ML17275B045

Forwards Amend 14 to FSAR

ML17275B045
Person / Time
Site:	Columbia
Issue date:	05/18/1981
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML17275B046	List: ML17275B045
References
GO2-81-100, NUDOCS 8106010544
Download: ML17275B045 (282)

Text

REGULATORY'ORMATION DISTRIBUTION SY EM (RIDS)

ACCESSION NBR :8106010544,, DOC ~ DATE: 81/Q5/18. NOTARIZED: YES DOCKET 50 397 NPPSS Nuclear FAC IL';.BYNAME Pro,lect'nit 2R washington Public Powe 05000397 AUTH AUTHOR AFFILIATION BOUCHEYiG ~ DE Nashington Public Power SuPPly System REC IP ~ NAHEi RECIPIENT AFFILIATION Of f ice of Nucl ear Reactor Regul ationi Director SC Hl'JENC ER s A ~ Licensing Branch 2 SUBJECT For Wards Ayend o FSAR DISTRIBUT DN'ODE: 8001 1

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DRIES RECEIVED:LTR. 'NCL SIZE:- '

8L TITLE': PSAR/FSAR AHDTS and Related Correspondence NOTES:PM: 2" copies of all material,i cy'.BHR LRG PM(CRIB) 05000397 RECIPIENT COPIES RECIPIENT

'4 COPIES'TTR ID LTTR ENCL ,ID CODE/NAME ENCL' A/D" LICENSNG LIC CODE/NAME'CTION:

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02 1 1 OELD 1 0 OP LIC, BR 1 1 POWER SYS BR 19 '1 1 PROC/TST REV 1 1 QA BR 2'1 1 1 BR22 2g'ESS 1 1-REAC SYS BR 23 1 1 01 1 SIT ANAL BR 24 1 STRUCT ENG BR25 1 EXTFRNAL: ACRS ?7 16 16 LPDR 03 1 1 NSIC Q5 1 1

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Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 May 18, 1981 G02-81-100

Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D. C. 20555 Attention: Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing Gentlemen:

Subject:

SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 FSAR AMENDMENT NO. 14 The Washington Public Power Supply System herewith submits sixty (60) copies of Amendment 14 to its Final Safety Analysis Report.

Pursuant to 10CFR2. 101, we will, within ten (10) days of filing, furnish to you an affidavit reflecting our distribution of this amendment to your designated distribution list.

Very truly yours, Director, Nuclear Safety GDB:CDT:kjf cc: HR Canter Burns 8 Roe J. Ellwanger - Burns & Roe - Woodbury RE Snaith - Burns & Roe JJ Verderber - Burns & Roe RC Root - Burns & Roe, Site FA MacLean - General Electric S. Smith - General Electric ND Lewis - EFSEC, Olympia WS Chin - Bonneville Power Admin.

M. Izaak - WPPSS, New York NS Reynolds - Debevoise & Liberman OK Earle - Burns 8 Roe - Hapo Bldg WNP-2 Files

STATE OF WASHINGTON)

Subject:

) ss COUNTY OF BENTON )

I, G. D. BOUCHEY, being duly sworn, subscribe to and say that I am the Director, Nuclear Safety., for the WASHINGTON PUBLIC POWER SUPPLY SYSTEI1, the applicant herein; that I have full authority to execute this oath; that I have reviewed the foregoing; and that to the best of my knowledge, information and belief the statements made in it are true.

DATED , 1981 G. D. BOUCHEY On this day personally appeared before me G. D. BOUCHEY to me known to be the individual who executed the foregoing instrument and acknowledged that he signed the same as his free act and deed for the uses and purposes therein mentioned.

GIYEN under my hand and seal this // day of , 1981.

Nota Pub ic n and for the State of Wash gton Residing at yf'

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1 1

,.j J 4 Pp* 8 4

SD I

-Q )7 NNP-2 AMENDMENT NO. 13 P5 '+< February 1981 S (r 8I'q)LIST EFFECTIVE PAGES OF it t tt [0 IttO( 0 Pa s not stated below are original issue:

CHAPTER 1 TEXT PAG AMENDMEN Chapter Ind 1 V1 1 1 1-1X 10 1.1-2 10 1.2-20 and 1.2-21 4 1.2-41 and 1.2-42 12 1 ~ 3 27 1 1.4-4 and 1.4-5 12 1.5-3 13 1.5-5 13 1.5-7 through 1.5-11 13 1.6-2 13 1.6-8 5 1.6-9 4 1.6-11 13 1.7-1 through 1.7-24 11 1.8-1 and 1.8-2 10 1.8-3 and 1.8-4 13 1.8-5 10 1.8-6 through 1.8-8 13 FIGURES 1.2-1a and 1.2 7 1 ~ 2 2 8 1.2-3 throu 1 . 2-14 7

1. 2-16 11 CHAPTER 2 TEXT AGES 2- 11 12 2 lv 5 XlV 7 2-xv and 2-xva 12 2-xvi and 2-xvii 12 2 X1X 13 2-XX11 12 2-xxiv and 2-xxiva 13 2-XXV 13 2-XXV1 1

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 2 (Continued)

TEXT PAGES AMENDMENT 2.1-1. and 2.1-1a 12 2.1-2 and 2.1-2a 12 2 ~ 1 7 12 2.1-9, 12 2 2-1

~

I 12 2.2-2'and 2.2-2a 12 2.2-3 through 2.2-12 12 2.2-13 through 2.2-I15 7 2.3-18 and 2.3-19 5 2~3 21 5 2 ~ 3 23 5 2 ~3 33 5 2.3-36 5 2 ~ 3 37 5

2. 3-4'6 2 2.3-47 through 2.3-48a 7

2. 4-1 13 2.4-2 5

2. 4-4 through 2. 4-11 13 2.4-12 5

2. 4-13 through 2. 4-20 13 2.4-24 through 2.4-29 13 2.4-30 and 2.4-30a 13 2.4-31 through 2.4-34 13 2.4-35 4 2.4-36 4 2.4-39 through 2.4-42 13 2.4-44 through 2'.4-53 13 2.5-136 13 2.5-144 5 2.5-157 5 (2.5F)14 5 (2.5F)24 5 (2.5F)28 5 2.5H-2 3 FIGURES 2 ~ 1 3 12 2.1-4 7 2 ~ 2 3 12 2.2-4 and 2.2-5 7 2~3 1 10 2.4-3 5 2.4-9 5 2.4-10 13

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 2 (Continued)

TEXT PAGES AMENDMENT 2.4-11 and 2.4-12 5

2. 4-13 13 2.4-20 through 2.4-23 13 2.4-26 13 2.5H-8b 3

2. 5H-10b 3

2. 5H-12b 3

2. 5H-14b 3 2.5H-16b 3 2.5H-16c 3

2. 5H-16d 3 CHAPTER 3 TEXT PAGES 3 11 12 3 Vill 8 3-lX 5 3-x and 3-xa 12 3-Xli 9 3-xiia 9 3-xv, 3-xva and 3-xvb 9 3-xlx 9 3-xx and 3-xxi 8 3-xxl 1 8 3 xxlla 1 3 xxlv 12 3-xxvi and 3-xxvia 8 3-xxlx 8 3-xxx and 3-xxxa 9 3-xxxvlll 4 3-xLi 9 3-xLviii 9 3-L and 3-La 9 3-Lviii 12 3-Lxi and 3-Lxii 12 3-Lxiii 5 3-Lxiiia 4 3-Lxiv and 3-Lxiva 9 3-Lxv and 3-Lxva 9 3-Lxvi 2 3-Lxxiii through 3-Lxxiiic 9 3-Lxxv 8 3-Lxxix 8 3-Lxxixa 13

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 3 (Continued)

TEXT PAGES AMENDMENT 3 ~ 1 2 1 12 3.1-25 12 3.1-48 12 3.1-49 12 3.1-69 13 3.1-74 12

3. 1-82 12 3 ~ 2 1 3

• 13

3. 2-18 7

3. 2-19 3 3 ~ 2 23 13 3.2-25 5 3.2-30 7 3 ~ 2 31 5 3 ~ 2 32 13 3 ~ 2 33 7 3.3-4 through 3.3-4b 8

3. 3-5 2 3 ~3 7 5 3.4-1 13 3.4-2 through 3.4-2b 5 3.4-4 1 3.4-5 and 3.4 -5a 5 3.4-7 5 3.5-5 1

3. 5-10 8

3. 5-11 8

3. 5-1 a1 12

3. 5-12 9 3.5-13 9 3.5-14 12

3. 5-16 9 3.5-19 5 3.5-22 7 3.5-23 3.5-24 8 3.5-26 9 3.5-28 and 3.5 -29 13 3.5-30 9 3.5-32 5 3.5-33 5 3.5-34 5 3.6-6 through 3.6-6m 9 3.6-8 2 3.6-9 1' 3.6-15 3.6-22 through 3.6-22g 9 3.6-25 9

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 3 (Continued)

TEXT PAGES AMENDMENT 3.6-27 9 3.6-28 and 3.6-28a 9 3.6-29 9 3.6-30 and 3.6-30a 9 3.6-31 *9 3.6-33 and 3.6-33a 9 3.6-42 9 3.6-48 9 3.6-57 and 3.7-57a 9 3.6-60 through 3.6-63 9 3.6-64 and 3.6-64a 9 3.6-65 through 3.6-67 9 3.6-69 through 3.6-71 9 3.6-72 and 3.6-72a 9 3.6-73 9 3.6-74 1 3.6-75 12 3.6-75a 9 3.6-76 9 3.6-88 through 3.6-100 9 3.7-4 12 3.7-5 8 3 ~7 1 2 8 3 ~7 1 3 and 3.7-13a 8 3.7-14 8 3.7-15 8 3.7-15a 9 3.7-16 8 3 ~ 7 1 7 8 3.7-18 7 3.7-20 1 3 ~7 23 1 3.7-26 7 3 ~7 27 8 3.7-28 8 3.7-29 8 3.7-30 8 3 ~7 31 and 3.7-31a 8 3 ~7 32 8 3 ~ 7 33 8 3.7-34 8 3.7-36 and 3.7-36a 8 3.7-38 through 3.7-38e 8 3.7-382 1 3.7-39 8 3.7-44 8

NNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 3 (Continued),

TEXT PAGES AMENDMENT 3.7-45 8 3.8-3 12 3.8-,3a and 3.8-3b 1 3.8-6 through 3.8-6c .8 3.8-9 8 3.8-,13 3

3. 8-13a 3

3. 8-14 3 3.8-17 8

3. 8-21 9=

3.8-42. and 3.8-42a 8 3.8-43 3 3.8-,44 through 3.8-44b 3' 3.8-45 3.8-46 and 3.8-46a -8 3.8-48 2 3.8-54 13 3.8-60 and 3.8-61 1 3.8-62 13 3.8-65 12 3.8-65a and 3.8-65b 1 3.8-66 13 3.8-71 13 3.8-72 13 3.8-74 2 3.8-76 12 3.8-85 13' 3.8,-86 3.8-87 8 3.8-88 13 3.8-88a 8 3.8-89 and'3.8-89a 8 3.8-90 an'd 3.8-90a 8 3.8-91 and 3.8-91a 8 3.8-92 5 3.8-93 2 3.8-94 through 3.8-96 5 3.8-99 12 3.8-100 1 3.8-103 and 3.8-103a 9

3. 8-104 -9 3.8-111 5 3.8-112 through 3.8-115 4 3.8-119a 7 3..8-119b 5 3.8-126 9

WNP-2 AMENDMENT NO. 1 3 February 1981 CHAPTER 3 (Continued)

TEXT PAGES AMENDMENT 3.8-127 9 3.8-129 and 3.8-129a 9 3.8-130 and 3.8-130a 9

3. 8-135 9 3.8-137 and 3.8-137a 12 3.8-137b 1

3. 8-141 10 3.8-144 and 3.8-144a 8 3.8-146 and 3.8-146a 12

3. 8-147 9

3. 8-161 12

3. 8-175 12

3. 8-176 13

3. 8-190 2

3. 8-191 12 3.8-194 13 3.8-199 8 3.8-201 9 3.8-202 9 3.8-203 9 3.8-204 9 3.8-205 8 3.9-4 9 3.9-17 9 3.9-22 9

3. 9-23 9 3.9-26 through 3.9-29 9

3. 9-36 9 3.9-38 9 3 . 9-39 4

3. 9-48 9 3.9-49 9 3.9-51 through 3.9-53 9 3.9-54 and 3.9-54a 9 3.9-59 8 3.9-66 3.9-67 3.9-69, 3.9-69a and 3.9-69b 3.9-70 3.9-71 3.9-92 3.9-155 3.9-160 through 3.9-162 3.9-163 and 3.9-163a 3 . 9-174 3.10-2

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 3 (Continued)

TEXT PAGES AMENDMENT 3.10-3 13 3.11-4 11 3 . 12-21 through 3.12-21b 9 3 ~ 1 2 31 1

3. 12-34 and 3.12-35 1 3.12-36 9 3.12-43 through 3.12-45 1 3.12-47 1 FIGURES 3.2-2 through 3.2-4 11 3.2-5 7 3.2-6 through 3.2-8 11 3 ~ 2 1 1 11

3. 2-16 11 3 ~ 2 1 7 7 3.2-20 7 3.2-25 11 3.5-9 3 3.5-41 4 3.5-50 7 3.5-51 and 3.5-52 4 3.6-6c and 3.6-6h 2 3.6-9a 2 3.6-11 9 3.6-12b and 3.6-12c 9' 3.6-13b and 3.6-13c

3. 6-14a through 3.6-14c- 9

3. 6-15b and 3.6-15c 9 3.6-16b through 3.6-16e 9 3.6-17b through 3.6-17e 9 3.6-18a through 3.6-18g 9

3. 6-19a through 3. 6-19c 9 3.6-20b 9 3.6-21b 9 3.6-22b 9 3.6-23a and 3.6-23b 9 3.6-24b 9 3.6-25b 9 3.6-26b 9 3.6-27b 9 3.6-28b 9 3.6-29b and 3.6-29c 9 3.6-30b through 3.6-30d 9 3.6-31a through 3.6-31c 9

WNP- 2 AMENDMENT NO. 13 February 1981 CHAPTER 3 (Continued)

FIGURES AMENDMENT 3.6-32b 9 3.6-33a through 3.6-33e 9 3.6-34b and 3.6-34c 9 3.6-35 5 3.6-36a through 3.6-36c 2 3.6-38 and 3.6-39 2 3.6-40a and 3.6-40b 2 3.6-116 through 3.6-146 9 3.7-27 and 3.7-28 8 3.8-2 2 3.8-6 through 3.8-8 2 3.8-10 2 3.8-13 through 3.8-15 1

3. 8-18 2 3.8.33 2 3.8-42 5 3.8-48 8 3.8-50 through 3.8-53 1 3.8-54 and 3.8-55 2 3.8-56 7 3.8-57 and 3.8-58 3 3.8-59 5 3.8-60 5 3.8-61 5 3.8-62a and 3.8-62b 13 3.8-63 13

3. 10C-1 2 CHAPTER 4 TEST PAGES 4-iv 12 4-iva 1 4-v 1 4-va 12 4-xii and 4-xiii 1 4.1-1 12 4.1-4 13

4. 1-9 1

4. 2-1 3

4. 2-2 1 4.4-1 through 4.4-5 1 4.4-6 12 4.4-7 through 4.4-13 1

4. 4-14 12

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 4 (Continued)

TEXT PAGES AMENDMENT 4.4-15 1

4. 4-16 and 4.4-17 10

4. 4-18 7

4. 4-19 12 4.4-20 and 4.4-21 1 4.4-22 7 4.4-23 through 4;4-25 1 4.4-26 12 4.4-27 through 4.4-31 1 4.4-32 13 4.4-33 7

4. 4-34 through 4.4-45 1 4.5-2 2

4. 6-2 12

4. 6-3 13

4. 6-21 12

4. 6-25 and 4.6-26 12 FIGURES 4.4-1 through 4.4-4 4.4-5 4.4-6 CHAPTER 5 TEXT PAGES 5-v 7 5-vi 12 5-vii and 5-viii 5 5-x 12 5-xiii 12 5-xv and 5-xva 12 5-xvii 12 5-xviii 12 5-xix 12 5.2-7 11 5.2-8 11 5.2-11 8

5. 2-12 4

5. 2-13 13 5.2-29 and 5.2-29a 7 5.2-30 through 5.2-35 7 5.2-36 8 5.2-37 13

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 5 (Continued) 4

\

TEXT PAGES AMENDMENT 5.2-39 13 5.2-40 8 5.2-41 and 5.2-42 13 5.2-45 and 5.2-46 ,13 5.2-47 7 5.2-48 10 5.2-52 and 5.2-52a 5 5.2-60 through 5.2-76 7 5.3-3 t hrough 5.3-13 5 5.3-17 and 5.3-18 13

5. 3-23 through 5.3-25 5 5.4-2 13

'.4-3 8

5.4-5 4 5.4-12 12 5.4-14 13 5.4-22 and 5.4-22a 8 5.4-24 8 5.4-25'.4-27 8 8

5.4-28 8 5.4-38 8 5.4-39 9 5.4-40 13 5.4-41 11 5.4-43 8 5.4-47 13 5.4-48 and 5.4-48a 13 5.4-49 and 5.4-49a 13 5.4-50 and 5.4-50a 12 5.4-54 12 5.4-57 through 5.4-59 4 5.4-60 9 FIGURES 5.1-1 3

5. 1-2 3

5. 2-4 11

5. 2-5 11

5. 3-4 5 5.3-5 5 5.4-3a 4 5.4-4a and 5.4-4b 8 5.4-5 4 5.4-14b 3

RNP-2 AMENDMENT NO. 13 February 1981 5 (Continued)

FIGURES

'HAPTER AMENDMENT

5. 4-14c '3 5.4-15a, b, c ',5 5.4-16 2 5.4-17c 13 5.4-19a and 5.4-19b '8-CHAPTER 6 TEXT PAGES 6-iii 6-v and 6-va 3 11 6-vi 5 6-via 10 6-vii and 6-viia 12 6-xi and 6-xii 9 6-xvi 12 6-xvii and 6-xviia 11 6-xviii 9 6-xix 11 6-xx .12 6-xxa 12 6-xxb 11 6-xxvii 12 6.1-5 and 6.1-5a 2.

6.1-10 and 6.1-11 2 6.2-2 12 6.2-3 13 6.2-6 3

6. 2-18 12 6 . 2-27 3 6.2-30 and 6.2-30a 3 6.2-31 and 6.2-32 12 6.2-33 5 6.2-33a 13 6.2-33b 8 6.2-33c 13 6.2-35 12 6.2-36 1 6.2-38 12 6.2-45 3 6.2-47 4 6.2-48 11 6.2-49 4 6.2-50 through 6.2-50e 11

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 6 (Continued)

TEXT PAGES AMENDMhNT

6. 2-51 and 6.2-51a 11 6.2-52 13 6.2-54 13 6.2-55 5 6.2-56 10 6.2-58 12 6.2-59 5

6. 2-61 and 6.2-61a 10 6.2-62 through 6.2-67 5 6.2-68 12 6.2-69 5 6.2-71 6.2-72 and 6.2-72a 11 6.2-73 13 6.2-73a 11 6.2-74 and 6.2-74a 11 6.2-75 8 6.2-76 5 6.2-77 and 6.2-78 2 6.2-79 5 6.2-80 13 6.2-85 4 6.2-88 12 6.2-88a 6.2-90 12 6.2-92 8 6.2-94 8

6. 2-107 11

6. 2-108 12

6. 2-109 4

6. 2-119 and 6.2-120 12

6. 2-121 5

6. 2-122 through 6.2-124 12

6. 2-125 and 6.2-125a 12

6. 2-126 12

6. 2-127 3

6. 2-128 5

6. 2-129 12

6. 2-130 through 6.2-132 5

6. 2-133 and 6.2-134 1'2

6. 2-135 and 6.2-135a 12

6. 2-136 and 6.2-137 3 6.2-138 5 6.2-139 12

6. 2-140 through 6.2-142 3

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 6 (Continued)

.TEXT PAGES AMENDMENT 6.2-143 through 6.2-146 12 6.2-147 through 6.2-152 3 6.2-153 through 6.2-160 8 6.2-161 through 6.2-170 11

6. 3-10 13 6.3-11 4 6.3-13 13 6.3-16 4

6. 3-19 5

6. 3-21 11 6.3-21a 5 6.3-22 13 6.3-27 and 6.3-27a 8 6.3-28 11 6.3-34 through 6.3-36 13 6.3-37 11 6.3-38 9 6.3-39 11 6.3-43 11 6.3-45 11 6.3-49 11 6.4-1 9 6.4-2 9 6.4-3, and 6.4-4 5 6.4-6 9 6.4-6a 11 6.4-7 9 6.4-7a 13 6.4-7b 9 6.4-8 9 6.4-9 through 6.4-9b 9 6.4-9c and 6.4-9d 11 6.4-9e and 6.4-9f 9 6.4-10 13 6.4-10a 9

6. 4-12 9 6.5-1 and 6.5-1a 11 6.5-3 11 6.5-6 through 6.5-8 5 6.5-9 and 6.5-9a 13 6.5-12 3 6.5-13 13 6.5-18 and 6.5-19 5 6.5-20 2 6.5-21 and 6.5-22 5 6.5-25 5

AMENDMENT NO. 13 February 1981 CHAPTER 6 (Continued)

TEXT PAGES AMENDMENT 6.6-1 through 6.6-15 7 6.7-2 5 6.7-4 8 6.7-5. '8 6.7-11 5 FIGURES 6.2-17a through 6.2-17c 3 6.2-21 5 6.2-23 5 6.2-26 5 6.2-30 5 6.2-31a through 6.2-31j 3 6.2-31k 9 6.2-311 and 6.2-31m 3 6.2-31n 13 6.2-31o through 6.2-31t 3 6.2-3lu 13 6.2-32 through 6.2-35 3 6.2-36 through 6.2-41 8 6.4-1 9 6.4-2 7 CHAPTER 7 TEXT PAGES 7-i through 7-xv 10 7.1-1 through 7.1-3 10 7.1-4 and 7.1-5 12 7.1-6 through 7.1-12 10 7  % 1 1 3 7.1-14 through 7.1-17 12 10 7.2-1 through 7.2-14 10

7. 2-15 12 7.2-16 10 7 ~ 2 1 7 7.2-18 through 7.2-21 12 10 7.2-22 and 7.2-23 12 7.2-24 through 7.2-32 10 7.3-1 through 7.3-12 10'2 7 ~ 3 1 3 7.3-14 through 7.3-18 10

WNP-2 AMENDMENT NO. '13 February 1981 CHAPTER 7 (Continued)

TEXT PAGES AMENDMENT 7.3-19 and 7.3-20 '12 7.3-21 through 7.3-24 7.3-25 7.3-26 and 7.3-25a through 7.3-28

'2 10 10 7 .3-29 12 7.3-30 13 7.3-31 through 7.3-79 10 7 .4-1 10 7.4-2 12 7.4-3 through 7.4-5 .10 7.4-6 and 7.4-6a 13 7 .4-7 10 7.4-8 12 7.4-9 and 7.4-,9a 12 7.4-10 10 7.4-11 and 7.4-11a 12 7.4-12 10 7.4-13 12 7.4-14 and 7.4-14a 12 7.4-15 and 7.4-16 12 7.4-17 and 7.4-18 10 7.4-19 and 7.4-20 12 7.4-21 through 7.4-23 10 7.5-1 through 7.5-10 10 7.6-1 through 7.6-10 10

7. 6-11 12 7.6-12 through 7.6-17 10 7.6-18 and 7.6-19 12

7. 6-20 10 7.6-21 12 7.6-22 and 7.6-23 10 7 . 6-24 12 7.6-25 13 7.6-26 through 7.6-48 10 7.7-1 through 7.7-16 10 7 ~ 7 1 7 12 7.7-18 through 7.7-28 10 7.7-29 12 7.7-30 through 7.7-52 10 FIGURES 7.2-1a through 7.2-1d 10 7.2-2 through -7.2-,7 10 7.2-8a through 7.2-8c 10 7.2-9 10

WNP-2 AMENDMENT NO. 13 February 198]

CHAPTER 7 (Continued)

FIGURES AMENDMENT 7.2-10a and 7.2-10b 10 7.3-1 through 7.3-3 10 7.3-4a through 7.3-4c 10 7.3-5 through 7.3-7 10 7.3-8a through 7.3-8c 10 7.3-9a through 7.3-9c 10 7.3-10a through 7.3-10f 10 7 ~ 3 1 1 10 7.3-12a and 7.3-12b 10 7.3-13a and 7.3-13b 10 7.3-14a through 7.3-14e 10 7.3-15a through 7-3-15g 10 7.3-16a through 7.3-16c 10 7.3-17a through 7.3-17s 10 7.3-18a through 7.3-18k 10 7.3-19a through 7.3-19j 10 7.3-20a through 7.3-20j 10 7.3-21a through 7.3-21d 10 7.6-1a and 7.6-1b 10 7.6-2 and 7.6-3 10 7.6-4a and 7.6-4b 10 7.6-5 10 7.6-6a through 7.6-6g 10 7.6-7 through 7.6-11 10 7 7~ 1 10 7.7-2a and 7.7-2b 10 7.7-3a through 7.7-3h 10 7.7-4a and 7.7-4b 10 7.7-5a and 7.7-5b 10 7.7-6 through 7.7-15 10 CHAPTER 8 TEXT PAGES 8-viii 12 8-ix 12 8-xii 8-xiii 12 2

8.2-11 3 8.3-8 8 8.3-13 and 8.3-14 2 8.3-17 13 8.3-23 2 8.3-27 13

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 8 (Continued)

TEXT PAGES AMENDMENT 8.3-29 and 8.3-29a 9 8.3-30 9 8.3-34 12 8.3-37 12 8.3-41 13 8.3-47 13 8.3-49 9 8.3-51 through 8.3-53 1 8.3-65 and 8.3-65a 8 8.3-66 8 8.3-70 2 8.3-72 2 8.3-78 1 8.3-79 and 8.3-80 1 8.3-96 2 8.3-99 through 8.3-103 1 8.3-10 6 and 8.3-107 1 FIGURES 8.'l-9a through 8.1-9c 11 8 . 1-9d 7

8. 1-10 7 8.3-1a through 8.3-1c 11 8.3-1d and 8.3-1e 7 8.3-1K 11 8.3-2 11 8.3-15 2 8.3-19 11 8.3-24a 5 8.3-24b 2 8.3-24c 5 8.3-29 1 CHAPTER 9 TEXT PAGES 9-i 5 9-v 12 9-viii and 9-viiia 8 9-xiii 11 9-xiv and 9-xiva 12 9-xvi and 9-xvii 7 9-xviii and 9-xviiia 12 9-xix 7 9-xx 12

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 9 (Continued)

TEXT PAGES AMENDMENT 9-xxi 12 9-xxia 8 9-xxii 12 9-xxiii 11 9.1-3 and 9.1-4 13 9.1-7 through 9.1-9a 5 9.1-12 12 9.1-14 1 9.1 16 5 9.1-22 13 9.1-24 5 9.1-27 5 9.1-29 1 9.1-39 1 9.1-42 13 9.1-54 through 9.1-56 1 9.1-62 13 9.1-63 and 9.1-63a 13 9.1-65 5

9. 2-7 2 9.2-9 13 9.2-12 13 9.2-13 and 9.2-13a 13 9.2-15 13 9.2-16 and 9.2-17 5 9.2-18 7 9.2-19 through 9.2-22 5 9.2-26 8 9.2-27 and 9.2-27a 8 9.2-30 13 9.2-31 13 9.2-33 1,3 9.2-36 through 9.2-39 5 9.2-40 through 9.2-42 11 9.2-44 3 9.2-46 through 9.2-48 5 and 9.2-50 '.2-49 11 9.3-2 11 9.3-4 11 9.3-4a 8 9.3-5 11 9.3-20 5 9.3-23 5 9.4-4 and 9.4-4a 9 9.4-5 and 9.4-5a 9 9.4-6 9

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 9 (Continued)

TEXT PAGES AMENDMENT 9.4-7 and 9.4-7a 9 9.4-8 and 9.4-8a 9 9.4-9 9

9. 4-10 9 9.4-12 and 9.4-12a 9 9.4-17 2 9.4-21 2 9.4-30 4 9 . 4-59 through 9 .'-61 a 11 9.4-75a through 9.4-75c 2 9.4-79 2 9 . 4-85 9 9.4-86 2 9.4-91 and 9.4-92 2 9.4-99 2 9.4-102 and 9.4-103 ~ 2

9. 5-3 13 9.5-9, and 9.5-9,a 9 9.5-10 ,13

9. 5-19 9

9. 5-27 13 9.5-31 and 9.5-31a 5 9.5-33 through 9.5-40 11 9.5'-41 8 9.5-42 8 9.5-43 through 9.5-52 7 9.5-53 13 9 .5-54 through 9;5-60 7 9.5-61 8 9.5-62 through 9.5-70 7 9.5-71 and 9.5-71a 9 9.5-72 through 9.5-77 7 FIGURES 9.1-4 11 9.1-17 5 9.2-1 through 9.2-4 11 9.2-5 and 9.2-6 '7 9.2-7a through 9.2-7d 5 9.2-8 5 9.2-9 and 9.2-10 11 9.3-1 and 11 through 9.3-10 9.3-2'.3-5 11 9.3-11 and 9.3-12 7 9.3-13 11

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 9 (Continued)

FIGURES AMENDMENT 9.3-14 4 9.4-1 through 9.4-3 11 9.4-4 7

,9. 4-5 11 9.4-6 7 9.4-7 through 9.4-12 11 9.5-1 11

9. 5-3 11

9. 5-4 11 CHAPTER 10 TEXT PAGES 10-i 12 10-iii 5 10-vi 13 10.1-1 12 10.2-2 7 10.2-4 13 10.2-5 7 10.2-6 13 10.2-6a 7 10.2-7 13 10.2-8 and 10.2-8a 13 10.2-9 13 10.2-10 3 10.2-11 9 10.3-2 13 10.4-2 7 10.4-3 13 10.4-4 7 10.4-4a 13 10.4-5 2 10.4-6 5 10.4-12 7 10.4-14 13 10.4-17 through 10.4-17c 5 10.4-24 13 10.4-25 8 10.4-28 4

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 10 (Continued)

FIGURES AMENDMENT 10.2-10 7 10.3-1 11 10.3-7 11 10.3-8 13 10.4-1 11 10.4-2 7 10.4-3 through 10.4-7 11 CHAPTER 11 TEXT PAGES 1 1-vi'ii, 12 11-xi and 11-xii 12 11-xiia 8 11.2-1 5 1 1 ~ 3 2 5 11.3-6 13 11.3-14 13 11.3-19 7 1 1 ~ 3 33 7 11.4-16 and 11.4-16a 2 11.5-2 2 11.5-16 5 11.5-18 and 11.5-18a 4 11.5-24 3 11.5-25 3 11.5-26 2'1 FIGURES 11.2-2 and 11;2-3 11.2-4a through 11.2-4c 11 11.4-1 11 11.5-10 4 CHAPTER 12 TEST PAGES 12-ii 5 12-ix 5 12-vi and 12-via 12 12-vii 5 12-viii and 12-ix 12

NNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 12 (Continued)

TEXT PAGES AMENDMENT 12.1-1 through 12.1-3 11 12.1-5 1 12.1-10 11 12.1-12 and 12.1-13 11 12.2-5 12 12.2-11 4 12.2-16a 5 12.2-29 4 12.2-35 5 12.3-5 4 12.3-10 and 12.3-11 4 12.3-13 and 12.3-13a 13 12.3-14 12 12.3-15 11 12.3-17 4 1 2.3-20 11 12.3-21 8 12.3-22 11 12.3-22a 11 12.3-22b 11 12.3-23 11 12.3-24 11 12.3-25 1 12.3-26 12 12.3-30 4 1 2 ~3 3 1 1 12.4-1 and 12.4-2 5 12.4-3 and 12.4-4 12 12.4-5 and 12.4-6 5 12.4-7 12 12.4-8 through 12.4-25 5 12.5-1 13 12.5-2 4 12.5-5 11 12.5-6 13 12.5-7 13 12.5-8 11 12.5-9 13 12.5-10 11 12.5-12 and 12.5-12a 11 12.5-14 and 12.5-15 4 12.5-16 11 12.S-17 and 12.5-18 4 12.5-20 4 12.5-22 and 12.5-23 4

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 12 (Continued)

FIGURES AMENDMENT 12.3-1 through 12.3-6 7 12.3-7 through 12.3-10 12 12.3-12 through 12.3-19 12 CHAPTER 13

'TEXT PAGES 13-iii 7 13-iv and 13-iva 7 13.2-1 and 13.2-2 7 13.2-4 through 13.2-11a 7 13.2-13 'and 13.2-14 7 CHAPTER 14 TEXT PAGES 14-i 7 14-ii 12 14-iii and 14-iv 7 14-v 12 14-vi 7 14-vii 12 14-viii and 14-ix 7 14-xi 7 14-xiii 12 14-xiv 12 14-xv 13 14-xvi and 14-xvii 7 14-xix and 14-xx 7 14.2-5 through 14.2-7 7 14.2-8 12 14.2-9 7 14.2-10 through 14.2-12 12 14.2-13 through 14..2-20 7 14.2-22 and'4.2-23 12 14.2-25 through 14.2-27 7 14.2-28 through 14.2-31 12 14.2-32 7 14.2-34 and 14.2-35 7

'14.2-37 through 14.2-41, 7

WNP- 2 AMENDMENT NO. 13 February 1981 CHAPTER 14 (Continued)

TEXT PAGES AMENDMENT 14.2-130 12

14. 2-132 12

14. 2-132a 13

14. 2-133 7

14. 2-134 13

14. 2-135 12 14.2-136 and 14.2-137 13

14. 2-138 12 14.2-138a 13 14.2-139 12 14.2-140 13 14.2-141 7 14.2-,142 and 14.2-142a 12 14.2-143 through 14.2-145 12 14.2-146 7.

14.2-147 through 14.2-149 12 14.2-151 through 14.2-159 7 FIGURES 14.2-1 and 14.2-2 7 14.2-5 and 14.2-6 7 CHAPTER 15 TEXT PAGES 15-xxi 8 15-xxii and 15-xxiia 8 15-xxxvii 5

15. 0-14 11

15. 0-15 11 15.0-20 11 15.1-31 12 15.2-48 through 15.2-50 11

.15.2-52 through 15.2-57 11 15.2-67 8 15.2-68 8 15.2-71 and 15.2-71a 8 15.2-72 through 15.2-76 8 15.2-80a through 15.2-80d 8 15.4-8 5 15.4-49 5 15.6-34 9 15.6-34a 11 15.6-39 10

WNP-2 AMENDMENT NO. 13 February 1981 CHAPTER 14 (Continued)

TEXT PAGES AMENDMENT 14.2-42 12 14.2-43 and 14.2-43a 12 14.2-43b 7 14.2-44 through 14.2-53 12 14.2-54 and 14.2-54a 12 14.2-55 12 14.2-56 12 14.2-57 13 14.2-58 through 14.2-61 12 14.2-62 13 14.2-63 13 14.2-64 and 14.2-64a 12 14.2-65 and 14.2-65a 12 14.2-66 through 14.2-70 12 14.2-71 7 14.2-72 through 14.2-78 12 14.2-80 through 14.2-87 12 14.2-89 and 14.2-90 12 14.2-92 12 14.2-94 12 14.2-96 7 14.2-98 12 14.2-101 12 14.2-102 7 14.2-103 through 14.2-105 12 14.2-107 12 1 4. 2-109 13 14.2-109a 12 14.2-110 7 14.2-112 12 14.2-113 7 14.2-114 12 14.2.115 13 14.2-116 12 14.2-116a 13 14.2-117 7 14.2-118 12 14.2-119 and 14.2-119a 12 14.2-120 12 14.2-121 7.

1 4 . 2-122 12 14.2-123 13 14.2-124 and 14.2-124a 13 14.2-125 13 14.2-126 12 1 4'. 2-128 12 1.4. 2-129 7

WNP-2 AMENDMENT NO. 13 February 1981 V

CHAPTER 15 (Continued)

TEXT PAGES AMENDMENT 15.6-40 9 15.6-41 10 15.6-42 9 15.6-44 through 15.6-47 10 15.6-48 9 15.8-3 12 FIGURES 15.A.6-8 11 15.A.6-12 and 15.A.6-13 11 15.A.6-14a and 15.A.6-14b 11 15.A. 6-15 11 15.A.6-20 11 15.A.6-22 through 15.A.6-31 11 15.A.6-38 through 15.A.6-40 11 15.A.6-42 11 15.A.6;43 through 15.A.6-45 11 15.A.6-51 through 15.A.6-53 11 CHAPTER 17 TEXT PAGES 17.1-1 through 17.1-4 5 17.1-5 through 17.1-14 13 1 7 ~ 2 1 11 APPENDIX C TEXT PAGES C.2-3 4 C.2-16 4 C.2-18 through C.2-20 13 C.2-40 13 C.2-44 through C.2-46 13 C. 2-55 13 C . 2-63 13 C.2-65 and C.2-66 5 C. 2-68 9 C.2-70 7 C.2-72 7 C.2-82 13 C.2-P6 13

WNP-2 AMENDMENT NO. 13 February 1981 APPENDIX C (Continued)

TEXT PAGES AMENDMENT C.3-5 5 C.3-9 13 5

'.3-11 C.3-14 12 C.3-17. 2 C.3-17a 2 C.3-18 13 C. 3-22 5 C.3-23 8 C.3-25 8 C.3-32 '

8

. 3-33. 8 C.3-34, 8 C.3-35 11 C.3-39 13 C.3-41 13 C.3-45 5 C.3-47 8 C.3-49 13 C . 3-49a 9 C.3-51 13 C.3-54, 8 C. 3-55 throu gh C.3-57 13 C. 3-59 and C .3-60 7 C;3-61 13

,C . 3-67 8 C.3-70 9 C. 3-74 9 C.3-75 13 C.3-76 11 C.3-81 4 C.3-82 8 C.3-84 13 C.3-91 13 C.3-92 9 C. 3-102 8 C.3-107 8 C. 3-111 8 APPENDIX D TEXT PAGES Title Page

WNP-2 AMENDMENT NO. 13 February 1981 APPENDIX E TEXT PAGES AMENDMENT Title Page 10 E.1-1 7 E.2-1 12 E.3-1 8 E.3-3 through E.3-5 12 E.3-5a 3 E.3-9 3 E.3-10 5 E.7-1 through E. 7-5 13 E. 9-1 8

,E. 9-2 5 E.9-3 and E.9-3a 8 E.10-1 5 E.11-1 8 E.12-1 12 E.12-2 12 APPENDIX F TEXT PAGES Title Page APPENDIX G TEXT PAGES Title Page APPENDIX H TEXT PAGES Title Page 7 H.O-vii and H.O-viia 12 H.1.2-1 12 H.2.1-4 10 H.2.3-5 and H.2.3-5a 12 H.4.3-1 and H.4.3-1a 12 H.A.O-i 10 FIGURES H.A-1a through h 10

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS AND'RESPONSES (See Separate List)

NRC QUESTION/FSAR SECTION CROSS REFERENCE, (See'eparate List)

EMERGENCY PLAN

.(Chapter 13.3) 3-28 4 5-4 and 5-5 5 5-6a 5 5-7 through 5-9 5 7-1 5 7-2a 7 7-3 7 APPENDIX IX (Emergency Plan) 10

TABLE OF CONTENTS (Continued) Pacae 1.2.2.12.14 Normal Auxiliary A-C Power System 1.2-48 1.2.2.12.15 Diesel Generator Fuel-Oil Storage and Transfer System 1.2-49

l. 2. 2. 12. 16 Auxiliary Steam System 1.2-49 1.2 ~ 3 COMPLIANCE WITH NRC REGULATORY GUXDES 1.2-49 1.3 COMPARISON TABLES 1~3 1 1.3.1 COMPARISONS WXTH SIMILAR FACILITY DESIGNS 1~3 1 1.3.1.1 Nuclear Steam Supply System Design Characteristics 1.3-1 1.3.1.2 Power Conversion System Design Characteristics l. 3-1 1.3. 1.3 Engineered Safety Features Design Characteristics 1~3 1

1. 3. 1. 4 Containment Design Characteristics l. 3-1 1.3.1.5 Radioactive Waste Management Systems Design Characteristics 1.3-1 1.3.1.6 Structural Design Characteristics 1~3 1 1.3.1.7 Electrical Systems'esign Characteristics 1.3-1 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 APPLICANT WPPSS 1.4-1

1. 4.2 ENGINEER AND CONSTRUCTION MANAGEMENT BURNS AND ROE, INC. 1. 4-1 1;4.3 NUCLEAR STEAM SYSTEM SUPPLIER GENERAL ELECTRIC COMPANY 1.4-1 1.4.4 TURBINE-GENERATOR SUPPLIER WESTINGHOUSE ELECTRIC CORP. 1.4-2

l. 4. 5 TECHNICAL CONSULTANTS 1.4-2 1.4.5.1 R. W. Beck and Associates 1.4-2

WNP-2 AMENDMENT NO. 7 November 1979 TABLE OF CONTENTS (Continued) ~Pa e 1.4.5.2 The S. M. Stoller Corporation 1. 4-3

.1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION l. 5-1 1.5.1 DEVELOPMENT OF BWR TECHNOLOGY 1. 5-1 1.5.1.1 ACRS Concerns 1. 5-1 1.5.1.1.1 Containment Design Features to Minimize Effects of Bypass Leakage 1.5.1.1.2 Pipe Whip Protection Provisions 1.5.1.1.3 Design Criteria for Inactive Pumps and Valves 1.5-1 1.5.1.1.4 Main Steam Line Leakage Control System 1.5-2 1.5.1.1.5 Mitigation of Consequences of Control-Rod Drop Accident 1. 5-2 1.5.1.1.6 Anticipated Transients Without Scram l. 5-2 1.5.1.2 Current Development Program l. 5-3 1.5.1.2.1 Loose Parts Detection 1. 5-3 1.5.1.2.2 Mark II Containment Suppression Pool Dynamic Loading 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS 1. 7-1 1.8 CROSS REFERENCE FOR PIPING AND INSTRUMENTATION DRAWINGS 1. 8-1

'1.9 ACRONYMS l. 9-1

WNP-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 APPLICANT WPPSS Washington Public Supply System is a municipal corporation, and a joint operating agency of the State of Washington,

'organized in January 1957, pursuant to Chapter 43.52 of the Revised Code of Washington as amended. WPPSS is composed of 19 operating public utility districts of the state of Wash-ington and the cities of Richland, Tacoma and Seattle, Wash-ington. WPPSS assumes the responsibility for safe operation and maintenance of the plant and for providing related ser-vices as described'n Chapter 13.'.4.2 ENGINEER AND CONSTRUCTION MANAGEMENT BURNS AND ROE, INC.

Burns and Roe, Inc. (B&R) has been retained by WPPSS to pro-vide engineering and construction management and quality assurance services for the design and construe'tion of the plant, integrating the major plant items furnished by General Electric Company and Westinghouse Electric Corporation.,

Burns and Roe was,also the engineering consultant and the construction manager for the Hanford No. 1 generating plant.

Burns and Roe has been continuously engaged in construction or engineering activities since 1935; Burns and Roe was founded in 1932, and was incorporated in 1935 as Burns and Roe, Inc. Burns and,Roe has been active in the fields of power generation and distribution, sea water and brackish water desalination, waste water renovation, en-gineering, design and/or construction management services for over 50 thermal power generating units representing more than 11,400,000 kilowatts of'ew generating capacity of which more than 4,800,000 kilowatts is nuclear.

1.4.3 NUCLEAR STEAM SYSTEM SUPPLIER GENERAL ELECTRIC COMPANY The General Electric Company (GE) has been awarded the con-tracts to design, fabricate, and deliver the direct cyclethe boiling water nuclear steam supply system, to fabricate first core'of nuclear andfuel, and to provide technical direc-tion of installation startup of this equipment. GE has engaged in the development, design, construction and opera-tion of boiling water reactors since 1955. Table 1.4-1 lists over 90 GE reactors completed, under construction, or on order. Thus'E has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation and startup of reactors.

1.4-1

~ WNP-2 1.4.4 TURBINE-GENERATOR SUPPLIER -, WESTINGHOUSE ELECTRIC CORP.

WPPSS has awarded a contract to Westinghouse Electric Corp.

to design, fabricate and deliver the turbine-generator for WNP-2 as well as to provide technical assistance for installa-tion and start-up of this equipment.

Westinghouse Electric Corp. has a long history in the appli-cation of turbine-generators in nuclear power stations going back to the inception of commercial electrical power pro-duction utilizing nuclear facilities. Westinghouse furnished the turbine-generator unit for'hippingport No. 1. This unit was shipped in 1956. Westinghouse also furnished the turbine-generator unit for Yankee Atomic Power Company Rowe No. l.

This unit was shipped in 1959. San Onofre No. 1 and Connec-ticut Yankee,,Haddam Neck No. 1 unit went into commercial operation in 1968. Westinghouse nuclear turbine-generators produced over 300 billion kilowatt hours of electricity through May, 1976, when twenty-five nuclear turbine-generators totaling over 16,500 megawatts were in service. By 1984, seventy-five Westinghouse nuclear turbine-.generators should be in service producing over 61,319 megawatts. Inlet steam

. pressures of, these units vary between 750 psig and 1000 psig and electrical outputs vary from 500, kW to 1,090,000 kW.

Westinghouse is therefore'competent to design, fabricate, deliver and erect the turbine-generator set and to provide technical assistance for the start-up of this equipment.

1.4.5 TECHNICAL CONSULTANTS In connection with this project, and in addition to the engineering consultant (BGR) describe'd above, WPPSS has engaged R.W. Beck and Associates as consulting engineer and the S. M. Stoller Corporation as nuclear fuel consultant.

1.4.5.1 R. W. Beck and Associates The independent consulting firm of R. W. Beck and Associates has been employed as the consulting engineer for Washington Public Power Supply System's Nuclear Project No. 2. This firm was also employed as a consulting engineer for Hanford No. 1. The current employment level of R. W. Beck and Associates is approximately 350 including approximately 200 professional engineers. Having extensive experience in preparing engineering feasibility and financing studies and reports necessary for the success of utility and civic im-provement, projects, the firm is well qualified for employ-ment as a consulting engineer and was chosen as a result of its experience.

l. 4-2

WNP-2 The duties of the consulting engineer are briefly summarized as follows: prepare estimates of plant capability, energy potential, usability within area loads and resources, the cost of power and energy output of the project, and generally determine the feasibility of the project. These duties will include assisting in preparation of a Bond Resolution, pre-paration of an engineering report, schedules for investment of funds, schedules for debt service payments, and other engineering services necessary to facilitate the financing of the project.

1.4.5.2 The S. M. Stoller Corporation The S. M. Stoller Corporation has been employed as the nuclear fuel consultant to the Washington Public Power Sup-ply System with principal responsibilities for preparation of bid specifications for nuclear fuel and evaluation of the technical and economical impact of the bids resulting from those specifications. The services of S. M. Stoller Corp.

have been required throughout the project as a consultant for assessment of core design, fuel cycle management, and waste management. The S. M. Stoller Corp. currently has approximately 30 senior engineers, well experienced in not only nuclear fuel management, but also in technological and environmental assessment of commercial nuclear generating fac-ilities. The S. M. Stoller Corp. has been a nuclear con-sultant for more than thirty of the nation's largest nuclear oriented utilities and other organizations.

1. 4-3

NNP-2 AMENDMENT NO ~ 12 November 1980 TABLE 1.4-1 COMMERCIAL NUCLEAR REACTORS COMPLETEDE UNDER CONSTRUCTION, OR IN DESIGN BY GENERAL ELECTRIC YEAR YEAR RATING OF OF STATION UTILITY (MWe) ORDER STARTUP Dresden 1 Commonwealth Edison 200 1955 1960 Humboldt Bay Pacific G&E 69 1958 1963 Kahl Germany 15 1958 1961 Garigliano Italy 150 1959 1964 Big Rock Point Consumers Power 70 1959- 1965 JPDR Japan 11 1960 1963 KRB Germany 237 1962 1967 Tarapur 1 India 190 1962 1969 Tarapur 2 India 190 1962, 1969 GKN Holland 52 1963 1968 Oyster Creek JCP&L 640 1963 1969 Nine Mile Point 1 Niagara Mohawk 625 1963 1969 Dresden 2 Commonwealth Edison 809 1965 1970 Pilgrim Boston Edison 664 1965 1972 Millstone 1 NUSCO 652 1965 1970 Tsuruga Japan 340 1965 1970 Nuclenor Spain 440 1965 1971 Fukushima 1 Japan 439 1966 1971 BKW KKM Switzerland 306 1966 1972 Dresden 3 Commonwealth Edison 809 1966 1971 Monticello Northern States 545 1966 1971 Quad Cities 1 Commonwealth Edison 800 1966 1972 Browns Ferry 1 TVA 1098 1966 1974 Browns Ferry 2 TVA 1098 1966 1975 Quad Cities 2 Commonwealth Edison 800 1966 1972 Vermont Yankee Vermont Yankee 514 1966 1972 Peach Bottom 2 Philadelphia Electirc 1065 1966 1974 Peach Bottom 3 Philadelphia Electric 1065 1966 1974 Fitzpatrick PASNY 821 1966 1975 Bailly NIPSCO 660 1967 Shoreham LILCO 819 1967 Cooper Nebraska PPD 778 1967 1974 Browns Ferry 3 TVA 1098 1967 1977 Limerick 1 Philadelphia Electric 1055 1967 Hatch 1 Georgia 786 1967 1975 Fukushima 2 Japan 762 1967 1974 Brunswick 1 Carolina P&L 821 1968 1977 Brunswick 2 Carolina P&L 821 1968 1975 Arnold, Iowa ELP 569 1968 1975 Fermi 2 Detroit Edison 1056 1968 Limerick 2 Philadelphia Electric 1055 1969 Hope Creek 1 PSE&G 1067 1969 Hope Creek 2 PSE&G 1067 1969 1.4-4

WNP- 2 AMENDMENT NO. 13 February 1881

Response

The consequences of an anticipated transient without scram (ATWS) are mitigated by tripping the recirculation pumps and by manual insertion of the control rods. (For more information, see 15.8.).

1.5.1.2 Current Development Program 1.5.1.2.1 Loose Parts Detection A Loose Parts Detection System for will be provided. See 7.7.1.12 a system description.

1. 5-3

1.5.1.2.2 Mark II Containment Suppression Pool Dynamic Loading The Washington Public Power Supply System, in conjunction with- other Mark II owner utilities, has submitted a desig'n basis document designated as "Mark II Containment Dynamic Forcing Function Information Report" (DFFIR), NED0-21061, and NEDE-21061P describing the suppression pool dynamic loading phenomena during a safety/relief valve actuation or LOCA event. The evaluation of that, design basis document against the current WPPSS Nuclear Project No. 2 design was prepared and submitted to the NRC.' verification'rogram to demon-strate the conservatism of the DFFIR has been sponsored the Mark II owners and is described in the "Mark II Con- by tainment Supporting Program Report", GE Document NEDO-21297.

1.5-4

WNP-2 AMENDMENT NO. 13 February 1981 Page 10 of 10 TABLE l.6-1 (Continued)

FSAR PORTIONS REPORT WHERE NUMBER TITLE REFERENCED 4

WAPD-T-416 WIGLE -',.'Program for the Solu- 4.3(10)*

tion of the Two-Group Space-Time Diffusion Equations in Slab Geometry (1964)

Irradiation Behavior of 4.2(6)*

Zircaloy-Clad Fuel Rods Con-WAPD-TM-629'PPSS-74-2-. taining Dished-End U02 Pellets (July 1967)

Washington Public Power Supply 3.8 R2 and Sup- System Sacrificial Shield Wall plements 1 WPPSS-74 R2A and WPPSS-74-2-R2B

,Report sub- Engineering Evaluation of the 3.8, 6.2 mitted on WNP-2 Sacrificial Shield Wall letter GO2-80-172, August 8, 1980 Report sub- Engineeri'ng. Evaluation of the 3.8, 6.2 mitted on ' 'WNP-2 Sacrificial Shield Wall, letter '

Supplement No. 1 GO2-80-1'82, August 19, 1980

• These FSAR sections do not reference these documents directly

'but indirectly via NEDE-20944P. NEDE-20944P and FSAR section numbers correspond. The numbers in parentheses are reference numbers from NEDE-20944P for the respective section numbers.

1. 6-11

WNP-2 AMENDMENT NO. 5 August 1979 TABLE OF CONTENTS (Continued) ~Pa e 3.4.1.3 Identification of Structures, Systems, and Components 3.'4-2 3.4.1.4 Description of Structures, Systems, and Components 3.4-2

3. 4. 1. 4. 1 Flood Protection Requirements 3.4-2

3. 4. 1. 4. 1. 1 External Flood Protection Requirements 3.4-2

l. l. 2

3. 4. 4. Internal Flood Protection Requirements 3.4-2a 3.4.1.4.2 Groundwater Protection Requirements 3.,4-3 3.4.1.5 Flood Protection Measures 3.

4-5'.4-5 3.4.1.5.1 External Flood Protection Measures 3.4.1.5.2 Internal Flood Protection Measures 3.4-5 3.4.1.6 Emergency Flood Protection 3.4-5a 3.4.2 ANALYSIS PROCEDURES 3.4-5a 3.5 MISSILE PROTECTION 3.5-1 3.5.1 MISSILE SELECTION AND DESCRIPTIONS 3.5-1 3.5.1.1 Internally Generated Missiles (Outside Containment) 3.5-1 3.5.1.1.1 Systems Available for Safe Shutdown 3.5-1 3.5.1.1.2 Ability of Structures, Systems, and Components to Withstand Missile Effects 3.5-2 3.5.1.1.3 Missile Selection 3.5-6 3.5.1.1.3.1 Valves 3.5-7 3.5.1.1.3.2 Thermowells and Sample Probes 3.5-7 3.5.1.1.3.3 Bolts 3.5-8 3.5.1.1.3.4 High Speed Rotating Equipment 3.5-8 3.5.1.2 Internally Generated Missiles (Inside Containment) 3.5-8 3.5.1.2.1 Systems Available for Safe Shutdown 3.5-8 3.5.1.2.2 Ability of Structures, Systems, and Components to Withstand Missile Effects 3.5-8 3 3x

WNP-2 AMENDMENT NO. 12 November 1980 TABLE OF CONTENTS (Continued) ~Pa e 3.5.1.2.3 Missile Selection 3.5-9 3.5.1.2.3.1 Valves 3. 5-10 3.5.1.2.3.2 Thermowells and Sample Probes 3. 5-10 3.5.1.2.3.3 Bolts 3. 5-10 3.5.1.2.3.4 High Speed Rotating Equipment 3. 5-10 3.5.1.2.4- Falling Objects 3.5-11 a 3.5.1.2.5 Secondary Missiles Postulated Generated by Primary Missiles 3.5-11 a 3.5.1.3 Turbine Missiles 3. 5-12 3.5.1.3.1 Turbine Placement and Orientation 3; 5-12 3.5.1.3.2 Missile Identif ication and Characteristics 3. 5-12 3.5.1.3.3 Low Trajectory Missiles 3. 5-'13 l

,3.5.1.3.4 High Trajectory Missiles 3. 5-13 3.5.1.3.5 Turbine Overspeed Protection System 3. 5-15 3.5.1.3.6 Inspection and Testing 3. 5-16 3.5.1.3.7 Turbine Characteristics 3. 5-16 3.5.1.4 Missiles Generated by Natural Phenomena 3. 5-16.

3.5.1.4.1 Tornado Generated External Missiles 3. 5-17 C

3.5.1.4.2 Tornado Generated Internal Missiles 3.5-20 3.5.1.4.3 Flood Generated Missiles 3.5-21 3.5.1.4.4 Protection and Design Procedures 3.5-21 3.5.1.5 Missiles Generated by Events Near the Site 3.5-21 3.5.1.6 Aircraft Hazards 3.5-22 3.5.1.6.1 DELETED 3.5-22 3-x

WNP-2 AMENDMENT NO. 12 November 1980 TABLE OF CONTENTS (Continued) ~Pa e 3.5.1.6.2 DELETED 3.5-22 3.5.2 SYSTEMS TO BE PROTECTED 3.5-22 3.5.3 BARRIER DESIGN PROCEDURES 3.5-24 3- XB

~ ~

i WNP-2

TABLE OF CONTENTS (Continued) Pacae .

3.6.2.2.3 Material Properties Under Dynamic Loads 3.6-42 3.6.2.2.3.1 Dynamic Yield Strength 3.6-42 3.6.2.2.3.2 Maximum Strain of Tension Members 3.6-42 3.6.2.2.3.3 Maximum Deformation of Flexural Members 3. 6-42 3.6.2.2.3.4 Materials and Proportions of Structural Shapes 3. 6-43 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability 3.6-43 3.6.2.3.1 Dynamic Analysis Methods for Jet Impingement Effects 3.6-43 3.6.2.3.2 Jet Impingement Effect 3.6-47 3.6.2.3.2.1 Physical Separation 3.6-47 3.6.2.3.2.2 Jet Impingement Evaluation 3.6-47 3.6.2.3.2.3 Postulated Pipe Rupture Locations Inside Containment 3.6-48 3.6.2.3.2.4 Signals from Primary Containment 3.6-48 3.6.2.3.2.5 Signals to the Primary Containment 3.6-48 3.6.2.3.2.6 Power Requirement Inside Primary Containment 3.6-48 3.6.2.3.2.7 Mechanical Engineered Safety Systems 3.6-50 3.6.2.3.2.8 Jet Impingement on Major Structures Inside Primary Containment 3.6-50 3.6.2.3.3 Pipe Whip Restraints 3.6-51 3.6.2.3.3.1 . Definition of Function 3. 6-51 3.6.2.3.3.2 Pipe Whip Restraint Features 3. 6-51

.3.6.2.3.3.3 Pipe Whip Restraint Loading 3.6-53 3-xvi3.

WNP- 2 TABLE OF CONTENTS (Continued) PacCe 3.6.2.3.4 Pipe Whip Effects on Safety Related Components 3.6-54 3 '.2.3.4.1 Pipe Displacement Effect on Components in Same Piping Run 3.6-54 3.6.2.3.4.2 Pipe Displacement Effects on Structures, Other Systems and Components 3.6-55 3.6.2.4 Guard Pipe Assembly Design Criteria for Dual Barrier Containment 3.6-56 3.6.2.5 Implementation of Criteria for Pipe Whip and Jet Impingment Protection 3.6-56 3.6.2.5.1 Piping Systems outside Primary Containment 3.6-56 3.6. 2 5. 2 Piping Systems Inside Primary Containment 3.6-56 S.e.2.5.3 System Requirements Subsequent to Postulated Pipe Rupture 3.6-57 3.6.2.5.3.1 Control Rod Insertion Capability 3.6-57 3.6.2.5.3.2 Core Cooling Requirements 3.6-58 3.6.2.5.3.3 Maximum Allowable Break Areas 3.6-58 3.6.2.5.3.4 Break Combinations 3.6-58 3.6.2.5.3.5 Required Cooling Systems 3.6-58 3.6.2.5.3.6 Containment System Integrity 3.6-59 3.6.2.5.4 System By System Description of Pipe Whip Protection 3. 6-59 3.6.2.5.4.1 Main Steam System 3.6-59 3.S 2.5.4.2 Reactor Feedwater System (Inside Primary Containment) 3.6-61 3 6.d.5.4.3 Reactor Water Cleanup System (RWCU) 3.6-62 3-xv iii

WNP-'2 AMENDMENT NO. 5 August 1979 LIST OF FIGURES (Continued)

Number Title 3.5-46 Diesel Generator Bldg Roof Plan and Sections 3.5-47 Diesel Generator Bldg Elevations 3.5-48 Standby Service Water Pumphouse Elevation and Details 3.5-49 Standby Service Water Pumphouses lA 6 1B Plans at El. 441'-0" and Roof 3.5-51 Radwaste and Control Building Control Room Fresh Air Intake and Exhaust Shield Walls, Slab and Hood for External Missiles 3.5-52 DELETE 3.6-1 Analytical Model 3.6-2 Required Resistance of Structures (R )

3.6-3 NOT USED 3.6-4 Flexible Type Pipe Whip Restraint Configuration 3.6-5 NOT USED 3.6-5a Rigid Type Pipe Whip Restraint Config-uration (Continues through Figure 3.6-5e.)

3.6-6 NOT USED 3.6-6a Pipe Whip Restraint Installation Main Steam System (Continues through Figs. 3.6-6h, 3.6-6j and 3.6-6k.)

3-Lxi ii

WNP-2 AMENDMENT NO. 4 June 1979 LIST OF FIGURES. (Continued)

Pipe Whip Restraint Installation RWCU,System (Continues through Figure 3.6-7c.

Pipe Whip Restraint Installation LPCS System Pipe Whip Restraint, Installation RHR'System (Continues through Fig. 3.6-9b.)

Pipe Whip Restraint. Installation RHR System 3-Lxiiia

TABLE 3.2-1 (Continued)

Quality Scope Group of Safety Loc- Classi- Quality Seismic Com-Principal Component Supply Class ation fication Class Category ments (1) (2) (3) (4) (5) (6) (7)

28. Primary Containment Cooling System (Figure 3.2-15)

.1 Piping and Valves up to outermost isolation valves, containment purge and ex-haust P CgR B

29. Standby Gas Treatment System (Figure 3.2-16)

.1 Filter Units P R B

.2 Fans P R B

.3 Piping and Valves P R B

30. Primary .Containment Atmospheric Control System (Figure 3.2-17)

.1 Piping and Valves P C,R B

.2 Equipment P R B

31. Other HVAC (Figures 3.2-18 to 20).

.1 Reactor Building (non-essential) P G R N/A IZ II (10)

.2 Reactor Building (essential).P 3 R N/A I I

.3 Turbine Building P G G N/A ZI II (28)

.4 Radwaste Building P G N N/A IZ IZ (28)

.5 Control Room, Critical Switchgear Area, Cable Spreading Area (non-essential) P N/A ZI 0

TABLE 3.2-1 (Continued)

{}uality Scope Group of Saf ety Loc- Classi- quality Seismic Com-Principal Component Supply Class ation fication Class 'Category ments D) (2) (.3). (4) (5) (6) (7)

~ 6 Control Room, Critical Switchgear Area, Cable Spreading Area (essential) P N/A I

~ 7 Diesel Generator Building P N/A Z. (29)

.S Standby Service Water Pumphouse P N/A (29) 32 Condensate Storage and

'1 Transfer (Figure 9~2-6)

Condensate storage tank

.2 Piping and valves P

P-G 0 C II II II II (20)

G OuTIRcW D

~ 3 Pumps P G 0 D II II

33. Instrument and Sample Lines See note 12

34. Fuel Storage Facilities

.1 Fuel Pool/Dryer Separator Liner P R N/A

~ 2 Storage Racks and Supports GE R N/A

35. Building Cranes

.1 Reactor Building P 3 R N/A I- I,

.2 Turbine Building P G T N/A II. ZZ

~ 3 Radwaste Building P G W N/A II.

.4 Standby Service Water Pumphouse P G P N/A. II'I 5 Miscellaneous Areas P G P,K,TsS N/A ZE'I

36. Instrument and Service Air (Figure 9.3-1}

.1 Piping and Valves P' G RWTgO D (10)

.2 Compressors G T D II II ZI II

~ 3 VesselS P G T D

3.5.1.4.3 Flood Generated Missiles The design basis flood elevation discussed in 3.4 and defined in 2.4, exceeds the flood levels associated with breaches of the Grand Coulee Dam. The final plant grade level is higher than the design basis flood. Therefore, flood generated missiles are not considered in the design of the Seismic Category I safety related structures and installations.

3.5.1.4.4 Protection and Design Procedures Systems protected from missiles generated by natural phenomena, and barrier design procedures are described in 3.5.2 and 3.5.3 respectively.

3.5.1.5 Missiles Generated by Events Near the Site Hazards due to missiles postulated in the design basis ex-plosions or accidents at nearby industrial plants, military facilities pipelines or storage facilities as discussed in 2.2 can be discounted because of their remote relationship to the WNP-2 plant.

Transportation facilities in the immediate vicinity of the plant consist of a railroad system owned and operated by DOE and a DOE owned Reservation Road System which connects the Reservation with two (2) nearby state highways. The DOE Railroad and Road Systems are restricted and used in support of the Hanford Operations and are discussed in 2.2.

There is no commercial river traffic passing the site on the Columbia River; only small pleasure craft normally use the river. For detailed discussion of current and potential future activities on the Columbia River, see 2.2.

I For an evaluation of potential accidents at the foregoing facilities, and determination of design basis events, see 2.2.3.1.

3.5-21

WNP-2 AMENDMENT NO. 7 November 1979 3.5.1.6 Aircraft Hazards All airports, commercial, private and military are located outside a ten mile radius of the plant. Projected operations at these. airfields are examined in 2.2.2. The frequency of, aircraft flights per year does not pose a threat. to WNP-2 operation.

Military installations do -not threaten the plant site since

'one are located within 20 miles. Refer to 2.2.2.

As discussed in 2.2.3, the probability of the aircraft crashing

=

into the plant is less than 1 x 10-7. Therefore, aircraft is discounted as a credible missile.

3. 5. l. 6. 1 DELETED: see 3.5.1.6.

3.5.1.6.2 DELETED: see 3.5.1.6.

3.5.2 SYSTEMS TO BE PROTECTED The structures, systems and components necessary for bringing the plant to a safe shutdown and the protection provided for these structures, systems and components is discussed in 3.5.1 and 3.5.2.

A description of the protection provided for the safety re- .

lated structures located outdoors, against tornado generated missiles, is furnished in 3.5.1.4; by turbine missiles, in 3.5.1.3; and by a sesimic event, in 3.8.4.

3.5-22

3.5-11 Re ulator Guide 1.14, Rev. 1, Reactor Coolant Pump F ywheel Integrity , August. 1975.

3.5-12 Miller, D. R. and Williams, W. A., Tornado Protection for the Spent Fuel Pool, General Electric Company, APED-5696, November 1968.

3.5-13 "Protection Against Pipe Breaks Outside Containment",

Burns and Roe, Inc., Hempstead, New York, Report No. WPPSS-74-2-R3, April, 1974.

3.5-14 "A Review of Procedures for the Anal sis and Desi n of'oncrete Structures to Resist Mis'sil'e'm act Effects R.P. Kennedy, Nuclear and Systems Scz.ences Group, Holmes and Narver, Inc., September 1975.

3.5-15 "Anal sis of the Probabilit of the Generation and Strike of Missiles from a Nuclear Turbine" March, 1974 by Westinghouse Electric Corporation Steam Turbine Division Engineering.

3.5-27

WNP-2 AMENDMENT NO. 13 February 1981 TABLE 3.5-1 SYSTEMS DESCRIPTION OUTSIDE CONTAXNMENT SYSTEMS SEISMIC &

AVAILABLE QUALITY FOR A SAFE SECTION FIGURES CLASSIFICATION SHUTDOWN FUNCTION FSAR 3.5 RESPECTIVELY RCIC MAXNTAIN 5.4.6 ~

9-14 I, I RPV WATER 7.4.1.1 INVENTORY HPCS MAINTAIN 6.3, 9-14 I, I RPV WATER 7.3.1.1.1.1 INVENTORY SSW HEAT 7.3.1'.1.6 9-14 I, I REJECTXON RHR A MAINTAIN 5.2, 1-8 I, I WATER'N- 7.3.1.1.1.4 C VENTORY & 6.3, 5.4.7 DECAYS HEAT REMOVAL CRD REACTIVITY 7.7.1.2 15, Ii I CONTROL RFW MAXNTAIN 5.4.9 9-14 RPV WATER INVENTORY LPCS MAINTAIN 6.3, 1-8 I/ I RPV WATER 7~3 ~ 1 ~ 1 ~ 1 3.

INVENTORY Note: Identification of missiles to be protected against, their sources, and bases 'for'election are discussed in 3.5.1.1.3. The ability of the structures, systems and components to withstand the effects of selected internally'enerated missiles is discussed in 3.5.1.1.2.

3.5-28

a. For breaks not, involving recirculation piping, at least two LPCI pumps or one core spray system is available for core cooling.

b,. For breaks involving recirculation piping, at lease one core spray line and 2 LPCI pumps, or 2 core spray lines, are available for core cooling.

c. For a LOCA with y total effective break area less than 0.7 ft , either the .HPCS or ADS is available. for reactor depressurization.

d. For liquid breaks, such as cleanup suction or the combination of liquid and steam breaks whose total break area is less than 0.7 ft2 the ADS system is required. for depressurization, at -least 6 ADS valves are available.

e. For breaks less than the equivalent flow area of one open ADS valve, at least 6 ADS valves are available. However, the required number of ADS valves is one- less for each additional steam break area equivalent to the area of one open ADS valve.

3.6.2.5.3.6 Containment System Integrity The following was considered in addressing the LOCA dynamic effects with respect to containment system integri'ty:

a ~ Leak tightness of the containment fission product barrier is assured throughout any LOCA.

b. For those lines which penetrate the containment and are closed during normal operation, the inboard isolation valves are as close as prac-ticable to the reactor pressure vessel. This arrangement reduces the length of pipe subject to a pipe break.

3.6.2.5.4 System by System Description of, Pipe Whip Protection 3.6.2 '.4.l Main Steam System a ~ System Arrangement The main steam system consists of four, 26-inch lines which are arranged inside primary contain-

3. 6-59

WNP-2 AMENDMENT NO. 9 April 1980 ment with mirror image symmetry about the 0'nd 180'orth-south azimuth. 'The lines exit the reactor pressure vessel on opposite sides of primary containment and drop down vertically in two parallel pairs to the main steam relief valve platform at elevation 541 ft. where they are routed horizontally, in parallel, in the northeast and northwest quadrants to the azimuth. At this point, the four lines 0'orth drop vertically in parallel, to an elevation just above the .diaphragm floor. The main steam isolation valves are located here. The four lines exit the containment nearest the north azimuth at elevation 500 ft. (approx.). The two feedwater piping loops are described in 3.6.2.5;4.2 and are routed near the main steam lines.

'b. Pipe Whip Protection The postulated pipe breaks and pipe whip restraints for the four main steam lines, are shown in Figures 3.6-12a through'.6-15a. Where pipe breaks are postulated inside primary con-tainment, the main steam lines are restrained to prevent the unacceptable motion of these pipes.

These restraints are mounted on the side of the sacrificial shield wall structure, as well as on radial beams which extend from the sacrificial shield wall to the primary containment vessel wall. A sliding beam seat at the primary con-tainment wall, permits the beam to grow axially and also permits the primary containment wall to move relative to the sacrificial shield wall.

A structural steel frame (see Figures 3.6-36a, 3.6-36b, and 3.6-36c) between the drywell diaphragm floor and the containment vessel, in the area of the main steam isolation val'ves, is provided for mounting of pipe whip restraints.

The structure is designed with vertically sliding connections at the containment vessel, to allow for differential thermal expansion between the containment vessel and the diaphragm floor.

c. Verification of Pipe Whip Protection Adequacy Sufficient pipe whip protection is provided for the main steam system to assure safety as defined

'. 6-60

(2) Equipment access hatch 1000 psf (3) Supression chamber access 150 psf hatch

'd ~ Live load on temporary construction scaffolds.

e. Operating weight of fluid in attached normally empty piping, headers and penetrations.

Head of water, 23'-6" high, on the refueling bellows seal with the containment vessel head removed and coincident hydrostatic pressure (under the Refueling Condition).

go Same as 3.8.2.3.4 (f) above except without the containment vessel head removed. (under the Flooded Condition).

3.8.2.3.5 Mechanical Piping Loads C

Mechanical piping loads consist of:

a ~ Piping reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition.

b. Pipe reactions under thermal conditions generated by a postulated break and including (a) above.

c~ Equivalent static load generated by the reaction of a broken high-energy pipe during a postulated break (and including an appropriate dynamic load factor to account for the dynamic nature of the load) .

d. Jet impingement equivalent static load generated by the DBA postulated break (and including an appropriate dynamic load factor to account for the dynamic nature of the load).

e. Pipe reactions and thermal conditions during an event causing external pressure.

3.8-25

A description of certain loads included among those listed above follows in detail:

3.8.2.3.5.1 Je't Forces in Drywell The drywell shell, personnel air lock, equipment hatch, jet deflectors and the removable top closure head are designed and constructed to withstand, in combination with other loads, jet forces consisting of either steam and/or water at 340 F and applied as follows:

Jet Force Area Sub'ected From the closure head flange to the top of the head 33 kips 26 sq. in.

From the closure head flange down to the drywell floor 534 kips 429 sq. in.

A jet force is considered to occur in any direction but is not considered to occur simultaneously with another'jet force; however, a jet force is considered to occur coincident with the drywell internal design press'ure of 45 psig and design temperature of 340 F. Local yielding may take place on the drywell shell from the jet force, but the shell will not rupture. On the top closure head and other areas, where the shell is not backed up by concrete, the primary stresses re-sulting from this combination of loads do not exceed the values specified in the ASME Code, Section IlX, paragraph NE-3131 (c) at a temperature of 340 3.8.2.3.5.2 Vent Pipe (Downcomer) Thrusts ~

The vent pipes (downcomers) and their connections to the dry-well floor are designed for the following loads:

a. Jet Blowdown Thrust A jet force of 20,000 lbs acting upward on each of the downcomers is considered to occur simul-taneously with the internal design pressure of 45 psig in the drywell and suppression chamber and the design temperature of 275 F in the dry-well, wetwell and drywell floor concrete.

3.8-26

b. Initial and Final Test Conditions A force equal to design pressure multiplied by the flow area of the vent pipe.

c. Accident Conditions Forces obtained from 3.8.2.3.5.2 (a) except that the temperature in the drywell is taken as 340oF and the temperature in the suppression chamber is taken as 275 F. The drywell floor concrete temperature is taken as 95oF.

3.8.2.3.5.3 Pipe Whip Pipe whip protection support rings, which are fully circum-ferential rings, are attached to the primary containment ves-sel at elevations 516'-6" and 542'-7-1/4". The basic function of these rings is to support pipe whip protection framework and to adequately distribute pipe whip loading into the vessel.

The pipe rupture loading is applied to the vessel through the support rings during the Normal Operating Condition at normal o crating temperature and at atmospheric pressure, as well as ope during an Incident Condition at maximum temperature o 340oF and at design pressure. The primary containment vessel anal-ysis includes the effects of a pipe rupture at any single location.

For further discussion on function and design of load trans-mitting members see 3.6.

3.8.2.3.6 Thermal Loads The thermal loads in the primary containment vessel steel are produced by the presence of temperature gradients within the containment and its appurtenances. Thermal effects and loads during normal operating conditions are based on the most crit-ical transient or steady state condition. Thermal loads are also considered under thermal conditions generated by a postu-lated pipe break.

3.8-27

3.8.2.3.7 Construction Loads

a. Wind load in the projected area of the steel primary containment vessel before the completion of the reactor building in accordance with re-ference 3.8-1, with a'basic wind of 100 mph as discussed in 3.3~

b. Snow loads before the completion of the reactor building. ~

3.8.2.3.8 Missile Loads There are no external missile loads considered since the pri-mary containment vessel is protected by the biological shield wall.

Potential internal missiles and protection provisions are discussed in 3.5.

3.8.2.3.9 Loss-of-Coolant Accident Loads The loss-of-coolant accident (LOCA) imposes pressure and thermal loads plus jet forces associated with coolant flow from any ruptured pipe within the containment. This LOCA loading condition is determined by analysis of the transient pressure and temperature effects which occur during a loss-of-coolant accident. .The governing'esign condition for the.

LOCA is discussed in Chapter 6.

3.8.2.3.10 Accident Recovery Loads Among the postulated loss-of-coolant accidents there may be an accident within the drywell that requires a "last ditch" contingency flooding of the pressure suppression chamber and the drywell to an elevation above the top of the active fuel zone in the reactor vessel as indicated in 3.8.2.3.12h; and, with the primary containment vessel head not removed,,

the reactor vessel cavity outside the primary containment vessel and above the refueling bellows seal flooded to a level above the refueling bellows seal noted in 3.8.2.3.12h.

The structural design criteria for the primary containment vessel are consistent with the provisions, of Regulatory Guide 3.8-28

WNP-2 AMENDMENT NO. 12 November 1980

3. 8-10 ACI 318-1971, "Building Code Requirements for Reinforced Concrete", American Concrete Institute (1971) .

3.8-11 AISC, "S ecification for Desi n, Fabrication and Erection of Structural Steel for Buildin s , American Institute o Steel Construction 1 9).

3.8-12 ANSI N45.2.5, "Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants",

Draft 3, Revision 1, January 1974, American National Standards Institute.

3. 8-13 Letter GI2-75-10, W. R. Butler to J. J. Stein, Transmittin Re uest for Additional Information, ate January 1, 5, on Drywell to Wetwe Leakage Study, Docket 50-397.

3.8-14 Letter GO2-76-156, D. L. Renberger to W. R. Butler, entitled WPPSS Nuclear Project No. 2, Dr well/Wetwell Leaka e Stud , transmitting response to request for a itional information in reference 3.8.3-10, dated April 23, 1976, Docket 50-397.

3.8-15 Deleted. Replaced with 3.8-23.

3. 8-16 Savin, G. N., Stress Distribution Around Holes, Translation of "Raspredeleniye Napryazheniy Okolo 0 tvers tiy", "N aukov a Dumk a" Press, 1968, National Aeronautics and Space Administration, NASA TT F-607, Washington, D. C., November 1970.

3.8-17 Roark, R. J., Formulas for Stress and Strain, Fourth Edition, McGraw-Hill Book Company, Inc., New York, 1965.

3. 8-18 Abbett, R. W., American Civil En ineerin Practicei John Wiley and Sons, Inc., New York, 1956.

3. 8-175

WNP-2 AMENDMENT NO. 13 February 1981 3.8-19 Shannon and Wilson, Inc., Soil Com action Evaluation-of Qualit Class I Backfill, Washington Public Power Supp y System, WPPSS Nuclear Project No. 2 (WNP-2).

3.8-20 Deleted. Replaced with 3.8-23.

I 3.8-21 "Primary Containment Vessel for Washington Public Power Supply System, Hanford No. 2, Jet Impingement Analysis," FIRL Technical Report F-C14121, May 21, 1975.

3.8-22 "HYBOS,"'IRL Users Manual, July 1973.

3. 8-23 Plant Design Assessment Report for SRV and LOCA Loads, Revision 2, Washington Public Power Supply System, August '1979, transmitted to NRC as Amendment No. 6 to the FSAR, September 19, 1979.

3.8-24 En ineerinn Evaluation of the Sacrificial Shield Wall, submitted to the NRC on WPPSS letter GOGe. 80 172'ugust 8, 1980.

3.8-25 'n ineerin Evaluation of the Sacrificial Shield Wall, Su lement No. 1, submitted to NRC on WPPSS letter G , August 19, 1980.

3. 8-176

Transient . Catecaor ~Ccles and return to normal operating temperature of 546oF at a rate of 100oF/hr. Pressure change from 1000 psig to 1180 psig to 240 psig and return to normal operating of 1000 psig.

Paragraph NB3552 of ASME III code excludes various for combining those'which trans-are not.

ients and provides means excluded. Review and approval of the equipment supplier's certified calculation provides assurance of proper accounting of the specified transients.

3.9.1.1.10 Recirculation Flow Control Valve Transients The following pressure and temperature transients were con-sidered in the design of the recirculation system flow con-trol valve:

Transient, CatecaoCr ~Ccles

a. Startup (100 F/hr normal/upset 300 heating rate 70 F to design temperature

b. Small temperature normal/upset'00 changes (29 F)

c. 50 F step changes normal/upset. 200

d. Safety/relief valve normal/upset. 30 blowdowns (single valve) (546 F to 375oF in 10 minutes)

e. Safety valve tran- normal/upset sient (110% of de-sign pressure)

3. 9-11

Transient CatecaCor r ~Colas

f. Installed hydrotests

a. 1300 psig testing 130

b. 1670 psig testing

g. Automatic blowdown 'mergency (546 F to 281 F in 15 seconds)

h. Improper start of emergency pump in cold loop (130 F step to 546oF for 15 ~

seconds) 3.9.1.1.11 Recirculation Pump Transients The following transients are listed in the design specifica-tion as a requirement for design considerations. However, a submitted certified analysis considering thermal stresses was not required. The vendor was required to submit a certification, of compliance. The submitted certified design calculations only considered pressure transient. Nozzle piping loads were considered in accordance with the following paragraph:

"The pump case shall be designed to withstand secondary stresses due to piping reactions in accordance with Paragraph 452.4b of the ASME Standard Code for Pumps and Valves for Nuclear Power (1968 Draft)."

Transients Ca~eqaor ~colas

a. Heatup and cooldown normal/upset 300 at 100oF/hr

b. +29 F temperature normal/upset 600 changes 3.9-12

3.

9.7 REFERENCES

3. 9-1 "Design and Performance of,G.E. BWR Jet Pumps,"

General Electric Company, Atomic Power Equipment Department,'PED-5460, July 1968.

3. 9-2 Moen, R. H., "Testing of Improved Jet Pumps for the BWR/6 Nuclear System,", General Electric Company, Atomic Power Equipment Department, NED0-10602, June 1972.

3. 9-3 General ELectric Company, "Analytical Model for Loss of Coolant Analysis in Accordance with 10CFR50 Appendix K," Proprietary Document, General Electric Company, NEDE-20566.

3.9-4 "BWR Fuel Channel Mechanical Design and Deflection,"

NEDE-21354-P, September 1976.

3. 9-5 "BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings," NEDE-21175-P, November 1976.

3. 9-6 "Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants," NEDE-24057-P (Class III) and NEDO-24057 (Class I), Octo er 77.

3.9-91

WNP-2 AMENDMENT NO. 4 June 1979 TABLE 3.9-1 PLANT EVENTS No. of

~Cclee Normal, U 'set, and Testin Conditions

a. Bolt Up* 123

b. Design Hydrostatic Test 130

c. Startup (100 F/hr Heatup Rate)** 120

d. Daily Reduction to 75% Power* 10,000

e. Weekly Reduction to 50% Power* 2,000

f. Control Rod Pattern Change* 400

g. Loss of Feedwater Heaters (80 Cycles Total): 80

h. Operating Base Earthquake'Event at Rated Operating Conditions

i. Scram:

1) Turbine Generator Trip, Feedwater on, Isolation Valves Stay Open 40

2) Other Scrams 140

3) Loss of Feedwater Pumps, Isolation Valves Closed 10

4) Single Safety or Relief Valve Blowdown Reduction to 0% Power, Hot Standby, Shutdown (100 F/hr Cooldown Rate)**

3. 9-92

NNP-2 AMENDMENT NO. 13 February 1981 discharge line to prevent drawing an excessive amount, of water up into the line as a result of steam condensation following termination of relief operation. Each. vacuum relief valve pair is situated with the valves in parallel, the discharge being routed to a common tee in the safety/relief valve discharge line. The safety/relief valves are located on the main steam line piping, rather than on the reactor vessel top head, primarily to simplify the discharge piping to the pool and to avoid the necessity of having to remove sections of this piping when the reactor head is removed for refueling.

In addition, valves located on the steam lines are more accessible during a shutdown for valve maintenance.

The nuclear pressure relief system automatically depressurizes the nuclear system sufficiently to permit the LPCI and, LPCS systems to operate as a backup for the high pressure core spray (HPCS) system. Further descriptions of the'peration of the automatic depressurization feature are found in 6.3, "Emergency Core Cooling Systems," and in 7.3.1.1.1, "Emergency Core Cooling Systems (ECCS) Instrumentation and Controls."

5.2.2.4.2 Design Parameters Table 5.2-3 lists design temperature, pressure, and maximum test pressure for the RCPB components. The specified operating transients for components within the RCPB are given in 3.9. Refer to 3.7 for discussion of the input criteria for design of Seismic Category I structures, systems, and components.

A summary of the number of cycles for transients used in design and fatique analysis is listed in Table 5.2-11 and categorized under the appropriate design conditions (i.e.,

normal, upset, emergency, and faulted).

The design requirements established to protect the principal components of the reactor coolant system against environmental effects are discussed in 3.11.

5.2.2.4.2.1 Safety/Relief Valve The discharge area of the valve is 16.117 square inches and the coefficient of discharge KD is equal to 0.966.

The design pressure and temperature of the valve inlet and outlet are 1250 psig 9 575'F and 550 psig 9 550'F, respectively.

5.2-13

The valves have been designed to achieve the maximum prac-tical number of acutations consistent with state-of-the-art technology. Cyclic testing has demonstrated that the valves are capable of at least 60 actuation cycles between required maintenance.

See Figure 5.2-l0 for a schematic cross section of the valve.

5.2.2.5 Mounting of Pressure Relief Devices The pressure relief devices are located on the main steam piping, header. The mounting consists of a special, contour nozzle and an over-sized flange connection. This provides a high integrity connection that accounts for the thrust, bending and torsional loadings which the main steam pipe and relief'valve discharge pipe are subjected to. This includes:

a. The thermal expansion effects of the connecting piping

b. The dynamic effects of the piping due to SSE c.'he reactions due to transient unbalanced wave forces exerted on the safety/relief valves during the first few seconds after the valve is opened and prior to the time steady-state flow.has been established.

d. The dynamic effects of the piping and branch connection due to the turbine stop valve closure.

In no case will allowable valve flange loads be exceeded nor will the stress at any point in the piping exceed code allowables for any specified combination of loads. The de-sign criteria and analysis methods for considering loads due to SRV discharge is contained in 3.9.3.3.

5.2.2.6. Applicable Codes and Classification The vessel'verpressure protection system is designed to satisfy the requirements of Section III, Nuclear Vessels, of the ASME Boiler and Pressure Vessel Code. The general requirements for protection against, overpressure as given in Article 9 of Section III of the Code recognize that reactor vessel overpressure protection is one function of the reactor protective systems and allows the integration of pressure relief devices with the protective systems of the nuclear reactor. Hence, credit is taken for the scram pro-tective system as a complementary pressure protection device.

Since there are no additives in the BWR coolant, leakage would expose materials to high purity, demineralized water.

Exposure to demineralized water would cause no detrimental effects.

5.2.3.3 Fabrication and Processing of Ferritic Materials 5.2.3.3.1 Fracture Toughness Fracture toughness requirements for the ferritic materials used for piping and valves (no ferritic pumps in RCPB) of the reactor coolant pressure boundary were as follows:

Safety/Relief Valves were exempted from fracture toughness requirements because Section III of the 1971 ASME Boiler and Pressure Vessel Code did not require impact testing on valves with inlet connections of 6 inches or less nominal pipe size.

Main Steam Xsolation Valves were also exempted because the Code existing at the time of the purchase order, April 1971/

did not require brittle fracture testing on ferritic pressure boundary components when the system temperature was in excess of 250oF at 20% of the design pressure.

Main Steam Piping was tested in accordance with and met the fracture toughness requirements of paragraph NB-2300 of the 1972 Summer Addenda to ASME Code,Section XII, the applicable code at the time of the purchase order, September 1972.

5.2.3.3.1.1 Compliance with Code Requirements The ferritic pressure boundary material of the reactor pres-sure vessel was qualified by impact testing in accordance with the 1971 Edition of Section XXI ASME Code and Addenda to and including the Summer 1971 Addenda. From an opera-tional standpoint, this Code would require that for any significant pressurization (taken to be more than 20% of Code hydrostatic test pressure = 312 psig) the minimum metal tem-perature of all vessel shell and head material be 100oF (NDTT +600F).

5.2-25

5.2.3.3.2 Control of Welding 5.2.3.3.2.1 Control of Preheat Temperature Employed for Welding of Low Alloy Steel Regulatory Guide 1.50 The use of low alloy steel is restricted to the reactor pressure vessel. Other ferritic components in the reactor coolant pressure boundary are fabricated from carbon steel materials.

Preheat temperatures employed for welding of low alloy steel meet or exceed the recommendations of ASME Section III, Subsection NA., Components were either held .for an extended time at preheat temperature to assure removal of hydrogen, or preheat was maintained until post weld heat treatment.

The minimum preheat and maximum interpass temperatures were specified and monitored.

All welds were nondestructively examined by radiographic methods. In addition, a supplemental ultrasonic examination was performed.

5.2.3.3.2.2 Control of Stainless Steel Weld Cladding of Low Alloy Steel Components Regulatory Guide 1.43 Regulatory Guide 1.43 does not apply to BWR components.

5.2.3.3.2.3 Control of Electroslag Weld Properties Regulatory Guide 1.34 No electroslag welding was performed on BWR components.

5.2.3.3.2.4 Welder Qualification for Areas of Limited Accessibility Regulatory Guide 1.71 There are few restrictive welds involved in the fabrication of BWR pressure boundary components. Welder qualification for welds with the most restricted access was accomplished by mock-up welding. Mock-ups were examined with radiography or sectioning.

5.2-26

TABLE 5.2-11 RCPB OPERATING THERMAL CYCLES No. of Events Normal, U set, and Testin Conditions Bolt Up* 123 Design Hydrostatic Test 130 Startup (100oF/hr Heatup Rate)** 120 Daily Reduction to 75-Percent Power* and Control Rod Pattern Change* 10,400 Weekly Reduction to 50-Percent Power* 2,000 Feedwater Heater Loss, Partial Heater Bypass 70 HPCS Operation (10), SLC Operation (10) 20 Scram:

Turbine Generator Trip, Feedwater On, Isolation Valves Stay Open 40 Other Scrams 140 Loss of Feedwater Pumps, Isolation Valves Closed 10 Reduction to 0-Percent Power, Hot Standby, Shutdown Unbolt (100oF/hr Cooldown Rate) ~* ill (each) 123 mer enc Conditions Scram:

Reactor Overpressure with Delay Scram, Feedwater Stays On, Isolation Valves Stay Open Automatic Blowdown Single Safety/Relief Valve Blowdown Improper Start of Cold Recirculation Loop Sudden Start of Pump in Cold Recirculation Loop Improper Startup with Reactor Drain Shut Off Followed by Turbine Roll and Increase to Rated Power Faulted Conditions Pipe Rupture and Blowdown ASME H drostatic Test 1.25 x Design Pressure Hydrostatic Test 10 (per NB 6222 and NB 3114)

• Applied to reactor pressure vessel only

• • Bulk average vessel coolant temperature change in any 1-hour period 5.2-59

WNP-2 AMENDMENT NO ~ 7 November 1979 TABLE 5 '-12 DELETED The itemized examinations to be conducted in accor-dance with the commitments herein are delineated in the WNP-2 Preservice Inspection Program Plan (Ref.

5. 2-6) for all preservice inspection commitments, and will subsequently bePlan delineated in the WNP-2 Inservice applicable to each inspection Inspection Program interval.

5 '-60

WNP-2 AMENDMENT NO. 4 June 1979 TABLE 5.4-3 (Continued)

Active/ Reference Location Inactive Valve No. Fi ure Flow Con- Inactive B35F060 5.4-2b trol Pump Dis- Inactive B35F067 5.4-2b charge RCIC Vessel Head In Active E51F066 5.4'-9a Active E51F065 5.4-9a Active E51F013 5.4-9a HPCS In Active E22F005 6.3-1 Active E22F004 6.3-1 Inactive E22F038 6.3-1 LPCS In Active E21F006 6. 3-5 Active E21F005 6. 3-5 Inactive E21F051 6. 3-5 Standby Active SLC-V-7 9.3-13 Liquid Active SLC-V-4 9.3-13 Control In Active SLC-V-6 9.3-13 Inactive SLC-V-8 9.3-13 Pum Descri tion Recircu- Active B35C001 5.4-2b lation Pump 5.4-59

WNP-, 2 AMENDMENT NO. 9 April 1980 TABLE 5. 4-4 SAFETY AND RELIEF VALVE DESCRIPTION Safety and/or Relief Valve Identification Descri tion B22F013 Main Steam Line Safety Relief Valve E51F017 RCIC System Suction line E51F018 RCIC Lube Oil Cooler Supply line E51F033 RCIC Vacuum Tank E12F055 RHR Condensing Mode Steam Supply line RHR-RV-9 5 RHR Condensing Mode Steam Supply line E12F036 RHR Condensing Mode Return line to RCIC E12F005 Shutdown Cooling Supply line E12F025 Shutdown Cooling Return line E12F088 Suppression Pool Supply for Loop C E12F030 RHR Flush line RHR-RV-1 RHR Heat Exchanger (Shell side)

RWCU-RV-1 RWCU Regenerative Heat Exchanger (Shell side)

RWCU- RV- 2 RWCU Non-Regenerative Heat Exchanger (Tube side)

RWCU- RV- 3 RWCU Regenerative Heat Exchanger (Tube side)

G33F036 RWCU Blowdown to Radwaste System or Condenser RWCU- RV- 2 62 RWCU Filter Demineralizer Influent E22F014 High Pressure Core Spray suction line E22F035 High Pressure Core Spray Pump discharge line E21F018 Low Pressure Core Spray Pump discharge 'line E21F031 Low Pressure Core Spray suction line C41F029 Standby Liquid, Control Pump discharge line RHR- RV- 9 8 RHR Shutdown Cooling Supply line

5. 4-60

TABLE OF CONTENTS (Continued)

Parte

6. 7. 4 INSTRUMENTATION APPLICATION 6.7-14 6.7.4.1 System Description 6.7-14 6.7.4 1.1 Main Steam Line Leakage Control System

~

(MSIV-LCS) 6. 7-14 6.7.4.1.1.1 MSIV-LCS Instrumentation and Controls 6.7-15 6.7.4.1.1.1.1 Power Sources 6.7-15 6.7.4.1.1.1.2 Equipment 6.7-15 6.7.4.1.1.1.2.1 General 6.7-15 6.7.4.1.1.1.2.2 Initiating Circuits 6.7-15 6.7.4.1.1.1.2.3 Logic and Sequencing 6.7-16 6.7.4.1.1.1.2.4 Interlocks 6.7-16 6.7.4.1.1.1.2.5 Redundancy 6.7-16 6.7.4.1.1.1.3 Specific Regulatory Requirements Conformance 6.7-16 6.7.4.1.1.1.3.1 Regulatory Guide 1.29, Revision 1 6.7-16 6.7.4.1.1.1.3.2 Regulatory Guide 1.75, Revision 0 6.7-17 6.7.4.1.1.1.3.3 Regulatory Guide 1.62, Revision 0 6.7-17 6.7.4.1.1.1.3.4 Regulatory Guide 1.96, Revision 1 6.7-17 6.7.4. 1.1.1.4 Conformance to General Functional Requirements 6.7-17 6.7.4.1.1.1.4.1 IEEE 279-1971 6.7-17 6.7.4.1.1.1.4.1.1 General Functional Requirements (IEEE-279 Para. 4.1) 6.7-17 6.7.4.1.1.1.4.1.2 Single Failure Criterion.(IEEE-279, Para. 4.2) 6.7-17 6-xiii

TABLE OF CONTENTS (Continued)

PacCe 6.7.4.1.1.1.4.1.3 Quality of Components and Modules (IEEE-279 Para. 4.3) 6.7-18 6.7.4.1.1.1.4.1.4 Equipment'Qualification (IEEE-.279 Para. 4.4) 6.7-18 6.7.4.1.1.1.4.1.5 Channel Integrity (IEEE-279 Para. 4.5) 6.7-18 6.7.4.1.1.1.4.1.6 'Channel Independence (IEEE-279 Para. 4.6) 6.7-18 6.7.4.1.1.1.4.1.7 Control and Protection System Interaction (IEEE-279 Para. 4.7) 6.7-18 6.7.4.1.1.1.4.1.8 Derivation of System Inputs (IEEE-279 Para. 4.8) 6.7-18 6.7.4.1.1.1.4.1.9 Capability of Sensor Checks (IEEE-279 Para. 4.9) 6.7-19 6.7.4.1.1.1.4.1.10 Capability of Test and Calibration (IEEE-279 Para. 4.10) 6.7-19 6.7.4.1.1.1.4.1.11 Channel Bypass or Removal From Operation (IEEE-,279 Para. 4.11) 6.7-19 6.7.4.1.1.1.4.1.12 Operating Bypasses (IEEE-279 Para. 4.12) 6.7-19 6.7.4.1.1.1.4 '.13 Indication of Bypasses (IEEE-279 6.7-19 Para. 4.13) 6.7.4.1.1.1.4.1.14 Access to Means For Bypassing (IEEE-279 Para. 4.14) 6.7-19 6.7.4.1.1.1.4.1.15 Multiple Set Points (IEEE-279 Para. 4.15) 6.7-20 6.7.4.1.1.1.4.1.16 Completion of Protection Action Once It Is4.16)

(IEEE-279 Para.

Initiated

6. 7-20 6.7.4.1.1.1.4.1.17 Manual Actuation (IEEE-279 Para. 4.17) 6.7-20 6-xiv

TABLE OF CONTENTS (Continued)

PacCe 6.7.4.1.1.1.4.1.18 Access to Set Point Adjustments, Calibration and Test Points (IEEE-279 Para. 4.18) 6.7-20 6.7.4.1.1.1.4.2 Not Used 6.7-20 6.7.4.1.1.1.4.3 IEEE-338, 1971 6.7-20 6.7.5 TESTS AND INSPECTIONS 6.7-20 6.7.5.1 Preoperational Tests 6.7-20 6.7.5.2 Operational Tests 6.7-21 6-xv

WNP-2 AMENDMENT NO. 12 November 1980 LIST OF TABLES NUMBER TITLE PAGE 6.1-1 Engineered Safety Features Systems'nd Related Systems Component Materials 6.1-7 6.2-1 Containment Design Parameters 6.2-89 6.2-2 Engineered Safety Systems Information for Containment Response Analyses 6.2-91 6.2-3 Accident Assumptions and Initial Conditions for Recirculation Line Breaks 6.2-95 6.2-4 Initial Conditions Employed in Containment Response Analyses 6.2-96 6.2-5 Summary of Accident Results for Containment Response to Limiting Line Breaks 6.2-98 6.2-6 Loss-of-Cool.ant Accident Long-Term Primary Containment Response Summary 6.2-100 6.2-7 Energy Balance for Design Basis Recirculation Line Break Accident 6.2-101

6. 2-8 Accident Chronology Design Basis Recircu-lation Line Break Accident 6. 2-103

6. 2-9 Reactor Blowdown Data for- Recirculation Line Bre ak 6. 2-104

6. 2-10 Reactor Blowdown Data for Main Steam Line Bre ak 6.2-105 6.2-11 Core Decay Heat Following LOCA for Contain-ment Analyses 6. 2-106 6.2-12 Secondary Containment Design and Performance Data 6. 2-107 6.2-13 Deleted See Table 6.2-16

6. 2-14 Containment Penetrations Subject to Type B Tests 6. 2-117

6. 2-15 Suppression Chamber/Reactor Building Differential Pressure Indicating Switch 6.2-118 Char ac ter is ties 6-xvi.

WNP-2 AMENDMENT NQ. 4 June 1979 I~

When vessel pressure reaches 200 psid* the system reaches rated core spray flow. The HPCS motor size is based on peak horsepower requirements.

The elevation of the HPCS pump is sufficiently below the water level of both the condensate storage tanks and the suppression pool to provide a flooded pump suction and to meet pump NPSH requirements with the containment, at atmos-pheric.pressure and the suction strainer 50% plugged. The available NPSH has been calculated in accordance with Reg-ulatory Guide 1.1, 1970, Rev. 0, and is 36'. I A motor-operated valve is provided in the suction line from the suppression pool. The valve is located as close to the suppression pool penetration as practical. This valve is used to isolate the suppression pool water source when HPCS system suction is from the condensate storage system and to isolate the system from the suppression pool in the event a leak develops in the HPCS System.

The HPCS pump characteristics, head, flow, horsepower, and required NPSH are shown in Figure 6.3-4.

The design pressure and temperature of the system components are established based on the ASME Section IXI Boiler and Pressure Vessel Code. The design pressures and temperatures, at, various points in the system can be obtained from the information blocks on the HPCS Process Diagram, Figure 6.3-2.

A check valve, flow element. and restricting orifice are pro-vided in the HPCS discharge line from the pump to the in-jection valve. The check valve is located below the minimum suppression pool water level and is provided so the piping downstream of the valve can be maintained full of water by the discharge line fill system (see 6.3.2.2.5) . The flow element is provided to measure system flow rate during LOCA and test conditions and for automatic control of the minimum low flow bypass gate valve. The measured flow is indicated in the main control room. The restricting orifice is sized during the pre-operational test of the system to limit system flow to acceptable values as described on the HPCS system Process Diagram.

• psid = differential pressure between the reactor vessel and the suction source.

6.3-11

WNP-2 A low flow bypass line with a motor-operated gate valve con-nects to the HPCS discharge line upstream of the check valve on the pump discharge line. The line bypasses water to the suppression pool to prevent pump damage from overheating when other discharge line valves are closed. The valve auto-matically closes when flow in the main discharge line is sufficient to provide required pump cooling.

C To assure continuous core cooling, primary containment iso-lation does not interfere with HPCS operation.

The HPCS system incorporates relief valves to protect the components and piping from inadvertent overpressure con-ditions. One relief valve, set to',relieve at 1100 psig with a capacity of 25 gpm is located on the discharge side of the pump downstream of the check valve to relieve thermally expanded fluid. A second relief valve is located on the suction side of the pump and's set at 100 psig with a ca-pacity of 10 gpm.

The HPCS components and piping are positioned to avoid damage from the physical effects of design basis accidents, such as pipe whip, missiles, high temperature, pressure, and humidity.

The HPCS equipment and support structures are designed in accordance with Seismic Category I criteria (see 3.2 .1) .

The system is assumed to be filled with water for seismic analysis.

Provisions are included in the HPCS system which will permit the HPCS system to be tested. These provisions are:

a. All active HPCS components are testable during normal plant operation and/or during shutdown downs~

as discussed in 6.3.1.1.2m.

b. A full flow test line is provided to route and to the condensate storage tanks without water'rom entering the reactor pressure vessel. The suction line from the condensate tanks also pro--

vides reactor grade water to fully test the HPCS including injection into the RPV during shut-

c. A full flow .test line is provided to route water from and to the suppression pool without entering the reactor pressure vessel.

6. 3-12

X O K I 3 A K 2 O O BHP Z

m O Y m

1 O tn

~

g

~ M I t/)

m 70 EFF ICIENCY X

I-g O

50 NPSHR Nozzle I C.L. Suet'on Ill z UJ LL O z

u. 40 HEAD Z

Ch 30 I-0 I-C/) z 20 20 GPM NPSH z 0 36 O 10 '7000 34 10 Ch 1000 2000 5000 8000 GALLONS PER MINUTE

~mKl

NNP-2

i. The MSIV-LCS is designed to permit testing of the operability of controls and actuating de-vices during power operation to the extent practical, and testing of the complete func-tioning of the system during plant shutdowns.

The MSIV-LCS is designed so that effects re-sulting from a sealing system single active component failure will not affect the integrity of the main steam lines or MSXV's.

6.7.1.3 Codes and Standards The detailed design and construction criteria are provided by published codes, standards andregulatory guides. All piping systems and components for the MSXV-LCS comply with the applicable codes, addenda, code cases and errata in effect at the time the equipment is procured. Currently in effect is the:

ASME Boiler and Pressure Vessel Code Section XII, Nuclear Power Plant Components. The piping and components at the point of connection to the main steam line including the reactor pressure retaining system valves are Class 1. All other piping and components are Class 2. Subsections NA, NB, and NC of the Code'pply to the MSIV-LCS.

The equipment and piping of the MSIV-LCS, in order to meet specified seismic capabilities, are designed to Seismic Category I requirements. This category includes all struc-tures,and equipment essential to the safe shutdown and isola-tion .of the reactor, or whose failure or damage could result in undue risk to the health and safety of the public.

Refer to Table 5.2-4, Reactor Coolant Pressure Boundary Materials for a presentation of the specifications which generally apply to the selection of materials used in the MSIV-LCS'. Nonmetallic materials such as lubricants, seals, packings, paints and primers, insulation, metallic materials, etc., are selected as a result of an engineering review and evaluation for compatibility with other materials in the system and the surroundings. with concern for chemical, radiolytic, mechanical and nuclear effects.

The MSIV-LCS is designed to'e in accordance with XEEE Std; 279-1971 (Criteria for Protection Systems for Nuclear Power 6.7-3

WNP-2 AMENDMENT NO. 8 I February 1980 Generating Stations) and IEEE 344-1971 (Guide for Seismic Qualification of Class I Flectric Equipment for Nuclear Power Generating Stations.)

6.7 ' .SYSTEM DESCRIPTION 6.7.2.1 General Description The MSIV-LCS is designed to minimize the release of fission products which could by-pass the standby gas treatment system (SGTS) after a LOCA. This is accomplished by directing leakage from the closed main steam isolation valves (MSIV's) through a bleed line and into an area served by SGTS. The flow is effected by a small blower which maintains the pres-sure in the steam lines negative with respect to atmosphere, thus ensuring that the MSIV leakage will pass through the blower and on into the SGTS prior to release to the atmosphere.

The flow diagram of the MSIV-LCS is shown in Figure 3.2-25.

As indicated, two independent systems (an outboard system and an inboard system) are provided to accomplish the leakage control function. The inboard system receives power from one division and the outboard system from the other division of the two redundant critical electrical power supply divisions.

The outboard system is connected to the segments of the main steam lines between fast closing MSIV's outside containment and the downstream block valves. The bleed line from each main steam line connects to a bleed header. The bleed header outlet is provided with two valves in, series to permit the main steam lines to be depressurized by venting, following a LOCA. A parallel set of valves is provided which are opened following depressurization to connect. the blower sunction to the steam lines. Pressure sensors are also used for depres-surization interlock control to prevent, any accidental actu-ation of the -system during normal reactor operation. Another pressure sensor is used for interlock control on the valves in the blower suction line to prevent. accidental actuation when pressure is appreciably greater than atmospheric. Pressure indicators are provided for monitoring the pressure in the main steam lines between the fast closing MSIV's outside con-tainment and the downstream block valves. The major flow blower suction is dilution air from the reactor building. to'he This dilution air greatly reduces the temperature of the MSIV leakage as it. passes through the blower.

6. 7-4

Malfunction Outboard MSIV 1 of 4 fails to Outboard LCS system func-close (excessive tions to collect leakage leakage) through 3 sets of 2 MSIV's in series and one inboard MSIV, and deliver it to the SGTS for treatment.

Inboard LCS is available to collect leakage from inboard MSIV's and deliver it to the SGTS for treatment.

Downstream Fails to close Outboard LCS functions but Block Valves (excessive leak- capacity is insufficient age) to positively control MSIV leakage. Inboard LCS functions to collect leak-age through inboard MSIV and deliver it to the SGTS for treatment.

Inboard LCS Fails to operate, Inboard LCS inoperative.

blower or loss of power Outboard LCS powered by Inboard LCS other electrical division emergency power functions to collect leak-Inboard LCS age through 4 sets of 2 initiation MSIV's in series and de-switch ment.

it liver to SGTS for treat-Outboard LCS Fails to operate, Outboard LCS inoperative.

blower or loss of power Inboard LCS powered by the Outboard LCS other electrical division power supply operates to collect leak-Outboard LCS age through 4 inboard initiation MSIV's and deliver it to switch the SGTS for treatment.

Outboard LCS bleed valve 6.7-13

Com onent Malfunction Inboard LCS Fails to open Inboard LCS collects leak-Bleed valve age from 3 of 4 inboard MSIV's. Outboard ZCS func-tions to collect leakage through 4 of 4 sets of 2 MSIV's in series.

Inboard LCS Fails to open, One main steam line between Depressuri- MSIV's will not depressurize zation Valve in time giving a false in-board MSIV excessive leak-age signal, and resulting in isolation of of in-

>4 board LCS. 3/4 of inboard LCS continues to function.

A Outboard LCS Fails to close Gas from the reactor build-Depressuri- ing volume is recirculated zation Valve 'hrough blower, decreasing its effective capacity.

The LCS continues to func-tion inboard.

Inboard LCS Fails to heat Leakage relative humidity heater will be high, condensation may plug pipe and decrease inboard LCS capacity. Out-board system continues to function.

6.7.4 INSTRUMENTATION APPLICATION Refer to Figure 3.2-25 for instrumentation and control infor-mation.

6.7.4.1 System Description 6.7.4.1.1 Main Steam Line Leakage Control System (MSIV-LCS}

The MSIV-LCS is manually initiated after a loss-of-cool-ant accident (LOCA) . The purpose of this system is to control and minimize the release of fission products which leak through closed main steam line isolation valves by directing leakage to the standby gas treatment system for processing prior to release to the'atmosphere.

6.7-14

The MSIV-LCS is divided into two independent subsystems.

The inboard (see Figure 3.2-25) subsystem is connected to the steam lines outside the primary containment between the in-board and outboard MSIV's. The outboard subsystem connec-tion is downstream of the outboard MSIV's.

6.7.4.1.1.1 MSIV-LCS Instrumentation and Controls 6.7.4.1.1.1.1 Power Sources The instrumentation and control of the MSIV-LCS are powered by 120 volts a-c Division 1 (inboard subsystem) and Division 2 (outboard subsystem) critical power supplies.

6.7.4.1.1.1.2 Equipment 6.7.4.1.1.1.2.1 General The instrumentation components for the MSIV-LCS are located outside the steam tunnel. Cables connect the sensors and transducers to control circuitry within the logic panel.

Functional test, on the system instrumentation can be performed during normal reactor power operation. However, the MSIV-LCS isolation valves can only be tested one at a time. Inboard and outboar'd subsystem control and instrumentation are electric-ally and mechanically separated to assure that no single fail-ure event can disable the MSIV-LCS. The MSIV-LCS is designed to operate from normal off-site auxiliary power sources or from divisional standby power supplies not available.

if off-site power is 6.7.4.1.1.1.2.2 Initiating Circuits The MSIV-LCS can be actuated manually after a LOCA has occurred, provided that the reactor and steamline pressure are below the pressure permissive interlock set points. The outboard sub-system is provided with one remote manual initiating switch, and the inboard subsystem is provided with one remote manual initiating switch.

The inboard subsystem has individually controlled process lines provided for each steam line.

When the inboard subsystem is initiated, the bleed valves and bypass valve open simultaneously, and the exhaust blower and the heater are energized. After one minute, if the steamline pressure is greater than 5 psig, the bleed valves will close.

6. 7-15

If the pressure main .open.

is 5 psig or less, the bleed valves will re-After another minute, the bypass valve is closed.

The flow is thus, routed .through the flow element. .After another 4 minute, the third timer closes the bleed valves if high leakage flow is present.

The outboard subsystem process lines,'from each main steam-line, are connected to a header leading to the depressuri-zation and bleed off branch.

When the outboard subsystem is initiated, depressurization valves open and the exhaust. blower is activated. When the steam lines have depressurized to approximately atmos-pheric pressure, the depressurization branch valves are closed and flow is diverted to the blower suction lines.

6.7.4.1.1.1.2.3 Logic and Sequencing A LOCA is indicated by high drywell pressure and low, low vessel water level. After a LOCA has occurred, the MSIV-LCS system can be manually initiated.

Indicators for both reactor and steamline pressures for the inboard and outboard subsystems are available in the main control room.

6.7.4.1.1.1.2.4 Interlocks Both the inboard and outboard subsystems are provided with reactor and steamline pressure interlocks to prevent inadver-tent system initiation during normal reactor power operation.

The inboard subsystem is in additio'n, provided with an inter-lock to prevent bleed valves from opening unless the respec-tive inboard MSIV is closed.

6.7.4.1.1.3..2.5 Redundancy The MSIV-LCS consists of two independent redundant subsystems; inboard and outboard. Either system may be manually initiated a LOCA. 'ollowing 6.7.4.1.1.1.3 Specific Regulatory Requirements Conformance 6.7.4.1.1.1.3.1 Regulatory Guide 1.29, Revision 1 All instrumentation and controls are tested and qualified 'to meet Seismic Category I requirements and remain functional after a ll seismic event.

6.7-16

6.7.4.1.1.1.3.2 Regulatory Guide 1.75, Revision 0 The instrumentation and control devices for each subsystem are completely separated and independent. The system raceway groupings. comply with the requirements of this regulatory guide. 'ach subsystem has a separate and independent control room panel.

6.7.4.1.1.1.3.3 Regulatory Guide 1.62, Revision 0 System initiation is manual from the control room. Interlocks are provided to prevent inadvertent manual initiation during normal reactor power operation.

6.7.4.1.1.1.3.4 Regulatory Guide 1.96, Revision 1 The MSIV-LCS is designed to comply with this regulatory guide, with the exception that the MSIV-LCS is not designed to reduce and control stem packing leakage or other direct leakage to the steam tunnel. Leakage from valve stem packing is addressed in. 6.7.3m., while other direct leakage to the steam tunnel is addressed in 6.7.3.1.

6.7.4.1.1.1.4 Conformance to General Functional Requirements 6.7.4.1.1.1.4.1 IEEE 279-1971 6.7.4.1.1.1.4.1.1 General Functional Requirements (IEEE-279 Para. 4.1)

After a LOCA, the MSIV-LCS can be manually initiated from the control room by plant personnel. After initiation, the system is shut down by high steam line or reactor pressure to prevent excessive leakage to the SGTS.

6.7.4.'1.1.1.4.1.2 Single Failure Criterion (IEEE-279 Para. 4.2)

The MSIV-LCS consist of two subsystem, inboard and outboard.

The two subsystems feature separate and independent sets of controls and instrumentation and meet the single failure criterion.

6.7-17

6.7.4.1.1.1.4.1.3 Quality of Components and Modules (IEEE-279 Para. 4.3)

Components used in MSIV-LCS have been carefully selected on the basis of suitability for the specific application. The=

logic relays have been selected to ensure against significant deterioration during anticipated duty over the lifetime of the plant.

A quality control and assurance program is required to be implemented and documented by equipment vendors with the in-tent of complying with requirements set forth in 10CFR50 Appendix B.

6.7.4.1.1.1.4.1.4 Equipment Qualification (IEEE-279 Para 4.4)

Vendor certification is required to ensure that components operate in accordance with the requirements of the purchase specification.

6.7.4.1.1.1.4.1.5 Channel Integrity (IEEE-279 Para 4.5)

The MSIV-LCS is required to be operable under the envi'ron-mental conditions noted in the design basis (6.7.1).

6.7.4.1.1.1.4.1.6 Channel Independence (IEEE-279 Para. 4.6)

Channel independence for sensors is provided by electrical and mechanical separation. Physical separation is maintained between inboard and outboard. subsystem to increase reliability of operation. The MSIV-LCS is sufficiently separated to give a high degree of reliability.

6.7.4.1.1.1.4.1.7 Control and Protection System Interaction (IEEE-279 Para. 4.7)

There is no control and protection system interaction.

6.7.4.1.1.1.4.1.8 Derivation of System Inputs (IEEE-279 Para. 4.8)

All input signals to the instrumentation and control systems are derived from direct measurement of system variables.

6.7-18

6.7.4.1.1.1.4.1.9 Capability of Sensor Checks (IEEE-279 Para. 4.9)

The sensors which are used for inputs to the MSIV-LCS can be checked one at a,time by application of simulated signals during normal plant operation.

6.7.4.1.1.1.4.1.10 Capability of Test and Calibration (IEEE-279 Para. 4.10)

All active components of the MSIV-LCS can be tested during plant operation. Valves can be tested by operating manual switches in the control room and observing indicating lights.

Operation of the blowers and heaters by manual switches can. be verified by dilution air flow and heater temperature indicators.

6.7.4.1.1.1.4.1.11 Channel Bypass or Removal from Operation (IEEE-279 Para. 4.11) k This system is manually initiated after a LOCA, and allows calibration and removal of one subsystem which will not affect the other subsystem.

6.7.4.1.1.1.4.1.12 Operating Bypasses (IEEE-279 Para. 4.12)

During system operation, the operating control system can be shutdown automatically by high pressure or high flow inter-locks in the inboard subsystem. Otherwise, the trip is manual.

6.7.4.1.1.1.4.1.13 Indication of Bypasses (IEEE-279 Para. 4.13)

Motor operated valve closure can be prevented by shutting off electric power to the motor starters. This action will be annunciated in the control room to indicate power failure.

6.7.4.1.1.1.4.1.14 Access to Means for Bypassing (IEEE-279 Para. 4.14)

Access,to switchgear and motor control centers is procedurally controlled by the following administrative means:

a. Lockable doors on the critical switchgear rooms

b. Lockable breaker control switch handles in the motor control centers 6.7-19

WNP-2 6.7.4.1.1.1.4.1.15 Multiple Set Points (IEEE-279 Para 4.15)

All set points are fixed.

6.7.4.1.1.1.4.1.16 Completion of Protection Action Once it is Initiated (IEEE-279 Para. 4.16)

The MSIV-LCS will not remain in operation after system if initiation excessive leakage, high reactor pressure, or high steamline pressure occurs.

6.7.4.1.1.1.4.1.17 Manual Actuation (IEEE-279 Para.. 4.17)

The MSIV-LCS can be initiated manually, at system level, from the control room provided interlocks permit.

6.7.4.1.1.1.4.1.18 Access to Set Point Adjustments, Calibra-tion and Test Points (IEEE-279 Para. 4.18)

The control system and its respective set points are accessible on the main control room panels. The sensors are accessible on the instrument racks during normal plant operation.

6.7.4.1.1.1.4.2 Not Used 6.7.4.1.1.1.4.3 IEEE-338, 1971 The MSIV-LCS is testable during reactor operation. The test completely tests each logic through the final actuators and demonstrates independence of subsystems. A failure of one subsystem while testing will not prevent the other subsystem from being initiated manually.

6.7.5 TESTS AND INSPECTIONS 6.7.5.1 . Preoperational Tests Preoperational tests are conducted prior to initial startup.

The tests ensure functioning of all controls, instrumentation and all active .components. Functional testing and flow measurements are accomplished with compressed air. Compressed air testing is conservative compared to operation of the sys-tem with steam blowdown under post accident conditions. For the hot steam blowdown case, compared to the preoperational test case with 70 F air, the space between MSIV's will depres-surize faster and the 5 psig trip setting has additional margin against reclosure. System reference characteristics such as timer setpoints and flow rates are documented during the pre-operational testing and are used as base points for measure-ments obtained in subsequent operational tests.

6.7-20

TABLE 7 3-27 MAIN STEAM LINE LEAKAGE CONTROL SYSTEM INSTRUMENTATION SPECIFICATIONS INSTRUMENT REQUIRED RESPONSE FUNCTION INSTRUMENT RANGE (1) TRIP SETTING (2) MARGIN (3) ACCURACY (4) TIME (4)

Reactor Press. Switch 35 psig Pressure (MSLC-PS20)

Low (MSLC-PS24)

(MSLC-PS&A-D)

(MSLC-FS7A-D)

MSLC Header Press. Switch 0 psig Pressure Low (MSLC-PS 25)

MSLC Header Press. Switch 5 psig Pressure Low (MSLC-PS 70A-D)

MSLC High Flow Switch 505 CFH Flow (MSLC-FS 3A-D) 8 <g PP 0 G)Q th OR 0

WNP-2 AMENDMENT NO. 1 0 July 1980 Page 2 of 2 NOTES FOR TABLE 7.3-27 (1) See Chapter 16, "Technical Specifications" for opera-tional limits.

The range for safety-related instrumentation is selected so as to exceed the expected range of the process variable being monitored.

(2) Trip settings shown are subject to change to comply with Chapter 16, "Technical Specifications".

The trip setpoint is located in that portion of an instrument's range which provides the required accuracy.

Initial trip setting values are established from operating experience with similar size plants, and back up with analysis as necessary.

(3) See Chapter 16, "Technical Specifications" for instrument setpoint margins.

(4) Values shown are subject to change to comply with Chapter 16, "Technical Specifications".

7.3-78

Tray/Conduit Inscription Marking Character'olors Background Division Characters Colors Power, Control Div. 1 Black Yellow

& Instrumentation Power, Control Div. 2 Black Orange

& Instrumentation Power, Control Div. 3 Black

& Instrumentation RPS Channel Al R Ch Al Red Lt. Blue RPS Channel A2 R Ch A2 Red Green RPS Channel Bl R Ch Bl Drk. Blue RPS Channel B2 R Ch B2 Brown Non-Class 1E equipment and cables have nameplates or tags with color coding as indicated below. Identification markers are of sufficient durability and at a sufficient number of points to facilitate initial verification that the installation is in conformance with the separation criteria. These cable tray and conduit markings are applied prior to cable installation.

Tray/Conduit Inscription Marking Character Background Division Application Characters Colors Colors Power, Control Div. A Black Yellow

& Instrumentation Power, Control Div. B Black Gold

& Instrumentation

Non-Class lE cables that are routed in Class lE raceways with Class lE cables become "Associated Circuits." They are routed in accordance with the separation criteria defined in 8.3.1.4.

Cable marking for associated circuits consists of a black inscription (cable number) on a composite background of horizontal bands of the background colors of both divisions.

For example, a background consisting of yellow and silver bands indicates that a non-Class 1E Division A cable is run in Division 1 cable tray or conduit, somewhere along its routing. Orange and gold bands indicate Division B cable in Division 2 tray or conduit.

Information concerning routing is given in Table 8.3-8 to illustrate the computer program used for cable identification and routing. Table 8.3-9 indicates sample cable routing schedules. Actual cable tray drawings for the reactor, control and radwaste buildings are shown in Figures 8.3 to 8.3-14 inclusive.

8.3.1.4 Independence of Electrical Divisions Refer to 7.0 for electrical separation within the PGCC.

Separation of other electrical equipment and wiring is such that=no single event will prevent any of the necessary safety functions from being performed.

Physical separation as a protection against common mode failure of critical power and instrumentation systems (Div-ision 1 to 7) is achieved by spatial separation and/or physical barriers between the equipment and raceways serving different divisions. Spatial separation is preferred and is utilized wherever possible.

Separate cable trays and conduit systems are provided for each of the standby ac power systems (Divisions 1, 2 and 3).

Reactor protection systems (RPS Divisions'4, 5, 6 and 7) are run in totally enclosed raceway systems, except for the area immediately underneath the reactor.

The physical independence of electrical systems complies with the requirements of IEEE Std. 279-1971, IEEE Std. 308-1974 (IEEE Std. 308-1972 for HPCS system), General Design Criter-ion 17 and 21 and NRC Regulatory Guide 1.6, Revision 0.

8.3-56

TABLE 8. 3-13.

Motor 'and Mo't'or Star'ter Vol'ta e Ze'irements Min. Re uirement Max. Ca abilit E ui ment Name late Volta e Voltage t NPV Voltage '

NPV 6.9kv Motors (Continuous Operation) 6600V 5940 90% 7260 110%

4. 16kv Motors (Continuous Operation) 4000V 3600 90% 4400 110%

480V Motors m (Continuous Operation) 460V 414 90% 506 110%

480V Starters (Pickup) 480V 374 78% 528 110%

480V Starters (Holding) 480V 264 55%

NOTES: 1. NPV indicates nameplate voltage of motors or starters for the particular voltage level.

TABLE 8.3-14 CLASS 1E AUXILIARYAC DISTRIBUTION SYSTEM (25kv MA'IN GENERATING UNIT SUPPLY)

EXPECTED VOLTA'GES'VER 'GENERATOR VOLTAGE 'RANGE Minimum 25kV Generator Voltage Maximum 25kV Generator Voltage (23.8kv) (26.3kv)

Steady State 4.16kV Motor Steady State 4.16kV Motor r

Motor Motor Motor Motor System Level Voltacle NPV Voltacle NPV Voltaqe NPV Voltacle NPV 4.16kV Swgr. 3761 94 3448 86 4399 109 3996 99 480 V Swgr. 424.8 92 389 84 Note 5 4

480 V Motor 422.8. 91.9 387 84 Note 5 Terminals c r NOTES: 1. NPV indicates nameplate voltage of motors for the particular voltage level.

2. Under plant maximum loading conditions.

3. Under plant minimum loading conditions.

4. Voltage at. terminals of "worst case" motor.

5. Analysis for this case performed with 480V system unloaded to obtain maximum 4.16kV Swgr. level voltage.

0

TABLE 8. 3-15 CLASS lE AUXILIARYAC DISTRIBUTION SYSTEM (230kV GRID SUPPLY)

EXPECTED VOLTAGESOVER GRID VOLTAGE RANGE Minimum 230kV Grid Voltage Maximum 230kV Grid Voltage (230kv) (240kv)

Steady State 4.16kV Motor Steady State 4.16kV Motor L'oading Motor Motor Motor Motor System Level Voltage 2 NPV Voltacle 2 NPV Voltacle NPV Vo1taqe NPV 4.16kV Swgr. 3732 93 3333 83 4189 10'4 3755 93 480 V Swgr. 416.7 90.5 372. 1 80. 8 479 104 429 93 480 V Motor 4 415.0 90.2 370.4 80-5 468 101 418 90.8 Terminals .

NOTES: 1. NPV indicates nameplate voltage of motors for the particular voltage level.

2. Under plant maximum loading conditions.

3. Under plant minimum loading conditions.

4. Voltage at terminals of "worst case" motor.

TABLE 8.3-16 CLASS 1E AUXILIARYAC DISTRIBUTION SYSTEM (115kv GRID SUPPLY)

EXPECTED VOLTA'GES'VER GRID VOLTA'GE MNGE Minimum 115kV Grid Voltage Maximum 115kV Grid Voltage

• (11'3kV) (122kV)

Steady State 4.16kV Motor Steady State 4.16kV Motor Motor Motor Motor Motor System Level Voltacle NPV Voltacle NPV Voltacle NPV Voltacle NPV 4.16kV Swgr. 4024 '00 3793 94 4432 110 4177 104 480 V Swgr. 454 98 426 92 505 109 476 103 480 V Motor 4 452 98 424 92 503 109 474 103 Terminals NOTES: 'l. NPV indicates nameplate voltage of motors for the particular voltage level.

2. Under plant maximum loading conditions.

3. Under plant minimum loading conditions.

4. Voltage at terminals of "worst case" motor.

N,R

~Ã CO +

ml 0

WNP-2 TABLE 8.3-18 (Continued) page 2 of 2 Sw r./MCC DC Control Power Source aortae 1 Division Panel 2 Division SL-61 480V Note 4 SL-62 480V Note 4 SL-63 480V DP-Sl-2C SL-71 480V DP-Sl-1D SL-73 480V DP-Sl-1D SL-81 480V DP-Sl-2D SL-83 480V DP-Sl-2D MC-Sl-1D 125VDC DP-Sl-1 MC-Sl-2D 125VDC DP-'Sl-2 MC-S2-1A 250VDC DP-Sl-1D MC-S2-1B 250VDC DP-Sl-lc-1 NOTES:

l. Voltage is ac, unless otherwise noted.

2. See Figure 8.3-19

3. Control power for 480 volt ac motor control centers is obtained from control power transformers located in the units.

4. Swgr. contains manually operated breakers.

All indication is remote powered via the non-Class lE supervisory system.

8.3-105

WNP-2 AMENDMENT NO. 1 July 1978

'ABLE

8. 3-19 CLASS "lE AUXIIiIARY'AC DISTRIBUTION SYSTEM (ALL SUPPLIES)

'EXPECTED VOLTA'GESUNDER 'DEGRADED':SUPPLY CONDITIONS De raded Class 1E .Bus Su 1 (2870V)

S stem Level Voltaile  % Mo'to'r 'NPV

4. 16kV Swgr. 2870 71.'8 480 V Swgr. 332 72.1 480 V Motor 323 70.2 Terminals NOTES: 1~ NPV indicates nameplate voltage of motors for the particular voltage level.

2. Under plant maximum loading condition.

3. Corresponds to 69% of nominal 4.16kV bus voltage (undervoltage relay setpoint.)

'8.3-106

WNP-2 9.2.6 CONDENSATE SUPPLY SYSTEM 9.2.6.1 Design Bases The condensate supply system (COND) is designed to:

a. Store and provide a condensate supply to the reactor core isolation cooling (RCIC) system, the high pressure core spray (HPCS) system, and the RHR loops.

b. Maintain an adequate level of condensate in the condenser hotwell.

c. Provide a condensate supply for the control rod drive pumps.

d. Provide makeup water to the spent fuel pool.

e. Provide condensate for various radwaste processes.

f. Facilitate testing and/or flushing of the high pressure core spray, low pressure core spray, residual heat removal, and the reactor core isolation system.

g. Receive and accommodate a surge volume for condensate returned to the storage tanks after treatment in the liquid radwaste system.

h. System piping is designed to ANSI B31.1. Con-densate storage tanks are designed to ASME

'ection III, Class 3 requirements. System piping inside the reactor building is Seismic Category I. All other piping and system pumps are designed to Seismic Category II require-ments. The radwaste building condensate supply pump and the condensate filter demineralizer backwash pump are designed to ASME Code,Section XII. The reactor building condensate supply pump is designed to the Standards of the Hydraulic Institute.

9. 2-23

9.2.6.2 System Description The demineralized water system and the liquid radwaste system are the primary sources of makeup water to the condensate storage tanks.

The condensate supply system is shown on Figure 9.2-9. The system consists of two storage tanks (COND-TK-1A, COND-TK-1B) each with a nominal capacity of 400,000 gallons and equipped with electric heaters, a xeactor building condensate supply pump (COND-P-3), a radwaste building condensate supply pump (COND-P-4), a condensate filter demineralizer backwash pump (COND-P-5), and necessary .piping and instrumentation. The tanks are manufactured with'a design pressure of atmospheric plus full static head and maximum design temperature of 140oF.

Minimum operating temperature of the tanks is 40oF. The tanks are designed to withstand a wind load of 20 psf on the vertical projected area of the tank and a snow load of 20 psfo The radwaste building condensate supply pump and the conden-sate filter demineralizer backwash pump each are designed to supply l535 gpm at 185 ft total head. The radwaste conden-sate supply pump has a secondary operating point of 500 gpm at 220 ft total head. The reactor building condensate supply ft pump is. designed to supply 200 gpm at 220 total head.

All three pumps are designed to operate satisfactorily at temperatures between 40oF and 104oF and humidity between 20% and 90%.

A minimum inventory of 135,000 gallons in the condensate storage tanks is reserved for the RCIC and HPCS pumps. This assures the immediate availability of a sufficient quantity of condensate for emergency core cooling and reactor shut-down as discussed in 9.2.6.3.

Makeup for the condenser hotwell is gravity fed from the stor-age tank. Bleedoff water from the condensate system is returned to the storage tanks from the discharge of the condensate demineralizer.

A separate line'is provided to supply the control rod drive pumps with condensate. Condensate is supplied for various reactor building services, including fuel pool makeup by the reactor building condensate supply pump. The condensate storage tank can be drained to the condenser hotwell. Inad-vertant overflow of the tanks is collected in the concrete retaining basin surrounding the tanks. This water can be drained to the radwaste system for processing if sampling indicates that the water is radioactively contaminated. Rain water collected in the retaining basin can be drained to the storm sewer or the radwaste system if necessary for processing.

9.2-24

WNP-2 10.3.6.2 Materials Selection and Fabrication The requirements for welding the main steam supply system com-ponents and steam piping from the reactor to the turbine gen-erator are in accordance with ASME Section III, October 1973.

The welding requirements for all other piping is in accordance with ANSI,B31.1, October 1973 (See 3.2).

For low alloy steel piping designed in accordance with ASME Section III or ANSI B31.1 (October 1973), ASME Section III (October 1973) Appendix D or Table 131 of ANSI B31.1 (October 1973),covering non-mandatory preheat procedures are applic-able to all classes of welds. The control, of. preheat tem-perature for welding of low-alloy steel are in accordance with Regulatory Guide 1.50.,

Procedure qualifications for welding of austenitic stainless steel components include the following requirements:

a. The welding procedure is designed to avoid sensitization of the weld joint area and is in accordance with Regulatory Guide 1.44.

b. The test weld closely simulates the heat transfer properties of the actual welds to be performed.

c. In addition to the evaluation of the test welds specified in ASME Section IX (October 1973),

the weld area is examined for susceptibility to intergranular corrosion in accordance with ASTM A 262 '(Practice E or A) or A 393 (October 1973).

d. Controls are exercised to assure that nonmetallic thermal insulation for austenitic stainless steel are in accordance with Regulatory Guide 1.36.

e. The control of stainless steel welding conforms to Regulatory Guide 1.31.

10.3-5

Cleaning of components in the main steam system is in accord-ance with ANSI N45.2.1 (October 1973) or ASTM A380-57 (October 1973) for stainless steel surfaces and Regulatory Guide 1.37.

Welding procedure qualifications and welder performance quali-fications are in accordance with ASME Code,Section IX and addenda (October 1973) and the requirements of ASME Section III (October 1973). In addition to the aforementioned code, the following requirements also apply. All welding is done to qualified procedures by welders qualified to these proced-ures. Welding is done with a weld filler material procedure which includes cleanup of unused weld filler material at the end of each shift,. All recent contracts, except for monitoring and certifying requirements pertaining to welder qualification for areas of limited accessibility, conform to Regulatory Guide 1.71. Contracts prior to Regulatory Guide 1.71 meet the intent of this guide. Where welding is per-formed on low or high alloy steel piping in an area where accessibility is limited such that the welder cannot view the weld melt puddle directly, or must use extension rods, or must bend the electrode, the procedure and performance qualifications require a mock-up to simulate this limitation.

10.3-6

WNP- 2 AMENDMENT NO. 8 February 1980

b. Airborne Radiation Monitors Airborne radioactivity monitoring for plant per-sonnel protection and surveillance utilizes:

fixed location, continuous particulate monitors which include continuous iodine samplers; por-table continuous particulate monitors with con-tinuous iodine samplers positioned at specific work sites; and particulate and iodine grab samples taken before and during specific jobs.

The installed continuous particulate monitoring system was designed for responsive personnel pro-tection and plant surveillance. The five installed particulate monitors measure par-ticulate activity levels in the Radwaste, Reactor and Turbine Building ventilation exhaust air, and furnish alarm and recording signals to the main control room. These monitors draw 5 cfm of ven-tilation air through a moving filter tape. A shielded beta detector is used with an efficiency of approximately 30% and an external background of less than 40 cpm/mr/hr. The resultant sen-sitivity of the system is 438 cpm above back-ground after one hour of sampling at a 1 x 10 pCi/cc concentration in air.

The actual ability of a ventilation exhaust moni-toring system to detect one MPCa in a specific space is dependent upon the following factors:

1. Flow rate ratio (flow of air from a specific confined space/flow rate of bulk ventilation system exhaust)

2. Particulate activity of ventilation system air

3. MPCa of the specific confined space (radio-nuclide composition)

Normal plant conditions are expected to yield a bulk ventilation exhaust air concentration (pri-marily short-lived fission product daughters and natural activity) of l. 3 x 10 Ci/cc. The MPCa for normal plant air is expected to be greater than 6 x 10 8 Ci/cc, and the ventilation monitor-ing system will be able to detect one MPCa hour in all locations for normal airborne mixture background concentrations. At high activity

12. 3-21

WNP-2 AMENDMENT NO. 11 September 1980 concentrations in the bulk ventilation exhaust air the ability to detect one MPCa concentration in areas with low ventilation exhaust flow rates may be compromised. Under these conditions, corrective actions including assessment by portable sampling system results and portable monitoring activities will establish activity levels in all normally occupied areas which have potential for abnormal airborne activity.

Table 12.3-2 illustrates the detection capabil-ities of the fixed location particulate moni-toring system for various ratios of isolated area ventilation flow rates compared to the bulk ven-tilation exhaust flow rate from the reactor building.

In the Radwaste Building, the potentially con-taminated areas normally entered by people would be those corridors adjacent to radioactive liquid and gaseous waste processing systems equipment such as demineralizers, concentrators, waste storage tanks, recombiners, driers, moisture separators and charcoal holdup vessels. Assuming that exfiltration from any one of the process systems to a normally entered corridor was suf-ficient to attain MPCa levels for Cs-137 in that corridor, the dilution ratio would approach a factor of 10 to 100. For the worst case (100 to 1 ratio of bulk ventilation flow rate to corridor flow rate), Cs-137 at MPCa would be detected within one hour on the continuous par-ticulate monitor. If exfiltration from a process vessel cubicle was sufficient to produce MPCa levels in an adjoining corridor, it is more probable that the normal cubicle flow rate input to the bulk ventilation flow would produce a prior distinguishable countrate ramp.

12.3-22

WNP-2 AMENDMENT NO. 1 1 September 1980 In the Turbine Building, airborne contamination is most likely to arise from nuclear steam leaks or of f-gas processing systems piping. Heater bay areas, steam jet air ejector rooms, turbine areas and steam driven feedwater pump areas have the potential for airborne contamination. The areas have individual ventilation exhaust rates in excess of 5000 cfm and Cs-137 MPCa concentra-tions originating in these areas would give a continuous air monitor response ramp which is distinguishable within one hour.

Each of the continuous particulate monitors has an associated iodine sampling cartridge which is counted weekly for baseline and surveillance information. This cartridge and iodine sampled/

collected with portable sampling devices will be analyzed in the plant laboratory counting facility when abnormal airborne activity levels are signaled by a continuous particulate moni-toring system. At a 2 cfm air flow rate through an iodine sampling cartridge, iodine present at an occupational MPCa concentration of 9 x 10 p Ci/cc would be quantitatively observable within a one-minute sample interval. Bases for this assessment are a ten-minute count time on a 12 15% Ge(Li) detector system having an overall efficiency of about 1% when source and geometry considerations are included. The information presented for detecting one MPCa concentration for Cs-137 in areas having a low ventilation flow rate can also be applied to the iodine case.

One MPCa of iodine can be ascertained within a one-hour sampling period with a dilution factor of greater than 100. The routine weekly analysis of integrated iodine samples of Radwaste, Reactor and Turbine Building ventilation air will permit observation 'of small iodine inputs. When these inputs are significant, a particulate and iodine sampling program is initiated to establish the source point.

With the exception of outage periods, continuous particulate monitoring is reasoned to be the most responsive personnel protection and internal plant surveillance mechanism available. In addi-tion, all tasks with potential for generating airborne contamination will be performed only when authorized by a Radiation Work Permit (RWP).

12.3-22a

WNP-2 AMENDMENT NO. 11 September 1980 The RWP assesses the radiological hazards, establishes additional monitoring and sampling requirements and, if necessary, specifies re-quired engineering control and/or respiratory protection.

During outages, the above airborne monitoring system will be augmented by additional iodine sampling (continuous and grab) on the refueling floor since airborne iodine concentrations can exist in the absence of particulate contamination.

12. 3-22b

Table 12.3-2 Particulate Concentration at Various Flow Rate Ratios Necessary To Increase Reactor Building Exhaust Activity By 25%

Count Rate, Bulk Reactor Including Exhaust,

'ldg.

External Concentration, pCi/cc (1) , at Listed Flow Rate Ratios (2)

~

Particulates Background to Produce a 25% Increase in Count Rate 0.10 0.05 0.02 0.01 0.004"'

x 10-11 250 7.5x10 1.5x10 3.8x10 7.5x10 1.9x10 1 x 10-10 550 2.5x10 5x10 l.lxl0 2.5xl0 6.3xlO 3 x 10 1445 7.5xlO 1.5x10 3.8x10 7.5x10 1,9x10 lx10 4580 2.5xlo 5x10 l.lxlO 2.5x10 6.3x10 (1) Cs-137 is used as the reference nuclide; MPC for soluble Cs = 6 x 10 -8 yCi/cc.

(2) Flow Rate Ratio ~ flow rate of specific area < flow rate of total Rx Bldg exhaust.

(3) A flow rate ratio of 0.004 corresponds to the CRD Room with a 400 cfm flow rate compared to a 97,000 cfm reactor building exhaust flow rate.

4Q

<wR R v

f 00 +

0

WNP-2 AMENDMENT NO. 1 July 1978 (BLANK)

15.6.5.5.1 Design Basis Analysis The methods, assumptions, and conditions used to evaluate this accident are in accordance with those guidelines set forth in the NRC Standard Review Plan 15.6.5 and Regulatory Guides 1.3 Rev. 0 and 1.7 Rev. 2. The specific models, assumptions, and computer code used, to evaluate this event based on the above criteria are presented in Reference 15.6-4. Specific values of parameters used in this evalu-ation are presented in Table 15.6-12.'5.6.5.5.1.1 Fission Product Release from Fuel It is assumed that 100% of the noble gases and 50% of the iodine are released from an equilibrium core operating at a power level of 3468MWt for 1000 days prior to the accident.

While not specifically stated in Reg. Guide 1.3 the assumed release of 100% of the core noble gas activity and 50% of the iodine activity implies fuel damage approaching melt conditions. Even though this condition is inconsistent with operation of the ECCS system (see 6.3),

cable for the evaluation of this accident. Of this release, it is assumed appli-100% of the noble gases and 50% of the iodine become air-borne. The remaining 50% of the iodine is removed by plate-out and condensation; therefore, it borne release to the, environment. The activity airborne in is not available for air-the containment is presented in Table 15.6-13.

15.6.5.5.1.2 Fission Product Transport to the'Environment The transport pathway consists of leakage from the containment to the secondary containment by several different mechanisms and is discharged to the environment through the standby gas treatment system (SGTS) at an elevated location. The SGTS filter efficiency for 'iodine removal is assessed at 99%. The mechanisms for leakage from the primary containment are dis-cussed below:

a. Containment Leakage The design basis leak rate of the containment and its penetrations (See 6.2.6) is .5% per day for the duration of: the accident. This leakage is to the secondary containment and from there to the environment via the SGTS. No credit is taken for mixing and holdup with'in the secondary containment.

15.6-33

AMENDMENT NO. 9 April 1980

b. Leakage from engineered safety feature (ESF) components outside the primary containment-all ESF equipment which circulates primary coolant or suppression pool water during the course of the postulated accident is located within the secondary containment so that any leakage from the pressure barriers for these systems is 'into the secondary containment atmosphere and is therefore processed by the SGTS prior to release to the environment. Due to the higher SSW pressure any leakage through the RHR heat exchangers would be from the service water side to the ECCS side.

c~ Hydrogen Purge Since the hydrogen recombining system consists of two 100% redundant recom-biners, no'hydrogen purge is required nor assumed throughout the post accident period.

d. Leakage from the main steam isolation valve leakage control system (MSIV-LCS). The MSIV-LCS routes any leakage through the MSIVs to an area serviced by the SGTS. Assuming the MSIVs leak at ll. 5 SCFH per valve, leakage past the inboard MSIVs is conservatively estimated to begin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident. The airborne fission products are assumed to be, uniformally mixed in the dr@well. air volume neglecting the suppres-sion pool air volume. For conservation this leakage is assumed to be in addition to the .5%

containment leakage discussed in paragraph "a" above. For maximum leakage through each main-stream line, (4 MSIVs or 46 scfh), this correla-tes to 0.23 weight %/day leak rate.

e. Bypass Leakage As described in 6. 2. 3. 2 and 6.2.3.3, 0.'74 scfh primary containment leakage is assumed to bypass the secondary containment and.

leak directly to the environment.

I Fission product release to the environment based on the above assumptions is given in Table 15. 6-14 and 15.

15. 6. 5. 5. l. 3 Results The calculated exposures for the design basis analysis are presented in Table 15.6-16 and are well within the guidelines of 10CFR100.

15. 6-34

TABLE 15.6-12 Page 1 of 2 LOSS-OF-COOLANT ACCIDENT PARAMETERS TABULATED FOR POSTULATED'CCIDENT ANALYSES Design Realistic Basis Basis I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level 3468MWt 3468MWt B. Burn-up NA NA C. Fuel damaged 100% 0 D. Release of activity by Table Table nuclide 15.6-14 a 15 15.6-19 6 20 E. Iodine fractions (1) Organic 4% 1%

(2) Elemental 91% 99%

(3) Particulate 5% 0 F. Reactor coolant activity before the accident NA 15.6.5.5.2.1 II. Data and assumptions used to estimate activity released

'. Primary containment leak rate (%/day) 0.5 0.5 B. Secondary containment leak rate (%/day) NA 100 C. Valve movement times NA NA D. Adsorption and filtration efficiencies (%)

(1) Organic iodine 99 99 (2) Elemental iodine 99 99 (3) Particulate iodine NA NA (4) Particulate fission products NA NA 15.6-37

WNP-2 TABLE 15.6-12 (Continued) Page 2 of 2 Design Realistic Basis Basis II. Data and assumptions used to estimate activity released E. Recirculation system para-meters (1) Flow rate (CFM) NA NA (2) Mixing efficiency NA NA (3) Filter efficiency NA NA F. Containment spray parameters (flow rate, drop size, etc.) NA NA G. Containment volumes NA NA H. All other pertinent data and assumptions None None III. Dispersion Data A. Boundary and LPZ distance (m) 1950/4827 1950/4827 B. X/Q's for time intervals of (1) 0-2 hr SB/LPZ 7.50xlO 5/ 7." 50xlO 2.80xlO 2.80xl0 5/

(2) 2-8 hr LPZ 2.80xl0-6 2.80xlO (3) 8-24 hr LPZ 3.45xlO. 3.45xlo 66 (4) 1-4 days LPZ 1.59x10-66 1.59xl0-6 (5) 4-30 days - LPZ 1.02x10 1.02xlO IV. Dose Data A. Method of dose calcula- Reference Reference tion 15.6-4 15.6-2 B. Dose conversion assump- Reference Reference tions 15.6-4 15.6-2 C. Peak activity concentra- Table Table tions in containment 15.6-13 15.6-18 D. Doses Table, Table 15.6-16 15.6-21 15 6-38

WNP-2 TABLE OF CONTENTS (Continued) PacCe Regulatory Guide 1.84, Rev. 7 C.2-83 Regulatory Guide 1.85, Rev. 7 C.2-84 Regulatory Guide 1.88, Rev. 2 C.2-85 Regulatory Guide 1.89, Rev. 0 C.2-86 Regulatory Guide 1.92, Rev. 1 C.2-87 C.3.0 BALANCE OF PLANT SCOPE= OF SUPPLY EVALUATION C. 3-1 Regulatory Guide 1.3, Rev. 2 C. 3-2 Regulatory Guide 1.5, Rev. 0 C.3-3 Regulatory Guide 1.6, Rev. 0 C.3-4 Regulatory Guide 1.7, Rev. 1 C.3-5 Regulatory Guide 1.8, Rev. 1-R C.3-6 Regulatory Guide 1.9, Rev. 0 C.3-7 Regulatory Guide 1.10, Rev. 1 C.3-8 Regulatory Guide 1.11, Rev. 0 C.3-9 Regulatory Guide 1.12, Rev. 1 C.3-10 Regulatory Guide 1.13, Rev. 1 C.3-11 Regulatory Guide 1.15, Rev. 1 C.3-12 Regulatory Guide 1.16, Rev. 4 C.3-13 Regulatory Guide 1.17, Rev. 1 C.3-14 Regulatory Guide 1.18, Rev. 1 C.3-15 Regulatory Guide 1.19, Rev. 1 C.3-16 Regulatory Guide 1.21, Rev. 1 C.3-17 Regulatory Guide 1.22, Rev. 0 C.3-18 Regulatory Guide 1.23, Rev. '0 C.3-19

TABLE OF CONTENTS (Continued) Pa<ac Regulatory Guide 1.25, Rev. 0 C. 3-20 Regulatory Guide 1.26, Rev. 3 C.3-21 Regulatory Guide 1.27, Rev. 2 C.3-22 Regulatory Guide 1.28, Rev. 0 C.3-23 Regulatory Guide -1.29, Rev. 2 C.3-24 Regulatory Guide 1.30, Rev. 0 C.3-25 Regulatory Guide 1.31, Rev. 2 C. 3-26 Regulatory Guide 1.32, Rev. 2 C.3-27 Regulatory Guide 1.33, Rev. 1 C.3-28 Regulatory Guide 1.34, Rev. 0 C.3-29 Regulatory Guide 1.35, Rev. 2 C.3-30 Regulatory Guide 1.36, Rev. 0 C.3-31 Regulatory Guide 1.37, Rev. 0 C.3-32 Regulatory Guide 1.38, Rev. 2 C.3-33 Regulatory Guide 1.39, Rev. 1 C.3-34 Regulatory Guide 1.40, Rev. 0 C.3-35 Regulatory Guide 1.41, Rev. 0 C.3-36 Regulatory Guide 1.43, Rev. 0 C.3-37 Regulatory Guide 1.44, Rev. 0 C.3-38 Regulatory Guide 1.45, Rev. 0 C.3-39 Regulatory Guide 1.46, Rev. 0 C.3-40 Regulatory Guide 1.47, Rev. 0 C.3-41

/ C. 3-42 Regulatory Guide 1.48, Rev. 0 .

Regulatory Guide 1.50, Rev. 0 C.3-43 C-iv

TABLE OF CONTENTS (Continued) Pacre Regulatory Guide 1.51, Rev. 0 C.3-44 Regulatory Guide 1.52, Rev. 1 C.3-45 Regulatory Guide 1.53, Rev. 0 C.3-46

\

Regulatory Guide 1.54, Rev. 0 C.3-47 Regulatory Guide 1.55, Rev. 0 C.3-48 Regulatory Guide 1.56, Rev. 0 C.3-49 Regulatory Guide 1.57, Rev. 0 C.3-50 Regulatory Guide 1.58, Rev. 0 C.3-51 Regulatory Guide 1.59, Rev. 1 C.3-52 Regulatory Guide 1.60, Rev. 1 C.3-53 Regulatory Guide 1.61, Rev. 0 C.3-54 Regulatory Guide 1.62, Rev. 0 C.3-55 Regulatory Guide 1.63, Rev. 1 C.3-56 Regulatory Guide 1.64, Rev. 2 C.3-57 Regulatory Guide 1.67, Rev. 0 C.3-58 Regulatory Guide 1.68, Rev. 1 C.3-59 Regulatory Guide 1.68.1, Rev. 1 C.3-60 Regulatory Guide 1.68.2, Rev. 0 C.3-61 Regulatory Guide 1.69, Rev. 0 C.3-62 Regulatory Guide 1.70, Rev. 2 C.3-63 Regulatory Guide 1.71, Rev. 0 C.3-64 Regulatory Guide 1.72, Rev. 0 C.3-65 Regulatory Guide 1.73, Rev. 0 C.3-66 Regulatory Guide 1.74, Rev. 0 C.3-67 C-v

TABLE OF CONTENTS (Continued) Parcae Regulatory Guide 1.75, Rev. 1 C.3-68 Regulatory Guide 1.76, Rev. 0 C.3-69 Regulatory Guide 1.78, Rev. 0 C.3-70 Regulatory Guide 1.80, Rev. 0 C.3-71 Regulatory Guide 1.82, Rev. 0 C.3-72 Regulatory Guide 1.84, Rev. 9 C.3-73 Regulatory Guide 1.85, Rev. 9 C.3-74 Regulatory Guide 1.88, Rev. 2 C.3-75 Regulatory Guide 1.89, Rev. 0 C.3-77 Regulatory Guide 1.90, Rev. 0 C.3-78 Regulatory Guide 1.91, Rev. 0 C.3-79 Regulatory Guide 1.92, Rev. 1 C.3-80 Regulatory Guide 1.93, Rev. 0 C.3-81 Regulatory Guide 1.94, Rev. 1 C.3-82 Regulatory Guide 1.95, Rev. 1 C.3-83 Regulatory Guide 1.96, Rev. 1 C.3-84 Regulatory Guide 1.97, Rev. 0 C.3-85 Regulatory Guide 1.100, Rev. 0 C.3-86 Regulatory Guide 1.101, Rev. 1 C.3-87 Regulatory Guide 1.102, Rev. 1 C.3-88 Regulatory Guide 1.103, Rev. 1 C.3-89 Regulatory Guide 1.104, Rev. 0 C.3-90 Regulatory Guide 1.105, Rev. 1 C.3-91 Regulatory Guide 1.106, Rev. 1 C.3-92 C-vi

Regulatory Guide 1.88, Rev. 2, October 1976 Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records.

Regulatory Guide Intent:

This guide describes an acceptable method of complying with the NRC's regulations for collection, storage, and maintenance of quality assurance records.

Application Assessment:

Assessed Capability in Design.

Compliance or Alternate Approach Statement:

The identified Boiling Water Reactor Quality Assurance Program utilized on this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or NRC-approved alternate position.

General Compliance or Alternate Approach Assessment:

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation

Reference:

Information was provided at the PSAR stage.

Similar Application

Reference:

Similar application has not been utilized on other projects.

C.2-85

WNP-2 AMENDMENT NO. 13 February 1981 Regulatory Guide 1.89,. Rev. 0, November 1974 Qualification of Class 1E Equipment for Nuclear Power Plants Regulatory Guide Intent:

Regulatory Guide 1.89 endorses both the requirements and recommendations of IEEE Standard 323-1974. "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations". Regulatory position stipula-tions are also included.

This regulatory guide is applicable to all Class 1E electrical equipment.

Application Assessment:

Assessed Capability in Design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design and/or equipment util'ized in this facility is in compliance'ith the intent of the subject regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

All environmental and seismic qualification testing of Class 1E equipment within the NSSS scope of supply was in compliance with IEEE 323-1971 and IEEE 344-1971 which were NRC accepted standards.

Specific Evaluation

Reference:

Refer to 3.10, 3.11 and 7.1.2.4.

Similar Application

Reference:

Similar application was utilized on Zimmer and LaSalle.

C.2-86

General Compliance or Alternate Approach Assessment (Cont.)

where R = Combined Response R.= Response i in the i.th mode n = Number of Modes considered in the analysis Closely spaced modes are not accounted for as required by the guide because the design was significantly developed prior to issuance of the guide.

Specific Evaluation

Reference:

Refer to 3.7.3.6 and 3.7.3.7.

Similar Application

Reference:

Similar application was utilized on LaSalle and Zimmer.

C.2-89

WNP-2 AMENDMENT NO. 7 November 1979 Regulatory Guide 1.68, Rev. 1, January 1977 Initial Test Programs for Water-Cooled Reactor Power Plants Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to the WNP-2 initial test program since Revision 0 of this regulatory guide is committed to in FSAR 14.2.7. However, WNP-2 complies with the intent of the'guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate A'pproach Assessment:

Refer to FSAR Chapter 14 for description of initial testing program and to 14.2.7 and Appendix C.2.0 for statements concerning compliance with Regulatory Guide 1.68, Rev. 0. Revision No. 1 of this guide in general clarifies Revision No. 0 and therefore there are no exceptions to the intent of this procedure.

Specific Evaluation

Reference:

FSAR 14.2.7 and Appendix C.2.0 discussion Reg. Guide 1.68, Rev. No. 0.

C. 3-59

WNP-2 AMENDMENT NO. 7 November 1979 Regulatory Guide 1.68, Rev. 1, January 1977 .Preoperational and Xnitial Startup of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants.

Compliance or Alternate Approach Statements:

WNP-2 complies with the intent of the guidance set forth in this Regulatory Guide by an alternate approach.

General Compliance or Alternate Approach Assessments:

The preoperational testing and the initial Startup testing as described in FSAR, Chapter 14, complies with the in-tent of this Regulatory Guide. However, due to the limitations of the auxiliary steam supply system, the confirmation that, the feedwater pumps satisfy required head, flow rate and suction head will not. occur until the startup phase of the initial test program when the normal steam supply is available to the feedwater pump turbines.

Specific Evaluation

Reference:

FSAR 14.2.12.1.1 C.3-60

WNP-2 AMENDMENT NO. 13 February 1981 Regulatory Guide 1.68.2, Rev. 0, January 1977 Compliance or Alternate Approach Statement:

WNP-2 complies with the intent of the guidance set forth in this Regulatory Guide by an alternate approach.

General Compliance or Alternate approach assessment:

The startup test described in FSAR 14.2.12.3.28 complies with the Regulatory Guide with the following exceptions:

a. The test will be initiated by scramming plant from the control room versus a location outside the control room as described in Section C.3 of the Regulatory Guide. This exception is made to better simulate the actual procedure which would be followed if a control evacuation were to occur. The capabil-ity to scram the reactor outside the control room exists; for example, tripping the RPS MG sets.

b. The Cold Shutdown Demonstration Procedure as described in Section C.4 of the Regulatory Guide may not be performed immediately following the demonstra-tion of achieving and maintaining safe hot standby from outside the control room. Rather this cooldown portion may be performed when cooldown is required during the course of the normal power ascension test program. Although this is an exception to Regulatory Guide 1.68.2, Rev. 0, Revision 1 of this Guide con-tains provisions for a delay in the demonstration of cooldown.

Specific Evaluation

Reference:

FSAR 14.2.12.3.28, 7.4.1.4.

C.3-61

Regulatory Guide 1.69, Rev. 0, December 1973 Concrete Radiation Shields for Nuclear Power Plants Compliance or Alternate Approach Statement:

WNP-2 complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment:

Although the regulatory guide was promulgated after design and specification implementation of the engi-neering criteria, the recommended design and construc-tion practices specified in the regulatory guide are documented in codes and specifications which were used in the development of the engineering criteria and contract specifications.

Specific Evaluation

Reference:

Refer to 12.3.2.

C.3-62

Regulatory Guide 1.89, Rev. 0, November 1974 Qualification of Class 1E Equipment for Nuclear Power Plants Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to WNP-2 since it applies to the evaluation of construction permit applications docketed after July 1, 1974.

However, WNP-2 complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Statement:

(To be provided at a later date)

Specific Evaluation

Reference:

(To be provided at a later date)

C.3-77

Regulatory Guide 1.90, Rev. 0, November 1974 In-Service Inspection of Prestressed Concrete Containment Structures with Grouted Tendons Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to WNP-2 because WNP-2 does not have a prestressed concrete containment structure with grouted tendons.

General Compliance or Alternate Approach Assessment:

Not, applicable.

Specific Evaluation

Reference:

Not applicable.

C.3-78

WNP-2 Regulatory Guide 1.97, Rev. 0, December 1975 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During 'and Following an Accident Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to WNP-2 since it, applies to the evaluation of construction permit applications docketed on or after August 1, 1976.

General Compliance or Alternate Approach Assessment:

WNP-2 provides sufficient instruments in the main control room to monitor plant variables and systems during and following an accident. The instrumentation is qualified to remain functional during the worse case environmental conditions that it must monitor. The indicators and recorders are not seismically qualified. Means are provided to monitor the primary containment atmosphere, the spaces containing components for recirculation of loss of coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from postulated accidents.

Specific Evaluation

Reference:

Refer to 7.5.

C ~ 3-85

WNP-2 Regulatory Guide 1.100, Rev. 0, March 1976 Seismic Qualification of Electric Equipment for Nuclear Power Plants Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to WNP-2 since it applies to the evaluation of construction permit applications docketed after November 15, 1976.

General Compliance or Alternate Approach Assessment:

(To be provided at a later date)

Specific Evaluation

Reference:

(To be provided at a later date)

C.3-86

WNP-2 AMENDMENT NO ~ 8 February 1980 Regulatory Guide 1.128, Rev. 0, April 1977 Installation Design and Installation of'Large Lead Storage Batteries for Nuclear Power Plants.

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to WNP-2 since it.

applies to the evaluation of construction permit applica-tions docketed after December 1, 1977. However, WNP-2 complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment:

Safety-related battery installation design criteria con-forms to IREE Standard 484-1975. In addition, HYDROCAPS (catalyst battery caps) are provided to preclude discharge of combustible gases into the battery room area. A Class lE ventilation system is also provided which is capable of limiting hydrogen concentrations (Neglecting HYDROCAPS) to 1%.

Storage prior to installation was not in strict com-pliance with subsection 5.1.3 "Storage" of the subject regulatory guide. However, preoperational tests will establish whether or not any damage or loss of capacity resulted from storage.

Specific Evaluation

Reference:

8.3 ~ 2~1~5 8~3~2~1~6 8 '.2.2 ~ 1 ~ 1 8 ~ 3 ~ 2 ~ 2 ~ 1.2 C.3-111

Regulatory Guide 1.129, Rev. 0, April 1977 Maintainance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants Compliance or Alternate Approach Statement:

(To be provided at a later date)

General Compliance or Alternate Approach Assessment:

(To be provided at a later date)

Specific Evaluation

Reference:

(To be provided at a later date)

C.3-112

WNP-2 AMENDMENT NO. 12 November 1980" Figure Engineering

. Number Title Dwg.No; 2.1-3 Overall Site Plan C025 2.1-4 Plant Plot Plan SS053 2.3-1 Overall Site Plan C020 E. 2-1

NNP-2 AMENDMENT NO. 13 February 1981 LIST OF NRC QUESTIONS NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 005.001 010.001 010.002 010.003 010.004 010.005 010.006 13 010.007 010.008 010.009 010.010 010.011 010.012 010.013 010.014 010.015 13 010.016 5,13 010.017 13 010.018 13 010.019 010.020 010.021 13

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130.024 8 130.025 130.026 130.027 130.028 130.029 130.030 130.031 130.032 130.033 130.034 130.035 130.036

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 130.037 130.038 '2 130.039 '7 8

130.040 130.041 130.042 130.043 130.044 130.045 12 130.046 12 130.047 12 130.048 12 130.049 12 210.001 211.002 211.003 211.004 211.005 211.006 211.007 211.008 211.009 211.010

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 211.011 211.012 211.013 211.014 211.015 211.016 211 ."017 211.018 211.019, 211.020'11.021 211.022 211.023 211.024 211'.025 211.026 211.027 211.028 211.029 211.030 211.031 8,10 211.032 211.033

WNP-2 AMENDMENT NO ~ 13 February 1981 NRC QUESTIONS'O.

QUESTIONS OF PAGES AMENDMENT 211.034 1 211.035 1 211.036 211.037 211.038 211.039 211 .040 211.041 211.042 211.043 211.044 211.045 211.046 211.047 1 211.048 2 211.049 211.050 211.051 13 211.052 211.053 211.054 211.055 211.056

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 211.057 211.058 211.059 211.060 211.061 211.062 11 211.063 11 211.064 211.065 211.066>>

211.067 211.068 211.069 211.070 211.071 211.072 211.073 211.074 211.075 211.076 211.077 211.078 211.079

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 211 .080 211.081 211.082 211.083 211.084 211.085 211.086 211.087 211.088 211.089 211.090" 211.091

• 211.092 211.093 211.094 211.095 211.096 211.097 211.098 211.099 211.100 211.101 211.102

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 211.103 211.104 211.105 211.106 212.001 212.002 212.003 212.004 221.001 221.002'21.003 221.004 221.005 221.006 221.007 221.008 221.009 221.010 2" 221.011 221.012 221.013 222.001 222.002

NNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 222.003 222.004 231.001 231.002 231.003 232.001 232.002 232.003 232.004 232.005 312.001 312.002 312.003 312.004 312.005 312.006 312.007 312.009 312.010 312.011 312.012 312.013 312.014

NNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 312.015 312.016 312.017 312.018 312.019 321.001 321.002 321.003 13 321.004 321.005'31.001 13 331.002 13 331.003 331.004 331.005 331.006 331.007 331.008 331.009 331.010 331.011 331.012 331.013

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 331.014 331.015 331.016 331.017 331.018 331.019 331.020 331.021 13 331.022 331.023 331 .024 360.001 360.002 360.003 360.004, 10 360.005 10 362.001 362.002 362.003 362.004 362.005 362.006 362.007

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS -NO'. OF PAGES AMENDMENT 362.008 362.00'9 371.001 371.002 371.003 371.004 371.005 371.006 371.008 371.009) 371.010 371.011 371.012 371.013 371.014 372.001 372.002 372.003 372.004 372.005 372.006 372.007 372.008

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF PAGES AMENDMENT 372.009 372.010 372.011 372.012 372.013 372.014 372.015 372.016 372.017 422.001 422.002 422.003 422.004 7 422.005 422.006 422.007 422-008 422.009 423.001 423.002 423.003 423.004 423.005

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS QUESTIONS NO. OF'PAGES AMENDMENT 423.006 423.007 423.008 423.009 423.010 423.011 423.012 423.013 423.014 423.015<

423.016 423.017 423.018 423.0l9 423.020 423.021 423.022 423.023 423.024 423.025 423.026 423.027 423.028

WNP-2 AMENDMENT NO. 13 February 1981 NRC QUESTIONS UESTIONS NO. OF PAGES AMENDMENT 423.029 432.001 '7 432.002 432.003 432.004 432.005 432.006 432.007 432.008 432.009 432.010 432.011 432.012 432.013 2 432. 014'."

432.015 432. 0'16 441.001 441.002 441.003 441.004 441.005

WNP-2 AMENDMENT NO. 1 1 September 1980 Q. 022.007 Provide the, secondary containment pressure time response for the design basis accident. List and discuss all assumptions made in this analysis.

~Res ense:

See revised 6.2.3.3 and Table 6.2-29.

022.007-1

WNP-2 AMENDMENT NO. 1 July 1978 Page 1 of 1 Q 022.8 Identify the manufacturer of the hydrogen recombiner and describe the test. program which demonstrates that the hydrogen recombiner will perform as required in the con-tainment environment following a postulated loss-of-coolant accident. Provide the systems quality group classification including that of the hydrogen analyzer.

Res onse:

The hydrogen recombiner was manufactured by Air Products and Chemicals, Inc. The tests are briefly discussed in 6.2. 5.4 and detailed, supplemental information is being provided by separate letter to the NRC. The analyzer, in-cluding quality group classification, is discussed in

7. 6. 1.13.8. All pressure containing equipment, including piping between components, is considered an extension of the containment and is classified code Quality Group 3, as dis-cussed in 6.2:5.2.3.

022. 8-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 Q 22.025 Identify the location of the hydrogen sampling points in the drywell and the suppression chamber. Identify the location of the suction and discharge points of the combus-tion gas control system with respect to local structures and equipment.

~Res onse Please refer to FSAR Table 7.6-12 and Figures 6.2-32 through 6.2-35 for the requested information.

022.025-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 Q 22.026 In accordance with Appendix J of 10CFR50 we require that con-tainment isolation valves for those systems not vented and drained during Type A tests, are to be tested in accordance with Section III.C of Appendix J and those results are to be reported to the Commission.

Res onse:

In general the containment isolation valves for those systems not vented and drained during Type A tests are being Type C tested. The exceptions to this are listed in the response to question 22.10 along with the justification. The results of these tests will be reported to the Commission as stated in FSAR 6.2.6.4.

022.026-1

. WNP-2 AMENDMENT NO. 13 February 1981 Q. 022.027 (6.2.6)

Augment Table 6.2-16 .to provide the information requested in Section 6.2.4.2, "Systems Design" of Regulatory Guide 1.70.

~Res onse Tables 6.2-13, 6.2-16 and 7.3-13 have been combined into one

.table, number 6.2-'16. This table has been further expanded to include all the information required in Section 6.2.4.2 of

,.Regulatory Guide 1.70. Chapter 6 has been revised to reflect the revised Table 6.2-16.'22.027-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 Q 22.028 Several of the loads presented in Table 3.4-1 of the Plant Design Assessment Report (DAR} have been generated using computer codes which have not been reviewed by the NRC staff.

Provide a complete description'f your method of analysis for all codes presented in Appendix D of the DAR.

~Res onse Three computer codes have been used to develop short. term LOCA hydrodynamic loads for WNP-2 plant assessment. The three codes are the downcomer vent clearing analytical model computer code VENT, the pool swell analytical model computer code SWELL, and the LOCA bubble charging analytical model computer code BUBBLE. Complete documentation on each of the above three codes is provided or referenced in Appendix D of the DAR (see below for specific references}.

Documentation on the WNP-2 load calculation procedure is provided in 3.2.1 of the DAR.

Vent Code a) Assumptions See Section D.2 of the DAR.

b) Equations See Figure D-1 of the DAR.

c) Methodology See Figure D-1 of the DAR.

2. Swell Code The assumptions, equations, and methodology for the Swell Code are identical to that described in Reference D-l.

3 ~ Bubble Code The assumptions, equations, and methodology for the Bubble Code are identical to that described in Reference D-7.

022. 028-1

NNP-2 AMENDMENT NO. 5 August 1979 Sheet 1 of 2 Q 22.043 Tables 6.2-13, 6.2-16, and 7.3-13 of the FSAR indicate that a check valve outside the containment is considered as a containment isolation valve for the minimum flow at the pumps in the reactor heat removal system (X-47, 48), vacuum relief from secondary containment (X-66, 67, 119) and a process sample line (X-69D). Provide justification for this design approach.

~Res onse Tables 6.2-13, and 7.3-13 have been deleted. See question 22.027 for revised Table 6.2-16.

There are check valves inboard of the isolation valves on the minimum flow line from the RHR pumps (X-47, X-48). These valves are built to the same standards as the isolation valves and will, if necessary, isolate the minimum flow line from the primary containment; however, these check valves are not considered containment isolation valves. Please see revised Table 6.2-16.

There are no check valves on the process sample line (X-69D).

The notation, C.V., which was previously used was not intended to designate check valve. Revised Table 6.2-16 now clearly designates the valve types for the isolation valves on penetration X-69D.

Both isolation valves on the reactor building to wetwell vacuum relief lines (X-66, X-67, and X-119) are located out-side the wetwell to improve valve operability (see Note 17 of revised Table 6.2-16). The reactor building to wetwell vacuum relief system is required to prevent excessive negative pressures in the primary containment under certain postulated conditions .(see 6.2.1.1.4). The disc in the check valve is maintained in the close position during normal operation by means of a spring actuated lever arm and magnets embedded in the periphery of the disc. The magnetic and spring forces are overcome, and the disc starts to open, when the pressure differential across the valve exceeds 0.2 psid. The check valves have position indication lights which can alert the operators to the fact that a check valve is not fully closed. The operator can then remotely shut the valve by means of a pneumatic operator. The operating switch is spring-return to neutral. The air supply to these valves is Quality Class I.

022. 043-1

WNP-2 AMENDMENT NQ. 5 August. 1979 Sheet 2 of 2 Revised Table 6.2-16 now lists check valves outside containment for CIA to the inboard MSIV's .and MS relief valves (X-56) and CIA and nitrogen backup to the ADS valves (X-89A and B).

These check valves are in all three cases inboard of motor operated, isolation globe valves. The check valves are located outside of the primary containment to improve valve operability as discussed in Note 17 of Table 6.2-16.

022.043-2

Page 1 of 1 Q. 031.001 (a)

The FSAR contains many conflicting statements and incompre-hensible statements which must be resolved prior to the start of our review. For each of the items below, provide a re-sponse which is responsive to the NRC staff's need for in-formation satisfying the requirements of the "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," (Revision 2).

a. Clarify the discrepancy between the designation of the main generator as "Unit 1" in 1.2.2.7.1 and the plant designation as NNP-2.

RESPONSE

The text of 1.2.2.7 has been revised to clarify this dis-crepancy.

031. 001 (a) -1

Page. 1 of 1 Q. 031.001 (b)

Clarify the discrepancy between- the reference to "three trip logics" in 3.1.2.3.2.1 and the description of the reactor protection system in 7.2.1.1.3.2 and Figures 7.2-3 and 7.2-11.

RESPONSE

3.1.2.3.2.1, third paragraph which refers to "three trip logics" is in error, and has now been revised.

031- 001 (b) -1

Page 1 of 1 Q. 031.001 (c)

Clarify the discrepancy between the 0.25 to 33 Hz seismic range given in 3.10.1.2.3.1 and the 5 to 33 Hz values given in 7.3.2.1.2.3.1.

RESPONSE

Some seismic vibration tests on certain equipment were per-formed starting from 5 Hertz because of test machine limi-tations. High test table displacements beyond the capacity of the testing machine are required to obtain higher input accelerations (g-levels) at frequencies below 5 Hertz. In all these cases it was shown by calculations or by a resonant frequency search test or based on similarities with previously qualified equipment that no resonant frequencies exist below 5 Hertz; consequently, tests are completed to required levels at 33 Hertz and at all resonant frequencies in between. This would qualify the equipment seismically. 7.3.2.1.2.3.1 has been revised to eliminate the discrepancy, and will cross-reference to 3.10 for seismic data.

031. 001 (c) -1

WNP-2 Page 1 of 1 Q. 031.001 (d)

Indicate whether the reference to Figure 7.2-4 which is made in 7.2.1.1.3.2, is intended to be Figure 7.2-3.

RESPONSE

7.2.1.1.4.3 has been revised to provide the correct figure references.

031. 001 (d) -1

WNP-2 Page 1 of 1 Q- 031.001 ( )

Clarify the discrepancies between the main steamline isolation valve (MSIV) response time given in 5.4.5.3, 6.3.3.3.1, and 7.3.2.1.2.3.1.5.2.1.1 and the accident analyses of Chapter 15.

RESPONSE

No discrepancies exist between the stated MSIV response times.

The main steamline isolation valve closure time is in the range from 3 to 5 seconds. As indicated in 7.3.2.2.2.3.1.1 the minimum MSIV closure time is 3 seconds. 5.4.5.3 assumes a maximum MSIV response time of 5.5 seconds.

This assumes the maximum closure time of 5 seconds for MSIV plus .5 seconds for instrument response to initiate MSIV closure. The analyses of Chapter 15 assume the worst case response of 3 seconds for MSIV closure. All closure times given are within the specified 3 to 5 second range of the MSIV.

The MSIV response time for full closure is set prior to plant operation. The minimum set.time for closure is 3.0 seconds and the maximum set time is 5.5 seconds (includes as much as 0.5 seconds for instrument response). The difference in re-sponse time reported in the various paragraphs is due to whether the minimum or maximum response time i's conservative to the results of the analysis under consideration. 5.4.5.3, 15.6.4 list the maximum value while 7.3.2.2.2.3.1.1 lists the minimum value. In summary, the conservative response time. for MSIV closure was used in each of the identified sections of the FSAR.

031. 001 (g) -1

WNP-2 Page 1 of 1 Q. 031. 001 (h)

Clarify the discrepancy between the description of the MSIV solenoid valves which is given on Page 6.2-55 in 6.2.4.2 and that presented in 7.3.1.1.2.2 and Figure 7.3-19.

RESPONSE

7.3.1.1.2.4 and Figures 7.3-2 and 7.3-4 are correct. MSIV solenoid valves are no longer discussed in 6.2.4.

031. 001 (h) -1

WNP-2 Page 1 of 2 Q. 031.001 (r)

The discussion in Chapter 7 regarding compliance with the requirements of 4.20 of IEEE Std. 279-1971 relating to the readout of information, is inadequate. Revise the FSAR to describe the equipment and systems which provide the operator with accurate, complete, and timely information pertinent to the status-of the information channel and to generating sta-tion safety.

RESPONSE

The information required to identify compliance with the requirements of 4.20 of IEEE Std. 279-1971, i.e., descriptions of the equipment and systems which provide the operator with accurate, complete and timely information, is in general pro-vided in the sections titled, "Reactor Operator Information,"

under the appropriate system in the FSAR. The FSAR has been revised to include the appropriate references within the sub-paragraphs specifically addressed to 4.20. Following is a list of paragraphs in Chapter 7, specificallly related to 4.20 and the paragraphs which are referred to:

Paragraph 7.2.2.1.2.3.1.20 (Reactor Protection System) refers to paragraph 7.2.1.1.6.1.

Paragraph 7.3.2.1.2.3.1.20 (ECCS) refers to paragraphs 7.3.1.1.1.3.11.2 (HPCS), 7.3.1.1.1.4.11.2 (ADS) i 7.3.1.1.1.5.11.2 (LPCS), and 7.3.1.1.1.6.11.2 (LPCI) .

Paragraph 7.3.2.2.2.3.1.20 (PCRVICS) refers to paragraphs 7.3.1.1.2.4.1.2.'9, 7.3.1.1.2.4.1.7.9.2, and 7.3.1.1.2.13.2.

Paragraph 7.3.2.3.2.3.1.20 (MSLIV-LCS).

Paragraph 7.3.2.4.3.1.20 (CSCS) refers to 7.3.1.1.4.12.2.

See 7.3.2.5.20 for SSW system information readout.

See 7.3.2.6.20 for main control room HVAC system information readout.

See 7.3.2.7.20 for reactor building ventilation and pressure control system information readout.

See Chapter 8 for standby power system information readout.

See 7.3.2.9.20 for SGTS information readout.

031.001(r) ,1

WNP-2 Page 2 of 2 See 7.3.2.10.20 for CIA system information readout.

See 7.3.2.11.20 for CAC system information readout.

Paragraph 7.4.2.1.2.3.1.20 (RCIC) refers to paragraph 7.4.1.1.5.2 ~

Paragraph 7.4.2.2.2.3.1.20 (SLCS) refers to paragraph 7.4.1.2.5.2.

Paragraph 7.6.2.8.2 (Recirculation Pump Trip System) refers to paragraph 7.6.1.8.5.2.

7.5.2, Safety Related Display System Analysis, contains further discussion of compliance with 4.20 of IEEE Std. 279-1971.

031. 001 (r) -2

Page 1 of 1 Q. 031.001 (u)

Clarify the references in 7.2.1.1.4 to Table 3.11-1 for the reactor and control building environments.

RESPONSE

7.2.1.1.5 of revised text states that the environmental con-ditions for the drywell, the containment, and the turbine building are given in Tables 3.11-1, 3.11-2, and 3.11-3.

Tables 3.11-1, 3.11-2 and 3.11-3 have been revised to provide the referenced information.

031. 001 (u) -1

WNP-2 Page 1 of 1 Q. 031.001 (v)

Clarify 7.3.1.1.2.3 to clearly state where the pressure, temperature, and water level sensors and racks are located.

RESPONSE

The text of 7.3.1.1.2.3 and Table 7.3-46 has been revised to incorporate the response to this question.

031.001(v)-l

WNP-2 Page 1 of 1 Q. 031.001 (w)

Clarify the discrepancy between the discussion of compliance with the requirements of 4.10 of IEEE Std. 279-1971 in 7.3.2.1.2.3.1.5.2.1.10 and the discussion of conformance with the staff positions in Regulatory Guide 1.22 which follows it.

RESPONSE

Even though the mainsteam line high temperature sensors are inaccessible during plant operation, they can be tested while the plant is operating by cross comparison between channels.

7.3.2.2.2.1.2 reads in part as follows:

The main steamline isolation logic, and sensor devices (except the MSL high temperature sen-sors) may be tested from the sensor device to one of the two solenoids. Both solenoids must be deenergized to verify that there are no obstructions to the valve stem at full power.

A reduction in power is necessary to avoid reactor scram before performing a valve closure using two, fast acting, main solenoids.

7.3.2.1.2.3.1.5.2.1.10 is replaced in the revised text of the FSAR by 7.3.2.2.2.3.1.10, which reads in part as follows:

All active components of the primary containment isolation control system, can be tested and calibrated during plant operation with the ex-ception of the main steamline high temperature sensors. By observing the contact action on an HFA type relay during a channel trip condit'ion, the actual drop-out can be verified when deenergized.

031. 001 (w) -1

Page 1 of 1 Q. 031.001 (x)

Several of the figures in the FSAR are illegible or are missing component. designations and are, therefore, unaccept-able. Revise the FSAR to eliminate both illegible and/or unintelligible figures.

RESPONSE

The FSAR has been revised to eliminate illegible or unintel-ligible figures.

031.001(x)-1

WNP-2 Page 1 of 1 Q. 031.001 ( )

Clarify the discrepancy in the response time of the reactor core isolation cooling system given in 7.4.1.1.3.1 and that given in 7.4.1.1.3.5 of the FSAR.

RESPONSE

Both 7.4.1.1.3.2 and 7.4.1.1.3.6 have been revised to provide a 30 second response time for the RCIC system.

031. 001 (y) -1

WNP-2 Page 1 of 1 Q. 031.001 (z)

Clarify the discrepancy between the description in 7.4.1.1.3.5 of the automatic transfer of the suction for the reactor core isolation cooling system which satisfies the design shown in Figure B-25 and the manual transfer shown in Figures 7.4-la and 7.4-2b which does not satisfy the assumptions made in the Operational Analysis contained in Appendix B.

RESPONSE

The operational analysis in Appendix B (now Appendix 15A) did not assume automatic transfer of the RCIC suction. 7.4.1.1.3.5 and Figure B-25 are replaced in the revised text of the FSAR by 7.4.1.1.3.6 and Figure 15.A.6-40 respectively. However, additional discussion about this transfer is provided in the response to Question 031.015.

031. 001 (z) -1

WNP-2 Page 1 of 1 Q. 031.001 (aa)

Resolve the contradiction between 7.4.1.1.3 and 7.4.2.2.2 so as to provide a clear statement of the conditions under which the isolation valves of the reactor core isolation cooling system will be required to operate and the seismic and environmental conditions for which these valves are qualified.

RESPONSE

7.4.1.1.3 and 7.4.2.1.2.3.1.4 have been revised to provide a clear statement of the conditions under which the isolation valves for the RCIC will be required to operate. The valves are qualified for operation during those environmental con-ditions listed in Tables 3.11-2 and 3.11-3. Table 3.2-1 states the seismic classification of the valves to be Category I.

031. 001 (aa) -1

Page 1 of 1 Q 031.001 (bb, cc) bb. Clarify the discrepancy between 7.6.1.1.3.1, which states that the refueling interlock system is single failure proof and the design of the reactor manual control sy-stem which has a single rod position input. path and a single refueling equipment, output path.

cc. Modify the FSAR to include concise definiti'ons of the worst case environmental conditons under which the re-fueling interlocks will be required to operate.

RESPONSE

bb. The refueling interlock system provides two independent channels of instrumentation where either loss of signal or trip signal from either channel will prohibit any further rod movement. The reactor manual control system (RMCS) design has two inputs, and the interlock status is merged into a serial data transmission. Even though there is only one rod position path, this information must be'n a precise format before any rod movement.

it is accepted for Xf the information is in the proper format, it is then determined if the coded information will allow rod movement. The transmitted information is echoed back by the hydraulic control units and compared to ensure proper, transmission. The RMCS pro-vides two refueling equipment output paths.

7.6.1.1.3.1 and 7.6.1.1.3.2 have been revised to clarify the discrepancy.

cc. The refueling platform is not required to operate during the run mode and would not be affected in an accident situation. The worst case environmental conditions under which the refueling interlocks will be required are the normal reactor building. conditions to'perate listed in Table 3.11-3.

031.001(bb,cc)-l

WNP-2 Page 1 of 1 Q. 031.001 (dd)

Clarify the discrepancy between 7.6.1.6.7.1.1.4 and the design of the reactor manual control system which is pre-sented in 7.7.1.1.

RESPONSE

The isolation, separation, and redundancy features discussed in 7.6.1.6.7.1.1.4 are features of the rod block monitor.

The RBM does interface with the reactor manual control sy-s".em, but is separate from it.

7.6.1.6.7.1.1.4 appears in Rev. 2 of the FSAR as 7.6.1.5.7.1.4.

A new subsection has been added which reads:

"7.7.1.2.3.2.3 Rod Block Interlocks The rod block functions are discussed in 7.6.1.5.7."

The reference in 7.7.1 to the appropriate subsection for the RBM will eliminate the discrepancy in 7.7.1.

031.001(dd)-1

WNP-2 (BLANK)

WNP-2 Page 1 of 1 Q. 031.001 (ee)

Identify and justify all design differences between the reference rod block monitor design described in 7.6-2 and that. associated with the new solid state reactor manual controi system which is described in 7.7.1.1.

RESPONSE

The reference (Hatch 2) rod block monitor (RBM) described in Reference 7.6-2 and the RBM associated with the reactor manual control system described in 7.7.1.1 are essentially identical.

The comparison of the reference (Hatch 2) and the Zimmer RBM designs were discussed at length between GE and the NRC on April 18, 1977, at GE, San Jose, Ca. The question is con-sidered to have been resolved for Zimmer.

The designs for the RBM and WNP-2 and Zimmer are identical.

031. 001 (ee) -1

WNP-2 Page 1 of 1 Q. 31.001 (ff)

Identify the specific bus which powers the reactor manual control system and the refueling interlock.

RESPONSE

The reactor manual control system may be powered from either Instrument bus A or B (PP-7A-A or PP-SA-A). Since the system is not safety re-lated, the choice of bus has no effect on per-formance of the system. See 7.7.1.2.2.

2. The busses which supply power to the refueling interlock are 120 VAC instrument busses. See 7.6.1.1.2.

031.001(ff)-1

Page 1 of 1 Q. 031.001 ( )

Quantify the recirculation system low water level interlock range, setpoint, and accuracy.

RESPONSE

The interlock can be set from instrument zero to 60 inches; instrument zero is at. 527.5 inches above invert (vessel bottom) .

The interlock setpoint is at, 31.5 inches above instrument zero.

The accuracy is + 0.5% of full scale.

031.001(gg)-1

WNP-2 Page 1 of 1 Q. 031.001 (hh)

Clarify the discrepancy between 7.7.1.2.3.3.9 and 7.7.1.2.4 of the FSAR with respect to the required motion of-the flow control valves under accident conditions.

RESPONSE

The recirculation flow control system is not required for safety purposes, nor required to operate during or following a design basis accident. However, during operation, the valve actuator has an inherent rate limiting feature that will limit the resulting rate of change of core flow and power to within acceptable limits in the event of an upscale or downscale failure of the valve position or velocity con-trol system.

See revised 7.7.1.3.3.4.9 and 7.7.1.3.4.

031.001(M1)-1

Page 1 of 1 Q. 031.001 (ii)

Indicate where in Chapter 7 the information concerning the instrumentation for the reactor building closed cooling water system is located.

RESPONSE

The text of 7.6.1.15 has been added to respond to this question.

031. 001 (ii)-1

Page 1 of 1 Q. 031.002 Tl.7-1 Provide process instrumentation and control. logic, wiring, and electrical schematic drawings for: (a) the overpressur-ization px'otection (relief) system; (b) the reactor pxotec-tion system; (c) the leakage detection temperature monitors; (d) the neutron monitor auxiliary trip units; and (e) the reactor protection system instrument, racks which are located in the reactor building.

RESPONSE

The following WNP-2 drawings are applicable:

GE MPL Drawing System Number Iden't'i'f ica't'ion A. ADS Logic (FCD) B22 1030 731E788 ADS Elementary B22 1060 807E180TC NBS P 6 ID B22 1010 732E103 B. RPS" IED" C72 1010 732E170AD RPS Elementary C72 1050 807E178TC C. Leak Detection IED E31 1010 732E191AD'07E154TC Leak Detection Elementary E31 1050 D. Neutron Mon. Aux. Trip Schematic 127D1861 Neutron Mon. Aux. Trip Wiring 195B9206 E. RPS Inst. Racks (Not in FSAR, but, available and auditable at GE)

Reac Wtr Lvl and Press Panel H22 P004 127D1827TC Connection H22 P005 127D1827TC Reac Wtr Lvl and Press Panel H22 P026 828E390TC Connection Reac Wtr Lvl Press Panel H22 P027 828E387TC Connection Logic diagrams and elementaxy (electrical schematic dxawings) for BOP systems are listed in Table 1.7-1.

031.002-1

WNP-2 Page 1 of 1 Q. 031.005 (RSP)

(3. 10)

(3.11.3)

(7.2.2.2)

A request for documentation of the seismic and environmental qualification of Class 1E equipment is contained in 3.10 and 3.11 of the Standard Format. This request is applicable to all engineered safety features, reactor protection systems and all supporting systems. It is not limited to those particular safety-related supporting systems supplied by General Electric. Accordingly, we require you to provide the information requested in 3.10 and 3.11 of the Standard Format for all Class lE systems in accordance with: (a) the NRC Staff positions stated in Attachments 1 and 2; (b) IEEE Std. 323-1971; and (c) IEEE Std. 344-1971. Identify and justify any exceptions.

RESPONSE

The requested seismic and environmental qualification data for Class IE equipment for all engineered safety features, reactor protection systems and all support systems is provided in the revised'ext and tables of 3.10 and 3.11.

For additional discussion refer to 7.2.1.2.'7 (RPS), 7.3.1.1.1 (ECCS), 7.3.1.1.2.12 (PC and RVIS), and 7.3.1.1.4.11 (RHR/

Containment Spray).

Additional discussion of conformance to IEEE 323-1971 is pro-vided in 7.2.2.1.2.3.4 (RPS), 7.1.2.5.4 and 7.3.2.1.2.3.1.4 (ECCS), and 7.3.2.2.2.3.1.4 and 7.1.2.5 (PC and RVIS) .

Additional discussion. of conformance to IEEE 344-1971 is pro-vided in 7.2.2.1.2.3.7 (RPS), 7.3.2.1.2.3.5 (ECCS), and 7.3.2.2.2.3.5 (PC and RVIS) .

031.005-1

WNP-2 AMENDMENT NO. 1 0 July 1980 Page 1 of 2 Q. 031.006 The staff requests that the following information regarding the qualification test program be provided for Class 1E equipment: (a) the equipment design specification require-ments; (b) the test plan; (c) the test set up; (d) the test procedures; (e) the acceptability goals and requirements; and (f) the test results.

Provide this information for each of the following Class 1E components: (a) the 4.16 kV switchgear SM 7; (b) the damper operator for WMA-V-52C; (c) the fan WMA-FN-52B; (d) the logic equipment for the standby gas treatment system; (e) the diesel-generator control equipment; (f) the 480 V ESS switch-gear MC-7A-A; and (g) the solenoid valve for the main steam line isolation valves.

~Res onse:

An extensive seismic and environmental review program is pres-ently underway encompassing BOP and NSSS scope, with a planned completion date in December 1980.

Within the BOP scope,. the equipment documentation has been extracted from the contract files, copied and categorized for easy retrieval. Within the NSSS scope, contract negotiations are underway with GE to perform a similar function.

A list of all Class 1E equipment including splices, terminal blocks, termination cabinets and connectors is presently being compiled. This list will contain the following information:

1. Equipment location

2. Safety functional requirement

3. Manufacturer 6 Model No.

4. Qualification Method (test-analysis)

5. Environmental Extremes

6. Identification and location of qualification documents 031.006-1

WNP-2 AMENDMENT NO. 10 July 1980 Page 2 of 2 The documentation will be reviewed to insure that the testing was adequate to meet the seismic and environmental extremes under which the equipment must either function or not fail.

A composite list will be included in the FSAR as equipment tables in 3.10 (seismic) and 3.11 (environmental).

The extensive review program underway will also satisfy the requirements of IE Circular 78-08, address the degree of compliance with NUREG-0588, and establish the conservativeness of seismic tests and analysis performed to IEEE-344, 1971.

The detailed results of this review will be made available to NRC SQRT and environmental review personnel during their site documentation reviews.

031.006-2

WNP-2 (BLANK)

Page 4 of 4 (f) The drive flow demand limiters are adjustable. The high signal limiter is to establish the maximum drive flow demand limit needed for the upper end of the automatic load-following range. The low signal limit is determined from a core stability criterion and defines the lower end of the automatic load-following range. There is no flow limit, and the valve can be closed to its minimum position when the master controller is in manual mode operation. (See 7.7.1.3.3.4.4).

(g) A limiting function is provided for one feedwater pump trip flow runback. An electronic limiter with reason-able range adjustment is provided in each main flow control loop. This limiter is normally held bypassed by auxiliary devices such as relay contacts. When one feedwater pump trip coincides with reactor low water level alarm, the main regulating valve control signal is limited to close the valve to the desired position.

(See 7.7.1.3.3.4.8).

See also 5.4.1.3.1.

031.010-4

WNP-2 Page 1 of 1 Q. 031.011 6.2.2.3 The description of operator actions is incomplete. Provide the following information: (a) describe the operator actions which are necessary to establish containment spray; and (b) describe the design provisions, if any, which prevent the operator from shutting down a pump too early or diverting flow too soon.

RESPONSE

(a), (b) Discrepancies have been present in the WNP-2 FSAR (see Question 031.001 (q)) in describing the operation of the containment spray system. These discrepancies have been corrected. One can now obtain the operator actions, procedural controls and design provisions for containment spray oper-ation in 6.2.2.2, 6.5.2.2 and 7.3.1.1.4. These sections have been revised. Figure 7.3-16 has been updated to show the correct initiation signals for containment spray.

031.011-1

Page 1 of 1 Q. 031.012

( .2. .2 Describe the pump motive source which is used to provide feedwater flow after the main steamline isolation valves close and which forms the basis for the assumptions in 6.2.4.2.1.2 of the FSAR.

RESPONSE

The text of 6.2.4.3.2.1.1.1 has been revised to incorporate the response to this question.

031.012-1

Page 1 of 1 Q. 031.013 (RSP) 6.3.1.3)

(7.3.1.1)

Provide justification for not testing the emergency core cool-ing system flow rate and the associated sensing networks dur-ing normal operation. Define the term "sensing network" as used in 6.3.1.3 of the FSAR. Identify each network which cannot be tested during normal operations. It is the staff's position that the WNP-2 design should provide engineered safety feature circuits which satisfy the guidance contained in Regulatory Guide 1.22. Accordingly, we require you to provide a revised design which conforms with the staff's position on this matter and to provide a description of these sensors and networks which provides the information requested in 7.3 of the Standard Format. Clarify the discrepancy between 6.3.1.3 and 7.3.1.1.1.2.1.2 with regard to the testability of the emergency core cooling system.

RESPONSE

6.3.1.3 was in error. The text of 6.3.1.1.2(m) and 7.3.2.1.2.3.1.10 have been revised to reflect that all active components of the ECCS are testable during normal operation.

031.013-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 0 31.014 (6.3.2.2)

(6.3.2.8)

(F 6.3-la)

The location of sensors LS N001 A, B, C, and D, as shown in Figure 6.3-1a, does not appear to meet Seismic Category I requirements. Revise the design of the WNP-2 to assure that.

the sensors controlling the transfer of suction to the suppression pool will be seismically and environmentally qualified for their location and environment.

~Res ense:

Condensate storage tank pressure sensors used for level switches are designed and qualified to Seismic Category I requirements and are environmentally qualified.

The pressure sensors will be mounted on the interior side of the concrete fluid retaining walls surrounding the condensate storage tanks. The sensors will be located such that postu-lated failures of the condensate storage tanks will not com-promise the sensors. The sensors are designed and located to withstand natural phenomenon, e.g., tornados and high winds, and will be freeze protected.

031.014-1

Page 3 of 3 However, in order to automate the manual transfer aspects of RCICs water source in the unlikely event the condensate storage tank inventory is unavailable for use if called upon, the applicant is modifying the present design to include the following features:

1) Automatic transfer circuitry equivalent. to HPCS auto transfer system will be provided.

2) Condensate storage tank site natural phenom-ena considerations will be taken into account in order to assure that the automatic trans-fer function is not negated.

The above plant modification is similar to that proposed for the Zimmer plant.

(b) The RCIC System has automatic initiation and isolation, and manual initiation and isolation.

Compliance with IEEE 279-1971, paragraph 4.17, is in 7.4.2.1.2.3.1.17.

(c) The design compliance for the present system is presented in 7.4.2.1.

(d) Since a, b, 6 c have been responded to, a response here is not applicable.

031.015-3

Page 1 of 2 Q. 031.016 (RSP) 6.3.3.

(7.0)

(7.2)

(15)

The discussions of response times and the testing of response times in 6.3.3.9, 7.2 and elsewhere in Chapter 7, are inade-quate. Chapter 15 does not provide the response times, accu-racies nor ranges for instrumentation and control systems which were assumed in the accident analyses. It is the staff's position that the design of systems which are required for safety shall include provisions for periodic verification of the minimum performance of instruments and controls that are not less than those assumed in the safety analyses. The bases for this position are General Design Criterion 21, 3.9 of IEEE Std. 279-1971, and IEEE Std. 339-1971. Accordingly, we require you to demonstrate:

a. The capability to periodically verify the mini-mum performance characteristics of all systems, including the appropriate procedures, which are required for safety. Testing shall include the entire system from, and including, sensor to actuator output.

b. Compliance with Branch Technical Position 24 of Appendix 7-A of the Standard Review Plan.

c. Compliance with the requests for information in 15 of the Standard Format.

RESPONSE

a The Technical Specifications for WNP-2 will re-

~

quire periodic response time testing of the C & I protection and safeguards systems from sensor through actuation systems. The procedures for initial system test and periodic surveillance testing will be developed (see 14.2.12.1.18 and Chapter 16) by WPPSS.

031. 016-1

Page 2 of 2

b. The requirements noted in a. comply with Branch Technical Position 24 of Appendix 7-A of the Standard Review Plan.

c. Chapter 15.0 is not the appropriate location for sensor response times, accuracies, and ranges for instrumentation and control system. Tables 7.2-1, 7.3-1, 7.3-2, 7.3-3, 7.3-4 and 7.4-1 have been revised to provide instrument ranges and accuracies on reactor protection systems and ESF systems instrumentation. Trip settings and re-sponse times used in Chapter 15.0 analysis have been submitted.

031.016-2

WNP-2 BLANK PAGE

WNP-2 Page 1 of 1 Q. 031.019 7.1.2.1 Demonstrate how the design of the main steamline isolation valve leakage control system satisfies the requirements of 4.19'.20, and 4.21 of IEEE Std. 279-1971.

RESPONSE

The main steamline isolation valve leakage control system satisfies the requirements of IEEE Std. 279-1971, 4.19, 4.20, and 4.21 as follows:

Identification of Protective'ction's (IEEE 279, ara ra h 4.19)

Initiation of the MSLIV-LCS is indicated in the control room.

Information Read-Out (IEEE 279, a'ra ra h 4.20)

Meters located in the control room provide indication of process variables necessary for the proper operation of the MSLIV-LCS. Indicator lights actuated by valve position switches provide valve position indication.

S stem Re air (IEEE 279-1971, ara ra h 4.21)

The system is designed to provide easy recognition of mal-functioning equipment through proper test procedures. Accessi-bility is provided for the sensors and, controls to facilitate repair or adjustment. MSLIV-LCS isolation valves are located in the steam tunnel and repair is made during shutdown.

See,also'.3.2.3.2.3.1.19, 7.3.2.3.2.3.1.20 and 7.3.2.3.2.3.1.21.

031.019-1

Page 1 of 1 Q. 031.020 (T7.1-1)

Table 7.1-1 is incomplete. Complete this table= by filling in all of the blank spaces so as to indicate if the cited systems are a design which is unique to the WNP-2 facility or is similar to those used on o'ther nuclear power plants.

RESPONSE

Table 7.1-1 has been replaced by revised Table 7.1-2.

031.020-1

WNP-2 Page 1 of 2 Q. 031.021 (Tj. 2-1) (T7. 3-1)

(Tj. 3-2) (T7. 3-3)

(T7. 3-4) (T7. 4-1)

The information which is presented in Table 7.1-6 and 7.2.1.2.9 is inadequate. The trip settings and margins for some instruments are missing as well as the range and accu-racy for other instruments. Additionally, the units of measurement for other instruments are missing while the units of measurements for other instru'ments are not consistent with the units for the setpoint. Provide the following informa-tion:

a. All trip settings which are required for safety, expressed in units which are consistent with the instrumentation range.

b. All ranges and accuracies for all instrumentation systems which have a safety function or which provide required safety system support.

c. The design criteria used in establishing the required range of instruments in the reactor protection systems, engineered safety feature systems, and other safety-related systems.

d. The design criteria used in determining which portion of the range of an instrument may be

'used'or automatic initiation of a protective function.

e. Where trip settings are to be based on operating experience, state the initial values which are to. be used and provide the bases.

f. Response times, provide both required and calculated.

Our concern is that, in previous designs of nuclear power plants, the setpoints have either drifted beyond the range of operability of the sensor or sensor foldover has occurred because the. setpoints were initially set too close to the extreme ends of the instrument range.

031. 021-1

WNP-2 Page 2 of 2

RESPONSE

The Tables provided for the following answers include instruments in addition to those required for safety.

(a, b, and f) The requested information is contained in Tables 7.2-1, 7.3-1 through 7.3-5, and 7.4-1.

Tables 7.2-1, 7.3-1, 7.3-2, 7.3-3, 7.3-4, 7.3-5, and 7.4-1 have been revised to provide nominal trip settings, instrument ranges, accuracy, and response times. Margin will be provided later, as available.

(c) The range for safety related instrumentation is selected so as to exceed the. expected range of the process variable being monitored. This may, as in the case of the neutron monitoring system, require more than one instrument to cover the expected range.

(d) The trip setpoint is located in that portion of an instrument's range which provides the re-quired accuracy. For example, although the drywell high pressure setpoint, is near the low end of the instrument's range, setpoint drift downward would be in a conservative direction.

Also, this instrument has been proven in other BWR's.

(e) Initial trip setting values are established from operating experience with similar size plants, and backed up with analysis as necessary.

The initial values are the trip settings listed in the tables referenced in (a, b, and f) .

See the response to Q. 031.037 for a discussion of setpoint drift. Revised 7.1.2.4 and the response to Q. 031.001(r) provide additional discussion regarding this question.

031. 021-2

WNP-2 BLANK PAGE

WNP-2 Page 1 of 1 Q. 031.022 (7.2.1.1) 7.2.1.1.3.1.1 of the FSAR is incomplete. Provide a description of the source range trips which provide rod block and scram protection for the initial startup of new cores.

RESPONSE

In addition to 7.2.1.1.4.2, the source range trips which pro-vide rod block and scram protection for the initial startup of new cores are described in 7.6.1.6. Functions for rod blocks are provided in 7.6.1.6.2. Figures 7.6-15a, 7.6-14 and 7.2-1 also provide descriptive information.

For the initial fuel load, high-high trip contacts from each SRM are combined to produce a 1:4 non-coincident reactor trip through the manual scram portion of the circuit. Following the initial fuel loading and startup, these latter contacts are permanently shorted to remove the SRM reactor trip func-tion.

031. 022-1

Page 1 of 1 Q. 031.027 (7.3.1.1)

The description of low pressure interlocks in 7.3.1.1.1.5 of the FSAR is incomplete. Describe the parameter sensed and the function of MO E12-F087, MO E12-F052, and MO E12-F051.

RESPONSE

7.3.1.1.1.7 has been revised to include:

RHR S stem T~e Valve Parameter Sensed Function Steam MO E12-F087 Steam pressure Provide low-Condensing pressure Mode supplementary flow MO E12-F052 None Block valve AO E12-F051 Steam pressure Maintain system pres-sure 031.027-1

WNP-2 Page 1 of 1 Q. 031.028 (RSP)

(7.3.1.1)

(7.3.2.1)

The discussion of environmental conditions such as that in 7.3.1.1 and 7.3.2.1 is unacceptable. It is the staff's position that all safety-related equipment, including cables, must be qualified for operation in the worst case environ-ment. Inside the containment, this design basis environment is established by postulated accidents. Equipment outside of containment must be qualified to the extremes of expected conditions which could result from the failure of other engineered safety features or equipment required to maintain a controlled environment such as plant heating systems.

Accordingly, we require you to demonstrate compliance with this staff position. Identify and justify all exceptions.

RESPONSE

All safety-related components including cables which are supplied by GE whether located in the containment or outside the containment are selected to meet the environmental condi-tions in 3.11. There are no exceptions taken.

See also revised 7.3.1.1.5.3, 7.3.1.1.6.3, 7.3.1.1.7.3, 7.3.1.1.10.3 and 7.3.1.1.11.3.

Qualification of balance of plant cables and components is covered in 3.11 and 8.3.1.2.3.

031.028-1

WNP-2 Page 2 of 2

b. The standard temperature for the control room instrumentation is 40-120oF and 90% relative humidity (maximum) . The range of temperatures and humidity over which the LPCI, LPCS, HPCS, ADS, and MSIV-LCS instrumentation and controls will meet their design basis is provided in Table 3.11-1, 3.11-2 and 3.11-3.

c. Probable maximum floods have no effect on Class lE systems (see 3. 4) .

d. Instrument response times used in the NNP-2 simulation in Chapter 15.0 will be provided in the response to Question 031.016.

Table 7. 2-1, 7. 3-1, 7. 3-2, 7. 3-3, 7. 3-4, 7. 3-5, and 7.4-1 will be revised to include instrument accuracies in response to Question 031.021.

e. The differential pressure sensors (level switches and hP transmitters) are designed for one side pressurization capability of up to 2000 psig without damage to diaphragm bellows.

031. 030-2

WNP-2 BLANK PAGE

WNP-2 Page 1 of 1 Q. 031.031 (7.3.2.1)

The description of the automatic depressurization system with respect to the requirements of IEEE Std. 279-1971, 4.19 and 4.20~ is inadequate. Provide the following information:

(a) Describe how the operator is made aware of items (a) through (d) under the discussion of compliance with 4.19 of IEEE Std. 279-1971.

(b) Provide justification for the use of the relief valve discharge pipe monitors and plant annunci-ators for pr'oviding information which forms the basis for operator action.

(c) Define the term "ADS level".

RESPONSE

1. Identification of Protective Actions is dis-cussed in 7.3.2.1.2.3.1.19.

2. The ADS is a backup system to HPCS. Its function is to depressurize the reactor automatically in the event of a LOCA if the HPCS system fails to maintain vessel water level. Depressurization allows the low-pressure ECCS to do their job.

No operator action is required. However, if, based on water level indications in the control room, the operator determines that the HPCS can restore water level without the aid of low-pressure systems, a reset switch is available to delay ADS initiation for 120 sec. The operator decision to delay ADS is made by looking at vessel water level indication and HPCS flow and pressure indication provided in the control room. See 7.3.1.1.1.4.5.

3. "ADS levels" refers to vessel water level indi-cations which are applicable permissives for ADS initiation. That is, ADS initiates on vessel low water level Trip 1 and Trip 3.

See 7.3.2.1.2.3.1.20.

031.031-1

WNP-2 Page 1 of 1 Q. 031.032 (RSP)

(7.2.1.1)

(7.3.2.1)

(15.2. 6)

(F7. 2-Z,)

(T15.2-13)

It is the staff's position that the use of Class 1E power supplies as alternate feeds for the reactor protection system buses as described in 7.2.1.1.2 of the FSAR is unacceptable since this prevents the required separation from the third division at level switch 1B21-N024A and N024C (GE Drawing 828E479TU). Accordingly, we require you to provide a re-vised design for the alternate power source for the reactor protection system which satisfies your electrical separation criteria. Identify and justify all exceptions. Clarify the discrepancy between the present design and the assump-tions of 15.2.6 and Table 15.2-13.

RESPONSE

Drawing 828E479TU is not applicable to WNP-2. See Figure 8.3-2.

031.032-1

Page 1 of 1 Q. 031.035 (7.3.2.1)

Describe the design features which provide assurance that. the main steamline isolation valves do not close in less than 3 seconds.

RESPONSE

The.MSIV does not close in less than 3 seconds due to an in-line hydraulic damper restricting the closure speed. A small vernier type control limiting the hydraulic flow from below the damper piston allows accurate speed setting of valve closure. This speed control valve (g7 of Figure 7.3-4),

having the capability to vary the MSIV closure speed from 3 to 10 seconds, can be set and locked for the desired speed.

The control valve is set, during production testing. MSIV closure speed are re-determined at pre-operational tests and plant shutdown surveillance tests and control valves re-set as required.

031.035-1

Page 1 of 1 Q. 031.036 (7-3.2.1)

(F7.3-8b)

Provide a drawing for the test control logic which shows how the main steam line isolation valve is tested for the re-c{uired response time limit (e.g., greater than 3 seconds but less than 5 seconds) at .rated steam flow.

RESPONSE

Figure 7.3-11 shows the test control logic for testing the MSIV closure time. In order to avoid scram during MSIV full closure testing, the reactor power is reduced to approxi-mately 75%.

The control room operator may time the valve closure while observing valve status lamps.

031.036-1

WNP-2 Page 1 of 1 Q. 031.037 (7.3.2.1 The statement that "All components used in the isolation sys-tem have demonstrated reliable operation in similar nuclear power plant protection system or industrial application," is unacceptable since: (a) this statement does not satisfy the requirements of IEEE Std. 323-1971; and (b) considerable problems have been experienced with sensor drift. Provide an amended discussion of compliance with the requirements of 4.4 of IEEE Std. 279-1971 which satisfies the requirements of IEEE Std. 323-1971 and which describes the methods used to reduce sensor drift to acceptable levels.

RESPONSE

Components were chosen, on the basis of being the best avail-able, vendor test and specs, and proven operational use in similar application. Operational experience has indicated some sensor drift, however, the frequency of surveillance check, test, and calibration and use of historical instrument data has consistently kept sensors within the limits of safe operation.

Conformance to IEEE Std. 323-1971 is discussed in 7.1.2.5.4.

Sensor drift is discussed in 7.1.2.4.

031.037-1

Page 1 of 1 Q. 031.038 (7.3.1.1)

(7.3.2.1)

Clarify the description of the temperature monitoring circuits which initiate containment and reactor vessel isolation. It is the staff's understanding that the BWR/5 and BWR/6 plants are equipped with a system using thermocouples. Include the following information in this clarification:

a. Describe the monitor system in the manner re-quested in 7.3 of the Standard Format.

b. Describe how the system satisfies the require-ments of IEEE Std. 338-1971 and General Design Criterion 21 and how it in Regulatory Guide 1.22.

conforms to the guidance

c. Clarify the discrepancy between 7.3.1.1.2.4.1.12.2.1 and 7.3.2.1.2.3.1.5.2.1.10 of the FSAR.

RESPONSE

(a) Main steamline space temperatures and differ-ential temperatures are measured by dual-element thermocouples, and the analog signals are transmitted from the sensors to the tempera-ture switch point modules located in the control room. This temperature measurement circuit arrangement is similar to Zimmer but not similar to Duane Arnold. Duane Arnold uses bimetallic temperature switches which are locally mounted.

'See 7.3.1.1.2.4.1.3.

(b) The main steamline space temperature detection system, can be tested during reactor operation.

Operability of the sensors (thermocouples), can be verified by comparing the readings during reactor operation. A complete check of the system, including the sensors, can be made during refueling and other planned shutdown periods.

(c) Responses to items (a) and (b) help to clarify any conceived discrepancies.

See 7.3.2.2.2.3.1.9, 7.3.2.2.2.3.1.10, 7.3.2.2.2.3.4, and 7.3.2.3.2.2.1.

031.038-1

Page 1 of 2 Q. 031.039 (RSP)

(7.3.2.2)

The methods proposed for testing of some safety functions are unacceptable. Accordingly, we require you to provide modified designs for all of the safety-related instrumentation and control systems so that these designs will satisfy the fol-lowing staff positions. Identify and justify any exceptions.

a. All portions of the protection systems shall be designed in accordance with IEEE Std. 279-1971, as required by 10CFR50.55a(h) . All actuated equipment that is not tested during reactor operation, should be identified and a dis-cussion of how this equipment conforms to the guidance contained in paragraph D.4 of Regula-tory Guide 1.22, should be submitted.

b. The use of jury-rigged bypasses such as tempor-ary jumpers, the removal of fuses, or removal of connectors is not an acceptable method for standard in-service testing.

c ~ The containment isolation valves for the WNP-2 facility shall be tested from the sensors through the actuating circuits and the valves themselves.

d. The methods which are provided for the testing of protection systems shall satisfy the require-ments of 4.11 of IEEE Std. 279-1971.

RESPONSE

a. Relative to IEEE 279-1971 and Regulatory Guide l.'22, which requires that actuated equipment be tested during reactor operation, all of the actuated equipment has the capability to be tested during reactor operation.

b. it In no instance will be necessary during test-ing of these circuits to either 'lift leads or remove fuses.

031. 039-1

WNP-2 Page 2 of 2

c. All isolation valves can be tested from the sensor through the actuating circuits.

d. See also revised 7.2.2.1.2.3.1.11, 7.3.1.1.5.3, 7 3 1 1 6 3I 7 ~ 3 1 1 7 3g 7 ~ 3 1 1 10 3f 7.3.1.1.11.3 and 7.3.2.2.2.3.1.11.

031.039-2

WNP-2 (BLANK}

Page 1 of 1 Q. 031.040 (7.3.2.1)

(7.4.2.2)

Provide justification for your position stated in 7.3.2.1.2.3.4 and 7.4.2;2.2.3 of the FSAR that only 2.1 and 2.2 of IEEE Std. 338-1971 are applicable to the design of the emergency core cooling system and the reactor core isolation cooling system. Identify and justify all exceptions to IEEE Std. 338-1971.

RESPONSE

The statements in the subject paragraphs are in error and have been revised to state that the ECCS and RCIC comply fully with the requirements of IEEE 338-1971.

031.040-1

Page 1 of 1 Q. 031.041 (7.3.2.4)

(7. 4)

The discussion in 7.3.2.4.4 and 7.4 of the conformance of the WNP"2 design with the present Regulatory Guides, is incomplete.

Revise these sections of the FSAR to include a discussion of how the WNP-2 design conforms with Regulatory Guide 1.29.

RESPONSE

NRC Regulatory Guide 1.29 has been addressed in revised sections of the FSAR. See revised 7.1.2.6.6.

031. 041-1

Page 1 of 1 Q. 031.042 F7.3-8b Revise all FSAR figures (such as Figure 7.3-8b) to include alpha-numeric area locators if. such figures are referenced by, or continued on, additional sheets or figures.

RESPONSE

FSAR figures have been revised to include alpha-numeric area locators.

031.042-1

WNP-2 Page 2 of 2 LPCI: In no event can failure of an automatic control circuit for equipment in one division disable the manual electrical control circuit for the other LPCI division. Single electrical failures cannot disable manual electrical control for the LPCI function. LPCI A has an armed manual initi-ation pushbutton in parallel with the automatic initiation logic which will also initiate LPCS.

The LPCI B and C systems have an armed manual initiation pushbutton in parallel with the auto-matic initiation logic.

Refer to 7.3.2.1.2.3.1.17 for the above descrip-tion.

(BLANK)

Page 1 of 1 Q- 031.045 (7.4.1.2)

It is the staff's understanding that the heating system for the standby liquid control system has been redesigned and now provides redundant heaters which are powered from Class lE buses. Accordingly, we request that you: (a) revise 7.4.1.2.2; and (b) provide process and instrumentation draw-ings and electrical schematics which include the revised heating system and its controls.

RESPONSE

a. The standby liquid control system heaters have not been redesigned. The system remains with two heaters, one of which is used for initial heating (when rapid heating is required), the other used to maintain the SLCS solution at required temperature. The heaters and their controls are not required for the initiation of the SLCS.

7.4.1.2.2 is correct and requires no updating.

b. There are no revised heating system drawings for the SLCS.

See Figure 7.4-3 and 7.4-4.

031.,045-1

WNP-2 Page 1 of 1 Q. 031.0'46 (7.4.2.4)

You indicate in 7.4.2.4.2 of the FSAR that the only Regulatory requirements applicable to the design of the residual heat removal (RHR) system are General Design Criteria (GDC) 34 and 61. However, GDC 34 states, in part, that the residual heat. removal system has a safety function and requires that..."suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities" be provided. Accordingly, revise your design to provide a leak detection capability and provide the .

appropriate information to demonstrate that the RHR system satisfies all the requirements of the General Design Criteria.

RESPONSE

3.1.2.4.5 contains descriptive material which provides a dis-cussion of RHR system compliance with General Design Criterion

34. 5.2.5 contains a discussion of the Leak Detection System and its application to the RHR System. 15.2.9 discusses a backup method for disposing of residual heat should the normal shutdown line become unavailable during shutdown.

031.046-1

AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 0 31.047 (7.5.2.4)

The seismic qualification of indicators and recorders for post-accident monitoring which is described in 7.5.2.4 is unacceptable. It is the staff's position that post-accident.

indicators and recorders must meet their minimum performance requirements before and after a seismic event without re-quiring adjustments*or repair. (The staff acknowledges that these electro-mechanical devices may not provide accurate readings during severe vibrational excitation). Accordingly, we require you to provide a revised design which satisfies the staff's position on this matter.

~Res ense:

The indicators and recorders are nonseismic. At the time WPPSS-2 was being designed, there was no IEEE standards or Regulatory Guide requirements for design of the subject instrumentation.

The instrumentation and readout devices are of high quality and from well-known manufacturers. This similar instru-mentation is used for post-accident monitoring in such licensed and operating plants as Duane Arnold and Brunswick 2, and in such plants as Zimmer and LaSalle presently in the late stages of review. Therefore, it is the position of the General Electric Company that the instrumentation provided is adequate.

031.047-1

WNP-2 Page 1 of 3 Q. 031.048 (7.4.1.4)

Identify the systems and functions controlled by the transfer switches referenced in 7.4.1.4.3e of the FSAR. Indicate the location of these switches.

RESPONSE

The transfer switches are located on panel C61-P001 (Remote shutdown panel) which is located outside the main control room. Selection of the location is based upon having no effect on the panel from the control room evacuation event.

See also revised'7.4.1.4.4.6.

The systems and functions controlled by the transfer switches are:

Reactor Core Isolation Coolin (RCIC) S stem The following RCIC System functions shall have control and transfer switches located at the remote shutdown control panel:

E51-F010 Motor operated valve (pump suction from condensate storage)

E51-F013 Motor operated valve (RCIC injection shutoff)

E51-F019 Motor operated valve (minimum flow to sup-pression pool)

E51-F022 Motor operated valve (test bypass to conden-sate storage)

E51-F031 Motor operated valve (pump suction from sup-pression pool)

E51-F045 Motor operated valve (steam to turbine)

E51-F046 -'Motor operated valve (lube oil cooling)

E51-F063 Motor operated-valve (steam supply line inboard isolation E51-F064 Motor operated valve (RHR heat exchanger steam line isolation) 031.048-1

Page 1 of 1

g. 031.051 0 ~ 6)

(T7. 1-1)

The descriptions of the systems and components presented in 7.6 of the FSAR are inadequate. Provide the information requested in 7.6 of the Standard Format, including a dis-cussion of all differences between the designs of these systems and BWR-5 designs such as Zimmer and LaSalle.

RESPONSE

The description of the systems and components present in 7.6 of the FSAR has been rewritten. Similarity to licensed reactors is illustrated by revised Table 7.1-2.

031. 051-1

Page 1 of 1 Q. 031. 052 (RSP)

(7. 1. 2. 1)

(7.6.2.6) ~

(15.4.2.2)

(T7.1-2a)

The accident analysis presented in 15.4.2.2 of the FSAR is based, in part, on the assumption that the rod block monitor (RBM) acts to mitigate the consequences of a continuous with-drawal of a control rod. Accordingly, it is the staff's position that the RBM is a protection system and must be de-signed, fabricated, installed, tested, and subjected to all of the criteria applicable to a reactor trip system. Ac-cordingly, we require you to revise your design to reflect the importance of the RBM. Identify and justify any excep-tions.

RESPONSE

The RBM is used to prevent the operator from withdrawing a rod in the power range so that fuel cladding integrity is always maintained.

separated, and isolated.

It has two channels that are redundant, It provides two alarms and then a block as the power goes up locally.

The rod block monitor (RBM) is integral with the neutron monitoring system which is presently identified in 7.1.2.1.4.

The RBM design basis is included in 7.1.2.1.4.6. The RBM is adequately described in terms of initiating circuits, logic, bypasses, interlocks, redundancy, diversity, and actuated devices in 7.6.1.5.7. It is to be emphasized that, the WNP-2 RBM is essentially identical to the RBM which has been licensed previously on all plants from Duane Arnold through Hatch-1 and Brunswick. The design of the RBM for WNP-2 is identical to the design of the RBM for Zimmer.

The Chapter 15 and Appendix 15A safety analysis involving the continuous rod withdrawal error, presents the event as an anticipated operational transient.

The RBM is not used in any Design Basis Accident (DBA) cate-gory event described in Chapter 15.

031. 052-1

Page 1 of 2 Q. 031.053 (7.7.1.1 (15)

(F7.7-2)

The reactor manual control system is assumed in Chapter 15 to function to mitigate or prevent several accidents. Therefore, it appears that the reactor manual control system, the reactor (sic) position indicator system and the reactor (sic) sequence control system are part of the reactor protection system.

Accordingly, provide the following additional information:

a. Provide the design bases and other appropriate information in the manner requested in 7.2 of the Standard Format for these three subsystems.

b. Provide a complete description of the scram time test, panel, in the manner requested in 7.7.2 of the Standard Format, including schematic and wiring diagrams.

RESPONSE

(a) The rod block monitor (RBM), the reactor manual control system (RMCS), the rod position indi-cation system (RPIS), and the rod sequence control system (RSCS) are not utilized to mitigate the consequences of accidents, and are not part of the reactor protection system (RPS) .

General Design Criterion 20 does not define these systems as reactor protection systems.

The RSCS does not initiate a scram or isolation

.signal and is therefore not a protection system.

It does, however, perform the power generator function of imposing strict limits on control system operation to prevent, out-of-sequence rod patterns. The allowable rod patterns are such as to limit the peak enthalpy in the fuel to less than 280 cal/g. The RBM considera-tions are discussed in response to Question 031.052.

031. 053-1

WNP-2 Page 2 of 2 (b) No functional changes in Figure 7.7-2 are anticipated in answer to Question 031.053.

(c) The rod scram time test panel is provided as a common location for obtaining control rod scram travel time data for all rods. A base for each time the rod passes an odd-numbered position is transmitted from the rod position information cabinet through a buffer to the scram time test panel plug-in jacks. Each jack carries infor-mation from 'one rod. Power for these on-off signals is provided by a power supply in the scram time test panel. A 29-channel recorder can be connected by patch cords to any 29 or fewer of the 137 plug-in jacks to record on a moving chart the passage of the rods through the odd-numbered positions. Scram time can be determined from the chart speed and pulse posi-tion rod. Each plug-in jack has an indicator above a

it to indicate which rod or rods are in test condition at the hydraulic control units.

This information is transmitted through the display memory module to the rod scram time test panels.

A contact of a relay across the scram backup solenoid is used to start the recorder chart.

A time-delay stops the chart after 20 seconds.

The schematic diagram is on Sheet 7 of 807E183TC, and the scram time test panel con-nection diagram is 828E183TC.

031. 053-2

WNP-2 (BLANK)

Page 1 of 1 Q. 031.054 7~7~1 ~

(8.3.1.1) ll The material presented in 7.7.1.11 of the FSAR is inadequate.

Provide the following additional information:

a0 Clarify the discrepancy between the statement in 7.7.1.11 that the reactor water cleanup system is fed from the plant instrumentation bus and does not require back-up power and that in 8.3.1.1.3 which describes these buses as Class 1E.

b. Since overpressure protection is a function of the reactor water cleanup system instrumentation, provide justification for not providing a Class lE system.

c~ Provide justification for not providing Class 1E equipment for the isolation functions listed in 7.7.1.11.1.1 (sic) of the FSAR.

RESPONSE

a~ There is no discrepancy existing between 7.7.1.8 and 8.3.1.1.3.

b. The RWCU system is not a safety-related system beyond the isolation valves. Overpressure pro-tection for the system is provided by primary pressure relief valves. Therefore, there are no instrumentation requirements relative to overpressurization of the RWCU system.

c ~ Class 1E equipment is provided for system iso-lation. The RWCU system isolation valves are part of the reactor coolant pressure boundary (RCPB) and as such are controlled by the primary containment and reactor vessel isolation control system instrumentation. These valves and piping are Seismic Category I, and the controls are Class lE as described in 7.3. The portion of the RWCU system outside the outer isolation valves is not part of the RCPB and not safety-related, and instrumentation for this portion is nonessential.

031. 054-1

WNP-2 AMENDMENT NO. 3 March 1979 Page 1 of 1 0 31.069 (3.8.2.1)

(6.2.1.1)

(031.001)

Your response to Item 031.001(p) is incomplete. Describe the air supply, pressure control, and position indication for the butterfly valves in accordance with the guidance provided in Section 7.3.1 of Regulatory Guide 1.70. Clarify the reference to 6.2.1.1.2 in the response to Item 031.001(p) since this response does not address the staff's concern regarding the position indication instrumentation..

~Res onse Please refer to 3.8.2.1.3, 6.2.1.1.2c and 7.3.1.1.2.9.1 for the information requested.

031.069-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 5 Q 31.070

~(RSP (6. 2. 2. 2)

(6. 5. 2. 2)

(7. 3. 1. 1)

(031.001)

(031.011)

It is the staff's position that insufficient time is available for the operator to reliably take the manual actions which are necessary to initiate suppression pool spray during a small break. The staff has established the requirement for automatic initiation of suppression pool spray for the Mark II containment. Accordingly, we require you to provide a Class IE automatic control system for each suppression pool spray system.

~Res onse The WNP-2 design meets the intent of the proposed CSB Branch Technical Position on "Steam Bypass for Nark II Containments".

The history of the questions of steam bypass on WNP-2 are ex-tensive, dating back to January 1972. Questions 5.4, 5.22, and 5.24 to the PSAR all respond to the concern. The SER (pp 63-65) summarized the NRC position on the issue at the CP stage and noted that WPPSS agreed to study additional means to mitigate the consequences or minimize the potential for bypass leakage. Thiswas formally documented as a Post CP item in the notes of a NRC-WPPSS meeting held on October 17-18, 1973 (Reference 1). In the notes WPPSS committed to submitting a report on the matter. In August 1974, Reference 2 transmitted the WPPSS report WPPSS-74-2-R5, "Drywell to Wetwell Leakage Study",'atisfying the commitment. The NRC requested additional information concerning the report in Reference (3). References (4) and (5) provide WPPSS responses to the NRC questions. Reference (6) indicated that Structural Engineering Branch found the applicable WPPSS responses acceptable. WPPSS has no record of feedback from Containment Systems Branch on the responses to its questions but assumed in Reference (5} that, in the absence of feedback, the post CP item was resolved. Accordingly, WPPSS has gone ahead with construction in these areas based on the above correspondence.

031. 070-1

WNP-2 AMENDMENT NO. 3 TJIARCH 1979 Page 1 of 1 0 31.071

~7. 3. 1. 1)

(031.001)

The primary containment. and reactor vessel isolation control system receives power from the reactor protection system motor generator sets. Describe how the reactor operator determines the position of each motor operated and each solenoid operated or controlled isolation valve after a loss of the motor generator sets. We are concerned that your present. design de-energizes these sets during a loss of off-site power and does not include automatic restart of the motors.

Response

Status indication for all motor operated containment isolation valves is powered from diesel generator buses and this is not dependent on the availability of the RPS M/G set buses.

In addition to valve position status located on the main bench board at the control switch for each isolation valve, a com-plete containment, isolation valve position display exists on Board S. This panel is located in the first row of control room back panels. The power sources used for the display valve position status are uninterruptible with diesel genera-tor backup. Thus, the solenoid operated isolation valves also have position indication that, is not dependent on RPS M/G set availability.

031.071-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 0 31.072 (7.6.1.5)

(F 7.7-1)

(F 7.2-5)

(031.007)

(031.045)

Your response to Item 031.001(t) is unacceptable. Clarify the discrepancies between 7.1-11, Figures 7.1-1 and 7.2-5, and 7.6.1.5. Specifically, clarify the number of instruments and the designation of these instruments with regard to their trip channel assignments. Provide justification for running redundant signals through the same penetrations.

~Res ense:

There are no discrepancies between Table 7.1-11, Fig.'.1-1 and 7.2-5, and 7.6.1.5. Also, redundant signals do not run through the same penetrations.

The neutron monitoring system trip outputs to the reactor protection system are derived from 6 APRM channels as follows:

RPS tri channel APRM Al A&E A2 C&E Bl B&F B2 D&F This combination which uses APRM channels E&F in redundant RPS trip logic allows an APRM channel in each trip division to be bypassed by the operator without the loss of ability to Scram on a high flux condition. See FSAR 7.2.1.1.4.2 for additional information concerning the neutron monitoring system inputs to RPS.

031.072-1

AMENDMENT NO. 3 MARCH 1979 Page 1 of 1 9 31.075 (7.7.1.3)

(7.7-7)

(031.008)

The response to Item 031.008 and the additional information which is presented in Figure 7.7-7 and 7.7.1.3 is incomplete.

In this regard, the staff notes that the design includes an interlock which prevents the transfer of one pump from high speed operation,to low speed operation when the second pump is operating at high speed and the control switch is moved to the motor generator position. Explain why this interlock is provided. Justify not providing a similar interlock in the pump start circuitry in order to prevent a similar occurrence under the same conditions (i.e., both pu'mps running at high speed) if the switch should be placed in the start position.

~Res ense:

The purpose of the recirculation pump interlock identified in 031.075 is to prohibit flow inbalance between the pumps, thereby minimizing jet pump vibration. A similar interlock in the pump start circuitry is not provided because, by design, if both pumps are running at high speed and the switch is placed in the start position, nothing will happen to affect the status of either pump.

031.075-1

WNP-2 AMENDMENT NO. 5 August 1979 Sheet 1 of 3 Q. 031.076 (RSP)

(6. 7)

(7. 3. 2. 3)

(Q. 031. 019)

It is the staff's position is provided in response to that neither the information which Item 031.019 nor the information which is presented in 6.7 and 7.3.2.3 provides sufficient information on the main steamline isolation valve leakage contol system. Describe this system in accordance with the guidance provided in Section 7.3 of Regulatory Guide I.70, including a process and instrumentation drawing, an electrical schematic, and a failure mode and effects analysis. which is sufficiently detailed to address failures at, theofcomponent level. For example, describe the consequences a spurious closing of the contacts of relay K4 under all plant-operating modes, including testing.

Res onse:

In addition to FSAR 6.7 and 7.3.2.3 and the MSIV leakage control system instrumentation and controls is described in 7.3.1.1.3. The system is shown diagramatically in -PAID form in FSAR Figure 3.2-25 with logic diagrams shown in FSAR Figures 7.3-18a-g.

Electrical schematics for the MSIV leakage control system (Drawings, E519, shts 30 6 31) have previously been submitted as part of FSAR 1.7.

A failure modes and effects analysis addressing'orst function case failures and consequences concerning the safety aspects of the MSIV Leakage Control System, i.e., following a LOCA, was submitted as part of the FSAR 6.7.3.1. This

'analysis was submitted previously to the NRC in Reference 1 as a response to a post-construction permit item. In Refer-ence 2 the NRC requested additional information and WPPSS responded in Reference 3. In Reference,4, the NRC stated the design was acceptable subject to the provision of an inter-lock preventing actuation of the MSIV-LCS In Reference agreed to if the inboard MSIV provide this inter-were open. 5 WPPSS

'ock. In any case, to respond to what is felt the intent of Question 031.076 is, an additional FMEA was performed which addressed failures which could occur during other plant operating modes.

031.076-1

WNP-2 AMENDMENT NO. 5 August 1979 Sheet 2 of 3 The following is a description of those failures identified having undersirable consequences.

Mode'quip.

Failure Effected by Failure Undersirable Remarks/

Mode Effects Results

1. Spurious clos- MOV's Both valves Reactor pressure ing of pres- MSLV-V-9, open simul- steam admitted sure switch 6 MSLV-V- taneously into low pressure PS-25 contacts 10 system piping and into the reactor building. Pos-sible piping dam-age and/or plant personnel hazard.

2. Spurious clos- Same as Same as Same as item 1 ing of relay item 1 item 1 above CR-3 contacts above above

3. Spurious clos- MOV's Same as Same as item 1 ing of relay MSLC-V-4 item 1 above CR-1 contacts MSLC-V- above 5

In order to prevent the events described in items 1, 2.and 3

. from occurring the logic design will be modified. An additional interlock will be added in series with the contacts of pres-sure switch PS-25, relay CR-3, and relay CR-1. The interlock will be provided from the system initiation control switch.

Thus, two active. component failures would be r'equired to cause a similar occurrence.

In addition, it was noted from the results of FMEA that events similar to those described in items 1, 2 and 3 above might occur from localized events in control panels and wireways.

This is due to inherent system designs requiring the several pairs of series motor operated valves to be controlled from a common safety division to meet single failure criteria. To preclude such situations certain key control devices and wiring associated with each valve of a series pair will be separated from the other. This will prevent localized events within a control panel, instrument rack, wireway, or motor control center from causing simultaneous opening of both valves during normal operating modes.

The MSIV leakage control system does not contain a relay K4.

031.076-2

WNP-2 AMENDMENT NO. 5 August 1979 Sheet 3 of 3

References:

WPPSS to NRC letter, G02-74-73, "Post Construction Permit Item Transmittal of Report WPPSS 74-2-'RG, Concept for Main Steam Isolation Valve Leakage Control System",

dated Dec. 3, 1974.

2. NRC to WPPSS letter, Butler to Stein, dated March 18, 1975.

3. WPPSS to NRC letter, G02-75-238, "Response to Request.

for Information MSIV-CCS", dated Aug. 18, 1975.

4. NRC to WPPSS letter, Parr to Stein, dated November 21, 1975.

5. WPPSS to NRC letter, G02-76-294, "MSIV-CCS", dated July 14,-1976.

031.076-3

WNP-2 AMENDMENT NO. 3 MARCONI 1 97 9 Page 1 of 1 Q 31.077 (031.021)

(031.037)

The responses to Items 031.021 and 031.027 are unacceptable because the accuracy of the instrumentation sensors is not provided. The information provided in Table 7.2-1 is labeled as that which is required in contrast to that which is actually provided. Additionally, the first item of Table 7.2-1 indi-cates that, a pressure switch which has an error of plus or minus 10 psi is required. In similar BWR-5 application, this instrument is stated to have accuracy of plus or minus one percent of full scale. Accordingly:

a. Provide an amended response to Item 031.021 which includes the accuracy of the sensors which are installed in your plant.

b. Provide an amended response to Item 031.037 which defines such terms as "adequate margin" and describes the criteria and procedures for determining and adjusting the instrument test frequencies.

~Res ense:

See the response to Question 31.063 (31.016).

031.077-1

WNP-2 AMENDMENT NO. 3 MARCH 1979 Page 1 of 2 0 31.78 (031.050)

(T 7.3-5)

Clarify the discrepancy between the response to Item 031.050(a) and Table 7.3-5. Specifically, explain how the temperature trips, which are usually set between 135 and 185 degree F, can reliably detect a 50 gpm leak from the reactor heat. removal system during a prolonged cold shutdown when the primary cool-ant temperature may be less than 135 degrees F.

~Res onse:

The equipment area temperature monitors are not intended to detect 50 gpm leakage from the emergency core cooling systems (ECCS) during the long-term recovery following the postulated loss-of-coolant accident. There is insufficient energy in the ECCS fluid (when the reactor's depressurized and cooled down) to heat the equipment area in the event of the postu-lated 50 gpm leak.

The design bases for the ECCS equipment area temperature moni-tors is to identify leakage and provide inputs for isolation in the event of leakage from high energy RCPB lines beyond the second isolation valve during normal plant conditions.

In addition to the temperature sensors, RHR flow, Reactor water level, HPCS and LPCS flow detectors listed in response to Q. 31.050, numerous drain flow indicators are provided.

These consist of:

a. Restricting orifice, placed directly in the collecting drain header. These orifices are designed and calibrated to pass 5 gpm with a static head of 6 inches. As soon as total flow in the collector exceeds this flow rate an electrode will sense the increase of fluid and activate an annunciator alarm.

b. Conductance type electrodes 6 inches long

,mounted in a suitable fitting threaded into the collector-header.

c. Mating control switch consisting of a solid-state electronic relay to operate controls.

031.078-1

WNP- 2 AMENDMENT NO. 10 Juiy 1980 Page 12 of 25

f. Markin Characteristic Code Tray/Conduit Inscription Characters P,C, I Div. 1 Black Yellow P,C,I Div ~ 2 Black Orange P~C, I Div. 3 Black Red RPS-Al R Ch. Al Red Lt. Blue NSSS-Al NMS-A RPS-A2 R Ch ~ A2 Red Green NSSS-A2 NMS-C RPS-Bl R Ch. B2 Red Dk. Blue NSSS-Bl NMS-B RPS-B2 R Ch. B2 Red Brown NSSS-B2 NMS-D P,C, I Div. A Black Silver or Silver/

Yellow Str ipe P,C,I Div. B Black Gold or Gold/

Orange Str ipe P Power C Control I Instrumentation Non-Class lE circuits receiving power from Class 1E power sources which are not shed by an accident signal shall be identified by the addition of checkered black/silver or black/gold markers indicating the Class 1E division (Division 1 or 2 respectively) from which the circuit receives its power and identified as A'1 or B'2 (respectively) in the com-puterized cable schedule.

031. 100-12

WNP-2 AMENDMENT NO. 10 July 1980 Page 13 of 25 S ecific Re uirements for Se aration of Cables for Nuclear Safe uards S stems Reactor Protection S stem (RPS, NSSS and NMS)

Reactor Protection System (RPS, NSSS and NSSS, and NMS fail-safe wiring:

a. Fail-safe wiring outside of the main protection system cabinets shall be run in rigid or flex-ible conduits and/or totally enclosed trays used for no other wiring and shall be conspicuously identified at all junction or pull boxes. IRM, LPRM input, and RPS Scram Group output cables may be combined in the same wireway provided that the four divisional separation is maintained.

b. Wires from both RPS trip system trip actuators to a single group of scram solenoids may be run in a single conduit; however, a single conduit shall not contain wires to more than one group of scram solenoids. Wiring for two solenoids on the same control rod may be run in the same conduit.

c. Cables through the primary containment penetra-tions shall be so grouped that failure of all cabling in a single penetration cannot prevent a scram. (This applies specifically to the neutron monitoring cables and the main steam isolation valves position switches.)

d. Power supplies to systems which de-energize to operate (so called "fail-safe" power supplies) require only that separation which is deemed prudent to give reliability (continuity of oper-ation). Therefore, the protection system fly-wheel motor generator (MG) sets and load cir-cuit breakers are not required to comply with the separation requirements of this Specifica-tion for safety reasons even though the load circuits go to separate panels.

e. Wiring for the four RPS scram group outputs and the NSM LPRM inputs must be routed as four separate divisions.

031. 100-13

WNP-2

'Page 1 of 1 Q. 040.9 7.5.2 of the FSAR references General Electric topical report NEDO-10466 which describes the power generator control complex. This report is not acceptable to the staff as a basis for licensing., Demonstrate the acceptability of the power generator control complex for the WNP-2 facility.

RESPONSE

General Electric topical report NEDO-10466 Rev. 1 was issued Oct. 1, 1977. The revised topical report, will demonstrate the acceptability of the power generator control complex for the WNP-2 facility. Following acceptance of the revised topical report by the NRC, the FSAR will be revised to refer to NEDO-10466 Rev. 1.

See also 7.5.2 which has been revised to incorporate the response to this question.

040. 009-1

Page 1 of 2 Q. 040.10 A review of licensee event reports related to the operation of diesel-generators has indicated that, in some cases, the information available to the control room operator regarding the operational status of the diesel-generators, may be im-precise and could lead to a misinterpretation by the operator.

This can be caused by the sharing of a single annunciator alarm to indicate: (1) conditions which would render a diesel-generator unable to respond to an automatic emergency start signal; and (2) abnormal, but not disabling, conditions.

Another cause can be the specific wording of an annunciator window which does not clearly indicate that a diesel-generator is inoperable (i.e., unable to respond to an automatic emergency start signal during the period the annunciator is sounded) when, in fact, it is inoperable for that purpose.

Provide the results of an evaluation of the alarm and control circuitry for the WNP-2 diesel-generators to determine how each condition that renders a diesel-generator inoperable, is alarmed in the control room. These conditions would include:

(a) the trips that lock out, the diesel-generator start, thereby requiring manual reset; (b) control switch or mode switch positions that block automatic start; (c) loss of control voltage; (d) insufficient starting air pressure or (e) insufficient battery voltage. Your review should. consider all possible operational conditions for the diesel-generator (e.g., test conditions and operation from local control stations). One area which is of particular concern to the NRC staff is the unreset condition following a manual stop at a local station which terminates a diesel-generator test.

Manual stops such as this prevent subsequent automatic opera-tion until the diesel-generator controls are reset. Your re-sponse should provide a detailed evaluation, including the results and your conclusions, and a tabulation of the fol-lowing items:

a. all conditions that could render the diesel-generator incapable of responding to an auto-matic emergency start signal for each operating mode as discussed above;

b. the wording on the annunciator alarm window in the control room for each of the conditions identif ied in item (a);

c. any other alarm signals not included in item (a) which also cause the same annunciator to alarm; 040.010-1

WNP-2 AMENDMENT NO. 7 November 1979 Q. 040. 043 Review the electrical control circuits for all safety-related equipment to provide assurance that disabling of one component does not, through incorporation in other inter-locking or sequencing controls, render other components inoperable. All modes of test, operation, and failure should be considered. Describe and state the results of your review.

Res onse:

The WNP-2 physical and electrical separation criteria does not require that the disabling of any safety related component does not render any other safety related component inoperable.

The, WNP-2 design is consistent, with Federal Regulations and Industry Codes and Standards which require that a sufficient number of circuits and equipment be maintained such that protective functions required during and following any design basis event, when taken with any single failure, can be accomplished.

This design approach does not preclude the disabling of safety system components, either intra-division or inter-division, from failure of other intra-division or inter-division components during all plant operating modes unless that failure can result in the loss of a protective function.

An analysis of this position is contained in 7.2, 7.3, 7.4 and 7.6 where compliance with IEEE 279-1971 paragraphs 4.2, 4.5, 4.6, 4.7 and 4.12 are addressed for each safety system.

040.043-1

AMENDMENT NO. 7 WNP-2 November 1979 Page 1 of 2 Q 40 44 During our review of the Hatch 2 application for an operating license, we identified certain potential problems that could be caused by the motor-generator sets of the reactor protection system (RPS).. These problems were related to the operating characteristics of these motor-generator sets which might exceed the envelopes of acceptable values of voltage and frequency, thereby adversely affecting the connected loads. Indicate whether the motor-generator sets in the WNP-2 RPS are similar to those in the Hatch 2 facility. If they are, provide: (1) a commitment to the generic resolution of this item; or (2) justification for the use of these motor-generator sets in the WNP-2 facility.

Response

WPPSS has the subject motor-generator set type and was made aware of the problem in reference (1). In reference (2), WPPSS committed to the generic resolution of thewill problem with General Electric. The motor-generator be supplied with Class IE qualified equipment to monitor and protect the connected loads from unacceptable values of voltage and frequency. The generic design and qualification plan supplied by GE has been approved by the NRC as satisfying the requirements of IEEE 379-1972,'ection 6.6.

040.044-1

AMENDMENT NO- 7 WNP-2 November 1979 Q. 40. 59 Describe the instrumentation, controls, sensors and alarms of the diesel engine combustion air intake and exhaust system which alert the reactor operator when the design parameters of this system are exceeded. Discuss the actions of the operator if this system annunciates an alarm in the control room. As before, our concern is the time available for an operator to take appropriate action. (Refer to Paragraphs II.1 and II.4 of Section 9.5.8, Revision 1, of the SRP).

Res onse:

Alarms are not provided on DG exhaust and intake parameters.

Other upset conditions are monitored and annunciated on local instrument panels and brought to the main control room opera-tors attention in the form of a single trouble annunciator.

The air filter for the turbocharged diesel engine is the panel type oil bath filter. This type filter is self cleaning during operation, and air restriction due to a clogged filter is not considered a relevant possibility. The panel type oil bath filters provide efficient air filtration with a minimum of maintenance. These filters are inspected, drained and cleaned periodically as recommended by the manufacturer.

The diesel exhaust is also an open flow system with no evident potential for development of restriction. Diesels are tested periodically as required by Technical Specification, there-fore, any unforseen degradations would become evident in the performance parameters upon which DG operability is based.

Emergency DG systems are redundant therefore no credit is taken for operator action for an assumed failure in a single unit.

040.059-1

WNP-2 AMENDMENT NO- 7 November 1979 Page 1 of 2 Q. 40.60 (10.2)

Expand your discussion of the turbine speed control and over-speed protection system. ~ Provide additional explanation of the turbine and generator electrical load following capabil-ity for the turbine speed control system with the aid of sys-tem schematics (including turbine control and extraction steam valves to the heaters). Tabulate the individual speed control protection devices (normal, emergency and backup),

the design speed (or range of speed) at which each device begins operation to perform its protective function (in terms of percent of normal turbine operating speed). In order to the adequacy of the control and overspeed protection 'valuate system provide schematics and include identifying numbers to valves and mechanisms (mechanical and electrical) on the schematics. Describe in detail, with references to the identifying numbers, the sequence of events in a turbine trip including response time, and show that the turbine stabilizes.

Provide the results of a failure mode and effects analysis for the overspeed protection systems. Show that a single steam valve fai,lure'cannot disable the turbine overspeed trip from functioning., (SRP10.2, Part III, items 1, 2, 3 and 4).

Response

a. Provide additional explanation of the turbine and generator electrical load following capabil-ity "for the turbine speed control'system with the aid of system schematics (including turbine control and extraction steam valves to the heaters).

Answer: See the response to question 40.63 and revised FSAR 7.7.1.5 (including Figures 7.7-9 and 7.7-10).

l

b. Tabulate the individual speed control protection devices (normal, emergency and backup), the design speed (or range of speed) at which each device begins operation to perform its protec-tive function (in terms of percent of normal turbine operating speed).

Answer: See revised FSAR page 10.2-5.

040.060-1

WNP-2 AMENDMENT NO. 7 November 1979 Q. 40.69

~10 . 4. 1 Indicate and describe the means of detecting radioactive leakage into and out of the main condenser. Indicate what provisions have been incorporated into the WNP-2 facility to preclude unacceptable accidental release of radioactivity to the environment. (Refer to Paragraph III.2.b of Section 10.4.1, Revision 1, of the SRP).

Response

Please see revised 10.4.1.3. The main condenser evacuation system maintains a vacuum to remove noncondensible gases from the condenser, including air and radioactive gaseous products originating in the reactor.. This effluent is discharged to the gaseous radwaste system. See 11.3 for a description of this system. Monitoring and control of release paths from

,the gaseous radwaste system and other potential release paths from the condenser (e.g., turbine building exhaust, circu-lating water, etc.) is described in 11.5.2. See also 10.4.2.3 and 10.4.2.5.

040.069-1

WNP-2 AMENDMENT NO. 7 November 1979 Q. 40. 70

~10..1 Discuss the operation of the main steam line isolation valves if there is a loss of condenser vacuum. (Refer to Paragraph III.3b of Section 10.4.1, Revision 1, of the SRP).

Response

See revised 10. 4. 1.5, 7. 3. 1. 1. 2. 4. l. 13, and 7.3. 2.2.2.3. l. 12 .

See also the response to 40.65.

040.070-1

WNP-2 AMENDMENT NO. 5 August 1979 Sheet 1 of 2 Q. 212.003 (6. 3)

Your discussion of single failure does not adequately address ECCS passive failures during long-term cooling. Accordingly, provide a response to the attached Reactor Systems Branch Technical Position regarding the leak detection requirements for passive failures in the ECCS piping.

REACTOR'YSTEMS BRANCH TECHNICAL POSITION Leak Detection Re uxrements for ECCS Passive Failures The passive failures to be considered are limited to leaks from valve stem packing and pump seals. The sum of these leak rates may range from essentially no leakage up to the equiva-lent of the sudden failure of the seal of the largest ECCS pump (e.g., about 50 gpm). It is the staff's position that detection and alarms be provided to alert, the operator of passive ECCS .failures during long-term cooling. The timing of these alarms should be "such that the reactor operator has sufficient time to identify and isolate the faulted ECCS line.

Provide the following information regarding the ECCS leak detection system:

a. An identification and justification of the maxi-mum leak rate;

b. The maximum allowable time for operator action, including a justification of the time interval;

c. A demonstration that the leak detection system will be sensitive enough to provide an alarm to the operator, subsequent identification by the operator of the faulted line, and, finally, per-mit the operator to isolate the faulted line prior to the leak creating any undesirable con-sequences such as flooding of redundant equip-ment. The minimum time to be considered for this sequence of events is 30 minutes.

d. A demonstration that. the leak detection system can identify the faulted ECCS train and that the leak is isolable.

Additionally, the ECCS lead detection system must meet the following standards: (1) control room alarm; and (2) IEEE-279, except single failure requirements.

212.003-1

WNP-2 AMENDMENT NO. 5 August 1979 Sheet.2 of 2

Response

a. The ECCS are capable of withstanding passive fail-ures of valve stem packings and pump seals fol-:

lowing a LOCA. The maximum leakage due to a failure of this nature could be as high as 23 gpm from an HPCS, LPCS or RHR pump seal failure.

Valve stem leakage would be significantly less than this.

b. The maximum allowable time for operator action is determined as the shorter of the time required to flood to the level of an ECCS pump motor in the secondary containment, or the time required to drain the suppression pool to a level below that required ECCS pump NPSH. The'maximum NPSH re-quired for any ECCS pump is 21 ft. (HPCS).

With a minimum NPSH available of 36 ft., calcu-lated in accordance with Regulatory Guide 1.1, and a leakage rate of 23 gpm, there is about 15 days of operator time available before NPSH becomes a problem. A Class IE level instrument will be installed in each'ECCS pump room and it will be mounted just above floor level. After the alarm in the control room, operator receives an there is at least 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> of operator time avail-able before the water level reaches the bottom of an ECCS pump, assuming. a 23 gpm leak rate into the smallest ECCS pump room (RHR C),.

c. The sensitivity of the leak detection system is not vital to the identification and subsequent isolation of the faulted line prior to any unde-sirable consequences with at least 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> available.

d. Any ECCS leak can be isolated, including any packing failure on any ECCS pump suction valve.

This packing can be isolated by closing the valve since the valves are double-seat, wedge knife gates. With a Class IE level instrument in each ECCS pump room, there is no problem with identification of the faulted ECCS train.

The leak detection system will have a control room alarm and meet IEEE 279, except single failure re-quirements.

212.003-2

NNP-2 AMENDMENT NO. 1 1 September 1980 Q. 312 ~ 17 (15. 6. 5)

Your statement-on Page 15.6-34 of the FSAR regarding the bypass, leakage of 0.11 percent per day is ambiguous. Indicate whether this leakage is in addition to, or is part of, the primary containment leakage. The bypass leakage, which will be a major contributor to offsite doses, is not listed as an assumption in Table 15.6-12 of the FSAR. Correct this omission.

~Res onse:

The statement given on Page 15.6-34 regarding bypass leakage provided a quantitative assessment of how much bypass leakage could hypothetically occur in addition to the primary to secondary containment leakage identified in 15.6.5.5.1.2 with the resulting radiological consequences not exceeding 10 CFR 100 criteria. This hypothetical leakage value is Oe10%/day.

The primary to secondary containment leakage would be pro-cessed by SGTS. Bypass leakage would not be processed prior to release. Since the bypass leakage value presented is a bounding value provided for information only, it appropriate to include this hypothetical leakage in the is not assumption identified in Table 15.6-12. If it were included, the radiological consequences would, by definition, just match the 10 CFR 100 criteria. The text on page 15.6-34 will be modified to clarify the statements.

No bypass leakage paths which would be a major contributor to offsite doses have been identified. The bypass leakage iden-tified in 6.2.3.3 (0.74 SCFH) is less than 1/27th of the hypothetical bounding leakage valve identified above. Since this contribution will not change the conclusion that the dose consequences would be a small fraction of 10 CFR 100 criterion, the leakage is not included in Table 15.6-12.

WNP- 2 AMENDMENT NO. 11 September 1980 Q. 312.018 (15.6.5)

Your lat'est response to Item 022.7 in your letter of November 21, 1978, indicates that you cannot provide the secondary con-tainment pressure response following postulated loss-of-coolant accident until February 1979. Since we need this information before we can begin our evaluationof the potential offsite doses, we request that you expedite your response on this matter. Our original request for this information was forwarded to you in our first acceptance review dated June 24, 1977.

~Res onse:

See the revised response to Question 022.007 which refers to revised FSAR 6.2.3.3.

hh 312. 018-1

WNP-2 AMENDMENT NO. 1 July 1978 Page 1 of 1 Q 321. 1 The description of the main condenser offgas treatment system is incomplete. Describe the hydrogen gas analyzers which are downstream of the recombiners and upstream of the delay portions of the system, including a brief description of analyzer checks and calibrations. We require that the analyzer(s) shall be nonsparking.

Res onse:

Please see 7.7.1.10.3.5 and 11.3.2.1.8.1 for the requested discuss3.on ~

321.001-1

WNP-2 AMENDMENT NO. 1 July 1978 Page 1 of 1 g 321.2 In 15.7.3 of the FSAR, you state that the accident analysis for postulated liquid waste tank failures will be supplied later. Provide an estimate of the date this information will be submitted.

Res onse:

Please see 15.7.2.5 and 15.7.3.5 for discussions of liquid waste tank failure accident analyses.

321.002-1

WNP-2

1. 1 ROLE OF DFFR/MARK 'II'ROGRAM Operating experience with foreign and domestic BWR plants has shown that significant dynamic loadings may be imposed on suppression chamber str'uctures and components during safety relief valve (SRV) discharge. from Hydrodynamic loads in the suppression chamber resulting postulated loss-of-coolant accident (LOCA) events have also been identified.

As a result of such concerns, a group was formed by the domestic utility owners of BWR plants with containments of Mark II,configuration (Mark II Owners Group) to address the concerns related to these dynamic loads. The specific purpose of the group was to develop an overall program with .

the objective of defining suppression pool hydrodynamic loads on a generic level, i.e., loads which could be used for the design or assessment of design adequacy of any containment of Mark II configuration. The results to date of the Mark II Owners Group efforts can be found in the Dynamic Forcing Function Information Report (DFFR) (Reference l-l) and related documents. Because of the continuing programs of the Mark II Owners Group, several revisions to the DFFR have been made future revisions may be required.

and The Mark II Containment Lead Plant Program Load Evaluation Report (LER) (Reference 1-2) was issued in October, 1978.

The report addresses the portion of the Mark II Owner's pro-gram that provides a generic methodology for establishing design basis LOCA and SRV loads for the lead Mark II facili-ties (Zimmer, Shoreham, and LaSalle) and follow-onanplants that wish to use the lead plant criteria. It includes evalua-tion of Mark II Owner's methodology, a description of load methodologies found acceptable by NRC and the basis for NRC conclusions. Table 1.1-1 lists the containment hydrodynamic loads in the format and order given in Table IV-1 of the LER.

Table 1.1-1 identifies:

a. where the review of loads conducted by NRC for lead plants is expected to be sufficient for WNP-2, i.e., the criteria given in the LER are acceptable for WNP-2 at this time,

b. where NRC review of generic items (for WNP-2 and other plants) remains to be completed,

c. where NRC review of WNP-2 plant unique inform-ation is required.

Revision 2

l. 1-1 August 1979

TABLE 1.1-1 CONTAINMENT HYDRODYNAMIC LOADS/LOAD EVALUATION REPORT SUhtKLRY NRC ITEM NUMBER IV-1) SUBJECT/LOAD STATUS OF LICENSING (FROM LER TABLE LOCA I.A SUBMERGED BOUNDARY VENT CLEARING LOAD NRC CRITERIA GIVEN ZN LER, NO ADDITIONAL NRC REVIEW FOR WNP-2 ANTICIPATED POOL SWELL ANALYTICALMODEL NRC CRITERIA GIVEN IN LER, NO ADDITIONAL NRC REVIEW FOR WNP-2 ANTICIPATED SUBMERGED BOUNDARY POOL SWELL LOAD NRC CRITERIA GIVEN IN LER; NO ADDITIONAL NRC RFVIEW FOR WNP-2 ANTICIPATED POOL SWELL IMPACT LOAD FOR SMALL STRUCTURES AND GRATING, REVIEW OF WNP-2 DAR METHODS REQUiRED WETWELL AIR COMPRESSION NRC CRITERIA GIVEN IN LER; NO ADDITIONAL NRC RFVXEW FOR WNP-2 ANTICIPATED ASYMMETRIC LOAD COMPLETION OF GENERIC REVIEW OF MK II ALTERNATE POSITION REQUIRED DOWNCOMER LATERAL LOAD COMPLETION OF GENERIC REVIEW OF PRETECH LOAD DFFINITION REQUIRED SUBMERGED BOUNDARY STEAM C.O.: NRC CRITERIA GIVEN IN LER; NO ADDITIONAL NRC CONDENSATION LOADS REVIEW FOR WNP-2 ANTICIPATED CHUGGING: COMPLETION OF REVIEW OF BSR CHUGGING LOAD DEFINITION REQUZRED; SEE REPORTS OF APRIL 13 AND JUNE 15 SRV II.A POOL TEMPERATURE LIMITS NRC CRITERIA GIVEN IN LERt NO ADDITIONAL NRC REVIEW FOR WNP-2 ANTICIPATED X-QUENCHER LOADS TO BE PRESENTED LATER IN WNP-2 DAR II.B .AIR CLEARING LOADS Ii.c.l QUENCHER ARM LOADS NRC CRITERIA GIVEN IN LER; NO ADDITIONAL NRC REVIEW FOR 'ÃNP-2 ANTICIPATED QUENCHER TZE DOWN LOADS NRC CRITERIA GIVEN IN LER; NO ADDITIONAL NRC REVIEW FOR WNP-2 ANTICIPATED

WNP-2 Figure 2.1-3 and enter the pool vertically within 3'-0" distance from the"containment as shown in Figures 2.1-6 through 2.1-8. The" d'esign of these pipes is governed by the pool swell impact load on the horizontal projection of the pipes above the initial pool surface combined with the drag load on submerged portions.

c. Piping Systems In Pool Swell Zone The pool swell zone is identified in Section 3.2.3 to be between the elevations of the initial pool surface (466'-4 3/4") and the maximum pool rise during a LOCA (design basis accident) (484'-4 3/4").

Piping systems in the pool swell zone include short projections into the chamber from the con-tainment at one access hatch and ten miscellan-eous piping systems as shown in Figure 2.1-3 and Figures 2.1-6 through 2.1-8.

d. Piping Systems Above the Pool Swell Zone Piping systems above the pool swell zone include short lengths of pipe entering at elevation 491'-0" and two penetrations for the wetwell spray header also at elevation 491'-0" as shown in Figure 2.1-4 and Figures 2.1-6 through 2.1-8.

These systems are not subjected to direct hydro-dynamic loads associated with a LOCA or with SRV actuation, only building response due to hydro-dynamic events would affect piping in this classification.

2.1.2 STRUCTURES, PIPING AND COMPONENTS INDIRECTLY AFFECTED BY POOL DYNAMIC LOADS Outside of the suppression chamber, that structures, piping, and it components has been postulated may be affected by pool dynamic loads. This has been postulated to occur as a result of loading applied to the suppression chamber boundary (basemat, pedestal, and containment shell) which would result in vibratory motion elsewhere in the reactor building. This is commonly referred to in this and other reports as. "building response". As discussed later in this report, studies now are underway of measured building response results at Caorso.

The results to date of the investigations indicate that building response effects are small and of secondary significance as

'2. 1-3 Revision 2 August 1979

WNP-2 compared to the effects of pool dynamic loads on structures within the suppression chamber directly affected by these loads. The structures, piping, and components which may be indirectly .affected, by pool dynamic loads will be covered in later revisions of this report and the FSAR after the com-pletion of the Burns. and Roe evaluation of Caorso test results.

2.1-4 Revision 2 August 1979

2.3

SUMMARY

AND CONCLUSIONS

'I

2. 3. 1

SUMMARY

OF CHANGES TO PRESERVE DESIGN MARGINS As noted in Section 2.1, structures, piping, and components which may be affected by pool, dynamic loads can be divided into two general categories, i.e., those directly affected by pool dynamic loads (those in and bounded by the suppression chamber) and those affected only indirectly by pool dynamic loads (outside the suppression chamber). This revision covers the structures, piping, and components in and bounded by the suppression chamber. For these structures several changes in design have been implemented as a result of consideration of SRV discharge and LOCA hydrodynamic loads. Table 2.3-1 provides a list. of the structures and components that have been covered in this report and the design changes that have been made. The steel containment structure has been rein-forced by the addition of several horizontal rows of tee stiffeners as shown in Figure 4.1-1. The downcomer bracing system has been redesigned from a system of radial beams to a pipe truss system. This bracing system also is designed to provide lateral restraint for the SRV discharge pipes.

Quenchers have been provided as exit devices for the SRV dis-change pipes. Additions and modifications of pipe supports for miscellaneous piping systems have been provided. Other miscellaneous changes are noted in Table 2.3-1.

For structures, piping and components affected only indirectly, by pool dynamic loads (those outside the suppression chamber),

changes in design to accommodate pool dynamic loads are not being made in view of the results being obtained from the Caorso tests. Data from the Caorso tests indicate that the actual effects of SRV discharge are small as compared to currently available load definitions. A more realistic load definition will be presented in a subsequent. revision of the NNP-2 DAR after the completion of the Burns and Roe evaluation of, Caorso test results.

2. 3. 2 CONCLUSIONS The assessment indicates that the modified design of the wet-well for NNP-2 is capable of withstanding the effects of the hydrodynamic loads resulting from SRV actuation and postulated LOCA events in conjunction with other applicable loads. This conclusion is based on evaluation using loads given in the NRC Load Evaluation Report (NUREG-0487, Reference 2-1) except as noted herein, and additional data and information such as is now available from the SRV tests performed at the Caorso plant in Italy (Reference 2-2).

2. 3-1 Revision 2 August 1979

WNP-2 This revision reflects new developments in-several areas of load definition. For example, chugging loads used in this report are based on work by Burns and Roe to account for fluid structure interaction effects and other effects peculiar to the 4T test facility which were not accounted for in the previous load definition (Reference 2-3). Also reflected in this revision is a LOCA jet load definition which accounts for the ring vortex character of the LOCA jet phenomenon (Reference 2-4) .

Work is continuing as noted in this report to finalize the re-maining load definitions. Of particular note is the work under-

=

way now at Burns and Roe to finalize SRV discharge load definition. The results of SRV testing carried out at the Caorso plant in Italy is being used to complete this work. The Caorso tests represents the most extensive SRV testing program to date with geometry and plant conditions similar to WNP-2. A detailed evaluation and definition of the SRV discharge load will be the final result of the work now underway and will be reported in a subsequent revision to this report. Although studies such as those for SRV loads remain to be completed, the assessments reported in this revision are expected to remain valid since the final loads are expected to be less than those used in this revision. This should allow NRC review of the adequacy of the WNP-2 containment system design to proceed in parallel with the continuing NRC generic and plant specific review of the remaining load definitions.

2~3 2 Revision 2 August 1979

WNP-2

2.4 REFERENCES

Mark II Containment Lead Plant Program Load Evaluation Report, NUREG-0487, United States Nuclear Regulatory Commission, October, 1978.

2-2 Hark II Containment Supporting Program Caorso SRV Dis-charge Tests, Phase I Test Report, General Electric Company, NEDE-25100-P, May, 1979.

2-3 "Chugging Loads-Improved Definition and Application Methodology to Mark II Containments," Technical Report, Burns and Roe, Inc., June, 1979.

2-4 Letter MFN-144-79 from L.J. Sobon, General Electric, to J.F. Stolz, Nuclear Regulatory Commission on "Technical Description of the Ring Vortex Model,"

May 22, 1979.

2. 4-1 Revision 2 August 1979

WNP-2 Revision 2 August 1979

WNP-2 3.1.2.3 Steam Condensation Loads After the water and air have been expelled from the SRV dis-charge line, high pressure, high temperature, high mass flux steam is discharged into the suppression pool. As the steam condenses and collapses, vibrations or small pressure fluctua-tions are produced in the water. Experiments employing a quen-cher device have exhibited no significant pressure fluctuations up to the pool boiling temperature of 212 F. As a result, significant pressure fluctuations in the WNP-2 suppression pool due to steam condensation from the quenchers are not expected.

This phenomenon is therefore not considered further.

3.1.3 SRV AIR CLEARING LOADS Testing and analytical efforts have been performed by the Mark II Owners Group and by other organizations to define the loads resulting from discharge through a quencher device upon actua-tion of -the SRV. The SRV testing carried out at the Caorso plant in Italy represents the most extensive test program to date with geometry and plant conditions similar to WNP-2.

An analytical effort has been undertaken by Burns arid Roe to evaluate the data taken during the Caorso Phase I and II tests. Work is in process to understand and quantify effects of SRV discharge on a Mark II containment. A detailed evaluation and definition of the SRV discharge load will be the final result of this ongoing work and will be reported in a subsequent revision to this report.

For the purposes of this report a conservative interpretation of preliminary results of the Caorso test data and the methods found in the DFFR (Reference 3-3) are used to define an SRV load specification for assessment. of the adequacy of WNP-2 wetwell structures and components. The details of this interim load specification are provided in the following sections.

3.1.3.1 Boundary Loads Data from scaled quencher experiments and in-plant quencher tests show that pressure fluctuations associated with the water and air clearing of the SRV line occur in the pool during the initial second or so after an SRV actuation.

The DFFR (Reference 3-3) provides a methodology for predic-ting the peak positive and negative boundary pressures. It has been found that the peak boundary pressures predicted by the DFFR methodology at conditions reflective of those en-countered the Caorso plant during Phase I and II tests are significantly higher than the loads recorded during'he actual tests. (Reference 3-2).

Revision 2

3. 1-3 August. 1979

WNP-2 The highly conservative peak pressure definitions found in the DFFR and the conservative definition for pressure dis-tributions and forcing function time histories have been used to evaluate the structural adequacy of the containment, basemat and pedestal. The DFFR based peak pressures are shown in Table 4. 1-1 and the results of the assessment of the contain-ment, basemat and pedestal are discussed in Section 4.0.

3.1.3.2 Building Response Building responses calculated using the DFFR (Reference 3-3) peak boundary pressure prediction and application methodology (Monte Carlo approach, Reference 3-21) are reduced by a factor of two in order to obtain interim building responses for this assessment. This reduction is justified in view of:

a) The much smaller boundary loads measured during the Caorso tests relative to the peak pressures recommended by the DFFR, (Table 3.1-2).

b) The much smaller pressure response at all frequencies for the Caorso measured pressure time histories relative to the pressure time histories for the idealized pressure time histories of the DFFR. Figure 3.1-1 shows a comparison between the two pr'essure responses. These re-sponse spectra were obtained by applying each of the pressure time histories as a forcing function on a single-degree-of-freedom (SDOF) system with 1% damping and computing the maximum force in the spring as the spectral value.

c) The much smaller building responses recorded during the Caorso tests in comparison with those predicted using the methods recommended by the DFFR, Figure 3.1-2.

These conservative interim building responses are used in the assessment to meet WNP-2 schedule requirements. The building response spectra are shown in Section 5.0. A more realistic load definition will be presented in a subsequent revision of the WNP-2 DAR after the completion of the Burns and Roe evaluation of Caorso test results.

3.1.3.3 Submerged Structure Loads The pressure and velocity fluctuations created in the pool during, the air and water clearing are expected to produce acceleration and standard drag forces on submerged structures.

Revision 2 3.1-4 August 1979