDE RADIOACTIVITIES FOR NORMAL EXPECTED CASE Reactor Coolant Steam Generator Isotope (Ci) Liquid (Ci)
Other Nuclides CR-51 2.5E-01 1.7E-05 MN-54 4. 0E-02 3.8E-06 FE-55 2.1E-01 1. 5E-0 5 FE-59 1.3E-01 1.1E-05 00-58 2.1E+00 1.5E-04 CO-60 2. 6E-01 1.7E-05 SR-89 4.6E-02 3.8E-06 SR-90 1. 3E-0 3 7.6E-03 SR-91 1.1E-01 4.7E-06 Y-90 1.7E-04 1.6E-08 Y-91M 7.4E-02 2.7E-06 Y-91 8.3E-03 5.7E-07 Y-93 5.6E-03 2.3E-07 ZR-95 7. 8E-03 7.6E-07 NB-95 6.5E-03 7.6E-07 16 O' NO-99 TC-99M 1.2E+01 8.4E+00 8.0E-04 7.3E-04 RU-103 5.9E-03 3.8E-07 RU-106 1.3E-03 7.6E-08 RH-103M 9. 2E-03 5.4E-07 RH-106 2.1E-03 1.1E-07 TE-125M 3.8E-03 1.9E-07 TE-127M 3.6E-02 1.9E-06 TE-127 1. 4E-01 7.0E-06 TE-129M 1.8E-01 1.1E-05 TE-129 3.2E-01 1.6E-05 TE-131M 3. 6E-01 2.1E-05 '
TE-131 2.3E-01 5. 5E-0 6 TE-132 3.7E+00 2.0E-04 BA-137M 3.4E+00 2.5E-04 BA-140 2.9E-02 1.9E-06 LA-140 2.1E-02 1.4E-06 CE-141 9.1E-03 7.6E-07 CE-143 5.BE-03 2.1E-07 CE-144 4.3E-03 3.8E-07 PR-143 6.6E-03 3.8E-07 PR-144 6.9E-03 5.5E-07 NP-239 1.7E-01 1.2E-05 O
4 of 4
HUP PSAR GENERAL TABLE OF CONTENTS (CONT 8D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS -
7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPIAY INSTRUMENTATION O 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONING , HEATIIG, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS O -
lii
WUP PSAR GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 10 STEAM AND POWER COINERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMEh"T V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL O MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS iv
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFFCTIVE PAGES Chapter 12 Page, Table (T) or Revision Piqure (F) Number 12-1 through 12-il 16 12-iii through 12-iv 14 12-v through 12-vi 16 12.1-1 through 12.1-2 14 12.1-3 2
'12.1-4 12.1-5 through 12.1-8b 16 12.1-9 9 12.1-10 through 12.1-12 -
12.1-13 2 12.1-14'through 12.1-14a 9 12.1-15 through 12.1-16 14 12.1-17 16 12.2-1 -
12.2-2 through 12.2-11 14 12.3-1 -
12.3-2 through 2a 16 12.3-3 2 12.3-4 through 12.3-6 -
T.12.1-1 9 T.12.1-2 through T.12.1-4 -
T.12.1-5 (page 1) 0 (page 2) 2 T.12.1-6 (2 pages) 2 T.12.1-7 (2 pages) 2 T.12.1-8 through T.12.1-12 2 T.12.2-1 through T.12.2-4 14 T.12.2-5 6 T.12.2-6 through T.12.2-B 14 F.12.1-1 16 F.12.1-2 11 F.12.1-3 through F.12.1-5 10 F.12.1-6 through F.12.1-11a 16 F.12.1-12 through F.12.1-15 2 F.12.1-16 through F.12.1-20 16 F.12.1-21 -
F.12.1-22 16 F.12.1-23 6 F.12.2-1 16 P.12.2-2 2 EP.12-1
.- . - - - . .-. . ~ .
WUP ' Amendment 16 PSAR 11/78
, CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS Section Title Page 12.1 SHIELDING 12.1-1 12.1.1 Design Objective .12.1-1 12.1.2 Design Description 12.1-3 12.1.2.1 Primary Shielding 12.1-8b 12.1.2.2 Secondary Shielding 12.1-9 12.1.2.2.1 Containment Structure Shielding 12.1-9 12.1.2.2.2 Auxiliary Equipment Shielding 12.1-10' 12.1.2.3 Fuel Handling Shielding 12.1-11 12.1.2.4 Accident Shielding 14.1-12 12.1.2.5 Control Room Shielding 12.1-12 12.1.2.6 Health Physics Station and Chemistry Laboratory Areas 12.1-12 12.1.3 Source Terms 12.1-13 12.1.4 Area Radiation Monitoring System 12.1-14 12.1.5 Operating Procedures 12.1-15 12.1.o Estimates of Exposure 12.1-16 12.2 VENTILATION 12.2-1
, 12.2.1 Design Objective 12.2-1 12.2.2 Design Description 12.2-1 12.2.3 Source Terms 12.2-2 12.2.4 Airborne Radioactivity Monitoring System 12.2-4 12.2.4.1 Containment Structure Monitoring System 12.2-5 12.2.4.1.1 Containment Structure Particulate Monitor 12.2-o 12-i
WUP Amendment 16 PSAR 11/78 CHAPTER 12 TABLE OF CONTENTS (CONT ' D)'
Section Title Page 12.2.4.1.2 Containment Structure Gas Monitor 12.2-b 12.2.4.2 Auxiliary and Waste Disposal Building -
Cubicle and General Area Air Monitors 12.2-6 12.2.4.2.1 Cubicle and General Area Particulate Monitors 12.2-8 12.2.4.2.2 Carbon Adsorber 12.2-8 12.2.4.2.3 Cubicle and General Area Gas Monitors 12.2-8 12.2.4.3 Sensitivity and Detectable Levels of the Cubicles and General Area Air Monitors 12.2-9 12.2.5 Operating Procedures 12.2-10 12.2.6 Estimates of Inhalation Doses 12.2-11 12.3 RADIATION PROTECTION PROGRAM 12.3-1 )
12.3.1 Organization and Objectives 12.3-1 12.3.2 Facilities and Equipment 12.3-2a 12.3.2.1 Controlled Area 12.3-2a 12.3.2.2 Radiation Protection Facilities 12.3-3 12.3.2.3 Special Shielding 12.3-4 12.3.2.4 Protective Clothing and Respiratory Equipment 12.3-4 12 .3.2.5 Portable and Laboratory Equipment 12.3-5 12.3.3 Personnel Dosimetry 12.3-5 12.3.3.1 Dosimetry Equipment 12.3-5 12.3.3.2 Processing and Recording 12.3-6 12.3.3.3 Internal Dosimetry 12.3-6 O
12-ii
WUP Amendment 16 PSAR 11/78 t
\~e CHAPTER 12, LIST OF FIGURES Figure Title 12.1-1 Radiation Zones, Access Control Points and Area Monitors Plot Plan 12.1-2 Radiation Zones, Access Control Points and Area Monitors Containment Structure El. 49'-7" 12.1-3 Radiation Zones, Access Control Points and Area Monitors Containment Structure El. 228-7" 12.1-4 Radiation Zones, Access Control Points and Area Monitors Containment Structure El. 08-6" 12.1-5 Radiation Zones, Access Control Points and Area Monitors Containment Structure El. 26 *-4 1/2" 12.1-6 Radiation Zones, Access Control Points and Area Monitors Auxiliary Bldg. Plan El. (-) 218 -6" 12.1-7 Radiation Zones, Access Control Points and Area
['-) Monitors Auxiliary Bldg. Plan Ej. 08-6" 12.1-8 Radiation Zones, Access Control Points and Area Monitors Auxiliary Bldg. Flan El. 268-4" 12.1-9 Radiation Zones, Access Control Points and Area Monitors Auxiliary Bldg. Plan El. 508-6" 12.1-10 Radiation Zones, Access Control Points and Area Monitors Auxiliary Bldg. Partial Plans 12.1-11 Radiation Zones and Area Monitors - Puel Building 12.1-11A Radiation Zones and Area Monitors - Fuel Building 12.1-12 Radiation Zones Access Control Points and Area Monitors Waste Disposal Building Plan El. O'-6" 12.1-13 Location and Shape ot Radiation Sources in the Containment Structure El. C'-6" 12.1-14 Location and Shape of Radiation Sources in the Containment Strdcture El. 26f-4 1/2"
/"} 12.1-15 Location and Shape of Radiation Sources in the
's,/ Containment Structure Section 2-2 12-v
WUP Amendment 8 PSAR b/10/75 CHAPTER 12 O1 l l
LIST OF FIGURLS (CONT'D)
I Figure Title l 12.1 16 Location of Airborne Radiation Monitors (ARM) and Shape of Radiation Sources in the Auxiliary Building (-) 22 '-6a 12.1-17 Location of Airborne Radiation Monitors (ARM) and Shape ot Radiation Sources in the Auxiliary Building El. On-6" 12.1-18 Location of Airborne Radiation Monitors (ARM) and Shape of Radiation Sources in the Auxiliary Building El . 26 '-4" 12.1-19 Location and Shape of Radiation Sources in the Auxiliary Building - Partial Plan Various Levels 12.1-20 Location and Shape of hadiation Sources in the Puel Building 12.1-21 Location and Shape of Radiation Sources in the Waste Disposal Building 12.1-22 Location and Shape of Radiation Sources Within the Plot Plan 12.1-23 Service Building Controlled Area Arrangement and Radiation Monitors (RM) 12.2 -1 Contalment Atmosphere Monitoring System 12.2-2 Radiation Monitoring of Reactor Plant, Containment St.ructure, Ebel Building Ventilation Systems O
12-vi
\
WUP Amendment 16 PSAR 11/78 m
( Penetrations, where practicable, are made high on shield walls to prevent direct streaming to normally occupied areas. Due to the l
j nature of the diffusion of radiation from penetrations, streaming effects are generally localized.
The bases for the selection of maximum radionuclide concentrations are described in Section 11. A radioactivity balance for the entire plant is first established with detailed computer modeling which uses mass flow data as input. This balance is used in assessing maximum expected releases from the plant. Once this balance is established, the sources in individual components are intentionally imbalanced to simulate worst case conditions. The levels of radioactivity under these conditions may thus be ten times the expected levels in some components. This approach is carried out on a component-by-component basis so that shielding in each local area is adequate for the worst conditions, e.g., radioactivity in a tank just prior to drainage. Dose rates computed under such conditions ensure that the maximum levels in each zone are not exceeded under normal desint onditions.
Major design f m as incorporated in plant design and layout to 0
minimize pei et exposure (Regulatory Guide 8.8) are as follows:
1.
Equipment and components, such as filters, tanks, pumps' l 16 and other process equipment, which are expected to be or
./
may be significant radiation sources, are located in separate shielded cubicles as far as practicable, and are shielded from other radiation sources to facilitate access for maintenance. Adequate laydown area is provided in and adjacent to cubicles to allow placement of temporary shielding if required. Access to cubicles is provided at all elevations for multi-level cubicles to allow rapid and convenient access and egress.
- 2. Instruments in radioactive systems which require in situ calibration are located as much as practicable on exterior walls of shielded cubicles to minimize exposure 0 of instrumentation and control personnel. Ins truments which cannot be located in this manner are located in the lowest practicable radiation area in the cubicles and are provided with convenient access. Where practicable, instruments are designed for removal to low radiation areas for calibration and maintenance.
l
- 3. Equipment and components located in cubicles are designed to be removable as far as practicable and adequate access is provided for removal or replacement.
As an example, pump cubicles are designed to allow fs removal of the pump to the lowest practicable radiation
( field.
12.1-5
WUP Amendment 16 PSAR 11/78
- 4. Diaphragm valves are used extensively in fluid systems to minimize maintenanc.s and packing requirements and to minimize maintenance time if diaphragm replacement is required. Where diaphragm valves cannot be used, valves are provided with leak-off connections and best available packing, to minimize maintenance requirements.
In specialized cases, bellows seal valves are used.
O Valves are located in separate shielded valve cubicles or areas outside equipment cubicles and pipeways as far I as practicable to minimize maintenance exposure. j
- 5. Design objectives for penetrations are described in Section 12.1.2 and include:
- a. Lines containing high levels of radioactivity do not penetrate shield walls in occupied areas as far gl as practicable.
- b. Piping and ductwork penetrations are provided with shielding where necessary and practicable to prevent streaming of radiation from pipeways to accessible areas.
- c. Shield walls are normally a part of the building structure and serve as internal supports in 0
addition to shielding.
- 6. Radiation sources are separated from normally occupied i
l areas by shield walls and cubicles as far as I practicable. Piping runs are located in shielded pipeways as far as practicable which are separated from equipment cubicles. Valves, as far as practicable, are located in shielded valve areas which are separated from equipment cubicles, pipeways, and areas of general access. Pipes containing highly radioactive sources are MI routed around, rather than through, normally occupied areas.
- 7. Sources of contamination from leaks or spills from components located in cubicles are prevented from spreading by cubicle entrance dikes or low point drains 0 to enclosed collection sumps. Gaseous leaks are prevented from spreading by the reactor plant ventilation system (Section 9) which maintains a slightly negative pressure inside equipment cubicles.
Air flow in all cases is designed to flow into individual pipeways and cubicles from general external M i areas. Floor surf aces and walls are sealed or painted as regaired with a protective chemically resistant 0
coating to provide a surface which is easily gl decontaminated. Systems containing radioactive fluids are fabricated of corrosion resistant saterials.
12.1-6
WUP Amendment 16 PSAR 11/78
('s 8. Where necessary and practicable, interior surfaces of piping and ductwork are designed to minimize contamination buildup. The location and layout of equipment cubicles and pipeways shown on Figs . 3.8. 4-1 through 3.8.4-10 allow piping layouts which minimize loops which could collect contamination. Pipe runs for spent resin sluicing are provided with five diameter bends rather than welded elbows to prevent accumulation 0 of resin fines and crud particles. Piping greater than 2-inch diameter is butt welded to minimize the potential for crud pockets. Design of ventilation systems incorporates long radius bends as necessary to minimize contamination buildup.
- 9. Means for flushing and draining of potentially highly radioactive tanks, lines, and other components are 16 considered in fluid system design. Waste collection tank design includes provisions for internal flushing ,
with spray nozzles to remove potential collections of particulate material. Flushing and vent connections are provided to allow flushing of piping systems for maintenance.
- 10. The ventilation systems are designed with suffi o capacity to control airborne radioactivity release .d
-~ concentrations during normal and maintenance cond a,
,j including periods of potentially increased flow .u th
'- open hatches or doorways. The ventilation flow through equipment cubicles is based on unrestricted air flow through the access to the cubicles. The design of ventilation systems ensures a positive flow from 16 noncontaminated areas to potentially contaminated areas to prevent the spread of airborne radioactivity.
- 11. The process and effluent, area and radiation monitoring systems indicate and annunciate locally and in the control room. Monitors are located throughout the plant as described in Sections 11.4, 12.1.4, and 12.2.4.
0
- 12. Ventilation system ductwork and equipment are located and arranged as far as practicable to provide for ease of maintenance, alteration, and filter changes .
Ventilation equipment is not expected to require decontamination but is located to enable decontamination or replacement.
- 13. As described in (6) above, and shown on Figs. 3.8.4-1 through 3.8.4-10, radiation sources are separated from normal and routine personnel access areas by shielding designed for maintaining doses ALARA. lg
/" 14. As described in Section 12.3.2.3, temporary shielding 0
inventories are maintained for use in manitenance or 12.1-7
WUP Amendment 16 PSAR 11/78 shielding of components based on results of radiation surveys and operating procedures. Convenient means for transport and placement are provided by hand or forklift 0 trucks.
- 15. As shown on Fig. 3.8.4-16, shielding of radioactive wastes in tanks, evaporators and other components and of 0 solidified waste in the waste disposal building is provided by equipment cubicles and waste storage areas gj designed to maintain ALARA.
- 16. Remote handling equipment is provided for filter 0
removal. Remote removal and replacement of ion exchange resins is provided. Miscellaneous remote handling tools for small sources are provided, as practicable. Other 16 remote handling equipment and controls are provided wherever appropriate in the design.
- 17. Shielding design is based on the design value of 0
1 percent fuel defects and primary to secondary leakage of 166 gpd at 120 F, where applicable and conservative.
- 18. Sampling is done at centralized sampling stations shielded from other radiation sources. Where piping to a centralized station is not practicable, local sample points are provided outside equipment cubicles and are 16l provided with suitable drain connections. Sample points in sample rooms are located inside an enclosed, ventilated hood with drainage to waste collection tanks.
O The sample hoods and normal access areas are shielded from sample cooling equipment. Local sample points and central sampling stations are located and designed to Ml maintain doses ALARA.
- 19. Provisions are made for controlled access for maintenance of equipment that may have high radiation levels to minimize exposure time during maintenance. In 16 addition, access to highly radioactive components is restricted and locked to prevent unauthorized access and exposure.
- 20. Shop and decontamination facilities for work on contaminated equipment are provided.
Plant, equipment, and piping layouts are developed under supervision of competr.nt personnel experienced in nuclear plant design, radiation protection, and operation in Applicants' and 0
architect-engineer's organizations. Changes to layouts are reviewed with regard to radiation exposure, maintenance, ul operability, and access. ALARA design concepts are used 0l throughout the design and construction of the plant.
O 12.1-8
l l
l WUP Amendment 16 PSAR 11/78 ALARA design reviews are scheduled in two phases: l
- 1. Conceptual design review
- 2. Detailed design review The conceptual design review begins during the early stages of conceptual plant layout and design. This review continues as the design is further aveloped. Detailed design reviews will be conducted concurrent.j with detailed design of plant systems and layout. The review will be performed as the design progresses to ensure that appropriate changes due to ALARA considerations can be considered and incorporated. Modifications occurring after completion of the detailed design review will be subjected to ALARA review of the modification itself and of any system it may impact. The Architect-Engineer documents findings, recommended actions, and resolutions. An ALARA review file is maintained and will be periodically reviewed to assure the resolution of any outstanding items. Since the relationship between the Applicants and the Architect-Engineer is highly interactive, conceptual design, layout design, and overall system design drawings are routed for the Applicants' approval. Radiation related items are subjected to ALARA review by the Applicants in the normal course of this routing. Responsibilities for ALARA review are as follows: 16
- 1. For the Architect Engineer, the Principal Nuclear fs Engineer is responsible for scheduling ALARA reviews.
Actual performance of the review is led by the Principal
(_) Nuclear Engineer and the Lead Radiation Protection Engineer, and is approved by the Radiation Protection Group Supervisor. Acceptance and implementation of the results of the review are by the Lead Power Engineer.
For project purposes, all Radiation Protection Group personnel are responsible to the Principal Nuclear Engineer. The Principal Nuclear Engineer is responsible to the Lead Power Engineer; the Lead Power Engineer reports directly to the Project Engineer, who is responsible for overall plant design. The Project Engineer is responsible to the client (Applicants) .
- 2. For the Applicants, ALARA reviews are accomplished in conjunction with design review by the Radiological Design Engineer and approval by the Senior Project Engineer. The Senior Project Engineer reports to the Project Administrator, who is responsible for overall plant design.
The minimum experience and education requir ements for the Architect-Engineer radiation protection personnel who participate in ALARA reviews are as follows:
- 1. The Radiation Protection Engineer holds a bachelor's O degree in a physical science or engineering and has k_ attended courses in radiation protection either as part 12.1-8a
WUP Amendment 16 l PSAR 11/78 of the degree program or in addition to it. This individual is familiar with radiation control instrumentation, exposure limits, shielding, dosimetry, and air sampling.
- 2. The Lead Radiation Protection Engineer meets all the requirements for Radiation Protection Engineer and, in addition, has at least 3 years of engineering experience requiring the implementation of radiation protection g principles. This experience should be in power plant systems similar to or associated with the systems being reviewed.
- 3. The requirements for Radiation Protection Group Supervisor are identical with those for Lead Radiation Protection Engineer, except that the experience requirement is at least 8 years.
For the Applicants, the educational background and experience of the Senior Project Engineer and the Radiological Design Engineer are provided in Sections 13.1.1.4.1.4 and 13.1.1.4.1.13, respectively.
12.1.2.1 Primary Shielding Primary shielding is provided to limit radiation emanating from the reactor vessel which includes neutrons diffusing from the core, prompt fission gammas, fission product gammas, and gammas resulting from the moderation and capture of neutrons.
The primary shield is designed to the following objectives:
- 1. Attenuate neutron flux to minimize activation of containment components and structures.
- 2. Reduce residual radiation from the core to a level which allows access to the region between the primary and secondary shields at a reasonable time after shutdown, and to permit limited access into the cubicle of an inactive steam generator during normal Operation.
l O
12.1-8b
-- _ --- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - __ _ _ _ __ A
- _- - - ._- ._ __- -.- ~. --
WUP Amendment 16 PSAR 11/78 Based on Wisconsin Electric Power Company operating experience, l 11 an estimate of annual man-rem exposure is given in Table 12.1-12. l0 The radiation shielding and equipment layout design bases are similar for Point Beach Nuclear Plant and the Wisconsin Utilities Project units. The following general design features minimize operator, maintenance and inservice inspe ction, personnel exposure, and are incorporated in both designs:
- 1. Equipment and components which are expected to be or may be significant radiation sources are located in separate shielded cubicles. Specific examples of this are the location of ion exchangers, filters, pumps, evaporators, and beat exchangers.
- 2. Valves are located in separate shielded areas to i
minimize operator and maintenance exposure. Diaphragm valves are used extensively in fluid systems.
- 3. Filters and ion exchangers require periodic renewal of filter cartridges and ion exchange resin. To minimize operator and maintenance personnel exposure, filters are designed and located for remote cartridge replacement. 16 Ion exchangers are designed and located for remote flushing and replacement of resins.
4 O 4. Components requiring periodic inservice inspection, such as steam generators, are provided with platforms and easy access as far as practicable to allow ease of ,
inspection. Inservice inspection is performed remotely as far as practicable to minimize inspection personnel exposure.
The plant staffing requirements for Point Beach and the Wisconsin Utilities Project units are nearly identical. Since both facilities incorporate the same major design features affecting radiation exposure, personnel exposure can be estimated from experience at Point Beach with allowances for a 3-loop versus a 2-loop unit where exposure can be expected to be loop-related.
The major loop-related exposures are inservice inspection of reactor coolant piping; refueling maintenance of valves and
- reactor coolant pumps; and inspection and/or repair of steam generators. For these exposures, values obtained from Point Beach experience have been increased by a factor of 3/2 in Table 12.1-12.
Reference
- 1. " Stone & Webster Radiation Shielding Design and Analysis Approach," Stone & Webster Engineering Report RP-8, May 1974. Corporation l11 12.1-17
_ . . - ._ _ _ _ _ ,_ _ _._ _. _._ _ . _ _. - . _ _ _ _ . - . ..___.._ _ _ .___.~ _.-
WUP PSAR l l
12.3 RADIATION PROTECTION PROGRAM 12.3.1 Organization and Obiectives Applicants' vorporate and plant organizations are described in Sections 13.1.1 and 13.1.2, respectively. The qualifications of Applicants' plant staff are stated in Section 13.1.3.2.
Applicants' management philosophy is to maintain radiation exposure of personnel at levels as low as practicable during normal operation and at levels within 10CFR20 at all times. This I is accomplished by implementation of an effective radiation protection program and review of the effectiveness of the program by upper management. ,
The plant organization includes the Chemistry and Health Physics Group which has overall responsibility for carrying out radiation protection activities in accordance with the radiation protection program. The radioctemical Engineer reports directly to the Plant Superintendent and exercises authority and supervisory control over the Chemistry and Health Physics Group. The Health Physicist normally 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.
The Health Physicist exercises supervisory control over all O~ radiation protection aspects of the plant and subject, to the Radiochemical Engineer's approval, exercises authority in these areas. The Health Physicist is normally assigned a Technical Assistant who provides technical support and performs supervisory duties as instructed by the Health Physicist.
Radiation surveys, removable contamination surveys (i .e . , smear surveys), air sampling, decontamination activities, and other radiation protection activities are performed by Radiochemical Technicians and Radiation Control Helpers as directed by the Health Physicist. Auxiliary Operators from the Operations Group are trained in radiation protection procedures by the Chemistry and Radiation Protection Group. These personnel perform radiation protection functions during periods when radiation protection personnel are not on site.
The Radiochemical Engineer and Health Physicist are responsible for initial radiation protection program formulation and obtaining the approval of the program by the plant Supervisory Staff. The approval program implementation is the responsibility of the Radiochemical Engineer and the Health Physicist and includes review of the effectivness of, and modifications to, the program.
The program is implemented in accordance with procedures e contained in the plant Radiation Protection Manual, a copy of which is given to all plant personnel.
12.3-1
WUP Amendment 16 PSAR 11/78 procedures addressing the practice of ALARA will be l Written included in the plant radiation protection manual and will address, at a minimum, the following topics:
(1) A firm management commitment to pract4 4 ALARA in the plant, (2) ALARA responsibilities of all plant employees, particularly the General Superintendent, the Radiochemical Engineer, and the Health Physicist, (3) ALARA responsibilities of corporate personnel in the Nuclear Projects Office, particularly the Radiological Design Engineer, (4) Provision for periodic ALARA audits by corporate personnel, (5) Provision for ALARA review of procedures, (6) Provision for ALARA review of proposed modifications at the early design stages, 16 (7) Provision for dose review by plant management and supervisory personnel, (8) Advice to employees on maintaining exposures ALARA.
The written program will address the bases for concluding that particular dose-reducing c'langes are not consistent with the ALARA principle. These bases will include:
(1) Cases where dose reduction in one area or situation has potential for increasing dose in another area or situation, (2) Changes which would result in a negligible reduction of I exposure, )
(3) Changes in which the benefit of exposure avoidance does not justify the costs of the change. Experience at Point Beach Nuclear Plant indicates that, in the ,
majority of ALARA considerations, cost-effectiveness or '
the lack thereof is immediately evident without detailed ;
analysis. In the event such conclusions are not ;
immediately evident, a cost analysis may be performed.
11 l Functions of tne Chemi stry and Health Physics Group include:
- 1. Ensuring that exposure limits in accordance with the approved radiation protection program are not exceeded by plant or visitor personnel on the site.
12.3-2 1
WUP Amendment 16 PSAR 11/78 9 2. Controlling radiological measures, radiation conditions (2) controlling exposure by:
and taking personnel (1) evaluating precautionary and equipment movement into and out of the Controlled Area, (3) ensuring proper use and care of special protective clothing and equipment described in Section 12.3.2, (4) conspicuously posting each area within the Controlled Area with appropriate caution signs and radiation conditions, and (5) administering and controlling conditions of radiation work permits for work in areas having high radiation and/or contamination levels in accordance with approved procedures.
- 3. Determining the requirements for and extent of use of personnel monitoring devices (Section 12.3.3) and maintaining records of personnel exposure. l l
- 4. Controlling and accounting for all radioactive material entering or leaving the plant site.
- 5. Establishing emergency evacuation plans (Section 13.3) . l
- 6. Training of plant and visitor personnel in radiation l rrotection policies and procedures, as required. l 12.3.2 Facilities and Equipment 12.3.2.1 Controlled Area The plant design establishes a controlled area which inicudes all areas in which radioactive materials are present or potentially present in quantities sufficient to require protective measures. l The controlled area includes all areas designed as Zone II or ,
greater (Section 12.1) . !
-O The normal access and egress point of the controlled area is through the health physics station corridor in the service building, to the auxiliary building, and is controlled by the O Chemistry and Health Physics Group. All other potential access points to the controlled area shown on Fig. 12.1-23 are kept locked or sealed. Temporary cc ntrolled access areas may belji established in the clean areat of the plant and are subject to all rules and procedures of the c ontrolled area. All materials .
are surveyed by radiation prote. '. ion personnel prior to removal from contolled access areas. A radiation monitor is provided at the access cont ol point and is 9
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l WISCONSIN UTILITIES PROJECT l l "6 ' N' #
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! l RADIATION ZONES, ACCESS CONTROL l
, POINTS AND AREA MONITORS l ELH 0- 6* AUXILARY BLOG. PARTIAL PLANS i PrJITIAL PL44 A WISCONSIN UTILITIES PROJECT l PRELIMIN ARY SAFETY AN ALYSIS REPORT '
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1 AMENDMENT 16 1
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. . RADIATION ZONES AND AREA MONITORS-PLAN EL42-C FUEL BUILDING i WISCONSIN UTILITIES PROJECT
( PRELIMIN ARY S AFETY AN ALYSIS REPORT
( AMENOMENT 16
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MONITORS (ARM) AND SHAPE OF U '[%S ~~ RADI ATION SOURCES IN THE AUXILI ARY
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r WISCONSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT ,
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LOCATION OF AIRBORNE RADIATION MONITORS (ARM) AND SHAPE OF ;
j uf 4**= RADIATION SOURCE IN THE AUXILI ARY
-g.h d BUILDING EL.O'-6" L __ . _6 --
n.( WISCONSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT ,
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-}- g l 1 --
l( MONITORS ( ARM) AND SHAPE OF R ADI ATION SOURCES IN THE AUXILI ARY f_ mt.'[. .l t_ _._-[
- J '"" ""' BUILDING EL.26'-4" l WISCONSIN UTILITIES PROJECT
< ue j PRELIMIN ARY SAFETY ANALYSIS REPORT l
(-e) f AWENOMENT 16
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= -j~' %,_,yn. mo LOCATION AND SHAPE OF RADIATION SOURCESIN THE AUXILIARY BLDG-PARTlAL PLAN VARIOUS LEVELS H 0-6 WISCONSIN UTILITIES PROJECT gy3 PRELIMINARY SAFETY ANALYSIS REPORT i
t
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SCA E - FEET FIG.12.1 - 20 LOCATION AND SHAPE OF RADIATION SOURCES IN THE FUEL BUILDING WISCONSIN UTILITIES PROJECT PRELIMINARY SAFETY ANALYSIS REPORT 1
l AMENDMENT 16
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..__ _ J dI gua FIG.12.1- 22 LOCATION AND SHAPE
- ---~
j 0F RA'DIATION SOURCES
/ WITHIN THE PLOT PLAN r
J WISCONSIN UTILITIES PROJECT
, , 7 .g ,p PRELIMNARY SAFETY ANALYSIS REPORT 5
%A.E f f11
+ = = AMENDMENT 16 '
rg WUP Amendment 16 i PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Chapter 13 Page, Table (T) or Revision Fge (P) Number 13-1 13 13-11 - 13-111 through 13-iv 8 13-v 14 13-vi -
13-vii 8 13.1-1 through 13.1-4 13 13.1-5 through 13.1-12 -
13.1-13 through 13.1-14 0 13.1-15 1 13.1-16 through 13.1-20 -
13.1-21 through 13.1-23 13 13.2-1 13 fr~g 13.2-2 through 13.2-3 -
- 13.2-4 3
\% ,- 13.2-5 -
13.3-1 through 13.3-2 -
13.3-3 through 13.3-4 16 13.3-5 through 13.3-8 -
13.4-1 through 13.4-2 -
13.5-1 through 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.13.1-4 -
- - - - . .~ . - - . - -
WUP Amendment 16 PSAR 11/78 j 13.3.3 Contacts with Agencies [
Contacts will be made with the following agencies to arrange for-assistance in emergencies:
- 1. The County Sheriff's Department ,
2.. Wisconsin State Police ,
- 3. The nearest General Hospital,
- 4. U. S. Coast Guard (if applicable),
- 5. Wisconsin Department of Health and Social Services ,#
- 6. U.S, Department of Energy, Region V Radiological l16 .
Emergency Assistance Team , '
- 7. U.S. Nuclear Regulatory Conmission, Region III Office of Inspection and Enforcement, 16
- 8. The nearest Fire Department,
- 9. The nearest County Civil Defense Department.
Means of coordination with these agencies, and interfaces with O
t individual agency mergency plans where applicable, will be !
included in applicable sections of emergency plans.
13.3.4 Emergency Measures Tb Be Taken Emergency measures to be taken will vary depending on the nature I of the incident. For incidents which might affect the safe '
condition of the plant an orderly shutt wn of the affected !
reactor will be made in accordance with technical specifications.
Emergency actions, other than plant shutdown, are anticipated to )
be as discussed below.
13.3.4.1 Maior Incidents Requiring Evacuation In the unlikely event of an incident requiring imediate evacuation of the plant site, the Emergency Evacuation Plan will be placed in effect. Nonessential plant personnel will proceed to an emergency center remote from the plant and remain until given further instructions. Essential plant personnel will proceed to the control room and will aid, as directed by the Duty Shift Supervisor, in evaluation of the incident and in directing site-activities. The Coordinator at the offsite emergency center will maintain direct communication with the control room and with offsite agencies responding to the incident.
The Emergency Evacuation Plan will include specific detailed actions to be taken to evacuate members of the public from the 13.3-3
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WUP Amendment 16 PSAR 11/78 site exclusion area, initiate site boundary monitoring activities and mobilize essential offsite personnel.
The Wisconsin Department of Health and Social Services will be notified and will respond in accordance with the Wisconsin Radiological Response Plan, which will include notification of the Wisconsin State Police and the County Sheriff's Department.
The U.S. Department of Energy Radiological Assistance ham and 16 the U.S. Nuclear Regulatory Commission Region III Office of Inspection and Enforcement will both be notified. Actions taken by each offsite agency will be prearranged and will be detailed in the Emergency Evacuation Plan.
13.3.4.2 Fire Emergency In the event of a fire on site, the Fire Emergency Plan will be placed in effect. The actions to be taken will depend on the severity and location of the fire and it is anticipated that plant fire brigade personnel and installed fire protection equipment will be adequate for any conceivable fire situation occurring within the plant restricted area. Personnel will be evacuated from the affected area and fire brigade personnel will carry out fire fighting actions using suitable protective clothing and equipment. In the event of need, or for a fire within the exclusion area but outside the restricted area, arrangements will be made with the nearest fire department to respond upon being notified.
It is not anticipated that offsite agencies other than the nearest fire department will be required to respond to plant fire emergencies unless the fire results in an incident within the scope of another plan (i.e. , an injured member of the plant fire brigade) .
13.3.4.3 Onsite Radiation Incidents If major spills or releases of radioactivity occur within the plant, the Onsite Radiation Incident Plan will be placed in
, effect. Installed plant ventilation, filtration, and collection systems are adequate for containment of major spills or releases
, of airborne radioactivity, except noble gases and significant l offsite releases will normally not result.
Personnel will be evacuated from the affected area and the area will be barricaded pending an evaluation by radiation protection personnel. Installed and portable radiation monitoring instrumentation will be used to evaluate and formulate recovery operations; and to evaluate the potential consequences of offsite releases, if any. In the unlikely event of general plant contamination, nonessential personnel evacuation may be initiated.
9 13.3-4
WUP Amendment 16 PSAR 11/78 I
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WUP Annde nt 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONT' D)
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WUP ,
PSAR GENERAL TABLE OF CONTENTS (CONT 'D) l 1
Chapter /
Section Title Volume ;
6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY O AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSI'IE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONING , HEATIIG, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS l
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GENERAL TABLE OF CONTENTS (CONT ' D)
Chapter /
Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM V 10.1
SUMMARY
DESCi'IPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL l MONITORING SYSTEMS j 11.5 SOLID WASTE SYSTEM i
11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM l
l 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS ,
1 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS .
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PRELIMIlOLRY SAFETY ANALYSIS REPOPT I
LIST OF w rm;TAVE PAGES Chapter 16 l
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17-iv 7 17-v 13 17-v1 7 17-vii 13 17.1 13 17.1-1 14 17.1-2 2 17.1-3 through 17.1-4 3 1 17.1-5 13 17.1-6 through 17.1-6a 14 .
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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)
WUP Amendment 14 PSAR 5/26/78 17.1.1.4 Procurement Document Control 17.1.1.4.1 General The implementation of measures for Procurement Document Control meeting the requirements of ANSI N45.2.13-1974 has been delegated l14 to the prime contractors for their respective scope of work.
Meanures are established and documented to provide assurance that appi.icable regulatory requirements, design bases, and other requirements which are necessary to assure adequate qual-ity are included or referenced in the WE and/or prime f
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\ making disposition, the design engineering representative must I determine that utilization of the item in the "as is" or repaired condition will not infringe upon the capability of the item to satisfactorily perform its intended function. Items dispositioned as scrap are deformed, defaced or clearly identified as unfit for their intended function and removed from the processing area as soon as possible.
Significant safety-related nonconformities detected during design, manufacturing, installation or operation, and which meet the criteria of 10CFR50.55 (e) will be identified and reported to the WE Project Administrator. Determination of the significance of the nonconformity shall be made by WE in accordance with Fig. 17.1.1-6.
Other nonconformities which have been dispositioned to "use as is" or " repair" but could conceivably adversely affect WE's use of the item with regard to expected life, performance, interchangeability, or interfaces, will also be identified and reported to the WE Nuclear Project Administrator. j 17.1.1.16 Corrective Action 17.1.1.16.1 General The implementation of measures for the identification of Os deficiencies and the implementation of corrective action has been delegated to the prime contractors for their respective scope of work. Procedures and practices are established and documented by the prime contractors, which provide assurance that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and eauipment, and nonconformances, are promptly identified, documented and corrected as soon as practicable, if action to correct the cause or condition is appropriate.
Corrective action documentation forms or formal letters are used to document the corrective action-related requests, responses and f ollow-up . Significant conditions adverse to quality, such as those which, if they were to remain uncorrected could adversely affect safety-related functions, are identified, the cause of the condition determined, and corrective action taken to preclude repetition. Procedures provide that such significant conditions, their causes, and the corrective action taken be documented and reported to the appropriate levels of the prime contractors' management.
Corrective action follow-up and close-out procedures provide for assuring that corrective action commitments are implemented in a syste natic and timely manner. These procedures are audited and verified by WE-QA personnel and/or their designated agent.
17.1-29
WUP Rmendment 16 PSAR 11/78 17.1.1.16.2 Recurrence Prevention WE and its prime contractors maintain surveillance on all aspects of the project during design, procurement, and construction to identify potential inherent weaknesses in the equipment or system which would lead to recurring problems. Weaknesses so identified will be analyzed by the cognizant organization for the cause and the required corrective action. Followup surveillance will be performed to ensure that the action taken has been effective in eliminating the problem.
17.1.1.17 Quality Assurance Records 17.1.1.17.1 General The implementation of measures for the assembly and storage of quality assurance records produced during the engineering, procurement, manufacture, construction, erection and testing phases has been delegated to the prime contractors for their respective scope of work and shall be commensurate with the U
requirements of ANSI N45.2.9, 1974 (including commitment to the regulatory position of Regulatory Guide 1.88). Procedures and practices are established and documented by WE and its contractors to provide assurance that sufficient records are prepared and accumulated as work is performed to furnish documentary evidence that the quality of items is satisfactory and that other closely related activities have been performed satisfactorily. Records shall be consistent with the requirements of the applicable codes, standard s , regulations, specifications, and shall be adequate for use in effective management of the applicable quality assurxnce programs.
17.1.1.17.2 Ouality Assurance Record Requirements The records include such documentation as the results of design reviews, inspections, tests, material analyses, audits, etc. The records also include closely related data such as qualifications of personnel, procedures, and equipment, and other documentation required by applicable codes and regulations. Inspection and test records, as a ndnimum, identify the date of inspection or test, the ins pector or data recorder, the type of observation, the results, the acceptability of the iten and the action taken in connection with any nonconformances noted. Records are identified and are retrievable while under the contractor and WE control for the retention period specified by the appropriate 13 regulations. Requirements and responsibilities for record generation, collection , storage, maintenance, retention and transfer to WE are established by procedures which are consistent with the applicable codes and standards, regulatory requirements, I and other requirements that may be established by WE. l For items of equipment important to safety for which there are no code requirements for record retention or transmittal, designated quality assurance records, such as: outline drawings; purchase 17.1-30 l
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i WUP Amendment 16 PSAR 11/78 meaningful and are effectively complying with the corporate policy and 10CFR50 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 are l meaningful and are effectively complying with the -l 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, j procedures and activities to assure compliance with all aspects of the QA program and procurement requirements.
- 4. Audits performed u2 the prime contractors, subcontractors, and vendors to verify and evaluate their l own and m1ppliers' QA programs, 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 O contents of this section can be found in the Stone & Webster Topical Report, SWSQAP 1-74A " Standard Nuclear Quality Assurance Program" Revision B dated August 1, 1976. QA Category I items l2 gg covered by this program are listed in Section 3.2.5 of this PSAR. l6 17.1.3 Westinghouse Nuclear Systems Divisions Quality Assural;ce Plan The contents of this section can be found in WCAP-8370, " Quality l2 Assurance Plan," Revision 7-A dated February 1975 with the lg 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 saf ety -related structures, systems, and components controlled by this program.
17.1-33
WUP Amendment 16 PSAR 11/78
- PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES Appendix A Page, Table (T) or Revision Fiqure (F) Number A-i through A-li -
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WUP Amendment 16 PSAR 11/78 O A.1-1.26 Quality Group Classifications and Standards (Regulatory Guide 1.26)
The classification system of ANSI 18.2 " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
as discussed in Section 3.2, is used as an alternate acceptable method of meeting the intent of Regulatory Guide 1.26.
A.1-1.27 Ultimate Heat Sink (Regulatory Guide 1.27)
The ultimate heat cink complies with Regulatory Guide 1.27, g Revision 2, issued January 1976.
Methods of compliance are discussed in Section 9.2.5.
A.1-1.28 Quality Assurance Program Recuirements (Design and Construction) (Regulatory Guide 1.28)
The design and construction of safety-related systems, structures, and components are subject to quality assurance g requirements that commit to meet the regulatory position of Regulatory Guide 1.28 issued June 1972. The Quality Assurance Program requirements that satisfy this position are described in Chapter 17.
A.1-1.29 Seismic Design Classification (Regulatory Guide 1.29) t The seismic classification of structures, systems, and components complies with the regulatory positions stated in Regulatory Guide 1.29, Revision 2, issued February 1976, as interpreted in l14 the following paragraphs.
Regulatory Position C.2 of this guide is not interpreted as requiring structures, systems, or components to be designated as Seismic Category I if their f ailure could adversely af fect the ;
performance of structures, systems, or components listed in !
Regulatory Position C.1. Rather, these items are designed and l constructed such that the safe shutdown earthquake (SSE) does not result in such interference due to failure. This does not necessarily imply that the items being considered must remain functional following the SSE, but only that their failure cannot adversely affect a Seismic Category I structure, system, or component. lH Regulatory Position C.3 is strictly interpreted to prevent a cascading effect in designating Seismic Category I items. Once the boundary of a Seismic Category I item is defined, the guide requires that the Seismic Category I requirements be extended to the first restraint beyond the boundary. This satisfies the requirement for the interface between Seismic Category I and nonseismic Category I structures, systems, or components. The interface conditions are therefore not considered a new Seismic
(
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WUP Amendment 16 PSAR 11/78 Category I boundary necessitating an additional Seismic h Category I interface.
Seismic classification is discussed in Section 3.2.
A.1-1.30 Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (Regulatory Guide 1.30)
The installation, inspection, and testing of all Class IE and IE electric power, instrumentation, control equipment and systems, g including auxiliary equipment and associated material, commit to meet the regulatory position of Regulatory Guide 1.30, issued August 1972. Specific quality assurance program requirements are defined in Charter 17.
A.1-1.31 Control of Stainless Steel Welding (Regulatory Guide 1.31)
Stone & Webster Scope of Supply The control of stainless steel welding will be in accordance with the interim NRC regulatory position on Regulatory Guide 1.31, Revision 1, pending formal revision.
Production welds (except: fillet welds having a throat dimension 3/8 inch or less; or where the thickness of the thinner member 2 being joined in full penetration welds is 1/4 inch or less; or where the nominal diameter of pipe is 2 inches and less; or for repair welding of austenitic stainless steel coatings; or for the welding of austenitic stainless steel to ferritic steel) are examined to verify that adequate delta ferrite levels are present, by magnetic measurement devices.
Westinghouse Scope of Supply Westinghouse complies with the intent of the requirements of Regulatory Guide 1.31 as discussed in Section 5.2.5.7. j A.1-1.32 Use of IEEE Std 308-1971 " Criteria for Class IE Electrical Systems for Nuclear Power Generating Stations" (Requlatory Guide 1.32)
The design of the Class IE Electrical Systems complies with Regulatory Guide 1.32, issued August 1972.
Methods of compliance are discussed in Section 8.3.1.2.
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() A.1-1.33 Quality Assurance Program Requirements (Operation)
(Regulatory Guide 1.33 A commitment to written procedure for items of the type described in E gulatory Guide 1.33, Revision 1, issued February 1977, is l14 j prov.3ed in Section 13.5. The operational quality assurance l program will be described in Section 17.2 of the FSAR. '
A.1-1.34 Control of Electroslag Weld Properties (Regulatory Guide 1.34)
Electroslag weld properties are controlled as specified in Regulatory Guide 1.34, issued December 1972, with the following modifications:
Position C.3.c (2) : Only one weld of each Class 2 vessel is tested in accordance with Regulatory Positions.C.3.a and C.3.b.
Discussion If the electroslag weld parameters identified in Regulatory Position C.1.a are controlled during production welding to the values qualified, then the production weld properties duplicate those of the qualification weld. Therefore, the qualified weld parameters are maintained during production welding and this Os provides the assurance required for Class 2 vessel welds without the additional testing required in Regulatory Position C.3.c (2).
A.1-1.35 Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures (Regulatory Guide 1.35)
Prestressed concrete is not used in the containment, therefore Regulatory Guide 1.35 does not apply.
, A.1-1.36 Nonmetallic Thermal Insulation for Austenitic Stainless Steel (Regulatory Guide 1.36)
Stone & Webster Scope of Supply Nonmetallic thermal insulation for austenitic stainless steel complies with Regulatory Guide 1.36 issued February 23, 1973, with the exception of packaging and shipping requirements of paragraph C.1 of this guide. In lieu of controlled packaging and shipping, receipt inspection and tests are required, by 4 specification. This consists of visuo inspection for physical or water damage to all cartons. Damaged cartons are segregated.
Potentially contaminated insul& tion is not accepted unless randomly selected samples from each carton are shown to be acceptable after b?ing resubjected to the production test outlined in Regulatory Guide 1.'6.
A.1-9
WUP Amendment 16 PSAR 11/78 8 Westinghouse scope of Supply Westinghouse practice meets the intent of Regulatory Guide 1.36 and is more stringent in several respects. The test for qualification specified by the guide (ASTM C692-71 or RDT M12-1T) allows use of the tested insulation material if no more than one of the metallic test samples crack. Westinghouse rejects the tested insulation material if any of the test samples crack.
The Westinghouse procedure is more specific than the procedures suggested by the guide, in that the Westinghouse specification requires determination of leachable chloride and flouride ions from a sample of the insulating material. In addition ,
Westinghouse experience indicates that only one of the three methods allowed under ASTM D512 and ASTM D1179 for chloride and fluoride analysis is sufficiently accurate for reactor applications. This is the " referee" method, which is used by Westinghouse (see Section 5.2.3.3) .
A.1-1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (Regulatory Guide 1.37)
Quality aesurance requirements for cleaning of fluid systems and g associated components commit to meet the regulatory position of Regulatory Guide 1.37, issued March 1973, with the following exceptions and modifications:
Position C.3: The water quality for final flushes of fluid systems and associated components is at least equivalent to the quality of the operating system water for the following:
- a. Reactor coolant pressure boundary,
- b. System required for reactor shutdown,
- c. Systems required for emergency core cooling,
- d. Reactor vessel internals that are relied upon to permit adequate core cooling for any mode of normal operation or under credible postulated accident conditions.
Other saf ety-related systems are flushed with water in accordance with ANSI N45.2.1-73, except that the quality of water is as close as practicable to that of the operating system water.
Position C.4: Contamination levels in expendable products are controlled and surfaces are subsequently cleaned to remove contaminants to final acceptable levels.
fi A.1-10
WUP Amendment 16 PSAR 11/78 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, l especially for stainless steel. These same systems are the only ones the NRC has singled out in Regulatory Guide 1. 4 4., regarding prevention of stainless steel sensitization. '
For other safety-related systems, it is adeguate 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 system operation. For example, " demineralized water"- of ANSI is used for systems that operate with demineralized / deionized /conden- ,
sated' 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. Contam!' ant levels are controlled such that sunsequent 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 Ouality Assurance Requirements for Packaging, Ship-ping, Receiving, Storage, and Handling of Items for ,
Water-Cooled Nuclear Power Plants (Regulatory i Guide 1.38) ]
l Quality assurance requirements for packaging, shipping, receiving, storage, and handling commit to meet the regulatory position of Regulatory Guide 1.38, Revision 2, issued May 1977, I6 l
with the following exceptions and modifications Position C.1.b: The test weight used for rerating is at least l4 105 percent of the joist lift weight in lieu of the specified 110 percent.
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.
O A.1-11
~ . . _ , - _ . . ._ . . . . _ - . _ . _ _ _ _ , _ _ - ~ _ . _ _ _ , ~ . . _ . _ , . . _ . . _ . _ _ _ _ . . . _ . _ . _ _ . _ . . , _ _
WUP Amendment 16 PSAR 11/78 Discussion 14l Position C.1.b: The established position is applied to the design and construction phases. The crane is proof tested to at least 105 percent of the gross weight of the load to be lifted. For a major lift (reactor vessel) or a series of major lifts (steam generators), where the gross weight of the lif t (s) may approach the rating shown by manufacturer's nameplate, the crane is retested to the above criteria within several weeks of when the heavy lifts are to be made. If the crane is to have a permanent nameplate showing a rating less than that shown by the temporary nameplate in place during the heavy lif t (s) , the permanent nmmeplate is not installed until the crane is
, tested at 125 percent of the permanent rating; the heaviest test lift under the temporary nameplate serves as the required test for the permanent nameplate if the test load equals or exceeds 125 percent of the permanent rating and if that lift is made within 2 or 3 months of when the permanent nameplate is installed.
g Position C.1.c See discussion C.4 of Section A.1-1.37.
and C.2.c:
A.1-1.39 Housekeeping Requirements for Water-Cooled Nuclear Power Plants (Regulatory Guide 1.39)
Quality assurance requirements defined in Chapter 17 for 4 housekeeping, commit to meet the regulatory position of Regulatory Guide 1.39, Revision 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 Preonerational Testing of Redundant Onsite Electric Power Systems to Verif y 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 preoperational testing program and also after major modifications or repairs.
f A.1-12
WUP Amendment 16 PSAR 11/78
() Position C:
- a. Surfaces The extent of coverage is as follows:
within the primary containment liner boundary:
- 1. For large surface area components the documents are submitted to Stone & Webster as required by ANSI N101.4-72. These components include such items as the polar crane, containment liner, structural steel (including miscellaneous steel and hand-rails), concrete, ductwork, uninsulated pipe, uninsulated portions of the neutron shield tank, exterior of uninsulated tanks and vessels, and 5 major ' equipment supports (the contents of which would be used directly or indirectly to facilitate post-accident cooling).
- 2. For manufactured equipment such as pumps, motors, pipe hangers, and pipe supports, the documentation required by ANSI N101.4-72 is maintained in the manuf acturer's files for the complete duration of the contract warranty / guarantee period. A certificate of compliance signed by responsible management personnel is furnished by the manuf acturer.
- b. For the interior lining of tanks, the contents of which y- could be used to facilitate post-accident cooling.
t Discussion Position C: The extent of coverage is a clarification of paragraph 1.2.4 of ANSI N101.4-72.
Westinghouse Scope of Supply The Westinghouse practice and recommendations are in compliance with the intent of requirements of Regulatory Guide 1.54.
A.1-1.55 Concrete Placement in Category I Structures (Regulatory Guide 1.55)
Placement of reinforcing bars and embedded items, design of the concrete structures, mixing of concrete, and placing of concrete meets or exceeds the requirements of Regulatory Guide 1.55, issued June 1973, with the following exception.
Shop detail drawings for the reactor containment mat, shell, and dome reinforcement are checked by the designer. All other reinforcing shop details are checked by engineers at the jobsite.
The ACI and ASTM specifications are supplemented as necessary with mandatory requirements relating to types and strengths of T concrete, minimum concrete densities, proportioning of ingredients, reinforcing steel requirements, joint treatments, testing requirements, and quality control.
A.1-23
WUP Amendment 16 PSAR 11/78 In the areas of the containment wall adjacent to the equipment and personnel hatches, when a large amount of reinforcing steel requires a more plastic concrete mix for placement, the maximum slump is increased to 4 inches.
A.1-1.56 Maintenance of Water Purity in Boilina Water Reactors (kequlatory Guide 1.56)
Regulatory Guide 1.56 is not applicable because a pressurized water reactor is utilized.
A.1-1.57 Desian Limits and Loading Combinations f or !%tal Primary Reactor Containment System Components (Regulatory Guide 1.57)
The plant utilizes a steel lined concrete containment structure, not a metal containment; therefore, Regulatory Guide 1.57, issued June 1973, is not applicable.
A.1-1.58 Qualification of !!uclear Power Plant inspection,
,Ez mination, and Testing Personnel (R egulatory Guide 1.58)
Quality assurance requirements defined in Chapter 17 for qualification of nuclear power plant inspection, examination, and 16 testing personnel, commit to meet the regulatory position of Regulatory Guide 1.58, issued August 1973, with the following exceptions.
The Westinghouse quality assurance program provides a system which meets ANSI N45.2.6 in the areas of non-destructive examination as applicable to component fabrication. For some peripheral tests and inspections for which there is no industry standard for personnel qualification, the Westinghouse program requires a formal system of personnel qualifications, but does not meet some of the more specific requirements of the ANSI standard, for example, documentation. These areas will be evaluated and, to the exte.c that the tests and inspections
, impact on the public safety, the formal system will be l supplemented to comply with the regulatory position.
Stone & Webster personnel perf orming inspections, examinations, and tests on safety-related systems, structures, components and services shall be trained, qualified, and requalitied by a
, combination of management evaluation and/or continuing education.
l l A.1-1.59 Design Basis Floods for Nuclear Power Plants 1 (Regulatory Guide 1.59) l l
l The determination of the design basis flood complies with Regulatory Guide 1.59, Revision 2, issued August 1977.
Methods of compliance are discussed in Section 2.4 of the Site Addendum.
A.1-24 l
l l
i
WUP Amendment 16 PSAR 11/78 5 A.1-1.60 Design Response Spectra for Seismic Design of
\
Nuclear Power Plants (Regulatory Guide 1.60)
The design response spectra are constructed in accordance with Regulatory Guide 1.60, Revision 1, issued December 1973.
Methods of compliance are discussed in Section 3.7.
A.1-1.61 Damping Values for Seismic Design of Nuclear Power Plants (Regulatory Guide 1.61)
The damping values used for the seismic design comply with Regulatory Guide 1.61, issued October 1973.
Methods of compliance are discussed in Section 3.7.
A.1-1.62 Manual Initiation of Protective Actions (Regulatory Guide 1.62)
The protection systems comply with Regulatory Guide 1.62, issued October 1973, as discussed in Section 7.2, except that no manual system level isolation is provided for steamline isolation.
Individual manual isolation is provided for each steamline.
Automatic trip of the residual heat removal pumps on low RWST level and a modified manual procedure for switchover from
( injection to recirculation is provided.
A.1-1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants (Regulatory Guide 1.63)
The design of the electric penetration assembly complies with Regulatory Guide 1.63, Revision 1, issued May 1977.
14 Methods of compliance are discussed in Section 3.8.1 and Section 8.3.1.2.
A.1-1.64 Quality Assurance Requirements for the Design of Nuclear Power Plants (Regulatory Guide 1.64) )
Regulatory Guide 1.64, issued October 1973, accepts ANSI Standard N45.2.11, " Quality Assurance Requirements for the Design of Nuclear Power Plants," (Draft No. 3, Rev. 1, July 1973) as providing an acceptable basis for meeting the pertinent requirements of Appendix B to 10CFR50. The quality assurance .
program, as described in Chapter 17, commits to meet the ig i regulatory position of Regulatory Guide 1.64. I A.1-1.65 Materials and Inspections for Reactor Vessel Closure Studs (Regulatory Guide 1.65)
Regulatory Guide 1.65 is followed except that the use of modified 2
SA540 B24 is specified in the ASME Boiler and Pressure Vessel l A.1-25
WUP Rmendment 16 PSAR 11/78 Code Case 1605 is permitted and a maximum ultimate tensile strength of 170,000 psi is not specified.
Both 45 ft lb and 25 mil lateral expansion is specified for control of fracture toughness determined by Charpy-V testing ,
required by the ASME Boiler and Pressure Vessel Code Section III Summer 1973 Addenda and NRC Regulation 10CFR50, Appendix G (July 17, 1973, Par. IV.A.4). These toughness requirements assure optimization of the stud bolt material tempering operation with the accompanying reduction of the tensile strength level when compared with previous ASME BSPV code requirements.
The specification of both impact and maximum tensile strength as stated in the guide results in unnecessary hardship in procurement of material without any additional improvement in quality.
The closure stud bolting material is procured to a minimum yield strength of 130,000 psi and a minimum tensile strength of 2 145,000 psi. This strength level is compatible with the fracture toughness requirements of 10CFR50 Appendix G, July 1973 (I .C) ,
although higher strength level bolting materials are permitted by the code. Stress corrosion has not been observed in reactor vessel closure stud bolting manufactured from material of this strength level. Accelerated stress corosion test data do exist for material of 170,000 psi minimum yield strength exposed to marine water environments stressed to 75 percent of the yield strength (given in Reference 2 of the guide) . These data are not considered applicable to Westinghouse reactor vessel closure stud bolting because of the specified yield strength differences and a less severe environment; this has been demonstrated by years of satisfactory service experience.
The ASME Boiler and Pressure Vessel Code requirement for toughness for Reactor Vessel bolting has precluded the guide's additional recommendation for tensile strength limitation, since to obtain the required toughness levels, the tensile strength levels are reduced. Prior to 1972, the Code required a 25-f t lb roughness level which provided maximum tensile strength levels ranging from approximately 155 to 178 kpsi. Af ter publication of the Summer 1973 Addenda to the Code and NRC Regulation 10CFR50, Appendix G, wherein the toughness requirements were modified to 45 it lb with 25 mil lateral expansion, all bolt material O
A.1-26
WUP Amendment 16 PSAR 11/78 n
/ ) 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 Qualification 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 with the requirements of IEEE-382-72 and thereby with Regulatory Guide 1.'3 positions with the exception that stem mounted switches are not environmentally tested along with the valve motor operator.
A.1-1.74 Quality Assurance Terms and Definitions (Regulatory Guide 1.74)
Stone S Webster Scope of Supply The Quality Assurance Program described in Chapter 17.1.2 commits g to meet the regulatory position of Regulatory Guide 1.74 issued 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 N February 1974.
gg l
l l' l N A.1-29
WUP Amendment 14 PSAR 5/26/78 14l A.1-1.75 Physical Independence of Electric Systems gl (Regulatory Guide 1.75) 14l Physical indenendence of electric system complies with Regulatory g Guide 1.75 issued January 1975. Methods of compliance are 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 components important to safety will at least meet the values o~f 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 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) ,
14 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 Hazardous Chem'Jal (Regulatory Guide 1.78) ;
1 The assumptions used for identifying chemicals, potentially hazardous to the control room in accordance with the provisions of Regulatory Guide 1.78 issued June 1974. Evaluation 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 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.
O A.1-30
HUP Amendment 14 PSAR 5/26/78
( The plant instrument air systems do have an interface with com-ponents that are part of safety-related systems. These air con-trolled components are individually tested to verify that upon loss of their nonsafety-related air supply, they will respond by 7
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 Mtylti-Unit Nuclear Power Plants (Regulatury 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, Revision 1, issued January 1975. lg 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 Steam Generator Tubes (Regulatory Guide 1.83) 4 Design of the steam generator support and restraint system will permit inservice inspection access to the steam generator manway 0 openings.
The program for steam generator tubing inservice incpection will 4 comply with the recommendations of Regulatory Guide 1.83, Revision 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 l14 months. Details of the program will be provided in the FSAR.
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, Revision 11, issued November 1977.
Because this guide is revised approximately every 4 months, later revisions of the guide need to be evaluated. Where the guide 14 limits the applicability of a particular Code Case by referencing an additional regulatory guide, the Wisconsin Utilities Project
(~T will comply to the extent specified by the referenced guide.
V A.1-33
WUP Amendment 16 PSAR 11/78 A.1-1.85 Code Case Acceptability ASME Section III 4
Materials (Regulatory Guide 1.85)
The Wisconsin Utilities Project will comply with the intent of Regulatory Guide 1.85, Revision 11, issued November 1977.
g Because this guide is revised approximately every 4 months, later 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 Wisconsin 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 Nisconsin 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 _ Reaulatory Guide 1.88 The Quality Assurance Program described in Chapter 17 commits to 5 meet the regulatory position of Regulatory Guide 1.88, Revision 2, issued October 1976.
O A.1-34
WUP Amendment 16 PSAR 11/78 A.2 OTHER DIVISION REGULATORY GUIDES A.2-3.2 Efficiency Testing of Air-Cleaning Systems Containing Devices for Removal of Particles (Regulatory Guide 3.2)
The design of the air cleaning systems permits testing the efficiency of these systems in accordance with Regulatory Guide 3.2, issued January 1973.
A.2-4.1 Measuring and Reporting of Radioactivity in the Environs of Nuclear Power Plants (Regulatory Guide 4.1)
Measuring and reporting of radioactivity in the environs complies 15 with Regulatory Guide 4.1, Revision 1, issued April 1975.
A.2-8.1 Radiation Symbol (Regulatory Guide 8.1)
Regulatory Guide 8.1, issued February 1973, is complied with without exception.
A.2-8.5 Immediate Evacuation Signal (Reaulatory Guide 8.5)
The unmediate evacuation signal complies with Regulatory Guide 8.5, issued February 1973.
A.2-8.8 Information Relevant to Ensuring that Occupational O Radiation Exposures at Nuclear Power Stations Will Be as Low as Is Reasonably Achievable (Regulatory 14 Guide 8.8)
The plant design is in agreement with the overall objectives of 114 Regulatory Guide 8.8, Revision 2, issued March 1977. Design consideraticas implemented by the Applicants to maintain occupational radiation exposur e as low as is reasonably achietable (ALARA) are discussed in Section 12.1. 16 ,
Exception is taken only to certain procedural details implied or ,
rr; commended for implementation in Regulatory Guide 8.8, as follows: M
- 1. The inclusion in the written ALARA program of detailed l16 analysis, formalized statistical measurement, and formalized cost-benefit studies. 14
- 2. The assignment of all ALARA responsibilities to a single individual designated Radiation Protection Manager. 16
- 3. The advance provision of . extensive electronic and artifical devices such as built-in monitoring s y s t e m s l 16 with centralized and computerized alarms and readouts, 14 portable TV cameras, and mock-up training.
A.2-1
I l WUP Amendment 16 PSAR 11/78 j isl Implementation of ALARA at the Haven Nuclear Plant will follow g' the same practices found to be successful at the Point Beach Nuclear Plant. The ALARA program consists of a management 16 l commitment and appropriate procedure assigning responsibilities for periodic review and audit. In order to maximize flexibility gi I and adaptability to particular concerns, reliance is placed in the professional judgment of appropriate individuals consisting 16 of the General Superintendent, the Radiochemical Engineer, and the Health Physicist. Other staff members contribute as required. Principal responsibilities for audit and design review g are assigned to appropriate individuals within the corporate Nuclear Projects Office. Particular attention is given to AIARA concerns at all phases. of Maintenance planning to ensure the participation of all individuals concerned. Further description 16 of ALARA efforts during design is provided in Section 12.1. The AIARA program to be implemented during plant operation is described in Section 12.3.
O O
A.2-2
WUJ Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT
\
LIST OF EFFECTIVE PAGES Appendix B Page, Table (T) or Revision Fioure (F) Number B -i ' 14 B.1-1 through B.1-3 14 B.2-1 through B.2-24 14 B.3-1 14 B.3-2' 15 l B.3-3 through B.3-11 14 B.3-12 through B.3-14 16 B.3-15 through B.3-16n 15 B.3-17 through B.3-18 14 B.3-19 15 B.3-20 through B.3-21 14 B.3-22 15 B.3-23 14 B.3-24 through B.3-24a 16 B.3-25 through 40 14 B.3-41 through B.3-43 15 EP.B-1
WUP Am ndment 14 !
PSAR 5/26/78 l
() postulated seismic event. Following the recorders will regain an operating status.
event, the
- 5. Implementation The NRC staff has failed to show that the installation of equipment to meet Regulatory Position C.3 in particular, and the whole guide in general, will provide
" substantial additional protection which is required for the public health and safety or the common defense and security (10CFR50.109) ." The provisions of Position C.3 will not be implemented.
D.15 Seismic Oualification of Electric Equipment for Nuclear Power Plants (Regulatory Guide 1.100)
Stone & Webster Scope of Supply All Class 1E electric equipment in the Stone & Webster scope of supply will be qualified in accordance with Regulatory Guide 1.100, Revision 1, issued August 1977., Methods of compliance are discussed in Section 3.10.2.
Westinghouse Scope of Supply O Westinghouse supplied Class IE saf ety-related electrical equipment will be seismically qualified in accordance with the methods described in WCAP 8587, Revision 1, " Methodology for Qualifying Westinghouse PWR-SD Supplied NSSS Safety-Related Electrical Equipment" and Supplement 1 to WCAP 8587,
" Equipment Qualification Data Packages." These documents describe the methods employed to satisfy the requirements of IEEE-344-1975 and Regulatory Guide 1.100.
D.16 Flood Protection for Nuclear Power Plants (Regulatory Guide 1.102)
The design of the plant complies with the flood protection guidelines of Regulatory Guide 1.102, Revision 1, issued September 1976.
D.17 Instrument Setpoints (Regulatory Guide 1.105)
The design of the instrumentation system and selection of instrument setpoints is in agreement with the regulatory positions of Regulatory Guide 1.105, Revision 1, issued November 1976, with the following comments and exceptions.
Technical specifications will provide margin from the nominal setpoint to the technical specification limit to for drift, when such information is available from
()
("% account suppliers and/or when measured at the rack during periodic B.3-11
-- . - . ~ . __
WUP Amendment 16 PSAR 11/78 testing. The allowances between the technical specification limit and the safety limit include the following items:
lh
- 1. The inaccuracy of the instrument,
- 2. Process measurement accuracy,
- 3. Calibration equipment accuracy, 4 The potential transient overshoot determined in the accident analyses (this may include compensation for the dynamic ef fect) , and
- 5. Environmental effects on equipment accuracy caused by postulated or limiting postulated events (only for those systems required to mitigate consequences of an accident).
Setpoints are chosen such that the accuracy of the instrument is adequate to meet the assumptions of the safety analysis.
The range of instruments is chosen based on the span necessary for the instrument's function. Narrow range instruments will be used where necessary. Instruments will be selected based on expected environmental and accident conditions. The need for qualification testing will be evaluated and justified on a case basis.
Administrative procedures coupled with the present cabinet alarms and/or locks provide sufficient control over the setpoint adjustment mechanism, such that no integral setpoint securing device is required. Integral setpoint locking devices will not be supplied.
The assumptions used in selecting the setpoint values in Regulatory Position C.1 and the minimum margin with respect to the technical specification limits will be documented.
D.18 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants (Regulatory Guide 1.108)
The diesel generator testing program will comply with Regulatory Guide 1.108, Revision 1, issued August 1977.
16 Compliance is discussed in Section 8.3.1.2.1.
O B.3-12
l l
t WUP Amendment 16 PSAR 11/78 16 l
l D.19 Protection Acainst Low-Traiectory 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 l 15 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 l 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 as interfaces between the electrical system and protective relays, does not degrade without changes in the steady state performance of the device. I l
l 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 that 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 fs response time adequacy. These response times will be
(' combined with system response time, which is measured independently of the sensor.
B.3-13 J
WUP Amendment 16 PSAR 11/78 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 l maintenance / operator error. The overall sensor reliability would be decreased as a result of failures introduced by testing. Repeated testing to dete ct response time !
degradation in sensors of proven historical performance l would only compromise the safety of plant operation. l 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. i i
- 1. The recommendations will be considered to be discretionary ;
in accordance with the Applicants' interpretation of the intent of IEEE 338-1971.
1
- 2. It is believed that testing design features without change proposed in the PSAR, supplemented by testing procedure and l surveillance requirements of the Technical Specifications, including those for response time testing, will be 1
I l
O B.3-14
WUP Amendment 14 PSAR 5/26/78 1
() D.35 Effects of Abnormal Grid Voltage Postion 1:
Additional Level of Under- or Over-Voltage Protection With a Time Delay Two levels of voltage protection for the safety-related buses are provided.
The first level of voltage protection detects extreme low voltage or loss of voltage. When this level of voltage protection is actuated, it disconnects the safety-related buses from the offsite power source, starts the diesel generators, and strips the safety-related buses of all loads except those in Load Block One of the diesel generator loading sequence. The time delay associated with this level of voltage protection is very short; it is only long enough to ignore momentary transients.
The second level of voltage protection detects grid voltage conditions which, if allowed to persist, could affect the ability of the safety related equipment to perform its safety function. This level of voltage protection is subject to the following:
- 1. The voltage and time set points are based on the voltage requirements of the safety-related loads at all distribution levels.
- 2. Coincidence logic is used to prevent spurious trips (e.g. a relay failure or a blown potential fuse) .
- 3. The time delay is selected to prevent spurious trips l
from short duration disturbances, and to allow sufficient time for the onsite power source to pi'k c up the safety-related loads, with margin, within the j maximum time allowed in the accident analysis.
l i 4. The actuation of this second level of voltage l protection will result in the disconnection of the safety-related bus from its of fsite power source.
- 5. The voltage sensors will meet the applicab]e requirements of IEEE 279-1971. -
l 6. The limiting conditions for operation, surveillance requirements, trip set points and allowable values for -
the voltage sensors and time delays will be included in
- the technical specifications.
(:) -
B.3-23 ,
WUP knendment 16 PSAR 11/78 Position 2:
Interaction of Onsite Power Sources With Load Shedding Feature Load shedding on the safety-related buses is initiated by the first level of voltage protection (loss of voltage).
Since this level of voltage protection has a very low set point, it is extremely unlikely that the voltage would dip low enough for a sufficient time to actuate the load shedding feature during diesel generator loading. Should the voltage decrease to a value low enough to cause load shedding while the diesel generator is supplying power to the safety-related bus, one of the following would be likely:
- 1. Diesel generator overload caused by excessive manual loading,
- 2. Diesel generator output circuit breaker trip, or
- 3. Diesel engine or generator problem.
In any case, the result would be stripping of the bus and subsequent reloading of safety-related loads through the load sequencer. '
An analysis will be made of all safety-related bus voltages for expected minimum and maximum grid voltages as well as during diesel generator loading, when the actual equipment parameters become available following vendor selections. The actual values of these equipment parameters are necessary to perform an accurate analysis. The parameters required include motor horsepower, starting and full load currents , transformer g impedance and regulation, diesel generator voltage regulator response time, and others.
As described in Section 8.3.1, the diesel generators will meet l the requirements of Regulatory Guide 1.9 which calls for a minimum voltage of 75 percent during the loading sequence. The voltage relays which are used for load shedding will be set substantially lower than 75 percent of normal; a typical setting would be 50 percent of normal bus voltage or less. j Position 3:
On Power Source Testinq l The onsite power testing program is discussed in Sections 8.3.1.1.2, 8.3.1.2.1, and 16.4.11.
O B.3-24
WUP Amnndment 16 PSAR 11/78 Position 4:
Optimization of Transformer Tap Settinos The transformer tap settings for all transformers associated with the safety-related buses are selected to provide proper voltage at all distribution levels for the expected high and low transmission grid voltage conditions. The analysis showing the optimum transformer tap settings will be available for review.
The tap selections will be verified by test during preoperational testing.
O O
B.3-24a
r WUP PSAR l GENERAL TABLE OF CONTENTS (CONT 'D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION ALL OTHER SYSTEMS REQUIRED FOR SAFETY
( 7.6 AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONIIG, HEATItG, COOLING ,
AND VENTILATION SYSTEMS .
l 9.5 OTHER AUXILIARY SYSTEMS
(-
iii
WUP PSAR GENERAL TABLE OF CONTENTS (CONT'D)
Chapter /
Section Title Volume 10 STEAM AND POWER COINERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS O
iv
1 WUP Amend.Jent 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFEC *.fIVE PAGES NRC OUESTIONS Page, Table (T) or Revision Page, Table (T) or Revision Fiqure (F) Number Fiqure (F) Number QO20.22-1 through QO20.25-1 2 ARQ i through 111 15 QO20.26-1 4 001.1-1 through Q01.4-1 1 F.020.26-1 4 002.1-1 through QO2.24-1 1 F.020.26-2 4 003.1-1 4 Q020.27-1 through QO20.27-3 5 003.2-1 6 QO20.28-1 through Q020.43-1 2 QO3.3-1 through QO3.4-1 1 0020.44-1 through 0020.48-1 6 Q11.1-1 . 1 QO20.49-1 15 Q11.2-1 through Q11.2-2 1 0020.50-1 through QO20.54-2 6 Q11.3-1 through Q11.8-1 1 QO20.55-1 through QO20.56-1 6 Q11.9-1 4 QO20.57-1 16 Q13.1-1 through A13.13-1 1 Q21.1-1 through Q21.2-1 1 Q042.1-1 10 Q 21. 3-1 5 Q042.1-2 12 Q21.4-1 through Q21.5-1 1 F.042.1-1 12 Q21.6-1 13 Q042.2-1 3 j Q21.7-1 through Q21.14-1 1 Q042.3-1 2 Q22.1-1 through Q22.3-1 1 Q042.4-1 through Q042.5-1 3 Q22.4-1 5 Q042.6-1 10 Q22.5-1 through Q22.6-1 1 Q042.7-1 2 j Q31.1-1 through Q31.2-1 1 Q042.8-1 4 O* Q31.3-1 A31.4-1 15 Q042.9-1 through Q042.10-1 Q042.11-1 2
5 Q31.5-1 1 0042.12-1 through Q042.12-2 8 l
l Q31.6-1 0 0042.13-1 10 l Q31.7-1 4 Q042.14-1 8 '
Q31.8-1 13 0042.14-2 10 Q31.9-1 through Q31.10-1 1 Q042.15-1 through Q042.16-1 8 Q31.11-1 6 Q042.17-1 2 j Q31.12-1 1 QO32.18-1 through Q042.19-1 8 Q31.13-1 2 0042.20-1 2 l Q31.14-1 through Q31.15-1 1 Q042.21-1 4 Q31.16-1 3 Q042.22-1 7 Q33.1-1 through Q33.9-1 0 Q042.23-1 2 Q41.1-1 1 Q042.24-1 5 Q42.1-1 through Q42.5-1 1 Q042.25-1 4 AEC i through 14 15 Q042.26-1 2 Q03.2-1 6 0042.27-1 through Q042.38-1 11 0010.1-1 through 0010.5-2 , 2 0042.29-1 through 0042.30-1 2 Q010.6-1 through QO10.6-2 8 0042.31-1 through Q042.31-2 11 Q010.7-1 7 Q042.31-3 through Q042.31-10 11 0010.8-1 through QO10.8-E 6 F.043.31-1 through F.042.31-2 11 Q0'0.9-1 6 0042.32-1 through Q042.39-1 6 0010.10-1 through QO10.11-1 16 Q042.40-1 through 0042.42-1 8 Q042.43-1 through 9042.46-1 6 QO20.1-1 15 Q042.47-1 16 Q020.2-1 through QO20.7-1 2 QO20.8-1 4 Q110.1-1 4 QO20.9-1 3 Q110.2-1 through Q110.3-1 2 0020.10-1 through QO20.10-2 2- Q110.4-1 through Q110.5-1 4 Q020.11-1 4 Q110.6-1 2 0020.12-1 2 Q110.7-1 4 Q020.13-1 6 Q110.8-1 2 0020.14-1 4 Q110.9-1 13 Q020.15-1 through Q020.20-1 2 Q110.10-1 6 Q020.21-1 15 Q110.11-1 7 EP.Q-1
1 l
WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONT'D)
NRC OUESTIONS Page, Table (T) or Revision Page, Table (10 or Revision Piqure fP) Number Fiqure (F) Nunttwr 9110.12-1 through Q110.16-1 6 Q214.22-1 through Q214.25-1 7 Q214.26-1 through Q214.27-1 8 Q120.1-1 2 Q214. 28 -1 9 l
Q120.2-1 through Q120.3-1 4 Q214.28-2 14 Q120.4-1 through Q120.4-4 2 Q214.29-1 through Q214.33-1 7 9120.4-5 through Q120.4-7 2 Q214.34-1 11 A120.5-1 2 Q214.35-1 through Q214.36-1 16 Q120.6-1 through Q120.11-1 6
! 9120.12-1 3 Q221.1-1 7 0120.13-1 through Q120.15-1 2 Q221.2-1 4 9120.16-1 5 Q221.3-1 through Q221.3-2 4 9120.17-1 through Q120.17-2 2 Q221.4-1 12 9121.1-1 through Q121.1-2 6 Q221.4-2 4 Q121.2-1 through Q121.5-2 6 Q221.5-1 7 Q121.5-2 9 Q221.6-1 through Q221.7-1 4 Q121.5-4 through Q121.5-8 6 Q221.0-1 3 Q121.6-1 through Q121.9-1 16 Q221.9-1 14 9122.18-1 7 Q221.10-1 3 Q122.18-3 through Q122.18-4 6 Q221.11-1 13 Q221.12-1 2 Q130.1-1 through Q130.5-1 2 Q221.13-1 through Q221.14-1 4 Q130.6-1 6 Q221.15-1 7 Q130.7-1 4 Q221.16-1 2 Q130.7-2 2 Q221.17-1 4 F.130.7-1 -
Q221.18-1 through Q221.20-1 2
( F.130.7-2 through F.130.7-4 2 Q221.21-1 7 l 9130.8-1 9 Q221.22-1 3 9130.9-1 4 Q221.23-1 2 Q130.10-1 7 Q221.24-1 2 0130.11-1 through Q130.12-1 2 Q221.25-1 2 9130.13-1 through Q130.13-3 4 Q221.26-1 through Q221.26-3 2 9130.14-1 2 Q221.27-1 through Q221.30-1 2 9130.15-1 through Q130.24-1 6 0221.31-1 6 Q221.32-1 through Q221.32-2 7 9211.1-1 through Q211.8-1 2 Q 221.33 -1 through Q221.36-1 6 9211.9-1 4 Q221. 37-1 7 l Q211.10-1 through Q211.11-1 2 Q221.3 8-1 through Q221.39-1 13 l 9211.12-1 through Q211.13-1 6 Q221.40-1 9 l Q214.1-1 through Q214.2-1 2 Q221.41-1 11 9214.3-1 3 Q221.42-1 7 Q214.4-1 5 Q221.43-1 through Q221.44-1 9 9214.5-1 13 Q221.45-1 6 9214.6-1 3 Q221.46-1 13 9214.7-1 2 Q222.1-1 through Q222.2-1 16 9214.8-1 3 9214.9-1 through Q214.9-4 5 Q240.1-1 16 l F.214.9-1 through F.214.9-4 5 Q241.1-1 through Q241.2-1 10 i
9214.10-1 through Q214.12-1 4 Q242.1-1 through Q242.2-1 2 l F.214.12-1 through F.214.12-2 7 9214.13-1 through Q214.14-1 4 Q312.1-1 through Q312.2-1 16 9214.15-1 through Q214.16-1 7 Q331.1-1 7 9214.17-1 through Q214.18-1 3 Q331.2-1 through Q331.12-1 2 F.214.18-1 7 Q 3 31. 'o -1 6 Q214.19-1 7 Q3 31.14 -1 15 9214.20-1 4 Q331.15-1 through Q331.16-1 6 0214.21-1 3 Q331.17-1 through Q331.17-2 -
EP.Q-2
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WUP Amendment 16 PSAR 11/78 PRELIMINARY SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES (CONP'D)
NRC OUESTIONS Page, Table (T) or Revision Fiqure (F) Number
.Q331.18-1 through Q331.21-2 6 Q331.22-1 through Q331.27-1 16 Q411.1 14 Q411.2-1 13
'Q411.3-1 through Q411.5-1 2 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 2411.13-1 .
6 Q411.14-1 through Q411.16-1 16 Q412.1-1 through Q412.1-2 3 Q412.2-1 3 Q412.3-1 through Q412.3-6 6 Q412.4-1 6 Q413.1-1 through Q413.2-1 3 Q413.1-1 through Q413.4-1 1 Q413.5-1 3 Q413.6-1 through Q413.6-3 3 CI Q413.7-1 through Q413.8-1 Q413.9-1 3
2 Q413.10-1 9 Q421.1-1 3 Q421.2-1 10 Q421.3-1 3 Q422.1-1 through Q422.1-6 5 F.442.1-1 through F.422.1-3 5 Q422.2-1 2 Q430.1-1 through Q430.2-2 3 Q431.1-1 through Q432.1-1 3 O
EP.Q-3
WUP Amendment 16 PSAR 11/78 l
I) AEC QUESTIONS AND RESPONSES V
i Haven Preliminary Safety Analysis Report j 1
Question No. Subiect Page 03.2 (6.2) Reactor Projects Q03.2-1 010.1 (11.2) Tank Overflows QO10.1-1 010.2 (11.3) Bed Depth Justification QO10.2-1 010.3 (11.2) Liquid Sampling in Turbine Building Sumps QO10.3-1 010.4 (RSP) Radioactive Waste Management and Blowdown Systems QO10.4-1 010.5 (15.24) Accident Spill QO10.5-1 010.6 (RSP) Liquid and Solid Waste System Design QO10.6-1 010.7 (2.4.13.3) Radionuclides Concentration Analy-sis QO10.7-1 010.8 (11.1.4.3) Charcoal Adsorbers QO10.8-1 010.9 (9.3.3.2) Turbine Building Sump Termination QO10.9-1 010.10 (RSP , A.1) ESF Ventilation and Cooling QO10.10-1 010.11(11.5.6) Solid Waste Storage Capacity QO10.11-1 020.1 (3.4) Equipment Below Plant Grade QO20.1-1
/- 020.2 (3.6) Moderate Energy Systems QO20.2-1
( ,g/ 020.3 (3.6) High or Moderate Energy Line Break Q020.3-1 020.4 (3.6) Potential for Flooding QO20.4-1 020.5 (8.3) Diesel Generator Exhaust System QO20.5-1 020.6 (9.0) Safety-Related Equipment Protection QO20.6-1 020.7 (9.0) Safety-Related Equipment Designed to Withstand SSE and Tornado Missiles QO20.7-1 020.8 (9.0) Conformance to Reg. Guide 1.29 QO20.8-1 020.9 (9.0) Containment of Pressurized Gas Q020.9-1 020.10 (9.0) Failure of Non-Category I Systems QO20.10-1 020.11 (9.1.4) Dropping of the Fuel Assembly QO20.11-1 020.12 (9.1.4 ) Dropping of Spent Fuel Cask Q020.12-1 020.13 ( 9 .1. 4 ) Major Refueling Equipment and Cranes QO20.13-1 020.14 (9.1.4) Geometric Changes of Load Position QO20.14-1 020.15 ( 9 .1. 4 ) Cask Handling Procedure QO20.15-1 020.16 (4 . 2.1) Service Water System 0020.16-1 020.17 (9.2.1) Two Train Service Water System Q020.17-1 020.18 (9.2.2) Component Cooling System 0020.18~1 020.19 (9.2.2) Chilled Water System QO20.19-1 020.20 (9.2.2) Chilled Water Systems QO20.20-1 020.22 (9.2.2) Seismic Classification of Water Storage System Pipes and Valves QO20.22-1 020.23 (9.2.2) CCWS Damage QO20.23~1 020.24 (9.2.2) CCWS Makeup Source QO20.24-1 020.25 (9.2.4) Isolation Valves in Potable Water Supply QO20.25-1 I"\ 020.26 (9.2.5) Essential Lines Between Seismic k_ Category I Structures QO20.26-1 020.27 (9.2.5) Ultimate Heat Sink Analysis QO20.27-1 AEC-i
WUP Amendment 16 PSAR 11/78 Question No. Subiect Page ll 020.28 (9.2.6) Flood Protection of Non-Seismic Structures QO20.28-1 020.29 (9.3.1) Air Operated Valves Q020.29-1 020.30 (9.3.4) Chemical and Volume Control System QO20.30-1 020.31 ( 9 . 4 .1 ) Automatic Detection of Smoke or Radiation QO20.31-1 020.32 (9.4.1) Seismic Classification of Control Building Heating Ventilation and Air Conditioning and Chilled Water Sys-tems QO20.32-1 020.33 (9.4.2) Auxiliary Building Ventilation System QO20.33-1 020.34 (9.4.5) Fuel Handling Building Isolation Dampers QO20.34-1 020.35 (9.4.8) Air Supply Exhaust System for Diesel Generators 0020.35-1 020.36 ( 9. 5.1) Polyvinyl Chloride QO20,36-1 020.37 (9.4.8) Oxygen Content of Combustion Air QO20.37-1 020.38 (9.5.1) bon-Seismic Portions of Fire Pro-tection System QO20.38-1 020.39 (9.5.4) Failure in Diesel Generator Fuel Oil Day Tank 0020.39-1 020.40 (9.5.5) Diesel Generator Cooling Water Sys-tem Q020.40-1 020.41 (9.5.6) Ciesel Generator Starting System QO20.41-1 020.42 ( 10. 3) Main Steam Line Stop Check Valve QO20.42-1 020.43 (10. 4 . 5) Circulating Water System QO20.43-1 020.44 (3.4) Flood Damage to Essential Electri-cal Lines QO20.44-1 020.45 (3.5) Protecting Safety-Related Equipment from Missiles QO20.45-1 020.46 (9 .1. 4 ) Separation of Spent Puel Pump and Cask Pool QO20.46-1 020.47 (9.1.4 ) Cask Drop Analysis QO20.47-1 020.48 (9.1. 4 ) Major Refueling Pools (Seismic Categories) QO20.48-1 020.49 (9.2.2) Piping and Isolation Valves QO20.49-1 020.50 (9.2.5) UHS Spray Systems QO20.50-1 020.51 (9.3.3) Flooding of Plant Drainage System QO20.51-1 020.52 (9.4.8) Diesel Generator Exhaust Duct 0020.52-1 020.5 s (9.5.1) Flood Protection in Fire Protection System Piping QO20.53-1 020.54 (9.5.5) Diesel Generator Cooling Water System Q020.54-1 020.55 (9.5.6) Diesel Engine Startups QO20.55-1 020.56 ( 10.4 . 5) Main Steam Stop Valves QO20.56-1 020.57 (RSP , B .3) Quality RC Pumps QO20.57-1 042.1 (6 .2.1. 3) Minimum Containment Backpressure Q042.1-1 042.2 (6 . 2 .1. 3) Mass and Energy Release Rates Q042.2-1 042.3 (6.2.1.3) Heat Sink Description Q042.3-1 042.4 (6.2.1.3) Mesh Spacing Q042.4-1 042.5 (6 . 2 .1. 3) Heat TranP'er Q042.5-1 SC-ii
. ~ . . . _ . _ _ . ._ . . ._ . __ ___ _ _ . _ - _ .
WUP Amendment 16 PSAR 11/78
(') Question No. Subiect Page 042.6 ( 6 . 2.1. 3 ) Containment Initial Conditions Q042.6-1 042.7 (6 . 2 .1. 3) Tagami Equation Q042.7-1 042.8 (6.2.1.3) Component Cooling Water Temperature Q042.9-1 042.9 ( 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 0042.15-1 042.16 (6.2.1.3) Subcompartments Not Located Near a Rupture Q042.16-1 042.17 (6.2.1.3) Volumes and Vent Areas for sub-compartment Pressure Response Analysis Q042.17-1 042.18 ( 6 . 2.1. 3) Surge Line Rupture Q042.18-1 042.19 (6 . 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 Q042.21-1 042.22 (6.2.2.2) Debris Size in RHR i
0, 042.23 (6.2.2.3) Trapped Reactor Coolant Blowdown Liquid and Spray Water Q042.22-1 Q042.23-1 4
042.24 (6.2.2.3) Sump and Spray Systems Q042.24-1 042.25 (6.2.2.4) 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 4
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 Pressur< Q042.37-1 042.38 (6.2.1) Fan Coc er it Removal Rate Q042.38-1 042.39 (6.2.1) Contain., ray System Q042.39-1 042.40 (6. 2.1) CUPAT Q042.40-1 042.41 (6.2.1) Inertial L..a . on Subcompartments Q042.41-1 0 042.42 (6. 2.1) 042.43(6.2.4.1)
L/A Valves ror Flsw Paths in Sub-compa rtments Hydrogen Recombiner Suction Lines Q042.42-1 Q042.43-1 AEC-iii
WUP Amendmant 16 PSAR 11/78 Question No. Subiect Page 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 Q042.47-1 110.1 (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 Components Q110.3-1 110.4 (3. 9.1. 3 ) Pre-operational Test Program Prototype Plant Q110.4-1 110.5 (3. 9.1. 5, Faulted Condition Loading 3.9.1.6) Combinations Q110.5-1 110.6 (3.10) Electrical and Mechanical Equipment Seismic Qualification 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 (3.6.2.1) 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 Q 110.14 -1 110.15 (3.10) Seismic Listing Methods Q110.15-1 110.16 (10.3.3) Main Steam Stop and Check Valves Q110.16-1 120.1 (4.2.3, Materials with Yield Strength 6.2, 6.3) Greater Than 90,000 psi Q120.1-1 120.2 (5.2.3, Regulatory Position 2 5.4, 5.5, A.1-150) Q120.2-1 120.3 (10.3, Class 2 and 3 Components 10.4, A .1-15 0) Q120.3-1 120.4 (4.2.2) Regulatory Guide 1.31 Q120.4-1 120.5 (4.2.2) Welder Qualification Q120.5-1 120.6 (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 AEC-iv
WUP Amendment 16 PSAR 11/78
() Question No. Subiect Page 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 Q121.7-1 121.8 (RSP,B.3) Regulatory Guide 1.99 P-T Limits Q121.8-1 121.9 (5.2.4.2) 10CFR50 Appendix GSH Q121.9-1 122.18 AUT 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 Frar._ag Q130.2-1 j 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 0 130.8 (3.7.2.14) Damping 130.9 (3.7.3.4) Model Responses Q130.8-1 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) Loading Combinations Q130.14-1 130.15 (3.4) Water Head Q130.15-1 130.16 Scabbing of Concrete Missile Barrier Q130.16-1 130.17 (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(3.7.4.2.3) Triaxial Rear Shock Recorder Q130.20-1 130.21 ( 3 . 8 .1. 3 ) Ambient Temperature Q130.21-1 130.22 ( 3 . 8 .1. 5 ) Tangetial Shear Stress Q130.22-1 130.23 (3.8.3) Local Combination Q130.23-1 130.24 (3.8.5.5) Buoyancy Q130.24-1 211.1 (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 1 0 211.6 (5.1.2) 211.7 ( 5.1. 2) 211.8 (6.2.2)
RHR System Loop Seal Drains Fuel Pool Cleaning and Cleanup Sys-Q211.6-1 Q211.7-1 tem Line Q211.0-1 AEC-v .
WUP Am:ndment 16 PSAR 11/78 Question No. Subiect Page 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 21'.10 (15.4.2.1) Location and Worth of Stuck Control Rod Q214.10-1 214.11 (15.4.2.1)DNBR Versus Time Curves Q214.11-1 21u.12 (15.4.2.2) Steam Flow Rates Q214.12-1 214.13 (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 Q214.25-1 214.26 (5.2.2) Overpressure Protection Calculations Q2 !u.26-1 214.27 (5.5.2) Preheater Box in Steam Generator Q2tt.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 Q7.14.31-1 214.32 ( 15.1. 9) Computer Codes Q214.32-1 214.33 (15.2.4) Boron Dilution Analysis Q214.33-1 214.34 (15.4.2) Steamline Break and Feedline Braak Analysis Q2'4.34-1 214.35 (B . 3) Overpressure Protection Q210.35-1 214.36 (5.5.7) Residual Heat Removal System and AFW Supply to RHR Q21h 36-1 221.1 (3.1.1) Single Failure Criteria 022f.1-1 AEC-vi
l WUP Amendment 16 PSAR 11/78
() Question No. Subiect Page 221.2 (3.11. 2) Safety Related Equipment Qualifi-cation Q221.2-1 22 1.3 (7.1) Standard Format Q221.3-1 221.4 (7.1. 2) Design Criteria for Instruments and Control Systems Q221.4-1 221.5 (7.5) Reactor Coolant Pump Coastdown Q221.5-1 221.6 (7. 9) IEEE-Std-279-1971 Q221.6-1 221.7 (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) Inadvertant Disabling of a Compon-ent Q221,.12-1 221.13 (7.3) Redundant Safety Systems Q221.13-1 221.14 (/.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 0 221.21 (8.2) 221.22 (8.2)
Grid Stability Analysis Interruption of Power to ESF Q221.20-1 Q221.21-1 Q221.22-1 221.23 ( 8 . 3.1.1) Emergency Bus Circuit Breakers Q221.23-1 21.24 ( 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 9221.32-1 221.33 ( 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 (15. 3. 6 ) 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 Q221.39-1 221.40 (7.2) Anticipatory Trips Q221.40-1 221.41 (7.3) Injection Mode to Recirculation
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Mode Switchover Q221.41-1 221.42 (7.3.2.2) ECCS Testing Q221.42-1 AEC-vii
WUP Amnndment 16 PSAR 11/78 Question No. Subiect Page 221.43 (7.5) Post-Accident Display Instru-mentation Q221.43-1 221.44 (3.11) Safety-Related Display Information Q221.44-1 221.45 ( 10. 3) Auxiliary Feedwater System Q221.45-1 221.46 (15.3.4) Underfrequency of Reactor Coolant Pump Breakers Q221.46-1 22 2.1 (RSP ,B .3) Regulatory Guide 1.108 Q222.1-1 222.2 (A.1, Regulatory Guide 1
- 8. 3 .1. 2) Q222.2-1 222.3(B.3) Onsite Power Supply LoLd Shedding Q222.3-1 240.1 (RS P , B . 3) Loose Part Monitoring Q240.1-1 241.1 (4.2) WCAP-8185 0241.1-1 241.2 (4.2) WCAP-8185 Q241.2-1 242.1 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) Q312.2-1 331.1 (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 ) Turbine Building as a Zone III ,
Region Q331.3-1 331.4 (12.1.4) Check Intervals for Plant Area ,
l 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 (T12.2-2) Concentration Calculations Q331.8-1 331.9 Layout Discrepancies (F12.1-1 through 12.1-12) 0331.9-1 331.10 (12.1-2 ) Layout Drawing of Service Building Q331.10-1 331.11 (12.1-2) Control Panels in Waste Disposal Building Q331.11-1 331.12 (12.1-3) Fuel Pool Area Location Q331.12-1 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 h (F12.1-12) Q331.20-1 1 331.21 Service Building Controlled Area AEC-viii
WUP Amendment 16 PSAR 11/78 l
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Question No. Subiect Page (F 12.1 -23) Arrangements Q331.21-1 331.22 (A.2) ALARA Program Q331.22-1 331.23 (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 l and Design and Review Q331.26-1 l 331.27 (A.2) Decommissioning Features Q331.27-1 411.1 (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 Training of QA Superintendent
( 17 .1.1.1 ) Q411.13-1 ri 411.14 (17.1.1) Quality Assurance Q411.14-1
(_)g 411.15(9.5.1.1) Fire Protection Organization Q411.15-1 411.16(9.5.1.3) Fire Protection Quality Assurance Q411.16-1 412.1 (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 Q412.4-1 413.1 (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 Procedur3s Q413.7-1 413.8 (14.1) Augmenting of Plant Staff During Initial Test Program Q413.3-1 413.9 (14.1) Unique Safety-Related Systems Q413.9-1 413.10 ( 14 .1) Regulatory Guide Compliance Q413.10-1 421.1 ( 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 Q421.3-1 422.2 (13.3.6) Hospitals Q422.2-1
- 430.1 (13.2) Number of Liscensed Operators Q430.1-1
(' 430.2 (13.2) Operating Supervisor Q430.2-1 431.1 (13.2) Training Program Evaluation Q431.1-1 432.1 (13. 5) Preliminary Schedule for Preparation of Procedures Q432.1-1 AEL-ix
WUP Amendment 16 PSAR 11/78
() OUESTION 010.10 (A.1) (RSP)
Concerning your response to position C.3.K of Regulatory Guide 1.52 (Revision 1) , you should identify which ESF ventilation systems will have provisions for cooling. It is our position that cooling mechanisms will be required for the Reactor Plant Ventilation System and the Puel Building Ventilation System.
Cooling mechanisms will not be required for the Control Building Ventilation System.
RESPONSE
Response for Reactor Plant Ventilation System is provided in Sections 6.5.1.2 and 6.5.1.3 and Tables 6.5-1 to 6.5-3. For the Fuel Building Ventilation System, the response is located in Sections 6.5.2.2 and 6.5.2.3 and Tables 6.5-1, 6.5-3, and 15.4.5-1.
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O 0010.10-1
WUP Amendment 16 PSAR 11/78 QUESTION 010.11(11.5.6)
Provide the storage capacity for packaged solid waste in terms of the maximum number of 50 ft.3 liners and-55 gal. drums that can be accommodated at one time.
RESPONSE
As discussed in Section 11.5.6, storage capacity is sufficient to hold the wastes generated in about 30 days of normal operation.
As shown on Figure 12.1-12, a shielded area is provided for the 50-ft.a liners which are expected to hold wastes with higher radiation levels. About 23 liners can be accommodated within the shielded area. Drums exhibiting low radiation levels are stored outside of the shielded area. The available storage area for drums is flexible, and the 60 drums generated in 30 days of normal operation can easily be accommodated. While this is the maximum anticipated for routine operations, an additional quantity of about 100 drums could be stored near the decontamination area of the building, if required for any reason.
Temporary shielding would be provided as necessary.
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WUP Amendment 16 PSAR 11/78
() QUESTION 020.57 (B.3) (RSP)
Your response to qualification review item D.43 is not acceptable. It is our position that you modify the component cooling water system design or qualify the reactor coolant pumps as stated in our qualification review letter of April 5, 1978.
RESPONSE
Should a loss of component cooling water to the reactor coolant pumps occur, the' Chemical and Volume Control System continues to provide seal injection water to the reactor coolant pumps. As discussed in the response to qualification review item D.43, the seal injection flow is sufficient to prevent damage to the seals with a loss of thermal barrier cooling. The loss of component cooling water to the motor bearing oil coolers will result in an increase in lube oil temperature and a corresponding rise in bearing metal temperature . Actual testing has shown that the manufacturer's recommended maximum bearing operating temperature will be reached in approximately 10 minutes. Therefore, as stated in the response to qualification review item D.43, the reactor coolant pumps will incur no damage with a component cooling water flow interruption of 10 minutes.
Should a loss of CCW to a Reactor Coolant Pump motor oil cooler occur, a low flow alarm will annunciate on the main control board Os to alert the operator. Note that there are three such alarms (one for each RCP) and should a complete loss of CCW occur, the operator would receive three independent indications. If the loss of CCW is due to a misalignment of the motor operated isolation valve (s) , monitor lights on the main control board ;
panel will provide additional information for the operator to assess the situation. These multiple indications are sufficient to warn the operator of the loss of CCW.
Further, assuming that no operator action is taken following the low flow alarms to the RCP motor oil coolers, a second set of different alarms (independent to each pump) would be initiated as a result of bearing metal temperature increase. At this time, only a single operator action, reactor trip, is required to continue safe plant conditions.
We believe that the instrumentation and alarms provided are sufficient to alert the operator and, as stated in the response to qualification review item D.43, that 10 minutes is an appropriate and conservative response time for this event during normal operation.
In our review, the alternatives given in NRC qualification review item D.43 constitute a significant change in regulatory requirements subject to the requirements set forth in NRC Office O
QO20.57-1
WUP Amendment 16 '
PSAR 11/78 Letter No. 16 dated January 31, 1978. Office letter No. 16 states:
" Effective immediately, all impact analyses are to assure that all significant alternatives and other considerations have been identified and weighed prior to NRC management approval, including RRRC consideration, and staff implementation of significant changes in regulatory requirements."
The instructions attached to that letter cites one of its purposes to be:
"To control (a tendency for escalating regulatory requirements through reinterpretation of rules, guides and review procedures), all significant deviations or departures should be subjected to value - impact analysis just as though they were proposed new guides or branch positions. The fact that they are applied on case reviews is not cause for exemption."
Thus, NRC Office Letter No. 16 establishes minimum requirements that must be met before significant new regulatory requirements can be implemented. Accordingly, until these requirements have been met and appeals by affected parties, if any, have been exhausted, it is inappropriate for NRC to request or for us to agree to a change in design to satisfy a significant new regulatory requirement, particularly when we can see no need for the change.
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WUP Amendment 16 PSAR 11/78 OUESTION 042.47 (B.3) (RSP)
Your response to qualification review item D.36 is not complete.
We require that all of the items concerning the containment leak testing program that were identified in our April 5, 1978 qualification review letter be addressed in the PSAR. In addition, we have the following comments concerning your proposed leak testing program:
(1) Paragraph III.D.2 of Appendix J states that air locks must be leak tested at six month intervals, and if they are opened during these intervals, they must be leak tested after each opening. When multiple openings of an airlock are necessary, it is not practical- to leak test the airlock after each opening. Therefore, we would find acceptable a commitment to leak te st an air lock within three days after being opened.
Furthermore, if dual seals are incorporated in the airlock door design, the application of test pressure between the door seals would fulfill the three-day test requirement. If the door seals cannot be pressurized to P, , a reduced pressure test to demonstrate seal integrity would be acceptable with proper justification of the test pressure and test acceptance criteria. Also, it is our position that the air lock door seal testing may not be substituted for the -
six-month air lock test at P . Therefore, discuss your plans for leak testing the personnel air locks.
(2) It is our position that for those penetrations which should be vented and drained for the Type A test but cannot be vented- and drained, the containment isolation valves should be locally leak tested and the results added to the upper 95 confidence limit of the Type A test result. Discuss your intentions to comply with this position.
(3) The statement of intent to conduct the Type A test in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not acceptable. It is our position that the !
minimum duration for a Type A test shall be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless l prior staff approval has been given to the test methods which ,
justifies reduced duration testing. Discuss your intentions l to comply with this position.
RESPONSE
1 The response to this question will be provided in a future '
amendment. j 1
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HUP PSAR GENERAL TABLE OF CONTENTS (CONT'D)
Chapter /
Section Title Volume 6.3 EMERGENCY CORE COOLING SYSTEM 6.4 HABITABILITY SYSTEMS 6.5 EMERGENCY FILTRATION SYSTEMS 7 INSTRUMENTATION AND CONTROLS III
7.1 INTRODUCTION
7.2 REACTOR TRIP SYSTEM 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 7.4 SYSTFE REQUIRED FOR SAFE SHUTDOWN 7.5 SAFETY-RELATED AND POWER GENERATION DISPLAY INSTRUMENTATION 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY O AND POWER GENERATION 7.7 CONTROL SYSTEMS 8 ELECTRIC POWER IV
8.1 INTRODUCTION
8.2 OFFSITE POWER SYSTEM 8.3 ONSITE POWER SYSTEM 9 AUXILIARY SYSTEMS IV 9.1 FUEL STORAGE AND HANDLING 9.2 WATER SYSTEMS 9.3 PROCESS AUXILIARIES 9.4 AIR CONDITIONIIG, HEATI!G, COOLING, AND VENTILATION SYSTEMS 9.5 OTHER AUXILIARY SYSTEMS iii
WUP PSAR GENERAL TABLE OF COtTTENTS (CONT'D)
Chapter /
Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM V 10.1
SUMMARY
DESCRIPTION 10.2 TURBINE-GENERATOR 10.3 MAIN STEAM SUPPLY SYSTEM 10.4 OTHER FEATURES OF JTEAM AND POWER CONVERSION SYSTEM 11 RADIOACTIVE WASTE MANAGEMENT V 11.1 SOURCE TERMS 11.2 RADIOACTIVE LIQUID WASTE SYSTEM 11.3 GASEOUS WASTE SYSTEM 11.4 PROCESS AND EFFLUEITT RADIOLOGICAL MONITORING SYSTEMS 11.5 SOLID WASTE SYSTEM 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 12 RADIATION PROTECTION VI 12.1 SHIELDING 12.2 VENTILATION 12.3 RADIATION PROTECTION PROGRAM 13 CONDUCT OF OPERATIONS VI
13.0 INTRODUCTION
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANTS 13.2 TRAINING PROGRAM 13.3 EMERGENCY PLANS O ,
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WUP Amendment 16 PSAR 11/78 I OUESTION 121.6 (5.0)
G Provide information on the fracture toughness characteristics of the primary components supports structures and the minimum operating temperature of these supports.
RESPONSE
The design specifications for each primary component support contains requirements for material fracture toughness in accordance with the applicable provisions of subsection NF-2300 of Section III of the ASME Boiler and Pressure Vessel Code. When impact testing is required the test temperature is defined in the design specification.
Test temperature is established based on the minimum temperature of the support expected during normal operation minus the number of degrees change in the nil-ductility transition temperature as a result of material exposure to neutron radiation. When neutron radiation effects are neglible this test temperature is selected at or below the minimum service temperature.
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WUP Amendment 16 PSAR 11/78
() OUESTION 121.7 (5.2)
It is indicated in PSAR Table 5.2-7, " Reactor Coolant Boundary Materials Class 1 Primary Components," that SA 533 Class 2 steel might be used as plate material in the pressurizer, steam generators, and reactor vessel (other than the core region) ,
SA-533 Class 2 is a low alloy ferritic steel having a minimum specified yield strength of 70 KSI. Appendix G of 10 CFR Part 50 states that the adequacy of the fracture toughness of ferritic steels having a minimum specified yield strength greater than 50 KSI shall be demonstrated to the Commission on an individual case basis.
We require that the information required by Note 1 of Code case 1528-3 (later incorporated into the ASME Code as paragraph G-2110 (b) , Appendix G of Section III) be provided in the Safety Analysis Report if any high strength (greater than 50 KSI minimum specified yield strength) ferritic material is to be used in a pressure-retaining component of the reactor coolant pressure boundary. Therefore, provide this information.
RESPONSE
Westinghouse has conducted a test program which demonstrates the Q
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adequacy of SA508 Class 2 (Code Case 1528) and SA533 Grade A Class 2 material. The results of the test program are documented in WCAP-9292, " Dynamic Fracture Toughness of ASME SA508 Class 2 and ASME SA533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals," March 1978.
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a WUP Amendment 16 I PSAR 11/78 {
OUESTION 121.8 (B . 3 ) (RSP)
Concerning your response to qualification review item D.3, it is our position that, unless an acceptable alternative ne thod is provided, the procedures recommended in Regulatory Guide 1.99 be used to evaluate the pressure-temperature limits for Haven.
State your intent with regard to this position.
RESPONSE
The initial set of operating pressure-temperature limit curves will be developed in accordance with the existing version of Regulatory Guide 1.99 at the time the curves are developed and provided for NRC approval. Assuming no future changes, this would be Regulatory Guide 1.99, Revision 1, issued April 1977.
Subsequent revisions to the curves will again require NRC approval. When the Haven Nuclear Plant pressure-temperature limit curves are revised they will be based upon the state-of-the-art at that time, the results of the actual reactor vessel materials surveillance program, and a method that is acceptable to NRC.
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WUP Amendment 16 PSAR 11/78
() OUESTION 121.9 (5.2.4.2)
Compare all tests, data, methods, and proposed programs as presented in the PSAR and referenced sources (e .g . , topical reports) on a point by point basis with the requirements of Appendix G, " Fracture Toughness Requirements", and Appendix H,
" Reactor Vessel Material Surveillance Program Requirement", of 10 CFR Part 50. Identify all possible areas of non-compliance to these Appendices.
RESPONSE
The requirements of Appendices G and H of 10CFR50 are met as discussed throughout Sections 5.2.4 and 5.4.3. There are no areas of non-compliance, and a point by point comparison is not necessary.
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WUP Amendment 16 PSAR 11/78 O
v QUESTION 214.35 (B . 3)
Your response to qualification review item D.33 is not specific enough to permit evaluation. Discuss your compliance with the following requirements:
(1) A system should be designed and installed which will prevent exceeding the applicable Technical Specifications and Appendix G limits for the reactor coolant system while operating at low temperatures. The system should be capable of relieving pressure during all anticipated overpressurization events at a sufficient rate to satisfy the Technical Specification limits, particularly while the Reactor Coolant System is in a water-solid condition.
(2) The system must be able to perform its function assuming any single active component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most limiting single active failure.
The cause for initiation of the event; e.g., operator error, component malfunction, will not be considered as the single active failure. The analysis should assume the most limiting allowable operating conditions and systems configuration at the time of the postulated cause of the overpressure event. All potential overpressurization events s must be considered when establishing the worst case event.
Some events may be prevented by protective interlocks or by locking out power. These events should be reviewed on an individual basis. If the interlock / power lockout is acceptable, it can be excluded from the analyses provided the controls to prevent the event are in the plant Technical Specifications.
(3) The system may be manually enabled, however, the electrical ,
instrumentation and control system must provide alarms to alert the operator to:
1 (a) Properly enable the system at the correct plant l condition during cooldown. The enable alarm would not I be permitted to clear until the system is fully i operational (i . e . , isolation valves open, new setpoint !
selected, etc.) , I (b) Indicate if a pressure transient is occurring.
(4) To assure operational readiness, the overpressure protection system must be tested in the following manner:
(a) A test must be performed to assure operability of the system electronics prior to each shutdown.
O Q214.35-1 1
WUP Amendment 16 PSAR 11/78 (b) A test for valve operability must, as a minimum, be conducted as specified in the ASME Code Section XI. lh (c) Subsequent to system, valve, or electronics maintenance, a test on that portion (s) of the system must be performed prior to declaring the system operational.
(5) The system must meet the design requirements of IEEE-279, Regulatory Guide 1.26, and Section III of the ASME Code.
(6) The overpressure protection system must be designed to function during an Operating Basis Earthquake.
(7) The overpressure protection system must not depend on the availability of of fsite power to perform its f unction.
(8) Overpressure protection systems which take credit for an acta ve component (s) to mitigate the consequences of an overpressurization event must include additional analyses considering inadvertent system initiation / actuation or provide justification to show that existing analyses bound such an event.
RESPONSE
The overpressure protection system for the reactor coolant system will meet the following criteria:
(1) The overpressure protection system will be designed and installed to prevent its exceeding the applicable technical specifications and Appendix G limits for the reactor coolant system while operating at low temperatures.
(2) The overpressure protection system will be designed to perform its function assuming any single active component failure. Analyses supporting this ca pability will be provided or referenced in the FSAR.
(3) The electrical instrumentation and control system will be designed to provide alarms that alert the operator to properly enable the system at the correct plant condition during cooldown and indicate if any increasing pressure transient is occurring when at low pressure. The enabling alarm would not be permitted to clear until the system is fully operational.
(4) The overpressure protection system will be tested to assure operability of the system electronics prior to each shutdown. A test for valve operability will be conducted as specified in the ASME Code Section XI. Subsequent to system, valve, or electronics maintenance, a test on that Q214.35-2
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WUP Amendment 16 PSAR 11/78 portion of the system will be performed prior to declaring it operational.
(5) The . overpressure protection system will meet the design requirements of Section III of the ASME Code. Some of the criteria presented in IEEE 279-1971 do not apply to the interlocks for RCS pressure control associated with this system. The degree of compliance to the criteria of IEEE 279-1971 is presently being discussed between the NRC and Westinghouse Electric Corporation, and the system will meet those agreed-upon criteria. The classification system of ANSI N18.2 is used as an alternate acceptable method of meeting the intent of Regulatory Guide 1.26.
(6) Low temperature operation exists for only a short period of time during plant operation. A coincident earthquake is a low probability event and not a credible design basis. For the Westinghouse standard design, an evaluation will be performed to determine the potential for an occurrence of an overpressure situation as a consequence of an operating basis earthquake. The result of the evaluation will determine 1) if the plant design will be modified to provide the overpressure event, or 2) the pressure relief system will be qualified to function following an operating basis earthquake. In either case, the overpressure protection system will not compromise the design criteria of any safety O- grade system.
(7) The overpressure protection system will not depend on the availability of offsite power to perform its function.
(8) Analysis considering inadvertent system operation or justification to show that existing analyses bound such an event will be provided or referenced in the FSAR.
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WUP Amendment 16 PSAR 11/78 O
g OUESTION 214.36 (5. 5. 7)
The Regulatory Requirements Review Committee has approved a new staff position (BZP RSB 5-1) for the Residual Heat Removal System (RHR). The technical requirements for your plant are described below. Please respond to these requirements in sufficient detail to enable the staff to review your compliance.
(1) Provide safety grade steam generator dump valves, operators, air and power supplies which meet the single failure criterion.
(2) Provide the capability to cooldown to cold shutdown in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> assuming the most limiting single failure and loss of offsite power or show that manual actions inside or outside containment or return to hot standby until the manual actions or maintenance can be performed to correct the failure provides an acceptable alternative.
(3) Provide the capability to depressurize the reactor coolant system with only safety grade systems assuming a single failure and loss of offsite power or show that manual actions inside or outside containment or remaining at hot standby until manual actions or repairs are complete provides an acceptable alternative.
(4) Provide the capability for borating with only safety grade systems assuming a single f ailure and loss of of fsite power or show that manual actions inside or outside containment or remaining at hot standby until manual actions or repairs are completed provides an acceptable alternative.
(5) Provide the system and component design features necessary for the prototype testing of both the mixing of the added borated water and the cooldown under natural circulation conditions with and without a single failure of a steam generator atmospheric dump valve. These analyses will be used to obtain information on cooldown times and the corresponding AFW requirements.
(6) Commit to providing specific procedures for cooling down using natural circulation and submit a summary of these procedures.
(7) Provide a seismic Category I AFW supply for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at Hot Shutdown plus cooldown to the RHR system cut-in based on the longest time (for only onsite or of f site power and assuming the worst single failure), or show that an adequate alternate seismic Category I source will be available.
O Q214.36-1
WUP Amendment 16 j PSAR 11/78 i
RESPONSE
The safe shutdown design basis is hot standby (as it is for all Westinghouse pressurized water reactors). Following any !
condition II, III, or IV event including loss of of f site power, the plant can be placed in and maintained in, for an extended period of time, a safe hot standby condition using safety grade systems only. In this condition, decay heat removal is via the auxiliary feedwater system and the steam generator safety valves.
In the event of a loss of offsite power, the reactor coolant pumps and main condenser are not operable. Therefore, Westinghouse recommends that the plant be maintained in the safe hot standby condition until of fsite power, and thua operability '
of this equipment, is restored. Once offsite power is restored, the operator can either return the plant to power operation or can conduct a normal plant cooldown. However, if the operator feels that it is desirable to cool the plant down, it can be cooled down to cold shutdown conditions without offsite power.
To accomplish the cooldown without offsite power requires that four critical functions be performed. These are circulation of the reactor coolant, removal of residual heat, boration, and depressurization.
For the first stage of the cooldown, circulation of the reactor coolant is accomplished by natural circulation, with the reactor core as the heat source and the steam generators as the heat sink. In the second stage, circulation of the reactor coolant is performed by the residual heat removal pumps.
For the first stage of cooldown, removal of residual heat utilizes the auxiliary feedwater system and the steam generator power operated relief valves. In the second stage, heat removal is via the residual heat removal system, the component cooling water system, and the service water system.
Boration is accomplished using a portion of the chemical and volume control system consisting of the boric acid tanks and boric acid transfer pumps, the centrifugal charging pumps, and piping and valves from the boric acid tanks to the reactor coolant system.
Depressurization is accomplished by using a portion of the chemical and volume control system. This is the centrifugal charging pumps, the auxiliary spray valve and the connecting piping to the reactor coolant system.
Makeup for contraction during the cooldown can be provided either from the boric acid tanks, or from the refueling water storage tank.
0214.36-2
WU? Amendment 16 PSAR 11/78 Some of these systems and equipment (i.e., the control circuits for the steam generator power operated relief valves) are not designed as safety grade systems since 1) they are not required to mitigate the consequences of an accident, 2) the safe shutdown design basis is hot standby and plant cooldown is not a safety requirement, and 3) the ability to maintain the plant in the safe hot standby condition for extended periods of time permits the use of operator action, outside the control room, to overcome any failures which would restrict or limit plant cooldown capability.
As discussed above, the safe shutdown design basis of the plant is hot standby. Extended operation at hot standby which is permitted by the service water-auxiliary feedwater cross connections, while manual actions are taken or repairs are made, is a safe condition. The following responses are based on this capability to remain at hot standby for an extended period.
(1) The steam generator dump valves (see Section 10.3) are safety grade. Valve operators are powered by the NNS instrument air system. These valves are not required to maintain the unit in the saf e hot shutdown condition. The main steam safety valves satisfy this requirement, hence the steam generator dump valves are not required to meet the single failure criterion. In the event cooldown is initiated using only safety grade equipment, operators in O
the individual main steam valve houses, in communication with and under the direction of the control room operator, will manually adjust these valves to achieve the desired cooldown rate.
(2) Since extended operation at hot standby is safe, operator actions (such as opening a valve by hand) or repairs (such as repair of a diesel generator) can be used to overcome any failures in the RHRS which would prevent eventual achievement of cold shutdown conditions.
l (3) Since extended operation at hot standby is safe, operator i actions (such as opening a valve by hand) or repairs (such i as repair of a portion of the air system) can be used to l overcome any failures in the CVCS or the pressurizer power j operated relief valves which would prevent eventual l achievement of cold shutdown conditions, i
I (4) Since extended operation at hot standby is safe, operator actions (such as opening a valve by hand) or repairs (such l as repair of a portion of the air system) can be used to l overcome any single failures in the CVCS which would prevent j eventual achievement of cold shutdown conditions.
(5) The NSSS and BOP designs permit a natural circulation cooldown test to be conducted if required.
N Q214.36-3
WUP Amendment 16 PSAR 11/78 (6) In the event that cooldown from the safe hot shutdown condition to RHR system capability is desired, the tollowing abbreviated procedure utilizing safety grade euuipment and assuming a loss of offsite power is provided.
Subsequent to the loss of offsite power, the plant is maintained in a safe condition at hot shutdown. Operation in the hot shutdown mode is discussed and analyzed in Sections 15.2.8 and 15.2.9 of the PSAR. During this time period, decay heat is removed using the steam generator safety relief valves. Boration is accomplished either by aligning the suction of the charging pumps to the boric acid tanks and pumps (see Section 9. 3.4.3.1) or by using the boron injection tank (Section 6. 3.2) . The normal charging line flow through the regenerative heat exchanger is terminated (reactor coolant pump seal injection is maintained) until a coolant letdown path is available.
Although not required for safety consideration, cooldown and depressurization may be accomplished utilizing only safety grade equipinent. Once boration is substantially completed, the coole>wn-depressurization sequence may be initiated if desired. Boration is confirmed by sampling of the reactor coolant system and by knowledge of the amount of boric acid added to the system from the boric acid tanks or from the boron injection tanks. If the decision to cooldown and depressurizc the reactor coolant system is made, operator action is required to 1) deenergize the pressurizer heaters,
- 2) confirm the normal charging line flow path, and
- 3) partially open the steam generator dump. valves. These valves are manually opened by operators in the individual main steam valve houses, who are in constant communication with and under the direction of the control room operator.
The rate of steam dumping is increased until approximately a 50 F per hour cooldown rate is established. If one steam generator relief valve is unavailable for use in the cooldown sequence, cooldown is accomplished using the remaining two relief valves and increasing the steam release. Flow to the reactor coolant system is adjusted to maintain a pressurizer level with makeup for contraction supplied from either the boric acid tanks or from the refueling water storage tank. The CVCS auxiliary spray line is manually adjusted to provide boration of the pressurizer and depressurization of the reactor coolant system (see Section 5.5.10.3.4). Depressurization of the reactor coolant system may also be accomplished by manual adjustment of the pressurizer power operated refief valve. During the cooldown depressurization sequence, various SI activation blocks are inserted, after the RCS pressure has been decreased below 1,900 psig (low pressurizer pressure block) and when the RCS temperature has been decreased belav 540 F (high steam flow block). When the RCS pressure reaches 1,000 psig, the SI accumulator isolation valves are closed.
Q210.36-4
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l WUP' Amendment 16-i PSAR 11/78
_ (~') Cooldown and depressurization operations are continued until
\j- the reactor, coolant- pressure and temperature have been decreased to approximately.400 psig 'and 350 F. From this point, cooldown is affected using' the RHR system (see Section 5.5.7). Steam bleeding may yet be utilized for-purposes of loop cooling and circulation.
(7) The auxiliary feedwater system is described in Section 10.4.10. Adequate water is available in the - auxiliary feedwater tank to supply water for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the hot shutdown condition followed by a 4 - hour cooldown to 350.F where the RHR system is utilized. The plant may be maintained indefinitely in the hot shutdown condition utilizing the service water system crosstie (Fig. 10.4.10-1) from the ultimate heat sink (see Section 9.2.5) . -
NNS water supplies (demineralized water storage tank, Section 9.2.6, and condensate storage tank via the condensate pumps, Fig. 10 . 4 . 7-1) - are also available in addition to the two seismic Category I sources.
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WUP Amendment 16 PSAR 11/78 QUESTION 222.1 (B . 3 ) (RSP)
The exceptions to positions C.2.a.3 and C.2.a.9 of Regulatory Guide 1.108 in your response to qualification review item D.18 are not- acceptable. We require full conformance to the provisions of Regulatory Guide 1.108 and request that you provide a modified response that meets our requirements.
RESPONSE
Qualification review item D.18 has been revised to conform with the provisions of Regulatory Guide 1.108.
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WUP Amendment 16 PSAR 11/78 QUESTION 222.2 ( A .1, 8.3.1.2)
Your response to qualification review item D.11 is not adequate.
Provide a description which indicates how the penetration ,
circuits * (c3 ass IE and nonclass IE) overload protection meets the recommendations of Regulatory Guide 1.63, Revision 1.
RESPONSE
Section 8.3.1.4.4 has been revised to include a description of how the penetration circuits' overload protection meets the recommendations of Regulatory Guide 1.63, Rev. 1.
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-Q222.2-1
WUP Amendment 16 PSAR 11/78 QUESTION 222.3 (B . 3)
Your response to qualification review item D.35, concerning the interaction of onsite power sources with the load shedding i feature, is not complete. In order to provide an adequate basis for retaining the load shed feature when loads are energized by the diesel generators, we require an analysis be performed to demonstrate that the safety-related bus voltage will not dip sufficiently to cause stripping of the buses and subsequent
- reloading through the sequence". Provide a modified response to meet our requirements.
RESPONSE
Qualification review item D . 35 L oeen revised to include the commitment to perform bus voltagt 'ip analysis during diese,1 ;
generator loading.
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WUP Amendment 16 PSAR 11/78 OUESTION 240.1 (B.3) (RSP)
Section E.8, Appendix B of the Wisconsin Utilities Project PSAR discust:es the proposed loose part detection program. The loose part detection program will meet the requirements of Regulatory Guide 1.133 with a few exceptions.
You have taken exception to the reporting requirements of Sections C.3.a.2.d and C.S .5 of Regulatory Guide 1.133. These exceptions are not acceptable. However, since reporting requirements need not be formalized at the PSAR stage, these items can be resolved at the FSAR stage using the latest available version of the Regulatory Guide.
Section E.8, also indicates that, "the loose part detection system need not be capable of performing its function following an OBE." This position is unacceptable.
We require the loose part detection system to be capable of performing its function following all seismic events that do not require plant shutdown, i.e. , up to and including the Operating Basis Earthquake (OBE). The system should be shown to be adequate for the OBE by analysis, test, or combined analysis and test. Recording equipment need not function wit; tout maintenance following the specified seismic event provided the audio or visual alarm capability remains functional. Please modify your 0
f response to meet this position.
RESPONSE
The Nuclear Regulatory Commission Staff has solicited industry comments on a proposed revision to Regulatory Guide 1.133, which clarifies the staff position on several areas in the September 1977 revision. The Applicants will review the next revision of Regulatory Guide 1.133 and will respond to the staff's concerns in a future PSAR amendment.
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WUP Amendment 16 PSAR 11/78 OUESTION 312.1 V
With respect to qualification of Class IE equipment to the radiation environment, please provide the following information :
a) Briefly explain how the integrated doses for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the duration of the accident was obtained including: a discussion of the source term assumed, the the beta and gamma dose rates vs. time, and whether credit was taken for internal shielding or activity reduction mechanisms such as sprays.
b) Indicate how much of the integrated dose equipment to be qualified to is due to gamma rays vs. beta rays.
Discuss how you plan to qualify class IE equipment as well as containment coatings and cables to beta rays (i.e . , by direct exposure by shielding within an enclosure, etc.) .
c) Describe how you proposed to qualify the above equipment, whether by analysis or by test, and if by test, briefly indicate the test procedure, such as dose rate, type of source and duration of exposure.
RESPONSE
() a) Refer to revised Section 3.11.
(Westinghouse Scope of Supply) b) The qualification plans for the Class IE equipment in the Westinghouse scope of supply are identified in Supplement 1 of WCAP-8587 Rev. 1. The Equipment Qualification Data Packages (EQDP 's) , contained in this supplement identify the s and 7 doses to be employed in the qualification tests.
(Stone and Webster Scope of Supply)
See revised Section 3.11 for type of exposure and qualification plans for Class IE equipment.
(Westinghouse Scope of supply) c) The EQDP's in Supplement 1 to WCAP-8587 Rev. 1 identify the proposed method of qualification (i .e . , test, analysis etc.) and summarize the test procedure to be employed for the Class IE equipment in the Westinghouse scope of supply.
(Stone and Webster Scope of Supply)
A t, See revised Section 3.11 for the method of qualification w
and test procedure for Class IE equipment.
Q312.1-1
WUP Amendment 16 PSAR 11/78 r
( QUESTION 312.2 (15.7.4 and B.3) (RSP) s The response to item D.30 of our qualification review, Fuel Handling Accident (inside of containment) , is not adequate. We require either that the radiation monitors assure prompt containment isolation (prior to release), or that the reactor building purge ventilation system be modified to purge through ESF Grade Filters. Please modify your response to meet this position.
RESPONSE
In item D.30,the staff observed that radiation exposure estimates were well within 10CFR100 guidelines but that confirmation of this evaluation and documentation of the factors involved were still required. Accordingly, the Applicants performed a bounding analysis which confirmed the staff's earlier observation. Applicants agree that realistic mixing in containment along with prompt closure of the purge valves is likely to provide substantial mitigation of the release.
However, since the bounding analysis produced results which are well within currently applicable regulatory limits, there is no justification for further refinement of the analysis. Similarly, there is neither any basis nor justification for further upgrading of filtration or radiation monitoring systems, since j the bounding dose estimates are satisfactory.
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V Q312.2-1
WUP Amendment 16 PSAR 11/78
( QUESTION 331.7 (Table 12.1-12)
Provide a detailed comparison of specific design features of Point Beach and Koshkonong used in making the man-rem estimate given in Table 12.1-22.
RESPONSE
The response to this question has been incorporated into Section 12.1.6.
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WUP Amendment 16 PSAR 11/78 OUESTION 331.22 (A.2)
A written ALARA program is an essential part of implementing the AIARA philosophy and concept. It should address the guidance in Regulatory Guide 8.8 as it relates to Haven. It need not be a separate document from the SAR; parts would logically be included in the station Radiation Protection Manual and in the SAR.
Commit to the development of such a written program, including specifically your bases for concluding that particular dose-reducing changes are not consistent with the ALARA principle.
You should note we haven't specifically requested in RG 8.8 or 1.70 formalized statistical measurement and formalized cost-benefit studies as part of the written ALARA program.
RESPONSE
The response to this question has been incorporated into Section 12.3.1 and Appendix A, item A.2-8.8.
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WUP Amendment 16 PSAR 11/78 OUESTION 331.23 (A.2)
Although the ALARA program involves inputs from a number of plant and corporate personnel, it is essential for consistency and completeness that responsibility for the program be coordinated by one individual (or committee) , preferably one assigned to the plant. Identify by title the individual (or group) to be responsiole for the ALARA program, and describe her or his qualifications and experience.
RESPONSE
Applicants agree that, while the ALARA program involves inputs from a number of plant and corporate personnel, primary responsibility for overall coordination and implementation of the program is properly assigned to a few specific individuals. At Haven Nuclear Plant, these individuals are the Ge neral Superintendent, the Radiochemical Engineer, and the Health Physicist. At the corporate level, the responsible individual is the Radiological Design Engineer. The qualifications and experience of these individuals are addressed in Section 13.1.
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WUP Amendment 16 PSAR 11/78
( QUESTION 331.24 (A.2)
Describe how you will assure that area radiation and airborne radioactive monitoring systems will communicate significant changes in radiation levels immediately to workers in the area affected, and to the control room.
RESPONSE
The area radiation monitoring system, described in Section 12.1.4, has readout and alarm f unctions both locally and in the control room. Similarly, installed airborne radioactivity monitoring systems described in Section 12.2.4 have readout and alarm both locally and in the control room.
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WUP Amendment 16 PSAR 11/78 QUESTION 331.25 (A.2)
Such things as portable TV cameras and mockup training are often effective means to reduce occupational exposures involved in specific operations, and should be considered for any operation potentially involving high individual and man-rem doses, justify an a priori decision not to consider all such available dose-reducing techniques in assuring that resulting doses will be ALARA.
RESPONSE
Applicants have not made an a priori decision not to consider all available dose-reducing techniques. Portable television cameras and other techniques have been used in Applicants' operating plants where the use of these techniques resulted in significant reductior. in radiation exposure. It is intended that these techniques will continue to be considered in appropriate situations.
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WUP Amendment 16 PSAR 11/78
() QUESTION 331.26 (A.2)
With regard to the review of the design process, and changes made as a result, in order to assure that occupational radiation exposures are as low as is reasonably achievable:
- a. Identify by title the individuals who have been and will be responsible for the radiation protection aspects of the design review, and describe how they relate to the individual responsible for the overall plant design.
- b. Provide a breakdown by title of radiation protection personnel who have been participating in such reviews, tabulating the health physics education and experience required of each.
- c. Describe formal arrangements and procedures for assuring that adequate radiation protection reviews are performed throughout the design and construction process and adequate records are kept to document the completion of each such review.
- d. Describe snecific examples of actual dose-reducing changes in design that have resulted from these radiation protection design reviews.
RESPONSE
a.,b.,c. The answers to these portions of the question have been incorporated into Section 12.1.2.
- d. Examples of dose-reducing changes in the conceptual design that have resulted from design review by the Applicants and the Architect-Engineer include most of the items listed in Section 12.1.2. The conceptual arrangement of layout and radiation zones as shown in the figures accompanying Section 12.1 similarly were developed in the course of such review.
It should be noted, however, that these reviews occurred before promulgation of the formal ALARA program. Procedures at the time of the reviews reflected the Applicants' commitment to avoid needless exposure and fulfilled the requirements of 10 CFR20.1 to maintain exposures as low as practicable (ALAP). The principles involved in the approach are similar to those implemented in the formal AIARA program except that documentation requirements were minimal. There have been no ALARA reviews of detailed designs, since this stage of design has not yet commenced.
O Q331.26-1
WUP Amendment 16 PSAR 11/78
() OUESTION 331.27 (A.2)
Discuss the features that you have incorporated into your design to assist the decommissioning crews to maintain occupational radiation exposures as low as is reasonatly achievable during the eventual decommissioning of your plants.
RESPONSE
Design characteristics that could expedite and simplify the eventual decommissioning of the Wisconsin Utilities Project nuclear plant have been a consideration of design since the inception of the project. A number of design considerations have been included to reduce or prevent personnel exposure during periodic plant maintenance activities and would be directly beneficial in limiting occupational radiation exposures during plant decommissioning. These features include provision for the removal of steam generators from the containment, utilization of removable block walls, provisions for hatches over components in individual cubicles to facilitate repair and removal, and allowances for access throughout the auxiliary / safeguards buildings. These features allow for accessibility to the equipment for easier removal during dismantlement and for rapid maintenance which minimizes occupational radiation exposure.
x The plant design also includes provisions for liquid radioactive t waste processing and solid radioactive waste processing. These systems, as discussed in Sections 11.2 and 11.5, respectively, are designed to be remotely operated and to minimize exposure to the operating staff. These operational design features would also minimize exposures to the decommissioning crews, since both the liquid and solid radioactive waste processing systems would be utilized during the plant decontaminating and dismantling phase. Use of these radioactive waste processing systems would also lead to a reduction in the bulk and volume of radioactive wastes generated during plant decommissioning, which will also aid in minimizing radiation exposure and assist the decommissioning crews.
O Q331.27-1
WUP Amendment 16 PSAR 11/78
() OUESTION 411.14 (17.1.1)
The response to item 411.1 (PSAR Amendment 14) deletes the WE commitment to comply with the Grey Book and Green Book. Section 17.1.1 now refers to Appendix A of the PSAR. Thus, there is no WE commitment to ANSI Standards N45.2.5 and N45.2.8. Also, the reference to Appendix A of the PSAR does not provide the clear commitment to the quality assurance guidance required by the staff. Please revise the PSAR to show a WE commitment to " meet the regulatory position of" Regulatory Guides 1.28 (June 1972 or later) , 1.30 (August 1972) , 1.37 (March 1973) , 1.38 (March 1973 or later), 1.39 (March 1973 or later) , 1,58 (August 1973), 1.64 (October 1973 or later), 1.74 (February 1974) , 1.88 (August 1974 or later), and to " comply with the requirements of" ANSI N45.2.5 (Draft 3, Revision 1, January 1974 or commitment as above to Regulatory Guide 1.94), N45.2.8 (Draf t 3, Revision 3, April 1974 or commitment as above to Regulatory Guide 1.116) , N45.2.12 (Draft 3, Revision 4, February 1974), and N45.2.13 (Draft 2, Revision 4, April 1974 or commitment as above to Regulatory Guide 1.123) or provide alternatives for our evaluation. Also, the commitment to ANSI N45.2.13-1974 on page 17.1-12a and the commitment to ANSI N45.2.9-1974 on page 17.1-30 should be changed to agree with the commitments resulting from the previous part of this request (411.14).
RESPONSE
Section 17.1.1 and the applicable sections of Appendix A have been revised in response to this request.
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Q411.14-1
WUP Amendment 16 PSAR 11/78 9UESTION 411.15 ( 9 . 5 .1.1)_
Section 9.5.1.1 does not provide adequate information on your fire protection organization for us to complete our review.
Therefore, please provide the following information: I
- a. Describe the upper level offsite management position that has the overall responsibility for the. formulation, implementation, and assessment of the effectiveness of the fire protection program. j
- b. Describe the offsite position (s) that has direct responsibility for formulating, implementing, and periodically assessing the effectiveness of the fire protection program for. the nuclear plant, including fire drills and fire protection training.
RESPONSE
The response to this request is provided in Section 9.5.1.8.
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Q411.15-1
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WUP Amendmnnt 16 PSAR 11/78 QUESTION 411.16 ( 9. 5.1. 3)
Mr. Vassallo's letter of August 29, 1977 on fire protection provides supplemental guidance on quality assurance. Please modify the PSAR so that it is responsive to the latest supplemental guidance on quality assurance for fire protection or provide an alternative for the staff's evaluation. We note that if the fire protection quality assurance program criteria are met
.as part of the quality assurance program under 10 CFR Part 50 Appendix B (PSAR Chapter 17) , it is not necessary to submit a detailed description for NRC review.
RESPONSE
The fire protection quality assurance program criteria of the Supplemental Guidance on Quality Assurance for Fire Protection shall be met using the applicable parts of the quality assurance program as described in Chapter 17 and Section 9.5.1.3 of the PSAR.
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O Q411.16-1
GENERAL INFORMATION VOLUME Instructions for Making Page Changes O Amendment 16 material for the General Information Section of the PSAR follows. Not all holders of the PSAR have been issued a general information volume. Discard material if this volume has
, not been issued to you.
The instructions below serve as a checklist for entering new pages. Enter the revised pages as listed, discarding superseded material.
Remove Old Page Insert New Page 1 through 14 1 throuah 14 Behind Tab, Exhibit D, insert:
Exhibit D, page 16/ blank Exhibit D, page 16/ blank O
1.
WISCONSIN ELECTRIC POWER COMPANY, O , WISCONSIN POWER AND LIGHT COMPANY and WISCONSIN PUBLIC SERVICE CORPORATION APPLICATION FOR LICENSES
- 1. GENF.RAL INFORMATION Wisconsin Electric Power Company, Wisconsin Power and Light Company and Wisconsin Public Service Corporation (referred to herein collectively as the
" Applicants") hereby apply, pursuant to the Atomic Energy Act of 1954, as .
amended (the "Act"), and the fluclear Regulatory Commission's Rules and Regula-tions (the " Regulations") for a construction permit and all licenses required in connection with the ownership, use and operation of a nuclear power plant (the " plant" or " facility") to be located at the Haven site in Sheboygan County.
The plant will be owned by the Applicants as tenants in common, with Wisconsin Electric Power Company assuming responsibility for design, construction and operation of the plant on behalf of all the Applicants.
The Information Requested by the Attorney General for Antitrust Review required under Section 50.33a and Appendix L to Part 30 of thn Regulations was submitted on February 21, 1974, and was docketed on August 9, 1974. The information required by item 9 of Appendix L was previously submitted on November 30, 1973.
The non-site related Preliminary Safety Analysis Report (the "PSAR") f containing the information required under sections 50.34(a) and 50.34a of the Regulations was submitted on May 29, 1974, and was docketed on August 9, 1974.
i A Site Addendum to the PSAR and the Environmental Report--Construction Permit Stage required under Section 50.30(f) and Part 51 of the Regulations O .
Amendment 16 November 9, 1978
I were tendered on June 25, 1974, for the Koshkonong site in Jefferson County, Wisconsin. The Koshkonong Site Addendum and the Koshkonong Environmental .
I Report were docketed on August 26 and September 5,1974, respectively. Site l Addenda to the PSAR and Environmental Reports--Construction Pennit Stage for the Wood site in Wood County, Wisconsin, and the Haven site in Sheboygan County, Wisconsin, were tendered on August 15, 1975, and were assigned project Nos.
540 and 541, respectively. The Site Addendum and Environmental Report in Docket Hos. STN 50-502 and STN 50-503 were subsequently amended in their entirety, by Amendments 15 and 9 respectively on December 16, 1977, to reflect a change from the Koshkonong site to the Haven site.
II. ORGANIZATION AND RULES A. ORGANIZATION Wisconsin Electric Power Company is a private public utility incorporated in the State of Wisconsin with its main corporate office located at 231 West Michigan Street, Milwaukee, Wisconsin. It is not owned, controlled or dominated by an alien, a foreign corporation or a foreign government and is not acting as agent or representative of any other persons in making this application. The names and post office addresses of its directors and principal officers, all of whom are citizens of the United States, are as follows:*
DIRECTORS Frederick M. Belmore 2727 W. Good Hope Rd. Milwaukee, Wisconsin 53209 Russell W. Britt Sol Burstein Harold F. Falk Lcurel Point Kilmarnock, Virginia 22482 Richard L. Johnson P. O. Box 367 Neenah, Wisconsin 54946 Charles S. McNeer George S. Parker 219 E. Court St. Janesville, Wisconsin 53545 John P. Reeve P. O. Box 359 Appleton, Wisconsin 54911 Jon G. Udell 5210 Barton Road Madison, Wisconsin 53711 O
- Except where otherwise noted, all addresses are P. O. Box 2046, U Milwaukee, Wisconsin 53201.
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0 O PRIHCIPAL OFFICERS i Charles S. McNeer, President and Chief Executive Officer
- Russell W. D, .t. Executive Vice President Sol Burstein, Executive Vice President i '
Thomas J. Cassidy, Senior Vice President Nicholas A. Ricci, Senior Vice President Robert H. Gorske. Vice President'and General Counsel John H. McLean, Vice President-Customer Relations Huberto R. Platz, Vice President-Engineering and Construction Philip G. Sikes, Vice President-System Operations
., Norman C. Storck, Vice President-Division Operations 1 John E. Speaker. Vice President-Communications Richard E. Skogg, Vice President-Operating Services H. L..Warhanek, Vice President and Secretary ;
Jerry G. Remmel, Treasurer !
Richard R. Piltz, Controller l Wisconsin Power and Light Company is a private public utility incorporated l
in the State of Wisconsin, with general' corporate offices at 222 West Washington {
Avenue, Madison, Wisconsin. It is not owned, controlled or dominated by an alien, a foreign corporation or a foreign government and is not acting as agent or representative of any other persons in making this application. The names '
and post office addresses of its directors and principal officers, all of whom ,
b are citizens of the United States, are as follows:** ,
DIRECTORS l Allan W. Adams P. O. Box 278 South Beloit, Illinois 61080 Dr. Bernard S. Adams Ripon College Ripon, Wisconsin 54971 >
George F. Kasten 777 E. Wisconsin Avenue Milwaukee, Wisconsin 53202 '
Henry C. Prange H. C. Prange Company Sheboygan, Wisconsin 53081 Carol T. Toussaint 129 Burning Woodway Madison, Wisconsin 53704 l James R. Underkofler Edward A. Wiegner Eugene 0. Gehl 3 Peter S. Van Nort
- Except where othemise noted, all addresses are P. O. Box 192, Madison, Wisconsin 53701.
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Amendment 16 i November 9, 1978 '
PRINCIPAL OFFICERS James R. Underkofler, President and Chief Executive Officer Edward A. Wiegner, Senior Vice President-Customer, Public and Financial Affairs Peter S. Van Nort, Senior Vice President-Division and System Operations Robert A. Carlsen, Vice President-Division Operations-Southern William L. Keeper, Vice President-Power Production Charles G. Kerndt, Vice President-Engineering and Procurement Edward F. Killeen, Vice President-Employee Relations Burton C. Peters, Vice President-Division Operations-Northern Erroll B. Davis Jr., Vice President-Finance Homer J. Vick, Vice President-Rates, and Consumer Services, and Secretary John C. Acomb, Vice President-Corporate Communications Fredrick A. Remeschatis, Treasurer G. A. Goff, Controller Thomas L. Consigny, Assistant Vice President-Public Affairs Edward M. Gleason, Assistant Treasurer D. L. Mossman, Assistant Secretary D. L. Van Brunt, Assistant Secretary Wisconsin Public Service Corporation is a public utility corporation organized under the laws of Wisconsin with principal executive offices at 700 North Adams Street, Green Bay, Wisconsin. It is not owned, controlled or dominated by an alien, a foreign corporation or a foreign government and is not acting as agent or representative of any other persons in making this application. The names and post office addresses of its directors and principal officers, all of whom are citizens of the United States, are a s follows:
DIRECTORS A. Dean Arganbright 220 Washington Avenue Oshkosh, Wisconsin 54901 Michael S. Ariens 655 W. Ryan Street Brillion, Wisconsin 54110 William V. Arvold 1122 East Crocker Street Wausau, Wisconsin 54401 E. W. James P. O. Box 1200 Green Bay, Wisconsin 54305 John M. Rose P. O. Box 1267 Green Bay, Wisconsin 54305 John S. Stiles Box 3472 Green Bay, Wisconsin 54303 Dr. Neil J. Webb Saint Norbert College DePere, Wisconsin 54115 P. D. Ziemer P. O. Box 700 Green Bay, Wisconsin 54303 James H. Liethen 700 North Adams Street Green Bay, Wisconsin 54305 l
O Amendment 16 November 9,1978
PRINCIPAL OFFICERS P. D. Ziemer, President and Chief Executive Officer E. W. James, Senior Vice President-Power Supply and Engineering J. H. Liethen, Vice President-Finance E. R. Mathews, Vice President-System Planning and Engineering L. M. Stoll, Vice President-Administration H. J. Van Groll, Vice President-Division Operations D. A. Bollom, Treasurer A. E. Pearson, Assistant Vice President R. H. Knuth, Secretary and Assistant Treasurer B. BUSINESS Wisconsin Electric Power Company is engaged principally in the business of the generation, transmission, distribution and sale of electricity to more than 766,000 customers in southeastern Wisconsin, including the metropolitan Milwaukee area, the east central and northern portions of Wisconsin and the Upper Peninsula of Michigan. A subsidiary, Wisconsin Natural Gas Company, furnishes gas service for over 191,000 customers in southeastern Wisconsin and in the Appleton, O
Wisconsin area. The two companies together provide electric and gas service in a total area oT approximately 12,600 square miles with an estimated population of 2,326,000. Wisconsin Electric's business activities are further summarized in its 1976 annual report to stockholders, designated Exhibit A, attached hereto and made a part thereof.
Wisconsin Power and Light Company is engaged in the business of the produc-tion, transmission, distribution and sale of electricity and the purchase, distribution and sale of natural gas in a 16,000 square mile area in central and southwestern Wisconsin with a total estimated population of 750,000. Almost 80 percent of its total revenues come from the sale of electricity to over 274,000 customers and the remainder of its revenues come from the sale of gas to more than 91,000 customers nd its water utility operations in two cities. Its business activities are further summarized in its 1976 annual report to stock-f)g holders designated Exhibit B, attached hereto and made a part hereof.
Amendment 16 November 9,1978
O Wisconsin Public Service Corporation is engaged in the business of generation, transmission, distribution and sale of electric energy and in the purchase, distribution and sale of gas in a service territory of about 10,000 square miles with a population of approximately 700,000 in northeastern Wisconsin and an adjacent part of upper Michigan. About two-thirds of its operating revenues are derived from the sale of electricity to over 253,000 customers with the remainder resulting from the sale of gas to about 139,000 customers. Its business activities are further described in its 1976 annual report to stockholders designated as Exhibit C, attached hereto and made a part hereof.
III. LICENSE APPLICATION A. CLASS OF LICENSES Pursuant to Section 103 of the Act and Part 50, Appendix N, of the Regulations, Applicants hereby apply for a permit to construct and license to own, use and operate a nuclear power plant, consisting of one utilization facility in the form of a pressurized water reactor, to be located at the Haven site in Sheboygan County, Wisconsin.
O Amendment 16 November 9,1978
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B. OTHER LICENSES In accordance with Section 50.31 of the Regulations permitting the combination of several license applications, Applicants also request that the Nuclear Regulatory Commission issue licenses to authorize them to engage in the following activities:
- 1. The transfer or receipt in interstate commerce, the transfer, production, delivery, acquisition, ownership or receipt of possession of or title to special nuclear material in connection with the construction and operation of the facility as authorized by Section 53(a) of the Act.
- 2. The possession, transfer or receipt in interstate commerce, transfer, delivery and receipt of possession or title to, source material, as authorized by Sections 62 and 63 of the Act; and
- 3. The possession, transfer or receipt in interstate commerce, manufacture, production, transfer, acquisition and ownership of by-product material as authorized by Sections 81 and 82 of the Act.
At the appropriate time. Applicants will pursue the foregoing license requests with the appropriate divisions of the Nuclear Regulatory Commission.
C. PERIOD OF LICENSES
' Applicants request that the operating license and all other special licenses for the proposed facility be for a period of forty (40) years from the date of issuance of the construction permit. l D. U3E OF THE FACILITY The proposed facility will include a pressurized water reactor which will produce heat, special nuclear material and by-product material. Applicants plan to use such heat in the generation of electricity and to transfer such special nuclear material, as authorized by law, during the period of operation of the facility.
O Amendment 16 November 9,1978
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G The facility will add approximately 900 megawatts electrical to i Applicants' electric generating capacity for use in meeting the ever increasing demand of the Applicants' systems for electric energy. Energy produced by the facility will be distributed through Applicants' transmission and distribution systems for use by customers of the Applicants.
Applicants' transmission systems are interconnected at various voltage levels with other utilities in Wisconsin, Illinois and Michigan. Applicants, together with Madison Gas and Electric Company and Upper Peninsula Power Company, form the Wisconsin Upper Michigan System. The members of such System are also members of the Mid-America Interpool Network which includes utilities in Illinois and Missouri as well as Wisconsin and the upper peninsula of Michigan.
Applicants' proposal to build this nuclear facility was made after careful consideration of many factors. Important among these were the national energy policy calling for increasing use of nuclear power to meet the expanding energy requirements of the country, the continuing economic advantage of nuclear power over coal for base load generation and reconfirmation by recent studies that nuclear plants have fewer adverse environmental and health impacts than coal plants.
E. SPENT FUEL ELEMENTS The spent fuel elements, after temporary storage and preparation for shipment as required by governmental regulations, will be shipped off-site to a reprocessing i
I plant, storage facility or ultimate repository as determined by governmental policy.
Any low level solid waste, after proper preparation, will be shipped to a l
l disposal area approved by the Nuclear Regulatory Commission.
O Amendment 16 November 9,1978
IV. FINANCIAL QUALIFICATIONS Applicants have the financial qualifications to carry out, in accordance with the Regulations, the activities for which the permit and licenses are i sought. Applicants possess or have reasonable assurance of obtaining the funds necessary to cover the estimated construction costs and related fuel cycle costs for the proposed facility.
Exhibit D, attached hereto and made a part hereof, shows each Applicant's estimated share of the total cost of the proposed facility itemized as to (a) total nuclear plant production costs, (b) transmission, distribution and general plant costs, and (c) nuclear fuel inventory cost for the first core, and is accompanied by a statement describing the bases from which the estimate is derived.
Applicants intend to finance the construction cost of the proposed facility from both internal and external sources. The principal internal sources of funds are treasury funds on hand, undistributed earnings and depreciation accruals. The principal external sources of funds are short term bank loans and the sale of securities. Applicants are considering acquisition of the fuel by lease or other arrangement than purchase.
The financial statements of Applicants contained in their published annual reports for the year 1976, attached hereto as Exhibits A, 8 and C, demonstrate that Applicants possess or have reasonable assurance of obtaining the funds necessary to cover the construction and related fuel costs of the proposed facility.
l Amendment 16 November 9,1978
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In accordance with Section 170 of the Atomic Energy Act, Applicants will obtain the required financial protection in the form of public liability and property insurance for the proposed facility and the fuel and will advise the Nuclear Regulatory Commission accordingly. At the appropriate time, Applicants will also enter into an indemnity agreement with the Nuclear Regulatory Commission.
V. COMPLETION DATES Applicants request issuance of a construction permit in time for commencement of construction of the facility on or before October 1980. The estimated earliest date for completion of construction of the facility is April 1,1987. The estimated latest completion date for the facility is October 1,1987.
O VI. REGULATORY AGENCIES The Federal Energy Regulatory Commission has jurisdiction under the Federal Power Act over Applicants' rates in connection with the transmission of sale at wholesale in interstate commerce of electric energy which will be generated by the proposed facility. The address of the Federal Energy Regulatory Commission is 825 North Capitol, N.E., Washington, D.C. 20426 The Public Service Commission of Wisconsin has jurisdiction under Wisconsin law over Applicants' retail electric rates and services in the State of .
Wisconsin. The address of the Public Service Commission of Wisconsin is Hill Farms State Office Building, Madison, Wisconsin 53072.
Amendment 16 l November 9,1978 I
The Michigan Public Service Commission has jurisdiction under Michigan law over Wisconsin Electric Power Companf s and Wisconsin Public Service Corpora-tion's retail electric rates and services in the State of Michigan. The address of the Michigan Public Service Commission is Law Building, Lansing, Michigan 48913.
VII. TRADE AND NEWS PUBLICATIONS The following is a list of trade and news publications which circulate in the areas of the sites under consideration for the proposed fecilities and which are appropriate to give reasonable notice of this application to those municipalities, private utilities, public bodies and cooperatives which might have a potential interest in the proposed facilities:
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TRADE PUBLICATIONS Electrical World,1221 Avenue of the Americas, New York, N.Y. 10020 Nucleonics Week, 330 W. 42nd St. , New York, N.Y. 10036 Nuclear Industry, 7101 Wisconsin Avenue, Washington, D.C. 20014 Atomic Industrial Forum, 7101 Wisconsin Avenue, Washington, D.C. 20014 Nuclear News, 224 E. Ogden Ave. , Hinsdale, Illinois 60521 National Journal, 1730 M St. N.W., Washington, D.C. 22036 Public Utilities Fortnightly, Suite 502,1828 L St. N.W. ,
Washington, D.C. 20036 Utility Spotlight, 74 Trinity Pl . , New York, N.Y. 10006 Wisconsin Utilities Association, 4369 S. Howell Ave. , Milwaukee, Wisconsin 53207 MAJOR NEWS MEDIA Milwaukee Journal, State Desk, 333 W. State St., Milwaukee, Wisconsin 53201 Milwaukee Sentinel, State Desk, 918 N. 4th St., Milwaukee, Wisconsin 53201 Minneapolis Tribune (Madison Bureau), P.O. Box 467, Madison, Wisconsin 53701 Chicago Tribune, 435 N. Michigan Ave. , Chicago, Illinois 60606 United Press International, 918 N. 4th St. , Milwaukee, Wisconsin 53203* ,
Associated Press, 918 N. Fourth St. , Milwaukee, Wisconsin 53203* !
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Reuters,141 Jackson Blvd., Chicago, Illinois 60604 Wall Street Journal, 200 W. Monroe St., Chicago, Illinois 60606 4
- Wire service to daily and weekly newspapers.
Amendment 16 November 9,1978
D G
REGIONAL NEWSPAPERS Racine Journal Times, 212 Fourth St. , Racine, Wisconsin 53403 Kenosha News, 715 - 58th St. , Kenosha, Wisonsin 53140 Capitol Times,1901 Fish Hatchery Rd., Madison, Wisconsin 53713 Wisconsin State Journal,1901 Fish Hatchery Rd. , Madison, Wisconsin 53713 Appleton Post Crescent, 306 W. Washington St., Appleton, Wisconsin 54911 Herald-Times-Reporter, 902 Franklin St., Manitowoc, Wisconf in 54220 Jefferson County Union, 28 W. Milwaukee Ave. , Fort Atkinson, Wisconsin 53538 Watertown Daily ',ime3, 155 W. Main St., Watertown, Wisconsin 53094 Janesville Gazette, One S. Parker Dr., Janesville, Wisconsin 53545 The Sheboygan Press, 632 Center Ave. , Sheboygan, Wisconsin 53081 The Green Bay News-Chronicle, P.O. Box 2467, Green Bay, Wisconsin 54305 Green Press Press Gazette, 435 E. Walnut, Green Bay, Wisconsin 54305 Stevens Point Daily Journal,1200 Third St., Stevens Point, Wisconsin 54481 Wisconsin Rapids Tribune, 2201st Ave. , Wisconsin Rapids, Wisconsin 54494 Grant County Herald Independent, P.O. Box 310, Lancaster, Wisconsin 53813 VIII. RESTRICTED DATA This application contains no Restricted Data or other defense information.
Applicants agree, pursuant to Section 50.37 of the Regulations, that they will not permit any individual to have access to Restricted Data until the Civil Service Commission shall have made an investigation and report to the Nuclear Regulatory Commission on the character, associations and loyalty of such individual, and the Nuclear Regulatory Commission shall have determined that permitting such person to have any access to Restricted Data will not endanger the common defense and security.
Amendment 16 November 9,1978
,3-O IX. COMMUNICATIONS It is requested that all communications issued by the Nuclear Regulatory Commission in connection with this application be mailed or delivered to:
Sol Burstein, Executive Vice President Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201 and the attorneys for the Applicants, Robert H. Gorske, Vice President and General Counsel Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53202 Gerald Charnoff Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
p Washington, D.C. 20036 V
X.
SUMMARY
Based on the statements heretofore set forth and as substantiated and amplified by the exhibits attached hereto and made a part hereof, Applicants allege that:
- 1. The proposed activities, including the construction, use and operation of the proposed facility, provide reasonable assurance that Applicants will comply with the Regulations (including Part 20 thereof) and that the public health and safety of the public will not be endangered;
- 2. Applicants are financially and technically qualified to engage in the proposed activities in accordance with the Regulations and to receive the construction permit and the operating and other lic.enses required to conduct such activities; tO V
Amendment 16 November 9,1978
1 O 3. The issuance of the construction permit and the operating and other licenses to Applicants will not be inimical to the common defense and security or to the health and ' safety of the public;
- 4. All applicable requirements of Part 51 of the >
Regulations were satisfied-upon submission of updated Applicants' Environmental Report-Construction Permit Stage for the Haven site; and
.5. The construction permit and operating and other licenses to be issued will not create or main-
.tain a situation inconsistent with the antitrust laws as specified in Section 105a of the Act.
Applicants therefore respectfully request a construction permit, and upon completion of construction, an operating license pursuant to Section 103 of the'Act, to own, use and operate one utilization facility at the Haven site.
Applicants also respectfully request licenses to authorize them to engcge in the following activities:
- 1. The transfer or receipt in interstate commerce, the transfer, production, delivery, acquisition, ownership or receipt of possession of or title to special nuclear material in connection with the construction and operation of the facility, as authorized by Section 53a of the Act;
- 2. The possession, transfer or receipt in interstate commerce, transfer, delivery and receipt of possession or title to source material, as authorized by Sections 62 and 63 of the Act; and
- 3. The possession, transfer or receipt in interstate commerce, manufacture, production, transfer, acquisition and ownership of by-product material as authorized by Sections 81 and 82 of the Act.
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(D
\s / EXHIBIT D TOTAL ESTIMATED COST OF PROPOSED FACILITY Category Amount (000)
(a) Total nuclear production plant costs $ 971,000 (b) Transmission, distribution and general plant costs 166,200 (c) Huclear fuel inventory cost for first core 90,800 Total estimated cost $1,228,000 EACH APPLICANT'S SHARE OF TOTAL ESTIMATED COST Applicant Sha re Share
(%) (000)
Wisconsin Electric Power Company 62.5 $ 767,500
() Wisconsin Power and Light Company ,
19.1 234,548 225,952 Wisconsin Public Service Corporation 18.4 Total 100.0 $1,228,000 This estimate of the total cost of the proposed facility, consisting of one pressurized water reactor a t the Haven site, is based upon September 1, 1977 costs, including indirect costs and an allowance for indeterminates, l
l l escalated at 65 per year for inflation in materials and labor, and assumes a commercial operation date of June 1987.
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- ; Arendment 16
's Noverber 9,1978 l
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