Text

Technical Specifications Bases Control Program
	ML041180501
Person / Time
Site:	Crystal River
Issue date:	04/15/2004
From:	Powell S; Progress Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	3F0404-03, ITS 5.6.2.17
	Download: ML041180501 (94)
	v • d • e

C4 Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: ITS 5.6.2.17 April 15,2004 3F0404-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Technical Specifications Bases Control Program

Dear Sir:

Florida Power Corporation, doing business as Progress Energy Florida, Inc., hereby submits the changes that were made to the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) Bases as required by ITS 5.6.2.17. The attachments provide revisions to the CR-3 ITS Bases that will update NRC copies of the ITS.

Attachment A provides the instructions for updating the CR-3 ITS Bases. Attachment B provides the CR-3 ITS and Bases Lists of Effective Pages. Attachment C provides the replacement pages for the CR-3 ITS Bases.

If you have any questions regarding this submittal, please contact Mr. Sid Powell, Supervisor, Licensing & Regulatory Programs at (352) 563-4883.

Sincerely, S. C. Powell Supervisor Licensing & Regulatory Programs SCP/ff Attachments:

A.

Instructions for Updating the Crystal River Unit 3 ITS and Bases B.

CR-3 ITS and Bases Lists of Effective Pages C.

Replacement CR-3 ITS Bases Pages xc:

Regional Administrator, Region II (w/o Attachment C)

Senior Resident Inspector (w/o Attachment C)

NRR Project Manager (w/o Attachment C)

Progress Energy Florida. Inc.

Crystal River Nuclear Plant 15760W. Power Line Street

-Acc

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Crystal River, FL 34428

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT A INST'RUCTIONS FOR UPDATING THE CRYSTAL RIVER UNIT 3 ITS AND BASES

INSTRUCTIONS FOR UPDATING THE CRYSTAL RIVER UNIT 3 IMPROVED TECHNICAL SPECIFICATIONS f

4/15/04 Page(s) to be Removed Page(s)

Revision Page(s) to be Added Pages(s)

ITS LOEPages (1-5)

ITS Bases LOEPages (1-8)

B 3.2-1 / B 3.2-2 B 3.2-9 / B 3.2-10 B 3.2-11/ B 3.2-12 B 3.2-15 / B 3.2-16 B 3.2-17 / B 3.2-18 B 3.2-19/B 3.2-20 B 3.2-23 / B 3.2-24 B 3.2-26 / B 3.2-27 B 3.2-36 B 3.2-38 / B 3.2-39 B 3.244 B 3.3-113 / B 3.3-114 B 3.3-115/B 3.3-116 B 3.3-125B / B 3.3-126 B 3.4-39 / B 3.4-40 B 3.4-41 / B 3.4-42 B 3.6-7 / B 3.6-8 B 3.6-17 / B 3.6-18 B 3.6-19 / B 3.6-20 B 3.6-21 /B 3.6-22 B 3.6-23 / B 3.6-24 B 3.6-25 / B 3.6-26 B 3.643/ B 3.644 B 3.7-71 /B 3.7-72 B 3.7-73 / B 3.7-74 B 3.7-75 / B 3.7-76 B 3.7-85 / B 3.7-86 B 3.8-5/B 3.8-6 B 3.8-7 / B 3.8-8 B 3.8-9 / B 3.8-10 B 3.8-1OA /B 3.8-1OB B 3.8-1OC /B 3.8-11 B 3.8-20/B 3.8-21 B 3.8-22 / B 3.8-23 B 3.8-28 / B 3.8-29 B 3.9-9 / B 3.9-10 B 3.9-11/B 3.9-12 B 3.9-13 B 3.9-23 / B 3.9-24 B 3.9-25 2/24/03 2/24/03 149/ 149 149/ 149 149/22 149/ 149 149/ 149 149/ 149 149/ 149 149/ 149 149 149/ 149 23 35/17 149/37 35 /22 29/29 29/29 37/ 156 37/ 156 38/26 156/ 11 11/ 156 156/11 34/34 193/193 37/ 193 193 /37 182/ 199 160/ 149 26/ 182 182 /26 182 15/ 149 163 /22 149/ 157 184 / 37 37/ 184 184 37/37 37 ITS LOEPages (1-5)

ITS Bases LOEPages (1-8)

B 3.2-1/B 3.2-2 B 3.2-9 / B 3.2-10 B 3.2-11/ B 3.2-12 B 3.2-15 / B 3.2-16 B 3.2-17 / B 3.2-18 B 3.2-19 / B 3.2-20 B 3.2-23 / B 3.2-24 B 3.2-26 / B 3.2-27 B 3.2-36 B 3.2-38 / B 3.2-39 B 3.244 B 3.3-113 / B 3.3-114 B 3.3-115/B 3.3-116 B 3.3-125B / B 3.3-126 B 3.4-39 / B 3.440 B 3.441 / B 3.442 B 3.6-7 / B 3.6-8 B 3.6-17 / B 3.6-18 B 3.6-19 / B 3.6-20 B 3.6-21 /B 3.6-22 B 3.6-23 / B 3.6-24 B 3.6-25 / B 3.6-26 B 3.643 / B 3.644 B 3.7-71 / B 3.7-72 B 3.7-73 / B 3.7-74 B 3.7-75 / B 3.7-76 B 3.7-85 / B 3.7-86 B 3.8-5 / B 3.8-6 B 3.8-7 / B 3.8-8 B 3.8-9 / B 3.8-10 B 3.8-1OA / B 3.8-1OB B 3.8-1OC / B 3.8-11 B 3.8-20 / B 3.8-21 B 3.8-22 / B 3.8-23 B 3.8-28 / B 3.8-29 B 3.9-9 / B 3.9-10 B 3.9-11 / B 3.9-12 B 3.9-13 B 3.9-23 / B 3.9-24 B 3.9-25 Revision 10/20/03 4/14/04 149/45 149/45 149/45 149/45 44/ 149 45/ 149 149/45 44/45 45 149 /45 45 35 /43 43 /43 46/22 47/29 47/29 37 /48 37/44 44/26 209/209 209/ 156 209 / 209 44/44 42/42 42/42 208 /42 182/49 160/43 207 /207 182/207 207 /207 207 / 182 43 /43 207 /43 149/43 208 /43 43 / 43 43 208/208 208

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT B CR-3 ITS AND BASES LISTS OF EFFECTIVE PAGES

CRYSTAL RIVER UNIT 3 IMPROVED TECHNICAL SPECIFICATIONS ITS LOEPage I of 5 10120103 List of Effective Pages (Through Amendment 211)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page ii iii iv v

vi vii 1.1-1 1.1-2 1.1-3 1.1-4 1.1-5 1.1-6 1.1-7 1.1-8 1.2-1 1.2-2 1.2-3 1.3-1 1.3-2 1.3-3 1.3-4 1.3-5 1.3-6 1.3-7 1.3-8 1.3-9 1.3-10 1.3-11 1.3-12 1.4-1 1.4-2 1.4-3 1.4-4 2.0-1 2.0-2 2.0-3 3.0-1 3.0-2 3.0-3 Amendment 182 161 182 182 182 182 182 149 149 149 205 149 205 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 211 201 149 149 149 149 Page Amendment 3.0-4 3.0-5 3.1-1 3.1-2 3.1-3 3.1-4 3.1-5 3.1-6 3.1-7 3.1-8 3.1-9 3.1-10 3.1-11 3.1-12 3.1-13 3.1-14 3.1-15 3.1-16 3.1-17 3.1-18 3.1-19 3.1-20 3.1-21 3.2-1 3.2-2 3.2-3 3.2-4 3.2-5 3.2-6 3.2-7 3.2-8 3.2-9 3.2-10 3.2-11 3.2-12 3.2-13 3.3-1 3.3-2 3.3-3 3.3-4 203 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 152

CRYSTAL RIVER UNIT 3 IMPROVED TECHNICAL SPECIFICATIONS ITS LOEPage 2 of 5 10/20/03 List of Effective Pages (Through Amendment 211)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page 3.3-5 3.3-6 3.3-7 3.3-8 3.3-9 3.3-10 3.3-11 3.3-12 3.3-13 3.3-14 3.3-15 3.3-16 3.3-17 3.3-18 3.3-19 3.3-20 3.3-21 3.3-22 3.3-23 3.3-24 3.3-25 3.3-26 3.3-27 3.3-28 3.3-29 3.3-30 3.3-31 3.3-32 3.3-33 3.3-34 3.3-35 3.3-36 3.3-37 3.3-38 3.3-39 3.3-40 3.3-41 3.3-42 3.3-43 3.3-44 Amendment 204 149 149 149 149 149 149 149 178 152 178 149 152 149 155 206 206 149 152 149 152 149 194 194 194 149 149 149 149 149 208 199 199 149 149 177 177 149 152 196 Page Amendment 3.4-1 3.4-2 3.4-3 3.4-4 3.4-5 3.4-6 3.4-7 3.4-8 3.4-9 3.4-10 3.4-11 3.4-12 3.4-13 3.4-14 3.4-15 3.4-16 3.4-17 3.4-18 3.4-19 3.4-20 3.4-21 3.4-21A 3.4-21 B 3.4-21 C 3.4-21 D 3.4-22 3.4-23 3.4-24 3.4-25 3.4-26 3.4-27 3.4-28 3.4-29 3.4-30 3.4-31 3.4-32 3.4-33 3.5-1 3.5-2 3.5-3 149 204 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 149 183 183 183 183 183 158 149 149 149 149 149 149 195 149 149 149 149 149 149 149

CRYSTAL RIVER UNIT 3 ITS LOEPage 3 of 5 10120/03 IMPROVED TECHNICAL SPECIFICATIONS List of Effective Pages (Through Amendment 211)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page Amendment Page Amendment 3.5-4 182 3.7-13 149 3.5-5 178 3.7-14 149 3.5-6 149 3.7-15 182 3.5-7 161 3.7-16 182 3.5-8 149 3.7-17 182 3.5-9 149 3.7-18 182 3.5-10 149 3.7-19 182 3.6-1 149 3.7-20 182 3.6-2 191 3.7-21 182 3.6-3 156 3.7-22 182 3.6-4 149 3.7-23 149 3.6-5 149 3.7-24 199 3.6-6 149 3.7-25 199 3.6-7 156 3.7-26 199 3.6-8 156 3.7-27 149 3.6-9 209 3.7-28 193 3.6-10 209 3.7-29 149 3.6-11 209 3.7-30 193 3.6-12 209 3.7-31 193 3.6-13 209 3.7-32 193 3.6-14 149 3.7-33 193 3.6-15 149 3.7-34 149 3.6-16 149 3.7-35 149 3.6-17 149 3.7-36 149 3.6-18 149 3.7-37 200 3.6-19 149 3.7-38 199 3.6-20 149 3.7-39 192 3.6-21 149 3.7-40 192 3.7-1 149 3.8-1 149 3.7-2 149 3.8-2 207 3.7-3 149 3.8-3 207 3.7-4 149 3.8-4 207 3.7-5 149 3.8-5 149 3.7-6 149 3.8-6 163 3.7-7 149 3.8-7 149 3.7-8 149 3.8-8 207 3.7-9 182 3.8-9 149 3.7-10 182 3.8-10 207 3.7-11 182 3.8-11 149 3.7-12 182 3.8-12 149

CRYSTAL RIVER UNIT 3 IMPROVED TECHNICAL SPECIFICATIONS ITS LOEPage 4 of 5 10/20/03 List of Effective Pages (Through Amendment 211)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page 3.8-13 3.8-14 3.8-15 3.8-16 3.8-17 3.8-18 3.8-19 3.8-20 3.8-21 3.8-22 3.8-23 3.8-24 3.8-25 3.8-26 3.8-27 3.8-28 3.8-29 3.8-30 3.8-31 3.8-32 3.8-33 3.8-34 3.9-1 3.9-2 3.9-3 3.9-4 3.9-5 3.9-6 3.9-7 3.9-8 3.9-9 3.9-10 3.9-11 3.9-12 4.0-1 4.0-2 4.0-3 5.0-1 5.0-2 5.0-3 Amendment 149 163 163 163 149 149 149 149 149 149 149 149 149 149 149 149 149 149 182 182 149 149 149 149 152 208 149 149 149 149 149 149 208 149 210 193 193 201 201 149 Page 5.0-4 5.0-5 5.0-6 5.0-7 5.0-8 5.0-9 5.0-10 5.0-11 5.0-12 5.0-13 5.0-14 5.0-14A 5.0-15 5.0-16 5.0-17 5.0-17A 5.0-18 5.0-19 5.0-20 5.0-21 5.0-22 5.0-23 5.0-23A 5.0-23B 5.0-24 5.0-25 5.0-26 5.0-27 5.0-28 5.0-29 5.0-30 5.0-31 Appendix B -Part II 1-1 2-1 3-1 Amendment 149 149 149 149 149 201 149 191 149 149 188 188 198 180 198 198 199 199 199 185 201 204 199 199 149 180 180 149 149 191 201 149 190 190 190

CRYSTAL RIVER UNIT 3 IMPROVED TECHNICAL SPECIFICATIONS ITS LOEPage 5 of 5 10120/03 List of Effective Pages (Through Amendment 21 1)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page Amendment Paae Amendment 3-2 4-1 4-2 190 190 190

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage I of 8 4/14/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Paae Revi si on Paae Revi si on B 2.0-1 34 B 3.1-16 23 B2.0-2 34 B3.1-17 149 B2.0-3 34 B3.1-18 149 B 2.0-4 34 B 3.1-19 149 B 2.0-5 201 B 3.1-20 149 B 2.0-6 201 B 3.1-21 149 B 2.0-7 34 B 3.1-22 149 B 2.0-8 34 B 3.1-23 149 B 2.0-9 37 B 3.1-24 149 B 2.0-10 37 B 3.1-25 149 B 3.0-1 149 B 3.1-26 149 B 3.0-2 149 B 3.1-27 149 B 3.0-3 149 B 3.1-28 149 B 3.0-4 149 B 3.1-29 149 B 3.0-5 149 B 3.1-30 149 B 3.0-6 149 B 3.1-31 149 B 3.0-7 149 B 3.1-32 149 B 3.0-8 149 B 3.1-33 149 B 3.0-9 149 B 3.1-34 149 B 3.0-1 0 149 B 3.1-35 35 B 3.0-11 149 B 3.1-36 35 B 3.0-12 149 B 3.1-37 149 B 3.0-13 28 B 3.1-38 149 B 3.0-14 28 B 3.1-39 149 B 3.0-15 28 B 3.1-40 149 B 3.0-16 28 B 3.1-41 149 B 3.0-17 203 B 3.1-42 149 B 3.0-18 203 B 3.1-43 149 B 3.0-19 203 B 3.1-44 149 B 3.1-1 149 B 3.1-45 149 B 3.1-2 149 B 3.1-46 149 B 3.1-3 22 B 3.1-47 149 B 3.1-4 17 B 3.1-48 149 B 3.1-5 149 B 3.1-49 149 B 3.1-6 149 B 3.1-50 149 B 3.1-7 149 K

B 3.1-51 149 B 3.1-8 149 B 3.1-52 149 B 3.1-9 149 B 3.2-1 149 B 3.1-10 149 B 3.2-2 45 B 3.1-11 149 B 3.2-3 149 B 3.1-12 149 B 3.2-4 22 B 3.1-13 149 B 3.2-5 149 B 3.1-14 149 B 3.2-6 149 B 3.1-15 149 B 3.2-7 149

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 2 of 8 4114104 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page B 3.2-8 B 3.2-9 B 3.2-10 B 3.2-11 B 3.2-12 B 3.2-13 B 3.2-14 B 3.2-15 B 3.2-16 B 3.2-17 B 3.2-1 8 B 3.2-19 B 3.2-20 B 3.2-21 B 3.2-22 B 3.2-23 B 3.2-24 B 3.2-25 B 3.2-26 B 3.2-27 B 3.2-28 B 3.2-29 B 3.2-30 B 3.2-31 B 3.2-32 B 3.2-33 B 3.2-34 B 3.2-35 B 3.2-36 B 3.2-37 B 3.2-38 B 3.2-39 B 3.2-40 B 3.2-41 B 3.2-42 B 3.2-43 B 3.2-44 B 3.3-1 B 3.3-2 B 3.3-3 B 3.3-4 B 3.3-5 B 3.3-6 B 3.3-7 Revi si on 149 149 45 149 45 22 149 149 45 44 149 45 149 149 149 149 45 149 44 45 149 149 149 149 149 149 149 149 45 149 149 45 149 149 149 149 45 37 07 07 07 07 07 07 Page Revision B 3.3-8 B 3.3-9 B 3.3-10 B 3.3-11 B 3.3-12 B 3.3-13 B 3.3-14 B 3.3-15 B 3.3-16 B 3.3-17 B 3.3-18 B 3.3-19 B 3.3-20 B 3.3-21 B 3.3-22 B 3.3-23 B 3.3-24 B 3.3-25 B 3.3-26 B 3.3-27 B 3.3-28 B 3.3-29 B 3.3-30 B 3.3-31 B 3.3-32 B 3.3-33 B 3.3-34 B 3.3-35 B 3.3-36 B 3.3-37 B 3.3-38 B 3.3-39 B 3.3-40 B 3.3-41 B 3.3-42 B 3.343 B 3.3-44 B 3.345 B 3.346 B 3.347 B 3.348 B 3.349 B 3.3-50 B 3.3-51 178 178 07 07 30 07 22 40 22 22 22 07 178 178 07 20 31 07 07 23 07 26 23 149 149 149 149 149 149 149 149 149 149 149 149 149 182 11 178 178 22 23 11 178

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 3 of 8 4/14/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Paae Revision Paae Revi si on B 3.3-52 178 B 3.3-96 31 B 3.3-53 27 B 3.3-97 194 B 3.3-54 178 B 3.3-98 23 B 3.3-55 11 B 3.3-99 23 B 3.3-56 23 B 3.3-100 149 B 3.3-57 07 B 3.3-101 17 B 3.3-58 16 B 3.3-102 149 B 3.3-59 23 B 3.3-103 149 B 3.3-60 23 B 3.3-104 149 B 3.3-61 149 B 3.3-105 149 B 3.3-62 16 B 3.3-106 17 B 3.3-63 37 B 3.3-107 17 B 3.3-64 149 B 3.3-108 149 B 3.3-65 149 B 3.3-109 149 B 3.3-66 149 B 3.3-110 149 B 3.3-67 149 B 3.3-111 17 B 3.3-68 149 B 3.3-112 149 B 3.3-69 202 B 3.3-113 35 B 3.3-70 39 B 3.3-114 43 B 3.3-71 206 B 3.3-115 43 B 3.3-72 157 B 3.3-116 43 B 3.3-73 17 B 3.3-117 17 B 3.3-74 07 B 3.3-118 37 B 3.3-75 07 B 3.3-119 199 B 3.3-76 24 B 3.3-120 199 B 3.3-77 24 B 3.3-121 199 B 3.3-78 07 B 3.3-122 199 B 3.3-79 07 B 3.3-123 199 B 3.3-80 24 B 3.3-124 22 B 3.3-81 23 B 3.3-125 22 B 3.3-82 07 B 3.3-125A 39 B 3.3-83 25 B 3.3-125B 46 B 3.3-84 07 B 3.3-126 22 B 3.3-85 17 B 3.3-127 22 B 3.3-86 17 B 3.3-128 22 B 3.3-87 07 B 3.3-129 178 B 3.3-88 07 B 3.3-130 178 B 3.3-89 07 B 3.3-131 22 B 3.3-90 17 B 3.3-132 22 B 3.3-91 17 B 3.3-133 22 B 3.3-92 17 B 3.3-134 22 B 3.3-93 20 B 3.3-135 22 B 3.3-94 23 B 3.3-136 22 B 3.3-95 07 B 3.3-137 174

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 4 of 8 4114/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Paqe Revi si on Page Revision B 3.3-138 174 B 3.4-28 22 B 3.3-138A 174 B 3.4-29 22 B 3.3-138B 174 B 3.4-30 14 B 3.3-138C 177 B 3.4-31 22 B 3.3-138D 177 B 3.4-32 22 B 3.3-139 11 B 3.4-33 149 B 3.3-140 11 B 3.4-34 149 B 3.3-141 11 B 3.4-35 149 B 3.3-142 11 B 3.4-36 149 B 3.3-143 23 B 3.4-37 29 B 3.3-144 23 B 3.4-38 29 B 3.3-145 07 B 3.4-39 47 B 3.3-146 196 B 3.4-40 29 B 3.3-147 07 B 3.4-41 47 B 3.3-148 07 B 3.4-42 29 B 3.3-149 23 B 3.4-43 161 B 3.3-150 196 B 3.4-44 149 B 3.4-1 149 B 3.4-45 161 B 3.4-2 204 B 3.4-46 149 B 3.4-3 149 B 3.4-47 149 B 3.4-4 149 B 3.4-48 149 B 3.4-5 149 B 3.4-49 149 B 3.4-6 149 B 3.4-50 149 B 3.4-7 149 B 3.4-51 149 B 3.4-8 149 B 3.4-52 161 B 3.4-9 149 B 3.4-52A 161 B 3.4-10 149 B 3.4-52B 183 B 3.4-11 183 B 3.4-52C 183 B 3.4-12 149 B 3.4-520 183 B 3.4-13 149 B 3.4-52E 183 B 3.4-14 149 B 3.4-52F 183 B 3.4-15 149 B 3.4-52G 183 B 3.4-16 183 B 3.4-52H 183 B 3.4-17 149 B 3.4-521 183 B 3.4-18 149 B 3.4-52J 183 B 3.4-19 149 B 3.4-52K 183 B 3.4-20 149 B 3.4-52L 183 B 3.4-21 149 B 3.4-53 10 B 3.4-22 149 B 3.4-54 37 B 3.4-23 14 B 3.4-55 158 B 3.4-24 149 B 3.4-55A 10 B 3.4-25 149 B 3.4-55B 10 B 3.4-26 149 B 3.4-56 10 B 3.4-27 22 B 3.4-57 30

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 5 of 8 4114/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Paae B 3.4-58 B 3.4-59 B 3.4-60 B 3.4-61 B 3.4-62 B 3.4-63 B 3.4-64 B 3.4-65 B 3.4-66 B 3.4-67 B 3.4-68 B 3.4-69 B 3.4-70 B 3.4-71 B 3.4-72 B 3.4-73 B 3.4-74 B 3.5-1 B 3.5-2 B 3.5-3 B 3.5-4 B 3.5-5 B 3.5-6 B 3.5-7 B 3.5-8 B 3.5-9 B 3.5-10 B 3.5-11 B 3.5-12 B 3.5-13 B 3.5-14 B 3.5-15 B 3.5-16 B 3.5-17 B 3.5-18 B 3.5-19 B 3.5-20 B 3.5-21 B 3;5-22 B 3.5-23 B 3.5-24 B 3.5-25 B 3.5-26 B 3.5-27 Revi si on 149 149 149 149 149 33 33 179 149 149 149 149 195 37 37 37 37 149 149 17 17 149 149 149 17 182 37 06 22 37 06 25 24 24 182 22 149 161 161 149 23 149 149 149 Paae B 3.5-28 B 3.5-29 B 3.5-30 B 3.6-1 B 3.6-2 B 3.6-3 B 3.6-4 B 3.6-5 B 3.6-6 B 3.6-7 B 3.6-8 B 3.6-9 B 3.6-10 B 3.6-11 B 3.6-12 B 3.6-13 B 3.6-14 B 3.6-14A B 3.6-15 B 3.6-16 B 3.6-17 B 3.6-18 B 3.6-19 B 3.6-20 B 3.6-21 B 3.6-22 B 3.6-23 B 3.6-24 B 3.6-25 B 3.6-26 B 3.6-27 B 3.6-28 B 3.6-29 B 3.6-30 B 3.6-31 B 3.6-32 B 3.6-33 B 3.6-34 B 3.6-35 B 3.6-36 B 3.6-37 B 3.6-38 B 3.6-39 B 3.6-40 Revision 149 149 22 156 37 156 191 37 156 37 48 26 01 01 01 156 156 37

. 11 156 37 44 44 26 209 209 209 156 209 209 156 37 30 01 01 149 149 149 23 23 165 34 34 17

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 6 of 8 4/14104 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through AmTiendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 License, only, and did not effect changes to the ITS LCOs or Bases.

Operating Page B 3.6-41 B 3.6-42 B 3.6-43 B 3.6-44 B 3.6-45 B 3.6-46 B 3.6-47 B 3.6-48 B 3.6-49 B 3.7-1 B 3.7-2 B 3.7-3 B 3.7-4 B 3.7-5 B 3.7-6 B 3.7-7 B 3.7-8 B 3.7-9 B 3.7-10 B 3.7-11 B 3.7-12 B 3.7-13 B 3.7-14 B 3.7-15 B 3.7-16 B 3.7-17 B 3.7-18 B 3.7-18A B 3.7-18B B 3.7-19 B 3.7-20 B 3.7-21 B 3.7-22 B 3.7-23 B 3.7-23A B 3.7-23B B 3.7-24 B 3.7-25 B 3.7-26 B 3.7-27 B 3.7-28 B 3.7-29 B 3.7-30 B 3.7-31 Revision 23 24 44 44 165 149 37 149 149 149 149 149 149 149 149 17 37 37 18 149 37 17 17 17 01 01 17 23 01 37 12 12 37 22 182 182 182 182 26 26 24 149 149 22 Page Revi si on B 3.7-32 B 3.7-33 B 3.7-34 B 3.7-35 B 3.7-36 B 3.7-37 B 3.7-38 B 3.7-39 B 3.7-40 B 3.7-41 B 3.7-42 B 3.7-43 B 3.7-44 B 3.7-45 B 3.7-46 B 3.7-47 B 3.7-48 B 3.7-49 B 3.7-50 B 3.7-51 B 3.7-52 B 3.7-53 B 3.7-54 B 3.7-55 B 3.7-56 B 3.7-57 B 3.7-58 B 3.7-59 B 3.7-60 B 3.7-61 B 3.7-62 B 3.7-63 B 3.7-64 B 3.7-65 B 3.7-65A B 3.7-65B B 3.7-66 B 3.7-67 B 3.7-68 B 3.7-69 B 3.7-70 B 3.7-71 B 3.7-72 B 3.7-73 24 16 149 149 149 149 24 24 22 182 182 182 24 24 149 182 182 24 22 149 182 163 24 24 17 149 38 149 199 199 199 199 199 199 199 199 37 25 37 193 23 42 42 42

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 7 of 8 4/14/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page B 3.7-74 B 3.7-75 B 3.7-76 B 3.7-77 B 3.7-78 B 3.7-79 B 3.7-80 B 3.7-81 B 3.7-82 B 3.7-83 B 3.7-84 B 3.7-85 B 3.7-86 B 3.7-87 B 3.7-88 B 3.7-89 B 3.7-90 B 3.7-91 B 3.7-92 B 3.7-93 B 3.7-94 B 3.7-95 B 3.7-96 B 3.7-97 B 3.7-98 B 3.8-1 B 3.8-2 B 3.8-3 B 3.8-3A B 3.8-3B B 3.8-4 B 3.8-5 B 3.8-6 B 3.8-7 B 3.8-8 B 3.8-9 B 3.8-10 B 3.8-10A B 3.8-10B B 3.8-10C B 3.8-11 B 3.8-12 B 3.8-13 B 3.8-14 Revi si on 42 208 42 25 37 37 37 17 17 149 149 182 49 200 182 182 192 25 192 25 192 39 39 39 39 149 17 163 163 163 17 160 43 207 207 182 207 207 207 207 182 182 182 182 Page Revision B 3.8-15 B 3.8-16 B 3.8-17 B 3.8-18 B 3.8-19 B 3.8-20 B 3.8-21 B 3.8-22 B 3.8-23 B 3.8-24 B 3.8-25 B 3.8-26 B 3.8-27 B 3.8-28 B 3.8-29 B 3.8-30 B 3.8-31 B 3.8-32 B 3.8-33 B 3.8-33A B 3.8-33B B 3.8-34 B 3.8-35 B 3.8-36 B 3.8-37 B 3.8-38 B 3.8-38A B 3.8-39 B 3.8-40 B 3.8-41 B 3.8-42 B 3.8-43 B 3.8-44 B 3.8-45 B 3.8-46 B 3.8-47 B 3.8-48 B 3.8-49 B 3.8-50 B 3.8-51 B 3.8-52 B 3.8-53 B 3.8-54 B 3.8-55 182 149 149 163 22 43 43 207 43 149 17 160 149 149 43 182 182 182 37 163 163 163 163 39 39 39 39 34 34 34 34 08 23 23 08 34 34 149 149 149 149 149 149 149

CRYSTAL RIVER UNIT 3 ITS Bases LOEPage 8 of 8 4114/04 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 211 and ITS Bases Revision 49)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189 and 190 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page B 3.8-56 B 3.8-57 B 3.8-58 B 3.8-59 B 3.8-60 B 3.8-61 B 3.8-62 B 3.8-63 B 3.8-64 B 3.8-65 B 3.8-66 B 3.8-67 B 3.8-68 B 3.8-69 B 3.B-70 B 3.8-71 B 3.8-72 B 3.8-73 B 3.8-74 B 3.8-75 B 3.8-76 B 3.8-77 B 3.8-78 B 3.8-79 B 3.8-80 B 3.9-1 B 3.9-2 B 3.9-3 B 3.9-4 B 3.9-5 B 3.9-6 B 3.9-7 B 3.9-8 B 3.9-9 B 3.9-10 B 3.9-11 B 3.9-12 B 3.9-13 B 3.9-14 B 3.9-15 B 3.9-16 B 3.9-17 B 3.9-18 B 3.9-19 Revi si on 149 149 149 17 149 149

-149 149 149 149 149 149 149 182 26 182 26 26 182 149 149 149 149 149 149 149 149 149 149 33 33 07 33 208 43 43 43 43 149 149 149 149 149 149 Page Revi si on B 3.9-20 B 3.9-21 B 3.9-22 B 3.9-23 B 3.9-24 B 3.9-25 149 149 149 208 208 208

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT C REPLACEMENT CR-3 ITS BASES PAGES

I;,

is Regulating Rod Insertion Limits B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1, Regulating.Rod Insertion Limits BASES I

: -, 1 17

-1.,,

I BACKGROUND The insertion limits of the regulating rods are initial conditiori'asstimptions 'used in all safety analyses that assume rod'insertion-upon reactor trip. The insertion limits directly~affect the core power distributions, the

'worth of.,'apotential ejected 'rod, the assumptions of

'available'SDM,,and'the initial'reactivity insertion rate.

.'!The applicabletcriteria for these'reactivity and power distribution design requirements are described in FSAR Section 1.4, Criterions 6, "Reactor-Core Design" and 29, "Reactivity'Shutdown Capability", (Ref. 1), and in 10 CFR 50:46; "Acceptance"Criteria.,for;Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

,Limits on regulating rod insertion have been established, and all-rod positions a're monitored and controlled during P

6ower o"pieration' to ensure that the'power distribution and

'reactivity:limits defined by the-design power peaking and

'SDM limits are! not violated.

.The' regulating.rod groups operate with'a predetermined amount of position overlap,in order :to approximate a linear relation 'between rod:worth and rod position (integral rod worth).

To achieve this approximately linear relationship, the regulating rod groups are withdrawn and operated, in a pre'dete rmince Pr ned sequence' The'automatic control'system nt'rolsreactivity by mioving th'e regulating rAbd groups in sequence within analyzed'r nges.

The [group sequence and overlap limits are'specified in'the COLR.

The regulating rods-are'-used for-precise reactivity control of the reactor.,The positions-of the regulating rods are

normally controlled automatically by the Integrated Control System (ICS) but can also be controlled manually.

They are i-

capable of adding reactivitylquickly compared with borating or'dilutingitheReactor Coolant System (RCS).

The power density at any point in'the core must be limited to maintai n'specified acceptable fuel design limits, '

including limits thatensure that the criteria specified in 10 CFR 50.46 (Ref. 2)*are not violated.' Together, (continued)

Crystal River Unit 3 B 3.2-1 Amendment No.

149

-I Regulating Rod Insertion Limits B 3.2.1 BASES BACKGROUND LCO 3.2.1, "Regulating Rod Insertion`-Limits," LCO 3.2'.2, (continued)

"AXIAL POWER SHAPING ROD (APSR) Insertion Limits,"

LCO 3.2.3,."AXIAL POWER IMBALANCE Operating Limits,":aand LCO 3.2.4, "QUADRANT POWER TILT.(QPT)," provide limits on control component operation,and on'monitored process variables toensure-that the'core,operates within the FQ(Z) and FN -limits in the COLR. Operation within the FQ(Z)

AH

'limits given'in theC LR prevents power peaks that would exceed the loss 'of coolant' accident' (LOCA) limits derived from the analysis oof the Emergency Core Cooling Systems

.(ECCS).,- Operation.,within the FNlimitsgiven in the COLR AH prevents departure from nucleate boiling (DNB) during a loss

of forcod reactor:cioolAnt'flw accident. -In addition to the

?

.F4

and F^N limits,.certain,,reactivity-limits are met by AH regulating rod insertion limits.

The regulating rod insertion"limits al'so'restrict-the ejected CONTROL ROD worth to the values assumed.in the safety-analysis and maintain the minimum required SDM-in MODES Wand 2.

-This LCO is required' to minimize fuel cladding failures that breach the primary fission product'barrier and release fission products into'the rea'ctor'coolant in the event of a LOCA, loss of fl w-4c6ident, ejected 'rod accident, or other

'postulated accidents 'requiring' termination by a Reactor Protection' System'trip function.

APPLICABL'E The fuel claddinig must not sus~taindamage as a result of SAFETY.ANALYSES', normal operation (Condition I)r'"6a"nti:cipated operational r

ioccurrenceserti ConditiAon II)n,' Th'e LCOs 'governing regulating

"','.',odi erSAXIAL,POWER,'IMBALANCE, and QPT preclude' core power distributions that.-violate the following fuel design criteria',

a.

'During a large break LOCA, the-peak cladding temperature must not exceed!22000F (Ref. 2);

,..b.:. During a,lossof..forced-reactor-coolant flow accident, there, must be.at-least 95%.probabi-lity at the 95%

confidence level (the 95/95.DNB criterion) that the hot fuel rod, in, the core.does.not experience a DNB

.....c.

',conditionCRe'ls.

7 'and 8);.and

c. -During an ejected rod accident, the fuel enthalpy must not'exceed 280 cal/gm'(Ref. 3).

(conti nued)

Crystal River Unit 3 B 3.2-2 Revision No.

45

ri Regulating Rod Insertion Limits BASES ACTIONS' D.1 (continued)

If. th6 regulating-rods-cannot be restored to within the

'acceptable.operating limits for -the original THERMAL POWER,

-or if the:power reduction cannot be completed within the

'associated Completion Time, then the plant must be is placed

..in a.MODE in-which-.this LCO does not.apply. This Action

,-,-ensures-that.the'reactor does not continue operating in viol ati on fthe peaking' limits, thelejected rod worth, the reactivity-insertion rate assumed'as initial conditions in the accident analyses,'or'the required minimum SDM assumed in the:'accident analyses'. To achieve'this status, the plant r

7must be'placed in at.least MODE 3.within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The i;;'allowed.Completion-7 me reaso'able,',based on operating

-. experience, to redch'the 'required plant conditions in an M

orderly manner' ahd withou 'ch'allengin'g'plant systems.

4 's0I

'SURVEILLANCE-'

SR 3.2.1.,'."

REQUIREMENTS' This Suryveilance':ensureswthat the'sequence and overlap

... limitsare' not.violated.. A Surveillance Frequency of

-.12.hours-or4 hours.-igs isrequired depending on-whether or not

-:l..,

the CONTROL ROD drivye sequenceialarm is OPERABLE.

Verification that 'thle sequenceand overlap are within limits at a 12 ho'ur-Frequency.is..sufficient to ensure these limits are preserved. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months :Completion.time is acceptable

- because little r6d"mctibnfl ccurs ii'4',

hours duetofuel burunup,'and thdpr6bob'ility of: a:"devi'tion occurring:

l"

simultaneously with an inoperable' sequence -mo'nitor in 'this ativel short.tirme'frame-is low.

Both Frequencies take

' ' ~':

intoc'acc4ount the level 'of information' available in the contrd 6

r0oom fo'mo nNitoring -the staitus`.'of the regulating rods.

/"..-

, '-,SR'3.2.1.2.

.:w':' '-

With'anOPERABLEreg'ulating rod insertion limit alarm,

-/

' verification of the 1regulati'nd 'rod insertion limits as

. '."'raspecified.'inA-the;COLR-'at'a'FFrequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'is

-sufficient t`6dete'ct'.regiulating 'rod banks that may be approaching theng'roUi-Insertion' limits, because little rod i

'.c.,;

.(conti nued)

Crystal River Unit 3 B 3.2-9 Amendrhent No.

149

r

.Regulating Rod Insertion Limits B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS motion due:to:fuel burnup occurs in.12:hours.

If the insertion limit alarmbecomes inoperable, verification of the regulating rod group-position.at a2Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is-sufficient to detect'wheth'er. the-regulating rod groups may:be'approaching or.,:exceedin4g 'their'group insertion

..-,l~imits.

Both Frequencies take intoaccount the level of information availabl:e in the control room for monitoring the status.of. the reguiating rods.'

SR3.2.'.3 Prior to-achievi.ng-,criticality',;.an estimated critical

.:: position forithe',CONTROL::RODS or.estimated critical boron

concentration-is>;determined. -Verification that SDM meets the minimum.requirements.ensures that sufficient SDM capability exists with the CONTROL RODS at the estimated critical position if it is necessary'to shut down or trip the reactor after criticality.'The`,Frequency of4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to criticality-provides sufficient time to verify SDM capability and m'ake ihe reactor' critical.

-REFERENCES

.. 1.

,. FSAR,, Section;1.4..

2. 10 CFR 50.46,.,

...- g.s

3.,FSAR, Section.14.2.2.4.:

4..

FSAR,. Section.l3.1.2.2.'

5.-

FSARj, Section 14.

6. CR-3 COLR

7.

BAW-10143P-A, Rev..1, "BWC,.Correlation of Critical Heat Flux", April 1985.

I. -.

8:.

BAW-10241P-OO;,BHTP;DNB'Correlation Applied with Crystal River Unit 3 B 3.2-10 Revision No. 45

,~

r'-.APSR Insertion Limits B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2-AXIAL POWER SHAPING ROD (APSR) Insertion Limits BASES

' BACKGROUND :

The insertion-limits of the APSRs are initial condition assumptions in all safety analyses that'are affected by core

'.power, distributions'-..The'applicable criterion for these "power distribution design requirements 'are FSAR Section 1.4, Criterion 6, "Reactor-Core Design", (Ref. 1), and

.- 10 CFR :50.46; "Acceptance Criteria for. Emergency Core

.Cooling'Systems.for Light Water Nuclear Power Plants" (Ref. 2).'

.A..PS..R5X in..ser i

h h

bee

.e.s Limits on-:APSR insertionhave been established, and all APSR

.positions a'r'e 'mon'itor6d and:controlled during power

.operatio'n-'to'ensure';thatrthe"p6we'r

distribution defined by the design' power peakii g limitsi'is maintained.

S I.';t,

S!,

-toa epta

-etfelcodesg msbelimits, The.power-density.at:-any.point in the core must be limited

... to maintain~s'pi'cif~iedlacceptabl'e-fuel..,design limits,

including limits.that. meet the criteria specified in Reference' 2.Together','LCO 3.2.1,. "Regulating Rod Insertion mi i ',"

LCO'3.2.'2, -"AXIAL'POWER SHAPING ROD (APSR)

Insertion.Limi.ts,"' LCO3.2.3,.. "AXIAL'.POWER IMBALANCE Operating Limits," 'and LCO 3.2.4,. "QUADRANT POWER TILT (QPT)," provide limifts'n' control component operation'and on

-monitored. process variables to ensure that the core operates within the FQ(Z) and' FN limits.in the COLR.

Operation AH within the FQ(Z)-li'mits given in'the COLR prevents power

--peaks thatexceed the loss;of. coolant accident (LOCA) limits derived from the analysis of the Emergency Core Cooling Systems (ECCS).

Operation within the FN limits given in AH the COLR prevents departure from nucleate boiling (DNB).

duringa loss-of forced'reactor coolant flow accident.

The APSRs are -not'requi red'for'.reactivity'insertion rate on trip

or SDM and, !

heireforthey'do not trip'upon a reactor trip.

This LCO is required to minimize fuel cladding failures that would 'breach..:the primary fission product barrier and release fission-products tothe r'eactor coolant in the event of a

-LOCA,'loss of flow accident, ejected'-rod accident, or other postulated'accident requiring termination by a Reactor Pro'tection-System-trip function.'

(continued)

Crystal River Unit 3 B 3.

2-1 1

Amendment No.

149

-I--

.. r 17; j ! i--. ' ', !

I

, i l

APSR Insertion Limits B 3.2.2 BASES APPLICABLE SAFETY ANALYSES The fuel cladding.must'not'sustain'damage'as a result of normal operation (Condition I) or anticipated operational occurrences (ConditionaII)." Acceptance criteria for the safety and.regulating rod,.inserti.on, APSR position, AXIAL

.POWER IMBALANCE, andQPT LCOs.preclude core power distributions thatyviolate-the following fuel design criteria:

a.

During a large break

-temperature must:not

e.

LOCA, the 'peak cladding exceed 2200'F (Ref. 2);

'I i* 1.'I!

-'b.

EDuring a`loss-`of forced'.reactor coolant flow accident,

there must -be'at least.95% probability at the 95%

confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB

-condition (Refs..- 4 and 5); and,

.-....:, -c.., Duringqan ejected rod accident-,,the fuel enthalpy must

.,p not exceed 280._cal/gm;,(Ref,.-:3)...

Fuel; cladding'darmage.does-nottoccur~when the core is operated outside,these: LCOs during.no'rmal operation.

However,'fuel. claddihg.damage could 'result should an accident-'occur'sim'ultaneously' with violation of one or more

.'ofithese LCOs'

-This potential for fuel cladding damage

.' 'exists because changes inth pow-r-distribution can cause increased power peaking and'-corresp'onding increased local linear heat rates.

Operation at the APSR' insertionlimits'may approach the

.maximum allowable linear heat generation rate or-.peaking factor.,with-the allowed QPT, present.

The.APSR. insertion limits-satisfy.Criterion 2 of the NRC Policy Statement"..,0;'

I I 1

.4.

I:.

I f

ILCO j.'

_ 1!,:-

':1 The l'imi s on APSR'physical ins'ertion. as defined in the COLR must b'e maintainedbecauslethey'serve the function of controlling-the' power'distribution within an acceptable

' range.

Measurement system-independent limits.for APSR insertion are provided-in the COLR'.'. Measurement-'system-dependent maximum allowable setpoints'are'.7der,ived by adjustment of the, measurement system-independent limits to allow for THERMAL POWER level'uncertainty and rod position errors.

(continued)

Crystal' River Unit 3

. - B ;.2-12 I Revision No. 45

APSR Insertion Limits B 3.2.2 BASES ACTIONS A.1 and A.2 (continued) depletion with thexAPSRs-in positions that have'not-been

' 'analyzddthereby limiting th'e potential for an adverse

' xenon redistribution.* This time limit also ensures that the

'intended burnup-distribution'is maintained, and allows the

'operator sufficient time to-r'eposition the APSRs to correct their.positions.

Because the APSRs are not operated by the automatic control system, manual action -by the operator is required to restore

- the.APSRs to the positions specified in the COLR.

B.1 4,

If the APSRs cannot be restored to their intended positions within'the'associated&Completion'Time,' then the plant must be placed'in"a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least

'MODE'3within 6'hours' This action ensures that the fuel doesjnot continue tod',be-depleted in an unintended burnup distributip`.iThe, hloWed'Completion Time is reasonable, ba's'bd on'operating experience, to reach the required plant condiaopns in an orderly manner. and without challenging pi plant-sy~stem-s',,- ;

',sz.

plat:sitem..

j SURVEILLANCE, SR 3.2.2.1 REQUIREMENTS,

t.

Verification that th'e'APSRs are within their insertion limits at a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is sufficient to ensure that

' ' th'eIAPSR insrtibn'iinits 'are preserVed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency required for' per'forming this verification is sufficient because APSRs are positioned by manual control and are normally moved-infrequently.- The-probability of a

~,,

,,.deviation occurring~simultaneouslywith a non-functioning

, APSR 9position c.omrputir. alarm is low in 'this relatively short

,.tjme'frame. Also, Pie'Frequency takes into account other information available'in the control room for monitoring the axial power distribution in the reactor core.

REFERENCES,

1. 'FSAR, Section 1.4.

i CF

.46.

2.

10 CFR 50.46.

(continued)

I Crystal River Unit 3 B -3.2-15 Amendment No.

149

-..i APSR Insertion Limits B 3.2.2 BASES I

REFERENCES (continued)

3. FSAR, Section 14.2.2.4.

-4. BAW-10143P-A, Rev. 1, !'BWC Correlation of Critical Heat Flux", April 1985.

5.

BAW-1O241P-OO,' BHTP DNB. Correlation Applied with LYNXT. l I

S

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. I

'.' -f ':

z..

CV I V 1.1:.:

.'. I:.-.-

I

, - I

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£ 1

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5) 5" l

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River Unit 3 S.,,,

I Crystal B 3.2-16 Revision No. 45

R, AXIAL POWER IMBALANCE Operating Limits B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL-POWER IMBALANCE Operating Limits 1

BASES BACKGROUND This; LCO is'required'to limit the core power distribution based on.'accident".initial condition criteria.

The power density at any point in the core must be limited

'tfo'maintain ispe cifi'd 'acceptable 'fiel design limits,

'including limits'that"satisfy the criteria specified in

.'10 CFRz5: 46:CRef;1):

This' LCO provides limits on AXIAL

-POWER IMBALANCE'to' ensure thatthe-core operates within the FQ(Z) and FN limits, given in the COLR.

Operation within

'AH the FQ(Z) limits given in the COLR prevents power peaks that exceed the lbss-of coolant'accident (LOCA) limits derived

"'from'the analysisIof the Emergency Core Cooling Systems

-,(ECCS).

Operation within.the,FN limits given in the COLR AH prevents departu're fr6m nucleate'boiling (DNB) during a loss

-'of forced-reactor coolant 'flow accident.

This LCO is;required-to limit-fuel cladding failures that

, ',breach-the primary fission product barrier and release fission productsjinto the reactor coolant in the event of a

-LOCA, loss of forced reactor coolant flow accident, or other

. - postulated; accident requiring. termination by a Reactor

.Protection-System-trip-function. This LCO limits the amount of damage to the.fuel cladding, during an accident by maintaining the validity of the assumptions in the safety

-- analyses related-to the-initial power distribution and

.,,reacti vi ty..,.g>..

Fuel,claddingfailure'during a postulated LOCA is limited by restricting the maximum linear'heat rate (LHR) so that the

.'Peak claddingb'-temperature doe's not'exceed 2200F (Ref. 1).

Peak'claddingltemperatures > '2200Fcaause severe cladding

'failure'by-oxidation'due to a zirconium water reaction.

Other criteria must also be-met (e.g.`, maximum cladding oxidation,,maximum.hydrogen,generation, coolable geometry,

; : and longi-terrm cooling).. However,'peak cladding temperature is 'usually most'limiting.

Proximity to'the-DNB'condition is expressed by the departure from nucleate boiling ratio (DNBR)-, defined as the ratio of

-the-cladding su'rface'heat-flux'requ'ir'ed to cause DNB to the actual cladding surface heat flux. -The minimum DNBR value (conti nued)

I Crystal River Unit 3 B 3.2-17

-. Revision No. 44

AXIAL POWER IMBALANCE Operating Limits

'B 3.2.3 BASES BACKGROUND during both no'rmal operationand anticipated transients is (continued) limited to the DNBR.correlation.limit. for the particular fuel design.in use.and is.accepted as~an appropriate margin to DNB. The' DNB.-correlation.'limit ensures that there i's at least 95% probability at the. 95% confidence level (the 95/95 DNB criterion).that the hot fuel-rod in the core does not experience DNB.

The measurement -system independent limits on AXIAL POWER IMBALANCE are determined directlyby,the reload safety evaluation analysis'without adjustment.for measurement system error

-and uncertainty.

Operation beyond these limits could invaljidaite'.the assumptions'used in the accident analyses regardin the core poweiK distribution.

Core protection from'tm e ffect's,:of-'Sxcessive AXIAL POWER IMBALANCE is accomplisheid'j'n a'-tiered approach. The operating envelope limits'addressed by this Specification

'formthe first -ine ofb'defense, are-monitored by the o

operator and compensatory.actions initiated when AXIAL POWER IMBALANCE is.. not. within,,theacceptable.region.

The AXIAL POWER IMBALANCE operating limit envelope contained in the

-.COLR represents-the.-,setpoint,,-for::which the core power

.. dis~tribution~would either exceed.,the LOCA LHR limits or cause a reduction in-.the DNBR'below-the Safety Limit during the loss of flow accident with the-maximum allowable QPT present.andwith the APSR positions consistent with the

, -imitations on APSR'withd rawald determined by the fuel cycle design!and specified,,by-.LCO. 3.2.Z.

t.

t-

The next line of defense is the Reactor Protection System

.(RPS)

Flux/Delta..Flux/Flow,,trip'setpoints.

The trip

setpoints are addressed in; LCO;13.31 andare developed by

.conservativelyadjusting the. AXIAL.OWER:IMBALANCE operating

.lTimits to account;'for measurement.:system inaccuracies and

.'other potential errors.'rThe: tri'p 'setpoints ensure the reactor-will be automatically shutdown:'prior to exceeding a

'Safety Limit.:

The last'line of defense'is the Safety Limit itself (Reference Section':2.0)..:If'the"AXIAL>POWER IMBALANCE protective limit specifi edin 'th"'COLR'is exceeded, then the

-operator must.takes'additional actions-to preclude conditions

.'.underawhich DNB.:.cbuld-'occur.-..Eve'n'with AXIAL POWER IMBALANCE at the Safety Limit, damage to the fuel will not occur.

(continued)

Crytal River Unit 3 B 3.2-18 Amendment No.

149

.AXIAL POWER IMBALANCE Operating Limits B 3.2.3 BASES

BACKGROUND;

!-..Actual alarm setpoints are more'restrictive than the maximum (continued)-. allowable setpoint'values to provide.additional :conservatism between,.the actual.-alarm setpoints and the measurement system independent.;(operating) limit.

I APPLICABLE

'The fuel. cladding must.,not sustain damage as a result of

'SAFETYANALYSES' normal operation '(Condition I),' 'and anticipated operational occurrences (Condition.II). 'The LCOs'based on power distribution,',LCO 3..2.1, "Regulating'Rod Insertion Limits,"

LcO 3.2.2,. "AXIAL'6POER SHAPING ROD,(APSR) Insertion Limits," LCO'3.2.3,',"AXIAL POWER IMBALANCE Operating L~imits"and LCO 3;.2.'4, "QUADRANT POWER TILT (QPT),"

preclude-core-power distributions that would violate the

~~~~~ '0X ;

olwjing-',fuelj desigrn crit'erita.,isl

a.

Duing-a larg'e break LOCA,'

peak'cladding temperature

'"must',not'exceed'2200-F (Ref.-1)','

'-Duri'ng 1ss'offorced reacto6rcoolant flow accident,

"':theremust' be1z'least a 95%'p'obability'at the 95%

',.c6ifide'nc'eleve' (the'9S/95':DNB criterion) that-the 2 'hotfuel rd 'in;hthe'core does'not'experience a DNB ncordition,'(Reff. 2,ra'id 3).

r I

-The 'regulating

.rod 06itions, lthe-APSR positions, the AXIAL

' 8",POWERrIMBALANCE,.and'nthe QPT-,are pro'cess variables that characte'rize'ands coht'r6l the three dimensional power distribution of.the 'reactor core..

j

n e'

§ ;..-

t Fuel-cladding damagedoes not occur.when the core is

-6 operated-outside thisrLCOiduring normal'operation.

However,

1 feli.,cladding.r!damageicould result-should an accident occur with-simultaneous violation of:one6or'more of the LCOs

.. governing;the;.four process variables-.cited above.

This t.',

potential forrfuel '.cladding damage exists because changes in

the p6wer-distribution'can causelincreased power peaking and corresponding increas'ed'local LHRs.

,','The:,'regulating'rod'insertion,'.the APSR.positions, the AXIAL POWER IMBALANCE, 'and.jthe'QPT are:.monitored and controlled during'power operation,tojensure that.'the power distribution is' within thebounds.set...by the bsafety analyses. The axial (continued)

Crystal -River Unit 3 B 3.2-19 Revision No. 45

eA

I AXIAL POWER IMBALANCE Operating Limits B 3.2.3 BASES APPLIC SAFETY (cont

' S.-.

ABLE.

'power distribution is maintained primarily by the AXIAL ANALYSES POWER IMBALANCE and the APSR position-limits; and the radial inued) power distribution is maintained primarily by the QPT limits.

The regulating rod insertion'limits affect both the radial'and axial power distributions..

Theddependence of the core power distribution on burnup, regulating rod insertion,APSR position, and spatial xenon

'distribUtion',is taken into account when the reload safety evaluation ahalysis'is perf6rmed.

'Operation at the-AXIAL POWER IMBALANCE.limitmust be interpreted;.as operatingd.the core at the maximum allowable F (Z)-or FN peaking'. factorsAssUmd as initial conditions the a cident.,

a y wit f6 r.thd-accidentanalyses with the allowed"QPT present.

AXIAL POWER'-IMBALANCE'satisfies,2 Criterion 2 of the NRC Pn1 i rv Ctntamnnt

-v

- y.J i

l;l,

~:

a

-, a

-LCO

.The power:distribution LCO'limits 'have been established

'based on-correla'tions between power peakigadesl measured process('ariables:- regulating rod position, APSR position, AXIAL POWER-IMBALANCE,':and'QPT.

The AXIAL POWER

.IMBALANCE.operatingilimit'envelope contained in the COLR represents the-set'point for which the-core power distribution would:'either exceed'the LOCA LHR limits or cause a' reductionin-.the DNBR-below the Safety Limit during the-loss of flow accident with the allowable QPT present and with the APSR positions' consistent with the limitations on APSR withdrawal: determined.b'. the! fuel cycle design and

.'.specifiedbyLCO3-;2-.2.

L~~~~

I',;..; '

APPLICABILITY In MODE 1, the limits on AXIALUPOWER IMBALANCE must be maintained when THERMAL POWER is >..40% RTP to prevent the core power distribution from exceedingq'the LOCA and loss of fl'ow assumptions used in the accident analyses.

...Applicability of these limits at <,40% RTP in MODE 1 is not

-required because the combination of AXIAL POWER IMBALANCE with the maximum ALLOWABLE THERMAL POWER level will not (continued)

Crystal River Unit 3 I B 3.2-20 Amendment No.

149

r I,

AXIAL POWER IMBALANCE Operating Limits B 3.2.3 BASES ACTIONS B.1 (continued) 4 Completion Time of;2'hou'rs is reasonable' based'on limiting a potentially adver'se.xenon-redistribution, the low probabil'ity'ofn anaccident occu'rring'in this relatively

-shorf timeperiod

-andon operating experience regarding the

- amount of time'required to reach 40% RTP from RTP without "challenging plant systems..

4' 4

I t bidb e

s 6ipi h

opeg'~'~

bra;#lt In SURVEILLANCE ' The AXIAL POWER *IMBALANCE'can'be monitored by both the

'REQUIREMENTS

.-Incore and,Exco're-Detector.Systems..'

The AXIAL POWER

'IMBALANCE'maximumallowable setpoints'in the operating procedure..are7derived:.from.their...corresponding measurement system independent.limits (in"the'COLR) by adjusting for

,. :both the systernobs r~abilit' errors and instrumentation errors.,, Although,:they~are based on the'same measurement system' independent-iinhits,' the setpoints for the different

..tsystems are'not'..identical because of differences in the

'errors applicable for each' of these systems. The

uncertaintyarialysisthat 'defines

the required error

'adjustm ent t'o cohvertrthe measurement-system independent limits.to alarmsetpoints'assumes that!75% of the detectors in each quadrant;'are,;OPERABLE; Detectors located on the

-,core major

-axes..ar'e-:assumed to'cbntribute one half of their

.output to,each'-quadrant;l.detecto'r~s.in.the center assembly

..,.are assumea to.contrtibute.one quarter.;of their output to

..;.,.eachquadrant.:ForAXIAL POWER IMBALANCE measurements using

,.-..< the Incor eDetector,.System,.the.Minimum Incore Detector

, 'System.consists.of detectors:configured as follows:

ra.;'

Nine~detectors.'shall :be arrangedsuch that there are

'three-detecto'rsiiea'chi'of-'hree>'strings and there are three detectors'lying in the same'axial plane, with one plane at-the-core midplane and one plane in each s;,axial.core half;;r,.;.

b The' axiai,lanes

'Jn each core half'shall be A

laboft,'the ecoren'midpflane; and

c.

The.det'ecto'r. 'srtrpg4 shall not have radial symmetry.

iI

'i; t-

',J'-

j!j*, r, (conti nued)

Crystal River Unit 3 i

. I

.. B 3.2-23 Amendmeni No.

149

I;,

I AXIAL POWER IMBALANCE Operating Limits I.

B 3.2.3 BASES SURVEILLANCE REQUIREMENTS (continued)

Figure B 3.2.3-1 (Minimum Incore Detector System for AXIAL POWER IMBALANCE Measurement) depicts an example of this

.configuration-..,This-arrangementis chosen to reduce the uncertainty inthe measurement of the AXIAL POWER IMBALANCE by' the Minimum Incore-Detector System. For example,' the re'quir'ement.forplacing:one detector.'of each of the three strings'at the. core midplane'puts three-detectors in the

central region of the core-where.the' neutron flux tends to be higher., It also',helps prevent measuring an AXIAL POWER IMBALANCE that is excessively large when the reactor is operating at--low THERMAL POWER..levels. 'The 'third'

.."requirement for placement of.detectors (i.e.j radial:.'

, asymmetry).'reduces..ur'certainty by-'meas'uring the neutron flux at core locations',that ar~e not radially symmetric.

' R.:;;

I.

. II

I i,

If the plant'compute'r.becom'es'inoperable, then the Excore

! ' System or-Minimum Incore Detector System may be used to

monitor thei.AXIAL POWER-IMBALANCE. 'Although these systems do.not provide:a.direct.calculation and-display of the AXIAL POWER-IMBALANCEi.:a 1,hour, Frequency:provides reasonable time

- between.calculations for detecting any trends in the AXIAL

'.POWER IMBALANCE thavtmay'exceed its. alarm setpoint and for corectveaction..

.'...- undertaking cor~rectiyeacio When'the AXIAL POWER IMBAL'ANCE'aldrm.is' OPERABLE, the operator receives1an alarm if the AXIAL POWER IMBALANCE i

increases to its'alarm"'setpoint.

Verification of the AXIAL POWER'IMBALANCE ind'i'cation every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ensures that the AXIAL'POWER--IMBALANCE'limits§ are not violated.

This Survei~iance Frequency' is;aicceptable because the mechanisms that can.-cause AXIAL POWER:IMBALANCE',"such as xenon redistribution.or CONTROL ROD drive mechanism malfunctions

.!that. cause,-slow8AXIAL:POWER-IMBALANCE;.increases, would likely be-discovered.-.by, the.operator-before the specified

.limits.are violated.

REFERENCES.

1.

10 CFR,'50.46.:.

'2.' BAW-iO143P-A;,Rev. 1, "BWC Correlation of Critical

- -Heat Flux"-,:;April '1985'.

3.'

BAW-10241P-00' BHTP DNB Correlation Applied with LYNXT.

I Crystal River Unit 3 B 3.2-24 Revision No. 45

-. ',.'i Q PT B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS

--B 3.2.4 QUADRANT--POWER-TILT- (QPT-)------.--

BASES BACKGROUND This LCO is required.to limit the core-power distribution based on',accident !-,ini.tial condition~criteria.

.The powerdensity-atany point inthe'core must be limited

,to maintain specifi'dacceptable fuel design limits, including limits'that,preserve thecriteria specified in 10'CFR 50.46 (Ref.'lj..

Together, LCO:3.2.1, "Regulating Rod Insertion Limits,"'.LCO,3.2.2, "AXIAL POWER SHAPING ROD (APSR).Insertion Limits,".LCO.3.2.3,-"AXIAL POWER IMBALANCE Operating Limits," andLCO 3.24, "QUADRANT POWER TILT (QPT)," provide',limits'on'control component operation and on

, monitored process variables to ensure-that the core operates

-fithinthe -F(Z) 'arndid li itsggiven in'.the COLR.

Operation within.the.F(Z) limits given in the COLR prevents power peaks.that exceed.the loss of coolant accident (LOCA) limitstderived by: Emergency Core Cooling Systems (ECCS) analysis. Operation:, ithin the.FN limits given in the COLR f '-': -.

preven deromncleate'boiling(DNB) during'a loss

'fo'rced'reactor.

coolant flow accident.

-"*¢;

t*

4)-= -.

rC

'.:This'1C-',is,',requiired'--to-'limit'fuel,'"cl'adding failures that br.eachthe.priay ission'productbarrdier and release

.fission.products tothe-reactor.coolant in the event of a LOCA;,loss.of.forcediceactor coolant~flow, or other accident requiring terminationiby,,a Reactor Protection System trip function,'.This"LCO imitsth'e.amount.of damage to the fuel 6

cladding.-during'an.actident by'maintaining the validity of

,.-:the ass-umptions-used

in the safety'anaiysis related to the initial.pow'er'distribution 'and reactivity.

'_.....Fuel'.cladding failure-_during a postulated LOCA is limited-by restricting the'maximuml linear'heat rate (LHR).so that the peak cladding temperature does-not exceed 2200°F-(Ref. 1).

Peak cladding t'emp'eratures,> 2200OF cause severe cladding

' failu're'b'y 'oxidation'.'due toa'.'zir'conium water reaction.

Other criteria' must also be met (e.g.',

maximum cladding

-oxidation,,maximum hyd~rogen generation, coolable geometry, and long termnicooling).

However;,'peak-cladding temperature is usually most ',-imiting'.

(continued)

Crystal River Unit 3

.B 3.2-26

-. Revision No. 44

BASES QPT B 3.

2.4 BACKGROUND

Proximity to the DNB. condition.is expressed by thedeparture (continued) from nucleate boiling ratio (DNBR),

defined as the ratio of the-cladding surface.rheat-.flux required to cause DNB to the actual cladding surface heat flux.

The minimum DNBR value during.both normal operati nwand anticipated transients is limited.to the DNBR correlati6n limit'for the particular fuel design-in use,'and is.accepted as an appropriate margin to.DNB.. The DNBR'correlation limitensures that there is al:

least'95%'probability at the 95%;confidence level (the 95/95.DNB criterion)'that'the hot fuel rod in the core does

.. notvexperience DNB

'The;measurement systemlindependent limits on QPT are determined directly by the reload'safetyevaluation analysis

-without.adjustment.-for'i'measurement-system-error and

'uncertainty.'

Operation beyond.these 'limits could.invalidate core power~distribution assumptions-used in the accident analysis. The error adjusted maximum allowable alarm setpoinits (measureme`nt syste dependent limits) for QPT are specified in the COLR.

APPLICABLE The fuel cladding'must not sustain damage as a result of SAFETY ANALYSES normal.operation.,(Condit.ion'I).and anticipated operational occurrences (Cofldition II):.The"LCOs based on power distribution (LCO 3.2.1, LCO 3.2.2,-LCO 3.2.3, and LCO 3.2.4) preclude-core power distributions that violate

" ' '"'-the"'foliowinglfueldel ign criteria:

a. During'a large break LOCA, the'.peak cladding temperature must not exceed 2200,0 F (Ref. 1).

b.. During a loss of forced reactor coolant flow accident, there must be, at'least'.95%,Oprobability at the 95%

confidence level (the 95/95 DNB criterion) that the hot.,fuelrod in the core does not experience a DNB condition.(Refs. 3'and 4).

QPT'is one of the process variables.that characterize and control-'the three dimensional power distribution of the reactor core.

' Fuel cladding damage does not occur when the core is operated outside this LCO during normal'operation.

However, fuel cladding damage could result'if an accident occurs with (continued)

I Crystal River Unit 3 2

B 3. 2-27 Revision No. 45

QPT B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 (continued)

-REQUIREMENTS After a THERMAL.POWER increase following restoration of the

. QPT to within the steady state limit, QPT must be determined to remain within the steady state limit at the increased THERMAL POWER'level. -'This is accomplished by monitoring QPT for 12;consecutiVe-hourly intervals or until verified acceptable at >:95%:RTP to determine whether the period of any oscillation due to xenon redistribution causes the QPT to exceed the steady.statelimit again;'- In case QPT exceeds the steady state limit forsmore than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or exceeds the transient limit (Condition A, B, or D),

the potential for this xenon redistribution -is greater.

6 REFERENCES'

1.

10OCFR.50.46

2.

BAW 10122A, Re..,

"Normal. Operating Controls",

May 1984..

3.

.BAW-10143P-A,..Rev. 1, "BWC Correlation of.-Critical Heat Flux", April 1985..

'4.'. BAW-4024io-OO,"'BHTP DNB Correlation Applied with

. ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~,

.z

-;.t hs.

I I*

~,

-I

.I Z?

I.

k6 I'

4

~17 I

.*2.

4.

, I B 3.2-36 Revision No. 45 Crystal River Unit 3

Power Peaking Factors B 3.2.5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5- -Power Peaking Factors BASES I,

L

, ~:

,~

~I

-. I I '

..1..

I I

, ..": I, j BACKGR61 7,

UND' The purpose' of'this.LCQ 'is to establish limits that constrain thecore poiwerdistrib'ution.within design limits during normal operation (Condition'l) and during anticipated operational.-occurr'ences' (Co'nditio'n II) such that accident initial.-.conditio'n -pro6ection'cr'iter~ia are preserved.

The accident initial condition 'criteria are preserved by bounding operation at THERMAL POWER Within specified acceptable fuel design limits.

FQ(Z) is a specified acceptable fuel design limit that preserves the initial:conditions for~the Emergency Core

' Cool ing' Systems (ECCS) analysis.; FQ(Z) is defined as the maximum local fuel 'rod linear';power 'density divided by the

average'fuel rod linear power'density, assuming nominal fuel pellet and rod dimensions.

Because 'F'(Z) is a ratio of (pellet)- power density in a fuel rod.

Operation within the FQ(Z) limits given'lin the'COLR prevents power peaking that would ex'ceed.,the-loss of'coolant'accident (LOCA) linear heat

-rate (LHR) limitstderived from the analysis of the ECCS.

The FN limit is a specified acceptable fuel design limit that preserves-the initial conditions.-for the limiting loss of flow transient. FN is defined as the ratio of the

..!AH integral of linearpower along the fuel rod on which the minimum departure from nucleate boiling ratio (DNBR) occurs to the average'fuel trod power. Because F Nis a ratio of integrated-powers, it is related to the maximum total power produced in a fuel rod.'w Operation within the FN limits AH given in the COLR prevents departure from nucleate boiling

'(DNB) during' apostulated loss 'of. forced reactor coolant flow accident.' ' ;.

Measurementiof..the core power'peaking factors using the

. Incore Detecto'r.~ Systemto obtain asthree dimensional power distribution map'provides direct: confirmation that FQ(Z) and F", are within: their limits, and may, be.'used to verify that the

-power peaking factors remain bounded when one or more normal operating parameters exceed their limits.

(conti nued)

Crystal River Unit 3 B 3.2-38 Amendment No. 149

i, -

r; 1

.. ,

i..

Power Peaking Factors B 3.2.!;

BASES (continued)

APPLICABLE SAFETY ANALY The limits on FQ(Z).are determined by the ECCS analysis in

'SES order to limit peak cladding temperatures to 22000F during a

.LOCA. -The-maximum.acc.eptable. cladding:temperature-is specified by.1O.CFR:50.46,(Ref..1).

Higher cladding.

temperatures.could -cause severefcladding failure by oxidation due toa a.irconium water reaction.

Other criteria must also be met (e.g., maximum cladding oxidation, maximum' hydrogen'generation,'coolable-geometry, and long term cooling).

However, peak cladding temperature is usually most-limiting'.

The limits on-F Nprovide protection from DNB during a AH limiting, loss.of flow.-transient.

Proximity to the DNB condition'is ex'pr'edsed by the'DNBR;,-definedas'theratio of thev'cladding-,surface heat-flux.required to cause._DNB to the actual cladding surface heat flux..The mi'nimum' DNBR value

.during.both normal-operation and anticipated transients is limited to theDNBR.correlation.limit for the particular

.fuel.design' in 'use, and is'accepted as an appropriate margin

!-to.

DNB.:' The DNBR correlation limit;ensures that there is at

'.least 95% probability at the".'95%',confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB.

This'LCO 'precludes core power distributions that violate the following fuel design'criteria:'

a.

During a large break LOCA, peak cladding temperature must not exceed 2200'F (Ref.- 1).

b.,During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95%

confidence level (the 95/95 DNB criterion) that the hot.fuel rod in the core does not experience a DNB condition (Refs. 2 and 3)'.

The',reload 'safety evaluation analysis determines limits on global'core parameters that characterize the core power distribution.. The primary parameters used to monitor and control the'core power distribution.are the regulating rod position, the APSR position, the AXIALPOWER IMBALANCE, and the-QPT... These parameters are' normally used to monitor and control the core power distribution because their measurements are' continuously'observable. Limits are placed on these parameters.to ensure:.that.the core power peaking factors remain bounded during operation in MODE 1. Nuclear (continued)

I Crystal River Unit 3 B 3.2-39 Revision'No. 411

Power Peaking Factors B 3.2.5 C-:

dI BASES

.j SURVEILLANCE

.SR -3.2.5.1 (continued)'

REQUIREMENTS

-for exceeding both-Zthe power peaking factors assumed in the accident analyses 'due't'o operation with unanalyzed core

'power'distributions'and spatial xenon distributions beyond their analyzed ranges.

The measured value of F is increased 'by 2.0% to account for manufacturing tolerances on'the fuel and further increased by 7.5% to account for measurement uncertainty. F N is AH increased by 5% f6r measurement uncertainty.

(,

\\

REFERENCES 1."

4 CFR '5.46.; !-

F'Q

-i

'a i

,j,.*

- a..,

I B.-C-Col'at

';2. BAW-10143P-A','Rev. ',

"BWC' Correlation of Critical

'Heat Flux,'April 1985.'

3. '.-,BAW 10241P-00, BHTP'DNB'Correlation Applied with LYNXT>

-~6>;-.............^Ni"Gr'..........

I

.r3 0'

g¶.

..

S

j 2

a a

.J

'2

'

B 3.2-44 Revision No. 45 Crystal River Unit 3

i I

I I

1.

EFIC-EFW-Vector Valve Logic B 3.3.14 BASES I

ACTIONS A.l (continued With one channel inoperable, the system cannot meet the single-failure criterion and still satisfy the dual functioinal criteria described'above. Therefore, when one vector.valve logic channel is-inoperable, the channel must be restored to OPERABLE status within:72 hours. This Condition is analogous to havingrone EFW train inoperable; wherein-a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time isprovided by the Required Actions of LCO 3.7.5, "EFW System."

As such, the Completion Time of 72:hours is based on engineering

,judgment.

B.1 and B.2--

If Required Action A.Icannot be met within the associated Completion Time,.the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant

-must.be placed-in at least MODE:3:within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ., The allowed Completion Times are reasonable, based-on operating experience, to reach the required plant conditions from full power conditions in an orderly~manner andwithout challenging plant systems.

SURVEILLANCE SR 3.3.141 '

REQUIREMENTS

-\\.-

SR 3.3.14.1 is the~performance of.a CHANNEL FUNCTIONAL-TEST every 31 days. This test demonstrates that the EFIC-EF--vector valve logic is capable of performing its intended function. The Frequency is based on operating experience,that demonstrates failure of more than one

- channel within thersame 31 day interval is unlikely.

REFERENCES None.

I Crystal River Unit 3 Revision No.

35 B 3..3-113

RB Purge Isolation-High Radiation B 3.3.15 B 3.3 INSTRUMENTATION B 3.3.15. Reactor Building (RB) purge Isolation-High Radiation BASES 4

4,

.4.

4 BACKGROUND The RB Purge Isolation-High'Radiation Function closes the RB purge and:RB mini-purge.valves'to isolate the RB atmosphere from the environment-and minimize releases of radioactivity in the event an accident occurs.

The'radiation monitoring system (RM-A1) measures the activity in a representative-sample.of-air drawn in succession through a particulate sampler, an iodine sampler, anda gas sampler. The sensitive-volume of the gas sampler

'is shielded With lead and monitored-by a Geiger-Mueller detector;. The 'ir sahple-i's taken-ffr6m'the center of the purge exhaust duct through a nozzle.installed in the duct.

.i 1

The monitor will aarm dnd initiate'cl6sure of the valves prior to:exceeding the noble'~'gas limits specified in the Offsite Dose Cal cul'ationManual".;:''

,-.s-I 1 S.,',

1

.f

-The closure of the purgd'anddminii-purge valves ensures the

RB remains as.a barrier' to-fission product release.

~~~.1.

1/4f.3,'

I APPLICABLE FSAR. Chapter 14 LOCA analysis.assumes RB purge and mini SAFETY ANALYSES-purge lines-are isolated.within,60-seconds following initiation of the event. Since the early 1980's, this isolation time'has only.been practically applicable to the mini-purge valves since the large purge-valves are required to. besealed :closd-during'thd MODES of-plant operation '(1, 2; '3, '-and.4) in whi-ch LOCAs are"po'stulated to occur. Even C.E (conti nued)

Crystal River Unit 3 IB 3.3-114 R6visio'n'No. 43,

I

-. 1,.

,! I !.

RB Purge Isolation-High Radiation B 3.3.15 BASES I,,.;.

APPLICABLE:

, for 'mini-'purge' -valves, design requirements on these'valves

.SAFETY.ANALYSES: require closure;.times.on the order of 5 seconds. Thus, the (continued I-I

.I II

.,;t purge isolation -time.of...the current plant design is.

f conservative.to.the~original safety analysis.

-..,The signal-;credited for initiating:purge isolation in the original.safety.aanalysis.is.'the RB.Pressure - High ESAS

,;..,.signal,:and notRBPurge.Isolation - High Radiation

'instrumentation.-Assuch, design basis LOCA mitigation is

.. not a:'basis :for.'including'this:.instrumentation.

I

RB purge isolation,,on.high radiationisolreuedt

-,maintain-10.-CFR.-20--limits-during-normal operations.---

However, this-is not a-basis for-requiring a Technical

-Specification.f ThergeforeejthisSpecification is not requirjed..in MODES,1,;2,2.3 and 4.-;

.2

{

--,*,-t*

Closure of.'the purge valves on high radiation is also not credited as pa'rt of,.the fuel handling-accident (FHA) inside containment.,:.The activity,-from the ruptured fuel assembly is as'sume'd to.bensantaneously released to the atmosphere in'the' form ofla;"puff"; type release.Therefore, this specification'is-not required if'moving fuel that has not been recently-Jirradiated-.(See B 3.9.3, APPLICABLE SAFETY

'.ANALYSIS for th'e cycle-specific definition of recently

.irradiated fuel)

'C-This specification is. only requi red to minimize dose;,if' moving fUel 'th at-has"'bee'. ecently irrdiated'.- -.

J

{

h i'.

has-nI etlyi diat LCO

'.One:chanrel of2RB.Purge Isolation-High.Radiation instrumentationjs required.to be.OPERABLE to ensure safety analysis assumptions regarding RB isolation are bounded.

Operability of 'the' instrumentation' includes proper operation

:of the'-sample.pump., ThisLCO'addresses only the gas sampler portion of 'the System.'.

.4 1 '

' ~ '~

,(continued)

Crystal River Unit 3 i, . B 3,..3-115

. a Re. Revision No. 43

___.1_J_

RB Purge Isolation-High Radiation B 3.3.15 BASES APPLICABILITY The*; RB: Purge Isolation-High Radi ationinstrumentation shall be OPERABLE whenever-movement of recentlyli'rradiated fuel within the RB is taking'place;.

These specifiedrconditions are indicative of.'those unde'r which the potential for a fuel handling accident,,and thus radiation release, is the greatest.- While i'rn MODES 5:and 6,- when handling of recently irradiated fuel in the RB is':not in progress, the isolation system does not. need. to be-OPERABLE because the potential for a significant. radioactive release is minimal and operator action is.sufficient to ensure post accident offsite doses are-maintained within the limits of 10 CFR 50.67 (Ref. 1).

I ACTIONS A.

1'

'Condition A applies to failure of the6high'radiation purge

,'isolation fu'nction'diuring movement.of recently irradiated

.fuel assemblies within containment..

.The imm'ediate Corp1etion.Time is consistent with the loss of

-RB isolation capability under conditions in which the fuel handling accidents.involving handling',recently irradiated fuel

,-V.L, are possible.and th'e.fiigh Oradiation.function is required to provide automatic attion.-to terminate the release.

SURVEILLANCG REQUIREMENT!

1 E SR 3.3.15.1 ; a;.:):<: K

.This.SR.is.the performance.of.the CHANNEL CHECK for,the RB

-purge isolation-high-.radiation-instrumentationonce every 12 -hours.-The CHANNEL CHECK is-.a.comparison of.-the.

parameter indicated.on.the..radiation.monitoring -

instrumentation chtannel to'"a'si'milar.parameter on other channels. It, is.based on-the assumption-that instrument channels monitoring-the-same parameter should read approximately' the same value., Significant deviations between two instrument, cha'nn-lss,`could'be an indication of,excessive instrument drift in one of the channels or of (continued)

Crystal River Unit 3

,B 3.3-116 Revision'No.

4:3

BASES PAM Instrumentation B 3.3.17 i~

I.'

A FUNCTION CHANNEL A CHANNEL B 15.,Steam.Generator Water OTSG A:

.SP-17-LI1 or OTSG A:

SP-18-LI1 Level. (Operating Range)'

f SP-17-LIR K DTSG B:

SP-22-LI1.

- :,, O T S G B :

S P - 2 1 -L I1 o r

m.

.'SP-21-LIR

16. Steam Generator,

OTSG A:

MS-106-PI1 or '

OTSG A:

MS-107-PI1 or Pressure M5-106-PIR, MS-107-PIR O

B

MS-110-PI1 or.

OTSG B:

MS-111-PI1 or MS-110-PIR MS-111-PIR

17. Emergency-Feedwater EF U

1 EF-99-LI1 Tank,Level' E

E 4

18a. Core'Exit.

Quadrant' E

-:Temperature WX 'IM-5G-TE/IM-6C-TE IM-7F-TE/IM-2G-TE

(:Thermocouple) '

X n

.IM-9E-TE/IM-13G4TE-IM-lOC-TE/IM-llG-TE YZ IM-9H-TE/IM-100-TE -

IM-1OM-TE/IM-13L-TE ZW 'IM-3L-TE/IM460-TE IM-4N-TE/IM-6L-TE i8b. Core Exit Temperature RC-171-TR RC-17Z-TR

, (Recorder)

19. Emergency Feedwater OTSG A:

EF-25-FI1 OTSG A:

EF-26-FI1 Flow OTSG B: EF-23-FI.:

.OTSG B:

EF-24-FI1 20; tow PressureInjection DH-1-FI2 Flow -[.

21. Degrees of Subcooling As Displayed onEMCO-38

As Displayed on EMCO-39 Entry into LCO 3.3.17 is Note:

Entry into LCO 3.3.17 is required if any. of the fo7o7wing i required if any of the fol7owing Hardware. or RECALL!4Points are Out of.

Hardware or RECALL Points are Out i

Service..*:

of Service.

d,

- j Hardware Multiplexers EMCO-17/18/19.

Multiplexers EMCO-26/27/28 Comm.'-HUBs:EMCO0'07/20, '.

Comm. HUBs EMCO-08/29 M'onitor EKCO-3, Monitor EMCO-39 Transmitter EMCO-72 Transmitter EMCO-74

.Receiver.:,EMCO-73-.

ReceiverEMCO-75 Recorder.RC-171-TR Recorder RC-172-TR RECALL Points.

K r::.'

RECALL Points RCS Pressure LR RECL-243 RCS Pressure LR RECL-40 RCS Pressure WR RECL-4 RCS Pressure WR RECL-S

_-____________________-_T-Hot RECL-17/2 39 T-Hot RECL-18/240 2i. Emergency Diesel tEGG-Wattmeter'SSF-AH Main EGDG-1B Wattmeter SF-AX Main

-Generator-kW Indication-controll'boardgirdicator control board indicator

23. LPI.Pump Run Status '-.I ESFA-X3 (Red i¶ghOt or

.;' H.

'ESFB-LX3 (Red Light) or

.ESFA-HU (ESLiqht.Matrix "A")

- ESFB-HU (ES Light Matrix *B")

24.

.DHV42 and DHV-43 f' *t;P'

ESFA-KN3:(Red L'ght) 2 '

ESFB-KN3 (Red Light)

Open Position Pump Run Status" p

Pump1A:,-j*

25. HPI

A HIPup C ESFA-:MF7 (Red Li~ght) or ESFB-MF7 (Red Light) or E5FB-AH (ES Light Matrix "B")

z

.t

'Y,-

.{;s, OR-,

OR HPI Pup1B:.:

HPI Pump lB:

ESFA-MN7. (Red'Light) or ESFB-MV7 (Red Light) or

,_.;,,,___,.,,,_,___,_,___ ESFA-AJ (ES Light Matrix "A).

ESFB-A3 (ES Light Matrix "B")

26. RCS Pressure RC-147-PI1 RC-148-P11 (Low Range)

I NOTES:

For Function 18a. each quadrant requires at least 2 OPERABLE detectors one from each channel.

OPERABILITY of only one detector for any quadrant constitytes entry Into Condition A of LCO 3.3.17.

Any quadrant with no OPERABLE detector constitutes entry into Condition C i LCO 3.3.17.

Separate Condition entry is allowed for each quad rant.

For Futrtin 25, OWOBIl of indicatiun is rewired only for the one ES selected WFI pap in each dmel.

.(c ontinued)

Crystal River Unit 3

- ' `-B-3.3-125B ReRevision No. 46

-I-

..1.. - -

--'i,

PAM Instrumentation B 3.3.17 BASES LCO (continued)

.The following listis a discussion of the specified instrument Functidns,listed in Table 3.3.17-1.

1.

Wide Rancte Neutron Flux Two wide-'range.neutron flux mnonitors are provided for

,post-accident:r'eactivity mon'it'ring over the entire range of. expected conditions: Each monitor provides indication over the range'of 10.8 to 100% log rated power covering..the'source, intermediate, and power ranges. Each-monitor utilizes a fission chamber neutron detectorto provide redundant main control board indication; A'single channel provides recorded information in the control room. The control room indication of neutron flux is considered-one of the primary indications.used bythe.operator following an accident. Following an event the neutron flux is monitored for reactivity control'.'The operator ensures that the reactor trips as.necessary and that emergencylbbration is initiated-if required. Since

^:

.the operator.,relies.-Upon'thi§ indication in order to

-take specified-manual action, the variable is included in this LCO.' Therefore, the LCO'deals specifically with this portion'of the string.

~~.

2.'

Reactor Coolaht-System (RCS)" Hot-Leg Temperature 11I

-1 Two wide range resistance temperature detectors.

(RTD's), one per. loop,, provide indication of reactor.

coolant system hot. leg-temperature' (T ) over the range of 1200:,to 92C0F.: Each TV measurement provides an input to.a control.room~indicator.

Channel B is also

,,recorded; in -the'control-room. Since the operator relies on theicontrol.room -indication following an accident, the LCO deals specifically with this portion of-the-string..,..

',Tis.'s a Type--A-variable on which the operator bases manual actions7;required for event. mitigation for which

.no.automatic; controls are.provided.

. I

'Following a steam'generator tube' rupture, the affected steam generator is to be isolated only after T falls below the saturation.temperature-corresponding to the pressure setpoint of the main steam safety valves.

For event mohitoringo.

nce'the:.RCP's are tripped, 'T is used along.with the,.core 'ex'it.t'emperatures and RCS' cold leg -temperature to.measure the temperature rise across the core for verification of core cooling.

.4.

(continued)

Crystal River Unit 3 B 3.3-126 Revision No.

22

I.

Pressurizer B 3.4.8

.....E I.

.~

1/}

BASES APPLICABLE,'

water. relief.through the pressurizer safety valves. 'Since SAFETY ANALYSIS ;prevention'of-Water;'relief is a goal for abnormal.,transient (continued)'

operation,:,rather-..than.-an acceptance-criterion, the nominal value for.the:press'irizer'level limit.is not required to be ad usted for-,instrument error.-,The analysis performed to stantiate' th '-290".;upper limit-.(Ref. 3) assumed the reactor -tripped.o6nhigh RCS pressure.(consistent with historical assumptions for this event).

Had the anticipatory'.'reactor-, trip'.(ART) on loss of both feedwater pumps been modeled,"'.thereactor.-would-have tripped much

-sooner, inthe event,.,'terminati'ng.the.nuclear chain reaction

-sooner, thereby,:limitin'g RCS heatup (and insurge). Thus, there is :marginin.'the analysis -to -substantiate the use of

.the nominalvalue'.as;acceptance-criterion, however additional: margin. isconservatively applied by administrativeXladju ting the'limit for leve' measure-ent lu-stg uncertainty(Ref.

4.- t

.u.

Eva:.ions'peipfor edfor.the design basis large break loss oftco6lant accident,*(LOCA),.also:.assume the maximum level assumed.fori.,the, LMFW~event The pressurizer level assumed

.for the LOCA is the'-partial'basis for: the volume of reactor coolant -.released tothe.containment following the accident.

.. :The containment-analysis performed using the mass and energy r

release-demonstrated that the maximum'resulting containment

-pressure was: within'design limits'., Parametric evaluations of this 'analysis indicate the sensitivity to pressurizer

.volume'-is.small; *1

.r The, requirement for.reduindant emergency power supplies is

.based on'NUREG-0578;r(Ref..-;.1): The intent is to all ow

-' maintaining'thei'reactor coolant'.in a'subcooled condition

.'.-i--with natural !circulation for-an-undefined, but extended,

.'time period after aioss`,of,-offsite power.

The maximum pressurizeriwater-level limit satisfies Criterion 2 of the NRC Policy Statement.

Although the'-

~

heaters are ;not'.specifically used in accident analysis, the

>..:,:.need to maintain sub?6oling in'the long term during loss of offsite'power,-asindicated in NUREG-0578 (Ref. 1), is the

'reason for providing a limit on this feature.

-:' *t-'

--4

,.-._..r.-.-

-- ^J-I

,,or,9.;t e

,p..re.

-ssurize -_

LCO,

'F orthepressurizerto3be OPERABLE, water level must be mainta'ined '<290 i'cH6s§'.and a minimum of 252 kW of pressurizer.heaters'are to'be capable'of beina powered from ach emer~gen'cy, powersupply..

Limiting the maximum operating water level,preserves,'the steam'space'for pressure control

'-'an'd ensures the' capability to establish and maintain pressure.co'ntrbl';for'.stfeady' sta'te operation and to minimize the'consequences. of potential;, overpressure transients.

(continued)

Crystal River Unit 3 I

B.. 3.4-39

. Revision No.

47

I I I -.,.

I Pressurizer B 3.4.8; BASES LCO The'minimum heater capacity required, is sufficient-to (continued)- 'maintain the systemadequately-subcooled when accounting

for heat losses through the: pressurizer insulation and minimal margin'for pressurizer steam space leakage. By maintaining the pressure near the operating conditions, a wide margin-to.subcoolin'g-can_:be obtained in the loops.

-The capability:to.,provide a minimum~of1252 kW of pressurizer heaters-from each emergency power supply requires.that both an adequate'amount of heaters and the circuits that-supply power from.each of the emergency power supplies to the pressurizer-heaters'are OPERABLE.

The

.480 V buses and supply breakers listed in Table B 3.4.8-1 must be available::in order. to satisfy:this LCO.

APPLICABIL ACTIONS ITY The need for pressure-control'is.smost pertinent when. core heat can'cause'the greatest effect on RCS temperature, resulting in the.'greatest effect'.on pressurizer level.and RCS'pressure control.

Thus, the:Applicability has been designated for-.MODES;1 and 2-.

For:-additional conservatism, the Applicabilityli's also extendedwdown to include MODE 3.

~~~~.

,*.;. e ;.;.

In MODES 1', 2', andr 3, thereis'the. need to maintain the availability-ofpressurizer heaterscapable of being powered from an emergency powert'supply,.- In'the event of a loss of thoffsite power,: t & initial: conditionis'of these MODES give the greatest demand for maintai'inng'the RCS in a hot pressurized condition with loop subcooling for an extended period.. For MODE 4,.5., or 6,I.it-.is not necessary to control pressure (by.heaters)- to'ensure loop subcooling for heat transfer. when.the Decay.Heat.Removal System is in service..Therefore the' LCO is not applicable in the MODES.

A..

';-~~~.:'.

'a!.4 S5i' 5';

With pressurizer water level in excess of the maximum limit, action must be taken, to restore pressurizer level to within the bounds assumed in the-analysis.- The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is considered reasonable for adjusting makeup. and" letdown or taking level control to..hand and decreasing.level to within limit.'

B. 1 I I I%

x,,

L- *.t.

I.

'If

'there'is,.<.252-kW of heaters, capable of being powered from each emergencypower'supply, restoration is required in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .: -.TheRCompletion Time.:of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering thelow-probability of a loss of offsite power during this period.

Pressure-control will be maintained during this time using normal non-1E powered heaters.

(continued)

I Crystal River Unit 3 1 B 3.4-40 Revision No.

29

.,~..

,,,Pressurizer B 3.4.8 BASES ACTIONS C.1 and C.2 (continued)

If pressurizer'hea'ter.capability or water level cannot be restored within-the'allowed Completion Time,.the plant must

-be placed ina MODE"in'which' the LCO does not apply. To

-achieve this.status,'.the';plant must'be-placed in MODE 3 within 6'hours-and'MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The Completion Times are reasonablet based on operating

.experience; to reach'the specified MODES from full power

,eplantcysems.

f:

conditions,.in-.an.orderly-manner-and-without challenging l

'* \\ ~~plant systems-..,,!.:>

K, In the case of'waterlevel reducingTHERMAL POWER and RCS Tave will tend-torestorelevel and also reduce the thermal energy of the reactor coolant mass for potential LOCA mass and energy releases.

'A; t

1.

. SURVEILLANCE-

.SR 3.4.8.1 3.!',

REQUIREMENTS,-

ThisSR requires thatvpressuirizer water level be monitored every.;12.hour's-in ordersto verify operation is maintained below the' nominal-upper-limit. -The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency has been shown,.by. operating experience to.be sufficient to

r

,;."'regularly..:asse~ss the.1'.eVel for deviations and trends, and verify that,,.ope~ration A s.within safety analyses assumptions.

'Al hrms are' also avajilable for early detection of abnormal

-'~~~~~

....'~vl',indi cati ohs.',,.;

Jq

. ~*

, a ;;

SR.

3 N.4.8 2

'This S verifies minimum redun'dant pressurizer heater capacity-iscapable of bein powered'from its associated

.emergencyvower supply.

(This may be done by testing the

.ower supply output and.-by-performing an electrical check on neater element continuity and resistance.) The Frequency of 24 months is considered adequate to detect heater

degradation and has been shown by operating experience to be acceptable.

. *z.';<

e t

-REFERENCES

'.;'1. ';NUREG'0578, 'July 1979, "TMI-2 Les'sons Learned Task Force Status Report and Short Term Recommendations."

-2.

NUREG 0737,-"Clarification of TMI Action Plan Requirements",' November, 1980.

- :3.

-B&W Topical Repo'rt-51-1200406-0O,.January 1991.

j a

l w_

+

4. Calculation M97-0064, Revision 3' "Pressurizer Level vs. Tave'for Power Operations."<--'

...1

~'I

.Crystal River Unit 3

.B 3.4-41 C. rUn4Revision No. 47

I

'L

,1.r'.

Pressurizer B 3.4.8 Pressurizer Heater Table B 3.4.8-1 Emergency Power'Source Circuit Components 3..

-TYPE"

-'"TRAIN A

'"TRAIN B

. r,

480 V Buses-Reactor Aux Bus 3A, Plant Aux Bus 3 Reactor Aux Bus 3B Breakers 3321,

3222 "i.3395 3312

,3355 3392 3396; 3356 Press Pressuriz'er' Heate'r Pressurizer Heater MCC Breakers, MCC 3A.

MCC 3B 1A Pressurizer 1A - Pressurizer Heater Control-Heater'Control

..ransformer A-1'i

__Transformer B-1 2A:-, -Pressurizer l,.1B

- Pressurizer

',Heater.Controlt,.

,,Heater Control Transformer, A-2'

.- Transformer B-2 1C -

Pressurizer 1D - Pressurizer Heater Group 7 Heater Group 10 2C - Pressurizer 2C - Pressurizer Heater Group.8 Heater Group 11.:

3C -

Pressurizer 3C - Pressurizer Heater Group 9 Heater Group 12 4C - Pressurizer Heater Group 13 Crystal River Unit 3 B 3.4-42 Revision No. 29

Containment Air Locks B 3.6.2 BASES

-l

APPLICABLE^. The, DBAs analyzed.for dose' consequences that result in a SAFETY ANALYSES release of radioactjvematerial within containment are a

.loss ;of. coolant'accident.(LOCA)- and.a-rod ejection accident

'(Ref. 2).In'theanalysis of'each of these accidents, it is

-assumed.that'c'ontainment is.OPERABLE-so that release of fission,products to 'theenvironment iscontrolled by the

-'rate of,-containment leakage. The.containment was designed

. with :an'allowable leakage rate of 0.25% of containment air weight'per day '(Ref. 3) This. leakag'e rate is defined in 10 CFR' 50, Appendix)'J (Ref.'1),,as L : the maximum allowable containment leakage rate at the calculated maximum peak containment pressure (P) following a DBA. This allowable leakage rate formsrthe-basis for the acceptance criteria imposed on the-'SRs`as'sociated with' the air lock. L is 0.25%

of containment air weight'per,'day. and P is 54.

'resulting' from th limiting ;design bas 2O pig ngfrmt-t bais LOCA.

The dose-acceptance-criteria applied to DBA releases of

-'mradioactive'material to the environment are given in 10 CFR The containment air locks satisfy Criterion 3 of the NRC

.Policy Statement:

LCO Eachi containme'nti ai rql ock forms part of the containment pressure.boundary.,-As'a part of containment, the air lock

-.'safety.funtc'tion'.is'related to control of the containment leakage rate res ultig 'from a DBA.',Thus, each air lock's structural integrity and leak tightnessare essential to the successful mitigation of such an event.

1 1'J (continued)

Crystal River Unit 3

,. B 3.6-7

.Revision-No.

37

I

, 7 Containment Air Locks B 3.6.2 BASES LCO Each air lock is required to be OPERABLE. For.the air lock (continued) to be considered OPERABLE;* the air lock interlock mechanism

-. must be:OPERABLE,. the air lock must be in'compliance with

...the Type:B air lock leakage test *and.both air lock doors M

must e OPERABLE.:.The interlock allows only one air lock door' of an air lock'to. be opened at one time (Ref. 5). This provision ensures that a gross breach'of containment does not exist'when'containment is required'to be OPERABLE.

Closure:of.a sing'e door in each air.lock is sufficient to

.providealeak.tight1.barrier following postulated events.

Nevertheless,'.,both doors are kept'closed when the air lock is not being, used for normal entry into and exit from containment:-

,j*,*

Each airJlock contains equalizationwvalves that are

....operatedjin conjunction with the associated air lock doors.

The equalization valves are-integral to the air lock design and are:driven by'the mechanical operating system associated!with th'e.:air.l'ckdoor.' OPERABILITY of an air

lock door 'requires' the 'associated'Tequalization valve to be OPERABLE. Therefore-OPERABILITY of:the equalization valves is addressed' by.this LCO.and not.,by LCO 3.6.3, "Containment Isolation Val'ves.".'A failure. of the equalization valve would result"in the'associated'a'irlock door being declared inoperable and LCO 3.6.2-Condition'"Awould be applicable.

p'a.: '-

.F r..

APPLICABILITY In-MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material 'to containment. In MODES 5-and 6 the probability and consequences.of these events are reduced due

'to the.:pressure-anrd temperature limitations of these MODES.

rTherefore, the 'containment air locks:are not required in MODE..5 to prevent.:leakage.,of.radioactive material from containment. The;r~equirements for: the.containment air locks

- dur'ing"MODE 6' are addressed, in' LCO.3.9.3, "Containment

" ~Pene'trations.;"

ACTIONS The ACTIONS are modified by a Note that allows entry and exit to perform~repairs~on the, affected air lock component or for emergencies-invol'ving personnel..safety. If the outer door is inoperable,. then it ma'be' easily accessed to repair. If the inner door is' the'onethat is inoperable, however, then a short time-exists when.-the containment boundary is not intact (during access.through the outer

'door). In this context, repairs include follow-up actions to an initial failure of the air lock door seal SR in order to determine which air lock door(s) is faulty. There are circumstances where an at-power containment entry would be required during the period of time that one air lock was inoperable. In this case, entry would be made through the OPERABLE air lock if ALARA conditions permit. However, the (continued)

Crystal River Unit 3

. B 3.6-8 Revision No.

48

Containment Isolation Valves B 3.6.3 BASES APPLICABLE The containment'isolation valve.LCO was der~ived from.the

SAFETY ANALYSES-'requirements related.to the c'ontrol of -leakage'from containment during major accidents.; This LCO'is-intended to ensure the'_containmnent:leakage rates do not exceed the

.-,valbes.assumedJin the safety analysis..As part of the contai.nment boundary,.-containment isolation valve OPERABILITY.supports leak tightness of-the containment.

)'Therefore, the safety"`aralysis of.-any event requiring

-:...cntai nment. i solaio.s.;ppicable7to: this LCO.

.,The DBAs analyzed.for dose consequences that result in a

.release of.radioactivematerial within'containment are a lossof.'coolant accident;(LOCA) 'and'a'.rod ejection accident (Ref.,-3). In the ¢halysi' for each of.'these accidents, it is

,assum'e'dt that-'containment isolatiornivaves are either closed

.or, funct-ion,to-close within.the.required isolation time following event initiation. This.e'nsure's that potential eaka'g'e:paths'-to.the.environment.through containment isol'ation valves (including containment purge valves) are minimized.,

The dose acceptance criteria applied to accidental releases CFR 50.67-(

o '

an l

d,'The DrA' haal w'tohi th a

n'.60 seconds after the accijdent, sisolat'ion of.the.containment.is complete and leaka'ge iterminated 'eicept fo'r, thede'sig'n leakage rate, L.

tainO n'tota response of 60 seconds TheDA' con nifiii.a6syjnes tht witherof60 secondsafeth

includes' ignal delayl

diesel generaor;.,startup (for loss of offsite power)", and co'ntainment'is'otlaion valve stroke ie.SR-3;.~3.tS 4 'address'es t he'"respons time testin requirements.

Theingle'failure criterion required in the safety analyses htms R **54adrse teepne tim testinglse

' 7was 'onsjdered in 'the Vriginal design' f the containment prg vala series oneach purge line

provide:assurance that both the supply and exhaust lines

could beisol'aed eve'nif a'singlee failure occurred. The

(continued)

Crystal River Unit 3

.B 3.6-17 RevisionNo. 37

'k.,

I I,

I,

I Containment Isolation Valves B 3.6.3 BASES APPLICABLE, inboard'and outboard'isolation' valves on each line."are SAFETr ANALYSES provided with diverse powersources,;.motoroperated:'and' (continued) pneumatically operated spring-closed, respectively. This arrangement was designed;.to preclude common mode failures from disabling both--valves on a purge line.

The containment pur e, valves'may be.unable to close in the

.environment following.a'LOCA'. Therefore, each of the 48

'inch-purge-valves.is: required to remain sealed closed during MODES 1, 2, 3, and[4'. In this case,<the single-failure

-criterion' remains applicableto the containment purge valves because of failure in the control.circuit associated with each valve. Again, the 48 inch purge system valve design prevents-a single failure from compromising containment OPERABILITY.as ;long as'the system'nis'soperated in a'ccordance with.the sdbjectL'COd

I I

.The containment'isolation valves'satisfy Criterion 3 of the NRC Policy Statement-

LCO f

j.t...

Containment.isolation valves form a part of the containment

. bounda'ry.- The containment isolation valve safety function is related to control of containment leakage rates during a

.DBA.

Theautomaticpswer'operated isolation valves are required

' to, have' isolation

-,times w thinnlimits and to actuate on an automatic isolation 'signal.

Tie 48 inch purge valves must be. mai'ntained sealed-closed in MODES 1, 2, 3 and 4. The valves coveredby"this LCO;arei sted in the FSAR (Ref.

4).

..The normal y.clos&IX isol ation.hvalves are considered OPERABLE when manual.valves-;are closed,.check valves have flow through the valve secured,;bl-indd flanges are in place, and

-closed systems. are intact.-

-Purge valves 'with rresilient:'seals-must"meet additional

'leakage rate requi'rements addressed,a's part of this Specification. All. other containment isolation valve

.leakage irate :testing is addressed by LCO 3.6.1,

-"Containment," as-part of Type'C testing.

I (continued)

Crystal River Unit 3

. B 3.6-18 Revision No.

44

t '.

I

. :. ;- -

1 ), 1; ;

.. i,..

Containment Isolation Valves B 3.6.3 A,

BASES LCO This LCO provides assurance that the containment isolation (continued) valves and purge valvesIwill perform their designated safety

. functions-to control-ieakage'from the containment during

,-accidents. '..,;s t's ts, i'

'l APPLICABILITY In MODES 1I,2, 3, and 4," a DBA could cause a release of radioactive material.,to containment... In MODES 5 and 6, the probability and consequences of these events are reduced due to'the-pressure and temperature'limitations of these MODES.

Therefore, the-containment isolation-valves are not required to be&OPERABLE-in MODE:-S. The requirements for containment isolation;.valves.'during MODE 6.are.addressed in LCO 3.9.3, "Containment Penetrations.,

..ACTIONS.,::

The following..terms aTre-defined for,,the purpose of

,implementing this"S,cification:....

penetration flowpath: The piping which passes through

the.RB liner-sucn that a portion of the system inside the RB can communicate with the portion outside the RB. A penetration passes through the imaginary plane

.established by. the RB liner.

unrisolated:'The'state ofa penetration flowpath whereby J'

i.i i,-the operating-fluid'-(liguid-origas) of the system is

.capable oftpassing;,freely through -the imaginary plane

.- - i-established.by te,RB liner.',,-

-penetration flow path with two-containment isolation

,.yalves: Penetrations where the primary flow path is l-a.,-isolated.by-valvestboth.inside~andoutside the conta'inme'nt'-buiilding.'..'.The 'primary,'flow path does not

-i

.includetestconnections, vents or-drains.

-penetration,,flowspath with:.only-.one containment

,- isolation:.valvej:andva -,closed system: Penetrations where

'the primaey-'flobWpath has a closed system on one side

'of containment-and is'isolated by-a valve on the other.

-44:The~primary flow path does-'not 'include test

.connections,.vents-or,.drains.;

-The ACTIONS 6are modified by a Note'allowing penetration flow

,paths,lexcept'for 48':inch purge valve'penetration flow paths, to be unisolated intermittently under administrative controls;.-These administrativetcontrols consist of stationing a-dedicated operator atthe valve controls, who

-:Jis n'contihu'ous-communication-with.the control room. In this.iWay, `the penetration can be rapidly isolated when a

" ' need for..containment-Misolation is;,indicated.

Due to the size.of. the-.containment-purge line-penetration and the fact

'that those'penetrations exhaust directly from the containment atmosphere to the environment, the penetration flow paths containing these valves may not be opened under (continued)

.-Crystal.- River Unit 3

.- -_~B 3.6-19 11 I RevisionNo. 44

-7!, *I*;

. I Containment Isolation Valves B 3.6.3 BASES ACTIONS

. (continued) administrative controls.

A single purge valve in a-

-penetration flow pathmay be opened.to.effect repairs to an inoperable valve, as.allowed'by'.SR-3.6'.3.1.

I Note 2 has been added-to provide!clarification that, for

.this LCO,.separate Condition.entry is-allowed for. each penetration flow"pa th.-

The; ACTIONS~are further-modi'fied by a Note 3, which ensures appropriate-.remedialactions are-taken, if necessary, if

the'affectedi'systems:are rendered; inoperable by an inoperable containment isolation valve.

I I

Inthe' event-pu'rge vaive leakage.'results in exceeding the

.overall.'containment-leakage..rate,,;.Note:4 directs entry into

-,the applicable.Conditions and:Required Actions of LCO 3.6.1.

A.1 and A.2,

',.In' the.event one conta nm' t,-isola invleionorme penetration. flow paths is inoperable;(except for purge valve leakage:.notw'iithini;limit)-;:the'-affected penetration flow

.path must be..isolated.,Theimethod 'of isolation must include the useof at0l'eastone isolatio'n'barrier that cannot be

-r.

adversely aff-ted by a~single' active failure.

Isolation

.-barriers thati.meet this:,-crterio-n-,are a closed and

-de-activated.,'automatic containhmentrisolation valve, a closed

.manualvalve

,a-blind flange, and:a check valve with flow m a nb a lr e r t h at -e t h s

~ e i n a e a c o e n

through the'valve,'secur'ed.'

For, a-pen-etration isolated in accordance, with' Required Action-A.1, the valve used to

,within the-4. hour: Completion Time..-'The specified time e

iolt hpne ration shou j~e the--closest available one period-is-reasonabli",'cons'ide'ri'ig-'the' time required to nisolatethe penetration fard the relatjve importance of supp6rting contai'nment"OPERABILITY du'ring MODES 1, 2, 3, and 4.

C For affected. penetration. flow-pathsithat cannot be restored to OPERABLE status within thel 4-Koar Completion Time and that have b'een.,'sol ated-in accordance'with Required Action 'A'.1,,'the..affected penetration flow paths must be verified to be"isolated on a periodic basis.

This periodic (continued)

Crystal River Unit 3

.- s; 3.6-20 RevisionNo.

26

-Containment Isolation Valves B 3.6.3 BASES ACTIONS A.1 and A.2 ~(continued) verification is.ie'cessary-to ensure that containment

penetrations-'requi red to be Aisolated following an accident and no longer'capable'of-being automatically isolated will be, -im n

iutheiisolation position should an' event occur. This Requ'ired Action-does-not.require' any.testingq or valve manipulation;Ratlher.,.:it.involves'-verification that those i'

isolation-devices'.'capable.of being.mispositioned are in the

correct position.-The Completion Time.of "once per 31 days for-'isolation de'vi;c'es boutside 'containment" is appropriate considering-the fact-that the'valves are operated under administrative'controls and the probability of their misalignment'is';low'-.For the'isol ation;devices inside

- containment,'the-time period specified as "prior to

-entering MODE-4:-from'MODE'5.if not.performed within the

'-r'*vious-.92:days'.';is bas9ed on'engqineering judgment and is t

')

.v L' '"

o'cosdrd'esnal~in-iwo-hsnccessibility of the

-isolation.devices)ainjdother-administrative controls that will ensure that;.isolation'device.misalignment is an

-.unlikely'possibility.

-- Condition A-has been!modified by a'Note -ndicatin this Condition is only. applicable to those:penetration flow paths With'f-two.containment3nisolation valves..For penetration flow

-paths withnonlyobne-'containment isolation valve and a closed system,.[.Conditibn Ci provides appropriate actions.

equ'A.2 odified'bytwo'Notes. Note 1 applis't isolation'

!evices located in high radiation areas and'all6ws-the-devi&`s'btobe verified by use of adminitstrat-ve mieaas.JAllowifig verification by

-- administrativermeansdis:considered'acceptable since access

-to'-these'.areas.is-typically.restricted.

Note 2 applies to isolation devices---that-are locked,.sealed, or otherwise

.secured in position an8'allows'th'ese. devices to be verified closed'by useofiadmi'iistrativemeans"'- Allowing

verificati by';dnii'tratiive'means 'is considered acceptablePssince-thifunction' oflocking, sealing or securing'componehts:.'is'to-ensure that 'these devices are not

.inadvertentlylrep6sitioned.

'Therefore the probability of

.t alibnme, i.thesrper's*devices, oncethey have been verified t-,b..-lnppsition-,.is small.

B.1 and B.2 i,....

~.'..-Wi-WithtaalAcontainment'isolation-valves in one or more

. penetration';flowipaths iin6perable-(except for 48 inch purge valve leakage.not'within limit).,'.the.affected penetration

.fl

.athmust be.Asobated-within'1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The method of

.<:'-isolation must'include-theuse-bf;at least one isolation (continued)

Crystal River Unit 3

--- -B 3.6-21

- 'Amendment No. 209

Containment Isolation Valves B 3.6.3 BASES ACTIONS B.1 and B.2 (continued) barrier that cannot failure.. Isolation. be adversely affected by adsingle'active barriers that'meet this criterion are a closed and de'-actiivated automatic valve, a closed manual valve, 'and a blind flange.'-The'.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consisteht with-the ACTIONS'of LCO.3.6.1. In the event the affected'penetratio'n is' isolatedain accordance with Required Action B.A, the affected penetration must be verified to be isolatedon a periodic'basis per'Required Action B.2.

This periodic verification is necessary to assure leak tightness

'of containment and that-penetrations requiring isolation

.following an accident-are isolated. The Completion Time of once per 31 days.fodr verifying each affected penetration

.flow path-is-isolated-is appropriate considering the fact that the-valves are.-operated under. administrative controls

,,';,4.

.,:andtheprobabilityoftheir.misalignment is low.

Condition..B is 'modified by-a-Note-indicating this Condition is only applicable-to.-penetration flow paths with two containment isolation valves or. those with one containment isolation-valve:and no-closed system.-Condition A of this Specification addresses the condition of one containment

isolation.1valve inoperable.inh-apenetiration flow path with

two containment is6lation;'valves.-.-

Required Action B.21is.modified by twoNotes. Note 1

.applies to isolation, devices locatedj ii, high radiation areas and 'allows the:'devyices..to b'e verified: by use of adiiniist'rativemeans:..Allowing.' ve'rification by

'dministrative"'means'is.'c'onsidered.acceptable since access

'to these areas is5 typically restricted-.

Note 2 applies to isolation devicesrt'Iat are'locked, sealed, or otherwise

.secured in position and allows these devices to be verified

' closed'by us'e of administrative'meahs'.'. Allowing verification'-by-administrati~ve means-Ais considered acceptable, 'since'&:the-function-of locking, sealing or

'securing components is--'to ensure-that-these devices are not inadvertently-repositioned

-'Therefore the probability of misalignment of-these devices,- once' they have been verified

. -to-be in the proper position, is-small; C.1 and C.2

. 1.

fA

.a

-With one or morepenetration'flow.paths with one containment isolation' val~ve inoperable:or-the~closed system breached, inoperable-valve.-must be restored.to 2

OPERABLE status or affected' in'et'rati6n.fl-ow path must-be-.isolated. The method of'isolationh-must include the use of at least one I

(conti nued)

Crystal River Unit 3

B 3.6-22 Amefidm'ent No'.

209

~~~~~~

~

rVCnanetIoainVle

-Containment Isolation Valves B 3.6.3 BASES ACTIONS C.1 and C.2 (contfinued)-

isol'ation barrier.'that cannot'be adversely affected by a single.active 'failure.'"Isolati6n barriers that meet this criterion'.are'a 'closed and de-activated automatic valve, a

- :closed manual valve,.and 'a blind-flange. A check valve may not be used t.isolatethe affected penetration. Required Action C.1'mus't'.b'e.co'p6leted within'the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion

Time.The specifi'edtimep".riod.is 'reasonable, considering the relative s'tabiliiy of thecl'osed'system (hence, reli-ability) toact.asapenetrationisolation boundary and the' relative importance' of suppo'rting containment

OPERABILITY-during MODES 1',

'2,':3,.and 4. In the event the affected penetration.is isolated in accordance with Required

-Action.'CI1,>theaffected:penetration-flow path must be verified'-toi~be-i1isolated,'bn.a periodic.basis. This periodic verification is necessary to assure leak tightness of

containment~and thait'*containment'penet'rations requiring iisolat-ion.followinglan-accident are isolated. The

-.X Completion Timet'of on'ce per'31 days'for verifying that each affected'penetrati'on':flow"path'is isolated is appropriate contaduinmen an~.d -th cothainment pe'stratix n rpeqirn consdrinth.fact~that the'valves-re operated under

'administ-ative controls and the pir6bability of their misalignienis'low!'-

.Condi tio fis

'm'odified bya'Note indicating that this Condi to thospenetration flow paths

. -with;only one 'con'ta nentisolation'valve and a closed system.' This.'Note: is..recessa'ry~since'this Condition is written.to 'specifically'a s

ion flow paths

' ..utilizing.ia clos de'edsystem;:'

e pe t

f paths

'Rqui*red.Action t.2 1`4is modified-bywo Notes. 'Note 1 pp sol ati on'devi ces 16cated in high'radiation'areas

.;:j:

and~allows-these! devi sst~everif~ied. by use of

.:administrative means.;:Allowing verification by administrative-.means:ais:,considered 'acceptable since access

.,--to these areas;is typically restricted.

Note 2 applies to

.. : -isolation devices. that'iare locked,; sealed, or otherwise secured in'position 'ard allows these devices to be verified

'closed by use of'administrative' means." Allowing

'verif.ication.by administrative means i's considered acceptabl e," since0 the' function'hof locking, sealing or seciring components i-s.to 'ensure that,'these devices are not repos

.-'Th'"ef'"'the probability of misalignment-of-the's'devicesonce' verified to be in the p-'

.-.-proper'jositi6n,'ismall (continued)

.Crystal River Unit 3

-B 3.6-23 Amendment No.

209

Containment Isolation Valves B 3.6.3 t

BASES ACTIONS (conti nu(

ad)

D.1 In the eventvone or"more containment 48 inch purge valves in one or more penetration flow'paths are not within the purge valve leakagelimits,-purge valve leakage must be restored' to within l`imits within 24 'hours. The specified time-is' a' reasonable pe'riod'fo'r restoring the valve leakage to' within limits,: provided 'overall containment leakage rate

-'remains within' limits.' With the purgei'valve seal degraded such.that..leakage exceeds',the limits,'there is an increased potential for thesame mechanism that caused the initial degradation to cause furtheridegradation. If left unchecked, this could result in'.

aloss-of containment OPERABILITY. Thus, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is necessary toilimit the length of timde'the plantican operate in this

-con'dition. -;;*' >i'

i i e *-,

-. J,.

E'.1 and E;2' v~~~~

I r

I

,, w.

-If the'Required.Actions and'associat'ed'Completion Times are not met,'the plant'-must -be^placed'in 'a MODE in which the LCO' does'not apply&':To a'chieve"th'is'st'atus, the plant must be placed in at'least'MODE 3 within'6'hours and in MODE 5

within.,36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .:-The'allowed,-Completion Times are

'reasonable,- based 'ooperating.'ex'perience, to reach the required plant'conditions from'full'- power conditions in an orderly manner and without challenging plant systems.

i I

SURVEILLANCi REQUIREMENT' E-,

SR.3.6.3.1 S.,

V

.,.Each 48,1inch containment-purge valve-~is required to be verified sealed c-losed-.at-.31 day.intervals. This

'.Sur'veillance.

'is.designed to ensure'that a gross breach of containment is not.'caused. byan'inadvertent or spurious

.opening.of.a containment purgevalve..Detailed analysis of the pu'rge valves failed to conclusively demonstrate their ability to close during a.LOCAjin-time'to maintairi offsite doses-to.within licensing,basiss,.Iimits. Therefore, these valves are.required to be'.in the.sealed closed position duringMODES-1,..2,' 3,. and 4.-A-,containment purge valve that is sealed closed must have motive power to the valve operator removed. This can be-accomplished by de-energizing

, (continued')

Crystal River Unit 3 B 3.6-24 Amendment No. 156 Crystal River Unit 3

.B 3.6-24

-..

Amendment. No.

156

,Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.1 (continued)

REQUIREMENTS the source of electric power.or by.removing the'air supply

-to: the valve'operator...In.thissapplication, the term nsealed".-has--no connotation-of-.leak tightness. The Frequency is a-resultof.:an NRC initiative, Generic Issue B-24-(Ref. 6),',related to containment purge valve use

-during unit.operations..In'the event-purge valve leakage requires:entry intoCondition D,-the Surveillance permits opening-,one-pur~ge.v~alve.in~a penetration flow path to

.:- :-.-.perform -repai~r~s.'.-.,-.-;-,,,..,

SR 3;6.3.2 hdg purge This,-,SR ensures that, the-6 inch post.accident hydrogen purge valves are closed as required..or;-,if open, open for an allowable reason. The SR is not required to be met when the post accident hydrogen purge valves are open for pressure control, ALARA or air quality considerations for personnel entry,-.or-.for,*Surve~illances that require the valves to be open. Thepost-iaccident-hydrogen.purge'valves are capable of. closing in.: ~the.environment following a LOCA. Therefore, thes'e-.valves are.allowed.:to.be open for limited periods of

,,time.;.The 31,day,Frequency for.verifying valve position is

.consistent.with';othe.r-containment isolation valve ssed jfi iSR; 3.6.3.3.

requirements-disc

'SR 3;6.3.3 3

This5R~requires..verification that-each containment

- isolation manual valve-and-blind:flange located outside containment.and-is'not'locked, sealed, or otherwise sec'ured

'--a and 'requied-to'b'ei'clo'ed'du'ring'accident conditions is closed.'The'SR-'helps t6 ensure thatipObst accident leakage

-0¢

'of eadi oacti'v- -'fl W dso'.-b ase s ioutsidde'the contai nment boundary'!is~within'design limits. This 'SR does not require

-anytestii'g§or valveOranipulation. Rather, it involves verification that those valves outside&containment and capableof being mispbsitioned'arein the correct position.

-.- -Siceverificatonofvalve'positin for valves outside containment;jis relatively easy',a 311day. Frequency, based on engineering judgment was-chosen to provide added assurance A -

( c.

(conti nued)

Crystal-River Unit 3 1 7

.'.. ;-B 3.6-25 t

R. Amendment No.

209

I~~1 Containment Isolation Valves B 3.6.3 BASES SURVEILLAt REQUIREMEO KCE SR 3.6.3.3 (continued) 4TS

of the correct positions..The.SR.specifies that valves open under administrati'v'e controls are not-required to meet the SR'during-the time.the valves are open.

. ANote modifies this"SR and applies to-valves and blind

flanges located in'hi'h radiation areas allowing these devices to be verified.closed by use of administrative means. Allowing verification-by administrative means is considered acceptable, s.ince.access to these areas is typically restricted during MODES:1,.,2, 3, and 4 for ALARA

..reasons. -Therefore, the probability.of misalignment of

....-these.,valves, once they havebeenm.verified to be in the proper position,-is~low.-

I

I ::,

. :,

I.I I

-SR 3.6.3.4

-,This SR-requires verification that.each containment isolation-manual.valve and.-blind.fl'ange that is located

.inside containment and.;is not-.locked-, sealed, or otherwise secured and required to be:.cl6sed during'accident conditions is closed. -The SR helps.:to-ensure that post accident leakage.of radioactive fluids.or gases outside the containment boundary is within.design'limits.

For valves inside containment, the Frequency-of "prior to entering MODE 4 from.MODE 5-.if not performed within the previous 92 days.".is appropriate, since these valves and flanges are typically inaccessible during reactor operation, are

-,operated under-admi.nistrative co'ntrol's and the probability of'their'misali'gnmnt 'is low'.

The.SR*'specifies that valves open under-administratiVe-controls 'are-not required to meet theSR during the-'time-they'-are'open.';

The-Note'allows valves -and-blinfdflan'ges located in high

.radiation areas to'be'verified cl'o'sed-by use of

'administrative means.-',Allow in'g v'er'fi'cation by administrative means'

-considered acceptable, since the dministrative means.isllowiii vdrifiainb

. access tothese areas is typicall-'res'tricted during

.MODES'1, 2 :3, inid4 for ALARA reasons.

Therefore, the probability of misalignment 'of these valves, once they have been verified to be in their proper position, is small.

(conti nued)

Crystal River Unit 3

.B 3. 6-26 Amendient No.

209

-:'. 'Reactor.Building:Spray and-Containment Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.2 (continued) :..

REQUIREMENTS i

occurring between sur~veiilancesland has been'shown to be

-acceptable through,operating experience.

-It Jis preferable~i6'ru:_the fans n'slow speed for this SR since,this.-provides.additional confidence the post-accident

-containment coolingntrain circuitry is,:OPERABLE.

-SR3.6.6.3.

Verifying that each -RB spray p'ump's.:developed head at the flow test point is greatervthan'or equal to the required developed head ensures t.hat'spray-pump'performance has not U1.degraded-du ri'ng' thecycle'. Flow'and-differential pressure

,are normal.tests -bf ifcentrifu'gal 'puimp':performance required by

.-.Section;XI of. the-.tASME Code-,(Ref.- 5).'zSince the RB Spray

-..- -System pumps.cannot.-be-tested-with,flow through the spray headers, they,;are-tested.on-recirculation flow. This test

.confirms-one point-on-:the pump design curve and is

-indicative of overall performance. Such'inservice tests confirm.-component;OPERABILI1TY;.trend performance, and detect

, !'i ncipient failures'..by:.indicating abnormal performance. The Frequency.-.of-this SR;is.Ain-accordance with the Inservice 0.Testingt.rogram.-:.d-iI;;.-

r

SR 3.6.6.4i.2f,*-

Verifying.an.emergency design cooling.'water flow rate of 2

.1780: gpm to-eachnrequired.;containment cooling system heat l

I

,exchangers, (fan

cooling.coils).ensures-the design flow rate assumed-inithe-: safetyrlanalysis is being achieved. The SR veriifies -that, with,the-SW-System-in the post-accident ES

.,alignmrent,.!adequate,,fiowj-s providedto the heat exchangers to remove.-the design,;basis-reactor-building heat load. The 24 month Frequency is based on the need to perform this

-_Surveilllance :unde

-theconditions.that,.apply during a plant

-, outage.w-While-theeheat:exchangers can be aligned to the SW

. System during normaloperations,.'other critical normal-running iSWi loads make.it impractical to verify accident flow

?rate:-to-the heat exchangers.with the.plant on-line.

On an ES: actuation,these.i.normal-running loads are isolated and r

. I, the-.SWflow!.f.

.w i

(conti nued)

CrystalRiver Unit 3 B 3.6-43 Revision No. 44

-A---

.~-1 ! It

{

I Reactor Building Spray and Containment Cooling Systems B 3.6.6 BASES SURVEILLANCE REQUIREMENTS SR 3.6.6.4 (continued) normally-supplied to them re-directed to the post-accident loads.' The 24"'month' Frequ'en'cy was also considered acceptable based upon-the existence of other Technical

'Specification Surveillance Requirements. A degradation in

-heat-exchanger performance between'performances of this SR would likely be seen as-an increase in RB temperature (monitored once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in accordance with SR -3.6.5.1).- The'1780 gpm cooling'water.fl'ow rate does not include the flow to.the motor cooler or any allowance for instrument' flow uncertainty.

I SR 3.6.6.65 and SR 3.6.6.6

... These SRs require verification that each automatic RB spray

'valvethatisnotlocked, sealed,'-or otherwise secured in the-correct position, actuates.to 'its correct position and l

that each RB spray pump 'starts upon receipt of an actual or

simulated-actuation'sigfial.. The-24, month Frequency is based on the need to perform these Surveillances under the conditions that'Ipp y during a-ant outage and the

'potential 'for anunplanned transient if the Surveillances were performed with'the'reactor :at'. power. Operating experience has shown that.these components usually pass the Surveillances when'performed at'the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.'

The:SR is modified by'a note'indicating the SR is not

.applicable;in'theidentified'MODE' -This is necessary in order to make the; requirements-for'automatic system response consistent with those for the actuation instrumentation.

SR 3.6.6.7

.This SR requires verification that each required containment cooling train actuates upon receipt-of-an actual or simulated actuation'signal. In the-event of a LOCA, the air steam mixture density is much higher than normal air

-density. The uni ts are not-designed to-handle the full flow rate at this condition. To operate-the unit at full flow (motor at high speed) at this condition, will cause the motor to overload and trip. To guard the motor from overloading, the volumetric flow rate must be cut (conti nued)

Crystal River Unit 3 B 3. 6-44 Revision No.

44

.SpentFuel Pool Boron Concentration B 3.7.14 BASES SURVEILLANCE SR 3.7.14.1 (continued)

REQUIREMENTS Operating experience has shown significant differences

'between boron,'meas'ured.near the' top of the pool and that

.measured elsewhere.;,Asl,-ong-asthis SR is met,- the analyzed events.are f u

The;7day Frequency is acceptable because no.major rjeplenishment-of.pool.water is expected to

.take place,over, this period of, time.-,

J REFERENCES

1.

Criticality Safety Evaluation of the Pool A Spent Fuel Storage Racks in'Crystal Rive.r Unit 3 With Fuel of 5.0% Enrichment,'S. E. Turner, Holtec Report HI-931111,. December -1993.-

2.'

Cr'iticiality'Safety Analysis'of the Westinghouse Spent FuelStorage'Raks-n:Pool B-of Crystal River Unit 3, S..E.Turner,'Holtec Report HI-992128, May 1999.

3.

Criti'cality SafetAn isoft Crystal River Unit 3.PoolA orStorageof 5%

id Mark B-11 Fuel in iChecke'riboardA'rr angemrent with Water Holes, Holtec

-Report_

HI-'992285' August 1999.*

~~~I 4

Criticality. Evluatjon '.of.CR3 Spent Fuel Pool Storage

'Racks viith Mark;,B-12 Fuel, Holtec Report HI-2022907,

-September 2002.

.5.., Progress.Energy..Engineering.Change EC No. 52456, Documentation ofAcceptability to Receive and Store Mk B/HTP,,Fuel

.C,

*1.*- I- ;,..._...

. ~

~

c i~

t ti

~..;!:

Crystal River Unit 3 t

B 3..7-71 Revision No.

42

~

- 4:.7 1 !

I I Spent Fuel Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage..

7..

I I !.

I

. I QArCFCr

'J.

I~.

BACKGROUND

'This do'cument describes' the, Bases for'the Spent Fuel Assembly Storage which' imposes storage requirements upon irradiated and unirradiated fuel assemblies'stored'in the fuel storage pools containing high density racks.'The-storalge areas, which are part of the Spent Fuel System, governed by this Specification are:.

-v a.,,

Fuel storage pol

b.

Fuel storage pool "A" and..

E B I

I t o:. %,

i,

In general1 the.fvnction~of the.storage racks is to support and' protectnew and.:pent fuel.f rom the time it is placed in

,the storage area.u'nti`l it. isshipped offsite.

Spent fuel 'is.stored 'underwater in,.either fuel storage pool A or. B. Only. fuel p6ol.A hasi.the.c'apability to store failed fuel 'in'containers. Spent fuel pool A features high density poison storage racks with a 10.1/2 inch center-to-center distance capable of storing 542',assemblies. Fuel pool A is capable of storing,.fuel with enrichments up to 5.0 weight percent-U-235'(Ref. 1, 6,.7 and 8) without exceeding the criticality' criteria of'Referenc e3 providing the fuel has sufficient burnup.. New fuelT.will be placed into pool A only.

I Spent -fuel 'pool, Balso contains high density racks having a 9.11 inch'center-to-center distance capable of storing 932 assemblies. Fuel po6ol0 B is capable of stdring fuel with' enrichments up to S.0 weight percentU-235 (Ref. 2, 7 and 8) without exceeding-the criticality criteria of Reference 3, providing the fuel has sufficient-burnup and required storage configuration. New fuel will not be placed into pool B.

I (continued)

Crystal River Unit 3 B 3.7-72 Revision No.

42

.~:

~

-Spent Fuel Assembly Storage B 3.7.15 BASES

-BACKGROUND Bothof the.spent-.fuel.pools are constructed of reinforced (continued) concrete and lined:with,-stainless steel plate. -They are located.in,the,fuel handling -area of the auxiliary building.

Newfuel-storage requirements are addressed inSection 4.0, "Design Features APPLICABLE "The function of the s'pent fuel storage racks are to support SAFETY'ANALYSES and protect spentffuel assemblies from the time they are placed in 'the pool until,they are shipped offsite. The spent fuel assembly;'storage'LCO'was'derived from the need to establish.limiting'con'ditions on'fuel storage to assure sufficient safety' 'margin exists'to prevent inadvertent criticality. The spent Ifuel assemblies are stored entirely underwater in a confiigurati'onthat has been shown to result in a reactivity,.of less'than or equai to 0.95 under worse case

'conditiors:.(Ref. 1-,' 2, 6, 7 an'd 8).-- The spent fuel assembly enrichn'ent'.'requirements in'this 'LCO are' required to ensure inadvertent'-Criticalit9 does rhot occurin the spent fuel pool?

ji

.p' fuel Inadvertent crtiticality'within the fuel storage area could result idi offsite radiation' doses xce'eding 10 CFR 50.67 limits.

i~~~~~

,4

r.

S+tn.

f { * ; i:

-, The spent fuel iss'embly~storage satisfies Criterion 2 of the NRC Policy Statement.'

LCO Limits-on the newandci.rradiated fuel';.assembly storage in high densi~ty.-racks were established to.ensure the assumptions of

... the,criticality-,safety analysis of~the spent fuel pools is

-mai ntai ned *. '

i.

?

,?

Limits on, initial fuel.:enrichmentand burnup for both new

, and forspent-fuel stored.in pool A have been established.

Two limits are defined:

1.

Initial fuel enrichment must be less than or equal to 5.0 weight percent U-235, and (continued)

Crystal-River Unit 3 I -, B 13-.7-73 Crya RRevision-No. 42

I

.., - z I.'

I.-

I I-I.,

I Spent Fuel Assembly Storage B 3.7.15 BASES LCO;

'(continued i 1. -

I I 1,

2.

For new, low.-irradiation, and spent fuel with initial I) '

enrichment lessthan' or'equal to-5.0'weight percent and greater than'dr 'equal to 3'.5 wveight percent, fuel burnup must'be within the.limits specified in Figure 3.7.15-1.

Figuire.3.7.15-i-presents.'two areas of required fuel assembly burnup.as a function 'of initial enrichment.

For fuel with'enrichment'-burntip, combinations in the area-above the cure',. there are no restrictions on where the 'fuel'..can.be stored.

For fuel with.

enrichment-burnup cdmbinations below the curve, the fuel must be. stored in a.one-out-of-two checkerboard configuration',with water cells that contain no fuel.

Thfe acceptability of storing this fuel in the hkerbdoard conf iguration is d6cumented in References Fuel enrchrent limiits are based on avoiding inadvertent criticality in the-spent fuel pool'..., The CR-3 spent fuel storage system was sinitially designed'to a maximum enrichment of 3.5'weight'-perdcent. 'Enrichments'of up to 5.0 weight percent are permissible for storage in spent fuel pool A as long as the fuel burnup is sufficient-to limit the worst case reactivity in the storage pool. toless than or equal to 0.95.

Fuel burn'up reduces,.the' reactivity 'of the fuel due to the accumulation of fission product poisons. Reference 1 documenrts-,that the'`requiired blurnup' varies linearly as a functibon"'of'en'richmnt'wieth 10500. megawatt days per metric ton tiraniuni (Mwd/m U) required for fuel with 5.0 weight pecent en'ri'chner 1'ad 0 burn'up-'required for 3.5 weight p'ercent'enri'ched fuel.'.

Similar types of restrictions have been established for Pool B.

r

1.

Initialfuel enrichmentimust be s 5.0 weight percent;U-235, and, I

2..

For'f Ul'with'.jnitia

-e'nrichment ! 5.0 percent and' i2'.-0',weight percent, fuel must be'within.the limrits.'specified in 3.7.15-2.

weight

'burnup Figure (continued)

Crystal River Unit 3 B 3.7-74 Revision No.

42

i., -

Spent Fuel Assembly Storage B 3.7.15 BASES LCO (continued)'

Fuel with 'burnup-enrichmentcombinations in the area'above the, upper curve' has Kno 'res rictions on where 'it can be stored. ' Fuel,;with' burnup-enrichment combinations in the

,area between.the rower anid upper.curves must be stored in the peripheral cells'of,theipool. -The peripheral cells are those that are ad,4jacent to' the walls of the spent fuel

'pool. 'Fuel with b~urhnp-enrichment combinations inxthe area below the lower curve'tcannot be stored in Pool B,.but must be stored' in PooA."' '

, ~~

~

_V _,_

A L{i

,.APPLICABILITY,

'In general, limiting eenrichment of stored fuel prevents inadVdrtentIcritica1ityjin the storage'pools. Inadvertent criticality is dependent. on'whether fuel is stored in the pools and is compl~etely'independent of. plant MODE.

Therefo're,thisiLCOis applicable whenever any fuel assembly is stored6 in, high.density.fuel storagpelocations.

o ~~~~

~I.;

1I J.

' C t5i ;i ACTIONS' A.

1 equi,

,*4 tla;;*

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' R equiredAction`A.1 is modified by a Note indicating LCO 3.0.3 does, not applV.jince.,the.design basis accident of concern in this Speci'ficationlisjan -'nadvertent.criticality, and since

..,S...

tepsibilit or.consequences :of this-event are independent of plant MODE, E'.there; isisn reasonto shutdown the plant if the LCO or Requi red Actions cannot' be met.

When the configuration of fuel assemblies stored in the spent fuel pool is-notlin accordance with Figure 3.7.15-1 or Figure;3'.7.'15-2,'inmediate:action must be taken to make the necessary fiiel.assembl'y movement(s) to'bring the configuration" into compliance. The Immediate Completion

'time underscoresthe'e'ic~essity of' re'storing spent fuel pool

,fuelol6ading'to'wthwint the-fiini'tial assumptions of the

':critic'dlity'd analysis I

1,.

11

-I. " -

, i '!

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s

. I

. I -..

. I,

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I (continued)

Amnden No.

0-,

8 Amendment No. 208 C a-R

e.

r UB Crystal River Unit 3 B 3,7-75

1

I I Spent Fuel Assembly Storage B 3.7.15 BASES ACTIONS A.1 (continued)

The ACTIONS do not specify a time limit for completing.

movement of the affected fuel assemblies to their correct location. This 'is' not' meant to. allow an. unnecessary delay in resolution, but is a-reflection of. the fact that the complexity of the cbrrective actions is unknown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS

.Verification by administrative means that initial enrichment and burnup of fuel-assemblies in accordance with Figure 3.7.15-1 and'Figure.3.7.157'2 is' required prior to storage of spent fuel in storageipool. A or pool B (as applicable).- This surveillance ensures that fuel enrichment limits, as specified in:the'criticaalit~' safdtya-n'yses (Ref. 1, 2, 6, 7 and. 8Y, are-not excedded. The;survreillance Frequency (prior

.to-storage.in highs:density. regioniof the fuel storage pool) is approp riate since the initial,fuel enrichment and burnup cannot, change after removal from-.the core.

REFERENCES

'1.

' Criticality' Satfety Evalu'ation'of'the Pool A-Spent Fuel Storage Racks'in Crystal River-Unit;3 with Fuel of 5.0%

Enrichment, S. E. Turner, Holtec Report HI 931111, December 1993.; -

2.

Criti'ality.Safet'y-A'ina ysis,:6f thi Westinghouse Spent

'-'FueV'Storage'Racks in PoolV'B of Crystal River Unit 3, S.

E'.'Turner, H6ltec-'Report'HI -992128', May 1999.

3.

NUREG 0800, Standard Review Plan, Section 9.1.1 and 9.1.2, Rev..2, July 1981..

4.

10 CFR 50.67.;.i..;

5..

CR-3 FSAR,.Section.,.9.6.

6. Critic'ality"Safety,Analysis of4lth'e;Crystal River Unit 3 'Pool, A for'-Storageof 5%V Enr ichled Mark B-11 Fuel in Checkerboard!'Arrangement Witli Water Holes, S. E.

- Turner;! Holtec Report HIL-992285;<August 1999.

7.

Criticality Evaluation of CR3 Spent Fuel Pool Storage Racks with Mark B-12 Fuel, Hol~tec Report HI-2022907,

'Septe'mber 2002.

8.

Progress Energy Engineerinq Change EC No.

52456, "Documentation of Acceptability to Receive and Store Mk-B/HTP Fuel".

I Crystal River Unit 3 B 3.7-76 Revision No. 42

Control Complex Cooling System B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Control Complex Cooling System BASES t-BACKGROUND; The Control Complex Cooling 'System provides temperature control for the'control room and other'portions of the Control-Complex containing-safety related-equipment.

The Co plex Cooling System-consists of two redundant r pum

-anCntrolCaralleex oolci

:.chillers, associated chilled water pumps,and parallel duct mounted air..heat;-exchangers'that can'receive chilled water from-either chilled lwatier pump.

A'train consists of a

chiller and associt aed chilledwaterjnimp as well as a duct

'- 'mountedheat exchanger-that provide cooling of recirculated

,'con lcromplexir.a-.i The-,,dsjgnof,-the'.Control Complex

,Cooin. System,contains features thatiallow either train chiller and ass'ociated'ch'illed vwater pump to provide cooling capability,to;ei ther -duct mounted heat exchanger.

-Red ndant-6hillers'sandi chilled'water pumps are provided for suitable temperature conditions in the control complex for

-. operating personnel-and safety related.control equipment. The ContraolrComple,,Cooling System maintains.the nominal...

, : :.tempeaturebetwen'7-F.,:'and80°F.."

' i

-~. '

.s

';;,1!

'A single chiller and asociated chilled water pump will provide the required heat removal,for either duct mounted

,: i,

.. e'a ecagr.

.The1Ci nT.olCbm`06.Cooling System

operation. to.manaintaifn>ontroIl 'c`mplextemperature is discussed in the FSAR, Section 9.7 (Ref...1).

APPLICABLE.

.i.The Control Complex Cooling System: .consists of redundant, SAFETY ANALYSIS safety related-components; with some'common piping. The Control 'CompleV Co6oli Th-System maintains the temperature

'between OF..'and 80F..A-,single active failure of a omple.x;ooligSystem; c mponent idoes not impair the i, itjy of,t-hsystenito perform as designed. The Control Complex :Coolving7System.is designed in accordance with

.:Seismic:Category -I-requirements. The. Control Complex CoolinSystem'is aable of;remooVing heat loads from the

- 'conthrol rooi anrd -Iother ~prtions 'ofthe"Control Complex

containing, safety related eqUipment," including consideration

.equipment hea.'tloads.and (continued)

I Crystal River Unit 3 B 3.7-85 Amendment No.

182 I

, t -, -,

Control Complex Cooling System B 3.7.18 BASES APPLICABLE

' personnel occupancy requirements, toensure equipment SAFETY ANALYSIS OPERABILITY.

(continued)

The'Control Complex'CoolingSystem satisfies Criterion 3 of the NRC Policy Statement.,,

LCO Two redundant trains of the Control Complex Cooling System are-required to be"OPERABLE -to ensureethat at least one

-train is available,;assuming a single failure disables one redundant-component.;

A Control Complex Cooling train consi~sts of a chiller and associated chilled water pump as

of recirculated control complex air., All components of an OPERABLE train mustibie,energized by.lthe same train

' 'eleciritcal bi

'.To+al'systemafailiurei!c'ould cause control

-complex equipment'to-exceed its operating temperature

limits. In'addition-;`the Control Complex Cooling System must be OPERABLE tothe extent that.'air circulation can be maintained (See-Specification :3.7,.12)..

'I

-I APPLICABILITY In MOES 1, 2, 3, 'and4, the Cont'rol Complex Cooling System must be OPERABLE to ensure.that the control complex

' temperature ewill not exceed equipment OPERABILITY

. 'requirements.

- 5

>1 v

t sF-.,

ACTIONS

A.1

. -I-.

The LCO requires the OPERABILITY of a-number of independent subsystems'.:' Due!'t6`the `redundartcy and' diversity of subsystems, the Ai'operabil-ity:'of. one component in a'train does not render the Control Complex'Cooling-System-incapable of performing its safety function.

Neither does

-the.'.inoperabi.l.ity.of two.,different-components, each in a different train, necessarily'result in a loss of function

" for the Control' Complex.-Cool-ing System:

The intent.'of-this Condition is to maintain-a',combination of equipment such that the.cool'ing c'apability'equivalent to 100% of a single

.- train:remains available and in'.operation.

The word available means: 'The functional system placed in service

'does not-have to meet the requirement.of an operable train.

The functional-system will have to provide at least 100% of the cooling capability of a single operable Control Complex Cooling train.

This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.

(conti nued)

Crystal River Unit 3 B 3. 7-86 Revision No.

49

AC Sources-Operating B 3.8.1 BASES

-a- -

LCO The 230 kV and 500 kV substations, while part of the' (continued)

'. offsite netwo'rk,_ are not considered part of-the circuit required.by-this-LCO. *The9OPERABILITY of the circuit is supported by thesubstation.,provided the substation is.

capable of supplying the required post accident loads. Each EDG.must be-capable-of.starting, accelerating to rated speed and voltage, and connecting to its respective ES bus on detection of bus undervoltage. This must be accomplished within 10 seconds. Each EDG must also be capable of accepting.required loads within the assumed loading sequence intervals,, andac'6nt'tinue to'op~erate until offsite power can be restored'to'th'e&ES buses..These capabilities are required to be met fromavariety of initial conditions, such as the EDG in sfandby.-with-the engine hot and the EDG instandb'y with'tke'een'gine at ambient conditions. Proper

I b,'

' ; "din shedding of non-essential

,... -'aly-e q u'isr;reqhir ed>,0functio nfor EDG OPERABILITY.

EDGcOPERABILITY requiresproper ventilation using EDG Air

- Handling Systemicooling';'fan(s)'forieach EDG in order to maintain :theitemperature:of ~the EDG engine room and EDG control room within manufacturer's limits. Based on

.. analysis, single-fan-or dual-fan operation is acceptable

-dependent'uponj'fan supply air.temperature.

,Th eAC sources -nW.one

.rain" mus be sep'arate and independent

- (to the extent possible) ofithe;AC.sources in the other tra'in'. For the EDGs'; separation and independence are complete. For,the offsite AC.sour~ces, separation and..

'independence exist to the extent practical. A circuit.may

~ --

ec

'inced tomore'tha'n one ES bus and notvilt separation criteria.A'cir~cibit'that is not connected to an ES.bus-is b;'abality for the operator to transfer' power..to.,the.ES' busesjina.order to be considered

- - aa OPERABLE'.:.'.'-.,:'a,.

-a~j

.', -1;.% 'fffI zi,,

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, I

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i

-8'i 1A

-APPLICABILITY, Twoaonsite;,:and,two.aofffsite-AC sources are required to be

-OPERABLE "in':MODES T1',92, 3, and 4 to-ensure that:

.,.f: '

.e

-- t '

i m a-. -

{

a tablenlimits and reactor coolant

.. uimits rrenot exceeded as a result

- of anticipatedoperational occurrences (A0Os) or

'.abnormal1,transients;

-and a

a o

a

b.

Adequate co're cooling is provided and containment OPERABILITY and 'other vital functions are maintained in the event of a postulated DBA.

-(continued)

Crystal River Unit 3 A B 3.8-5

'Amendment No. 160

^A in AC Sources-Operating B 3.8.I BASES APPLICABILITY' AC power requirements for MODES 5 and: 6 are (continued) addressed in LCO 3.8.2, "AC Sources-Shutdown."

ACTIONS A.1

..To ensure a-highly reliable power source remains with one offsite circuit in'operable,'it is.'necessary to verify the OPERABILITY of the. remaining required offsite circuit on a more frecueni basis...

,Since'the Require& Action only specifies "perform," a failure.of SR,3.8.1..l.acceptance criteria does not result in a Required.Actionnot met (Condition F). However, if the remaining required circuit fails SR 3.8.1.1, the second

..offsite circuit is'inoperable,.and Condition C, for two offsite-circuits inoperable, is.

entered.

.A2,

2 Required.Action.A.2, which only applies if the train cannot be powered from an offsite, source,' isintended to provide

'assurance that '.'an'event.:coiincident 1 with a single failure of the`.associated. EDG-.will`,`not.,`res'ult;,in. a complete loss of safety o fjdund antrequired features. These features are powered-from the redundant AC electrical power train. Single train systems (from an electrical perspective),..such

.as.the..turbine driven'emergency feedwater pump,!are notincluded.:.

-The Completion Time for'Required-Action A.2 is intended to allow the operator,time to evaluate and repair any discovered i'noperab ii'ties. This Completion Time also allows for an exception.to`th-e.nor'aI'a "time zero". for.

beginning the all6wed.outtag'e time "cloick." In this Required Action, the.Compl-etion Time'.6nl y'begins on discovery that both:

a.

-The train has no offsite power supplying its loads; and.

b.

A required feature on the other train is inoperable.

I (continued)

-Revision' No. 43 Crystal,'.River Unit 3C l RB 3.8-6

AC Sources-Operating B 3.8.1

T.

BASES dI t..

ACTIONS A.2 (continued)

..,II

.1.

~

If t-aytimedu' the existence ofCondition A (one offsiite ci'rcuit inoperable)both".'a and Wb above become mnet','.this 'Coin"I iiion'-Time begins'to-be tracked.

-The 'remainingi OPERABLE off'it crut and EDGs are adequate to-supply. electrjcal po~wer :to Train A'and Train B of the onsite Class 1Ed istr.ibutinsse.Te2horCmltn remdundantsin'to~ acco~n tecmnntOEAITYof the redndntc6onteirpri~tiolthe~inoperable required feature.

Additionally,-the, 24ihoUr C'ompletion-Timie takes into account the capacity and.capability-of the remaining AC sources, a

'.,rasonAble;-time for rearand the low probability of a

.2DBA occurring -during.this period.

~ cA.

3J?

I I

According to the recommendations of Regulatory Guide 1.93 (Ref. 6),, operatioin with',one' required-offsite circuit inoperable should be limited to~a period of time not to exceed 72 houirs-J.._h this.-.conditibn(,,the reliability of the off'site.-syste~m'.i's.de'6raded,"anid the'-potentiaI for a loss of offsiit6e,'owerjsnres with atte~ndant potential for a chal epge,to`.)the" un iftisaf et y systems",,,.However, the

'remnainin-6:OPERABLE'o'ffs'i-e circuitand'EDGs are adequate to su'pply-electr.hai power to t-he ont'Cass lE distribution system'

~*

The,72 hour: Completion.Time ~takes-into~account the capacity and capability of-thef remaining AC:sources, a reasonable time for repairs, and the low probability of a DBA occurring

during this period.

Th~6'dy l'days~withthe"'alternate"AC source available) t~otojfi:fi

Required Action A.'3 establishes a limit AC. power;. sources to be'inop'erable.dueihg any single con iguou '-occUrre-nc64bf:fai1lure-to. meet the LCO. If Condition-Als entered~while,- for instance, an EDG is inoperabile.;and~7hat,-EDG js:subsequently returned to upto-'14,days.' 'This could lead:'1to..a_`t'o'ta1 of 17 days, since initial fa'ilure-to meet the LCO, to restore the I'

.~

offsite circuit.:

I (continued)

Crystal RiVer Unit 3

-. B:-3.8-7 Amendment.-No. 207

1

AC Sources-Operating B 3.8.1 BASES ACTIONS A.3 (continued)

The 6 day and i7 day4Complehtioril Times provide limits on the time allowed'in a specified'codition

'after discovery of failure to meet the LCO>' 'This'limit is considered reasonable for'situati'ons i which-Conditions A and B are entered concurrentli.:-

wh o

As in Required Action A.2, the Completion Time allows for an exception'to the normal "time zero" for beginning the allowed outage'time "clock." This: will'result in establishing the "time zero" atithe time that the LCO was initially'not met,-;instead of'at the time Condition A was entered'.

ig;ly'rei'1, w~'

,,c

,i_.n I;l B.1l

-To ensure a highly reliable power 'source in the event one EDG islinoperable, it is necessary' tfo'verify the availability of the'-OPERABLE'offsite ci'rcuits on a more frequent basis. Sirice the Requireid'Action only specifies "perform,!a failure of SR3.81.1'acceptance criteria does not result in a Required'Action being not met (Condition F).

However,-

if'alcircuit fails to passSR3.8.1.1, it is inoperable.' Upon'-offsite'circuit i'noperability, additional Conditions and Required Actions mstithen be entered.

B.2 Required A'ction B.B2`is intended to pr6vide assurance that a loss of-offsite pdwer,; during the' perif'd that a EDG is inoperable;Adoes-not res'ul'tijn a`'c6mplete loss of safety function of critical red ndan.-reired features. These features are designed with redundant safety related trains.

'Redundant required feature failures consist of inoperable

-features associated with a train,-'redundant to the train that has an'inoperable EDG. Single train systems (from an electrical perspective), such as the turbine driven emergency feedwater pump, are not included.

(continued)

Crystal River Unit 3 B 3.8-8 Amendment No. 207

AC Sources-Operating B 3.8.1 BASES ACTIONS B.2 (continued)

. -Z I S I

-The.Completion;T~ime.,for Required-Action B.2 is intended to

-,allow-,the operator..,time to evaluate and repair any discovered,inoperabilties.,This Com'pletion Time also allows 'foran exception-to the normal "time zero" for

'beinning.the'allowed,outage' time "clock." In this Required

.',Actionthe CompletionTimeonlybegins on discovery that Abothin,.,,-,..,,tion.. -

.a. A EDG isinoperable;.and

b.

A required feature on the other. train is inoperable.

I

'If at:any tiime during ihe existence of 'this Condition (one EDG inoperable) a required feature,subsequently becomes inoperable, this Completion'Time'begins to be tracked.

Declaring.,,the.requiredrfeatures inoperable within four hours from thedi'scovery.of items.'a' and tb'l. existing concurrently is*-acceptable because it~minimizes risk while allowingtimefor restoration.,befor~e.subjecting the plant to

.transients associatedwith shutdown.

,;In,'this,;Condition,-th~e-,remaining OPERABLE EDG and offsite

'circuits 'are adequate,.to supply electr~ical power to the

.. 'onsiteClass.1E distr.ibution system::-Thus, on a component basis, single-failure protection for-the required feature's

-function.may havebeen'lost; however,' function has not been

'lost.' The 4:hour'Crirhpletion Time tak'es.into account the OPERABILITY of.the redundant counterpart to the inoperable

,requir,ed featu'rie'.Add-i-tionally '*the'e4:hour Completion Time takes into account-th,,e,'~capacity and capability of the remainingACsources,,'areasonabletime for repairs, and the low probability of a.,,DBA occurring during this period.

J; tJ S-,,::,,

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,>v;it5, Crystal River Unit 3 B 3.8-9 Crystal-River Unit 3 I- - -B 13.8-9 (continued)

Amendment No. 182

!. - --. I I

AC ourcs-Opeatin

.m AC Sources-Operating B 3.8.1 BASES ACTIONS B.3.1fand B'.3.2 Required Action.B.3.1 provides an option to testing the OPERABLE EDG in order to-avoid unnecessary testing. If it can.be determined that.the.cause.of'.the inoperable EDG does not.:exist on the OPERABLE1EDG,: SR 3.8.1.2 does not have to be performed. If the'cause,of inoperability exists on the other EDG' the'other EDG would be.decared inoperable upon discovery and Condition E of LCO'3.8.1 would be entered.

If the conmmoncause"failure evaluation is indeterminate (the cause of the initial inoperable EDG cannot be confirmed not exist on the remaining'EDG),--performance of SR 3.8.1.2 is adequate to provide assurance of continued OPERABILITY of that EDG.;-

.The Completion -Time of: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 'is.reasonable to confirm

- ':.:t....

-thatthe OPERABLE EDG is not-affected by the same problem as the.-inoperable.,EDG and isbased'on the recommendations of Generic Letter 84-15. (Ref-. 7).:..

~~~~*

r B.4.

'Accordi'ng. to -the recommendations"of'.Regulatory Guide 1.93 (Ref. 6),-operation"with'dne& EDG inoperable should be limited to a per'io'd' not to exee'd "72.hours.

The completion time'may'be'extendid to14' days'if 'ftaternate AC (MC) power is av'ailable'o'r a one'time basis'as described in the footnote to'the Comnpletion'Time. -:The'alternate AC source must be...capable-of;.being.'aligned-to:the same bus as the inoperable EDG and must be'capable.of supporting loads required for safe shutdown of the reactor.

In Condition.B,'.the remaining OPERABLE.EDG, MAC source and offsite circuits are' adequate.to supply electrical power to the onsite ClasslE distribuition'..system. The Completion Time take's into' acco'unt'th'e'capacity'and'capability of the remaining AC sources, a reasonable time for repairs, the ability to perform online preventative.maintenance, and the low probability of a DBA occurrihng during this period.

During'online preventative maintenance that is planned to take over 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the following compensatory measures will be put in place prior to initiating the activity:

(continued)

Crystal River Unit 3 B 3.8-10 Amendment No. 207 I

. 'AC Sources-Operating B 3.8.1 BASES ACTIONS B.4 (continued)

(continued)-

-Availability will be assured during an extended EDG-AOT

.-,- +

by the following:.-

i.

, ~~~~*.

e,go. 'r,'a;A X'!...'

. Starting-the MAC:.and assuring-:proper operation prior to

..removing the.EDG. fromservice, A.

Verifying-every. 72hoursthat-a-24-hour fuel supply is onsite,,and

Ensuring:,the AAC.;s -electrically.'and mechanically ready

' for manual ope'ration-and can..be-aligned to supply the applicable safety-:related' buswith:simple operator

,action ev'ery"'72-,hours'.

CR-3 will perform procedure CP-253,."Power Operation Risk Assessment and Management," which requires both a

~. ;. ~.-

.. :.,.determiniisti~c.,and:.probabilistic-;evaluation of risk for the performance.of all'.maintenance -activities. This procedure

.. :. uses the Level.1.PSA.model to-evaluate'the impact of maintenance activities'on CDF.t CR-3 will not plan any maintenance that results in "Higher Risk" (Orange Color Code) during EDG maintenance.

ECCS equipment, emergency.feedwater,.Control Complex Cooling and auxil iary.'feedwater(FWP-7and'MTDG-1) will be

desi gnated,,adiini str'a!t~j v'el,- -as '"protec'ted" (no planned maintenance,or, discretinary equipment manipulation).

The e"discretionaqryquipment, manipulation" is not intended to preclude manipulations requiredfor normal operation of

. -;the plant,- required surveillances or.operator response to

,abnormal: conditions. Jt,!,;..

masfi gne ve-ya "proetd (no pl ianned, *.,*

Prior to initiating a planned EDG outage, CR-3 will verify

: i -the-availAbiliey'i.of '-ffsite',poiw'er' to 'the 230 kV switchyard and&einsiure-that'-the apability td '0iower both ES busses is available f rom -each -of 1the two ES'offsite power transformers

,*,;,;,;, *, i

( O P

, B T ) W r

.~

Act

-i

,d.,.

BEST CR-3 w~i~l.,not ii.itate. an EDG extended preventive maintenance outage if adverse weather, as designated by Emergency reredness-procedures, *is anticipated.

"4.,,.l Ei A.,

j 1,

r No elective maintenance will;be scheduled in the switchyard that would challenge the availability of offsite power to the ES busses.

(continued)

Crystal River Unit 3

-B 3.8-bOA Amendment No. 207

-I--

t' ;;

B.

BASES AC Sources-Operating B 3.8.1 ACTIONS i

. (continued)

B.4 (continued)

A periodic fire'watch'will be established-in fire areas thai:

are considered risk-significant by the IPEEE, affect both EDGs or have increased risk significance due to EDG maintenance...: -The fire areas are listed in Table B 3.8.1-1.

The 17-day'Completion Time'for Requ'ired Action B.4 -

establishes a limit on the maximum time allowed for any combination of required'AC'powe'r'sources to be inoperable during any single -'cntiguous occurrence of failure to meet

-the-LCO'. Refer'to the'Bases'for Required Action A.3 for additional information on'this Completion Time.

- I -,

I.

... 4. :. -

,., :-'l

','-'I I

I..

"I.

- I '

-, ! 1,. ;

' " '.. 1

1.

j

v J.,'

o'%.

-t i

_rs a R i e U n i 3

_ ~

O (continued)

Amendment No. 207 Crystal, River Unit 3 I I e B 3.8-10B

AC Sources-Operating B 3.8.1 Table B 3.8.1-1 g, pw ZONEZN DSRITO FRE ZNE,GGED AUO IGNITION 0RAI IFREQUENCV

? '

.. et-Pipe 1

`AB-11976A

' x x NORTH HALLWAY Sprinkler-dual hour, 9.73E-05

?

j 1',

evel,,3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months I

?Wet-Pipe

, 1

_AB-119__E_

x EASTHALLWAY Sprinkler-dual hour, 1.73E-04

.vel 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

.:AB-1319.6).C3):

xt ENTRALHALLWAY-,

1dtPp 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 2.36E 04 z2.z....:,,Sprinkler

.A,196r3

, -?xj~,RLHLWY;,,t';.

'Sr r.......

.36E-04 A --

AB-l9,-6K (2) l x DECONTAMINATION ROOM

,et-Pipe 1.02E-04

?

AB11-7

! EMERGENCY DIESEL~GENERATOR~j

,Pre-Action

-: hou 1.73E-04 CONTROL ROOM 313 Sprinkler 1hor13E0

,'AB-119 7B (3) l lEMERGENCY DIESEL GENERATOR ROOM Pre-Action 5.30E-03.

Li-1 B S rnk --

ler ---- ---

'EMERGENCY DIESEL GENERATOR ROOM Pre-Action AB-119-8A (2) x

,CONTROLROOM 3A.Sprinkler 1.02E-04

~~~~~~~........

EMERGENCY DIESEL-GENERATOR: ROO Pre-Action 500

.AB-119 8A (2) x

[CONROL ROM3`

prinkler CC-108-103 x

x ;PLANT BATTERY ROOM 3B None

'3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 9.73E-05 CC-108-104 J

.H x TPLANTOBATTERY ROOM 3A None

,3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 9.73E-05 CC-108-105

x, x PATB CHARGERY ROOM 3B' None 13 hour1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months 4.03E-04 (108 '

I P1ANBATTERY 3 hor 1x x BATTERY CHARGER 'ROOM 3A None 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 3.68E-04

!CC-108-106 I

..14160V ES SWITCHGEAR BUS NROOM 3B one

3 hour 2.27E-04 1 CC-108-108 I x l x l4160V ES SWITCHGEAR BUS ROOM 3A None 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 2.60E-04 i_.._ !,_

CC-108-109 N1 x

x INVERTER ROOM3B None 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 2.14E-04

, ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~...............

CC-108-110 x

x IINVERTER,ROOM 3A one

,3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 1.90E-04 ICC-124-111 I

Wet-Pipe hu, 50E0 1

x,, x,CRD t& COMMUNICATION EQUIP, ROOM Sprinkler 3hour 5.06E-04ou

,(

-ESIHABUS ROOM 3B None 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 1.90E-04 C1224 1480V ES SWITCHGEAR BUS ROOM 3A None 2.04E-04

__.__.4 lCC13 x

x :CABLE SPREADING ROOM oal Foodin 9.73E-05

.Halon Room....

_._._i_.__.........................._.........__....___.__.........__.___,_.............,.,,,,,,,.,,,,,.,,,,,,.,,,,,,,,,,,,,,,,,,,,.

C-145-118B 1.24E-04

_x x CONTROL ROOM

.one 1__4___

(1) Fire zone (2)

Fire zone (3)

Fire zone identified as risk significant per IPEEE may have increased significance when EGDG-1B is in maintenance may have increased significance when EGDG-1A is in maintenance I

-(continued)

Crystal~

Rient3.

~81C.mnmn o

0 Crystal River Unit 3 L 1. -,B 3.8-10C 4:Amendment No. 207

a.. ~.

AC Sources-Operatinc B 3.8.1 BASES ACTIONS C.1 and C.2 2 (continued)'

Required Action C.1, which applies when both required offsite circuits are inoperabl6,is'intended to provide' assurance that'a DBA, coincide'ntwith a worst-case single failure,' willnot result in-a complete loss of redundant required safety functions. The Compl'etion Time for declaring the redundant required'features inoperable is

.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; reduced from that allowed for one train without offsite power. (Required Action A.2). The rationale for the

'gureduction.

t 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is.

that Regulatory Guide 1.93 (Ref. 6) allows.a CompltionTimeof 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for two required offsite circuits-inoperable, based.upon the assumption that two' complete safety trains are.OPERABLE. When a concurrent redundant required.feature'failure.exists, this assumption

-is.

no.longer valid,:.anda. shorter Completion Time'6f

. :12hurs'isappropri'atet',These features are powered. from redundant AC safety'trains. Single'train features (from an electrical perspective),-'such'as-'the turbine driven

emergency.feedwatee"pump, arenot.Jincluuded.

The Completion Time for Required Action C.1 is intended to allow,,the operatorntime to'evaluate and repair any.

discovered inoperatilities. This-Completion Time also Gallows for an-exception;to-thenormal "time zero" for.

beginning:the allowed outage;time-."clock." In this Required Action; the--CompletioonrTimeonly begins on discovery-that both:.,.

~~~

- ..*.e..-

a. All requi'redo'ffsite circuits are inoperable; and

b. A required fe tur e'is in6perable.

If at:any'tiedtiring 'the existence of Condition C (two.

offsite circuitts ihn6perablj)' a'required feature becomes

-inoperable, this Completion-.,Time begins to be tracked..

I.

7.:

I (conti nued)

Crystal River Unit 3

> B 3.8-11

-Amendment No.

182

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.8 (continued)

REQUIREMENTS In order to ensure thatithe EDG is tested under load conditions that`arezas close.to design'basis conditions as

-.possible; testing.must-be performed using a power factor 0.9. This power factor is chosento be representative of the'actual.,design basis..inductive loading that the EDG would

.experience.-

'i""

t t

-This-SR is modified'by two Notes.'The reason for Note 1 is that during'power.operation, performance of this SR could cause perturbations.to, 'the electrical distribution systems

,that could'challeng&e'continued steady state operation and,

'as a result, safetysystem's.. This restriction from normally performing the'-Surveillance in'MODE'1 or 2 is further

.-.amplified..tom.al.10w'.the',YSurveiliafice;toibe performed for the purposesof reestablishingOPERABILITY: .(e.g., post work testing folowing-periodic governor replacement, corrective

.-maintenance, corrective modification,,deficient or incomplete'su'rveillan'c1.testing, and other unanticipated

'OPERABIIY;'concerns)-provided an assessment determines plant safety is maintained-or-enhanced.-This assessment

hAll, as'a-mihimLim,-'onsider the potential outcomes and trans'ients ass6cia'ted with'a failed -Surveillance, a successful.Surilance 'and a perturbation of the offsite

or'onsite'lsysterm.wher"-they,'ar&-tied-'together or operated independently.for.theSuSrveillance,as well as the operator procedures'available6irto'cope with these outcomes. These shall be measured against the avoided.Frisk of a plant shutdown and startup to determine that plant safety is maintained..or. enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be

- used..for this.8'assesment:".However',the Note recognizes that

.'shduld an dfplinred event occur in'MODES 1 or 2, following

'verification that theacceptance'.criteria of the SR are met, i' ` the.6vent

.bel ce'difeA a'successful performance of this'SR.

aote'2'acknThledges this SR may be performed using

.'-..omoenlods~or it may b form"'d.'by paralleling the EDG.with offsite power.- When-the SR is.performed with the EDG carrying*the'-4160 Volt ES:bus, the power factor of the EDG is a'function 'of thexreactive component of the loads powered from'it, and'as such, is not under direct control of the operator.,

4 v

SR 3.8.1.9 9

Per the recommendations-of Regulatory Guide 1.108 (Ref. 9),

paragraph.2.a.(2),^ each EDG is-required to demonstrate (continued)

Crystal River Unit 3

. B 3.8-20

-Revision No. 43

AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued) proper operation for the DBA loading sequence to en'sure that voltage and freq'uency are maintained within the required limits. Under accident conditions prior to conne`cting.the EDGS'-to their: respective', buses, all loads aret she'd except load block 1' feeder'breakers that power Class lE loads (referred todas_"perhmanently connected" loads)'. Upon reaching the requiired'voltage and frequency, the-EDGs'are'auto-connected to their respective 4160 V buses. Loads are then 'sequentially con'niected to the bus by the automatiic load's'equencing' relays'.'

The sequencing logic controls the'permissive and starting signals to motor breakers to prevent'overloading of the EDGs due to high motor'starting currents.;

'" " ;The 10% lo

's'eq'uence'time interval tolerance ensures that

'sufficient 'time exists fdr the EGG to restore frequency and voltage prior, to ap'pying'the' next load and that safety analysis,assumptions regarding ES equipment response times are not violated.: Reference 2'provides'a summary of the

automatic loading ofES buses.

-The'Frequtelicy:'6f 24f months takes'.into' consideration plant codnditions ne'eded'"to6pe'r rmf t h"'eSu`rv illance and is intended ito b'e conis9itenet wi't'i th eexpected fuel cycle

' ~~len'gth.

a g :*,,,,;...............

~~*

.!A;;,;f SR 3.8. 1. 10 the'evntof aDBA coincident with 'a loss of offsite power, the EDGsa required to supply the necessary power to ES systems so that the -fuel, RCS,'"and containment design limits are not exceeded.-'

This Surveillance-demonstr ates 'theiEDG' operation during a loss dfoffsite p6wer actuati'6n test signal in conjunction

- with an ES actuati.n signal.

n lieu 'of' actual demonstration of connectionind loadingof loads, testing that adeqhuately shows the'capability of the EDG to perform

' these functions is' acceptable." This testing may include any series'of sequential, overlapping, bor total steps so that the entire connection and loading sequence is verified.

The:Frequency of 24 months takes into consideration plant conditions'needed to perform the Surveillance.

(continued)

Crystal'River Unit 3

-* B 3.8-21

'Revision No. 43

-,AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.10 (continued).,

REQUIREMENTS

...This.SR.is modified.by three Notes..The reason for Note 1

~-,;..-,..,,,.,

nis.to.'minimize',wear, and tarion the'EDGs during testing.

Forithe purpose.o i testing, he'EDGs may be started from standby conditions, 'that.is, with the engine coolant

-'..'.'ndolcntinuosycruIlated andtemperature maintained

.consistent withi mannfacturer recommendations for EDGs. The

,.reason foi.Note';'.2 is -that performing the Surveillance would remove.a~ requi red offsite,:circuit from,.service, perturb the electrical distribu'tion system,.and, potentially challenge

",.'safety systems'..However;, Note 2 acknowledges that should an unplanned'vent'oc tr in MODES 1, 2 or 3, following

. verification thai-e..cceptance criteria of the SR are met, the event"carinbe credited as asuccessful performance of this SR. Note-3:is'an`SR 3.0.4 type allowance to place the

-plant in6.MODE/,4,,for-th'e purposes 'of-performing this SurveiV ancee. This is.necessarjyin'order to establish the pre-,requisite plant6.onfiguration needed to perform the SR.

SR 3.8.1 li ThisSurveillance.de'monstrates 'the EDGs are capable of

.synchronizing-andaccepting a load greater than or equal to

.the axmsdte'>dy state accident loads, which are

automatically-connected accident-'loads and required

'mauly ap'plie'd accident.loads. However, the upper limit of the 200 h6ur"'service rating is 'still available for flexibility in post'accident EDG load.management, including short duration loads. The test'load'band is provided to ravoid routine.verloading'of'the EDGs. 'Routine overloading

, maytresult in moreTfrvequent.tearidown inspections, in accordance with vendojr r.ecommendations,. in order to maintain EDG OPERABILIT.Y. 2

.The.60.minute'run.time'is provided to stabilize the engine

. te ensyrest cooling and lubrication are adequateforextended

)periods of o'peration.

'... The',24 month ',Frequency takes'into consideration plant conditions :reqguired 'to'perform the'.Surveillance and is J

.tintendeddt'be.consistent with expe'cted fuel cycle lengths.

This' Su'rveillan&e isf mo'dified by two N6tes. Note 1 states

,that momentary -transients due.to changing bus loads do not invalidate this test.'The reason for Note 2 is that during (continued)

Crystal Rfver Unit 3 B 3.8-22

.Amendment No. 207

A

-.t~j,--

.AC Sources-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.11 (continued)

REQUIREMENTS' operation with'the reactor. critical,performance of this Surveillance could cause perturbations to the electrical distribution systems' that cduld challenge continued steady state.operation and, as 'a result,'plant safety systems.

;This restriction from-'nor'mally performing the Surveillance in MODE 1 or 2.is further.amplified to allow the Surveillance.to be'performed, for..the purpose of

reestablishing OPERABILITY (e'.g.,' post work testing following periodic:-governor replacement, corrective

-maintenance, corrective modification, deficient or incomplete surveillance testing,. and other unanticipated OPERABILITY concerns) provided.an assessment determines

~plant safety is maintained or.enhanced. This assessment

'shall as a minimum',' consider the potential outcomes and transients associated with a-failed'Surveillance, a successful.Surveil

ance, and aperturbation of the offsite

,or.ons.ite system..when.theyare.tied-itogether or operated

'independent]

yfor.the'Surveil.lance;.as well as the operator procedures available to cope with these outcomes. These shall be measured against the-avoided risk of a plant shutdown and startup to determine that'plant safety is maintained or enhanced when theSurveillance is performed in MODE 1 or.2. Risk insights.or deterministic methods may be used'for this assessment.: However,'.the Note acknowledges thati cedit may be:taken for.tnplanned events that satisfy this SR.'.

REFERENCES" 1.'

10

-CFR 5'0, Appendix A, GDC 17.

2.

FSAR, Chapter 8.

3. -Regulatory. Guide'1.9, Rev. 3, July 1993.

'4. FSAR, Chapter 6.

5.

FSAR, Chapter 14.

6.

Regulatory Guide 1.93, Rev. O, December 1974'.'

7.

Generic Letter 84-15..

8.

10 CFR 50, Ap'pendix A, GDC 18.

9.

Regulatory Guide 1'.108, Rev. 1,: August 1977.

10.

Regulatory Guide 1.137, Rev. 1, October 1979.

11.

ANSI C84.1-1982.

12.

ASME, Boiler and Pressure Vessel Code,Section XI.

13.

Deleted.

Crystal River Unit 3 B 3.8-23 Revision No. 43

AC Sources -

Shutdown B 3.8.2 BASES ACTIONS '77 A.2.1.A.2.2'

..A.2.3.. B.1a.

B.2. and B.3 (continued)

With the offsite circuit not available to all required:

trains',-the'option'would still exist.to declare all required

'features-:inoperableSince this.6option-may involve undesi red 'Administrative effo'ts,'-alternative conservative action's are'provided.,. With the required offsite circuit

-: ' inoperble'!the minimum required diversity of AC power sources'is not vavailable.`In*'this'condition it is required to take actions' t6 omiinimize the'probbability or the occurrence f"postulated events.'This is done by suspending

-,-,CORE ALTERATIONS and:initiating7action-to suspend operations rinvolving positivere'activity:additiohs. Suspension of

'these activities'-d6es'not:;preclude-completion of actions to establish -a'safe.c6r

-ervative co6diti6n. Additionally, the Recuired Action'toinitiate action to,suspend positive

' ~

-reactivity additions' does not,'preclude:-actions to maintain

--orreduce RCSatemp ratur e6,

.-to'r trintdin or increase RCS inventory' r'viaed the, required

-SDM 'is maintained.

-' -'Notwithstandirig-performance of th e'conservative Required Actions,'theplant:'isstill without sufficient AC power surce'-to o'perate 'i-nasafe manne r.'.Therefore, action must

.beinitiatedtorestrethe minimum 'required AC power sources and continue until'the LCOdrequirements are

-restored.-The restoration of the required AC electrical

,,power. sources-should bet.completed ina timely.manner in order to minimize'-the ;jimeduring.which plant safety systems may be-without 'suffi-cient power.

t...................

The immediate'Compietion Time of these ACTIONS is consistent withothers requi ring prompt attention SURVEILLANCE'.xSR 3.8.2.1 ae:, [

-REQUIREMENTS- -

-r; I;

SR3.8.2.1 requires'the SRsfrom'Specification 3.8.1 that

' othare necesay orensring the,'OPERABILITY of the AC sources n

'other than,, DES L 2,2,-3, and4.'.SR:3.8.1.7 is not requ"i'red t eobe m'et', bescause-with only one offsite circuit requiored to."be'OPERABLE, there'.will not be an alternate t -- a -

1.~

- b-r

~ ~

~

U (continued)

Crystal River Unit 3

.--.B 3.8-28 Amendment No. 149

1

AC Sources - Shutdown B 3.8.2 BASES SURVEILLANCE SR-3.8.2.1 (continued)

REQUIREMENTS source to transfer to. SR 3.8.1.9 and SR 3.8.1.10 are also not required-to be met since the ability to respond to ES actuations in'other'than MODES 1o 2, 3 or 4 is not a requirement-for-CR-3.

-This SR.is modified-by a-Note. The NOTE indicates SRs-3.8.1.3,' 3.8'.1.8,'and 3.'8'.1.1iare not'required to be performed to 'comply. with SR.3.8'.2.1; -The reason for the

'Noteiis to preclude.situitions in which the OPERABLE EDG.

would'be'paralleled with"'the offsite power network or

-r rendered inoperable during performance.of SRs. With limited AC-sources available, a single~event could compromise both the required--offsit'e circuit and-the EDG. It is the intent

that these SRs,must still be capable of being met, but actual performance is not required. Refer to the Bases for LCO 3.8.1 for'a discussion of e'ach SR..

Addi-ionally,;the impact of:enter.ing the ACTIONS of this Specification during-Surveillance-testing was evaluated as part of ITS implementation and a unique situation was discovered. relativeto the, monthlyEDG testing performed as

.,part of'surveillanceprocedures-SP-354A and SP-354B and the MODES and-Conditionsaddressecdby',.this'Specification.

The EDG tear-down-and inspection may.,beiperformed during MODES 5

'and 6-ouiages.

Thus, it"was-concludedCR-3 could

-coincidentally have.onlyooner'OPERABLE'EDG and be required to pe'rform SP-354.'Thi's created a.. concern.since certain portions'of SP-354:'render'the EDG;'inoperable, requiring entry'into.the--ACTIONS.of.thi's'Specification.

Required Action B.3. requires action be initiated immediately to restore the'EDG'to OPERABLE'statu's. Thus, a situation is established in'whirch-SP-354 is performed, the ACTIONS are

~

"entered, SP-354:'is i'mmediately terminated (in order to restore OPERABILITY of"the EDG).

In' this case, SP-354 could never be performed.

,, -v-A The need to perform SP-354 was'evaluated in light of this concern. It was decided that since the periods of. -

inoperability for.jthe performance-.of-the portions ofSP-354

-associated.with SR-3.8;1-;2 and-3.8.-1.6:were brief and the SRs-provided c6nfidence.inthe OPERABILITY of the EDG, the SP should,.be performed... Expeditiously 'performing the portionsof SP735'4:assoc'iated wiith-SR.'3.8.1.2 and 3.8.1.6

' during.,these'MODES satisfies theiint'ent of Required Action B.3 and'i's therefore'acceptable. -Fo6rT.IODE 5 and 6 surveillance testing performed in accordance with SP-354 to meet SR 3.8.1.3 w-1l be performed at the frequency of SR 3.8.1.3 only when.both EDG trains are operable.

REFERENCES None.

Crystal River Unit 3 B 3. 8-29 C

- Revision No. 43

~.--.; -;....rK;'.,i Containment Penetrations B 3.9.3 B.3.9 REFUELING OPERATIONS

- B 3.9.'3 Containment'Penetrations BASES BACKGROUND

An -accidenrt wh'ich ccurs during movement of recently

iadiated fuel ass'emblies within containment will have any released-'radioactivityiimited 'frdm escaping to the environment. In :MODEJ6, the: potential for containment pressurization as'.-a result 'of'anaccident is not likely;

,therefore,9th'e'requirement~to6isolate the containment from the 'outside atmo'sphere.is'.less.stringent than those

established'for MODES.1 through'.4. In'order to make this distinction, the'penetration requirements are referred to as "containment closur e" rather.;than "containment OPERABILITY."

Contai'hnment'closure means that all'potential escape paths

'closed-or' capable of being closed by an' OPERABLE cbntainiment purge or mini-purge valve.

Q.~

The containment equipment hatch.or outage equipment hatch

'(OEH)p'r'ovides'-a mean fo'r'movinig large equipment and components;in't6-and!'but of containme'nt.

During movement of

r;ecetly'irr'adiated'sfuel' assembl-ies within containment, the equipmentihatch br 'OEH.musit'be':'held

ini'place by at least

. four.bolts.-:The' required '.number 'of bolts is based on dead

,weig ht;and is'acceptable-due"to'the'l'ow likelihood of a pressur ization.eve'nt.Good4 ehgineering practice dictates

,.',that.theboltsrequired b

-y.this LCO,be approximately equall'y spacedd'.DJUri'ng 'movement,of,'recently irradiated fuelassemblies'jithin containment,-containment closure is required; tierefore/.the'door.in the"OEH (if installed)

,,ust.always r O'emair'cipsed..

The containment air locks provide a imeans for personnel

:1 access-during)'MODESt1,; 2,.3;. and 4in" accordance with LCO 3.:6:2, 3'Cohtain'meit -Air Locks.-..Each air lock has a

do6 at both' ends..Therdoors are normally interlocked to prevent s'imultaieous o6pening when containment OPERABILITY is

-.d recuired.,Howevert,during periods of unit shutdown when

';.!con taihiet'-'OPERABILITY is not required, the door interlock mechanism ma.

be disabled, allowing both doors of an tall'e'd'arock'to'reain open for extended periods when During ~~~~~dv~anment'nf re estyraite ulasmbiswti

frequent containment ingress and egress is necessary.

.,Duri'g mo'veme'nt of re'ce'ntly. irradiated fuel assemblies within

'containment,'containment.,closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door-must always remain closed.

(continued)

CrystalRiver Unit 3

.: - B 3.9-9 r..

,Amendment.No.

208

Containment Penetrations B 3.9.3 BASES BACKGROUND (continued)

I The requirements. on,,-containment penetration closure ensure

. that a release of fission product radioactivity to. the environment from the containment-will be limited.

The, closure restrictions are sufficient to limit fission -

product radioacti'vity release.from containment due to a fuel handling accident involving handling recently irradiated fuel duri`ng refueling'..

In.

MODE 6,' it 'is'ndcessaryto periodically recirculate/

exchangeRB atmosphere in order to.'minimize radiation uptake

.'during the conduct.. of refueling operations.

The 48 inch purge valves are'normally'used for'this purpose, but the mini-purge valv'es'may.be 'relied 'upon as well.

Both valve types are automatically.isolated: on. a unit vent-high

,. radiation-signal:'(fromiRM-A1).;;So6il'ong as one valve in the flow' path is OPERABLE; these'lJines' may remain unisolated

'during the subject pl'ant'conditions'.

-rment penetratins'that provide direct access fromcontainmentaitm6osphereto outside atmosphere must be-isolated by a minimum of one isolation device.

Isolation may be achieved by'an automatic or manual isolation valve, blind. flange-, or equivalent.

Equivalent isolatiion'meth'ods'iinclude'useofa material (e.g., temporary sealant),that can provide a'temporary,-,atmospheric pressure ventilation. barriey for tihe-other containment penetrations

,durin'g fuel movements handling recently

'irradiated fuel'- se I

'g

,

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-'. -,

,.. I I

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{

(continued)

Crystal River Unit 3

. : -'B 3.9-1.0 C

" Revision No. 43

Containment Penetrations B 3.9.3 BASES

-APPLICABLE.

During movement-of,.,fuel assemblies within containment,..

.,SAFETY, ANALYSES the most-severe,radiological,consequences result from a

-fuel handling.accident involving handling recently

' 'rradiat'ed fuel..For Cycle 13' (including Refueling Outage

13) and future.Cy'cls'(including Refueling Outages) that are, operated at a RATED-THERMAL.POWER'of 2568 mWt, recently irradiated fuel,.is,the fuel,that hasoccupied part of a critical-reactorcore.:within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The fuel handling 'acciden't is'a'postulated'event that involves damage'to'-irradiated 'fuel '(Ref. '-13 Fuel handling accidents

.include dropping a'sin'gle. irradiated fuel assembly and handling tool or a heavy object-onto -other -irradiated fuel assemblies.

The requirements of LC0 3.9.6,."Refueling Canal Water Level,"'jn'conjunction'with 'the administrative limit on minimum decay t'i'me'-p'ri'o'rt'ot irr'adiated'fuel movement ensure

- -@'that'.the'.releaisetfission'prodtict

'radioactivity subsequent to

.; '..:..afuel handling "accident'results: in' doses that are within the requirements specifiediin 10 CFR'50;67 even without

.. ;containment closure'.-

.1,Containment-penetrations.satisfy. Criterion 3 of the NRC

, Policy Statement..,,

LCO

.,'..This LCO limits teth consequences of.a'fuel handling accident

'i ecently-irradiated fuel in containment

" by iimitin'g the'.potehtial '

path's for fission product radioactivity from c'on ainment'.

The 'LCO requires any

penetration 'pr6oidin-' direct access 'from the containment

--atmosphere to the-outside'atm'osphere-,'including-the

equipment hatch or the Outage Equipment Hatch, to be closed except for penetrations containing an'OPERABLE purge or

mini-purge valve.

For the containment,purge and mini-purge valves to be'considered OPERABLE,- at least one valve in each_:'penetrat ion must:be-au to'matically isolable on an RB

Putrge-high radiation isolation -signal.

-The

'definition zof "di rect 'access from.the containment

-: atmosphere to the outside atmosphere", is any path that would

allow for.transport~of containment atmosphere to any

'j-atmosphere-locatedoutside-the containment structure. This

- includes the AuxiliaryBuilding..,As,a general rule, closed or.pressurized systems. do not constitute a direct path (continued)

.. Crystal River Unit 3 I

Cr, stl Ri,

,,Revision No. 43

p.

..,1

r

... A Containment Penetrations B 3.9.3 BASES I

LCO between the RB and-outside environments. All.peermanent'and (continued) temporary-penetration closures'should *be evaluated to.assess

. -the possibility fora' 'release path toithe outside environment.-'For,'the;'purpose of determining what constitutes' a"direct'access" path, no failure mechanisms should be applied to'create a scenario.which results in a

'"direct-access". path.' For exa'mple,'.line breaks, valve failures,.power losses or, natural.phenomenon should not be postulated'as part of the evaluation process.

, APPLICABILITY...Thecontainment penetration requirements are applicable during' movement

'of recently--irradiated fuel assemblies within containment becau se'this.i's when there is a

..poentia forthe ii fuel.'handli'ng accident.

In MODES' 1,:2, -3, an&47' containment penetration requirements are addres'sed by'LCO 3.6.1',."Contaiinment." In MODES 5 and 6, when movement of irradiated fuel assemblies within containment is not being conducted;

.th'e potential for a fuel handling accident does' nrot-exist:` Additionally, due to. radioactive.decay, a.fuel handling accident involving fuel. that has not been recently irradiated will result in

.dosesthat are wel l within theguideline values specified in 10 CFR' 50';67 even

`without containment closure capab~iIity._...,There'f6re, under these conditions no requirements are.,paced: on. contai'nment penetration status.

,1

' 1.

.i ACTIONS A:1'I4 With the containmente-quipmeft'.hatch,I OEH, air locks, or any containment penetration thatprovides direct access from the;containment atmosphere-tothe outside-atmosphere--not in the required status,.including,,containment purge or.mini-purge valve'penetrations not capable of automatic isolation when the penetrations are unisolated,the plant must be placed in.a condi'tion in which the isolation functiop is not needed.

This is accomplished by'immediately suspending movement'of recently irradiated fuel'assemblies within containment.

Performance of these'actions shall not preclude moving a component to a safe position.

I (conti nued)

Crystal River Unit 3 B 3.9-12 Revision No. 43

p -., *-. -

-. ~.

I..

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I:

Containment Penetrations B 3.9.3 BASES SURVEIL REQUIRE LANCE.SR 3-9.3.1...

MENTS..

This Surveillance demonstrates thatreach of the containment penetrations required to be in.its.'closed position is in that position..-

he Surveillanceis performed every 7 days during movement of recently.irradiated fuel-assemblies within the containment.

-. -The Surve'illance interval -is selected.to be commensurate with the normal duration of time to complete fuel handling operations..

As'such, thiss'urveillance'ensures that a postulated fuel

' handling accidentirnvolving handling 'recently irradiated fuel that-releases :fissioh product radioactivity within contai nment' wll'; Ait res'ul tni'a re. ease of significant fission product radioactivity to the' environment.

2~~

..j_.;.

SR 3.9.3 2"' '

This Su'rveillanc'e de6m6nstrates~thit each containment purge

and mini-p~uvrgeyalvie actuates to its isolation position on an actual or "simulated hi'gh'radiatibn;signal.

The 24 month requency is 'onsisgekt

'with other 'similar instrumentation

-' ' and valve test'ing rduirements.' The Surveillance ensures

'tha't 'thevai are cap'able of':closing after a postulated fuel handling accident involving handling recently irradiated fuel to lmit a re ease ot fission product radioactivity from the containment.

SR 3.6.3.5 demonstrates -that the isolation time of each valve is in accordance'w'ith the Inservice

,Testing Program:-requijrements.

l

,¢;l i;.

n'

.J.

REFERENCES' 1

FSAR,; Sect'ion '14.2 32.3.'

3w,

,;w§ j

i i*

I Revision No. 43 Crystal River Unit 3 B 3. 9-13

Refueling Canal Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Canal Water Level BASES BACKGROUND.

-The:move'ment'of irradiated fuel assemblies within

'-,,containment requires:a minimum refueling canal water level of 156 ft plant datum.' This maintains sufficient water level above the fuel contained in the vessel and the bottom of the fuel-transfer canal;-and the spent fuel pool to ensure iodine fission product activity'is retained in the water to a level consistent with the dose analysis of a fuel-handling accident (Ref. 4).

Sufficient-iodine

.activity would :be retained-to limit.offsite doses from the

accident to well.within 10 CFR 50.67 limits (Ref. 3).

APPLICABLE.

During-movement of Irradiated fuel assemblies, the water SAFETY ANALYSES 1level Ain therefueling canal is' an assumed initial condition in the analysis of thelfuel,handling accident in containment. This relates to the assumption that 99% of the total-iodine-released from-the fuel is-retained by the

,.,.,refueling canal water.

There are postulated-drop scenarios where'therels'i< 23'ft-above the top.of'the fuel bundle and a,

,the surface..Inparticular,,this'is the case for the period of. time during which the assembly travels between the cavity and-the deep 'end'of the, refueling canal.

During this time, therejis)$otentially,21 feet of water between thereactor vessel flange-(135 ft plant datum) and the surface of ihe p'ol.'-"The iodine retention factors used in the'dose assessment are'still cons'rvative at water levels of 21 feet above'the' daiaged'fuel, (Ref.' 4). The 156 ft value was chosen to'be consistentwith the level specified for LCO 3.7.13, "Fuel Storage Pool 'Water Level" and plant configuration.

i

)

~(conti nued)

Crystal River Unit 3 B 3.9-23

-Amendment No. 208

-- I.

J

.i%

i t;

Refueling Canal Water Level B 3.9.6 BASES APPLICABLE The fuel handling accident analysis inside containment'is SAFETY ANALYSES described in Reference 4. With a minimum water level of (continued) 23 ft above the stored fuel, and the administrative limit on minimum decay time: of 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 's prior to movement of irradiated fuel in the.vessel, analyses'.demonstrate that the iodine release:'due to' a.postulated fuel handling accident is,

adequately.-captured-by the'water.such that offsite doses

.-are maintained within' allowablelimits. (Ref. 3).

Refueling canal'.water level 'satisfies Criterion 2 of the NRC Policy Statement. -

LCO A minimum-refueling 'canal water level,'of 156 ft plant datum

'is require'dto ensur'e,'that the radiological consequences of a postulated fuel handling accident inside'containment are within acceptable limits. This minimum level also ensures an-adequate operational window between.-the surface of the p;,.,Pool andtthe transfer winch-for theSRB-fuel handling",

equipment....

APPLICABILITY:

,ThisSpifiation;,is'applicabie'when'moving irradiated

'fuel'.,assemblies:.'.withii 6the containment'., The LCO minimizes the' apotentia

'of a,.,fuel', andli""'ga""ident in containment which results 'in offsite doses greater than those calculated.-by the safety analysis. If'irradiated fuel is not.present in containment, there can be no significant radioactivity release.'as a result;of a'postulated fuel handlingaccident. Water level requirements for fuel handling accidents postulated 'to; occur in the spent fuel pool are addressed by LCO 3.7.13, "Fuel Storage Pool Water Level."

ACTIONS A. 1.

With a refueling canal water level of < 156 ft plant datum, all movement of irradiated fuel assemblies shall be suspended immediately to preclude a fuel handling accident from occurring. The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.

(conti nued)

Crystal River Unit 3 B 3.9-24 Amendment No.

208

Iw IWj

,'t

'Refueling Canal Water Level B 3.9.6 BASES

.ACTIONS

-A.2-.'

In addition-to-immediately suspending movement of irradiated fuel,.actions to..

restore refueling canal water

.,..level must be initiated~immediately. The immediate Completion Time is:-based on engineering judgment. When

.increasing refueling canal water level the boron concentration of -the make-up and the effect of this concentration on the minimum specified in the COLR (Ref. LCO 3.9.1) must.be-considered.-

I I

SURVEILLANCE REQUIREMENTS

A SR 3.9.6.1

.Verification of a minimum refueling canal water level of

-156:ft -plant datum-ensures..that thedesign basis for the

postulated fuel handlin'g accident analysis during refueling operations.is met. 'Water-at the required level above the

.top of--the.;reactor.vessel.f~lange.limits the consequences of damaged fuel;'rods that are assumed to.result from a postulated fuel handling accident-inside containment (Ref. 2).

The Frequency.of 24.hours is.based onengineering judgment

.and c'onsidere'd adequate in view of'the large'volumeof wateranid;:tkhe'normal'-procedural'co'ntrols of valve positions,

.:w signifcant unplanned level'changes unlikely.

X !

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REFERENCES;:. -'1. -.Deleted.-,

p s

L r -.

2.

-FSAR-SectionK14 2-.23.

3.

10 CFR 50.67.

4. -FPC-Calculation-N-00-0001.

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Crystal River Unit 3 Amendment No. 208 B 3.9-25

3F0404-03, Technical Specifications Bases Control Program

Contents

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Subject:

Dear Sir:

2.4 BACKGROUND

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3F0404-03, Technical Specifications Bases Control Program

Text

Subject:

Dear Sir:

2.4 BACKGROUND

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