ML20136A525
| ML20136A525 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/20/1985 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-5-153, TAC-62942, NUDOCS 8601020050 | |
| Download: ML20136A525 (539) | |
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,"* 3o DeYe$br20, 85 United States Nuclear Regulatory Cannission Washington, DC 20555 ATTENTION: Mr. Harold R. Denton Office of Nuclear Reactor Regulation
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Technical Specification Pre-Issue Revision No.1
REFERENCES:
(a) NRC letter fran Mr. G. Knighton to Mr. J. Carey, dated September 18, 1984 (b) DLC letter 2NRC-4-144, dated Septenber 14, 1984 Gentlemen: Duquesne Light Conpany (DLC), acting on its own behalf and as agent for The Cleveland Electric Illuminating Canpany, Ohio Edison Canpany, and The Toledo Edison Canpany, is filing herewith three (3) signed originals plus six (6) copies of Pre-Issue Revision No.1 to the proposed Technical Specifica-tions for the Beaver Valley Power Station, Unit No. 2 (BVPS-2) in accordance with 10CFR50.36 and Reference (a). This revision to the set originally submitted per Reference (b) is
-needed to incorporate recent plant design changes, to provide the necessary values and criteria for previously unfilled setpoints and specifications, to closer model the BVPS-2 Technical Specifications to be like the BVPS-1 Tech-nical. Specifications (as per Reference [a]), to correct typographical errors, and to incorporate changes resulting fran the resolution ~ of BVPS-2 Final Safety Analysis Report (FSAR) review conments fran NRR. These changes to the proposed BVPS-2 Technical Specifications have been derived fran the analyses and evaluations included in the BVPS-2 FSAR, and anendments thereto, submit-ted pursuant to 10CFR50.34.
Please insert the Pre-Issue Revision No.1 into the BVPS-2 Technical Specifications binders (provided by Reference [b]) as per the Revision Instruction Sheet. In Reference (a), the NRC indicated the staff review of the BVPS-2 Technical Specifications would conmence approximately 12-18 months prior to our expected date for issuance of our operating license. Since only 16 months exist before our expected loading of fuel at BVPS-2 and the SER for the BVPS-2 FSAR has been - issued, DLC requests that the review of the proposed BVPS-2 Technicai Specifications be conmenced as soon as possible to facili-tate the non-standard review which will be used on the BVPS-2 Technical Specifications. A longer review and conment resolution time is expected since the BVPS-1 Technical Specifications are going to be used for the first AD - J. Knight (ltr only) B601020050 8D1220 M/f ye'Nsal s i ! PDR ADOCK 00000412 esa (cAnn m.) l PDR ass (stuunctn> A 3 FOB (BENAROVA) l' 1
United States Nuclear Regulatory Commission Mr. Harold R. Denton Technical Specifications Pre-Issue Revision No.1 Page 2 i time as the model (in accordance with Chapter 16.0 of the SER) for Technical Specifications developnent. DLC believes that if the Technical Specifications review is not started within the next two months, time for an adequate review ' and resolution of conments oy both the staff and DLC would not be available. OLC stands ready to support the NRC as needed in order to get the BVPS-2 Technical Specifications review started at the earliest possible monent. If you have any questions on this issue, please do not hesitate to contact Mr. R. W. Fedin at (412) 923-1960. DUQUESNE LIGHT COMPANY l 1 By , M QJ Carey Vice. President RWF/wj s Attachment i cc: Mr. L. S. Rubenstein (w/o) Mr. B. K. Singh, Project Manager (w/o) Mr. G. Walton, NRC Resident Inspector (w/o) COMMONWEALTH OF PENNSYLVANIA )
) SS:
COUNTY OF ) On this M /) day of / h ey/v , /((6 .before me, a Notary Public'in ano for said C6nmonwealth and County, personally appeared J. J. Carey, and said that (1) he is Vice tao beingLight, President of Duquesne duly (sworn,
- 2) he is deposed duly authorized to execute and file
, the foregoing Submittal on behalf of said Company, and '(3) the statenents set l forth in the Submittal are true and correct to the best of his knowledge. LL/'t & - L b7] Notary Public shelf A U. FATIORE. I'01/Rf FUSilC SHf UlwCRT ECM, EUlin CC'J'jiY UY CORilWON Eli:TdS CLI 23. W3
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INSERTION INSTRUCTIONS FOR TECHNICAL SPECIFICATION PRE-ISSUE REVISION NO. 1 NOTE: Since the revision is very extensive, a conplete copy of the Beaver Valley Unit No. 2 proposed Technical Specifications is enclosed. INSTRUCTIONS: Renove all old pages and instet the enclosed Technical Speci-fication pages behind the appropriate tabs. This transnittal letter, along with this insertion instruc-tion, should either be filed or entered in front of Volume 1. e
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O BEAVER VALLEY POWER STATION UNIT 2 TECHNICAL SPECIFICATIONS APPENDIX "A" O LICENSE NO. l I
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PRE-ISSUE REVISION No. O ': Pages September 14,1984 No.1 See List of Ef f ective Pages December 20,1985 i
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES O Page, Table, or Figure Pre-Issue Revision Number Title Page O Pre-Issue Revislon Table 1 Effective Page Listing I Index Title Page 0 i 0 ii 0 iii 0 iv 0 v 1 vi 1 vii 1 viii 0 ix 0 x 0 xi - 0 xii 0 xiii 0 xiv 0 xv 0 xvi 0 xvii 0 xviii 1 Section 1 Title Page 0 O 1-1 1 1-2 1 1-3 1 1-4 0 1-5 1 1-6 1 1-7 0 18 0 1-9 0 1-10 0 Section 2 'Jitle Page 0 2-1 0 2-2 0 2-3 0 2-4 0 2-5 0 2-6 0 2-7 1 2-8 1 2-9 0 2-10 0 0 PicE 1 Or 12
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) O Page, Table, or Figure Pre-Issue Revision Number 2-11 0 2-12 1 2-13 0 2-14 0 B2-1 1 B2-2 0 B2-3 1 B2-4 1 B2-5 0 B2-6 0 B2-7 0 B2-8 0 B2-9 0 B2-10 0 Section 3 & 4 Title Page 0 3/4 0-1 1 3/4 0-2 0 3/4 0-3 0 3/4'l-1 0 3/4 1-2 0 3/4 1-3 0 3/4 1-4 0 3/4 1-5 0 1 0 3/4 1-6 3/4 1-7 0 0 3/4 1-8 0 3/4 1-9 0 3/4 1-10 0 3/4 1-11 0 3/4 1-12 0 3/4 1-13 0 3/4 1-14 0 3/4 1-15 1
-3/4 1-16 1 i 3/4 1-17 0 3/4 1-18 0 3/4 1-19 0 3/4 1-20 1 l 3/4 1-21 0 3/4 1-22 0 3/4 1-23 0 3/4 1-24 1 3/4 1-25 0 l 3/4 1-26 0 '
L O e4ct 2 08 12 l
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, T able, or Fi F ure Pre-Issue Revision Number l 3/4 2-1 1 3/4 2-2 1 3/4 2-3 1 3/4 2-4 0 3/4 2-5 1 3/4 2-6 1 3/4 2-7 0 3/4 2-8 0 3/4 2-9 0 3/4 2-10 0 3/4 2-11 0 3/4 2-12 0 3/4 2-13 0 3/4 2-14 1 3/4 2-15 0 3/4 3-1 1 3/4 3-2 1 3/4 3-3 0 3/4 3-4 1 3/4 3-5 1 3/4 3-6 0 3/4 3-7 1 O 3 /4 ' 3-8 3/4 3-9 0 1 3/4 3-10 0 3/4 3-11 1 3/4 3-12 0 3/4 3-13 1 3/4 3-14 1 3/4 3-15 1 3/4 3-16 0 3/4 3-17 1 3/4 3-18 1 3/4 3-19 1 3/4 3-20 1 3/4 3-21 1 3/4 3-22 1 3/4 3-23 1 3/4 3-24 1 3/4 3-25 1 3/4 3-26 0 3/4 3-27 1 3/4 3-28 1 3/4 3-29 1 PAGE 3 OF 12 i
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BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue Revision Number 3/4 3-30 1 3/4 3-31 1 3/4 3-32 1 3/4 3-33 1 3/4 3-34 1 3/4 3-35 1 3/4 3-36 1 3/4 3-37 0 3/4 3-38 1-3/4 3-39 1 3/4 3-40 0 3/4 3-41 1 3/4 3-42 1 3/4 3-43 1 3/4 3-44 1 3/4 3-45 1 3/4 3-46 1 3/4 3-47 1 3/4 3-48 ' 1 3/4 3-49 1 3/4 3-50 1 3/4 3-51 1 3/4 3-52 1 3/4 3-93 1 3/4 3-54 1 3/4 3-55 1 3/4 3-56 1
'3/4 3-57 1 3/4 3-58 1 3/4 3-59 0 3/4 3-60 '
0 , 3/4 3-61 0 3/4 3-62 0 3/4 3-63 0 3: 3-64 1 3/4 3-65 l' 3/4 3-66 1 3/4 3-67 1 3/4 3-68 0 3/4 3-69 1 3/4 3-70 1 3/4 3-71 0 3/4 3-72 1
- '3/4 3-73 1 O PicE 4 or 12
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BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES' (Cont.) O Page, Table, or Figure Pre-Issue Revision Number
-3/4 3-74 1 3/4 4-1 0 3/4 4-2 1 3/4 4-3 0 3/4 4-4 1 3/4 4-5 0 3/4 4-6 1 3/4 4-7 0 3/4 4-8 0 3/4 4-9 1 3/4 4-10 1 3/4 4-11 0 4
3/4 4-12 0 3/4 4-13 0 3/4 4-14 0 3/4 4-15 0 3/4 4-16 0 3/4 4-17 0 3/4 4-18 0 3/4 4-19 0 3/4 4-20 0 3/4 4-21 0 3/4 4-22 0 0 3/4 4-23 1 3/4 4-24 1 3/4 4-25 1 3/4 4-26 1 3/4 4-27 1 3/4 4-28 1 3/4 4-29 1 3/4 4-30 1 3/4 4-31 1 3/4 4-32 0 3/4 4-33 0 3/4 4-34 0 3/4 4-35 0 3/4 4-36 0 3/4 4-37 1 3/4 4-38 1 3/4 4-39 1 3/4 4-40 1 3/4 5-l' O 3/4 5-2 0 3/4 5-3 1 PAGE 5 OF 12
BVPS-2 Technical Specifications LISr OF EFFECTIVE PAGES (Cont.) O Page. Table, or Figure Pre-Issue Revision Number 3/4 5-4 1 3/4 5-5 1
~ 3/4 5-6 1 3/4 5-7 0 3/4 5-8 0 3/4 6-1 0 3/4 6-2 1 3/4 6-3' 1 3/4 6-4 1 3/4 6-5 0 3/4 6-6 0 3/4 6-7 0 3/4 6-8 0 3/46-9 0 3/4 6-10 , 0 3/4 6-11 0 3/4 6-12 1 3/4 6-13 1 3/4 6-14 1 3/4 6-15 0 3/4 6-16 1 3/4 6-17 1 3/4 6-18 1 0 3/4 6-19 3/4 6-20 1
1 3/4 6-21 1 3/4 6-22 1 3/4 6-23 1 3/4 6-24 1 3/4 6-25 1 3/4 6-26 1 3/4 6-27 1 3/4 6-28 1 3/4 6-29 1 3/4 6-30 0 3/4 6-31 1 3/4 6-32 1 3/4 6-33 0 3/4 6-34 1 3/4 6-35 1 3/4 7-1 0 3/4 7-2 0. 3/4 7-3 0 3/4 7-4 1 O >>cz 6 or 12
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue _ Revis ion Number
~3/4 7-5 0 3/4 7-6 1 3/47-7 0 3/4 7-8 1 3/4 7-9 0 3/4 7-10 0 3/4 7-11 0 3/4 7-12 1 3/4 7-13 1 3/4 7-14 1 3/4 7-15 1 3/4 7-16 1 3/4 7-17 1 3/4 7-18 1 3/4 7-19 1 3/4 7-20 1 3
3/4 7-21 1 3/4 7-22 1 3/4 7-23 0 3/4 7-24 0 3/4 7-25 1 3/4 7-26 0 0 3/4 7-27 3/4 7-28 1 0 3/4 7-29 0 3/4 7-30 1 3/4 7-31 1 3/4 7-32 1 3/4 7-33 1 3/4 7-34 1 3/4 7-35 1 3/4 7-36 1 3/4 7-37 0 3/4 7-38 1 3/4 7-39 0 3/4 7-40 0 3/4 7-41 0 3/4 7 1 3/4 7-43 1 3/4 7-44 0 3/4 7-45 1 3/4 8-1' 1 3/4 8-2 1 3/4 8-3 0 PAGE 7 0F 12
BVPS-2 Technical Specifications i LIST OF EF7ECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue Revision Number 3/4 8-4 1 3/4 8-5 0 3/4 8-6 0 3/4 8-7 1
-3/4 8-8 0 3/4 8-9 0 3/4 8-10 1 3/4 9-1' 1 3/4 9-2 0 3/4 9-3 0 3/4 9-4 0 3/4 9-5 1 l
3/4 9-6 0 3/4 9-7 1 3/4 9-8 0 3/4 9-9 0 3/4 9-10 .0 3/4 9-11 0 3/4 9-12 1 3/4 9-13 1 3/4 9-14 0 3/4 9-15 0 f 3/4 10-1 3/4 10-2 1 1 . 3/4 10-3 1 j 3/4 10 -4 1
- 3/4 10-5 1 3/4 10 -6 0
, 3/4 11-1 1 3/4 11-2 1 3/4 11-3 0 3/4 11-4 1 3/4 11-5 1 3/4.11-6 1 3/4 11-7 0 3/4 11-8 1 3/4 11-9 1 3/4 11-10 1 3/4 11-11 1 3/4 11-12 1 3/4 11-13 0 . 3/4 11-14 1 3/4 11-15 0 3/4 11-16 0 ! PAGE 8 0F 12 l 1
. _ . _ _ . _ _ . __ _. -, - . . _ . . . . . , . _ .. _,. _ , y __ , , , , , _ . . _ . _ , _ , . . . _ _ .
I BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue Revision Number 3/4 11-17 0 3/4-11-18 1 3/4 11-19 0 3/4.11-20 1 3/4 11-21' 1 3/4 12-1 1 3/4 12-2 1 3/4 12-3 0 3/4 12-4 0 3/4 12-5 0 3/4 12-6 0 3/4 12-7 0 3/4 12-8 0 3/4 12-9 0 3/4 12-10 1 3/4 12-11 1 3/4 12-12 1 B 3/4 0-1 0 B 3/4 0-2 0 B 3/4 0-3 0 B 3/4 0-4 0 B 3/4 1-1 1 O V B 3/4 B 3/4 1-2 1-3 1 1 B 3/4 1-4 1 B 3/4 1-5 0 B 3/4 1-6 0 B 3/4 1-7 1 B 3/4 2-1 0 B 3/4 2-2 1
-B 3/4 2-3 1 B 3/4 2-4 0 B 3/4 2-5 0 B 3/4 2-6 0 B 3/4 2-7 0 B 3/4 3-1 0
- B 3/4 3-2 0 B 3/4 3-3 1 B 3/4 3-4 1 B 3/4 3-4 1 B 3/4 3-5 1 B 3/4 3-6 1 B 3/4 3-7 1 B 3/4 3-8 1 PAGE 9 OF 12
T" _ BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue Revision Number B 3/4 3-9 1 B 3/4 3-10 1 B 3/4 3-11 1 B 3/4 3-12 1 B 3/4 4-1 1 B 3/4 4-2 0 B 3/4 4-3 1 B 3/4 4-4 0 B 3/4 4-5 0 B 3/4 4-6 1 B 3/4 4-7 0 4 B 3/4 4-8 0 B 3/4 4-9 0 B 3/4 4-10 0 j B 3/4 4-11 0 i B 3/4 4-12 0 B 3/4 4-13 0 B 3/4 4-14 1 4 B 3/4 4-15 0 B 3/4 4-16 1
-B 3/4 4-17 0 B 3/4 4-18 0 O = 3/' '-19 B 3/4 4-20 o
0 B 3/4 4-21 0 B 3/4 5-1 0 B 3/4 5-2 0 B 3/4 6-1 'l B 3/4 6-2 1 B 3/4 6-3 1 B 3/4 6-4 1 B 3/4 6-5 1 B 3/4 6-6 1 B 3/4 6-7 1 B 3/4 6-8 0 B 3/4 6-9 0 B 3/4 6-10 1 B 3/4 7-1 1 B 3/4 7-2 1 B 3/4 7-3 1 B 3/4 7-4 1 B 3/4 7-5 1 B 3/4 7-6 1 B 3/4 7-7 0 4 PAGE 10 0F 12
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) O Page, Table, or Figure Pre-Issue Revision Number B 3/4 7-8 0 B 3/4 7-9 0 B 3/4 7-10 0 B 3/4 7-11 0 B 3/4 7-12 0 B 3/4 7-13 1 B 3/4 7-14 0 B 3/4 7-15 0 B 3/4 7-16 0 B 3/4 7-17 0 B 3/4 7-18 1 B 3/4 7-19 0 B 3/4~ 7-20 0 B 3/4 8-1 0 B 3/4 8-2 - 0 B 3/4 9-1 0 B 3/4 9-2 0 B 3/4 9-3 0 B 3/4 9-4 0 B 3/4 9-5 0 1 B 3/4 9-6 0 B 3/4 9-7 0 O B 3/4 9-8 0 V B 3/4 9-9 0 B 3/4 9-10 0 B 3/4 9-11 0 B 3/4 9-12 0 B 3/4 10-1 0 B 3/4 10-2 0 B 3/4 .10-3 0 B 3/4 10-4 0 B 3/4- 10-5 0 i B 3/4 10-6 0 B 3/4 11-1 1 B 3/4 11-2 1 B 3/4 11-3 1 B 3/4 11-4 1 B 3/4 11-5 1 B 3/4 11-6 i B 3/4 11-7 1 B 3/4 11-8 1 B 3/4 11-9 1 B 3/4 11-10 1 B 3/4 11-11 0 PAGE 11 0F 12 t L_
BVPS-2 Technical Specifications LIST OF EFFECTIVE PAGES (Cont.) Page, Table, or Figure Pre-Issue Revision Number I B 3/4 11-12 0 B 3/4 12-1 0 B 3/4 12-2 0 B 3/4 12-3 0 5-1 0 5-2 0 1 5-3 0 5-4 0 5-5 0 6 0 5-7 0 5-8 0 6-1 1 6-2 0 6-3 0 6-4 0 6-5 0 6-6 0 ) 6-7 1 6-8 0 6-9 0 6-10 0 O 6-11 6-12 1 0 6-13 1 6-14 0 6-15 1
; 6-16 1 6-17 0
- 6-18 1
!' 6-19 1 6-20 1 6-21 1 i 6-22 1 6-23 1
- 6-24 1 6-25 1 6-26 1 6-27 1 6-28 1 i
O PAGE 12 0F 12 f
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4 1 O INDEX l l O I i i i BEAVER VALLEY POWER STATION i l i O UNIT NO. 2 l t I
INDEX DEPIN IT ION S SECT ION PAGE 1.0 DEFIN IT ION S Defined Terms . . ............ . . . . . . . . . . . 1-1 Thermal Power . . ........................ 1-1 Rated Thermal Power . ...................... 1-1 Operational Mode. ........................ 1-1 A c t io n. . . . . . ........................ 1-1 Operable - Operability. ..................... 1-2 Reportable Event ........................ 1-2 Cont ainment Integr ity . ..................... 1-2 Channel Calibration . ...................... 1-3 Channel Check . ......................... 1-3 (} Channel Functional Test ..................... 1-3 Core Alteration . ........................ 1-3 Shutdown Margin . ........................ 1-3 Identified Leakage. ....................... 1-4 Unidentified Leakage. ...................... 1-4 Pressure Boundary Leakage . ................... 1-4 Controlled Leakage. ....................... 1-4 Quadrant Power Tilt Ratio . ................... 1-5 Dose Equivalent I-131 . ..................... 1-5 Staggered Test Bas is. ...................... 1-5 Frequency Notation. ....................... 1-5 Reactor Trip Response Time. ................... 1-5 Engineered Safety Feature Response Time . . . . . . . . . . . . . 1-6 BEAVER VALLEY UNIT 2 i
IND EX DEFIN IT ION S SECTION PAGE 1 1.0 D EFIN ITION S Axial Flux Dif ference . . . . . . . ............... 1-6 Physics Test. . . . . . . . . . . .. .............. 1-6 E-Average Disintegr ation Energy . . ............... 1-6 Source Check. . . . . . . . . . . . ............... 1-6 Process Control Progr aw . . - . . . . . ............. 1-6 Solidific at io n. . . . . . . . . . . ............... 1-7 Of f-Site Dose Calculation Manual (ODCM) ............. 1-7 Gaseous Radwaste Treatment System . ..,............ 1-7 Vent ilation Exhaust Treatment System. .............. 1-7 Purge - Purging . . . . . . . . . . ....,.......... 1-7 Vent ing . . . . . . . . . . . . . . ..........;.... 1-8
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Major Changes . . . . . . . . . . . ............... 1-8 Members (s) of the Public. - . . . . . ............... 1-8 l Operational Modes (Table 1.1) . . . ............... 1-9 Frequency Notation. . . . . . . . . ............... 1-10 f I J , r BEAVER VALLEY UNIT 2 il 1 4
INDEX SAFETY LIMIT S AND LIMITING SAFETY SYSTDi SETTINGS SECT, ION PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i 2.-l SAFETY LIMIT S
< REACTOR CORE. . . . . . . . . . . . . . . . . . . . . . . . . 2-1 REACTOR COOLANT SYST EM PRESSIRE . . . . . . . . . . . . . . . 2-4 2.2 LIMITING SAFETY SYSTEM SETTINGS t REACTOR TRIP SYST EM INSTRUMl!NTATION SETPOINT S . . . . . . .. 2-5 l
i i O - SECTION 2.1 SAFETY LIMIT S REACTOR CORE. . . . . . . . . . . . . . . . . . . . . . . . . B2-1 REACTOR C001 ANT SYST EM PRESSJRE . . . . . . . . . . . . . . . B2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS ) REACTOR TRIP SETPOINT S. . . . . . . . . . . . . . . . . . . . B2-4
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l 1 O ' 4 BEAVER VALLEY UNIT 2 iii
INDEX LIMITING CONDITIONS FOR OPERATION AND SJRVEILLANCE REOUIREMDiT S SECTION 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . . . . . 3/4 0-1 3/4.1 REACTIVITY C0ffrROL SYSTEMS 3/4.1.1 BORATION C0ffrROL 200*F . . . . . . . . . . . . . . . 3/4 1-1 Shutdown Margin - T avg >< 200*F . . 3/4 1-3 Shutdown Margin - T avg . . . . . . .. . . . . . Boron Dilution. . . . . . . . . . . . . . . . . . . . . . . 3/4 1-4 Moder ator Temper atur e Coef ficient . . . . . . . .. . . . . 3/41-5 Minimum Temperature for criticality . . . . . . . . . . . . 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown . . . . . . . . .. . . . . . . . . 3/4 1-7 Flow Paths - Oper ating. . . . . . . . . . . . .. . . . . . 3/4 1-8 Charging Pump - Shutdown. . . . . . . . . . . . . . . . . . 3/4 1-10 Charging Pumps - Operating. . . . . . . . . . . . . . . . . 3/4 1-11 Boric Acid Trans fer Pumps - Shutdown. . . . . . . . . . . . 3/4 1-12 O neric ^ cia Tr n r P=mP - oP et . . . . . . . . . . . 3/4 i-13 Borated Water Sour ces - Shutdown. . . . . . . . . . . . . . 3/4 1-14 Borated Water Sources - Operating . . . . . . . . . . . . . 3/4 1-15 3/4.1.3 MOVABLE CONTROL ASSEMBLIES , Group Height. . . . . . . .. . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Operating . . . . . . . . . . 3/4 1-20 Position Indication System - Shutdown . . . . . . . . .. . 3/4 1-21 Rod Drop Time . . . . . . . . . . . . . . . . . . . .. . . 3/4 1-22 Shutdown Rod Insertion Limit. . . . . . . . . . . . . . . . 3/4 1-23 Control Rod Inser tion Limits. . . . . . .. . . . . . . . . 3/4 1-24 1 4 O iv BEAVER VALLEY UNIT 2 l 1
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IND EX l R U LIMITING CONDITIONS FOR OPERAT ION AND TRVEILIANCE REQUIREMENT S SECTION PAG E 3/4.2 POWER DISTRIBUTION LIMIT S I 3/4.2.1 AXIAL FLUX DIFFERENCE . . . . . . . . . . . . . . . . . . . 3/4 2-1 3/4.2.2 HEAT FLUX HUT CHANNEL FACTOR. . . . . . . . . . . . . . . . 3/42-5 t 3/4.2.3 NUCLEAR BTTRALPY HUT CHANNEL FACTOR . . . . . . . . . . . . 3/4 2-9 3/4.2.4 QUADRANT POWER TILI RATIO . . . . . . . . . . . . . . . . . 3/4 2-11 3/4.2.5 DNB PARAMET ERS. . . . . . . . . . . . . . . . . . . . . . . 3/4 2-14 ' 3/4.3 IN STRUMENTATION L 3/4.3.1 PR7T ECTIVE INSTRUMENTATION. . . . . . . . . . . . . . . . . 3/4 3-1 l 3/4.3.2 ENGINEERED SAFETY FENTURE INSTRUMENTATION . . . . . . . . . 3/4 3-13 l 3/4.3.3 MONITORING INSTRUMENTATION ! Radiation Monitoring. . . . . . . . . . .. . . . . . . . . . 3/4 3-37 [ Movable Incore Detectors. . . . . . . . . . . . . . . . . . 3/4 3-43 i Seismic Instrumentation . . . . . . . . . . . . . . . . . . 3/4 3-44 }r Meteorologicat Ins trume nt at ion. . . . . . . . . . . . . . . 3/4 3-48 Remote Shutdown Instrumentation . . . . . . . . . . . . . . 3/4 3-51 ; () Fire Detection Instrumentat' ion. Chlorine Detection Systems. . . . 3/4 3-54 3/4 3-59 i Accident Monitoring Instrumentation . . . . . . . . . . . . 3/4 3-60 Radioactive Liquid .Ef fluent Monitoring Instrument ation. . . 3/4 3-63 Radioactive Gaseous Ef fluent Monitoring Instrumentation . . 3/4 3-68 r r 3/4.4 REACTOR COOLANT SYSTEM r 3/4.4.1 REACTOR COOLANT LOOPS Nc mal Operation. . . . . . . . . . . . . . . . . . . . . . 3/4 4'I l Hot Standby . . . . . . . . . . . . . . . . . . . . . . . . 3/44-4 ' Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5 - Isolated Loop . . . . . . . . . . . . . . . . . . ... . . . 3/44-7 , Isolated Loop Startup . . . . . . . . . . . . . . . . . . . 3/4 4-8 , 3/4.4.2 SAFETY VALVES - SHUTDOWN. . . . . . . . . . . . . . . ' . . . 3/4 4-9 I 3/4.4.3 SAFETY VALVES - OPERATING . . . . . . . . . . . . . . . . . 3/4 4-10 i 3/4.4.4 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-11 r 3/4.4.5 ST EAM G ENERATORS. . . . . . . . . . . . . . . . . . . . . . 3/4 4-12 E i i BEAVER VALLEY UNIT 2 v L b
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LIMITING CONDITIONS FOR OPERATION AND SJRVEILIANCE REQUIREMDIT S
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SECT ION PAG E 3/4.4 REACTOR COOLANT SYSTEM (Continued) 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems . . . . . . . . . . . . . . . . 3/4 4-20 Operational Leakage . . . . . . . . . . . . . . . . . . . . 3/4 4-21 RCS Pressure Isolation Valves . . . . . . . . . . . . . . . 3/4 4-23 l-3/4.4.7 CHEMIST RY . . . . . . . . . . . . . . . . . . .. . . .. . 3/4 4-26 3/4.4.8 SPECIFIC ACTIVrrY . . . . . . . . . . . . . . . . . . . . . 3/4 4-29 3/4.4.9 PRESSURE /T EMPERATURE LIMIT S Reactor Coolant System. . . . . . . . . . . . . . . . . . . 3/4 4-33 Pressurizer . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-38 l 1 3/4.4.10 STRUCTURAL INT EGRITY ASME Code . Class 1, 2 and 3 Com, onents . . . . . . . . . . . 3/4 4-39 l 3/4.4.11 RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-40 l 3/4.5 EMERGDiCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATOR S. . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5 2 ECCS SUBSYST EMS - T ayg > 350*F . . . . . . . . . . . . . . 3/4 5-3 3/4.5.3 ECCS SUBSYST EMS - aT yg T 350*F , . . . . . . . . . . . . . 3/4 5-6 ' (_/ BEAVER VALLEY UN TI 2 vi 4
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IND EX ( LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMDiT S SECT ION PAG E 3/4.6 CONTAINMDIT SYSTEMS 3/4.6.1 PRIMARY C0KTAINMDiT Cont ainme nt Integrity . . . . . . . . . . . . . . . . . . . 3/4 6-1 Containment Leekage . . . . . . . . . . . . . . . . . . . . 3/46-2 Containment Air Locks . . .. . . . . . , . . . . . . . . . . 3/4 6-4 Internal Pressure . .. . . . . . . . . . . . . . . . . . . 3/4 6-6 Air Temperature . . .. . . . . . . . . . . . . . . . . . 3/4'6-8 ' Containment Structural Integrity. . . . . . . . . . . . . . 3/46-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System . . . . . . . . . . . . . . 3/4 6-10 Containment Recirculation Spray System. . . . . . . . . . . 3/4 6-12 Chemical Addition System. . . . . . . . . . . . . . . . . . 3/4 6-14 3/4.6.3 CONTAINMETr ISOLATION VALVES. . . . . . . . . . . . . . . . 3/4 6-16 3/4.6.4 COMBUSTIBLE GAS COWrROL Hydrogen Analyzers. . . . . . . . . . . . . . . . . . . . . 3/4 6-32
)
Electric Hydrogen Recombiners . . . . . . . . . . . . . . . 3/4 6-33 Hydrogen Purge System . . . . . . . . . . . . . . . . . . . . 3/4 6-34 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYST Dt Steam Jet Air Ejector 3/4'6-35
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. . . . . . . . . . . . . . . . . . . l3l 3 /4'. 7 PLANT SY ST EMS 3/4.7.1 TURBINE CYCLE Safety Valves . . .. . . . . . . . . . . . . . . . . . . . 3/4 7-1 Auxiliary Feedwater Pumps . . . . . . . . . . . . . . . . . 3/47-5 Primary Plant Demineralized Water . . . . . . . . . . . . . 3/4 7-8 Act iv ity. . . . . . . . . . . . . . . . . . . . . . . . . . 3/47-9 Main Steam Line Isolation Valves. . . . . . . . . . . . . . 3/4 7-11 t
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IND EX LIMITING CONDITIONS FOR OPERATION AND SJRVEILIANCE REQUI(EMMTS SECT ION PAGE 3/4.7 PIANT SYSTEMS (Continued) 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION . . . . . . 3/4 7-12 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYST EM. . . . . . . . . . . 3/4 7-13 3/4.7.4 SERVICE WAT ER SY ST EM. . . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.5 ULTIMKT E H EAT SINK. . . . .. . . . . . . . . . . . . . . . 3/4 7-15 3/4.7.6 FLOOD PROT ECT ION . . .. . . . . . . . . . . . . . . . . . 3/4 7-16 3/4.7.7 C0trTROL BOOM EMERGENCY HABITABILITY SYSTEMS . . . . . . . . 3/4 7-17 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYST EM . . . . . . 3/4 7-19 3/4.7.9 SEAL SOURCE CONTAMINATION . . . . . . . . . . . . . . . . . 3/4 7-21 3/4.7.12 SNUBBERS . . . . . . . . .. . . . . . . . . . . . . . . . 3/4 7-25 3/4.7.13 ST ANDBY SERVICE WAT ER SYST EM. . . . . . . . . . . . . . . . 3/4 7-30 3/4.7.14 FIRE SUPPRESSION SYST EMS Fire Suppression Water Systems. . . . . . . . . . . . . . . 3/4 7-31 Spr ay Systems . .. . . . . . . . . . . . . . . . . .
. 3/4 7-34 Low Pressure CO2 System . . . . . . . . . . . . . . . . . 3/4 7-36 Fire Hose Stations. . . . .. . . . . . . . . . . . . . . . 3/4 7-38 Halon Syotees . . . . . . .. . . . . . . . . . . . . . . . 3/4 7-42
() 3/4.7.15 FIRE RKT ED ASSEMBLIES . . . . . . . . . . . . . . . . . . . 3/4 7-43 3/4.7.16 T ERRESTRIAL ECOLOGICAL SURVEY . . . . . . . . . . . . . . . 3/4 7-45 3/4.8 ELECTRICAL POWER SYSTEMS 3 3/4.8.1 A.C. SOURCES Operating . . . . . . . . .. . . . . . . . . . . . . . . . 3/4 8-1 Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-4 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating . . . . . . . . . . . . . . . 3/4'8-5
. 3/4 8-6 A.C. Distr ibution - Shutdown. . . . . . . . . . . . . . .
D.C. Distr ibution - Operating . . . . . . . . . . . . . . . 3/4 8-7 D.C. Distr ibution - Shutdown. . . . . . . . . . . . . . . . 3/4 8-10 i i i 1 4 BEAVER VALLEY UNIT 2 viii
IND EX LIMITING CONDITIONS FOR OPERATION AND SJRVEILLANCE REQUIRENENT S SECT ION PAGE 3/4.9 REFUELING OPERATION S 3/4.9.1 BORON CONCENTRATION . . . . . . . . . . . . . . . . . . . . 3/4 9-1 3/4.9.2 INST RUMENT ATION . . . . . . . . . . . . . . .. . . . . . . 3/4 9-2 3/4.9.3 DECAY TIME. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRKTIONS . . . . . . . . . . . . . 3/4 9-4 3/4.9.5 - COMMUNICATION S. . . . . . . . . . . . . . . . . . . . . . . 3/4 9-6 3/4.9.6 MANIPULATOR CRANE OPERABILITY , . . . . . . . . . . . . . . 3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING . . . . . . 3/4 9-8 3/4.9.8 CO0lANT CIRCULATION Residual Heat Removal and Coolant Circulation . . . . . . . 3/4 9-9 Low Water Leve l . . . . . . . . . . . . . . . . . . . . . . 3/4 9-10 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOIATION SYSTEM. . . . . . . 3/4 9-11 3/4.9.10 WATER LEVEL-REACTOR VESSEL. . . . . . . . . . . . . . . . . 3/4 9-12 3/4.9.11 WKTER LEVEL-STORAGE POOL . . . . . . . . . . . . . . . . . 3/4 9-13 3/4.9.12 FUEL BUILDING VENTILATION SYST EM - FUEL MOVEMENT. . . . . . 3/4 9-14 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE . . . . . . 3/4 9-15 O 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SRUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . 3/4 10 -1 3/4.10.2 GROUP HEIGKT , INSERTION, AND POWER DISTRIBUTION LIMIT S . . 3/4 10-2 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS - REACTOR CRITICALITY. . . 3/4 10 -3 3/4.10.4 PHYSICS T EST . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 NO FLOW T EST S . . . . . . . . .. . . . . . . . . . . . . . 3/4 10-5 3/4.10.6 POSTTION INDICATION SYST EM-SHUTDOWN . . . . . . . . . . . . 3/4 10-6 3/4.11 RADIOACTIVE EFFLUENT S 3/4 I1.1 LIQUID EFFLUENT S Concentr ation . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-1 Dose. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-7 Liquid Waste Treatment. . . . _ . . . . . . . . . . . . . . . 3/4 11-9 Liquid Holdup Tanks . . . . . . . . . . . . . . . . . . . . 3/4 11-10 O BEAVER VALLEY UNIT 2 ix
INDEX
. LIMITING CONDITIONS FOR OPERATION AND SJRVEILLANCE REQUIREMENT S SECTION PAG E 3/4.11 RADIOACTIVE EFFLUENT S (Continued) 3/4.11.2 GASEQUS EFFIDENT S Dose Rate . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-11 Dose - Noble Gases. . . . . . . . . . . . . . . . . . . . . 3/4 11-15 Dose - Ralioiodines , Particulates , and Radionuclides Other than Noble Gases. . . . . . . . . . . . . . . . . . 3/4 11-16 i Gaseous Radwaste Treatment . . . . . . . . . . . . . . . . . 3/4 11-17 Gas Waste Stor age Tanks . . . . . . . . . . . . . . . . . . 3/4 11-18 Explosive Gas Mixture . . . . . . . . . . . . . . . . . . . 3/4 11-19 3/4.11.3 SOLID RADIOACTIVE WASTE . . . . . . . . . . . . . . . . . . 3/4 11-20 3/4.11.4 TOTAL DOSE. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-21 3/4.12 RADIOIDGICAL DIVIRONMl!NTAL MONITORING 3/4.12.1 MONITORING PROGRAM. . . . . . . . . . . . . . . . . . . . . 3/4 12-1 O 3/4.12.2 LAND USE Cl!N SUS . . . . . . . . . . . . . . . . . . . . . .
3/4.12.3 INTERLABORATORY COMPARISON PROGRAM. . . . . . . . . . . . . 3/4 12-10 3/4 12-12 i i i O BEAVER VALLEY UNIT 2 x y + - - ~ ~ ~ . -nn . .,,-., - - -- .--.,.~.-- - - - . - . . - ,- -
IND EX () BA SES SECTION 3/4.0 APPLICABILITY . . . . . . . .. . . . . . . . . . . . . . . B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL. . . . .. . . . . . . . . . . . . . . . B 3/4 1-1 3/4.1.2 BORATION SYST EMS. . . . . .. . . . . . . . . .. . . . . B 3/4 1-5 3/4.1.3 MOVABLE 00KTROL ASSEMBLIES. . . . . . .. . . . . . . . . B 3/4 1-7 3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.1 AXIAL FLUX DIFFERENCE . .. . . . . . . . . . . . . . . . B 3/4 2-1 3/4.2.2 and 3/4.2.3 HFAT FLUX AND NUCLEAR WTHALPY HOT QlANNEL FACTORS. . . . B 3/4 2-4 () 3/4.2.4 QUADRANT POWER TIII RATIO . . . . . . . . . . . . . . . . B 3/4 2-6 3/4.2.5 DNB PARAMET ERS. . . . . . .. . . . . .. . . . . . . . . B 3/4 2-7 3/4.3 IN STRUMDrTATION 3/4.3.1 PRUT ECTIVE DiSTRUMENTATION. . . . . . . . . . . . . . . . B 3/4 3-1 3/4.3.2 DIGINEERED SAFETY FEATURE INSTRUMDITATION . . . . .. . . B 3/4 3-1 3/4.3.3 MONITORING INSTRUMDITATION Radiation Monitoring. . .... . . . .. . . . . . . . . B 3/4 3-3 Movable Incore Detectors. . . . . . . . . . . . . . ... B 3/4 3-4 Seismic Instrumentation . .. . . . . . . . . . . . .. . . B 3/4 3-5 Meteorological Instrumentation. . . . . . . . . . . . . . B 3/4 3-6 Remote Shutdown Ins trument at ion . - . . . . . . . . .. . . B 3/4 3-7 Fire Detection Instrumentation. . . . . . . . . . . . . . B 3/4 3-8 Chlorine batection Systems. . . . . . . . . . . . . . . . B 3/4 3-9 Accident Monitoring Instrumentation . . . . . . . . . . . B 3/4 3-10 Radioactive Liquid Effluent Monitoring Instrumentation. . B 3/4 3-11 Radioactive Caseous Effluent Monitoring Instrumentation . B 3/4 3-12 O BEAVER VALLEY UNIT 2 xi
IND EX () s SECT ION 3/4.4 REACTOR COOLANT SY ST EM 3/4.4.1 REACTOR COOLANT LOOPS. . . . . . . . . . . . . . . . . . .B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES. . . . . . . . . . . . . . . . . . . . . . .B 3/4 4-3 3/4.4.4 PRESSURIZER. . . . . . . . . . . . . . . . . . . . . . . .B 3/4 4-4 3/4.4.5 sr EAM c EN ERATOR S . . . . . . . . . . . . . . . . . . . . .B 3/4 4-5 3/4.4.6 REACTOR COOLANT SY ST EM LEAKAGE Leakage Detection Systems . . . . . . . . . . . . . . . .B 3/4 4-7 Oper ational Leakage . . . . . . . . . . . . . . . . . . .B 3/4 4-8 3/4.4.7 CH EMIST RY. . . . . . . . . . . . . . . . . . . . . . . . .B 3/4 4-9 3/4.4.8 SPECIFIC ACTIVITY. . . . . . . . . . . . . . . . . . . . .B 3/4 4-10 3/4.4.9 PRESSURE /T ENTERATURE LIMET S. . . . . . . . . . . . . . . .B 3/4 4-11 3/4.4.10 STRUCTURAL IKT EGRITY . . . . . . . . . . . . . . . . . . .B 3/4 4-20 3/4.4.11 RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . .B 3/4 4-21 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS . . . . . . . . . . . . . . . . . . . . . . .B 3/4 5-1 3/4.5.2 a nd 3/4.5.3 ECCS .SUB SYST EMS. . . . . . . . . . . . . . . . . . . . . .B 3/4 5-2 3/4.6 CONTAINMDrr SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT. . . . . . . . . . . . . . . . . . . .B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYST EMS . . . . . . . . . . .B 3/4 6-6 3/4.6.3 CONTAINMENT ISOLNTION VALVES . . . . . . . . . . . . . . .B 3/4 6-8 (O
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INDEX BA*ES bG _ 3/4.6 CONTAINMmT SYSTEMS (Continued) 3/4.6.4 COMBUSTIBLE GAS CONTROL. . . . . . . . . . . . . . . . . .B 3/4 6-9 3/4.6.3 SUBATMOSPHERIC PRESSURE CONTROL SYSTDI . . . . . . . . . .B 3/4 6-10 3 /4. 7 PLANT SY ST EMS 3/4.7.1 TURBINE CYCLE Safety Valves . . . . . . . . . . . . . . . . . . . . . .B 3/4 7-1 Auxiliary Feedwater System . . . . . . . . . . . . . . . .B 3/4 7-3 Primary Plant Demineralized Water . . . . . . . . . . . .B 3/4 7-4 Activity . . . . . . . . . . . . . . . . . . . . . . . . .B 3/4 7-5 Main Steam- Line Isolation Valves . . . . . . . . . . . . .B 3/4 7-6 3/4.7.2 STEAM GMERATOR PRESTRE/TD(PERATURE LIMITATION. . . . . .L 3/4 7-7 3/4.7.3 COMPONMT COOLING WATER SYST DI . . . . . . . . . . . . . .B 3/4 7-8 3/4.7.4 SERVICE WAT ER SYST EM . . . . . . . . . . . . . . . . . . .B 3/4 7-9 3/4.7.5 ULT IMAT E H EAT SINK , . . . . . . . . . . . . . . . . . . .B 3/4 7-10 3/4.7.6 FIDOD PROT ECTION . . . . . . . . . . . . . . . . . . . . .B 3/4 7-11 3/4.7.7 ' CONTROL ROOM D(ERGMCY HABITABILITY SYST Di . . . . . . . .B 3/4.7-12 3/4.7.8 SUPPLEMMTAL LEAK COLLECTION AND RELEASE SYST EM. . . . . .B 3/4 7-13 3/4.7.9 SEALED EURCE CONTAMINATION. . . . . . . . . . . . . . . .B 3/4 7-14 3/4.7.12 HYDRAULIC 'NUBBERS . . . . . . . . . . . . . . . . . . . .B 3/4 7-15 3/4.7.13 STANDBY SERVICE WAT ER SYSTEM . . . . . . . . . . . . . . .B 3/4 7-17 3/4.7.14 FIRE ' WPPRESSION SYST EMS . . . . . . . . . . . . . . . . .B 3/4 7-18 3/4.7.15 FIRE RAT ED ASSEMBLIES . . . . . . . . . . . . . . . . . .B 3/4 7-19 i 3/4.7.16 T ERRESTRIAL ECOLOGICAL SURVEY . . . . . . . . . . . . . .B 3/4 7-20 3/4.8 ELECTRICAL POWER SYST EMS 3/4.8.1 A.C. EURCES . . . . . . . . . . . . . . . . . . . . . . .B 3/4 8-1 3/4.8.2 ONSITE POWER DIST RIBlTTION SY ST EMS. . . . . . . . . . . . .B 3/4 8-1 - L
) i BEAVER VALLEY UNIT 2 xiii ,
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IND EX BASES SECTION l 3/4.9 REFUELING OPERATION S 3/4.9.1 BORON CONCMTRATION. . . . . . . . . . . . . . . . . . . .B 3/4 9-1 3/4.9.2 IN STRUMMTAT ION . . . . . . . . . . . . . . . . . . . . . .B 3/4 9-2 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . .B 3/4 9-3
,B 3/4 9-4 3/4.9.4 CONTAINMMT BUILDING PMETRATIONS. . . . . . . . . . . .
3/4.9.5 COMMUNICAT ION S . . . . . . . . . . . . . . . . . . . . . .B 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY. . . . . . . . . . . . . . .D 3/4 9-6 3/4.9.7 CRAhE TRAVEL - SPmT FUEL STORAGE BUILDING . . . . . . . .B 3/4 9-7 3/4.9.8 C001 ANT CIRCUIATION. . . . . . . . . . . . . . . . . . . .B 3/4 9-8 3/4.9.9 CONTAINMMT PURGE AND EXHAUSE SYST Di . . . . . . . . . . .B 3/4~9-9 g 3/4.9.10 and V 3/4.9.11 WAT ER LEVEL - REACTOR VESSEL AND STORAGE POOL. . . . . . .B 3/4 9-10 3/4.9.12 FUEL BUILDING .VmTIIATION SYSTEM-FUEL MOVEMET . . . . . .B 3/4 9-11 3/4.9.13 FUEL BUILDING VmTIIATION SYSTEM-FUEL STORAGE . . . . . .B 3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SIUIDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . .B 3/4 10-1 3/4.10.2 GROUP HEIGHT , INSERTION AND POWER DISTRIBUTION LIMIT S. . .B 3/4 10-2 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS - REACTOR CRITICALITY . .B 3/4 10-3 3/4.10.4 PHYSICS TEST S. . . . . . . . . . . . . . . . . . . . . . .B 3/4 10-4 3/4.10.5 NO FIDW TEST S. . . . . . . . . . . . . . . . . . . . . . .B 3/4 10-5 3/4.10.6 POSITION INDICATION SYSTEM-91UTDOWN . . . . . . . . . . .B 3/4 10-6 O BEAVER VALLEY UNIT 2 xiv
INDEX BA SES SECT ION 3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.1 LIQUID EFFLUENT F . . . . . . . . . . . . . . . . . . . . .B 3/4 11-1 3/4.11.2 GASEOU S EFFLUENT S. . . . . . . . . . . . . . . . . . . . .B 3/4 11-5 3/4.11.3 SOLID RADIOACTIVE WASTE. .. . . . . . . . . . . . . . . .B 3/4 11-11 3/4.11.4 TOTAL DOSE . . . . . .. . . . . . . . . . . . . . . . . .B 3/4 11-12 3/4.12 RADIOIDGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM . . . . . . . . . . . . . . . . . . . .B 3/4 12-1 3/4.12.2 IAND USE CEN SUS. . . ... . . . . . . . . . . . . . . . .B 3/4 12-2 3/4.12.3 IffrERLABORATORY COMPARISON PROGRAM . . .. . . . . . . . . .B 3/4 12-3 4 0 O BEAVER VALLEY UNIT 2 xv
,IND EX DESIGN FEATURES
(~ SECT ION 5.0 DESIGN FEATURES 5.1 SrT E Site Boundary for Gaseous Ef fluents. . . . . . . . . . . . . . . . .5-1 Site Boundary for Liquid Ef fluents . . . . . . . . . . . . . . . . .5-1 Exclusion Area . . . . . . . . . . . . . . . . . . . . . . . . . . .5-1 Low Population Zone. . . . . . . . . . . . . . . . . . . . . . . . .5-1 Flood Control. . . . . . . . . . . . . . . . . . . . . . . . . . . .5-1 5.2 CONTAINMEKT Configur at ion. . . . . . . . . . . . . . . . . . . . . . . . . . . .5-2 Design Pressure and Temperature. . . . . . . . . . . . . . . . . . .5-2 Pe netr at io ns . . . . . . . . . . . . . . . . . . . . . . . . . . . .5-2 5.3 REACTOR CORE Fuel Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . .5-3 Control Rod Assemblies . . . . . . . . . . . . . . . . . . . . . . .5-3 C 5.4 REACTOR COOIANT SY ST EMS Design Pressure and Temperature. . . . . . . . . . . . . . . . . . .5-3 Volume . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5-3 5.5 EMERGENCY CORE COOLING FYST EMS . . . . . . . . . . . . . . . . . . . .5-4 5.6 FUEL STORAGE Cr it ic ality. . . . . . . . . . . . . . . . . . . . . . . . . . . . .5-4 Dr ainage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5-4 Capacity . . . . . . . . . . . . . . . . . . . . . . . . . ... . . .5-4 5.7 SEISMIC CLASSIFICATION . . . . . . . . . . . . . . . . . . . . . . .5-4 5.8 MET EOROLOGICAL TOWER LOCATION. . . . . . . . . . . . . . . . . . . .5-4 i n) BEAVER VALLEY UNIT 2 xvi
4 t IND EX : I L
\ -- ADMINISTRATIVE CONTROLS i t
SECT ION : t 6.0 ADMINISTRATIVE CONTROLS r 4 6.1 RESPONSIBILITY , . . . . .. .. . . . . . . . . . . . . . . . . . .6-1 l 6.2 ORGANIZATION t Of fsite . ... . . . . .... . . . . . . . . . . . . . . . . . .6-1 t Unit Staff .. . . .. . . . . . . . . . . . .. . . . . . . . . .6-1 ! i 6.3 FACILITY STAFF QUALIFICATION S. . . . . . . . . . . . . . . . . . . .6-7 6.4 T RAINING . . .. . . . . ... . . . . . . . . . . . . . . . . . . .6-7 , t J !
; 6.5 REVIEW AND AUDIT i l
l 6.5.1 ONSIT E SAFETY COMMITTEE (OSC) ! I I Funct io n. ... . . .. . ... . . . . . . . . . . . . . . . . . .6-7 Composition' . .. . . . . . ... . . . . . . . . . . . . . . . . .6-7 l Alter nates . .. . . . . .. . . . . . . . . . . . . . . . . . . . .6-8 l Meeting Frequency . . .. . . . . . . . . . . . . . . . . . . . . .6 l
, Quorum. .. .. . . . . ... . . . . . . . . . . . . . . . . . . .6-8 i Res po ns ib ilit ies . . ... . . . . . . . . . . . . . . . . . . . . .6-8 !
Authority . .. . . .. . . . . . . . . . . . . . . . . . . . . .
.6-9 5 Records .... . . . . . . . . . . . . . . . .. . . . . . . . . .6-9 {
l 4 t 6.5.2 0FFSrT E REVIEW COMMITT EE (ORC) ' i Funct ion. ... . . .. ... . . . . .. . . . . . . . . . . . . .6-10 . Composition . .. . ... .. . . . . . . . . . . . . . . . . . . .6-10 f Alternates. .. . . . . . .... . . . . . . . . . . . . . . . . .6-11 4 Consultants . . . . .. ..... . . . . . . . . . . . . . . . . .6-11 . Meet ing Frequency . . . .. . . . . . . . . . . .. . . . . . . . . .6-11 ! j Quorum. . .... . . . .. . . . . . . . . . . . . . . . . . . . .6-11 j
! Review. ..... . . . . .. . . . . . . . . . . . . . . . . . . .6-11 , ; Audits. . .. .. . . . . .. . . . . . . . . . . . . . . .. . . .6-12 i Author ity . .. . . . . ..... . . . . . . . . . . . . . . . . .6-13 Records . . .. . . . .. .. . . . . . . . . . .. . . . . . . . .6-13 5
k
; 6.6 REPORTABLE OCCURRENCE ACTION . . . . . . . . . . . . . . . . . . . .6-13 [
i i i BEAVER VALLEY UNIT 2 xvii : t
?
.I [ I [
IND EX ADMINISTRATIVE CONTROLS
\ __ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _
! SECTION 6.7 SAFETY LIMIT VIOLATION . . . . . . . . . . . . . . . . . . . . . . .6-13 6.8 PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6-14 6.9 REPORTING REQUIREMDrf S 6.9.1 ROUTINE AND REPORIABLE OCCURRENCES St artup Report . . . . . . . . . . . . . . . . . . . . . . . . . . .6-15 Annual Reports. . . . . . . . . . . . . . . . . . . . . . . . . . .6-15 Monthly Operating Report . . . . . . . . . . . . . . . . . . . . . .6-16
~
Annual Radiological Environmental Report. . . . . . . . . . . . . .6-17 Semi-Annual Radioactive Ef fluent Release Report . . . . . . . . . .6-19 Radial Peaking Pactor Limit Report . . . . . . . . . . . . . . . .6-20 Bi-Annual Environmental Operating Report . . . . . . . . . . . . .6-20 6.9.2 SP EC IA L REP 0KI S . . . . . . . . . . . . . . . . . . . . . . . . . .6-21
.~
6.10 RECORD RET ENTION. . . . . . . . . . . . . . . . . . . . . . . . . .6-22 U) 1 6.11 RADIATION PRUT ECTION PROGRAM. . . . . . . . . . . . . . . . . . . .6-24 6.12 HIGH RADIATION ' AREA . . . . . . . . . . . . . . . . . . . . . . . .6-24 . 6.13 PROCESS CONTROL PROGRAM (PCP) . . . . . . . . . . . . . . . . . . .6-25 6.14 0FFSTT E DOSE CALCULATION MANUAL . . . . . . . . . . . . . . . . . .6-25 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYST'.MS. . . . . . . .6-25 6.16 RADIOlDGICAL ENVIRONMDrTAL MONITORING PROGRAM . . . . . . . . . . .6-28 i ~h (V BEAVER VALLEY UNIT 2 xviii
u _ m 2 a .e- a w a. + a -m-, m --m-, m,--- --_- I
- l lO 1
I f I l l l l f
-(
SECTION 1.0 i DEFINITIONS I 4 l 1 A 1O I 1 4 I I 1 i I. I i i 1 0 9 1 l0 I l l l
1.0 DEFINITION S k 1.1 DEFINED TERMS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specification. 1.2 THERMAL POWER: THERMAL POWER shall be the total reactor core heat trans fer rate to the reactor coolant. 1.3 RATG THERMAL POWER: RATE THERMAL POWER shall be a total reactor core heat trans fer rate to the reactor coolant of 2652 MWt. 1.4 OPERATIONAL MODE: An OPERATIONAL MODE shall correspond to any one inclusive combination y of core reactivity condition, power level, and average reactor coolant temperature specified in T able 1.1. 1.5 ACTION ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the spec ific at ions . O BEAVER VALLEY UNIT 2 1-1
1.0 DEFINITION S 1 1.6 OPERABLE - OPERABILITY: A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified func-tion (s) . Implicit in this definition shall be the -assumption that all necessary attendant instrumentation, controls, normal and emergency l[ electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable 'of performing their related support funct ion ( s) . 1.7 REPORTABLE EVI!NT: A REP 0ltTABLE EVENT shall be any of those conditions specified in Sec-tion 50.73 to 10 CFR Part 50. 1.8 C0ffrAINMENT INTERITY: CONTAINMl!NT INTERITY shall exist when: 1.8.1 All penetrations required to be closed during accident l.J' conditions are either: d a) Capable of being closed by an OPERABLE containment automat ic l } isolation valve system, or b) Closed by manual valves , blind flanges , or deactivated auto- l
}
matic valves secured in their closed positions, except as provided in T able 3.6-1 or Specification 3.6.3.1. 1.8.2 All equipment hatches are closed and sealed, l 1.8.3 Each air lock is OPERABLE pursuant to Specificat ion 3.6.1.3, l 1.8.4 The containment leakage rates are within the limits of Speeification 3.6.1.2. I O BEAVER VALLEY ' UNIT 2 1-2
1.0 DEFINITION S O 1.9 CHANNEL CALIBRATION A GANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. l} The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and shcIl include the Channel Functional Test. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. 1.10 CHANNEL CHECK: A CHANNEL WECK shall be ':he qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from inde pe nde nt ins tr ume nt channels measuring the same parameter. 1.11 CHANNEL FUNCTIONAL TEST A GANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to O- verify OPERABILITY including alarm, interlocks , and/or trip func-t io ns . 1.12 CORE ALTERATION: CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALT ERATION shall not prec lude completion of movement of a component to a safe conservative position. 1.13 sitTTDOWN MARGIN: 9ttTIDOWN MARGIN shall be the inseantaneous amount of reactivity by which the reactor is suberitical or would be subcritical from its prese nt condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be i fully withdrawn. i I O BEAVER VALLEY UNIT 2 1-3 1
]
l .0 DEFINITION S 1.14 IDDrfIFIED LEAKAGE: IDDITIFIED LEAKAGE shall be:
- a. Icakage (except C0ffrROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Im akage into the cont airunent atmosphere from sources that are both specifically located and known either not to inter fere with the operation of leakage detection systems or not to be PRESERE BOUNDARY LEAKAGE, or
- c. Reactor coolant system leakage through a steam generator to the secondary system.
i 1.15 UNIDDrrIFIED LEAKAGE: UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. I 1.16 PRESERE BOUNDARY LEAKAGE:
- PRESERE BOUNDARY LEAKAGE shall be leakage (except stema generator tube leakage) through a nonisolable f ault in a Reactor Coolant System component body, pipe wall, or vessel wall.
1.17 CONTROLLED LEAKAGE:
- CONTROLLED LEAKAGE shall be that seal water flow supplied to the
! reactor coolant pump seals, i l l O BEAVER VALLEY UNIT 2 1-4
1.0 DEFINITION S l0 - - 1.18 QUADRAW POWER TILT RATION: QUADRAYT POWER TILT RATIO shal1 ~be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one (1)excore detector inoper ab le, the remaining three (3) detectors sham be used for lI l computing the average. 1.19 DOSE QUIVALENT I-131: DOSE MUIVALDrT I-131 shall be that concentration of I-131 (micro-curie /gran) which alone would produce the same thyroid dose as the quanticy and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109, 1977. l 1.20 STAGGERED TEST BASIS: A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, tr a ins , or ot he r O' designated components obtained by dividing the specified *.ast interval into n equal subintervals,
- b. The testing of one (1) system, subsystem, train, or other l designated compone nt at the beginning. of each subinterval.
1.21 FRMUDICY NOTATION: The FREQUDiCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. 1.22 REACTOR TRIP SYSTEM RESPONSE TIME: The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel ! sensor until loss of stationary gripper coit voltage. I BEAVER VALLEY UNIT 2 1-5
. _ - . .- . ~ . . - - . ...- - --. -- . , - - .. _ -- _ .. - .- .,
i I 4 1.0 DEFINITIONS < O. v . 1.23 WGINEERED SAFETY FEATURE RESPONSE TIME: I ! The WCINEERED SAFETY FEATURE RESPONSE TIME shall be that time inter-j val from when the monitored parameter exceeds its ESF actuation set-point at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their re-4 quired positions, pump discharge pressures reach their required values, etc.). Times shall include ~ diesel generator starting and ] sequence loading delays where applicable. 1.24 AXIAL FLUX DIFFERBCE: ! 4 : $ AXIAL FLUX DIFFERWCE shall be the dif ference in normalized flux sig-l - nals between the top and bottom halves of a two section excore neutron j detector. ! l
+
i j 1.25 PHYSICS TEST S: l l !
! PHYSICS TESTS shall be those teste performed to measure the fundame n-l tal nuclear characteristics of the reactor core and related instrumen- !
l tation (1) described in Chapter 14.0 of the FSAR, (2) authorized under l l the provisions of 10 CFR 50.59, or (3) otherwise approved by the Com-j p mis s ion. i \J , 1.26 Y - AVERAGE DISIlfrBGRATION WERGY: l T shall .be the average sum (weighted in proportion to the concentra- l} } tion of each radionuclide in the reactor coolant at the time of samp1- ; ing) of the average beta and gamma energies per disintegration (in
~
j . ~ MeV) for isotopes, other than iodines, with half lives greater than 15 I sinutes, making up at least 95% of the total noniodine activity in the : coolant. ! ! i l 1.27 SOURCE CHECK: ! I I A S)URCE OtECK shall be the qualitative assessment of channel response 1 l when the channel sensor is exposed to a radioactive source. [ r 3 i 1.28 PROCESS CONTROL PROGRAM: 7 l ! A PROCESS CONTROL PROGRAM (PCP) shall be the manual or set of operat- , ing parameters detailing the program of sampling, analysis, and evalu- l ation by which SOLIDIFICATION of wet radioactive wastes is assured. lI l Requirements of the PCF are provided in Specification 6.14. ; e l { O BEAVER VALLEY UNIT 2 1-6 i I i i' l__ _
1.0 DEFINITIONS lO I l 1.29 SOLIDIFICATION: I !DLIDIFICATION shall be the conversion of radioactive wastes fr om liquid systems into a form that meets shipping and burial ground re-quir eme nt s . 1.30 0FFSIT E DOSE CALCULATION MANUAL (ODCM): An OFFSITE DOSE CALCUIATION MANUAL (ODCM) shall be a manual cont aining i the methodology and parameters to be used in the calculation of off-site doses due to radioactive gaseous and liquid ef fluents and in the calculation of gaseous and liquid ef fluent monitoring instrumentation al arm / tr ip se tpoint s . Requirements of the ODCM are provided in Speci-fication 6.15, 1.31 GA SEOU S RADWA ST E TREATMENT SYST EM: A GASEDUS RADWA ST E TREATMENT SYST EM is any systes designed and in-atailed to reduce r adioactive gaseous ef fluenta by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total r adioactivity prior to release to the enviromsent.
~1.32 VDrTIIATION EXHAUST TREATMENT SY ST EM:
A VINTI!ATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or r adioactive material in particulate form in ef fluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing lodines or particulates from the gaseous exhaust strees prior to the release to the envirorunent (such a system is not con-sidered to have any ef fect on noble gas ef fluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are r.ot considered to be VINTI!ATION EXHAUST TREATMENT SYST EM component s . 1.33 PURG E-PURGING : PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, con-centration or other operating conditions, in su:h a manner that re-placeme nt air or gas is required to purify the :onfinement. O BEAVER VA1. LEY UNIT 2 1-7 l L
i i l 1.0 DEFINITIONS 1.34 VDTTING : 4 VENTING is the controlled process of dischargirg air or gas fr om a
; confinement to maintain temper ature, pressure, humidity, concentr at ion t j or other operating conditions, in such a manner that replacement air 4
or gas . is not provided or required during VENTING. Vent, used in sys-tem names , does not imply a VBfTING process . l i
} 1.35 MAJOR CHANGES:
MAJOR OtANGES to radioactive waste systems, as ad&essed in Paragr aph 6.16.2, (liquid, gaseous and solid) shall include the following: i ; i 1. Major changes in process equipment , components , structures and effluent monitoring instrumentation from those described in the l Final Safety Analysis Report (FSAR) or the Hazards Summary Report i and evaluated in the staff's Safety Evaluation Report (SER) (e.g., deletion of evaporators and installation of domineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);
; 2. Major changes in the design of radwaste treatment systese (liquid, I
- O d tid) es e cooid < tric eir t c<
ties or activity of ef fluents released or volumes of solid waste es a et-i stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement); ] , j 3. Changes in systes design which may invalidate the accident analy-sie as desceibed in the SER (e.g., changes in tank capacity that
! would alter the curles released); and ) 4. Changes in system design that could potentially result in a sig-I nificanc increase in occupational exposure of operating personnel (e.g. , use of tempor ary equipment without adequate shielding pro- , vis io ns ) . ; 1.36 MEMBER (S) 0F THE PUBLIC i
i MEMBERS OF THE PUBLIC shall include all persons who are not occupa-j tionally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also ex-cluded from this category are persons who enter the site to .;ervice equipment or to make deliveries and persons who tr averse portions of j the site as the consequence of a public highway, railway, or waterway [ 4 located within the confines of the site boundary. This category does , include persons who use portions of the site for recreational, occupa-
- tional, or other purposes not associated with the plant.
!O BEAVER VALLEY UNIT 2 1 , 1-8
,e +w.-n.. .w---w- pge,pwr,r ---h--+'--"'u'* w'esw' ----=v-=w-*"ww --
my---3ri t'f'7v"- yw "- rwr v mw w verw'*w-ywemTg*r's'tr'--Tvn**1rr wei-='w"**"*v'W"fNwe'==99Y-MWW*
._. - _ - . _ - - - - -. . _. =- . - . -
d TABLE 1.1 l () OPERNTIONAL MODES RFACTIVITY % RAT ED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER
- T DtPERATURE eff i 1. POWER OPERATION 2,0.99 .>$% >350* F
- 2. STARTUP >0.99
<5%
)350' F
- 3. HOT STANDBY <0.99 0 >350' F
- 4. HOT SHUTDOWN <0.99 0 350' F > T svg
' >200* F
! 5. COLD SHUTDOWN <0.99 0 -
<200* F 4
- 6. REFU ELING** <
_0.95 0 _!40' F i i i O I t
, t 2
l I i l t i , l
- Excluding decay heat. [
! ** Reactor vessel head unbolted or removed and fuel in the vessel. L l l ; i ! (:) BEAVER VALLEY UNTT 2 1-9 i i t i i
. - .- . . . . _ . -- - . _ . _ ~ - . - - _ . - .
' TABLE 1.2 () i j FREQUENCY NOTATION r f NOT AT ION FREQUENCY I , S At least once per 12 hours. 7 1 ' j D At least once per 24 hours. 1 i ' W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. j At least once per 184 days. ) SA l R At least once per 18 months. i S/U Prior to each reactor startup. P Completed prior to each release. (} N/A Not applic ab le , i i i I
?
l 1 i , d t l 1 i t }
- t I r i .
I a f i 4 i I I i () BEAVER VALLEY UNIT 2 1-10 , I l
-+ -. . . - - u.-- . - --. - - . -
4 u - - I I I l i i 1 i 1 l I 1 SECTION 2.0 4 SAFETY LIMITS I MD LIMITING SAFETY SYSTEM SETTINGS 1 I O t i i I i 4 O l I I
- - . - . - - - - _ - - , - - - , , , , - . - -. - - ~ - - -,
- - - - - - _ . - - . - _ . - . . . . . - - - - - , , , , - - - - - - , , , , - . , , . . , - - _ - _ _ - . . - - , . - . _ . ~ , . . , - - . -
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYST EM SETTING S 2.1 SAFETY LIMIT S E REACTOR COR_E 2.1.1 T he combinat ion of THERMAL POWER, pres sur izer pr es sur e , anl the highest operating loop coolant temper atur e (T ,yg) shall not exceed the lim-j its shown in Figure 2.1-1 for 3 loop operation and Figure 2.1-2 for 2 loop operation. APPLICABILITY: MODES 1 and 2. ACT ION: Whe never the point defined by the combination of the highes t operating loop aver age temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. 1 O J J BEAVER VALLEY UNIT 2 2-1 i i
1 l 6es , Q 660 . Unacceptable P"" " 655 , 650 . 2400 PSIA i 645 <- 640 - 2250 PSIA 635 - 630 - 625 2000 PSIA _ 620 , u 615 , w
$ 610 ,
E05 , 1775 PSIA 600 , O SS5 590 , Acceptable Operation
.585 -
580 > 575 "
- 570 '
565 j 0 .1 .2 .3 .4 .5 .6 .7 .8 .9 1.0 1.1 1.2 Power (fractionofnominal) Figure 2.1-1 Reactor Core Safety Limit - 3 Loops in Operation j t i O ! BEAVER VALLEY UNIT 2 2-2 i N _ _ , _ - - . _ . . - _ , . _ _ - . . - - . - - . . , . - , - . - . _ . , . . _ - - . . _ . - . , ~ , , , . . - - - - , - . . . - _ _ . ~ . - . . . . . . . - , - . - _ . , , , . . . - , , _ _ . . - - .
l l l 665 - 'O 660 , 655 ,, Unacceptable Operation 650 ,, 645 - 2400 PSIA 640 " 2250 PSIA 635 630 625 620 ., 2000 PSIA 615 , { 610 " l p 605 , 1775 PSIA e
- 600 < ,
O SS5 " 590 585 , , Acceptable Operation ! 580 , 575 " 570 565 560 . j 0 .1 .2 .3 .4 .5 .6 .7 .8 .9 1.0 Power (Fraction of Nominal) Figure 2.1-2 Reactor Core Safety - 2 Loops in Operation O BEAVER VALLEY UNIT 2 2-3 l f I
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTING S 2.! SAFETY LIMIT S REACTOR _COOLA!rr SYST EM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: i MODES 1, 2, 3, 4, and 5. l ' ACT ION : MODES I and 2: Whe never the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MOD E S 3, 4, and 5 : Whe never the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactot Coolant System pr es sur e to within its limit within j five minutes. O l I 4 1 1 b l i O BEAVER VALLEY UNIT 2 2-4 i
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYST EM SETTINGS REACTOR TRIP SYST EM INSTRUMDITATION SETPOINT S LIMITING CONDITION FOR OPERATION y- - - ;._-- - - - 2.2.1 The reactor trip system instrumentation setpoints shall be set cons is te nt with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACT ION: With a re actor tr ip system ins tr ume nt at io n se t point le s s conserv at ive than the value shown in the Allowable Values column of Table 2.2.1, de-clare the channel inoperable and apply the applicable ACTION s t ateme nt requirement of Specification 3.3.1 until the channel is restored to OPER-ABLE status with its trip se tpoint adjus ted cons is te nt with the Trip Se tpoint value. O BEAVER VALLEY UNIT 2 2-5
. _ _ _ _ ____ _ . . _ _ _ . _ _ . - _ _ - . _ . _ .. .._._ . _ _ . . _ _ _ _ _ . _ _ __ _ ._ _ - ___=..__ __ _ _ _ _ _ _
i O O O TABLE 2.2-1 R EACT OR T R I P SY ST DI IN ST RUN ENT AT ION T R I P SET PO INT S FUNCTIONAL UNIT TRIP SETPOINT ALIDWABLE VALUES l 1. Manual Reactor Tr ip Not Applicable Not Applicable Low Setpoint - < 26% of RAT ED
- 2. Power Range, Neutron Flux low Se t point - < 25% o f RAT ED l '
THERMAL POWER THERMAL POWER l 1 High Setpoint - < 109% o f RAT ED High Setpoint - < 110% of RAT ED l ! THERMAL POWER THEIMAL POWER i
- 3. Power Range, Neutron Flux, High < 5% of RATED THERMAL POWER with < 5.5% of RAT ED THERMAL POWER with Posit ive Rate a time constant 12 second a t ime cons tant 12 secoral
- 4. Power Range, Neutron Flux, High < 5% of RAT ED THERMAL POWER with < 5.5% of RAT ED THERMAL POWER with Negative Rate a time constant 12 second a t ime cons tant > 2 second
- 5. Intermediate Range, Neutron Flux < 25% of RATED THERMAL POWER < 30% of RAT ED THERMAL POWER
- 6. Source Range, Neutron Flux < 105 counts per second < l.3 x 105 counts per second
- 7. Overtersperature Delta T See Note 1 See Note 2
- 8. Overpower Delta T See Note 3 See Note 4
- 9. Pressur izer Pressure--Low > 1945 psig > 1935 psig
- 10. Pressurizer Pressure--High < 2385 psig < 2395 psig
!! . Pressur izer Water level--High < 92% of instrument span < 93% of instrument s pa n
- 12. Loss of Flow > 90% of design flow per loop * > 89% of design flow per loop *
- Design flow is 88,500 gpm per loop.
SEAVER VALLEY UNIT 2 2-6
. - - - -- --. . - . . - __ _ - . - ~ - _ _ -
O O O TABLE 2.2-1 (Cont inued ) REACTOR TRIP SYSTEM INSTRUMBfTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALIhWARLE VALUES
- 13. Steam Generator Water level--Iow-Iow > (later) of narrow range ins trume nt > (later) of narrow range i ns trume nt span-each steam generator span-each steam generator 14 Steam /Feedwater Flow Mismatch and Iow -< 40% of full steam flow at RAT ED THERMAL < 42.5% of full steam flow at RAT ED Steam Generator Water Level POWER coincident with steam generator THERMAL POWER coincident with steam water level gener ator water level
> 25% of narrow range ins t rume nt s pa n-- > 24% of narrow range i ns t rument each steam generator span--each stema generator
- 15. Undervaltage-Reactor Coolant Pumps 75% $ of nominal bus voltage- > 70% of nominal bus volt age-each I each bus bus
- 16. Under frequency-Reactor Coolant Pumps 58.0 H, + 0.1 Hz -each bus > 57.5 H, -each bus
- 17. Turbine Trip
- a. Emergency Trip Header Low 1000 psig + 42 psig
- b. Turbine Stop Valve > 1% open > II open
- 18. Safety Inject ion Input from ESF Mot Applicable Not Applicable 19 Reactor Coolant Pump Breaker Not Applicable Not Applicable I
Posit ion Tr ip BEAVER VALLEY UN TT 2 2-7 l
....__..___..m._____..___.____.._.___ . _ _ . . _ . . . . _ . . _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _._...m._ . . . - . _ . - _ . . _ _ _ . . _ . - _ . O O O TARLE 2.2-1 i (Cont inued) REACTOR TRIP SYSTDI INSTRUMDITATION TRIP SETPOIItTS FUMCTIONAL UNIT TRIP SETPOIItr ALLOWABLE VALUES , 20 Reactor Trip System Interlocks (Based on Ascending Power) A. Intermediate Range Neutron > I x 10-10 Amps > 6 x 10-II Amps Flux, P-6 B. Power Range Neutron Flux, < M% RAT ED THERMAL POWER < 31% RAT ED THERMAL POWER ' P-8 C. Power Range Neutron Flux, < 70% RATED THERMAL POWER < 71% RAT ED THERMAL POWER P-9 . D. Power Range Neutron Flux, 10% RAT ED THERMAL POWER > 9% and < 12% RAT ED THERMAL P-10 (Input to P-7) POWER E. Turbine Impulse Chamber < 66 PSIC < 72 P SIG Pressure, F-13 (Irput to P-7) , i s ! t l i t
*7 ;
DEAVER VALLEY UNIT 2 2-8 l I
.,----_.__.-_,.m. -_ _._,..-___..._,_~,,_,--_.-__-..,...---.-_,_-.,-,--.....~,3
_ . - , _ , . ._. -, ,,,___,.-_y.. _.-. .--
- _ ~ - _ - - - - . - - - . - - - - , - - . _ _ - _ - - - - . - - - - .- _ ~ . - - . - . _ _ ~ -- .. _ -_-_.-_~.- . O O O TABLE 2.2-1 (Cont inued) REACTOR TRIP SYSTDt INSTRUMENTATION TRIP SETPOINTS NOTAT ION l , l NOT E 1: OVERTEMPERATURE Deita T l (1 + t i s) 1 (1 + c as) I i l Delt a T ( ) < De l t a T O El-K2 [T ( )- T I] + K3 (P - P I) - f g (De l t a I) l (1 + t 8 2) I*E38 II + E S) S 1+t68 ! Where: Delt a T = Measured Delta T by RTD Manifold Instrumentation; l I + tgS !
= Lead-lag compensator on measured Delta T ;
I+t35 l tg, t2 = Time constants utilized in lead-lag controller for Delt a T , t g = 8 seconds, t2 ".3 seconds; ; i 1
= Lag compensator on measured Delta T ;
I+t3s t3 = T ime cons t ant s ut ilized in t he lag compensator for Delta T , t3 = 0 seconds;
~
Delta TO = Indicated Delta T at RATE THERMAL POWER; I l i Kg = 1.2006 for 3 loop operat ion 1 K2
= 0.01747; BEAVER VALLEY LMIT 2 2-9 1
I 4 1 I . _ . , . _ _ - _ -
._ - , , - - - - ._ . _ ~ , _ , , - _ . - . - - - - - - - - - _ . . . - --- -, - - , .-,____.__--____ _ _
_. ._ _ _ _ - - - _ ~.- -. , O O O TABLE 2.2-1 (Cont inued) REACTOR TRIP SYSTDt INSTRUMIMTATION TRIP SETPOINTS NOTAT ION (Cont inued) NUT E 1: ( Cont inued) 1+t48 = The function generated by the lead-lag controller for T ,y dynamic caspe nsat ion; t4, t5 = Time constant s ut ilized in the lead-lag controller for T ,y , t4 = 30 seco nd s , t5" 4 seconds;
- T = Average temperature, *F; 4
1 = Lag compensator on measured T,y ; , I + t6S t6 = T ime cons t ant utilized in the measured T,y lag compensator , t6 = 0 seconds; 1 > ! TI -
< 576.2*F (Nominal T avg at RAT ED THERMAL POWER);
i f K3 = 0.000823; P = Pressurizer pressure, psig; ! pl = 2235 psig (Nominal RCS operating pressure); and ! B EAVER VA LLEY UN IT 2 2-10 I i i____., _
O O O TABLE 2.2-1 (Co nt inued ) REACTOR TRIP SYSTEM INSTRIIMENTATION TRIP SETPOINTS NOTAT ION (Cont inued) NUT E 1 : (Continued) . S = Laplace trans form operator , second-I-and f g (Delta I) is a funct ion of the indicated dif ference between top and bottom detectors of the powe r-r a nge 1 nuclear ion chambers; with gains to be selected based on measured ins tr ume nt response during plant startup tests
- such that
- (i) For qt 9bbetween -34% and + 10% fg (Delta I) = 0, where q and qb are Per ce nt RAT ED THERMAL POWER
, in the top and bottom halves of the core respectively, and q t
- 9b ts total THERMAL POWER in percent
, of RAT ED THERMAL POWER; (ii) For each percent that the magnitude of qt ~9 b exceeds -34% the Delta T Trip Setpoint shall be automatically reduced by 1.439% of its value at RAT ED THERMAL POWER; and (iii) For each percent that the magnitude of q q exceeds +10% the Delta T Trip Setpoint shall be b
- automatically reduced by 1.789% of its value at RAT ED THERMAL POWER.
i 1 ^ NOT E 2 : The channel's maximus Tr ip . Setpoint shall not exceed its computed Trip Setpoint by more than 3.8%. i l REAVER VALLEY UNIT 2 2-11
O O O TABLE 2.2-1 , (Cont inued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1 NOTAT ION (Cont inued) , NUT E 3: OVERPOWER Delt a T 1 + tgS I t 78 I I I Delta T ( )( )< Delta TO K4-K5( )( )T-K6 [T ( -) - T II) - f2 (Delta I) 1+t2 S 1+t3S I+t7 S 1+t6S 1+t6S Where: Delta T = As defined in Hote 1; I + tgS
= As defined in Note 1; I+t 2S i , tg, t2 = As defined in Note 1;
} f I
= As defined in Note 1; t
t 3
= As defined in Note 1; Delta TO -
= As defined in Note 1; K4
= 1.0781 1
R EAV ER VA LLEY UN IT 2 2-12 1 i
)
l
. . -_ _ _ - ~_ . .-_-
l O O O TABLE 2.2-1 (Co nt i nued) REACTOR TRIP SYST EM INSTRUMFNTAT ION TRIP SETPOINTS NUTAT ION (Cont inued) NUT E 3: (Continued) K5
= 0.02/*F for incraasing average temperature and 0 for decreasing average temperature; t7S
=
The function generated by the rate-lag controller for T, d ynamic canpe ns a t io n; t7
=
T ime constants ut ilized in the rate-lag controller for T , , t7 = 10 secoals; i 1
= As defined in Note 1; 1+t6S 4
4 4 t6 = As defined in Note 1; [ K6
= 0.00ll5/*F for T > T II a nd K6 = 0 for T < T II; T = As defined in Note 1; i
! T II = As indicated T at RAT ED THERMAL POWER (Calibrat ion temperature for Delta T instrument at io$ 8< 576.2*F); ) BEAVER VALLEY UN IT 2 2-13 1 5 1
O O O : i 4 TABLE 2.2-1 ! (Co nt inued ) a REAcr0R TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS j NorATION i (Cont inued) { NOT E 3: (Continued) i i S = As defined in Note 1; and 1 4 4
= 0 for all j f2 (Delta I) I i
1 j NOT E 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip .Setpoint by more than 3.0%. l 1 4 1 i } } J J i i i i l i i a i REAVER VALLEY UNIT 2 2-14 ) i 1 1 i
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT S 2.1.1 REACTOR CORE BASES The restrictions of this safety limit prevent overheating of the fuel ami pos sible cladding perforation which would result in the rele ase of fis-sion products to the reactor coolant . Overheating of the fuel cladding is pr eve nt ed by res tr ic t ing fue l ope rat io n to within the nuc le ate boiling regime where the heat tr ans fe r coef ficient is large and the cladding surface temper atur e is slightly abov e the coola nt satur at io n tempe r a-ture. Operation above' the upper boundary of the nucle ate boiling regime could r es ult in excessive cladding te:speratures because of the onset of depar-ture fr om rucleate ' boiling (DNB) a nd the res ult ant sharp red uct io n in he at tr ans fe r coe f ficie nt . DNB is not a directly measurable par ase te r during operation and therefore,- THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the R-grid correla-t io n. The R-grid DNB correlation has be e n developed to predict the DNB flux and the loca t io n of DNB for axially uniform and rion-uniform, heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal oper-O ational tr ans ient s , and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent conf ide nce level that DNB will not occur and is cho'se n as an appropriate margin to DNB for all operating conditions. The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pres sure and average temperature for which the minimum of DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. S The curves are based on an e nt halpy hot channel factor F'i M' f 1.55 and a reference cosine with a peak of 1.55 or axial power shape. An allowance is included for an incre ase in H, at reduced power based on the expression: F H
= 1.55 [1 + 0.2 (1-P)]
where P is the fraction of RAT E THERMAL POWER l l O BEAVER VALLEY UNIT 2 B 2-1 i
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYST EM SETTINGS 2.1 SAFETY LIMIT S
. - REACTOR CORE BA SES These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdr awn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (Delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance ef fect on the Overtemperature Delta T trip will reduce the se tpoint to provide protection consistent with core safety limits.
, I I
O BFAVER VALLEY l! NIT 2 B 2-2 l l 1
l 2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS l 2.1 SAFETY LIMIT S l 2.1.2 REACTOR COOLANT ' SYST EM PRESSURE BA SES The restriction of this safety limit protects the integrity of the Reac-tor Coolant System from overpressurization and the r eb y pr eve nt s the re-lease of radionuclides contained in the reactor coolant fr om re ac hing t he cont ainment atmosphere. T he reactor pressure vessel and pres sur izer are des igned to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transi-e nt pressure of 110 percent (2735 psig) of de s ign pr es sur e . T he safety-related portion of the Reactor Coolant System piping and fittings are de s igned to ASME III which permit a maximum transient pres sure of 120 1 per ce nt (2985) psig of component des ign pr es sur e . The Safety Limit of 2735 psig is therefore, cons is tent with the ' ' des ign criteria and associated code requirements. The entire Reactor Coolant System is hydr otes ted at 3107 psig to demon-s tr ate integrity prior to initial operation. O o\ BEAVER VALLEY UNIT 2 B 2-3
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYST EM SETTING S 2.2 LIMITING SAFETY SYSTEM SETTINGS O (j 2.2.1 REACTOR TRIP SYSTEM IN STRUMDITATION SET POINT S 1 BASES REACTOR TRIP SET POINT S: The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a Trip Setpoint less conservative than its Setpoint Limit but within its specified Allowable Value is accept able on the basis that each Allowable Value is equal to or less than the drif t allowance assumed to occur for each trip used in the accident analyses. MANUAL REACTOR TRIP: The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. POWER RANGE, NEUTRON FLUX: (} The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low setpoint provides redundant protection in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 9 per ce nt of RAT ED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RAT ED THERMAL POWER), POWER RANGE, NETTRON FLUX, HIGH RAT ES: The Power Range Positive Rate trip provides protection agains t rapid flux increases which are characteristic of rod ejection events fr om any powe r level. Specifically, this trip complements the Power Range Neutron Flux High -and low trips to ensure that the criteria are met for rod ejection from partial power. O b B 2-4 BEAVER VALLEY UNIT 2
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMDrTATION SETPOINT S BA SES The Power Range Negative Rate trip provides pr ot ect io n to ensure that the minimum DNBR is maintained above 1.30 for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maint ained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurrirs by tripping the reactor for all single or multiple dropped rods. INTERMEDIATE AND SOURCE RANGE, NUCLEAR FLl'X: The Intermediate and Source Range,_ Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant channels protection to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes ac t ive . T he Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATl!D THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation o f t he tr ip s associated with either the Intermediate or Sour ce Range Channels O. in the accident analyses; however , their functional capability at t he specified trip settings is required by this Specification to enhance the over all reliability of the Reactor Protection System.
~
OVERT EMPERATURE DELTA T : The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power , coolant temper ature, and axial power distribution, provided that the tr ansient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and he at capacity of water with temperature and dynamic campensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figures 2.1-1 and 2.1-2. If axial peaks are greater than des ign, as indicated by the dif ference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the not at io ns in T able 2.2-1. O BEAVER VALLEY UNIT 2 B 2-5
1 2.0 SAFETY LIMIT S AND LIMITING SAFETY SYST EM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINT S BASES
. . . . . I Operation with a reactor coolant loop out of service below the 3 loop P-8 setpoint does not require reactor protection system setpoint modi-fication because the P-8 setpoint and associated trip will prevent DNB d ur i ng 2 loop operation exclusive of the Overtemperature Delta T setpoint. Two loop operation above the 3 locp P-8 setpoint is permissible af ter resetting the K1, K2, and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 2 loop value. In this mode of operation, the P-8 inter-lock and trip functions as a High Neutron Flux trip at the reduced power level.
OVERPOWER Delta T: The Overpower Delta T reactor tr ip pr ov ide s as sur ance of fue l integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip. Tre setpoint includes corrections for changes in de ns ity and he at capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detec-tors. No credit was taken for operation of this trip in the accident analy- / ses; however, its functional capability at the specified trip setting is \ required by this Specification to enhance the overall reliability of the Re ac tor Protection System. PRESSURIZER PRESSURE: The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore, set lower than the set pr es s ur e for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pr e s s ur e . PRESSURIZER WATER LEVEL: The Pressurizer High Water Level trip ensures pr ot ect io n agains t Reactor Coolant System overpressurization by limitirg the water level to a volute sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however , its functional capability at the specified trip setting is required by this Specification to enhance the over-all reliaoility of the Reactor Protection System. O BEAVER VALLEY UNIT 2 B 2-6
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTING S 2.2 LIMITING SAFErY SYSTEM SETTINGS REACTOR TRIP. SYSTEM IN SrRUMENTATION SETPOINT S BASES 1 DSS OF FLOW: The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps. Above 11 percent of RAT ED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90 percent of nominal full- loop flow. Above 31 percent (P-8) of RAT ED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90. percent of nominal full loop flow. This latter trip will prevent the minimuim value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when 2 loops are in operation and the Overtemperature Delta T trip set point is adjusted to the value specified for all loops in oper ation. With the Overtemperature T trip set point adjusted to the value specified for 2 loop operation, the P-8 trip at 70 percent RAT ED THERMAL POWER with a loop stop valve closed will prevent the minimum value of the DNBR from going below 1.30 during normal operational tr ansients and antici-pated tr ansients with 2 loops in operation. ST EAM GENERATOR WATER LEVEL: The Steam Generator Water Level low-low trip provides core protection by preventing oper ation with the stean gener ator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be suf ficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system. O BEAVER VALLEY UNIT 2 B 2-7
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS , 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMDrTATION SETPOINT S BA SES STEAM /FEEDWATER FIDW MISMATCH AND LOW STEAM GDiERATOR WATER LEVEL: The Steam /Feedwater Flow Mismatch in coincidence with a Stears Generator low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip cettings and thereby enhance the over all reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Iow trip. The Steam /Feedwater Flow Mismatch portion of thts trip is activated when the steam flow exceeds the fe edwater flow by
>1.55 x 106 pounds / hour. The Steam Generator Low Water level portion of the trip is activated when the water level & ops below 25 percent, as indicated by the narrow range instrument. These trip values include suf ficient allow-ance in excess of normal operating values to preclude spurious trips, but will initiate a reactor trip before the stean generators are dry. T her e for e ,
! the required capacity and sta ting time requirements of the a2xiliary feed- ! water pumps are reduced and the resulting thermal transient on the Re actor Coolant System and steam generators is minimized. UNDERVOLTAGE AND UNDERFREQUDICY - REACTOR C001 ANT PUMP BUSSES: O The Undervoltage and Underfrequency Reactor Coolant Pump bus i trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump. The speci-fied setpoints assure a reactor trip signal is generated before the low flow l tr ip setpoint is reached. Time delays are incorporated in the underfrequency ; l and undervoltage trips to prevent spurious reactor tr ips from momentary 1 electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall'not exceed 0.9 seconds. For under frequency, the delay is set so that the time required for a signal to reach the reactor trip breakers af ter the under frequency trip setpoint is reached shall not exceed 0.3 seconds. i O BEAVER VALLEY UNIT 2 B 2-8
- m _ _ - -
l.
- 2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYST EM SETTINGS
- REACTOR TRIP SYSTEM INSTRUMDrTATION SETPOINT S a
l l BA SES TURBINE TRIP: 7 A Turbine Trip causes a direct reactor trip when operating above P-9. Each of the turbine trips provide turbine protection and reduce ! the severity of the ensuing transient. No credit was taken in the accident 1 analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the 1 Reactor Protection System. l ) SAFETY INJECTION INPITT POR ESF: { If a reactor trip has not already been generated by the reac-
] tor protective instrumentation, the ESF automatic actuation logic channels !
4 will initiate a reactor trip upon any signal which initiates a safety injec-l' tion. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels whict initiate a safety injection signal are shown in Table 3.3-3. REACTOR COOLANT PUMP BREAKER POSITION TRI_P: The Reactor Coolant Pump Breaker Position Trips are antici-patory trips which provide reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. These trips are blocked below P-7. The open/close position trips assure a reactor trip signal is generated before the low flow trip set point is reached. No credit was taken in the accident analyses for operation of these trips. Their functional ' capability at the open/close position settings is required to enhance. the 4 over all reliability of the Reactor Protection System. ;
- REACTOR TRIP SYSTEM INTERIDCKS The Reactor Trip System Interlocka perform the following l functions
- l 1 P-6 Above the setpoint P-6 allows the manual block of the Sour ce Range ;
i reactor trip and de-energizing of the high voltage to the
; detectors. Below the setpoint Source Range level trips are i automatically reactivated. and high voltage restored.
4 P-7 Above the setpoint P-7 automatically enables reactor trips on low 1 flow or coolant pump breaker open in more than one primary coolant loop, reactor coolant pump bus undervoltage and underfrequency,
~
pr essur izee low pressure and pressurizer high level. Below the l setpoint the above listed trips are automatically blocked. I i ) i j O BEAVER VALLEY UNTT 2 B 2-9 l l i t l l l t l _-,___,..__.___.--_m_. ., , , - - , _ . , . . . _ . . _ . _ _ _ _ _ , _ . . _ _ _ , _ _ _ _ . _ . _ _ - _ , - _ _ . , _ , . _ - , _ , , , - _ . , _ - - . _ _ . -
2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM f,ETTINGS 2.2 LIMITING SAFETY SYST EM SETTING S REACTOR TRIP SYST EM IN STRUMDITATION SET POINT S BA SES P-8 Above the setpoint P-8 automatically enables reactor trip on low flow in one or more primary coolant loops. Below the setpoint P-8 automatically blocks the above listed trip. P-9 Above the setpoint P-9 automatically enables a reactor trip on turbine trip. Below the setpoint P-9 automatically blocks a reactor trip on turbine trip. P-10 Above the setpoint P-10 allows the manual block of the Intermediate Range reactor trip and the low setpoint Power Range reactor tr ip; and automatically blocks the Source Range reactor trip and de-energizes the Source Range high voltage pcwer. Below t he setpoint the Intermediate Range reactor erip and the low setpoint Power Range reactor tr ips are automatically reactivated. Pr avides input to P-7. P-13 Pr ovides input to P-7. O O BEAVER VALLEY UNIT 2 B 2-10 1
O SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS O l l O
3/4.0 APPLICABILITY LIMITING CONDIT ION NR OPERATION g () -- Co.:pliance with the Limiting Conditions for Operation 3.0.1 contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operat ion, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements tre not met within the specified time int e rv al s . If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:
- 1. At le as t H OT STANDBY within the next 6 hours , {
?. At least HUT SHUTDOWN within the following 6 hours, and
- 3. At le as t COLD SHUTDOWN within the subsequent 24 hours .
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting p) (_ Condition for Operation. Exce pt io ns to these requirements are stated in the individual Specifications. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not preve nt passage through OPERATIONAL MODES as required to comply with ACTION requirements. Except io ns to these requirements are stated in the individual Specifications. 3.0.5 When a system, subsystem, train, component or device is l , determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoper ab le , it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, pr ovided: 1) its corresponding normal or emergency power source is OPERABLE; and 2) all of its redundant system (s), subsystem (s), tr ain(s ), component (s) , and device (s) are OPERABLE or likewise satisfy the requirements of this Specification. Unless both O O BEAVER VALLEY UNIT 2 3/4 0-1 t l
3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION conditions 1) and 2) are satisfied within 2 hours, action shal~1 be initiated
< to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply, by placing it, as applicab le , in:
j 1. At le as t HOT STANDBY within the next 6 hours ,
- 2. At least HOT SHUrDOWN within the fo11owing 6 hours, and
- 3. At least COLD SRUTDOWN within the subsequent 24 hours .
This Specification is not applicable in MODES 5 or 6. i i
- SURVEILLANCE REQUIREMl!NT S 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless othe'rwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, and 1
l () b. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance int erval . I Failure to perform a Surveillance Requirement within the 4.0.3 ' < specified time interval shall constitute a failure to meet the OPERABILTTY requirement s for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. Surveillance Requirements do not have to be performed on inoperable equipment. i , 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated I surveillance interval or as otherwise specified. l f ! ($) BEAVER VALLEY UNIT 2 3/4 0-2 i 1
- m. _ . _ _ . . - _
_ _ . . _ _. _ _ . _ . _ _ . ~ ..__ ._ _ _ . . - _ , _ _ _ _ - . . _ _ . _
3/4.0 APPLICABILITY m SURVEILLANCE REQUIREMDIT S 4.0.5 Surveillance Requirements for inservice inspection and tes t ing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Se ct ion 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Code Required frequencies for and applicable Addenda terminology for per forming inservice inspect ion
! inservice inspection and testing activities and testing activities Weekly At least once per 7 days Monthly At least once per 31 days
() Quarterly or every 3 months Semiannually or every 6 months At least once per 92 days At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2. are applicable to the above required frequencies for performing inservice ir.4pection and testing act ivit ies .
- d. Performance of the above inservice ins pect ion and tes ting act ivit ies shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
O 3/4 0-3 BEAVER VALLEY UNIT 2 l l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL s SHUTDOWN MARGIN - T avg
> 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 1.77% Ak/k.
APPLICABILITY: MODES 1, 2* , 3, and 4. ACTION: With the SHUTDOWN MARGIN < l.77% Ak/k, immediately initiate and continue boration at > 30 gpm of 7000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 1.77% Ak/k: ,
- a. Within one hour af ter detection of an inoperable control rod (s) and at
(} least once per 12 hours thereaf ter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
- b. When in MODES 1 or 2**, at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.
- c. When in MODE 2***, at least once during control rod withdrawal and at least once per hour thereafter until the reactor is critical,
- d. Prior to initial operation above 5 percent RATED THERMAL POWER af ter [
each fuel loading, by consideration of the factors of "e" below, with , the control banks at the maximum insertion limit of Specification t 3.1.3.6. ;
* -See Special Test Exception 3.10.1 :
** With Ke f f > 1.0
*** With K,gg < 1.0 lBEAVERVALLEYUNIT2 3/4 1-1 l t
t
3/4.1 REACTIVITY CONTROL SYST EMS l 3/4.1.1 BORATION CONTROL ( SitTTDOWN MARGIN - T,y > 200* F SURVEILLANCE REQUIREMDIT S
- e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1. Reactor coolant system boron concentr at ion,
- 2. Control rod position,
- 3. Reactor coolant system average temper ature , .
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon conce ntr ation, and
- 6. Samarium concentr ation.
- f. The Reactor Coolant System shall be borated to at le as t the cold shutdown boron concentration prior to manually blocking the low Pressurizer Pressure Safety Injection Signal and shall remain at this boron concentration or greater at all times during which this signal l V is blocked.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstr ate agreement within + 1% Ak/k at least once per 31 Ef feetive Full Power Days (EFPD). This comparison shalI consider at le as t those f actors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Ef fective Full Power Days af ter each fuel loading. O BEAVER VALLEY UNIT 2 3/4 1-2
3/4.1 REACTIVITY CONTROL SYST DIS - 3/4.1.1 BORATION CONTROL EITTDOWN MARGIN - T < 200*F LIMITING CONDITION -FOR OPERATION i i i 3.1.1.2 The SHlTTDOWN MARGIN shall be > 1.0% Ak/k. APPLICABILITY: MODE 5. ACT ION : With the SHITTDOWN MRGIN < l.0% Ak/k, inusediately initiate and continue boration at > 30 gpm of 7000 ppe boric acid solution or equivalent until the required SitTTDOWN MARGIN is restored. SURVEILIANCE REQUIREMDiT S II I I 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be > 1.0% Ak/k:
- a. Within one hour af ter. detection of an inoperable control rod (s) and at
. O. le as t once per 12 hours thereafter while the rod (s) is inoper able. If the inoperable control rod is immovable or untrippable , the SHITTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the inunovable or untrippable control rod (s).
- b. At least once per 24 hours by consideration of- the following factors:
- 1. Reactor coolant system bi on concentr ation,
- 2. Control rod position,
- 3. Reactor coolant system average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentr ation, and
- 6. Samarium concener ation.
O BEAVER VALLEY UNIT 2 3/4 1-3
- 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL -
. BORON DILITTION LIMITING CONDITION EDR OPERATION 3.1.1.3 The flow rate of reactor coolant tirough the core shall be
> 3000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.
APPLICABILITY: All MODES. ACTION: With the flow rate of reactor coolant tk ough the core ! < 3000 gpe, immediately suspend all oper ations involving a reduction in boron concentr ation of the Reactor Coolant System. SURVEILIANCE REQUIRDtENT S i 4.1.1.3 The flow rate of reactor coolant tirough the core shall be O determined to be > 3000 gym prior to the start of ~ and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:
- a. Verifying at least one reactor . coolant pump is in operation, or
- b. Verifying that at le as t one RHR pump is in operation and supplying
> 3000 gpa through the core.
1 a O i BEAVER VALLEY UNIT 2 3/4 1-4 I
3/4.1 REACTIVITY CONTROL SYST EMS 3/4.1.1 BORATION CONTROL MODERATOR TEMPERATURE COEFFICIDir (MIC) LIMITING CONDITION FOR OPERATION - 3.1.1.4 The moderator temper ature coef ficient (MTC) shall be:
- a. Less positive than 0 x 10-4 delta k/k/*F,
- b. Less negative than -3.9 x 10-4 delta k/k/*F at RAT ED THERMAL POWER.
APPLICABILITY: MODES l' and 2*f. ACrION: With the moderator temperature coef ficient outside any one of the above limits, be in HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENT S
- 4.1.1.4.1 The MTC shall be determined to be within its limits by 1
confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits. 4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER :onditions during each fuel cycle:
- a. Prior to initial operation above 5 percent of RAT ED THERMAi, POWER, after each fuel loading,
- b. At any THERMAL POWER, within 7 EFPD af ter reaching a RAT ED THERMAL POWER equilibrium boron concentration of 300 ppe.
I
- i With Keff > 1.0.
# See Special Test Exception 3.10.4.
C. BEAVER VALLEY UNIT 2 3/4 1-5 l l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITIONiPOR OPERAf10tV 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (Tavg) s hall be > 541
- F whe n t he r e actor is critical.
APPLICABILITY: MODES I and 2*#. ACT ION : With a Reactor Coolant System operating loop temperature
) to within its limit within (T,yg) < 541*F, 15 minutes restore or be in (T,yhBY within the next 15 minutes.
HOT STAN SURVEILIANCE REQUIREMDIT S i s. . ii . .. . 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be > 541* F. A U a. Within 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant system T ,yg is less than 551*F with the t (T,yg) deviation alarm not reset.
- See Special Test Exception 3.10.3.
# With Keg g > 1.0.
O BEAVER VALLEY UNIT 2 3/4 1-6 l l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH S - SIUTDOWN LIMITING CONDITION FOR OPERATION 3.1. 2 .1 As a minimum, one of the following boron injact ion flow paths shall be OPERABLE:
- a. The flow path from the boric acid storage system via a boric acid transfer pump to a charging pump to the Reactor Coolant System if only the boric acid storage tank in Specification 3.1.2.7.a is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the r efueling water stor age tank in Specification 3.1.2.7.b is OPERABLE.
APPLICABILITY: MODES 5 and 6. ACT ION : With none of the above flow paths OPERABLE, suspend all oper at ions involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. SURVEILIANCE REQUIRDIENT S i 4.1.2.1 At least one of the at ove required flow paths shall be demons te ated OPERABLE:
- a. At least once per 7 days by:
- 1. Cycling' each testable power operated or automatic valve in the flow path ttrough at least one complete cycle of full travel.
- 2. Verifying that the temperature of the heat tr aced port ion of the flow path is > 65'F when a flow path from the boric acid tanks is used and the ambient air temperature of the Auxiliary Building is
< 65*F.
- b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct pos it io n.
O , BEAVER VALLEY UNIT 2 3/4 1-7
3/4.1 REACTIVITY CONTROL SYFTEMS 3/4.1.2 BORATION SYST EMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the folluwing boron injection flow paths shall be OPERABLE:
- a. The. flow path from the boric acid tanks via a boric acid trans fer pump and one charging pump to the Reactor Coolant System, and
- b. The flow path from the refueling water storage tank via one charging pump to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3, and 4 ACT ION :
- a. With the flow path from the boric acid tanks inoperable, restore the inoper able flow path to OPERABLE status within 72 hours or .be in at least HOT STANDBY and borated to a SHITTDOWN MARGIN equivalent to at le ast 1 percent 4k/k at 200*F within the next 6 hours; restore the O flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
- b. With the flow path from the refueling water storage tank inoperable ,
restore the flow path to OPERABLE status within one hour or be in at least HCTT STANDBY within the next 6 hours and in COLD SIUTDOWN within the following 30 hours. SURVEILIANCE REQUIREMENT S 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
t l O BEAVER VALLEY UNIT 2 - 3/4 1-8 1 I I 1 I 1 1 I
3/4.1 REACTIVITY CONTROL SYSTEMS ; 3/4.1.2 BORATION SYFT EMS FIDW PATH S - OPERATING f I SURVEILIANCE REQUIREMENT S j i 4 i
- 2. Verifying that the temperature of the he at tr aced portion of the flow path from the boric acid tanks is > 65'F when the ambient air temperature of the Auxiliary Building is < 65'F.
t
- b. At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
i
- c. At least once per 18 months during shutdown by cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle
! of full travel.
4 t l i O 1 j .l 4 i i i i i i O BEAVER VALLEY UNIT 2 3/4 1-9 l l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYST EMS
-O
(/ CHARGING PUMP - SHITTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 As a minimum, one charging pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus. APPLICABILITY: MODES 5 and 6. ACTION : With no charging pump OPERA 3LE, suspend all operations involving CORE ALTERATION S or positive reactivity changes until one charging pump is restored to OPERABLE status. SURVEILIANCE REQUIREMENTS 4.1.2.3 T$e above required charging pump shall be demonstrated OPERALLE at least once per 31 days by:
- a. Starting (unless already operating) the pump from the control room,
- b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2437* psig, and
- c. Verifying pump operation for at le as t 15 minutes .
bl~ l r
)
i
' :j
*To be verified during Pre-Operati$nal Testing
~
BEAVER VALLEY UNIT 2 *3/4 1-10 4'
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION MR OPERATION 3.1.2.4 At le as t two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION : With only one charging pump OPERABLE, restore at le as t two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 48 hours or be in COLD SHITTDOWN within the following 30 hours. SURVEILIANCE REQUIREMENT S 4.1. 2 . 4 Two charging piimps shall be demonstrated OPERABLE at le as t once per 31 days on a STAGGERED TEST BASIS by:
- a. Starting (unless already operating) the pump from the control room,
- b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2437* psig, and
- c. Verifying pump operation for at le as t 15 minutes .
d i t t I f 1 *To be verified during Pre-Operational Testing O BEAVER VALLEY UNIT 2 3/4 1-11
, , m , -- - - . - - ---,,y - - , , - , g , - - - , -,, . > - , ---,+--r, +w- - - - - , .-n-- n -
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYST EMS BORIC ACID TRANSFER PUMPS - SilfrDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 One boric acid tr ans fer pump shall be OPERABLE and capable of being powered frcm an OPERABLE emergency bus if only the flow path through the boric acid transfer pump of Specification 3.1.2.1.a is OPERABLE. APPLICABILITY: MODES 5 and 6. ACT ION : With no boric acid trans fer pump OPERABLE .as required to complete the flow path of Specification 3.1.2.1.a suspend all oper ations involving CORE ALTERATIONS cr positive reactivity changes until at le as t one boric acid trans fer pump is restored to OPERABLE status. SURVEILIANCE REQUIREMDiT S 4.1.2.5 The above required boric acid tr ans fer pump shall be demonstrated OPERABLE at least once per 7 days by:
- a. Starting (unless already operating) the pump from the control room,
- b. ' Verifying, that on recirculation flow, the pump develops a discharge pressure of > 107* psig, and
- c. Verifying pump operation for at le as t 15 minutes .
4
*To be verified during Pre-Operational Testing BEAVER VALLEY UNIT 2 3/4 1-12
f a 3/4.1 REACTIVITY CONTROL SYSTEMS l 3/4.1.2 BORATION SYST EMS BORIC ACID TRAN SFER_ PUMPS -_ OPERATING LIMITING CONDITION FOR OPERATION ; i 3.1.2.6 At least one boric acid transfer pump in the boron injection flow path required by Specification 3.1.2.2.a shall be OPERABLE and capable _ of being powered from an OPERABLE emergency bus if the flow path ttrough the i boric acid pump in Specification 3.1.2.2.a is OPERABLE. i APPLICABILITY: ! i MODES 1, 2, 3, and 4. ACTION : : With no boric acid tr ans fer pump OPERABLE, res tore at le as t i one boric acid transfer pump to OPERABLE STATUS within 72 hours or be in at leas t HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent- to 1 percent ok/k at 200*F; restore at least one boric acid ! tr ans fer pump to OPERABLE status within the next 7 days or be in COLD SHITTDOWN within the next 30 hours. I SURVEILLtJCE REQUIRD(ENT S O 4.1.2.6 At le as t the above required boric acid pump shall be i demons trated OPERABLE at least once per 7 days by, i
- a. Starting (unless already operating) the pump from the control room, i
- b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 107* psig, and '
l
- c. Verifying pump operation for at leas t 15 minutes. -
l l t
*To be verified during Pre-Operational Testing O BEAVER VALLEY UNIT 2 3/4 1-13 !
r
3/4.1 REACTIVITY CONTROL SYSTEMS l 3/4.1.2 BQRATION_ SYST EMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the followi borated water sources shall 4 be OPERABLE:
- a. A boric acid storage system with:
- 1. A minimum contained volume of 2315 gallons,
- 2. Between 7000 and 7700 ppm of boron, and
- 3. A minimum solution temperature of 65'F.
l b. The refueling Gater storage tank with: l 1. A minimum contained volume of 217,000 g allo ns , s j 2. A minimum boron' concentration of 2000 ppm, and
- 3. A minimum solution temperature of 45'F.
APPLICABILITY: MODES 5 and 6. 1 ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status. SURVEILIANCE REQUIREMl!NTS 4.1.2.7 The above required borated water source shall be demonserated OPERABLE:
.a. At least once per 7 days:
- 1. Verifying the boron concentr ation of the water , '
- 2. Verifying the water level of the tank, and I
l i
- 3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.
j O b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside ambient air temperature
- is < 45'F.
BEAVER VALLEY UNIT 2 3/4 l'I'
3/4.1 REACTIVITY CONTROL SYSTEMS a 3/4.1.2 BORATION SYSTEMS BORATED WATER SOURCES - OPERATING [ LIMITING CONDITION FOR OPERATION 2- - . __== 3.1.2.8 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: I
- a. A boric acid storage system with:
- 1. A minimum contained volume of 13,390 gallons ,
) 2. ~Between 7000 and 7700 ppm of boron, and
- 3. A minimum solution temperature of 65'F. I I
- b. The refueling water storage tank with: ,
! 1. A minimum contained volume of 859,248 gallons of water ,
t
- 2. A minimum boron concentration of 2000 ppa, and J
- 3. A minimum solution temperature of 45*F.
1 APPLICABILTTY: i 1 . MOD E S 1, 2 , 3 , and 4. i
, ACT ION : i
- s. With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STAND- l BY and borated to a SHUTDOWN MARGIN equivalent to at le as t 1 i percent delta k/k at 200*F within the next 6 hours; restore the boric i j acid storage system to OPERABLE status within the next 7 days or be in !
! COLD SHUTDOWN within the next 30 hours. l b. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within ! the next 6 hours and in COLD SIUTDOWN within the following 30 hours. 4 l 1 r 1 i i l i i ! BEAVER VALLEY UNIT 2 3/4 1-15 i
+
1
?
I_. -__ - . _ - . . _ - - . _ . - _ . _ - _ . -- _ . . _ _ _ _ _ _ _ _ _ _ __ . . - _ _ -
- 3/4.1 REACrIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS BORAT ED WATER SOURCES - OPERATING l
l SURVEILIANCE REQUIREMDiT S _ _sp 4.1.2.8 Each borated water source shall be demons trated OPERAB.LE:
- a. At least once per 7 days by:
- 1. Verifying the boron concenteation in each water source ,
- 2. Verifying the water level in each water source , and
- 3. Verifying the boric acid storage system solution temperature when it is the source of borated water,
- b. At least once per 24 hours by verifying the RWST temperature when the RWST ambient air temperature is- < 45'F.
I O
\
- O BEAVER VALLEY UNIT 1 3/4 1-16
) i 1 3/4.1 REACTIVITY CONTROL SYSTEMS J 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 1 () GROUP HEIGHT LIMITING CONDITION FOR OPERATION i ? 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (indicated position, as determined in accordance with Specification 3.1.3.2) corresponding to their respective group demand counter position. 4 APPLICABILITY: i MODES 1* and 2*. i ACTION:
' a. With one or more full length rods inoperable due to being immovable as j
a result of excessive friction or mechanical interference or known to be untrippable, determine that .the SHUTDOWN MARGIN requirement of
- i Specification 3.1.1.1 is satisfied within I hour and be in NOT STANDBY f within 6 hours.
I b'. With more than one full length rod inoperable or misaligned from the i group demand counter position by more than + 12 steps (indicated {' () position determined in accordance with SpecTfication 3.1.3.2), be in HOT STANDBY within 6 hours.
- c. With one full length rod trippable but inoperable due to causes other
, than addressed by ACTION a, above, or misaligned from its group demand I counter position by more than + 12 steps (indicated position j detensined in accordance with Specification 3.1.3.2), POWER OPERATION j may continue provided that within one hour either: r 1 i 1. The rod is restored to OPERABLE status within the above alignment j requirements, or 1 2. The rod is declared inoperable and the remainder of the rods in the i group with the inoperable rod are aligned to within + 12 steps of 4 the inoperable rod while unintaining the rod sequence and insertion limits of Figures (3.1-1) and (3.1-2); the THERMAL POWER level shall be restricted pursuant to Specification (3.1.3.6) during subsequent operation, or i
- See Special Test Exceptions 3.10.2 and 3.10.4 l
1 i i i I O
! BEAVER VALLEY UNIT 2 3/4 1-17 i
i. f s
.-e-.~,~,-,,- ~, , -y.,, ,,y,,.gm_.,.,, _ ,,...,.,_.,,,,.,,,,,,,_,.,.,,,.,n. -
,. - , - - . .-._,,...n , _ -,_ . ,,.v.,_.J
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVEABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION EUR OPERATION
- 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
- a. The THERMAL POWER level is reduced to less than or equal to 75 percent of RATED THERMAL POWER within the hour and, within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85 percent of RATED THERMAL POWER.
- b. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours.
- c. A power distribution map is .ibtained from the movable incore N
detectors and Fq (Z) and FAH are verified to be within their limits within 72 hours,
- d. A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that-the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 Each shutdown and control rod not fully inserted in the core shall be determ.ned to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. 4.1.3.1.2 Each full length rod position shall be determined to be + 12 steps of the associated group demand counter by verifying the individual rod position at least once per 12 hours except during intervals when the Rod Position Deviation monitor is inoperative, then verify the group position at least once per 4 hours. l O BEAVER VALLEY UNIT 2 3/4 1-18
TABLE 3.1-1 ACCIDENT ANALYSES REQUIRINC REEVALUATION IN THE EVENT OF AN INOFERABLE FULL LENGTH ROD 1 O Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolent From Small Ruptured Pipes or from Cracked Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) Major Secondary Systems Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) O 1 1 I l
)
O BEAVER VALLEY' UNIT 2 3/4 1-19 i
3/4.1 REACTIVITY CONTRJL SYSTEMS 3/4.1.3 MOVABLE C0KTROL ASSEMBLIES POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown and control rod position indication system and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within + 12 steps. APPLICABILITY: MODES 1 and 2*. ACT ION :
- a. With a maximum of .one rod position indicator per bank inoperable e ither:
- 1. Determine the position of the non-indicating rod (s) indirectly by the movable incore detectors at le ast once per 8 hours and immediately af ter any motion of the non-indicating rod which exceeds 24 steps in one direction since the las t determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50 percent. of RAT ED THERMAL POWER O~ within 8 hours,
- b. With a maximum of one demand position indicator per bank inoperable either:
- 1. Verify that all rod position indicators for the af fected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at le as t once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50 percent of RAT ED THERMAL POWER within 8 hours.
SURVEILIANCE REQUIREMENTS 4.1.3.2.1 Each rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position l1 indication system agree within 12 steps at le as t once per 12 hours except during time intervals when the Rod Position Deviation Monitor is ino per ab le , then compare the demand position indication system and the rod position e indication system at le as t once per 4 hours.
*With the reactor trip system breakers in the closed position. 5 BEAVER VALLEY UNIT' 2 3/4 1-20
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES POSITION INDICATION SYSTEM - SIUTDOWN 1 LIMITING CONDITION EUR OPERATION 3.1.3.3 The group demand position indicators shall be.0PERABLE and capable of determining within + 12 steps the demand position for each j shutdown or control rod not fuT1y inserted. APPLICABILITY: MOD ES 3* , 4* , and 5* . ACTION : With less than the above required position indicator (s)
- OPERABLE, inusediately open the reactor trip system breakers.
SURVEILIANCE REQUIREMDET S 4.1.3.3 Each of the above required group demand position indicator (s) shall be determined to be OPERABLE by movement of the associated control rod p at le as t 10 steps in any one direction at least once per 31 days when the () reactor coolant system pressure is greater than 400 psig. 1 J
- With the reactor trip system breakers in the closed position.
I V BEAVER VALLEY UNIT 2 3/4 1-21
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE COffrROL ASSEMBLIES "oo o"o" ""e + O LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be f 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
- a. T,yg > 541* F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODE 3. ACTION :
- a. With the drop time of any full length rod determined to exceed the
! above limit, restore the rod . drop time to within the above limit pr ior l to proceeding to MODE 1 or 2.
- b. With the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided THERMAL POWER
- O is restricted to i 65 percent of RAT ED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed. ;
i ) 91RVEILLANCE REQUIREMENT S I i~ 4.1.3.4 The rod drop time of full length rods shall be demons tr ated the ough me asur ement prior to reactor eriticality: I a. For all rods following each removal of the reactor vessel head. I i ' b. For specifically af fected individual rods following any maintenance on j or modification to the control rod drive system which could af fect t he i drop time of those specific rods. )
- c. At le as t once per 18 mont hs .
i i r i a BEAVER VALLEY UNIT 2 3/4 1-22 t i
3/4.1 REACTIVITY CONTROL SYSTEMS $ r i I 3/4.1.3 MOVABI.E CONTROL A SSEMBLIES j l SitTTDOWN ROD IN SEltTION LIMIT l i
. LIMITING CONDITION FOR OPERATION [
l _
! 3.1.3.5 All shutdown rods shall be fully withdrawn. l t
I L APPLICABILITY: >
! r
- i. MOD E S 1
- a nd 2*# .
t ,' ACTION: : t With a maximum of one shutdown rod not fully withde awn, except f
! for surveillance testing pursuant to Specificat ion 4.1.3.1.1, within one hour ;
i either : l l i
- a. Fully withdr av the rod, or l
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
t i SURVEILIANCE REQUIREMENT S 4.1.3.5 Each shutdown rod shall be determined to be fully withdr awn by use of the group demand counters, and verified by the rod position indicators. t ,
- a. Within 15 minutes prior to withdrawal of any rods in control banks A, !
' B, C. or D during an approach to reactor criticality, and I b. At le as t once pe r 24 hour s t her e af ter . j i i
}
i l
- See Special Test Exception 3.10.2 and 3.10.4
! I
# With keft > 1.0 l,
U 1 F !. I l O BEAVER VALLEY UNIT 2 3/4 1-23 l i l \ l l l ) i
3/4.1 REACTIVITY CONTROL SYST EMS 3/4.1.3 MOVABLE C0trTROL ASSEMBLIES CONTROL R0_D IN SERTION LIMIT S 1 l LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as i shown in Figures 3.1-1 and 3.1-2. i APPLICABILITY: MODES 1* and 2*d. ACT ION : I With the control banks inserted beyond the above insert ion { i limits, except for surveillance testing pursuant to Spec ificat ion 4.1.3.1.1, either: a I a. Restore the control banks to within the limits within 2 hours, or i l
- b. Reduce THERMAL POWER within 2 hours to lose than or equal to that fraction of RATED THERMAL POWER which is allowed by the group j position using the above figures, or
- c. Be in at le as t HUT STANDBY within 6 heurs.
- O ,
S1RVEILIANCE REQUIREMENT S , l _ i -
-- .=
! 4.1.3.6 When the Rod Insert ion Limit M'iitor is OPERABLE, the deviation between the position indicated by the individual rod position ins tr ume nt channel and the position indicated by the corresponding group demand indication shall be checked manually for each rod at least once per 24 ' hours. When the Rod Insertion Limit Monitor is (noperable, the deviation between indicated positions shall be checked manually at least once per 4 hours. lI i
- See Special Test Exception 3.10.2 and 3.10.4 !
# W it h k, g g > 1.0 l i
t !O BEAVER VALLEY INIT 2 3/4 1-24 l ] i 1 1 9
I FIGURE 3.1-1 ROD CROUP INSERTION LIMITS VERSUS THERMAL POWER THREE LOOP OPERATION v
)
(FULLY WITHOR AWN) [' f I t I - I l l I ' . i ' I:. ._t
.. , , i . . . . . . . . . .u
...: % ='; =- - a f . . .. *.. L L:. m
. .o i :*
.a L . I I -
gp - 1:. - I . l . .. : . l . I:.as-l:=; =;p r !.: n...:l. , 200 ~ b 5.~..-!. M " l
-- - .i . i lNI.k
'-# 1 hl' - uI I:i _!iE-l:#.lY:'Il= .a
.;\ l --
1 ::. ,- ::.: . . : t' BANK C ::_ L_..;:; . . ._j r_ . : :=--
- = :- :t : I . l . .f - l:
[ 'J- I ' _1 l [- l l l l! !?:bh:[,. ~
.I..l
. I 1 I i .
. . t==.:.:. =_:: L- - .: f :. =: : .
150 f + -! - l i:: I=y._I- r- 1 c t:7g=y.L:t= c-- - l
. . : a _ . .__ :.2y_ __: ;_.:::: : -= ...,. = .-*L: ::::..
i E. ::-
- . =. +' . . .. .
g = ._i.i. . -
.=
_ m. :=;: :. .s_ - t... -,:.-
.3..=. ;: g_: ..:.;-=, -- . - . _: p. .; . .. .=. . .- : ,: - . .: r.:
~
- r m ,_. . .
- l: .:.,p:. , ,;, ,,
. __ j ...;._.au : :
_ .=.: - , __ ._ - ,- _ . p; ._. _. . ; t_._ . .=. _.::
; ., i . . ., .
t
- :.l .. *. =
.=: -
=__ _- _ h.. :-.. . ... . , . . i . =.::; :- =.:=. . l:=.:. . . ;, === = ._ [- 2 .=.: .: =i =:#.==_: l-_.=-:-
y _.._ ....'._ f. L. .. .I.._ _~.; _- :; _u.l.: ... _- . h
...; ..-: - .: - r 5
t . . . _
. .. . f. S AN K O --
_ g. i .; - l 1 :__-p. - . i. ._- -:._
- 4. _. . . .n. . _m. =. . ,.=._= . p=.. = , .___._=_.- . . . p__. . , .
-- - ,- :: = g . . . . . . '".. _ u =- t . . _ ,
. L._ :_. . r .=::. . , = :_: = , . _ .1.= w __ I: =.- :. : = . :.
9_ . . .
=e_ -I'*....._.r.._.__,-
_ .._ _. - =
=..t p;
- i_w
- l. P-
= = t . .::.:- =
- . .J . = .= . . ;._. :: _._.:-
..__.=.j . . t..c j.=. v._ .
. . . . .g. v..- . - {._" 3_. l . ._
.p.....
_a _ .__. . A.1=:..- --l =_ =.=_=
-- :. . . . . :. . : r
.:._: .:.a:.::ug:._: __ =. g. =_-- .. ..
= _ l ._. r. .-
.a : . . - e =. i . .. . G: . g: ... .. .: _.- :_ -
i F ty.d[ '-
'l../;;l -l3 j...:[ :j-Qi: {.Olff.*: :f ,- .:.-{ :F . .
if I l' I l - l T -l[ i l i i.- E.*.1:ilE E Of;I'*-
.:i l f. .. l , l l
.l l... ., . l . ...g.
.. j...
l , l , l L. l- t. ,
. u : .t.-
g
- .J_:- [ .
-_ } l t .j : l c. .I.N!._";._ .
J.. i l-0 .2 .4 .8 .3 1.0 (PULLY INSERTIDI PRACTION CP R ATED THERMAL POWER
\'
REAVER VALLEY UNIT 2 3/4 1-25
6 FIGURE 3.1-2 ROD GROUP INSERTION LIMIT S VERSUS THERMAL POWER TWO LOOP OFERATION 4 1 a r l (PULLY Wl?NCR AWNI 23 .. _. , . . . . .. .. ._ 3, 9 .
- # . 5 E R I_: : M 7 i ~. E .:i5 3 - _... . . . .... _
l j
. . . +-'8. _ .3. 2. . _ .*. . _ _: .".". :
I . . :- f SAN K C :. ._ - : ;.=:n :. . .- . :
.=
- I ::
- 7. *. ".. .
- 00 _ . . _ _
l ::.:#. J : .2-..J ..._ =......_.._ =:=.. . .. . =
- .=--
1 I
; _==. ._- . .. .:.==.=: - . _u -
? _
j 92..'"."...~. .. _ . .Z.* 6-f l;; "I'="* .- I.
~ ~ - .. - -a.
TSO ...._. qst . .
==
.--=
- i. .e _ - _ . . . _ . . . . .
b :; ;. ,._._.= ...
. - = , m,e .. _ ..._ . . . -
d f k ..
- _SANKO .,, . _'* =' . ;
z ... --- 3 -- _- . 1
. g.g. .
z.... _.- g g .. 3 z .. . . ). 4 . 1- . _ .. . .. l _-- .a.. t c'= - '. - , < 90 __ 1 - [ a , i i P'""'"" ._ 1 t
= ._ .
.i, I -,4 1 o-_ _ . ._ ._. . _ . . _ . _._ . ; 0.4 i I O 0.2 0.s * (PULLY IN88RTEDI : i i PR ACTICN CP R A710 THEFMAL 'CWER I t, f I ! I I
*' l i I i i i ,
i s 1 I I f
- BEAVER VALLEY UNIT 2 3/4 1-26 !
, t , i i l l > 6 1 5 , f (I
.--___,,.m...-.._._ _ _ _ . -. _ __. - _. - .- _ ._ _ __, . ,. - .... . - , _ , - . . _ - _ , . , _ , , . . . . . _ _ , , _ _ . . - _ . _
3/4.2 POWER DIS *RIBUTION LIMIT S 3/4.2.1 AXIAL FLUX DIFFERDiCE (AFD) O LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERDICE (AFL) shall be maintained within a + 7 perce nt target band (flux dif ference units) about the t arge t flux dif7 fe re nce . l} i APPLICABILITY: MODE 1 ABOVE 50 PERCDfT RAT E THERMAL POWER *. ACTION :
- a. With the indicated AXIAL FLUX DIFFERDICE outside of the 1 7 perce nt l}
target band about the target flux dif ference and with THERMAL POWER:
- 1. Above 90 percent cf RAT E THERMAL POWER, within 15 minutes:
- a. Either restore the indicated AFD to within the t ar ge t band lim-its, or
- b. Reduce THERMAL POWER to less than 90 percent of RAT E THERMAL POWER.
- 2. Between 50 percent and 90 percent of RAT E THERMAL POWER:
- a. POWER OPERATION may continue provided:
- 1. The indicated AFD has not been outside of the + 7 per ce nt l}
target band for more than 1 hour penalty deviatioiE cumulative during the prevLous 24 houra and
- See Special Test Exception 3.10.2 O BEAVER VALLEY UNIT 2 3/4 2-1
I 3/4.2 POWER DISTRIBUTIQN LIMIT S 1 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) O- LIMITING CONDITION FOR OPERATION
- 2. T he indicated AFD is within the limits shown on Figure 3.2-1.
Ot herwise , reduce THERMAL POWER to les s than 50 percent of ; RAT ED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoint s to < 55 percent of RAT ED THERMAL POWER within the next 4 hours,
- b. Surveillance testing of the Power Range Neutron Flux Channels may be per formed pursuant to Specificat ion 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-
- 1. A total of 16 hours operation may be accumulated with the AFD outside of the targe t band during this testing witho ut penalty deviation.
- b. THERMAL POWER shall not be increased above 90 percent of RAT ED THEIDtAL POWER unless the indicated AFD is within the + 7 percent t arge t band l7 and ACTION a.2.a.1 above, has been satis fied.
I
- c. THERMAL POWER shall not be increased above 50% of RAT ED THERMAL POWER unless the indicated AFD has not bee n out s ide of the + 7 per ce nt t ar- l11 get band for more . than 1 hour penalty deviation cuveulative during the '
previous 24 hours. 51RVEILIANCE REQUIREMET S 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its i limits dur ing POWER OPERATION .above 15 per ce nt of RAT ED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1. At le as t once per 7 days when the ' AFD Monitor Alarm is OPERABLE, and
- 2. At le as t once per hour for the firat 24 houre af ter restoring the AFD Monitor Alarm to OPERABLE status.
L. Monitoring and logging the indicated AXIAL FLUX DIFFERWCE for each OPERABLE excore channel at leas t once per hour for the first 24 hours and at le as t once per 30 minutes thereaf ter , when the AXIAL FLUX DIF-FERENCE Monitor Alarm is inoperable. The logged values of the indi- ' cated AXIAL FLUX DIFFERENCE shall be as sumed to exist d ar ing the interval preceding each logging. O BEAVER VALLEY UNIT 2 3/4 2-2 ; 1
- 3/4.2 POWER DISTRIBUTION LIMIT S 3
3/4.2.1 AXIAL FLUX DIFFERENCE ( AFD) 1 l g SURVEILIANCE REQUIREMDiTS 4.2.1.2 The indicated AFD shall be considered outside of its + ~7 percent t arge t l band when at le as t 2 of 4 or 2 of 3 OPERABLE excore channels are indica- )' ting the AFD to be outside the t arge t ba nd . POWER OPERATION outside of the + 7 per ce nt target band shall be accumulated on a time basis of: lj j a. One minute penalty deviation for each one minute of POWER OPERATION i outside of the tar ge t band at THERMAL POWER levels equal to or above 50 percent of RAT ED THERMAL POWER, and
- b. One-half minute penalty deviation for each one minute of POWER OPERA-l TION outside of the targe t band at THERMAL POWER levels below 50 l per ce nt of RAT ED THERMAL POWER.
1 4.2.1.3 The t ar get flux dif ference of each OP ERABLE excore channel shall be determined by measurement at least once per 92 Ef fective Full Power Days. The provisions of Specification 4.0.4 are not applic ab le .
; r l 4.2.1.4 e
The t ar ge t flux dif ference shall be updated at le as t once per 31 Ef fec- { 1 t ive Full Power Days be either de termining the t ar ge t flux dif fer ence ! ! pursuant to 4.2.1.3 above or by line ar interpolation between the most recently measured value and 0 percent at the end of the cycle life. The provisions of Specifiestion 4.0.4 are not ap plicab le . 4 l 1 : I 1 t l i } } I i i : ! i
- l I
!o ! BEAVER VALLEY UNIT 2 3/4 2-3 i
, i 6
i _ _ _ _ _ - _ . _ _ - - _ _ _ . _ _ - _ - - . . _ _ _ . _ . . _ . _ _ _ . _ _ , _ . _ _ _ _ ~ . _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER i I _ . . .- .=;- -- - - - -- - == - : . . . : - - = ; . ,; : . : . . . . = _- : . . :- . .:
.. ,= - -. _- .= = _.- .-- ._ -: :.
j =_= 5_0 ; g_ . _=. ,_. _ s: = C=6 Qi$
=52
-. s, = a =_ __-
1C0 s _UNACCIFAELij(.11,3Cl. (11,5Cl=UNACCI? TABLE
~
' ~ CPEMATtCN _ CP era ~'C N
=.Sm*=.*
1 J
~
80 -- ;
=- _
i i 4 _ ! -3ACC17T A8LiiCP!M ATION l SQ . i i M (.31'.50) (31.501' : E i I i l l l l j ' l ' I I 40 - , f I l I - i i l, . ._ _ _ . .)
! 9 i '
'*Z ;
.- ll . M . . . , _ . , _
.. i
.. . - . - ' = ,
0 ' l 50 40 30 20 10 0 10 20 cc 40 Sc j i l
' . Pt.UX CIFF1MENCE (1ll P. i i
i I I. i P i f i. .i t t 1 i t i i BEAVER VALLEY UNIT 2 3/4 2-4 i i i I i t 1 I i ! t
- _ _ - . ~ . - . . _ _ _ . . - - - _ _ , . _ . _ _ , _ _ _ _ _ . _ _ , _ _ _ _ _ , _ . _ , - _ _ . _ . _ _ _ m-. . _ _ , .
3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.2 HEAT FLUX HOT GANNEL FACTOR - F9 (Z) O v LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships: Fq (Z) < [2.18] [K(Z)] for P>0,5 P F9 (Z) < [(4.36)] [K(Z)] for P<0.5 where P = THERMAL POWER RAT ED THERMAL POWER and K(Z) is the function obt ained from Figure 3.2-2 for a given core height lo cat io n. APPLICABILITY: MODE 1 A CT ION : With F (Z) exceeding its limit: 9
- a. Reduce THERMAL POWER at le as t 1 pe r ce nt for e ac h I pe r ce nt F0 (Z) exceeds the limit within 15 minutes and s imili ar ly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a tot al of 72 hours; subsequent POWER OPERATION may proceed provided the Over-power delta T Trip Setpoint s have been reduced at le as t 1 pe r ce nt !I for each I per ce nt Fn (Z) exceeds the limit. The Overpower delta T Trip Setpoint redbetion shall be performed with the reactor l 7 suberitical.
- b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may the n be increased J pr ovided Fo(Z) is demons tr ated thr ough incore mapping to be within its limit.
SURVEILIANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not ap plicab le . 4.2.2.2 F shall be evaluated to determine if F9 (Z) is wYlhin its limit by: O BEAVER VALLEY UNIT 2 3/4 2-5
3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.2 HEAT FLUX HOT OlANNEL FACTOR - F9 (Z) O SURVEILIANCE REQUIREMDrTS
- a. Using the movable incore detectora to obtain a power distribution map at any THERMAL POWER greater than 5 percent of RAT ED THERMAL POWER.
- b. Increasing the measured F compone nt of the power distribution map by 3 percent to accouEl for manufacturing tolerances and further increasing the value by 5 percent to account for measurement unce rt ant ies .
- c. Comparing the F computed (F*I ) obtained in b, above to-
*I
- 1. The F xy limits for RAT ED THERMAL POWER xy(FRTP) for the
; appropriate naasured core planes given in e and f below, arti i 2. The relationship:
F xy
= F xy
[1+0.2(1-P)]
"here F is the limit for fractional THEINAL POWER operation expressed as a function of F y and P is the fraction of RAT ED THERMAL POWER at which F was me as ur ed .
- d. Remeasuring F,y according to the following schedule:
- 1. When F xy is greater than the F limit for the appropriate xy measured core plane but less than the F y r e lat ionship, additional power distribution maps shall be taken and F compared to L U F and F :
xy xy
- a. Either within 24 hours af ter exceeding by 20% of RAT ED THERMAL POWER or greater, the THERMAL POWER at which F was las t 1 xy determined, or O BEAVER VALLEY UNIT 2 3/4 2-6 i
4 L
3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F9 (Z) O SURVEILIANCE REQUIRDfENT S
- b. At least once per 31 EFPD, whichever occurs first.
MP
- 2. When the F,Cy is less than or equal to the F limit for the appropriate measured core plane, additional power distribution maps shall be taken and F compared to F xy and F at le as t xy xy once per 31 EFPD.
- e. The F xy limit for Rated Thermal Power (F xy ) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.14 limits of e, above, are not applicable in the following
- f. The coreF,[ane p regions as measured in percent of core height from the bottom of the fuel:
- 1. Lower core region from 0 to 15 percent , inclusive.
- 2. Upper core region from 85 to 100 percent , inclusive.
- 3. Cr id plane regions at 17.8 + 2 per ce nt , 32.1 + 2 pe r ce nt ,
46.4 + 21, 60.6 + 2%, and 7T.9 + 2%, inclusive.
- 4. Core plane regions within + 2 percent of core height
(+ 2.88 inches) about the Tank demand position of the bank "D" control rods. C L
- g. With F y exceeding F,7 , the of facts of F on Fq(Z) shall be evaluated to determine ifq F (Z) is within its limit.
4.2.2.3 When nF (Z) is measured pursuant to Specification 4.10.2.2, an overall nessured F (Z) shall be obt ained from a power distribution map and increased Sy 3 percent to account for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty. i O BEAVER VALLEY UNIT 2 3/4 2-7 : l 4
Figure 3.2-2 K(Z) - NORMALIZED F (Z) - AS A FUNCTION OF CORE HEIGHT l () 1.5000
; 1.2500 i
l , 1.0000 i 1 I 0.7500 . i i 0.5000 TOTAL Fq(Z) ] 2.180
- CORE HEIGHT K(Z) l
() 0.2500 0.000 6.000 1.000 1.000 11.044 0.937
- 12.000 0.688 i
- \ 0.0 0.0 2.0000 4.0000 6.0000 8.0000 10.0000 12.0000 4
4 CORE HEIGHT (FEET) l i i I l l i f i BEAVER VALLEY UNIT 2 3/4 2-8 i s
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.3NUCLEARENTHALPYHOTCHANNELFACTOR-F$g O LIMITING CONDITION FOR OPERATION 3.2.3 F H shall be limited by the following relationship: N kHi1.55[1+0.2(1-P)] where P = THERMAL POWER RATED THERMAL POWER APPLICABILITY: I MODE 1 ACTION: N Withg(Z)exceedingitslimit:
- a. Reduce THERMAL POWR to less than 50 percent of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to i 55 percent of RATED THEkMAL POWER within the next 4 hours, N
- b. Demonstrate through in-core mapping that F is within its limit O within 24 hours af ter exceeding the limit or redu e THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 2 hours, and
- c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION may proceed provided N
that F is demonstrated throuRh in-core mapping to be within its limit at a nNinal 50 perceSt of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75 percent of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours af ter attaining 95 percent or greater RATED THERMAL POWER. O lBEAVERVALLEYUNIT2 3/4 2-9
- 3/4.2 POWER DISTRIBlTTION LIMIT S 3/4.2.3 NUCLEAR DrTHALPY HOT CHANNEL FACTOR -FjH O
SURVEILIANCE REQUIRDIENTS 4.2.3.1 [ shall be determined to be within its limit by using moveable iNore detectors to obtain a power distribution map:
- a. Prior to operation above 75 percent of RAT ED THERMAL POWER af ter each fuel loading, and j
- b. At least once per 31 Ef fective Full Power Days.
4.2.3.2 The measured F of 4 shall be increased by 4 percent for meNuremen.2.3.1 t above, uncertainty. O O BEAVER VALLEY UNIT 2 3/4 2-10 i
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO O LIMITING CONDITION FOR OPERATION
- 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE I above 50 percent of RATED THERMAL POWER *. ACTION: I
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but ,
i1.09:
- 1. Within 2 hours: ;
i i
- s. Either reduce the QUADRANT POWER TILT RATIO to within its limit,
! or $
1 I b. Reduce THERMAL POWER at least 3 percent for each 1 percent of I indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip [ Setpoints within the next 4 hours, f 2. Verify that the QUADRANT POWER TILT RATIO is within its limit ;
! within 24 hours af ter exceeding the limit or reduce THERMAL POWER !
I to less than 50 percent of RATED THERMAL POWER within the next 2 , hours and reduce the Power Range Neutron Flux-High Trip setpoints 1 to i 55 percent of RATED THERMAL POWER within the next 4 hours.
- 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above i 50 percent of RATED THERMAL POWER may proceed provided that the ,
- QUADRANT POWER TILT RATIO is verified within its limit at least ;
l once per hour until verified acceptable at 95 percent or greater RATED THERMAL POWER. i i ' l i i
- See Special Test Exception 3.10.2.
k I i i I O l t j lBEAVERVALLEYUNIT2 3/4 2-11 i
--.---,,,-n e.-n-- - - - - - -- ,. .n-.. , . - - - - -- - - --,~---v. -- - . - , , , - , -- - - - . , ------e, -
3/4.2 POWER DISTRIBUTION LIMIT S i 3/4.2.4 QUADRANT POWER TILT RATIO
- O LIMITING (DNDITION PUR OPERATION f'
I b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to ' mis aligrunent of either a shutdown or control rod: l
- 1. Reduce THERMAL POWER at le as t 3 per ce nt for each I percent of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minut es .
j 2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within the next 2 hours j, ~ , and reduce the Power Range Neutron Flux-Migh Trip Setpoints to j f 55 percent of RAT ED THERMAL POWER within the next 4 hours. l a f 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at le as t once per hour until verified acceptable at 95 percent or greater RAT ED THERMAL POWER. I
- c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misaligrunent of either a shutdown or control 4
O rod: o
- 1. Reduce THERMAL POWER to less than 50 percent of RAT ED THERMAL POWER ;
< within 2 hours and reduce the Power Range Neutron Flux-High Trip '
Setpoints to i 55 percent of RAT ED THERMAL POWER within the next 4 i hours. j.
- 2. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above ,
i 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TIII RATIO is verified within its limit at le as t
~
< once per hour until verified at 95 percent or greater RAT ED THERMAL ; I POWER. p 4 4 > i i i
- I
- I I
I i i f i i O BEAVER VALLEY UNIT 2 3/4 2-12 i i
\
.., - ,.,_ -..-. - - - . - - - - .)
3/4.2 POWER- DISTRIBlTTION LIMIT S ! l 3/4.2.4 QUADRANT POWER TILT RATIO ' O SURVEILIANCE REQUIREMENT S l 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50 percent of RAT ED THERMAL POWER by
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE.
- b. Calculating the ratio at le ast once per 12 hours during steady state operation when the alarm is inoperable.
- c. Using the movable detectors to determine the QUADRANT POWER TILT RATIO at least once per 12 hours when one Power Range Channel is inoper ab le ,
and THERMAL POWER is > 75 percent of RAT ED THERMAL POWER. l t i f i O l l I T i L i i I i O- BEAVER VALLEY UNIT 2 3/4 2-13 ; i
l l l 3/4.2 POWER DISTRIBUTION LIMrr S l l 3/4.2.5 DNB PARAMETEF S i LIMITING (X)NDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on T able 3.2-1.
- a. Reactor Coolant System T avg
- b. Pressurizer Pressure {
- c. Reactor Coolant System Total Flow Rate APPLICABILITY:
MODE 1 ACTION : With any of the above parameters exceeding its limit, r es to r e the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 4 hours. SURVEILIANCE REQUIREMDrr S (]) ==tr - 4.2.5.1. Each of the parameters of Table 3.2-1 shall be verified to be indicating within their limits at le as t once per 12 hours. 4.2.5.2 The Reactor Coolant System total flow rate shall be Jetermined to be wtthin its limit by measurement at least once per 18 months. L I I BEAVER VALLEY UNIT 2 3/4 2-14
O O O TABLE 3.2-1 DNB PARAMFT ERS LIMIT S 2 loops In Oper at ion. 3 Ioops In and Isolated loop Par amet er Oper at ion Stop Valves Closed Reactor Coolant System T , < 581*F < 570*F Pressur izer Pressure > 2220 psia * > 2220 psia
- Reactor Coolant System Total Flow Rate > 265,500 gpm > 187,800 gpm
- Limit not applicable during either a THERMAL POWER r amp increase in excess of 5 percent RATl!D THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10 percent RAT I!D THEllMAL POWER.
i E l i l-i- a f BEAVER VALLEY UNIT 2 3/4 2-15 i' I
3/4.3 IN STRUMLNTATION 3/4.3.1 REACTOR TRIP SYSTEM IN STRUMENTATION LIMITING CONDITION MR OPERATIGN 3.3.1.1 As a minimum, the reactor trip system instrumentation channels l7 and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in T able 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACrION: As shown in T able 3.3-1. SURVEILLANCE REQUIREMENT S 4.3.1.1.1 Each reactor trip system instrumentation channel shall be l demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL WNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1. 4.3.1.1.2 The logic for the interlocks shall be demons tr ated OPERABLE l{ during the at power CHANNEL FUNCTIONAL TEST of channels af fected by interlock oper at io n. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel af fected by interlock operation. 4.3.1.1.3 The REACIOR TRIP SYST EM RESPONSE TIME of each reactor trip l function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that bot h logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at leat once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the '*rotal Number of Channels ' column of T able 3.3-1. O BEAVER VALLEY UNIT 2 3/4 3-1
-~ ~ , . , . _ -....~,.,,,----y.. .,_,-,e.n, , - , , . , , . . . , , . . , , . , _ . . , . , , , - - , ,e e. - - - -, , - . , , - , . .
(~)%
% (%J~1 TABLE 3.3-1 react 0R TRTO SYSTEH INStRUNENTATION
- MINIMUM TOTAL NO. CHANNELS CHANN ELS AP PLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OP ERABLE MODES ACT ION _ 4 i 1. Manual Reactor Trip 2 1 2 1, 2, 3*, 4*, 5* 12
- 2. Power Range, Neutron Flux 2 3 2 2
- a. High Setpoint 4 2 3 1,1),
1( 2 2 l 1! j-
- b. Low Setpoint 4
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2
. High Positive Rate 4 2 3 1, 2 2
- 4. Power Range, Neutron Flux, High Negative Rate
) 1(I), 2, 3*, 4*, 5* 3
- 5. Intermediate Range, Neutron Flux 2 2 1
- 6. Source Range, Neutron Flux (Below P-10) g ,
- a. St ar tup 2 1 2 2(2), 3*, 4*, 5* 4 2 0 3, 4, and 5 5
- h. Shutdown 1
- 7. Overtamperature Delta T 2
- a. Three Loop Operation 3 2 2 1, 2 1** 2 1, 2 9
- b. Two Loop Operat ion 3
' 8. Overpower Delta T 2 1, 2 2
- a. Three Loop Operation 3 2 1** 2 1, 2 9
; b. Two Loop Operation 3 2 2 1, 2 7 4
- 9. Pressurizer Pressure-Low 3 l (Above P-7)
B EAVER VA LLEY UN IT 2 - 3/43-2 i f s
..-,, _-,#-,c- _... ,.-,,,,-,_.,- -
., ,,e,--- -e._, ..,.,_-m--,-. ~ ~. ,, -,,,,-- .-,. ,_,- m , -
--,..mc. .....,,-,_m- ,---,. m._w-n,, .. . - , , _ , . , , - - -
.== -. . -
O O O TABLE 3.3-1 REACTOR TRIP SYST FN INSTRUMENTATION MININUM TOTAL NO. CHANN ELS CHANN ELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACT ION
- 10. Pressur izer Pressure--High 3 2 2 1, 2 7
- 11. Pressur izer Water leve l--H igh 3 2 2 1, 2 7 (Above P-7)
- 12. Loss of Flow - Single Loop 3/ Loop 2/ loop in 2/ loop in 1 7 (Above P-8) any oper- any oper-ating loop at ing loop
- 13. Loss of Flow - Two Loops 3/ Loop 2/ loop in 2/ loop each 1 7
( Above P-7 and below P-8) two oper- oper at ing sting loops loop
- 14. Steam Gener ator Water Level--Low-Low. 3/ Loop 2/ loop in 2/ loop in 1, 2 7 (Loop Stop Valves Open) any oper- each oper-ating loops ating loop
- 15. Steam /Feedwater Flow Mismatch and 2/ loop / level 1/ loop-level 1/ loop-level 1, 2 7 Low Steam Generator Water Level and coincide nt and 2/ loop-flow with 2/ loop-flow ,
mismatch 1/ loop-flow mismatch or mismatch in 2/ loop-level same loop a nd 1/ loop-flow mismatch
- 16. Undervoltage-Reactor Coolant Pumps 3-1/ bus 2 2 1 7 (Above P-7)
- 17. Under frequency-Reactor Coolant Pumps 3-1/ bus 2 2 1 7 (Above P-7)
BEAVER VALLEY UN IT 2 3/43-3
O O O ' TABLE 3.3-1 REACTOR T R I P SYST EM IN ST RUM ENT AT ION MIN IMUM TOTAL NO. CHANN ELS CH ANN ELS APPLICA3LE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE __ MODES _ ACT ION
- 18. Turbine Trip (Above P-9)
- a. Emergency Trip Header Low Pressure 3 2 2 1 7 l1
- b. Turbine Stop Valve Closure 4 4 4 1 8
- 19. Safety Injection Input from ESF 2 1 2 1, 2 1
- 20. Reactor Coolant Pump Breaker 1/ breaker 2 1/ breaker 1 11 Position Trip per oper-(Above P-7) .at i ng loop
- 21. Reactor Tr ip Breakers 2 1 2 1, 2, 3*, 4*, 5* 1
- 22. Automatic Trip Logic 2 1 2 1, 2, 3*, 4*, 5* 1
- 23. Reactor Trip System Interlocks l a. Intermediate Range Neutron 2 1 1 2 3 Flux, P-6
- b. Power Range Neutron 4 2 3 1 12 Flux, P-8
- c. Power Range Neutron 4 2 3 1 12 Flux, P-9
- d. Power Range Neutron 4 2 3 1 12 Flux, P-10 i e. Turbine Impulse Chamber 2 1 1 1 12 Pressure, P-13 BEAVER VALLEY UNIT 2 3/43-4 1,
T able 3.3-1 -() TABLE NOTATION
- With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.
** The channel (s) associated with the protective functions derived from the out-of-service Reactor Coolant Loop shall be placed in the tripped .
condit ion. (1) Trip function may be manually. bypassed 'in this mode above P-10. (2) Trip function may be manually bypassed in this mode above ?-6. l1 ACTION STATEMDIT S ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement , be in HOT STANDBY within 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:
- a. Less than or equal to 5 ~ percent of RAT ED THERMAL POWER, place t he inoperable channel in the tripped condition within I hour
- s. and restore the inoperable channel to OPERABLE status within 24 hours af ter increasing THERMAL POWER above 5 percent of RATED THERMAL POWER; otherwise reduce THERMAL POWER to less than 5 percent RATED THERMAL POWER within the following 6 hours,
- b. Above 5 percent of RATED THERMAL POWER, operation may continue provided all of the following conditions are satis fied:
- 1. T he inoperable channel is placed in the tripped condition within I hour.
- 2. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.
- 3. Either THERMAL POWER is restricted to < 75 percent of RKTED -
THERMAL and the Power Range, Neutron FIux trip setpoint is reduced to < 85 percent _ of RAT ED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours. h t t O BEAVER VALLEY UNIT 2 3/43-5 , c
T able 3.3-1 TABLE NOTATION ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 setpoint.
- b. Above P-6 but below 5 percent of RATE THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5 per cent of RAT E THERMAL POWER.
- c. Above 5 percent of RATE THERMAL POWER, POWER OPERATION may continue.
ACTION 4 - With the number of channels OPERABLE one less than :f equired by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below P-6, restore the inoper able channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 setpoint ,
- b. Above P-6, operation may continue.
O ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specificlation 3.1.1.1 or 3.1.1.2, as applicable within I hour , and at least once per 12 hours ther eaf ter . ACTION 6 - Not Applicable. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:
- a. Less than or equal to 5 percent of RAT E THERMAL POWER, place the inoperable channel in the tripped condition within I hour; restore the inoperable channel to oper able status within 24 hour s after increasing THERMAL POWER above 5 percent of RAT E THERMAL POWER; otherwise reduce THERMAL POWER to less than 5 percent of RAT ED THERMAL POWER within the following 6 hours ,
- b. Above 5 percent of RATE THERMAL POWER, place the inoper able channel in the tripped condition within I hour ; operation may continue until performance of the next required CHANNEL FUNCT IONAL T EST .
BEAVER VALLEY UNIT 2 3/4 3-6 i
T able 3.3-1 TABLE NOTATION ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in the tripped condition within I hour; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST. ACTION 9 - With a channel associated with an operating loop inoperable, !I restore the inoperable channel to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours; however, one channel associated with an operating loop may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1. ACTION 10 - Not applicab le. ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within I hour. ACTION 12 - With the number of channels OPERABLE one less than required by. the Minimum Channels OPERABLE requirement, restore the. inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY
. within the next 6 hours and/or open the reactor trip breakers.
CJ '\_J ; BEAVER VALLEY UNIT 2 3/43-7 ,
O O- O TABLE 3.3-2 RFACTOR TRIP SYST FM INSTRUMENTATION _ RESPONSE T IMES FUNCfIONAL UNIT RESPONSE TIME I. Manual Reactor Trip Not Applicable
- 2. Power Range, Neutron Flux (High and Iow Se t po int) f 0.5 seconds *
- 3. Power Range, Neutron Flux, High Positive Rate Not Applicable
- 4. Power Range, Neutron Flux, High Negative Rate f 0.5 seconds *
- 5. Intermediate Range, Neutron Flux Not Applicable
- 6. Sour ce Range, Neutron Flux Not Applicable
- 7. Over temperature Delta T f 4.0 seconds *
- 8. Over power Delt a T Not Applicable
- 9. Pressur izer Pressure--Iow f 2.0 seconds
- 10. Pr essur izer Pressure--High . I 2.0 seconds
!I . Pressur izer Water Level--High Not Appiicable
- Neutron detectors are exempt from response time testing. Response time shall be measured from detector output or input.
of first electronic component in channel. BFAVER VALLEY UNIT 2 3/43-8 .-2__ _ _ _ _ - - _ _ _ _ - - _ _ _ __
-- -. - - . - - . - - _ _ - ~ _ . - - . . . . . - _ _ _ -
l l O O O l TABLE 3.3-2 REACTOR TRIP SYSTDI INSTRUMDrrATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 12. Loss of Flow - Single Loop ( Above P-8) f 1.0 seconds
- 13. ~oss of Flow - Two Loopa ( Above P-7 and below P-8) f 1.0 reconds
- 14. Steam Generator Water Level--Low-Iow f-2.0 seconds
- 15. Steam /Feedwater Flow Mismatch and Low Steam . Not Applicable Generator Water Level
- 16. Undervoltage-Reactor Coolant-Pumps f 1.5 seconds ,
- 17. Under frequency-Reactor Coolant Pumps 1 0.9 seconds
- 18. Turbine Trip 4
- a. Emergency Trip Header Low Pressure Not Applicable Not Applicable l1
- b. Turbine Stop Valve ,
Not Applicable ' 19 Safety Injection Input from ESF l
- 20. Reactor Coolant Pump Breaker Position Trip Not Applicable
- 21. Reactor Trip Breakers '
Not Applicable
- 22. Automatic Trip Logic Not Applicable ,
- 23. Reactor Trip System Interlocks Not Applicable ,
l t SEAVER VALLEY UNIT 2 3/43-9 i 4
, , . _ _ _ _ . _ , _ . _ - _ . __ .y . , _ _ - _
_..r , . - ,
O O O Tr 1LE 4.3-1 l f REALTOR TRIP SYST EN INSTRUMDfrATION SURVEILIANCE REQUIREMENT S CHANN EL MODES IN 1AIICH CH ANN EL CHANN EL FUN CT IONAL SURV EI LLANC E FUNCTIONAL UNIT CH ECK CALIBRATION TEST REQUIR ED i
' . Manual Reactor Tr ip N/A N/A S/U(1) N/A
?ower Range, Neutron Flux
- c. High Setpoint S D(2), M(3) and Q(6) 'M 1, 2
- h. Low Setpoint S N/A S/U(1) 2
- 3. Power Range, Neutron Flux N/A R M 1, 2 High Positive Rate
- 4. Power Range, Neutron Flux, N/A R M 1, 2 High Negative Rate
- 5. Intermediate Range, Neutron Flux S N/A S/U(1), M(7) 1, 2 , 3 * , 4* , 5*
- 6. Source Range, Neutron Flux (Below P-10) N/A N/A S/U(1), M(8) _ 2 , 3* , 4 * , 5 *
- 7. Over temper ature Delta T S R M 1, 2
- 8. Overpower Delta T S R M 1, 2
- 9. Pressur izer Pressure-Low (Above P-7) S R M 1, 2
- 10. Pr essur izer Pressure--H igh S R M 1, 2
- 11. Pr essur izer Water Level--High (Above P-7) S R M 1, 2
- 12. Loss of Flow - Single loop S R M 1
- 13. Loss of Flow - Two Loops S R N/A 1
- 14. Steam Gener ator Water Level--Low-Low S R M 1, 2 BFAVER VALLEY UNIT 2 3/4 3-10
O O O TABLE 4.3-1 REACTOR TRIP SYSTDI INSTRUMENTATION SURVEILIANCE REX)UIREMINTS CHANNEL MODES IN WH ICil CHANNEL CHANN EL FUNCrIONAL - SURV EI LLANC E FUNCrIONAL UNIT CH ECK CALIBRAT ION TESr REQUIRED
- 15. Steam /Feedwater Flow Mismatch and S R M 1, 2 Low Steam Generator Water level ,
- 16. Undervoltage - Reactor Coolant Pumps N/A R M 1 (Above P-7)
- 17. Under frequency - Reactor Coolant Pumps N/A R M 1 (Above P-7)
- 18. Turbine Trip (Above P-9)
- a. Faergency Trip Header Low Pressure N/A N/A S/U(1) I, 2 l1
- b. Turbine Stop Valve Closure N/A N/A S/U(1) I, 2 19 Safety Injection Input from ESF N/A N/A M(4) 1, 2
- 20. Reactor Coolant Pump Breaker Position N/A N/A R N/A Trip
- 21. Reactor Trip Breaker N/A N/A M(5) and-S/U(1) 1, 2, 5*
- 22. Automat ic Tr ip Logic N/A N/A N(5) 1, 2, 5*
- 23. Reactor Trip System Interlocks
- a. P-6 N/A N/A M(9) 1, 2
- b. P-8 N/A N/A M(9) I
- c. P-9 N/A N/A M(9) I
- d. P-10 N/A N/A M(9) I
- e. P-13 N/A R M(9) 1 l.
SEAVER VALLEY UNIT 2 3/4 3-11
TABLE 4.3-1 NCTTAT ION O * - With the reactor trip system breakers closed and the control rod drive ! system capable of rod withdr awal. (1) - If not performed in previous 7 days, j (2) - Heat balance only, above 15% of RATED THERMAL POWER. t l (3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute dif ference > 3 percent. (4) - Manual ESF functional input check every 18 months. (5) - Each train tested every other month. (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) - Below P-10. (8) - Below P-6. (9) - Required only when below Interlock Trip Setpoint. O 1 I 1 i ' l I l I l , 4 BEAVER VALLEY UNIT 2 3/4 3-12 k _ _ - _ . , - _. _ ., , , _ _ . - , , . . . _ _ . ~ . _ . , . _ _ . . _ . , - , _ _ , , . . _ , . -
. - - - - -. - -_ . . _ - _ _ _ _ - _ - - . -- - - - - - .- _. - . _ - ~ _
3/4.3 IN ST RUMENT AT ION
! 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION O LIMITING CONDITION FOR OPERATION
~.
3.3.2.1 The engineered safety feature actuation system instrumentation l} i channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint I column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. l 1 l ACT ION : ;
- a. With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the-Allowable Values column of Table 3.3-4, declare the channel inoperable j and apply the applicable ACTION requirement of Table 3.3-3 until the I
channel is restored to OPERABLE status with the trip setpoint adj us ted ' j consistent with the Trip Setpoint Value. ; l i b. With an engineered safety feature actuation system instrumentation i channel inoperable, take the action shown in Table 3.3-3.
;1 i
i SURVEILLANCE REQUIREMENTS O 4.3.2.1.1 Each engineered safety feature actuation system l} l
! instrumentation channel shall be . demonstrated OPERABLE by the performance of the CHANNEL OlECK, OIANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST j operations during the modes and at the frequencies shown in Table 4.3-2. ;
i I 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE I l during the at power CHANNEL FUNCTIONAL TEST of channels effected by interlock I i operation. The total interlock function shall be demonstr ated OPERABLE at ! ! least once per 18 months during CHANNEL CALIBRATION testing of each channel ;
; af fected by interlock operation.
- 4. 3 .2 .1. 3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF l9 '
j function shall be demonstrated to be within the limit at least once per 18 months. Each test shal1 include at least one logic train such that both logic ) trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N
! is the total number of redundant channels in a specific ESF function as shown l l, in the " Total No. of Channels" Column of Table 3.3-3.
i i
!O ! BEAVER VALLEY UNIT 2 3/4 3-13 I
O O O TABLE 3.3-3 i ENGINEERED SAFETY FEATURE ACTUATION SYSIDI INSTRUMENTATION Total Minimum No. of Channels Channels Applicable Funct ional Unit Channels to Trip Oper ab le Modes Ac t io n I. SAFt.TY INJECTION AND FEEDWATER I SOLAT ION
- a. Manual Initiation 2 1 2 1, 2, 3, 4 18
- b. Automatic Actuat ion Logic 2 1 2 1, 2, 3, 4 13, 36 ]
- c. Containment Pressure - High 3 2 2 1, 2, 3 14
- d. Pressurizer Pressure - Low 3 2 2 1, 2, 3# 14 .
- e. Low Steam 1ine Pressure (Loop Stop Valves Open)
Three Loops Operating 3/ loop 2/ loop ~ 2/ loop 1, 2, 3f 14 any loop any loop Two Loops Operating 3/ loop 2/ loop 2/ loop 1, 2, 3# 15 any operat ing any operating l/ loop loop BEAVER VALLEY UNIT 2 3/4 3-14
O O O TABLE 3.3-3 ENCINEERED SAFETY FEATURE ACTUATION SYSTD( IN ST RUMDfTAT ION Total Minimum No. of Channels Channels Applicab le Funct ional Unit Channe ls to Trip Oper able Modes Ac t io n 1.1 SAFETY INJ ECTION - TRAN SFER FROM INJECTION TO THE RECIRCULATION MODE
- a. Automatic Actuation Logic 2 1 2 1, 2, 3 18 Coincident with Safety Injection Signal
- b. Refueling Water Storage Tank 4 2 3 1, 2, 3 16 Level - Ext reme Low I I
BEAVER VALLEY UNIT 2 3/4 3-15
O O O TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION Total Minimum No. o f Channels Channels Applicable Functional Unit Channels to Trip Operable Modes Action
- 2. CONTAINMENT SPRAY
- a. Manual
- 2 sets I set of 2 switches 2 sets 1, 2, 3, 4 18
- b. Automatic Actuation Logic 2 1 2 1,2,3,4 13
- c. Containment Pressure - High High 4 2 3 1,2,3 16
- 3. CONTAINMENT ISOLATION
- a. Phase "A" Isolation
- 1) Manual 2 1 2 1,2,3,4 18
- 2) From Safety Injection 2 1 2 1,2,3,4 13 Automatic Actuation Logic
- b. Phase "B" Isolation
- 1) Manual
- 2 sets I set 2 sets 1, 2, 3, 4 18 (2 switches / sets)
- 2) Automatic Actuation Logic 2 1 2 1, 2, 3, 4 13
- 3) Containment Pressure - 4 2 3 1, 2, 3 16 High Iligh
- M nual actuation of containment spray is accomplished by -actuating either of two sets (two switches per set). Both switches in a set must be actuated to obtain a manually initiated containment depressurization signal per train.
BEAVER VALLEY UNIT 2 3/4 3-16 I
l l O O O l i TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYST DI INSTRUMDfTAT ION - Total Minimum No. of Channels Channels Applicab le Functional Unit Channels to _Tr ip Ohyrable Modes Action
- 4. STEAM LINE ISOLATION I
i
- c. Manual 1/ loop
- 1/ loop 1/ loop
- 1, 2, 3, 4 18 I b. Automatic Actuation Logic 2 1 2 1, 2, 3, 4 13
- c. Containment Pressure -
Intermediate High High 3 2 3 1, 2, 3 14
- d. Low Steamline Pressure I (Loop Stop Valves Open) p
l Three Loops Operating 3/ loop 2/ loop 2/ loop 1, 2, 3# 14 any loop any loop i l Two Loops Operating 3/ loop 2/ loop 2/ loop 1, 2, 3# 15 l any operating loop any operat ing loop
- e. High Steam Pressure Rate 3/ loop 2/ loop 2/o per at ing 3f f, 4 37 .
any loop loop l2 l
- Additionally, there will be two sets of control switches (two momentary controls per set) on the main control board, j Operat ing either set will actuate all three main steamline stop and bypass valves at the system level .
l BEAVER VALLEY UNIT 2 3/4 3-17
- . . . . - - - , . . . - - - . -_ ._. _ . . _ . _ _ - . - . . . - _ - - - _ _ . - ~_ . - .
O O O TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMDITATION Total Minimum No. of Channels Channels Applicable Funct ional Unit - Channels to Trip Operable Modes Ac t io n
- 5. TURBINE TRIP & FEEDWATER ISOLAT ION
- a. Steam Generator Water 3/ loop 2/ loop in any 2/ loop in each 1, 2, 3 14 Level - High-High P-14 operating loop operating loop
- 6. LOSS OF POWER
- a. 4.16kv Bus
- 1) Loss of Voltage (trip feeder) 1/.. 16kv Bus 1/4.16kv Bus 1/4kv Bus 1, 2, 3, 4 33 1 :
- 2) Loss of Voltage (start diesel) 1/4.16kv Bus 1/4.16kv Bus 1/4kv Bus 1, 2, 3, 4 33
- b. Crid Degraded Voltage 2/4.16kv Bus 2/ Bus 2/ Bus 1, 2, 3, 4 34 (4.16kv Bus)
- c. Crid Degraded Voltage 2/480v Bus 2/ Bus 2/ Bus 1, 2 , 3, 4 34 (480v Bus)
BEAVER VALLEY UNIT 2 3/4 3-18
O O O TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTDI INSTRUMENTATION Total Minimum , No. of Channels Channels Applicab le Funct ional Unit Channels to Trip Oper ab le Modes Ac t io n
- 7. AUXILIARY FEEDWAT ER
- a. Steam Generator Water Level - Low-Low
- i. Start Turbine Driven Pump 3/ steam 2/ steam generator 2/ steam 1, 2, 3 14 ge nerator any steam generator generator i
li. Start Motor Driven Pump 3/stema 2/ steam generator -2/ steam 1, 2, 3 14 ge nerator any 2 steam generator ge ner ator
- b. Undervoltage-RCP Start (3)-1/ bus 2 2 1 14 Turbine Driven Pump i
- c. S. I. St art Motor-Driven Pumps See I above (all S.I. initiating functions and requirements)
- d. Turbine-Driven Pump Discharge (2)-1/ train 1 1 1, 2, 3 18 Pressure Low With Steam Valves Open ( Start Motor-Dr iven Pumps)
- e. Trip of Main Feedwater Pumps 1/ pump 1 1 1, 2, 3 18 St at t Notor-Driven Pumps ,
1: t BEAVER VALLEY UNIT 2 3/4 3-19 l r
- . . . _ . - . ~ . - . _ _ . . . . - . . _ . . . . . . --
}
O O O l
}
l TABLE 3.3-3 4 ! ENGINEERED SAFETY FEATURE ACTUATION SYSTEM IN ST RUMEKTAT ION Total Minimum
- No of Channels Channels Applicab le
- Funct ional Unit Channels to Trip Operable Modes Ac t io n
. I t 8. ESF INTERihCKS l a. Reactor Tr ip, P-4 2 1 2 1, 2 , 3 - 38 1
l b. Pressurizer Pressure, P-il 3 2 2 1, 2, ~ 3 38
- c. Low-Low T avg., P-12 3 2 2 1, 2, 3 38 lI[
1 I y I i ' I i 4 r f t 5 i l i 4 i BEAVER VALLEY UNIT 2 3/4 3-20
)
i 1 j
TABLE 3.3-3 (Continued) TABLE NOTATION
# Trip function may be bypassed in this MODE below P-ll. ## Trip function automatically bypassed above P-11, and is bypassed below P-11 when Safety Injection on low steam pressure is _ not manually by-passed.
ACT ION STAT EMENT S ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the _ following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing in accord-ance with Specification 4.3.2.1.1, provided the other channel is l} operable. ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels:
- a. Below P-ll or P-12, place the inoperable channel in the trip-ped condition within 1 hour; restore the inoperable channel to OPERABLE status within 24 hours af ter exceeding P-11 or P-12; otherwise be in at le as t HOT STANDBY within the following 6 hours.
- b. Above P-11 and P-12, place the inoper able channel in the trip-ped condition within I hour; operation may continue until performtnce of the next required CHANNEL FUNCTIONAL TEST.
ACTION 15 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT SHITTDOWN within the following 12 hours; however, one channel associated with an' operating loop may be bypassed for up to 2 hours for surveillance testing in accordance with Specifica-t ion 4.3.2.1.1. l; ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels:
- a. Below P-ll or P-12, place the inoperable channel in the bypass condition; restore the inoperable channel to OPERABLE status within 24 hours af ter exceeding P-ll or P-12; otherwise be in at least HOT SIUTDOWN within the following 12 hours.
- b. Above P-ll or P-12, demonstrate that' the Minimum Channels OPERABLE requirement is met within I hour; operation say con-tinue with the inoperable channel bypassed and one additional channel may bne bypessed for up to 2 houra for surveillance testing in accordance with Sepcification 4.3.2.1.1 l O BEAVER VALLEY UNIT 2 3/4 3-21
i i 4 ! TABLE 3.3-3 (Continued) ACTION 17 - With less than the Minimum Channels OPERABLE, operation may con-tinue provided the containment. purge and exhaust valves are main-tained closed. ACTION 18 - With the number of OPERABLE Channels one lees than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY vithin the t next 6 hours and COLD SIUTDOWN within the following 30 hours. ACTION 33 - With the number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kv Bus shall be declared inoperable and the ACTION i Statements for Specifications 3.8.1.1 or 3.8.1.2, as appropriate, shall apply, i s' ACTION 34 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed
! until the performance of the next required Channel Functional Test provided the inoperable channel is placed in the tripped condition within I hour.
j ACTION 36 - The block of the automatic actuation logic introduced by a reset
.; of safety injection shall be removed by resetting (closure) of
! the reactor trip breakers within one hour. of an inadvertent ini-tiation of safety injection providing that all trip input signals i have reset due to stable plant condit ions. Otherwise, the requirements of action statement 13 shall have been met. l} ACTION 37 - With the number of OPERABLE channels one less than the Total Number of channels, STARTUP and/or POWER OPERATION may proceed i provided the following conditions are satisfied: 1
- a. The inoperable channel is placed in a tripped condition with-in one hour.
I
- b. The Minimum Channels OPERABLE requirements .is met; however, the inoperable channel may be bypassed for up to 2 hours for i surveillance testing of other channels per specification l 4.3.2.1.1. I I ACTION 38 - With less than the Minimum Number of Channels OPERABLE, within
! one hour determine by observation of the associated permissive annunciator window (s) (bistable status lights or computer checks) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
O BEAVER VALLEY UNIT 2 3/4 3-22
. - . . _ - _ ~_ - _ - _ -.-.-. . - _ _ - . . - . _ . - . , - -_ -
O O O TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYST EM IN ST RUNENTAT ION TRIP SETPOINT S Functional Unit Tr ip Setpoint Allowable Values
- 1. SAFETY INJECTION AND FEEDWATER ISDLATION
- a. Manual Initistion Not Applicable Not Applicable ,
, b. Automat ic Actuat ion Logic Not Appiicable Not Applicable l I
- c. Containment Pressure - High 1 1.5 psig < 2.0 psig l
- d. Pressurizer Pressure - Low > 1,875 psig > 1,866 psig.
l{
- > 510 psig
- e. Steam Line Pressure - Low > 525 psig steam line pressure steam line pressure 1.1 SAFETY INJECTION - TRANSFER FROM INJECTION TO THE RECIRCULATION MDDE
- a. Automatic Actuation Logic Coincident with Not Applicable Not Applicable i Safety . Inject ion Signal
- b. Refueling Water Storage Tank Level - Extreme Low 37'2" 37'2" + 1.9" .l~;
i 4 3 BEAVER VALLEY UNIT 2 3/4 3-23 s
O O O TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOIKT S Funct ional Unit Trip Setpoint Allowable Values
- 2. COMTAINMENT SPRAY
- a. Manual Initiation Not Applicable Not Applicable
- b. Automat ic Actuat ion Logic Not Applicable Not Applicable
- c. Cont ainme nt Pressure - Iligh-Iligh < 8.0 psig
- 3. COVTA INMENT ISOLATION
< 8.8 psig l}
- a. Phase "A" Isolation
- 1) Manual Not Applicable Not Applicable
- 2) From Safety Inject ion Automat ic Not Applicable Actuat ion Logic Not Applicable
- b. Phase "B" Isolat ion
- 1) Manual Not Applicable Not Applicable
- 2) Automatic Actuation Logic Not Applicable Actuat ion Logic '
Not Applicable
- 3) Containment Pressure - Iligh liigh < 8.0 psig < 8.8 psig lJ B EAVER VA LLEY UN IT 2 3/4 3-24
- - - - - - - - - - - _ m --, :u , _ , , _ _ _ , a _4 _. . --4_ am. E n-
~~
O O O
~,
TABLE 3.3-4 (continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM _ INSTRUMENTATION TRIP SETPOINTS 1 Functional Unit Tr ip Setpoint Allowable Values
- 4. Sr EAM LINE ISDLATION
- a. Manual Not Applicable Not Applicable
- b. Automatic Actuat ion Logic Not Applicable Not Applicable
~
' c. Cont ainment Pressure - Intermediate High liigh < 3.0 psig < 3.8 psig lj
- d. Steam Line Pressure - Low > 525 psis > 510 psig l steam line pressure steam line pressure
- e. Iligh Negative Steam Pressure Rate < 100 psi with a time < 110 psi with a time cons t ant > 50 seconds 2 i
cons t ant > 50 seconds , 5. TURBINE TRIP AND FEEDWATER ISDLATION
- a. Steam Generator Water Level - liigh-High < 75% of narrow range < 76% of narrow range Ins tr ume nt span each Ins tr ume nt spa n e ac h stema generator stema generator a
I 'f BEAVER VALLEY UNIT 2 3/4 3-25 1
)
f
. . . . _ _ -. . . - . - - . . , - ... , -, .~ ,,.. .- , ..m. .. ,--._, . . .... .
O O O TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip Setpoint Allowable Values
- 6. LOSS OF POWER
- a. 1) 4.16kv Emergency Bus Undervoltage 75% [jf of nominal bus > 70% of nominal bus (Loss of Voltage) (Trip Feed) volt age with a 1 + 0.1 volt age with a 1 + 0.1 second time delay ~ second time delay ~
- 2) 4.16kv Emergency Bus (Start Diesel) > 70% of nominal bus voltahe,ofnominalbus75% [I{voltage, 20 cycles + 2 cycles 20 cycles + 2 cycles
- b. 4.16kv Emergency Bus Undervoltage 90%[ffofnomianibus > 89% of nominal bus (Degraded Voltage) voltage with a 90 + 5 voltage with a 90 + 5 second time delay - second time delay -
- c. 480v Emergency Bus Undervoltage 90% [ff of nominal bus > 89% of nominal bus (Degraded Voltage) voltage with a 90 + 5 voltage with a 90 + 5 second time delay ~ second time delay -
) BEAVER VALLEY UNIT 2 3/4 3-26
O O O TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMElfrATION TRIP SETPOINT S Functional Unit Tr ip Setpoint A'llowable Values
- 7. AUXILIARY FEEDWATER
- a. Steam Generator Water Level - Low-low > (later) of narrow > (later) of narrow l 7ange ins tr ume nt span 7ange ins tr ume nt spa n each stema generator each steam generator t
- b. Undervolt age - RCP 75%1hofnominalbus voltage
> 70% of nominal RCP bus voltage
.l }
- c. S. I. See 1 above (all S. I. Se tpoint s )
- d. Turbine-Driven Auxiliary Feed Pump Disch pressure > 468 psig Disch pressure > 452 psig l Discharge Pressure with steam inlet valves open with steam inlet valves open
- e. Tr ip of Main Feedwater Pumps Not Applicable Not Applicable l 1
- 8. ESF INTERLOCKS
- a. P-4 N/A N/A
- b. P-11 < 2000 psig < 2010 psig
- c. P-12 > 541*F > 539'F l
i BEAVER VALLEY UNIT 2 3/4 3-27
..e -
r ._yy ,._,_m,__ -
,,w. _
,-. - r,,. _ . -,, ._ - -
TABLE 3.3-5 (D DIGINEERED SAFETY FEATURES ~ RESPONSE TIMES Initiating Signal and Function -Response T ime in Seconds
- 1. Manual
- a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation - Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable
- b. Containment Quench Spray Pumps Not Applicable Containment Quench Spray Valves Not Applicable Containment Isolation - Phase "B" Not Applicable
- c. Containment Isolation - Phase "A" Not Applicable
- d. Control Room Ventilation Isolation Not Applicable
- 2. Contairunent Pressure - High
- a. Safety Injection (ECCS) 1 27.0*
- b. Reactor Trip (from SI) 12.0
- c. Feedwater Isolation i7.0(1)
- d. Containment Isolation - Phase "A" 161.5(4)/115.5(5) {
- e. Auxiliary Feedwater Pumps Not Applicable
- f. Service Water System 1 72.5(2)/211.5(3) 1 BEAVER VALLEY UNIT 2 3/4 3-28
1 i TABLE 3.3-5 (Continued) O ENGINEERED SAFETY FEATURES RESPONSE TIMES Initiating Signal _and Function Response Time in Seconds
- 3. Pressurizer Pressure - Low
- a. Safety Injection (ECCS) i27.0*/12.0#
- b. Reactor Trip (from SI) i2.0
- c. Feedwater Isolation i7.0(1)
- d. Containment Isolation - Phase "A" 161.0(4)/115.0(5) l
- e. Auxiliary Feedwater Pumps Not Applicable
- f. Service Water System i 72.0(2)/211.0(3) l7
- 4. Stean Line Pressure - low
- a. Safety Injection (ECCS) 127.0*/12.0f
- b. Reactor Trip (from SI) 12.0
- c. Feedwater Isolation i 7.0(1)
- d. Containment Isolation - Phase "A" 1 61.0(4)/115.0(5)
- e. Auxiliary Feedwater Pumps Not Applicable
- f. Service Water System i72.0(2)/211.0(3) g.' Steen Line Isolation i 7.0
- 5. Contairunent Pressure - High-High i
- a. Containment Quench Spray 185.5 l
- b. Contairunent Isolation - Phase "B" Not Applicable
- c. Control Room Ventilation Isolation 1(later)
O BEAVER VALLEY UNIT 2 3/4 3-29 l i 1
, -- - - - - - ., - - . . - . - . . - , . , - - - - . , - , . - ..e ,e.,,.,-ve.. , v--,---m--- -
-~
i i TABLE 3.3-5 (Continued) O ENGINEERED SAFETY FEATURES RESPONSE TIMES Initiating Signal and Function Response Time in Seconds 4
- 6. Steam Generator Water Level - High-High i a. Turbine Trip j[ 2.5
- b. Feedwater Isolation j[ 7 0(I )
- 7. Containment Pressure - Intermediate High High i j[7.0
- a. Steam Line Isolation
- 8. Steamline Pressure Rate - High-High j < 7.0
- a. Steam Line Isolation
- 9. Loss of Power
- a. 4.16kv Emergency Bus Undervoltage (Loss of Voltage) j:, 10 s b. 4.16kv and 480v Emergency Bus Undervoltage (Degraded Voltage ;< 10
- 10. (Intentionally blank.)
- 11. Steam Generator Water Level - Low-Low
- a. Motor-driven Auxiliary Feedwater Pumps ** ;[60.0
$ b. Turbine-driven Auxiliary Feedwater Pumps *** j[60.0 t NUT E: Response time for Motor-driven Auxiliary Feedwater Pumps on all S.I. signal starts. i ! ** on 2/3 in 2/3 Steam Generators
*** on 2/3 any Steam Generators
() BEAVER VALLEY UNIT 2 3/4 3-30 t I
, -,- , - , , - . - - - . - . -- -m, yy.-.,, , , _ , ., - , ,rm._, .~_.,--.-_._ep- -.y -,-. .-,--,-, -- .. - . - -- ,- . _
__ ~_
^ w
't TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES Initiating Signal and Function Response Time in Seconds
- 12. Undervoltage RCP
- a. Turbine-driven Auxiliary Feedwater Pumps i60.0
- 13. Trip of Main Feedwater Pumps l}
- a. Motor-driven Auxiliary Feedwater Pump i60.0
- 14. Turbine - Driven Auxiliary Feed Pump Discharge Pressure Low I
- a. Steam Driven Auxiliary Feed Pumps < 60.0 NUTE: Response time for Motor-driven Auxiliary Feedwater Pumps on all O s t ta t c <c -
** on 2/3 in 2/3 Steen Generators *** on 2/3 any Steam Generators O BEAVER VALLEY UNIT 2 3/4 3-31
TABLE 3.3-5 (Continued) O TABLE NOTATION
- Diesel generator starting and sequence loading delays ~ included. Response time limit includes ' opening of' valves to establish SI path and attainment 4
of discharge pressure for centrifugal charging pumps and Low Head Safety Injection pumps. $ # Diesel generator starting and sequence loading delays not included. Of fsite power available. Response time limit. includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
## Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
i I (1) Feedwater system overall response time shall include verification of valve stroke time applicable to the feedwater valves shown for pe ne tr at ions 76, 77, 78, 79, 80, and 83 shown in Table 3.6-1. l (2) Diesel generator starting and. sequence loading delays included. Res po ns e time limit includes attainment of discharge pressure for service water l Pumps. (3) Diesel generator starting and sequence loading delays not included. Response time limit only includes opening of valves to establish the I flowpath to the diesel coolers. (4) Diesel generator starting and sequence loading delays not included. Of fsite power available. Rasponse time limit includes operation of valves / damper s . (5) Diesel generator starting and sequence loading delays included. Res po ns e - time limit inicudes operation of valves / dampers. i O BEAVER VALLEY UNIT 2 3/4 3-32 i k
O O O TABLE 4.3-2 DIGINEERED SAFETY FEATURE ACTUAT ION SYST EM IN ST RUMENTAT ION SURV EILLANCE . RMU IREMENT S Channel Modes in Which Channel Channel Funct io nal Surveillance Functional Unit Check Calibration Test Required I. SAFETY INJECTION AND FEEDWATER ISDLATION
- a. Manual Initiat ion N/A N/A M(1) 1, 2, 3, 4
- b. Automatic Actuation Logic N/A N/A M(2) 1, 2, 3, 4
~
- c. Cont ainment Pressure - High S R M 1, 2, ~ 3
- d. Pressurizer Pressure - Low S R M 1, 2, 3
- e. Steam Line Pressure - Low S R M i , 2, 3 1.I SAFETY INJECTION-TRANSFER FROM INJECTION TO c THE RECIRCULATION MODE
- a. Automatic Actuation Logic Coincident N/A N/A M(2) 1, 2, 3 with Safety Injection Signal
- b. Refueling Water Storage Tank Level - Extreme Low S R M 1, 2, 3 lt
- 2. CONTAINMENT SPRAY
- a. Manual Initiation N/A N/A M(1) 1, 2, 3, 4
- b. Automatic Actuation Logic N/A N/A M(2) 1, 2, 3, 4
- c. Contain Pressure - High High S R H 1, 2, 3 BEAVER VALLEY UNIT 2 3/4 3-33 l
O O O TABLE 4.3-2 (Cont inued) ENGINEERED SAFETY FEATURE ACTUATION SYSTDt INSTRUMDrrATION SJRVEILLANCE REQU_IREMENT_S Channel Modes in Which Channel Channel Funct ional Surve il lance Functional Unit Check Calibration _ Test _ Required _ ,,
- 3. COptrAINMENT I SDLAT ION
- a. Phase "A" Isolat ion '
- 1) Manual N/A N/A M(1) 1, 2, 3, 4
- 2) From Safety Injection Automatic N/A N/A M(2) -1, 2, 3, 4 1
Actuat ion Logic
- b. Phase "B" Isolation
- 1) Manual N/A N/A M(1) 1, 2, 3, 4
- 2) Automatic Actuation Logic N/A N/A M(2) 1, 2, 3, 4 1
Actuat ion Logic l 3) Containment Pressure - High-High S R M 1, 2, 3
- 4. Sr EAM LINE ISOLATION
- a. Manual N /A' N/A M(1) 1, 2 , 3, 4
- b. Automatic Actuation Logic , N/A N/A M(2) 1, 2, 3, 4 i^
- c. Containment Pressure - Intermediate High-High S R M 1, 2, 3 i
- d. Steam Line Pressure - Low S R -M 1, 2, 3
- e. Steam Line Pressure Rate - High S R M 1, 2, 3
- 5. TURBINE TRIP AND FEEDWATER ISDLATION
- a. Steam Generator Water Level - High-High S R M 1, 2 , 3 i HEAVER VALLEY UNIT 2 3/4 3-34
-- . _ - - .- __ _ . -_ . _ _ . . ~ - ...
O O O TABLE 4.3-2 (cont inued) ENGINEERED SAFETY FEATURE ACTUAT ION SYST EM INSTRUMENTAT ION ! SJRVEILLANCE REQUIREMENT S Channel 'fodes in Which Channel Channel Functional Surve il lance Functional Unit Required ,,, _ Check Calibration _ Test 6 LOSS OF POWER
- a. 4.16kv Emergency Bus Undervoltage (Loss of Voltage) N/A R M 1, 2, 3, 4 Trip Feed and Start Diesel
- b. 4.16kv and 480v Emergency Bus Undervoltage N/A R M. 1, 2, 3, 4 (Degraded Voltage)
- 7. AUXILIARY FEEDWAT ER
- a. Steam Generator Water Level - Low-Low S R M 1, 2, 3
- b. Undervoltage - RCP S R M I, 2
- c. S. I. See 1 above (all S. I. surveillance requirements)
- d. Turbine Driven Auxiliary Feed Pump Discharge Pressure Low N/A R R 1, 2, 3 l 1
- e. Trip of Main Feedwater Pumps N/A N/A R 1, 2, 3
- 8. ESF IlfrERLOCKS
- a. P-4 N/A N/A R 1, 2, 3
- b. P-Il N/A R M 1, 2, 3
. c . P- 12 N/A R M 1, 2, 3 BEAVER VALLEY UNIT 2 3/4 .3-35 l
O O O TABLE 4.3-2 (Cont inued) DIGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMDfrATION SJRVEILLANCE RR)UIREMDir S Table Notation (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL RJNCTIONAL TEST at least once per 31 days. (2) Each train or logic channel shall be tested at least every other 31 days. SEAVER VALLEY UNIT 2 3/4 3-36 gy
3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified limits. APPLICABLE: As shown in Table 3.3-6. ACTION:
- a. With a radiation monitoring channel alarm / trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable,
- b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL i CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.
O BEAVER VALLEY UNIT 2 3/4 3-37
_ _ _ _ _ _ _..__.____.__._.._.m . _ . . _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ . . ___..__.__._.-___________.__m . _ _ _ _ _ __ _. O O O , f TABLE 3.3-6 RADIATION MONITORING SYSrEM Minimum Channels Applicable Alarm /Tr ip Measur ene nt Instruments Operable Modes Setpoint Range _ Ac t io n I. ARFA MONITORS ,
- c. Fuel Storage Pool Area (2RMF-RQ202) 1 * <l5mR/hr 10~I to 10 0mR/hr 19
- b. Containment Area (2RMR*RQ206 & 207) 2 1,2,3,4 (later) I to 10 7R/hr 36
- 2. PROCESS MONITORS
- a. Containment
- i. Caseous Activity (XeI33)
RCS Isakage Detection (2RMR*RQI 303) 1 1,2,3,4 (later) 10 4 to 10-luci/cc 20' j ii. Particulate Activity (II3I) RCS teakage Detect ion (2RNR*RQI 303) 1 1,2,3,4 (later) 10 10 to-10-5u ci/cc 20
- b. Fuel Building Vent
- i. Caseous Activity (Kel33)
(2RMF-RQI 301) 1 ** (<2xbackgr ound) 10-6 to 10-lu ci/cc 21 ii. Particulate (II3I) (2RMF-RQI 301) 1 ** (<2 xbackgr ound ) 10-10 to 10-5u ci/cc 21 i l REAVER VALLEY UNIT 2 3/4 3-38 : _ . . _ , . . _ _ _ . _ _ - _ _ . _ _ _ . _ _ _ _ . . _ . . , . _ . _ . _ _ . ~ , _ . _ . , _ _ _ _ . _ _ _ , _ , _ _ _ _ _ _ _ . . _ , _ . _ _ . , _ _ _ . . _ - _ _ . .__ _ _ _ _
(x p () O V V TABLE 3.3-6 (cont inued) RADIAT ION MONITORING SYST EM Minimum Channels Applicable Alarm / Trip Me as ur eme nt I n.7.t r ume nt s Operable Modes Set po int Range Ac t io n
- 3. PDCESS MONITORS (CONTINUED)
- c. Noble Gas & Ef fluent Monitors
- i. Supplementory Leak Collection and Release System (2RMR-RQI301)
- 1) Particulate (1131) I 1,2 ,3 ,4 .(later) 10-10 to 10-6uci/cc 36
- 2) Gaseous (Xel33) I 1,2 ,3 ,4 (later) 10-6 to 10-lu ci/cc 36 I ii. Containment Purge Exhaust (2HVR*RQIl04 A & B)
- 1) Gas (Xel33) 1 6 (<2 xb ackgr ou nd) 10-6 to 10-luci/cc 22 iii. Main Stears Discharge (2MSSS*RQI 101) 1/ SG 1.2 . 3 .4 (later) 10 -2 to 10-3uci/cc 36 RF.WER VALLEY UNIT 2 3/4 3-39
[ l f I L L ) l TABLE 3.3-6 (Continued) l RADIATION MONITORING SYSTEM TABLE NOTATION Action 19 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, pe r f orm area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours. Action 20 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1 Action 21 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the applicable ACTION requirements of Specifications 3.9.12 and 3.9.13. Action 22 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9. Action 36 - With the Number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel (s) to OPERABLE status within 72 hours, or:
- 1) Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and
- 2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days following the event outlining the action taken, the cause of the inoperability and the plans and sched ule tor restoring the system to OPERABLE status, o With fuel in the storage pool or building 00 With irradiated fuel in the storage pool.
BEAVER VALLEY UNIT 2 3/4 3-40
O O O TABLE 4.3-3 RADIATION MONITORING SYSTEM SURVEILLANCE Channel Modes in Which Channel Channel' Funct ional Surve il lance leatrument Check Calihr at ion Test Required I. AREA MONITORS
- a. Fue1 Storage PooI Area (2RMF-RQ202) S R M *
- b. Contairunent Area (2RMR*RQ 206 & 207) S R M 1, 2, 3, 4
- 2. PROCESS MONITORS
- a. Containment
- i. Caseous Activity RCS leakage Detection (2RMR*RQI 303) S R M 1, 2, 3, 4 ii. Particulate Activity I RCS Imakage Detection (2RMR*RQI 303) S R M 1, 2, 3, 4
- b. Fuel Building Vent
- i. Caseous Activity (2RMF-RQI 301) S R . M **
ii. Particulate Activity (2RMF-RQI 301) S R M **
- With fuel in the storage pool or building
** With irradiated fuel in the storage pool BEAVER VALLEY UNIT 2 3/4 3-41 1
. - . - .- .. ._ - . - . . . . . - . . ~ - - - - . _ , - - . - - . . _ . . ~ _ _ . _ . _ . _ - _ . . _ .
. . . _ - . . . - . - ~
d O O O TABLE 4.3-3 (Continued) RADIATION MONITORING SYSTEM SURVEILLANCE Channel Modes in Which Channel Channel Funct ional Surve il lance Instrument Check Calibration Test Required
- 2. PDCESS MONITORS (continued)
- c. Noble Cas Ef fluent Monitors
- i. Supplementary Leak Collection and Release System (2RNR-RQI301) S R M 1, 2, 3, 4 g (i. Cont ainment Purge Exhaust (2HVR*RQIl04 A & B) S R M 6 iii. Main' Steam Discharge (2MSSS*RQI 101) S R M 1, 2, 3, 4 REAVER VALLEY UNIT 2 3/4 3-42
-- ._. - ~ _ _ _- __ _ _- -
3/4.3 IN ST RUMDITAT ION 3/4.3.3 MONITORING IN STRUMDirATION O MovAntE 1NCORE otrEcTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:
- a. At le as t 75 percent of the detector thimbles,
- b. A minimum of 2 detector thimbles per core quadrar.t , and
- c. Suf ficient movable detectors, drive, and readout equipment to map these thimbles.
I APPLICABILITY: When the movable incore detection system is used for:
- a. Recalibration of the axial flux of fset detect ion system, i
- b. Monitoring the QUADRAffT POWER TILT RATIO, or N
j
- c. Measurement of FAH and Fq(Z) .
ACTION: O With the movable incore detection system inoperable , do not j use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applic ab le . SURVEILLANCE REQUIREMDrTS I j 4.3.3.2 The incore movable detection system shall be demonstrated OPERABLE by normalizing each detector output to be used within 24 hours pr(or to its use when required for:
- a. Recalibration of the encore axial flux of fset detection system, or
- b. Monitoring the QUADRAlfT POWER TILT RATIO, or N
- c. Measurement of FAH and Fq(Z).
BEAVER VALLEY UNIT 2 3/4 3-43
3/4.3 IN ST RUMENT AT ION 3/4.3.3 MONITORING IN STRUMERATION SEISMIC IN STRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILITY: At all times ACT ION :
- a. With the number of OPERABLE seismic monitoring instrumencs less than required by T able 3.3-7, restore the inoperable instrument (s) to OPERABLE status within 30 days.
- b. With one or more seismic monitoring ins tr ume nt s inoper ab le for more than 30 days, prepare and submit a Special Report to the Commiss ion pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status,
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not ap plic ab le .
'd SURVEILLANCE REQUIRE;(ENT S 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the per formance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL WNCTIONAL TEST operations at the frequencies shown in Table 4.3-4.
4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status and a CHANNEL CALIBRATION performed within 24 hours following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum, and resultant ef fect upon f acility features important to safety. I) x- l g-BEAVER VALLEY UNIT 2 3/4 3-44
l
\ . TABLE 3.3-7 ,
l SEISMIC MONITORING IN ST RUMENT AT ION Minimus Me asureme nt Ins t rume nt s
- Instrument and Sensor Locations Range Operable
- 1. TRIAXIAL TIME-HISTORY AND RESPONSE SPECTRUM RECORDER lI
- a. Contain Mat el. 692'-11" + Ig 1*
i b. Contsin Operating Floor + 13 1* . el. 767'-10" a
- c. Switchyard + 13 1*
- 2. TRIAXIAL PEAK ACCELOGRAPH i
- a. On top of RMS heat +23 1 l -
exchanger (2R9t-E21A)
- b. Six inch SI pipe + 23
~
1 l j (2 SIS-006-269-1( A) el 741'-5"
- c. MCC*2-E03 el. 755'-6" (PAB) + 53 1
O 3- tarixtt' sirsaic switc= 4 a. Containment est N/A 1* 1 . i
- 4. TRIAXIAL TIME HISTORY ACCELOCRAPH t
l t l l
- a. Steam Generator Support 0-1 3 1 i j cubicle No. I el. 718'-6" ;
I l l
- b. Center of PAB el. 710'-6" 0-13 1
[ i t ! c. McC*2-E03 el. 755'-6" 0-1 1 3 f l !
- d. Containment eat (accelerometer 0-1 g 1 from the Time History Accolograph) t
]
- With reactor control room indication. l 4
1 : I !n U BEAVER VALLEY UNIT 2 3/4 3-45 I ! i I r 1 t
T ABLE 4.3-4 SEISMIC MONITORING IN STRUMENTATION SJRVEILLANCE RMUIREMDIT S j Channel Channel Channel Funct ional Instrument and Sensor Locations Check Calibr at ion Test
- 1. TRIAXIAL TIME-HISTORY AWD I RESPONSE SPECTRUM RECORDER
- a. Containment Mat, el. M* R SA i 692'-11"
- b. Containment Operating M* R SA j floor el. 767'-10" f c. Switchyard M* R SA
- 2. TRIAXIAL PEAK ACCEL 0 GRAPHS
; a. On ' top RHS heat N/A R N/A exchanger (2RHS-E21A)
- b. Six inch SI pipe N/A R N/A (2 SI S-006-269-1( A))
< el. 741'-5"
- c. McC*2 E03 el 755-6" N/A R N/A
- 3. TRIAXIAL SEISMIC SWITCH
- a. Containment met N/A N/A R I
l )
- Except seismic trigger.
I ( i J J l t j ] 1 BEAVER VALLEY UNIT 2 3/4 3-46 1 1 l i
TABLE 4.3-4 (Continued) SEISMIC MONITORING IN STRUMENTATION SURVEILLANCE REQUIREMl!NT S Channel Channel Channel Funct ional Instrument and Sensor Locations Check Calibration Test
- 4. TRIAXIAL TIME-HISTORY 1 ACCEL 0 GRAPH
- a. Steau Generator Support N/A R N/A Cubicle No.1 el. 718'-6"
- b. Center of PAB el. 718'-6" h/A R N/A
- c. MCC*2-E03 el . 755'-6" N/A R N/A
- d. Contairunent Mat N/A R N/A (accelerometer from the Time-History Accolograph)
O O BEAVER VALLEY UNIT 2 3/4 3-47 l ;.
. . _ _ . - - _ _ . _ . _- - ._ - .-_.. . _ _ _ _= - _ ..
3/4.3 IN ST RUMl!NTAT ION 2 3/4.3.3 MONITORING INSTRUMDITATION I MET EOROLOGICAL IN STRUMDfTATION l LIMITING CONDITION FOR OPERATION i 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3'.3-8 shall be OPERABLE. 1 APPLICABILITY:
)
At all times. l 4 ACTION:
- a. With the number of OPERABLE meteorological monitoring channels less than required by Table 3.3-8, suspend all release of gaseous radioactive material from the radwaste gas decay tanks until the y
}
- inoperable channel (s) is restored to OPERABLE status.
i 1 b. With one or more required meteorological monitoring channels I inoperable for more than 7 days, prepare and submit a Special Report i to the coussission pursuant to Specification 6.9.2 within the next 10 i days outlining the cause of the malfunction and the plans for restoring the channel (s) to OrEra::LI: atetus, i
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applic able .
t
!KIRVEILIANCE REQUIREMDrT S 4 '
l i 4.3.3.4 Each of the above meteorological monitoring instrumentation ! channels shall be demonstrated OPERABLE by the performance of the CHANNEL l CHECK and CHANNEL CALIBRATION operations . at the frequencies shown in Table j 4.3-5. . 1 I i i 4 l } ! i 4 i < l0 BEAVER VALLEY UNIT 2 3/4 3-48 l1 t
4 I TABLE 3.3-8
, METEOROLOGICAL MONITORING IN STRUMENTATION Ins trume nt Minimum Minimum Instrument Location Accuracy Operable
- 1. WIND' SPEED
- a. Nominal Elevation 500' +0.5 mph
- Any
- b. Nominal Elevation 150' +0.5 mph
- 2 of 3
- c. Nominal Elevation 35' +0.5 mph
- i 2. WIND DIRECTION
- a. Nominal Elevation 500' +5* Any
) , f b. Nominal Elevation 150' 1* 5 2 of 3 i
! c. Nominal Elevation 35' +5*'
- 3. AIR TEMPERATURE Delta T
- a. Delta T Elevation 500'-35' +0.l*C Any l1 O
- b. Delt a T Elevat ion 150'-35' +0.l'c 1 of 2
' l1 I
l t 1 l i
- Starting speed of anemometer shall be < 1 mph.
i f I () BEAVER VALLEY UNIT 2 3/4 3-49 l.
) TABLE 4.3-5 4 METEOROIDGICAL MONITORING IN SrRUMENTATION
- O SURVEILLANCE REQUIREMEKIS Channel Channel Instrument Check Calibr at ion
- 1. WIND' SPEED
- a. Nominal Elevation 500' D SA
- b. Nominal Elevation 150' D SA
- c. Nominal Elevation 35' D SA
- 2. WIND DIRECTION l
- a. Nominal Elevation 500' D SA
- b. Nominal Elevation 150' D SA
- c. Nominal Elevation 35' D SA
- 3. AIR TEMPERATURE Delta T i
, s. Delta T Elevation 500'-35' D SA l O b. Delta T Elevation 150'-35' D SA I i 1 i j BEAVER VALLEY UNIT 2 3/4 3-50 l[ 1 . 1
----4, - - ._--c.__,.. . ___g,__. _ . _ . _ _ . , , . . , _ . . , , , _ , , _ , _ _ , _ _ . , , _ , , _ ,
l 3/4.3 IN ST RUMDITAT ION
- l. 3/4.3.3 MONITORING INSTRUMENTATION O ar"or s sautoo= 1" =r au"='T^r to"
. LIMITING CONDITION FOR OPERATION
- 3. 3.3. 5 The remote shutdown monitoring instrument ation channels shown l! in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.
) APPLICABILITY: Modes 1, 2, and 3 i
! ACT ION :
f I With the number of OPERABLE remote shutdown monitoring l channels less than required by Table 3.3-9, either: ;
- a. Res tore the inoperable channel to OPERABLE status within 30 days, or
) b . Be in HUT SIUTDOW within the next 12 hour s .
l SURVEILIANCE REQUIREMENT S l 1 , p 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall L v be demonstrated OPERABLE by per formance of the CHANNEL OlECY. and CHANNEL ; CALIBRATION operations at the frequencies shown in Table 4.3-6. { i I I 1 l i i i i 1 I
- t I
i i
! I 3
l 1 i l I
- L iO i
BEAVER VALLEY UNIT 2 3/4 3-51 t ; l i i ! t i 1
TABLE 3.3-9 REMOTE SHlTTDOWN MONITORING IN STRUMENTATION O MEA SUREMl!NT MINIMUM CHANNELS IN ST RUMENT S* RANG E OPERABLE
- 1. Intermediate Range Nuclear Flux 10-II to 10-3 g,p, i
- 2. Intermediate Range Startup Rate -0.5 to +5 dpm 1 l]
- 3. Source Range Nuclear Flux 10* -106 ep, g
- 4. Source Range Startup Rate -0.5 to +5 dpa 1 l}
- 5. Reactor Coolant Temperature
- Hot Leg 0-700*F. 1
- 6. Reactor Coolant Temperature 1
- Cold Leg 0-700*F 1 j
i 7. Pressurizer Pressure 1700-2500 psig 1 1 1 8. Pressurizer Level 0-100% 1
- 9. Steam Generator Pressure 0-1400 psig 1/Stema generator
{ ] 10. Steam Generator Level 0-100% 1/ Steam generator ! 11. RHR Return to Loop Temperature 50*-400*F 1 1 4
- 12. Auxiliary Feedwater Flow 0-400 gpm 1/ Steam generator l
i i
?
i l 1
- Emergency Shutdown Panel BEAVER VALLEY UNIT 2 3/4 3-52 l1 4
I l
, _ . - - - - - _ . - - , . - , , - . - . , - - , .-,...m.. . . . , . - , . . _ . . . . .-,...m. . , ,
TABLE 4.3-6 REMar E SHtTfDOWN MONITORING INSTRUMDITATION SURVEILIANCE O CHANN EL CHANN EL CA LIBRAT ION IN Sr RUMEiT S* CH ECK
,5
- l. Intermediate Range Nuclear Flux M N/A
- 2. Intermediate Range Startup Rate M N/A
- 3. Source Range Nuclear Flux M N/A
- 4. Source Range Startup Rate M N/A
- 5. Reactor Coolant Temperature M R
- Hot Leg
- 6. Reactor Coolant Temperature M R
- Cold Leg
- 7. Pressurizer Pressure M R I
- 8. Pressurizer Level M R
- 9. Steam Generator Pressure M R
- 10. Steam Generator Level M R
- 11. RHR Return to Loop Temperature M R
- 12. Auxiliary Feedwater Flow S/U(I) R NOT E 1: Channe1 Check to be performed in conjunction with surveillance requirement 4.7.1.2.a.9 following an extended outage.
- Emegency Shutdown Panel BEAVER VALLEY UNIT 2 3/4 3-53 l
l 3/4.3 IN ST RUMENT AT ION l 3/4.3.3 MONITORING IN STRUMETT ATION FIRE DETECTION INSTRUMDITATION O LIMITING CONDITION NR OPERATION 3.3.3.6 As a minimum, t he fire detection instrumentation for each fire detection zone shown in Table 3.3-10 shall be OPERABLE. APPLICABILrrY: Whenever equipment in that fire detection zone is required to be OPERABLE. ACT ION: With the number of OPERABLE fire detection instruments less l than required by Table 3.3-10:
- a. Within I hour, establish a fire watch patrol to ins pect the zone (s) with the inoper able instrument (s) at least once per hour, unless the
' instrument (s) is located inside the containment , then ins pect t he {
cont ainme nt at least once per 8 hours or monitor the containment air temperature at le as t once per hour at the locat io ns listed in Specification 4.6.1.5 and
- b. Res tore the inoperable instrument (s) to OPERABLE status within 14 days O,s or, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the act ion taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status,
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILLANCE REQUIREMDIT S 4.3.3.6.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at le as t once per six months by performance of CRANNEL FUNCTIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPElaBLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SRVIDOWN exceeding 24 hours unless performed in the previous six months. 4.3.3.6.2 The NFPA Code 72D Class A supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at le as t once per 6 months. 4.3.3.6.3 The non-supervised circuits between the local panels in Specification 4.3.3.6.2 and the control room shall be demonstrated OPERABLE at le as t once per 31 day . (_) BEAVER VALLEY UNIT 2 3/4 3-54 -
. ~ . -. .
T able 3.3-10 FIRE DET ECTION INSTRUMENT S Instrument Location Minimum Instruments Operable Smoke Fl ame
- 1. Reactor Containment
- Zone 64 (cable penetration area) 10
- Zone 65 (2ER S*P21A ,B) 1
- 2. Control Building
- Zone 11 (comm., ins tr ume nt ,
and rely below floor) 8
- Zone 10 (instrument and I relay room ~ ceiling mount) 7
- Zone 23 (cable spreading room) 12
- Zone 17 (control room) 6
- Zone 18 (computer room) 1
- Zone 19 (equipment room) 1
- Zone 20 (equipment room duct mount) 1
- Zone 14 (west comm.
room)- 1 O - Zone 12 (MCC room) 1 l l
- 1 s
iI 1 U BEAVER VALLEY UNIT 2 3/4 3-55 i l l l ! i i i t
Table 3.3-10 (Continued) FIRE DETECTION INSTRUMENT S Instrument Loc at ion Minimum Instruments Operable Smoke Flame
- 3. Service Building
- Zone 01 (emergency swgr west) 12
- Zone 02 (emergency swgr e as t) 11
- Zone 03 (battery rooms 2-1 to 2-4) 4 per room )
- Zone 4A (service building cable tr ays north) 12
- Zone 4B (service building cable trays south) 12
- 4. Diesel Generator Building (EG2-2)
- Zone 62 (die. generator nort h) 2
- 5. Diesel Generator Building (EG2-1)
() - Zone 62 (die. ~ generator south) 2
?
i
?
4 5 I l l l < h
-h L
[ () l l BEAVER VALLEY UNIT 2 3/4 3-56 l1 i i
Table 3.3-10 (Continued) FIRE DET ECT ION IN ST RUMENT S s
' Detectors Instrument Location Minimum Instruments Operable Smoke Flame
- 6. Safeguards Building
- Zone 26 (pump area north) 3
- Zone 27 (pump area south) 3
- Zone 28 (swgr north) 1
- Zone 28 (swgr north,
- duct mount) 1
- Zone 29 (svgr south) 1
- Zone 29 (svgr south,
- duct mount) 1
- 7. Fuel and Decontadination Building 1
- Zone 66 (2FNC*P21A,B) 1
- 8. Primary Auxiliary Building
- Zone 51B (charging pumps) 7 l - Zone 51 A (2.CCP*P21A ,B,C and storage area) 10 O' - Zone 52B (supp. leak-collection filters) 10 ;
I I i i i J l BEAVER VALLEY ITN TT 2- 3/4 3-57 } l
T able 3.3-10 (Continued) FIRE DET ECTION IN STRUMDiT S Instrument Location Minimum Instruments _ Operable Snoke Fl ame
- 9. Cable Vault and Rod Control Area
- Zone 30 (MCC*2-E3, MCC*2-E05) 7
- Zone 31 (MCC*2-E14, MCC*2-E06) 5
- Zone 32 (2RCP*H2A,B,D,E) 11
- Zone 53 ( Auxiliary Building relay room 2
- Zone 52 ( Auxiliary Building cable tunnel) 3
- Zone 50 (Auxiliary Building cable area) 3 1
- 10. Cable Tunnel
- Zone 16 (cable tunnel '
nor th) 4
- Zone 36 (cable tunnel r.or t h) 1
- Zone 22 (cable tunnel sout h) 4 0- - Zone 15 (cable tunnel south) 2 a
BEAVER VALLEY UNIT 2 3/4 3-58
3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTURMENTATION- , 1 CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 Two independent chlorine detection systems, with their alarm / trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppe, shall be OPERABLE. APPLICABILITY: Modes 1, 2, 3, 4. ACTION:
- a. With one chlorine detection systems inoperable, restore one of the inoperable detection systems to OPERABLE status within 7 days, or within the next b hours, initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operations.
- b. With no chlorine detection system OPERABLE, within 1 hour initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.
() c. The provisions of Specification 3.0.4'are not applicable. SURVELLLANCE REQUIREMENTS 4.3.3.7 Each chlorine detection system shall be. demonstrated OPERABLE by performance of a CHANNEL FUNCTION TEST at least once per 31' days and a CHANNEL CALIBRATION at least once per 18 months. O l BEAVER VALLEY UNIT 2 3/4 3-59
3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3 i ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in the channel designation column in Table 3.3-11, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours except for the PORV(s) which may be isolated in accordance with Specification 3.4.11.a.
- b. With the number of OPERABLE accident monitoring instrumentation channels less the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-11, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE' REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. O BEAVER VALLEY UNIT 2 3/4 3-60 l
TABLE 3.3-11 O. ACCIDENT MONTTORING IN STRUMENTATION Required Number of Minimum Channels Instrument Channels Operable Operable
- 1. Pressurizer Water Level 3 2
- 2. Auxiliary Feedwater Flow Rate 1 per steam generator 1 per steam generator
- 3. Reactor Coolant System Subcooling Margin Monitor 1 0
- 4. PORV Position Indicator 2/ valve 1/ valve
- 5. PORV' Block Valve Posit ion Indicator 1/ valve 0/ valve
- 6. Safety Valve Position Indicator 2/ valve 1/ valve
- 7. Safety Valve Temper ature Detector 1/ valve. 0/ valve
(} BEAVER VALLEY UNIT 2 3/4 3-61 , I l
IABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Channel Check Channel Calibration
- 1. Pressurizer Water Level M R
- 2. Auxiliary Feedwater Flow Rate SU(I) R
- 3. Reactor Coolant System Subcooling Margin Monitor M R
- 4. PORV Position Indicator M R
- 5. PORV Block Valve Position Indicator M R
- 6. Safety Valve Position Indicator M R
- 7. Safety Valve Temperature Detector M R
- O
?
7
.i l
1 i (I) Channel Check to be performed in conjunction with Surveillance i Requirement 4.7.1.2.a.9 following an extended plant outage. . BEAVER VALLEY UNIT 2 3/4 3-62 l i
3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION O LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid ef fluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip septoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM) . APPLICABILITY: During releases through the flow paths ACTION:
- a. With a radioactive liquid ef fluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification immediately suspend the release of radioactive liquid ef fluents monitored by the af fected channel or correct the alarm / trip setpoint.
- b. With one or more radioactive liquid effluent monitoring instrumentation channels INOPERABLE, take the action shown in
() Table 3.3-12 or conservatively reduce the alarm setpoint. Exert best effort to return the instruments to OPERABLE status within 30 days. If unsuccessful, explain in the next Semiannual Radioactive Ef fluent Release Report why the INOPERABILITY was not corrected in a timely manner,
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12. l (v~h lBEAVERVALLEYUNIT2 3/4 3-63 I
4 O O O TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUPNr MONITORING INSTRUMENTATION i Minimum Channel s Oper able Action Instrument
- 1. Gross Radioactivity Monitor Providing Alarm I and Automat ic Terminat ion of Release a) Liquid Waste Process Ef fluent Monitor (2SGC-RQIl00) 1 23 l
- 2. Gross Radioactivity Monitors Providing Alarm But Not l i
4 Providing Termination of Release I None
- 3. Flow Rate Measurement Devices i a) Liquid Radwaste Effluent (2SGC-FS100 and HIC 100) 1 25 lI b) Cooling Tower Blowdown Line (2CWS-PT101) 1 25 l
- 4. Tank Level Indicating Devices
- (For tanks outside plant buildings)
a) RWSr (2QSS*LT 100A ,B) 1 26 i i I 9 BEAVER VALLEY UNIT 2 3/4 3-64 i i j
O O O t , TABLE 3.3-12 RADIOACTIVE LIQUID EFFIAJENT MONITORING INSTRUMENTATION TABLE NOTATION
. ACTION 23 - With the number of channels OPERABLE 'less than required by the Minimum Channels OPERABLE requirement, ef fluent releases may be resumed provided that prior to initiating a release:
]
- 1. At least two independent samples are analyzed in accordance with Specificat ion 4.! ! . l .1.1 and; h
- 2. At _least two technically qualified members of the Facility Staff independently verify the release rate calculat io ns and discharge valving; l otherwise, suspend release of radioactive effluents via this pathway.
ACTION 24 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least one per 8 hours grab sangles are analyzed for gross radioactivity (BETA or GAMMA) at a Lower Limit of Detection (LLD) of at le as t 10- uCi/ml. ACTION 25 - With the~ number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fl ue nt releases via this pathway may continue provided the flow rate is estimated at least one per 4 hours during actual
- releases. Pump curves may be used to estimate flow.
i i ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, l iqu id I addit ions to this tank may continue provided the tank liquid level is estimated during all liquid additions to l. !
. the tank.
1 1 i l REAVER VALLEY UNIT 2 3/4 3-65 1
O O O TABLE 4.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUNENTATION SURVEILIANCE REQUIRENENTS Channel Sour ce Channel Channe l Instrument Check Check Calibration Funct iona l
- 1. Cross Radioactivity Monitor Providing Alarm and Automat ic Termination of Release a) Liquid Waste Process Ef fluent (2 SGC-RQIl00) D- P(5) R(3) Q(1)
- 2. Cross Radioactivity Monitors Providing Alarm y
But Not Providing Automatic Termination o f ' Re lease None
- 3. Flow Rate measurement 'dev ice s a) Liquid Radwaste Effluent (2SGC-FS100 and HIC 100) D(4) N/A R Q b) Cooling Tower Blowdown Line D(4) N/A R Q l}
- 4. Tank Level Indicating Devices a) RWST (2QSS*LT100A,B) D* N/A R Q
- During liquid additio. s to the tank HEAVER VALLEY UNIT 2 3/434
1 5 TABLE 4.3-12 (Cont inued) T A BLE ,NOTAT I_0N r 1. The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic l isolation of this pathway and Control Room Alarm Annunciation occurs if l any of the following conditions exist: l 1. Instrument indicates measured levels above ' the alarm / trip setpoint. ! 2. Downscale failure, j 1 3. Instrument controls not set in operate mode, i [ 2. The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm a annunciation occurs if any of the following conditions exist: i
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Downscale failure.
a l 3. Instrument controls are not set in operate mode.
- 3. The initial CHANNEL CALIBRATION for radioactivity measurement j instrumentation shall- be performed using one or more of the reference ,
standards certified by the National Bureau of Standards or using standards j that have been obtained from suppliers that part icipate in measurement p assurance activities with NBS. These standards should permit calibrating i v the system over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per , 18 months.. This can normally be accomplished during refueling outages. (Existing plants may substitute previously established calibration j procedures for this requirement.) i ) 4. CHANNEL CHECK shall consist of verifying indication of flow during periods j of release. CHANNEL OlECK shall be made at least once dai-ly on any day on . I which continuous, periodic, or batch releases are made, i 4 l S. A source check may be performed utilizirs the installed means or flashing I the detector with a portable source to obtain an upscale increase in the existing count rate to verify channel response. ;
- 6. The Channel Functional' Test - shall also demonstrate that automatic isolation of this pathway and Control Room Alarm Annunciation occurs when l the instrument indicates measured levels above the Alarm / Trip Setpoint.
I The Channel Functional Test shall also demonstrate that Control Alarm ! Annunciation occurs if any of the following conditions exist: { ' i 1. Downscale Failure ; i l 2. Instrument controls are not set in operate mode, f I ! BEAVER VALLEY UNIT 2 3/4 3-67 l I
- a. ;
i
3 /4.3 IN ST RUMENT AT ION 3/4.3.3 MONITORING IN STRUMENTATION RADIOACTIVE GA SEOUS EFFLUENT MONITORING IN STRUMENTATION O LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous ef fluent monitoring instrumentation chancels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined ^ in accordance with the methodology and parameters in the OFFSITE DOSE i CALCULATION MANUAL (ODCM). APPLICABILITY: i During releases through the monitored flow paths. 1 ACT ION :
- a. With' a radioactive gaseous ef fluent monitoring instrumentation channel j alarm / trip setpoint lese conseryative than a yalue which will enoure that the limits of 3.11.2.1 are met , immediately suspend the release
- of radioactive gaseous effluents monitored by the af fected channel or correct the alarm / trip set point.
i j b. With one or more radioactive gaseous ef fluent monitoring instrumentation channels INOPERATIVE, take the ACTION shown in Table 3.3-13 or conservatively reduce the alarm set . point. Exer t be s t l efforts to return the instruments to OPERABLE status within 30 days ! ' and, 'if unsuccess ful, explain in the next Semiannual Radioactive ,
. Effluent Release Report why the INOPERABILITY was not corrected in a [
timely manner. l i i
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicab le .
i SURVEILLANCE REQUIREMENTS i [ 4.3.3.10 Each radioactive gaseous ef fluent monitoring instrumentation i channel shall be demonstrated OPERABLE by performance of the CHANNEL OlECK, l SOURCE OlECK, QlANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the requirements shown in Table 4.3-13.
.[
l I i BEAVER VALLEY UNIT 2 3/4 3-68 1
O O O TABLE 3.3-13 RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument _ Operable Applicability Parameter Action
- 1. Caseous Waste / Process Vent System (Rm-CW-108 A & B) [BVPS Unit 1]
- a. Noble gas activity monitor 1
- Radioactivity Rate Measurement 27,30***
- b. Iodine Sampler Cartr idge 1
- 32
- c. Particulate Activity Monitor
- 1
- 32 '
- d. System Ef fluent Flow Rate Measuring Device !
- System Flow Nate Measurement 28 (FR-CW-108)
- e. Sampler Flow Rate Measuring Device .I
- Sampler Flow Rate Measurement 28
- 2. Ventilation Vent (2HVS-RQIl01)
- a. Noble gas activity monitor I
- Radioactivity Rate Measurement 29,30***
- b. Iodine sampler 1
- 32
- c. Particulate sampler I
- 32
- d. Process flow monitor 1
- Process Flow Rate Measurement 28 I
- e. Sampler flow monitor 1
- Sampler Flow Rate Measurement 28
- 3. Elevated Release (2HVS*RQIl09 B & C)
- a. Noble gas activity monitor 2
- Radioactivity Rate Measurement 29,30***
- b. Iodine Sampler 1
- 32
- c. Particulate Sampler 1
- 32
- d. Process flow rate monitor 1
- Process Flow Rate Measurement 28
- e. Sampler flow rate monitor i
- Sampler Flow Rate Measurement 28
- 4. Decontamination Building Vent (2 RMQ-RQI301)
- a. Noble gas activity monitor 1
- Radioactivity Rate Measurement 29
- b. Iodine sampler I
- 32
- c. Particulate sampler 1
- 32
- d. Process flow rate monitor 1
- Process Flow Rate Measurement 28
- e. Sampler flow rate monitor 1
- Sampler Flow Rate Measurement 28 i
BEAVER VALLEY UNIT 2 3/4 3-69 I t
O O O TABLE 3.3-13 RADI0 ACTIVE GASEQUS EFFLUDIT MONITORING INSTRUMENTATION Minimum Channels Instrument Oper able Applicability Parameter _ Action
- 5. Condensate Polishing Building Vent (2HVL-RQIll2)
- a. Noble gas activity monitor 1
- Radioactivity Rate Measurement 29
- b. Iodine Sampler I
- 32
- c. Part iculate Sampler 1
- 32
- d. Process flow rate monitor 1
- Process Flow Rate Measurement 28-
- e. Sampler = flow rate monitor i
- Sampler Flow Rate Measurement 28 1
- 6. Waste Gas Storage Vault (2 RMQ-RQI303)
- a. Noble gas activity monitor 1
- Radioactivity Rate Measurement 29
- b. Iod ine sampler 1
- 32
- c. Particulate sampler 1
- 32
- d. Process flow rate monitor 1
- Process - Flow Rate Measurement 28
- e. Sampler flow rate monitor 1
- Sampler Flow Rate Measurement 28
- 7. Waste Gas Holdup System Explosive Gas Monitoring
- a. Oxygen Monitor 2 ** 0xygen 31
~ (2DWS-0A 100 A & B)
- During Releases via this pathway.
- During Waste Gas Decay Tank filling operation.
- During purging of reactor containment via this pathway BEAVER VALLEY UNIT 2 3/4 3-70
O O O TABLE 3.3-13 RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATION ACTION 27 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to init iat ing the rele ase:
- 1. At least two indepe ndent samples of the tank's content are analyzed, and
- 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculat io ns and discharge' valve lineup.
Otherwise, suspend release of radioactive ef fluents via this pathway. ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may cont inue provided the flow rate is estimated at least once pe r 4 hour s . ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via. this pathway may continue provided grab samples are taken at least once per 8 hours ard these samples are analyzed .for gross activity within 24 hours. ACTION 30 - With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, immediately suspend PURGING of Reactor Containment via this pathway. ACTION 31 - With the number of channels OPERABLE one 'less than required by the MINIMUM Channels OPERABLE requirement, operation of this system may continue provided grab samples are obtained every 4 hours and analyzed within the following 4 hours during additions. to a rank. ACT ION ' 32 - Wit h t he number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are cont inuously collected with auxiliary sampling equipment as required in Table 4.11-2 or sampled and analyzed once every 8 hours. ACTION 35 - See Surveillance 4.11.2.5.1. BEAVER VALLEY UNIT 2 3/4 3-71
p. m TABLE 4.3-13 RADIOACTIVE GASEOUS EFFLUINT MONITORING IN ST RUMDITAT ION Channe1 Channel Sour ce Channel Funct ional Instrument Check Check Calibration Test
- 1. Gaseous Waste / Process Vent System (RM-GW-108 A & B) [BVPS Unit 1] ,
- a. Noble gas activity monitor P P(5) R(3) Q(1) i
- b. Iodine sampler cartridge W(6) N/A N/A N/A
- c. Particulate activity monitor W N/A N/A N/A
- d. System ef fluent flow rate measuring device P N/A R Q (FR-GW-108)
- e. Sampler flow rate measuring device D* N/A R Q
- 2. Ventilation Vent (2ilVS-RQIl01) g.
- a. Noble gas activity monitor D M(5) P(5)=** R(3) Q(2)
- b. Iodine sampler W N/A N/A N/A
- c. Port ible sample W N/A N/A N/A
- d. Process flow rate monitor D N/A. R Q
- e. Sample flow rate monitor D N/A R Q
- 3. Elevated Release (2HVS-RQIl09 8 & C)
I i M(5) P(5)***
- a. Noble gas activity monitor D R(3) Q(2)
- b. Iodine Sampler . W N/A N/A N/A
- c. Portible sampler W N/A N/A N/A
- d. Process flow rate monitor D N/A R Q
- e. Sampler flow rate monitor D N/A R Q 1
1 j BEAVER VALLEY UNIT 2 3/4 3-72 1 I.
O O O TABLE 4.3-13 , RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION Channel Channel Sour ce Channel Funct ional Instrument Check check Calibration Test
- 4. Decontamination Building vent (2RMQ-RQI301)
- a. Noble gas activity monitor D M(5) R(3) Q(2)
- b. Iodine sampler W N/A N/A N/A
- c. Particulate sampler W N/A N/A N/A
- d. Process flow rate D' N/A R Q
- e. Sample flow rate. D N/A R Q
- 5. Condensate Polishing Building Vent (2HVL-RQIll2 )
- a. Noble gas activity monitor D M(5) R(3) Q(2)
- b. Iodine Sampler W N/A N/A N/A
- c. Particulate sampler W N/A N/A N/A I
- d. Process flow rate monitor D N/A R Q
- e. Sample flow rate monitor D N/A R Q i
- 6. Waste Gas Storage Vault (2RMQ-RQI303 )
- c. Noble gas activity monitor D M(5) R(3) Q(2)
- b. Iodine sample W N/A N/A N/A
- c. Particulate sample W N/A N/A N/A
- d. Process flow rate monitor D N/A R Q
- e. Sample flow rate monitor D N/A R Q i 7. Waste Gas Holdup System Explosive Gas Monitoring
- e. Oxygen Monitor D N/A Q(4) M (2DWS-OA 100 A & B) t x
BEAVER VALLEY UNIT 2 3/4 3-73
O O O TABLE 4.3-13 RADI0 ACTIVE CASEOUS EFFLUfNT MONITORING INSTRUNENTATION
- Dur ing releases via this pathway I
** During waste Cas Tank filling operations ! *** During purging of reactor containment via this pat hway (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that aut omat ic isolation of this pathway and control room alarm cnnunciat ion occurs if any of the following condit ions exists:
- a. Ins tr ume nt indicates measured levels above the alare/ trip setpoint.
2
- b. Downscale failure, c . Ins tr ume nt controls not se t in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demoastrate that control room alarm annunciation occurs if any of t he following conditions exists: I
- a. Instrument indicates measured levels above the alarm / trip setpoint.
- b. Downscale failure,
- c. Ins tr ument controls not se t in operate mode.
4 (3) The initial CHANNEL CALIBRATION for rulioactivity measurement instrumentat ion shall be per formed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obt ained fr om
! cuppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the j cystem over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have
, been related to the initial calibration shall, be used, at intervals of at least once per eighteen months. This can f normally be accomplished during refueling outages. 1 (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: 1 l' a. One volume percent oxygen, balance nitrogen, ' and
- b. Four volume percent oxygen, balance nitrogen.
l (5) A source check may be performed utilizing the installed nears or flashing the detector with a port able source to obtain ' an upscale incre ase in the existing count rate to verify channel response, j (6) Compliance is demonstrated in Table 4.11-2. REAVER VALLEY UNIT 2 3/4 3-74 I k' _
3/4.4 REACTOR COOLANT SYST EM 3/4.4.1' REACTOR COOLANT IDOPS NORMAL OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation . APPLICABILITY: MODES I and 2*. ACTION : Above P-7, comply with the following ACTIONS:
- a. With one reactor coolant loop and associated pbsp not in operation, subsequent STARTUP and POWER OPERATION above 25 per cent of RAT ED THERMAL POWER may proceed provided:
- 1. The following actions have been completed with the reactor suberitical:
3 a) Reduce the overtemperature AT trip setpoint to the value specified in Specification 2.2.1 for 2 loop operation.
- See Special Test Exception 3.10.5.
O O BEAVER VALLEY UNIT 2 3/4 4-1
3/4.4 REACTOR COOLANT SYS1 tM 3/4.4' 1 NORMAL OPERATION O LIMITING CONDITION FOR OPERATION
- - I b) Place the following reactor trip system and ESF instrumentation channels, associated with the loop not in operation, in their tripped conditions:#
- 1. Overpower Delta T channel.
- 2. Overtemperature Delta T channel, c) Change the P-8 interlock setpoint from the value specified in Table 2.2-1 to < 70 percent of RATED THERMAL POWER when the ly reactor coolant stop valves in the non-operations loop are closed.
- 2. Thermal Power is restricted to < 65 percent of RAT ED THERMAL POWER when the reactor coolant stop valves in the non-operating loop are closed.
O 9
# These channels may be placed in the bypass condition for up to 8 hours during surveillance testing of the overpower and overtemperature Delta T channels of the active loops.
O BEAVER VALLEY UNIT 2 3/4 4-2
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.1 NORMAL OPERATION LIMITING CONDITION FOR OPERATION Below P-7:
- a. With K,f f > 1.0, operation below P-7 may proceed provided at le as t two 7eactor coolant loops and associated pumps are in operation.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applic ab le .
SURVEILIANCE REQUIREMENT S 4.4.1.1.1 With one reactor coolant loop and associated pump not in oper at ion, at least once per 7 days determine that:
- a. The applicable reactor trip system channels specified in the ACTION statements above have been placed in their tripped conditions, and
- b. If P-8 interlock setpoint was reset for 2 loop operation, the P-8 interlock setpoint is < 70 percent of RAT ED THERMAL POWER when the reactor coolant stop valves in the non-operating loop are closed.
4.4.1.1.2 The power to each of the reactor coolant system loop stop O, valves shall be verified to be removed at least once per 31 days during operation in Modes 1 or 2. l l 1 BEAVER VALLEY UNIT 2 3/4 4-3 1
3/4.4 REACTOR COOLANT SYST EM 3/4.4.1 REACTOR COOLANT LOOPS HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2
- a. At least two reactor coolant loops and associated steam generators and reactor collant pumps shall be in operation
- when the rod control system is capable of control bank rod withdrawal.
- b. At le as t two reactor coolant loops and associated steam generators and reactor coolant pumps shall be OPERABLE and one reactor coolant loop shall be in operation
- when the rod control system is incapable of control bank rod withdrawal.
APPLICABILITY MODE 3 ACTION :
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT 91UTDOWN within the next 12 hours.
- b. With less than two reactor coolant loops in operation, immediately de-energize all control rod drive mechanisms, or align the rod control 1 system so that it is incapable of control bank rod withdrawal.
("] c. With no reactor coolant loop in operation, suspend all operations \s/ involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation. SURVEILIANCE RIQUIREMDIT S 4.4.1.2.1 With the rod control system capable of rod withdrawal, at least two cooling loops shall be verified to be in operation and circu-lating reactor - coolant at least once per 12 hours. 4.4.1.2.2 With the rod control system incapable of rod withdrawal, at least two cooling loops, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.3 With the rod control system incapable of rod withdrawal, at least one cooling loop shall be verified to be in operation and circulat- 1 ing reactor coolant at least once per 12 hours.
- All reactor coolant pumps may be de-energized for up to I hour pro-vided (1) no operations are permitted that would cause dilution of the reac-tor coolant system boron concentration and (2) core outlet temperature is maintained at least 10*F below saturation temperature. This does not preclude natural circulation cooldown under abnormal cooldown conditions.
O v BEAVER VALLEY UNIT 2 3/4 4-4
i i 3/4.4 REACTOR COOLANT SY ST EM 3/4.4.1 REACTOR COOLANT LNPS
~ SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3
- a. At le a;t two of the coolant loops listed below shall be OPERABLE:
- 1. Reactor Coolant Loop (A) and its associated steam generator and reactor coolant pump *,
- 2. Reactor Coolant Loop (B) and its associated steam generator and reactor coolant pump *,
- 3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump *.
- 4. Residual Heat Removal Pump (A) and a heat exc h ang e r ** ,
- 5. Residual Heat Removal Pump (B) and a second heat exchanger **.
- b. At least one of the above coolant loops shall be in operation ***.
APPLICABILITY: () MODES 4 and 5. A CT ION :
- a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SitTTDOWN within 20 hours.
- A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to (later)*F unless the secondary water temper ature of each steam generator is less than (later)*F above each of the RCS cold leg temperatures.
** The normal or emergency power source may be inoperable in MODE 5.
*** All reactor coolant pumps and Residual Heat Removal pumps may be de-energized for up to I hour provided: 1) no operations are permitted that would cause dilution of the reactor coolant system boron-conce ntr at ion, and 2) core outlet temperature is maintained at leas t 10 *F below satur at ion temperature.
O BEAVER VALLEY UNIT 2 3/4 4-5 i
3/4.4 REACTOR COOLANT. SYST EM 3/4.4.1 REACTOR COOLANT LOOPS , SHUTDOWN LIMITING CONDITION FOR OPERATION
- b. With no coolant loop in operation, . suspend all operat ions involving a reduction in boron concentration of the Reactor Coolant System and imediately initiata corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMDIT S 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5, and by verifying that each residual heat removal pump develops a differential pressure of > (later) psi on recircula-tion flow. 4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level equivalent to (later) percent narrow range } at le as t once per 12 hours. 4.4.1.3.4 At le as t one coolant loop shall be verified to be in operation and circulating reactor coolant at least ' once per 12 hours. O BEAVER VALLEY UNIT 2 3/4 4-6
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.1 REACTOR COOLANT IDOPS ! ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.4 The boron concentr ation of an isolated loop shall be main-tained greater than or equal to the boron concentration of the operating loops, except when the loop is de sined for maintenance. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACT ION : With the requirements of the above specification not satis-fied, do not open the isolated loops stop valves; either increase the boron concentr ation of the isolated loop to within the limits within 4 hours or be in at le as t HOT STANDBY within the next 6 hours with the unisolated portion of the RCS borated to a SHUTDOWN MARGIN equivalent to at le as t 1 percent
! delta k/k at 200*F.
SURVEILIANCE REQUIREMENT S 4.4.1.4 The boron concentr ation of an isolated loop shall be deter-mined to be greater than or equal to the boron concentration of the operating loops within 30 minutes prior to opening either the hot leg or cold leg stop valves of an isolated loop. O BEAVER VALLEY UNIT 2 3/4 4-7
3/4.4 REACTOR COOLANT SYST EM 3/4.4.1 REACTOR COOLANT LOOPS ISOLATED IDOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.5 A reactor coolant loop shall remain isolated until:
- a. The' isolated loop has been operating on a recirculation flow of
> 125 gpm for at le as t 90 minutes and the temperature at the cold leg of the isolated loop is within 20*F of the highest , cold leg temperature of the operating loops.
- b. The reactor is suberitical by at le as t 1 percent delta k/k.
APPLICABILITY: All Modes
- ACT ION :
With the requirements of the above specification not satisfied, suspend startup of the isolated loop. SURVEILIANCE REQUIREMDIT S 4.4.1.5.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest ~ cold leg temper ature of the oper ating loops within 30 minutes prior to opening the cold leg stop valve. 4.4.1.5.2 The reactor shall be determined to be suberitical by at le as t 1 percent delta k/k within 30 minutes prior to opening the cold leg stop valve.
- With fuel in the reactor vessel BEAVER VALLEY UNIT 2 3/4 4-8
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES - SHITTDOWN O LIMITING CONDITION FOR OPERATION l __ l 1 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 psig + 1 percent. APPLICABILITY: MODES 4 and 5. A CT ION : With no pressurizer code safety valve OPERABLE, immediately suspe nd all oper at io ns involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. 1 SURVEILLANCE REQUIREMDIT S 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4.3. i i O BEAVER VALLEY UNIT 2 3/4 4-9 i
3/4.4 REACTOR COOLANT SYST EM 3/4.4.3 SAFETY VALVES - OPERATING LIMITING CONDITION FDR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 psig + 1 percent. APPLICABILITY: MODES 1, 2, and 3. ACT ION : With one pressurizer code safety valve inoperable , either restore the inoperable valve to OPERABLE status within 15 minutes or be in' HOT SHUT-DOWN within 12 hours. 1 SURVEILLANCE REQUIREMENT S 4.4.3 Each pressurizer code safety valve shall be demonstrated OPER-AB~.E with a lift setting of 2485 psig +- 1 percent, in accordance with Speci- I fication 4.0.5. O O BEAVER VALLEY UNIT 2 3/4 4-10 l i r P
-- - -- - , , , - , , - - , - ., -,% ,-, ,,,ye ---y--,-, , ,- ,-w. y -e.-
3/4.4 REACTOR COOLANT SYST EM 3/4.4.4 PRESSURIZER O LIMITING CONDITION FOR OPERATION 3.4.4 The pres surizer shall be OPERABLE with at le as t 150 kw of pressurizer heaters and with a stean bubble. APPLICABILITY: MODES 1, 2, and 3. ACT ION : With the pressurizer inoperable due to less than 150 kw of heaters supplied by an emergency bus, be in at least HOT STANDBY vithin the next 6 hours and in HOT SUTDOWN within the following 12 hours. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT WITTDOWN within the following 6 hours. ERVEILIANCE REQUIREMENT S 4.4.4.1 The power supply for these pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by energizing the heaters and measuring circuit current to verify capacity. O BEAVER VALLEY UNIT 2 3/4 4-11 ) I l I 1
l 3/4.4 REACTOR COOLANT SY ST EM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION l ! 3.4.5 Each steam generator shall be OPERABLE.
- APPLICABILITY
MODES 1, 2, 3, and 4. l , ACT ION : E With one or more steam generators inoperable, restore the ! inoperable generator (s) to OPERABLE status prior to increasing T **I above 200*F. e 4 ] SURVEILIANCE REQUIREMENT S
. 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam j generator shall be determined OPERABLE during shutdown by selecting and
- inspecting at le as t the minimum number of steam generators specified in T ab le 4.4-1.
4.4.5.2 Steam Generator Tube Sample Selection and Inspection - Tre steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. O The inservice inspection of steam generator tubes shall be performed at the ] frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubes shall be examined in accordance with the method i prescribed in Article 8 " Eddy Current Examinat ion of Tubular Pr oducts ," as contained in A SME Boiler and Pressure Vessel Code, Sect ion V "Non-destructive Examination," and referenced in A SME Boiler and Pressure Vessel i- Code - Appendix IV of the 1980 Edition through Winter '1980 Addenda of Section XI " Inservice Inspection of Nuclear Power Plant Components." The , tubes selected for each inservice inspection shall include at least 3 percent of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
- a. Where experience in similar plants with similar water chemistry j indicates critical areas to be inspected, then at least 50 percent of the tubes inspected shall be from these critical areas.
l i b. T he fir s t inservice inspection (subsequent to' the preservice inspection) of each steam generator shall include: i J 1. All nonplugged tubes that previously had detect able wall penetr ations (>20%), and 1
- 2. Tubes in those areas where experience has indicated potential
- pr ob lems .
,' BEAVER VALLEY UNIT 2 3/4 4-12 _ . - - . _ _ _ - _ _ _ _ . . _ . _ . , _ _ _ , . ~ . , _ . . . _ _ _ _ _ _ _ . . _ - _ . _ - _ _ . , _ . , _ _ - - _ - . - -
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.5 ST EAM GENERATORS O-SURVEILLANCE REQUIREMENTS
- c. The second and third inservice- inspections may be less than a full tube inspection by concentrating (selecting at least 50 percent of the tubes to the inspected) the inspection on those areas of the tube sheet arr ay and on those portions of the tubes where tubes with imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results C-1 I4ss than 5 percent of the tot al tubes inspected are degr aded tubes c and none of the inspected tubes are de fect ive . C-2 One or more tubes, but not more than 1 percent of the total tubes inspected are defective, or between 5 percent and 10 percent of the j total tubes inspected are degr aded
.O- t ses.
C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective. NOT E: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the I above percent age calculationa. 4.4.5.3 Inspection Frequencies - The above required inservice 4 inspections of steam generator tubes shall be per formed at the following freque ncies:
- a. The first inservice inspection shall be performed af ter 6 Ef fective i Full Power Months but within 24 calendar months of initial
, criticality. Subsequent inservice inspections shall be per formed at
+
intervals of not less than 12 nor more than 24 calendar months af ter the previous inspection. If two consecutive inspections following service under All Volatile Treatment (AVT) conditions, not including the preservice ins pection, result in all inspection results falling into the C-1 category or if two consecutive inspections demons trate that previously observed degradation has not continued and no i additional degr adation has occuered, the inspection interval may be ! i extended to a maximum of once per 40 months. I BEAVER VALLEY UNIT 2 3/4 4-13
- _ _ _ . _ _ .. ____ _ _ _ _ _ _ . __ ~ __
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.5 STEAM GENERATORS O SURVEILIANCE REQUIREMENT S j b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose l results fall in Category C-3, the inspection frequency shall be reduced to at least once per 70 months. The reduction in inspection frequency shall apply until a subsequent ins pection demonstrates that a third sample inspect ion is not requir ed. ]
- c. Additional, unscheduled inservice inspections shall be performed on each ste am gener ator in accordance with the first sample inspection i'
specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.
; 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, ! 2. A seismic occurrence greater than the Operating Basis Earthquake,
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or
- 4. A main steam line or feedwater line break.
(} l 4.4.5.4 Acceptance Criteria i i s. As used in this Specification:
- 1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by f abrication dr awings or spe cificat ions . Eddy-curre nt testing indications below 20 percent '
, of the nominal tube wall thickneen, if detect able, may be ! considered as imper fe ct ions .
- 2. Degradation means a service-induced cracking, wastage, wear or
. general corrosion occurring on either inside or outside of a tube.
- 3. Degraded Tube means s' tube containing imper fections > 20 percent of -
the nominal wall thickness caused by degradation. 4
- 4. % Degradation means the percentage of the tube wall thickness af fected or removed by degradation.
1 h t
'I ! (:)
l BEAV1R VALLEY UNIT 2 3/4 4-14
,- .-_. . ~ . . _ - _ . . _ - , . . , _ . . , _ . . . . . . . . . . . , _ . _ , ~m,,,.,___,,._,_.._.,._._-, , _ _ _ _ , , _
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.5 STEAM GDIERATORS SURVEILIANCE RQUIREMENT S
- 5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current ins pe ct ion probe shall be deemed a defective tuba.
- 6. Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service or repaired because it may become unserviceable prior to the next inspection and is equal to 40 percent of the nominal tube wall thickness.
- 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to af fect its structur al integrity in the event of an Operating Basis Earthquake, a loss-of-coolant ac c ide nt , or a steam line or feedwater line break as specified in 4.4. 5.3.c , above .
- 8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
- b. The steam generator shall be determined OPERABLE af ter completing the corresponding actions (repair all tubes exceeding the plugging limit O-s and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Re por t s
- a. Following each inservice inspection of steam generator tubes , the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
- b. The complete results of the steam generator tube inservice ins pect ion shall be included in the Annual Operating Report for the per iod in which this inspection was completed. This report shall include:
- 1. Number and extent of tubes ins pected .
- 2. Location and percent of wall-thickness penetr ation for each indication of an imper fection.
- 3. Identification of tubes plugged.
1 O BEAVER VALLEY UNIT 2 3/4 4-15 _ - . _ . . - . _ - , , ~ . . . _ , . , . , ,_m. , . . - . - , - _ _ . _ _ - ._.__r,
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.5 STEAM GENERATORS O SURVEILIANCE REQUIREMENT S
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to Specification 6.6 prior to resumption of plant operation. The written followup of this report i shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent r ecurr e nce.
O I O BEAVER VALLEY UNIT 2 3/4 4-16 l
A O V U TABLE 4.4-1 MINIMUM NUMBER OF SrEAM CF)lERA_ TORS TO BE INSPECTED DURINC INSERVICE INSPECTION I Preservice Inspection No Yes Number of Steam Generators per Unit Two Thr ee Four Two Tlr ee Four First Inservice Ins pect ion All One Two Two Second & Subsequent Inservice Inspections Onel Onel One2 One3 TABLE NUTATION: l . T he inservice inspect ion may be limited to one stems generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous ins pect io ns ind ica t e that all steam generators are performing in a like manner. Note that under some circumst ances, t he oper at ing cond it ions in one or more steen generators may be found to be more severe than those in other stema generators . Under such circumstances, the sample sequence shall be modified to inspect the most severe condit ions.
- 2. The other stema generator not inspected dur ing the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in I above.
- 3. Each of the other two steam gener ators not inspected dur ing the first inservice inspect ions shall be inspected dur ing the second and third ins pect ions . The fourth and subsequent ins pect io ns shall follow the instruct ions described in I above.
RFf.VER VALLEY UNIT 2 3/4 4-I7 3 1
O O O TABLE 4.4-2 ST EAN GENERATOR TUBE INSPECTION ist SANPLE INSPECTION 2 nd SANPLE INSPECTION 3rd SANPLE INSPECT ION Sample Size Result Act ion Required Result Act ion Required Result Act ion Required A minimum of C-! None N/A N/A N/A N/A S Tubes per S.C . C-2 Repair defect ive C-1 None N/A N/A Tubes and ins pect addit ional 2 S tubes C-2 Plug defect ive tubes and C-1 None in this S.C. ins pect additional 4S tubes in this S.C . C-2 Plug defective tubes C-3 Per form act ion for C-3 result of first s ample C-3 Per form act ion for C-3 N/A result of first sample N/A A C-3 Ins pect all tubes in All t his S.C. , repair ot her defect ive tubes and S.C.'s ins pect 2 S tubes in ar e C-1 None N/A N/A each other S.C. Some Per form act ion for C-2 S.C. 's result of second sample N/A N/A C-2 but no add ' l Pr ompt not if ic at io n S.C. are to NRC pursuant to C-3 Specificat ion 6.9.1 Add'l I ns pect all tubes in S.C. is e ac h S.G . and plug C-3 defect ive tubes. Prompt not ificat ion to NRC N/A N/A pursuant to l Specificat ion 6.9.1 R EAV ER VA LLEY UN IT 2 3/4 4-18
i O O O , TABLE 4.4-2 (Cont inued) I STEAN CENERATOR TUBE INSPECTION i N S= 3 - % Where N is the number of steam generators in the unit, and n is the number of stemi generators n inspected dur ing an inspect ion, t i f i t i I k REAVER VALLEY UNIT 2 3/4 4-19
3/4.4 REACTOR COOIANT SY ST EM 3/4.4.6 REACTOR C001 ANT SY ST EM LEAKAG E LEAKAGE DETECTION SYST EMS O LIMITING CONDITION FOR OPERATION 3.4.6.1 The following reactor coolant system leakage detection systems shall be OPERABLE. 4
- a. The containment atmosphere particulate radioactivity monitoring system,
- b. The containment sump discharge flow measurement system or narrow range level instrument, and
- c. Containment atmosphere gaseous activity monitoring system.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION : l
- a. With one of the above radioactivity monitoring leakage detection systems inoper able, oper ations may continue for up to 30 days i pr ovided:
- 1) The other two above required leakage detection systems are I OPERABLE, and
- 2) Appropriate grab samples are obtained and analyzed at least once i per 24 hours:
otherwise, be in at least MOT STANDBY within the next 6 hours and in COLD SHITTDOWN within the following 30 hours,
- b. With the containment sump discharge flow measurement system and narrow range level instrument inoper able, res tore at least one inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY i wtthin the next 6 hours and in COLD SHITTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable in Modes 1, 2
- and 3.
RJRVEILIANCE REQUIREMENT S 4'.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
- a. Cont ainment atmosphere particulate and gaseous monitoring syatem by per forming a CHANNEL OlECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies specified in T able 4.3-3.
- b. Containment sump discharge flow measurement system by per forming a CHANNEL CALIBRATION TEST at least once per 18 months.
- c. Logging the narrow range level indication every 12 hours, i O BEAVER VALLEY UNIT 2 3/4 4-20 l
l i
- - - - _ _ - . , , _ . - ~ . _ . _ , , . ,- ,, .-. _ , _ . _,-mr._,- ,_~___-__-m,. _~,__-__.,_-_-.--,--v_,
I 3/4.4 REACTOR COOLANT SY ST EM 1 3/4.4.6 REACTOR COOLANT SY ST EM LEAKAG E , OPERATIONAL LEAKAGE I O LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. I gpm UNIDl!NTIFIED LEAKAGE,
- c. I gpm total primary-to-secondary leakage through all steam generators i
not isolated from the Reactor' Coolant System and 500 gallons per day i through any one steam generator not isolated from the Reactor Coolant l System.
. l l
d.10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and 5
- e. 28 gpm C0ffrROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUrDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSIRE BOUNDARY LEAKAGE, reduce the le akage ,
rate to within limits within 4 hours or be in at least Har STANDBY ' within the next 6 hours and in COLD SHUTDOWN within the following 30
~
hour s . i O BEAVER VALLEY UNIT 2 3/4 4-21
- . .__ - _ - - .- _ =_ - .
F 3/4.4 REACTOR COOLANT SY ST EM 3/4.4.6 REACTOR C.00LANT SYSTEM LEAKAGE OPERATIONAL LEAKAG_E a O- , SURVEILIANCE REQUIREMENT S c 4.4.6.2 Reactor Coolant System leakage shall be demonstr ated to be within each of the above limits by: l' a. Monitoring the containment atmosphere gaseous and part iculate r adioactivity monitor at least once per 12 hours,
- b. Monitoring the containment sump discharge at le as t once per 12 hour s .
- c. Measurement of the COffrROLLED LEAKAGE to the reactor coolant pemp seals when the Reactor Coolant System pressure is 2235 + 20 psig at (
least once per 31 days with the modulating valve fully open.
- d. Performance of a Reactor Coolant System water inventory balance at least once par 72 hours during steady state operation, and 1
- e. Monitoring the reactor head flange leakof f system at least once per 24 hours.
- .O i
4 I O t i BEAVER VALLEY UNIT 2 3/4 4-22 i 1
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.6 REACTOR COOLANT SY ST EM LEAKAG E 1 O LIMITING 00NDrrION FOR OPERNTION 3.4.6.3 Reactor coolant system pressure isolation valves shall be oper at ional. I APPLICABILrrY: l H0 DES 1, 2, . 3, and 4. i ACTION:
- 1. All pressure isolation valves listed in T able 4.4-3 shall be func-tional as a pressure isolation device, except as specified in 2.
Valve leakage shall not exceed the amounts indicated.
- 2. In the event that integrity of any pressure isolation valve specified in Table 4.4-3 cannot be demonstrated, reactor operation may con-l tinue, provided that at le as t two valves in each high pressure line having pondinga tonon-functional valve are(ai , and remain in, the mode corres-the isolated condition ,
- 3. In Specification 1 and 2 cannot be me t , an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition I within 24 hours.
- O 4. The provision of Specification 4.0.4 is not applicable for entry into Mode 3 or 4.
a r 3 t (a) Motor operated valves shall be placed in the closed position and s power supplies de-energized. 1 I f O BEAVER VALLEY UNIT 2 3/4 4-23 i
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYST EM LEAKAGE O SURVEILLANCE REQUIREMENT 4.4.6.3.1 Periodic leakage testing (a) on each valve listed in Table 4.4-3 shall be accomplished prior to entering Mode 1 af ter every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours if testing has not been accomplished in the preceeding 9 months, and prior to returning the valve to service af ter maintenance, repair or replacement work is performed. 4.4.6.3.2 Whenever integrity of a pressure isolation valve listed in Table 4.4-3 cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily. t 1 O , (a) To satisfy ALARA requirements, leakage may be measured indi- l rectly (as from the performance of pressure indicators) if accom-plished in accordance with approved procedures and supported by com-putations showing that the method is capable of demonstrating valve compliance with the leakage criteria. l l I I I
?
I) BEAVER VALLEY UNIT 2 3/4 4-24 ! r
TABLE 4.4-3 REACTOR C001 ANT SYSTEM PRESSURE ISOLATION VALVES Maximum (a)(b) System Valve No. Allowable Leskage Loop 1, cold leg, LH SI 2 SIS *548 j,5.0 gpm i Loop 2, cold leg, LHSI 2 SIS *550 j,5.0 gpm Loop 3, cold leg, LHSI 2 SIS *552 j,5.0 gpm Common, ccid leg, LH SI 2 SIS *132 <5.0 gpm 2 SIS *133 35.0 gpm (a) 1. Leakage rates less than or equal to 1.0 gpa are considered acceptable. I
- 2. Leakage rates greater than 1.0 gpa but less than or equal to 5.0 gpa are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount
() that reduces the margin between measured leakage rate and the maximum persaissible rate of 5.0 gpm by 50 percent or greater. i
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpa are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpa by 50 percent or greater.
- 4. Leakage rates greater than 5.0 gpa are considered unacceptable.
(b) 1. Minimum test dif ferential pressure shall not be less than 150 paid.
- t J
BEAVER 7 ALLEY UNIT 2 3/4 4-25
3/4.4 REACTOR COOLANT SYSTEM ! 1 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION ____ :: =:- ---- - -- : 3.4.7 The Reactor Coolant System chemistry shall be maintained with-in the limits specified in Table 3.4-1. I APPLICABILITY: At all times. ACT ION : MODES 1, 2, 3, and 4:
- a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the Parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHlTTDOWN with'in the following 30 hours,
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHtTIDOWN within the following 30 hours.
AT ALL OTHER TIMES: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to < 500 psig, if applicable, and perform an analysis to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations prior to increasing the pressurizer pressure above 500 psig or' pr for to pro-ceeding to MODE 4. SURVEILIANCE Rl!QUIREMDrT S 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at t he frequencies speci-fied in T able 4.4-10. 1 BEAVER VALLEY UNIT 2 3/4 4-26
1 TABLE 3.4-1
- - REACTOR C00lMT ' SYST.EM CHEMISTRY LIMITS l l.
l ST EADY STATE TRAN SIDIT , PARAMET ER LIMIT LIMIT Dissolved Oxygen <0.10 ppm * <1.00 ppm
- i Chloride <0.15 ppe <1.50 ppe Fluoride <0.15 ppe <1.50 ppe i
t i j I J
!
- Limit not applicable with T <250*F.
i avr i ~ 1 O l t i I ) i i i 4 i l I
- O BEAVER VALLEY UNIT 2 3/4 4-27
TABLE 4.4-10 y REACTOR C001AlfT SYST EM { CHEMISTRY LIMITS SURVEILIANCE REQUIREMENTS t 1
! MINIMUM MAXIMUM TIME CONTAINMDIT SAMPLING FREQUENCIES BETWE!!N SAMPLES ,
i
- Dissolved Oxygen 3 times per 7 days
- 72 hours Chloride 3 eines per 7 days 72 hours 3 Fluoride 3 eines per 7 dsys 72 hours l
.+ ,
1 ! r t i t
- Not required wit.h T avy<250*F.
i i P i i j - i, I i
- .i i i i
f I t l l i i O BEAVER VALLEY UNIT 2 3/4 4-28 1 i 1 t
3/4.4 REACTOR COOIANT SYST EM 3/4.4.8 SPECIFIC ACTIVITY O LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited I to: i a. < 1.0 uCi/ gram DOSE EQUIVALENT I-131, . and e
- b. < 100/E uCi/ gram.
l] APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACT ION : I MOD E S 1, 2 and 3* : r
! a. With the specific activity of the primary coolant > 1.0 uCi/gran DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left i of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applic ab le .
t ! b. With the specific activity of the primary coolant > 1.0 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time inter val or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY wit h T,, < 500*F within 6 hours,
- c. With the specific activity of the primary coolant > 100ff uCi/ gram, be in ly HUT STANDBY with T < $00*F within 6 hours,
- avg MODES 1, 2, 3, 4, and 5
i
; s. With the specific activity _of the primary coolant > 1.0 uCt/ gram DOSE EQUIVALENT I-131 or > 100/E uCi/ gram, per form the sampling and analysis requirements of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to within its limits. Submit a Special Report -
to th= Commission within 30 days pursuant to Specification 6.9.2 contain ing the results of the specific activity analyses together with the fo t i loving information:
- With T > $00*F.
avg -- BEAVER VALLEY UNIT 2 3/4 4-29
, - .-.-.-- . - - - , - - , .n-- - - ~ ~ . , - - - - - , , - - - , . , . , . , - . - - . -. . - - - - . - - - , - . - , , , . , , , , , - , - . . , - - - , , , ,
. . . . . _ _ _ _ _ . _ . _ _ _ _ _ . ~ . _ _ _ . _ _ . . . _ . -
c 3/4.4 REACTOR COOLANT SYST EM l 3/4.4.8_ SPECIFIC ACTIVITY O ' 1 LIMITING ' CONDITION FOR OPERATION i
=-= _ _
- 1. Reactor power history starting 48 hours prior to the first sample
! in which the limit was exceeded, I
l 2. Fuel burnup by core region, 1 i
- 3. Clean-up flow hiseory atatting 48 houra prior to the first sample f I in which the linic was exceeded,
- 4. History of de-gaseing operations , if any, etatting 48 houra prfor to the first sample in which the limit was exceeded, and
- 5. The time duration when the specific activity of the primary coolant
].
ex:ecded 1.0 uCi/ gram DOSE EQUIVALENT I-131. SUlt'iEILLANCE REQUIREMENT S 4 l 4.4.8 The specific activity of the primary coolant shall be deter-I mined to be within the limits by performance of the sampling and analysis
! program of Table 4.4-12. 1, !O l
i i . i I i i s l l 1 O~ BEAVER VALLEY UNIT 2 3/4 4-30 t 4 ,. i
O O O TABLE 4.4-12 1 PRIMARY COOIAltr SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM MODES IN WHICH TYPE OF MEASJREMElfr AND ANALYSIS MINIMUM FREQUENCY SURVEILIANCE RISQUIRED
- 1. Cross Activity Determination 3 times per 7 days with a maximum 1, 2, 3, 4 time of 72 hours between samples.
- 2. Isotopic Analysis for DOSE EQUIVALigtr 1 per 14 days 1 1-131 Concentr at ion
- 3. Radiochemical for E Determination I per 6 months 1
- 4. Isotopic Analysis for Iodine Including a) Once per 4 hours, whenever the 1#, 2#, 3#, 4# 5#
I-131, I-133, and I-135 specific activity exceeds 1.0 uCi/ gram DOSE EQUIVALBtr I-131 or 100/E uCi/ gram, and _. b) One sample between 2 & 6 hours 1, 2, 3 following a THERMAL POWER change exceeding 15 percent of the RATED THENGAL POWER within a one hour period. i Until the specific activity of the primary coolant system is restored within its limits. BEAVER VALLEY UNIT 2 3/4 4-31
FIGURE 3.4-1 O
\!
, . _ 1 i ie i i I ii iii. i_ i i ! ...
_- - i * ' i - ' E T - ' ' ' ' ' ' g . i't 'T
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- s. . , . . . . . , . , , ii.i i . . i ,
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3 . i 16 . I i . i 6 e i ! 6 .
' ' i U
. 4 . .ii .N i
, ,se UNACCEPTABLE i
, i ' i.
4 . .. ,ii i . .x OPERATION i o . , .! . . ! i e i T i t i , i t [ r
! l , i , i e i 'LIi6 . 6 I i ! t
= . t , l 56 .
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' . iI i l \ t t 5 4 i 4 i . . i i -< , . .
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- a. : i i i t i
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* . + i i i i i i .
. I L l
' . I 6 i o h i p .
z i' ACCEPTABLE "i 50
!.'.. OPERATION !l i
3> t , 6 i ! i 5 ', i. o us i ' i 6 , i
' t gg '
e it ' e I ! I I i f I i 0 20 33 40 50 80 70 30 90 100 PERCENT OP RATED THERMAL POWER DOSE EQUIVALDTT I-131 Primary Coolant Specific Activity Limit Versus Percent of RAT ED THERMAL POWER with the . Pr imary Coolant Specific Activity
> 1.0 uCi/gr am Dose Equivalent I-131.
O BEAVER VALLEY UNIT 2 3/4 4-32
- < , e .e- - -, -L --M_-
- A- -. --e =- m --A- - - -- a i
3/4.4 REACTOR C001 ANT SY ST EM 3/4.4.9 PRESSURE / TEMPERATURE LIMIT S l REACTOR COOLANT SY ST EM i LIMITING CONDITION FOR OPERATION i
- 3.4.9.1 The Reactor Coolant System (except the ptessurizer) !
temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and j inservice leak and hydrostatic testing with: f
- a. A maximum heatup of 100*F in any one hour period,
- b. A maximum cooldown of 100'1 in any one hour period, and
. c. A maximum temperature change of < 5*F in any one hour period,* during I hydrostatic testing operations aFove system design pressure. t APPLICABILITY: MOD E S 1, 2 * , 3 , 4, a nd 5. ACT ION : ! With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an analysis to 4 determine the ef fects of the out-of-limit condition on the fracture
- toughness properties of the Reactor Coolant System; determine that the
] Reactor Coolant System remains acceptable for continued operations or be in i at least HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within th$#Eollowing 30 hours . i SJRVEILLANCE REQUIREMENTS ' i 4.4.9.1 : i a. The Reactor Coolant System temperature and pressure shall be I l determined to be within the limits at least once per 30 minutes during ! system heatup, cooldown, and inservice leak and hydrostatic testing ! oper at ions .
- 1
- b. The Reactor Coolant System temperature and pressure conditions shall
, be determined to be to the right of the criticality limit line within ! 15 minutes prior to achieving reactor criticality. 1
- See Special Test Exception 3.10.3.
- O 3/4 4-33 BEAVER VALLEY UNIT 2 I
i l . _ _ . . _ . _ _._ _._ ,
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.9 PRES!URE TEMPERABURT LIMIT S REACTOR COOLANT SY ST EM SURVEILIANCE REQUIREMB(T S
- c. The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these exaninations shall be used to update Figures 3.4-2 and 3.4-3.
O O BEAVER VALLEY UNIT 2 3/4 4-34
4 FIGURE 3.4-2 BEAVER VALLEY UNIT 2 REACTOR C001 ANT SYSTDI HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY O MAT ERIAL PROPERTY BASIS Controlling Material : Plate Metal Copper Content : Conservatively Assumed to be 0.10 wt% Phosphorus Content : 0.010 wt% RT Initial : 60*F. RT A f ter 10 EFPY : 1/4T, 139'F. 3/4T, ll4*F CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AtlD C0!iTAI!iS MARGIfiS OF 10*F Arid 60 PSIG FOR POSSIBLE I!iSTRUMEliT ERRORS me .. I I il liillilil litilflit til'iljil tiilliit' I l ' ll lillill 's l't!Illit lij ilill ilijllitt i I ' I Illill i li lillill ill iltil liilitii; i ' il i I tilll ll I lililli i l ill i lifillill
' IjIIIIi I III hES I II 'II I Il IIIIIIII' LIMIT lli W III! I II I '/l / ! Il! - '
lil lil!!
;l I lli Thil '
l/ i / II' lli ill I ll1 ill I i I V I
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'1 1l lll Il i lill!l Il '
l ' l f. ' (1) llillilii
- ll i l liiill I i I/ / I I 11ll.11111
= 1 I i ll l l
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- r i J
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,a_ey p-- CRITICALITY LIMIT
~ ~ ~'~ ; ' '
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gg79p_ . f, , I I I BASED Off INSERVICE CURVE i I > / -- - 1 I I" I Ii HYDR 03TATIC TEST TEMPERATURE (279'F) li l ii FOR THE SERVICE l ! Ic l l1 PERIOD UP TO 10 EFFY
! i l' l '
1 I L 1
,,, l i j e.* m.e . m.e m.e see.. m. , 1 Q BEAVER VALLEY UNIT 2 Imercarts igw anatunt ots.F:
3/4 4-35 l 1 l l
FIGURE 3.4-3 BEAVFR VALLEY UNIT 2 REACTOR C00LAfff SYSTDi COOLDOWN LIMITATION S APPLICABLE UP TO 10 EFPY O MATERIAL PROPERTY BASIS Controlling Material : Plate Metal Copper Content : Conservatively Assumed to be 0.10 wt% Phosphorus Content : 0.010 wt% RT NUT Initial : 60*F. RT A f ten. 10 EFPY : 1/4T, 139'F. NUT 3/4T, 114*F CURVE APPLICABLE FOR C00LOCWN RATES UP TO 100*F/HR FCR THE: PERIOC UP TO 10 EFFY Afic CONTAIN! MARGINS OF 10*F AND 60 PSIG FO - l l P05SIBLE INSTRUMENT ERRORS llll1iliiI tii1'tIlll 1I ifeii ililiilifl*I i t ! ] IIilili!i ilii 6III IIIi'i!lI I IIIIilll'l i t i ii' lillilti; lilllllli Illilleil ftilliti: {.t- t il I lilllit tillillll 11ilitill 1111111I' Iili'sf fI lll111 Illlillll lllll1ll1 tilllilli lill*lIll llllll lililllli 1llllll11 li l l l l l l l ! 'l!! ill! 11 lllilll llll11111 1illlllll fillllllli lifl.!!ii i ili til l14111111 l111111ll; lillfillfil's!'f.: i llIll illl 111841111 1816 6 l l 11/ lillllilllillLli!Il Illillill 111ll1111 tiltlill) 11ll16l111116t'111! 3 !! Ililll 11:1111111 l 1 1 1 l l l l/e llillllli lIllllIili I l lilli ll l Illt i lilli Il Ill1lllll lillllill
; i 11 1111111 til l l' Il tilll/ I 8illllli Illlillli 2 - l ll Il il lit ll jil / I l11111111 lilllilli E i l l 'l i l til li Il I t il /11 i fililli illfifill l- -
ll l lIlli l l l i l/ Ill l 1; Il lll Illllli!l S i ' jill ! 11/ ill if I ll1 lititilti I l ! l /lilll il ll l i l- 11111111I w l l 111lllll1 3 l l [ lIllt il ll I , l l l l/lllilli il lll l 1 111ll16lll 1 i fi l 11 Illllll'l lillllliI l // I ll lillll l lilll;ll! ,
! LdN ' l'Ill' I Ill'Illil '
CQQ M H MTE* III I ME il I J'I I lilllllii 181111111 (*F/HR) 0 g_ 1==4 ' MR,# Ii -"# llll I I l I'
'Ill
''I Il1!ll}} !'i'I'.
(l1IIIiil i !Ill 181333111 till'illt 40 " ' 'Td
]JC 'l Illi i 161111161 Illililti 60'E '
.- " I Il fililli i il l' i 100 l '
t lil I i 18 I I 'I litillill iliiiiiii il l8 ll Ill I I il I I i fililliti iililllii e.e see.e l ses.e ses.o ses.s een.e e.e IsetCaf ts 7tfeetnafust t StG.F j BEAVER VALLEY UNIT 2 3/4 4-36 I l l
O O O TABLE 4.4-5 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE ESTIMAT ED CAPSULE VESSEL LEAD CAP alLE FLU ENC E-
'IDCAT ION FACTOR WITHDRAWAL TIME ( EFPY) ( 2 NUMBER. __N/CM )
U~ 343* 3.5 I ST Refueling 0.8 x 10I9 l V 107* 3.5 3 1.9 x 1019 X 287* 3.5 6 3.9 x 10 19(*} W l10* 2.9 11 5.9 x 10 gg(b) Y 290* 2.9 20 10.7 x 1019 Z 340* 2.9 St andby --- Approximate fluence at 1/4T vessel wall thickness at e nd-o f-l i fe . ( Approximate fluence at vessel inner wall at e nd-o f-l i fe . BEAVER VALLEY UNIT 2 3/4 4-37
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMIT S PRESSURIZER LIMITING CONDITION FDR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of 100*F in any one hour period,
- b. A maximum cooldown of 200*F in any one hour period, and
- c. A maximum spray water temperature dif ferential of 320*F.
APPLICABILITY: At all times. ACT ION : With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an analysis to determine the ef fects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. O SURVEILIANCE REQUIREMl!NT S 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at le ast once per 30 minutes during system heatup or cooldown. The spray water temperature dif ferential shall be determined to be within the limit at least once per 12 hours during steady state operations. BEAVER VALLEY UNIT 2 3/4 4-38
b 3/4.4 REACTOR COOLANT SY ST EM . 3/4.4.10 STRUCTURAL INT EGRrrY l C:) l LIMITING CONDITION FOR OPERATION
\
3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 com- i ponents shall be maintained in accordance with Specification 4.4.10. l APPLICABILITY: i All MODES. l ACT ION :
- a. With the structural integrity of any ASME Code Class I component (s) not conforming to the above requirements, restore the structural integrity of the af fected component (s) to within its limit or isolate i the af fected component (s) prior to increasing the Reactor Coolant !
System temperature more than 50*F above the minimum temperature required by NDT considerations. !
- b. With the structural integrity of any ASME Code Class 2 component (s) i not conforming to the above requirements, restore the structural i integrity of the af fected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
( c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the .af fected component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicab le .
SURVEILLANCE REQUIREMl!NT S 4.4.10 Each ASME Code Class 1, 2, and 3 component shall be demon-l strated OPERABLE in accordance with Specification 4.0.5. l s i i BEAVER VALLEY UNIT 2 3/4 4-39 l1 i i
3/4.4 REACTOR C001 ANT SY ST EM 3/4.4.11. RELIEF VALVES (3' LIMITING CONDITION FOR OPERATION 3.4.11 Two power operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. A CT ION :
- a. With less than 2 PORV(s) operable, within I hour eith?.c restore two PORV(s) to OPERABLE status or close the associated block valves (s) and remove power from the block valves (s); otherwise, be in at le as t HOT STANDBY within the next 6 hours and in COID SHITTDOWN within the fol-lowing 30 hours,
- b. With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to OPERABLE status or close the block valves (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SRUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENT S 4.4.11.1 Each PORV shall be demonstrated OPERABLE:
- a. At least once per 31 days by performance of a CRANNEL DiECK of the position indication, excluding valve operation, and
- b. By performance of a CRANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
4.4.11.2 Each block valve shall be demonstrated OPERABLE at le as t once per 92 days by operating the valve through one complete cycle of full travel. 4.4.11.3 The power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves 1 l through a complete cycle of full travel. O BEAVER VALLEY UNIT 2 3/4 4-40 11
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS O LIMITING CDNDITION FDR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. Between (later) and (later) gallons of borated water,
- c. Between 1900 and 2100 ppm of boron, and
- d. A nitrogen cover-pressure of betscen (later) and (later) psig.
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
O b. With one accumulator inoperable due to the' isolation valve being clos'ed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1. Verifying, by the absence ~ of alarms, the contained borated water water volume and nitrogen cover pressure in the tanks, and
- 2. Verifying that each accumulator isolation valve is open.
*. Pressurizer Pressure above 1000 psig.
O BEAVER VALLEY UNIT 2 3/4 5-1 2
. - _ . . . - - . . _ _ _ . .,, ,, -n.,-, ,
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS O SURVEILLANCE REQUIREMENTS
=
- b. At least once per 31 days and within 6 hours af ter each solution volume increase of greater than or equal to 1 percent of tank volume by verifying the boron concentration of the accumulator solution.
- c. At least once per 31 days when the RCS pressure is above 2000 psig by verifyir.g that power to the isolation valve operator control circuit is disconnected by removal of the plug in the lock out jack from the circuit.
- d. Verifying at least once per 18 months that each accumulator isolation valve opens automatically under each of the following conditions:
- 1. When the RCS pressure exceeds 2000 psig.
- 2. Upon receipt of a Safety Injection test signal.
4.5.1.2 Each accumulator water level and pressure alarm channel shall be demonstrated OPERABLE:
- a. At least once per 31 days by the performance of a CHANNEL P1JNCTIONAL TEST.
O
\/ b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
4 O BEAVER VALLEY UNIT 2 3/4 5-2
i 3/4.5 EMERGDiCY CORE COOLING SYST EMS 3/4.5.2 ECCS SUBSYST EMS - T > 350*F
._.__ avg - ,
LIMITING CONDITION mR OPERATION 3.5.2 Two separate and indepe ndent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Low head safety injection pump,
- c. One OPERABLE recirculation spray pump
- capable of supplying the safety injection flow path during recirculation phase, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suc-t io n to the cont ainme nt sump during the recirculation phase of opera-tion.
i APPLICABILITY: MODES 1, 2, and 3. ACT ION : s
- a. With one ECCS subsystem inoperable , res tore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours ,
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and s'ubmitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated .actu-ation cycles to date.
- Recirculation spray pump 2RSS*P21C or 2RSS*P21D. y BEAVER VALLEY UNIT 2 3/4 5-3
. - - _- _ . . _ . ~ . _ _ _
i 3/4.5 EMERGENCY CORE COOLING SYST_ EMS 3/4.5.2 ECCS SUBSYST EMS - T g
> 350*F I
SURVEILIANCE REQUIREMDIT S i L 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: lI'
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operator control circuits disconnected by removal of the plug in the lock out cir cuit from each circuit. t l
r Valve Number Valve Function Valve Position _
- a. 2 SIS *MOV8889 IHSI to hot legs Closed 3
- b. 2 SIS *MOV869A HMSI to hot leg Closed !
- c. 2 SIS *MOV869B HHSI to hot leg Closed
- d. 2 SIS *MOV836 HHSI to cold leg Closed i
- e. 2 SIS *MOV841 HHSI to cold leg Ope n
- f. 2CHS*MOV8132A HHSI pump disch X-conn Ope n
- g. 2CH S*MOV8132B HMSI pump disch X-conn Open i
- h. 2CH S*MOV8133A HHSI pump disch X-conn Open
- i. 2CHS*MOV8133B HHSI pump disch X-conn Open
- b. At least once per 31 days on a STAGGERED TEST BASIS by:
(Q./
- 1. Verifying that each centrifugal charging pump:
- a. Starts (unless already operating) from the control room,
- b. Develops a discharge pressure of > 2437* psig on recirculation flow,
- c. Operates for at le as t 15 minutes.
- 2. Verifying that each low head safety injection pump:
- a. Starts (unless already operating) from the control room,
- b. Develops a discharge pressure > 103* psig on recirculation flow,
- c. Operates for at le as t 15 minutes .
- 3. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
*To be verified during Pre-Operational Testing BEAVER VALLEY UNIT 2 3/4 5-4
3/4.5_ EMERGDiCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T > 350 *F avg - O SURVEILIANCE REQUIREMDIT S
- 4. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in pos it io n, is in its correct ' position.
- 5. Verifying that each ECCS subsystem is aligned to receive electrical power from separate OPERABLE emergency buses.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1. For all accessible. areas of the cont ainment prior to establishing containment integrity, and
- 2. Of the areas af fected within containment at the completion of each cont ainment entry when containment integrity is established,
- d. At- least once per 18 months by:
- 1. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens , etc.) show no evidence of structural distress or corrosion.
- e. At le as t once per 18 months, during shutdown, by:
- 1. Cycling each power operated -(excluding automatic) valve in the flow path that is not testable during plant oper at ion, through at le as t one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection signal.
- 3. Verifying that the centrifugal charging pump and low head safety injection pumps start automatically upon receipt of a safety injection signal.
- f. The containment recirculation spray subsystem shall be demonsteated 1 OPERABLE per the applicable portions of Specification 4.6.2.2.
O BEAVER VALLEY UNIT 2 3/4 5-5
3/4.5 EMERGDICY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYST EMS - T,y < 350*F O LIMITING CONDITION FT)R OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE low head strety injection pump,
- c. One OPERABLE recirculation spray pump
- capable of supplying the safety injection flow path during recirculation phase, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION : () a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours.
- b. In the event the ECCS is actuated and injects water- into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total -accumulated actu-ation cycles to date.
I l 1 SURVEILLANCE REQUIREMEWTS l l 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements. of 4.5.2. l l
- Recirculation spray pump 2RSS*P21C or 2RSS*P21D. ]
BEAVER VALLEY UN TT 2 3/45-6
3/4.5 EMERGENCY CORE COOLING SYSTEMS O LIMITING CONDITION FOR OPERATION 3.5.4 This Technical Specification intentionally blank. 4 O a { i h 1 1 l i i l .i O BEAVER VALLEY UNIT 2 3/4 5-7 i
- - - , . . , , . . . - . , . - . . . ~ - ,- ,.,,...,,,n. -n,, . , ~ , ,, - - , , ,,, _ . . . , , . --,- , . , , - - , , . , , . , -, , ' '
3/4.5 EMERGENCY CORE COOLING SYSTEMS O LIMITING CONDITION FOR OPERATION 3.5.5 This Technical Specification intentionally blank O . O BEAVER VALLEY UNIT 2 3/4 5-8
3/4.6 CONTAINMDIT SYSTEMS . 3/4.6.1 PRIMARY CONTAINMENT CONTAINMDIT INTERITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMDIT INT ERITY shall be maint ained. APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION : Without primary CONTAINMDfr INT EGRITY, restore CONTAINMDIT INT EX;RITY within one hour or be in at least HOT STANDBY within the next 6 hour s and in COLD SitTTDOWN within the following 36 hours. SURVEILIANCE REQUIREMDiT S 4.6.1.1 Primary (X)NTAINMENT INTEGRITY shall be demonstr ated:
- a. At least once per 31 days by verifying that:
I
- 1. All penetr ations not capable of being closed by OPERABLE cont ainsent automatic isolation valves and required to be closed during accident conditions are closed by valves , blind flanges, or deactivated automatic valves secured in their positions, except as provided in T able 3.6-1 of Specification 3.6.3.1.
- 2. All equipment hatches are closed and sealed,
- b. By verifying that each containment air lock is OPERABLE per Specificat ion 3.6.1.3.
i a BEAVER VALLEY UNIT 2 3/4 6 .
3/4.6 CONTAINMDIT SY ST EMS 3/4.6.1 PRIMARY CONTAINMDIT n v CONTAINMDIT LEAKAGE LIMITING CONDITION EUR OPERATION 3.6.1.2 Cont ainment leakage rates shall be limited to:
- a. An overall integrated leakage rate of:
i
- 1. Less than or equal to La, (0.10) percent by weight of the I containment air per 24 hours at Pa (44.7 psig), or
- b. A combined leakage rate of less than or equal to 0.60 La for all .
penetrations and valves subject to Type B and C tests, as ident ified in Table 3.6-1, when pressurized to Pa (44.7 psig). APPLICABILITY: f MODES 1, 2, 3, and 4. t ACTION : With either (a) the measured overall integrated containment leakage rate. exceeding 0.75 La, or (b) with the measured combined leakage l () rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the leakage rate (s) to within the limit (s) prior to increasing the Reactor Coolant System temperature above 200*F. .i SURVEILIANCE REQUIREMDIT S 1 4.6.1.2 The containment leurage rates shall be demonstrated at t he following test schedule and shall be determined in conformance with the criteria specified in Appendix J. of 10 CFR 50* using the methods and provisions of AN SI N45.4-1972: i
- a. A Type A test (Overall integrated Containment Leakage Rate) shall be conducted tc 40 + 10 month intervals during : hutdown at Pa (44.'
~
psig).
- b. If any periodic Tyve A test fe ls to meet .7 5 La , t he te s t schedule for subsequent Type A tests shall be reviewed and approved by the Commis s ion. If two consecutive Type A tests- fail to meet . 7 5 La , a Type A test shall be performed at le ast every 18 months until two consecutive Type A tests meet .75 La, at which time the above test schedule .may be resumed. ,
. t J
- Exemption to 10CFR50 Appendix J, Section III.D l(a) i
- BEAVER VALLEY UNIT 2 3/4 (~2 1
i
3/4.6 CONTAINMDIT SYSTEMS 3/4.6.1 PRIMARY CONTIANMENT CONTAINMDir LEAKAGE i SURVEILLANCE REQUIREMENT S __p _
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A test data is within 0.25 La.
- 2. Has a duration suf ficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
- 3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at Pa (44.7 psig).
- d. Type B and C tests shall be conducted with gas at Pa * (44.7 psig) at l}
intervals no greater than 24 months except for tests involving.
- 1. Air locks, d 2. Penetr ations using continuous leakage monitoring systems, and
- 3. Valves pressurized with fluid from a seal system.
- e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
- f. Leakage from isolation valves that .are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J.
Section III.C.3, when determining the combined leakage rate provided
'the seal system and valves are pressurized to at le ast 1.10 Pa (49.2 psig) and the seal system capacity is adequate to maintain system pressure for at least 30 days.
- g. All test leakage' rates shall be calculated using observed data converted to absolute values. Error analysis shall be performed to determine the inaccuracy of the measured leakage rates due to maximum me asur eme nt accuracy and instrument repeatability; the measured leakage rates shall be adjusted to include the measurement error.
- Applicable valves may be tested using water as the pressure fluid in }
accordance with the Incervice Testing program.
- BEAVER VALLEY UNIT 2 3/46-3
3/4.6 CONTAINMENT SY ST EMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMDIT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:
- a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at le ast one air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 La at Pa, (44.7 psig).
APPLICABILITY: Modes 1, 2, 3, and 4. ACT ION :
- a. With one containment air lock door inoperable:
- 1. Maintain the associated OPERABLE air lock door closed and either restore the associated inoperable air lock door to OPERABLE status within 24 hours or lock the associated OPERABLE air lock door
() closed.
- 2. Operation may then continue entil performance of the next required overall air lock leakage test provided that t he associated OPERABLE air lock door is verified to be locked closed at least once per 31 7
days.
- 3. Otherwise , be in at leas t HOT STANDBY within the next 6 hours and in COLD SHVIDOWN within the following 30 hours.
- 4. The provisions of Specification 3.0.4 are not applicab le .
- b. With a containment ale lock inoperable , except as the result of an inoperable air lock door, maintain at le as t o ne air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, i
BEAVER VALLEY UNIT 2 3/4 6-4 I I
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT AIR LOCKS SURVEILLANCE-REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours following each containment entry, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying no detectable seal leakage when the gap between the door seals is pressurized for at least 2 minutes to:
- 1. Personnel airlock > 44.7 psig
- 2. Emergency air lock > 10.0 psig or, by quantifying the total air lock leakage to insure the requirements of 3.6.1.3.b are met.
- b. By conducting overall air lock leakage tests, at not less than P, (44.7 psig), and verifying the overall air lock leakage rate is within its limit:
- 1. At least once per 6 months, f and O 2. Upon completion of maintenance which has been performed on the air lock that could affect the air lock sealing capability.*
- c. At least once per 18 months during shutdown by verifying:
- 1. Only one door in each air lock can be opened at a time, and
- 2. No detectable seal leakage when the volume between the emergency air lock shaft seals is pressurized to greater than or equal to 44.7 psig for at least 2 minutes.
# The provisions of Specification 4.0.2 are not applicable.
- Exemption to Appendix J of 10 CFR 50.
n v BEAVER VALLEY UNIT 2 3/4 6-5
3/4.6 CONTAINMENT SY ST EMS 3/4.6.1 PRIMARY CONTAINMENT INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary contairunent internal pressure shall be maintained greater than or equal to 9.0 psia and within the acceptable operation range (below and to the lef t of the RWST water temperature limit lines) shown on Figure 3.6-1 as a function of RWST water temperature and service water temper atur e . APPLICABILITY: MODES 1, 2, 3, and 4. ACTION : With the contairunent internal pressure less than 9.0 psia or above the applicable RWST water temperature limit line shown on Figure 3.6-1, restore the internal pressure to within the limits within 1 hour or be in at i least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours . SURVEILIANCE REQUIREMENT S I I I 4.6.1.4 The primary contairunent internal pressure shall be determined to be within the limits at least once per 12 hours. l
- O BEAVER VALLEY UNIT 2 3/46-6 i
i
l FIGURE 3.6-1 l MAXIMUM ALLOWABLE PRIMARY CONTAINMINT AIR PRESSURE l l VERSUS SERVICE WATER T EMPERATURE AND tlWST WAT ER T EMPERATURE l
~
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. g j "il Q)88TAINHENT TDtPERATURE DOES NOT LIE BEIDW l g '85'F OR ABOWE 105'F
/A/
a _. y i % r SET POINT VAIER IN CONTAINMENT %ACUUM AI. ARM _ /M \ - .i.__ -:. - e INSTRUMDITAT10el EHOUI.D BE SET AT 0.25 PSI 3 t \ .
. . BELOW THE BWST TEMPERATUPE LlHIT CURVE A!41) y 39.0 rsIA j,2 .%._ i
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i f !* t i il i e:' iN NY 1 -{! . i i, , serv 1Cs WATER TEMPERATURE (*F) (3 V BEAVER VALLEY UNIT 2 3/4 6-7 i i
3/4.'6 CONTAINMDIT SYST EMS 3/4.6.1 PRIMARY CONTAINMENT O ^'a 'z"Pra^tuat LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall be maintained less than or equal to 105'F and greater than or equal to 85'F. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION : With the containment average air temperature greater than 105'F or below 85* , restore the aver age air temperature to within the limit
! shown within 8 hours or be in at least Har STANDBY within-the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILIANCE REQUIREMPNTS l 4.6.1.5 The primary containment average air temperature shall be the
- arithmetical average of any five temperatures from the following locations and shall be determined at least once per 24 hours
Incation Elevation RC Annulus . 730'-7" RHR Cub 801'-6" SG 21C Cub 701'-5" SG 21C Cub 745' -6" SG 21C Cub 865'-0" , Pzr Cub 802'-0" RC Annulus 740'-7" Pzr Cub Stairway 746'-0" SG 21B Cub 701'-6" i SG 21B Cub 865'-0" SG 21B Cub 730'-0" RC Annulus 736'-11" Reactor Head Stor age Area 802'-0" RC Annulus 701'-6" RC Annulus 777'-4" SG 21A Cub 701'-6" SG 21A Cub 865'-0" SG 21A Cub 726'-6" RC Annulus 740'-10" SG 21C Cub 727'-0" O BEAVER VALLEY UNIT 2 3/46-8
- - - - . , , -n..r-r,- r--,- ---w_, - -,, , -r,---,.v-n--,
3/4.6 CONTAINMDR SY ST EMS 3/4.6.1 PRIMARY CONTAINMDU CONTAINMDIT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structur al integrity of the cont airunent shall be maintained at a level consistent with clw acceptance criteria in Specification 4.6.1.6.1. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION : With the structural integrity of the containment not conforming to the above requirements, restore the structur al integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILIANCE REQUIREMDIT S i i i si i i i n , 4.6.1.6.1 Liner Plate and Concrete - The structural integrity of the containment liner plate and concrete shall be determined during the shutdown for each Type A contairement leakage rate test (reference Specification ! 4.6.1.2) by:
- a. A visual inspection of the accessible surf aces and verifying no apparent changes in appearance or other abnormal degr adat ion.
- b. A visual inspection of accessible containment liner test channels prior to each Type A containment lerkage r ate test.
! c. A visual inspection of the done area prior to each Type A contairunent l leakage rate test to insure the integrity of the protective coating. ,
- 4. 6.1. 6. 2 Reports - An initial report of any abnormal degr adation of the contairusent structure detected during the above required tests and inspections shall be made within 10 days af ter cumpletion of the . surveillance !
requirements of this specification, and the detailed reports shall be ! submitted- pursuant to Specification 6.9.2 within 90 days af ter cornpletion. 3 This report shall include a description of the condition of the liner plate i
,and concrete, the inspection procedure, the tolerances on cracking and the coreeetive actions taken.
O BEAVER VALLEY UNIT 2 3/46-9 i
. _ _ ._. .= -
3/4.6 CONTAINMENI SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS O: CONTAINMENT QUENCH SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two separate and independent containment quench spray subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: With one containment quench spray subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment quench spray subsystem shall be demonstrated OPERABLE: + O' '
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Starting each s} ray pump.
- 2. Verifying, that on recirculation flow, when tested in accordance with the requirements of Section 4.0.5, each quench spray pump develops a discharge pressure of > 172* psig at a flow of > 3000*
g Pm.
- 3. Verifying that each spray pump operates for at least 15 minutes.
- 4. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
- 5. Verifying that each valve (manual, power operated or automatic) in l the flow path that is not locked, sealed, or otherwise secured in position, is in its correct posit ion.
- 6. Verifying the temperature of the borated water in the refueling water storage tank is within the limits shown on Figure 3.6-1.
*Will be verified during Pre-Operational Testing.
l BEAVER VALLEY UNIT 2 3/4 6-10
3/4.6 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS O CONTAINMENT QUENCH SPRAY-SYSTEM SURVEILLANCE REQUIREMENTS
- b. At least once per 18 months during shutdown:
- 1. Cyciing each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
- 3. Verifying that each spray pump starts automatically on a test signal. ,
- c. At least once per 5 years by performing an air or smoke flow test through -each spray header and verifying each spray nozzle is unobstructed.
- O a
l l O BEAVER VALLEY UNIT 2 3/4 6-11
3/4.6 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYST EMS O cosr^ a">=r arctacutir'o" sea ^v srsrex LIMITING CONDITION FOR OPERATION 3.6.2.2 Four separate and independent cont ainme nt recirculat ion spray subsystems, each composed of a spray pump, associated heat exchanger and flow I path shall be OPERABLE. APPLICABILITY: 1 MODES 1, 2, 3, and 4 ACTION:
- a. For subsystem containing recirculation spray pump 2RSS*P21A or 2RSS* P21B: with one containment recirculation spray subsystem inoperable , restore the inoperable subsystem to OPERABLE status within 7 days or be in at le as t HOT STANDBY within the next 6 hours; restore the inoper ab le i spray system to OPERABLE status within the next 48 hours or be in COLD ,
SHUTDOWN within the following 30 hours. -
- b. For subsystem containing recirculation spray pump 2'RSS*P21C or 2RSS*P21D: see Specification 3.5.2 or 3.5.3.
1 (i . \_/ SURVEILIANCE REQUIREMENT S 4.6.2.2 Each containment recirculation spray subsystem shall be demonstrated OPERABLE.
- a. At least once per 31 days on a STAGGERED TEST BASIS, by:
I 1. Manually starting each spray pump and verifying the pump shaf t rotates.
- 2. Verifying correct position of all accessible manual valves not locked, sealed or otherwise secured in position, and all remote or automatically operated valves in each recirculation spray subsystem flow path.
- 3. Cycling each testable power-operated or automatic valve in the flow path through at least one complete cycle of full travel'.
- m
- 4. Verifying that each valve (manual, power-operated or automatic) in the flow path that, is not locked, sealed or otherwise secured in posit ion, is in its correct pos it ion.
O BEAVER VALLEY UN rr 2 3/4 6-12
3_/4.6 CONTAINMENT SY ST EMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT RECIRCULATION ' SPRAY SYSTEM SURVEILLANCE REQUIREMENT S i
- b. At least once per 18 months during shutdown, by verifying that on a Containment Pressure HI-3 (High/High-High) test signal, each recirculation spray pump' starts automatically af ter a 628 + 3 second ,
a delay. a
- c. At least once per 18 months, during shutdown, by verifying that on recirculation flow, each recirculation spray pump develops a discharge pressure of }; 96* psig at a flow of }; 3500* gpe,
- d. At le as t once per 18 months during shutdown, by:
- 1. Cycling each power-operated (excluding automatic) valve in the flow path not testable during plant operation, through at least one
.I complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
-3. Initiating flow through each Service Water subsystem and its two x associated recirculation spray heat exchangers, and verifying a I flow rate of at le as t 8000 gpm.
- e. At least once per 5 years, by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobs tr uct ed.
i
- To be verified during pr'e-operational testing, O'~' BEAVER VALLEY UN TT 2 3/4 6-13
l 3/4.6 CONTAINMEWT SY ST EMS _3/4.6.2 DEPRESSURIZATION AND COOLING SYST EMS ; CHEMICAL ADDITION SYST EM t LIMrTING 00NDITION FOR OPERATION I i 3.6.2.3 The chemical addition system shall be OPERABLE with: l
- a. A chemical addition tank containing at least 8,500 gallons of between 23 and 25 percent by weight NaOH solution, and
- b. Two chemical injection pumps each capable of adding NaOH solution from [
the chemical addition tank to a containment quench spray system pump ' flow, i APPLICABILITY: l MODES 1, 2, 3, and 4. ! I ACT ION : ; With the chemical addition system inoperable, restore the system to OPERABLE status within 72 hours or be in HOT STANDBY within the " next 6 hours; restore the chemical addition system to OPERABLE status within t he next 48 hours or be in COLD SHVIDOWN within the next 36 hours. i l SURVEILLANCE REQUIREMEKIS 4.6.2.3 The chemical addition system shall be demonstrated OPERABLE: r a.- At le as t once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed,
- or otherwise secured in position, is in its correct pos it io n.
- b. At least once per 31 days on a STAGGERED TEST BASIS by-
- 1. Starting each injection pump.
- 2. Verifying that each injection pump operates for at le as t 15 minutes.
- 3. Cycling each testable power-operated or automatic valve in the flow path through at least one complete cycle of full travel.
- 4. Verifying that on recirculation, each injection pump develops a flow between 55 and 60 gpm. l{
O BEAVER VALLEY UNIT 2 3/4 6-14 f
3/4.6 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS () CHEMICAL ADDITION SYSTEM SURVEILLANCE REQUIREMENTS ' c. At least once per 6 months by: l 1. Verifying the contained solution volume in the tank, and
- 2. Verifying the concentration of the NaOH solution by chemical analysis,
- d. At least once per '18 months, during shutdown, by:
- 1. Cycling each valve in the chemical addition system flow path that is not testable during plant operation, through at least one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
1 I O i
?
l i 4 BEAVER VALLEY UNIT 2 3/4 6-15
- _ __- . _ . _. ___._.~-
3/4.6 CONTAINMDIT SY ST EMS 3/4.6.3 CONTAINMDIT ISOLATION VALVES O V LIMITING CONDITION FOR OPERATION 3.6.3.1 The containment is'olation valves specified in T able 3.6-1 l} shall be OPERABLE with isolation times as shown in Table 3.6-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION : With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable, either:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate the affected penetration within 4 hours by use of at le as t one deactivated automatic valve secured in the isolation position, or
- c. Isolate the affected penetration within 6 hours by use of at le as t one closed manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours and in COLD O satrroow" witate the reitewi = so ae r -
SURVEILIANCE. REQUIREMDIT S 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be lI
, demonstrated OPERABLE:
- a. At least once per 92 days by:
- 1. Cycling each OPERABLE power-operated or automatic valve testable during plant operation through at least one complete cycle of full tr avel .
- 2. Cycling each weight or spring loaded check valve testable during plant operation through one complete cyc'le of full travel and verifying that each check valve remains closed when the dif ferential pressure in the direction of flow is < 1.2 psid and opens when the dif ferential pressure in the direction of flow is
> 1.2 psid but less than 6~0 paid.
't O BEAVER VALLEY UNIT 2 3/4 6-16 4 I
3/4.6 CONTAINMDIT SY ST EMS 3/4.6.3 CONTAINMENT ISOLATION VALVES O V SURVEILIANCE REQUIREMDrr S
- b. Immediately prior to returning the valve to service af ter maintenance ,
repair, or replacement work is per formed on the valve or its associated actuator, control or power circuit by performance of the applicable cycling test, above, and verification of isolation' time. 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 shall be . l1 demonstrated OPERABLE during COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a Phase A concainment isolation test signal, each Phase A isolation valve actuates to its isolation posit ion.
- b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
- c. Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation posit ion.
- d. Cycling each power operated or automatic valve through at least one complete cycle of full travel and measuring the isolation time.
- e. Cycling each weight or spring loaded check valve not tes t ab le dur ing
()
\ss plant operation, through one complete. cycle of full travel and verifying that each check valve remains closed when the dif ferential pressure in the direction of flow is < l.2 psid and opens when the differential pressure in' the direction of flow is > 1.2 psid but les s than 6.0 paid.
- f. Cycling each manual valve not locked, sealed, or otherwise secured in the closed position through at least one complete cycle of full tr ave l, p
() BEAVER VALLEY UN rr 2 3/4 6-17 i w - ,
O O O TABLE 3.6-1 CONTAINMENT PENETRATION S Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve T ime ( Sec ) - 1 Comp Cool from Res (1)(B) 2CCP*MOV157-2 50 (1)(B) 2CCP*MOV157-1 28 Heat Exch 2CCP*RV105 N/A l1 2 Comp Cool to Res (1)(B) 2CCP*MOV150-2 50 (1)(B) 2CCP*MOV150-1 28 Heat Exch 2CCP*RV102 N/A l 4 Comp Cool to Res (1)(B) 2CCP*MOV151-2 50 (1)(B) 2CCP*MOV151-1 28 Heat Exch. 2CCP*RV103 N/A 5 Comp Cool from Res (1)(B) 2CCP*MOV156-2 50 (1)(B) 2CCP*HOV156-1 28 Ileat Exch 2CCP*RV104 N/A l1 6 SPARE 7 High Head (3)(2) 2 SIS *83 N/A (3)(2) 2 SIS *MOV869A 10 (4) l Safety Injection 9 SPARE 11 Instrument Air (A) 2 LAC *HOV133 30 (A) 2 IAC*MOV134 30 13 SPARE 14 Chill & Service Wtr (B) 2 SWS*MOV153-2 22 (B) 2 SWS*HOV153-1 22 to Cont . Air Recirc 2 SW S*RV15 3 N/A Cooling Coils 15 CHARGING ( 3)(2) 2 CH S*31 N/A (3)(2) 2CH S*MOV289 10 l! 16 SPARE HEAVER VALLEY UNIT 2 3/4 6-18
,- . - _ _ , -,wy,w, -
, , , , , , - . - - - , , , - .c-. . ~ ,- ,, .
4 O O O TABLE 3.6-1 (Cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve Time ( Sec) 17 High Head (3)(2) 2 SIS *84 N/A (3)(2) 2 SIS *MOV8698 10 (4) l1 Safety Injection 19 Seal Water from (A) 2CHS*MOV378 10 ( A) 2CH S*MOV381 10 l Re -tor Coolant 2CH S*473 N/A 3 7 :,s es Safet) . .j . c.t ion 2 SI S*42 N/A (1) 2 SIS *41 N/A Accumef4 : Makeup 2 SI S*RV130 N/A 1 21 *
. t11 a Service Wtr (B) 2 SWS*MOV155-2 22 (B) 2SWS*MOV155-1 22 from Cont . Air 2 SWS*RV15 5 N/A Recire Cooling Coils 22 SPARE 23 SPARE 24 Residual Heat Removal 2 RH S*107 N/A 2 RH S*15 N/A to Refueling Water Tank 2 RH S*RV100 N/A 25 Chill & Service Wtr (B) 2 SWS*MOV154-2 22 (B) 2 SWS*MOV154-1 22 fr om Co nt . Air Recirc 2 W S*RV154 N/A Cooling Coils l
27 Chill & Service Wtr - ( B) 2 SWS*MOV152-2 22 (B) 2 SWS*MOV152-1 22 to Cont. Air Recire 2 WS*RV152 N/A Cooling Coils l 28 Reactor Coolant (A) 20lS*A0V200A 10 (A) 2OIS*A0V204 10 Letdown (A) 2CH S*A0V200B 10 (A) 2CHS*A0V200C 10 (1) 2CHS*HCV142 40 2CH S*RV203 N/A E BEAVER VALLEY UNIT 2 3/4 6-19
! O O O TABLE 3.6-1 (Cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Ident i ficat ion /Descr ipt ion Valve Time ( Seg) Valve T ime ( Sec) . 29 Pri Dr Trans ( A) 2 DG S*A0V108A 1.5 ( A) 2DCS*A0V108B 1.5 Pump Disch 2 DG S*RV i l 5 30 SPARE 31 SPARE 32 SPARE 33 SPARE i 1 34 High Head (3)(2) 2 SIS *94 N/A (3)(2) SIS *MOV836 10 (4) Injection Line (3)(2) 2 SIS *MOV840 15 35 Inj Seal Wtr to Reactor (3)(2) 2CHS*474 N/A (3) 2CHS*MOV308A 10 (4) lg 36 Inj Seal Wtr to Reactor (3)(2) 2CHS*476 N/A (3) 2CHS*MOV308B 10 (4) l i 37 Inj Seal Wtr to Reactor (3)(2) 2CHS*475 N/A ( 3) 2 CH S*HOV308C 10 (4) l 38 Sump Pump Discharge (A) 2DA S*A0V100A 1.5 (A) 2DA S*A0V1008 1.5 2DA S*RV110 39 St Cen Blowdown Closed System N/A (2) 2BDG*AOV100A-1 10 l l 40 St Gen Blowdown Closed System N/A (2) 2BDG*A0V100B-1 10 lI 1 l 41 St Gen Blowdown Closed System N/A (2) 2BDG*A0V100C-1 10 i i t BEAVER VALLEY UNIT 2 3/4 6-20
C O O I TABLE 3.6-1 (Cont) i i Maximum Maximum Penet. Inside Stroke Outside Stroke 4 No. Area Identification / Description Valve Time ( Sec) _ Valve Time ( Sec) 42 Service Air 2 SA S* 15 N/A 2 SA S*14 N/A 43 Air Monitor Sample 2 CVS*93 N/A (A) 2CVS*SOV102 1 44 Air Monitor Sample (1)(A) 2CVS*SOV153B 8 (1)(A) 2CVS*SOV153A 8 ! 45 Pr imary Grade Water 2RCS*72 N/A ( A) 2 RCS*A0V519 10 2RCS*RV100 N/A 46 Loop Fill (3)(2) 2 Cil S*472 N/A (3)(2)(1) 20lS*FCV160 10 1 , 47 SPARE 48 Primary Vent Header '(A) 2VRS*A0V109A-2 1.5 ( A) 2VRS*A0V109A-1 1.5 j 49 Nitrogen Supply Manifold 2RCS*68 N/A ( A) 2RCS*A0V101 10 l 1 50 SPARE 51 SPARE 52 SPARE 53 Nitrogen Manifold (A) 2CNS*A0V101-2 10 (A) 2CNS*A0V101-1 10 ]
}
1 I i BEAVER VALLEY UNIT 2 3/4 6-21 e
O O O TABLE 3.6-1 (Cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve T ime ( Sec) 55 Leakage Detection (2) 2LMS*SOV953 8(4) lI Press Relief Tank (1)( A) 2 SSR*SOV130A-1 0.3 (!)( A) 2 SSR*SOV130A-2 0.3 Accumulator Water (A) 2 SSR*A0V109A-1 0.3 (A) 2 SSR*A0V109A-2 0.3 Sample 2 SSR*RV117 N/A Hydrogen Analyzer (1) 2HCS*SOV136A 4(4) (1) 2HCS*SOV136B 4(4) l 56 Cold Leg Sample ( A) 2 SSR*A0V102A-1 0.3 (A) 2 SSR*A0V102A-2 0.3 l 2 SSR*RVi l8 N/A Hot Leg Sample (1)(A) 2 SSR*SOV128A-1 0.3 ( t )( A) 2 SSR*SOV128A-2 0.3 2 SSR*RV120 N/A Pressurizer Liquid ( A) 2 SSR*AOV100A-1 0.3 ( A) 2 SSR*A0V100A-2 0.3 Space Sample 2 SSR*RV119 N/A l Blo.idown Sample Closed System N/A (2) 2 SSR*A0Vll7A 2 l 57 Leak Detection (2) 2LMS*SOV950 8(4) l Blowdown Sample Closed System N/A (2) 2 SSR*A0V117B 2 l Pressurizer Vapor (A) 2 SSR*A0Vil2A-1 0.3 (A) 2 SSR*A0Vil2A-2 0.3 Space Sample 2 SSR*RV121 N/A Hydrogen Analyzer (1) 2HCS*SOV135A 4(4) (1) 2HCS*SOV!35B 4(4) ll 59 Instrument Air Containment 2IAC*22 N/A (A) 2 IAC*MOV130 30 1 BEAVER VALLEY UNIT 2 3/4 6-22
-_ -. . . - - _ - - . . - , ._ - . - _ . - - = . _ . . - - - , - - - . , . _ - . _ . .. - . . .
O O- O TABLE 3.6-1 (cont ) Maximum Maximum Pe ne t . Inside Str oke Outside Stroke No. Area Ident i ficat ion /Deser ipt ion Valve Time ( Sec) ' Valve , T ime ( Sec) 60 Iow Head Safety (3)(2) 2 SIS *132 N/A (3)(2) 2 SIS *Mov8888B 15(4) Injection Discharge 61 Low Head Safety (3)(2) 2 SIS *130 N/A (3)(2) 2 SIS *MOV8889 15(4) Inject ion Discharge 62 Low Head Safety (3)(2) 2 SIS *133 N/A (3)(2) 2 SIS *MOV8888A 15(4) Injection Discharge 1 63 Quench Pump Discharge 2Q SS*4 .N/A (B)(1) IQSS*MOV101 A 60 (4) 2Q SS*RV101 A N/A 64 Quench Pump Discharge 2Q SS*3 N/A (B)(1) 2QSS*MOV101B 60 (4) 2Q SS*RV101 B N/A 65 Fuel Transfer Tube (7) Flange . 66 Recire Spray Pump Suction (B)(2) 2RSS*MOV155A 22 (4) 67 Recire Spray Pump Suction (B)(2) 2RSS*MOV155C 22 (4) 68 Recire Spray Pump Suction (B)(2) 2RSS*MOV155D 22 (4) i 69 Recire Spray Pump Suction (B)(2) 2RSS*MOV155B 22 (4) f 70 Recirculat ion Pump (2) 2RSS*29 N/A (B)(2) 2RSS*MOV156A 60 (4) Discharge (6)(2) 2RSS*RV156A N/A 71 Recirculat ion Pump (2) 2RSS*31 N/A (B)(2) 2RSS*MOV156C 60 (4) Discharge (6)(2) 2RSS*RV156C N/A l BEAVER VALLEY UNIT 2 3/4 6-23
s O O O TABLE 3.6-1 (cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Ident i ficat ion /Descr ipt ion Valve T ime_ ( Sec ) Valve T ime ( Sec ) 73 Main Steam System "A" Closed System N/A (2) 2 MSS *HYV10!A 5 Closed System N/A (2) 2 MSS *A0V102A - Closed System N/A (2) 2 MSS *SOV105A 8(4) Closed System N/A (6) 2 MSS *SV101 A N/A Closed System N/A (6) 2 MSS *SV102A N/A . I closed System N/A (6) 2 MSS *SV103A N/A Closed System N/A (6) 2 MSS *SV104A N/A , Closed System N/A (6) 2 MSS *SV105A N/A Steam Drains System Closed System N/A (2) 2 SDS*A0 Vill A-1 2.6 Closed System N/A (2 ) 2 SD S*A0V1298 1 Steam Vent System Closed System N/A (6) 2 SVS*PCV101 A 10.5 I Closed System N/A (6) 2 SVS*HCV104 14.5 ; BEAVER VALLEY UNIT 2 3/4 6-24
_ = _ _ _ _ _ . - _ _ _ _ . , . _ . _ . _ - - . _ . _ _ . _ _ - _ _ _ . _ . ~ . . _ _ _ _ _ _ _ . . . . - _ - - - _ _ O O 0: TABLE 3.6-1 (cont) Maximum Maximum Pe ne t . Inside Stroke Out s i de - Stroke No. Area Identification / Description Valve Time ( Sec) Valve Time ( Sec) e 74 Main Steam System "B" Closed System N/A (2) 2 MSS *HYV101B 5 Closed System N/A (2) 2 MSS *AOV102B - Closed System N/A (2) 2 MSS *SOV105B 8(4) Closed System N/A (6) 2 MSS *SV101B N/A , Closed System N/A (6) 2 MSS *SV102B N/A E
! Closed System N/A (6) 2 MSS *SV103B' N/A i
Closed System N/A (6) 2 MSS *SV104B N/A { Closed System N/A (6) 2 MSS *SV1058 N/A i Steam Drains System Closed. System N/A (2 ) 2 SD S*A0Vil l B-1 2.6-Closed System N/A (2) 2 SDS*A0V1298 1 1 Steam Vent System closed System N/A (6) 2 SVS*PCV101B 10.5 I Closed System N/A (6) 2 SVS*HCV104 14.5 4 I i l L 4 i BEAVER VALLEY UNIT 2 3/4 6-25 l
O O O TABLE 3.6-1 (Cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve Time ( Sec) 75 Main Steam System "C" Closed System N/A (2) 2 MSS *HYV101C 5 Closed System N/A (2) 2 MSS *A0V102C - Closed System N/A (2) 2 MSS *SOV105C 8(4) Closed System N/A (6) 2 MSS *SV101C N/A Closed System N/A (6) 2 MSS *SV102C N/A I Closed System N/A (6) 2 MSS *SV103C N/A Closed System N/A (6) 2 MSS *SV104C N/A Closed System N/A (6) 2 MSS *SV105C N/A Steam Drains System Closed System N/A (2 ) 2 SD S*A0Vl l lC-1 2.6 Closed System N/A (2) 2SDS*A0V1298 1 1 Steam Vent System Closed System N/A (6) 2 SVS*PCV101C 10.5 } Closed System N/A (6) 2 SVS*HCV104 14.5 76 Feedwater "A" Closed System N/A (2) 2FWS*HYV157A 5 (2) 2 FWS*28 N/A i 77 Feedwater "B" Closed System N/A (2) 2FWS*HYV!5 7B 5 (2 ) 2 FW S*29 N/A 78 Feedwater "C" Closed System N/A (2 ) 2 FW S*H YV15 7C 5 j (2) 2 FWS*30 N/A 1 REAVER VALLEY UNIT 2 3/4 6-26
. __ _. _ __ -1
O O O TABLE 3.6-1 (Cont) i Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve T ime ( Sec ) 79 Aux Feed "A" (2) 2FWE*99 N/A (2) 2FWE*HCV100E 40 (2) 2NE*HCV100F 40 (2) 2FWE*42A N/A (2) 2FWE*42B N/A r 80 Aux Feed "B" (2) 2WE*100 N/A (2) 2WE*HCV100C 40 (2) 2WE*HCV100D 40 (2) 2FWE*43A N/A (2) 2 FWE*43B N/A 83 Aux Feed "C" (2) 2FWE*101 N/A (2) 2FWE*HCV100A 40 (2) 2WE*HCV1008 40 (2) 2FWE*44A N/A (2) 2WE*44B N/A 87 Hydrogen Recombiner Discharge 2HCS*120 N/A (1) 2HCS*MOV117 30 88 Hydrogen Recombiner Discharge 2HCS*119 N/A (1) 2HCS*MOVil6 30 89 SPARE 90 Purge Duct Exhaust (5) 2HVR* MOD 2 3B 10 (5) 2HVR* MOD 23A 10 . , 91 Purge Duct Supply (5) 2HVR* MOD 25B 10 (S) 2HVR* MOD 25A 10 (5) 2HVR*DMP206 N/A 92 Hydrogen Recombiner (1) 2HCS*SOVil4B 8 l Isolat ion (1) 2HCS*SOVil5B 8 - Reactor Cont. Vacuum (A) 2CVS*SOV1518 8 Pump Suct ion (A) 2CVS*SOV152 8 8 BEAVER VALLEY UNIT 2 3/4 6-27 5
O O O TABLE 3.6-1 (Cont) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time ( Sec) Valve Time ( Sec) 93 Hydrogen Recombiner (1) 2HCS*SOVil4A 8 lI Isolat io n (1) 2HCS*SOVil5A 8 Reactor Cont. Vacuum (A) 2CVS*SOV151A 8 Isolation (A) 2CVS*SOV152 A 8 94 Ejector Suct ion 2CVS* 151 N/A 2CV S*151-1 N/A 96 SPARE 97 Leakage Detection (2) 2LMS*SOV952 8(4) Blowdown Sample Closed System N/A (2) 2 SSR*A0Vil7C 2 Liquid Sample - Cont. (1)(A) 2 SSR*SOV129A-1 0.3 (1)(A) 2 SSR*SOV129A-2 0.3 Sump & RH S 2 SSR*RV122 N/A l Hydrogen Analyzer (1) 2HCS*S0V133B 4(4) (1) 2HCS*SOV134B 4(4) 98 SPARE l l 99 Hose Rack Supply 2FPW*761 N/A (A) 2FPW*A0V206 12 100 SPARE 101 Reactor Cont. Deluge - ' 2FPW*753 N/A (A) 2FPW*A0V205 12 , Cable Penet. Area & RHS Pumps ) 103 Reactor Cavity Purif Inlet 2FNC*121 N/A 2FNC*38 N/A 104 Reactor Cavity Purif Outlet 2 FNC*122 N/A 2FNC*9 N/A BEAVER VALLEY UNIT 2 3/4 6-28
O O O TABLE 3.6-1 (Cont ) Maximum Maximum Penet. Inside Stroke Outside Stroke No. Area Identification / Description Valve Time (Sec) Valve Time ( Sec) 105 Leak Detection (2) 2LMS*S0V951' 8(4) l Leak Detection 2 LMS* 51 N/A 2 LMS*52 N/A i Hydrogen Analyzer (1) 2HCS*SOV133A 4(4) (1) 2HCS*S0V134A 4(4) I Post Accident Sampling (A)(1) 2 PAS *SOV105A-1 4 (A)(1) 2 PAS *SOV105A-2 4 106 Safety Inj. Test Line (A) 2 SIS *MOV842 10 (A) 2 SIS *A0V889 10 2 SI S*RV175 N/A i 108 SPARE i 110 SPARE 113 Boron Injection (3)(2) 2 SIS *95 N/A (3)(2) 2 SIS *MOV867C 10 (4) (3)(2) 2 SIS *MOV867D 10 (4) 114 Recirculation Pump (2) 2RSS*32 N/A (B)(2) 2RSS*MOVl56D 60 (4) j Discharge (6)(2) 2RSS*RV156D N/A
'15 Recirculat ion Pump (2) 2RSS*30 N/A (B)(2) 2RSS*MOV156B 60 (4)
Discharge (6)(2) 2RSS*RV156B N/A 8 i 11' Fire Protection HVR Filter B 2FPW*388 N/A (A) 2FPW*A0V221 1.5 117 Fire Protection HVR FiIter A 2FPW*382 N/A (A) 2FPW*A0V204 1.5 l 118 Quench Spray System 2Q SS*267 N/A (B)(1) 2QSS*SOV100A 7 (4) j (B)(1) 2QSS*S0V100B 7 (4) i i
- BEAVER VALLEY UNIT 2 3/4 6-29 f
4 2 t
O O O TABLE 3.6-1 -( Cont ) Maximum Maximum Pe ne t . Inside Str oke Outside Str oke No. Area Ident i ficat ion /Descr ipt ion Valve Time ( Sec) Valve T ime ( Sec) Pr imary Containment Personnel Air Lock 2PH S-PAL 1 Equalizing Valve (7) 2PH S*ll2 N/A Equalizing Valvt (7) 2PH S*ll3 N/A Equalizing Valve ( 7 ) 2 PH S*101 N/A ; Equalizing Valve (7 ) 2 PH S*110 N/A Equalizing Valve (7) 2PH S*ll! N/A Equalizing Valve (7) 2PH S*100 N/A Fmergency Containment Air Lock 2PHS*EALI Equalizing Valve ( 7 ) 2 PH S*202 N/A Equalizing Valve (7) 2PH S*201 N/A k l r i j BEAVER VALLEY UNIT 2 3/4 6-30
- rw = = , __
.i O O O
- TABLE 3.6-1 (Cont)
NUT ES: (A) Containment Isolation Phase A. (B) Contairment Isolat ion Phase B. (1) May be opened on an intermittent basis under administrative control . (2) Not subject to Type C leakage tests. (3) May be leakage tested with water as the test fluid. (4) Maximum opening time. (5) Applicability: During CORE ALTERNTION S or movement of irradiated fuel within containment. The provisions of Specification 3.0.4 are not applicab le . The containment Purge Exhaust and Supply valves will be locked shut during operation in modes 1, 2, 3, and 4. (6) Not subject to the requirments of Specification 3/4.6.3. Listed in TABLE 3.6-1 for in format ion only. (7) Tested under Type "B" testing. (8) Temporarily removed and penetration plugged. (9) Auto open on Safety Injection recirculation signal. } (10) This valve is not required to be Type C leak tested due to the double barrier seal arrangement on the f ue l ! trans fer tube inside containment isolat ion flange. i k BEAVER VALLEY UNIT 2 3/4 6-31 l 1 l
3/4.6 CONTAINMENT SYST EMS 3/4.6.4 COMBUSTIBLE GAS CONTROL r HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two separate and independent cont ainment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACT ION : With one hydrogen analyzer inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours. SURVEILLANCE REQUIREMDrr S 4.6.4.1 Each hydrogen analyzer shall be' denonstrated OPERABLE at le as t once per 92 days on a STAGGERED TEST BASIS by:
- a. Performing a CHANNEL CALIBRATION using sample gas containing:
- 1. One volume percent hydrogen, balance nitrogen, and l 1
- 2. Four volume percent hydrogen, balance nitrogen.
l 1 O ts BEAVER VALLEY UNIT 2 3/4 6-32
3/4.6 CONTAINMENT SY ST EMS 3/4.6.4 COMBUSTIBLE GAS CONTROL ELECTRIC HYDROGEN RECOMBINERS LIMITING (DNDITION FOR OPERATION I I 3.6.4.2 Two separate and independent cont airunent hydrogen recombiner systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION : With one hydrogen recombiner system inoperable, restore the i inoperable system to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours. SURVEILLANCE REQUIRENENT S l , I \ \ 4.6.4.2 Each hydrogen recombiner system shall be demonstrated j OPERABLE: l
- a. At least once per 6 months by verifying during a recombiner system
- O- functional test at a flow r ate of > 50 scfm that the heater outlet temperature increases to > 700*F wIthin 90 minutes and is maintained
, for at le as t 2 hour s .
j b. At least once per 18 months by: I
- 1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation j and control circuits.
- 2. Verifying through a visual exanination that there is no evidence of abnormal conditions within the recombiners (i.e., loose wiring or structur al connections, deposits of foreign materials, etc.).
- 3. Verifying during a recombiner system functional test using containment atmospher ic air at a pressure of f 13 psia and a flow rate of > 50 sefs, that the heater temperature increases to
> 1,100*F within 5 hours and is maintained for at least 4 hours.
- 4. Verifying the integrity of all heater electrical circuits by performing a continuity and resistance to ground test innsediately l following the above required functional test. The resistance to ground for any heater phase shall be > 10,000 ohms.
.i O BEAVER VALLEY UNIT 2 3/4 6-33 i i i
, . - -- , - , . . - - - - - , , , , - - - , , --. .<y.~, , . . - , . , . . . . - . . ,
.,_,,-c _n.e.,--n _n,-- , -.- n -, e- , -m-,
3/4.6 CONTAINMENT SY ST EMS 3/4.6.4 COMT,USTIBLE GAS CONTROL HYDit0CDi PURGE SYST EM LIMITING CONDITION FOR OPERATION 3.6.4.3 A cont ainment hydrogen purge system shall be OPERABLE and capable of being powered from a minimum of one OPERABLE emergency bus. APPLICABILITY: MODES 1 and 2. 4 l ACT ION : With the containment hydrogen purge system inoperable , res tore the hydrogen purge system to OPERABLE status within 30 days or be in HOT STANDBY within 12 hours.
! SURVEILLANCE REQUIREMDrr S 4.6.4.3 The hydrogen purge system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that the purge fan operates for at le as t 15 minutes at a flowrate of 50 CFM + 10%.
- b. The Supplementary Leak Colleetion and Release System (SLCRS) is OPERABLE in accordance with Specification 3.7.8.
4 l i O BEAVER VALLEY UNIT 2 3/4 6-34 i 4
3/4.6 C0!rTAINMENT SY ST EMS 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM f STEAM JET _ AIR EJECTOR LIMITING CONDITION EUR OPERATION 4 3.6.5.1 The inside and outside manual isolation valves in the steam jet air ejector suction line .shall be closed. APPLICALILITY: MODES 1, 2, 3, and 4. ACT ION _: With the inside or outside manual isolation valve in the stean jet air ejector suction line not closed, restore the valve to the closed position within I hour or be in at le as t HOT STANDBY within the next 6 hours and COLD - SHITIDOWN within the following 30 hours. SURVEILIANCE REQUIREMDrr S J _- 4.6.5.1.1 The steam jet air ejector suction line outside manual isolation valve shall be determined to be in the closed position by a visual O inspection prior to increasing the Reactor Coolant System temperature above . 350*F and at least once per 31 days thereaf ter. 4.6.5.1.2 The steam jet air ejector suction line inside manual isolation
! valve shall be determined to be sealed or locked in the closed porf tion by a visual inspection prior to increasing the Reactor Coolant System temperature above 350*F.
d I l ) O BEAVER VALLEY UNIT 2 3/4 6-35 I
3/4.7 PLANT SYSTEMS _ 1 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each , steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. I ACTi g
- a. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed, provided ' that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With 2 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, O~ and 3 may proceed provided that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable .
SURVEILIANCE REQUIREMENTS 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section II of the ASME boiler and Pressure Vessel Code,1980 Edition. i h k I BEAVER VALLEY UNIT 2 3/4 7-1 l [ -. . .-. -- . -
TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEITTRON FLUX HIGH SEIPOINT WITH INOPERABLE ST EAM LINE SAFETY VALVES DURING 3 LOO _P OPERAT ION
. O Maximum Number of Inoper able Maximum Allowable Power Range j
Safety Valves On Any Neutron Flux High Setpoint 3 Oper ating Steam Gener ator (Percent of RAT ED THERMAL POWER) l 1 87 l 1 l 65
. 2 4
l 3 43 i i l 1 !O i i l t I. e i 1 i l O BEAVER VALLEY UNIT 2 3/47-2 i I b i
. - - , - - ,-.-,..n. , _ _ , . , ,. ..---,..-.-,,-.<,_n.,-, s--..-_,n..,--a.,__.,,w-nyn,-,-,,.,n,-w..--,--.-n..
TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOIffr WITH INOPERABLE ST EAM LINE SAFETY VALVES DURING 2 LOOP OPERATION O Maximum Number of Inoper able Maximum Allowable Power Range Safety Valves On Any Neutron Flux High Setpoint
~
Oper ating Steam Gener ator* . (Percent of RATED THERMAL POWER) i 4~ 1 56 2 42 3 28 i 1 l i O i i i I i At least two safety valves shall be OPERABLE on the non-operating steam ge ner ator . I l } i O BEAVER VALLEY UNIT 2 3/4 7-3 l , 1 l l
i TABLE 4.7-1 l STEAM LINE SAFETY VALVES PER LOOP l
)
O Valve Number Lift Setting (+ 1%) Orifice Dianeter
- a. 2MS S-SV101 A , B , & C 1075 psig 4.515 inches b . 2 MSS-SV102A, B, & C 1085 psig 4.515 inches
- c. 2 MSS-SV103A, B, & C 1095 psig 4.515 inches ]
- d. 2 MSS-SV104A, B, & C 1110 psig 4.515 inches
- e. 2 MSS-SV105A, B , & C 1125 psig 4.515 inches O
O BEAVER VALLEY UNIT 2 3 /4 4
4 3/4.7 PLAM SY ST EMS 3/4.7.1 TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three steam generator auxiliary feedwater pumps and associated flow paths sna11 be OPERABLE with:
- a. Two feedwater pumps, each capable of being powered from separate emergency busses, and
- b. One feedwater pump capable of being powered from an OPERABLE ,tean supply system.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one auxiliary feedwater pump inoperable, restore the three auxiliary feedwater pumps (two capable of being powered from separate emergency busses and one capable of being powered by an OPERABLE stean supply system) to OPERABLE status within 72 hours or be in HOT SLUTDOWN within the next 12 hour s .
O b. With the motor driven auxiliary feedwater pump supplying the redundant he ader inoper able, realign the two remaining auxiliary feedwater pumps to the separate headers within 2 hours. SURVEILIANCE REQUIREMl!NT S 4.7.1.2 Each auxiliary feedwater pump shall be demons tr sted OPERABLE:
- a. At least once per 31 days by:
- 1. Starting each pump from the control room.
O ! BEAVER VALLEY UNIT 2 3/4 7-5 I l l
3 /4. 7 , PLANT SY ST EMS 3/4._7.1_ TURBINE CYCLE O 'd AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENT S
- 2. Verifying that:
- a. Each motor driven pump develops a discharge pressure of > 1335*
psig on recirculation flow, and
- b. The steam turbine driven pump develops a discharge pressure of
> 1335* psig on recirculation flow when the secondary stean pressure is greater than 600 psig.
- 3. Ver: fying that each pump operates for at le as t 15 minutes.
- 4. Cycling each testable power operated or automatic valve in the flow l path through at least one complete cycle of full travel. g
- 5. Verifying that each valve (manual, power operated or automatic) in l the flow path that is not locked, sealed, or otherwise secured in pos it ion, is in its correct position.
- 6. Reverifying the requirements of Technical Specification surveill-ance 4.7.1.2.a.5 by a second and independent operator.
O 7. Establish and maintain constant communications between the control room and the auxiliary feed pump room while any normal discharge valve is closed during surveillance testing.
- 8. Verifying operability of each Service Water System auxiliary supply valve by cycling each manual Service Water System to Auxiliary Feedwater System valve through one complete cycle.
- 9. Following an extended plant outage verify Auxiliary Feedwater Flow from TK-210 to the Steam Generators with the Auxiliary Feedwater Valves in the ir norni= A alignme nt .
- b. At le as t once per 18 months during shatdown by:
- 1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at le as t one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal. l[
*To be verified during Pre-Operational Testing O BEAVER VALLEY UNIT 2 3/47-6
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE O iUx1t1*Rr rEEowiTER SvSTE* SURVEILIANCE REQUIREMENTS
- 3. Verifying that each pump starts automatically upon receipt of a i test signal.
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L 4 O BEAVER VALLEY UNIT 2 3/4 7-7 t s J
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3/4.7 PLANT SYST EMS 3/4.7.1 TURBINE CYCLE PRIMARY PLANT _ DEMINERALIZED WAT ER (PPDW) LIMTTING 00NOTTION FOR OPERATION ~
- ~;
3.7.1.3 The primary plant demineralized water storage tank shall be OPERABLE with a minimum usable volume of (later) gallons. 1 APPLICABILITY: MODES 1, 2, and 3. ACT ION : With less than 140,000 gallons of water in the PPDW storage tank, within 4 hours either:
- a. Restore the water volume to within the limit or be in HOT SIUTDOWN within the next 12 hours, or
- b. Demonstrate the OPERABILITY of the reactor plant service water system as a backup supply to the auxiliary feedwater pumps and restore the PPDW storage tank water volume to within its limit within 7 days or be in HOT SHITTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENT S _m______ _ _ _ ______=: __ ________ _ _a 4.7.1.3 The PPDW storage tank shall be demonstrated OPERABLE at le as t once 'per 12 hours by verifying the water level.
-O BEAVER VALLEY UN ET 2 3/47-8
l l l l 3/4.7 PLANT SY ST EMS l 3/4.7.1 TURBINE CYCLE j i ACT IVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be
<0.10 uCi/gr am DOSE EQUIVALENT I-131.
APPLICABILITY: i MODES 1, 2, 3, and 4. ACT ION : With the specific activity of the secondary coolant system
>0.10uci/gr am DOSE EQUIVALENT I-131, be in at least HOT STANDBY within ~ 6 hours and in COLD SIUIDOWN within the next 30 hours.
SURVEILLANCE REQUIRDfENT S 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sanpling and analysis progr am of T able 4.7-2. O J BEAVER VALLEY UNIT 2 3/4 7-9
- l TABLE 4.7-2 SECONDARY COOLA!G SYSTDi SPECIFIC ACTIVITY ,
SAMPLE AND ANALYSIS PROGRAM Type of Measurement and Analysis Minimum Frequency
- 1. Gross Activity Determination 3 times per 7 days with a ,
maximum time of 72 hours between sanples. j
- 2. Isotopic Analysis for DOSE a) I per 31 days, whenever the EQUIVALENT I-131 Concentration gross activity determination indicates iodine l concentr ations greater than l 10% of the allowable limit. .
b) I per 6 months, whenever the gross activity determination I indicates iodine. i concentrations below 10% of ; the allowable limit. i O . 4 i L J 4 i r l O BEAVER VALLEY UNIT 2 3/4 7-10 l l 1
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3/4.7 PLANT SY ST EMS 3/4.7.1 TURBINE _ CYCLE MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FDR OPERATION 3 3.7.1.5 Each main steam line isolation valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACT ION : MOD E S 1 - ' With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoper able valve is either restored to OPERABLE status or closed within 4 hours: Otherwise , be in Har SIUIDOWN within the next 12 hour s . MOD E S 2 - With one main steam line isolation valve inoperable , and 3 subsequent operation in MODES 1, 2 or 3 may proceed after: a) The inoperable isolation valve is restored to OPERABLE status, or O t ^
- 6) t et et tv t e t a cie di Ot her wi se , he in Hf7T MitTTDOWN within the next 12 honra, SURVEILIANCE REQUIREMENT S 4.7.1.5 Each main steam line isolation valve that is open shall be demonstr ated OPERABLE by:
- a. Part-stroke exercising the valve at least once per 92 days, and
- b. Verifying full-stroke closure within 5 seconds on any closure actuation . signal, while in HUT STANDBY with T -> 515F durire each reactor shutdown except that verification **Ef full closure within -
5 seconds need not be determined more of ten than once per 92 days. O BEAVER VALLEY UNIT 2 3/4 7-11
l i I 3/4.7 PLANT SYST EMS 3 /4. 7. 2 ST EAM GDIERATOR PRESSURE / TEMPERATURE LIMITATION i i \d LIMITING CONDITION MR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in l1 the steam generators shall be > 70*F when the pressure of either coolant in the steam generator is > 200 psig. APPLICABILITY: At all times. ACTION : With the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side to < 200 psig within 30 minutes, and
- b. Perform an analysis to determine the ef fect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains ac cept ab le for continued operation prior to increasing its temperatures above 200*F.
SURVEILLANCE REQUIREMDrrS 4.7.2.1 The pressure in each side of the steam generator shall be - lI determined to be < 200 psig at le ast once per hour when the temperature of either the primary or secondary coolant in the steam generator is < 70
- F .
O
's BEAVER VALLEY UNIT 2 3/4 7-12
3/4.7 PLANT SY ST EMS 3/4.7.3 PRIMARY COMPONDIT COOLING WATER SYST EM _ , - () LIMITING CONDITION FOR OPERATION i 3.7.3.1 OP ERABLE. At le as t two primary component cooling water loops shall be l} APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: i With less than two primary component cooling water loops OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD 91UTDOWN within the following 30 hours. SURVEILIANCE REQUIREMElfT S 4.7.3.1 At le as t two primary component cooling water locps shall be l j demonstrated OPERABLE:
- 4. At least once per 31 days on a STAGGERED TEST BASIS by:
i ( l. Verifying-that each pump develops the required differential pres- { aura a nd flow rare when tested in-acenrdance with t he requirements of Section 4.0.5.
- 2. Cycling each testable power operated or automatic valve servicing l safety related equipment through at least one complete cycle of full travel.
- 3. Verifying that each valve. (manual, power operated, or automatic) l t^
servicing safety related equipment that is not locked, sealed, or otherw~ise secured in pos ttion, in its correct pos itio n.
- b. At least once per 18 months during shutdown, by cycling each power operated valve servicing safety related equipment that is not tes t able during plant operation, through at least one complete cycle of full travel.
4 I e
-O BEAVER VALLEY UNIT 2 3/4 7-13 i
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3/4.7 PLANT SY ST EMS 3 /4. 7.4 SERVICE WATER SYSTEM ( SWS) LIMITING CONDITION FOR OPERATION 13.7.4.1 At le as t two service water loops supplying safety related i equipme nt shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION : With less than two SWS loops OPERABLE, restore at le as t two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILIANCE REQUIREMDIT S 4.7.4.1 At least two SWS loops shall be demonstrated OPERABLE: l
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Verifying that each pump develops the required dif ferential pres-(g"' sure and flow rate when tested in accordance with the requirements I of Section 4.0.5.
- 2. Cycling each testable power operated or automatic valve servicing l safety related equipment through at le as t one complete cycle of full travel.
- 3. Verifying that each valve (manual, power operated, or automatic) l servicing safety related equipment that is not- locked, sealed, or otherwise secured in position, is in its correct position.
- b. At le as t once per 18 months during shutdown, by cycling each power operated valve servicing safety related equipment that is not t es t able during plant operation, through at least one complete cycle of _ full travel.
<s U BEAVER VALLEY UNIT 2 3/4 7-14
3/4.7 PLANT SYST EMS 3/4.7.5 ULTIMATE HEAT SINK .- OHIO RIVER O b LIMITING CONDIIION FOR' OPERATION i 3.7.5.1 The ultimate heat sink shall be OPERABLE with: l} l!
- a. A minimum water level at or above elevation 654 Mean Sea Level, at the l intake structure, and
- b. An average water temperature of <86*F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION : With the requirements of the above specification not satisfied, be in at' least HOT STANDBY within 6 hours and in COLD SHUTDOWN within-the following 30 hours. SURVEILIANCE REQUIREMDrrS 4.7.5.1 The ultimate heat sink shall be determined OPERABLE at le as t l O once per 24 hours by verifying the average water temperature and water level to be within their limits. O BEAVER VALLEY UNIT 2 3/4 7-15
3/4.7 PLANT SY ST EM S 3/4.7.6 FIp0D PROrECTION LIMITING CONDITION EUR OPERATION 3.7.6.1 Flood' protection shall be provided for all safety related l1 systems, components and structures when the water level of the Ohio River - exceeds 695 feet Mean Sea Level at the intake structure. APPLICABILITY: At all times. ACT ION : With the water level at the intake structure above elevation 695 feet Mean Sea Level:
- a. Be in at le as t HOT STANDBY vithin 6 hours and in COLD SIUTDOWN within the following 30 hours, and
- b. Initiate and complete within 8 hours, the following flood protection measures:
' 1. Install and seal the flood doors in the int ake structure.
SURVEILIANCE REQUIREMl!NT S 4.7.6.1 The water level at the intake structure shall be determined to l1 ! be within the limits by:
- a. Measurement at least once per 24 hours when the water level is below elevation 690 feet Mean Sea Level, and i
- b. Measurement at least once per 2 hours when the water level is equal to to or above elevation 690 feet Mean Sea Level.
F BEAVER VALLEY UNIT 2 3/4 7-16
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3/4.7 PLANT SYST EMS 3/4.7.7 CONTROL ROOM EMERGDiCY HABITABILITY SYST EMS O t "1Ttwo co"o1T1os roa orsa^T os 3.7.7.1 The following control room emergency habitability systems l1 shall be OPERABLE:
- a. The emergency ventilation system, and
- b. The Lottled air pressurization system.
APPLICABILITY: MODES 1, 2, 3, and 4. ACrION: With one control room emergency habitability system inoper-able, restore the system to OPERABLE status within 7 days or be in at le as t HOT STANDBY within the next 6 hours and in COLD SIUTDOWN within the following 30 hours. SURVEILIANCE REQUIREMDIT S 4.7.7.1 The emergency ventilation system shall be demonstrated OPERABLE:
- a. At le as t once per 12 hours by verifying that the control room air temperature is less than or equal to 104*F.
- b. At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for l 15 minutes. I
- c. At least once per 18 months and 1) af ter each complete or partial l replacement of a HEPA filter or charcoal adsorber bank, or 2) af ter any structural maintenance on the HEPA filter or charcoal adsorber housings, or 3) following painting,. fire, or chemical release in any ventilation zone communicating' with the system by:
- 1. Verifying that the charcoal adsorbers remove > 99.95% of a halogen- l ated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1980 while operating the ventilation l -]
system at a flow rate of 1000 CFM + 10%.
- 2. Verifying that the HEPA filter banks remove > 99.95% of the DOP when they are tested in place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 1000 CFM +_
10%. O BEAVER VALLEY UNIT 2 3/4 7-17 I l , __ . - _ . _ _ _ _
3/4.7 PLANT SY ST EMS 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYST EMS SURVEILLANCE REQUIREMENT S
- 3. Subjecting the carbon contained in at least one test canister or at le as t two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal ef ficiency of > 99% for radioactive methyl iodide at an air flow velocity of 0.7 ft/sec + 20% with an inlet methyl iodide concentra- 1 tion of 1.5 to 2.0 mg/m7, 95% relative humidity, and >176*F; other test conditions shall be in accordance with USAEC RUT '5tandard M-16-1T , June 1972. The carbon samples not obtained from test canisters shall be prepared by either:
a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at le ast two inches in diameter and with a length equal to f 5 thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at le ast two inches in diameter and with a length equal to the thickness of the bed.
- 4. Verifying a system flow rate of 1000 cfm + 10 percent during system oper at ion.
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5.6 inches Water Gauge while operating the ventilation system at a flow rate of 1000 cfm +_ l1 10 percent.
- 2. Verifying that on a containment isolation signal, the system auto-matica11y starts in 60 minutes and diverts its inlet flow through the HEPA filters and charcoal adsorber banks.
- 3. Verifying that the system maintains the' control room at a positive pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere during system operation.
4.7.7.2 The' bottled air pressurization system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that the system contains a minimum of 10 bottles of air, each pressurized to at least 1825 psig.
b..At least once per 18 months by verifying that the sp tem will supply at least 800 cfm of air to maintain the control room at a positive , pressure oi greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere during system operation. BEAVER VALLEY UNIT 2 3/4 7-18
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l 1 3/4.7 PLANT SY ST EMS l l 3/4.7.8 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYST EM '( SLCRS) , l l ( LIMITING CONDITION FOR OPERKTION 3.7.8.1 Two SLCRS exhaust air filter trains shall be OPERABLE. l1 APPLICABILrrY: MODES 1, 2, 3, and 4. ACTION : With one SLCRS exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD 91UTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENT S 4.7.8.1 Each SLCRS exhaust ak filter train shall be demonstrated l OPERABLE:
- a. At least once per .31 days by initiating, from the control room, flow through the standby HEPA filter and charcoal adsorber train and veri- I fying that the train operates for at le as t 15 minutes.
s ("'/ N- b. At least once per 18 months and (1) af ter each complete or partial l replacc=ent of a HEPA filter er charec=1 adscrber bank, or (2) af ter any structural maintenance on the HEPA filter or charcoal adsorber housings, or (3) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the charcoal adsorbers remove 2,99.95 percent of a I halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1980 while operating the l ventilation system at a flow rate of 59,000 cfm j; 10 percent.
- 2. Verifying that the HEPA filter banks remove 2,99.95 percent of the DOP when they are tested in-place in accordance with ANSI N510-1980 while- operating the ventilation system ~at a flow rate of 59,000 cfm j; 10 percent.
4 O BEAVER VALLEY UNIT 2 3/4 7-19
-_ __ _ . _ - - _ .___..u...._.__
1 1 3/4.7 PLANT SY ST EM S !
-3/4.7.8' SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYST EM ( SLCRS) l l
SURVEILLANCE REQUIREMENT S
- 3. Subjecting the carbon contained in at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal efficiency of >99 percent for radioactive methyl iodide at an air 1 flow velocity of 0.7 ft/see f; 20 percent with an inlet methyl iodide concentration of 1.5 to 2.0 mg/m , 95 percent relative 3
humidity, and >176*F; other test conditions shall be in accordance wit h U SAEC RDT , Standard M-16-1T , June 1972. The carbon samples not obtained from test canisters shall be taken with a slotted tu'oe sampler per ASTM E-300,
- 4. Verifying a system flow rate of 59,000 cfm f; 10 percent during system operation.
- c. At le as t once per 18 months by:
- 1. Verifying.that the pressure drop across the combined HEPA filters and charcoal adsorber banks is <6.8 inches Water Gauge while opera- l ting the ventilation system at a flow rate of 59,000 cfm j; 10 per-cent.
- 2. Verifying that the exhaust from the contiguous area is diverted s through the SLCRS filtcr train on a Containment Isolation - Phase
{d "A" signal. I
- d. Verifying that the air flow distribution to each HEPA filter and char-coal adsorber is within + 20 percent of the averaged flow per unit after initial installation and af ter any maintenance af fecting the flow distribution.
l 9 s O BEAVER VALLEY UNIT 2 3/4.7-20.
3/4.7 PLANT SY ST EM S 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in l1 l excess of those quantities of by product material listed in 10 CFR 30.71 or j
> 0.1 microcuries of any other material, including alpha emitters, shall be i free of > 0.005 microcuries of removable contamination. ,
1 APPLICABILITY: At all times. ACTION :
- a. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:
- 1. Either decontaminated and repaired, or
- 2. Disposed of in accordance with Commission Regulations ,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILLANCE REQUIREMDIT S O - 4.7.9.1.1 Test Requ i ----- - - Each =caled source shall be te=ted for l1 leakage and/or contamination by:
- a. The licensee, or
- b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at le as t 0.005 microcuries per test sample. 4.7.9.1.2 Test Frequencies - Each category of sealed sources shall be l tested at the frequency described below,
- s. Sources in use (excluding startup sources previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive materials.
- 1. With a half-life greater than 30 days (excluding Hydrogen 3), and l 2. In any form other than gas.
i O BEAVER VALLEY UNIT 2 3/4 7-21
3/4.7 PLANT SYST EMS 3/4.7.9 SEALED SOURCE CONTAMINATION SURVEILLANCE REQUIREMENT S
- b. Stored sources not in use - Each sealed source shall be tested prior to use or transfer to another licensee unless tested within the pre-vious six months. Sealed sources tr ans ferred without a certificate
. indicating the last test date shall be tested prior to being placed into use.
- c. Startup sources - Each sealed startup source shall be tested prior to being subjected to core flux and following repair or maintenance to the source.
4.7.9.1.3 Reports - A Special Report shall be prepsred and submitted to the Commission pursuant to Specification 6.9.2 within 5 days if source leak-age tests reveal the presence of > 0.005 microcuries of removable contamina- l1 tion. 8 O BEAVER VALLEY UN TT 2 3/4 7-22
3/4.7 PLANT SYST EMS LIMITING CONDITION NR OPERATION l, 3.7.10 This Technical Specification intentionally blank. j
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I 3 /4. 7 PLAhT SY ST EMS LIMITING CONDITION FOR OPERATION O 3.7.11 This Technical Specification intentionally blank. 4 i 1 d I a O i 4 i t i O BEAVER VALLEY UNIT 2 3/4 7-24
3/4.7 PLANT SYST EMS 3/4.7.12 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.12 All snubbers shall be OPERABLE. The only snubbers excluded from this requirement are those installed on non safety-related systems and 1 then only if their failure or failure of the system on which they are installed would have no adverse .ef fect on any safety-related system. APPLICABILITY: MODES 1, 2, 3, and 4. (MODES 5 and 6 snubbers located on sys-tems* required OPERABLE in those MODES.) ACT ION : With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engi-neering evaluation per Specification 4.7.12.c on the supported component or declare the supported system inoperable and follow the appropciate ACTION s t atement for that system. SURVEILIANCE REQUIREMDrTS 4.7.12 Each snubber shall be demonstrated OPERABLE by performance of O the following augmented inservice inspection program and the requirements of Specification 4.0.5.
- a. Visual Inspections The first inservice visual inspection of snubbers shall be performed after four months, but within 10 months of commencing POWER OPERATION and shall include all snubbers. If less than two (2) snubbers are gI found inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months + 25 per cent from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:
- These systems are defined as those portions or subsystems required to preve nt relesses in excess of 10CFR100 limits.
l ( BEAVER VALLEY UNIT 2 3/4 7-25 I i s
. . . - . . . , .-- . - , - . _ , , _ . . _ . . . . _ . ,~ , . . . _ . - . . _ , . . . . . - , . , _ _ _ . . . . _ , . . , - - _ , , . - .
'3/4.7 PLANT SYSTEMS 3 /4. 7.1_2 SNUBBERS SURVEILIANCE REQUIREMENT S No. Inoperable Snubbers Subsequent Visual Per Inspection Period Inspection Per iod**# i 0 18 months j; 25%
1 12 months + 25% 2 6 months }[ 25% 3, 4 124 days + 25% 5,6,7 62 days 325% 8 or more 31 days j;25% The enubbers may be categorized into two groups; those accessible and those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule.
- b. Visual Inspection Acceptance Criteria Visual inspections shall verify:
- 1. that there s e no visible indications of damage or impaired OPERABILITY,
- 2. attachments to the foundation or supporting structure are secure, O and
- 3. in those locations where snubber movement can be manually induced without disconnecting ti;e snubber, that the snubber has freedom of movement and is not frozen up.
Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that:
- 1. the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible, and l
- 2. the af fected snubber is functionally tested in the as-found l condition and determined OPERABLE per Specification 4.7.12.d or 4.7.12.e as applicable .
~ ** T he inspection interval shall not be lengthened more than one step at a time.
i # The provisions of Specification 4.0.2 are not applicab le . O BEAVER VALLEY UNIT 2 3/4 7-26
3/4.7 PLANT SYSTEMS 3/4.7.12 SNUBBERS O SURVEILIANCE REQUIREMENTS However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined inoper41e and cannot be determined OPERABLE via functional testing for the purpose of estab-lishing the next visual inspection interval.
- c. Functional Tests At least once per 18 months during shutdown, a representative sample (of at le as t 10 snubbers or at le ae: 10 per ce nt, whichever is less) of the total of each type of snubber in use in the plant, shall be func- ,
tionally tested either in place or in a bench test. For each snubber 4 that does not meet the functional test acceptance critera of Specifi-cation 4.7.12.d or 4.7.12.e, an additional 10 anubbers or at le as t 10 percent, whichever is less of that type of snubber shall be funct io n-ally tested. The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At le as t 25 per cent of the - snubbers in the representative sample shall include snubbers from the following three categories:
- 1. The first snubber away from each reactor vessel nozzle; l
l 2. Snubbers within 5 feet of heavy equipment (valve, pump, turbine, ; motor, etc.) l
- 3. Snubbers within 10 feet of the discharge from a safety relief valve.
Snubbers that are esycially difficult to remove or in "High Radiation l1 Zones During Shutdown" shall also be included in the representative sample.* i i i 4 i
- Permanent or other exemptions fece functional testing for individual anubbers inthese categories may be granted by the Commission only if a justifiable basis for exemption is presented and/or snubber life destructive testing was performed _ to qualify snubber operability for all design conditions at either the completion or their fabrication or at a subsequent date.
I ? r O BEAVER VALLEY UNIT 2 3/4 7-27
l
.3/4.7 PLANT SYST EMS l
3/4.7.12 SNUBBERS (:) SURVEILLANCE REQUIREMENT S If a spare snubber has been installed in place of a failed snubber , the spare snubber shall be retested. Test results of this snubber may not be included for the re-sampling. If any snubber selected for functional testing either fails to lockup or f ails to move, i.e., frezen in place, the cause will be evaluated and if caused by manufacturec or design deficiency, all snubbers of the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated above for snubbers. not i meeting the functional test acceptance criteria. For the snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber (s). The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely af fected by the inoperability of the snubber (s) in order to ensure that the sup-ported component remains capable of meeting the designed service. 4 d. Hydraulic Snubbers Functi_onal Test Acceptance Criteria i. The hydraulic snubber functional test shall verify that:
- 1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.
- 2. Snubber bleed, or release rate, where required, is within the spec-l ified range in compression or tension. For snubbers specifically
! required to not displace .under unidirectional dynamie load, the ability of the snubber to withstand load without displacement shall be verified. ] e. Mechanical Snubbers ' Functional Test Acceptance Criteria The mechanical snubber functional test shall verify that:
- 1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.
- 2. Activation (restraining action) is achieved within the specified range of velocity or acceleration -in both tension and compression.
- 3. Snubber release rate, where required, is within the specified range in compression or te ns ion. For snubbers specifically required not to displace under unidirectional dynamic load, the ability of the snubber to withstand load without displacement shall be verified.
O BEAVER VALLEY UNIT 2 3/4 7-28 l l l 4
._.. , - . _ - , . - , _ _ - . . . - ..-,,,,,_m--~,---,,_ , , , _ , _ _ . , _ . - .
3/4.7 PLANT SYST EMS 3/4.7.12 93UBBER S O SURVEILIANCE REQUIREMENT S
- f. Snubber Service Life Monitoring A record. of the service life of each snubber , the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by specification 6.10.2.m.
Concurrent with the firs t inservice visual inspection and at le as t once per 18 months thereaf ter , the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. .If the indicated service life will be exceeded pr for to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records. O O l BEAVER VALLEY UNIT 2 3/4 7-29 l l
3/4.7 PLANT SYST EMS 3/4.7.13 STANDBY SERVICE WAT ER SYSTEM ( SSWS) LIMITING CONDITION FOR OPERATION 3.7.13.1 At least one Standby Service Water System shall be OPERABLE. l1 APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION : With less than one SSWS system OPERABLE, restore at least one system to OPERABLE status within 7 days or be in at least HOT STANDBY within
~
the next 6 hours and in COLD SiUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENT S 4.7.13.1 At least one SSWS system shall be demonstr ated OPERABLE. l1
.a. At least once per 31 days by:
- 1. Starting each pump from its control station.
- 2. Verifying that each pump develops at least (later) psig discharge pressure while pumping through its test flow line.
- 3. Verifying that each pump operates for at le as t 15 minutes .
- 4. Cycling its power operated discharge valve through at least one complete cycle of full travel.
- b. At -least once per 18 months during slutdown by starting a Standby Service Water System pump, shutting down one Service, Water System, pump, and verifying that the Standby Service Water Subsystem provides at le ast 7,589 gpa cooling water to that portion of the Primary Ser- l1 vice Water System under test for at le as t 2 hours .
(V9 B'. AVER VALLEY UNIT 2 3/4 7-30 l l I l --. - --- - . - . - . - ..- - .
3/4.7 PIANT SY ST EMS 3/4.7.14 FIRE SUPPRESSION SYST EMS < FIRE SUPPRESSION WATER SYST EM O LIMITING CONDITION FOR OPERATION I 3.7.14.1 The fire suppression water system shall be OPERABLE with:
- a. An OPERABLE flow path capable of taking suction from the Ohio River I and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser required to be 3
OPERABLE per Specifications 3.7.14.2 and 3.7.14.4.
- b. The booster fire pump, with a capacity of 625 gpa with its discharge l aligned to Unit 2 hose rack stations for the safety-related equipment ar e as .
- c. An OPERABLE flow path capable of taking suction from the Service Water l1 System and transferring the water through distribution piping to the first valve ahead of the water flow alarm device on each hose stand-pipe required to be OPERABLE per Specification 3.7.14.4.
APPLICABILITY: At all times. ACT ION :
- a. With one 2500 gpa pump inoperable, verify the operability of the booster fire pump, and restore the inoperable equipment to OPERABLE status within 7 days or, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss.
of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applic able .
- b. With the booster fire pump inoperable verify 'the operability of two 2500 gpa pumps, and restore the inoperable equipment to OPERABLE
, statue within 7 days or, prepare and submit a Special Report to the l Commission pursuant to Specification 6.9.2 within the next 30 days 4 outlining the' plans and procedures to be used to provide for the loss of redundancy in this ' system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable . l l BEAVER VALLEY UNIT .2 3/4 7-31 I
3/4.7 PLANT SYST EMS 3/4.7.14 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WAT ER SYSTEM LIMITING CONDITION FOR OPERATION
- c. With the fire suppression water system otherwise inoper able:
- 1. Establish a backup fire suppression water system within 24 hours, and,
- 2. Submit a Special Report in accordance with Specification 6.9.2:
a) By telephone with 24 hours , b) Confirmed by telegraph, mailgram or facsimile transmission no later than the first working day following the event, and c) In writing within 14 days following the event , outlining the action taken, the cause of the ino per ability, and the plans and schedule for restoring the system to OPERABLE status. SURVEILLANCE REQUIREMDIT S 4.7.14.1.1 The fire suppression water system shall be demonstrated OPERABLE:
- a. At least once per 31 days by starting the booster fire pump and !I operating it for c le == 15 minutes,
- b. At least once per 31 days by verifying that each valve (manual, power l operated, or automatic) in the flow path is in its correct pos it ion,
- c. At least once per 12 months by performance of a system flush to be l performed in conjunction with the Unit I system flush, (Unit 1 Technical Specification Surveillance Requirement 4.7.14.1.1.c).
- d. At least once per 12 months by cycling each testable valve in the flow l path through at least one complete cycle of full travel.
- e. At least once per 18 months by: l
- 1. Verifying that the booster fire pump develops at le as t 625 gpm at 'a system head of 250 feet, A
V BEAVER VALLEY UNIT 2 3/4 7-32 l l i
3/4.7 PLANT SY ST EMS 3/4.7.14 FIRE SUPPRESSION SYST EMS FIRE SUPPRESSION WATER SYST EM , O SURVEILIANCE REQUIREMDiT S t
- 2. Cycling each valve in the flow path that is not tes table during i plant operation through at least one completa cycle of full travel, and
- 3. Verifying that the booster fire pump maintains the fire suppression water system pressure > 90 psig. .
- f. At least once ' per 3. years by performing a flow test of the system in accordance with Section 16, Chapter 8 of the Fire Protection Handbook, '
15th Edition, published by the National Fire Protection Association. , i 1! i i i i l i t BEAVER VALLEY UNII 2 3/4 7-33 l
\
3/4.7 PLANT SYSTEMS 3/4.7.14 FIRE SUPPRESSION SY ST EMS SPRAY SY ST EMS ' LIMITING CONDITION FOR OPERATION 3.7.14.2 The following spray systems shall be OPERABLE:
- a. Auxiliary Feed Pumps; 2FWS*P22, P23A&B, Safeguards Building.
- b. Component Cooling Pumps; 2CCP*P21A,B, and C, Auxiliary Building. i
- c. Residual Heat Removal Pumps; 2RES*P21A and B, Re ac tor Co nt ainment .*
- d. Charcoal Pilters; 2HVS*FLIA 205A and B, 208A and B, Auxiliary Build-ing.
APPLICABILrrY: Whenever equipment in the spray protected areas is required to be OPERABLE. , 1
- ACTION
- l C
- a. With one or more of the above required spray systems inoperable, with-in one hour establish a roving fire watch with backup fire sappression i equipment for those areas in which redundant systems or components {
O v could be damaged, such that the area is checked hourly when the system has to be operable . Restore the system to OPERABLE status within 14 days or, prepare and submit a Special Report to the Commission pursu- l ant to Specification 6.9.2 within the next 30 days outlining the ' action taken, the cause of the inoperability and the plans and sched- ; ule for restoring the system to OPERABLE status. '
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicab le .
SURVEILLANCE REQUIREMDrr S 4.7.14.2 Each of the above required spray systems shall ba demonstr ated OPERABLE: '
- a. At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path accessible during plant o pe r-ation is in its correct pos itio n.
- b. At least once per 12 months by cycling each testable valve in the flow l path through at least one complete cycle of full travel.
*With a containmect area spray system inoperable, check this area during scheduled containment entries in Modes 1-4 and once per shift in Modes 5 and BEAVER VALLEY UNIT 2 3/4 7-34
3/4.7 PLA!TT SYST EMS 3/4.7.14 FIRE SUPPRESSION SYST EMS SPRAY SY ST EMS SURVEILLANCE REQUIREMDIT S
- c. At least once per 18 months:
- 1. By performing a system functional test which includes simulated automatic actuation of the system, and:
a) Verifying that the automatic valves in the flow path actuete to their correct positions on a manual test signal, and' b) Cycling each valve in the flow path that is not tes t ab le during plant operation through at least one complete cycle of full travel.
- 2. By a visual inspection of the dry pipe spray headers to verify t heir integrity, and
- 3. By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.
- d. At le as t once per 3 years by performing an airflow test through each l1 open spray header and verifying each open spray nozzle is unobs tr ucted.
I BEAVER VALLEY UNIT 2 3/4 7-35
3/4.7 PLANT SY ST EMS 3/4.7.14 FIhE SUPPRESSION SYSTEMS 1DW PRESSURE C03 SYST EM O LIMITING CONDITION mR OPERATION 3.7.14.3 The 10 ton low pressure CO2 system (2-FPD System No. 2) serv-ing the following areas shall be OPERABLE with a minimum level of 91 percent and a minimum pressure of 295 psig in one of the two 10 ton storage tanks.
- a. Cable Spreading Area Control Building
- b. Instrumentation Room Control Building
- c. Communication Room Control Building
- d. Cable Tunnel Between Control and Auxiliary Buildings l
- e. Cable Tunnel Auxiliary Euilding i
- f. Cable Vault and Rod Control Areas
- g. Cable Spreading Area Service Building
- h. Baerg=ncy Dierel Generator Building Room APPLICABILITY:
Whenever equipment in the low pressure CO2 Protected areas is required to be OPERABLE. ACT ION :
- a. With one or more of the above required low pressure CO2 systems inop-erable, establish a continuous fire watch with backup fire suppression equipment for the unprotected ares (s) within 1 hour; restore the sys-tem (s) to OPERABLE status within 14 days or, prepare and submit a l1 Special Report to the Commission pursuant to Specification 6.9.2 with-in the next 30 days outlining the action taken, the cause of the inop-erability and the plans and schedule for restoring the system to OPERABLE status. ,
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicab le .
i i BEAVER VALLEY UNIT 2 3/4 7-36 L
l 3/4.7 PIANT SY ST EMS 3/4.7.14 FIRE SUPPRESSION SYST EMS IDW PRESSURE CO2 SY ST EM SURVEILLANCE REQUIREMINT S 4.7.14.3 The above required low pressure CO2 system shall be demons trated OPERABLE:
- a. At least once per 7 days be verifying the two CO2 storage tank levels and pressures, and
- b. At least once per 18 months by verifying:
i
- 1. The system valves and associated ventilation dampers actuate manually and automatically, upon receipt of a simulated actuation signal, and
- 2. Flow fr om e ac h no zz le dur ing a "Pu f f T es t ."
O BEAVER VALLEY UNIT 2 3/4 7-37
3/4.7 PLANT SYST EMS j 3/4.7.14 FIRE SUPPRESSION SYSTEMS l FIRE HOSE STATIONS t O LIMl'ING CONDITION NR OPERATION 3.7.14.4 The fire hose stations shown in Table 3.7-5 shall be OPER-ABLE. APPLICABILITY: Whenever equipment in the ar eas protected by the fire hose ! s t at ions is required to be OPERABLE. { ACT ION : !
- a. With one or more of the fire hose stations shown in T able 3.7-5 inop- ,
erable, route an additional equivalent capacity fire hose to the ! unprotected ares (s) from an OPERABLE hose station within one hour (4 l hours for containment hose stations) if the inoperable fire hose is the primary means of fire suppression; otherwise, route the additional l hose within 27+ hours. Restore the fire hose station to OPERABLE ! status within 14 days or submit a Special Report to the Commission ! pursuant to Specification 6.9.2 within the next 30 days, outlining the I' action taken, the cause of the inoperability, and plans and schedule for restoring the station to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applic ab le . l l
O suavzt^"cz aiinu'^r"='r s t i 4.7.14.4 Each of the above required fire hose stations shall be demon- ; strated OPERABLE: '
- a. At least once per 31 days by visual inspection of the fir e hose s t a- l tions accessible during plant operation to assure all required equip-ment is at the station.
- b. At least one per 18 months by:
- 1. Visual inspection of the stations not accessible during plant oper- .
ations to assure all required equipment is at the stat ion.
- 2. Removing the hose for inspection and re-racking, and ;
- 3. Inspecting all gaskets and replacing any degraded gaskets in the l1 couplings.
- c. At least one per 3 years by:
- 1. Partially opening each hose station valve to verify valve OPERA- !
BILITY and no flow blockage. j w 2. Conducting a hose hydrostatic test at a pressure at le as t 50 psig i above maximum fire main operating pressure,
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j BEAVER VALLEY UNIT 2 3/4 7-38 , l
. TABLE 3.7-5 FIRE HOSE EATION S
() Location Isolating Valve No. Elevation Hose Rack Number Decon Bldg 182 735' 6" 201 Fuel Bldg 181 752' 5" 202 Fuel Bldg 180 752' 5" 203 Fuel Bldg 179 735' 6" 204 Diesel 2-2 Bldg 523 732' 6" 205 Diesel 2-1 Bldg 526 732' 6" 206 Cable Vault /Contig Area 566 718' 6" 207 l i Cable Vault /Contig Area 568 718' 6" 208 Cable Vault /Contig Area 558 735' 6" 209 Cable Vault /Contig Area 564 735' 6" 210 Cable _ Vault /Contig Area 560 755' 6" 211 Cable Vault /Contig Area 562 755' 6" 212 {' } Cable Vault /Contig Area 570 7 73' 6" 213 ^ Service Bldg 138 7 80' 6" 214 Service Bldg 137 760' 6" 215 Service Bldg 135 760' 6" 216 Service Bldg 139 745' 6" 217 Service Bldg 142 745' 6" 218 1 Service Bldg 144 730' 6" 219 Scevice Bldg 146 730' 6" 220 Auxiliary Bldg 195 710' 6" 241 Auxiliary Bldg 199 773' 6" 242 Auxiliary Bldg 316 755' 6" 243 O BEAVER VALLEY UNIT 2 3/4 7-39 l
TABLE 3.7-5 FIRE HOSE STAT ION S (CONT ' D. ) () Location Isolat ion
. Valve No. Elevation Hose Rack Number Auxiliary Bldg 317 735' 6" 244 Auxiliary Bldg 320 718' 6" 245 Auxiliary Bldg 324 773' 6" 246 Auxiliary Bldg 325 755' 6" 247 Auxiliary Bldg 326 735' 6" 248 Auxiliary Bldg 327 718' 6" 249 Auxiliary Bldg 331 773' 6" 250 Auxiliary Bldg 132 755' 6" 251 Auxiliary Bldg 333 735' 6" 252 Auxiliary Bldg 340 718' 6" 253 Cont ainme nt 7 72 735' 6" 258 735' 6"
{} Cont ainment 774 259 Cont ainme nt 770 735' 6" 260 Containment 777 717' 0" 261 Containment 779 -717' 0" 262 Cont ainme nt 776 717' 0" 263 Containment 783 692' 11" 264 Cont ainment 785 692' 11" 265 Cont ainment 781 692' 11" 266 Cont ainment 764 767' 10" 267 Cont ainment 766 767' 10" 268 Cont ainment 768 767' 10" 269 712' 6" Auxiliary Bldg 577 270 O BEAVER VALLEY UNIT 2 3/4 7-40 L
TABLE 3.7-5 FIRE HOSE STAT ION S (CONT ' D. ) () Location Isolat ion Valve No. Elevation Hose Rack Number Auxiliary Bldg 579 712' 6" 271 Cable Tunnel 575 735' 6" 272 Relay Room 5 73 735' 6" 273 Control Bldg 549 725' 6" 274 Control Bldg 551 725' 6" 275 Control Bldg 553 712' 10" 276 Control Bldg 555 707' 6" 277 Control Bldg 545 712' 6" 278 Safeguards Area 516 737' 6" 287 Safeguards Area 514 737' 6" 288 Safeguards Area 518 718' 6" 289 () Safeguards Area 5 12 718' 6" 290 i BEAVER VALLEY UNIT 2 3/4 7-41
3/4.7 PLANT SY ST EM S 3/4.7.14 FIRE SUPPRESSION SYSTEMS HALON SY ST EMS d LIMITING CONDTTION FOR OPERATION 1 3.7.14.5 The following Halon systems shall be OPERABLE:
- a. Control Building Computer Room 1
- b. Control Building West Communications Room
; APPLICABILITY:
i Whenever equipment protected by the Halon system is required to be OPERABLE. ACT ION :
- a. With one or more of the above required Halon systems inoperable , with- ,
in I hour establish a continuous fire watch with backup fire suppres-sion equipment for those aceas in which redundant systems or compon-ents could be damaged. Restore the system to OPERABLE STATUS within lg 14 days or , prepare and submit a Special Report to the Commission pur suant to Specification 6.9.2, within. the next 30 days outlining the action taken, the cause of the inoperability and the plans and sched-ule for restoring the system ot OPERABLE status. j O b. The provisions of Specifications 3.0.3 and 3.0.4 are not applic ab le . 1 ! SURVEILIANCE REQUIREMENTS t _ - __ 4.7.14.5 Each of the above required Halon systems shall be demonstrated OPERABLE:
- a. At le as t once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path is in its correct pos it io n.
- b. At least once per 6 months b'y verifying Halon storage tank weight to be at least 95% of full charge weight and pressure to be at le as t 90%
of full charge pressure. I ! c. At least once per 18 months by:
- 1. Verifying the system, including the associated ventilation dampera and fire door closing mechanisma, actuate-manually and automatic- I ally upon receipt of a simulated actuatio'n signal.
- 2. Visually inspect each header and nozzle . to verify their integr ity .
f l d. At least once per 36 months by performance of a flow test thr ough y i headers and nozzles to assure no blockage. . BEAVER VALLEY UNrr 2 3/4 7-42 l
- I l
l
3/4.7 PLANT SYST EMS 3/4.7.15 FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 1 3.7.15 All fire rated assemblies (walls, floor / ceilings, cable tr ay enclosures and other fire barriers) separating safety related fire areas or , separ ating portions of redundant systems import ant to safe shutdown within a fire area and all sealing devices, in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable and piping penetration seals and ventilation seals) shall be OPERABLE. l APPLICABILITY: At all times. ACrION:
- a. With one or more of the above required fire rated assemblies and/or ataling devicer, it. operable, witnin one hour either esta'olish a contin-uous fire watch on at least one side of the af fected assembly, or l verify the OPERABILITY of fire detectors on at le ast one side of the inoperable assembly and establish an hourly fire watch patrol until the functional capability of the barrier is restored. Restore the inoperable fire rated assembly and sealing device to OPERABLE status or prepare and submit a Special Report to the Coasais- 1 within 7 days,t TSpecification 6.9.2 within the next 30 days outlining sion pursuant p
v the action _ taken, the cause of inoperability and the plans and sched-ule for restoring to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILIANCE REQUIREMDIT S 4.7.15.1 At least once per 18 months the above required fire rated l1 assemblies and penetration sealing devices shall be. verified OPERABLE by:
- a. Performing a visual inspection of the exposed surfaces of each fire rated assemblies,
- b. Performing a visual inspection of each fire window / fire damper and associated hardware.
- c. Performing a visual inspection of at leas t 10 perce nt of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10 per-cent of each type of sealed penetration shall be made. This ins pec-tion process shall continue until a 10 percent sample with no apparent changes in appearance or abnormal degradation is found.
BEAVER VALLEY UNIT 2 3/4 7-43
3/4.7 PIANT SYSt EMS 3/4.7.15 FIRE RATED ASSEMBLIES O Soavetttisce arou asns rS 4.7.15.2 Each of the above required fire doors
- shall be verified OPERABLE by inspecting the automatic holdopen, release and closing mechanism.
and latches at le as t once per 6 months, and by verifying:
- a. The position of each closed Fire door at least once per 24 hours.
- b. That doors with automatic holdopen and release mechanisms- are free of obstructions at le ast once per 24 hours.
O
- Security alarm fire doors are not included in the above surveillance requirements,'since they are checked per security requirements.
O BEAVER VALLEY UNIT 2 3/4 7-44 l
._ _ _ _ =_ _ . _ _
3/4.7 . PLANT SYST EMS 3/4.7.16 T ERRESTRIAL ECOLOGICAL SURVC.' 1 LIMITING CONDITION FDR OPERATION i 3.7.16 The ef fects of cooling tower drif t on terrestrial biota shall g
- be determined and report submitted as required by Sect ion 6.9.1.15. l l APPLICABILITY:
l At all times. ACT ION : An assessment of the data, including copies of the latest photographs, gathered in this program element shall be submitted at the end of each alternate year in accordance with Section 5.6.1. SURVEILIANCE REQUIREMl!NT S 4.7.16 ! a. The Terrestrial Ecological Survey program element shall assess the
! potential impact of the cooling tower drift on the terrestrial veget a-tion of the site and its vicinity by comparison and evaluation of infrared aerial photographs taken once every two years. Locat ion, O exte nt , and severity of any stressed areas shall be documented and 4
related to the meteorological data. Incipient impacts of cooling tower drift on the terrestrial vegetation of the site and its vicinity shall be detected with color infrared aerial photography. Interpretation of i t he infrared photographs shall include ground reconnaissance of selected areas. Aerial photos shall be taken on color infrared film at a scale of 1 inch = 400 feet. Photos shall be taken between 11 A.M. and 2 P.M. EDT during the middle of the growing season and as close as possible , to the same date during alternate years. The flight direction shall remain the same as preoperational flights. Photogr.aphs shall be free of cloud shadows. The film processor will insure that processicg meth-ods and conditions -shall remain the same throughout the study. l A flight log shall be compiled when the photographs are taken and processed.
! b. Areas with the greatest and least potential for being af fected by cooling tower drift shall be selected and compared. The location, l extenc, and severity of any stressed area shall be documented and i
related to meteorological data. The possible role of cooling tower dr if t in the development of ' stressed areas 'shall be as se s sed. l () BEAVER VALLEY UN TT 2 3/4 7-45 l 4
,_m--- . y_ _ - . , _ ._ . . _ . , _ , _ _ . _ _ _ _ , , _ , _ _ _ . , . _ _ , . . . , ..___,_,,, .,_ __. _ ,__ . , , , , _ . - . _ . _ , . . _
3/4.8' ELECTRICAL POWER SYST EMS 3/4.8.1 A.C. SOURCES OPERATING O LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the of fsite transmission network and the onsite Class IE distribution system, a nd
- b. Two separate and independent diesel generators each with:
- 1. A separate day tank containing a minimum of 222 gallons of fuel, l1
- 2. A separate fuel storage system containing a minimum of 53,225 gal-lons of fuel, and
- 3. A separate fuel transfer pump.
APPLICABILITY: MODES 1, 2, 3, and 4. ACT ION :
% a. With either an offsite circuit or diesel generator of the above
("J b required A.C. electrical power sources inoper able, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at least. once per 8 hours thereaf ter; restore at le as t two offsite cir-cuits and two diesel generators to OPERABLE status within 72 hours or be in COLD SitTTDOWN within the next 36 bours.
- b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoper able, demonstrate the OPERABILETY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at least once per 8 hours thereaf ter; restore at least one of the inoper-able sources to OPERABLE status 'within 12 hours or be in COLD SRUTDOWN within the next 36 hours. Restore at le as t two of fsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in COLD SRUIDOWN within the next 36 hours,
- c. With two of the above required of fiste A.C. circuits inoperable, dem-onstrate the OPERABILITY of two diesel generators by performing Ser-l}
veillance Requirement 4.8.1.1.2.a.5 within one hour and at le as t once per 8 hours thereafter, unless the diesel generators are already BEAVER VALLEY UNIT 2 3/4 8-1 l
l 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERAT ING
)
i LIMITING CONDITION FOR OPERATION operating; restore at le as t one of the inoperable of fsite sources to OPERABLE status within 24 hours or be in at le as t HUT STANDBY within the next 4 hours. With only one of fsite source restored, restore at le as t two offsite circuits to OPERABLE status within 72 hours fr om time of initial loss or be in COLD SHUTDOWN within the next 36 hours,
- d. With two of the above required diesel generators inoperable , demon-strate the '0PERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at le as t once per 8 hours thereaf ter, restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in COLD SHUTDOWN within the next 36 hours. Restore at le ast two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in COLD SHUTDOWN within the next 36 hours.
SURVEILLANCE REQUIREMDrr S 4.8.1.1 Two physically independent circuits between the of fsite trans-() mission network and the onsite Class lE distribution system shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months by trans ferring (manually and automatically) unit power supply from the unit circuit to the system circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Verifying the fuel level in the day tank,
- 2. Verifying the fuel level in the fuel storage tank,
- 3. Verifying that a sample of diesel fuel from the fuel storage tank is 'within the acceptable limits specified in Table 1 of ASIM D975-68 when checked for viscosity, water and sediment ,
- 4. Verifying the fuel trans fer pump can be started and trans fers fuel from the storage system to the day tank, J
ll
- 5. Verifying the diesel starts from ambient co nd it io n, BEAVER VALLEY UNIT 2 3/4 8-2 l
t .
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OP ERAT ING O SURVEILIANCE REQUIREMENT S . 6. Verifying the generator is synchronized, loaded to greater than or equal to 4,238 kW (continuous rating) and operates for at least 60 minutes , and
- 7. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
- b. At le as t once per 18 months, during shutdown by:
- 1. Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with it's manufacturer's '
recommendat ions for this class of standby service.
- 2. Verifying the generator capability to reject a load of greater than or equal to 825 kW (largest single emergency load, a standby service water pump) without tr ipping ,
- 3. Simulating a loss of offsite power in conjunction with a safety injection signal, and:
- a. Verifying de-energization of the emergency busses and load shedding from the emergency busses,
- b. Verifying the diesel starts from asbient condition on the auto-s t art signal, energizes the emergency busses with permanently connected loads, energizes the auto-connected etwegency loads 4
through the' load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
- 4. Verifying that on a loss of power to the emergency busses, all diesel generator trips, except engine overspeed, generator dif ferential and overcurrent , are automatically disabled,
- 5. Verifying the diesel generator operates for > 60 minutes while
~
loaded to > 4,2 38 kW, 6.-Verifying that the auto-connected loads to each diesel generator do not exceea the 2000 hour rating of 4,535kW,
- 7. Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within + 10 percent of its required value.
O BEAVER VALLEY UNIT 2 3/4 8-3
. . _ _ _ ~. _ _ _ _ _ _ _ ._ _ _ ... _ .__._ _ _ _ _ , . . . _ _ _ _ _ .
3/4.8 ELECTRICAL POWER SYST EMS 3/4.8.1 A.C. SOURCES . SHUTDOWN O LIMITING CONDITION FUR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit between the of fsite transmission network and the onsite Class 1E distribution system, a nd
- b. One diesel generator with:
- l. A day tank containing a minimum of 222 gallons of fuel. l1
- 2. A fuel storage system containing a minimum of 53,225 gallons of I
fuel, and
- 3. A fuel trans fer pump.
APPLICABILITY: MODES 5 and 6. A CT ION : < l With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status. SURVEILIANCE REQUIREMl!NT S 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Require- l ment s of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2 . a .6. i I i l BEAVER VALLEY UNIT 2 3/4 8-4 i
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERA 1 IJG LIMITING CONDITION FOR OPERATIO'l 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open between redundant busses: 4160 volt Emergency Bus #2AE and 480 volt Emergency Bus #2N 4160 volt Emergency Bus #2DF and 480 volt Emergency Bus #2P 120 volt A.C. Vital Bus #I 120 volt A.C. Vital Bus #II 120 volt A.C. Vital Bus #III J 4 120 volt A.C. Vital Bus #IV - APPLICABILITY: I MODES 1, 2, 3, and 4. ACT ION : l
}
With less than the above complement of A.C. busses OPERABLE, RESTORE THE INOPERABLE BUS TO OPERABLE status within 8 hours or be in at leas t HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILIANCE REQUIREMENT S 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power evailability. i i O BEAVER VALLEY UNIT 2 3/4 8-5 l t i i t
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYST EMS i A.C. DISTRIBUTION - SHUTDOWN O LIMITING CONDITION EVR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel gener ator but aligned to an OPERABLE diesel generator. 1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Bus 2 - 120 volt A.C. Vital Busses , APPLICABILITY: MODES 5 and 6. ACT ION : With less than the above complement of A.C. busses OPERABLE and energized, establish CONTAINMENT INT EGRITY'within d hours. SURVEILIANCE REQUIREMENT S 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability. 1 1 ) i 1O BEAVER VALLEY UNIT 2 3/4 8-6
3/4.8 ELECTRICAL POWER FYST EMS 3 /4. 8. 2_ ON SITE POWER DISTRIBUTION SYSTEMS fs- D.C. DISTRIBUTION - OPERATING d LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be energized and OPERABLE. T RAIN "A" (orange) consisting of 125-volt D.C. busses No. 2-1 & 2-3, 125-volt D.C. battery banks 2-1 & 2-3 & charger 2-1 & inverter 2-3. T RAIN "B" (purple) consisting of 125-volt D.C. busses No. 2-2 & 2-4, 125-volt D.C. battery banks 2-2 & 2-4 & charger 2-2 & inverter 2-4. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION : l
- a. 'With one of the required battery banks inoperable , restore the inoper-able battery bank to OPERABLE status within 2 hours or be in at le as t HOT STANDBY within the next 6 hours and in COLD SIUTDOWN within the
{} following 30 hours.
.b. With one of the required full capacity chargers inoperable , demon-strate the OPERABILITY of its asociated battery bank by performing Surveillance Requirement 4.8.2.3.2.a.1 within one hour, and at le as t once per 8 hours thereaf ter. If any Category A limit in T able 3.8-1 is l}
not met, declare the battery inoperable. SURVEILIANCE REQUIREMENT S 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability. 4.8.2.3.2 Each 125-volt battery bank, charger, and inverter shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The parameters in Table 3.8-1 meet the Category A limits, and
- 2. The total battery terminal voltage is greater than or equal to 127.8 - volts on float charge.
O BEAVER VALLEY UNIT 2 3/4 8-7 1 L
3/4.8 ELECTRICAL POWER SYSTEMS
-3 /4. 8. 2 ONSIT E POWER DISTRIBITTION SYSTEMS D.C. DISTRIBITTION - OPERATING SURVEILIANCE REQUIREMENT S
- b. At least once per 92 days and within 7 days af ter a battery discharge l
with battery terminal voltage below 110 - volts, or battery overcharge 4 with battery terminal voltage above 150 - volts, by verifying that:
- 1. The parameters in T able 3.8-1 meet the Category B limits.
- 2. There is no visible corrosion at either terminals or connectors or the connection resistance of these items is. less than 150 x 10-5 j ohms, and
- 3. The average electrolyte temperature of eu sry tenth cell of connect-ed cells is above 60*F.
- c. At least once per 18 months by verifying that:
- 1. The cells, cell plates, and battery racks show no visual indication of physical danage or abnormal deterioration,
- 2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material.
O 3. The resistance of each cell-to-cell and terminal connection is les s than or equal to 150 x 10-6 ohns; and
- 4. The battery charger will supply at least (100) amperes at 140 i -volts for at least (4) hours.
; d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the 2-hour design
] duty cycle when the battery is subjected to a battery service test.1 l l l
- e. At least once per 60 months, during shutdown, by verifying that the battery cspacity is at least 80 percent of the manufacturer's eating when subjected to a performance discharge test. Once per 60 month 1 interval, this performance discharge test may be performed in lieu of the battery service test.
- f. At least once per 18 months, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85 percent of the service life i expected for the application. Degradation is indicated when the bat-tery capacity drops more than 10 percent of rated capacity from its aver age on previous performance tests, or is below 90 percent of the manufacturer's rating.
3Load testing conducted pursuant to IEEE 450-1980. BFAVER VALLEY UNIT 2 3/4 8-8
TABLE 3.8-1 BATT ERY SURVEILIANCE REQUIREMBT S O CNT EGORY ' A(1) CNT EGORY B(2) P ar ameter Limits for each Limits for each A11owable(3) designated pilot connected cell value for each cell connected cell Ele ctr olyt e > Minimum level > Minimum level Above top of Level indication mark, indicat ion mark, plates, and and < 1/4" above and < 1/4" above not over flowing maximum level maximum level indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (c) > 2.07 volts Not more than
.020 below the average of all
_. 1.195 connected cells Specific ~
>1.200(b) Aver age of all Average of all Gravity (a) connected cells connected cells
> 1.205 > 1.195(b)
() ~ ]! (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than (2) amps when on charge. (c) Corrected for average electrolyte temperature. (1) For any Category A par ameter(s) outside the limit (s) shown, the battery may
- be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. .
(2) For any Category B parameter (s) outside the limit (s) shown, the batery may l be considered OPERABLE provided that the Category B par ameters are within j t he ir allowable values and provided the Category B parameter (s) are i restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an i inoper able battery. Numbers in parentheses assume a manufacturer's recommended full charge specific gr avity of 1.215. O BEAVER VALLEY UNIT 2 3/4 8-9 l
3/4.8 ELECTRICAL POWER SYST EMS 3/4.8.2 ONSITE POWER DISERIBUTION SYSTEMS D.C. DISrRIBUTION - SHITTDOWN
)
^
LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE: 2 - 125-volt D.C. bus systems, and 2 - 125-volt battery bank and chargers / inverters associated with the above D.C. bus systems. APPLICABILITY: MODES 5 and 6. ACT ION : With less than the above complement of D.C. equipment and bus system OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours. SURVEILIANCE REQUIREMENTS O 4.8.2.4.1 The above required 125-volt D.C. bus system shall be deter-mined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability. 4.8.2.4.2 The above required 125-volt battery bank and chargers / invert-ers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2. l1 O BEAVER VALLEY UN rr 2 3/4 8-10
l 3/4.9 REFUELING OPERATION S 3/4.9.1 BORON CONCDITRATION O LIMITING CONDITION FOR OPERATION l 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the-l refueling canal shall be maintained uniform and sufficient to ensure that the j more restrictive of the following reactivity conditions is met: l l
- a. Either a K of 0.95 or less, which includes a 1 percent de lt a }
k/k conser,gkive va allowance for uncert aint ies , or
- b. A boron concentration of 2000 ppe, which includes a 50 ppa conservative allowance for uncertainties.
APPLICABILITY: MOD E 6 * . l ACrION: With the requirements of the above specification not satisfied, inneediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 30 gem of 7000 ppe boric acid solution or its equivalent until K is reduced to
< 0.95 or the boron concentration is restored to > 2006ggppa, whichever is the O -
more restrictive. The provisions of Specification 3.0.3 are not applicab le . SURVEILLANCE REQUIREMDrr S I I I 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and I
- b. Withdrawal of any full length control rod in excess of 3 feet from its -
fally insertsd position. 4.9.1.2 The boron concentration of the reactor coolant system and the i refueling canal shall be determined by chemical analysis at le as t 3 t imes per
- 7. days with a maximes time interval between samples of 72 hours.
- The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or rcmoved.
t lO BEAVER VALLEY UNIT 2 3/4 9-1 l
I 3/4.9 REFUELING OPERATION S 3/4.9.2 IN ST RUMDITAT ION O LIMITING CONDITION FUR OPERATION. l l 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containnent and control room. APPLICABILITY: MODE 6. ACTION : With the requirements of the above specification not satis fied, immediately suspend all oper ations involving CORE ALTERATION S or positive reactivity changes. The provisions of Specification 3.0.3 are not applic ab le. SURVEILIANCE REQUIRl!Ml!NT S 4.9.2 Each so s ce range neutron flux monitor shall be demonstr ated OPERABLE by performance of:
- a. A GANNEL FUNCTIONAL TEST at least once per 7 days , and
- b. A GANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. A GANNEL GECK at lest once per 12 hours during CORE ALT ERATIONS.
i - 4 t BEAVER VALLEY UNIT 2 3/4 9-2 i i - 1
3/4.9 REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR opt. NATION [ , 3.9.3 The reactor shall be subcritical for at le as t 150 hour s . APPLICABILITY: t During movement of ireadiated fue1 in the reactor pressure vessel. ACT ION : With the reactor suberitical for less than 150 hours, suspend all operati1ns involving movement of irr adiated fuel in the reactor pressure vessel. The vovisions of Specification 3.0.3 are not applicab le . 4 I SURVEILIANCE REQUIREMENT S 1 s 4.9.3 The reactor shall be determined to have been suberitical for at le as t 150 hours by verification of the date and time of suberiticality prior to movement of irradiated fuel in the reactor pressure vessel. i I I, i i i l i i O BEAVER VALLEY UNIT 2 3/4 9-3 l I l
- 1
. . _ _ _ . . _ _ _ . - . . . _ _ _ _ _ . _ _ , , _ _ _ , . . - _ - _ . - _ , - . . _ . , - ~ _ . . _ _ . _ . _
3/4.9 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PDIETRATION S
~
LIMITING CONDITION FOR OPERATION 1 ! 3.9.4 The contairunent building penetr ations shall be in the following status: < a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and l
- c. Each penetr ation providing direct access from the cont airunent atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or
- 2. Exhausting at less than or equal to 7500 cfm through OPERABLE Contairunent Purge and Exhaust Isolation Valves with isolation times as specified in Table 3.6-1 to OPERABLE HEPA filters and charcoal adsorbers of the Supplemental I4ak Collection and Release System
( SLCR S) . I APPLICABILITY: j During CORE ALT ERATIONS or movement of irradiated fuel within the contairunent. O ACT ION : i With the requirement of the above specification not s at is fied , immediately suspend all cperations involving CORE ALTERATIONS or movement of irradiated fuel in the contairunent. The provisions of Specificar. ion 3.0.3 are not applic ab le . SURVEILIANCE REQUIREMENT S 4.9.4.1 Each of the above required containment penetr ations shall be determined to be in its above required condition within 150 hours prior to the start of and at least once per 7 days during CORE ALT ERATIONS or movement i t of irr adiated fuel in the contairunent. i \ Q V BEAVER VALLEY UNIT 2. 3/4 9-4 i
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3/4.9 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS SURVEILLANCE REQUIREMENTS 4.9.4.2 The containment purge and exhaust system shall be demonstrated OPERABLE by:
- a. Verifying the flow rate through the SLCRS at least once per 24 hours when the system is in operation,
- b. Testing the Containment Purge and Exhaust Isolation Valves per the applicable portions of Specification 4.6.3.1.2, and 1
- c. Testing the SLCRS per Specification 4.7.8.1.
O 4 b e s O BEAVER VALLEY UNIT 2 3/4 9-5
3/4.9 REFUELING OPERATION S
.3/4.9.5 COMMUNICAT ION _S _
LIMITING CONDITION FOR OPERATION j 3.9.5 Dir ect communications shall be maintained between the control room and personnel at the refueling station. APPLICABILITY: Dur ing CORE ALT ERATIONS. ACT ION : When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The govisions of Specification 3.0.3 are not applicab le , f SURVEILIANCE REQUIRDIENT S s 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the s t ar t of and at leas t once per 12 hours during CORE ALT ERATIONS. 4 O t i O BEAVER VALLEY UNIT 2 3/4 9-6 n- . ,., - - . , _. - , . _ , _ . _ _ _ , , , _ _ _ _ _ _ _ _ __ ____ __ . , _ _
3/4.9 REFUELING OPERATION S 3/4.9.6 MANIPULATOR CRANE OPERABILITY O t'*'t1"o co"otrio" >oa oesa^tto" 3.9.6 The manipulator crane and auxiliary hoist shall be used for moveme nt of control rods or fuel assemblies and shall be OPERABLE with: l}
- a. The manipulator crane used for movement of fuel assemblies having:
- 1. A minimum capacity of 3250 pounds, and l 2. An overload cutoff limit less than or equal to 2850 pounds.
t j b. The Auxiliary hoist used for latching and unlatching drive rods having:
- 1. A minimum capacity of 700 pounds, and
- 2. A load indicator which shall be used to prevent lif ting loads in excess of 600 pounds.
APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor pressure vessel. ACT ION : With the requirements for crane and/or hoist OPERABILITY not satisfied, suspenri use of any inoperable manipulator crane and/or auxiliary hoist fron operations involving the movesent of drive rods and fue1 assemblies within the reactor pressure vessel. The provisions of i Specification 3.0.3 are not applicab le . SURVEILLANCE REQUIREMENT S s , 1 4.9.6.1 Each manipulator crane used for movement of fuel assemblies \ l' within the reactor pressure vessel shall be demonstrated OPERABLE within 150 hours prior to the st >rt of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2850 pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used for moveme nt of control rods within the reactor pressure vessel shall be l[ demonstrated OPERABLE within 150 hours prior to the start of such operations ' by performing a load test of at least 700 pounds. * ' i 4 lO l BEAVER VALLEY UNIT 2 3/4 9-7 , 1 t i
3/4.9 REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING O t1sITINc CoNotTroN roR oeEaiTroN 3.9.7 Loads in excess of 3000 pounds shall be prohibited from travel over fuel assemblies in the storage pool. APPLICABILITY: With fuel assemblies in the storage pool. ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. . SURVEILLANCE REQUIREMENTS i 4.9.7 Crane interlocks and physical stops which prevent crane travel with loads in excess of 3000 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereaf ter during crane operation. O i l 1 (:) lBEAVERVALLEYUNIT2 3/4 9-8 r
. , - - - , , - - --- m -
_# . , . - - - . - ~ . - . - - . - , - , - - _ . , , - . ,.,->---w-- - - - - r ,, - - - - . . . .-.-- , -,-
3/4.9 REFUELING OPERATIONS 3/4.9.8 COOLANT CIRCULATION . O RsStoUit riEiT asaovit ^*o cootist c1acutirios LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be in operation. APPLICABILITY: MODE 6. ACTION:
- a. With less than one residual heat removal loop in operation, except as provided in b below, suspend all operations involving, an increase in the reactor' decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. ,
- b. The residual heat removal loop may be removed from operation for up to I hour per 8 hour period during the performance of CORE ALTERATIONS in j the vicinity of the reactor pressure vessel (hot) legs.
- c. The residual heat removal loop may be removed from operation for up to 4 hours per 8 hour period during the performance of Ultrasonic In-service Inspection inside the reactor vessel nozzles provided there is at least 23 feet of water above the top of the reactor vessel flange.
- d. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at a flow rate of > 3000 gym at least once per 4 hours when making boron dilution changes and >~1000 gpm for decay heat removal when the Reactor Coolant System is in the drained down condition within the loops. 1
-O BEAVER VALLEY UNIT 2 3/4 9-9
3/4.9 REFUELING OPERATIONS l 3/4.9.8 COOLANT CIRCULATION , LOW WATER LEVEL i LIMITING CONDITION FOR OPERATION 3.9.8.2 Two Residual Heat Removal (RHR) loops shall be OPERABLE.* ' APPLICABILITY: 1 MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet. t ACTION: t
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status
, as soon as possible.
- b. The provisions of Specification 3.0.3 are' not applicable, i
SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined
- OPERABLE per Specification 4.0.5.
4 I t
* . The normal or emergency power source may be inoperable for each RHR !
loop. t l l O BEAVER VALLEY UNIT 2 3/4 9-10 ;
3/4.9 REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE AND EXRAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the Containment Purge and Exhaust Isolation System inoperable, close each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 1 4.9.9 The Containment Purge and Exhaust Isolation System shall be demonstrated OPERABLE within 150 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment Purge () and Exhaust isolation occurs on manual initiation and on a high-high radiation test signal.from each of the containment radiation monitoring instrumentation channels, t (:) BEAVER VALLEY UNIT 2 3/4 9-11 6
- ., , _ _ _ _ , _ . . ._ - . . - _ _ - ,-_ , _ _ _ _ . -m , ,,m_., ._, . ..r,.- , _ _
3/4.9 REFUELING OPERATION S l 3/4.9.10 WATER LEVEL - REACTOR VESSEL O ttatTt"o cosotrios roa ortaatto" 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange. APPLICABILITY: During movement of irradiated fuel assemblies or c'ontrol rods { within the reactor pressure vessel while in MODE 6. ACT ION : With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMDrr S 4.9.10 The water level shall be determined to be at le as t its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereaf ter during movement of fuel assemblies or control rods. O O BEAVER VALLEY UNIT 2 3/4 9-12
3/4.9 REFUELING OPERATION S 3/4.9.11 WAT ER LEVEL - STORAGE POOL O - trattrac coaoir1o# von ortaarto" 3.9.11 At least 23 feet of water shall be maintained over the top of the irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whe never irradiated fuel assemblies are in the storage pool. ACT ION: With the requirements of the above specification not satis fied, suspend all movement of fuel assemblies and crane operations with l loads in the fuel storage areas and restore the water level to within its limit within 4 hours. The provisions of Specification 3.0.3 are not applicable. SURVEILIANCE REQUIRDIDrr S I 4.9.11 The water level in the storage paol shall be determined to be at le ast its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. l{ O O BEAVER VALLEY UNIT 2 3/4 9-13
3/4.9 REFUELING OPERATIONS 3/4.9.12 FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT i LIMITING CONDITION FOR OPERATION 3.9.12 The fuel building ventilation system shall be operating and discharging through at least one train of the SLCRS HEPA filters and charcoal adsorbers during either:
- a. Fuel movement within the spent fuel storage pool, or
- b. Crane operation with loads over the spent fuel storage pool.
APPLICABILITY: When irradiated fuel which was decayed less than 60 days is in i the fuel storage pool. 1 ) ACTION: . With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.12 The fuel building ventilation system shall be verified to be operating with all building doors closed within 2 hours prior to the , initiation of and at least once per 12 hours during either fuel movement within the fuel storage pool or crane operation with loads over the fuel storage pool. l l 1 t .l . r O i lBEAVERVALLEYUNIT2 3/4 9-14 l l _ . _ . . _ _ _ _ _ _ _ . _ . - _ _ . . _ .J
3/4.9 REFUELING OPERATIONS 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE O t1"ttrac co otr1o* ro oezair1o* 3.9.13 The Supplemental Leak Collection and Release System (SLCRS) portion of the fuel building ventilation system shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION: Without the SLCRS portion of the fuel building ventilation system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one fuel building ventilation system is restored to OPERABLE status. The provisions of Specification 3.0.3 are not applicable . t SURVEILLANCE REQUIREMENTS 4.9.13 The SLCRS portion of the fuel building ventilation. system shall be demonstrated OPERABLE by: p 1. Verifying that on a high-high radiation signal, the system v automatically directs its exhaust flow through the HEPA filters and charcoal adsorber banks of the SLCRS at least once per 18 months, i
- 2. Testing the SLCRS per Specification 4.7.8.
l i i O ! BEAVER VALLEY UNIT 2 3/4 9-15 l J i I
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SitTTDOWN MARGIN O LIMITING CONDITION EUR OPERATION 3.10.1 The SHUTDOWN margin requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACT ION :
> 1.0) and with less than the
- a. With the reactor critical above reactivity equivalent (K'shatTable for trip insertion, immediately initiate and continue boration at > 30 gpa of 7000 ppe boric acid
~
solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored,
- b. With the reactor suberitical (K f 1.0) by less than the above reactivity equivalent , immediatefy < initiate and continue boration at i
> 30 gpm of 7000 ppe boric acid solution or its equivalent until the 91UTDOWN MARGIN required by Specification 3.1.1.1 is res tored.
O SURVEILIANCE REQUIREMBfT.S 4.10.1.1. The position of each full length rod either partially or fully l withdrawn shall be determined at least once per 2 hours. 4.10.1.2. Each full length rod not fully inserted shall be deinons trated l capable of full insertion when tripped from at le ast the 50 percent withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGlit to less than the limits of Specification 3.1.1.1. f I O BEAVER VALLEY UNIT 2 3/4 10 - 1
._~
3/4.10 SPECIAL TEST EXCEPT ION S 3/4.10.2 GROUP HEIGHT , INSERTION AND POWER DISTRIBUTION LIMIT S O LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insert ion and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be sus pe nded during the performance of PHYSICS TEST S provided:
- a. The THERMAL POWER is maintained <- 85 percent of RAT ED THERMAL POWER, and
- b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE l'. ACT ION : With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either: O a. Reduce THERMAL POWER sufficient to satisfy the ACTICN requirements of Specifications 3.2.2 and 3.2.3, or
- b. Be in HUT STANDBY within 6 hours.
SURVEILIANCE REQUIREMENT S 4.10.2.1. The THERMAL POWER shall be determined to be < 85 percent of RAT ED THERMAL POWER at least once per hour during PHYSICS TESTS. l I 4.10.2.2. The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS: g
- a. Specification 4.2.2, at le as t once per 12 hours .
l
- b. Specification 4.2.3, at le as t once per 12 hours.
l { O BEAVER VALLEY UNIT 2 3/4 10-2
3/4.10 SPECIAL TEST EXCEPTION S 3/4.10.3 PRESSURE / TEMPERATURE LIMITATION - REACTOR CRITICALITY O LIMITING CNDITION FOR OPERATION
'3.10.3 The minimum temperature and pressure conditions for reactor criticality of Specifications 3.1.1.5 and 3.4.9.1 may be suspended during low temperature THYSICS TEST S provided:
- a. The THERMAL POWER does not exceed 5 percent of RAT E THERMAL POWER,
- b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at < 25 per ce nt of RATE THERMAL POWER, and
- c. The Reactor Coolant System temperature and pressure relationship is maintained within the acceptable region of ~ operation shown on Figures 3.4-2 and 3.4-3.
APPLICABILITY: MODE 2. ACTION :
- a. With the THERMAL POWER > 5 percent of RAT E THERMAL POWER, immediately O ePe the reacter trip 6 ea*ers-
- b. With the Reactor Coolant Syst.em temperature and pressure relationship within the unacceptaole region of operation on Figures 3.4-2 and 3.4-3, immediately open the reactor trip breakers and res tore the temperature pressure relationship to within its limit within 30 minutes; perform the analysis required by Specification 3.4.9.1 prior to the next reactor criticality.
SURVEILIANCE REQUIREMl!NT S 4.10.3.1. The Reactor Coolant System shall be verified to be within the l acceptable region for operation of Figures 3.4-2 and 3.4-3 at least once per l hour. 1 4.10.3.2. The THERMAL POWER shall be determined to be < 5 percent of
~ l 1 RAT E THERMAL POWER at le as t once per hour.
4.10.3.3.- Each Intermediate and Power Range Nuclear Channel shall be l subjected to .a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating low temperature PHYSICS TEST S. O BEAVER VALLEY UNIT 2 3/4 10-3
3/4.10 SPECIAL T EST EXCEPTION S 3/4.10.4 PHYSICS TEST S O LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TEST S provided:
- a. The THERMAL POWER does not exceed 5 percent of RAT E THERMAL POWER, and l I
- b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at i 25 percent of RAT E THERMAL POWER. l APPLICABILITY:
MODE 2. ACT ION : With the THERMAL POWER > 5 percent of RAT E THERMAL POWER, immediately open the reactor trip breakers. SJRVEILIANCE REQUIREMENT S O 4.10.4.1. The THERMAL POWER shall be determined to be 15 percent of l RATE THERMAL POWER at least once per hour during PHYSICS TEST S. 1 4.10.4.2. Each Intermediate and Power Range Channel shall be subjected to l a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating PHYSICS TEST S. O BEAVER VALLEY UNIT 2 3/4 10- 4
3/4.10 SPECIAL TEST EXCEPTION S 3/4.10.5 NO FLOW TEST S O LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint , arxi
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set at < 25 per ce nt of RAT ED THERMAL POWER.
APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACT ION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, inanediately open the reactor trip breakers. SURVEILIANCE R1!QUIREMENT S p 4.10.5.1. The THERMAL POWER shall be determined to be less than the P-7 l Interlock Setpoint at least once per hour during startup and PHYSICS TESTS.
}
4.10.5.2. Each Intermediate and Power Range Channel and P-7 Interlock l shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating startup or PHYSICS TESTS.
~
l BEAVER VALLEY UNIT 2 3/4 10-5
3/4.10 SPECIAL T EST EXCEPTION S 4 3/4.10.6 POSITION INDICATION SYSTEM - SitTTDOWN O LIMITING CONDITION FOR OPERATION 3.10.6 T he limitations of Specification 3.1.3.3 may be suspended during the per formance of individual full length (shutdown and control) rod drop time measurements provided:
- a. Only one shutdown or control bank is withdr awn from the fully inserted position at a time, and
- b. The rod position indicator is OPERABLE during the withdr awal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time me as ur eme nt s . ACT ION : With the position indication system inoperable or with more than one bank of rods withdr awn, immediately open the reactor trip breakers. SURVEILIANCE REQUIREMENT S 4.10.6 The above required rod position indication systems shall be determined to be OPERABLE within 24 hours pr for to the start of and at le as t once per 24 hours thereaf ter during rod drop time measurements by verifying the demand position indication system and tne rod position indication systems agree within 12 steps when the rods are stationary. l
- This requirement is not applicable during the initial calibration of the rod position indication system provided (1) Kegg is '
maintained less than or equal to 0.95, and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one-time. O BEAVER VALLEY UNIT 2 3/4 10-6
3/4.11 RADIOACTIVE EFFLUDIT S 3/4.11.1 LIQUID EFFLUDIT S CONCDIT RAT ION LIMITING CONDITION FOR OFERATION 3.11.1.1 The correncration of radioactive material released from the site (See Figure 5.1-1) shall be limited to the concentration specified in l y 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the conce ntr ' ? ' .n sh : ! ' be limited to 2 x 10-4 uCi/mi total activity. APPLICABILIT f: h all times. ACT ION :
- a. With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, immediately restore concentration within the above limits.
- b. Submit a Special tapet to the Commission within 30 days in accordance I with Specification 6.9.2
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
O suavrt*"c= ai!autai!"=rr s 4.11.1.1
- 1. Radioactive liquid wastes shall be sampled and. analyzed according to the sampling and analysis progras of Table 4.11-1*.
- 2. The results of radioactive analysis shall be used in accordance with l the meth'ods of the ODCM to assure that the concentration at the point 1 of release are maintained within the limits of Specification 1 3.11.1.1.
l
- Radioactive liquid discharges are normally via batch modes. Turbine Building Drains shall be monitored as specified in paragraph' 4.11.1.1.3. Recirculation Drain Pump discharge shall be monitored as specified in paragraph 4.11.1.1.4.
O-BEAVER VALLEY UNIT 2 3/4 11-1
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.1 LIQUID EFFUUENT S CONC ENT RAT ION O SURVEIL 1ANCE REQUIREMDIT S 4.11.1.1.3 When the activity of the secondary coolant is greater than l 10-5 uCi/ml and the Turbine Building Trans fer Pumps (2DGS-P42, 2DBS-P43, I 2DBS-P44) are not pumping their sumps to the steam generator blowdown tank (2 SGC-TK21B) gras samples shall be taken for each sump discharge from the turbine building. The sample shall be analyzed for gross activity at a sensitivity of at le as t 10-7 uCi/mi and recorded in the plant records. Water volume discharged shall be estimated from the number of pump operations unless alternate flow or volume instrumentation is provided. l 4.11.1.1.4 Prior to the Recirculation Drain Pump (s), 2DAS-P215A/B discharging to catch basin number 16 a grab sample will.be taken. Ttw samples will be analyzed for gross activity at a sensitivity of at le as t 10-7 uCi/ml and recorded in the plant records. Water volume discharged shall be estimated from the number of pump operations unless alternate flow or volume ins tr ume n-tation is provided. l 1 O 1 l s 1 ($) BEAVER VALLEY UNIT 2 3/4 11-2
TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit O Liquid Mimimum Analysis Type of Activity of Detect ion (LLD) Release Sampling T ype Frequency Frequency Analysis (uCi/ml) (a) Principal A. Batch P Gamma Waste P Rele ase Each Each T anks Batch (h) Batch (h) Bnitters (f) 5 x 10-7
-(d)
I-131 1 x 10-6 Dissolved and Entrained Gases P (Gamma One Batch /M M Baitters) 1 x 10-5 (h) P Each Batch (h) M H-3 1 x 10-5 Composite (b) Gross Alpha 1 x 10-7 P Sr-89, 1 Each Batch (h) Q Sr-90 5 x 10-8 l I composite (b) Fe-55 1 x 10-6 , B. Continuous Grab W ' Principal Gamma Release Sample (g) Composite (c) Emitters (f) 5 x 10-7 (e) (g) I-131 1 x 10-6 Grab M Dissolved and Sample (g) Entrained Gases (Gamma Cases) 1 x 10-5 Grab M H-3 1 x 10-5 Sample (g) Composite (c) Gross Alpha 1'x 10-7 , Grab Q Sr-89 Sample (g) Composite (c) Sr-90 5 x 10-8 Fe-55~ l x 10-6 l ! BEAVER VALLEY UNrr 2 3/4 11-3 (
TABLE 4.11-1 NOTAT ION
; a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95 percent probability with 5 percent probability of f alsely concluding that a blank observation represents a "real" s ignal.
For a particular measurement ' system (which may include radiochemical sep ar at ion) LLD = 4.66 Sb (E) (V) (2.22) (Y) exp (-3 Delta T) where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume); sb is the standard 'deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute) E is the counting ef ficiency (as counts per disintegration); V is the sample size (in units of mass or volume) 2.22 is the number of disintegration per minute per picoeurie; Y is the fractional radiochemical yield (when applicable); 3 is the radioactive decay constant for the particular radionuclide; Delta T is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant ef flue nt samples). The value of sb used in the calculation of the LLD for a detection system shall be based on the actual ' observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on a unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g. potassium it milk samples). Typic 71 values of E, V, Y, and Delta l9 T should-be used a the calculations. The LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (af ter the fact) limit for a particular measurement. BEAVER VALLEY UNIT 2 3/4 11-4
TABLE 4.11-1 NOTAT ION l i (continued)
- b. A composite sample is one in which the quantity of liquid sampled is l }
proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in 'a specimen which is representative of the liquids released.
- c. To be representative of the quantities and concentration of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of' the ef fluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite smople to be representative of the ef fluent release,
- d. A batch release exists when the discharge of liquid wastes is from a discrete volume. Prior to sampling for analysis, each batch shall be isolated, and then thoroughly mixed to assure representative smopling.
! e. A continuous release exists when the discharge of liquid wastes' is from a nondiscrete volume; e.g. from a volume of a system having an input flow during the continuous release. This is applicable to the Turbine Building drains when the secondary coolant gross radioactivity (beta and gamma) is 4 greater than 10-5 uCi/ml. l f. The principal gamma emitters for which the LLD specification will apply l are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, ! p Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean v that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, ' shall also be identified and reported. Nuclides which 'are below the LLD for the analyses should be reported as "less than" the nuclide's LLD, and should not be reported as being present ~at the LLD level for that nuclide. The "less than" values should not be used in the required dose calculations. When unusual circumstances result in LLD's higher than required the reasons shall be documented in the semi-annual Radioactive ! Effluent Release Report.
- g. Whenever there is primary to secondary leakage occuring, sampling is done for turbine building drain ef fluents by means of grab samples ta'cen every 4 hours during the period of discharge and analyzed for gross l
' radioactivity (beta and gaussa) at a sensitivity of at le as t 10-7 uCi/ml and recorded in the plant records, along with the flow rate. Primary to i secondary leakage is considered to be occurring whenever measurements 1 indicate that secondary coolant gross radioactivity (beta and gmama) is greater than 10-5 uCi/ml. In addition, two-(2) plant personnel shall l check release calculations to verify that the limits of 3.11.1.1 and 3.11.1.2 are not exceeded. I O. BEAVER VALLEY UNIT 2 3/4 11-5
TABLE 4.11-1 NOTATION (continued)
- h. Whenever the Recirculation Drain Pump (s) are discaerging to catch basin 16 sampling will be performed by means of a grab sample taken every 4 hours during pump operation. lI ,
l 1 [ O : i BEAVER VALLEY UNIT 2 3/4 11-6
. .- . . = - . . - -. . _ - - - - ---
1 ! 3/4.11 RADIOACTIVE FFFLUETT S 3/4.11.1 LIQUID EFFLUENT S DO SE - LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose committments to MEMBER (S) 0F THE PUBLIC from radioactive materials in liquid ef fluents released from the s ite (see Figure 5.1-2) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 arem to the total '
body and to less than or equal to 10 mrem to any organ. I I APPLICABILITY: f l Ac all times. ACTION :
- a. With the calculated dose from the release of radioactive materials in liquid ef fluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the i limit (s) and defines the corrective actions to be taken to reduce the releases, and the proposed corrective actions to be taken to assure
() the subsequent releases will be within the above limits. (This Special Report shall also include 3
- 1. the results of radiological analyses of the drinking water source
- and,
- 2. the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act).*
- b. The provi~sions of Specifications 3.0.3 and 3.0.4 are not applicab le ,
i
- Applicable only if drinking water supply is taken Brom the receiving water body within 3 miles of the plant discharge. (3 miles down stream only).
l l l C:) BEAVER VALLEY UNIT 2 3/4 11-7 t
--.---w1- - , . , e ,s -y g--,,-- --.n--,m,- ..,n. , , ,e,._n - . , ,----en. - - - , - , , , , , ,,-r-- . , , , ,- , -n - ,---, + - - ---.c., --- -
3/4.11 RADI0 ACTIVE EFFLUDIT S 3/4.11.1 LIQUID EFFLU ST S DO SE SURVEILLANCE REQUIREMDIT S 4.11.1.2.1 Dose Calculations - Cumulative . dose contribut ions fr om l 1 liquid ef fluents shall be determined in accordance with the ODCM at le as t once per 31 days. O BEAVER VALLEY UNIT 2 3/4 11-8
I 3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.1 LIQUID EFFLUENT S LIQUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be used to reduce the radioactive materials in each liquid waste batch prior to its discharge when the projected doses due to liquid ef fluent rele ases fr om t he s ite ( See Figure 5.1-1) when averaged over 31 days would exceed 0.06 mrem to the total l I I body or 0.2 mrem .to any organ. APPLICABILITY: At all times. ACTION :
- a. With liquid waste being discharged without tr e atme nt and exceeding the limits specified, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2, a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restate the inoperable equipment to operational status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILIANCE REQUIREMENT S 4.11.1.3.1 Doses due to liquid releases shall be projected at le as t l g once per 31 ' days, in accordance with the ODCM. O BEAVER VALLEY UNIT 2 3/4 11-9
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.1 LIQUID EFFLUELRT S LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to < 21.4 curies, excluding tritium and l dissolved or entrained noble gases,
- a. Refueling Water Storage Tank
- b. Miscellaneous temporary outside radioactive liquid storage tanks.
APPLICABILITY: At all times. A CT ION:
- a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of l radioactive material to the tank and within 48 hours reduce the tank contents to within the limit,
- b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities planned and/or taken to reduce the contents to within the
(} specified limits.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILIANCE REQUIREMENT S II I 4.11.1.4 The quantity of radioactive material contained in each of the l above tanks shall be determined to be within the above limit by analyzing a i I representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. l BEAVER VALLEY UNIT 2 3/4 11-10
3/4.11 RADIOACTIVE EFFLUENT S 3/4.I1.2 CASEOUS EFFLUENT S DOSE RAT E LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate in the unrestricted areas (see Figure 5.1-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:
- a. The dose rate limit for noble gases shall be < 500 mrem / year to the total body and j:, 3,000 mrem / year to the skin, and
- b. The dose rate limit, inhalation pathway only, for I-131, tritium and all radionuclides in particulate form (excluding C-14) with half-lives greater than 8 days shall be < 1,500 mrem / year to any organ.
APPLICABILITY: At all times. ACT ION :
- a. With the dose rate (s) exceeding the above limits, immediately decrease the release rate to comply with the above limit (s).
(} b. Submit a Special Report to the Commission within 30 days pursuant to Specificat ion 6.9.2.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are' not applicab le .
SURVEILLANCE REQUIREMENT S
~
4.11.2.1.1. The dose rate due to noble gaseous ef fluents shall be l determined to be within the above limits in accordance with the methods and , procedures of the ODCM. 1 4.11.2.1.2. The dose rate, inhalation pathway only, for I-131, tr itium l and all radionuclides in particulate form (excluding C-14) with half-lives greater than 8 days in gaseous ef fluents, shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in T able 4.11-2. O BEAVER VALLEY UNIT 2 3/4 11-11 e
O O O TABLE 4.11-2 RADI0 ACTIVE CASEDUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Analysis Type of Detect ion (LI.D) Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml) A. Was te Gas Storage T ank P P Each Tank Each Tank Principal Gamma Emitters E 1 x 10-4 1 Grab Sample H-3 1x 10-6 4 J B. Containment Purge P P b jj Each Purge b Each Purge Principal Gamma Emittersg 1 x 10-4 Grab Sample H-3 1 x 10-6 g C. Vent ilation Systems Mb ,c.e gb Principal Gamma Emitters 1x 10-4 Gr ab
- 1. Process Vent Sample H-3 1 x 10-6
- 2. Elevated Release Point
- 3. Ventilation Vent
- 4. Aux. Bldg. Exh. Hood ( Emergency Use Only)
- 5. Cord. Polish. Bldg. Vent . Exh.
- 6. Decon. Bldg. Vent. Exh.
- 7. Waste Gas Storage Vault Vent. Exh.
- 8. Turbine Building Exhaust Release from' Rad io iod ine and Part iculates ( Airborne) ;
may be limited to the Inhalation Pathway only. BEAVER VALLEY UNIT 2 3/4 11-12
O O O TABLE 4.11-2 (Continued) RADIDACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM dinimum Lower Limit ot Sampling Analysis Type of Detection (LLD) Caseous Release Type Frequency Frequency Activity Analysis (uCi/ml) D. All systems listed aoove which Continuousf W" 1-131 1 x 10-12 produce continuous release.* Charcoal bample 1-133 1 x 10-10 Continuousf W Principal Gamma Emitters 8 Particulate (1-131, Others) 1 x 10-11 Sample Continuousf M Gross alpha 1 x 10-11 Composite Par t icula te Sample Continuousf Q Sr-89, $r-90 1 x 10-11 Composite Particulate Sample Continuousf Noble Gas Noble Cases 1 x lu-D Monitor Gross Beta and Ga===
- Release fron radioiodine and particulates ( Airuorne) may be limited to the inhalation pathway only.
BEAVER VALLEY UNIT 2 3/4 11-13 .
O O' O
; TABLE 4.11-2 (Cont inued)
TABLE NMATION
, s t
- a. The Lower Limit of Detect io n (LLD) is de f ined in Table N ot at ion (a) of Table 4.11-1 of Speci fic a t io n 4.11.1.1.
- ;b . . Whe n reactor coola nt system act ivity exceeds t he ' limit.w's t ated in Specification 3.4.8, analyses shal l be performed once every.24 hours during startup, shutdown and 25 percent load charges 'aM J2 hours af ter ach iev- -
ing the maximd& r.teady , state-power operation unless cont inuous monitoring is pr ov ided . v q f
- c. Trit ium gr'ab' samples . shalt be taken at least once per 24 hours when the refueling canal is flooded.
,j I
- d. Samples shall be changed at le as t once per 7 days and analyses shall be comple ted within 48 hours after
' chang ing (or after removal from sampler) . Sampling and analyses shall also be performed at le as t once per 24 hours, during st ar t u p, shutdoisa and 25 percent }oad changee~ and 72 hours after a(h iev ing the maximum steady state power opolation when RCS ~ activity exceeds"che limits in Specification 3.4.8 unless cont inuous monitor-
~
ing is >provided. When samples collect 6d for 24 hours are analyzed, the corresponding LLD's may be 'jncreased
. by a factor of 10 1
> r .
,~~
- e. Tritium grab samples shall be taken at leas t once ter 7 days from the ventilation exhaust fr om the spe nt fue l
~ [ pool _ area, whenever spent ' fuel in iri the spent fuel pool. -s
- f. The average ratio of the sample flow rate to the sampled stream flos rate shall be known for the t ime period covered by each dose or dose rate calculation made in accordance with Specification 3.11.2.1, 3.11.2.2, aM 3.I1.2.3.
g . T he principal gamma emitters . for which the LLD Speci ficat ion ill apply are exclusively the following rad io-nuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous . emiesions and Mn-54, Fe-59, Co-58, co-60, Zn-65,'Mo-99, Cs-134, Cs-137, Ce-141, Ce-144c for par t icu late emis s io ns . This list does not mean that only these nuclides ar e to be detected and reported. Ot her pe aks which are measur ab le and idept if i d) le , together with the above nuclides, shall also be ident ified and reported. Nuclides which are below tne LLD for t he analyses should not be reported as be'ing . prese nt at the LLD level for that emc lide. When unusual cir cum-stances result in LLD's high than required, the re asons shall be docume nted iii the semi-a nnu a l ef fluent i report. >
~
- h. Only when this release path is in use.
, y D BFAVER VALLEY UNIT 2 ,
3/4 11-14 . d y a W
-/
! 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS
() ' DOSE, NOBLE GASES LIMITING CONDITION FOR OPERATION
> 3.11.2.2 The air dose in unrestricted areas (See Figure 5.1-1) due to r.oble gases released in gaseous ef fluents shall be limited to the following:
2
- a. During any calendar quarter, to j[ 5 mrad for gamma radiation and
} j( 10 mrad for beta radiation.
- b. During any calendar year, to < 10 mrsd for gamma radiation and j[ 20
~
t mrad for beta radiation. APPLICABILITY: At all times. ACTION: 4
- a. With the calculated' air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special
! Report which identifies the causes(s) for exceeding the limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be-taken to assure that subsequent C) relesses will be within the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
I ] ..e i SURVELLLANCE REQUIREMENTS
'3 4.11.2.2 Dose Calculations'. Cumulative dose contributions shall be determined in accordance with the ODCM at least once every 31 days.
4 1 O - l BEAVER VALLEY UNIT 2 3/4 11-15 8
'* * ^ = * * ,.,n- a , , ,- . , , . - - . . . ....,,,,,,.,,e , , , , , , . _ , _ , _ . , , , ,, ,,, . . , _ _ ,, _ _ _ _ _ , , , , _ , ,
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE, RADIO 10 DINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM, O AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITIO1 MR OPERATION 3.11.2.3 The dose to MEMBER (S) 0F THE PUBLIC from ra, ioiodines and radioactive materials in particulate form (excluding C-14), and radionuclides (other than nobles gases) with half-lives grea:.er than 8 days in gaseous effluents released from the site (See Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter to < 7.5 area to any organ, and
- b. During any calendar year to < 15 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form (excluding C-14), and radionuclides (other than noble gases) with half lives greater than 8 days, in gaseous ef fluents exceeding any of the above limits, prepare and i
() submit to the Commission with 30 days, pursuan't to Specification 6.9.2, a special Report, which identifies the causes(s) for exceeding the limit and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUICMENTS 4.11.2.3 Dose Calculations. Cumulative dose contributions shall be determined in accordance with the ODCM at least once every.31 days, i O lBEAVERVALLEYUNIT2 3/4 11-16
i 3/4.11 RADIOACTIVE EFFLUENTS I 3/4.11.2 GASEOUS EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4- The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-1), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation.- The appropriate portions of the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous ) effluent releases from the site (see Figure 5.1-1) when averaged over 31 days would exceed 0.3 mrem to any organ. APPLICABILITY: At all times. ACTION:
- a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days,
' pursuant ' to Specification 6.9.2, a Special Report which includes the following information:
O 1. Identification of the inoperable equipment or subsystems and the
. reason for inoperability, i
- 2. Action (s) taken to restore the inoperable equipment to operational ;
status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. )
SURVEILLANCE REQUIREMENTS i 4.11.2.4 Doses due to gaseous releases from the site shall be projected i at least once per 31 days, in accordance with the ODCM. I l l I
- i. ,
I (:) l, BEAVER VALLEY UNIT 2 3/4 11-17
-. ., . , . _ _ . _ . _ . _ _ . , , _ _ _ . _ _ _ _ , . _ , , , . , ._-r
, + - - -
- r ---T'
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.2 GA SEOU S EFFLUDIT S 1 CA SEOU S WA ST E STORAG E T ANE J l LIMITING CONDITION FOR OPERATION
)
i 3.11.2.5 The quantity of radioactivity cont ained in each gaseous waste storage tank shall be limited to 19,000 curies noble gases (considered as Xe- l] 133). APPLICABILITY: At all times. ACT ION :
- a. With the quantity of radioactive material in any gaseous vaste storage tank exceeding the above limit, inanediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.
- b. Submit a Special Report to the Conumission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are .not applicab le .
O SURVEILIANCE REQUIREMEh7 S 4.11.2.5.1. The quantity of radioactive material contained in each l gaseous waste storage tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank. 1 O BEAVER VALLEY UNIT 2 3/4 11-18
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS O EXPLOSIVE GAS MIXTURE G LIMITING CONDITION FOR OPERATION 3.11.2.6- The concentration of oxygen in the waste gas noldup system shall be limited to < 2 percent by volume whenever the hydrogen concentration exceeds 4 percent by volume. APPLICABILITY: At all times. ACTION:
- a. With the concentration of oxygen in the vaste gas holdup system > 2 percent by volume, innsediately suspend all additions of waste gases to the gaseous waste decay tank and reduce the concentration of oxygen to
< 4 percent within 48 hours.
- b. With the concentration of oxygen in the waste gas holdup system greater than 4 percent by volume and the hydrogen concentration greater than 2 percent by volume, innsediately suspend all additions of waste gases to the af fected tank and reduce the concentration of oxygen to less than or equal to 2 percent by volume within 12 hours.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVELLLANCE REQUIREMENTS 4.11.2.6 The concentrations of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required dERABLE by Table 3.3-13 of Specification 3.3.3.10 or monitoring in conjunction with its associated action statement. O l BEAVER VALLEY UNIT 2 3/4 11-19
3/4.11 RADIOACTIVE EFFLUDTT S 3/4.11.3 SOLID RADI0 ACTIVE WASTE O LIMITING CONDITION FOR OPERATION l 1 3.11.3.1 The solid radwaste system shall be used, as applicable , to solidify and package radioactive wastes, and to ensure meeting the l1 ' requirements of 10 CFR 20 and of 10 CFR 71. Methods utilized to meet these requirements shall be described in facility procedures and in the Process Control Program (PCP) . APPLICABILITY: At all times. ACTION :
- a. With the applicability requirements of 10 CFR 20 and 10 CFR 71 not satisfied, susperd affected shipments of solid radioactive vastes from the site
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicab le .
SURVEILLANCE REQUIREMl!NT S 4.11.3.1. Prior to shipment , solidification shall be verified in l accordance with Station Operating Procedures. 1 4.11.3.2. The semi-annual Radioactive Effluent Release Report in l Specification 6.9.1.12 shall include the following information for each type of ' solid waste shipped of fsite during the report per iod:
- a. Container volume;
- b. Total curie quantity (determined by measurement or estimate);
- c. Principal radionuclides (determined by measurement or estimate);
- d. Type of was te (e .g. , s pe nt resin, compacted dry waste, evaporator bottoms);
- e. Type of container (e.g. , LSA, Type A, Type B, Large Quantity); and
- f. Solidification agent ( e .g . , ceme nt , urea formsidehyde). l1 O 3/4 11-20 BEAVER VALLEY UNIT 2 l
3/4.1I RADIOACTIVE EFFLUENT S 3/4.11.4 TOTAL DOSE LIMITING CONDITION' FOR OPERATION 3.11.4.1 The dose or dose commitment to MEMBER (S) OF THE PUBLIC from all f acility releases is limited to 125 mrem to the total body or any organ l7 ; (except the thyroid, which is limited to i 75 mrem) for a calendar year. l APPLICABILITY: At all t imes . ACT ION :
- a. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specific at ions 3.11.1.2. a , 3.11.1.2.b , 3.11.2.2. a. 3.11.2.2.b ,
3.11.2.3.a, or 3.11.2.3.b, prepare and submit a special report to the commission pursuant to Specification 6.9.2 defining the corrective action and limit the subsequent releases such that the dose or dose commitme nt to MEMBER (S) 0F THE PUBLIC is limited to < 25 meem to the total body or any organ (except thyroid, which is limited to < 75 arem for a calendar year. This special report shall describe the sIeps to ) be taken or modifications necessary to prevent a recurrence. Otherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR 190 Standard. i b. The provisions of Specification 3.0.3 and 3.0.4 are not applicab le . SURVEILLANCE REQUIREMDIT S 4.11.4.1 Dose Calculations. Cususulative dose contributions from liquid l y and gaseous ef fluents shall be determined in accordance with Specifications 3.11.1.2. a , 3.11.1.2.b , 3.11.2.2. a. 3.11.2.2.b , 3.11.2.3. a, and 3.11.2.3.b , and in accordance with the ODCM. l l BEAVER VALLEY UNIT 2 3/4 11-21
. . _ . _ _.-~ . . . _ _ _ _ _ _ _ _ _ _ . . _ - - _ _ _ _ _ _ _ _ . . _ . .
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. ACTION:
- a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Report , a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of cutomatic sampling equipment. If the latter, every ef fort shall be made to complete corrective action prior to the end' of the next sampling period.)
- b. With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 3.12-1 exceeding the limits of Table 3.12-2 when averaged over any calendar quarter,
-() prepare and submit to the Commission within 30 days from the end of af fected calendar quarter a Report pursuant to Specification 6.9.2, l] which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 3.12-2 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if: Concentration (1) + Concentration (2) +....)"" 1.0 Limit Level (1) Limit Level (2). O BEAVER VALLEY UNIT 2 3/\ 12-1
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITINC CONDITION FOR OPERATION
- c. With milk or fresh leafy vegetable samples unavailable from the required number of locations selected in accordance with Specification 3.12.2 and listed in the ODCM, obtain replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1 and the ODCM provided the loca t ions from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations , if available.
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applic ab le .
SURVEILLANCE REQUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be l} collected pursuant to Table 3.12-1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1. O O- BEAVER VALLEY UNIT 2- 3/4 12-2 7 , , ,,. , - - , . - . - . _ - - , - . . -
O O O TABLE 3.12-1 RADIOIDCICAL ENVIRONMENTAL MOITORING PROCRAM i 1 l Expozure Pathway Number of Samples Sampling and T ype of Fr eq ue ncy ( a) snd/or Sample and Locations ** Collection Frequency of Analysis 3
- 1. AIRBORNE
- a. Radioiodine and 5 Locations Continuous operat ion of Each r adioiodine canister ,
i Par t ic ulates sampler with sample Analyze for I-131;
- l. One sample form a collect ion at least control locat ion 10-20 weekly. Part iculate sampler ,
miles distant and in t he gr oss bet a le as t prevalent wind AnalyzIbh;r weekly - Perform dir e ct ion, gamma isotopic aa elysis compos ite (by locat ion).
- 2. One sample from vicinity Sample at least quarterly.
of community having the highest calculated annual average ground level D/Q.
- 2. DIRECT RADIATION 40 Locations. > 2 TLD or a Continuous measurement Camma dose , quarter ly.
pressurized ion chamber at with collection at le as t each location. quar ter ly. (a) Analysis frequency same as sampling frequency unless otherwise specified. (b) Par t iculate samples are not counted for 2; 24 hours after filter change. Per form gamma isotopic analysis on each sample whe n gr os s be t a is > 10 times yearly mean of control samples.
** Sample locations are given on figures and table in Of fsite Dose Calculation Manual (ODCM).
REAVER VALLEY UN IT 2 3/4 12-3
O O O TABLE 3.12-1 (Cont inued ) RADIOLOGICAL ENVIRONMENTAL MOITORING PROGRAM Exposure Pathway Number of Samples Sampling and Type of Frequency (a) and/or Sample and Locat ions ** Collection Frequency of Analysis
- 3. WAT ERBORNE
- a. Sur f ace 2 Locations Composite
- sample collected Gamma isotopic analysis over a period not to exceed of each composite sample;
- 1. One sample upstream, one month.
Tr it ium analys is ~ of
- 2. One sample downstream. composite sample at least quarterly.
- b. Dr inking 2 Locations Composite
- sample collected I-131 analysis of each over a period not to exceed compos ite sanple; 2 weeks Gamma isotopic- analys is of composite sample (by location) monthly; Tr it ium analysis of composite s ample quarterly,
- c. Groundwater N/A - No wells in lower elevat ions
- between plant and r iver .
- d. Sed iment fr om Shor e line I ' Locat ion Semi-Annually Gamma Isotopic analysis sem i-a n nual ly.
- Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours.
**- Sample locations are shown on figures and t ables in 'the Of fsite Dose Calculation manual (ODCM).
1 { (a) Analysis frequency same as sampling frequency unless otherwise specified. BEAVER VALLEY UNIT 2 1
O O O TABLE 3.12-1 (Cont inued ) RADIOIDGICAL ENVIRONMENTAL MOITORING PROGRAM Exposure Pathway Number of Samples Sampling and Type of Frequency (a) and/or Sample and Locat ions ** Collect ion Frequency of Analysis
- 4. ING EST ION
- a. Milk 4 Locations (c) At least bi-weekly when - Gamma isotopic and I-131 animals are on pasture; at analysis of each sample.
- 1. Three samples selected least monthly at other times.
on basis of highest potential thyroid dose using milch census data.
- 2. One local large dairy.
- b. Fishing 2 Locations Semi-annual. One sample of Gamma isotopic analysis on available species. ed ib le por t io ns .
- c. Food Products 4 Locations Annually at time of harvest. Gamma isotopic analysis and (Leafy vegetables) 1-131 analysis on edible
- 1. Three Locations within por t io n.
5 miles.
- 2. One control location.
** Sample locations are shown on figures and tables in the Offsite Dose Calculation manual (ODCM).
(a) Analysis frequency same as sampling frequency unless otherwise specified.
- .(c) Other dair ies may be included as control station or for histor ical continuity. These would not be modified on basis of milch animal census.
REAVER VALLEY UNIT 2 3/4 12-5
_ _ _ _ _ _ _ _ _ _ . _ . . . _ . _ -. __ _ _ . _ . ._ _. . _ _ _..._ __ _ . _ _ ._ _ _ _ _ _ . . . - _ . _ . _ . _ _ .-_.__m _ _. ._ O O O
, TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONNENTAL SAMPLES j Reporting IAveIs '!
4 l Water Airborne Part iculate Fish Milk Br oad Le af Vege t ab le s j Analysis pCi/l or Cases (pCi/m3) (pCi/Kg. wet) (pCi/1) (pCi/Kg. wet) H-3 2 x 104 58) Mn-54 ' 1 x 103 3 x 104 Fe-59 4 x 102 1x 104 Co-58 1 x 103 3 x 104 l Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 Zr-Nb-95 4 x 102 i 1-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x 103 60 1x 103 ! Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102 3 x 102 i h i (') For. dr inking water samples. This is a 40 CFR 141 value. BEAVER VALLEY UNIT 2 3/4 12-6
O O O TABLE 4.12-1 MAXIMJM VALUES FOR THE IDWER LIMITS OF DETECTION (LLD)a 1 Water Airborne Part iculate Fish Milk Food Pr oduct s Sed ime nt j Analysis pCi/l or Gas (pci/m 3) (pCi/KE. wet) (pci/l) (pci/Kg. wet) (pCi/Kg. dry) Gross Beta 4 1 x 10-2 1 H-3 2000 , Mn-54 15 130 Fe-59 30 260 co-58, 60 15 130 Zn-65 39 260 i {, Zr 30 NB-95 15 I-131 1 7 x 10-2 1 60 l Cs-134' 15 5 x 10-2 130 15 60 150 Ca-137 18 6 x 10 2- 150 18 80 18 0 1 C Ba-140 60 60 l La-140 15 15 i NOT E: This list does not mean that only these nuclides are to be detected and reported. ! Other peak's which are measurable and ide nt if iab le , together with the above nuclides , shall be identified and reported. { REAVER VALLEY UNIT 2 3/4 12-7
TABLE 4.12-1 (Continued) TABLE NOTATION
- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95 percent probability with 5 percent ~ probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation): 4.66s b LLD = (E) (V) (2.22) (Y) exp ( "A Delta T) where: LLD is the lower limit of detection as defined above (as pCi per unit mass or volume); S, is the standard deviation of the background counting rate of of the counting rate of a blank sample as appr>priate (as counts l ! per minute); E is the counting efficiency (as counts per disintegration); O vt =* et i-(1 te
- or 1 >>
2.22 is the number of disintegrations per minute per picoeurie; Y is the fractional radiochemical yield (when applicable);
"A is the radioactive decay constant for the particular radionuclide; Delta T.is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).
The value of Sh use.d in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rett or of the counting rate of the blank samples (as appropriate) rather than en an unverified theoretically predicted variance. In calculating 'this LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other- radionuclides nomally present in the samples (e.g., pottasaium 40 in milk samples). Typical values of E, V, Y and delta T should be used in the calculations. O BEAVER VALLEY - UNIT 2 3/4 12-8
i TABLE 4.12-1 (Continued) O The LLD is defined as an a priori (before the fact) limit ; j representing the capability of a measurement system tnd not as a ~ posteriori (after the fact) limit for a p.i. ticular measurement. 4 i b. LLD for drinking water. 1
; c. If parent and daughter are totaled, most restrictive LLD should be l applied.
1 4 I i i i O l 1 l r .f
?
I i l i O BEAVER VALLEY UNIT 2 3/4 12-9 i l
. , , , . _ _ . _ _ . _ _ - - - - . . . _ . . . . - , . . _ . _ - . _ . _ . . _ _ _ _ . , - . , _ , _ , _ _ . . _ . . , _ _ , - . . ~ . _ , _ . _ . _ _ _ _ . _ . , _ . _ . . _ _
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS f)
\J LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. (For elevated releases defined in Regulatory Guide 1.111, (Revision 1) July 1977, the land use census shall also identify the locations of all milk animals and
, all gardens of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of three miles.) APPLICABILITY: At all times. ACTION:
- a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the new location (s).
- b. With a land use census identifying a milch animal location (s) which yields a calculated dose or dose commitment (via the same exposure
() pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1 prepare and submit to the Commission within 30 days, pursuant to
. Specification 6.9.2, a Special Report, which identifies the new location. The new location shall be added to the radiological l J 'i environmental monitoring program within 30 days, if possible. The milk sampling program shall include samples from the three active milch animal locations, having the highest calculated dose or dose commitment. Any replaced location may be deleted from this monitoring program after (October 31) of the year in which the land use census was conducted.
I
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applic ab le .
- Broad leaf vegetation sampling may be performed at the site boundary i in the direction sector with the highest D/Q in lieu of the garden census.
l l I ! I i
' BEAVER VALLEY UNIT 2 3/4 12-10
, - - - . , ~ . . = . , . - . - . - , - , _ , , , . - - , , - , - -
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS SURVEILLANCE REQUIREMENTS 4.12.2.1 Ths land use census shall be conducted at least once per 12 l1 months between the dates of June 1 and October 1 using that information which will provide the best results, such as by a door-to-door survey,* aerial survey, or by consulting local agriculture authorities. 1 0 i. O l Y i t l 1
- Confirmation by telephone is equivalent to door-to-door. ,
I 6 BEAVER VALLEY UNIT'2 3/4 12-11 I r l
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the-corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report,
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.3.1 The results of analyses performed as part of the above l1 required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report. O l 1 O BEAVER VALLEY UNIT 2 3/4 12-12 l 1
l O BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O O
3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Condition for Operation and Surveillance Requirements within Section 3/4. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification' is applicable. 3.0.2 This specification defines those conditions necessary to consti-tute compliance with the terms of an individual Limiting Condition for Oper a-tion and associated ACTION requirement. 3.0.3 This specification delineates ~ the ACTION to be taken for circum-stances not directly provided for in the ACTION statements and whose occur-rence would violate the intent of the specifi.:ation. For example, Specif ic a-tion 3.5.1 calls for each Reactor Coolant System accumulator to be OPERABLE and provides explicit ACTION requirements if one accumulator is inoperable. Under the terms of Specification 3.0.3, if more than one accumulator is inoper ab le , the unit is required to be in at least HOT STANDBY within 1 hour and in an least HOT SHUTDOWN within the following 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spr ay Systems, to be OPERABLE and provides explicit ACTION r equirements if one spr ay system is inoper ab le: Under the terms of Specifi ation 3.0.3, if both of the required Containment Spr ay Systems are inoper able, n e unit is required to be in at O least HOT STANDBY within 6 hours, in at least HOT at0TDOWN within the follow-ing 6 hours and in at least COLD SHUTDOWN in the next 24 hours. It is assumed that the unit is brought to the required MODE within the required times by promptly initiating and carrying out the appropriate ACTION state-me nt . 3.0.4 This specification provides that entry into an OPERABLE MODE, or other specified applicability condition must be made with (a) the full com-plement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements. The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoper able or other spec-ified limits being exceeded. Exceptions to this provision beve been provided for a limited number of specifications when startup with inoper able equipment would not af fect plant safety. These exceptions ar e s t at ed in t he A CT ION s t at eme nt s o f t he appr o-priate specifications. O BEAVER VALLEY UNIT 2 B 3/4 0-1
f 3/4.0 APPLICABILITY BASES l 3.0.5 This specification delineates what additional conditions must be satisfied to permit operation to continue, consistent with the ACTION seatenenes for power sources, when a normal or emergency power source is not ' OPERA BLE. It specifically prohibits operation when one division is inoperable because its normal or emergency power source is inoperable and a system, subsystem, train, component or device in another division is inoper able for another reason. l The provisic ns of the specification permit the ACTION statements associated with individual systems, subsystems, trains, components, or devices to be ce nsistent with the ACTION statraents of the associated alectr ical power source. It~ allows operation to be governed by the time limits of the ACTION statement associated with the Limiting Condition for Operation for the normal or emergency power source, not the individual ACTION statements for each system, subsystem, train, component or device that is determined to be inoper able solely because of .the inoperability of its normal or emergency power source. 1 l For example, Specification 3.8.1.1 requires in part that two emergency diesel gener ators be OPERABLE. The ACTION statement provides for a 72 hour l out-of-service time when one emergency diesel generator is not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems subsystems, trains, components and devices { O PPtt 4 67 e 6 t P 6t cr P <c 14 t 6 tP<6t-This would dictate invoking the applicable ACTION statements for each of the l applicable Limiting Conditions for Operation. However , t he pr ovis io ns o f Specification 3.0.5 permit the time limits for continued operation to be consistent with the ACTION statement for the inoper able emergency diesel generator instead, provided the other specified conditions are satisfied. In this case, this would mean that the corresponding normal power source must be OPERABLE, and all redundant systems, subsystems, tr ains, components, and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e., be capable of performing their design function and have at least one normal or one emergency power source OPERABLE). If they are not satisfied, action is required in accordance with this specification. As a further example, Specification 3.8.1.1 requires in part that two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system be OPERABLE. The ACTION statement provides a 24 hour out-of-service time when both required.offsite circuits are not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable normal power sources, both of the offsite circuits, would also be inoper able. This would dictate invok-O BEAVER VALLEY WIT 2 B 3/4 0-2 L
3/4.0, APPLICABILITY BA SES
- I ing the applicable ACTION statement for the inoper able normal power sources instead, provided the other specified conditions are satisried. In this case, this would mean that for one division the emergency power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, tr ains , components and devices in the other division must be OPERABLE,, or likewise satisfy Specification 3.0.5 (i.e. , be capable of per forming their design functions eind have an emergency power source OPERABLE). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, tr ains , com-ponents and devices in both divisions must also be OPERABLE. If these condi-tions ar e not satisfied, action is required in acccedance with this specifi-cation.
In MODES 5 or 6 specification 3.0.5 is not applicable , and thus the indi-vidual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adhered to. 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other condition, for which the Limiting Con-ditions for Operation are applic ab le . Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other _ conditions are provided in the individual Surveillance Requirements. O s <vetti ce nea treme es for sPeciai rest cePtions eed o 17 de Performed when the Special Test Exception is being utilized as an exception to an indi-
- vidual specification.
4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval.. These tolerances are necessary to provide operational
; flexibility because of scheduling and performance considerations.
i The tolerance values, taken either individually or consecutively over 3 test intervals, are sufficiently restrictive to ensure that' the reliability , associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. l 4.0.3 The provisions of this specification sat forth the criteria for . determinaton of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment , systems or compo-nents are assumed to be OPERABLE if the associated surveillance activities have been satisf actorily per formed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements. O
- BEAVER VALLEY UNIT 2 B 3/4 0-3
__ _ _ _ . _ . . . ~ _ _. __ . _. _ . _ _ _ -- ___ 1 l 3/4.0 APPLICABILITY BA SES
; O l 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been per formed within the specified time interval prior to entry into an OPERATIONAL MODE or other j applicable condition. The intent of this provision is to ensure that sur-veillance activities have been satisfactorily demonstested on a cureent basis as required to meet the OPERABILITY requirements of the Limiting Condition l for Operation.
Under the terms of this specification, for exanple, during initial plant startup or following extended plant outages, the applicable surveillance i activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. 4.0.5 This specification ensures that inservice inspection of A9fE Code
- Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and
.i 3 pumps' and valves will be per formed in accordance with a periodically
; updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above r equir e-j ments has been provided in writing by the Commission and is not a part of i
these technical specifications. l This specification includes a clarification of the frequencies for per-4 forming the inservice inspection and testing activities required by Section
- A XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. _ This U clarification is provided to ensure consistency in surveillance intervals thr oughout these Technical Specifications and to remove any mibiguities rela-tive to the frequencies for performing the required inservice inspection and i testing activities.
I ' Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedance over the ASME Boiler and Pressure Vessal Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into l an OPERATIONAL MODE or other specified applicability. condition takes prece-dence over the ASME Boiler and Pressure Vessel Code provision which allows l ! pumps to be tested up to one week af ter return to normal opeation. And for i example, the Technical Specification definition of OPERABLE does not grant a gr ace period before a device that is not capable of performing its specified function is declared inoperable and takes precedance over the ASME Boiler and Pressure Vessel Coda provision which allows a valve to be incapable of per-forming its specified function for up to 24 hours before being declared inoper ab le . i 4 4 O BEAVER VALLEY UNIT 2 B 3/4 0-4 t
- -_ .- - ~ - , - ,.. - - - . . _ , , , . . - - - - _-_,,,n. ._.-.,__-.nn, -- - , - - , _ , , . . , _ . . , , , - - ."
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL () 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN f1 BA SES
--2 A sufficient SHUTDOWN MARGIN ensures that:
- 1. The reactor can be made suberitical from all operating conditions,
- 2. The reactivity transients associated with postulated accident conditio ns we controllable within acceptable limits, and
- 3. The reactor will be maintained suf ficiently suberitical to preclude inadver te nt criticality in the shutdown condition.
SIUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T The most restrictive condition occurs at EOL, with T at no load *ol4 rating temperature, and is associated with a postulat.ed'sEeam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this ac cide nt , a minimum SHUTDOWN MARGIN of 1.77 percent k/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. With T < 200
- F, the re ac t ivity tr ans ie nt s resulting from a postulated $ team line break cooldown are minimal and a O 1 percent k/k shutdown margin provides adequate protection.
The purpose of borating to the cold shutdown boron concentration prior to blocking safety injection is to preclude a return to criticality should a steam line break occur during plant heatup or cooldown. l i O BEAVER VALLEY UNIT 2 B 3/4 1-1
l l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL l' O 3 1 1 3 aoao" ottor'o" BA SES 1 A minimum flow rate of at least 3000 gpm provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant Sys tem. A flow rate of at least 3000 gpa will circulate an equivalent Reactor Coolant j System volume of 9370 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron reductions will therfore be within the capability for operator recognition and control. O l BEAVER VALLEY UNIT 2 B 3/4 1-2
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) l} BASES The limitations on Moderator Temperature Coef ficient (MIC) are i provided to ensure that t he as sumpt io ns used in t he acc ide nt and transient analyses remain valid through each fuel cycle. The Surveillance Requirement for measurement of the MTC. at the beginning and near the end of each fuel cycle is adequate to confirm the MTC value since this coef ficient change s slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. l h r .1 f j ! i i i 1 : I l I C:) B EAVER VALLEY UN IT . 2 B 3/4 1-3
3/4.1 ,__ REACT IVITY CONTROL SYST EMS 3/4.1.1 30 RATION CONTROL 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY l} BA SES This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure:
~
- 1. The moderator temperature coef ficient is within its analyzed temperature. range,
- 2. The pressurizer is capable of being in an OPERABLE status with a steam bubble,
- 3. The reactor pressure vessel is above its minimum RTNUT temperature , and
- 4. The protective instrumentation is within its normal operat,ing range.
O - O BEAVER VALLEY UNIT 2 B 3/4 1-4 i i
3/4.1 REACIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYST EMS A
.g BASES The boron injection system ensures that negative reactivity control is available during each mode of f acility operation. The components required to per f'o rm this function include:
- 1. Borated water sour ces ,
- 2. Charging pumps,
- 3. Separate flow paths,
- 4. Bor ic ac id tr ans fe r pumps ,
- 5. Associated heat tr acing systems , and
- 6. An emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to e nsure single functional capability in the event an assumed failure renders one of the systems inoper able. Allowable out-of-service periods ensure that minor () component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. The boration capability of either system is suf ficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.0 percent delta k/k af ter zenon decay and cooldown to 200*F. The maximum boration capability requirements occur at EOL from full power equilibr ium xenon condit ions and requires ~13,390 gallons of 7000 ppm borated water from the boric acid storage tanks or 58,965 gallons of 2000 ppm borated water from the ' refueling water stor age tank. The associated technical specification limit on the refueling water storage tank volume of water while operating has been established at 859,248 gallons 'to account for reactivity considerations , t he NP SH requirements of the BCCS system, and the water required for containment s pr ay oper at io n. With the RCS temperature below 200*F, one injection system is acceptable' without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERKTIONS and positive reactivity change in the event the s ingle injection system becomes inoper ab le . P BEAVER VALLEY UNIT 2 B 3/4 1-5
- r n -
. - - . - , , , . , , .,.,.~,.....,.,..n.,.n.
, , _ . . _ , . , _ .,. , .,_.,,,,n__,.,___
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SY ST EMS T
] BA SES The boration capability required below 200*F is suf ficient to provide a SHUTDOWN MARGIN of 1 percent delta k/k . af ter xenon decay and cooldown from 200*F to 140*F. This condition ' requires either 2315 gallons of 7000 ppm borated water from the boric acid storage tanks or 10,196 gallons of 2000 ppm borated water from the refueling water storage tank. The associated technical specification limit on refueling water etor age tank volume of water while below 200*F has been established at 217,000 gallons to account for reactivity considerations and the NPSH requirements of the EECS system.
The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of bor ated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron i concentration ensure that:
- 1. Suf ficient water is available within cont ainment to permit recirculation cooling flow to the core, and
- 2. The reactor will remain suberitical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.
These as sumptions are cons istent with the LOCA analyses .
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O BEAVER VALLEY UNIT 2 B 3/4 1-6
3/4.1 REACTIVITY CONTROL SYST EMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES BASES The specifications of this section ensure that:
- 1. acceptable power distribution limits are maintained,
- 2. the minimum SRUTDOWN MARGIN is maint ained, and
- 3. the potential ef fects of rod misalignment on associated accident analyses are limited.
OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within + 12 steps (indicated position), of the respective group demand counter position. OPERABILITY of the control rod position indicators is required to determine control rod position and thereby ensure compliance with the control rod alignme nt and insertion limits. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires me asurement of peaking f actors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses af fected by a misaligned rod are re-evaluated to confirm that the results remain valid during future ope r at io n. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Me asur eme nt with T greater than or equal to 541*F and with all reactor coolant pumps operN$ng ensures that the measured drop times .will be representative of insertion times experienced during a reactor trip at operating conditions. Continuous monitoring of rod position with respect to insert ion limits and rod deviation is provided by the rod insertion limit monitor and rod deviation monitor , respect ivity. OPERABILITY of the rod deviation monitor is verified by comparison of the indicated positions versus the respective group y demand counters at least once per 12 hours. . If the rod deviation monitor or the rod insertion limit monitor is INOPERABLE, the frequency of manual comparison of indicated rod position is increased to an interval of at le as t once per 4 hours. O BEAVER VALLEY UNIT 2 B 3/4 1-7
, , - - - _ - , _ . . . . . , -~. - - - . . .,
3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) () BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents 'of Moderate Frequency) events by: a.' Maintaining the minimum DNBR in the core > 1.30 during normal opera-tion and in short term tr ansients , and
- b. Limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Conditions I events provides assurance that t he initial conditions asumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.
The definitions of hot channel factors as used in these Specifications are as follows: F9 (Z) Heat Flux Hot Channel Factor , is defined as the maximum local heat flux on the surf ace of a fuel rod at core elevation Z divided by the aver age fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods. N (~Y f Nuclear Enthalpy Rise Hot Channel Factor, is defined as
\~/ H the ratio of the integral of linear power along the rod with the highest integrated power to the aver age rod power.
The limits on AXIAL FLUX DIFFERENCE assure that the Fq (Z) upper bound envelope of 2.18 times the normalized axial peaking f actor is not exceeded during either normal operation or in the event of re non redis tr ibution following power ch'anges. Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their ree pective insertion limits and should be inserted near the ir normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by t he fr ac t ion of RAT ED TH ERMAL POWER is t he t ar ge t flux difference at RAT ED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by nultiplying the RATED THERMAL POWER value bvy the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect cora burnup considerations. O BEAVER VALLEY UNIT 2 l B 3/4 2-1 ( F
.-. -- . . . ,, - , e , - ~ - - -
j 3/4.2 POWER DISTRIBUTION LIMIT S 3/4.2.1 AXIAL FLUX DIFFERENCE ( AFD) O BASES Although it is- intended that the plant will be operated with I the AXIAL FLUX DIFFERENCE within the + 7 percent target band about the target I i flux dif fer e nce , during rapid plant THERMAL POWER reduct ions , cont r ol rod motion will cause the AFD to deviate outside of the t arget band at reduced THERMAL POWER IAvels. This deviation will not af fect the me non red is tr ibut io n suf ficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target
- band) provided the time duration of the deviation is limited. Accordingly, a j 1 hour penalty deviation limit cumulat ive during the previous 24 hours is provided for operation outside of the targe t band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50 percent and 90 percent i
of RAT ED THERMAL POWER. For THERMAL POWER levels between 15 percent and 50 i perce nt of RAT ED THERMAL POWER, deviat ions of the AFD outside of the t arge t j band are less significant. The penalty of. 2 hours actual time reflects this reduced significance. J Provisions for monitoring the AFD on an automatic basis are derived from the plant proce s s computer through the AFD Monitor Alarm. T he computer determines the one minute aver age of each of the OPERABLE excore 4 detector outputs and provides an alarm message immediately if the AFD for at le as t 2 of 4 or 2 of 3 OPERABLE excore channels are outside the t arge t band and the THERMAL POWER is gr e ate r than 90 per ce nt of RAT ED THERMAL POWER. O During operation at THERMAL POWER levels between 50 percent and 90 percent and 15 perce nt and 50 per ce nt RAT ED THERMAL POWER, the computer outputs an i alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band near the beginning of core life. i t i ) i l O BEAVER VALLEY UNIT 2 8 3/4 2-2
FIGURE B 3/4 2-1 . TYPICAL INDICATED AXIAL FLUX OIFFERENCE VERSUS THERMAL POWER AT BOL O V 7Y 71 100% .! l lt- . l. . l:. T i :h = .
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-20% -10% 0 +10% +20%
O~ INDICATED AXIAL FLUX OIFFERENCE Beaver Valley Unit 2 B 3/4 2-3
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F q (Z) N 3/4.2.3. NUCLEAR ENTHALPY FACTOR - F AH BASES ] The limits on heat flux and nuclear enthalpy hot channel factors ensure that:
- 1. The design limits on peak local power density and minimum DNBR are not exceeded, and
- 2. In the event of a LOCA, the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200*F.
Each of these hot channel factors are measurable, but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is suf ficient to insure that the hot channel f actor limits are caintained provided:
- a. Control rod in a single group move together with no individual rod insertion dif fering by more than + 12 steps from the group demand position.
O b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.
- c. The control rod insertion limie.s of Specifications 3.1.3.4 and 3.1.3.5 are maintained,
- d. The axial power distribution, expressed in terus of AXIAL FLUX
' DIFFERENCE is maintained within the limits.
i N The relaxation in F as a function of THERMAL POWER allows changes in the radial powebhape for all permissible rod insertion N limits. F will be maintained within its limits provided conditions a throughd$bove, are maintained. O BEAVER VALLEY UNIT 2 B 3/4 2-4
3/4.2 POWER DISTRIBUTION LIMrT S 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) O 3/4.2.3 NUCLEAR ENTHALPY FACTOR - FAH BA SES When an OF measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate experimental error allowance for a full core map taken with the incore detector flux mapping system and 3 percent is the appropriate allowance for manuf acturing tolerance. N The specified limit of F contains an 8 percent allowance uncertainties which means that normal, kUll power, three loop oper ation will N result in F is less than or equal to 1.55/1.08. AH Fuct rod bowing reduces the value of DNB ratio. Credit is
~
available to offset this reduction in the generic margin. Tre generic design margins , totaling 9.1% DNBR, completely of fset any rod bow penalties (less than 3% for the worst case which occurs at a bur nup of 33,000 MWD /NEU). This margin includes the following:
- 1) Design limit DNBR of 1.30 vs.1.28
- 2) Grid Spacing (Ks) of 0.046 vs. 0.059
- 3) Thermal Dif fusion coef ficient of 0.038 vs. 0.059
- 4) DNBR Multiplier of 0.365 vs. 0.88 5
- 5) Pitch reduction ,
. i The radial peaking f actor F assurance that the hot channel IIc(Z) tor,isFmeasured periodically (Z), remains within its to provide limit.
9 J The F limit for Rated Thermal Power (F ) as provided in the RadiaIIPeaking Factor Limit Report per Spleification 6.9.1.14 was determined ; from expected power control maneuvers over the full r ange of bur rup ' conditions in the core. ! l ; 1 r f f ! ($) B 3/4 2-5 BEAVER VALLEY UNIT 2 1 h [ l
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO BASES The Quadrant Power Tilt Ratio limit assures that the radial power distribution satisfies the design values used in the power capability analy1is. Radial power distribution measurements are made during startup testing and periodically during power operation. The-Limit of 1.02 at which corrective action is required ' provides DIG and linear heat generation rate protection with x-y plane power tilts. The two hour tune allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event i such action does not correct the tilt the margin for uncertainty on F i is reinstated by reducing the maximum allowed, power by 3 percent for9 each percent of tilt in excess of 1.0. i i t i l t r
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O l BEAVER VALLEY UNIT 2 B 3/4 2-6
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS ro BASES The limits on the DNB related parameters assure that each of the parameters are asintained within the normal steady state envelope of l operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically
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' demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored i within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis. 9 () l 3 i I O lBEAVERVALLEYUNIT2 a 3/4 2-7
l 3/4.3 IN ST RUM ENT AT ION 3/4.3.1 and 3/4.3.2 PRUTECTIVE AND ENGINEERED SAFETY FEATURES (ESP) i IN ST RUMDrTAT ION l O RASES The OPERABILTTY of the Engineered Safety Feature Actuation System instrumentation and interlocks ensure that:
- 1) The associated action and/or reactor trip will be initiated when the par ameter monitored by each channel or combination thereof reaches its setpoint,
- 2) The specified coincidence logic is maint ained,
- 3) Sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and ,
- 4) Suf ficient system functional capability is available from diverse p ar ame te r s .
The OPERABILTTY of these systems is required to provide the j over all reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and tr ansient conditions. Tte integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. () The surveillance requirements specified for these systems ensure that the over all system functional capability is maintai*.ed compar able to the original design standards. The periodic surveillance tests performed at the minimum frequencies are suf ficient to demonstrate this capability. The measurement of response time at the specified frequencies provides assurance that the protective and ESF actuation function associated with each channel is completed within the time limit assumed in the accident
! analyses. No credit was taken in the analyses for those channels with response times indicated as not applic ab le .
Response time may be demonstr ated by any series of sequent ial, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: ! 1) In place, onsite, or offsite test measurements or
- 2) Utilizing replacement sensors with certified response times.
I BEAVER VALLEY UN TT 2 B 3/4 3-1 ! I ; [
i i 3/4.3 IN ST RUM ENT AT ION i l 3/4.3.1 and 3/4.3.2 PRUT_ECTIVE AND ENGINEERED SAFETY FEATURES (ESF) IN ST RUM ENTAT ION j i (2) BA SES j l i The Engineered Safety Feature Actuation System interlocks per form the following funct ions: P-4 Reactor tripped - actuates turbine trip, closes main feedwater valves on T evg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam gener ator water level signal, allows safety injection block so i that components can be reset or tr ipped, j i' Reactor not tripped - prevents manual block of safety injection. I ! l P-11 Above the setpoint P-11 automatically reinstates safety injection [ actuation on Low pressuriser pressure, automatically blocks t steamline isolation on high steam pressure rate, enables safety l j injection and steauline isolation on (Loop Stop Valve Open) with low steamline pressure, and enables auto actuation of the pressurizer l l i j PORVs. Below the setpoint P-11 allows the martaal block of safety injection i actuation on low pressuriser pressure, allows manual block of safety l injection and steamline isolation on (Loop Stop Valve Open) with low
' steamline pressure and enabling steamline isolation on high steam i pressure rate, automatically disables auto actuation of the l j pressurizer PORVs unless the Reactor Vessel Over Pressure l
Protection System is in service. ! P-12 Above the setpoint P-12 automatically reinstates an arming signal to l t he steam dump system. Below the setpoint P-12 blocks steam dump i and allows manual bypass of the steam dump block to cooldown with
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condenser dump valves.
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[ k i l i 1 I i
- \
i i i I f s l C) BEAVER VALLEY UNIT 2 B 3/4 3-2 I
b 3/4.3 IN ST RUMDITAT ION 3/4.3.3 MONITORING IN STRUMDIT ATION 3/4.3.3.1 RADIATION MONITORING ! BA SES The OPERABILITY of the radiation monitoring channels ensures that:
- 1) The radiation levels are continually measured in the areas served by I the individuals channels; ,
- 2) The alarm or automatic action is initiated when the radiation level j trip setpoint is exceeded; and 1 ,
j 3) Suf ficient information is available on selected plant parameters to 4 monitor and assess these variables following an accident. This ' capability is' consistent with the recommendations of NUREG-0737,
" Clarification of the TMI Action Plan Requirements," October,1980.
i l 1 I i ,1 O .
** y 1
l 3 1 . l
- e l l i
t ! l l f I . O BEAVER VALLEY UNIT 2 8 3/4 3-3 l 4 i $ ] l t [
3/4.3 INSTRUMINTATION 3/4.3.3 MONITORING IN STRUMLNTATION 3/4.3.3.2 MOVABLE INCORE DET ECTORS BA SES The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spat ial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. For the purpose of measuring FQ (Z) or F N H, a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 19 76, may be used in recalibration of the excore neutron flux detection system, a nd full incore flux maps or symmetric incore thimbles may be used for monitoring the Quadrant Power Tilt Ratio when one Power Range Channel is inoper ab le . O O BEAVER VALLEY UNIT 2 8 3/4 3-4
.- . - _ = . _. . .~ - - _ . . . _ _ _ . - . _ ._ . _. - _ .. .. _ -- . . _ . . -
t i t 3/4.3 IN STRUMENTATIO!1 3/4.3.3 MONITORING INSTRUMDiTATION r 3/4.3.3.3 SEISMIC IN STRUMDITATION I> BASES _n j The OPERABILITY of the seismic instrument ation ensures that suf ficie nt capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those fe atures import ant to 3 j safety. This capability is required to permit comparison of a measured response to that used in the design basis for the f acility and is consistent with the recoussendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes." ,
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1 3/4.3 IN ST RUMENT AT ION i 3/4.3.3 P.'ONITORING IN STRUMENTATION J (} 3/4.3.3.4 MET EOROLDCICAL IN STRUMENT ATION g f EASES j I The OPERABILETY of the meteorological instrumentation ensures I that sufficient meteorological data is available for estimating potential i radiation doses to the public as a result of routine or accidental release of I radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health ! and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programa." i I l , t i
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l j i 3/4.3 IN ST RUMDiTAT ION l 4, 3/4.3.3 MONITORING IN ST RUMENTATIOR 3/4.3.3.5 RDETTE SHUTDOWN IN STRUMDITATION 1 ] BASES I - n :-. -_ _ - - l The OPERABILITY of the remote shutdown. instrumentation ensures j that suf ficient capability is available to permit shutdown and tsaintenance of l Har STANDBY of the f acility from locations outside the control room. This j capability is required in the event control room habitability is lost and I this capability is consistent with GDIERAL DESIGN CRIT ERIA 19 of 10 CFR 50. 1 l I 1 a 4 5 I lO : l l l a ! 1 i i i 1 + U
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J lO 1 BEAVER VALLEY UNIT 2 B 3/4 3-7 i i i i
3/4.3 IN ST RUMENTAT ION 3/4.3.3 MONITORING IN STRUMENTATION 3/4.3.3.6 FIRE DET ECTION l O BASES OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in the ir early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall f acility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols or in-cont ainment air temperature nonitoring in the af fected areas is required l} to provide detection capability until the inoperable instrumentation is restored to OPERABILITY, I i I O l l 1
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i f BEAVER VALLEY UNIT 2 B 3/4 3-8
3/4.3 IN ST RUMENT AT ION 3/4.3.3 MONITORING INSTRUMDITATION 3 /4.3. 3. 7 CHLORIN E DET ECT ION SY ST EM S 1
.O o BA SES The OPERABILITY of the chlorine detection system ensures that suf ficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the ~
reconumendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," January, 1977. I O 1 l i i l O BEAVER VALLEY UNIT 2 B 3/4 3-9 1 i I
3/4.3 IN ST RUMENT AT ION 3/4.3.3 MONITORING IN STRUMENTATION 3/4.3.3.8 ACCIDENT MONTTORING IN STRUMENTATION I BASES x_-- - ._-_ _ = ------- The OPERABILITY of the accident monitoring ins tr ume nt at io n ensures that suf ficient information is available on seleted plant par ameter s to monitor and assess these variables during and following an accident . This capability is consistent with the recommendations of Regulatory Guide 1.97,
" Instrumentation for Light-Water-Cooled Nuclear Plants to assess Plant Conditions During and Following an Accident", December 1975 and NUREG 0578,
'*TMI-2 Lassons Learned Task Force Status Report and Short-Tern Recommendations."
O i \ BEAVER VALLEY UNIT 2 8 3/4 3-10 l
3/4.3 IN ST RUMENTAT ION 3/4.3.3 MON TIORING IN ST RUMENT AT ION l 3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l1 (:) BASES _- - 7 i The radioactive liquid ef fluent instrumentation is provided to I monitor and control, as applicable, the release of radioactive materials in j liquid ef fluent during actual or potential liquid releases. The alarm / trip l setpoint for these . instruments shall be calculated in accordance with the i procedures in the ODCM to ensure that the alarm / trip will occur prior to
! exceeding the limits of 10 CFR 20. The OPERABILETY and use of this l j instrumentation is consistent with the requirements of GENERAL DESIGN i CRITERIA 60, 63 and 64 of Appendix A to 10 CFR 50.
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1 3/4.3 IN ST RUMENTAT ION 3/4.3.3 MONITORING IN STRUMENTATION 3/4.3.3.10 RADIOACTIVE GASEDU S EFFLUENT MONITORING IN STRUMENT ATION 1 BASES The radioactive gaseous ef fluent instrumentation is provided J to monitor and control, as applicable, the release of r adioactive materials in gaseous ef fluents during actual or potential releases of gaseous
, ef fluents. The alarm / trip setpoints for these instruments shall be calculated, i in accordance with the procedure in the ODCM to ensure the alarm / trip will i occur prior to exceeding the limits of 10 CFR 20. The OPERABILTTY and use of ' this instrumentation is consistent with the requirements of GENERAL DESIGN
{ CRETERIA 60, 63 and 64 of Appendix A to 10CFR 50. 1 4 i i i i i 1 I ( i I l i : i i l I I t I ! i i ? f
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i I I ! i i () BEAVER VALLEY UNIT 2 B 3/4 3-12 l l L ! + 1 l
3/4.4 REACTOR COOLANT SYST EM 3/4.4.1 REACTOR COOLANT LOOPS BASES The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations 4 and anticipated transients. In MODES 1 and 2, with one reactor coolant loop 4 not in operation, THERMAL POWER is restricted to < 30 percent of RATED THERMAL POWER until the Overtemperature T trip II reset. Either act ion ensures that the DNBR will be maintained above 1.30. A loss of flow in two loops will cause a reactor trip if operating above P-7 (10 percent of RAT ED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (30 percent of RATED THERMAL POWER). In MODE 3, a single reactor coolant loop provides suf ficient heat removal capability for removing decay heat; however , due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a 1 suberitical condition, two operating coolant loops are required to meet the DNB design basis for' this condition II event. In MODES 4 and 5, a single reactor coolant loop or RHR l subsystem provides suf ficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus , if the reactor coolant loops are not OPERABLE, this Specification requires two RHR loops to be OPERABLE. The operation of .one Reactor Coolant Pump or one RHR pump O' - provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the , Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. r The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 275*F are provided to prevent RCS
, pressure transients, caused by energy additions from the secondary system, ,
which could exceed the limits of Appendix G to 10 CFR 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCP's to when the secondary water i temperature of each steam generator is less than 50*F above each of the RCS : cold leg ceaperatures. l 4 O BEAVER VALLEY UNIT 2 B 3/4 4-1 i
3/4.4 _ REACTOR COOLANT SYST_EM 3/4.4.1 REACTOR COOLANT LOOPS BA SES The requirement to maint sin the boron concenteation of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop. Verification of the boron concentration in an idle loop iimmediately prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at leas t 90 minutes pr ior to opening its stop valves ensures adequate mixing of the coolant in this loop and prevents any reactivity ef fects due to boron concentr ation s tr at i ficat io ns . Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delays isolated loop startup until its temperature is within 20*F of the operating loops. Making the reactor suberitical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient. O i O BEAVER VALLEY UNIT 2 8 3/4 4-2
l I j 3/4.4 REACTOR COOLANT SYST EM 1
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j 3/4.4.2 and 3/4.4.3 SAFETY VALVES , I i l i BASES ] -- __ _- _ _. i The pressurizer code safety valves operate to prevent t he RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve . is designed to relieve 345,000 pounds per hour of saturated steam at the l1, valve setpoint. Tre relief capacity of a single safety valve is adequate to j relieve any overpressure condition which could occur during shutdown. In the l eve nt that no safety valves are OPERABLE, an operating RHR loop, connected to > the RCS, provides overpressure relief capability and will prevent RCS j overpressure. j I During operation, all pressuriser code safety valves must be 'j OPERABLE to prevent the RCS from being pressurised above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater l
- than the maximum surge rate resulting from a complete loss of load _ assuming [
no reactor trip until the first Reactor Protective System trip setpoint is j reached (i.e. , no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or i steam' dump valves. I Demonstration of the safety valves' lift settings will occur i only during shutdown and will be performed in accordance with the provisions of Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Code, dated July 1974. l i i 2 d 4 h i i i t l I l a i i 1 b i j BEAVER VALLEY UNIT 2 8 3/4 4-3 i I l
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.4 PRESSURIZER BA SES The requirement that 150 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of of f site power condition to maintain natur al circulation at HOT STAND BY. e O 4 I O BEAVER VALLEY UNIT 2 B 3/4 4-4
t 3/4.4 REACTOR COOLANT SY ST EM ! 3/4.4.5 ST EAM GENERATORS j () SASES I t
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l One OPERABLE steam generator in a non-isolated reactor coolant i loop provides suf ficient heat removal capability to remove decay heat after a i r e actor shutdown. The recuirement for two OPERABLE stea= generators, combined l with other requirements of the Limiting Conditions for Operation ensures ! adequate decay heat removal capabilities for RCS temperatures greater than [ 350*F if one steam generator becomes inoper able due to single failure ; considerations. Below 350*F, decay heat ~ is removed by the RHR system. I l1 The Surveillance Requirements for inspection of the steam [ gener ator tubes ensure that the structur al integrity of this portion of the RCS will be maintained. The progr am for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision ;
- 1. Inservice inspection of steam generator tubing is essential in order to i maintain surveillance of the conditions of the tubes in the event t hat t her e !
is evidence of mechanical damage or progressive degr adation due to design, { manuf acturing errors, or inservice conditions that le ad to cor r os ion. [ l Inservice inspection of steam generator tubing also provides a means of I char acterizing the nature and cause of any tube degr adation so that I corrective measures can be taken. f 1 The plant is expected to be operated in a manner such that the r () secondary coolant will be maintained within those par ameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary I f coolant chemistry is not maintained within these par ameter limits, localized ! corrosion may likely result in stress corrosion cracking. The extent of , cracking during plant operation would be limited by the limitation of steam l generator tube leakage between the primary coolant system and the secondary [ coolant system (pr imary-to-secondary leakage = 500 gallons per day per steam ! generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the [ loads imposed during normal operation and by postulated accidents. Operating j plant have demonstrated that primary-to-secondary leakage of 500 gallons per ; day per steam generator can readily be detected by radiation monitors of , steam gener ator blowdown. Leakaee in excess of this limit will require plant i shutdown and an unscheduled inspection, during which the leaking tubes will [ be located and plugged. 3 I r l f i t B 3/4 4-5 BEAVER VALLEY UNIT 2-P e
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3/4.4 REACTOR COOLANT SYST EM 3/4.4.5 STEAM GDIERATORS l O BASES Wastage-type defects are unlikely with the All Volatile Treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice stean generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40 percent of .the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetr ated 20 per cent of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the l~ Connaission pursuant to Specification 6.6 prior to resumption of plant oper at ion. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis , laboratory examinations , tests, additional eddy-current inspect ion, and revision of the technical specif icat ions , if necessary. O o U BEAVER VALLEY UNIT 2 B 3/4 4-6
3/4.4 REACTOR COOLANT SYST EM 3/4.4.6 REACTOR COOLANT SYST EM LEAKAGE LEAKAGE DETECTION SY ST EMS BA SES The RCS leakage detection systems required by this ) specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary leakage Detection Systems." f O
- BEAVER VALLEY UNIT 2 B 3/4 4-7 1
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.6 REACTOR C00LAh*r SYSTEM LEAKAGE OPERATIONAL LEAKAGE RASES Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this le akage can be reduced to a threshold value of less than 1 gpm. This threshold value is suf ficiently low to ensure early detection of additional leakage. The 10 gpm IDENTIFIED LEAKAGE ' limitation provides allowance for a limited amount of leakage from known sources whose presence will not inter fere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 28 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2,235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses. The total steam generator tube leakage limit of I gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of P ar t 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in s the analysis of these accidents. The 500 gpd leakage limit per s te am s generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure bound ar y. Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolat ion removes the source of potential failure. O BEAVER VALLEY UNIT 2 B 3/4 4-8 l
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.7 CH EMI ST RY BA SES i The limitations on Reactor Coo. int System chemistry ensure ! that corrosion of the Reactor Coolant System is minimized and reduces the ! potential for Reactor Coolant System leakage or f ailure due to str ess corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated ef fects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with cont aminant concentration levels in excess of the Steady State Limits, up to the Tr ansient Limitc, for the specified limited time intervals without having a significant ef fect on the structural integrity of the Reactor Coolant system. The time interval permittirg continued operation within the restr ictions of the Transient Limits provides time for taking corrective actions to restore the contaninant concentrations to within the Steady State Limits. The Surveillance Requirements provide adequate assurance that conce ntr at io ns in excess of the limits will be detected in suf ficient time to take corrective action. o O BEAVER VALLEY UNIT 2 B 3/4 4-9
3/4.4 REACTOR COOLANT SY ST EM 3/4.4.8 SPECIFIC ACTIVITY O. BA SES The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not
- exceed an appropriately small fr action of Part 100 limits following a steam gener ator tube rupture accident in conjunction with an assumed steady state pr imar y-to-secondary ste am generator leakage rate of 1.0 gpm.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uCi/gr am DOSE EQUIVALEKT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 uCi/gr an DOSE EQUIVALDTT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. Reducing T to < 500*F prevents the release of activity should a steam generator fu$e rupture since t he satur at ion pr e~s sur e of the primary coolant is below the lift pressure of the atmospheric steam relief v alve s . The Surveillance Requirements provide adequate assurance that g-- excessive specific activity levels in the primary coolant will be detected in (_) suf ficient time to take corrective action. Information obtained on iodine spiking will be used to assess the par ameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. O BEAVER VALLEY UNIT 2 B 3/4 4-10
3/4.4 REACTOR COOIANT SYST EM 3/4.4.9 PRESSURE / TEMPERATURE LIMIT S O BASES All components in the Reactor Coolant System are designed to withstand the ef fects of cyclic loads due to system temper ature and pressure changes . These cyclic loads are introduced by normal load transients, reactor tr ips , and startup and shutdown' operations. The various categories of load cycles used for design purposes are provided in Sect ion 3.9 of the FSAR. During startup and shutdown, the rates of . temper ature and pressure changes are limited so that the maximum specified heatup and cooldown rates ar e consistent with the design assumptions and satisfy the stress limits for cyclic operation. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary 6com compressive at t he inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as t he governing location. The heatup analysis also covers the determination of pressure-temper atur e limitations for the case in which the outer wall of the vessel p becomes cl.e controlling location. The thermal gradients established during (/ heatup produce tensile stresses at the outer wall of the vessel. These s tr es se s ar e 9dditive to the pressure induced tensile stresses which are already present. The thermal induced stresses 'at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that descr ibed for the heatup of the inner wall cannot be de fined . Subseque nt ly, for the cases in which the outer wall of the vessel becomes the stress controlling - location, each heatup rate of interest must be analyzed on an individual basis. The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with eitter the inside or outside wall controlling, for any heatup rate up to 60*F per hour . The cooldown lbsit curves Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception- that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce ternile stresses while producing compressive stresses at the outside wall.{The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference 2 temperture at t he e nd of 10 EFPY. I BEAVER VALLEY UNIT 2 B 3/4 4-11
3/4.4 REACTOR COO 1 ANT SY ST EM 3/4.4.9 PRESSURE /T EMPERATURE LIMIT S BA SES The reactor vessel materials have been tested to determine the ir initial RTNUT; the results of these tests are shown in Table B.3/4.4-1. Reactor oper at ion and r esult ant fast neutron (E>l Mev) irr adiat io n will cause an increase in the ENUT. Therefore, an adjusted reference temper ature, based upon the fluence, copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4~.4-1 and Regulatory Guide 1.99, Revision 1, "Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" or the Westinghouse copper trend curves shown by Figure B 3/4.4-2. The heatup and cooldown limit curve s Figures 3.4-2 and 3.4-3 include predicted adjustments for this shif t in ENDT at the end of 10 EFPY as well as adjustments for possible errors in t he pr es s ur e a nd temper atur e se ns ing ins tr ume nt s . Heatup and cooldown limit curves are calculated using the most limiting value of RINUT (reference nil-ductility temperatur e) . ENUT increases as the material is exposed to fast-neutron radiation. T hu s , to find the mos t limiting RINDT at any time period in the r e ac t or 's li fe , RTNUT due to the radiation exposure associated with that time period must be added to the original unieradiated RINUT. The ' extent of the shift in RTNUT is enhanced by certain chemical elements (such as copper) present in reactor vessel steels. The Westinghouse trend CJ delta RTNUT for reactor vessel steels we shown in Figure B 3/4.4-2. Given the copper content of the mos t limiting material, the radiation-induced delta RTNUT can be estimated from Figure B 3/4.4-2. Fast-neutron fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are given as a function of full power
~
service life in Figure B 3/4.4-1. The data for all other ferritic materials in the reactor- coolant pressure boundary are examined to insure that no ot her component will be limiting with respect to RTNUT - The preirradiation fracture-toughness properties of the Beaver Valley Unit 2 reactor vessel materials are presented in Table B 3/4.4-1. The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary we determined in accordance with the 1972 Sunuser Addenda to Section III of the ASME Boiler and Pressure Vessel Code. ~ BEAVER VALLEY UNIT 2 B 3/4 4-12
- - - ~ ' - - - - - - - - - - - - - - - -
w_. - -.a _.s - A-3/4.4 REACTOR C00LAlff SY ST EM 3/4.4.9 PRESSURE / TEMPERATURE LIMIT S O BASES The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total str ess intens ity factor , K for the combined thermal and pressure stresses at any time during heaku,p and cooldown cannot be greater than t he r e fer e nce s tr es s intensity factor , K for the metal temperature at that time. K is obtained . from the refe,rence fracture toughness curve, defined in IR Appendix G to the A SME Code.2 T he ~ K IR curve is given by the equation: K IR
= . + 1. eXP N.W M NUT where K is the reference stress intensity factor as a function of the metal tImperature T and the metal reference nil-ductility temperature RT the heatup-cooldown analys is is debe. dThus, the governing in Appendix G to the equation ASME Code for 2 as follows:
CKg+KItiKIR -2) where K is' the stress intensity factor caused by membrane (pressure) skr es s K is the stress intensity factor caused by the thermal R gr adient s K is a function of temperature relative to the RT TR NUT the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical. 2 ASME Boiler and Pressure ' Vessel Code, Section III, Division I-l Appendix, " Rules f or Construction of' Nuclear Vessels ," Appendix .G.
" Protection Against Non-ductile Failure," pp. 559-569, 1980 Ed it io n, American Society of Mechanical Engineers, New York,1983.
O BEAVER VALLEY. UNIT 2 B 3/4 4-13
. - - _ _ _ _ _ _ _ . _ . _ . _ _ _ __. . ~ __ _ . _ _ _ _ . _ .
3/4.4 REACTOR COOLANT SYSrEM i 3/4.4.9 PRESSURE /T EMPERATURE LIMIT S 1 O BASES 1 -- At any time during the heatup or cooldown transient, KIR is determined I by the metal temperature at the tip of the postulated flaw, the appropriate I. value for RIMr , and the reference fracture toughness curve. The thermal i stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors , K It, for the reference flaw are computed. From equation (4-2), the pressure stress intensity factors are obtained and, from these, the allowable
; pressures are calculated, l For the calculation of the allowable pressure-versus-coolant i temperature during cooldown, the Code reference flaw is assumed to exist at
- the inside of the vessel wall. During cooldown, the controlling location of
- the flaw is always at the inside of the wall because the thermal gradients 4
produce tensile straisses at the inside, which increase with increasing l cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations , composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement 4 of reactor coolant temperature, vaereas the limiting pressure is actually depe ndent on the material tempersture at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adj acent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value ,
- of KIR at the 1/4 T location for finite cooldown rates than for steady };
j state operation. Furthermore, if conditions exist such that t he increase in KIR exceeds kit, the calculated allowable pressure during cooldown 4 will be greater than the steady-state value. t i The above procedures we needed because there is no direct ; I control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of .the composite' curve eliminates this problem and insures conservative operation of the system for i l the entire cooldown period. i t , I i t ! i- f l O BEAVER VALLEY UNIT 2 B 3/4 4-14 ! i k f 1 t l l
- _ . - .- ~ . . _ . . - - . - - .. --
r 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE /T EMPERATURE LIMIT S-BA SES
- Three separate calculations are required to determine the limit curves for finite heatup rates. As in done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gr adients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses rcoduced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack is lower than the KIR for the 1/4 T crack during steady-state conditions at t he s ame coolant . temper atur e .
During heatup, especially at the end of the tr ans ient , conditions may exist such that the ef fects of compressive thermal stresses and lower KIR's do not offset each other , and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that a: any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite 'heatup rates is obtained. ! The second portien of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 7 O deep outside surface flaw is assumed. Unlike the situation at the vessel inside sur f ace, . the thermal gr adients established at the outside sur f ace during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup camp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature cur'ves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by point comparison of the stesdy-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches fr om t he inside to the outside and the' pressure limit must at all times be based on analysis of the most critical criterion. lthen, composite curves for the heatup rate data and the cooldown rate data are adju'sted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective Curves. BEAVER VALLEY UNrr 2 B 3/4 4-15
, . . . , - , , , . - + -
3/4.4 REACTOR COOLANT SYST EM 3/4.4.9 PRESSURE /T EMPERATURE LIMIT S f'
\
SASES The actual established periodically shif t operation during in RINQT by of removing the vesseland material will bein evaluating, accordance with 10 CFR 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at t he irr adiation samples and vessel inside radius are essentially identical, the measured transition shif t for a sample can be. applied with confidence to the adj acent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the delta RI NDT determined from the surveillance capsule is higher than the } calculated delta RTNDT for the equivalent capsule radiation exposure. l The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticalilty and for inservice leak and hydrostatic testing. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR 50. The limitations . imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. I b 4 O BEAVER VALLEY UNIT 2 B 3/4 4-16
O O O TABLE B 3/4.4-1 REACTOR VESSEL TOUCHNESS 50 Fr/LB MNT ERIA L TNDT 35 MIL RrNDT USE COMPON ENT CODE NO. SPEC. NO. C4 % P% % T EMP % % FT/LBS. Closure Head Dome B9008-1 A533B, CL.1 .13 .013 -20 50 -10 137 Closure Head Flange B9002-1 A508 CL.7 ---
.012 -10 <40 -10 1 36 Vessel Flange 59001-1 A508 CL.7 ---
.010 0 <10 0 132.5 Inlet Nozzle B9011 -1 A508 CL.7 ---
.006 0 <10 0 10 4 Inlet Nozzle B9011-2 A508 CL.7 ---
.010 10 <10 10 115 Inlet Nozzle B9011-3 A508 CL.7 ---
.009 20 <40 20 122 Outlet Nozzle B9012-1 A508 CL.7 ---
.007 -10 0 -10 137 Outlet Nozzle 59012-2 A508 CL.7 ---
.006 -10 0 -10 121 Outlet Nozzle B9012-3 A508 CL.7 ---
.008 -10 0 -10 112 Nozzle Shell B9003-1 A5338, CL.1 .13 .008 -10 110 50 91 Nozzle Shell B9003-2 A533B, CL.1 .12 .009 0 120 60 79.5 Nozzle Shell B9003-3 A533B, CL.I .13 .008 -10 110 50 97.5 Inter. Shell B9004-1 A5335, CL.I .07 .010 0 120 60 83 Inter. Shel l B9004-2 A5335, CL.I .07 .007 -10 100 40 75.5 Lower Shell B9005-1 A533B, CL.1 .08 .009 -50 88 28 82 Lower Shell B9005-2 A533B, CL.1 .07 .009 -40 93 33 77.5 Bottom Head Torus B9010-1 A533B, CL.1 .15 .007 -30 56 -4 97 Bot tom He ad Dome 59009-1 A533B, CL.1 .14 .007 -30 35 -25 116 Weld (Inter , & Lower She'll Long. Se ams & G ir t h Se am) * .08 .008 -30 <30 -30 144.5 HAZ (Plate B9004-2) --- ----
-80 40 -20 76
- Same heat of wire and lot of flux used in all seams including surveillance weldment.
BEAVER VALLEY UNIT 2 B 3/4 4-17
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FIGURE B 3/4.4-1 FA ST N EUT RON FLU ENCE (E > 1 Mev AS A FUNCTION OF FULL POWER SERVICE LIFE) O 1020
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I O 1017 1016 0 5 10 15 20 25 30 EFFECTIVE FULL POWER (YEARS) O BEAVER VALLEY UNIT 2 B 3/4.4-18 i l l -- - _._- - . _ _ - _ . _
i 1 i-j FICURE B 3/4.4-2 4 1 EFFECT OF FLUENCE AND COPPER ON 911Fr OFART FOR REACTOR NITT VESSEL STEELS EXPOSED TO IRRADIATION AT 550*F 900 uuusimimlit!ilaUHlinillitsimi!jin # .ttl E L ll [j: g ig ig ii: . :!p .;
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1 1' Fast Neutron Fluence (n/cm2, E> I Mev) BEAVER VALLEY UNIT 2 B 3/4 4-19 l i 4 _ . . _ _ _ _ . _ _ . - . _ _ _ , . . _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . . , _ _ _ . _ . . , _ _ _ . . . _ _ - -
3/4.4 REACTOR C001 ANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY
- O BA SES The inservice inspeetion and testing progr ams for ASME Code Class I, 2, and 3 components ensure that the structur al integrity and operational readiness of these components will be maint ained at an accept able level throughout the life of the plant. These progr ans are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g)(6)(i).
i I l D 4 l 3 I , f O BEAVER VALLEY UNIT 2 B 3/4 4-20 I j 1 I
3/4.4 REACTOR C00LAffr SYSTEM 3/4.4.11 RELIEF VALVES BA SES The relief valves have remotely 'aperated block valves to provide a positive shutof f capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path. O i i
- C
- )
1 BEAVER VALLEY UNIT 2 B 3/4 4-21 i b
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS O BASES a The OPERABILITY of each of the RCS accumulators ensures that sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into 'the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure, ensure that the assumptions used for accumulator injection in the accident analysis are met. The limit of one hour for operation with an inoperable accumulator minimizes the time exposure of the plant to a LOCA event occurring concurrent with f ailure of an additional accumulator which may result in unacceptable peak cladding temperatures. O 1 O BEAVER VALLEY UNIT 2 B 3/4 5-1
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS O BASES The OPERABILITY of two separate and independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying suf ficient. core cooling to limit the the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the large=t RCS cold leg pipe downward. In addition, each'ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the accident analyses are met and that subsystem OPERABILITY is maintained. O O BEAVER VALLEY UNIT 2 B 3/4 5-2
3/4.6 CONTAINMENT SY ST EMS 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY l l'J'] BA SES Primary CONTAINMENT INT EGRITY ensures that the release of' radioact-ive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction. with the leakage rate limit at io n, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions. 'I O I J o l f i i i l i BEAVER VALLEY UNIT 2 B 3/4 6-1 ' r l'
3/4.6 CONTAINMENT SY ST EMS i f 3/4.6.1 PRIMARY COKTAINMENT (k 3/4.6.1.2 CONTAINMENT LEAKAGE BASES i r The limitations on containment -leakage rates ensure that t he . total containment leakage volume will not exceed the value assumed in the accident analysis at the peak accident pressure, Pa. As an added : < conservatism, . the measured overall integrated leakage rate is fur t her limited ! to less than or equal to 0.75 La during performance of the periodic test to ! account for possible degradation of the containment leakage barriers between [ 1eakage tests. ' The surveillance testing for measuring leakage rates ar e I consistent with the requirements of Appendix J of 10 CFR 50. t The exemption to 10 CFR 50 Appendix J, III.D.1 (a) allows Type A tests to be conducted on a 40 + 10 month schedule, not in conjunction with
~
I[. [ any ISI tests. t t t f i 4 O BEAVER VALLEY UNIT 2 8 3/4 6-2 4 P P P g . m.. .. - - _ _ , _ . . - . . . - _ . , , . . , .,.,,,,, .,_ ._ ,,,--,,._c.,, . _ , - . , , . . .
1 3/4.6 CONTAINMENT SYST Df S 3/4.6.1 PRIMARY CONTAINMEhT 3/4.6.1.3 CONTAINMENT AIR LOCKS BASES The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and cont ainment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. O I ( i [ I t I i
- I O BEAVER VALLEY UNIT 2 B 3/4 6-3 r I
f r 4 l
3/4.6 CONTAINMDIT SYST EMS 3/4.6.1 PRIMARY CONTAINMDIT 3/4.6.1.4 & 3/4.6.1.5 INT ERNAL PRESSURE / AIR TEMPERATURE 1 BA SES The limitations on containment internal pressure and average air temperature as a function of RWST and service water temperature ensure that:
- 1) The containment structure is prevented from exceeding its design negative pressure of 8 psia,
- 2) The containment peak pressure does not exceed the design pressure of 45 psig during IDCA conditions, and
- 3) The containment' pressure is returned to subatmospheric conditions
- e. following a LOCA.
The containment internal pressure limits shown as a function of RWST- and service water temperature describe the operational envelope that will:
- 1) Limit the cont ainment peak pressure to less than its design value of 45 psig, and
- 2) Ensure the containment internal pressure returns subatmospheric within 60 minutes following a LOCA.
The limits on the parameters of Figure 3.6-1 are consistent with the assumptions of the accident analyses. l p I t O BEAVER VALLEY UNIT 2 B 3/4 6-4 : i t i f
3/4.6 CONTAINMDIT SYST EMS 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.6 CONTAINMDIT STRUCIURAL INTEURITY 1 BASES i ' This limitation ensures that the structural integrity of the containment vessel will be maintained comparable to the original design standards for the life of the facility. . Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 44.6 psig in a the event of a LOCA. The visual and Type A leakage tests are sufficient to j demonstrate this capability. ,
- j 7
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O BEAVER VALLEY UNIT 2 B 3/4 6-5 i l l l
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i 3/4.6 CONTAINMENT SYST EMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYST EM 3/4.6.2.1 CONTAINMENT QUDICH SPRAY SYSTEM 3/4.6.2.2 CONTAINMENT RECIRCULATION SYSTEM I BA SES t The OPERABILITY of the containment spray systems ensures that containment depressurization and subsequent return to subatmospheric pressure will occur in the event of a LOCA. The pressure reduction and resultant termination of containment leakage are consistent with the assumptions used in the accident analysis. I i 4 j i e t E r I
- O BEAVER VALLEY UNIT 2 B 3/4 6-6
3/4.6 CONTAINMENT SY ST EMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS () 3/4.6.2.3 CHEMICAL ADDITION SY FT EM BA SE S The OPERABILITY of the chemical addition system ensures that suf ficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH minimum volume and concentration ensure that:
- 1. The iodine removal efficiency of the spray water is maintained because of the increase in pH value, and
- 2. Corrosion ef fects on components within containment are minimized.
T he se as sumpt ions are consistent with the iodine removal ef ficiency assumed in the accident analyses. C:) t o l i I i BEAVER VALLEY UNIT 2 B 3/4 6-7
, 3/4.6 CONTAINMENT SYSrEMS 3/4.6.3 CONTAINMDIT ISOLATION VALVES O BA SES The OPERABILITY of the cont airunent isolation valves ensur es that the contairunent atmosphere will be isolated from the outside envirorument in the event of a release of radioactive material to the contairunent atmosphere or pressur ization of the contairunent. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the 'assumpt ions used in the analyses i of a lhCA. C
- O 3
l
,t I
h I i f e t i O BEAVER VALLEY UNIT 2 8 3/4 6-8 j i I .a i y-3,r-,, -
, y-v.---. - ., _ . _ _ _ - _ _ , . , . . , _ _ - _ _ . , _ , , , , , , , _ , . , _ . , , , , , _ _ . _ . , ,
3/4.6 CONTAINMENT SY ST EMS 3/4.6.4 COMBUSTIBLE GAS CONTROL BA SES ;
)
i The OPERABILITY of the equipment and systems required for the , detection and control of hydrogen gas ensures that this equipment will be l available to maintain the hydrogen concentr ation within cont airunent below its ! flammable limit dur ing post-LOCA condit ions. Either recombiner unit or the : purge system is capable of controlling the expected hydrogen generation 3 { associated with:
- 1. Zirconium-water reactions, ;
i
- 2. Radiolytic decomposition of water ,
4
- 3. Corrosion of metals within containment. t These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrat ions in Cont ainment Following a LOCA."
1 O l i i t i b f l t 4 O . BEAVER VALLEY UNIT 2 B 3/4 6-9 i ! i r y , ,.,-,y,----.--. .-- -.,v. -- . _ , , , - - , - - ~ - - - ,,,a- ---,-rwy4-e . --- _ . - - - er
i 3/4.6 CONTAINMENT SYST EMS 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.5.1 STEAM JEr AIR EJECTOR I BASES __ _ - - ~_ _ The closure of the manual isolation valves in the suction of the stean jet air ejector ensures that:
- 1. the containment . internal pressure may be maintained within its operation limits by the mechanical vacuum pumps, and
- 2. the containment atmosphere is isolated from the outside envirorument in the event of a 1DCA.
These valves are required to be closed for containment isolation . f f P l O ! i f r 4 4 i f I i , i r \ ? BEAVER VALLEY UNIT 2 B 3/4 6-10 ; l
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i t
- - . - . . , , - - . - , .em.,.,,,___,,,,m.... . _- ... r ...,.._,--,,__m%--.,.r. m., , .,_.,m.r,xym
3/4.7 TURBINE CYCLE 3/4.7.1 TURBINE CYCLE () 3/4.7.1.1 SAFETY VALVES BASES
- - - = . - - - - _ -.-
The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100 percent RAT ED THERMAL POWER coincident with an assumed loss of condenser heat s ink ( i .e . , no s t e am byp as s to the condenser). The specified valve .lif t settings and relieving capacities ar e in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition and Winter 1972 Addenda. The total relieving capacity for all valves on all of the steam lines is 110 per cent of the total secondary steam flow at 100 percent RAT ED THERMAL POWER A minimum of 2 OPERABLE safety valves per operable steam generator ensures that su f ficient relieving capacity is available for the allowable THERMAL POWER restriction in T able 3.7-2. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of ' the reduction in secondary system steam flow and. THERMAL POWER required by , I'} 5' the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following bases: For. 3 loop operation: i (X) - (Y)(V) SP = x (109) X l For 2 loop operation: (X) - (Y)(U)
- SP = x (70) ,
X l Wher e: SP = Reduced reactor trip setpoint in percent of RAT ED THERMAL POWER V = maximum number of inoperable safety valves per steam line U = maximum number of inoperable safety valves per operating steam line L k'_')T BEAVER VALLEY UN TI 2 B 3/4 7-1 i
- i i
3/4.7 TURBINE CYCLE
'3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES 1 BASES (109) = Power Range Neutron Flux-High Trip Setpoint for 3 loop operation (70) = maximum percent of RATFD THERMAL POWER permissible by P. 8 -
setpoint 'for 2 loop operation with stop valves closed X = Total relieving capacity of all safety valves per steam line in pounds / hour Y = Maximum relieving capacity of one safety valve in pounds / hour 1 0 i i I 1 l I i I O BEAVER VALLEY UNIT 2 B 3/4 7-2 l P
3/4.7 PIANT SY ST EMS 3/4.7.1 TURBINE CYCLE F 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM 1 V) BASES
. The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of of f-site power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 350 gpm at a pressure of 1133 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is. capable of delivering a total feedwater flow of 700 gpm at a pressure of 1133 psig to the entrance of the steam generators. This . capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removsl System may be placed into operation.
- O O BEAVER VALLEY LHIT 2 B 3/4 7-3
3/4.7 PLANT SY ST EM S 3/4.7.1 TURBINE CYCLE 3/4.7.1.3 PRIMARY PLANT DEMINERALIZED WAT ER (PPDW) ) BASES The OPERABILITY of the PPDW storage tank with the minimum water volume ensures that suf ficient water is available for cooldown of the Reactor Coolant System to less than 350*F in the event of a total loss of off-site power. The minimum water volume is suf ficient to maintain the RCS at HOT STANDBY conditions for 9 hours with steam discharge to atmosphere. O I 4 i O' BEAVER VALLEY INIT 2 B 3/4 7-4 l l i
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)
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE l O 3/4.7.1.4 ACTIVITY BASES 1 1 The limitations on the secondary coolant system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the ef fects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the af fected steam line. These values are consistent with the assumptions used in the accident
- analyses.
O 4 O BEAVER VALLEY UNIT 2 B 3/4 7-5
3/4.7 PLANT SY ST EMS 3/4.7.1 TURBINE CYCLE O 3/4.7.1.5 MAIN ST EAM LINE ISOLATION VALVES l} v BASES
= = - _ _ _ _ - - _ ._ --
The OPERABILITY of the main steam line isolat ion valves ensures that no more than one stema generator will blowdown in the event of a steam line rupture. This restriction is required to:
- 1. Minimize the positive reactivity ef fects of the Reactor Coolant Sys tem cooldown associated with the blowdown, and
- 2. Limit the pressure rise within the containment in the event the stean line rupture occurs within the contairusent.
The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. i lO i i 1 l l I i 1 i i O BEAVER VALLEY UNIT 2 8 3/4 7-6 l
,, . . , - . _ . _ . _ - _ _ - , . , _ _ . _ . _ , . _ . , _._._...----_-,.,1 - . . , _ _ , - _ , . . . - . . , -
3/4.7 TURBINE CYCLE 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 0 BA SES The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the stean generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on stean ' generator average impact values taken at 60*F and are suf ficient to prevent brittle fr acture. 4 i j O i 'I
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4 l J J i, I j O BEAVER VALLEY UNIT 2 B 3/4 7-7 l t
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3/4.7 PIANT SY ST EMS 3/4.7.3 PRIMARY COMPONDrr COOLING WATER SYSTEM
, BA SES The OPERABILITY of the primary component cooling water system ensures that suf ficient cooling capacity is available for continued operation of safety related equipment during normal and accident cond it io ns . T he r edundant cooling capacity of this system, assuming a single f a ilur e , is cons is te nt with the assumptions used in the accident analys is .
O 1 l i 4 .t BEAVER VALLEY UNIT 2 B 3/4 7-8 1
3/4.7 PLANT SY ST EM S 3/4.7.4 SERVICE WATER SYSTEMS BASES The OPERABILITY of the service water system ensures that suf ficient cooling capacity is available for continued oper ation of safety related equipment dur ing normal and accident conditions . The redundant cooling capacity of this system, assuming a single failur e , is consistent with the assumptions used in the accident condit ions .
, O t
O BEAVER VALLEY UNIT 2 B 3/4 7-9
3/4.7 PIANT SY SI LMS i 3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER BASES I I The limitations on the ultimate heat sink level and temper ature ensure that sufficient cooling capacity is available to either:
- 1) Provide normal cooldown of the facility, or
- 2) To mitigate the ef fects or . accident conditions within acceptable limits.
The limitations on minimum water level and maximum temperature are based on providing a 30 day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuc le ar Plant s ." O e O BEAVER VALLEY UNIT 2 B 3/4 7-10
3/4.7 PLANT SY ST EMS 3/4.7.6 FLOOD PROTECTION O V BA SES The limitation on flood level ensures that facility operation will be terminated in the event of flood conditions. The limit of elevation 695 feet Mean Sea Level was selected on an arbitrary basis as an appropriate flood . level at which to terminate further operation and initiate flood protection measures for safety related equipment , i f O I' 4 1 J l h O BEAVER VALLEY UNIT 2 B 3/4 7-11
3/4.7 PIMT SY ST EMS 3/4.7.7 CONTROL ROOM EMERGDICY HABITABILITY SYSTEMS BASES The OPEPABILITY of the control room ventilation system ensures that:
- 1. The ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentat ion cooled by this system, and
- 2. The control room will remain habitable for operations personnel during and following all credible accident operations.
The OPERABILITY of this system in conjunction with control room design provisions, is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. O O BEAVER VALLEY UNIT 2 B 3/4 7-12
3/4.7 PLANT SY ST EMS 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM ( SLCRS) BA SES The OPERABILITY of the SLCRS provides for the filtering of
; postulated radioactive ef fluents resulting from a Fuel Handling Accident (FRA) and from leakage of LOSS OF CDOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containement building, such as Engineered Safeguards Features (ESP) equipment, prior to their release to the e nvironme nt. This system also collects potential leakage of LOCA activity from the Reactor Containment building penetrations into the contiguous areas ventilated by the SLCRS. The operation of this system was assumed in calculating the postulated offsite doses in the analysis for a FHA. System- 1 ; operation was also assumed in that portion of the Design Basis Accident (DBA) '
IDCA analysis which addressed ESF leakage following the LOCA, however, no credit for SLCRS operation was taken in the DBA LOCA analysis for collection
! and filtration of Reactor Containment building leakage even though an unquantifiable maount of contiguous area penetration leakage would in fact be collected and filtered. Based on the results of the analyses , the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA and leakage' resulting from a FHA will no;. exceed 10 CFR 100 limits.
O 1 i j t 1 1 1 i BEAVER VALLEY UNIT 2 B 3/4 7-13
3/4.7 PLANT SY ST EMS 3/4.7.9 SEAL 8I) SOURCE C0!frAM1 NATION BA SES The limitations on sealed source removable contasination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source j material. The limitations on removable contamination for sources requiring ! le ak tes t ing , including alpha emitters, is based on 10 CFR 70.39(c) limits i for plutonium. Ieakage of sources excluded ft on the requirements of this i I specification represent less than one maximum permissible booy burden for total body irradiation if the source material is inhaled or ingested. J i J I I ! I
- O i
T l i i 1 4 , j { O 8FAVER VAlt.EY UNIT 2 8 3/4 7-14 l' l I i
l L ' 3/4.7 PLANT SYSTEMS 3/4.7.12 !MUBBERS ' I BA SES i 1 1 j All snubbers are required OPERABLE to ensure that the I structural integrity of the reactor coolant system and all other safety- l j related systems is maintained during and following a seismic or ocher similar I ! event initiating dynamic loads. Snubbers excluded from this inspection , j program are those installed on nonsafety-related systems and then only if [ j t heir failure or failure of the system on which they are installed, would -i have no adverse of feet on any safety-related system. The visual inspection frequency is based upon maintaining a r
- cons t ant level of snubber protection to systems. Therefore, the required i
! inspection interval varies inversely with the observed snubber failures and i l is determined by the number of inoperable snubbers found during an [ i ins pe ct io n. Inspections performed before that interval has elapsed, may be j
! used as a new reference point to determine the next ins pect ion. ;
When the cause of the rejection of a snubber is clearly I established and remedied for that snubber and for any other snubbers that may i be generically susceptible, and verified by inservice functional testing, p j that snubber may be exempted from being counted as inoper able. Generically l l susceptible snubbers we those which we of a specific make or model and have j the same design features directly related to rejection of the enubber by j visual inspection, or are similarly located or exposed to the same ; enviromnental conditions such as temper ature, r adiation and vibration. ! i j When a snubber is found inoperable, an engineering evaluation l is performed, in addition to the determination of the enubber mode of j failure, in order to determine if any safety-related component or system has
- been adversely affected by the inoperability of the anubber. The engineering
! evaluation shall determine whether or not the enubber mode of failure has i imparted a eignif(cant ef fact or dearadat(on on the supported component or .' eystem. i I To provide assurance of snubber functional reliability, a i representative sample of the installed enubbers will be functionally tested during plant ehtdowns at refueling or 18 month intervals not to exceed two } ye ar s . Observed failures of these emeple snubbers shall require functional ! testing of additional units. j- Hydr mu1ic enubbers and mechanical snubbers may each be treated i j as a dif ferent entity for the above surveillance progress. 1 i , i O ! BEAVER VALLEY UNfr 2 B 3/4 7-15 1 i
3/4.7 PLANT SYSTEMS 1 j 3/4.7.12 SNUBBERS BA SES The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber , seal replaced, spring replaced, in high radiation area, in high temper ature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in i view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant oper at ion. O i l () BEAVER VALLEY UNIT 2 5 3/4 7-16
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3/4.7 PLANT SY ST EMS
- 3/4.7.13 STANDBY SERVICE WATER SYSTEM ( SSWS)
BA S ES ' The OPERABILITY of the SSWS ensures that su f ficient cooling , capacity is available to bring the reactor to a cold shutdown condition in the event that a barge explosion at the stat ion's int ake structure or any other extremely remote event would render all of the normal Service Water System supply pumps inoper able . 1 i l l
- O i
2 i a lO BEAVER VALLEY UNIT 2 B 3/4 7-17
-- . . _ - . _ - _ _ . . _ . _ . - . . . _ _ . . . , . . . _ _ . _ . . _ , . . _ . _ . , . . . . ~ . , . _ . . - _ . . _ _ . _ , . _ . . - , _ . , , ,
3/4.7 PLANT SY ST EM S i 3/4.7.14 FIRE SUPPRESSION SYSTEMS 4 BASES The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment i is located. The fire suppression systems consist of the water system, , sprinklers, CO2 Halon and' fire hose stations . ' The collective capability of the fire suppression systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program. In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the af fected areas until the inoperable equipment is restored to i service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. l1 The surveillance requirements provide assurance that the minimum OPERABILITY requirements of the fire suppression systems are ' met. An allowance is made for ensuring a suf ficient volume of Halon in' the Halon storage tanks by verifying the weight of the tanks. The Halon systems ar e indoce systems not susceptible to outdoor weather conditions. The systems are dry pipe (rust is not expected) gas suppression systems. In the event the fire suppression water system becomes inoper ab le, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. The requiremer.t for a 24 hour report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant. i O BEAVER VALLEY UNIT 2 B 3/4 7-18
3/4.7 PLANT SY ST EMS . 3/4.7.15 FIRE RAT ED A SSEMBLIES BASES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. The design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier pene tr at ions for conduits, cable tr ays , and piping, fire windows , fir e damper s , and fir e doors are periodically inspected to verify their operability, i
- O i
i a lO l BEAVER VALLEY UNIT 2 B 3/4 7-19
3/4.7 PLANT SY ST EM S 3/4.7.16 TERRESTRIAL ECOLOGICAL SURVEY BA SES The objective of the Terrestrial Ecological Survey is to determine the effects of cooling tower drift on terrestrial biota. The Terres tr ial Ecological Survey progr am element involves long term monitoring. Cooling tower drif t could lead to ecological ef fects that would apper as vegetation stresses on infr ar ed aerial photogr aphs. I l O I I O BEAVER VALLEY UNIT 2 B 3/4 7-20 l 4- --- - - - - - _ -
, ,_ . . _ _ , _ . _ _ _ _ _ . , , , _ , _ _ _ , _ _ - , . _ _ _ _ , _ _ . . _ _ . , . . , . . _ - _ _ , , _ , _ . , _ , , , . , y 7,m, , , , . , _ _ - . _
l 3/4.8 ELECTRICAL POWER SYSTEMS i 3/4.8.1, 3/4.8.2 A.C. SOURCES, D .C. SOURCES AND ONSIT E POWER DI ST RIBUT ION SY ST EMS () BASES r The OPERABILITY of the A.C. and D.C. power sour ces and associated distribution systems during operation ensures that sufficient
; power will be available to supply the safety related equipment required for:
) 1. The safe shutdown of the facility, and t
- 2. The mitigation and control of accident conditions within the facility.
j The _ minimum specified independent and redundant A.C. and D.C. power sour ce s l and distribution systems satisfy the requirements of General Design { Criterion 17 of Appendix "A" to 10 CFR 50. i
- The ACTION requirements specified for the levels of
! degradation of the power sources provide restriction upon continued f acility ;
operation commensurate with the level of . degradation. The OPERABILITY of tim
! power sources are consistent with the initial condition assumptions of the !
l safety analyses and are based upon maintaining at le ast one redundant set of onsite A.C. 'and D.C. power sources and associated distribution systems i 1 OPERABLE during accident conditions coincident with an assumed loss of l offsite power and single failure of the other onsite A.C. source. l i The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling . ensures that ; l 1. The facility can be maintained in the shutdown or refueling condition for extended time periods, and
- 2. Sufficient. instrumentation and control capability is available for
- monitoring and maintaining the unit status.
! The Surveillance Requirement for demonstr ating the OPERABILITY l of the Station batteries are based on the recommendations of Regulatory Cuide I
1.129, " Maintenance Testing and Replacement of Large Lead Storage Batteries i for Nuclear Power Plants," February 1978, and IEEE Standard 450-1980, "IEEE ! Recommended Pr actice for maintenance , Teeting, and Replacement of Large Lead ' l Storage Batteries for Gener ating Stations and Substations." i Verifying average electrolyte temperature above the minimum i for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the ef fectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at j that time with the r ated capacity. ?,
- (1)
! BEAVER VALLEY UNIT 2 B 3/4 8-1 i i i i l
3/4.8 ELECTRICAL POWER SYSTEMS A.C. SOURCES, D .C. SOURCES AND ON SIT E POWER , DI ST RIBLTT ION SY ST EMS BA SES Table 3.8-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gr avity. The limits for the des ignat.ed pilot cells float voltage and specific gravity, greater than 2.13 volts and .015 below the manuf acturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is char acteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than .G2L below the manuf acturer's full charge specific gr avity with an average specific gr avity of all the connected cells not more than .010 below the manuf acturer's full charge specific gravity, ensures the OPERABILITY and capability of the bat ter y. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 3.8-1 is permitted for up to 7 days. During this 7 day period:
- 1. The allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability;
- 2. The allowable value for the average specific gravity of all the cells, tot more than .020 below the manuf acturer's recoassended full charge
}
specific gravity, ensures that the decrease in rating will be les s than the safety margin provided in sizing; j 3. The allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than .040 below the manuf acturer's full charge specific gr avity and that the over all capability of the battery will be maintained within an acceptable limit; and
! 4. The allowable value for an individual cell's float voltage, greater j than 2.07 volts, ensures 'the battery's capability to per form its design function. ,
l l l i BEAVER VALLEY UNIT 2 B 3/4 8-2 l _-________-_--_______--_____-____-____---____-__-_______--_________-______\
3/4.9 REFUELING OPERATION S 3/4.9.1 BORON CONCDTTRATION O BA SES The limitations on minimum boron concentration (2000 ppm) ensure that:
- 1. The reactor will remain suberitical during CORE ALT ERATIONS, and
- 2. A uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
The limitation on K,gg of no greater than 0.95 which includes a conservative allowance for uncertanties , is suf ficient to prevent reactor criticality during refueling operations. J l 4 O I I i O BEAVER VALLEY UNIT 2 B 3/4 9-1
_ _ - -. - . ~ . _ .- _ - _ _ _ _ _ _ .- -. 3/4.9 REFUELING OPERATION S
} 3/4.9.2 IN STRUMDITATION
- O BA SES 1
The OPERABILITY of the source range neutron flux monitors i ensures that redundant monitoring capability is available to detect changes l in the reactivity condition of the core. l i l .] 1 O I i 1 i I l t t i i ,. j O BEAVER VALLEY UNIT 2 5 3/4 9-2 a
3/4.9 REFUELING OPERATIONS 3/4.9.3 DECAY TIME BA SES The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that suf ficient time.has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the .ac cide nt analyses. O e O BEAVER VALLEY UNIT 2 5 3/4 9-3
3/4.9 REFUELING OPEaAT10N b
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O # Es Tne requirements en containment penetrat ton closure and OPEkABILITY of the containsent.: parge and exhaust system HEPA tilters and l charcoal ~4 soroers ensure that a' release at racioactive matertal witnan containment will be restricted from Leakage to'the environment or tiltereo I Lnrough the HEPA tilters and charcoal adsorpers prior to disenarge to the atmosphere within 10CFR100 Limits. The OPERABILITY and closure restrictions are surtteient to restrictdadfoactive material release trom a tuel element ruptyke based upon the lack'of containment pressurization potential while in one RiJUELING MODE. N ,
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j 3/4.9 REFUELING OPERATION S 3/4.9.5 COMMUNICATIONS BA SES 2 The requirements for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALT ERAT ION S. f
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I h l { i i O 2 BEAVER VALLEY UNIT 2 B 3/4 9-5 l i l
3/4.9 REFUELING OPERATION S 3/4.9.6 MANIPULATOR CRANE OPERABILITY BA SES The OPERABILITY requirements for the manipulator crane and auxiliary hoist ensure that:
- 1. Manipulator crane and auxiliary hoist will be used for the movement of fuel assemblies and drive rod assemblies,
- 2. T he cr ane has suf ficient load capacity to lift a fuel assembly and the hoist has suf ficient capacity to lift a drive rod assembly,
- 3. The core internals and pressure vessel are protected from excessive lifting force in the event they we engaged during lifting oper at io ns .
O f i V BEAVER VALLEY UNIT 2 B 3/4 9-6 .
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3/4.9 REFUELING OPERATION S 3/4.9.7 CRANE TRAVEL _- SPEKr FUEL STORAGE POOL BUILDING BA SES The restriction on movement of loads in excess of the normal weight of a fuel assembly over other fuel assemblies ensures that no mor e than the contents of one fuel ' assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident analys is . O 4 i V BEAVER VALLEY UNIT 2 B 3/4 9-7
3/4.9 REFUELING OPERATIONS 3/4.9.8 RESIDUAL H_EA_T REMOVAL AND COOLANT CIRCUIATION (D BA SES The requirement that at least one residual heat removal (RHR) loop be in operation ensures that:
- 1. Suf ficient cooling capacity is available to remove decay heat and maintain the. water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and
- 2. Sufficient coolant circulation is maint'ained throughout the reactor core to minimize the ef fect of a boron dilution incident and prevent boron str atification.
The requirement to have two (RHR) loops OPERAB12 when there is less than 23 feet of water above the reactor pressure vessel flange ensures-that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core. O O BEAVER VALLEY UNIT ' 2 B 3/4 9-8
3/4.9 REFU ELING OPERATION S 3/4.9.9 CONTAINMDrr PURGE AND EXHAUST ISOLATION SYST EM O >^ses The OPERABILITY of this system insures that the cont ainnent vent and purge penetrations will be automatically isolated upon detection of high-high radiation levels witnin the containment . The integrity of the containment penetrations of this system is required to restrict t he r e le ase of r adioactive material from the cont ainment atmosphere to' accessible levels which a e less .than those listed in 10 CFR 100. Applicability in MODE 5, although not an NRC requirement, will provide additional protection against small releases of radioactive material from the containment during maintenance acti rities. 4 O
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1 O BEAVER VALLEY UNIT 2 - B 3/4 9-9
- - - - - - - _ _ . . _ . _ . _ _ - , _ . ~. . . . _ _ - _ - , .
3/4.9 REFUELING OPERATION S 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL BA SE S The restrictions on minimum water level ensure that su f fic ie nt water depth is available to remove 99 percent of tha assumed 10 percent iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analys is . 9 O BEAVER VALLEY UNIT 2 B 3/4 9-10
3/4.9 REFUELING OPERATION S 3/4.9.12 FUEL BUILDING VENTILATION SYST EM - FUEL MOVEMENT (,) BA SES , s The limitations on the storage pool ventilation system ensures that all r adioactive material released from an irr adiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are . consistent with the assumptions of the accident analysis. O i i l O BEAVER VALLEY UNIT 2 B 3/4 9-11 I
3/4.9 REFUELING OPERATION S f 3/4.9.13 FUEL BUILDING VB'TILATION SYST EM - FUEL STORAGE BA SES The spent fuel pool area ventilation system is non-safety related and only recirculates air throughout the fuel building. The SLCRS portion of the ventilation filters the fuel building air upon receipt of a high-high r adiation signal. O Y 4 BEAVER VALLEY UNIT 2 B 3/4 9-12 l
3/4.10 SPECIAL TEST EXCEPTION S 3/4.10.1 91UTDOWN MARGIN S O
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BA SES 4 This special test exception provides that a minimum arnount of control rod worth is immediately available for reactivity control when tests are performed fer control rod worth measurement. This special test except io n is required to permit the periodic verification of the actual rersus predicted : ore reactivity condition occuring as a result of fuel burnup or fuel cycling operations. O i M O BEAVER VALLEY UNIT 2 B 3/4 10-1
3/4.10 SPECIAL T EST EXCEPTION S 3/4.10.2 GROUP HEIGHT , IN SERTION AND POWER DISTRIBUTION LIMIT S BA SES This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TEST S as those required to:
- 1. Measure control rod worth, and
- 2. Determine clie reactor stability index and damping factor under xenon oscillation conditions.
O O BEAVER VALLEY UNIT 2 B 3/4 10-2 4 4 -- -,
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3/4.10 SPECIAL TEST EXCEPTION S 3/4.10.3 PRESSURE / TEMPERATURE LIMITATION S - REACTOR CRITICALITY O BASES l This special test exception permits the reactor to be critical at less than or equal to 5 percent of RAT ED THERMAL POWER during low temperature. PHYSICS TEST S required to measure such parameters as control rod worth and SHUTDOWN MARGIN. O O BEAVER VALLEY UNIT 2 B 3/4 10-3 i
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3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.4 PHYSICS TESTS C) BASES I 4 This special test exception permits the physics tests to oe , performed at less than or equal to 5 percent or RATED THERMAL POWER and is required to verify the fundamental nuclear characteristics ot the reactor l core and related instrumentation. 4 i 4 4 O O l BEAVER VALLEY UFIT 2 - B 3/4 10-4
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.5 NO FLOW TEST S O BA SES This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and physics tests while at low THERMAL POWER levels. i 1 1 4 BEAVER VALLEY UNIT 2 B 3/4 10-5
3/4.10 SPECIAL TEST EXCEPTION S 3/4.10.6 POSITION INDICATION SYSTEM - 91UTDOWN O BA SES This special test exception permits the position indication systems to be inoper able during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time is derived from the induced voltage in the position indicator coils as the rod is dropped. This induced . volt age is small compared to the normal voltage and, therefore, cannot be observed if the position indication systems remain OP ERABLE. i i i i O l l l l h f ( i O BEAVER VALLEY. UNIT 2 B 3/4 10-6 T L I
3/4.11 RADIOACTIVE EFFLUDIT S _3/4.11.1 LIQUID EFFLUENT S 3/4.11.1.1 CONCENT RAT ION g BASES This Specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluent to UNRESTRICT3D AREAS will be less than the concentration levels specified in 10 CFR 20 Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within:
- 1. The Section II A design objectives of Appendix I,10 CFR 50 to a individual, and
- 2. The limits of 10 CFR 20.106e to the populat io n.
Th'e concentration limit for dissolved or entrained noble gases is based upon
~
the assumption that Xel35 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent . concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. f M e BEAVER VALLEY UNIT 2 B 3/4 11-1 l
l 3/4.11 RADIOACTIVE EFFLUDIT S 3/4.11.1 LIQUID EFFLUDIT S 3/4.11.1.2 DOSE 1 V BA SE S This specification is provided to implement the requirement s of Sections II.A. III.A. and IV.A of Appendix I to 10 CFR 50. The Limiting Condition for Operation implements the guides set forth in Section II.A cf Appendix I. The Action statements provides the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in liquid ef fluents will be kept "as low as reasonably achievable." Also, for fres h water sites with drinking water supplies that can be potentially af fected by plant operation, there is reasonable assurance that the operation of the f acility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the dose due to the actual release rates of radioactive materials in liquid ef fluent are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of g, Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, " Revision 1, October,1977, and Regulatory Guide 1.113 "Estimatin p Aquatic Dispersion of Ef fluents from Accidental and Routine Reactor Releases v for the Purpose of Implementing Appendix I", Revision 1, April,1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113. This Specification applies to the release of liquid ef fluents from Beaver Valley Power Station, Unit No. 2. For units with shared radwaste treatment systems, the liquid ef fluents from the shared system are proportioned among the units sharing that system. N BEAVER VALLEY UNIT 2 B 3/4 11-2
3/4.11 RADIOACTIVE EFFLUDrr s 3/4.11.1 LIQUID EFFLUDrr S - O 3/4.11.1.3 LIQUID WASTE TREATMENT I' BASES _ . - ~ ~ . . .: The . requirement s that are the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid ef fluents will be kept "as low as is reasonably achievable." This Specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 ami design objective given in Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the liquid radwsste tre atme nt system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendiox I,10 CFR 50 , for liquid ef fluents. This Specification applies to Beaver Valley Power Station - Unit 2. O O BEAVER VALLEY UNIT 2 B 3/4 11-3
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.1 LIQUID EFFLUENT S 3/4.11.1.4 LIQUID HOLDUP TANKS g-BASES j _=- _ -- 3 Restricting the quantity of radioactive material contained in the specified tank provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be les s l than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted i area. 4 1 I 1 4 O v 4 i I i O BEAVER VALLEY UNIT 2 B 3/4 11-4 f
- f
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.2 GASEOUS EFFLUENT S 3/4.11.2.1 DOSE RAT E ( ,) I BA SES _ -- -~~ -- This specification is provided to ensure that the dose at anytime at the site boundary from gaseous ef fluents from all units on the site will be within the annual dose limits of 10 CFR 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, T able II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous ef fluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II, of 10 CFR 20 (10 CFR 20.106(b). For individuals who may at times be within 'the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to < 500 mrem / year to the total body or to < 3,000 mrea/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate dbove l background to an infant via the cow-milk-infant pathway to < 1,500 arem/ year-for the neares t cow to the plant . This specification applies to the release of gaseous ef fluents O from Beaver Valley Power Station, Unit No. 2. For units with shared radwaste treatment system, the gaseous ef fluents from the shared system are i proportioned among the units sharing that system. e 6 9 I b I 0 BEAVER VALLEY UNIT 2 B 3/4 11-5 I
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i 3/4.11 RADIOACTIVE EFFLUENT S l 3/4.I1.2 GA SEOU S EFFLUENT S 3/4.11.2.2 DOSE, NOBLE GASES BASES This specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I,10 CFR 50. The Limiting Condition for Operation implements the guides set forth in Section II.B 'of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of r adioactive material ~ in gaseous ef fluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially under-estimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of r~idtoactive noble gases in gaseous ef fluents are consistent with the methodology provided in
~
Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef flue nt s in Routine Rele ase s from Light-Water-Cooleo reactors ," Revision 1, July,1977. The-0DCM equations provided for determining the air doses at the N exclusion area boundary are based upon the historical average atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111. This specification applies to the release of gaseous effluents from Beaver Valley Power Station, Unit 2.
O BEAVER VALLEY UNIT 2 B 3/4 11-6
- 3/4.11 RADIOACTIVE EFFLUDIT S 3/4.11.2 GA SEOU S EFFLU ENT S
> 3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULAT E FORM, } AND RADIONUCLIDES DIHER TRAN NOBLE GASES BASES This specification is provided to implement the requirements of Sections II.C, III.A, and IV. A of Appendix I to 10 CFR 50. The Limiting Condition for Operation are the guides set forth in Section II.C of Appendix I. The Action statements provide the required operating 4 flexibility and at the same time impleme nt the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in ! gaseous effluents will be kept "as low as is reasonably schievable." The ODCM > calculational methods specified in the surveillance requirements impleme nt the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and , data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided [ in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Rector Ef fluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision I, October 1977 and Regulatory Guide 1.111, ' 3 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef flue nts in Routine Releases from Light-Water-C6oled Reactors," Revision 1, . July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form, - and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which ; are examined in 'the development of these calculations are. I
- 1. Individual inhalation of airborne radionuclides, s
- 2. Deposition of radionuclides onto green leafy vegetation with I
subsequent consumption by man, ,
- 3. Deposition onto grassy areas where milk animals and meat producing !
animals graze with consumption of the milk and meat by man, and r
- 4. Deposition on the ground with subsequent exposure of man. This ;
specification applies to radioactive staterial in particulate form and ( radionuclides other than noble gases release'd from Beaver Valley Power j Station, Unit 2. l f i O BEAVER VALLEY UNIT 2 B 3/4 11-7 t [ i
3/4.11 RADIOACTIVE EFFLUENT S 3/4.11.2 GA SEOU S EFFLU ENT S 3/4.11.2.4 GASEOUS RADWAST E TREATMENT l* BASES
. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous ef fluente will be kept "as low as is reasonable achievable." This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and design objective Section II.D of Appendix I to '0 CFR 50. The specified limits governing the use of appropriate portione cf the systems were specified as a suitable fraction of the dose design ob;ectives set forth in Sections II.B and II.C of Appendix I,10 CFR 50, for itaseous ef fluents. This specification applies to gaseous radwaste from Beaver Valley Power Station, Unit 2.
O-O BEAVER VA' L LEY UNIT 2 B 3/4 11-8
3/4.11 RADIOACTIVE EFFLUDIT S 3/4.11.2 GASEOUS EFFLUDIT S 3/. 1.2.5 GA SE0tJ S WA ST E STORAG E TANK S U BASES Restricting the quantity of radioactivity contained in each gaseous waste storage tank provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting total body exposure to an individual located at the nearest exclusion area boundary for two hours innsediately following the onset of the release will not exceed 0.5 rem. The specified limit restricting the quantity of radioactivity contained in each gaseous waste storage tank was specified to ensure that the total body exposure resulting from the postulated release remained a suitable fraction of the reference value set forth in 10 CFR 100.11(a)(1). O V O BEAVER VALLEY UNIT 2 B 3/4 11-9
3/4.11 RADI0 ACTIVE EFFLUDiT S, 3/4.11.2 GA SEOU S EFFLU EKT S 3/4.11.2.6 EKPLOSIVE GAS MIKTURE O EASES
=- ----- __ - _ _ _ -
This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Isolation of the af fected tank for purposes of purging and/or discharge permits the flammable gas concentrations of the tank to be reduced below the lower explosive limit in a hydrogen rich system. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR 50. O BEAVER VALLEY UNIT 2 B 3/4 11-10 i
3/4.11 RADI0 ACTIVE EFFLUBT S 3/4.11.3 SOLID RADIOACTIVE WASTE () BA SE S This specification implements the requirements of 10 CFR 50.36a and General Design Criteria 60 of Appendix A of 10 CFR 50 and requires the system be used whenever solid radwastes require processing and packaging pr ior to being shipped offsite. The process parameters used in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to was te type , waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil cont e nt , waste principal chemical constituents, mixing and curing times. i e I { O J O BEAVER VALLEY UNIT 2 B 3/4 11-11 1 J
3/4.11 RADIOACTIVE EFFLUENT S i 3/4.11.4' TOTAL DOSE BASES This Specification is provided to meet the dose limitations of 40 CFR 190. The Specification requires the preparation and submittal of a Special Report , in lieu of any other report , whenever the { calculated doses from plant radioactive ef fluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 nuclear reactors, it is highly unlikely that the resultant dose to MDtBER(S) 0F THE PUBLIC will ' exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to MEMBER (S) 0F THE PUBLIC for the calender yea- to within the 40 CFR 190 limits. For the purpose of the Special Report, it may be assumed that the dose commitment to MEMBER (S) OF THE PUBLIC from other uranitas fuel cycle sources is negligible, with the exception that dose contributions from ocher nuclear' fuel cycle
'!acilities at the sm site or within a radius of 5 miles must be cons idered.
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i i l I T i l l j O BEAVER VALLEY UNIT 2 B 3/4 11 ! i i 1 I i l 1
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3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM BASES The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways for those radionuclides which lead to the highest potential radiation exposures of MEMBER (S) 0F THE PUBLIC resulting from the
; station operation. This monitoring program thereby supplements the i radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the ef fluent measurements and modeling of the environmental exposure pathways, he initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operation experience.
he detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in -industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141. i 4 O i l BEAVER VALLEY UNIT 2 B 3/4 12-1 I i i I
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N 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS BASES (}.
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This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the Leonitoring programs are made if required by the results of this census. The best survey information from the door-to-door survey,* aerial survey, or by consultig with local agriculture authorities shall be used. This census saiisfies the vaquirements of Section IV.B.3 of Appendix I to 10 CFR 50. Restrict ing the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified al.d moni.'ored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used:
- a. That 20 percent of the garden was used for growing broad leaf vegetation (i.e., similar$to lettuce and cabbage), and
- b. A vegetation yield of 2 kg/ square neter.
O
- Confir44 tion by telephone is equivalent to door-to-door.
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hm O BEAVER VALLEY UNIT'2 B 3 /4 12-2
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM _ BASES The requirement for participation in an Interlaboratory Comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed a part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid. O
- O BEAVER VALLEY UNIT 2 B 3/4 12-3
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SECTION 5.0 l . DESIGN FEATURES I O , I i I O 1
5.0 DESIGN FEATURES 5.1 SIT E: SITE BOUNDARY FOR CASEOU S EFFLUENT S: 5.1.1 The site' boundary for gaseous ef fluents shall be as shown in Figur e 5.1-
- 1. Release points are shown on Figure 5.1-2.
SITE BOUNDARY FOR LIQUID EFFLUENT S: 5.1.2 The site boundary for liquid ef fluents shall be as shown in Figure 5.1-1. Release points are shown on Figure 5.1-2. EXCLUSION AREA: 5.1.3 The exclusion area shall be as shown in Figure 5.1-3. LOW POPULATION ZONE: 5.1.4 The low population zone shall be as shown in Figure 5.1-4. FIDOD CONTROL: 5.1 5' The flood control provisions (dikes, levees, etc.) shall be designed and maintained in accordance with the original design provisions contained in Section 3.4.1 of the FSAR. O BEAVER VALLEY UNIT 2 5-1
5.0 ' DESIGN FEATURES S .2 . CONTAINMENT : CONFIGURATION: 5.2.1 The reactor cont ainme nt building is a steel lined, reinforced concrete building of cylindeical shape, with a dome roof and having the following s de s ign - fe atur es :
- a. Nominal inside diameter = 126 feet.
- b. Nominal inside height = 185 feet.
- c. Minimum thickness of concrete walls = 4.5 feet,
- d. Minimum thickness of concrete roof = 2.5 feet,
- e. Minimum thickness of foundation mat = 10 fe e t ,
~
- f. Nominal thickness of vertical portion of steel liner = 3/8 inch.
- g. Nominal thickness of steel liner, dome portion = 1/2 inch,
- h. Net free volume = 1.73 x 106 cubic feet .
O ots1c= >=rssuat ino r >rairuar: 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 45 psig and a temperature of 369'F. PEN ETRAT ION S: 5.2.3 Penetr at ions tir ough the reac tor cont ainme nt building are designed and shall be maintained in accord ance with the original de s ign pr ovisions contained in Section 6.2.4 of the FSAR with allowance for normal degrada-t io n pur su' ant to the applicable Surveillance Requirements, b BEAVER VALLEY UNIT 2 5-2
5.0 DESIGN FEATURES l O 5 3 ==^croa coat: FUEL ASSEMBLIES: 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assem-bly containing 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal ac t ive fuel length of 144 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enr ichment of 3.2 weight per ce nt U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.3 weight percent U-235. CONTROL ROD ASSEMBLIES: 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies. The full length control rod as semblies shall contain a nominal 142 inches of absorber material. The absorber material shall be hafnium. All control rods shall be clad with stainless steel tubing. 5.4 REACTOR COOLANT SYST EM: DESIGN PRESSJRE AND TEMPERATURE: 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the appli-cable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME: 5.4.2 The total water and steam volume of the reactor coolant system is 9370 + ~ 100 cubic feet at a nominal T avg of 576*F. O BEAVER VALLEY UNIT 2 5-3
1 5.0 DESIGN FEATURES , 5.5 EMERGENCY CORE COOLING SYSTEMS: J The emergency core cooling systems are designed and shall be maint ained 4 in ac cord ance with the original design provis ions cont ained in Sect ion 6.3 of the FSAR with allowance for normal degr adation pursuant to the I applicable Surveillance Requirements. 5.6 (UEL STOR6GE l CRITICALITY: . 5.6.1 T he spe nt fuel storage racks are designed and shall be maintained with a minimum of 10.4375 inch center-to-center distance between fuel assemblies placed in the storage racks. With spent fuel of a maximum enrichment of 3.6% ' by weight 2UO . the fuel pool filled with pure water at 32*F, the fuel stored in the worst feasible geometric configuration, and with the worst case seismic deflection, Kgg will be lese than .95. DRAIN _AG E: 4 5.6.2 The spe nt fuel storage pool is designed and shall be maintained to pre-l ve nt inadver t e nt dr aining of the pool below elevation c 742'-1". 1
! CAPACITY:
i j 5.6.3 j The spe nt fuel s tor age pool is des igned and shall be maintained with a storage capacity limited to no more than 1,088 fuel elements. l 5.7 SEISMIC CLASSIFICATION l 5.7.1 Those s tr uctur es , systems and components ident ified as Category I items in Section 3.7 of the FSAR shall be designed and maintained to the origi-nel design provis ions with allowance for normal dege adation pursuant to the applicable surveillance requirements. 4 5.8 METEOROLOGICAL TOWER IDCATION 5.8.1 i i The meteorological tower shall be located as shown on Figure 5.1-1. !O BEAVER VALLEY UNIT 2 5-4 l f L I
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l FIGURE 5.1-3 EXCLUSION AREA BOUNDARY - BEAVER VALLEY PO'a'ER STATION UNIT NO. 2
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O SECTION 6.0 ADMINISTRATIVE CONTROLS i O i a O
ADMINISTRATIVE CONTROLS _ _ _ _ _ _ _an____ - - - ----:-- 6.1 RESPONSIBILITY The Superintendent of Operations shall be responsible for 4 6.1.1 overall f acility operation and shall delegate in writing the succession to l' this responsibility during his absence. 6.2 ORGANIZATION OFFSIT E:
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6.2.1 The of fsite organization for unit management and technical support shall be as shown in Figure 6.2-1. UN IT STAFF: 6.2.2 The Unit (s) organization shall be as shown in Figure 6.2-2 and:
- a. Each on duty shift shall 1e composed of at le as t the minimum shif t crew composition shown in Table 6.2-1.
- b. At le as t one licensed Operator shall be in the Control Room when fuel is in the reactor. I
- c. At le as t two licensed Operators shall be in the control room during l reactor startup, scheduled reactor shutdown and during recovery from reactor trips,
- d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor,
- e. All CORE ALT ERAT IONS af ter the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
- f. A Fire Brigade of at least 5 members shall be maintained on site at all times. The Fire Brigade shall not include 3 members of the minimum shif t crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.
i O BEAVER VALLEY UNIT 2 6-1 i
i ADMINISTRATIVE CONTROLS
- g. Administrative procedures shall be developed and implemented to limit ,
() the working hours of unit staf f who per form safety-related functions; , senior reactor operators, reactor operators, radiation control technicians, auxiliary oper ators , meter and control repairmen, and all personnel actually per forming work on safety related equipment. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is oper at ing. However , in the event that unforseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines onall be followed:
- a. An individual should not be permitted to work more than 16 hours s tr aight , excluding shif t tur nover time.
- b. An individual should not be permitted to work more than 16 hours in any 24-hour per iod, nor more than 24 hours in any 48-hour per iod, nor more than 72 hours in any seven day period, all excluding shif t turnover time.
- c. A break of at least eight hours should be allowed between work per iods , including shift tur nover t ime .
- d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staf f on a shift.
[) Any deviation from the above guidelines shall be authorized by the Plant Superintendent or predesignated alternate, or higher levels of management. Authorized deviations to the working hour guidelines shall be documented and available for NRC review. t I r [ r P BEAVER VALLEY UNIT 2 6-2
.i l FIGURE 6.2-1 I I 1 4 l' i i i i
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@ BEAVER VALLEY UNU 2 6-4
TABLE 6.2-1 MINIMUM BVPS UNIT 1 and UNIT 2 SHIFT CREW COMPOSITION
- Applicable Modes (#
License Category Both Units One Unit One Unit Both Units Qualification _l, 2, 3, 4 1,2,3,4 5, 6 5, 6
- dW Senior Reactor Operator (SRO) 2 2 1 8
Reactor Operator (RO) 3 3 2 8 8 3(8 Nonlicensed Auxiliary Operator 3 3 i Shif t Technical Advisor (STA) I Ih)(I) 1(h)(I) 0 1(h) g(h) g(h) ! Individual Qualified in Radiation Protection Procedures Rad / Chem Technician 1 1 1 NOT ES: (a)Except for the Shif t Supervisor , the shif t crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shif t crew members pr o-O vided immediate act io n is taken to restore the shift crew composition to within the minimum requirements. This provision does not permit any shift crew pos it ion to be unmanned upon shift change due to an oncoming shift creissan being late or absent. * (b) Includes the licensed SRO serving as the Shif t Supervisor. Operational Mode Definitions: I MODE 1 - Power Operation MODE 2 - Star t-up MODE 3 - Hot Standby
, MODE 4 - Hot Shutdown 1
MODE 5 - Cold Shutdown
! ODE 6 - Refueling (d)Does not include the SRO assigned during MODE 6 to directly supervise core i o per at io ns , ,
Dur ing any absence of the 9hift Supervisor from the Control Room while the , unit is in Mode 1, 2, 3, or ), an individual (other than the Shif t Technical Advisor) with a valid SRO license shall be designated to assume the Control 4 O BEAVER VALLEY UNIT 2 6-$
l, TABLE 6.2-1 (continued) O Room command function. During any absence of the Shif t Supervisor from the Control Room while the unit is in Mode 5 or 6, an individual with a valid I SRO or R0 license shall be des ignated to as sume the Control Room command f unct io n. (f) Minimum of 2 individuals for each unit; each individual may fill the same position on both units if qualified on both units. I Minimum of 2 individuals for each unit; one of two individual's may fill the same position on both units if qualified on both units. ( Minimum of 1 individual for each unit; one individual may fill the same position on both units if qualified on both units. ; ( One of two required individuals filling the SRO positions may also fill the STA position, if qualified. , O l { O BEAVER VALLEY UNIT 2 6-6
l l ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATION S 6.3.1 Each member of the facility and Radiation Protection steif l shall meet or exceed the minimum qualifications of A13SI N18.1-1971 for comparable positions , except for the Radiological Operations Coordinator who shall meet or exceed the qualifications of Regulatory Guide 1.8, beptember 1975, and the Shif t Technical Advisor who shall have a bachelor's de gree or l equivalent in a scientific or engineering discipline with specific training l in plant design and response analysis of the plant for trans ients and l accidents. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Director Nuclear Division Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. - 6.4.2 A Training Program for the Emergency Squad shall be maintained under the direction of the Director Nuclear Division Training and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976. { 6.5 REVIEW AND AUDIT 6.5.1 Onsite Safety Consmittee (OSC) FUNCTION: 6.5.1.1 The OSC shall function to advise the Station Superintendent on all matters related to nuclear safety. COMPOSITION: 6.5.1.2 The OSC shall be composed of the: Chairman: Chief Engineer Member: Senior Licensed Operator Member: Radiation Control Foreman Member: Maintenance Engineer Member: Project Engineer - Nuclear Engineering Department Member: Senior Testing or Study Projects Coordinator Member: Shif t Technical Advisor Member: Chemist Member: Quality Control Engineer Member: I&C Supervisor l l lO BEAVER VALLEY UNIT 2 6-7
ADMINISTRATIVE CONTROLS ("'y (_/ NUT E: The chairman of the OSC shall appoint an individual fr om each of the above listed job categories to serve as a member of the OSC for a per iod of at le as t 6 mont hs . NUT E: OSC members shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for compar able positions. The SRO shall meet the qualifications of Section 4.2.2 and the Maintenance Engineer will meet the qualifications of Section 4.2.3. A LI ERNNT ES: 6.5.1.3 All alternate members shall be appointed in writing by the OSC Chairman to serve on a tempor ary basis; however, no more than two alternates shall participate as voting members in OSC activities at ~any one time. MEETING FREQUENCY: 6.5.1.4 The OSC shall meet at le as t one: per calendar month ami as convened by the OSC Chairman or his designated alternate. QUORUM: O V 6.5.1.5 A quorum of the OSC shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES: 6.5.1.6 The O SC shall be res pons ib le for:
- a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Super int e nde nt to af fect nucle ar s a fe ty, i b. Review of all proposed tests and experiments that af fect nucle ar s a fe ty.
- c. Review of all proposed changes to Appendix "A" Technical Spec i ficat ions ,
- d. Review of all _ proposed changes or modifications to plant systems or equipment that affect nuc le ar s a fe t y.
BEAVER VALLEY UN rr 2 6-8
ADMINISTRATIVE CONTROLS s () e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to pr eve nt recurrence to the Manager of Nuclear Operations and to the Chairman of the Offsite Review Committee,
- f. Review all REPORTABLE EVENT S,
- g. Review of facility operations to detect potent ial safety hazards ,
- h. Per formance of special reviews , investigations or analyses and reports thereon as requested by the Chairman of the Offsite Review Committee.
AUTHORIT Y: 6.5.1.7 The OSC shall:
- a. Recommend to the Plant Superintendent written approval or disapproval of items considered under 6.5.1.6 (a) through (d) ab ov e .
- b. Render determinations in writing with regard to whether or not e ach item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the Manager of Nuclear Operations and the Of fsite Review Committee of disagreement between the OSC and the Plant Superint e nde nt ; however , the Plant Su per inte nde nt shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above'.
RECORD S: 6.5.1.8 The OSC shall maintain written minutes of each meeting and copies shall be provided to the Manager of Nuclear Operations and Chairman of the Of fsite Review Committee. I O BEAVER VALLEY UNIT 2 6-9
ADMINISTRATIVE CONTROLS 6.5.2 Offsite Review Committee (ORC) FUNCTION: 6.5.2.1 The ORC shall function to provide indepe nde nt review and audit of designated activities in .he areas of:
- a. Nuclear power plant operations ,
- b. Nuclear engineering,
- c. Chemistry and radiochemistry,
- d. Met allurgy, e . Ins tr ument at ion and contr ol,
- f. Radiological safety I g. Mechanical and electrical engineering,
- h. Quality assur ance practices .
COMPO SITION: i
, 6.5.2.2 The ORC shall be composed of the:
Chairman: Vice President , Nucle ar Division l Vice Chairman: Manager , Nuclear Safety and Licens ing q Member: Manager , Nucle ar Engineering Member: Manager , Nuclear Operations Member: Manager, Regulatory Affaire, Beaver Valley Power Station Unit No. 2 Member: Senior Project Engineer , Nuclear Engineering < Depar tme nt Member: Manager , Nuclear Support Services j Member: Site Service Manager , Westinghouse Electric Corpor at ion Member: Manager , Quality Assur ance Member: Director , Envirornient al and Radiological Safety Progrmas Member: Outside Conau1 tant , Chemistry and Radtochemistry O BEAVER VALLEY UNIT 2 6-10 i I
ADMINISTRATIVE CONTR0LS ( ALT ERNAT ES: 6.5.2.3 All alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in ORC activities at any one time, r . CON SULTArrr S: 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC. lI MEETING FREQUENCY: 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at le as t once per six months thereaf ter. S
, QUORUM:
I 6.5.2.6 A quorum of ORC shall consist of the Chairman or his designated alternate and at least 4 members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the f acility. l REVIEW: 6.5.2.7 The ORC shall review:
- a. The safety evaluation for 1) changes to procedures, equipment or systems, and 2) tests or experiments completed under the provision of Sect ion 50.59, 10 CFR, to verify that such actions did not cons e itute an unreviewed safety question.
- b. Proposed changes to procedures, equipment or systems which involve an i unreviewed safety question as defined in Section 50.59, 10 CFR.
- c. Proposed tests or- experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes in Technical Specifications or licenses,
- e. Violations of applicable statutes, codes, regulations, orders ,
Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance,
- f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety,
- g. All REPORTABLE EVENT S.
O BEAVER VALLEY UNIT 2 6-11
ADMINISTRATIVE CONTROLS
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structure systems, or compone nt s .
- i. Reports and meetings minutes of the OSC.
- j. The results of the eniironment al monitoring proge am pr for to submittal of the Annual Environnental Operating Report.
AUDIT S: 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the ORC. These audits shall encompass:
- a. The conformance of facility operation to provisions cont ained within the Technical Specifications and applicable license conditions at le as t once per 12 mont hs .
- b. The performance, tr aining and qualifications of the ent ire facility staff at least once per 12 mo nt hs .
- c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of oper ation that af fect nucle ar safety at least once per 6 months.
O v
- d. The performance of activities required by the Quality Assurance Progr am to meet the criteria of Appendix "B",10 CFR 50, at least once per 24 mont hs ,
- e. The Facility Dnergency Plan and implementing procedures at le as t once per 12 months.
- f. The Facility Security Plan and implementing procedures at le as t once per 12 months.
g.- Any other area of f acility operation cons idered appropr iate by the ORC or t he Vice Pr e s ide nt , Nuc le ar .
- h. The Facility Fire Protection Program and implementing procedures at least once per 24 months,
- i. An independent fir e pr ot ect io n and lo s s pr eve nt io n pr ogr am ins pe ct io n and audit shall be performed at least once per 12 months utilizing either qualified of f-site licensee personnel or an outside fire protect ion firm. l
- j. An inspection and audit of the fire protection and loss prevention !
program shall be performed by a qualified outside fire consultant at lease once per 36 months. l r b BEAVER VALI.EY UNIT 2 6-12 l ' 1
i k i ADMINISTRATIVE C0trTROLS I
/
i f AtTTHORITY: i l 6.5.2.9 The ORC shall report to and advise the Vice President, Nuclear lg , ] on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8. ! RECORD S: I
- 6.5.2.10 Records of ORC activities shall be prepared, approved aM 1 distributed as indicated by the following
- a. Minutes of each ORC meeting shall be prepared for and approved by the ORC Chairman within 14 days following each meeting, j b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved, and forwarded to the ORC Chairman within 14 days
- following completion of the review.
I c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded i to the Vice President, Nuclear and to the management posit ions l j responsible for the areas audited within 30 days af ter completion of i the audit. I d. The Vice President, Nuclear shall review all recommendations of the lg , ! ORC. I 6.6 REPORTABLE OCCURRENCE ACTION
]
i 6.6.1 The following actions shall be taken for REPoltTABLE i OCCURRDICES: l
- a. The Commission shall be notified in accordance with 10CFR50.72 aM/or j a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50, and t l b. Each REPORTABLE EVENT shall be reviewed by the OSC, and the results of g this review shall be submitted to the ORC. I i
- 6.7 SAFETY LIMIT VIDIATION 6.7.1 The following actions shall be taken in the event a Safety l Limit is violated
j i
- a. The unit shall be placed in at least Har STANDBY within one hour, j b. The Safety Limit violation shall be reported to the Commission, the Manager of Nuclear Operations and to the ORC within 24 hours. !
j , O , BEAVER VALLEY UNIT 2 6-13 i
ADMINISTRATIVE CONTROLS
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the On-Site Safety Committee (OSC). This report shall describe 1) applicable circumstances preceding the violation, 2) ef fects of the violation upon f acility components, systems or structures, and 3) corrective action taken to prevent r ecur r e nce .
- d. The Safety Limit Violation Report shall be submitted to the Commission, the ORC and the Manager of Nuclear Operations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be est'blished, implemented, and maintained covering the activities referenced below;
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November ,19 72.
- b. Refueling oper ations ,
- c. Surveillance arvi test activities of safety related equipment.
- d. Secur ity Plan implement er lon.
- e. Dnergency Plan implement ation.
O
- f. Fire Protect hn Progr am implementation,
- g. PROCESS CONTROL PROGRAM implement ation.
- h. OFFSIT E DOSE CALCUIATION MANUAL implement ation.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the OSC and approved by the Plant Supe r int e nde nt , predesignated alternate or a predesiginated Department Manager to whom the Plant Superintendent has assigned in writing the responsibility for review and appecval of specific subjects considered by the corsmittee, as app li c ab le . 6.8.3 Tempor ar y changes to pr ocedur es of 6.8.1 above may be made provided:
- a. The intent of the original procedure is not altered,
- b. The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior Reactor Oper ator's Licenae on the unit af fected.
- c. The change is documented, reviewed by the OSC and approved by the Plant Superintendent within 14 days of implementation.
O BEAVER VALLEY UNIT 2 6-14
i l I ADMINISTRATIVE CONTROLS O 6 9 atro rtac ==outar"='r s 1 ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports sha11 be submitted to the Director of the Regional Of fice of Inspection and Enforcement unless otherwise noted. 4 STARTUP REPORT S: 6.9.1.1 A suusary report of plant startup and power escalation testing will be submitted following 1) receipt of an operating license, 2) amendment to the license involving a planned increase in power level, 3) installation of fuel that has a dif ferent design or has been manuf actured by a dif ferent fuel supplier, atui 4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. 6.9.1.2 The startup report shall address each of the tests ide nt ified in the FSAR and shall include a description of the measured values of the 1 operating conditions or characteristics obtained during the test progran and a comparison of these values with design predictions and specifications. Any l corrective actions that were required to obtain satisfactory operation shall i also be described. Any additional specific details requested in license i conditions based on other couritments shall be included in this report. O 6.9.i.3 completion of the startup teet program, 2) 90 days following resumption or StartuP re, orts shait be sebsitted within i) 90 dars fe11 wieR coweencement of consercial power operation, or 3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three event s (i.e. , initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), suppleme nt ary reports shall be submitted at least every three months until all three events have been completed. ANNUAI, REPORT SI ! 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year The initial report shall be submitted prior to March 1 of the year following initial criticality. 1 l 1 A eingte submittal may be made for a mult(pie unit site. The submittal j should combine those sections that are conson to all units at the j site. l i BEAVER val. LEY UNIT 2 6-15 1 1
ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation of the number. of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem / year and their associated man-rem exposure according to work and job functions 2 (e.g. , reactor oper ations and surveillance, inservice inspection, routine maintenance , special maintenance [ describe maintenance], waste processing, and refueling). The dose as s ignme nt s to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for . In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions,
- b. Documentation of all challenges to the pressurizer power operated relief valves (PORV's) or pressurizer safety valves.
1 MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Of fice of Manageme nt Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Of fice, submitted no later than the 15th of each month following the calendar month covered by the report. 2 This tabulation supplements the requirements of Section 20.407 of 10 CFR Part 20. O
- BEAVER VALLEY UNIT 2 6-16
,L 1 ADMINISTRATIVE CONTROL 6.9.1.7 T his item inter,t ionally blank.
6.9.1.8 This item intentionally blank. 6.9.1.9 This item intentionally blank. ANNUAL RADIOLOGICAL DIVIRONMDITAL REPORT 3: 6.9.1.10 Routine radiologiesi environmental operating reports covering the operation of the unit duJiru the previous calendar year shall be submitted prior to May 1 of each year and will include reporting any deviations not reported under 6.9.2 with respect to the Radiological Ef fluent Technical Specifications. 6.9.1.11 The annual radiological envirormental reports shall include suussar ies , inter pret at io ns , and statistical evaluation of the results of the radiological environmental surveillance activities for the report per iod , including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports, and an assessment of the observed impacts of.the plant operation on the environment. The reports shall also include the rescits of the land use censuses required by Specification 3.12.2. If harmful ef fects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of, sction to alleviate the problem. O The annual radiological environmental operating reports shall gj include summarized and tabulated results in the format of Table 6.9-1 of all radiological environmental samples taken during the report period. In the eve nt that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. 3 A single submittal may be mado for a multiple unit site. The submittal should combine those sections that are conmon to both units. O . BEAVER VALLEY UNIT 2 6-17
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ADMINISTRATIVE CONTROLS The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions fr om one reactor; and the results of licensee participation in the Interlaboratory Comparison Program required by Specification 3.12.3. SEMI-ANNUAL RADIOACTIVE EFFLUD(T RELEASE REPORT 4-6.9.1.12 Routine radioactive ef fluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. 6.9.1.13 The radioactive ef fluent release reports shall include a summary of the quantities of radioactive liquid and gaseous ef fluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, Revis ion 1, June ,1974, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light'-Water-Cooled Nuclear Power Plants," with data summarized l. on a quarterly basis following the format of Appendix B thereof. In addition, the radioactive ef fluent release report to be submitted 60 days af ter Jan sary 1 of each year shall include an annual summary of hourly meteorological data collected collected over the previous year. This annual suussary may be either in the form of an hour-by-hour listing of wind speed; wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous caltadar year. This report shall also include an assessment of the radiation dcses from radioactive ef fluents to , MEMBER (S) 0F THE PUBLIC due to thair activities inside the site boundary (Figure 5.1-1) during the report period. All assumptions used in making these l assessments (e.g., specific activity, exposure time and location) shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM). l 4 A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to all units at the site; however, for units. with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. O BEAVER VALLEY UNIT 2 6-19 l}
ADMIN I Sr RAT IV E CONT ROLS The radioactive ef fluent release report to be submitted 60 days af ter January 1 of each year shall also include an assessmant of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1. The SKYSHINE' Code (available from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16. The radioactive ef fluent release reports shall include an as ses sme nt of radiation doses from the radic active liquid and gaseous ef fluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21. In addition, the unrestricted area boundary maximum noble gas ganna air and beta air doses shall be evaluated. The assessment of radiation doses shall be performed in accordance with the ODCM. The radioactive ef fluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month period. RADIAL PEAKING FACTOR LIMIT REPORT: RTP 6.9.1.14 The F xy limit for Rated Thermal Power (Fxy ) shall be provided to the Director of the Regional Of fice of Inspection and Enfor ceme nt , with a copy to the Director , Nuclear Reactor Regulation, Os Attention Chief of the Core Performance Branch, United States Nuclear Regulatory Commission, Washington, DC 20555 for all core planes cont aining bank "D" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days pr ior to the date the limit would become ef fective unless otherwise exempted by the Commission. RTP Any information needed to support F xy will be by request fr.om the NRC and need not be included in this report. BI-ANNUAL ENVIRONMDITAL OPERATING REPORT: 6.9.1.15 A report on the terrestrial environmental surveillance program for the previous 24 months of operation shall be submitted to the Regional Director of the Of fice of Inspection and Enforcement (with copy to Director of Nuclear Reactor Regulation) within 90 days af ter January 1 of each year. The period of. the first report shall begin with the date of initial criticality. The report shall include sununaries, interpretations, and statistical evaluation of the results of the nonradiological and environmental surveillance activities (Section 3.7.16) for the report pe r iod . O BEAVER VALLEY UNIT 2 ta 20 lg
ADMINISTRATIVE CONTROLS A comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports, and an as se s sme nt of .the observed impacts of the plant operation on the environment shall be provided. If harmful effects or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to alleviate the problem. SPECIAL REPORT S: 6.9.2 Special reports shall be submitted to the Dii,.ctor et thc Of fice of Inspection and Enforcement (Regional Of fice) within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification: l
- a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Seismic Monitoring Instrumentation, Specificat ion 3.3.3.3.
- c. Inoperable Meteorological Monitoring Instrumentation, Specificat io n 3.3.3.4.
- d. Seismic event analysis, Specificat ion 4.3.3.3.2.
- e. Sealed source le akage in excess of limits, Specificat ion 4.7.9.1.3.
O f. rire oetection 1nstrumentatien. Specificatien 3.3.3.6.
- g. Fire Suppression Systems, Specificat ions 3.7.14.1, 3.7.14.2, 3.7.14.3 and 3.7.14.5.
- h. Miscellaneous reporting requirements specified in the Action Statements for Radiological Effluent Technical Specifications.
- i. RCS Specific Activity, Specificat ion 3.4.8.
I j . Cont ainme nt Inspection Report , Specificat ion 4.6.1.6.2. BEAVER VALLEY UNIT 2 6-21
ADMINISTRATIVE CONTROLS O 6 io ascono arrs=11o" 6.10.1 The following records shall be retained for at leas t. five (5) years:
- a. Records and logs of facility operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities , ins pect io ns ,
repair and replacement of principal items of equipment related to nuclear safety.
- c. ALL REPORTABLE EVENT S.
- d. Records of surveillance activities, ins pect io ns and calibrations
< required by these Technical Specifications,
- e. Records of reactor tests and experiments.
I f. Records of changes made to Operating Procedures.
- g. Records of radioactive shipments.
- h. Records of sealed source leak tests and results.
- i. Records of annual physical inventory of all sealed source material of
(} record. i BEAVER' VALLEY UNrr 2 6-22 !
ADMINISTRATIVE CONTROLS
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6.10.2 The following records shall be retained for the duration of (#') the Facility Operating License:
- a. Records and drawing changes reflecting facility design _ modifications made to systems and equipment described in the Final Safety Analysis Report,
- b. Records of new and irradiated fuel inventory, fuel trans fers and assembly burnup histories,
- c. Records of facility radiation and contamination surveys,
- d. Records of radiation exposure for all individuals entering radiation control areas.
- e. Records of gaseous and liquid radioactive material released to the environs.
- f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles,
- g. Records of training and qualification for current members of the plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications,
- i. Records of Quality Assurance activities required by the QA Manual,
- j. Records of reviews performed for changes made to procedures or equipme nt or reviews of tests and experiments pursuant to 10 CFR 50.59.
i
- k. Records of meetings of the OSC and the ORC.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
- m. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and acsociated y'
installation and maintenance records. L
- n. Records of analyses required by the Radiological Environmental !
Monitoring Program. i
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BEAVER VALLEY UNIT 2 6-23 l1' t l I i
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ADMINISTRATIVE CONTROLS i O e ti a^o it'o" >"or"cr'o" raoca^" Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA' 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 arem/ hour, but less than 1000 arem/ hour, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit
- or Radiological Access Control Permit. Any individus1 or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a facility health physics supervisor in the Radiological Work Permit or Radiological Access Control Permit.
6.12.2 The requirements of 6.12.1, above, also apply to each high radiation area in which the intensity of radiation is greater than 1000 mren/ hour. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shif t Supervisor on duty and/or a facility health physics supervisor. 6.13 THIS IT EM INTENTIONALLY BIANK
- Health physics personnel, or personnel escorted by health physics personnel in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement during the performance of their radiation protection duties, provided they comply with approved -
radiation protection procedures for entry into high radiation areas. O BEAVER VALLEY UNIT 2 6-24 1I
! ADMINISTRATIVE CONTROLS
__ = __ _ _ 6.14 PROCESS CONTROL PROGRAM (PCP) l FUNCT ION : } 6.14.1 The PCP shall be a manual containing the processing steps and ! a set of established process parameters detailing the progran of sampling, analysis, and evaluation within which solidification of radioactive wastes is assured, consistent with Specification 3.11.3.1 and the surveillance requirements of these Technical Specifications. LICDISEE INITIATED WANGES: 6.14.2 Shall become ef fective upon review and acceptance by the OSC. l 6.15 0FFSITE DOSE CALCUIATION MANUAL (ODCM) FUNCT ION: 6.15.1 The ODCM shall describe the methodology and parameters to be; ( used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid ef fluent monitoring instrumentation alare/ trip setpoints consistent with the applicable LCO's contained in these Technical . Specifications. Methodologies and calculational procedures acceptable to the Comunission art contained in NUREG-0133. LICDISEE INITIATED MANGES: Shall become effective upon review and acceptance by the OSC. 6.15.2 l1 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASEOUS, l AND SOLID) j FUNCTION: 6.16.1 The radioactive waste treatment systems (liquid, gaseous and l solid) are those systems described in the f acility Final Safety Analysis , Report or Hazards' Summary Report , and anendments thereto, which are used to maintain that control over radioactive acerials in gaseous and liquid ; ef fluents and in solid waste packaged for of fsite shipment required to meet the LCO's set forth in Specifications 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.1.4, , 3.11.2.1, 3.11.2.2, 3.11.2.3, 3.11.2.4. , 3.11.2. 5, 3.11.2.6, 3.11.3.1, ard t 3.11.4.1. l 6.16.2 Major changes as defined in Section 1 to time radioactive waste l l systems (liquid, gaseous and solid) shall be made by the following method: f I i O : BEAVER VALLEY UNIT 2 6-25 {I !
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ADMINISTRATIVE CONTROLS () A. Licensee initiated changes:
- 1. If a permanent facility change is made to a radioactive treatment systen that could result in an increase in the volume or activity discharged, the Commission shall be informed by the inclusion of a suitable discussion of each change in the Annual 10 CFR 50.59 Report for the period in which the changes were made. The discussion of each change shall contain:
- a. A summary of the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59);-
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- b. Suf ficient detailed information to totally support the reason i for the change without benefit of additional or supplemental information;
- c. A detailed description of the equipment, components and processes involved and the inter f aces with other plant systems;
- d. An evaluation of the change will be submitted which shows t he -
predicted increase of releases of radioactive materials in liquid and gaseous ef fluents and/or quantity of solid waste from those previously predicted in the license application and amendments thereto; p e. An evaluation of the change which shows the expected increase in v the maximum exposures to an individual in the unrestricted area
~from those previously estimated in the license application and amendments thereto;
- f. A comparison of the predicted increase of releases of radioactive materials in liquid and gaseous effluents and in solid waste to the actual releases for the period the changes were made;
- g. An estimate of the exposure to plant operating personnel as a
- result of the change; and
- h. Documentation of the fact that the change was reviewed and found acceptable by the OSC.
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- 2. The change shall become effective upon review and acceptance by the O SC. ;
i l BEAVER VALLEY UNIT 2 6-26 r . t
ADMINISrRATIVE CONTROLS (-)s 6.16.3 Background of what cons titutes " major changes" to radioact ive l1 waste systems (liquid, gaseous, and solid),
- a. Background
- 1. 10 CFR 50, Section 50.34a(a) requires that each application to construct a nuclear power reactor provide a description of the equipme nt installed to maintain control over radioactive material in gaseous and liquid ef fluents produced during normal reactor oper at io ns including operational occurrences.
- 2. 10.CFR 50, Section 50.34a(b)(2) requires that each application to construct a nuclear power reactor provide an estimate of the quantity of radionuclides expected to be released annually to unrestricted areas in liquid and gaseous effluents produced during.
normal reactor operation.
- 3. 10 CFR 50, Section 50.34a(3) requires that each application to cons truct a nuclear power reactor provide a description of the pr ovis ions for packaging, storage and shipment offsite of solid waste containing radioactive materials resulting from treatment of gaseous and liquid ef fluents and from other sources.
- 4. 10 CFR 50, Section 50.34A(3)(c) requires that each application to operate a nuclear power reactor shall include 1) a description of
(} the equipment and procedures for the control of gaseous and liquid ef fluents and for the maintenance and use of equipment installed is radioactive waste systems and 2) a revised estimate of the information required in (b)(2) if the expected releases and exposures dif fer significantly from the estimate submitted in the application for a construction permit.
- 5. The Regulatory staf f's Safety Evaluation Report and ame ndme nt s thereto issued prior to the issuance of an operating license contains a description of the radioactive waste systems installed in the nuclear power reactor and a detailed evaluation .(including estimated releases of radioactive materials in liquid and gaseous waste and quantities of solid waste produced from normal operation, es timated annual maximum exposures to an individual in the unrestricted area and estimated exposures to the general population) which shows the capability of these systems to meet t he appropriate regulations.
t { l BEAVER VALLEY UNIT 2 6-27 l} l l
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ADMINISTRATIVE CONTROLS "o""o"'"o ""oc "^" O 6 17 a^o to'oc 'c^' -
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l1 The Manager of Nuclear Safety and Licensing delegates the responsibility for the Radiological Environmental Monitoring Progran to the Director, Environmental and Radiological Safety Programs of his designated alternate. The Director , Environmental and Radiological Safety Programs is responsible for administering the of fsite Radiological Environmental ~ Monitoring Program. He shall determine that the sampling program is being implemented as described to verify .that the environment is adequately protected under existing procedures. He shal1 also have the responsibility for establishing, implementing, maintaining and approving offsite environmental . program sampling, analysis and calibration procedures. I O BEAVER VALLEY UNIT 2 6-28 \}}