ML20073C109: Difference between revisions
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B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Baltimore Gas & Electric Company (1) Pursuant to Section 104b of the Act and 10 CTR Part 50, " Licensing of Production and Utilization racilities " to possess, use, and operate the facility at the designated location in Calvert . | B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Baltimore Gas & Electric Company (1) Pursuant to Section 104b of the Act and 10 CTR Part 50, " Licensing of Production and Utilization racilities " to possess, use, and operate the facility at the designated location in Calvert . | ||
County, Maryland, in accordance with the procedures and limitations set forth in this licenre; (2) Pursuant to the Act and 10 CTR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission de-tectors in amounts as required for reactor operation; (3) Pursuant to the Act and 10 CTR Part 30 to receive, possess, use at any time 100 mil 11 curies each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration; 500 millicuries of byproduct material in the form of equipment that is radioactively contaminated by radio- | County, Maryland, in accordance with the procedures and limitations set forth in this licenre; (2) Pursuant to the Act and 10 CTR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission de-tectors in amounts as required for reactor operation; (3) Pursuant to the Act and 10 CTR Part 30 to receive, possess, use at any time 100 mil 11 curies each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration; 500 millicuries of byproduct material in the form of equipment that is radioactively contaminated by radio- | ||
. isotopes with atomic numbers within the range 3 to 83, for inspection and maintenance of the facility, as described in the licensee's s letter-application for license amendment dated. November 13, 1974, as amended by letter dated November 15, 1974, and Sodium-24, in liquid form, not to exceed 500 mil 11 curies for tracer measurements for steam turbine acceptance testing. | . isotopes with atomic numbers within the range 3 to 83, for inspection and maintenance of the facility, as described in the licensee's s letter-application for license amendment dated. November 13, 1974, as amended by {{letter dated|date=November 15, 1974|text=letter dated November 15, 1974}}, and Sodium-24, in liquid form, not to exceed 500 mil 11 curies for tracer measurements for steam turbine acceptance testing. | ||
(4) Pursuant to the Act and 10 CTR Parts 40 and 70 to receive, possess and use at any time 100 milligrams each of a.ny source or special nuclear material without restruction to chemical er physical form, for sample analysis or instrument calibration; (5) Pursuant to the Act and 10 CTR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. | (4) Pursuant to the Act and 10 CTR Parts 40 and 70 to receive, possess and use at any time 100 milligrams each of a.ny source or special nuclear material without restruction to chemical er physical form, for sample analysis or instrument calibration; (5) Pursuant to the Act and 10 CTR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. | ||
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ENVIRONMENTAL I!@ACT APPRAISAL BY THE OPTICE 07 Nt' CLEAR REACTOR RE::ULATION SUFFORTING AMEhTME*ti NOS. 23 AND 7 TO LICENSE NOS. DPR-53 AND DPR-69 BALTIMORE GAS & ELECTRIC CCtOANY CALVERT CLTTTS STCLEAR PO'1ER PLANT UNIT NCS.1 AND 2 DOCKET Nos. 50-317 AND 50-318 Description of Procesed Action By letter dated January 3, 1977, and supplement thereto dated June 7, 1977, the Baltimore Gas & Electric Company (DC&E) requested an amendment to racility Operating License No. DPR-53 for the Calvert Cliffs Nuciant Power l | ENVIRONMENTAL I!@ACT APPRAISAL BY THE OPTICE 07 Nt' CLEAR REACTOR RE::ULATION SUFFORTING AMEhTME*ti NOS. 23 AND 7 TO LICENSE NOS. DPR-53 AND DPR-69 BALTIMORE GAS & ELECTRIC CCtOANY CALVERT CLTTTS STCLEAR PO'1ER PLANT UNIT NCS.1 AND 2 DOCKET Nos. 50-317 AND 50-318 Description of Procesed Action By {{letter dated|date=January 3, 1977|text=letter dated January 3, 1977}}, and supplement thereto dated June 7, 1977, the Baltimore Gas & Electric Company (DC&E) requested an amendment to racility Operating License No. DPR-53 for the Calvert Cliffs Nuciant Power l | ||
Plant Unit No. 1. Following discussions with and agreement by BG&E, the reques . | Plant Unit No. 1. Following discussions with and agreement by BG&E, the reques . | ||
was med1fied to include Tacility Operating License No. DPR-69 for the Calvert Cliffs Nuclear Power Plant Unit No. 2. The request, as modified, vould replace the Appendix B (Environmental) Technical Specifications (ETS) of both Unit Nos.1 and 2, in their antirety, with a common set of ITS. The requested content and format of the new common CTS is very closely reisted | was med1fied to include Tacility Operating License No. DPR-69 for the Calvert Cliffs Nuclear Power Plant Unit No. 2. The request, as modified, vould replace the Appendix B (Environmental) Technical Specifications (ETS) of both Unit Nos.1 and 2, in their antirety, with a common set of ITS. The requested content and format of the new common CTS is very closely reisted | ||
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e By letter dated June 4, 1976, the licansee providcd addittensi . | e By {{letter dated|date=June 4, 1976|text=letter dated June 4, 1976}}, the licansee providcd addittensi . | ||
'information pursuant to Appendix I to 10 CPR Part 30. After we complete our evaluation of this information we intend to revise the Technical Specifications to reflect the requirements of Append 1:: I. | 'information pursuant to Appendix I to 10 CPR Part 30. After we complete our evaluation of this information we intend to revise the Technical Specifications to reflect the requirements of Append 1:: I. | ||
To redute confusion, the word " site" has been inserted to designate | To redute confusion, the word " site" has been inserted to designate | ||
Latest revision as of 17:28, 27 September 2022
| ML20073C109 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/16/1991 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20073C093 | List: |
| References | |
| NUDOCS 9104250106 | |
| Download: ML20073C109 (974) | |
Text
{{#Wiki_filter:- - - - - - - M ANUAL NO. 77 ~ TECHNICAL SPECIFICATIONS
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MANUAL .. THIS MANUAL IS CONTROLLED
,BY THE NUCLEAR POWER DEPT. .
TECHNICAL LIBRARIAN
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ASSIGNED TO Pe 4cuoj .. ORGANIZATION esta
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BALTI!! ORE CAS & ELECTRIC COMPA*Y ' l DOCKET No. 50-317 (Calvert Cliffs Nuclear Power Flant, Unit 1) l TACILITY OPEPATING LICENSE License No. DPR-53
- 1. The Atomic Energy Commission (the Comission) having found that:
A. The application for license filed by Baltimore Gas & Electric Company (the licensee complies with the. standards and requirements of the Atomic , Energy Act of 1954, as acended (the Act), and the Comission's rules and regulations set forth in-10 CTR-Chapter I and all required notifications 4
, to other agencies or bodies have been duly madei B. Construction of the Calvert Cliffs Nuclear Power Plant Unit 1. (facility) has been substantially completed in conformity with Construction Permit No. CPPR-63 and the application, as amended, the provisions of the Act and the rules and regulations of the Comission C. The facility will operate in conformity with the application, as amended.
the provisions of the Act, and tha rules and regulations of the Comission; D. There is reasonable assurancet (1) that the activities authorized by this operating license can be conducted without endangering the health and safety of the. public. 'and _(ii) that such activities will be conducted in compliance with the rules and regulations of the Comission; I E. The'11censee.is technically and financially qualified to engage in the . !. activities authorized by this operating license in accordance with the rules and regulations of the Comission: T. The licensee has satisfied the applicable provisions of.10 CFR Part 140, "Tinancial Protection Requirements and Indemnity Agreements," of the
- Comission's regulations:
G. The issuance of this operating license will not be inimical to the comon defense and security or to the health and safety of the~ public;
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H. After weighing the environmental, economic, technical, and other benefits t of the facility against environmental. costs-and considering available ! alternatives the issuance.of Tacility Operating License No. DPR-53 l- (subject to the conditions for protection of the environment set forth l herein). is in accordance .with 10 CFR Part 50, Appendix D, of the Comission's regulations and all applicable requirements of said Appendix D have been satisfied; and i 6
. .g. ;
I. The receipt, possession, cnd use of source, byproduct end special nuclear material as authorized by this license vill be in accordance with ths Commission's regulations in 10 CTR Parts 30, 40, and 70, including 10 CTR Section 30.33. 40.32, and 70.23 anu 70.31. II. Tacility Operating License No. DPR-53 is hereby issued to the Baltimore Gas and Electric Company to read as follows: A. This license applies to the Calvert Cliffs Nuclear Plant Unit 1 a pressurized water reactor and associated equipment (the facility), ovned by Baltimore Gas & Electric Company. The facility is located in Calvert County, Maryland and is described in the "Tinal Safety Analysis Report" as supplemented and amended (Amendments through ), and the En-vironmental Report as supplemented and amended (Amendments No. I and 2). B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Baltimore Gas & Electric Company (1) Pursuant to Section 104b of the Act and 10 CTR Part 50, " Licensing of Production and Utilization racilities " to possess, use, and operate the facility at the designated location in Calvert . County, Maryland, in accordance with the procedures and limitations set forth in this licenre; (2) Pursuant to the Act and 10 CTR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission de-tectors in amounts as required for reactor operation; (3) Pursuant to the Act and 10 CTR Part 30 to receive, possess, use at any time 100 mil 11 curies each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration; 500 millicuries of byproduct material in the form of equipment that is radioactively contaminated by radio-
. isotopes with atomic numbers within the range 3 to 83, for inspection and maintenance of the facility, as described in the licensee's s letter-application for license amendment dated. November 13, 1974, as amended by letter dated November 15, 1974, and Sodium-24, in liquid form, not to exceed 500 mil 11 curies for tracer measurements for steam turbine acceptance testing.
(4) Pursuant to the Act and 10 CTR Parts 40 and 70 to receive, possess and use at any time 100 milligrams each of a.ny source or special nuclear material without restruction to chemical er physical form, for sample analysis or instrument calibration; (5) Pursuant to the Act and 10 CTR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
3 C. This 81 cense shall be doemed to conta6n and is subject to the conditions specified in the following Commission regulations in 10 CTR Chapter 1: Part 20 Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part $0, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated belovt (1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 mega-watts (thermal). (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amencment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-(3) The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.21 of the NRC's Fire Protection Safety Evaluation (SE), dated September 14, 1979 for the facility. These modifications vill be cocpleted in accordance with the schedule in Table 3.1 of the SE. If any modification cannot be completed on schedule, the licensee shall submit a report explaining the circumstances and propose, for staff approval, a revised schedule.
In addition, the licensee shall submit the additional information identified in Table 3.2 of the SE in accordance with the schedule contained therein. If the information cannot be submitted on schedule, the licensee shall submit a report explaining the circumstances together with a revised schedule. The licensee is required to implement and maintain the administrative controls identified in Section 6 of the NRC's Pire Protection Safety Evaluation on the facility dated September 14, 1979. The administrative controls, with the exception of fire fighting strategies and quality assurance procedures, shall be in effect by Novsaber 1, 1979. The quality assurance procedure shall be in effect by January 1,1980, and the fire fighting strategies by August 1, 1980. I
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(4) physical protection Tho licensoe shall fully implement and ma6nBain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the MAscellaneous Amendments and Search Requirements revisions 1 to 10 CTR 73.55 ($1 TR 27817 and 27822) and to the authority I of 10 CTR 50.90 and 10 CTR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Calvert Cliffs Nuclear Pover Plant Physical Security Plan," with revisions submitted through Tebruary 17, 1988; .
"Calvert Cliffs Nuclear Power Plant Guard Training and Qualification Plan " with revisions submitted through November 1, 19851 and "Calvert Cliffs Nuclear Power Plant Safeguards Contingency Plan." with revisions submitted February 9,1988.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein. (5) Secondarv k'ater Chemistry Honitoring program The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall includes
- a. Identification of a sampling schedule for the critical parameters and control points for these parameters,
- b. Identification of the procedures used to quantify parameters that are critical to control points.
- c. Identification of process sampling points.
- d. Procedure for recording and management of data,
- e. Procedures defining corrective actions for off control point chemistry conditions; and
- f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.
Amendment #132
5 D. Pursuant to the Agreement doted April 25, 1973, between the licenses, Chesapeake Environm:ntal Protection Association and the Commission, the licensee shall, prior to the operation of Unit No. 1 above 60% capacity, but in no event later than six (6) months af ter the date of commercial operation, have selected an optimum cooling tower system for both Units Nos. 1 and 2 and vill have completed the following pre-construction design and engineering of said system (this effort would provide a ready design in the event conversion to an alternate cooling system proves necessary at a later date because of unexpected, detrimental environmental effects): (1) Civil Engineering (a) Site investigation and soil studies (b) Shoreline design change studies and preparation of license application for dredging and filling. (c) Prelimincry layout drawings of circulating water piping. (d) Toundation studies (2) Electrical Engineering Preliminary single line drawings and design review of auxiliary electrical capacity. (3) Mechanical Layout and Engineering (a) Cooling tower sizing and selection (b) Cooling tower specifications (c) Preliminary equipment sizing of pumps, piping and buildings (d) Large pump specifications (e) Additional site meteorological studies pertaining to cooling tower drift. E. This license is effective as of the date of issuance and shall expire at midnight July 31, 2014. g g FOR THE ATOMIC ENERGY COMMISSION I ' TA. Giambusso Deputy Director DW6 CN , for Reactor Projects NDirectorate of Licensing ' g g [ O XC) M
Attachment:
Appendices A & B Technical Specifications Date of Issuance: July 13, 1974
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[ 0 ) { )' X [ 5 "O _ CEME 10. ))2 - 53 AMENDMENT NO. 90, FEBRUARY 11,1977 i e 4 5.4
'c5UED BY THE UNITED STATES NUCLEAR REGULATORY COMMIS$10N
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/ P)EN7 X "A" TO l.lCENSE NO. DPR - 53 AMENDMENT NO.90, FEBRUARY 11,1977
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'cSUCD BY THE UNITED STATES NUCLEAR REGULATORY COMMISSION ,
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CALVERT CLIFFS NUCLEAR POWER PLANT UNIT 1 TECHNICAL SPECIFICATIONS APPENDIX "A" . TO LICENSE NO. DPR-53
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!DF:N!**0NS SECTION PAGE i1. 0 CEFIN!TIONS Defined Ter.s............................................... 11 Thermal Power............................................... 1-1 Rated Thermal Power......................................... 11 O p e r a t i o n a l Mo d e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -l A:tien...................................................... 1-1 Operable - Operability...................................... 1-1 Reportable [ vent............................................. I-2 Containment Integrity....................................... 1 2
- C h a n ne l C a l i b ra t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 Channel Check................................................ 13
- Channel Functional Test..................................... 1 3 I
CoreAlteration.............................................1-3 S hu t d ewn Ma rg i n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 y i d e n t i fi e d L e a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -4 iy U ni c e nt i f i e d l e a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -4 P res s u re Bounda ry L e a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -4 Q; Controlled Leakage.......................................... 1 /. 1 A:imuthal E0.wer 7i1t ....................................... CseL0uivaient I .31........................................ i 1-4 l * =, l - Avera ge Di 5 i nteg ra*iCn Ene rgy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -5 Staggerec Test Basis........................................ 1-5 FreQuem y NCsM iC0.a......................................... 1-5 Axial Shace Incex............................................ 15
}' Unroeded Planar Radial Peaking Factor - F 1-5
$' Rea ct:r Trio System Respons e Time. . . . . . . . xy. ................
.................... I6 4 En9ineered Safety feature ResCOnse I1m0...................... 1-6 P hy s i c s T e s t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6
. Unre: ed Integrate: Racial Peaking facter - F 1 -6 ,
, Ga s ecus Ra cwa s te T rea tment 5ys em. . . . . . . . . . . ................ r. . . . . . . . . . . . . . 1-6 i
, MeC er($) Of the EUDlic..................................... 1-6 '
' Of f site Des e Calculation Manual (0DCM) . . . . . . . . . . . . . . . . . . . . . . 16 Proc a is Control Program (PCP ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -7 l Eur9e-Purjing................................................ l-7 i S i t e S o u n c a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l - 7 i
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SCIiCifiCatiCn.............................................. 1-7 ! ScurceCheck................................................1-7 ' Unres*?ictedArea...........................................1-7 l l Ventilati0n [Khaus t Trea tment 5ystem. . . . . . . . . . . . . . . . . . . . . . . . , I-7 ! Ven*in$..................................................... 1-8 , { i DALVERT CLIFF 5 - M:7 1 4 ,e : Acenement No. 21, y , 100 t
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SECTION PAGE P. 9 2.1 $ATETY LIMITS Reactor C0re................................................ 2+l Reactor Coola nt Sys tem Pres sure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 2.2 LIMITING SAFETY $YSTEW SE~'INGS
- React 0r Trip Setpoint3...................................... 26 I
BASES SECT:0N DaGE
.g 2.1 SAFE *Y L!WI*5 Reacter ~ ~ Cere..........................................'...... B21 Reactor Coolant System Pressure............................. B 2-3
_2. 2 LIMITING SAFETY SYSTEw SEn NGS Reacter Trio 5etpoints...................................... 3 2.A
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I LIMITING CONDITIONS FOR OPEPATION AND $URVE!LLANCE REQUIREMENTS PAGE
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} . 3/4.0 AP PL I CAB I L I T,Y,,. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-1 t
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3/41-3* t B o ron D il u t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/41-4 L Moderator Temperature Coef ficient . . . . . . . . . . . . . . . . . . . 3/415 Mi nimum Tenperature for Criticality. . . . . . . . . 6. . . . . . . 3/4 1-7 3/4.1.2 BORATION SYSTEMS
; Fl ow Pa th s - S hu tdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/41-8 '
Fl ow Pa ths - Opera ting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/41-g Charging Puso - 5hutdown............................. 3/4 1-10 i Charging Pumps - Operating........................... 3/4 1-11 l Boric Acid Pumps - Shutdown.......................... 3/4 1-12 Boric Acid Pumps - Operating......................... 3/4 1-13 4 Borated Water Sources - 5hutdown..................... 3/4 1-14 Borated Water sourc es - Opera ting . . . . . . . . . . . . . . . . . . . . 3/4 1-16
. 3/4.1.3 MOVABLE CONTROL ASSE2 LIES Full Length CEA Position. . . . . . . . . . . . . . . . . . . . . . . . . . . .i 3/4 1-17 i .
Position Indicator Channels.......................... 3 /4 1 - 21 CIA Drop T1me...............n........................ 3/4 1-23 Shutdown CEA Insertion Limitl....................... 3/4 1-24 Aegul a ting CEA Insertion L1mits . . . . . . . . . . . . . . . . . . . . . . 3/4 1-25 t
!!! Anandment No. 32 CALVERT CLIFFS - UNIT 1 ... .. .
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U. 6 i TA6Lt N M @DT9_ID t , i LIMITING CONDITIONS FOR OPERATION AND SURVE!LLANCE RE0V!REMENTS ; r SECTION PAGE, 3/4.2 POWER O!STRIBUTION LIMITS { 3/4.2.1 L I NEAR HEAT RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2 1 i 3/4.2.2 TOTAL PLANAR RADIAL PEAK!NG FACT 0R.................... 3/4 2-6 I 3/4.2.3 TOTAL INTEGRATED RADIAL PEAK!NG FACTOR. . . . . . . . . . . . . . . . . 3/4 2 9 j 3/4.2.4 A21MUTHAL POWER T1LT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 2-12 3 i 3/4.2.5 DKB PARMETE RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 2 -13 l 3/4.3 INSTRUMENTATION j 3/4.3.1 REACTOR PROTECTIVE INSTRUMEKTAT!0N. . . . . . . . . . . . . . . . . . . . 3/4. 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM J INST RUMENT ATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3 10 3/4.3.3 MONITORING INSTRUMEKTATION I Radiation Monitoring Instrumentation. . .. . .. . .. . . .. . . .. 3/4 3 25 8 l
. I nco re De t e c to rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 3 2 9 Seismi c Instrumentati on. . . . . . e. . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-31 Meteomlogica l Ins trumenta tion. . . . . . . . . . . . . . . . . . . . . . . . 3/4 3 34 L
Remo te Shtttdown Ins trumenta tion. . . . . . . . . . . . . . . . . . . . . . . 3/4 3-37 i Post Accident Instrumentation......................... 3/4 3-40 L Fire De tection Instrumenta tion. . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-43 Radioactive Gaseous Effluent Monitoring l Instrumentation...............c...../..!.li......... 3/4 3-48 ( j Radioactive Liquid Effluent Monitoring
- Instrumentation..................................... 3/4 3-53 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION................. 3/4 4-1 S ta rtu p a n d Powe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4 -1 Hot S ta n dby . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 4- 2 S hutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 2 a 3/4.4.2 SAFETY YAL VES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 - 3 3/4.4.3 REL I E F VAL V ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 4-4 )
CALVERT CLIFFS - UNIT 1 IV Amendment No. 39.53.55.95./ # 9 105
$ 6 L l[ D k' D M f O D
-aatr LIMITING CONDITIONS FOR OPERAT!0N AND SURVE!LLANCE REOUIREMENTS SECTION PAGE 3/4.4.4 PRESSURIZER.............................................. 3/4 4 5 3/4.4.5 STEAM GENERATORS......................................... 3/4 4 6 3/4.4.6 RE. ACTOR COOLANT .tYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4 13 Reactor Coolant System Leakage........................... 3/4 4-14 3/4.4.7 CHEMISTRY................................................ 3/4 4-16 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-19 3/4.4.9 PRESSURE / TEMPERATURE LIMITS .
Reactor Coolant System................................... 3/4 4-23 Pressurizer............,.................................. 3/4 4-26 Overpres sure Protection Systems . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-26a 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Cl as s 1, 2 and 3 Components. . . . . . . . . . . . . . . . . . . . 3/4 4-27 3/4.4.11 CORE BARREL M0VEMENT..................................... 3/4 4-29 3/4.4.12_ LETDOWN LINE EXCESS FL0W. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31 3/4.4.13 REACTOR COOLANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2 ECCS_ SUBSYSTEMS - Tavg 1 300'F........................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 300* F. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-6 3/4 S.4- RE FUEL I N G WAT E R - T AN K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 5- 7 l l CALVERT CLIFFS - UNIT 1 V Amendment No. 34,119 l
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TA6Lt CA~ CONtT M LIMITING CONDITIONS FOR OPERATION AND SURVE!LLANCE REQUIREMENTS SECTION PAjE, 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT C on ta i nme n t I n teg ri ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6 1 Co n ta i nme n t L e a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 6 2 Conta i nment Ai r Loc ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 ... 64 Internal Pressure....................................... 3/4 4as Ai r Tempe ra tu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6. / Conta inment Struc tura l Integrity. . . . . . . . . . . . . . . . , . . . . . . . 3/4 C-8 Co n ta i nme n t Pu rg e Sy s t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/> r, 9 d Con t a i nme n t Ven t Sys t.em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3/4 % 9e 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Conta i nment Spray Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-10 Conta inment Cooling Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-12 3/4.6.3 IODINE REMOVAL SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . .3/4 . . .6. 13 3/4.6.4 CONTAINMENT ISOLATION VALVES............................ 3/4 6 17 3/4.6.5 COMBUSTIBLE GAS CONTROL Hyd ro g e n An a l y z e rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 6 2 6 Electric Hydrogen Recombiners - W....................... 3/4 6-27 3/4.6.6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSilM.3/4 ....6..28 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves........................................... 3/4 7-1 Auxilia ry Feedwa ter Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-5 i Condens a te Stora ge Tank. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 7-6 ' L A c t i v i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ._ . . . 3/ 4 7- 7 Main Steam Line Isolation Va1ves........................ 3/4 7-9 I 4 CALVERT CLlfr5 - UN F 1 VI Amendment No. S $
u - - - - - - - - - - - - - - - - - - - - - - - - _ _ - _ __ Q LIMITING CONCITIONS FCR OPERAT!0N AND SURV!!LLANCE REOUIR!ugy;$ PAGE SECTION
- 3/4 7 13 I 3/4.7.2 STEAM GENERATOR PRE!SURE/ TEMPERATURE L:M!TATION. .. ....
3/4 7 14
, 3/4.7.3 COMPCNENT COOLING WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . .
3/4 < 1 i-3/,.7.4 ....-. S c s,.,. -..: .-.t- e.,S.:o..................................
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3 / 4 7 - 16 3/4.7.5 S AL T W AT E R S Y ST EM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7 17 3/4.7.6 CONTROL ROOM EMER3ENCY VENTILAT
- 0N SYSTEM. 3/4 7-21 3/4.7.7 ICCS PUMP ROCH EXKAUST AIR FILTRATION SYSTEM.........
3/4 7 25 3/4.7.8 SNUBBERS............................................. 3/4 7 63 3/4.7.9 SEALID SOURCE CONTAMINAT;0N.......................... 3/4 7 65 3/4.7.10 WATERTIGHT DC0RS..................................... 3/4.7.11 FIRE SUFPRESSION SYSTEMS 3/4 7-66 Fire Suppressien Water System........................ 3/4 7 69 Spray and/or Sprinkler Systems....................... 3/4 7-72 Halen System......................................... l 3/4 7 73 i Fire hose Stations................................... 3/4 7 75 I Yard Fire Hycrants and Hycrant Hese Houses........... 3/4 7-77 3/4.7.12 PENETRATION FIRE BARRIERS............................ . . I"- 3/4.8 ELE TR: CAL POWER SYSTEMS .- 3/4.8.1 A.C. SOURCES 3/4 6-1 Operating............................................ 3/4 6-5 Shutdown............................................. 3/4.5.2 ONSITE POWER C157R:BUT 0N SYSTEMS Operating........................ 3/4 B-6 H A.C. Districutien - 3/4 6-7 A.C. Distribution - Shutdown......................... 3/4 6-8 0.C. Di stribution 0;erating. . . . . . . . . . . . . . . . . . . . . . . . 3/4 6 11 D.C. Distribution - Shutdown......................... t VII Amencment No. 26. 57 f 4, 99, ;;3 CALVERT CLIFFS - UNIT i l O
~ g1 CO OTt.ON
, ,a' I LIMITING j_ CON 0!TIONS ::t OFEUTION ANC SUR'/EILLANCE RE%
'SECTION DAGE
,,3/4.9 .
RErt!~ N OFE:.ATIONS 3/4.9.1 BCRON 3/4.9.2 CONCENTRATION..................................... 3/4 9-1 3/4.9.3 INSTRUMENTATION....................................... DECAYT!ME.............................................3/492 J/4.9.4 CONTAINMENT . 3/4 9 3 3/4.9.5 FENETRAi!0NS............................... 3/4 9 1 COMMUNICATIONS......................................... . 3/4 9-5 3/4.9.6 REFUELING MACNINE 3/4.9.7 0PERASILITY........................... 3/4 9-5 3/4.9.6 CRANE TPAVEL - SPENT FUEL STORAGE P00L BUILOING. 3/4 9-7 ........ 3/4.9.9 SHUT 00WN COOLING ANO COOLANT CIRCULATION. 3/ 4 9 6 ........... CONTAINMENT PURGE VALVE !$0LAT10N SYSTEM................ 3/4 9 9 3/4.9.10 WATER LEVEL 3/4.9.11 REACTOR VESS EL. . . . . . . . . . . . . . . . . . 3/ . .4. 9. .10. . . . . SPENT FUEL FOOL WATER 3/4.9.12 LEVEL............................. 3/4 9 11 SPENT FUEL POOL VENT!LATION 3/4.9.13 SYSTEM...................... 3/4 9 12 3/4.9.14 SP ENT FUEL CAS K KANCL ING CRANE . . . .3/4
. .9.16 .. ...........
CONTAINMENT VENT !$0LATION VALVES....................... 3/4 9 17 3/4.10 SPECIAL TEST Ex Ec?!CNS l3/4.10.1 S HUTOOWN PARG I N . . . . . . . . . . . 3/4.10.2 MOCE?ATOR . . . . . . . .3 ./ 4. .10. i.. . . . . . . . . . . . . TEMPERATURE COEFFICIENT, CEA INS 3/4.10.3 AND POWER DISTRIBUTION LIMITS.............ERi!CN ........ 3 / 4 10-2 3/4.10.4 N O. F L O W T E S T 5 . . . . . . . . . . . . . . . . . . . 3/4. .10-3 CENTER CEA H!$ ALIGNMENT,.........................'....... 3/4 10 ' 3/4.10.5 COOL ANT C I RCULAT ION . . . . . . . . . . . . . . . 3. /. 4. 10-! ...............
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eCALVERT CLIFFS UNIT 1 l VIII Amencment No. 25, !!. g i _ _ _ _ _ _ _ _ _ _ - _ - - - - - - - " ^ ~
magLt of CDrTrearTT5-tmE LIMITING CONDITIONS FOR OPER ATION AND SURVETLL ANCE REOUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LlQUID EFFLUENTS Concen tra tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4111 Dose........................................ 3/4 11-3 Liquid Radwaste Treatment System ................. 3/4 11-6 3/4.11.2 GASEOUS EFFLUENTS Dose Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 114 Dose-Noble Gase s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/41111 Dose-lodine 131, and Radionuclides in Particulate Form ................ 3/4 11-12 Gaseous Radwaste Treatment . . . . . . . . . . . . . . . . . . . . . . 3/41113 Explosive Gas Mixture . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11 14 Gas Storage Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/41113 l 1 3/4.11.3 SOLID RADIO ACTIVE W ASTE . . . . . . . . . . . . . . . . . . . . . 3/4 11-16 3/4.11.4 TOT A L D OS E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-17 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 M ONITORING PR OGR AM . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-1 l 3/4.12.2 LAND USE CF.NSUS ........................... 3/4 12 12 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM . . . . . . . 3/4 12-13 CALVERT CLIFFS UNIT 1 VW N Amendment No. 100.105 gv.
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- lo-ble. of Con 4er1IS BASES SECTION PAGE 3/4.0 APPLICABILITY .............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL ......................................... B 3/4 1-1 3/ 4.1.2 BORATION SYSTEMS ......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ............................... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE ......................................... B 3/4 2-1 3/4.2.2, 3/4.2v3, and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTOR'S - FT
( POWER TILT - T 4..........M.ANDF{TANQAZINUTHAL
..... ..................... B 3/4 2-1 3/4.2.5 DNB PARAMETERS ................................. ......... B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES INSTRUMENTATION ............................... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 3-1 e.>
CALVERT CLIFFS - UNIT 1 + Amendment No. 27, 55 ( C'LVEF ?_i m "6 5- I h -- r.; 5 . ;, 33 .
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g .c Th6L.L Ch g-COM8ON BASES SECTION PAGE 3/4.4 REACTOR COOLANT-SYSTEM 3/4.4.1 COOLANT. LOOPS AND COOLANT CIRCULATION..................... B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-1
] 3/4.4.3 RELIEF VALVES............................................. B-3/4 4-2 3/4.4.4 PRE SSU RI Z ER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 4- 2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2
- 3/4.4.6 REACTOR COOLANT - SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.7 CHEMISTRY................................................. B 3/4 4-4 3/4.4.8 SP ECI FIC ACTIVITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.9- PRESSURE / TEMPERATURE LIMITS...............................-B 3/4 4 .
3/4.4.10 STRUCTURAL INTEGRITY...................................... B 3/4 4-12 3/4.4.11 CORE BARREL H0VEMENT...................................... B 3/4 4-12 l [ 3/4.4.12 LETDOWN LINE-EXCESS FL0W ................................. B 3/4 4-12 1 3/4.4'.13 REACTOR COOLANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-13 3/4'. 5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SATFTY INJECTION TANKS.................'................... B 3/4 5-1 3/4-5.2 and-3/4.5.3' ECCS-SUBSYSTEMS.....<.......................... B 3/4 5-1 3/4.5.4- REFUEL ING WATER TANK ( RWT ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B - 3/4 5- 2, 4 4 CALVERT CLlfrS - UNIT 1 ~JC Amendnent No. ;?,EE, lb
g iAb ,
" O@hID i BASES SECTION PAGE 3/4.6 CONTAINME'4T SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.................................. B 3/4 6-1 3/4.6.2 DEPRESSUR!2ATION AND COOLING SYSTEMS,................ B 3/4 6-3 2/4.6.3 IODINE REMOVAL SYSTEM................................. B 3/4 6-3 3/4.6.4 CONTAINMENT ISCLATION VALVES......................... B 3/4 6-3 3/4.6.5 COMBUSTIBLE GAS CONTR0L.............................. B 3/4 6-4 3/4.6.6 PENETPATION ROOM EXHAUST AIR FILTRATION SYSTEM,...... B 3/a 6-4 CALVERT CLIFFS - UNIT I X1f l
l 6A6Lt- @ ,_ CDF'TTEJYT.S
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p fIBASES 1 j$ECTION i . PAGE ha,7~ D'.4N7 $Y$*[V$ . l3/4.7.1 IURBINE CYCLE........................................ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMI 3/4.7.3- COMFONENT COOLING WA7ER SYSTEM....................... B 3/4 7-4 3/4.'7.4 S ERVICE WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . 3/4.7.5 SALT. WATER SYSTEM..................................... B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENQY VENTILATION SYSTEM.......... B 3/4-7-4 l.3/4.7.7-ECCS PUMP-ROOM EXHAUST AIR FILTRA7!ON BSYST . . . . , M.. 3/4 7-4
'3/ 4. 7. S'~
SNUBBEIS............................................. B 3/4 7 5 3/4_. 7. 9 = SEALED SOURCE CONTAMINATION. . . . . . . . . . . . . . . . . .
- 3/4.7.10 -WATERTIGHT D00RS..................................... B 3/4 7 6 3/4. 7.11 FIRE SUPPRESSION SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . , , . . B 3/4 3/4.7.12. PENETRATION FIRE BARRIERS...........................,
B 3/4 7 7 J 3/4.8 , EL E CTRI CAL POWER SYSTESS . . . . . . . . . . . . . . . . . . . . . . 3/4.9 REFUELING OPERAT!ONS-3/4.9.1-o. BORON CONCENTRATION. . . . . . . . . . . . . . . . . . . . . . . . , , . . . . . . . B 3/4 g-3/4.9.2 INSTRUMEN"ATION......................................_B 3/4 g-1' g . 3/449.3 OECAY O
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TIME........................................... B 3/4 9-1 13/4.9.4 . CONTAINMENT PENETRATIONS............................. B 3/4 9-1. L
- ' CALVERT* CLIFFS 1 -
' UNIT'l' XII i Amencment No; 21,,?9,n3
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TA6Lt_ DF cD3TtlTr5
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dkttr BASES . SECTION P.L.E 3/4.9.5 COMMUNICATIONS........................................ B 3/4 9-1 3 /4'. 9. 6 REFUELING IMCHINE OPERABILITY. . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 , 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING. . . . . . . . . .! . B 3/4 9-2 3/4.9.B SHUTDOWN COOLING AND COOLANT CIRCULATION. . . . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM. . . . . . . . . . . . . . B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL. . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.9.12 SPENT FUEL "GL W.NTILATION SYSTEM.................... B 3/4 9-3 3/4.9.13 W hT FUEL CASK HANDLING CRANE........................ B 3/4 9-3 3/4.9.14 CONTAINMENT VENT ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....................................... B 3/4 10 1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION L I MI T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 10- 1 3/4.10.3 NO FLOW TEST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 10- 1 3/4.10.4 CENTER CEA MI SAL IGNMENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 10-1 3/4.10.5 COO LANT C I RCULAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 10- 1
~
f' 3/4.11 RADIOAC IVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......s............................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS..................................... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-5 3/4.11.4 TOTAL 00SE............................................ B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM............................'........ B 3/4 12-1 3/4.12.2 LAND USE CENSUS....................................... B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM. . . . . . . . . . . . . . . . . . . . B 3/4 12-2 CALVERT CLIFFS - UNIT 1 -k!ff Amendment No. 26,55,88,700, 105
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-96Lt -Bbt** -
cf CorritfsTS I DESIGN FEATURES SECTION PAGE 5.1 SITE . Map Defining the Site Boundary and Effluent Re l e a s e P o i n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 Low Po p ul a t i o n Zo ne . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1
- 52. CONTAINMENT Configuration................................................. 5-1 Design Pressure and Temperature............................... 5-4
. 5.3 REACTOR CCRE F u e l As s e mb l i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 4 ~
Co nt rol El eme n t As s embl i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature............................... 5-4 ) Vo1ume........................................................ 5-5 5.5 METEOROLOGICAL TOWER LOCATION................................. 5-5 5.6 FUEL STORAGE
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- C r i t i c a l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 Drainage...................................................... 5-5 Capacity....................................................... 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 CALVERT CLIFFS - UNIT 1 Amendment No. 100 1
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g'Q 6 mWyT}
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ADP!iNISTRATIVE CONTROLS SECTION g 6.1 RESPONSIBIL ITY 61 6.2 ORRANIZATION 6.2.1 DNSITE & OFFSITE ORGANIZATIONS 61 6.2.2 UNIT STAFF 61 6.3 FACILITY STAFF OUALIFICATIONS 66
,6.4 TRAINING 66 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS AND SAFETY REVIEW ComfTTEE (POSRC)
Function 66 Composition 66 Alternates 66 i Heeting Frequency 67 Quorum 67 , Responsibilities t 67 Authority 68 Records l 6-8 l 6.5.2 0FF-SITE SAFETY REVIEW Com ITTEE (OSSRC) l 1 i Function 68 : ! Composition 1 69 l Alternates 69 Consultants 69 Heeting Frequency 69 l Quorum 69 Review 6 10 Audits 6 11 l Authority 6-12 l Records 6 12 l 1 4 CALVERT CLIFFS - UillT 1 XVL Amendment No JE, Jff, 131 l j l l
W ADMfNISTPATIVE CONTROL $ PAGE - ! SECTION-6-12 6.6 REPORTABLE EVENT ACT10N.................................... 6- 13 6.7 S AF E T Y L IMIT V 10L AT 10 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 13 6.8 PROCEDURES................................................. 6.9 REPORTING REQUIREMENTS _ 6.9.1 ROUTINE REP 0RTS...........................................-6-14 6 18 6.9.2 S P E C I AL RE P 0 RT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19 6'.10 RECORD RETENT10N,.......................................... PROGRAM..,............................ 6-20 6.11 RADIAT10N' PROTECTION 6- 20 6.12 HI GH R AD I AT I ON AR E A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 6.13 SYSTEM INTEGRITY........................................... 6 6.14 IODINE MONITORING.......................................... 6- 21 6.15 ' PO ST ACC I DE NT S AM P L I N 3. .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l 6- 2 2 [; 6.16 P ROC E S S CONTRO L P R0G RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.17 0FFSITE DOSE CALCUL ATION MANUAL . . . . . . . . . . . . . . . . . . . .l . . . . . . . 6.18 MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS AND SOL I O W A5T E T RE ATMENT SY ST EM5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,l6- 23 Amendment No. 29,53,9#,705,795, 113 CALVERT CLIFFS - UNIT 1 XVIL
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e SECTION 1.0 DEFINITIONS 1
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M OLLA - 1.0 DEFIN!TIONS ~~~ NLULT nwyb 1 DEFINED TER.*5 t {.\ i b he DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. THERMAL F0WER I ,3'3 MF THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
~
RATED THERMAL POWER l ,;;L5 35' RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt. OPERATIONAL MCCE
- l. &O cr An OPERATIONAL MODE shall correspond to any one inclusive comoination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.
ACTION L, g M < ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
+4 OPERABLE - OPERAB:LITY 1.l4 t Xr system. subsystem, tra'n, i component or device shall-be OPERABLE or have OPERABILITY when it is capable of perfoming its specified function (s).- Implicit in this definition shall be the assumption that all necessary attendant instrtmentation, controls, nomal and emergency e electrical power sources, cooling or-seal water, lubrication or other required auxiliary equipment that are required for the system, sub-system, train, component er device to perfom its function (s) are also capaele of performing tneir related support function (s). .
CALVERT CLIFFS-UNIT ; 11 Amendment No. 84 100 w-- ,-, - ,. - - - - - - , . , - - . ,
i DEFINITION 5 t& M REPORTABLE EVENT l . &'l -h1 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. CONTAINMENT INTEGR!TY 1.0 1-8 CONTAINMENT INTEGRITY shall exist when: 1.8.1 All penetrations recuired to be closed during accident conditions are either:
- a. Cacable of being closed by an OPERABLE containment automatic isolation valve system, or l! b. Clesed by manual valves, blind flanges, or deactivates ll automatic valves secured in their closed positions, ,.
l: exceot as providee in Table 3.5-1 of Specification ij 3.6.4.1. j 1.8.2 All etuipment hatenes are closed and sealec. lI lj 1.8.3 Eacn airlock is OPERASLE pursuant to Soecification 3.6.1.3, I 1.B.4 The centainment leakage rates are witnin the limits of
, Specifica:1cn 3.6.1.2, and
.i
;l 1.8.5 The sealing mechanism associatec with each cenetration i' (e.g., weles, cellows or 0-rings) is OPERABLE.
. ,_ l
! CHANNEL CALIERATION
- 15 1:sr A CHANNEL CALIBRATION shall be the adjusement, as nece1sary, of the channel output suen that it rescends witn the necessary range and accura:y to known values of the parameter which tne channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor Cnd alam and/or trio functions, and shall incluce the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be cerfomed by any series of secuen-tial, overlapping or total enannel stees such that the entire enannel is i calibrated.
.l .
4 i . 4 1 CALVERT CU F 5-UNIT 1 1-2 Amencment No. 2&', 94 I
DEFINITIONS M d1LA CHANNEL CHECK 4:iw A CHANNEL CHECK shall be the qualitative assessment of channel Q 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 in.'ependent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST l.7 MT A CHANNEL FUNCTIONAL TEST shall be: .
- a. Analog channels - the injection of a simulated signal into the channel as close to the primary senscr as. practicable to verify OPERABILITY including alarm and/cr trip functions.
- b. Bistable channels - the injection of a simuitted signal into the channel sensor to verify OPERABILITY including alarm and/cr trip functions.
CORE ALTERATION b (O h=l=2 CORE ALTERATION shall be the movement or manipulation of any component pithin the reactor pressure vessel with the vessel head removed and fuel in the vessel . Suspension of CORE ALTERAT:0N snall not preclude completion of movement of a component to a safe conservative position. SHUTDOWNMARGH h44- SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by 1 9% which the reactor is suberitical or would be suberitical from its present condition assuming all full length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully wi,thdrawn. 00 1-3 Amendment No. CALVERT CLIFFS-UNIT 1
i DEFINITIONS
&{g.
W , IDENTIFIED LEAGGE . l.i(t 4:14
- IDENTIFIED LEAKAGE shall be:
- a. Leakage (exceptCONTROLLEDLEAXAGE)ini.oclosedsystems,such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAXAGE, or
- c. Reactor coolant system leakage through a steam generator to the secondary system.
UNIDENTIFIED LEAKAGE ION -h=k3 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAXAGE or CONTROLLED LEAXAGE. I PRESSURE BOUNDARY LEAKAGE l'aQ iHB PRESSURE BOUNDARY LEAKAGE shall bE leakage (exce0t steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. CONTROLLED LEAKAGE IM 4:.:N > CONTROLLED LEAXAGE shall be the water flow from the reactor coolant pump seals. AZIMUTHAL POWER TILT - T o
- 1. 4- .b:M- AZIMUTHAL POWER TILT shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower)~ of the core.
DOSE ECUIVALENT I-131 WT:iL DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (uC1/ gram) 1.l( which alone would produce the same thyroid dose as the quantit? and isotopic
+ mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The '
thyroid dose conversion factors used for this calculation snall be those listed in Table .Ill of TID-14844, " Calculation cf Distance Factors for Power and Test Reactor Sites " CALVERT CLIFFS - UNIT 1 14
.j.
. DEFINITIONS W \=
- T - AVERAGE 01SINTEGRAT10N ENERGY
- l .'l & Wf shall be the average (weighted in proportion to the coni:entration of each radionuclide in- the reactor coolant at the time of sampling) of the sum of the average. beta and gama energies per disintegration (in MEY) for isotopes, other than todines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
STAG 3ERED TEST BASIS j ,ba -3zt1 _R STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train or other designated
- comoonent at the beginning of each subinterval.
FREQUENCY NOTATION 1N 4??? 'The FREQUENCY NOTATION specified for the perfomance of Surveillance Recuirements shall c:rrespond to the intervals defined in Table 1.2. ' AXIAL SHAPE INDEX b) ., hd5 The ~ AXIAL SHAPE INDEX (Y ) is the power level detected by. the lower excere nuclear instrument detebtors (L) less the power level detected by the- upper excore nuclear instrument detectors (U) divided by the sum of
~
these power level s.- The AXIAL SHAPE INDEX (Y,) used for the trip and _ pretrip signals in the reactor protection system is the above value -(Y ) modified by an appropriate multiplier (A) and a constant (B) to detemine the true core axial power distribution for that channel. Yg g Yg = AYg *B UNRODDED PLANAR RADIAL PEAXING FACTOR - F ,
- 1.37 ~
tin The UNRODDED PLANAR! RADIAL PEAKING FACTOR is the maximum ratio of-the peak to average power density of the individual fuel rods _i,n any of the unrodded horizogtg planes, excluding tilt. - ! CALVERT CLIFFS - UNIT 1 15 Amencment No. 21
CEFINITIONS hM REACT 0r. TRIP SYSTEM RESPONSE TIME l .Gh
-M5 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time intervai from when the renitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.
ENGINEERED SAFETY FEATURE RESPONSE TIME 1,g 3
-h26 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interva from when the monitored parameter exceeds its ESF actuation setpoint at the I channel sensor until tne ESF equipment is capable of perfoming its safety function (i.e., the valves travel to their required positions, pumo discharge pressures reach their required values, etc. ), Times shall include diesel generator starting and sequence loading delays where applicable.
PHYSICS TESTS .
.),;)) lG PHYSICS TES"$shall be those tests performed to measure the fundamental
, nuclear characteristics of the reactor core and related instrumentation anc
- 1) descrited in Chapter 13.0 of the FSAR, 2) authorized under the provisions ij of 10 CFR 50.59, or 3T otherwise aporoved by the Commission, fUHRODDE3INTEGTsATEDRAOIALPEAKINGFACTOR-F,.
,1 jg -hi!L The UNR00DED INTEGRATED RADIAL PEAKING FACTOR is the ratto cf the peak pin power to the aversge pin poner in an unroeded core, excluding tilt. )
SASEOUS cACWASTE bEA'" MENT 3YSTE'i ! 1.t5 l-t:$t A GASIOUS RADWASTI TREATMENT SYSTEM is any system designed and installed
.l cffgases from tne primary system anc providing for delay or holdu 4;emer,:.pu" pose of reducing tne total radioactivity prior to release to the environ. .
l ! ! MEMBER (S1 0F TME PUBLIC \ .
~
117 430 ' MEMBER ($) 0F THE PUBLIC shall include all persons who are not occuca~- i g tionally associated with tne plant. This category does not include employees !
; of ne utility, its contractors or vendors. Also excluded from this category
; are eersons wno enter the site to service equipment or to make celiveries.
L
~l' OF75:TE DOSE CALCULATION MANUAL (00CM) '
G MT'HMT1 (The OFFSITE DOSE CALCULATION MANUAL shall contain the curr M ology and parameters used in the calculation of offsite doses cde to ract:- , ijactivegaseousandliquideffluents,inthecalculation-ofgaseousandlicuic L affluen monitoring alarm / trip setooints, and in the conduct of the environ-
! mental radiological monitoring program, ! -
CAL'/ERT CLIFFS '; NIT 1 1-6 i Amendmen; No. 21, .*;;
Nes) wec DEFINRT80NS ( PROCESS CONTROL PROGRAM (PCP) I .D3 *
-hse- The PROCESS CONTROL PROGRAM shall contain the current fonnula, sampling, analyses. tests, and determinations to be made to ensure that the processing
' and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 71 and Federal and State and local regulations governing the disposal of the radioactive weste, PURGE - PURGING I94
-hsa- PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement, SITE BOUNDARY
/.G . *
'.A"i- The SITE BOUNDARY shall be that line beyond which the land is neither-owned, nor leased, nor otherwise controlled by the licensee.
SOLIDIFICATION E.30 4,45- SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK I,3I
+:46-- A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
" UNRESTRICTED AREA
/ , ay
- 1. ; An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY
- access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
H VENTILATION EX'AUST TREATMENT SYSTEM
/. a
-hse- A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in perticulate j j
form in effluents by passing ventilation or vent exhaust gases through char- , coal adsorbers and/or HEPA filters for the purpose of removing iodines or ; particulates from the gaseous exhaust stream prior to the release to the ' environment. Such a system is not considered to have any effect'en ncble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanuo systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM comoonents. . 1-7 Amenement No.100 I,CAtvERTCLIFFS-uNI _ _ _ _ . _ _ _ _ _ _ ~ - - - ~ l
( Si j fatICFINITIONS ( VENTING . i
.&39, VENTING is the controlled process of discharging air or gas from a
\off confinement to maintain temperature, pressure, humidity, concentration or i
etner operating concition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. i
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ll l i! " ll I: . 1 11 I; ji . ii I; CALVERT C!.!FF5 - UNIT 1 1-8 Amendment No.100 l
.. _. __ . _ . . , . . . . . ~ . - - . _ _ _ _ _ . - . _ . . ~ . . . _ . - _ . . - - ~ - _ . - -
T AB'.E - 1.1 . ODERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MOD,:-
. CONDITION, K,g THERMAL POWER *- TEMPERATURE 1._ POWER OPERATION- 1 0.99 > 5% 1 300'F _
- 2. STARTUP 1 0.99 15% 1 300'F
- 3. HOT STACEY < 0.99 0 1 300'F t
- 4. - . HOT SHUTOOWN < 0.99 0 300'F> T '
> 200'F avg
- 5. COLD SHUT 00Wff < 0.99 0 < 200'F
- 6. REFUELING ** 1 0.95 O < 140'F j- Excluding decay. heat.
Reacter vessel head unbolted or removed and fuel in the' vessel. l l-j .. i-l. l CALVERT CLIFFS - UNIT 1 1-9 Amenpment No. 100
-l
TABLE 1.2 FRE0VENCY NOTATION I 1 NOTA'!ON FRE0VENCY
$ At least once per 12 hours.
O At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. O At least once per 92 days. SA At least once per 6 months. R At l' east once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release, s-NrAT A/A Not applicable. iRffurlirg!$farvs1
. At least once per 24 months. )
I CALVERT CLIFFS UNIT 1 1-10 Amendment No. 199,13,
SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS t / '
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'2. 0 SAFETY' LfMITS AND LIMITING SAFETY SYSTEM SETTINGS -
2.1 SAFFTY-LIMITS ' REACTOR CORE W '
,.. + .e .. . . -
2.1.1 The combination of THERF/.L POWER, pressurizer pressure, and highest ' ' operating loop colo leg coolant temperature shall not exceed the limits shown-in Figures 2.1-1( 2.1-2, 2.1-3 and 2.1-4 for the various combinations of two, three and four reactor coolant pump operation. .. [ APPLICABILITY: MODES i and 2. .,
~' ' " - * ' ' - -
ACTION: 1 .
~~
WheneYer the point defined by the com61 nation of the hifhest operatirig L loop cold leg temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDEY within 1 hour. f l REACTOR COOLANT SYSTEM PRESSURE , i 2.1.2' The Reactor Coolant System pressure shall not exceed 2750 psia.
- j. .
APPLICABILITY: MODES 1, 2, 3, 4 and 5. .
.' ACTION: ..
MODES 1,and'2 , Whenever the Reactor Coolant System pressure has exceeded 2750 ' ' . ... p'ia, s be in HOT STANDBY with the Reactor Coolant System pressure . within its , limit within 1 hour.- , MODES 3,;4 and 5 .
.Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.
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Three React:r Coolant Putos Operatinc. CALVERT CLIFFS - UNIT 'l 23
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This page left blank pending NRC approval of ECCS analysis for two pump (opposite loops) operation 1 e t + Figure 2.1-4' Reactor Core Themal Margin Safety Limit - Two Reactor Coolant Pumps Operating Opposite Loops
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_ CALVERT CLIFFS - UNIT 1 2-5
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I SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setroints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
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CALVERT CLIFFS - UNIT 1 2-6
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TABLE 2.2-1 (Cont'd) '
!.;] RE CTOR PROTECTIVC INSTRUMENTATION TRIP SETPOINT LIMITS ~t 9 -
TRIP SETPOINT ALLOWABLE VALUES _
*;; FUNCTIONAL UNIT ,
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b 3. Reactor Coolant flow - Low (1) . .
> 95% of design reactnr coolant > 95% of design re' actor coolant " a. Four Reactor Coolant Pumps T10w with 4 pumps- operating * -
Operating Tiow with 4 pumps operating * ~ ~
> 72% of design reactor coolant > 72% of design reactor coolant
- b. Three Reactor Coolant pumps T10w with 4 pumps operating
- Operating flow with 4 pumps operating *
> 47% of design reactor coolant > 47% of des.ign reactor coolant
- c. Two Reactor Coolant Pumps Tlow with 4 pumps operating
- Operating - Same Loop , T10w with 4 pumps operating
- l 7
cn > 50% of design reactor coolant l Two reactor Coolant Pumps > 50% or design reactor coolant
- d. T10w with 4 pumps operating
- T10w with 4 pumps operating
- Operating - Opposite Loops : :
^
l l _ ,
^
Design reactor coolant flow with 4 pumps operating is 370,000 gpm.
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TABLE 2.2-1 (Cont d) 9 r-REACTOR PROTECTIVE INSTRIMIENTATION TRIP SETPOINT LIIIITS g n ALLOWABLE VALUE5_ TRIP SETPOINT I- ~ FUNCTIONAL UNIT A 1 2400 psia 1 2400 psia
- 4. Pressurizer Pressure - High Containment Pressure - liigh < 4 psig 1 4 psig E- 5.
> 635 psia 16115 psia l
- 6. Steam Generater Pressure - Low (2). _
1 10 inches below top 1 10 inches below top
- 7. Steam Generator Water Level - Low -
of feed ring. of feed ring. Trip setpoint adjusted to Trip setpoint adjusted to
- 8. Axial flux offset (3) not exceed the limit ifnes not exceed the limit lines of Figure 2.2-1. of Figure 2.2-1.
- 9. Thermal Margin / Low Pressure (1) ,
Trip setpoint adjusted to Trio setpoint adjusted to
- a. Four Reactor Coolant Pumps be not less than the larger Operating ', not exceed the limit ifnes of Figures 2.2-2 and 2.2-3. of (1) the value calculated from figures 2.2-2 and 2.2-3 and (2) 1875W psin_ . .
- b. Steam Generater Pressure i 135 psid ,1 135 psid Difference - High (1) g e
I J. Loss of Load _W OA m .5 0 A E R . E 1 2 6 decades per minute 2 11. Rate of Change of Power - fligh (4) 1 2 6 decades per minute N TABLE NOTATION
~
N
* (1) Trip may be bypassed bejow 10'4% of RATED THERMAL POWER; bypass shall be automatically removed wher.
TIIERMAL POWER is 1 10~ % of RATED TilEPJfAL POWER.
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I 4 3.5 (.0.6,1.4) 14 - 1.3 N - A3 = .0.5 x A51 + 1.10 A3 1.2 N e (.0.2.1.2), (+0.6.1.2), h A e 21 N g(,= 4.5m A$1* 0.9 10 QDb'8 * ^1
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' CALVERT CLITTS . WIT 1 2 12 Jtmendment No. 21, 28. 3f.130 l
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y 2892 x Q ,+ 17.16 x Yg 10682 - D.10 0.00 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.5 0.9 1.0 1.1 1.2 FRACTIOR OF RATED TEERNAL POWER 7100RE 2.2-3 Thermal Margim/Inw Frassure trip setpolat Part 2 (Fraction of ALTED TIERNAL 70rIR 9Persus g) CALVERT C3.!FFS - IMIT 1 2-13 { hht No. !!, 25, 39,130 6 v . ~ --- , - - - - - --- - -.e - - - - - - - -__ - - --
J BASES FOR SECTION 2.0 SAFETY LIMITS m AND - LIMITiliG SAFETY SYSTEM SETTINGS 4
- a W
l NOTE
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The sumary statements contained in this section provide the bases for the specifications of section 2.0 and are not considered a part of these technical specifications as provided in 10 CFR 50.35. s
2.1 SartTY L1M1TS BA5E5 2.1.1 RiatTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which could result in the release of fission products to the reatter coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 22.0 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Dverheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the vpper bondary of the nucleate boiling regime could result in excessive cladding temperatures because of the unset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer toefficient. DNB is not a directly measurable parameter during operation and,
- therefore, THERMAL POWER and Reactor toolant Temperature and Pressure have been reisted to DNB through the CE-1 torrelation. The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non uniform heat flux distributions. The local DNB heat flux ratta, DNBR, defined as the ratio of the heat flux that would reuse DNB it a .
particular to DNB. core location to the local heat flux, is indicative of the margin The minimum value of the CNBR during steady state operation, normal operational transients, and anticipated transients is limited to the DNB SAFDL of 3.15 in conjunction with the Ertended Statistical Combination of Uncertainties (ESCU). This DNB SAFDL assures with at least a 95 percent probability at a 95 percent confidence level that DNB will not occur. The curves of Figures 2.1-3, 2.3-2, 2.3-3 and 2.1 4 show conservative , loci for points of THERMAL POWER, Reactor Coolant System pressure and maximum told leg temperature of various pump combinations for which the DNB SAFDL is not violated for the family of axial shapes and corresponding radial peaks shown in Figure B2.1-3. The limits in Figures 2.3 1, 2.1-2, 2.1-3, and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperature is not a safety limit;.however, operation above 580'F is not possible because of the actuation of the main steam 14e safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 130% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified.in CALVERT Cl1FFS - UNIT 3 B 2-3 Amendment No. JJ//fJ////,
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5ATETY tlyll} Ba5[$ Table 2.3. l. 'The area of safe operation is below and to the lef t of these lines. The conditions for the Thermal tiargin Safety Limit curves in Figures 2.1-1, 2.3 2, 2.3 3, and 2.J 4 to be valid are shown' en the figures. The reactor protective system in combination with the limiting Conditions for Operation, fs desiggd to prevent any anticipated combination of transier,t 8' conditions for peactor toolant bystem temperature, pressure, and THERML POWIR level that would result in a DNBR of less than 1.15, in conjunction with the 15C0 methodology, and preclude the existence of flow instabilitiet. 2.1.2 REACTOR COOLANT SYSTEM PRES $URE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressuritation and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and pressurizer are designed to Section 111, JS67 Idition, of the ASHI Code for Nuclear Power Plant Cor00nents which permits a maximum transient pressure of 230% The Reactor Coolant System piping, valves and(2750 fittings, arepsia) of design designed pressure. to ANS! B 33.7, Class I,1969 Edition, which permits a za.ximum iranstaat prAssure of 110% (7750 psia) of component design pressure. The Safety Limit of 2750 psia is, therefore, consistent with the design criteria and associated code requirements. Ji&S The entire Reactor Coolant System is hydrotasted at 4219 psia to demonstrate integrity prior to initial operation. CAtVERT CLIFFS - UNIT 1 B23 Amendment No. JJ//J)//M , 7U/JP/JE/M,130
' , 2. 2 LIM'Tlt4G SAFETY SYSTEM SITTit<GS BASIS 2.2.1 RE! TO: TRIP STTF01t.'TS The Reacter Trip 5etpoints specified in Table 2.21 are the values at which the Reactor Trips are set for each parameter. The Tri Setpoints have been selected to ensure that the reactor core andbactor colant 4-5 system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip 5etpoint but within its speci-fitd A110watie Yalue is acct; table on the basis that the difference tetween the trip set;cint and the Allomble Value is equel to or less than the drift allowance assumed for each trip in the safety analyses.
Manual Rea:ter irit The Manual Reactor Trip is a ttoundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. power level-Hich The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thereal Margin / Low Pressure trip. The power Level-High trip setpoint is operator adjustable and can be g set no higher than 10% above the indicated THERyAL POWER level. Operator pg action is required to increase the trip setpoint as THERMAL POWER is g increased. The trip setpoint is automatically decreased as THEWAL M OWEV.- T decreases. The trip setpoint has a maximum value of 107.0% of RATED THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 110t of RATED l THEPyAL POWER, which is the value used in the safety analyses. Reactor Coolant Flow-tow The Reactor Coolant Flow-Low trip provides core protection to prevent DNS in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to pemit I CALVERT CLIFFS - UNIT 1 B 2-4 '- Amendment No. AG, n l
L101 TING SAFETY SY5 TEM SETTIN35 BASES operatien of the reactor at reduced power if one or two reactor coolant put;s are taken out of service. The low flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above the DNB SAFDL of 1.15, in cohjunction with the ESCU methodolcgy, under normal operation and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant flow Low trip setpoints, the .*ower Level High trip setpoints, and the 7hermal Margin / Low pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two or three pump position. Changing these trip sepoints during t and three pum; operation prevents the minimum value of DNER from going bel B SAFDL of % 1.35, in conjunction with the ISCU methodology, during normal cperational transients and anticipated transients when only two or three reactor coolant pumps are operating, pressur ker Pressure Hieh The Pressurizer Pressitte-High trip, backed up by the , pressurizer code Jafety valves and main steam line safety valves, provides Seactor(ceolant -
'pystem protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lif t setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power operated relief valves avoids the undesirable rperation of the pressurizer code safety valves.
Cor.tainment Pressure-Nich The Containment Pressure High trip provides assurance that a reactor trip is initiated prior to, or at least concurrently with, a safety injection. Steam Generator pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant- The setting of 685 psia is sufficiently below the full-load operating point of 850 psia so as not to interfere with nomal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of i 85 psi which was based on the main steam line break event inside containment. CALVERT CLIFFS - UNIT 1 B25 Amendment No. 73//fE//7J, 111/117139/19/95,1so
_ . - _. w. - __ _ __ _ ._ __ _ .__ _ _ _ _ _ _ _ _ - - _ _ _ _ . _ - _ _ . _ _ _ . _ LIMITING SArETY SYSTIM 5LTT1HG$ BASE 5 . i steam Gererator Vater level The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adeogate heal removal capacity ahd assures that the / pressure of the Venetor toolant 'gestem will not exceed its Safety Limit. The x specified setpoint in combination with the auxiliary feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following a loss of main feedwater flow event. Avial Fluy OUset The axial flux offset trip is provided to ensu're that excessive axial peaking will not cause fuel damage. The axial flux offset is determined from the axially split excore detectors. The trip setpoints ensure that neither a DNER of less than the DNB SAFDL of 1.15, in conjunction with ESCU methodology nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip setpoints were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the
- uncertainty associated with the excore to incore axial flux offset relationship.
Thermal Marcin/ Low Pressure I The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than the DNB SAFDL of 1.15, in conjunction with ESCU methodology. The trip is initiated whenever the bactor bolant hstem pressure signal drops below either 1875 psia or a connted value as described below, wh is higher. The computed value is a function of the higher of a T power or neutron power, reactor inlet temperature, and the number of reactor coolant pumps operating. The minimum value of reactor coolant flow rate, the maximum AIIMtrrHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function'. In addition, CEA group sequencing in accordance with Specifications 3..l.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level High trip is assumed. I I CALVERT CLIFFS - UNIT 1 B26 Amendment No. JJ//JJ///E, 711/til/19/J9/Ri,130
. i LIMITfNG SAFETV SYSTEM SETTINGS BASES l The Thermal Margin / Low Pressure trip setpoints include allowances for equipment response time, measurement uncertainties, processing error and a further allowance of 40 psia to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
Asymmetric Steam Generator Transient Protection Trio Functicn ( ASGTPTF) i The ASGTPTF utilizes steam generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trip for those Anticipated Operational Occurrences associated with secondary system malfunctions which result in asymmetric primary loop coolant temperatures. The most limiting event is the loss of load to one steam generator caused by a single Main Steam Isolation Valve closure. The equipment trip setpoint and allowable values are calculated to account for instrument. uncertainties, and will ensure a trip at or before reaching the analysis setpoint. Loss of Load A Loss of Load trip causes a direct reactor trip when operating above 15% of RATED THERitAC POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lif ting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability is required to enhance overall plant equipment service life and reliability, i Rate of Change of Power-High The Rate of Chenge of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for l operation of this trip. Its functional capability st the specified trip l setting is required to enhance the overall reliability of the Reactor j Protection System. I l l CALVERT CLIFFS . UNIT 1 B27 Amendment No. 27, 32, 39, AB, 7/,, 3 3 1
SECTIONS 3.0 AND 4.0 - l LIMITING CONDITIONS FOR OPERATION j AND
- 1 SURVEILLANCE REQUIREMENTS 1
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Cencitier.s for 0:eratien, t*,e also: tete: A*T!0N *etairements shall te eet. i l l3.0.2 Nor.coepiian:t with a see:' fi*ation s' ail exist *ner th? reevirements :f
. . the '.iritin; Con:ition for 0:eratien eat este:dete: A;;: 0N re:virerer:s are rct j i me witr.in tne see:ified time inte veis. If :ne Liti;in; :encition fer
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ii3.0.4 Entry into en OPERATIONAL M00E cr other specifie c o r.d i t ' c a ins'.i n:: lIwitroutrelianceonc*visiers:entaine:l te madein unless the ACTION the :enditic-s recuiremer of the Liriting
. This Conditie l provisien shall no: preven; passe;e :aro;;h OPERATIONAL MODE! es re:ci-et to i: comply with ACTION recuire ents. Ex:c:ti:ns to these recuiretetts are stated
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3 I ; 3.0.5 Wher a sys:em, subtystem, train, eersooner: er device is cetereir.ec tt
. ce inoperable scieiy be:euse its ems egen y power seur:e is in:perec*te, or
!!solela because its normai power source is inopertole, it may be cor.sidered
!! OPERABLE fer the purpose of satisfyin; ne retairements of its aof.tenbie (1) its corresponcing norma' er le.ititingConditiceforOperation,prcvidet:
.. mergen:y power source is 0 FED.AELE; ene (t) a*i e' its redu'nten: s yt:et(s ) ,
1: subsyster(s ), tr6 int:), :omeonent(si an: oevice(s) are OPERABLE, or iikewise l satisfy the recuirements cf inis s;ecificati;n. Uni ets ooth cor.ditio".s (1) and (2) are satisfiec witrin ? hour:. ACT10'; steil oc initiatec to pia 2 the unit in a MODE in which the appli:aele Limiting Cenoition for Ojeration coes
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,!CALVERTCLFFS-UN71 3/4 0 1 Arenamen: 4c .h>2
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l APPLICABillTY LIMITING CONDITION FOR OPERATION
- 1. At least HOT STANDBY within the next 6 hours,
- 2. At least HOT SHUTDOWN within the following 6 hours, and
- 3. At least COLD S WTDOWN within the subsequent 24 hours.
This specification is not applicable in MODES S or 6. SgtlLLANCEREQWREMENTS 4.0.1 Surveillance Requirements shall be appitcable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable - at the time it is identified that a Surveillance Requirement has not been performed. However, this time of applicability may be delayed for up to
' 24 hours to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours.
Surveillance Requirements do not have to be performed on inoperable equipment. 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 surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL N00ES as required to comply with ACTION requirements.- 4.0.5 Surveillance Requirements for inservice inspection and testing'of ASME Code Class 1, 2,_and 3 components shall be applicable as follows: i i \ I. . l CALVERT CLIFFS UNIT 1 3/4 0 2 Amendment No JJ/yJ,150
7.-,m.~um JUoVEf ttaW(LE@lRrMEN15,f tentinued) , ,
- a. InserviceinspectionofASMECodeCiass1,2,and3componentsand* I inservice testing of ASME Code Class 1, 2, and 3 pumps and valves !
shall be perfomed in accordance with Section XI of the ASME Boiler I and Pressure Vessel Code and applicable Addenda as required by i 10 CFR 50, Section 50.55a(g), except where specific written relief l
. has been granted by the Commission pursuant to 10 CFR 50, 1 Section 50.55a(g)(6)(i).
. b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Coile 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 l follows in these Technical Specifications: , l ASME Boiler and Pressure Yessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing -
)
inseection and testine activities activities Weekly- At least once per 7 days - Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semi annually or every 6 months 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 frecuencies for performing inservice inspection and testing activities.
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements, a
- e. Nothing in the ASME Boiler and Pressure Vessel Code shali be
- construed to supersede the requirements of any Technical Specification. *
. CALVERT CLIFFS - UNIT 1 3/4 0 3 Amendment No. J2141 p qi a ;TT: '"::? ? 1
. - -t*=353M l
I I 3/4.1 REACTIVITY CONTROL SYSTEMS i 3/4.1.1 BORATION CONTROL l SHUTDOWN MARGIN - T > 2000F avo l l LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN line of Figure 3.1-lbpGIN shall be equal to or greater than the limit APPLICABILITY: MODES 1, 2**, 3, and 4. { 1 ACTION: I With the SHtJTDOWN MARGIN less than the limit line of Figure 3.1 lbY immediately initiate and continue boration at 2 40 gpm of 2300 ppm boric j acid solution or equivalent until the required SHUTDOWN MARGIN is ! restored, j E RVEILLANCE REOUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than the limit4of Figure 3.1 lbT l tino
- a. Within one hour after detection of an inoperable CEA(s)+ and at least once per 12 hours thereafter while the CEA(s)+ is inoperable. If the inoperable CEA is immovable or untrippable, I
the above required SHUTDOWN MARGIN shall be increased by an i amount at least equal or untrippable CEA(s),to the withdrawn worth of the immovable 8
- b. When in MODES 1 or 2 , at least once per 12 hours by verifying that CEA group withdrawal + is within the Transient Insertion l Limits of Specification 3.1.3.6.
- c. When in MODE 288, within 4 hours prior to achieving reactor I critica1 position {tybyverifyingthatthepredictedcriticalCEA is within the limits of Specification 3.1.3.6.
l
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of(e)below, with the CEA groups at the Transient Insertion Limits of l Specification 3.1.3.6.
t Adherence to Technical Specification 3.1.3.6 as specified in l Surveillance Requirement ( 4.1.1.1.1 assures that there is sufficient available shutdown margin to match the shutdown margin requirements
- of the safety analyses.
See Special Test Exception 3.10.1.
# With K ff 1 1.0 ff l.0
+ WithK, Exclud f
inh <thecenterCEAduringCycle10. l l l CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No. 2J/72//S/7J/EE/ Jpf#;p,151 ,
REACTIVITY CONTROL SYSTEMS l SURVEILLANCE REQUIRDtENTS (Continued)
- e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1. Reactor stem boron concentration,
- 2. CEA pos %polant.
ion,
- 3. Reactorkoolanthstem average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1.0% A k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors 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 Effective Full Power Days after each fuel load;ng, i CALVERT CLIFFS - UNIT 1 3/4 1-2
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{ CALVERT CLITTS . UNIT 1 I 3/4 1 2a Amendment No.130 I
REACTIVITY CONTROL SYSTEMS i SHlfiDOWN MRGIN - T ayg i 200 0F LINITING CONDITION FOR OPERATION 3.1.1.2 The SHlfTDOWN MRGIN shall be 2 3.0% 6 k/k. APPLICABILITY: MODE 5
- a. Pressurizer level 2 90 inches from bottom of the pressurizir.
- b. Pressurizer level < 90 inches from bottom of the pressurizar and all sources of non borated water 188 gpm.
ACTION:
- a. With the SHlfTDOWN MRGIN < 3.0% 6 k/k, immediately initiate -
and continue boration at 2 40 gpm of 2300 ppm boric acid solution or equivalent until the required SHUTDOWN MRGIN is restored.
- b. With the pressurizer drained to 5 90 inches and all sources of non borated water > 88 gpm, immediately suspend all operations involving positive reactivity changes while the SHUTDOWN MRGIN is increased to compensate for the additional sources of non borated water or reduce the sources of non borated water to 1 88 gpm.
SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHl1TDOWN MRGIN shall be determined to be 2 3.0% 6 k/k:
- a. Within one hour after detection of an inoperable CEA(s)+ and at least once per 12 hours thereafter while the CEA(s)+ is inoperable. If the inoperable CEA+ is immovable or untrippable, the above required SHUTDOWN MRGIN shall be increasedbyanamountatleastequgltothewithdrawnworthof the immovable or untrippable CEA s l
- b. At least once per 24 hours by icon (s)deration of the following factors:
- 1. Reactoricoolantfystemboronconcentration,
- 2. CEA position,
- 3. Reactor'Goolantfystemaveragetemperature,
- 4. Fuel burnup based on grcss thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.2.2 With the pressurizer drained to s 90 inches determine:
- a. Wit (inonehgurandtvery12hoursthereafterthatthelevelin the Yeactor toolant f/ stem is above the bottom of the hot leg nozzles, and
- b. Within one hour and every 12 hours thereafter that the sources of non borated water are s 88 gpm or the shutdown margin has compensated for the additional sources.
0 + Excluding the center CEA during Cycle 10. CALVERT CLIFFS - UNIT 1 3/4 1-3 Amendment No. # / S ,151 i
i REACilVliY CON 70L SYSTEWS EDRON DILUilCN LIMITING C001T10N FCA OPERAi10N J.l.1.3 The flow rate of reactx coolant through thehneterGoolant Dystem shall be > 3000 gpm whenever a reduction in P.eactor Coolant Syste- boron conTentration is being trade. APPLICABILITY: ALL MDDES. ACTION: With the flow rate of retetor toelant through the batter bolart3yste?
< 3000 gpm, imediately suspend all operations involving a reduction 4._
in baron concentration of the Reactor Coolant System. SURVEILLANCE RE0iJIREMENTS 4 The flow rate of reactor coolant through the bactorbolant @c
$y.1.1.3 stem shall t>e determined to be 1 3000 gpm within one hour 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 least one low pressure safety injec ion pump is in op3rntion and supplying y,3000 gpm through the actor tipolant hystem.
h e CALVERT CLIFFS-UNIT 1
, :CALVRT=CL4FFM*t7=2, 3/4 1-4
REACTIVITY CORTROL systems MODERATOR TEMPERATURE COEFFic1ERT kJM.111NGCONDI.TIONFOROPERATION 3.1.1.4 The moderator teeperature coefficient (MTC,) shall be: a. Less positive than the limit line of figure 3.1 la, and
- b. Less negative than 2.7 x 10*4 0 21 k/k/ F at RATED THERMAL POWER.
APPLICABillTY: MODIS J and 2*# ACTION: With the moderator temperature coefficient outside any one of the above J limits, be in at least HOT STANDBY within 6 hours. l
$URVEILLANCE RIDUIREMEVTS '
4.1.1.4.1 The measurements. MTC shall be determined to be within its limits by confire.atory permit-direct comparison with the above limits.HTC measured values shal l With Keff 2 1.0. 1 See Special Test Ixception 3.10.2. 1 l l l } CA1 VERT CLIFFS - UNIT 1 3/4 1-5 Amendment No. /E//EE//JEf,130 l l
WACCEPTAM E OPDAT10ii F1010N POSITTVE MTC UNIT UNI 0.70 c b/ - (0.7,0.7) w N A 0.60 - N { 0.50 - E 0.40 - - A;CITTA1'.2 3 OPU.ATION .
- u c10s (1.0,0.3) b m
E 0.20 - 5 .E D.10 - 0.00 ' O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 RACTION OT a Avtn 'mrmg41,pmT.R TIECEI 3.1 14 Praction vs. A11ovable of R.ated Positive MTCThersal Povey 2.init (10* Ap /*F) CA1. VERT CLITTS . ImIT 1 3/4 1 5a hda h m 1
t , REACTIVITY CONTROL SYSTEMS
$yRVEILLANCERE002REF.EN75(Continued) 4 4.1.1.4.2 The MTC shall be detemined at the followir.g frequencies and THER'tAL POWIR conditions-during each fuel cycle:
- . a. Prior to initial operation above 5% of RATED THER*AL POWIR. af ter ~
each fuel loading.
- b. - At any THERMAL POWER above 90% of RATED THERMAL P0kTR, within 7 l EFPD af ter initially reaching an equilibrium condition at or I above 90k of RATED THIPyAL POWER.
- c. At any THERFAL POWER, within 7 EFPD of rea'ching a PATED THERF.AL l POWER equilibriun boren concentration of 300 ppm.
7 4 h i J f9 4 t yd I u 4 l l1 i: l 1 cAtvERT cuFrs - unit 1 3/4 1-s mendment no. or, 227 I
PEACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURF FOR CRITICALITY I llHITING CONDITION FOR OPERATION 3.1.1.5 The Reactor (T,yg)shallbe2515goolantSystemlowestoperatinglooptemperature F when the reactor is critical. APPLICABILITY: MODES 1 and 26 ACTION: 0
) < 515 F, With restoreaTReactor Coolant to within its limitSystem within 15operating minutes orloop be intemperature (T,NOT withinth$yRext15 minutes.
SURVEILLANCE RE0VIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be 2 5150F:
- 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 avg is less than 5250F.
# With K,ff 2 1.0.
l CALVERT CLIFFS - UNIT 1 3/4 1 7 Amendment No. 145
- . _ - - - .~ . - - - _ . ~ _-.
i PELCTIVITY CONTROL SYSTEMS - 3/4.1.? B0rtT10N Sv5TEMS FLOW PATHS . SHUTDOWN MMITING CONDIT!ON F0c ODERATION 3.1.2.) As a minimum, one of the following boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:
- a. A flow path from the boric acid storage tank via either a boric acid pump or a gravity feed connection and charging pump to the Reactor Coolant System if only the boric acid storage tank in -
Specification 3.1.2.7a is OPERABLE, or
- b. The fics path from the refueling water tank via either a charging pump or a high pressure safety injection pump
- to the Reactor Coolant System if only the refueling water tank in Specification 3.1.2.7b is OPERABLE.
APPLICABillTY: MODES 5 AND 6. ACTION: With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least (. one injection path is restored to OPERABLE status. SURVEILLANCE RE0UIREMENTS
~4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is above the temperature liniit line shown on Figure 3.11 when a flow path from the concentrated boric acid tanks is used.
- 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. At 327PF and less, the required OPERABLE HPSI pump shall be in pull to lock and will not start automatically. At 327 F0 and less, HPSI pump use will be conducted in accordance with Technical Specification 3.4.9.3. CALVERT CLIFFS - UNIT 1 3/4 1 8 Amendment No. Jf),146
MACTIVITY CONTROL SYSTEMS FLOW PATHS . OPERATINJ LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:
- a. Two flow paths from the boric acid storage tanks required to be OPERABl.E pursuant to Specifications 3.1.2.B and 3.1.2.9 via either a boric acid pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and
- b. The flow path from the refueling water tank via a charging pump to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at lea:t 3% Ak/k at 2000 F within th next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEllLANCE RE0VIREMENTS 4.1.2.2 At least two of the above reouired flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the concentr M ed boric acid tanks is above the temperature limit line shown on Figure 3.1 1.
- b. At least once per 31 days 'oy verifying that each valve (manual,
- l. power operated or automatic) in the flow path that is not i
locked, sealed, or otherwise secured in position, is in its correct position,
- c. At least once perNfueling Nterval by verifying on a SIAS test signal that:
(1) each automatic valve in the flow path actuctes to its correct position, and (2) each boric acid pump starts. CALVERT CLIFFS - UNIT 1 3/4 1-9 Amendment No fp/Jpf/JJp,MA 145 l l
l PEACTIVITY CONTROL SYSTEMS [FARGING PUMP SMUT 00WN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or one high pressure safety injection pump
- in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.
APPLICABillTY: MODES 5 and 6. ACTION: With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by Specification 4.0.5, c L i 4 At 327'F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically. At 327 0F and less, ' HPSI pump use will be conducted in accordance with Technical Specification 3.4.9.3. CALVERT CLIFFS UNIT 1 3/4 1-10 Amendment No. J/J.146 I 1
REACTTVITY CONT 1tOL SYSTEMS ,, ,_ , .. , ( OLARGING PUMPS - OPER ATING LIMTTTNG CONDII1Q1L FOR OPER ATION I - 3.1.2 A At leut two charging pumps shall be OPERABLE.* APPLICABILITY: MODES 1, 2, 3 and 4 ACT10N. ' With only one charging pump OPERABLE, restore at leut two charging' pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at leut 3% A k/k at 2000F within the next 6 hours; restore at leut two charging pumps to OPERABLE status within the cext 7 days or be in COLD SHUTDOWN within the next 30 hours. SIIRVETLL ANCE REOUTREMENTS 4.1.2 A At least two charging pumps shall be demonstrated OPERABLE:
- a. At least once perMfueling%terval by verifying that each charg-ing pump staru automatically upon receipt of a Safety injection .
,. Actuation Test Signal.
- b. No additional Surveillance Requirements other than those required by $pecification 4.0.5.
Above 30% RATED THERMAL POWER the two OPERABLE charging pumps shall have independent power supplies. CALVERT CLIFFS - UNIT 1 3/4 1-!! Amendmer.t No. //, /k, //[.128
REACTIVITY CONTROL SYSTEMS t BORIC ACID PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the flow path through the boric acid pump in Specification 3.1.2.la above, is OPERABLE. APPLICABILITY: MODES 5 and 6. ,
, ACTION:
With no boric acid pump OPERABLE as required to complete the flow path of Specification 3.1.2.la suspend all operations involving CORE ALTERA. TIONS or positive reactivity changes until at least one boric acid pump is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.5 No additional Surveillance Requirements other than those required by Specification 4.0.5. i CALVERT CLIFFS - UNIT 1 3/4 1-12 I
T -T "T D tt'F O L sySt p's 3 0:.:
- 10 :0 DL"@ S . 0:!R ATI NS L:$' T NG C0 ri?10N OR ODEU TION 3.1.2.6 At least the beric acid pump (s) in the beror, inje: tion flow path (s) required OPERABLE pursuant to See:ification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow Dath through the boric acid pump (s) in Specification 3.1.2.2a is OPERAILI.
A:o'!CABIL!TY: M00!S 1, 2, 3 and 4 o ACTION: With ont boric acid pump required for the boron inje: tion flow path (s) pursuant to Soe:ification 3.1.2.24 inoperable, restore the boric acid pump to OPERAELE status within 72 hours or be in at least HDT STANDBY within the next 6 hours and borated to a SHUTOOWN KAR3:N etuivalent to
- i. a: least 35 's/k at 200'F: restore tne above re;uired boric a:id ov-o(s) .
- Ofi;AE'E. status w1:nin tne next 7 cays or be in COLD SnUT00<h wi:nin
- ne next 30 h grs.
SUP.VE!LLANCE REDUIREY.INTS 4.1.2.6 No additional Surveillance Recuirements other than those recuired by Spe:ifi:ations 4.0.5 and 4.1.2.2. l I I e
, j 4 se g ( ,
1 1
E:C':V:~Y C0'iTE:t SYS EMS BORATED WATER SOUP.CES - Suv7:0WN LIM: TING CONOITION F0; 00 ERAT: 0N - 3.1.2.7 As a minimum, one of the following berated water sources shali be.0PERABLE:
- a. One boric acid storage tank and one associated heat tra:ing circuit with the tank centents in accordance with Figure 3.1-1.
- b. The refueling water tank with:
- 1. A minimum contained berated water volume of 9,844 gallons, l
- 2. A minimum boren con:entration of 2300 ppe, and j
- 3. A minimum soluti:n temperature of 3E'F. '
APPLICAEILITY: MODES 5 and 6. ACTION: With no borated water sources 0:ERABLE, suspene all operations involving CORE ALTERATIONS or positive ro40tivity changes until at least one borated water source is restored to OPERABLE status. SURVEILLANCE RE00:REMENTS
-C 4.1.2.7 The abOve required berated water source shall be den. nstrated CPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the boren concentratien of the water,
- 2. Verifying the c:ntained berated wa er volume of :ne tank, and
- 3. Ve-ifying the boric acid storage tank soluticn tem:erature when it is tne source of berated water.
- b. At least once ;er 24 hcurs by verifying the RWT temperature when it is the source of berated water and the outside tir tem: era ure is < 3E'F.
CAL'lERT CLIFFS - Uh'T 1 3/4 1-14 Amen:mer: Nc. 2l>;3 l
S 170
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!I* -
l i EOD TEDMTEo 50';tCES .' OPE:.17:N3 s:
= jb!IMI*:N3,. -::NO."::N 70; 0:IR**:0N Ell
- l i .1.2.8 3 A: least one f the f:llowing two combina:icns :f bora:ed water ,
l
-;I;. sources'shall=be 0 ERAELE: -
- a. Two boric _ acid storage tank (s) and one associate: heat tracing circuit.-per tank witn the contents of the tanks in ac:ordan:e with Figure 3.1-1 and the boren con:entration limited to < 8%, or- _
i g b. Eeri: A:id Storage Tank 12 OPERABLE per Speci'i:ation 3.1.2.6.a an :ne refueling wa:er tanP witn 11 q_ 1. A minimum contained borated water volume of 400,000 gallons,
, 2. A boren con:entration of between 2300 and 2700 po ,
- 3. A minimum solu icn temperature of 400F, and s- 4 A maximum solution' temperature cf 10 0 F in MODE 1. .
1 A :LICAEIL*~v- MODE 1 > ED'4 of RATED THERM 4 POWEk.
- ACTION:.
a. I_ With neither combination of berated water sources OPE:AELE but at-
. leas:Etwo cf the' individual berated water sources GFERAELE, restore at leastione of'the combinations defined in Spe:ification 3.1.2.8 to OPEPAELE status- within 72 hours- or reduce power to less than 80' J cf RATED THERMAL POWER within the next 6 hours.
' b. -With only one berated water source OPEPAELE, within 1 hour either
-restore at least two of :he individual borated water sources to OPERABLE status or reduce power below 80% of RATED THERMAL POWER-and comply with Specification 3.1.2.9. s SURVEILLANCE REOL'IREMENTS
.h4.1'.2.8 A . least two corated water sour:es sna11. be demon
~
ql a. It least on:e per_7 days by:
;; 11 .
Verifyin; :ne~ boren cen entrati:n in es:hlwa:er sour:e, 2.- Verifying the contained berated water volume in.eac,n water source, and- - [ 3. Verifying the beric acid storage tank solu:i:n temperature. " ' b. A: least once oer 24 hours by va-if
,outside air :emperature is < 40 F0 ying :ne R,7 temoe-a:ure w.*er _:r,e o y
CALVERT CLIFFS - V C7 1 39 l-16 Amendment No. 42, !!. Ice _ s . . . _ . _ _ _ _ . , . _ .__ _-_ _ , _ . . _ ~ . .
@5ACTIIITY CONTR0' SYSTEM . 3 il j ' BORa ED '.-ltTIt. 50;:.t!S - 0 E:3?:NG ll LIM:7:N3CONOITIONFOROPERAT*0N I
3.1.2.9 l OPEFABLE:At least two of the folie;ing three corated water sources shall be t l
- a. Two boric acid storage tank (s) and one associated heat tracing circuit per tank with the contents of the tanks in accordance with Figure 3.1-1 and the boron concentration limited to 1 8'i, and
!! b, The refueling water tank with:
l'
- 1. A minimum contained berated water volume of 400,000 gallons,
- 2. A boron concentration of between 2300 and 2700 ppm,
- 3. A minimum solution temperature of 400F, and
- 4. A maximum solution temperature of 100 F 0 in MODE 1.
!! A :'L : : AE '. ".~ YF.00ES
- P , 2, 3 and 4 ,
ACTION: Witn only one borated water scurce OPEPAELE, restore at leest two berated water sources to OPERABLE status within 72 hours or be in at least HOT STANOSY within the next & hours and borated te a SHUTOOWN FARGIN equivalent to at least 2". 3.k/k at 200*F; restore at least two borated water sources to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REOUIREMENTS 4.1.2.9 At least two bort.ted water sources'shall be demonstrated OPEPASLE: l
- a. At least once per 7 days by:
- 1. Verifying the boren concentration in each water source,
- 2. Verifying tne contained borated water volume in each water source, and
- 3. Verifying she boric acid storage tank solution tem;erature.
l l.
- b. At least ence per 24 hours by verifving the RW~ temperature when the outside air temperature is < 40 DF.
'At 1 80% cf RATED THERMAL F0WER. j i
CALVERT CLIFFS - UN:71 3/4 1-164 Amencment No. /E, !!,104 t
REACTIVITY CONTROL SYSTEMS l 3.4.1.3 MOVABLE CONTROL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.1.3.1 The CEA Motion Inhibit and all shutdown and regulating CEAs+ l shall be OPERABLE with each CEA of a given group positioned within 7.5 inches (indicated position) of all other CEAs in its group. APPLICABILITY: MODES 1* and 2* ACTION:
- a. With one or more CEAs+ inoperable due to being immovable as a l result of excessive friction or mechanical interference or known to be untrippable, be in at least HOT STANDBY within 6 hours,
- b. With the CEA Motion Inhibit inoperable, within 6 hours either:
- 1. Restore the CEA Motion Inhibit to OPERABLE status, or
- 2. Place and maintain the CEA drive system mode switch in either the "Off" or any ' Manual Mode" cosition and fully withdraw all CEAs in groups 3 and 4 and withdraw the CEAs+
in group 5 to less than 5% insertion, or
- 3. Be in at least HOT STANDBY.
- c. With one CEA inoperable + due to causes other than addressed by ACTION a, above, and inserted beyond the Long Term Steady State Insertion Limits but within its above specified alignment requirements, operation in MODES 1 and 2 may continue for up to 7 days per occurrence with a total accumulated time of 114 days per calendar year,
- d. With one CEA inoperable + due to causes other than addressed by ACTION a, above, but within its above specified alignment requirements and either fully withdrawn or within the Long Term Steady State Insertion Limits if in CEA group 5, operation in MODES 1 and 2 may continue.
s
- See Special Test Exceptions 3.10.2 and 3.10.4.
+ Excluding the center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 1-17 Amendment No. U 7, 151 l
REACTIVITY CONTROL SYSTEMS LINITING CONDITION FOR OPERATION
- e. With one or more CEAs+ misaligned from any other CEAs in its !
group by more than 7.5 inches but less than 15 inches, operation in MODES 1 and 2 may continue, provided that within one hour the misaligned CEA(s) is either:
- 1. Restored to OPERABLE status within its above specified alignment requirements, or
- 2. Declared inoperable. After declaring the CEA+ inoperable l operation in MODES 1 and 2 may continue for up to 7 days per occurrence with a total accumulated time of 114 days per calendar year provided all of the following conditions are met:
- a. The THEPJ4AL POWER level shall be reduced to 170% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination within one hour; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used.
- b. WithinonehourafterreducingtheTHERMALPOWERas required byga) above, the remainder of the CEAs+ in the group w th the inoperable CEA+ shall be aligned to within 7.5 inches of the inoperable CEA+ while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.
- f. With one CEA+ misaligned from any other CEA+ in its group by l 15 inches or more, operation in MODES 1 and 2 may continue, provided that the misaligned CEA+ is positioned within 7.5 inches of the other CEAs+ in its group in accordance with the time allowance shown in Figure 3.1-3. The pre-misaligned Ff value used to determine the allowable time to realign the CEA+ from Figure 3.1-3 shall be the latest measurement taken l within 5 days prior to the CEA misalignment. If no measurements were taken within 5 days prior to the
_ ..._ __- __ -_l"Isalignment, a pre-misaligned Ff of 1.65 shall be assumed,
- g. With one CEA+ misaligned from any other CEA+ in its group by 15 inches or more at the conclusion of the time allowance permitted in Figure 3.1-3, immediately start to implement the l following actions:
- 1. If the THERMAL POWER level prior to the misalignment was greater than 50% of RATED THERMAL POWER, THERMAL POWER shall be reduced to less than the greater of:
+ Excluding the center CEA during Cycle 10.
)
CALVERT CLIFFS - UNIT 1 3/4 1-18 Amendment No. U7, 151
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION a) 50% of RATED THERMAL POWER b) 75% of the THERMAL POWER level prior to the misalignment within one hour after exceeding the time allowance permitted by Figure 3.1-3.
- 2. If the THERMAL POWER level prior to the misalignment was s 50%
of RATED THERMAL POWER, maintain THERMAL POWER no higher than the value prior to the misalignment. If negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Within one hour after establishing the appropriate THERMAL POWER as required above, either:
- 1. Restore the CEA+ to within the above specified alignment l requirements, or
- 2. Declare the CEA+ inoperable. After declaring the CEA inoperable, POWER OPERATION may continue for up to 7 days per occurrence with a total accumulated time of s 14 days ser calendar year provided the remainder of the CEAs+ in tie group l with the inoperable CEA are aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and
' insertion limits shown on Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation,
- h. With more than one CEA+ inoperable or misaligned from any other CEA in its group by 15 inches (indicated position) or more, be in at least HOT STANDBY within 6 hours.
- i. For the purposes of performing the CEA+ operability test of TS 4.1.3.1.2, if the CEA has an inoperable position indication channel, the alternate indication system (pulse counter or voltage dividing network) will be used to monitor position. If a direct position indication-(full out reed switch or voltage dividing network) cannot be restored within ten minutes from the commencement of CEA motion, or CEA withdrawal exceeds the surveillance testing insertion by > 7.5 inches, the position of the CEA shall be assumed to have been > 15 inches from its group at the commencement of CEA motion.
+ Excluding the center CEA during Cycle 10.
i l CALVERT CLIFFS - UNIT 1 3/4 1-19 Amendment No. JE#f#/7,151 i
REACTIVITY CONTROL SYSTEMS E gy1ILLANCE RE0UIREMENTS 4.1.3.1.1 The position of each CEA+ shall be determined to be within 7.5 inches (indicated position) of all other CEAs in its group at least once per 12 hours except during time intervals when the Deviation Circuit and/or CEA Motion Inhibit are inoperable, then verify the individual CEA positions at least once per 4 hours. 4.1.3.1.2 Each CEA+ not fully inserted shall be determined to be l OPERABLE by inserting it at least 7.5 inches at least once per 31 days. 4.1.3.1.3 The CEA Motion Inhibit shall be demonstrated OPERABLE at least once per 31 days by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit also prevents any CEA from being misaligned from all other CEAs in its group by more than 7.5 inches (indicated position).+
+ Excluding the center CEA during Cycle 10.
1 I l CALVERT CLIFFS - UNIT 1 3/4 1-19A Amendment No. #7,151
O C.
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9 iiiiiiiii t ilit tiiilitiiiiiiilit tit isiiltititiiiilis titis tilisis titi liittitin e. e e e e e o c e m .e e m N - e (5310N!W) Y33 NDI1V3Y 01 3WII CALVERT CLIFFS - UNIT 1 3/4 1-195 Amendment No. 127
I I I l This page intentionally lef t blank.
)
I CALVEF,T CLIFFS-UNIT 1 3/4 1-20 Amendment No. 27, 22. 77, hn , 127
REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS LIMITING CONDITION FOR OPERATION 3.1.3.3 At least two of the following three CEA position indicator channels shall be OPERABLE for each shutdown and regulating CEA+:
- a. CEA voltage divider reed switch position indicator channel, c:pable of determining the absolute CEA position within 1.75 inches;
- b. CEA " Full Out" or " Full In" reed switch position indicator l channel, only if the CEA is fully withdrawn or fully inserted, as verified by actuation of the applicable position indicator; and
- c. CEA pulse counting position indicator channel, i
l APPLICABILITY: MODES 1 and 2. ACTION:
- a. With a maximum of one CEA+ per group having its voltage divider reed switch position indicator channel or its pulse counting position indicator channel inoperable and the CEA(s) with the l inoperable position indicator channel partially inserted, either:
l 1. Within 6 hours a) Restore the inoperable position indicator channel to OPERABLE status, or b) Be in.at least HOT STANDBY, or c) Reduce THERMAL POWER to $ 70% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Operation at or below this reduced THERMAL POWER level may continue provided that within the next 4 hours either:
- 1) The CEA group (s)+ with the inoperable position indicator is fully withdrawn while maintaining the withdrawal sequence required by.
l Specification 3.1.3.C and when this CEt group i reaches its fully withdrawn sosition, the " Full Out" lisait of the CEA with tie inoperable position ilokator is actuated and verifies this CEA to be fully withdrawn. Subsequent to fully withdrawirg this CEA group (s), the THERMAL POWER l level may be returned to a level consistent with
.all other applicable specifications and g$ operation may continue per Specification 3.1.3.3 above; or
- 2) The CEA group (s)+ with the inoperable position l indicator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL POWER level required by Specification 3.1.3.6 and when this CEA group reaches its fully
+ Excluding the center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 1-21 AmendmentNo.JJHJ/Ef#El JEE,^151
REACTIVITY CONTROL SYSTEMS I POSITION INDICATOR CHANNEL 1 l LINITING CONDITION FOR OPERATION inserted position, the " Full In" limit of the CEA with the inoperable indicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6, and may continue per 4 Specification 3.1.3.3 above.
- 2. for,'llf the failure existed before entry into MODE 2 or occurs prior to an "all CEAs out" configuration, the CEA groups (s)+ with inoperable position indicator channel must be moved to the " Full Out" position and verified to be fully withdrawn via a " Full Out" indicator. These actions must be completed within 10 hours of entry into MODE 2 and prior to exceeding 70% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. The provisions of Specification 3.0.4 are not applicable. Once these actions are completed, operation may continue per Specification 3.1.3.3 above.
- b. With more than one CEA+ per group having its CEA pulse counting l position indicator channel and either (1) the " Full Out" or
" Full In" position indicator, or (2) the voltage divider position indicator channel inoperable, operation in MODES 1 and 2 may continue for up to 24 hours provided that for the affected CEAs, either:
- 1. The CEA voltage divider reed switch position indicator channels are OPERABLE, or
- 2. The CEA " Full Out" or " Full In" reed switch position indicator channels are OPERABLE, with the CEA fully withdrawn or fully inserted as verified by actuation of the applicable position indicator.
>wlERLANCE RE0VIRENENTS 4.1.3.3.1 Each required CEA+ position indication channel shall be I determined to be OPERABLE by determining CEA positions as follows at least once per 12 hours, by:
- a. Verifying the CEA pulse counting position indicator channels and the CEA voltage divider reed switch position indicator channels agree within 4.5 inches, or
- b. Verifying the CEA pulse counting position indicator channels and the CEA " Full Out" or " Full In" reed switch position indicator channels agree within 4.5 inches, or
- c. Verifying the CEA voltage divider reed switch position indicator channels and the CEA " Full Out" or " Full In" reed switch position indicator channels agree within 4.5 inches.
4.1.3.3.2 During time intervals when the deviation circuit is inoperable, the above verification of required CEA+ position indicator l channels shall be made at least once per 4 hours.
+ Excluding the center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 1-22 Amendment No. EN7 M M S , . 151
REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA+ drop time, from a fully withdrawn position, shall be 13.1 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:
- a. T avg 2 5150F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION:
- a. With the drop time of any full length CEA+ determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.
- b. With the CEA drop times within limits but de:.ennined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.
SURVEILLANCE REOUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs+ shall be demonstrated through measurement prior to reactor criticality:
- a. For all CEAs following each removal of the reactor vessel head,
- b. For specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and
- c. At least once per refueling interval.
+ Excluding the center CEA during Cycle 10.
l CALVERT CLIFFS - UNIT 1 3/4 1-23 Amendment No. 3209/ U #29, 151
REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMLI LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 129.0 inches. APPLICABILITY: MODES 1 and 2*#. EllQ!i: With a maximum of one shutdown CEA withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, to less than 129.0 inches, within one hour either:
- a. Withdraw the CEA to at least 129.0 inches, or
- b. Declare the CEA inoperable and apply Specification 3.1.3.1.
EURVEILLANCE REOUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least 129.0 inches:
- a. Within 15 minutes prior to withdrawal of any CEAs+ in regulating groups during an approach to reactor criticality, and
- b. At least once per 12 hours thercafter.
See Special Test Exception 3.10.2.
# With K ff2 1.0
+ Excluding,thecenterCEAduringCycle10.
l I l CALVERT CLIFFS - UNIT 1 3/4 1-24 Amendment No. U ,151
r REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating CEA+ groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3.12 (regulating CEAs are considered to be fully withdrawn in accordance with Figure 3.1-2 when withdrawn to at least 129.0 inches) with CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits restricted to:
- a. 1 4 hours per 24 hour interval,
- b. 15 Effective Full Power Days per 30 Effective Full Power Day interval, and
- c. 114 Effective Full Power Days per calendar year.
APPLICABILITY: MODES 1* and 2*#, ACTION:
- a. With the regulating CEA+ groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:
- 1. Restore the regulating CEA groups to within the limits, or
- 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using Figure 3.1-2.
- b. With the regulating CEA+ groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals > 4 hours per 24 hour interval, except during operations pursuant to the provisions of ACTION items c.
and e. of Specification 3.1.3.1, operation may proceed provided either:
- 1. The Short Term Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or
- 2. Any subsequent increase in THERMAL POWER is restricted to 5 5% of RATED THERMAL POWER per hour.
See Special Test Exceptions 3.10.2 and 3.10.4. ff2 1.0.
+ With The cenX,ter CEA may be excluded from the determination of Bank 5 position during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 1-25 Amendment No. JE#EE,151 _ _ _ __- _ _ - - - - _ . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - '~
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
- c. With the regulating CEA+ groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion limits for intervals > 5 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, except during operations pursuant to the provisions of ACTION items c. and e. of Specification 3.1.3.1, either:
- 1. Restore the regulating groups to within the Lon Steady State Insertion Limits within two hours,gorTerm 2.
Be in at least HOT STANDBY within 6 hours. EIRVEILLANCE REOUIREMENTS 4.1.3.6 The position of each regulating CEA+ group shall be determined to be within the Transient Insertion Limits at least once per 12 hours l except during time intervals when the PDIL Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated times during which the regulating CEA groups are inserted beyond the Steady State Insertion Limits but within the Transient insertion Limits shall be determined at least once per 24 hours.
+
The center CEA may be excluded from the determination of Bank 5 position during Cycle 10. CALVERT CLIFFS - UNIT 1 3/4 1-26 Amendment No.151
A q
- 1;00. 4.00,G, 5 9 351 -
1 i i e l. l l
, 0.90 g ,; 4.90,Gp $ 9 35%
N 0.80 l \ - 4 1 . 0.75,G, 5 9.505' l 9 0.70 . .70,Gp 5.9 60% 1 Q65,Gp5.985% j ,
.56,Gp 4 9.50%
l Lk Te5 ShTot' Term i D.40 . gSteady Stat e -Steady State - - - - i Insertion Insertion l Limit ' O.30' Limit - -- u I* Gr> $ 9 25% ; Grp 4 9 20% i 1 0.20 8 "- - - - --
.20,Gp 3 9 60'- -
0.10 0.00 , . 0.00, Gp 3 9 60* - -
'l
! AllowableilAS$$ .Grp 5 9 $5% Operating Region' l Groups: 5 3 - 1 ,, -t ( e 3 _e I 1 e t_ I a .a . I a e a i a a_ e
! 0 20 to 60 80 100 0 20 40 60 80 100 0 20 40 60 80 100
? 136.0 108.8 81.8 54.4 27.2 0 136.0 108.8 81.4 54.4 27.2 0 136.0 108.8 81.6 54.4 27.2 0 i
*z 4 2 o 8. I a.
a e I e a _ l 1 _a t u
- O 20 40 60 80 100 0 20 40 60 80 100 P 136.0 108.8 81.6 54.4 27.I O 136.0 108.8 81.6' 54.4 27.2 0 w
1 CEA IkSERTIOlt' u INCHES CEA WITHDRAWil
~ Figure 3.1-2 I P CEA GROUP IftSERTION l.IMITS VS. FhACTION OF ALLOWARLE THERMAL POWER [
~
FOR EXISTIffG RCP COMBINATION g t i
3/4.2 POWER DISTRIBUTION LIMITS ($k,Q.ILINEARHEATRATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2 1. APPLICABILITY: MODE 1. ACTION: With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:
- a. Restore the linear heat rate to within its limits within one hour, or
- b. Be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REOUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable. 4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system, 4.2.1.3 Excore Detector Monitorina Sttigm - The excore detector monitoring system may be used for monitoring the core power distribution by:
- a. Verifying at least once per 12 hours that the full length CEAs+
are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6.
- b. Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.
+ Excluding the center CEA during Cycle 10.
1 CALVERT CLIFFS - UNIT 1 3/4 2-1 Amendment No. JJ/JE,151
. O'JE:t 3: I'0*B'JT:01 '. M T5
. I i
' !! *.'E V E : '. '.
- N:I ?!?):Riv!'i S (Centinue T l 4
l
. .i
- c. Verifying at least once Nr 31 days that the AXIAL SHAPE -!.NDEX is ;
maintaine: wi:nin the limits of Figure 3.2-2, where 100 per:en; of ' the allowable power represents the maximum THERMAL POWER allowed by the following expressi'on: MxN l where:
. 1. M is the maximum allowable THERMAL POWER level fer the existing Ret: tor Coelant Pump combinatien.
- 2. N is tne maximum allowlble fra: tion of RATED THERMAL POWER
, asdeterminedbytheFj ycurve of Figure 3.2-35.
, e
!4. 2.1. 4 Incore Detector Moriterine System The intore de ector meni;0-ir; .
ilsyster may :e use: fer monitoring :ne core power distribution by verifying
' nat :ne incore de:e::Or Local Pcwer Density alarms:
- a. Are adjusted to satisfy the requirements of the core power.distribu-tion map which shall be uodated at least once per 31 days of a::umulated cperation in MODE 1.
- b. Have their alarm setpcint adjusted to less than or equal to the limits snown on Figure 3.2-1 when the fellowing factors are appro-priately in:1uded in the setting of these alares:
1
- 1. A measurement-calculational un:ertainty facter of 1.C52, :
- 2. An engineering uncertainty factor cf 1.03, -i 3.. A linear heat rate uncertainty factor of 1.002 due to axial '
,, fuel censif t:ation anc thermal expansion, an:
-4 A THERMAL POWER measurement uncertainty fa::cr of 1.02.
il t
,,CALVERT CL FFS - UN:Ti 3/4 2 2 Amencment N:. 22, 23, 33, 77,104
('
E. 3
- B E
E z m e 9 2 $ C o Q. _9 e Q k e w
=
a: w 9 h 5c 5 - o O N g w o a "z d $ d E* H m t 6 sl 0 6
- w Es u b k %#
.O.
< O u b a.
2 < $ e 3 g 3
=
i k e I @ 3 Q, (WQ1YM300W+ 0y13+ 13nd) id/AOI *H.LVM AYJH MY2M1 )lV3d K18VM011V 4 CALVERT CLIFFS - Ut(IT 1 3/4 23 Amendment No. 27,48
1.10 e i i i i 1.00 - ( 0.06,1,00) * * (0.12,1.00) - UNACCPETA31.E 1:KACCEPTABLE 0.90 -- OPERATION OPERATION - RIC 0S R.EGION 0.80 - 6 [ . 0.70 - ( 0.3, 70) ? ACCEPIAILE OPERATION o (0.3,.70) - j ' RIGIOS , 5 g 0.60 - 3 O.50 - ( 0.3,.50) * -
)
o g 0.40 - t3 0.30 - 0.20 -
'* ( .45,,15) 0.10 -
0.05 ' ' ' ' '
-0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 PERIPHERAL AIIAL SHAPE INDEX, Y 7
Tigure 3.2 2 linear Heat Rate Axial Flux Offset Control Limits CALVIRT CLIFFS-MG\Y Y 3/4 2 4 Amendment No. 22,24,22,23 I 19, 48, 72, 130 l l
n G m 1.0 (1.54, 1.01 ' 5 .
- n. ,
E. 4 n vi
=
0.9 - c .. i
.m.
d g a - (1.785,0.8) t O.8 - t N t to ACCEPTABLE VALUE
- 0.7 -
. t to .I.,
t 0.6 - N . a "y . k O.5 ' ' ' ' ' ' '
- 1.45 r+ - 1.50 1.55 1.60 1.65 1.70 1.75 1.80 -
y T I Fxy w w .
"
- Figure 3.2-3b i
. TOTAL PLANAR RADIAL PEAKING FACTOR vs N
I
- i i
1 i n 4
. i l
1 1 l t Uyc.j.. 4L D I t t e (n* .."v .; ". . ei - L..t t g - go,
- n. ,
3 i a / .,, 4 . .; (-ie n "~.* h
- NO. / P, .10 t,
POWER DISTRIBUTION LIMITS ry/4.,g.2TOTM. PLAKAR RADIAL PEAXING FACTOR - F}y LIMITING CONDITION FOR OPERATION __ 3.2.2.1 ThecalculatedvalueofF}y,definedasF}y-F xy(1+Tq ), shall be limited to s 1.70. APPLICABILITY: MODE 1*. ACTION: WithF}y>1.70,within6hourseither:
- a. Reduce THERMAL POWER to bring the combination of THERMAL POWER andF}y to within the limits of Figure 3.2 3a and withdraw the full length CEAs+ to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or
- b. Be in at least HOT STANDBY.
EURYEILLAELREQUIREMENTS 4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.1.2 F}y shall be calculated by the expression F}y - Fxy(1+T ) and q F}y shall be determined to be within its limit at the following intervals:
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,
- b. At least once per 31 days of accumulated operation in H0DE 1, and
- c. Within four hours if the AZIMU1RAL POWER TILT q(T ) is > 0.030.
See Special Test Exception 3.10.2.
+ Excluding the center CEA during Cycle 10.
l CALVERT CLIFFS - UNIT 1 3/4 2-6 Amendment No. JJ/JJ//E/7J/EE, 151
~
]
l POWER DISTRIBl1 TION LIMITS
$1lRVEILLANCE Rf0VIREMMISJContinued) 4.2.2.1.3 F xy shall be determined each time a calculation of F[y is required by using the incore detectors to obtain a power distribution map with all full length CEAs+ at or above the Long Term Steady State Insertion limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects.
4.2.2.1.4 T q shall be determined each t!me a calculation of F[y is required and the value Tq usedtodetermin0F[y shall be the measured value of Tq .
+ Excluding the center CEA during Cycle 10. ~
l CALVERT CLIFFS - UNIT 1 3/4 2-7 Amendment No. 2J/J1/71. 151 a .
l 9
- i E UNACCEPTABLE E OPERATION .-
(1.7 ,1. M REGION ! N,- e- 1.0 us ' M,, 3 o ! . en n. l Fh LIMIT CURVE i G m w 0.9 I ? H
- O eas
~
m ACCEPTABLE Jt1.785,0.801
+
u_ 0.8 -- - 0 OPERATION M Z REGION . a o -
~ i-e ;
u- 0.7 - ; m . E t ! m i ! < l 3 ' o 8 i , s
,Y, < 0.6 -
E l 1.65
~
t 1.70 I 1.75
< 1 1.80 l
. pr ;
- XY l y
' % 1 OS . = Figure 3.2-3a TOTAL PLANAR RADIAL PEAKING FACTOR vs ALLOWABLE FRACTION OF RATED THERMAL POWER -
POWER.DISTRIBllTION LIMITS TOTALPLANARRADIALPEAKINGFACTOR-F[y LIMITING CONDITION FOR OPERATION 3.2.2.2 The value of N presently used in Specification 4.2.1.3 shall be in accordance with Figure 3.2 3b. APPLICABILITY: MODE 1 when operating in accordance with Specification 4.2.1.3. ACTION: With the value of N presently used in Specification 4.2.1.3 exceeding the limit shown in Figure 3.2-3b, within 6 1ours either:
- a. Reduce the value of N used in Specification 4.2.1.3 to within the limits of Figure 3.2 3b; or
- b. Be in at least HOT STANDBY.
SURVEILLANCE REQUIREMENTS 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2.2 F[y shallbecalculatedbytheexpressionF}y-F (1+T ) and q N shall be determined to be within its limit by monitoring Fxy at the following intervals:
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,
- b. At least once per 3 days of accumulated operation in MODE 1.
4.2.2. F xy shall be determined each time a calculation of F[y is required by using the incore detectors to obtain a power distribution map with all full length CEAs+ at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid
. effects.
4.2.2.2.4 T y shallbedeterminedeachtimeacalculationofF[y is required and the value of Tq usedtodetermineF[y shall be the measured value of Tq . l
+ Excluding the center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 2 8 Amendment No. /J 151
,..,--..m_.4 - . - - - - - . - - - - . . - - --- - -. - y
'3M,a@POWERDISTRIBUTIONLIMITSTOTALINTEGRATEDRADIALP LIMIIIELCQHDITION FOR OPERATION 3.2.3 The calculated value of Ff, defined as Ff = Fr(1+Tq ), shall be limited to s 1.650.
APPLICABILITY: MODE 1*. ACTION: WithFf>1.650within6hourseither:
- a. Be in at leasi i;0T STANDBY, or
- b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and Ff to within the limits of Figure 3.2.3c, withdraw the full length CEAs+ to or beyond the Long Term Steady State Limits of Specification 3.1.3.6,andinsertnewvalueofFfinBASSS;or
- c. Reduce THERMAL POWER to bring the combination of THERMAL POWER and Ff to within the limits of Figure 3.2-3c and withdraw the full length CEAs+ to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from Figure 3.2-3c shall then be used to establish a revised upper THERMAL POWER level limit on Figure 3.2 4 (truncate Figure 3.2 4 at the allowable fraction of RATED THERMAL POWER determined by Figure 3.2-3c) and subsequent operation shall be maintained within the reduced acceptable operation region of figure 3.2-4.
SURVEILLANCE RE0VIREMENTS 4.2.3.1 The provision of Specification 4.0.4 are not applicable. 4.2.3.2 FT shall be calculated by the expression Ff F1
. _ shall be determined to be within its limit at the followin(g+T 18tervals: ) and Ff
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, l b. At least once per 31 days of accumulated operation in MODE 1, l and
- c. Within four hours if the AZIMUTHAL POWER TILT q(T ) is > 0.030.
~
- See Special Test Exception 3.10.2.
+ Excluding the Center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 2-9 Amendment No J2/JJ/JS//E/7J/ 88, 151 / - _ - - - . - . - - - .
SMayflLLANCE REDUIRDtEKTS (Continued) 4.2.3.3 Fr shall be determined each time a calculation of F[ is required by using the incore detectors to obtain a power distribution map with all full length CEAs+ at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. 4.2.3.4 T qshall be determined each time a calculation of Ff is required and the value of Tq used to determine Ff shall be the measured value of T. q
+ Excluding the center CEA during Cycle 10.
l r CALVERT CLIFFS - UNIT 1 3/4 2-10 Amendment No. JJ/JJ, 151
0.65,t m e
- 1.0 9
- UNACCEPTABLE i
G , e OPERATION g g -. REGION
-i p o a_
7 4 - 0.9 - T 2 E 6 . U h . Fr LIMIT CURVE o us W - l
- < 0.8 - .
[ ACCEPTABLE o OPERATION
- l. z ' REGION 9
e-O M < 0.7 - (1.72. 0.70) .
> m to i
c, "I 4 3 . O, 0.6 - 4 F a,
' I f 0.5 E 1.60 1.65 1.70 . 1.75
# T l
g - F, L i m 4 !. m Figure 3.2-3c l- . TOTAL INTEGRATED RADIAL PEAKING FACTOR vs ALLOWABLE FRACTION OF RATED THERMAL POWER
I~ 1.10 , , , , , 2.00 - (.0.1.1.00) ; * (0.15,1.00) - TNACCETTA3tI W ACCIFTA31.I 0FIRATION OPERATION 0.90 - RICION AICION - 0.80 - (.0.3. 30) y ACCIM.AILE 4 (0.3. 80) - OPERATION RICION 5E E ;- 0.70 - , i d*< n. 0.60 - -
%a
~
E
- - 0.30 -
(.0.3. 50) * - gI g l 0.40 -- -
'Ud E4 0.30 - -
0.20 - - o ( .45. 15) 0.10 - - 0.05 '
-0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 JDIRERA1. AIIAL SMAFI INDEX. Y g Tigure 3.2 4 DNA Axial Flux Offsat Control Limits CALVERT CLITTS - Qn\T \ 3/4 2 11 Amendment No. 22, 24. 32. 32.
! jf. 48. 72. 184. 130 L l L --
_ POWER D?STR1B'."10N LIMITS 1 g,g AZIMUTHAL POWER TILT - T, [1MITINGCOCITIONFOROPERATION 4 3.2.4 The AZIMUTHAL POWER TILTq(T ) shall not exceed 0.030. APPLICABILITY: MODI 1 above SD!i of RATED THERMAL POWER.* ACTION: ,
- a. With the indicated AZIMUTHAL POWER TILT determined to be > 0.030 l but 1 0.10, either corre:t the power tilt within two hours or detemine within the next 2 hours and at least once per subse-quent 8 hours, that the TOTAL PLANAR RADIAL PEAKING FACTOR (FI) y and the TOTAL 1hTEGRATED RADIAL PEAKING FACTOR (Ff) ,
the limits of Specifications 3.2.2 and 3.2.3.
- b. With the indicated A2IMUTHAL POWER TILT determined to be > D.10.
operation may proceed for up to 2 hours provided that the TOTAL T INTEGRATEDRADIALPEAKINGFACTOR(F)andTOTALPLANARRADIAL 7 I PEAKING FACTOR (Ffy) are within the limits of Specifications 3.2.2 and 3.2.3. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to < 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2 The AZIMUTHAL POWER TILT shall be deterwined to be within the limit by: l i a.. Calculating the tilt at least once per 12 hours, and
- b. Using the hcore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours when one excore channel is l . inoperable and THERMAL POWER IS > 75% of RATED THERHAL POWER.
l l See Special Test E.xception 3.10.2. I , CALVERT CLIFPS - UNIT 1 3/4 2-12 Amendment No. 21, 32
+
4 POWER DISTRIBUT10N LIMITS , 4 gg,q DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:
- a. Cold Leg Temperature ,
- b. Pressurizer Pressure
- c. Reactor Coolant System Total Flow Rate
- d. AXIAL SHAPE INDEX, Core Power APPLICABILITY: MODE 1. ,
ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or retuce THERMAL POWER to less than 1 5% of RATED THERMAL POWER within,the ner.t 4 hours , i SURVEILLANCE REQUIREMENTS ) 4 . 2. 5.1 Each of the parameters of Table 3.21.sha11 be verified to be within their limits at least once per 12 h6urs. ,
~
4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 shonths. I I l l CALVERT CLIFFS-UNIT 1 3/4 2-13 AmendmentNo.39.EA71
._. m _. _. --n
.*~. -
TABLE 3.2-1 o
# . DNB PARAMETERS p LIMITS l
4 - l 3 Four Reactor Coolant Pumps Three Reactor Coolant Pumps Two Reactor Coolant Pumps Two Reactor Coolant Pwnps a Parameter Operating _ Operating Operating-Same Loop Operating-Opposite Loop _ ). 4 5 Cold leg Temperature < 548'F ** ** ** f Pressurizer Pressure > 2200 psia
- 1
! Reactor Coolant System ! Total Flow Rate .> 370,000 gpm
~
AXIAL SHAPE INDEX - R~ l 7
- Limit not applicable during either a THERMAL POWER rag increase in excess of 5% of hTED THERMAL POWER I per minute or a TERMAL POWER step increase of greater than 10% of RATED THERMAL POWE4.
**These values left blank pending NRC approval of ECCS analyses for operation with less than four reactor coolant pumps operating.
4 ***The AXIAL SHAPE INDEX. Core Power shall be maintained within the limits established by the Better Axial Shape Selection System (BASSS) for CEA insertions of the lead bank of < 55% when BASSS is k OPERABLE, or within the limits.of FIGURE 3.2-4 for CEA insertions specified by FIGURE 3.1-2. a . . P. . a - E 5 s- .. 4 L,-. . ._, _ . . . - - . _ _ _ _ . . .- ,m=--. ---._.-m_:_.m - - .- _,_.,__y,., _,,_,,m , , . _. _ . _ - . _ _ - . _ ____.o
3, . , g . .,.
- t. e.. q =. = ... .. ; *. .*. ... q 2
- .3.1 EEa: :: :t: !: :lI :';STRUMENT*TI:'.
I
'.: '! Tit:3 COS:: :C'; FCE 0:ER T:0N
- 3. 3.1.1 As a tinimu~., the reactor prcte:tive instrumentation channels and bypasses of Table 3.31 shall be OPERAB'.E with RESPONSE TIMES as snown in Taoie 3.3-2. .
ACPLICAEILITY: As shown in Table 3.3-1.
' ::T:0N:
l.
'{
I As snoWB ir. Table 3.3*l . s l'
! . S '. :'/ E :LL:'.*E :!!U::E'T. I 4
4 . 3.1.1.1 Each re' actor prote::ive instrumentation channel shall be ost.cnstrate 0:ERAELE ty the performan:e of the CHANNEL CHECK, CHANNEL CAL 1EEAT10N an: CHANNEL FUNCTIO"AL TEST ccerations during the MODES, and a: :ne fretuen:ies sh:ar in Table 4.3-1. 4 . 3.1.1. 2 The iopic for the byDasses shall be demonstrated OPERAELE l ;rict 0 ea:". "ea:10" startup unless pe*forret during the preceding 92 cays. The tetti tjpass function shall be cemonstrate: OPERABLE at least once per 1E r:htns cerin; CHANNEL CALIBRATIO" testing of each channel affected by tystss c:eration.
- 4. 3.1.i . 3 ine REACTO:, TRi? SYSTEM RESPONSE TIME of eacn reactor trip fun::icn sr.all de ce ons; rated to be within its limit at least on:e per Ea:n test snail include at leas one channei per function
[1Econths. such that all cnannels are tested at least once every N times 18 months where it is tne total nur.ber of redundant channels in a specific reactor trip function as shown in the " Total No of Channels" column of Table 3.3-1. I t CALVERT CLIFFS . UNIT 1 3/4 3 1
TAIILE 3.3-1 F f.AC T 0,R,_ PRO T E C,T_Iy_IJi,5 T__R_WTfl I A l ipt
'A HiNinm 1OTAL NO. CIIANfill.S CilAfmlI.S APPLICAlllC P _TUNCTIONAL l! NIT TO T_it_I_P.
_O_f _C_II_AN_N_f_t._S_. O.PritAftlL
. . . M_O_D_ES .A.C.i.l oti 5 1. Manual Reactor Trip 2 1 2 1, 2 and
- 1 g 2. Power Level - Iligh 4 2 -
3(f) 1, 2 2f , a [ 3. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1, 2 (c) 2f i
- 4. Pressurizer Pressure - Ilielb 4 -
2 3 1, 2 27 l
- 5. Containment Pressure - Iligh 4 2 3 1, 2 2f !
- 6. Steam Generator Pressure - I.ow 4/SG 2(b)/SG 3/5G 1, 2 27
- 7. Steam Generator Water
{ Level - Low 4/SG 2/SG 3/5G 1, 2 ?# { 8. Axial Flux Offset 4 2(c) 3 1 7#
- 9. i . thermal :tergin/ Low Pressurc 4 Z (.3 ) 3 1, / (c) zi
- h. Steam Generator Pressure Di f ference - liigh 4 2(a) 3 I, 2 (c) 2# .
e M 10. Loss of Load 4 2(c) 3 1 27 i, S g i b ! E N 1 cm . L e .2 ! i i
i 3 n TABLE 3.3-1 (Continued)
? -
i M REACTOR PROTECTIVE INSTRUMENTATION 1 % ' ! P
- MINIMUM 5 . TOTAL NO. CHANNELS CHANNELS APPLICABLE TO TRIP OPERABLE MODES ACTION 7 FUNCTIONAL UNIT . OF CHANNELS
'C
= 11. Wide Range Logarithmic Neutron .
-4 Flux Monitor
- a. Startup and Operating--Rate ;
of Change of Power - High 4 -2(d) 3(f) 1, 2 and
- 2f I .
0 2 3, 4. S 3 l
- b. Shutdown . 4
- 12. Reactor Protection System . 6 1 6 1, 2
- 4 Logic Matrices. -
3/ Matrix 4/ Matrix 1, 2* 4 f
- 13. Reactor protection System 4/ Matrix R Logic Matrix Relays t 8 1, 2* 4 w 14. Reactor Trip Breakers 8 6 .
i l 4 [ i ~ 4 i t i I
+-
i i I I'
. i TABLE 3.3ol (Continued)
TABLE NOTATION
*With the protective system trip breakert in the closed position and I the CEA drive system capable of CEA withdrawal. l I The provisions of Specification 3.0.4 are not applicable. f of RATED THERPAL POW !
(a) Trip may be bypassed below 10'4be automatically removed!
~
of RATED THERRAL POWER. , (b) Trip may be manually bypassed below 785 psia; bypass shall be automatically removed at or above 785 psia. (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > ~ 15% of RATED THERMAL POWER. (d) Trip may be bypassed below 10'#% and above 12% of' RATED THERuAL POWER. (e) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3. ! (f) There s' hall be at least two decadei of overlap betwee'n the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels. ACTION STATEMENTS ACTION 1 - 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 protective system trip breakers. ACTION 2 .- With the number of OPERABLE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: I
- a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour. For the purposes
- of testing and maintenance.'the inoperable channel may be bypassed for up to 48 hours from time of initial loss
!. of OPERABILITY; however, the' inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. I CALVERT CLIFFS - UNIT 1 3/4 3-4 Amendment No. AB, fN,
-- - . - ,n. .,- . , , . , . --
. . _ _ - . . - . - _ - - - - . . . - . - - _ _ = _ _ _ . . . . _ - .-
. TAELE ".*-1 I cetinued)
- ACTION STATEMENTS
- b. Within one hour, all functional units-receiving an input from the inoperable channel are also placed in
. the same condition (either bypassed or tripped, as applicable) as that required by a. above for the
. inoperable channel,
- c. -The Minimum Channels OPERABLE recoirement is met; however, one additional channel may be bypassed for up to 48 hours while performing tests and maintenance on that channel provided the other inoperable channel is claced in the tripoed conditjen.
ACTION 3 - With the numt-er of channels 0:ERAELE cre less than recuired b;> the Minimum Channels OPERAELE reevirement, verify c:::it-ance with the SHUTDOWN MAR 31N requirements of Specification 3.1.1.1 or 3.1.1.2,-as applicable, witnin 1 hour and at least once per 12 hours thereaf ter. ACTION 4 - With the number of channels OPERABLE one less than reevired by the Minimum Channels 0FEkASLE. requirement, be in HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 1 hour.for surveillance testing per Specification 4.3.1.1. l L CALVET,7 CLIFFS - UNIT 1 3/4 3-5 ; l
n . TABLE 3.3-2 N - y REACTOR PROTECTIVE _ INSTRUMENTATION,,RE,SPON,SE_T.I_ME_S 1
~* ,
n I {n _rUNCTIONAL UNIT RESPCNSE TIME [ 1. Manual Reactor Trip Not Appifcable E 2. Power Level - High . __ 0.40 seconds *f and _< 12.0 secon.ds if
. 3. Reactor Coolant Flow - Low < 0.50 seconds I j 4. Pressurizer Pressure - High
. _1 0.90 seconds l . 5. Containment Pressure - High - 0.90 seconds i
- i
- 6. Steam Generator Pressure - Low - 0.90 seconds
- 7. Steam Generator Water Level - Low y -.0.90 seconds
- 8. 5xial Flux Offset < 0.40 seconds *f and < 12.0 seconds if !
') . a . Thermal Margin / Low Pressure - 0.90 seconds *f and 12.0 secnnds #8
- h. Steam Generator' Pressure Di f ference - Ifigh
] 0.90 seconds .
i
- 10. Loss of Load Not 3ppilcable I[
.r E 11. Wide Range Logarittwic Neutron flux Monitor Nnt Applicable i
z
- Neutron detectors are exempt from response time testing. Response time of the neutron flux sign 11 portion P {
of the channel shall be measured from detector output or input of first electronic component in thannel. '
# Response time does not include contribution of RIDS.
n ifRTil response time only. This value is equivalent to the time interval required for the RIDS output i to achieve 63,2% of its total change when subiected to a step change in RTD temperature. ' M ;_
.. . . . - - . _ . - - - . . . ~ . . - . ~ - . . . - - - - . - _ - . . . . _ _ - - . . - - . -
~.:.
-- ' ~.
i O si - 5 E ~- m C' C.
~ C . ,* ,
'.V
-- v. N N N N t., N
- C h
- 6. Q,
.g=, .
- .- - - - - - .- - =
4
. G, n ,
s MI,
==
s- n n z! -r - s E., s. w ,- = =- s ,
%z . v 5 m mz z r z z z z = = m , :
ei w 5' , l u, - 6.: , C.' *
. .S. g u
.m.
v
.d_ ', g Q=,
C =. *. x. w,
- h, E =. N 'r aC E 5:.c : ETE' = = = = = u = = a?
.. =
, .: s.. --
c,, et . Ti; .C ,
"8 j ,
, w l 7: E' <c.* E .. l b
. JMl R.
E V; N. V.- - r, . g! 6 =c: m% re e a m e e r '., m z\ E.
. v W. ,
2: u W8 v. l- > c i
-l u <
I- >=, - 6 c
. v i, -
n - l gi e :.- - c
~- - a b
~ ~~~a i E O
.E m. o u 8 a = = E u
a-a v i i -i i m : e U a Cl
- w. n u o m
e L a n' u El :. C 6 L es c A
.- .= - a a E <
% 6 j L mL 6 m e L J; =J
= C
>= = m M U s
= e 8
E E b e b w e e, ed m
- l. .e., w 6 6 O 4 A A ed u
iAL - c a
=
w m E us o e r: o e
-o m
ma
=
W
-6 om
>0
- 8W o N
a I w 3 - e f=D o
.J
. y a. u n a. L 6 -
n w. a .= m -w
.s o m= c
- - 3 >= .c, = =
w=- u m <: %' y ad. m v
. m e .=, 5 5o -
m m 5 m a w, .s .- - .m m- . . o E e .v.
. x e i
z L 4 .2 E L U W .M eC J
.: .c .
.e
- es en - m s cc e
. wl. -
CALVEF.T CLIFFS.- UNIT 1 3/437 /r,endment No.48. T3 l
n TARLE 4.3-1 (Contintred) N M REACTOR PROTEf.f..!VE INS.T.Rl! MENTATION .SilR.VrIII. A. .liC.E RE.Qtil.RTHENTS CllANNEL MODES IN HillCII 4 CilAt!NEL CilAHfift IllNCTIONAL SIIRVEltl AflCE FUNCTIONAL UNIT Clll CK CAI IllRAll0N TEST __ Rt(lIllRtli { _ E 11. Wide Range Lngaritimilc Neutron S !!(5) 5/U(1) 1, 2, 3, .4, G Flux Monitor S anel *
- 12. Reactor Protection System logic Matrices N / /\ A H,I[]4 t! and S/U(1) 1, 2
- 13. Reactor Proterduo System logic Matrix b 3ays ,lkI. iib ,td. O@ H and S/U (1) 1, 2
{ 14. Reactor Trip Breakers ,R<n. IH - N,f. Oq (1 1, 2 and
- Y
= o W
TABLE 4.3 1 (Continued) TABLE NOTATION
* - With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
(1) - If not performed in previous 7 days. (2) - Heat balance only, above 15% of RATED THERMAL POWER; adjust
" Nuclear Power Calibrate" potentiometer $ to make the nuclear power signals agree with calorimetric calculation if absolute difference is > 1.5%. During PHYSICS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.
(3) - Above 15% of RATED THERMAL POWER, recalibrate the excere detectors which monitor the AX1AL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to < 90% of the maximum allowed THERMAL POWER level with the existTng Reactor Coolant Pump combination. (4) - Above 15% of RATED THERMAL POWER, adjust "6T Pwr Calibrate" potentiometers to null " Nuclear Pwr - AT Pwr". During PHYSICS. TESTS, these daily calibrations of nuclear power and 4T power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next test power plateau. (5) - Neutron detectors may be excluded from CHANNEL CALIBRATION. , i CALVERT CLIFFS - UNIT 1 3/4 3-9
, , - _.y,- -y , w, , - - - , . -c.-, , .,m, , - . . , _ _ . _ . . _ _ _,_ . _m ___ -m_ _ __ _ _ _ _m, _ . _ _ _ _ _ _ _ _ . _ _ ,2,m____,.,___________.__.,_.__
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION l LIMITING CONDITION FOR OPERATION l 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE ! with their trip setpoints set consistent with the values shown in the i Trip Setpoint 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. ACTION: . j a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values , L column of Table 3.3-4, declare the channel inoperable and 1 apply the applicable ACTION requirement of Table 3.3-3 until I the channel is restored to OPERABLE status with the trip set- ,l l point adjusted consistent with the Trip Setpoint value. '
- b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3. !
SURVEILLANCE REQUIREMENTS 1 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2. 4.3.2.1.2. The logic for the bypasses shall be demonstrated OPERABLE l during the at power CHANNEL FUNCTIONAL TEST of channels affected by L bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. CALVERT CLIFFS - UNIT 1 3/4 3-10
1 i .
- g TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTimTION SYSTEM INSTMtmENTATION
~
i -p
~ -
RINinM 7 TOTAL NO. CHMWELS CHMWELS APPLICABLE 7 FUNCTIONRL UNIT OF CHMMELS TO TRIP OPERABLE MODES _ ACTION E 1. SAFETY INJECTION (SIAS)9 , t U a. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6
- b. Containment Pressure - High 4 2 3 1,2,3 7*
- c. Pressurizer Pressure - tow 4 2 3 1,2,3(a) P
- i. 2. CONTAllmENT SPRAY (CSAS) a.. Manual (Trip Buttons) 2 1 2 1, 2, 3, 4 6 i
! b. Containment Pressure - High -4 2 3 1, 2, 3 11 w t 1 3. CONTAINENT ISOLATION (CIS)I 1 w- -
.a. Manual CIS (Trip Buttons) 2 1 2 1, 2, 3. 4 6
- i
~
- b. Containment Pressure - High 4 2 3 1, 2, 3 7*
[ , i l E. l
~
# Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and
.1.c). !
g 9 When the RCS temperature is. (a) Greater than 350'F. the required OPERABLE HPSI pumps must be able to start automatically upon g receipt of a SIAS signal, ' li g (b) Between 350*F and 327'F, a transition region exists where the OPERABLE HPSI pump wI11 he placed in g pull to-lock on a cooldown and restored to automatic status on a heatup, r g (c) At 327'F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start g automatically. m j i
m TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION n
#- MINIMUM M TOTAL NO. CHMtNELS CHANNELS APPLICABLE
$ FUNCTIONAL UNIT Of EHANNELS TO TRIP OPERABLE MODES ACTION i b 4. MRIN STEAM LINE ISOLATION l'
?
- a. Manual (MSIV Hand Switches i and Feed Head Isolation Hand E Switches) 1/ valve 1/ valve 1/ valve 1, 2, 3, 4 6
- b. Steam Generator Pressure -
l~ Low 4/ steam 2/ steam 3/ steam I, 2, 3(c) 7* generator generator generator ! t
- 5. CONTAlfMENT SUMP RECIRCULATION
] (RAS) ;
Y a. Manual RAS (Trip Buttons) 2 1 2 1, 2, 3, 4 6
~ l
- b. Refueling Water Tank - Low 4 2 3 1, 2, 3 7*
l I n a R
?.
- 5 b
4.79 I
1 i p LABLE 3,3-3 fCont_]puedl G, , 9 DIGINEERED 5AFETY FEAftfat ACillAiION SYE EN INSYNtlNDITATleft . [p. p - 5 ( L k i. h 5 TI A I
;; usurMai n c o-s s n g S TOTAL NO. C9tMOIELS CMMWIELS APPLI E)
. D98tTIemt. teIIT IF CMasunts 1o TRIP erEnnsti Mootsscifois 1
- '6. CONTAIIDENT FURGE VALVES ISOLATI9ll
~
! a. Manual (Purge Valve l Control Switches) 2/ Penetration 1/ Penetration 2/ Penetration 6** 8 i
- b. Containment Radiation -
i High Area Nealter 4 2 3 6** 8 ! 7. w L955 SF POWER 1 y a. 4.16 kv Emergency Bus Underveltage (Less of
" Voltage) 4/ Bus 2/8us 3/ Bus I, 2, 3 1*
i
- b. 4.16 kw Emergency Bus Underveltage (Degraded Voltage) 4/ Bus 2/Bes 3/ Bus 1, 2,- 3 7*
U$I
.~ ~ , _
a i G2 ** l =3 Must be SPERABLE only in MODE 6 when the valves are required SPDtABLE and they are open.
, 5 g .
i q ^ 0 t . t m
l' TABLE 3.3-3 (Continued) I [>- ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
;;i i :o
- MINIMUM l
TOTAL NO. CHANffELS CHANNELS APPLICABLE- { FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION j $ 8. CVCS ISOLATION i c:
- a. Manual (CVCS 1/ Valve 1/ Valve 1/ Valve 1.2,3,4 5 Isolation Valve 6 j
[ Control Switches)
- b. West Penetration 4 2 3
~
j Room / Letdown Heat 1, 2, 3, 4 7* Excharger Room i Pressure - High ! = 9. AUXILIARY FEEDWATER w ACTUATION SYSTEM (AFAS)
- a. Manual (Trip 2 sets of 2 1 set of 2 -
l Buttons) per S/G 2 sets of 2 1,2,3 6
.per S/G per S/G
}
- b. Steam Generator 4/SG 2/SG 3/SG 1, 2, 3 7 Level - Low 3" c. Steam Generator 4/SG
{ 2 aP High 2/SG 3/SG 1, 2, 3 7 a O 3 e
TABLE 3.3 3 (Centinued) TABLE NOTATION _ (a) Trip function may be byussed in this MODE when pressurizer pressure is < 1600 psia; bypass shall be automatically remond when pressurizer pressure is ?.1800 psia. (r.) Trip function may be bypassed in this MODE below 785 psia; bypass shall be automatically removed at or above 785 psia.
- The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel
- to OPERABLE status within 48 hours or be in at least
- HOT STANDBY within the next 6 hours and in COLD SHUTDOWN w thin the following 30 hours. i ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided
- the following conditions are satisfied:
- a. The inoperable channel is placed in either the bypassed -
or tripped condition within 1 hour. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.
- b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel,
- c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.
3/4315 Amendment No.AB, f/}', 8 $ CALVERT CLIFFS - UNIT 1
TABLE 3.3-3 (continued) ACTION 8 - With less than the Minimum Channels OPERABLE, opi7ation may continue provided the containment purge valver,,are i maintained closed. ACTION 11 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels 0PERABLE requirement is demonstrated within 1 hour; one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, CALVERT CLIFTS - UNIT 1 3/4 3-16 Amendment No. log I
TABI.C 3.3-4 ENGINEERED _ SATETY FEATURE AClilAtl0H SYS1[M INSTRUMENTATION TRIP VALUES "n
--4 Alt 0WABl0 P VALUES M TUNCTIONAL UNIT T_ RIP SEIPOINT_ <n . 1. SAFETY INJECTION (STA5)
Not Applicable Not Applicabic c- a. Manual (Trip Duttons) 5 1 4.75 psig 1 4.75 psig
~* b. Containment Pressure - liigh
; 1725 psia 3 1725 psia
- c. Pressurizer Pressure - inw E. CONTAINMENT SPRAY (CSAS)
Not Applicable Not Applicable
- a. Manual (Trip Buttons)
- b. Containment Pressare -- liinh 1 4.75 psig i 4.75 psig
- 3. CONTAIPMENT ISOLATION (CIS) g .
Manual CIS (Trip Buttons) Not Applicable Not Applicable M a. u Containment Pressure - iligh _ 4.75 psig 1 4.75 psig i' b.
- 4. MAIN STEM 1 LINE ISOLATION -
- a. Manual (MSIV Hand Switches and Feed llead Isolation Not Applicable Not Applicable y lland Switches)
S (65 psia g fib 5 psia l 3 b. Steam Generator Pressure - tow P f Containment isolation of non-essential penetrations is also initiated by SIAS (functional E units 1.a and 1.c). w t . OO
TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUMENTATION TRIP VALUES
'biIo"
--I -4 VP ALLOWABLE VALUES
[O
; F1MCTIONAL UMil TRIP YALUE e b 5. CONTAlfWlENT SIMP RECIRCULATION (RAS) 55
-
- Not Applicable Not Applicab7e
- a. Manual RAS (Trip Buttons)
W" 1 24 inches above
- b. Refueling Water Tank - Low > 24 inches above tank bottom tank bottom
- 6. CONTAlletENT PURGE VALVES ISOLATION Manual (Purge Valve Control Switches) Not Applicable Not Applicable e a.
1 Containment Radiation - High w b. Area Monitor 1 220 mr/hr s 220 mr/hr CO
- 7. LOSS OF POWER
- a. 4.16 kw Emergency Bus Undervoltage 2450 i 105 volts with a 2450 i 105 volts with a (Loss of Voltage) 21 0.2 second time delay 2 1 0.2 second time delay
! b. 4.16 kv Emergency Bus Undervoltage 3628 1 25 volts with a 3628 1 25 volts with a (Degraded Voltage)
~
8 1 0.4 second time delay 8 1 0.4 second time delay iR ll3 n ". - ii D
?A Lt
?> C; 1, CD l
- 1 s
TABLE 3.3-4 (Continued),
,9 ENGINEERED SATETY FEATURE ACTU_ATION SYSTEll INSTRUMINTATI_0N TRIP VALUES, E ALLOWABLE FUNCTIDNAL UNIT TRIP VALUE VALUES
[ r-
$ 8. CVCS ISOLATION Y West Penetration Room / <0.5 ps19 1 0 5 psig c Letdown lleat Exchanger y Room Pressure - liigh ~
- 9. AUXILIARY FEEDWATEP ACTUATION SYSTEM (ATAS)
- a. Manual Trip ttons) Not Applicable Not Applicable w b. Steam Generator (A or B) -149 inches to -194 -149 inches to -194 2 Level - Low inches (inclusive) inches (inclusive) w i c. Steam Generator AP - liigh -<l35.0 psi -<135.0 psi (SG-A > SG-D) c
- d. Steam Generator AP - Ifig's -<135.0 psi -<135.0 psi 5 I (SG-B > SG-A)
I l 1 E a 9 EO 3 o . ... o - N C
~
g
- .. - .- . - . _ - . - . - . . . _ - -.- . . . . . . . - . - . . _ . - ~ - - . - . - ~ . - - _.-. ~ ..... ... .
I T ABtr 3 3-S' i'- - ENGINTERED S ATETY TE ATt*RES RESPONSE Ti\tES TNITI ATING SIGN AL AND' TLNC" TION' RESPONSF TI\fE iN SECONDS 1i hi1!.223 a; SIAS Safety Injection (ECCS) Not Applicable ' b. CSAS Containment Spray Not Appli:able
- . CIS -
Containme:t- Isolation Not App 11:able d, . RAS Containment Sump- Re:irculation Not Applicable e', AFAS Auxiliary Feedwater Initiation Not Appli:able 2; Prettur!'er P'ettu*e-L ew
- a. Safety inje: tion (ICCS) s 30*/30 " 1'
.3.- cents!nment Prettuee-Wk
- a. . Safety injection (ECCS) -s 30*/30 "
- b. Containment Isolation s 30
- c. Containment Fan Coolers s 35*/10 " .
- 4. Certainmem Pressure-Hkk Lai Containment Spray .
s 60*/60 " (1)
- 5. Centainment Radintlem Wh
- a. L Containment Purge Valves Isolation -
.s 7 4
- CALVERT CLIFFS - UNIT l' 3/4 3 20 . Amendment No. //, f/, ff,127
..,- - m+ - e = , . , . - - . . . . . - - , . . - - , . - .
TABLE 3.3 5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 6. Steam Generator Pressure-tow
- a. Main Steam Isolation b.
s 6.9 Feedwater Isolation s 80
- 7. Refuelina Water Tank-low
- a. Containment Sump Recirculation s 80
- 8. Reactor Trio
- a. Feedwater Flow Reduction to 5% s 20
- 9. Loss of Power
- a. 4.16 kv Emergency Bus Under-voltage'(LossofVoltage) 1 2.2***
- b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) s 8.4***
- 10. Steam Generator level-Low
- a. Steam Driven AFW Pump 1 180
- b. Motor Driven AFW Pump s 180
- 11. Steam Generator a P-Hioh
- a. Auxiliary Feedwater Isolation s 20.0 TABLE NOTATION Diesel generator starting and sequence loading delays included.
i: Diesel generator starting and sequence loading delays D21 included.
]
Offsite- power available. Response time measured from the incidence of the undervoltage condition to the diesel generator start signal. l (1) Header fill time not included. l l I CALVERT CLIFFS - UNIT 1 3/4 3-21 Amendment No. /S/J//7J/7J/ EE 1,49
T,ABIE 4.3-2 g ENGINEERED SAFEIY FEAVURE ACIDAIION SYSitM INSTRINtNIAll0N SURVEllLANCE REf]UIREMENT [G H CNANNEL MODES IN lallCil CilANNEL CilANNfL FUNCT*0NAL p IUNCTIONAL UNIT CllLCK CAllllRA110N 1EST SultVEILLANCE
.* ,; ltEf)UIRfD
" 1. SAFETY INJECTION (SIAS) 7 a. -Manual (Irip buttons) jffn.6% d.M R fff OA c b. Containment Pressure - liigh 'S 'N
- c. M. "I, 2, 3 21
-8 Pressurizer Pressure - Low S R M I , 2, 3
- d. Automatic Actuation Logic 1 - Jki.f W &W. fM M(1)(3) 1, 2, 3
- 2. CONTAll#KNi SPRAY (CSAS)
- a. Manual (Irip buttons) MX. G4 18d GA
- b. Containment Pressure-Iligh 5 R JtX ON 2
- c. Automatic Actuation Logic R M i 2, 3 VA.GR - Jki.OA- M(1)(6) I , 2, 3 M
l
- 3. CONTAllWENT ISOIAil0N (CIS)f
'r a. Manual CIS (Trip buttons) tt#. GA IMC Ota R O b. Containment Pressure-liigh S R 4
M 3X. 1, 2, 3 G A'
- c. Automatic Actuation Logic N/.OA p. DA - M(1)(4) 1, 2, 3
- 4. MAIN STEAM LINE ISOLATION (SGIS)
- a. Manual SGIS (flSIV llamt Switches and Fe d llead Isolation lland Swltches) 8(n. O A N,(. OA - R fin.9k y ~b. Steam Generator Pressure-Lew 'S X.
M ?, 3 2 c. Automatic Actuation Logic 1 a Jta.f\A JJf.1% M(I)(5) I, 2. J B f S g
- # Containment isolation of coni-essential penetrations is also initiated by SIAS (functicasi units 1.a s
, and 1.c).
GB 1 EIb i
; \ ,
t
?
TABLE 4.3-2 (Continued) 9 Q ENGINEERED SAFETV FEATURE _ ACTUATION SYST_E5_INSTRtNtEW7ATION SURVEILLAf0CE REQUIRENENTS 9 CHAINIEL MODES IN WHICM O CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE q FUNCTIONAL UNIT CHEJK CALIBRATION IEST REQUIRED E
. 5.- CONTAlfelD K SUNP RECIRCULATION (RAS)
- a. Manual RAS (Trip Buttons) NA NA R NA
- b. Refueling Water Tank - Low NA R M 1, 2, 3
*-~
- c. Automatic Actuation Logic NA NA M(1) I , 2, 3 -
- 6. CONTAllWIDIT PURGE VALVES ISOLATION
- a. Manual (Purge Valve Control Switches) NA NA R NA
- b. Containment Radiation - liigh S R M 6**
w Area Monitor 1 w 7. LOSS OF POWER
- a. 4.16 kw Emergency Bus Undervoltage NA R M 1, 2, 3 (Loss of Voltage)
- b. 4.16 kw Emergency Bus 8fndervoltage (Degraded Voltage) NA R M 1, 2, 3
.> 8. CVCS ISOLATION
".2 a West Penetration Room / Letdown y Heat Exchanger Room Pressure - High 4 MA R M 1, 2, 3, 4 3
2 9. AUXILIARY FEEDWATER o w a. Manual (Trip ' Buttons) NA NA R NA R b. Steam Generator Level - Low S R M 1, 2, 3 h c. Steam GeneratorAP - High S R M 1, 2, 3 R d. Automatic Actuation Logic NA NA M(1) I, 2, 3 D D ** Must be OPERABLE only in It0DE 6 when the valves are required DPERABLE and they are open.
TABLR 4.3-2 (Continued) -j
+
TABLE NOTAT!ON (1) The logic circuits shall be tested manually at least once per 31 days. (3) SIAS logic circuits A-5, B-5, A-10 and B-10 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. (4) CIS logic circuits A-5 and B 5 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. (5) SGIS logic circuits A-1 and B-1 may be exempted from testing during
, operation; however, these logic circuits shall be tested at least -
once per 18 months during shutdown. (6) CSA5 logic circuits A-3 and B-3 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. y i L l . I l-l l-( .' l' l CALVERT-CLIFFS - UNIT 1 3/4 3-24 Amendment No. 3, g g 9 e i
INSTRUMENTATION 3/4;3.3 MONITORING-INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION 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. APPLICABILITY: 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 radiction 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 REOUIREMENTS
~
4.3.3.1 Each radiation _ monitoring instrumentation channel shall be I demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3. L L 1 q CALVERT CLIFFS - UNIT 1 3/4 3-25 ,
4 1
..=
LTA8t E 3.3-6 RADIATION MONITORING INSTRUMENTATION i 'U-
.
- MINIMUM G CilANNELS . APPLICABLE ALARM / TRIP : MEASUREMENT-l j INSTRUMENT-OPERABLE: MODES SETPOINT RANGE ~ ACTION' I ~ l. AREAMONITOR3 l
- a. Containment
'@O '
\ '
i g 1. Purge'& Exhaust Isolation- 3 '6: s[mr/hr - 10-I - 104mr/hr . '6' 1 , [ b'. ' Containment Area High Rar.ge' 2 1, 2, 3, & 4 5 10 R/hr 1 - 108.R/hr 30.- !
- 2. PROCESS MONITORS a.
Containment _ q* 1. Gaseous Activity-y a). RCS Leakage Detection j 1 1, 2, 3, & 4 Not Applicable 10I - 106 cg; .34 ra >
- 11. Particulate Activity .
a): RCS Leakage Detection' 4 1 1, 2, 3, & 4 Not Applicable 10I - 106 cp, 34; CL . t
.lo b. Noble' Gas Effluent Monitors g 1. Main Vent Wide Range .I 1, 2, 3, & 4
- 10-7 to 100 pC1/cc 30 11.. Main Steam Header ~ 2 1, 2, 3, & 4
- 10-2 to 105 R/hr 30; b
.w R -
;$ * -- Alarm setpoint to be.specified in a controlled document (e.g., setpoint control manual). I D
y a
% i O
_ - , _ _. _ _ -, - . _ a
l TABLE-3.3-6 (Continued) TABLE NOTATION l . ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels 0FERABLE requirement, comply - with the ACTION requirements of Specification- 3.4.6.1. ! ACTION 16 - 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. s i j- ACTION 30 - With the number of channelt OPERABLELless than required L by the Minimum Channels' 0FERABLE requirement -initiate L the-preplanned alternate method of monitoring the appro-priateparameter(s),within72 hours,and:
-1) either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or .
12) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days follow-i ing the event, ott11ning.the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status. . 1 i e a db t 4 CALVERT CLIFFS - UNIT 1 3/4 3-27 l Amendment No. 99 1
A e TABLE 4.3-3 _g RADIAJIDH_ MONITORING 1,NSTRtlMENTATION StfRVEILLANCE REf)flIRD1ENTS C-m . E CilANNEL CllANNEL MODES IN WillCII CilANNEL FUNCTIONAL StfRVEILLANCE INSTRtlMENT CilECK - CALIBRAIION TEST REQti! RED 3 1. AREA MONITORS 4 [ a. Containment-5 1. Purge & Exhaust .
] Isolation S R M 6
- b. Containment Area litt,h Range S R M 1. 2, 3, & 4 y 2. PROCESS MONITORS y a.. Containment E 1. Gaseous Activity a) RCS Leakage Detection S R M 1, 2. 3, & 4
- 11. Particulate Activity p a) RCS Leakage g -Detection 5 R M 1, 2, 3, & 4 I
g b. Noble Gas Effluent i Monitors P 1. Main Vent Wide Range S R M 1, 2, 3, & 4 w w
, 11. Main Steam lleader 5 R M 1, 2, 3, & 4 w
h . p. 8 _ _ _ _ ___________________m__ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ - . . _ _ _ _ _ . _ _ _ . _ _ _ . - _ _ _
1 IltITESMATI(Mt - INCORE DETECTORS N Foe OPERATION _ 3.3.3.2 The incore detection system shall be OPERABLE with at least one l OPERABLE detector segment in each core quadrant on each of the four axial' clavations containinj, incore detectors and as further specified below:
- a. For monitoring the AZ11UTMAL POWER TILT: .
At least two quadrant symmetric incore detector segment groups at each of the four axial elevations containing incere detectors in the outer 184 fuel assemblies with suff cient OPERABLE detector segments in these detector groups to compute at least two AZIlWTHAL POWER
. TILT values at each of the four axi 1 elevations containing incore detectors.
- b. For recalibration of the excore neutron flux detector system:
- 1. At least 75% of all incore detector segments, '
l
- 2. A minimum of 9 OPERABLF. incore letector segments at each detector segment level, and
- 3. A minimum of 2 OPERABLE detector segments in the inner 109 fuel assemblies and 2 OPERABLE segmeets in the outer 108 fuel assemblies at each segment level,
- c. For monitoring the UNR000ED PLAMAR RADIAL PEAXING FACTOR,tne UNR000ED INTEGRATED RADIAL PEAXING FACTOR,.tr the linear heat rate:
- 1. At least 75% of all incere detector locations,
- 2. A minimum of 9 OPERABLE incere detector segments at each detector segment level, and
- 3. A minimum of 2 OPERABLE detector segments in the inner 109 fuel assemblies and 2 OPERABLE segments in the outer 108 fuel assemblies at each segment level.
An OPERABLE incere detector segment shall consist of an OPERABLE rhodium detector constituting one of the segments in a fixed detector string. An OPERABLE incere detector location shall consist of a string in which at least three of the four incore detector segments are OPERABLE. CALVERT CLIEFS - tl NIT 1 3/4 3-29 Amendment No. 34 II6, 129
INSTRUMENTATION l LIMITING ColeTTION FOR OPERATION (Continued) i An OPERA 8LE quadrant syimmetric incore detector segment group shall consist of a minimum of three OPERABLE rhodium incore detector segments in 900 syimnetric
-- fuel assemblies.
APPLICABILITY: When the incore detection system is used for:
- a. Nonitoring the.AZImlTHAL POWER TILT,
- b. Recalibration of the excore neutron flux detection system, or
- c. Nonitoring the UNR000ED PLANAR RADIAL PEAKIM FACTOR, the UNR000ED INTEGRATED RADIAL PEAKIM FACTOR, or the linear heat rate.
AGIlatt: With the incore detection system inoperable, do not use the system for the i3bove applicable monitoring or calibration functions. The provisions of gpecificatior43.0.3 and 3.0.4 are not applicable.
- SURVEILLANCE REQUIREMENTS
.4.3.3.2 The incore detection system stia11 be demonstrated OPERABLE:
- a. By performance of a CHAMEL CHECK within 24 hours prior to its use and at least once per 7 days thereafter when required for:
- 1. Monitoring.the AZINlmiAL POWER TILT.
- 2. Recalibration of the excore neutron flux detection system.
lL 3. Monitoring the UNR000ED PLANAR RADIAL PEAKING FACTOR, the 1 UNR000ED INTEGRATED RADIAL PEAKING FACTOR, or the linear I heat rate. ,
- b. At least once per refueling interval by performance of a CHANNEC CALIBRATION operation which exempts the neutron detectors but l i includes all- electronic components. The neutron detectors shall be
-calibrated prior to installation in the reactor core.
i l L CALVERT CLIFFS - UNIT 1 3/4 3-30 Amendment No. Jp,129
l INSTRUMENTATION SEISMIC INSTRUMENTATION l g LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic moaitering instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more seismic monitoring instruments inoperable for
! more than 30 days, prepare and submit a Special Report to the )
- - Comission pursuant to Spe
- ification 6.9.2 witnin the next-10 l days outlining.the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status, i
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not .
l , applicable. , SURVEILLANCE REOUIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies j shown in Table 4.3-4 4.3.3.3.2 Each of the above seismic monitoring instruments actuated i during a seismic event shall be_ restored to OPERABLE status within 24 l_ hours and a CHANNEL CALIBRATION perfortned within 5_ days following the seismic. event. Data shall be retrieved from actuated instruments and l analyzed to detemine the magnitude of the vibratory ground motion. A ! Special Report shall be ' prepared and submitted to the Commission pursuant L to Specification 6.9.2 within 10 days describing the magnitude, frequency 1 spectrum and resultant effect upon facility features important to safety. l l l I CALVERT CLIFFS - UNIT 1 3/ 4 3- 31
I i TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENT
-INSTRUMENTS AND SENSOR- LOCATIONS RANGE OPERABLE
- 1. 1 Triaxial Time-History Strong Motion Accelographs s.
- a. 0-YE-001 Unit 1 Containment Base 0-19 1-l
- b. 0-YE-002 Unit -1 Containment 69' 0-19 1 1
- c. 0-YE-003 Auxiliary Bldg. Base L -
0-19 1
- d. 0-YE-004 Intake Structure 0-1g 1
- e. 0-YE-005 Free Field 0-1 9 1
- 2. Triaxial- Seismic Switches I
- a. 0-YS-001 Unit l' Containment Base ;,pfX7fy4: 1
- b. 0-YS-002 Unit 1 Containment 69' )k%.OA 1 d
- 3. Seismic Acceleraticn Recorder
- a. 0-YRC-001 Control Room-JRfA.fd4 1
- b. 0-YR-001 Control Room
,3<A. (gA ' .1-i CALVERT CLIFFS - UNIT 1 3/4 3-32
TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONA'. INSTRUMENTS ANO SENSOR LOCATIONS CHECK ** CALIBRATION TEST
- 1. Triaxial Time-History Strong Motion Accelographs
- a. 0-YE-001 Unit 1 Contaiment Base M* R SA
- b. 0-YE-002 Unit i Containment 69' M* R SA
- c. 0-YE-003 Auxiliary Bldg. Base M* R SA
- d. 0-YE-004=. Intake Structure M* R SA
- e. 0-YE-005 Free Field M* R SA 2.. Triaxial Seismic Switches a.L O-YS-001 Unit'l Containment Base M .R SA
- b. 0-YS-002 Unit- 1 Containment 69' M R, SA u 3.
SeEmit AccejemAion Eecorder
- a. 0-YRC-001 Control Room M R SA
- b. 0-YR-001 Control Room M R SA
*Except seismic trigger -
** Verify instrument energized ,
4 CALVERT CLIFFS - UNIT l' 3/4 3-33
l 4
- 'l .
INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION-FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in - Table 3.3-8 shall be OPERABLE.
- APPLICABILITY: At all times.
ACTION:
- a. With one:or more required meteorological mocitoring; channels inoperable for more .than 7 days, prepare and submit a- Special Report:to,the Comission pursuant to Specification 6.9.2 within the- next 110 days outlining the cause of the malfunction f and-the plans for restoring the.chsnnel(s) to OPERABLE status,
- b. The provisions of. Specifications 3.0.3 and 3.0.4 are not=
applicable.- I. SURVEILLANCE REQUIREMENTS i
- 4. 3. 3, O Each.of the above meteorological monitoring: instrumentation-
- channel:1 shall' be demonstrated OPERABLE.by the performance.-of the CHANNEL CHECK and CHANNEL CALIBRATION operations at- the frequencies shown in Table 4.3 5.
l n b-l:..
.CALVERT CLIFFS - UNIT 1 3/4 3-34 i
. ; .. ...-1.-J.,,-'-. -
n v e'
. m.. : i..: . + 2. ::.
.g. . . .e n. z.e,
.. n.
o ..- r . .L . v.e.
. . .. . - : . . . ........o ..,.e.;.,,,,,.,..,-...,e..,,
- o. .
e
.u. . ...e..u. -
C.s.
. .. e . e........
. . . . e . .. . : . .~
.s..
,. , . . n. g . c..:...,:.....e
- 1. WIND : SPEED a . -- Ncminal Elev. 10M 1 i s' i
-b, Neminal Elev. 6CM 1 l
p
-2, .WIND DIRECTION
- a. Ncminal-Elev. 1CM -
1 l
- b. Nominal Elev. 6CM 1 .
l .
- 3. . AIR TEMPEFAT'JRE DELTA T (10M 60M) ) .l 4- .
e 1
- .h
. \i-CAL'/ERT CLIFFS -~UN!T 1 3/4 3 3E Amencment NC. 103
l
= - - .--
2:
. W C.= i ;
s;m. E <; C aC ac < g
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I CALVERT CLIFFS - UNIT 1 3/4 3 36 Amencmen: No'.103 i
INSTRUMENTATION ; REMOTE SHUTDOWN INSTRUMENTATION
, LIMITING CONDITION FOR OPERATION t
3.3.3.5 The remote shutdown monitoring instrumentation channels shown u in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY: MODES.1, 2 and 3. ACTIO,N:
- s. With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either restore the
!= inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours,
- b. The provisions of Spe:ification 3.0.4 are not applicable. -
i. SURVEILLANCE REOUIREMENTS - I . 4,3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by , performance of the CHANNEL CHECK and CHANNEL , CALIBRATION operations at the frequencies shown in Table 4.3-6. , E ,.- 1 i l'.-- ll I CALVERT CLIFF 5 --UNIT 1 3/4 3-37
~
l TABLE 3.3-9 19 g REMOTE SIRITDOWft MONITORING INSTRtlMENTATIOff S *- i n g . MINIMUM.
. READOUT. MEASUREMENT l
km ' INSTRUMENT LOCA110N RANGE CilANNEL5 OPERABLE 7 1. Wide Range Neutron Flux IC43' l O.1 cps-200% power
- l' l E 2. Reactor Trip Breaker ' Cable Spreading-
-e Indication OPEN-CLOSE- . 1/ trip breaker Room
- 3. Reactor Coolant Cold leg Temperature . 1C43 212-705*F 1-
- 4. Pressurizer Pressure IC43 0-4000 psia 1
{ 5. Pres'surizer Level' 1C43 0-360 inches I g, 6. Steam Generator Pressure IC43 0-1200 pstg 1/ steam generator
~
- 7. Steam Generator Level 1043 -
401 to 463.5 inches 1/ steam generator g *When the 1C43 instrumentation is inoperable, the wide range neutron flux monitors located in the
. auxillary feedwater pump room may be utfilzed to meet this requirement. During the period when the R instruments are utilized to meet the above requirement, they will be subject to the surveillance 2 requirements of Table 4.3-6.
fr
.O
~
. e e
-. . -- - --m._~_ _ - , , s ...%.-_m- .mm m--u--- m m-- .-.:- tem _:. m-_e-e-_ _
+ +ee ef
~- TAlltE 4. 3 g r-
;5 REMOTE SIRjTDOWN MONITORING'INSTRtjMENTATION StfRVllllANCE REf]tilREMfMTS
?! CilANNEL - CllAffMEL ClllCK CALIBRATION INSTRilMENT 4 '
M DO
- 1. Wide Range Neutron Flux 7
E - 2. . Reactor Trip Breaker Indication M- y<(. ()g :
~-e .
- 3. Reactor Coolant Cold Leg Temperature M R Pressurizer. Pressure 'M R 4.
- 5. Pressurizer level M R
- 6. Steam Generator Level (Wide Range) M R .
M R
- 7. Steam Generator Pressure Y
M t 1 9 - i. o. On a L
- c. 1
)
. I
' INSTRUMENTATION L POSI-ACCIDENT INSTRUMENTATION-a .
LIMITING CONDITION FOR' OPERATION l 1
,glI ' 3.3.3.6 The post accident monitoring instrumentation channels shown in Table 3. A;10 shall be OPERABLE. !
l APPLICABILITY: MODES 1, 2 and 3.
- ACTION:
- a. As shown in Table 3.3-10.
- b. The provisions of Specification 3.0.4 are not a:olicable.
~
, 1 SURVE:LLANCE REQUIREMENTS 4.3.3.6 }
Each post-accident' monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at'the frequencies shown in Table 4.3-10. [ e s i E l i-l: CALVERT CLIFFS UNIT 1 3/4 3-40 Amendment No.117 0 r ,- - - - . . .
'TABIE 3.3-10 ,
POST-ACCIDENT MONITORING INSTRUMENTATION' 3
,g MINIMUM N CliANNELS n INSTRUMENT OPERABLE ACTION 3 1. . Containment Pressure 2 31
- 2. Wide Range Logarithmic Neutron flux Monitor 2 31
$- '3. Reactor Coolant Outlet Temperature 2 31-
'[ 4. Pressurizer Pressure 2 31'
- 5. Pressurizer level 2 31
- 6. Steam Generator Pressure 2/ steam generator 31
- 7. Steam Generator Level (Wide Range) 2/ steam generator 31 q 8. , Auxiliary feedwater Flow Rate 2/ steam generator 31
[ 9. 'RCS Subcooled Margin Monitor
. 1 31 b 10. PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve 31 .
- 11. PORY Solenoid Power Indication 1/ valve 31
- 12. Feedwater Flow 2 31
- 13. Containment Water Level (Wide Range) 2 32, 33
- 14. Reactor Vessel Water Level 2 34,'35 E
g 15. Core Exit Thermocouple System 2 locations / core quadrant 31 ' E gg > tg A channel has eight sensors in a probe. A channel is operable if four or more sensors, one or more in the upper three and three or more in the lower five, are operable. L t* EE b
TABLE 3.3 10 (Continued) ACTION STATEMENTS ACTION 31 - With the number of OPERABLE post accident monitoring channels less-than required by Table 3.3 10, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours. ACTION 32 - With the number of DPERABLE Jost accident monit.oring channels one less than the(mGimum efia~n~riehoperablDr~~ requirement in Table 3.3-10. operation may proceed provided the inoperable channel is restored to OPERABLE status at the g next outage of sufficient duration.- ACTION 33 - With the number-of OPERABLE post-accident monitoring # channels two less than required by Table 3.3-10, either _ 6j restore one inoperable channel to OPERABLE status within +- , 30 days or be in HOT SHUTDOWN within the next 12 hours.
-A g
' ACTION 34 - With the number of OPERABLE # Post-MecjIlent. Monitoring /
Channels one less than thet1rirnimudMLn,els0PULABL6 requirement in Table 3.3-10, either restore the system to OPERABLE status within 7 days if repairs are feasible
.without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for
( restoring the system to OPERABLE status. ACTION 35 - With the number of OPERABLEghannels two less than required by Table 3.3-10, either restore the inoperable channel (s) to OPERABLE status within 4B hours if repairs are feasible without shutting down or:
- 1. Jpitiate an alternate method of monitoring for core and yeactor Goolant ) stem voiding;
- 2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event, outlining the action taken, the
-cause _of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and 3.- Restore the-system to OPERABLE status at the next-scheduled refueling.
l l Il CALVERT CLIFFS - UNIT 1 3/4 3-41a Amendment No. JJ7, 147
yh%?'w Q & l l0 ,~ at & %* IMAGE EVAL.UATION ,C [g// jp %[ $[fIf TEST TARGET (MT-3)
/ [ g% fg
%'7Y y,W' h?p f
+ <e 1.0 b ;'! OA la ad :l["2,2 I [lri
': ::. g2R I.l O!M l.25 1.4 1.6 4 150mm >
4 6" > <h,tQ oy
$6 ,,-l'k 4 4
+ -. - _ - - _ - - .
--e e-mm E O A W^!^
b> Th IMAGE EVALUATION /% N gpffk[f tb x
/[//j/ /
3 TEST TARGET (MT-3) /.j? p 3 g/, yV' %,,, ,fts* a
+ <te a
] l.0 a lt yl Ill M l.1 E "' E M l.25 1.4 1.6
. - - , . = ,
- 150mm >
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.y
/ ' p' b
,, q,,,,i,$ h> , gy _
y;
fk
, t? w Ag-E l0 4 IMAGE EVALUATION , t' b
//g// 'i $ t G# VF[#' TEST TARGET (MT-3)
/ e k f
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,.o :s a us a m
t-- Jan ll T E4 m=La t = 1.25 1.4 ' __ j .l.6_ 4 150mm > 4 6" > ef v% . Oy i
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- l
'l _ _ , _ . . _ __ __ __ ____ _____ - _ _ Q ___- _ _ _ _ _
TABIE 4.3-10 POST-ACCIDENT MONITORlNG INSIRUMENTATION St]RVEltt ANCE REJillRLMENTS
=
n CilANNEL CilANNEL C- INSTRUMENT _CilECK CALIBRATION l 1. Con'tainment' Pressure M R g 2. Wide Range Logarithmic Neutros finx Monitor M ~ Nglh4 Z 3. Reactor Coolant Outlet Temperaiva M R
~
- 4. Pressurizer Pressure' M R
- 5. Pressurizer Level M R
- 6. Steam Generator Pressure M R
- 7. Steam Generator Level (Wide Range) -
M R 1 8. Auxiliary feedwater Flow Rate M R i { 9. RCS Subcooled Margin Monitor M R
- 10. PORV/ Safety Val've Acoustic Monitor M. ,M R
- 11. PORY Solenoid Power Indication M.OA 34. D4
- 12. Feedwater Flow M R
- 13. Containment Water Level (Wide Range) M R
{ 14. Reactor Vessel Water level M fpt' 11/2 y 15. Core Exit Thermocouple System M R* E o . Q The performance of a channel calibration operation exempts the Core Exit Thermocouple
$$ but includes all electronic remponents. The Core Exit Thermocouple shall be calibrateil
$$ prior to installation in the reactor core.
*R U
~
.- i 1 i
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.. . . . .. . . . . . . ...e..............y I
.n. s. . ... . . . . . . . . . . . .
LIM 1-'NC. 00nC1 i*N 20; C:!:JTION 3.3.3.7 As a mini um, tne fire detecti:n ins:rumen;aden f:r en:n fire de:e::icn
- ene sh:wn in Ta:ie 3.311 shall te CEERAELE.
1 AP9t! A!!LITY: Whenever equi Pment in that fire detection :ene is re:vired t; te I C r. . r.s
. . .cL :. .
I ACTION: l 1 With one er more of :ne # ire dete: i0n irs rument(s) shewn in Ta:le 3.3-11 in ;e atie: i
- t. *lithin i P.our es u:lis
. a 'i t wat:n : ate: 1 :: ins:e: :ne ::ne(s) I i at:n the d n::e a:i i instru en ;s) a: 'eas :n=e :er r:va, u n'. e s s
- ne ins:rumer:(s) is ion :e: insi:e : e ::nuinment. ren ins:e::
l l l' :ne ::ntainment a inas; :n:e :e S 5:'.rs :r :n3 :r :ne ::nutn. L I
. e r. : air e.:e a:.re a: 1eas: :t:s :er neur a; :ne 10:::izns itste:
i in I:e:ift:sti:r 2. 5.1. i; Or .r'. as s :ne ir s .ru en l s ) is '.::a:e: l in 'tre de:e::: n ::nes e:vi::e: ut:n av : .ad: at: pi:e 5:riesler , syste s alarme: an: s t.:e rv i s e: :: :ne C:ntrol :.ccm, :nen itnin '._ ' h:gr aad a: leas'. '. 2: hcurs :nereafter, ins *e:: :ne : re(s) with
- , in
- : era:'.e instrumer:s and verify :na: :ne aut ma ic s;rint'.er system, in:iucing the water flow alarm an: supervisory system, is ,
opera:ie y :nannel fun::ional test.
- b. Restore :ne ino:eraole instrumen:(s) t: OPEF. tile s:atus oi:nin
~
iA ays or precare anc su mit a Special Re:or: :o ne Ccemissien l ' pursuant to 5:ecification 5.0.2 wi:hin :ne nex: 30 cays outlining -
- ne ac:i:n :sken. :he cause of the ine:erability and the clans an: sene:ule f r res : ring :ne instrumen:(s) to OPERAILE sta us.
- c. The previsions Of S:ecifications 3.0.3 and 3.0.4 are no: a;;1i:stie.
5'URVEILLANCE REOU!:E"ENTS 4.3.3.7.1 At leas: on:e per 5 months, at least 25 of the above recuired fire ! detection ins;ruments wnicn are ac:essible during plan operation shall be i- cemenstrated OPERABLE ty performance of a CHANNEL FUNCTIONAL TEST. Detectors i: selected for testing snail be selected on a rotating bas".s such that all detectors will be tested over a two year perioc. It_in any detection zone there are less than four detectors, a; least one differen; deteet:r in that
- ene shall be tested every six months. For eacn detector found ino:eratie during functional testing, at leas; an accitional 10% ef all ce:ect:rs er 10 cetect:rs, wnicnever is less, shall also be testec. Fire detecters wnien are
! inac:essible curing plant c eration snali ce cem:nstrated CPERACLE by :ne l;erformance of a MANNEL FUNCT 0NAL TEST during caen COL 3 SHUT 3OWN excee:ing 24 5:urs unless ;erf:rmed furing the :revicus six moe:ns.
t
. on..:-
.,,,-.. .... e..re.
- n. n ., . ,.
.f. 3.-
ent en: No 2 6, :. 7 . t l 1 l l
l IN5?tU"ENTAT*CN i SURv!!LL ANCE :EOL'1RE' TENTS Iced rue i 3.3.7.2 with the detector alarms of each of the a: evefire re;uireThe NF A Coce dete:; ion instru-ments shall be dem:nstrate: 0 EF.ABLE at least once per 6 months. 4.3.3.7.3 The non-supervised circuits, associated with detect:r alarms, te:,,een the instrument per 31 days. anc the centrol re m shall se dem:nstrate: OFERA!LE at leas: once l I il li I 9 t t CALVERT CL*FFS - UNIT 1 3/4 3-44 A.mendment No . 25, 94
_ _ _ _ . _ _.~. _-_ __ - _ . _ _ . l TABI,E 3. 3-17 FIRE DE ECTION INSTRUMENTS UNIT 1 MINIMUM INSTRUMENTS OPERABLE' ! ROOM / AREA-AUX BLDG INSTRUSENT-LOCATION HEAT FLAME SMOKE 100/103/ 104/Wr mi Corridors-Elev(-)10"-0" 5 110 Coolant Waste Rec & Mon. Tk Pp Rm. 2 j 111 Waste Processing Control Rm M. '
'112/114 Coolant Waste Rec Tank 4 *
.[,
'113 Misc. Waste Receiver Tank Room 1
'115 Charging Pump Room 3 ,
118/122 ECCS Pump Roo.n i l 115/123 ECCS Pump Room 7 Corridors, & l 200/202 { 209/210 Corridors & ' 212/219 Corridors - 13 207/208 Waste Gas Ecuip Rm -
-3 216 Reactor Coolant Make-up Pumps 1 217 Boric Acid Tank & Pump Room- 2 218 Volume control Tank Room 1
. (- 220 Degasifier Pump Room 1 221/326 West Piping Penetration Room 2 3 222 Hot Instrument Shop 2 223- Hot Machine Shop 4 224 12 MSIV Hyd Area 10
-225 Rad Exhaust Vent Equip Rm . 4 226 Service Water Pump Rm 3 6 227/316 East, Piping Penetration Rm 3 5 228 Component Cooling Pump Rm 8 301/304/300 Battery Room & Corridor -3 306/1C
- Cable Spreading Rm & Cable Chase" 2 '10 308 N/S Corridor _ 6
.315 Main Steam Piping Area -
6 317- Switchgear Room, Elev 27'-0"" 6 318 Purge Air Supply Room 2 31g/325 West Passage and Vestibule 6
'320 - Spent Fuel Heat Exchange Room 3 323 Passage 27' Val.ve Alley &; Filter Rm '
3 324' Letdown Heat Exchanger Am 1 Elev. 27'-0" Switchgear Vent Duct -1A Cable Chase 1A ~1
.18 Cable Chase 18 1-405- ' Control. Room 6 410 .
N/S Corridor 4. 417/418 Solid Waste Processing 2 -3 F . CALVERT C1.IFFS - UNIT 1 3/4 3-45 Amendment No. 75,f7,75,109
l TABLE 3.3-11 (Continued) I FIRE DETECTION INSTRUMENTS UNIT 1 MINIMUW INSTRUMENTS OPEU ELE
- ROOM / AREA AUX BLOG INS *RUMENT t.0C' TION HEAT FLAME SMOKE 413/419/420 -Cask and Equip Loading Aren &
424/425/426 Cask and Equip Loacin Area 3 22 - l 421 Diesel Generator No. 12)** 2 ' '
-422 Diesel Generator No. 11)** 2 ,
423 West Electrical Pen Rm '3 428 East Piping Area 7 429 East Electrical Pene Rm 3 l 430 Switchgear Room Elev 45'-0"** a J l 439 Refueling Water Tank Pump Rm 2 l 441 Spen Resin Metering Tank Rm 1 Elev 45' 0" Switchgear Vent Duct 1 . Elev 69'-0" Control Room Vent Duct "A" 1 Elev 69'-0" -Cable Spreading Room Vent Duct 1 512 Control Room HVAC Equipment a 586-590, Radiation Chemistry Area, g 592,593 Radiation Chemistry Area,
- 595-597, Radiation Chemistry Area & !
.521,523 Corridors 20 520 Spent Fuel Pool Area Vent Equip Rm 2-524 Main' Plant Exhaust Ecuip Rm- 8 525 Cntmt Access Area 3 529 Electrical Ecuip. Roem 3 530/531/533 Soent Fuel. Pool Area 5 17 1 536/537 Misc Waste Evaccrator & Equip Rm 3 Elev 83'-0" Caele Tunnel 4 603 -Auxiliary Feedwater Pump Rm 2
Containment B1 des U-1 RCP: Bay East
- 16 U-1 RCP Bay West'. . .
1 U-1 East Electric-Pen Area
- U-1 West Electric Pen Area *
-Intake St'ructure Elev 3'-0" Unit 1 Side 7 24-
'* Detection instruments located within the containment are not required to be OPERABLE during the perfomance of Type A Containment Leakage' Rate Tests.
** Detectors.whien autornatically actuate fire suppression systems.
= Monitored.by four protecto wires.
-CALVERT CLIFFS - UNIT 1 3/4 3-46 Amendment No. 25,57,96, 109
m *
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i t . . L I' ' ' THIS PAGE-INTENTIONALLY LEFT BLANK {.. L . II ' p
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CALVERT CLIFFS - UNIT 1 3/4 3-47 Amenenent'No. 99.109 L - o l-
. , , . ~ . . - - - . . - -
L INSTRUMENTATION g ll RADIC ACTIV2 C ASEOUS EFFLUENT MON! TOR!NC INSTRUM2NTAT10N l LIMITING CONDITION FOR OPER AT'ON l
) I 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in I i
I Table 3.bl2 shall be OPERABLE with their alarm / trip setpoints set to ensure that the I , limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these l l channels shall be determined and adjusted in accordance with the methodology and I partmeters in the ODCM. l L APPLIC A BILITY: As shown in Tar..e 3.3-1: A C~10N I l a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, without celay suspend the release of radioactive gaseous l l effluenu monitore: by the affected channel, or oeclare the channe! : l Inoperable, or enan5e the setpoint so it is acceptacly conservative. i 1
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take tne ACTION snown m L
t
,, Table 3.312. Exert best efforts to return the instruments to OPERASLE
{~ status within 30 oays and, if unsuccessful, explain in tne next Semiannual l Radioactive Effluent Releue Report wny the inoperability was not. l' correctec in a timely manner,
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not appilcaole. J i
1 SURVE!LLANCE REOUIREMENTS \' I i'
** 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel sha!! be 1 demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL . CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.312 -
t l i
.l l-l L l l- :
K CALVERT CLIFFS UNIT 1 3/4348 Amencment No.105 l . t l c
- ~ .
z
- TABLE 3.3-12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRtIMENTATION P
d MINIMUM CilANNELS OPERAf)LE APPLICARfLITY ACTION I ., INSTRUMENT l v 9 1. WASTE GAS HOLDUP SYSTEM O Noble Gas Activity monitor - p a. Providing Alarm and Automatic
- 3) l Termination of Release (I)
- b. Etfluent System Flow Rate * %
Measuring Device (1)
- 2. MAIN VENT SYSTEM
- 37
- a. Noble Gas Activity Monitor (I) y-
- b. . 5. (l>
(1) 38
- c. Particulate Sampler i
l a
- y. .
t E t . c he O US .
~ ~ * ^ ^ ^ ' ~ ^
TAB'LE '33 '12 (Csntinue di TABLE NOTATION , I
- At all times.
ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environments ,
- a. Using the main vent monitor as a backup and r'ecording RMS readings every 15 minutes during the release, or
- b. Provided that prior to initiating the release, at le ast two independent samples of the tLWs contents are analyzed, and at' least two technically quallfled members of the Facility Staff independently verify the release rate calculations and two quallfled operators verify the discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway., ACTION 36 - With the number of channels OPERABLE less than required by the
- Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once
- per 4 hours. ACTION 37 -- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided either (1) grab samples are taken and l analyzed for gross activity at least once per 24 hours, or (2) an equivalent monitor is provided. ACTION 38 - With the number of channels OPERABLE less than required,by the
. Minimum Channels OPERABLE requirement, effluent releases via the -
af fected pathway may continue provided samples are continuously collecte d as required in Table 4.11-2 with auxiliary sampling e quipment. 9 CALVERT CLIFFS UNIT 1 3/43-50 Amendment No.105-I 5
e 4
\
g TABLE 4.1-11 O I; R ADIOACTIVE GASEOtl5 EFFLtIENT MONITORING INSTRIIMENTATION SIIRVEII. LANCE REQt flREMENTS [1 CilANNEL MODES IN WillCII
;lf CilANNEL SO(IRCE CIIANNEL FilNCTIONAL SilRVEll. LANCE INSTR 11 MENT Cll2CK CllECK cal.IllH A TION TEST RFQti!Hlin el 1. WASTE GAS IIOLDtlP SYSTEM ie
- a. Noble Gas Activity Monitor -
Providing Alarm and Automatic. Termination of Release P P R(3) SA(1) *
- b. Effluent System Flow Rate Measuring Device D(fe) 134. fy% R %.A. ng *
- 2. MAIN VENT SYSTEM
.g.
L'l a. Noble Gas Activity Monitor D M R(3) SA(2) =
- b. Bodine Sampler W Ng()q [.(1/9 ,NI. O A' *
- c. Particulate Sampler W fkA. ll/} N'A.[)f} Ng. Q{ *
- M m
I b . . n h$
??
m. a
-- . . - - _ . .. . - - . . - . . - - . _ . . - - . _ - . ~ . - . . - .
TABLE 4.bli (Continued) TABLE NOTATION l At all times other than when the Une is valved out and locked. (1)' The CHANNEL FUNCTIONAL TEST shau also demonstrate the automati:1 solation of this pathway and/or control room alarm annunciation occurs !! me appropriate following condition (s) exists:
- 1. Instrument indicates measure levels above the alarm / trip setpoint.
- 2. Circuit fauure. - ,
- 3. Instrument indicates a downscale failure.
(2) The CHANNEL FUNCTIONAL TY.ST shall also demonstrate that control room alarm annunciation occurs if any of sne fouowing conditions exisus-
- 1. ~ Instrument indicates measured levels above the alarm setpoint.
x . 2. Circuit failure. ,
.- 3. Instrument indicates a downscale failure.
(3) m The initial CHANNEL CALIBRATION shau be performed using one or more of the
' reference standarcs traceable to the National Bureau ; of Standares or using
- standaras that have been obtained from suppuers that participats in measurement, I ,'
assurance activities with NES. These stanaards shall permit caubrating the system within lts intended range of energy and measurement range. For subsequent l CHANNEL CALIBRATION, sources mat have_ been related to the initial calibration can be used. (4) The CHANNEL CHECK anau consist of verifying indication of flow dJring periods of release and shall be made at least once per 24 hours on days on which effluent releases are mace. 4' 1 9 4 CALVERT CLIFFS UNIT 1 '3/4 b 32 Amendment No. 105 L I p i i I l
INSTRUMENT ATION R AD10 ACTIVE LICUlO EFTLUENT MONITORINC INSTRUMENTATION
- LIMITINC CONDIT10N FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the Umiu of Specification 3.11.1.1 are not exceeded. The ajarm/ trip setpoints of these $
channels shall be determined tnd adjusted in accordance with the methodology anc l parameters in the CFFSITE DOSE CALCULATION MANUAL (ODCM). ' APPLICABILITY: ' At all times. ACTION: ( a. With a radioactive Uquid effluent monitoring instrumentation cht,nnel l alarm / trip setpoint less conservative than required by the aoove specificauon, without delay suspend the release of radioactive liquid ! effluents monitored by the affected channel, or declare the channel Inoperable, or change the setpoint so it is acceptably conservative.
- i-.
- b. With less than the minimum number of radioactive !! quid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3
- 13. - Exert best efforts to return the instruments to OPERABLE status witnin 30 days and, if unsuccessful, explain in me next Semlannual Radioactive i l
Effluent Release Report why the inoperabluty was not corrected in a timely l manner. 1
- c. The provisions of Specificat!ons 3.0.3 and 3.0.4 are not applicable.
SURVETLLANCE REOUTREMENTS
+* -
4.3.3.10 Each radioactive Uquid effluent monitoring instrumentation channel shan be . demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TF.ST operations at the ; frequencies shown in Table 4.3-12.
- l l
l , l- . n I
.i CALVERT CLIFFS UNIT I 3/4 3-53 Amendment No. 105 i
i I
4-TABLE 3.3-13 i RADIOACTIVE LIQUID EFFLifENT MONITORING INSTRIJMENTATION I ih
)
4/ - v MINIMUM p INSTR (JMENT CllANNELS l
!l Ol' Eft Ant E_
ACTION
- ,, 1.
GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
- a. Ligeld Radwaste Effluent Line (1) 28 b.
Steam Generator Dlowdown Elfluent Line (t) 29 2. FLOW R ATE MEASUREMENT DEVICES T*- '
- a. Ligsid Radwaste Elfluent Line (1) 30 b.
Steam Generator Blowdown Effluent Line (1) 30 c
.if m
~
= men-
1 TABl.E 3J-13 (Continued) TABLE NOTATION
; (-
ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a releases l-l a. At least two independent samples are analyzed in accordance with j Specification 4.11.1.1.1, and
- b. At least two technically quallfled members of the Facility Staff independently verify the release rate calculations and two quallfled operators verify the discharge valve line up.
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 analyzed for gross radioactivity (beta or gamma) at the lower limit of detection defined in Table 4.11-1:
- a. At . least once per 12 hours when the speelfic activity of the j
'
- secondary coolant is greater than 0.01 microcurle/ gram DOSE EQUIVALENT l-131.
- b. At least once per 48 hours when the specific. activity of the secondary coolant is less than or equal to 0.01 microcurle/ gram DOSE EQUIVALENT l-131, u ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releans via this pathway may cor.tinue provided the flow rate is estimated at least once per 4 hours during-actual releases. Pump performance curves may be j used to estimate flow.
l 4 L 4 - I l (. CALVERT CLIFFS UNIT 1 3/4 3-55 Amendment No.105 l l
4-TABLE 4.3-12
\
q 8, RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATK)N SURVEILLANCE REQUIREM y , f- CHANNEL SOURCE CHANNEL INSTRUMENT FUNCTIONAL e CitECK CifECK CALIBRATION TEST
- t. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE '
- a. . Liquid Radwaste Effluent Line D t
P R(2) SA(1) ;
- b. Steam Generator Blowdown Effluent Line !
D P R(2)
~
SA(I) 2.
} FLOW RATE MEASUREMENT DEVICES y a. Liquid Radwaste Effluent Line W D(3) =M.A. R ara h N/4
- b. Steam Generator Blowdown Effluent Line "^
D(3) f//} R
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i 1 f . . . D a o,
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-__am7,,n__ - , _ = _ _ ._- .___,w
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~. _ _ ~ _ _ . . _ . _ . _ _ _ . . . _
i TABLE 4.312 (Continued) j TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and/or control room alarm annunciation occur if the appropriate . following condition (s) existu
- 1. Instrument Indicates measured levels above the alarm / trip setpolpt.
- 2. Circuit failure. -
- 3. Instrument indicates a downscale failure.
(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the refererae standards traceable to the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activltles with NB5. These standards shall permit calibrating the system within its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration can be used.
, (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which effluent releases are made. .
s 4 CALVERT CLIFFS UNIT 1 3/4 3-57 Amendment No.105 -
3/4.4 REACTOR COOL AN? SVSTEM p,4;iC00LANT LOOPS AND COOL ANT CIRCUL ATION STARTUP AND POWER OPERATION j LIMITING CONDITION FOR OPERATION 43.4 3
-fs/4.1.1 Both reactor coolant loops and both reactor coolant pumps in i each loop shall be in operation. I 1
APPL IC ABillTJ: MODES 1 and 2*. ACTION: I With less than the above required reactor coolant pumps in operation, be in at least HOT STANDBY within I hour. SURVEILLANCE REOUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. ( l-L See Special Test Exception 3.10.3. l-CA'LVERT CLIFFS - UNIT 1 3/4 4 1 Amendment No. 55 146
y
- REACTOR COOLANT SYSTEM MLANT 109$4ND t00TTEC1RNt1TTt%
HOT STANDBY ! i LIMITING CONDITION FOR OPERATION 3.4.1.2 a. The reactor coolant loops listed below shall be OPERABLE:
- 1. Reactor Coolant Loop til and at least one associated reactor coolant pump.
- 2. Reactor Coolant Loop #12 and at least once associated reactor coolant pump.
- b. At least one of the above Reactor Coolant Loops shall be in operation *. .
APPLICA91LITY: MODE 3** EllM:
- 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 SHUTDOWN within the next 12 hours.
.t b. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration-of the Reactor Coolant System and initiate corrective action to return the required-loop to operation within one hour.
SURVEILLANCE RE0VIREMENTS-4.4.1.2.1 At least the above . required reactor coolant pumps, if not in operation,.shall be determined to be OPERABLE once per 7 days by. verifying correct breaker alignments and indicated power availability, i 4.4.1.2.2 At least one cooling loop shall be verified to be in operation-and circulating reactor coolant at least once per 12 hours..
= All reactor coolant pumps may be de energized for up to I hour (up-to' 2 hours for low flow test) provided (1) Ao operations ara permitted that would'cause dilution of the Maactor 16eolanthystem boron concentration, 0
and (2) core outlet temperature is. maintained
, at least 10 F below saturation temperature.- <
A reactor coolant pump shall not be started with the RCS temperature less than or equal to 3270F unless (1)' the pressurizer water level is less than or equal to 170-inches and (2) the secondary water temperature of each ~ steam generator is less than or equal to 30 0F above the-RCS-temperature, and (3) the pressurizer-pressure is less than or equal to 290 psia. CALVERT CLIFFS UNIT 1 3/4 4-2 Amendment No. f5/JJ 146
REACTOR COOLQNT SYSTEM
-M' ff. ^ ^^^: r:rf0MWUtaW4 SHUTD0VN LIMITING-CONDITION FOR OPERATION 3.4.1.3
- a. At least two of the coolant loops listed below shall be OPERABLE:
- 1. Reactor Coolant Loop #11 and its associated steam generator and at least one associated reactor coolant pump,
- 2. Reactor Coolant Loop #12 and its associated steam generator and at least one associated reactor coolant pump,
- 3. Shutdown Cooling Loop #11*,
4 Shutdown Cooling Loop #12*.
- b. At least one of the above coolent loops shall be in operation".
APPLICABILITY: MODES 4***# and 5***#, L L e,C1103: L a. With less than the above required coolant loops OPERABLE, initiate corrective action to return the required coolant loops to OPERABLE status within one hour or be in COLD SHUTDOWN within 24 hours. i L b. With no coolant loop in operation, suspend all operations i: involving a reduction in boron concentration of the Reactor l 1 - Coolant System and initiate corrective action to return-the l required coolant loop to operation within one hour. The normal or emergency power source may be inoperable in MODE 5. All reactor coolant pumps and shutdown cooling pumps may. be de energized for up to I hour provided (1) ao operations ara permitted that would cause dilution of the Weactor tioolant:3ystem boron concentration, and (2) core outlet temperature is maintained ?
- - at least 10 0F below saturation temperature. *
*** A reactor coolant pump shall not be started with the RCS temperature less than or equal to 3270 F unless (1) the pressurizer water level ,
is less than or equal to 170 inches, and (2) the secondary water temperature of each steam generator is less than or equal.to 30 F- 0
-above the RCS temperature, and (3)-the pressurizer pressure is less than or equal to 290 psia.
#- See Special Test Exception 3.'10.5, I CALVERT CLIFFS UNIT 1 3/4 4 2a Amendment No. HD/J,146 i
w
- u-
- .. _ -_ . .- . -~ - _ - . . - . _ - - - _
. REACTOR COOLANT SYSTEM C6DLt&LifmS M40Al ANT GRCULATIDN SHUTOOWN
$URVEILLANCE REOUIREMENTS l 4.4.1.3.1 The required shutdown cooling loop (s), if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability for pumps and shutdown cooling loop valves.
4.4.1.3.2 The required steam generator (s), if it is being used to meet 3.4.1.3.a, shall be determined OPERABLE by verifying the secondary side water level to be above -50 inches at least once per 12 hours. 4.4.1.3.3 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. 1 i l l l I l j l l ) l i; CALVERT CLIFFS - UNIT 1 3/4 4-2b Amendment No. EE,145
REACTOR COOLANT SYSTEM SAFETY VALVES LIMIT!NG CONDITION FOR OPERATION 3.4.2.1 The following pressurizer code safety valves'shall be'0'PERABLE: Valve lift Settinos ( 1%) RC-200 2500 psia RC-201 2565 psia APPLICABILITY: MODES 1, 2 and 3. ACTION: hith one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours.. 3.4.2.2 At l' east one of tne above pressurizer code safety valves shall be OPERABLE:' APPLICABILITY: MODES 4 and 5. ACTION: , With no pressuri7er code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation. SURVEILLANCE REOUIREMENTS ' , 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.
- Both valves may be removed in MODE 5 provideo at least one valve is
! replaced b a spool piece which allows the pressurizer to relieve directly. I to the querc h tank. . CALVERT CLIFFS - UNIT 1 3/4 4-3 Amendment No.34, 53 I 9
i REACTOR COOLANT SYSTEM g34,4., $ RELIEF VALVES l i LIMITING CONDITION FOR OPERATION
\
~ '
3.4.3 Two power operated relief valves (POR.'s) and the'ir assoc'iaked block valves.shall be OPERABLE, APPLICABILITY: MODES 1, 2. and 3. ACTION:
- a. With one or more PORV(s) inoperable, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise. be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours,
- b. With one or more block valve (s) inoperable, within 1 hour either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); ctherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUT-DOWN within the following 30 hours.
- c. .With one er more block valve (s) closed and power removed from the block valve (s) to satisfy a. or b above, the provisions of Specifica-tion 3.0.4 are not applicable. .
SURVEILLANCE REOUIREMENTS 4.4.3.1 Each PORY shall be denonstrated OPERABLE:
- a. At l' east once per 31 days by performance of a CHANNEL FUNCTIONAL #
TEST, in accordance with Table 4.3-1, Item 4 -
- b. At least once per 18 months by performance of a CHANNEL CALIBRATION.
4.4.3.2 Each block valve shall be demonstrated OPERABLE at least one'e per 92 days by operating the valve through one complete cycle of full travel. This demonstration is not required if a PORY block valve is closed and power removed to meet Specification 3.4.3 a. or b. CALVERT CLIFFS - UNIT 1 3/4 4-4 Amendment No. 53,75 b
RE ACTOR C00'. A"i SYSTE'i PRES $URITER-53/4AA LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble and with at least 150 kw of pressurizer heater capacity capable of being supplied by emergency. power. The pressurizer level shall be maintained within an operating band between 133 and 225 inches except when three charging pumps are operating and letdown flow is less than 25 GPli. If three charging pumps.are operating and letdown flow is less than 25 GPM pressurizer l 1evel shall be limited to between 133 and 210 inches. APDLICASILITY: MODES 1 and 2. ACTION:
- a. With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDO N within the following 12 hours,
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY '
with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. I j j,URVEILLANCEREQUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within the above band at least once per 12. hours. CALVERT CLIFFS - UNIT 1 3/4 4-5 Amendment No. 33. S0,117 0 4
REACTOR COOLANT SYSTEM N :3 TEAM GENERATORS o LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200'F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the require - ments of Specification 4.0.5. 4.4.5.1 Steam Generator Samnle Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in } Table 4.4-1. l 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The' steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at L the frequencies.specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Speci-fication 4.4.5.4 The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes. selected for these inspections shall be selected on a random g basis except: L a, 'Where experience in similar plants with similar water chemistry ' indicates critical areas to be inspected, then at least 50% of-the tubes inspected shall be from these critical-areas.
- b. ' The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
L l 1. All nonplugged tubes that previously had detectable wall penetrations (>20%), and. CALVERT CLIFFS - UNIT 1 3/4 4-6 ) l W q = - - --r--m me '6 M ,..e-e,- e ,-g,- --- u- -- -w
- q.
REACTOR COOLANT SYST N
~
SURVEILLANCE RE001REMENTS (Continued) .
- 2. Tubes'in those areas where experience has indicated o
potential problems.
- c. The second and third inservice inspections may be less than-a full tube inspection by concentrating.(selecting at least 50%
of the tubes to be inspected) the inspection on those areas of the tube sheet array 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: Cateoory Insoection Results
" C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the in-spected tubes are defective.
i C-2 One or more tubes, but not more than it of j
-. -e the total tubes inspected are defective, or '
' between 51 and 10% of the total tuta inspected are degraded tubes. C-3 Mo're than 10% of the total tubes inspected ' are degraded tubes or more than 1% of the inspected tubes are defective.- Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations. 4.4.5.3- Inspection Frecuencies - The above required inservice inspections of storm generator tuoes snall be performed at the following frequencies: ' a. The first inservice inspection shall be perfonned after 6 Effective Full _ Power Months but within' 24 calendar months of. initial criticality. Subsequent inservice inspections shall- be perfonned at intervals of not.less.than 12 nor more than 24 calendar months after the previous inspection. If two consecu-tive inspections.following service under AVT conditions, not including the preservice inspection,Lresult,in all inspection - results falling'into the C-1: categny or if two consecutive - inspections- demonstrate that previously observed degradation has= not= continued and no additional degradation has occurred, the-inspection interval may be extended to a maximum of once per 40 months.-- e CALVERT CLIFFS - UNIT 1 3/4 4-7 l L
? . . . . . . .
i f;--- - .. .- -
REACTOR COOLANT SYSTEMS
~
~
SURVEll. LANCE REOUIREMENTS (Continued)
- b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4 2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1. Primary-to-secondary tube leaks (not including leaks I 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 .
g
- 4. A main steam line or feedwater line break.
$.4.5.4 Acceptance criteria l
- a. As used in this Specification:
- 1. Imperfection means an exception to the dimensions, finish .
or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
- 2. Deoradation means a service-induced cracking, wastage, wear or general cornston occurring on either inside or outside of a tube.
3.- Decraded Tube means a tube containing imperfections >20% of tne nominal wall thickness caused by degradation.~' 4 5 Deoradation means the percentage of the tube wall tnickness affected or removed by degradation. CALVERT CLIFFS - UNIT 1 3/4 4-8 Amendment No. 103 i l I l
l 1 REACTOR C00LAtT SYSTEM SURVEILLANCE REQUIREMENTS (Centinued)
- 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 inspection probe shall be deemed a defective tube.
- 6. Plueoine Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equel to 40% of the nominal tube wall thickness.
- 7. Unserviceable describes the condition of a tube if it leaks or contains a cefect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, 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 tne 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 detemined OPERABLE after completing
.the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by. .
Table 4.4-2. 4.4.5.5 Reports ( .
- 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 pursuant to Specifica:; ion 6.9.2.
b.. The complete results of-the steam generator tube inservice inspec-tion shall be included.in the Annual Operating Report for the period in which this inspection was completed (pursuant to Specification 6.9.1.5 b). This report shall include:
- 1. Number and extent of tubes inspected.
- 2. -Location and percent of wall-thickness penetration for i; each indication of an imperfection.
- 3. Identification of-tubes plugged, t-
. -CALVERT CLIFFS - UNIT 1 3/4 4-9 Amendment No.119 i
.__.l-_ _ _ _ _ _ _ _ _
REACTOR COOLANT SYSTEM g SU}VEILLANCEREQUIREMENTS (Continued)-
- c. Results of steam generator tube inspections which fall into :
Category C-3 require verbal notification of the NRC Regional ! Administrator by telephone within 24 hours prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence and shall be submitted within the next 30 days pursuant to Specification 6.9.2. l p l j 4 L .
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CALVERT CLIFFS UNIT 1 3/4 4-10 Amendment No. 9/,119 i 1
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- 1
l . 2 T AfstE 4.4 -7 n. 3 r-
' -c j' '" STE AM GT NEI1ATOft itfDE INSf'ECTION
--4 7ND SAMrti sNSPt_CTION 3HI) SAMPEC IN";PECif 0N O IST SAMet.E INSPECitO*4 Sample Sete fleserit Actea lie pewe f fle**ft Ats*** l'"Tu"" fies"'t Act" I'"t"""8 None N/A F4/A N/A N/A
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RE ACTOR COOL A'C SYSTEM 3/4.A.5 REA TOR COOLANT $YSTEM LEAKAGE LEAKAGE O!TECT!0N $YST!Y$ LIMITING CONDITION FOR ODEUTION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPEPAELE:
- a. A containment atmosphere particulate radioactivity monitoring system,
- b. The containment sump level alam system, and
- c. A containment atmosphere gaseous radioactivity monitoring system.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With only two of the above recuired leakage detection systems 8 OPEPABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours when either the required gaseous or l particulate radioactivity monitoring system is inoperable; otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With only one of the above recuired leakage detection systems l OPERABLE, operation may continue for up to 7 cays proviced snat:
- 1. Grab samples of the containment atmosphere are obtained and ' '
analy: Ad at least once per 12 hour: and
- 2. The Reacto: Coolant System water inventory balance of Surveillance ReWrement 4.4.6.2.c is perfomed at least once per 24 hours.
Otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVE!LLANCE RE0V!REMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
- a. Containment atmosphere gaseous and particulate monitoring systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL l FUNCTIONAL TEST at tne frequencies specified in Table 4.3-3, and 1
- b. Containment sume level alam system-performance of CHANNEL CALIBPATION at least once per 18 months.
CALVERT CLIFF 5 - UN!T 1 3/4 4-13 Amendment Nc.107 l l l l
PEi!TOD C00L AN* Sv$7!W. RE ACTOR COOL ANT SYSTEu LEAGaE LIY:T:N3 CON!!T:CN FOR Ot! UT!ON 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAGGE,
- b. 1 G:M UN10ENT!FIED LEAGGE, *
- c. 1 G:M total primary-to secondary leakage through steam generators, an
- d. 10 GPM ICENTIFIED LEAMGE from the Reactor Coolant System.
APDL!CAE!LITY: M00ES 1, 2, 3 and 4 ACT!ON:
- a. With any PRESSURE BOUNDARY LEAGGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTCCWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRES $URE BOUNDARY LEAGGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDEY witnin the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
_SUPVEILLANCE REOU!:EuENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of tne acove limits by:
- a. Either:
- 1. Monitoring the containment atmosphere particulate or gaseous radioactivity at least once per 12 hours, or
- 2. With the gaseous and particulate monitors inoperable, conducting the containment atmosphere grab sample analysis in accordance with the ACTION requirements of T.S. 3.4.6.1.
- b. Monitoring the containment sump discharge frequency at least once per 12 hours, when the containment sump level alarm system is OPEPABLE, CALVERT CLIFFS - UNIT 1 3/4 4 14 Amendment No. 107
~
e.aw.usum ~, - f![$;c'g'g/LAf'$YSTEM ~
.I.
iltb2VE!LLANCE REOUIREMENTS (Centinued)
- c. Perfomance of a Rea:ter Coolant System water inventory balan:e at least once per 72 hours during steady state operatien and at least once per 24 hours when re:;uired by ACTION 3.4.6.1.b, except when operating in the shutdown cooling mode, and
- d. Monitoring the rea:ter vessel head closure seal leakage dete: tion system at least once per 24 hours.
CALVERT CLIFFS - UNIT 1 3/4 4-15 Amendment No.107 5
- i
REACTOR COOLANT SYSTEM
$4.9 CHEMISTRY LIMIT!NG CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1.
APPLICABILITY: At all times. '
#CTION:
MODES 1, 2, 3 and 4
- 4. With any one or more chemistry parameter in excess of its ;
Steady !! ate Limit but within its Transient Limit, res!9re the i parameter to within its Steady State Limit within 24 hours or I be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within 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 SHUTDOWN within the following 30 hours.
MODES 5 and 6 With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Cteady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer s pressure to i 00 5 psia, if applicable, and perform an engineering evaluation to detemine the effsets of the out-of-limit condition on the . structural integrity of the Reactor Coolant System; detemine s that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure,above 500 , psia er prior to proceeding to MODE 4. SURVE!LLANCE RE0UIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be detemined to be within the limits by analysis of those parameters at the frequencies specified in. Table 4.4-3. i CALVERT CLIFFS - UNIT 1 3/4 4 16 l L
i TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT LIMIT LIMIT PARAMETER._ _ 1 0.10 ppm i 1.00 ppm DISSOLVED OXYGEN
- 1 0.15 ppm 1 1 50 ppm CHLORIDE 1 0.15 ppm 1 1.50 ppm FLUORIDE
- Limit not applicable with T,yg i 250'F.
CALVERT CLIFFS - UNIT 1 3/4 4-17 ;
TABLE 4,4 3 j i REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS PARAMETER ANALYSIS FRE0VENCY DISSOLVED OXYGEN
- At least once per 72 hours CHLORIDE At least once per 72 hours FLUORIDE At least once per 72 hours Not required with T,yg 3,250'F i
}
CALVERT CLIFFS - UNIT 1 3/4 4 18 l l
t , l J 1 REACTOR COOLANT SYSTEM M.i SPECIFIC ACTIVITY L:MITIN3 CONDITION FOR OP!UTION 3.4.8 The specific activity of the primary coolant shall be limited to:
- a. 1 1.0 uti/ gram OOSE EQUIVALENT l-131, and
- b. 1 100/Tuti/ gram.
AFDLICABILITY: MOOES 1, 2, 3, 4, and S. y ACTION: I
'00!S 1, 2 and 3*:
- a. With the specific activity of the primary coolant > 1.0 uti/ gram
- DOSE EQU: VALENT l 131 but within the alloweble limit (belcw and
- to the left of the line) shown_ on Figure 3.4-1, cceratien .ay continue for_ uo to 100 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 applicable. . ;
- b. With the specific activity of the primary coolant > 1.0 uti/ gram DOSE EC'.t! VALENT I 131 for more than 100 hours during ene centin-vous tite interval or exceeding the limit line shown on Figure
. , , 3.4-1, te in at least HOT STANDBY with Tavg < 500'F within 6 hours. ,
- c. . With the specific activity of the primary coolant > 100/I '
uti/ gram. be in at least HOT ST!.NDBY with Tavg c 500'T within t 6 hours, MODES 1, 2, 3, 4 and S: l
- d. With the specific activity of the prit:ary coolant > 1-.0 uti/ gram DOSE EQUIVALENT 1-131 or > 100/T vCi/ gram, perform the samoling
, and-analysis requirements of item 4 a) of Table 4.4-4 until- the
, specific activity of the primary coolant is restored to within L
its limits. Whenever the specific activity of the primary coolant exceeds 1.0 uCi/ gram DOSE EQUIVALENT l-131 for in excess of 50-hours for- one continuous time interval or S percent of the unit's total yearly operating time pursuant to ACTION a above, a Special Report shall be prepared and submitted to the C r. ission pursuant to. Specification 6.g.2 within the next 30 days. This report shall-contain the= results of the specific activity analyses-together
- With the following information:
~
'Witn ieyg >,,i;0'F.
CALVERT' CLIFFS -_ UNIT 1 3/4 4 1g Amendment No. g: s-y www *-, ,---,.....v,m-re*.--en ev-,-,,-,.-w-- .--+r-.v-eme-- n-,--- v=-,-er-r -,,-e,
- - - ' ~
REACTOR C00L37 SYSTEM
~
ACT*0N: (C:ntinued) f
- 1. Reactor power history starting 48 hours prior to the first i
' samole in which the limit was exceeded, \
- 2. Fuel burnup by core region, l l
3. Clean up flow history starting AB heurs prior to the first sample in which the limit was exceeded. 4 History of de gassing operation, if any, starting 48 hours prict to the first sample in which the limit was exceeded, and 5. The time duration when the specific activity of the pri-mary coolant exceeded 1.0 ,:Ci/ gram DCSE EQUIVALENT l-1 l SURVE!LtANCE RECUIRES'ENTS
~
4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the samp , gram of Table 4.4-4 i l l 1 1 I UNIT 1 3/4 4-20 CAlvERT CLIFFS 4
.- rp- ~, n . - - , ,- -, , ,-~ -.u-- - - - . - - - - - - - - . - . - - ~ - - - - - - - - - - - -
TABLE 4.4-4 r-PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE M 5 AND ANALYSIS PROGRAM n MODES IN Mf!CH SAMPLE SAMPLE AND 4 TYPE OF MEASUREMENT ANALYSIS FREQUENCY AND ANALY3IS REQUIRED AND At:ALYSIS 7 1, 2, 3, 4 Gross Activity Determination At least once per 72 hours E 1. 1 Isotopic Analysis for DOSE 1 per 14 days
- 2.
EQUIVALENT I-131 Concentration 1 1 per 6 months *
- 3. Radiochemical for E Detennination a) Once per 4 hours, 17,27,37,47,57
- 4. Isotopic Analysis for Iodine whenever the DOSE Including I-131, I-133, and I-135 EQUIVALENT I-131 Y exceeds 1.0 pCi/ gram,
[ and 4 1,2,3 b) One sample between 2 and 6 hours following a TilERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.
# Until the specific activity of the primary coolant system is restored within its ilmits.
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subtritical for 48 hours or longer.
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0 ' 20 30 40 50 -80 70 80 90 100 PERCENT OF RATED THERMAL POWER 1 FIGURE 3.41 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activ}ty > 1.OuCl/ gram Dose Equivalent 1131 1 l I CALVERT CLIFFS - UNIT 1 3/4 4 22
PEAf700 COOLANT SYS9EM - 3/4.a.9 PRESSURE /TEMPERATUDE LIMITS REACTOR COOLANT SYSTEP LIMITING (QO1 TION F00 OPE m ION _ _ . 3.4.9.1 TheReactorCoolantSystem(exceptthepressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4 2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of:
Marimum Allowable Heatuo Rate RCS Temeerature 40UF in any one hour period 10 F in any one hour period 70*0F to 313* 314Fto327[F 600 F in any one hour period 0
> 327 F
- b. A maximum cooldown of:
Marimum Allowable Cooldown pate RCS Tercerature 10ffinanyonehourperiod > 2 0VF 250gFto170F 20 F in any one hour period 0 100 F in any one hour period < 1700F
- c. A maximum temperature change of 05 F in any one hour period, during hydrostatic testing operations above system design pressure.
APPLICABILITY: At all times. ACIR: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out of limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acce) table for continued operations or be in at least NOT STAND 8Y witiin the next 6 hours and reduce the RCS T and pressure to less than 2000F and 300 psia, respectively,wildnthefollowing30 hours. SURVEILLANCE RE0VIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 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 examinations shall be used to update Figure 3.4 2. L CALVERT CLIFFS UNIT 1 3/4 4 23 Amendment No. J/J,146 l
FIGURE 3.a ta C ALVEQT CLIFFS UNIT i HE ATUP CURVE,12 EF PY QE ACTOQ COOL ANT SYSTEM PCESSURE TEMPER ATURE LIMITS 2500 .,.,,,,, ,, , _, ,,,_,,, i , ; u , j _ - . . . - - , , =;,,; ,3 =,._ _ ;;_,,_. l - ---1
.y,a : :: ~. - - 6 in?= HE AT UP M--l. .---J- _
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)
M ,! \ 2000 ~.'_ : _ l
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= TEMPER ATURE
- 51s0*F :
/ ./
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( h1000 L - e'
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.- _/ f' RCS TEMP. H/U RATE
? :
-.,/ 70*F TO 313'F 314*F TO 32PF
$4PF/1 HR 800 ,-' _
$1PF/1 HR
# >32PF $6PF/1 M a . MIN. 80LTUP TEMP,7FF; MAXIMUM PRESSURE ,3 -
FOR SDC OPERATION
- 0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE TC '*I
' The mirdmum boltup temperature b the temperature of the reactor veuel Dange, not the coolant temperature i CALVERT CLIFFS UNIT 1 3/4 a.24 Amendment No. J/J 146 I
j FIGURE 3.4 2b C ALVER7 CLIFFS UNIT 1 COOLDOWN CURVE,12 EFPY RE ACTOR COOL ANT SYSTEM PRESSURE TEMPERATURE LIMITS 2500 _i, , ,; g 9.,
,; _ , _ , g,j j; ==cgq 7,5,--- ----- - --,
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! _' COOLDOWN : m
? TEMPER ATURE _ _
- +-
m 1500 3160'F A m _ {. -- ,' c
- m. l.
l 1000 , ':. ' /
/
JtCS TEMP; CfD. RATE O ' w. l Y ,.- ,/ > 2 50*F S100T/1 WL,
. -' 250*F TO 170'F- 520T/1 HR!
'f'
< 17 0*F - 510*F/1 HR b~~ MIN. BOLTUP TEMP. 70 'F MAXIMUM PRESSURE-FOR S_DC OPER ATION-0 100 200 300 400: 500 600 INDICATED REACTOR COOLANT TE.MPERATURE T C'T'
- The mirdmum boltup temperature b the temperature of the reactor veuel nange, not the coolant temperature CALVERT CLIFFS UNIT 1 Amendment No. J/5146 l 3/4 4 24a l
IAf1".4.4_5 REACTOR VESSEL MATERIAL IRRADIATION SttRVElllANCE SCHEDULE SPECIMEN REMOVAL INTERVAL
- 1. Capsule No. 1 5 years
- 2. Capsule No. 2 14 years
- 3. Capsule No. 3 23 years
- 4. Capsule No. 4 30 years
- 5. Capsule No. 5 35 years
- 6. Capsule No. 6 40 years
'l i
i CALVERT CLIFFS UNIT 1 3/4 4-25 Amendment No. 145
i RE ACTOR COOL ANT SYSTEM PRESSUR12ER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to: 0
- a. A maximum heatup of 100 F in any one hour period, 0
- b. A maximum cooldown of 200 F in any one hour period, and
- c. A maximum spray water temperature differential of 4000F.
APPLICABILITY: At all times. ACTION: 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 engineering evaluation to determine the effects 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 STANDCY within the next 6 hours and reduce the pressurizer pressure to less than 300 psia within the l following 30 hours. SVRVEILLANCE RE0VIREMENTS_ 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. CALVERT CLIFFS - UNIT 1 3/4 4 26 Amendment No,145
PIAC700 C00ttNT SYSTEM OVERPRESSURE PROTECTION SYSTEMS (1HITING CONDITION FOR OIIRATION _ 3.4.9.3 The following overpressure protection requirements shall be met:
- a. One of the following three overpressure protection systems shall be in place:
- 1. Two power operated relief valves (PORVs) with a lift setting g 430 psia or
- 2. A single PORV with a lif t setting of 1 430 psia and a Reactor Coolant System vent of 21.3 square inches, or
- 3. A Reactor Coolant System (RCS) vent 1 2.6 square inches,
- b. Two high pressure safety injection (HPSI) pumps' shall be disabled by either removing (racking out) their motor circuit breakers from the electrical power supply circuit, or by locking shut their discharge valves,
- c. The HPSI loop motor operated valves (MOVs)* shall be prevented from automatically aligning HPSI pump flow to the RCS by placing their hand switches in pull to override.
- d. No more than one OPERABLE high pressure safety injection pump
( with suction aligned to the Refueling Water Tank may be used to inject flow into the RC! and when used, it must be under manual control and one of tha following restrictions shall apply:
- 1. The total high pressure safety injection flow shall be limited to 1 210 gpm OR
- 2. A reactor coolant system vent of 2 2.6 square inches shall exist.
APPLICABillTY: When the RCS temperature is s 3270F and the RCS is vented to < B square inches.
- ACTION:
- a. With one PORY inoperable, either restore the inoperable PORV to OPERABLE status within 5 days or depressurize and vent the RCS through a 2 1.3 square inch vent (s) within the next 48 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status,
- b. With both PORVs inoperable, depressurize and vent the RCS through a 2 2.6 square inch vent (s) within 48 hoursi maintain the RCS in a vented condition until either one OPERABLE PORV and a vent of 2 1.3 square inches has been established or both PORVs have been restored to OPERABLE status.
EXCEPT when required for testing. CALVERT CLIFFS UNIT 1 3/4 4 26a Amendment No. 7//J/),146 i
, FLACTOR COOLANT SYSTEM L1H1 TING CONDITION FOR OPERATION (Continued) I
- c. In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and su>mitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.
- d. With less than two HPSI pumps 8 disabled, place at least two HPSI pump handswitches in pull to lock within fif teen minutes AM disable two HPSI pumps within the next four hours, 8
- e. With one or more HPSI loop HOVs not prevented from automatically aligning a HPSI pump to the RCS, immediately place the MOV handswitch in pull to override, or shut and disable the affected HOV or isolate the affected HPSI header flowpath within four hours, Ad implement the action requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as applicable.
- f. With HPSI flow exceeding 210 gpm while suction is aligned to the RWT and an RCS vent of < 2.6 square inches exists,
- 1. Immediately take action to reduce flow to less than or equal to 210 gpm.
- 2. Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the given RCS temperature on Figure 3.4 2a or Figure 3.4 2b.
- 3. If a pressure limit was exceeded, take action in accordance with Specification 3.4.9.1.
- g. The provisions of specification 3.0.4 are not t.pplicable.
f EXCEPT when required for testing. CALVERT CLIFFS - UNIT 1 3/4 4 26b Amendment'No. # ,145 l
ELACTOR COOLANT SYSTEM
$RyllttANCE RE001REMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of a CHANNEL FUNCTIOKAL TEST on the PORY actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
- c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection,
- d. Testing in accordance with the inservice test requirements for ASME Category C valves pursuant to Specification 4.0.5 4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
4.4.9.3.3 All high pressure safety injection pumps, except the above OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours by verifying that the motor circuit breakers have been removed from their electrical power supply circuits or by verifying their discharge valves are locked shut. The automatic opening feature of the high pressure safety injection loop HOVs shall be verified disabled at least once per 12 hours. - Except when the vent pathway is locked, sealed, or otherwise secured in the open position, then verify these vent pathways open at least once per 31 days. e CALVERT CLIFFS UNIT 1 3/4 4 26c Amendment No. J/,145
REACTOR COOLANT SYSTEM l 3/4.4.10 57RUCTURAL INTEGRITY l l ASME CODE CLASS 1. 2 AND 3 COMPONENTS LIMITING CONDITION FOR OPERATlM 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1. APPLICABILITY: ALL MODES. ACTION:
- a. With the structural integrity of any A3ME Code Class I components the structura (s)l integrity of the affected components (snot conformin withinitslimitorisolatetheaffectedcomponents(s))to prior to in 50 greasing the Reactor Coolant System temperature more than F above the minimum temperature required by NDT considerations,
- b. With the structural integrity of any ASME Code Class 2 components (s) not conforming to the above requirements, restore the structural integrity of the affected components (s) to withinitslimitorisolatetheaffectedcomponents(s)priogF to increasir.g the Reactor Coolant System temperature above 200 .
- c. With the structural integrity of any ASME Code Class 3 components the structura (s)l integrity of the affected components (s) tonot confo within its limit or isolate the affected components (s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEllLANCE REOUIREMENTS 4.4.10.1.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated:
- a. Per the requirements of Specification 4.0.5, and
- b. Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2.
CALVERT CLIFFS - UNIT 1 3/4 4 27 Amendment No.143
REACTOR C00L Atn SYSTEM 51%YI1LLANCE RE0VIREMENTS (fontinued) , in addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1. August 1975.* 4.4.10.1.2 Auemented Inservice Inseettien Procram for Hain Steam and Main Feedwater Pioine - The unencapsulated welos greater than 4 inches in nominal diameter in the main steam and main feedwater piping runs located outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice inspection program using the applicable rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code, Section XI 1983 Edition and Addenda through Summer 1983, for Class 2 components. Each weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined, cumulatively, during each 10 year inspection interval. The welds to be examined during each inspection period shall be selected to provide a representative sample of the conditions of the welds, if these examinations reveal unacceptable structural defects in one or more welds, an additional 1/3 of the welds shall be examined and the . inspection schedule for the repaired welds shall revert back as if a new interval had begun. If additional unacceptable defects are detecttd in the second sampling, the remainder of the welds shall also be inspected. Reactor coolant pump flywheel inspections for the first inservice inspection interval may be completed during Unit 1 Refueling Outage No. 10 in conjunction with the reactor coolant pump motor overhaul program.
- I ~~
CALVERT CLIFFS - UNIT 1 3/4 4 28 Amendment No. #!/J25!/J29,143 l l
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1 4/. -- . v 4.e n:s t a.: .NO. 37
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) REACTOR COOLANT SYSTEM 3/4 ETDOWN LINE EJCESS FLOW LIMITING CONDITION FOR OPERATION 3.4.12 Toe bypasi valve for the excess flow check valve in the letdown line shall be c14 sed. APPLICABILITY: MODES 1. 2, 3 and 4. ACTION: With the above bypass valve open, restore the valve to its closed posi-tion within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.4.12 The bypass valve for the excess flow check valve in the letdown line shall be detemined closed within 4 hours prior to entering MODE 4 from HODE 5. CALVERT CLIFFS - UNIT 1 3/4 4-31 l
- - - - - - - - ~ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
_ _ _ _ _ _ __ _ _ _ _ _ ~ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ . _ . _ _ _ . REA TOR COOL ANT sv$TEw
,Ak.3f.lb REA TOR COOL A'iT Sv$'E9 VENTS l
LIMIT!N3 CONDITION FOR OPERATION 3.4.13 One rea: tor coolant syste vent path consisting of two solenoid valves in series shall be OPERABLE and closed at each of the following locations:
- a. Reactor vessel head
- b. Pressurizer vapor space APPLICABILITY: MODES 1 and 2 ACTION:
i
- a. With the reactor vessel head vent path inoperable, maintain the inoperable vent path closed with power removed from the actuator of the scienoid valves in the inoperable vent path, and:
- 1. If the pressurizer vapor space vent path is also inoperable, restore both inoperable vent oaths to OPEMBLE status within 72 hour
- or be in at least HOT STANDBY within 6 hours, or
- 2. If the pressurizer vapor space vent path is OPERASLE, restore the inoperable reactor vessel head vent path to OPERABLE status within 30 days or be in at least HOT STANDSY within 6, hours,
- b. With only the pressurizer vapor space vent path inoperable, maintain the inoperable vent path closed with power removed from the valve actuator of the solenoid valves in the inoperable vent path, and:
- 1. Verify at least one PORV and its associated flow path is OPEMBLE within 72 hours and restore the inoperable pressurizer vapor skace vent path to OPEMBLE status prior to entering MODE 2 following the next HOT SHUTDOWN of sufficient duration, or
- 2. Restore the inoperable pressurizer vapor space vent path to-OPEMBLE status within 30 days, or be in at least HOT STANDBY within 6 hours.
I
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS-4.4.13.1 Each reactor coolant system vent path shall be demonstrated OPERABLE by testing each vt1ve in the vsnt path per Specification 4.0.5. , CALVERT CLIFFS - UNIT 1 3/4 4 32 Amendment No. 119 1
~ l l RE ACTOR COOL ANT SYSTEM RE ACTOR COOL ANT SYSTEM VEN"T5
$URVEILL ANCE REOU1REMTNTS ICemiaueO 4.4.13.2 Ea*h en: tor c lant s) stem vent path shall be demonstrated OPERABl.E at least once p , fueling terval by:
- a. Verifying all manual isolation valves in each vent path are locked in the open position. , ,
- b. Verifying flow through the b:torbotanthstem vent paths with the vent vahes open. . ,
e 1 S 4 a
3/a,$ EuteGENS Cost COOL!NO sysTEws (Eets) S/4,51 Sarf'?Y IN3EF10N TAWS LIMITING COND1i10N FOR OPEDaTION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with;
- a. The isolation valve open,
- b. A contained berated water volume of between 1113 and 1179 cubic feet of berated water (eauivalent to tank levels of between 12,7 2nd 199 inches, respectively),
- c. A boren concentration of between 2300 and 2700 ppm, and
- d. A nitrog(n cover pressure of between 200 and 250 psig ADDLitacilliv: MODES 1, 2 and 3.
ACTION:
- a. With one safety injection tank inoperable, exceot as a result of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN witnin the next 12 hours.
I
- b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY witnin one hour and be in HOT SHUTDOWN within the next 12 hours.
SucVElttaNCE DE001 dew!NT! 4.5.1 Each safety injection tank shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1. Verifying the contained borated water volume and nitrogen cover pressure in the tanks, and
- 2. Verifying that each safety injection tank isolation valve is open.
CALVER,i CLIFF 5 . UNIT 1 Amendment No. #2, J5,140 3/4 5 1
l l EMERGENCY CORE COOLING SYSTEMS SURVEILL ANCE REOUTREMENTS 'Centinuo9
- b. At leut once per 31 days by verifying the boron concentration of the safety injection tank solution,
- c. At least once per 31 days when the RCS prenure is above 2000 psis, by verifying that power to the isolation valve operator is removed by main.
ts.ining the feeder break r open under administrative control.
- d. Within 4 hours prior to increasing the RCS pressure above 1750 psia by verifying, via local indication at the valve, that the tank isolation valve is open,
- t. At leut once perNfueling $terval by verifying that each safety injection tank isolation valve opens automatically under each of the following condi-tions:
- 1. When the RCS pressure exceeds 300 psia, and
- 2. Upon receipt of a safety injection test signal,
- f. Within one hout prior to each increase in solution volume of a 1% of normal tank volume by verifying the boron concentration at the operating high presme safety injection pump discharge is between 2300 and 2700 ppm.
e S I CALVERT CLIFFS - UNIT 1 34 5. Amendment No. //, h.128
EMERGENC'Y CORE COOLING SYSTEMS 0 3/& 5.9, ECCS SUBSYSTEMS T,yg 2 300 F-LIMITING CONDITION FOR OPERATION 3.5.2 Trio independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE high pressure safety injection pump,
- b. One OPERABLE low pressure safety injection pump, and c.- An OPERABLE flow path capable of taking suction from the refueling water-tank on a Safety Injection Actuation Signal and automatically transferring suction to'the containment sump on a Recirculation Actuation Signal.
I l APPLICABILITY: MODES 1, 2 and 3*. ACTION: , a.- With one ECCS subsystem inoperable, restore the inoper able subsystem to OPERABLE status within 72 hours or be in WOT SHUTDOWN within the next 12 hours.
- b. In the e sent the ECCS is actuated and injects water inte the Reactor Coolr.nt System, a Special Report shall be prepared and
- ubmitted to the Commission pursuant to Specification 6.9,2 within 90 days describing the circumstances of the actuation and the total ac umulated actuation cycles to date.
With pressurizer pressure 2 1750 psia, i i CALVERT CLIFFS - UNIT 1 3/4 S-3 Amendment No. 145
I l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE *:
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position
- 1. MOV 659 Mini flow Isolation Open
- 2. MOV-660 Mini flow Isolation Open
- 3. CV 306 Low Pressure SI Open Flow Control
- b. At least once per 31 days by:
l l 1. Verifying that upon a Recirculatioh Actuation Test Signal, the containment sump isolation valves open.
- 2. 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.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing,etc.)ispresentinthecontainment which could be transported to the containment sump and cause restriction of the aump suctions during LOCA conditions. This visual inspection sia11 be performed:
- 1. For all accessible areas of the containment prior to establishing CONTAl*ENT INTEGRITY, and
- 2. Of the areas affected within containment at the completion of containment entry when CONTAl m ENT INTEGRITY is established,
- d. Within 4 hours prior to increasing the RCS pressure above 1750 psia by verifying, via local indication at the valve, that CV-306 is open.
Whenever flow testing into the RCS is required at RCS temperatures of 3270F and-less, the high pressure safety injection pump shall recirculate RCS water (suction from RWT isolated) or the controls of Technical Specification 3.4.9.3 shall apply. i i CALVERT CLIFFS - UNIT 1 3/4 5-4 Amendment No. Jf)146
EhfERGENCY CORE COOLING SYSTEMS SURWTLL ANCE REOUTRE AfENTS (Centinues
- e. At least once perkfuelingMterval by:
- 1. Verifying automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psia.
- 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (truh racks, screens, etc.) show no evidence of structural distress or corrosion.
- 3. Verifying that a minimum total of 100 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP) is con-tained within the TSP storage bukets. '
- 4. Verifying that when a representative sample of 4.010.1 grams of TSP from a TSP storage buket is submerged, without agitation, in 3.5 t 0.1 liters of 77 g 100 F borated water from the RWT, the pH of the mixed solution is raised to r. 6 within 4 hours.
- f. At lent once per$fuelingMterval, during shutdown, by:
- 1. Verifying that each automatic valve in the flow path actu-ates to its correct position on a Safety injection Actuation test signal.
- 2. Verifying that each of the following pumps start automati.
cally upon receipt of a Safety injection Actuation Test Signal:
- a. High-Pressure Safety injection hump.
- b. Low-Pressure Safety injection hump.
4
~/'
CALVERT CLIFFS - UNIT 1 3/4 5-5 Amendment No. /[, h, (([,128
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) i 9 By performing a flow balance test during shutdown following i completion of HPSI system modifications that alter system flow characteristics and verifying the following flow rates for a single HPSI pump system *:
- 1. Tne sum of the three lowest flow legs shall be greater than 470** gpm.
- h. By verifying that the HPSI pumps develop a total head of i 2900 ft. on recirculation flow to the refueling water tank when tested pursuant to Specification 4.0.5.
I
- A HPSI pump system is a HPSI pump and one of two safety injection headers.
**These limits contain allowances for instrument error, drift or fluctuation.
CALVERT CLIFFS - UNIT 1 3/4 5-Sa Amendment No. 31,75,70/, 117 {
?
[ EMERGENCY COPE C00t1NG SYSTEv$ j- 0 3/4 6 3 ECCS SUBSYSTEMS Tavg < 300 F LIMITING CONDITION FOR OPEDATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: 8
- a. Ont OPERABLE high pressure safety injection pump, and
- b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety' Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.
APPLICABILITY: MODES 3* and 4,
- 8. LIM:
l
- a. With no ECCS subsystem OPERABLE, 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 describing the circumstances of the actuation and the total accumulated actuation cycles to date.
SURVEILLANCE RE001REMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the appilcable Surveillance Requirements of 4.5.2. l With pressurizer pressure <-1750 psia. 0 0
# .Between 350 F and 327 F, a transition region exists where the .
l OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and restored to automatic-status on a heatup. At 3270F and less, the l required OPERABLE HPSI pump 0 shall be in pull to-lock and will not start automatically. At 327 F and less, HPSI pump use will be conducted in accordance with Technical Specification 3.4.9.3. l 1 CALVERT CLIFFS - UNIT 1 3/4 5 6 Amendment No. 7//Jfp/Jf),146
EMERGENCY CORE C00llNG SYSTEM 3//4,rg,4 REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:
- a. A minimum contained borated water volume of 400,000 gallons,
- b. A boron concentration of between 2300 and 2700 ppm,
- c. A minimum water temperature of 400 F, and
- d. A maximum solution temperature of 1000F in H0DE 1.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE001REMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:
- a. At'least once per 7 days by:
- 1. Verifying the contained borated water volume in the tank, and
- 2. Verifying the boron concentration of the water,
- b. At least once per 24 hours by verifying the RWT temperature when the outside air temperature is < 40 0F.
1 CALVERT CLIFFS - UNIT 1 3/4 5 7 Amendment No. /2,46145 )
3/4.6 CONTAlf4 MENT SYSTEMS 4 3 /a . 6.1 PRifiARY CONTAINf1ENT - CONTAitdENT INTEGRITY [fMITING CONDITION FOR OPERATION l
- 3. 6.1.1 Primary CONTAIN!!ENT IjiTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3 and 4. - ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS
- 4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
-a. -At least once per 31 days by verifying that all penetrations
- not capable of being closed by-OPERABLE containment 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 Table 3.6-1 of Specification 3.6.4.1.
- b. By verifying:that each containment air lock is OPERABLE per Specification 3.6.1.3.
s
- c. By verifying that the equipment hatch is closed and sealed, ,
pr ior to entering itode 4 following a _ shutdown where the equipment hatch was opened, by conducting a Typt. B test per Appendix J to 10 CFR Part 50.
*Except valves. blind flanges, and deactivated automatic valves which are located inside the containment-and are locked, sealed, or-otherwise secured in the closed position. These penetrations shall be verified closed during , each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days, p .
I CALVERT C'LIFFS - UNIT 1 3/4 6-1 Amendment No.-.yh ,117 1 [ ~. * . e
l COMTAINMENT SYSTEMS CONTAINMENT LEAKAGE . LIMIT 1HG CONDITION FOR OPERATION 3.6.1.2 Containment leskege rates shall be limited to:
- a. An overall integrated leakage rate of:
- 1. <La (346.000 SCCM) 0.20 percent by weight of the containment air per 24 hours at P,, 50 psig, or
- 2. <Lg (61.600 SCCH), 0.058 percent by weight of the containment air per 24 hours at a reduced pressure of Pg . 25 psig.
- b. A combined leakage rate of 1 0.60 L (207.600 SCCM), for all penetra-tions and valves subject to Type B $nd C tests when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L SCCM) or 0.75 L (46.200 SCCM), as applicable, or (b) with the Sea (259.500sured combined leakage fate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L restore the leakage rate (s) to within the limit (s) prior to increasing the ke, actor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the follow-ing test schedule and shall be detennined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using the methods and provisions l of ANSI N45.4 - 1972:
- a. Three Type A tests (overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at either Pa (50 psig) or at7t (25 psig) during each 10-year service period.
l CALVERT' CLIFFS - UNIT 1 3/4 6-2 Amendment No. 75 $d/,112 l 1
CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)
- b. If any periodic Type A test fails to meet either .75 L (259,500 SCCM) or .75 Lt (46,200 SCCM), the test sched0le for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A Tests fail to meet either .75 L, (259,500 SCCM) or .75 Lt (46,200 SCCM), a Type A test shall be performed at least every 18 months until two g consecutive Type A tests meet either .75L L t(46,200SCCM)atwhichtimetheaboveles(259,500 SCCM) or 75 t schedule may be resumed.
- c. The accuraev of each Type A test shall be verified b,< a supplemental test Uhich:
- 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 L, (86,500 SCCM) or 0.25 Lt (15,400 SCCM).
i 2. Has a duration sufficient to establish accurately the i change in leakage between the Type A test and supplemental l 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 rate at Pa (50 psig) or Pt (25 psig).
- d. Type B and C tests shall be conducted with gas tt Pa (50 psig) atintervalsnogrea}erthan24monthsexceptfortests involving air locks. l
- e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
~f. All test leakage rates shall be calculcted using observed data converted to absolute valucs. Error analyses shall be performed to select a balanced integrated leakage measurement system.
- g. -Containment purge isolation valves shall be demonstrated OPERABLE any time upon entering MODE 5 from power operation modes, unless the last surveillance test has been performed
- - - " ~ ~ within the past 6 months or any time after being opened and prior to entering MODE 4.from shutdown modes by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Technical Specification 4.6.1.2.d for A one-time extension has been granted for CVC-515. The test due i March 23, 1991 has been extended to June 21, 1991.-
CALVERT CLIFFS - UNIT 1 3/4 6-3 Amendment No. 7MS, 152 i
CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) { all other Type B or C penetrations, the combined leakage rate is less than or equal to 0,60 L The leakage rateforthecontainmentpurgeiso(207,600SCCM). lation valves shall also be compared to the previously measured leakage rate to detect excessive valve degradetion,
- h. The containment purge isolation valve seals shall be replaced with new seals at a frequency to ensure no individual seal remains in service greater than 2 consecutive fuel reload cycles.
4 l l CALVERT CLIFFS - UNIT 1 3/4 6-3a Amendment No. 7 HUE,152
! CONTAINMENT SYSTEMS CONTAINMENT 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 least one air lock door shall be closed, and
- b. An overall air lock leakage rate of 5 0.05 La (17,300 SCCM) at Pa , 50 psig.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With an air lock inoperable, except as a result of an inoperable door gasket, restore the 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,
- b. With an air lock inoperable due to an inoperable door gasket:
1.. Maintain the remaining door of the affected air lock closed and sealed, and
- 2. Restore the air lock to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: a* -After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours by verifying that the seal leakage is < 0.0002 La (69.2 SCCM) as determined by precision flow measurement when the volume between the door seals is pressurized to a constant pressure of 15 psig, Exemption to Appendix "J" of 10 CFR 50. CALVERT CLIFFS - UNIT 1 3/4 6-4 Amendment No. 75
CONTAINMENT SYSTEMS
. SURVEILLANCE RTOUIREMENTS (Continued)
- b. At least once per 6 months by conducting an overall air lock leakage test et P (50 psig) and by verifying that the overall air lock leakage fate is within its limit, and
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
n a lC.*iVERTCLIFFC - UtilT 1 3/4 6-5
I CONTAIN'4ENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPEP.ATION R3.5.1.4 Primary containment internal pressure shall be maintained between
-1.0 and 1.8 PSIG.
A00LICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment internal pressure outside of the limits.above, restore the internal pressure to within the limits within I hour or be in at least HOT STANDB) within the next 6 hours and in COLD SHUTDOWN within ' the following 30 hours. m SURVEILLANCE REOUIREMENTS
. hh d
4.6.1.4 The primary containment internal pressure shall be determined to jae; within the limits at least once per 12 hours. 1 t a CALVEP.T CL:FFS . UNIT 1 3/4 6-6
' CONTAINMENT SYSTEMS AIR TEMPERATURE
. LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall. net exceed '
120'F. APPLICABILITY: MODES 1, 2. 3 and 4. ACTION: With the containment average air temperature > 120'F. reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at the following locations and shall be determined at least once per 24 hours: , location . . . . . . _
- a. Containment Dome
- b. Containment Reactor Cavity .
3/4 6-7 - - CALVERT CLIFFS - UNIT 1 .
CONTAINMENT SYSTEMS I CONTAINMENT STRUCTURAL INTEGRITY LIMITING COND7 TION FOR ,0PERATION 3.6.1.6 The_ structural integ. ity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With the containment structure exhibiting evidence of possible abnomal degradation per Specification 4.6.1.6.1, perform an engineering evaluation
- demonstrating the ability of the containment structure to continue to perform its design function. If contineed containment integrity cannot be assured by engineering evaluation within 90 days of the surveillance test, be in COLD SHUTOOWN within 36 hours. The reqbframents of Specifica-j tion 3.0.4 are not applicable.
- b. With the structural integrity of the containment not conforming at a level consistent with the acceptance criteria of Specification 4.6.~1.6.2 or L
4.6.1.6.3, restore structural integrity or complete an engineering evalua-i tion that assures structural integrity prior to increasing Reactor Coolant System Temperature above 200'F. ; I t SURVEILLANCE RE0VIRDIENTS p .. . l 4.6.1.6,1
'" - Containment Tendons. The containment tendons' structural integrity shall be demonstratec at five year intervals. The tendons' structural i
integrity shall be demonstrated by:
- a. Detennining that for a representative sample of at least 9 tendons
.. . (3. dome, 3 vertical, and 3 hoop), each tendon has a normalized lif t-off force equalling or exceeding its lower limit expected range for the time of the test -(see Figures 4.6-1, -2, and -3). If the normalized lift-off force of any one tendon in a group lies between the icwer limit expected range and the lower bound individual, an ad,jacent tendon on each side shall be checked for lift-off force. .. -, . . . If both of these tendons are found acceptable, the surveillance program may proceed considering the single deficiency as unique and
- - . . . acceptable. If either of the adjacent , tendons is found unacceptable, it shall be considered as evidence of possible abnormal degradation of the containment structure. In addition, more than one unacceptable-tendon out of those' selected for surveillance (from all three tendon groups) shall be considered as eviden:e of possible abnormal degrada-tion of the containment structure.
. If- the normalized lift-off force of any single tendon lies below the lower bound individual, the occurrence should be considered as evidence of possible abnormal degradation of the containment structure.
l y . CALVERT CLIFFS - UNIT 1 3/4 6-8 knendment No. 86-
1, CONTAINMENT SYSTEMS S.,UkVEILLANCEREOUIREMENTS(Continued) i In addition, detemining that the average of the normalized lif t-off forces for each suple population (hoop, vertical, donO is equal to or greater than the required average prestress level; SU. kips for hoop tendons, 622 kips for vertical tendons, and 555 kips for dome tendons (reference Figures 4.6-1. -2, and -3). If the average is below the required average prestress force, it shall be considered as evidence of possible abnormal degradation of the containment structure. j
- b. Removing one wire from each of a dome, vertical and hoop tendon checked I for lift off force, and detemining over the entire length of the wire:
i 1. The extent of corrosion, cracks, or other damage. The presence of abnormal corrosion, cracks or other damage shall be considered evidence of possible abnormal degradation of the containment
;. . structure. . ; 2. A minimu'n tensile strength value of 240 Ksi (guaranteed ultimate strength of the tendon material) for at least thtte wire. samples l (one from each end and one at mid-length) cut from each removed I wire. Failure of any one of the wire samples to meet the minimum tensile strength test is evidence of possible abnormal degradation ', of the containment structure.
I c. Perfom a chemical analysis to detect changes in the chemical properties of the sheath filler grease. Any unusual changes in physical appearance or chemical properties that could adversely Effect the a_bility of the filler grease to adhere to the . tendon wires or othe' vise inhibit corro-
- sion shall be reported to the Comission pursuant to Specification 6.9.2 I within the next 30 days.
( 4.6.1.6.2 End Anchorages and Adjacent concrete Surfaces. The structural integ-rity of the and anchorages and acjacent concrete surfaces shall be demonstrated by determining through inspection of a rapresentative sample of tendons (refer-ence Specification 4.6.1.6.1) that no apparent changes have occurred in the visual appearance of the end anchorages or their adjacent concrete exterior
- surfaces. Also, inspections of the pre-selected concrete crack pattems adja-cent to end anchorages shall be performed during the Type A containment leakage l rate tests (reference Specification 4.6.1.2) while the containmant is at its maximum test pressure. ,' , .
i ,
- 4.6.1.6.3 containment Surfaces. The exposed accessible interior and exterior surfaces of tne containment, including the liner plate shall be vi,sually I inspected during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2). This inspection shall be performed prior to the Type A containment leakage rate test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak tightness. .
- 4. 6.1. 6. 4 Reports. Any abnormel degradation of the containment structure detected during tne above required tests and inspections shall be reported to the Comission pursuant to Specification 6.9.2 within the next 30 days. This report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective actions taken.
CALVERT CLIFFS - UNIT 1 3/4 6-9 Amendment No. H .H . 109 ____._m____ ______
{ . t I, j , 1
- t 0 800 -
t i 700 - i
. . Expxted 4- 671 i .
k
} Lower Limit , Expected Range e
635
==. % 636 1 ====
8 *"""" an=== f k *mumme g .
- L 1 % ****= eam.,
a I x 600 - ""== 600
. f f.
Required Average -
, M6 a
536 E Y7 ""'"" % ' I 500 M *8'""d ' *=======
""*"'=
Individual
- 4. -
i =====. % N 482 1 i 1 10 40 100
, Years Figure 4.61 Normalized Prestrums Hoop Tendons CALVERT CLIFFS - U".IT 1 3/4 6-9a Amendment No. 86 e
i ee me see e eum enemuum eeS ee e tuump
- em ese en e 9 WWuMWWD W 4Eppe nM e a q, g qgm4g
, ,. . . _ . _ . . _ _ _ _ - _ _ . - . - = . . . . . _ - . . . . . _. ...
4 goo . 700 - Expected 680 -
~
- Lower Limit .
Expected Range 643 g- '% , , 645 g Required AveragD % %
- a. 622 % %
' 622
% 608 .
600 - -.
* ==
595
' == * -
- Lower Bound % . . .
Individual *
====-
. - 560 .
- 500 - .
i l _ __ i i
- 1 10 40 100 I Yun
- - Figure 4.6 2 Normalized Prestress Vertical Tendons -
- - -
l . l l
~
.. CALVERT CLIFFS 'JNIT _1 ,
3/4 6-9b Anen'dment' No 66 l-
800 I e 700 - Expected - 654 c Lower Lirnit H Expected Range t 618 k % % 613 2 600 - %
% 577 Recuired Average 555 555
~
541 % Lower Bound % %
.I.ndiv. idual m
500 - D % 500 1 -- l . 1 10 40 100 Years Figure 4.6 3 Normalized Prestress Dome Tendons - l CALVERT CLIFFS - UNIT 1 , 3/4 6-9c kneninent No. 86 I
CONTAINw!NT SYST!vs CON a!NMENT DURSE svs Ev L!u!T!NG CON 0!T!0N FOR OPESA !?N 3.6.1.7 The containment purge supply and exhaust isolation valves shall be closed by isolating air to the air operator and maintaining the solenoid air supply valve deenergi:ed. APDLICAS!LITY: MODE 51, 2, 3 and 4 ACTION:
- a. With one containment purge supply and/or one exhaust isolation valve open, close the open valve (s) within one hour or he in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With one containment purge supply and/or one exhaust isolation valve inocerable cue :: hign leakage, recair the valve (s) within 24 hours or be in COLD SHUT 00WN within the following 30 hours.
SURVEILLANCE RESU!RE"ENTS 4.6.1.7 The aB inch containment purge supply and exhaust isolation valves shall . be determined closed at least once per 31 days, by verifying that power to ne r:lencic valve .is removed. ,. CALVERT CLIFF 5 - UNIT 1 3/a 6-91 Amendment No.g 7 ,ppp , g g 1
. . ~ . .
1 iCONTAINv!NT SYSTEMS kONTAINMENTVENTSYSTEM LIMITING CON 0! TION FOR OPEUT!ON
- 3. 6.1,8 The containment vent isolation valves MOV 6900 and MOV. 6901 shall be maintained closed by tagging the motor power supply breakers coen and main-taining tr.e keyed hand switches locked in the closed position.'
'APPLICABIL!TY: MODES 1, 2, 3 and 4 ACTION:
With one or both containment vent isolation valves open, elese the open valve (s) within one hour or be in at least HOT STAN05Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, EURVE!LLANCE REOUIREMENTS I 4,6,1.8 The containment vent isolation valves shall be determined closed at - # least once per 31 days by verifying inat power to the motor operators is removed anc tne valves indicate shut. *
.'These requirements shall be deleted upon initial operability of the CRS isolation signal input to MOV 6900 and MOV 6901, l
CALVERT' CLIFF 5 UNIT 1 3/4 E-Ser Amencment No, g 3, 115 o l l
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURi1ATION AND COOLING SYSTE.MS CONTAINMENT SPRAY SYSTg1 LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWT on a Containment Spray Actuation Signal and Safety Injection Actuation Signal and autematically transferring suction to the containment sump on a Recirculation Actuation Signal. Each spray system flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger. APPLICABILITY: MODES 1, 2, and 3*, ]__ ACT'ON: With one containment spray system inoperable, restore the inoperable spray system 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 spray system shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1. Verifying that upon a Recirculation Actuation Test Signal, the containment sump isolation valves open and that a recirculation mode flow path via an OPERASLE ,
shutdown cooling heat exchanger is established.
- 2. -
Verifying)that automatic each valve (manual, power coerated orin th from the RWT on a Containment Pressure-High tes; signal.
'With pressurizer pressure > 1750 psia. .
1 CALVERT CLIFFS - UNIT 1 3/4 6-10 I
con"TANMEN"r SYSm13 EttRVEftl ANCE REOt7REMEN'TS (Ce"itue9
- b. At least once perhfueling Nterval, during shutdown, by:
- 1. Verifying that each automatie valve in the flow path actu-stes to its correct position on Safety Injection Attuation test signal.
2, Verifying that each spray pump stktts automatically on a Containment Spray Actuation test signal.
- c. At least once per 5 years by perfcrming an air or smoke flow test through ea:h spray header and verifying each spray nozzle is unobstructed.
i l l l l CALVERT CLIFFS - (* NIT 1 3'2 6-11 Amendment No.12il 139
1 L CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM ! LIMITING CONDITION FOR OPEPjTION 3.6.2.2 Two independent groups of containment air recirculation and cooling units shall be OPERABLE with two units to each group. APPLICABILITY: MODES 1, 2 and 3. ACTION:
- a. With one group of required containment air recirculation and cooling units inoperable and both containment spray systems OPERABLE, restore the inoperable group of air recirculation and cooling units to OPERABLE status within 7 days or be in at least HDT SHUTDOWN within 12 hours,
- b. With three required containment air recirculation and cooling units inoperable ard both containment spray systems OPERABLE, restore at least ene required air recirculation and cooling unit to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within 12 hours. Restore both above required groups of contain-ment air recirculation and cooling units to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within 12 hours,
- c. With one group of recuired containment air recirculation and cool- l ing units inoperable and one containment spray system inoperable, restore the inoperable containment spray system to OPERABLE status within 72 hours or be in at least HOT SHUT 00WN within 12 hours.
Restore the inoperable group of containment air recirculation and cooling units to OPERABLE status within 7 days of initial loss or be in at least HOT SHUTDOWN within 12 hours. SURVEILLANCE REQUIREMENTS 4.6.2.2 Each containment air recirculation and cooling unit shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Starting each unit from the control room.
- 2. Verifying that each unit operates for at least 15 minutes.
- 3. Verifying a cooling water flow rate of > 2000 gpm to each cooling unit when the full flow service _ water outitt valves are fully open.
. b. At least once per 18 months by verifying that each unit starts automatically on a Containment Soray Actuation test signal. j CALVERT CLIFFS - UNIT 1 3/4 6 12 Amendment No. 93 139
)
CONTAINMENT SYSTEMS 3/4.6.3 IODINE REMOVAL SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3.1 Three independent containment iodine filter trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one iodine 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 SHUTDOWN within the following 30 hours. SURVE1LL.AJiCE REOUIREMENTS 4.6.3.1 Each iodine filter train shall be demonstrated OPERABLE';
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and. verifying that the train operates for at least 15 minutes,
- b. At least once per kefuelin'g Ntervek or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the charcoal adsorbers remove 199% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 20,000 cfm 10%.
- 2. Verifying that the HEPA filter banks remove 2 99% of the DOP when they' are tested in-place in accordance with Regulatory Positions C.5.a and C.S.c of Regulatory Guide 1.52 Revision 2 March 1978 while operating-the filter train at a flow rate of 20,000 cfm 10%.
CALVERT CLIFFS - UN:T 1 3/4 6-13 Amendment No. J2E,142
(ONTAINMENT SYSTEMS SURVElllANCE RE001R[MENTS (Continued)
- 3. Verifying within 31 days after removal that a laborato;y analysis of a carbon sample removed from one of the l charcoal adsorbers demonstrates a removal efficiency of 2 95% for radioactive elemental iodine when the 0 sample is tested in accordance with ANSI H510 1975 (130 0 95% R.H.).
The carbon samples not obtained from test canisters shall be prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections.
- 4. Verifying a filter train flow rate of 20,000 cfm i 10%
during system operation when tested in accordance with ANSI N510 1975.
- c. After every 720 hours of charcoal adsorber operation by:
1 Verifying within 31 Jays after removal that a laboratory analysis of a carbon sample demonstrates a removal efficiency of 2 95% for radioactive elemental iodine when the sample is tested in accordance with ANSI H5101975 (1300C, 95% R.H.). Samples are prepared by emptying a representative sample from an adsorber test tray section, mixing the !1 sorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections. UNIT 1 3/4 6-14 Amendment No. E), 142 CALVERT CLIFFS l 1
CONTAINMENT SYSTEMS SURVtfLLANCE RE001REMENTS (Continued) Subsequent to reinstallino the adsorber tray used for obtaining the carbon sample, the filter train shall be # demonstrated OPERABLE by also verifying that the charcoal 4 adsorbers remove 199% of a halogenated hydrocarbon & refrigerant test gas when they are tested in-place in accordance with Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52 Revision 2 March 1978 while
; Q"S operating the filter train at a flow rate of 20,000 cfm ?
i 10%.
- d. At least once peMfuelingNfitervaf by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the filter train at a flow rate of 20,000 cfm i 10%.
- 2. Verifying that the filter train starts on a Containment Isolation test signal.
CALVERT CLIFFS - UNIT 1 3/4 6-15 Amendment No. 33/E3/12E.142
LQNTAINMENT SYSTEM 1 l iiREElllANCE RE0VIREMENTS (Continuid)
- e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 1 99% of the DOP when they are tested in place in accordance with Regulatory Positions C.S.a and C.5.c of Regulatory Guide 1.52 Revision : March 1978 while operating the filter train at a flow rate of 20,000 cfm 104.
- f. After each coraplete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 1 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52 kevision 2 March 1978 while operating the filter train at a flow rate of 20,000 cfm 1 10%.
- g. Af ter maintenance affecting the air flow distribution by testing in place and verifying that the air flow distribution is uniform within i 20*4 of the average flow per unit when tested in accordance with the provisions of Section 9 of
' Industrial Ventilation" and Section 8 of ANSI H510 1975.
4 4 CALVERT CLIFFS - UNIT 1 3/4 6 16 Amendment No. 72.142
CONT A1NAirNT SYSTT A15 3 /d.6 4 CONT AINAirNT 1901 ATION \' Al\'rS List 1 TING CONDITION FOR OPER ATION 3.6.4.1 The containment isolation vnhes spe:ified in Tatte 3 61 shall be OPER ABLE with isolation times as shown in Tabte 3.61. APPLIC ABillT Y: blODES 1, 2, 3, and 4. ACTION With one or more of the isolation valve (s) spe:ified in Table 3.61 inoperable, either;
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate en:h affe:ted penetration within 4 hours by use of at least one dea:tivated automati: vahe se:ured in the isolation position, or
- c. Isolate the affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- d. Be in at least HOT STANDDY within the next 6 hours and in COLD SHL*TDOWS within the following 30 hours.
- e. The prosisions of Specification 3.0.4 are not applicable provided that the affe ted penet/ation is isolated St*RT*T111 ANCE RTOl'IRT\1rNTS 4.6 1.1.1 The isolation tabes specified in Table 3,6-1 shall be demonstrated OPER APLE prior to returning the vahe to service after maintenan:e, repair, or repla:ement work is performed on the valve or its asso:iated a:tuator, control, or power cir:uit by performe%:e of a cy: ling test and strifi:stion of isolation time, CALVERT CLIFFS UNIT 1 3 4 6 li Amendment No. M i3C
CONT *INNirNT ST STTNIS i SL*PVilLL ANCE RFOL' IRE \1f NTS IC" tiered' 46J.l.: Ea:h is:!stion vahe specified in Table 3.61 thall be demonstrated PERABLE durips the COLD SHUTDOWN or REFUELING MODE at least on:e per , fuelind interva!7 L y-
- a. Verifying that on ea:h containment isolation Channel A or Channel B test signal, ea:h required isolation valve actuates to its isolation posi-tien.
- b. Verifying that on en:h Containment Radiation High Test Channel A or Channel B test signal, both required containment purge valves actuate to their isolation position.
- c. Verifying that on each Safet> Inje: tion Actuation Channel A or Channel B test signal, ca:h required isola ion vahe a tuates to its isolation position.
4.6.4.1.3 The isolatien time of en:h power ortrated or automati: vahe of Ta-ble 3.61 shall te determined to be within its imit when tested pursuant to Te:hni-cal Spe:ift:ation 4.0.5. 9 CALVERT CLIFFS UNIT 1 3: 6 11, Amendment No. [/, f, //411:.
'9 n - .
TABLE 3.6-1 N
;5 s
/
k L s CONTAINNENT ISOLATION VALVES 5 n PENETRATION ISOLATION ISOLATION VALVE . ISOLATION g NO. ___CHANN[t$- IDENTIFICATION NO. FUNCTION TIME (5tCOND5]_ I E 1A SIAS A PS-5465-CY- R.C. and Pressurizer Sampling <7 SIAS A PS-5466-CY ~1 g SIAS A PS-5467-CV 77
- SIAS B PS-5464-CY 77 18 SIAS A WGS-2180-CY Containment Vent Header to Waste <7 SIAS 8 WGS-2181-CY Gas 77 IC SIAS A CVC-506-CV REP Seals Controlled Bleedoff <7 SIAS 8 CVC-505-CY 77
- 10 NA PS-6529-SY* Post Accident Sampilns gg L - tiquid Return to RC Drafn Tank
! 2A SIAS A CVC-515-CY Letdown Linc <13 SIAS 8 CYC-516-CV 713 NA CVC-105 ~NA NA CVC-103 NA N 1 g- 28 MA CVC-517-Cy Charging Line NA i 3 NA CYC-518-CY NA E CVC-519-CY NA
& MA CYC-435-RY NA
( NA CVC-184 MA 4 *. 4 =
=
i J
TAOLE 3.5-1 (Continued) O Q CoffTAIMMUtr 150LATIGIl VALVES 9 P q POIETRATICII ISOLATICII ISOLATICII VALVE ISOLATICII C NUMBER QIAINIELS . ]DDITIFICATI0lfNo. FUleCTION 0 TIME [5EC_ND]5 h 7A R' Blind Flange ILRT m 5-MA ILRT-I m w
- 78 M Blind Flange ILRT m M ILRT-2 MA t
8 SIAS A EAD-5462-MOV Contal.went Norwal Sump S 13 m SIAS 8 EAD-5463-MOV g 13 h
$o 9 m 51-340 Containment Spray MA .
R 51-326 m 10 NA 51-330 Containment Spray MA M 51-316 NA 2 , R 13 CRS A CPA-1410-CV(3) Purge Air Inlet $7" 2 CRS 8 , CPA-1411-CV(3) 5" 7 E
~
t
> (
.w
- n. i j TABLE 3.6-1 (Continued) !' 'G- CONTAllWIDIT !$8 TAT 19N VALVES ! 9.
~
.? POICTRAY1911 ISOLATIGII ISOLATION VALVE ISOLATION
~
IIUNBR 95WWEL$ ISENTIFICATISIE NO. FINICTIsII TInC [5CCONO_SJ m -
. .14 CRS A CPA-1412-CV(3) PurTe Air Outlet $ **
7
- g CRS 5 CPA-1413-CV(3) $1 ** ~
a
~
i 15 SIAS A RE-5291-CY Purge Air Nonitor s7 , SIAS B RC-5292-CV 57 1-i 16 CIS A CC-3832-CV Component Cooling Mater Inlet 5 13
- w l 1 18 CIS S CC-3833-CY Component Coeling Water Outlet s 18 l ?
~
i
~
if 19A M IA-33T Instrument Air M CIS A IA-2000-NOV $ 13 198 m PA-1940* Plant Air M g M PA-1044* M l E-m j 2 204 m N,-344 Nitrogen Supply M A M -612-CV* M
-z IIR N -622-CV*- M P M N -632-CV* M g M N -642-CV* M i k
?
?
$ G
i i J' . e ) ,., TABLE 3.6-1 (Continued) i N g CONTAINMENT T50 TAT 10N YALVE5
- 5 j PENETRATION ISOLATION ISOLATION VALVE r, ISOLATION NO. CHA8NIEL IDENTIFICATION NO. FUNCTION C, TIffE (SECONDS)
E 208 - NA N2-389 Nitrogen Supply M NA N2-345 NA , E _, 20C NA N2-346 Nitrogen Supply M NA N2-392 NA t 23 SIAS A RCW-4260-CV R.C. Drain Tank Drains ;7 i w 24 SIAS B j PS-6531-SY 0xygen Sample line 17 i os l b 37 NA P5W-1019 Plant Water NA l NA P5W-1000 M i 38 NA DW-5460-CV* Denineralized Water NA
- y 39 NA 51-463 Safety Injection Tank Test Line NA l NA SI-455 NA R
~
41 NA SI-652-MOV (2) , Shutdown Cooling NA i 5
' MA SI-G51-MOV (2) MA
?
5 m s -
TAfflE 3.6-1 (Continued). C0fliAINfKtfi 1501AT10f' VALVE 5
] I 1501ATION ;
N PitttIRATIOM 150tATIOM 1501ATION VALVE NO. CIITMMil. IIIENTIFICATION HO. IllMCTIUM TIM d 5f(CN05}_ MA e 44 NA FP-141-A fire Protection NA c HA IP-141-B HA 5-e ftA TP-6200-MOV* a Ilyifrngen Sample Outlet MA 47A NA PS-65405-5V
- NA MA PS-650TA-SV*
- Ilydrogen sample Outlet NA 4 78 NA PS-6540E-SV*
NA Y NA PS-6507E-SV* . T NA El 47C NA PS-6540F-5V* Ilydrogen Sample Outlet NA NA F5-650/f-5V* PS-6540G-5Y* liydrogen Sample Return MA 41D !!A PS-6507G-SY* NA NA N
= 4nA SIAS-D llP-6900-MifV(4) Contalrunent Vent Isolation < 15 y SIAS A IIP-6901-Hf)V(4) [15
~
_W 9-w h W
n n 4 E TAttE 3.6-1 (Contimmed) CONTAIIMENT 150t AT1918 VALVts 8 Piet[1RA11006 . 150tAff0R 3501AilGR WALVE. IsotA1104 j NO. _CHANNf L IDENTIF1CAil04110._ FtRICTION IIMt (5 ICON 05) e 488 lea ISP-ID4 Mydrogen Parge Inlet ! e: Ig4 HA HP-6903-fWV NA. j .-a
~
49A IIA ' PS-65408'SV* Nydrogen Sample IIA
' NA: -PS-65018-5V*
, NA 498 IIA PS-6540C-5V* Ifydrogen Sample 91 4 PIA PS-65010-5V* MA e-e 49C nA r5-65400-SV* nydrose. Sample 4, MA nA le
.s PS-65019-SV*
IIA 1 50 NA Blind Flange ILRT IIA IIA . Blind Flamje nn g 59 IIA STP-110 Refueling reel Inlet MA a: IIA STP-ill , MA R -
=
l T. 60 IIA ES-144 Steam to Reae. tor Need Laydown NA l- 2 NA [5-142 .. . ? 1s4 , 5. . h"
'5
- ~ r a . - en w -
MI TABLE 3.6-1 (Continued) h CONTAINMENT ISOLATION VALVES
;3 ISOLATION ISOLATION VALVE ISOLATION PENETRATION P NO. CilMetEL IDENTIFICATION NO. FUNCTION TIME (SECONDS)
G 61 NA SFP-176 Refueling Pool Outiet MA MA STP-174 NA MA SCP-172 NA c , SFP-189 NA 5 a NA 62 SIAS A PH-6579-MOV Containment Ileating Outlet < 13 64 NA Pil-376 Containment lieating inlet NA M T 7 (1) Want or remote manual valve which is closed during plant operation. (2) May be opened below 300*F to establish shutdown cooling flow. (3) Containment purge valves will be shut in MODES 1, 2, 3 and 4 per TS 3/4 6.1.7. l k
- May be open on an intermittent basis under administrative control.
E 2 ** Containment purge isolation valves isolation times will only apply in MODE 6 when the valves are
?+ reoutred to be OPERABLE and they are open. Isolation time for containment purge isolation valves z is MA for MODES 1, 2, 3 and 4 per TS 3/4 6.1.7 during which time these valves must remain closed.
? ,
(4)- Containment vent isolation valves shall be opened for containment pressure control, airborne
=
radioactivity control, and surveillance testing purposes only.
-[
w
%b G
CONTAINMENT SYSTEMS l 3/4.6.5 COMBUST!BLE GAS CONTROL H_YDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICA3_1LITY: MODES 1 and 2. ACTION:
- a. With one hydrogen analyzer inoperable, restore the inoperable analyzer to ,
OPEPABLE status within 30 days or:
- 1. Verify containment atmosphere grab sampling capability and prepare and submit a special report to the Comission pursuant to Specification 6.9.2 within the following 30 days, outlining the ACTION taken, the cause for the inoperability, and the plans and
- tchedule for restoring the system to OPERABLE status, or
- 2. Se in at least HOT STANDBY within the next 6 hours. 4
- b. With both hydrogen analyzers inoperable, restore at least one inoperable l analyzor to OPERABLE status within 72 hours or be in at least HOT STANDBY j within the next 6 hours.
- c. Specification 3.0.4 is not applicable to this requirement. I
- s SURVEILLANCE REOUIREMENTS
; i 4.6.5.hl Each hydrogen analyzer shall be demonstrated OPERABLE at least bi-weekly on a STAGGERED TEST BASIS by drawing a sample from the waste gas '
system through the hydrogen analyzer. f.a 4.6.5.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by perfonning a CHANNEL CALIBRtTION '. using sample gases in accordance with manufacturers' recomendations. 1 CALVERT CLIFFS UNIT 1 3/4 6.?6 A*ndment No. $9i74.8),297.707,119 117, q l l l l
CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS W LIMITING CONDITION FOR OPERATION _ 3.6.5.2 Two independent containment hydrogen recombiner systems shall be OPERABLE. APPLICABillTY: MODES 1 and 2. i ACTION: With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY l within the next 6 hours. l SURVElllANCE REOUIREMENTS 4.6.5.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying during a recombiner l system functional test that the minimum heater sheath temperature increases to 2 7000 F within 90 m'inutes and is I maintained for at least 2 hours. l
- b. Atleastonceper%fuelird)Mtervakby:
- 1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits.
- 2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiners (i.e.,loosewiringorstructuralconnections,depositsof foreign materials, etc.)
- 3. Verifyingduringarecombinersystemfunctionalgestthat the heater shcath temperature increase to 2 1200 F within 5 hours is maintained for at least 4 hours.
l 4. Verifying the integrity'of the heater electrical circuits l- by performing a continuity and resistance to ground test ! following the above required functional test. The resistance to ground for any heater phase shall be 1 2 10,000 ohms. l i l l CALVERT CLIFFS UNIT 1 3/4 6 27 Amendment No. JJS,139,142 i --
CONTAINMENT SYSTEMS ' 3/4.6.6 PENE7RA?!ON ROOM EYHAUST AIR FILTRATION SYSTEM LIMITING CQNDITION FOR OPERATION 3.6.6.1 Two independent containment penetration room exhaust air filter trains shall be OPERABLE. APPLICABILIT1: HODES 1, 2, and 3. ACTION: With one containment tenetration room exhaust air filter train inoperable, restore tie inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILL ANCE REMIREMENTS 4.6.6.1 Each containment penetration room exhaust' air filter train shall be demonstrated OPERABLE:
- a. At least once per 31 days on STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the charcoal adsorbers remove 199% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 2000 cfm 10%.
CALVERT CLIFFS UNIT 1 3/4 6 28 Amendment No.142 l l
CONTAINMENT SYSTEMS SURVfittaNCE RE001REMENT$ (Continued) __
- 2. Verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in place in accordance with
,, Regulatory Positions C.5.a and C.5 d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 2000 cfm i 10%.
- 3. Verifying within 31 days after removal that a laboratory analysis of a carbon sample removed from one of the l charcoal adsorbers demonstrates a removal efficiency of 2 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI H5101975 (300 0, 95% R.H.). l The carbon samples not obtained from test canisters shall be prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections.
- 4. Vertfying a system flow rate of 2000 cfm i 10% during system operation when tested in accordance with ANSI N510 1975,
- c. After every 720 hours of charcoal adsorber operation by:
- 1. Verifying within 31 days after removal that a laboratory analysis of a carbon sample demonstrates a removal efficiency of 2 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI N510 1975 (3000, 95% R.H.). Samples are prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections.
cal. VERT CLIFFS UNIT 1 3/h. 6 29 Amendment No. E3,142 4
CONTAINMENT SYSTEMS SURVEllLANCE RE001REMENTS (Continued) , subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the filter train shall be demonstrated OPERABLE by verifying that the charcoal adsorbers remove 199% of the halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S.a and C.5.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 2000 cfm i 10%.
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the filter train at a flow rate of 2000 cfm i 10%.
- 2. Verifying that the filter train starts on Containment Isolation Test Signal.
l { l CALVERT CLIFFS - UNIT 1 3/46.30 Amendment No. JS/SJ,142
~ n y y -
-, ,,.,y ow- wom+t.c
CONTAINMENT SYSTEMS i@l[1LLAME RE0VIREMENTS (Continued 1
- e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in place in accordance with Regulatory Positions C.S.a and C.5.c of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 2000 cfm i 10%.
- f. After each com>1ete or partial reslacement of a charcoal adsorber bank 1)y verifying that tie charcoal adsorbers remove 1 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S.a and C.5.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 2000 cfm i 10%.
- g. After maintenance affecting the air flow distribution by testing in place and verifying that the air flow distribution is uniform within i 20% of the average flow per unit when tested in accordance with the provisions of Section 9 of ;
" Industrial Ventilation' and Section 8 of ANSI N510 1975.
l l l l l 1 I CALVERT CLIFFS UNIT 1 3/4 6 31 Amendment No JE,142
1 l i j!3/4.7 PLU'T SYSTEMS
; 3.4.7.1 TV:B:N! CYCLE g.
I SAFETY VALVES I i . l 4 LIM:T!NG Cd CITION FOR OPED.ATION 3.7.1.1 All main steam line codt safety valves shall be OPER.A!LE*. j l MODES 1, 2 and 3.
,; APPLICABILITY:
,i ACTION:
- a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves
, inocerable, operation in MODES 1, 2 and 3 may pro:eed ;roviced I
that, within a hours, either the in ;erable valve is restered to
' OPEPABLE status or the Power Level-High trip setpoint is reduced
.I per Table 3.7-1; otherwise, be in at least HOT STANOBY within the
., folloning 30 nogrs.
1
- b. With ene reactor :colant loop and asse:iated steam generator in operati0n and with one or more main steam line code safety valves associated with the operating steam generator inoperable, opera- '
q tion in MODES 1, 2 and.3 may proceed provided:
- 1. That at least 2 main steam line code safety valves on the non-operating steam generator are OPERAELE, and
- 2. . That within 4 hours, either the inoperable valve 15 .sstored to OPEPABLE status or the Power Level-High trip setcoint is !
redu:ed per Table 3.7-2; otherwise, be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the i following 30 hours,
- c. The provisions of Specification 3.0.4 are not applicable.
5 RVEI'. LANCE REQUIREMENTS 4 i.
! : 4. 7.1.1 No additional Surveillance Retuirements other than these reevired I; by 5:ecification 4.0.5 are applicaole for the main steam line c:de safety
.: valves of Taole 4.7-1.
I l ' Entry into MODE 3 is ;ermitted to cetermine c:erability of mair steam l'ne l ( .: c:te safety valves. During :nis time, at least 2 main team line ::de sa'e:y I g valves per steam generator shall ce ::erable. j t , CALVERT CL FFS - L'N:71 3/4 7-1 Amencment No.104 - i
4
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'CALVER,7 CLIFFS-UN:7 1 3/4 7 2 l
Y r +'ww-c e~ w w e w =* = ~ee e e n w , "
n . -TABL E 3. 7-2 i' d PRX1Mim Alt 0MRRLE POWER LEVtt.-iflGI 1 RIP '.E1P0lNT WITil INDl'tRABi t
- , .s._i_t_W1.iC5Af tliVhi.Vts tiiRiiii OPi Rh_l_l.l.M iillil'(m(' Si_t_/N_. P.tNE_Rh1..(Nt_
l' C a T.;
' Nxism Allowahic Power
.2 ' Nximm Ntaber of Inoperable Safety level-!!igh Trip Setpoint
-_E Valves'on The Ogyrating_ Steams Generator (Percent of' HAf tII littRMAL NMN R)
Ei
~
l-1 go 2 35 I i 3 a . i l N I ; he l O O . : 00 , e e f 1 e _ r
- w'l4*-+ . - _ _ _ _ _ _ _ _ .___A., . _ . _ _ ___ _ m._. , s _ __ , _ __,,
r
- n. TAntE 4.7-1 p .*
,1, - STEAM LINE_ SAFETY VALVCS PIR LOO _P __
U p -VALVE LITT SETTINGS
- ALLOWABLE ORirIEr 5170 4;
? -a. . 'RV-3992/4000 935-995 ps19 R h h. RV-3993/4001 935-995 psig R M
- ] c. RY-3994/4002 935-1035 psig R
- d. RY-3995/4003 935-1035 psig R
- c. RV-3996/4004 935-1065 psig R
- f. RV-3991/4005 . 935-1065 psig ,
R
- 5. 9 RY-3998/4006 935-1065 psig R
[ h. RY-3999/4007 935-1065 psig R
*Lif t settings ~ for a given steam line are also acceptahic if any 2 valves lif t between 935 anil 995 p,ig,
& any 2 other valves lif t between 935 and 1035 psig anel the 4 remaining valves lif t between 935 arul 1065
} ps19.
ll 3 g
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- - - - - - + - - - - -' -- -- d - - m
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{ , l * ' o. 1 ' j PLXNT SYSTEMS AU11LIARY FEEDWATE:t SYSTEM LIMITING CONDITION F0tt OPERATION _
' 3 .7.1.2 Two auxiliary feedwater trains consisting of one steam driven and one s etor-driven pump and associated flow paths capable of automatically initiating flow .shall be OPERABLE. (An OPERABLE steam-driven train shall consist of one pump aligned for automatic flow initiation and one pump aligned in standby.)*
i APPLICABILITY: MODES 1, 2 and 3. ACTION:
;* - a. With any single pump inoperable, perform the following:
f 1. With No.12 motor-driven pump inoperable: (a) Align the standby steam-driven pump to automatic initiating I status within 72 hours or be in HOT SHtTTDOWN within the next 12 hours, and l i Restore No.13 setor driven pump to CDERABLE status within the I (b) i i next 7 days or be in HOT SHlfiDOWN vithin the next 12 hours,
- 2. With one steam-driven pump inoperable:
(a) Align the OPERABLE stet:4-driven pump to automatic initiating status within 72 hours or be in HOT SHtTTDOWN within the next 12 hours, and - I (b) Restore the inoperable steam-driven pump to standby status (or I automatic initiating status if the other steam-driven pump is l
~
to be placed in standby) within the next 7 days or be in HOT
' SHUTDOWN within the next 12 hours.
I b. With any two pumps inoperable:
- 1. Verify that the remaining pump is aligned to automatic initiating ~
status within one hour, arLd,
- 2. Verify within one hour that No. 23 setor-driven pump is OPERABLE and valve 2-CY-4550 has been exercised within the last 30 days, and
- 3. Restore a second pump to automatic initiating status within 72 hours [
or be in HOT SHtTTDOWH within the next 12 hours.
'A standby pump shall be available for operation but aligned so that automatic flow initiation is defeated upon AFAS actuation.
3/4 7-5 Amendment No. 5A,67,78,88,109 CALVERT CLIFFS - UNIT 1 l
PLANT SYSTEMS l l f AUXILIARY FEEDWATER $YSTEM _ LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) . I. c. Whenever a tubsystem(s) (a subsystem consisting of one pump. piping, valves and controls in the direct flow path) required for operability is inoper. I able for the performance of periodic testing (e.g. manual discharge valve I closed for g operator (s)will pump Total Dynamic be stationed at theHand localTest or Logic station Testing) (s) with direct acomuni-dedicated cation to the Control Room. Upon completion of any testing, the sub-
! system (s) required for operability will be returned to its proper status and verified in its proper status by an independent operator check.
? d. The requirements of Specification 3.0.4 are not applicable whenever one 1
motor and one steam-driven pump (or two steam-driven ptsnps) are aligned t for automatic flow initiation. I t [- SURVEILLANCE REQUIREMENTS V [ 4.7.1.2 Each auxiliary feedwater flowpath shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
[ I
- 1. Verifying that each steam-driven pump develops a Total Dynamic Head of >2600 ft. on recirculation flow (if verification must be demonstrated during startup, surveillance testing shall be r perfonned upon achieving an RCS tanperature 1300',F and prior to entering MODE 1).
- 2. Verifying that the ector-driven pump develops a Total Dynamic L Head of 13100 ft. on recirculation flow, f
- 3. Cycling each testable, remote-operated valve that is not in its operating position through at least one complete cycle.
I
- 4. Verifying that each valve (manual, power operated or automatic) in the direct flow path is in its correct position,
- b. Before enterino MODE 3 after a CCLD SHUTDOWN of at least 14 days by ' .
, completing a flow test that verifies the flow path from the condensate '
storage tank to the steam generators. .
- c. At least once per 18 months by: i
- 1. Verifying that each automatic valve in the flow path actuates .
to its correct position (verification of flow-ecdulating CALVERT CLIFFS - UNIT 1 3/4 7-Sa Ame dment No. $7,3 g I l l
PLANT SY$TEMS AUXILIARY FEEDSATER $YSTEM h SURVEILLANCE REQUIREMENTS (Continued) I characteristics not required) and each auxiliary feedwater pump automatically starts upon receipt of each AFAS test signal, and
- 2. Verifying that the auxiliary feedwater system is capable of providing a minimum of 300 gpm nominal flow to each flow leg.* l 1
i i s 1 Y
*This surveillance may be performe'd on one flow leg at a time.
I i i CALVERT-CLIFF 5 - UNIT 1 3/4 7-5b Amendn.ont No. O ,88, MA.118 I _,. - , . .,__.. _._-.. ._. . _ . - . . . . _ _ . . . . - _ . . . ~ _ _ _ . . _ . . . _ _ _ . - - - - . . . - . _ _
DELETED FIGURE 3.7-1 4 CALVERT CLIFFS - UNIT 1 3/4 7-Se Amendment No. 67, 78. 8 8 i
PLAN 7 SYSTEM,S, CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The No. 12 condensate storage tank (CST) shall be OPERABLE with a minimum contained water volume of 150,000 gallons per unit. APPLICABILITY: MODES 1, 2 and 3. ACTION: With the No. 12 condensate storage tank inoperable, within 4 hours either: ,
- a. Restore the CST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours, or
- b. Demonstrate the OPERABILITY of the No. 11 condensate storage tank as a backup supply to the auxiliary feedwater pumps and restore the No. 12 condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hour:.
SURVEILLANCE REQUIREMENTS
- 4. 7.1. 3.1 The No.12 condens'te a storage tank'shall be demonstrated "
OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps. 4.7.1.3.2 The No.11 condensate storage tank shall be demonstrated OPERABLE at least o'nce per 12 hours by verifying that the tank contains a minimum of 1505 000 gallons of water and by verifying that the flow path for taking & ction from this tank is OPERABLE with the manual l valves in this fisw path open whenever the No.11 condensate storage l tank is the supply source for the auxiliary feedwater pumps. l CALVERT CLIFFS-UNIT 1 3/4 7-6 g
l l PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPEPATION 3.7.1.4 The specific activity of the secondary coolant system shall be
<_ 0.10 uti/ gram DOSE EQUIVALENT l-131.
APPLICABILITY _: MODES 1, 2, 3 and 4 ACTION: With the specific activity of the secondary coolant system > 0.10 uCi/ gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2. CALVERT CLIFFS-UNIT 1 3/4 7 7 ,
i TABLE 4.7 2 l SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY l SAMPLE AND ANALYSIS )ROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYS!$ AND ANALYSIS FRE0VENCY
- 1. Gross Activity Determination At least once per 72 hours
- 2. Isotopic Analysis for DOSE a) 1 per 31 days, whenever EQUIVALENT 1-131 Concentration the gross activity determina-tion indicates iodine con-centrations greater than 10%
of the allowable limit. ' b) I per 6 months, whenever the gross activity determination indicates iodine concentre. tions below 10% of the i allowable limit, i l. u { . l
?
I CALVERT CLIFFS-UNIT 1 3/4 7-8 ( M t
,_~~i--,a._._.....,~,.---.,_.,-...._.,-_ - , , - - - - . . , , , . - . - . , , . .- . _ - , - - . -
PLANT SYSTEMS , KAIN STEAM L!NE ISOLATION VALVFS LIMITING CONDITION FOR OPERATION l 3.7.1.5 Each main steam line isolation valve shall be OPERABLE. APPLICABILITY: MODES 1. 2 and 3. ACTION
- l HDDE 1 -
With one main steam line isolation valve inoperable, i POWER OPERATION may continue provided the inoperable i valve is either restored to OPERABLE status or closed i within 4 hourst otherwise, be in HOT SHUTDOWN within the next 12 hours.
- MODES 2 -
With one r.ain steam line isolation valve inoperable, and 3 subsequent operation in MODIS 1, 2 or 3 may proceed : provided:-
- e. The isolation valve is maintained closed.
- b. The provisions of Specification 3.0.4 are not applicable.
1 Otherwise, be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS
- 4. 7.1. 5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure in less than 5.2 seconds when tested pursuant to Specification 4.0.5.- l_
)
t i CALVERT CLIFFS - UNIT 1 3/4 7-9 Amendment No. /J. 72,126
I i i 4 I Deleted F 1 J I CALVERT CLIFFS - UNIT 1 3/4 7-10 Amendment No. 34. 59 i
w .. .. . .. . N . . . . . . .
,%* J . .' * :
l I i I l i 1 t Delete:
.e
,r I.
i i
- i It i
i CAL'!ERT CLIFFS - UNIT 1 3/4 7-11 - fcend.en: "0. 34, 59 - e I
N ir:s- . . . t i i. I i Celeted I i i i CALVERT CLIFFS - UNIT 1 3/4 7-12 knendment No. 3/ , ~9 4 I
, l
~------s_..m-.,,._ _, __ ___ _ . PL ANT SY ST EWS_
3/4.7.2 STEAM GENERAT0*. PRESSURE / TEMPERATURE LIM 1]AT10N_ l LIMITING CONDITION FOR OPERATION 3 .7.2.1 The temperatures of both the primary and sec0ndary coolants in i t he steam generators shall be > 80'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 6pplicable side to
< 200 psig within 30 minutes, and
- b. Perfom an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the stean generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200'F. '
SURVE!LLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant < 80*F. l 3/4 7-13_ Amendment No. 82 139 CALVERT CLIFFS-UNIT 1 f
1 I PL ANT SYSTEMS 3/4.71 COMPONINT COOLINO W ATER SYSTEM { LIMITINO COVDmON FOR OPER ATION 3.7.3.1 At least two component cooling water loops shall be OPERABLE. At least one component cooling water heat exchanger shall be operating and the remaining component cooling water heat exchanger may be in standby. { 1 APPLIC A BILITY: MODES 1, 0, 3 and 4. I ACTION: { With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE hours and in COLD status within 72within SHUTDOWN hours the or be in 30 following at hours. least HOT STANDBY within the 3 StrRVETLL A NCE REOliR EMTNTS I 4 7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:
- a. t At least on:e per 31 days by verifying that each valve (manual, power j
operated or automati:) in the flow psth that is no! locked, sealed, or otherwise secured in position. is in its corre:t position, b. At least on:e per$ekuelirk hntervsf during shutdown, by setifying that ea:h automati: valve servi:ing safety related equipment a:tuates to its f corre:t position on a Safety injection Actuation test signal. I CALVERT CLIFFS - UNIT 1 34 -14 Amendment No 22E 139
PL ANT SYSTEMS
- 3 /4.7.4 SERVICE W ATER SYSTEM LIMITING COVDTTION FOR OPER ATION 3.7.4.1 At least two independent servi
- e water loops shall be OPERABLE.
APPLICABrLrrY: MODES 1, 2, 3 and 4 ACTION: With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at letst HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVETLL ANCE REOUTRE AENTS 4.7.4.1 At least two service water loops shall be demonstrated OPERABLE
- a. At least on:e per 31 days by verifying that each valve (manual, power operated or automatie) in the flow path that is not locked, sealed, or otherwise se:ured in position, is in its correct position.
- b. At least once per hfuelinbntervah during shutdown, by verifying that ea:h automati: vahe servicing safety related equipment actuates to its correct position on Safety inje: tion A:tuaticn and Containment Spra)
A:tuation test signals. CALVERT CLIFFS - UNIT 1 3'4 -15 Amendment No. , 428 139
PL ANT SYSTEMS 3/4.7.5 S AL*fTATER SYSTEM LP.frTTNG COVDTTION FOR OPER ATION 3.7.5.1 At least two independent saltwater loops shall be OPERABLE. A PPLIC ABTLITY: MODES 1, 2, 3 and 4 ACTION: With only one saltwater loop OPERABLE, restere at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHL'TDOWN within the following 30 hours. SURVETLL ANCE REOUTREMENTS 4.7.5.1 At least two saltwater loops shall be OPERABLE. l
- a. At least once per 31 days be verifying that each valve (manual power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once perkr$fuelirbnterva during shutdown, by verifying that > '
each automatic valve servicing safety related equipment actuates to its correct position on a Safety injection Actuation test signal.' Y i e l CALVERT CLIFFS - L' NIT I 3 t 7-16 Amendment No. ((. M 1M l
PLANT SYSTEMS 3/J .'f. 6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM j,141 TING CONDITION FOR OPERATION ___ 2.7.6.1 The control room emergency ventilation system shall be OPERABLE with:
- a. Two filter trains,
- b. Two air conditioning units,
- c. Two isolation valves in each control room outside air intake duct,
- d. Two isolation valves in the common exhaust to atmosphere duct, and
- e. One isolation valve in the toilet area exhaust duct.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: ,
- a. With one 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 SHUTDOWN within the following 30 hours,
- b. With one air conditioning unit inoperable, restore the inoperable unit to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- c. Withoneisolationvalvepercontrolroomoutsyeairintake duc.t inoperable, operation may continue.providapthe other isolation valve in the same duct is maintained closed; p
otherwise, be in at least HOT STANDBY within 6 hours and in
. COLD SHUTDOWN within the following 30 hours.
L' d. With one common exhaust to atmosphere duct isolation valve inoperable, restore the inoperable valve to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in' COLD SHUTDOWN within the following 30 hours.
. e. With the toilet area exhaust duct isolation valve inoperable, I restore the inoperable valve 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.
! p L CALVERT CLIFFS - UNIT 1 3/4 7 17 Amendment No. 5S/7E,92 ,142 L W w r.v-,.-sr*+w, -
;macc>wmw SURVElllANCE RE001REMENTS 4.7.6.1 The control room emergency ventilation system shall be demonstrated OPERABLE:
%d@M
- a. Atleastonceper62 days,onah2eredtestbasis)by deenergizing the backup control room air condTIT57ter and verifying that the emergency control room air conditioners maintain the air temperature s 1040 F for at least 12 hours when in the recirculation mode.
- b. At least once per 31 days by initiation flow through each HEPA filter and charcoal adsorber train and verifying that each train operates for at least 15 minutes.
- c. At least once per 18 nonths or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housing, or (2) following painting, fire or chemical release in any ventilation zone communicating with the,systen by:
- 1. Verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 2000 cfm 10%.
- 2. Verifying that the HEPA filter banks remove 199% of the DOP when they are tested in. place in accordance with Regulatory Positior,s C.S.a and C.S.c of Regulat'ory Guide 1.52 Revision 2 Harch I'978 while operating the ventilation system at a flow rate of 2000 cfm i 10%,
- 3. Verifying within 31 days after removal that a laboratory analysis of a carbon sample removed from one of the l
charcoal adsorbers cemonstrates a remov61 efficiency of 2 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI N510 1975 (30 0C, 95% R.H.). The carbon samples not obtained from test canisters shall be prepared by emptyfng a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections. CALVERT CLIFFS - UNIT 1 3/4 7 18 Amendment No. EMEP,142
PLANT SYSTEMS SURVElllANCE RE0VIREMENTS (Continued!
- 4. Verifying a system flod rate of 2000 cfm i 10% during system operation when tested in accordance with ANSI N510 1975,
- d. After every 720 hours of charcoal adsorber operation by: l Verifying within 31 days afcer removal that a laboratory -
analysis of a carbon sample demonstrates a removal efficiency of 190% for radioactive methyl iodide when the sample is tested in accordance with ALSI N510 1975 (3000, 95% R.H.). Samples are prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at lei:st two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections. Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the filter train shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52 Revision 2 March 1978 whila operating the ventilation system at a flow of 2000 cfm
- 10%.
CALVERT CLIFFS - UNIT 1 3/4 7-19 Amendment No. U ,142
PLANT SYSTEMS SURVEILLANCE RE001REMENTS-(Continued)
- e. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4 inches Water Gauge while operating the ventilation system at a flow rate of 2000 cfm i 10%.
- 2. Verifying that on a control room high radiation test signal, the system automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks and that both of the isolation valves in each inlet duct and common exhaust duct, and the isolation valve in the toilet area exhaust duct, close.
, f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in place in accordance with Regulatory Positions C.5.a and C.S.c of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter system at a flow rate of 2000 cfm i 10%.
- g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 1 99% of a halogenated hydrocarbon refrigerant test gas when :
they are tested in place in accordance with Regulatory Positions C.5.a and C.5 d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter system at a flow rate of 2000 cfm 2 10%. CALVERT CLIFFS UNIT 1 3/4 7-20 Amendment No. 142
PLANT $YSTEMS 3/4.7.7 ECCS PUMP ROOM EXHAUST AIR FfLTRATION SYSTEM LIMfTING CONDITION FOR OPERATION j 3.7.7.1 The ECCS pump room exhaust ventilation system shall be OPERABLE l with one HEPA filter and charcoal adsorber train and two exhaust fans. APPLICABillTY: MODES 1, 2, 3 and 4. ACTION:
- a. With one ECCS pump room exhaust fan inoperable, restore the inoperable fan to OPERABLE status within 7 days or be in at l
least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN l within the following 30 hours. l- b. With the ECCS exhaust filter train inoperable, restore the i~ filter train 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, SURVEILLANCE RE0VIREMENTS l 4.7.7.1 The ECCS pump room exhaust ventilation system shall be demonstrated OPERABLE: l
- a. At least once per 31 days by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and
- verifying that each exhaust fan operates for at least 35 l minutes. i
~
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
l l l l l CALVERT CLIFFS - UNIT 1 3/4 7-21 Amendment 142 l l l -
PLANT systems L SURVEllLANCE RE001REMENTS (Continued)
-1. Verifying that the charcoal adsorbers remove 1 99% of a !
halogenated hydrocarbon refrigerant test gas when they are i tested in place in accordance with Regulatory Positions o C.S.a and C.S.d of Regulatory Guide 1.52 Revision 2 March i 1978 while operating the filter train at a flow rate of l 3000 cfm t 10%.
- 2. Verifying that the HEPA filter banks remove 199% of the DOP when they are tested-in place in accordance with l l= Regulatory Positions C.S.a and C.S.c of Regulatory Guide l l
1.52 Revision 2 March 1978 while operating the filter I train at a flow rate of 3000 cfm i 10. !
- 3. Verifying within 31 days after removal that a laboratory i analysis of a carbon sample removed.from one of the l l
- j. charcoal adsorbers demonstrates a removal efficiency of l L 2 90% for_ radioactive methyl iodide when the sample is I tested in accordance with ANSI N510 1975 (300 0, 95% R.H.). I L The carbon samples not obtained from test canisters shall I
be prepared by emptying a representative sample from an l: adsorber test tray section, mixing the adsorbent L -thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the l . bed. Successive samples will be removed from different test tray sections. -l L 4. Verifying a system flow rate of 3000 cfm 10% during l L system operation when tested in accordance with ANSI i N510 1975. i: L c. After every 720 hours of charcoal adsorber operation by: l e [ Verifying within 31 days after removal that a laboratory l anaiysis of a carbon sample demonstrates a removal efficiency of 2 90% for radioactive methyl iodide when the sample is L- tested in accordance with ANSI N510 1975 (3000, 95% R.H.). # l Samplas are prepared by emptying a representative sample from an adsorber test tray section, mixing-the adsorbent-thoroughly, ' and obtaining samples at least two inches-in diameter and with a-length equal to the thickness of.the bed. - Successive samples will be removed from different test tray sections. t' [ CALVERT CLIFFS
- UNIT 1 3/4 7 22 Amendment No. EJ/E3, 142 l
i L i
PLANT SYSTEMS SURVEILLANCE RE001REMENTS (Continued) Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the filter train shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove a 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory I'ositions C.5.a and C.S.d of Regulatory Guide 1.52 Revision 2 liarch 1978 while operating the ventilation system at a flow rate of 3000 cfm i 10%.
- d. At least once per 18 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is
< 4 inches Water Gauge while operating the filter train at a flow rate of 3000 cfm i 10%.
t CALVERT CLIFFS - UNIT 1 3/4 7-23 Amendment No. SE/EJ, 142
PLANT SYSTEMS
$URVEILLANCE REOUIREMENTS (Continued)
- e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in place in accordance with Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52 ;
Revision 2 March 1978 while operating the filter train at a flow rate of 3000 cfm i 10%.
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S.a and C.S.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the filter train at a flow rate of 3000 cfm 10%.
- g. After maintenance affecting the air flow distribution by testing in place and verifying that the air flow distribution is uniform within 20% of the average flow per unit when tested in accordance with the provisions of Section 9 of
" Industrial Ventilation" and Section 8 of ANSI N510-1975.
l l l l i l l l I 1 i CALVERT CLIFFS - UNIT 1 3/4 7 24 Amendment No. JE, 142
_PL ANT SYSTEMS 3/4.7.8 SNUBBERS _ LIMITING COUITION FOR OPERATION I 3.7.8.1 All safety related snubbers shall be OPERABLE. I APPLICABILITY: MODES 1, 2, 3 and 4 (HODES 5 and 6 fc snubbers located on systems required OPERABLE in those tiODES. ACTION: With one or mere snubbers inoperable, within 72 hours replace or restere the inoperable snubber (s) to OPERABLE status, and perform an engineering evalua-tion
- per Specification AL1i4and c on the supporting component or declare the supported system inoperabletand follow the appropriate ACTION statement for that system. ,
} ~ " " '
SURVEILLANCE REOUIREMENTS - ' 4.7.8.1 Each snubber shall be demonstrated OPERABLE by performance of the followinD augnented inservice inspection program and the recuirements of Specification 4.0.5. As used in tnis Spe:ification, type of snubber shall mean snubbers of tne same design and manufacturer, irrespective,of capacity.
- a. Visual inspectienf Visual inspecticns shall be performed in accordance with the following schedule: .
No. Inoperable Snubbers of Subsequent Visual ** Each Type per Inspection period inspection Period
- 0 18 months + 25%
1 12 months 7 25% 2 6 months T 25% 3, 4 124 days + 25% 5,6,7 62daysi25'. 8 or nore 31 days + 26*, Tne snubbers may be further categorized into two groups: Those accessible and those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule. 1 Safety related snubbers include those snubbers installed on safety related systems and snubbers on non-safety related systems if their failure or the failure of the system on which they are installed would have an adverse effect on any safety related system. A documented, visual inspection shall be sufficient to meet the requirements for an engineering evaluation. Additional analyses, as needed, shall be completed in a reasonable period of time.
** The inspection interval shall not be lengthened more than two steps at a time.
f The provisions of Specification 4.0.2 are not applicable. CALVERT CLIFF 5 - UNIT 1 3/4 7-25 Amendment No. f # ,772,125
PLANT SYSTEMS j SURVEILLANCE REQUIREMENTS (Centinued)
- b. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that tnere are no visible indica-tions of damage or impaired OPERAB]LITY, and (2) that the snubber installation exhibits-no visual indications of detachment from foundations or supporting structures. Snubbers which appear inoper-able as a result of visual inspections may be determined OPERABLE for the gurpose of establishing the next visual inspection interval,
,providfg that (1)-the cause of the rejection is clearly established and' remedied for that particular snubber and.for other snubbers tnat may be generically susceptible; and/or (2) the affected snubber is
- functionally-tested in the as found condition and determined OPERABLE per Specification Adss;W;s.as applicable. When the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined inoperable unieis it can be determined OPEPASLE via functional testing for the purp se of establishing the next visual inspection interval. ,
For the snubber (s) found inoperable, an engineering evaluation shall be performed on the component (s) which are supported by the snubber (s). The scope of this engineering evaluation shall be consistent with the licensee's engineering judgment and may be limited to a visual inspection of the supported componeni.(s). I-The purpose of this engineering evaluation shall be to determine if the component (s) supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure. that the supported component remains capable of meeting the designed service.
- c. Functional Tests At least once per 18 months during shutdown, a-representative sample of 10% of each type of snubbers in use in the-plant shall l be functio _nally tested either in place or in a bench test.' For each snubber that does not meet the functional test acceptance criteria of Specification 4r2 5:6 an additional 5% of that type l-l L
snubber shall be functionally test d until no-more failures are found or until all snubbers of th9 - type have been functionally f.18.l.d s bers 1-63-13 ough 1-63 28 nepd not be . unctihally
- teste'd The Steam Generator snu(iga,pedollo ' -M30,- 198~
untR tpt reTuel sh - i CALVERT CLIFFS - UNIT 1 3/4 7-26_ . Amendment No. SE,y,N), llE
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Snubbers identified as "Especially Dif ficult to Remove" or in "High Exposure Zones" shall also be included in the representative sample.' In addition to the regular sample, snu,bbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested during the next test , period, Failure of these snubbers shall not entail functional testing l
.of additional snubbers. ,
{
, If any snubber selected for functional testing either fails to lock up or fails to move, i.e. frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all generically ;
susceptible snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be L indeoendent of the requirements stated above for snubbers not meeting j the functional test acceptance criteria. For the snubber (s) found inoperable, an engineering evaluation shall be performed on the component (s) which are supported by the snubber (s). The scope of this engineering evaluation shall be consistent with the licensee's engineering judgment and may be limited to a visual inspec-tion of the supported component (s). The purpose of this engineering evaluation shall be to determine if the component (s) supported by the snubber (s) were adversely affected by the inopere'ility of the snubber (s) in order to ensure-that the supported .omponent remains capable of meeting the designed service.
- d. Hydraulic Snubbers Functional Test Acceptance Criteria l 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.
l ' 2. . Snubber bleed, or release rate, where required, is within the specified range in compression or' tension. For snubbers specifically required to not displace under continuous load, the ability of -the snubber to withstand load without displacement shall be verified.
- Permanent or other exemptions from functional testirig for individual snubbers in these categories may be granted by the Commission only if a justifiable
- l. basis for exemption is presented and/or snubber life destructive testing was l performed _ to cualify snubber operability for all design conditions at either 1: the completion of'their fabrication or at a subsequent date.
CALVERT CLIFFS - UNIT 1 3/4 7-26a Amendment No./ 6f'l/,125 t
-s- S5 4 +r + , - - t--e -- , - = - - e = ,,-e- -
P_LANT' SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. Snubber Service Life Monitoring 1
A record of the service life of each snubber, the date at which the designated service life comm.ences and the installation and maintenance records on which the designated service life is based shall be main-tained as required by Specification 6.10.2.ni. At least once per 18 months, the installation and maintenance records for ~each safety related snubber shall be reviewed to verify that the I indi'cated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review.* If the j
. indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated 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.
l s The provisions of Specification 4.0.2 are applicable. I CALVERT CLIFFS - UNIT 1 3/4 7-26b Amendment No. 5f,79),125
-s-- --
t PAGES 3/4 7-27 TH'!OUG8 3/4 7-62 WERE DELETED BY AMENpMENT N3 / 'f . CALVERT CLIFFS - UNIT 1 3/47-27 through 3/4 7-62 Amendment No. 125
PLANT SYSTEMS 3/4.i.9 SEALED SOURCE CONTAMINATION LIMITIN6 CONOITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gama emitting material or 5 microcuries of alpha (mitting material shall be free of 3,0.005 - microcuries o'f removable contamination. APPLICABILITY: At all times. ACTION:
- a. Each sealed source with removable contamination in excess of the above limit shall be imediately withdrawn from use and:
- 1. Either decontaminated and repaired, or -
- 2. Disposed of in accordance with Comission Regulations.
- b. The provisions of Specificatiers 3.0.3 and 3.0.4 are not appli-cable.
6 SURVEILLANCE REQUIREMENTS 4.7.9.1.1 Test Recuirements - Each sealed source shall be tested for leakage and/or contamination by:
- a. The licensee, or
- b. Other persons specifically authbrized by the Comission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. 4.7.9.1.2 Test Frecuencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequencies described below,
- a. Sources in use - At least once per six months 'for all sealed sources containing radioactive material:
CALVERT CLIFFS - UNIT 1 3/4 7-63 Amendment No. ::5 I
PLANT SYSTEP.S SURVEILLANCE REQUIREMENTS (Continued) ,
- 1. With a half-life greater than 30 days (ext,1uding Hydrogen 3),and
- 2. In any form other than gas. .
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another -
licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use,
- c. Startuo sources and fission detectors - Each sealed startup source and fission detector snall be tested within 31 days
~
prior to being subjected to core flux or installed in the core and following repair or maintenance to the source or detector. 4.7.9.1.3 Reoorts - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detecter leakage tests reveal the presence of > 0.005 microcuries of removable l contam[ nation. I CALVERT CLIFFS-UNIT 1 3/4 7-64 knendment No. 85 .
PLANT SYSTEMS 3/4.7.10 WATERTIGHT DCORS LIMITING CONDli10N FOR OPERATION 3.7.10 The following watertight doors shall be closed except when the door is being used for normal entry and exit:
- a. ECCS Pump Room Doors (4).
- b. Service Water Pump Room to Heater Bay Doors (2),
- c. Auxiliary Peed Pump Room to Heater Bay Doors (2),
- d. ET.ergency Escape Hatch, Service Water Pump Room from Penetra-tien Room,
- e. Main Steam Fiping Area from Pipin5 Penetration Room Coor,
- f. Passage to Main Steam Piping Area Door.
Warehouse to Intake Structure Door, Elevation 12'.
- g. .
~
- h. 0utside to Intake Structure Door. ,
- i. Warehouse to Intake Structure Door Elevation 29'.
APOLIC ABILITY : MODES 1. 2, 3 and 4 . ACTION: With one er. more of tne above doors open, restore the door to its closed position within 24 hours or be in at least HOT STAND 5Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REDUIREMENTS 4.*7.10 The above watertight doors shall be determined closed at least
.once per 12 hours. -- - --- ..,g.
CALVERT CLIFFS - UNIT 1 3/4 7-65 l
l' j j I PLANT: SYSTEMS ,_ ., l 3/a.7.11 FIRE SUPPRESSION SYSTEMS-F!RE SUPMESS!ON WATER SYSTEM LIMITING CONDITION FOR OPERAi!ON r 3.7.11.1 The fire suppression water system shall bt OPERABLE with: a; Two high pressure pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header, b, Two water supplies, each with a minimum contained volume of i i 300,000 gallens , and
- c. .An 0FEUBLE flow path capable of taking suction from the Pretreated Water Storage Tanks Numbers 11 and 12 and transferring the water ,
through distribution piping with OPERABLE sectionalizing control or isolation valves to the yar_d hydrant curb valves and the first L valve ahenc of the water flow alarm device on each sprinkler, hese
. stand;ipe or.5; ray system riser required to be OPERABLE per ,
l
$ peci fic a tio ns 3.7 .11. 2, 3. 7.11. 4, a nd 3.7.11. 5.
AFPLICA!!LITY: At all times. 3
~ -
i*- AC710N: 3.- With one pumo and/or one water supply
- inoperable, restore the inoperable equipment- to _0PERABLE status within 7 days or prepare l r
and submit a Soe:ial Report to the-Commission pursuant-to Specifi-E cation.6.9.2 within the neit 30 days outlining the plans and . L
- procedures to be used to provide for the loss of redundancy.in this I system. The provisions of: Specifications 3.0.3 and 3.0.~4 are not applicable. o u
-b. J With tne fire suppression water system otherwise inocerable:
. -1.- Establish a backup f. ire suppression ' water system within 24 hours , and_-
.2. Submit- a Special' Report in -accordance with Specification 6.9.2:
o ~ La) By telephone within 24 hours, I-b) -Confirmed by telegraph, mailgram.or facsimile transmission-I no later than the first working day following the event, and l-c)-- - In writing:within 14 days following the event, outlining the-l action:tiken, the cause of- the inoperability and.the plans -. L and schedule for restoring the system to OPERABLE status. E ( CALVERT CLIFFS - UHli_I
-3/4 7-66 knendment No. 25, y, ga ,
L
.. .. _ _ . . . _ _ ~ __ . _ .
PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS 4.7.11.1.1 The fire su;pression water system shall be demonstrated OPERABLE:
- a. At least er.ce per 7 days by verifying the contained water supply volume.
- b. At least once per 31 days on a STAGGERED TEST BASIS by starting the electric motor driven pump and operating it for at least 15 minutes.
This test shall be performed on a STAGGERED TEST BASIS with the test required by 4.7.11.1.2.a.2.
- c. 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.
- d. At.least once per 12 months by performance of a system flush of the filled portions of the system.
- e. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.
- f. At least once per 18 months by performing a system functional test r which includes simulated automatic actuation of the system l throughout its operating sequence, and:
l I
- 1. Verifying that each :utomatic valve in the flow path actuates to its correct position,
- 2. Verifying that each pump develops at least 2500 gpm at a discharge pressure of 125 psig,
- 3. Verifying that each high pressure pump starts (secuentially) l to maintain the fire suppression water system pressure _2 80 psig. s
- g. At least once per refueling interval by: (1) performing a flow test l of-the system in accordance with Chapter S. Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association, and (2) performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence and cycling each valve in the flow path.that is not testable during plant operation through at least or,e complete cycle of full travel.
l l 4 CALVERT CLIFFS UNIT 1 3/4 7 67 Amendment No. 2f!/fl!/97,129
I:L :',- 5(situ! , I i *! S'JDV!I ELLEEAN: L'::ESE'i S f:o't4 % e:' ii i,
' ; 4. 7.11.1. 2 Tr.e fire pum; citta'. engine sna'.1 te de onstrate: 0 !R15.!:
i. l{ a. A, least once ser 31 cays :y verifying: l l' l 1. Tne diesel fuel oil day st: rage tant contains at least 174 gallons of fuel, and
- 2. The diesel starts from ambient conditions and operates f:r i a: lets: 30 reinutes. This test shall be perfonned on a i STAGGERED TEST BASIS with the test required by Specification l 4.7.II.l.l.b.
\
lj b. At least once per 92 days by ve'ifying that a sam 0le of diesel fuel fron the fuel storage tank, obtained in accordance witt ASTM-0270 65,
'f li is within tne a::e;;acle limits spe:ified in Table 1 of ASTM 0975-74 wnen ene:ket for vis:osity, water and sediment.
l
, c. At least on:e per 15 mor.ths , during shutdown, by:
l,' - l. Subje: ting the diesel to an inspe: tion in accordan:e with
-l procedures prepared in conjun: tion with its ranuf a:turer's
' re:cmmen:ations for the class of service, and l
j' 2. Veri'ying the diesel starts from ambient conditions en :ne
, aut:-start signal and operates for 3 20 minutes v.ile loaded
! with the fire pum;.
! 4,7.11.1.3 ine fire pumD diesel startir; 24 volt battery bank and energer ll snali te demor.s; rate: OPERA 5LE:
l' a. A: least 6nte per 7 days ty verifying that:
- 1. The ele:trolyte level of each battery is above the plates, ll , and I
- 2. The overall battery voltage is 1 24 volts,
- b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.
- c. At least once per 18 months by verifying that:
- 1. The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioriation, and I
( 2. The battery-to-battery and terminal conne:tions are clean, i l tiga.t free of corrosion and coated with anti-corrosion material. I 3/4 7-6B I liCALVERTCLIFFS-U.'i:7 I Anendment he . 2f . /fil, 9 7 l
plint systry
$3 RAY ANS/CD SPRINKLER SYS?f"$
llWITING CONDIT10N FOR OPEpATION 3.7.11.2 The spray and/or sprinkler systems shown in Table 3.7 5 shall be OPERABLE: ApPLICABillTY: Whenever equipment in the spray / sprinkler protected areas is required to be OPERABLE. ACTION:
- a. With one or more of the required spray and/or sprinkler systems inoperable, within one hour establish a continuous fire watch with backu:
fire suppression equipment for those areas in which redundant safe shutdown systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlinin;
- the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, b The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
ipRVElllMCE RE0VIREMENTS a.7.11.2 Each of the above required spray and/or sprinkler systems shall be cemonstrated OPERABLE: .
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path, not locked, sealed or otherwise secured in position, is in its correct position.
- b. At least once per 12 months by cycling each valve in the flow path through at least one complete cycle of full travel.
- c. At least once per 18 months
- 1. By perfoming a system functional test which includes simulated automatic actuation of the system, and verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test signal, t
- CALVERT CLIFFS - UNIT 1 3/4 7 69 Amendment No Jf//J1//H,129 l
, . - , . . . . ~ . . . . - . . _ . - . . - - - . . . . - - . . . - . _ . . . - . . - - . - . . - . .
PLANT SYSTEWS- . i SURVEILLANCE REOUIREMEWTS (Centinued) l By a visual inspection of the ares in the vicinity of each 2. nozzle (s) to verify the spray pattern will not be obstructed.
}
1 I V. l' l L$ CALVERT CLIFFS - UNIT 1 3/4 7 70 Amendment No. 26//JJ,129
TABLJ 3,7 5 . 1 F19E PROTECTION SPP!NKLER$ UNIT 1 m I l SPRJ,K,J ER_ LOCAT10N CONTROL WALVE EL[yAT!0N j 11 Diesel Generator 45'-0" 12 Diesel Generator 45' 0" Unit 1 East Pipe Pen Room 227/316* 5' 0" Unit 1 Aux Feed Pump Room 603* 12' 0" Unit 1 East Piping Area Room 428' 45'-0" Unit 1 East Electrical Penetration Room 429' 45'-0" , , Unit 1 West Electrical Penetration Room 423' 45'-0" l Unit 1 Main Steam Piping Room 315* 45'-0" Unit 1 Component Cooling Pump Room 228' 5' 0" ; Unit 1 East Piping Area 224* 5' 0" Unit 1 Radiation Exhaust Vent Equipment Room 225' 5'-0" Unit 1 Service Water Pump Room 226' 5' 0" Unit 1 Boric Acid Tank and Pump Room 217* 5'-0" Unit 1 Reactor Coolant Makeup Pump Room 216* 5'-0"
. Unit 1 Charging Pump Room 115' (-)10'!0" Unit 1 Misc Waste Mon Room 113* (.)10'0" Cask and Espt Leading Area Rooms 419, 420, 425 & 426* 45' 0" Solid Weste Processing
- 45'-0" Corridors 200, 202, 212 and 219' 5' 0" Corridors 100,103 and 116* (-)10' 0" Cable Chase 1 A* 45'-0" Cable Chase 1B* 4 5 '-0" Unit 1 ECCS Pump Room 119* (-)15'-0" Hot Instrument Shop Room 222* 5'-0" Hot Machine Shop Room 223* 5' 0"
- Sprinklers required to ensure the OPERABILITY of redundant safe shutdown equipment..
CALVERT CLIFFS - UNIT 1 3/4 7-71 Amendment No. N 61
plain SYSTEMS 1 l HALON SY((MS . l LIMITING CONDITION FOR OPEP.ATION ,, 3.7.11.3. The following Halon systems shall be OPEMBLE with the storage tanks havi 4 at 14ast gA of full charge weight (or level) and 90% of full charge pretsure,
- a. Cable spreading rooms total flood system, and associated vertical cable chase 10. Unit 1.
- b. 4150 volt switchgear rooms 27 & 45' elevation Unit 1. l APPLICABILITY: Whenever equipment protected by the Halon system is required to be OPEMBLE.
. ACTION:
- a. With both the primary and backup Halon systems protecting the areas inoperable, within one hour establish an hourly fire watch with backup fire suppression equipment for those areas protected by the inoperable Halen system. Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperabil-ity and the plans and schedule for restoring the system to l OPERABLE status,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMEtn$ 4.7.11.3 Each of the above required Halon systems shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.
- b. At least once per 6 months by verifying Halon storage tank weight (level) and pressure.
- c. At least once per 12 months by performing a visual inspection of the noIZ1e(s) and visible flow paths for obstructions,
- d. At least once per 18 months by verifying the system, including associated ventilation dampers and fire door release mechanisms, actuates manually and automatically, upon receipt of a simulated actuation signal, and
- e. Following completion of e'ajor maintenance or modifications on the system (s), within 72 hours by performance of a flow test through headers and nozzles to assure no blockage.
I CALVERT CLIFFS - UNIT 1 3/4 7-72 Amendment No. 26, 6), H,lDI, E __._.___._.m___ _-....
ttav sysT!w! Fit! @$! STAT 10NS LIMITINO COCITION FOR OPEPJTION , 3.7.11.4 The fire hose stations shown in Tatie 3.7 6 shall te OPEPABLE. ADDLIC4ElllTY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. LLT,13:
- a. With one or more of the fire hose stations shown in Tabit. 3.7 6 inoperapie, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour. Restore the fire hose station (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the piens and schedule for restoring the fire hose station (s) to OPERABLE status.
The provisiersof Specificatiers3.0.3 and 3.0.4 are not applicable. y b. SURVEllLANCE RE001REMEWTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7 6 shall be demonstrated OPERABLE;
- a. At least once per 31 days by visual inspection of the station to assare all required equipment is at the statisn. Hose stations located in the containment shall be visually inspected on each scheduled reactor shutdown, but not more frequently than every 31 days,
- b. At least once per 18 months for hose stations located outside the containment and once per refueling interval for hose stations inside the containment by. .
- 1. Removing the hose for inspection and re racking, and
- 2. Replacement of all degraded gaskets in couplings,
- c. At least once per 3 years for hose stations located outside the containment and once per refueling interval for hose stations inside the containtnent by:
- 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.
- 2. Conducting a hose hydrostatic test at a pressure at least 50 psig greater that the maximum pressure available at that hose statien or replacement with a new hose.
g CALVERT CLIFFS UNIT 1 3/4 7 73 Amendment No. 2f//fj//H ,129
*11.! ? *.!
r::t w;st 5 : :: .$
- N Et!.'i*
- 0N NUw!!* Or d 1! ?* *::'.:
- 1. Centtie e t 10' 2 45' 2 69' 2
- 2. Asriliary Bui'.cing -15 1" l 10'. 7" l
$' 6 27' 3 45' 5 69 4
- 3. Turbine Building. Heater Bay Out: :e Service Water Pump Rooms ano a.4 Ftecwater Pump Rooms 12' 3
.}
Outside Switchgear Room 27' 2 Outside Switchgear Room 45' 3 4 Intake Stru:ture 10 1
' Fire Hose Stations required for primary protection to ensure the OPERABILITY cf safety related equipment. , " hose Stations which serve both Units I and 2.
CALVERT CLIFFS - UNIT 1 3/4 7 74 Amendment No. 26. g, 97 1
PLRNT SYSTEMS iYtRD FIRE HYORA'iT$ AC HCEANT F0$E 400$!$ LIM! TING CON 0171CN FOR OPERATION 3.7.11.5 The fo11 ewing yard fire hydrants and associated hydrant hose houses shall be OFERABLE. APPLICABILITY: Whenever equipment in the areas protected by the yard fire hycrants is required to be OPERABLE. 4 l 4. #6 yard nydrant and associated hydrant hose house, which provides primary prctection for Unit 2 AWT blocknouse.
- b. 87 yard hydrant and associated hydrant hose house, which provides primary protection for Unit 1 RWT tiockhouse. , ,
1 ACTION: .
- a. With cne or more of the yard fire hycrants or associated hydrant l" hose hovses inoperable, within 1 hour have suf ficient atti. tie'nal lengths of 21/2 inch ciameter hose located in an adjacent CPERABLE I hydrant hose-house to provige service to the unprotected area (s) if the inoperable fire hydrant or associated nydrant hose house is the primary means cf fire suppression. Restore the hydrant or hese house to OPERAELE status within 14 days or prepare and submft a j Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of f
the inoperability, and the plans.and schedule for restoring the hydrant or hose house to OPERABLE status t ,
+
- b. The provisions of Specifications 3.0.3 and -3.0.4 are net applicable.
SURVE!LL ANCE REOUIRE'/ENTS I 4.7.11.5 Each of the yard fire hydrants and associated hydrant hose houses , shall be denonstrated CPERABLE:
- a. At least once per 31 days by. visual inspection of the hydrant hose
. house to assure all required equipment is at the hose house.
- b. At least once per 6 months-(once during March April or !4ay and once du' ring September, October or November) by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry -
- and that the hydrant is not damaged. _
ICALVERT CLIFFS - UN!T 1 3/4 7 75 Amendment No.'E1 f/ , p _-...,..-..,-.,_-.J.-.-. . . - . _ .. ,, _ _._ _._...1._._______ , , . . ._ . _ _ . . . . - .
i PLANT SYSTEMS SURVEILLANCE RE0V!Riv!NTS (Centieved)
- c. At least once per 12 months by:
- 1. Conducting a hose hydrostatic test at a pressure at least 50 psig g. enter than the maxiraum pressure available at any yard fire hydrant.
- 2. Inspecting all the gaskets and replacing any degraded gaskets in the cr uplings.
- 3. Perfer.ing a flow check of etc5 hydrant to verify its C F E ?.A!!!. l T Y .
,CALVERT CLIFFS - UNIT 1 3/4 7-76 knentment No. 61 m .,a- , - , - -,--r--- - - , - ~-n,, ---s.-- -;- -. +e- -s , -m-,-m-n.-r,
PLANT SYST!v$ 1 3/a.7.1? PENETRATION FIRE BARR:ERS L!v! TING CON 0' TION! FOR OPERATION I 3.7.12 All fire barrier ;enetrations (i.e., cable penetration barriers, fire-ccors anc fire careers), in fire :ene teuncaries, protecting safe shutdoon areas shall be OPERABLE. APOLICABILITY: At all times. ACTION: .
- a. With one or more of the above recuired fire barrier penetrations inoperable within one hour either establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERAEILITY of fire detectors on at least one side of the inoperable fire barrier and establish an hourly fire watch patrol; or verify the operability of automatic sprinkler systems (including the water flow alarm and supervisory system) on both sides of the inoperable fire barrier. Restore the inoperable fire barrier i penetration (s) to operable status within 7 days or prepare anc {
submit a Special Report to the Commission pursuant to Specifica-tion 6.9.2 witnin the next 30 days outlining the action taken, the cause of the inoperable penetration and plans and schedule for restoring the fire tarrier penetration (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not aop11 cable.
SURVE:LLANCE REOUIREWENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be OPERABLE:
- a. At least once per 18 months by a visual inspection,
- b. Prior to returning a fire barrier penetration to functional status following repairs or maintenance by performance of a visual inspec-tien of the affected fire barrier penetration (s).
CALVERT CLIFFS - UNIT 1 3/4 7 77 Amendment No. 57, 94,113
3/4.8 ELERTtt! CAL POWER SYSTEP,$_ l 3/t. 8.1 A.C. $0'JRCES l 0,P,iMT I NG l gMITING C001T 0N FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be 0?EMBLE:
- a. Two physically independent circuits between the offsite transmission network and the ensite Class 1E distribution system consisting of either:
- 1. Two 500 kV offsite power circuits, or as necessary
- 2. The 69 kV SMECO offsite power circuit described in the Janva- 14, 1977 Safety Evaluation and one 500 kV offsite powercircuitig
- b. 'Twoseparateandindependentdieselgenerators(oneofwhichmay be a swing diesel generator capable of serving either Unit 1 or Unit 2) with: I
- 1. Separate day fuel tanks containing a minimum volume of 375 gallons of fuel for each diesel generator,
- 2. A comen fuel storage system consisting of two independent storage tanks each containing a minimum. volume of 18.250 gallons of fuel, and
- 3. A separate fuel transfer pump for each diesel generator.
APPLICABILITY: MODES 1, 2. 3 and 4. ACTION:
- a. With two offsite circuits of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
sources by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours thereafter; and 4.8.1.1.2.a.4 within 24 hours, unless the diesel generators are already operating. Restore at least two offsite circuits 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.
. b. With one diesel generator inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by perfominD Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours thereafter, and Surveillance Requirement 4.8.1.1.2.a.4 within 24 hours. Restore two diesel generators to OPEPABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and i'n COLD SHUTDOWN within the following 30 hours.
CALVERT CLIFFS - UNIT 1 3/4 6-1 Amendment No. JB,52,JJJ, MV.123 _______._-___.___s
ELECTRICAL POVER SYSTEMS LIMITINGCONDITIONFOROPEFy!0N (Continued) . ACTION: (Continutd)
- c. With two offsite circuits end one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERAEILITY of the remaining A.C. sources by perfoming Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours thereaf ter and Surveillance Requirement 4.8.1.1.2.a.4 within 8 hours, unless the diesel generators are already operating. Restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- d. With three of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by perfoming Surveillance Requirement 4.8.1.1.2.a.4 within 8 hours unless the diesel generators are already operating; restore at least one of the inoperable of f site sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one off-site source restored, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the follow-ing 30 hours. g
- e. With two of the above required diesel generators inoperable, demonstrate the OPIPABILITY of two offsite A.C. circuits by perfoming Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours thercafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDPY within the nsxt 6 hours and in COLD SHUTDOWN within the following 30 hours.
Restore at least two diesel generators to OPERABLE status within 72 hours .from time of initial loss or be in at least HOT STANDBY within s the next 6 hours and iri COLD SHUTDOWN within the following 30 hours.
- f. With one Diesel Fuel Oil Storage Tank inoperable, demonstrate the OPERABILITY of the remaining tank by: 1) perfoming Surveillance Requirement 4.8.1.1.2.a.2 (verifying 36,500 gallons) within one hour and at least once per 8 hours thereafter, and 2) verifying the flow-path from the OPERABLE fuel oil storage tank to the diesel generators within one hour. Restore two storage tanks 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. (NOTE: If the tank is inoperable, maintain an 8,000 gallon alternate fuel source onsite.)
This Action Statement is not applicable for the following conditions:
- 1) when #21 Diesel Fuel 011 Storage Tank is inoperable during the period from hril 1 tN ough Septrher 30 che to the higher Fobald10-i (Al VI, El (
- 2 t i S * 11 l d/R !!* 2 kmet M 91 L .. ,j" ,[] g )) / ,Y/r' ; 12 3 I
ELECTRICAL p0WER SYST FS LIMITING CONDITION FOR OPERATION (Continued) oftornadooccurrencesduringthistimeframeand2)whent6/oDiesel Fuel Oil Storage Tanks are inoperable. Action statement e is applicable for these conditions. SURVE!LLANCE RE0VIREMENTS 4.8.1.1.1. Each required independent circuit between the offsite transmission network and the onsite Class 1E distribution system shall be:
- a. Demonstrated OPERABLE, as follows:
- 1. For each 500 kV offsite circuit, at least once per 7 days by verifying correct breaker alignments and indicated power avail-ability.
- 2. For the 69 kV SMECO offsite power circuit, within one hour of substitution for a 500 kV offsite power circuit, and at least once per 8 hours thereafter during use by verifying correct breaker alignments and indicated power availability; and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by manually transferring unit power supply from the nomal circuit to .
the alternate 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 fuel tank.
~
- 2. Verifying the fuel level in the fuel storage tank.
- 3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank.
- 4. Verifyir,g the diesel starts and accelerates to at least 900 rpm with generator voltage and frequency at 4160 1 420 volts and 60 1 1.2 Hz. respectively,*
- 5. Verifying the generator is synchronized, loaded to 1 1250 kW, and operates for 1 60 minutes.
- 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
I
'All engine starts for the purpose of this Surveillance Requirement may be preceded by an kngine prelube period and/or other wamup procedures reconcended by the unuf acturer so that mechanical wear and stress on the diesel engine is minimized.
CALVERT CLIFFS - UNIT 1 3/4 8-3 Amendment No. JCS,UJ,m,123 l
l ELECTRICAL POWER SYSTEM 3 l l SURVEILL ANCE RE0V!REMMTS (Continued) l '
- 7. Verifying that the automatic load sequencer timer is OPEPABLE with design theinterval.
interval between each load block within + 10t of its
~
- b. At least once per 92 days by verifying that a sample of diesel fuel fr.c the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-81 when checked for viscosity, water and sedicent.
- c. At least once per 184 days by verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in 3 10 seconds,
- d. At least once per 18 months by:
- 1. Subjecting the diesel to an inspection in accordance with proce-dures prepared in conjunction with its manufacturer's recornenda-tions for this class of standby service.
- 2. Verifying the geners, tor capability to reject a load of > 500 hp without tripping.
- 3. Simulating a loss of offsite power in conjunction with a safety injection actuation test signal, and:
g a) VeEfying de-energization of the emergency busses and long shedding from the emergency busses. b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with permanently connected loads, energizes the auto-connected emergency loads through the load sequencer and operates for
> 5 minutes while its generator is loaded with the emergency Toads.
c) Verifying that the high jacket coolant temperature and low jacket coolant pressure trips are automatically bypassed on a Safety Injection Actuation Sig.31,
- 4. Verifying the diesel generator operates for 1 60 minutes while loaded to 1 2500 kW.
S. i Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour rating of 2700 kW. k 4./ YI bi (.? i ' i .
- k' 3[4 h i l )
[7,($dgM N$, )f f ,)h , i 1
ELECTRICAL POWER SYSTEMS SHUTOOWN
.' IMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and
- b. One diesel generator with:
- 1. A day fuel tank containing a minimum volume of 375 gallons of fuel,
- 2. A fuel storage system containing a minimum volume of 18,250 gallons of fuel, and
- 3. A fuel transf,er pump.
APPLICABILITY: MODES 5 and 6. ACTION: 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. SURVEILLANCE REQUIREMENTS
- 4. 8.1. 2 The above required A.C. electrical power sources shall be l
demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2a.5. 1 CALVERT CLIFFS-UNIT 1 3/4 8 5 i
1 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. O!STRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 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 # 11 4160 volt Emergency Bus # 14 480 volt Emergency Bus # 11A or 14B 480 volt Emergency Bus # 14A or 11B - I 120 volt A.C. Vital Bus # 11 120 volt A.C. Vital Bus # 12 120 volt A.C. Vital Bus # 13 120 volt A.C. Vital Bus # 14 APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: 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 least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment and indicated power availability. CALVERT CLIFFS-UNIT 1 3/4 8-6 1
ELE *TACALPCWERSYJI!ws A.C. O!!'R:BU':N SHU~00WN LIMITING CON 0! TION FCR OD! RATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energi:ed from sources of power other than a diesel gene. rator but aligned to en CPERABLE diesel generator: 1 4160 volt Emergency Bus 1 450 volt Emergency Eus 2 - 120 voit A.C. Vital Busses APPLICA!!LITY: MODES 5 and 6 ACTION: With less than the ateve cceplemen* Of A.C. busses OPERABLE and ener;ize . establisn CONTA:NMENT INTEGRITY within 8 hours. SURVEILLANCE REQUIREv!NTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE anc energized from A.C. sources other than the ciesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability. CALVERT CLIFFS. UNIT 1 3/4 8 7 l l
I
!LECTRICAL POW!c SYS*!us D.C. ?!!?RIBUT CN OP! RAT'N{
(!MITING CON 017:0N FCR OP! RATION 3.S.2.3 The felicwing D.C. bus trains shall te energize and OP! USLE:
- a. 125 volt 0.C. bus No. 11, the asscciated 125 volt 0.C. battery ;
bank or as necessary the Reserve Battery, and one associatec full capacity charger,
- b. 125-volt D.C. bus No. 12, the associated 125 volt 0.C. battery bank or as necessary the Reserve Battery, and one associate:
full capacity charger,
- c. 125-volt D.C. bus No. 21, the associated 125 volt 0.C. battery
- bank or as necessary the Reserve Battery, and one associated full capacity charger, j
- d. 125 volt D.C. bus No. 22, the associated 125 volt D.C. battery l
bank or as necessary the Reserve Battery, anc one associated i full capacity charge.. APcLI;AB:LITY: M00!5 1, 2, 3 and 4 ACTION:- l a .- With one 125-volt bus incoerable, restore the inoperable bus :: OPERABLE status within 2 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
. b. With one 125-volt 0.C. battery inoperable and the associated 125-volt D.C. bus not being supplies by the Reserve Battery except during surveillance testing per Specification 4.8.2.3.2.c.1: '
- 1. + Restore the inoperetle battery to OPERABLE status within 2 #
hours, or replace the inoperable battery with the OPERASLI Reserve Battery within the next 2 hours, or 2. De in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. , With both 125-volt battery chargers frem the same O.C. bus-ino:erable:
- 1. Except when necessary during surveil _ lance testing per Speci-1.
- fication 4.8.2.3.2.d.1, restore at- least one 125 volt 0.C.
battery charger to OPERABLE status within 2 hours or be-in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within' the following 30 hours. L
- 2. During surveillance testing per Specification 4.8.2.3.2.d.i.
restore at least one 125-volt 0.C. battery charger to OP!RABLE I status within 4 hours or te in at least HOT STANCBY witnin 6 hosts.and in COLD SHUTDOWW.within the.following 30 nours. CALVERT CU FF3 - UN!T 1 3/4 8-B Amendment No M ,6 ,52,72, M4
ELECTRICAL POWER SYSTEMS LIMIT!NS CONDITION FOR OPERATION (Continued)
- d. I With single cells having a voltage decrease of more than 0.10 volts 4 from the previous perfonnanJe discharge gest (4.8.2.3.2.f.) value, s but still > 2.10 volts per burveillance Navirement 4.8.2.3.2.b.1.,M either restore / replace cells or replace the affected battery with
- the Reserve Battery within 24 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.7.3.1 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 and charger and the Reserve Battery shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The electrolyte level of each pilot cell is between the minimum and maximum level indication marks.
- 2.
- The pilot cell specific gravity, corrected to 77'F and full electrolyte level is > 1.200.
3. The pilot cell voltage is > 2.10 volts. 4 The overall battery voltage is ,> 125 volts.
- b. At least once per 92 days by verifying that:
- 1. The voltage of each connected cell is > 2.10 volts under
. float charge and has not decreased mori than 0.10 volts from the value observed during the latest performance discharge test (4.8.2.3.2.f).
- 2. The specific gravity, corrected to 77'F and full electrolyte level, of each connected cell is > 1,200 and has not decreased more than 0.02 from the value obsirved during the previous test.
- 3. The electrolyte level of each connected cell is between the minimum and maximum level indication marks.
- c. At least once per 18 months by verifying that:
- 1. The cells, cell plates and battery racks show no visual indication of physical damage or deterioration.
- 2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material.
CALVERT CLIFFS - UNIT 1 3/4 8-9 Amendment No pp,gg,92.114
Ilt(7RICAL POWER SYSTEMS I SURVE!LLANCE REQUIREMENTS (Continued) d. At least once per 18 months by verifying that the battery capacity, with the charger disconnected, is adequate to either: 1. Supply and maintain in OPERABLE status all of the actual emergency loads for at least 2 hours when the battery is subjected to a battery service test. At the completion of this test, surveillance 4.8.2.3.2.e shall be performed for the affected battery. The battery shall be charged to at least 95% capacity in i 24 hours, or , 2. Supply a dumy load simulating the emergency loads of the design duty cycle for at least 2 hours while maintaining the battery terminal voltage 1105 volts. At the completion of this test, the battery shall be charged to at least 95% capacity in i 24 hours, excluding the stabilization time. Tne emergency loads of the design duty cycle shall be documented and updated, as appropriate, in the system description contained in FSAR Chapter 8. and updated in accordance with 10 CFR 50.71(e), e. At least once per 18 months, the battery charger
- shall be demonstrated capable of recharging the battery at a rate of 1 400 amperes while supplying normal 0.C. loads or equi. valent or greater dumy load. ,
f. At least once per 60 months by verifying that the battery capacity is at lesst 80% of the manufacturer's rating when subjected to a performance discharge test. 'This perfomance discharge test shall be performed subsequent to the satisfactory completion of the required battery service test.
*Not applicable to the charger associated with the Reserve Battery.
CALVERT CLIFFS - UNIT 1 3/4 8 10 Amendment No. M M.f4.N . M . W
l I l ELECTRICAL DOWER MYST!v$ D.C. DISTR!!UTION SvviDOWN II L1MITING CON 0!?!ON FOR OPERAT!CN
- 1 I
l3.3.2.4 shall be As a minimum, energize: the feilening C.C. electrical e;ui; ment and busses anc OPEFAELE: 2 - 125 volt 0.C. busses, and 2 125 volt battery banks, ene of wnien may be the Reserve Battery, and Che ass 0ciated charger per bank supplying the above O.C. busses. . AFDLICA!!LITY: MODES 5 and 6. ACTION: W th less than the above ccm:lement cf 0.C. e:vicment and busses OP!UELE, e establish CONTAINMINT INTEGRITY witnin 8 neurs. SURVIILLANCE REOUIREMENTS - 4.8.2.4.1 The aseve re;uired 125-volt 0.C. busses shall be detemined
, OPEMBLE and energi:ed at least once per 7 days by verifying correct breaker alignment anc indicated power availability. ,
4.B.2.4.2 The above re:uired 125-yelt battery banks and chargers shall be demonstra:ec OPERABLE per Surveillance Requirement 4.8.2.3.2. l l
/ CALV!RT CL:FF5 UNIT I 3/4 6-11 Amendment No.;p I
t l
3/4.9 REFUELING OPERATIONS - DYv d o0RON CONCENTRATION LIMITING CONDITION FOR OPERATION 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 refueling pool shall be maintained uniform and sufficient to ensure that the more restrictivenof following reactivity conditions is met: N
- a. Either a K vative a115!# anceoffor'0.95 or less, which uncertainties, or includes a 1% ak/k conser-
- i
- b. A boron concentration of 1 2300 ppm, which includes a 50 ppm l conservative allowance for uncertainties.
APPLICABILITY: MODE 6*. ACTION: With the requirements of the above specification not satisfied, ime.diately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 40 gpm of 2300 ppm l boric acid solution or its equivalent until K ~ is reduced to < 0.95 - or the boron concentration is restored to 123@ ppm, whichever is l the more restrictive. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be detemined prior to:
- a. Removing or unbolting the reactor vessel head, and i
- b. Withdrawal of any full length CEA in excess of 3 feet from its fully inserted position.
- 4. 9.1. 2 Theboronconcentrationofthebeactor olant3ystemandthe Sr refueling pool shall be determined by chemical analysis at least 3 ,
times per 7 days with a maximum time interval between samples of 72 hours. l .
*The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.
CALVERT CLIFFS - UNIT 1 3/4 9-1 Amendment No. 48 1
, REFUELING OPERATIONS pk INSTRUMENTATION ,
LIMITING CONDITION FOR OPERATION 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 containment and control room. APPLICABILITY: MODE 6. ACTION: With the requiremehts of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable. 4 SURVEILLANCE REOUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by perfonnance of:
- s. A CHANNEL FUNCTIONAL TEST at least once per 7 days,
- b. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. A CHANNEL CHECK at least once per 12 hours during CORE ALTERATIONS.
CALVERT CLIFFS - UNIT 1 3/4 9-2
+
REFUELING OPERATIONS g ,4S DECAY TIME ,
^
LIMITING CONDITION FOR OPERATION , 3.9.3 The reactor shall be suberitical for at least 72 hours. APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel. ACTION With the reactor suberitical for less than 72 hours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been suberitical for at least 72 hours by verification of the date and time of suberiticality prior to movement of irradiated fu61 in the reactor pressure vessel. s CALVERT CLIFFS - UNIT 1 3/4 p.3 , w&-r-e *er--t- v v m-+*4-eer= w -
*ew--- vP --,e: --- m .- w w - * - .- +- -- v r- w er----ti--, r4
, )
REFUELING OPERATIONS MSA CONTAINMENT PENETRATIONS , LIMITING CONDITION FOR OPERATION l 3.9.4 The containment penetrations shall be in the following status l 1
- a. The equipment door closed and held in place by a minimum of !
four bolts, j
- b. A minimum of one door in each airlock is closed,' and l
. c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual l
- valve, or-
- 2. Se capable of being closed by an OPERABLE automatic contain-ment purge _ valve. ;
i APPLICA81LITY: During CORE ALTERATIONS or movement of irradiated fuel within the contatnment. M: I With the requirements of the above specification not satisfied, ismediately suspend all operations involving CORE ALTERATIONS or movement of irradiated ! fuel in the containment. The pmvisions of Specification 3.0.3 are not J applicable. l
$URVE!LLANCE REQUIREMENTS __
4.9.4 Each of the above required containment penetrations shall be-determined to be either in its closed / isolated condition or capable of-being closed by an OPERABLE automatic containment purge valve within 72 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment by: ,
- a. Verifying the penetrations are in their closed / isolated condition,-or
- b. Testing the containment purge valves per the applicable portions of-
~
Specification 4.6.4.1.2.
- The emergency escape hatch temporary closure device is an acceptable replace-ment for that airlock _ door.
CALVERT CLIFFS - UNIT 1 3/4 9-4 Amendment No.108 l
. _. _ . . _ _ . _ . _ . _ _ _ _ _ _ _ _ _ . __ _ . ~ . . _ _ _ __ _ . . .
- r. " . -
e m c 3 , ji" . '! . a f..,
. .u 4
AEFUELING OPER.ATIONS F" 7" ((30KMUN! CATIONS ,
.=..
... ' 0:;
L,lMITING CONDITION FOR OPER.ATION )!!@ . g;- ,- :
. EF" ~
sm:: . 3.9.5. Direct comunications shall be maintained between the control [ET room and personnel at the refueling station. '!.
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Shen direct comunications between the control room and personnel at the $5itt refueling station cannot be maintained, suspend all CORE ALTERATIONS. . ftM The provisions of. Specification 3.0.3 are not applicable. . 19 :. .. ..
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i SURVE!l. LANCE REOUTREMENTS '. ":. 4.9.5 Direct comunications between the contro1* room and personnel at 2:; .: . .. the refueling station shall be demonstrated within one hour prior to the E. ;,i start of and at:1 east once per 12 hours during CORE ALTERATIONS. :Ef:m ,
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4 i REFUELING OPEPATIONS I REFUEL 1tI3 PACHINE OPEPABILITY g o,M LIMITING CONDITION FOR OPEPATION i S 3.9.6 The refueling machine shall be used for movement of CEAs and/or I fuel assemblies:
- a. The main hoist shall be used for the movement of fuel assemblies and shall be OPERABLE with: ,
- 1. A minimum capacity of 1550 pounds, and
. p
- 2. An overload cutoff limit of 1 3000 pounds. -
The auxiliary hoist shall be used for the movement of CEAs b. which are being removed from or inserted into fuel assemblies s in the core and shall be OPERAELE with:
- 1. A minimum capacity of 1000 pounds, and ;
n "
- 2. A load indicator which shall be used to prevent lifting loads in excess of 1000 pounds. l APPLICAEILITY: Ouring movement of CEAs or fuel assemblies within the reactor pressure vessel. :
ACTION: : With the requirements for refueling machine OPERABILITY not satisfied, '
~
suspend its use from operations involving the movement of CEAs and fuel assemblies within.the reactor pressure vessel. The provisions of - Specification 3,0.3 are not applicable, , y SURVEILLANCE REOUIREMENTS _ 4.9.6.1 The main hoist of the refueling machine shall be demonstrated , OPEPABLE within 72 hours prior to the start of movement of fuel assemblies within the reactor pressure vessel by performing a load test e of at least 1550 pounds with the refueling pooi dry or at least 1420 pounds with the refueling pool flooded and demonstrating an automatic ' load cut off,when the crane lead exceeds 3000 pounds. 4.9.6.2 The auxiliary hoist and associated loti indicator of the refueling machine shall be demonstrated OPERAELE within 72 hours prior I to the start of movement of CEAs within the reac'.or pressure vessel by perfoming a load test of at least 1000 pounds. CALVERT CLIFFS - UNIT 1 3/4 9-5 Amencment No. 3 3
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ii 11 l3.9.7 Leeds in excess of 1500 poun:s shall te ;r: 11:itet f :: tra.>ei eser ' ifuel esse.blies in :ne s : rage peel.
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SURVIILLANCE REOUIRE'iENTS i;4.9.7 The wei;ht of et:P. Iced, 0:5t* thaa a fuel esse .bl.v int CI' s'a'i be '
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- 3.3.5.1 At leas: :ne snu:d:an ::: ling 1:cp sna11 :e in :;eration.'
- lt APPLICAE!L'~Y: t'00E 5 a* all rea:: r water levels. . ..'.
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l a. a'ith less : nae :ne sr.utt:wn : colin; 1:0; in ::ert:1cn', sus:en: l all oCerati*ns inv01v# *; an increase in *he reac*:r de:ay*hea* . load er a **ducti:n in :e-en ::n: tat-eti:n of the Reici:r / Cool a n: Ssstem ene, 1:e:1fi:tily, :ne :nar;1n a ce.ener;1:e* ard the cnarging f':w ;aths $Paiy, ;um;s te clased. snail :e C:se all containmen ;ere:rati:ns pr:vicin; :i e:: a::ess f*: r. e contair. ment timespr.ere :: :ne Outside atmos;nere aitnin 4 ho.rs. The snutd:an c:ci-ir.; ;*.mps m.ay be de-energi:ed durin; 'he ti e /- intervals required fer 1::31 leak rate testin; of contain an .
@- ;enetration number 'l ;ursuant to the re:virements of 5;e:ifict- .
.tion 4.5.1.2.d and/or to permit maintenance on valves loca::: *i in the commen shutd:en c;old e; suction line, ;r:vided (1) ;1 'g
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- eeeraticas are :e-mitted whi:5 could cause dilution of t e ,
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- ne :narging pum:s snail de ce energi:ec an the char;ing f1:* -
g paths shall be close , (2) all CORE Ati! RATIONS are suster:ee, s (3) all c:n' air. ment :enetrations providin; cire:: at:ess fr:m ... lI- *ne *o*:ain~en; a tm*..:nere *D **e *u tsi:e a t:0s;nert ar e *ain- ' tained : :se:, anc N, the kater level a :ve :te ::: :' the ik irradiate: fuel is greater than 23 feet. ,,~
- t. The pr: visions-of 5:e:ification 3.0.3-are no: a;oli:a:;e. -
SURVEILLANCE 2.!!VIREv!NTS e-
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4.9.8.1 A snu;cown :: cling loco shall be determined to be in cpeasti:n anc 0 circulating rea: tor coolant at a flow rate of 1 3000 g;m " a: least Once per i 4 hours. . .s M
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- The shutdown. cooling loop may be removed from operation for _up to 1 -hour per "s;
e Sihourf period 'during the performance _ of CORE ALTERATIONS .in. the_ vicinity of :ne 4( > L reactor pressure vessel hot legs. ::7 ' A..
"1 1500 gpm when the Reactor Coolant System is drained : a level below the .t_
L micplane of the not leg. . . 0 e CALVERT CLIFFS - UNIT 1 3/4 9 8 Amencment No. 3B, 35. II
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RErVELING OPERATIONS l l 5 HUT??WN COOLING AND COOL ANT CIRCL'L ATION i L!**T!NG CON 0!T!CN FOR OPERATION li -- fl3.9.e.2 Two (2) shutdown cooling loops shall be OPERA!LI*.e. I APPLICABILITY: Mode 6 when the water level ab0ve the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet. ACTION:
- a. With less than the required shutdown cooling loops OPERABLE, initiate corrective action to return icops to OPERABLE status within one hour.
- b. The provisions of Specification 3.0.3 are not applicable.
SUPNEILLANCE RE0V!REv!NTS I 14.9.8.2 No additional Surveillance Requirements other than those required by 15eecification j 4.0.5. a l* Normal or emergency power source may be inoperable for each shutdown cooling l loop.
*0ne shutdown cooling loco may be replaced by one spent fuel tool cooling leep when it is lined up to provide cooling flow to the irradiatec .uel in the reactor core and the heat generation rate of the core is bel'sw the heat removal capacity of the spent fuel pool cooling loop.
CALVERT CLIFFS - UNIT 1 Arendment No. FF*103 l
. REFUELING OPERATf0NS 'N/4.U CONTAINMENT PURGE VALVE ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The containment purge valve isolation system shall be OPERABLE.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the containment purge valve isolation system inoperable, close each of the penetrations providing direct access from the containment at:nesphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REOUIREMENTS 4,9.9 The containment purge valve isolation system shall be demonstrated OPERABLE within 72 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment purge valve isolation occurs on manual initiation and on a high radiation test signal from the containment radiation monitoring instrumentation channels. e CALVERT CLIFFS - UNIT 1 3/4 9-9 Amendment No. 62
. REFUEL!NG OPERATIONS
\0 i 13 pt "4ATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel.
APPLICABILITY: During movement of fuel assemblies or CEAs within the reactoF" pressure vessel while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies.or CEAs within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.
.~. _
SURVEILLANCE REOUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or CEAs, CALVERT CLIFFS - UNIT 1 3/4 9-10 . l
' REFUELING OPERATIONS p,Q, ll SPENT FUEL POOL WATER LEVEL
.w LIMITING CONDITION FOR OPERATION .c 3.9.11 As a minimum, 211s feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel cool. ACTION: With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with 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.
~ . SURVEILLANCE RE001REMDITS 4.9.11 The water level in the spent fuel pool shall be detennined to be at least its mininum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pool. l CALVERT CLIFFS - UNIT 1 3/4 9-11 Amendment No. 108 i l
REFUELING OPERATIONS l SPENT FUEL P00L VENTILATION SYSTEM QC\M _ il
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LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel pool ventilation system shall be OPERABLE with:
- a. One HEPA filter bank,
- b. Two charcoal adsorber banks, and
- c. Two exhaust fans.
APPLICABIL ITY: Whenever irradiated fuel is in the storage pool. ACTION:
- a. With one charcoal adsorber bank and/or one exhaust fan inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided an OPERABLE exhaust fan is in operation and discharging through an OPERABLE train of HEPA filters and charcoal adsorbers.
- b. With the HEPA filter bank inoperable, or with two charcoal-adsorber banks inoperable, or with two exhaust fans inoperable, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage l pool until at least one charcoal adsorber bank, at least one exhaust fan, and the HEPA filter bank are restored to OPERABLE status, c.. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
O SURVEILLANCE REQUIREMENTS 4.9.12 The above required spent fuel pool ventilation system shall be demonstrated'0PERABLE: a .- At least'once per 31 days by initiating flow through I- the HEPA filter bank and both charcoal adsorber banks and verifying that each charcoal adsorber bank and each exhaust fan operates for at least 15 minutes. CALVERT CLIFFS - UNIT 1 3/4 9-12 l l l l l . .. .
REFUELING OPERATIONS SURVE1LL ANCE REQUIREMENTS (ContinufJlD
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housing, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the charcoal adsorbers remove 199% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.S a and C.S.d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 32,000 cfm i 10%.
- 2. Verifying that the HEPA filter banks remove 2 99% of the 00P when they are tested in place in accordance with Regulatory Positions C.S.a and C.S.c of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 32,000 cfm i 10%.
- 3. Verifying within 31 days after removal that a laboratory
. analysis of a carbon sample iemoved from one of the l charcoal adsorbers demonstrates a removal efficiency of 2 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI N510 1975 (3000, 95% R.H.). l The carbon samples not obtained from test canisters shall be prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections. 4 Verifying a system flow rate of 32,000 :fm i 10% during system operation when tested in accordance with ANSI N510-1975. 4 CALVERT CLIFFS - UNIT 1 3/4 9-13 Amendment No. E3,142
REFUElfNG OPERATIONS SURVEILLANCE RE001REMENTS (Continued)
- c. After every 720 hours of charcoal adsorber operation by: l Verifying within 31 days after removal that a laboratory analysis of carbon sample demonstrates a removal eff.ipiency of 2 90% for radioactive methyl iodide when the samplepidee' tested-Sb" in accordance with ANSI H510 1975 (300 0, 95% R.H.). Samples are prepared by emptying a representative sample from an adsorber test tray section, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Successive samples will be removed from different test tray sections.
Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the filter train shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with Regulatory Positions C.5.a and C.S.d of Regulatory Guide Revision 2 March 1978 while operating the ventilation system at a flow rate of 32,000 cfm i 10%. 1! f59-CALVERT CLIFFS - UNIT 1 3/4 9 14 Amendment No. E3,142
REFUELING OPERATIONS ELRy[llLANCE RE0VIREMENTS (Continued)
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4 inches Water Gauge while operating the ventilation system at a flow rate of 32,000 cfm 10%.
- 2. Verifying that each exhaust fan maintains the spent fuel storage pool area at a measurable negative pressure relative to the outside atmosphere during system operation.
- e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in place in accordance with Regulatory Positions C.5.a and C.S.c of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 32,000 cfm 210%.
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Regulatory Positions C.5.a and C.5 d of Regulatory Guide 1.52 Revision 2 March 1978 while operating the ventilation system at a flow rate of 32,000 cfm : 10%.
- g. After maintenance affecting the air flow distribution by testing in-place and verifying that the air flow distribution is uniform within 20% of the average flow per unit when tested in accordance with the provisions of Section 9 of
" Industrial Ventilation" and Section 8 of ANSI N510-1975.
i CALVERT CLIFFS - UNIT 1 3/4 9-15 Amendment No J//JE/JJE,142
REFUEllNG O9ERATIONS 3/4,Ci.)) SPENT FUEL CASK HANDllNG CRANE UMITING CONDITION FOR OPERATION 3.9.13 Crane travel of the spent fuel shipping cask crane shall be restricted to prohibit a spent fuel shipping cask from travel over any area within one shipping cask length of any fuel assembly.
- l APPllCABillTY: 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. SURVElllANCE REOUIREMENTS 4.9.13 Crane interlocks and physical stops which restrict a spent. fuel shipping cask from passing over any area within one shipping cask length of any fuel assembly shall be demonstrated OPERABLE within 7 days prior
-to crane use* and at least once per 7 days thereafter during crane operation. I These conditions are modified to permit shipping cask travel to and from the cask pit in the presence of fuel within one cask length radius of the pathway provided the boric acid concentration in the spent fuel pool is greater than or equal to 1000 ppm AND the following criteria are met by all assemblies within one cask length radius of the pathway:
- 1) Initial enrichment less than or equal to 4.1 w/o U 235, 2) Burnup greater than or equal to 28,000 MWD /MTU, and 3) Greater than 440 days elapsed from the shutdown of the last operating cycle in which the assembly was present in the core. Crane interlocks and physical stops which restrict a spent fuel shipping cask from passing over any area within one-shipping cask length of any fuel assembly not satisfying-the above criteria shall be demonstrated OPERABLE within 24 hours prior to using the crane for moving a cask within one length of fuel assemblies meeting the above criteria. These modifications are applicable only to the shipment of fuel rods supporting the EPRI sponsored hot-cell work for the shipment of a reactor vessel weld material surveillance capsule.
t-CALVERT CLIFFS - UNIT 1 3/s' 9 16 Amendment No. 144
1 REFUELING OPERATIONS
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CONTAINMENT VENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.9.14 The containment vent isolation valves shall be closed.
, APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION: . With one or more containment vent isolation valves open, shut the valve (s) within one hour or suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel within the containment. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.14 The containment vent isolation valves shall be determined to be closed within 72 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or, movement of irradiated fuel within the containment, CALVERT CLIFFS - UNIT 1 3/4 9-17 Amendment No. B B
3/4.10 SPECIAL TEST EXCEPTIONS
.3[Q 60.\ 21UTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of S aecification 3.1.1.1 may be suspended for measurement of CEA worth and slutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).
APPLICABillIX: MODE 2. ACTION:
- a. With any full length CEA+ not fully inserted and with less than l the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at 2 40 gpm of 2300 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored,
- b. With all full length CEAs+ inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at 2 40 gpm of 2300 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE RE0VIREMENTS 4.10.1.1 The position of each full length CEA+ required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA+ not fully inserted shall be demonstrated capable of l full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.
+ Excluding the center CEA during Cycle 10.
CALVERT CLIFFS - UNIT 1 3/4 10 1 Amendment No. JJ//E/Jpf, 151
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il ll3.10.2 ine mecera :r temreatture .'.effi:ie .:, tSe CIA inser ict en: *e l';c.e cis:ribution limits of Spe:if t:ations 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.~.6, 3.2.2, 3.2.3, and 3.2.4 may be suspended curing the performan:e of PHYS: 5 TESTS ;roviced: i
- a. The inERMAL POW'EP, is restri::ed :: below E55 cf RATED '
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l b. The limits of 5;e:ifi:ation 3.2.1 are maintaine: an: . ce: ermined as spe:ified in Spe:ification 4.10.2.2 below.
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3.2.4 are sus ende:, eitner:
- c. Redu:e THERMAL POWER sufficiently to satisfy the reesire.
tents Of Spe:ification 3.2.1. or
- b. be in H:7 STAN:EY aithin 5 hours.
4 5URVEILL ANCE RE0212.EMENTS 4.10.2.1 The THERMAL POWER shall be cetercined at least on:e per hour curing PHYSI:5 TESTS in which the requirements of Specifications 3.1.1. 4, 3.1. 3.1, 3.1.3.5, 3.1.3.5, 3.2.2, 3.2.3 or 3.2.4 are suspenced and shall be verified
.to be witnin the test power plateau. l 4.10.2.2 The linear heat rate shall. be determined to be within the limits of So,ecifi:ation 3.2.1 by monitoring it continueusly with the Incore Dete:ter
".M:ni:oring System pues;an to the requirements cf Spe:ifications 4.2.1.3 iland 3.3.3.2 curing PHYS:05 TEST 5 above 5% ef RATID THERMAL POWER in whi:n :ne requirements of Soe:ifi:ations 3.1.1.4,3.1.3.13.1.3.5,3.1.3.6,3.2.2, .
,i3.2.3 or 3.2.4 are saspended.
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- AL7EP.T ;;:FFS - C.:7 '. 3/4 10-2 Anenemen: N: 2f, 55
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-w SPECIAL TEST EXCEPTIONS NO FLOW TESTS - LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.4.1 may be suspended during the performance of startup and PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and
- b. The reactor trip setpoints of the OPERABLE power level channels are set at < 15t of RATED THERMAL POWER.
APPLICABILITY: During startup and PHYSICS TESTS. ACTION: With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately trip the reactor. SURVEILLANCE REOUIREMENTS ,, 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during startup ani PHYSICS TESTS. 4.10.3.2 Each wide range logarithmic and power level neutron flux s monitoring channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating startup or PHYSICS TESTS. CALVERT CLIFFS - UNIT 1 3/4 10-3
SPECI AL TEST Ex :E ~.;0NS A CENTER CEA MISA T HENT p r< p
- OR OPERATION LIMITING CONDIT~~.',
3.10.4 The rec.irecrents of Specifications 3.1.3.1 and 3.1.3.6 may be suspended durin; the performance of PHYSICS TESTS to determine the isothermal temps-atu: e coefficient and power coefficient provided:
- a. Only : .e canter CEA (CEA #1) is misaligned, and
- b. The li .i s of Specification 3.2.1 are maintained and deterMne:: as specified in Specification 4.10.4.2 below.
l APPLICABILITY: DDE
. 1 and 2.
ACTION: With any of the limi .s of Specification 3.2.1 being exceeded while the requ<icements of Spec fications 3.1.3.1 and 3.1.3.6 are suspended, either: ,
- a. Reduce THEMAL POWER sufficiently to satisfy the requirements of Specific.ation 3.2.1, or
- b. Be in HOT ETANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended and shall be verified to be within the test power plateau. 4.10.4.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended. CALVERT CLIFFS - UNIT 1 3/4 10-4 Amendment No. F g i
SPECTAt TEST EXCEpT10NS
.a;jg,c,5) COOLANT CIRCULATION
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LIMITING CONDITION FOR OPERATION 3.10.5 The reactor coolant circulation requirements of Specifica ion 3.4.1 may be suspended and all reactor coolant pumps and shutdown cooling pumps may be de-energized during the time intervals required 1) for local leak rate testing of containment penetration number 41 pursuant to the requirements of Specifica-tion 4.6.1.2.d and 2) to permit maintenance on valves located in the comon I shutdown cooling suction line or on the shutdown cooling flow control valve (CV 306) provided:
- e. No operationJt are permitted which could cause dilution of .
the DactorQioolant $ystem boron concentration, and specifically, the charging pumps shall be de-energized and the charging flow paths shall be closed.
- b. The xenon reactivity is < 0.1% Ak/k and is approaching stability, and
- c. The SHUTDOWN MARGIN requirement of Specification 3.1.1.2 is verified at least once per 8 hours when no shutdown cooling or reactor coolant pumps are in operation.
APPLICABILITY : MODES 4 and 5. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving local leak rate testing of containment penetration number 41, maintenance on valves located in the connon shutdown cooling suction line, and maintenanceuen valve CV-306. SURVEILLANCE REOUIREMENTS 4.10.5.1 The charging pumps shall be verified de-energized and the charging j flow paths shall be verified closed at least once per hour. 4.10.5.2 The xenon reactivity shall be determined to be < 0.1% Ak/k and l approaching stability within 1 hour prior to suspending reactor coolant circulation. CALVERT CLIFFS - UNIT 1 3/4 10-5 Amendment No. 48,106 L,
, 3/4.11 R ADIO ACT1VE EFFLUENTS
\b L10UID EFFLUENTS CONCENTR ATION LIMITING CONDITION FOR OPER ATION .
3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. APPLICABILITY: At all times. ACTION:
- a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeutng the above limits, without delay restore the concentration to within the above limits,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILL ANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1. 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the
*, methodology and parameters in the ODCM to assure that the concentrations at the point
,, , of release are maintained within the limits of Specification 3.11.1.1.
l CALVERT CLIFFS UNIT 1 3/4 11-1 Amendment No. 105 i
0 TABLE 6.!!.1 ; RADIOACTIVE LlOUID WASTE SAMPLING AND ANALYSIS PROGR AM l
. Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity' Type (LLD)a Frequency Frequency Analysts (uC1/mi)
A. BatchWyte P P Releases Each Batch Each Batch Principal Gamma 5x10*7 Emitter:c 1 131 lx104 Mo.99, Ce.144 2x10 4 P M H.3
- 1x10-3 Each Batch Compos!te d Grou Alpha 1x10-7 Each Batch Comksite d Sr49, Sr.90 3x10 4 i B. Turbine Building M M Principal Camma Sump 5x10-7 EmittersC 1 131 1x10 4
.p. Mo.99, Ce.144
- 2x104 CALVERT CLIFFS UNIT 1 3/4112 Amendment No,105
, TABLE 4.!!.1 (Continued) , TABLE NOTATION a The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5%' probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radlochemical separation LLD= .66sb l E* Y
- 2.22 x 10 6 . y .
exp (-lat) Where LLD is the "a prior!" lower limit of detection as defined above, as microcuries 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 efficiency, as counu per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when appilcable, 11s the radioactive decay constant for the particular radionuclide, and a t is the elapsed time between sample collection, or end of the sample collection 4, period, and time of counting.
Typical values of E V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori(a.f ter the fact) limit for a particular measurement. b i Prior to sampling Reactor Coolant Waste and Miscellaneous Waste for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling. I CALVERT CLIFFS UNIT ! 3/4 11-3 Amendment No. 105 l 1
TABl.E 4.11-1 (Continuee) i TABLE NOTATION C The principal gamma emitters for which the LLD specification app!!es exclusively are the following radlonuc!! des: Mn-54, Fe-59, Co-58, Co-60, In-65, Cs-134. Cs-137 and Ce-141. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above~nuclides, shall also be analyzed and reported in the Semlannuaj Radioactive Effluent Release Report pursuant to Specification 6.9.1.8. d A composite sample is one in which the quantity of !! quid sampled is proportional to the quantity of !! quid waste discharged in which the method of sampling employed results in a specimen that is representative of the liquids released.
=+'
CALVERT CLIFFS UNIT 1 3/411-4 Amendment No.105 l
RADIOACTIVE EFFLUENTS DOSE
!.1MITING CONDITION FOR OPER AT10N 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBl.lC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be lir
- td:
- a. During any calendar quarter to less than or equal to 3.0 mrems to the total body and to less than or equal to 10 mrems to any organ, and
- b. During any calendar year to less than or equal to 6 mrems to the total body and to less than or equal to 20 mrems to any organ.
APPLICABILITY: At all times. ACTION: '
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceding any of the above !!mits, prepare and submit to the Commission wittin 30 days, pursuant to Specification 6.9.2, a Special Report that identilles t5e cause(s) for exceeding the limit (s) and defines the corrective actions, that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Speelfications 3.0.3 and 3.0.4 are not applicable.
{URVEILLANCE REQUIREMENTS
-e- "
4.11.1.2 Monthly cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and paramsters in the ODCM at least once per 60 days. CALVERT CLIFFS UNIT 1 3/4 11-5 Amendment No,105
RADIOACTIVE EFFLUENTS LIOUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the calculated doses due to the liquid effluent to UNRESTRICTED AREAS exceeds 0.36 mrem to the total body or 1.20 mrem to any organ in a 92 day period. l. APPLICABILITY: At all times. ACTION:
- a. With radioactih liquid 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 that includes ' the
. following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the Inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE 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. l 4* ,, SURVEILL ANCE REQUIREMENTS 4.11.1.3 Monthly doses due to liquid releases to UNRESTRICTED AREAS shall be l calculated at least once per 60 days in accordance with the methodology and parameters in the ODCM. I CALVERT CLIFF 5 UNIT 1 3/4 11-6 Amendment No.105 s
R ADIOACTIVE EFFLUENTS
, 4 GA5EOUS EFFLUENTS E
l
'p.O' DOSE R ATE -
LIMITING CONDITION FOR OPER ATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from I the site to areas at and beyond the SITE BOUNDARY (see Figure 3.1-1) shall be limited to the foDowing !- a. For noble gases: Less than or equal to 300 mrems/yr to the total body and l less than or equal to 3000 mrems/yr to the skin, and 1
- b. For lodine 131 and for all radionuclides in particulate form with half lives L greater than 8 days: Less than or equal to 1500 mrems/yr to any organ. )
APPLICABILITY: At all times. ACTION:
- a. With the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s), l
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS 4.11.2.1.1 - The dose rate due to noble gases in gaseous effluents shall be determined to ,
be within the above limits in accordance with the methodology and parameters in the ' W- CDCM. 4.11.2.1.2 The dose tate due to lodine-131 and all radionuclides in particulate form with l half lives greater than 8 days in gaseous effluents shall be determined to be within the l' above limits in accordance with the methodology and parameters in the ODCM by l obtaining representative samples and performing analyses in accordance ~' with the L samp!!ng and analysis program specified in Table 4.112. 7 p CALVERT CLIFFS UNIT 1 3/4 11-7 - Amendment No.105 I I l
-t TABLE 4.11-2 4
RADIOACTIVE GASEOUS WASTE SAMPLING AND ANAt,YSIS PROGRAM Minimum Loa.- Limit of l{ Analysis Type of Detection (LLD)a E Sampling (uCi/ml) Frequency Frequency Activity Analysis a Gaseous Release Type P P b 4
? Principal Gamma Emitters 1x10 E A. Vaste Gas Storage I'E Each Tank Each Tank (Gaseous Emissions Only)
Tank Grab Sample g P P b d 1x10-4 Each Batch C Each PatchC Principal Gamma Emitters g B. Containment Purge (Gaseous Emissions Only) and Vent Grab Sample Mc Principal Gamma Emitters b ix104 C. Main Vent M - Grab SampleC (Gaseous Emissions Orly) 1x10 4 d M H-3 Continuous T d Ix10-12 Continuous W 1-131 Charcoal Sample
- g b d W Principal Gamma Emitters lx10-II Continuous Particulate (I-131, Others)
Sample
- d M Gross Alpha 1x10-II Cont h Composite Particulate Sample d Sr-89, Sr-90 1x10-II Continuous Q
~ Composite = -
Particulate Sample 2r W d Noble Cases Ix10 4 Continuous Noble Gas fL Monitor Gross Beta or Gamma [ Jxfo-7
* ** Each Batch I Each atch I Principal' Gamma Emitters r .
l TABLE 4.11-2 (Cen91nued) TABLE NOTATION a The LLD is defined, for purposes of these speelfications, as the smallest concentration of radioactive material in a sample that will yleid a net count, above system background, that will be detected with 95% probability with only 3%, probability of falsely concluding that a blank observation represents a "real" signal For a particular measurement system, which may include radlochemical separation: LLD= 4.66s3 E' Y
- 2.22 x 10 6 . y .
exp ( Aat) Where: LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume,
- sb is the standard deviation of the background counting rate or of the counting rate ol a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, Y !s the sample size in units of mass or volume, 6
2.22 x 10 is the number of disintegrations per minute per microcurie, Y ls the fractional radiochemical yleid, when applicable, 1 is the radioactive decay constant for the particular radionuclide, and At for plant effluents is the elapsed time between sample collection or end of the
** , sample collection period, and time of counting.
Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori(after the fact) limit for a particular measurement. CALVERT CLIFFS UNIT ! 3/4 11-9 Amendment No.105
TABLE 4.!!.2 (Con?lnued) TABLE NOTATION b The principal gamma emitters for which the LLD specification applies exclusively are the fouowing racionuclides: Kr-87, Kr-SS, Xe-133, Xe-133m, Xe-133, and Xe-138 for gueous emissions and Mn-54, Fe-59, Co-58, Co-60,2n-63, Mo-99, Cs-13,4, Cs-137, Ce-41 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Seminannual Radioactive Effluent Releue Report pursuant to Specification 6.9.1.8. C Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT l-131 concentration in the primary coolant has not increased more than a factor of 5, and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3. 4 The ratio of the sample flow rate to the sampled strer.m flow rate shall be known for the time period covered by eac:h dose or dose rate calculation made in accordance with Specifleations 3.11.2.1, 3.11.2.2 and 3.11.2.3.
- Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or sJter removal from sampler. When sample collection time is less than seven days, :he corresponding LLDs may be increased by a l proportional- factor. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT l-131 concentration in the primary coolant has not increased more than a factor of 5, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
I Waste Gas Decay Tank samples shall be collected and analyzed for exygen once per week on the in-service tank and following tank isolation. Waste Gas Surge Tank samples shall be collected and analyzed dally during power escalation from MODE 6
- p. through MODE 3, and once per week at all other times.
8 Collect sample *and analyze daily for total Curie content per Specification 3.11.2.6
- when the Reactor Coolant System specific activity of Xe-1331s greater.than 150 uC1/ml.
h incinerated o!! may be discharged via points other than the main vent (e.g., Auxillary Boiler). Releases shall be accounted for based on pre-release grab sample data. , t i Samples for incinerated all releases shall be collected from and representative of I filtered oil in liquid form. ! CALVERT CLIFFS UNIT 1 3/4 11-10 Amendment No.105
R ADIOACTIVE EFFLUENTS DOSE - NOBLE G ASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the followings
- a. During any calendar quarter: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation and,
- b. During any calendar years Less than or equal to 20 mrads for gamma radiation and less than or equal to 40 mrads for beta radiation.
APPLICABILITY: At all times.
. ACTION:
. 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 Specla! Report that identifles the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILi.ANCE REQUIREMENTS
- q. 4.11.2.2 Monthly cumulative dose contributions for the current calendar quarter and current calendar year for noble gases sha!] be determined in accordance with the methodology and parameters in the ODCM at least once per 60 days.
CALVERT CLIFFS UNIT .! 3/4 11-11 Amendment No.105 l l
MDIOACTIVE EFFLUENTS l DOSE -IODINE-131 AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPER AT10N 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from lodine-131 and a0 radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following
- a. During any calendar quarter: Less than or equal to 15 mrems to any organ and,
- b. During any calendar year: Less than or equal to 30 mrems to any organ,
- c. Less than 0.!He of the limits of 3.ll.2.3(a) and (h) as a result of burning contaminated oil.
. APPLICABILITY: At all times, m ACTION:
- a. With the calculated dose from the release of lodine-131 and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents I exceedmg any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identlfles the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compilance with the above limits.
- b. The provisions of Speelfications 3.0.3 and 3.0.4 are not applicable.
SURVE'iLLANCE REQUIREMENTS 4.11.2.3 Monthly cumulative dose contributions for the current calendar quarter and current calencar year for lodine-131 and radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 60 days. CALVERT CLIFFS UNIT 1 3/4 11-12 Amendment No.105 i l
R ADIOACTIVE EFFLUENTS CASEOUS RADT ASTE TRE ATMENT SYSTEM LIMITING CONDITION FOR OpER AT10N
~
3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shau be used to reduce radioactive materials in gaseous waste prior to their dlacharge when the gaseous effluent air doses due to gaseous effluent releases, to aress at and beyond the SITE BOUNDARY (see Figure 3.1 1) exceeds 1.20 mrad for gamma radiation and 2.4 mrad for beta radiation in a 92 day period. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the calculated doses due to gaseous effluent releases, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) exceeds 1.8 mrem to any organ in a 92 day period. 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 Speelfication 6.9.2, a Special Report that includes the following informattom ,
- 1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and 4 3. Summary descriptbn of action (s) taken to prevent a recurrence.
- b. The provisens M Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4 Monthly doses due to gueous releases shall be calculated at least once per 60 days in accordance with the methodology and parameters in the ODCM. CALVERT CLIFFS UNIT 1 3/4 11-13 Amendment No. 105
R ADIO ACTfVE EFFLUENTS l EXPLO5tVE CAS MIXTURE (Hydrogen rich systems not designed to withstand a hycrogen explosion) LIMITING CONDITION FOR ODERATION . 3.11.2.3 The concentration of crygen in the waste gas holdup system shall be limited to less than or equal to 4% by volume. APPLICABILTTY: At all times ACTION:
- a. With the concentration of oxygen in a waste gas decay t.tnk greater than 4%
by volume immediately suspend all additions of waste gases to that tank and reduce the concentration of oxygen to less than or equal to 4% by volume without delay.
~
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILL ANCE REQUIREMENTS .
}
4.11.2.5 The concentrations of oxygen in the waste gas holdup system shall be determined to be within the above ilmits by the sampling program described in Table 4.11-2. Y . CALVERT CLIFFS UNIT 1 3/4 11-14 Amendment No- 105
R ADIOACTiVE EFFLUENTS - _ GAS STORAGE TANKS LIMITING CONDITION FOR OPER ATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 58,500 curies noble gases (considered as Xe 133). . APPLICABILITY: At all times. ACT10Nt
- a. With the quantity of radioactive materialin any ga storage tank exceeding the above limit, immediately suspend all additions of radioactive material to that tank and within 48 hours reduce the tank contenu to within the limit,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not app!! cable.
SURVE!LL ANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in the in-service gas storage tank shall be determined to be within the above limit at least once per 24 hours when the Reactor Coolant System specific activity of Xe-133 is greater than 150 uC1/ml. 4-CALVERT CLIFFS UNIT 1 3/4 11 15 Amendment No.105
R ADIO ACTIVE EFFLUENTS 3)W,7 SOLID R ADIOACTIVE TA$TE , I l LIMITING CONDITION FOR OPER ATION,'___ 3.!!.3 The solid radwaste program shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet ahlpping and burial ground requirements. APPLICABILITY: At a.11 times. ACTION:
- e. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.
- b. The provision of Specifications 3.0J and 3.0.6 are not applicable.
$URVE!LL ANCE REQUIREMENTS 4.!!.3 THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive warte (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions). I y.
.af CALVEkT CLIFFS UNIT 1 3/4 11-16 Amendment No.105
\
l
R ADIO ACTIVE EFFLUENT 5 l l \b TOTAL DOSE 1.lMITING CONDITION FOR OPER ATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 23 mrems to the total body or any orga.n, except the thyrold, which shall be limited to less than or equal to 73 mrems. APPLICABILITYt At all times. ACTION:
- a. With the calculated tlhses from the release of radioactive matcritis in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2.a, 3.1.l.2.b 3.11.2.2.a, 3.ll.2.2.b, 3.ll.2.3.a or 3.ll.2.3.b, calt.ulatlans shall be made including direct radiation contributions from the reactor units end outside storage tanks to determine whether the above limits of Specification
,, 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrance of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Specia.1 Report, as defined in 10 CFR Part 20.40$c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, inclWing; all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive materlaj involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in
** ,, ,. accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted unt!! staff action on the request is complete.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
_SURVE!LLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liculd and gaseous effluents shall be determined in accordance with Specifications 4.im.l.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM. 4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and outside storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Specification 3.!!.4.a. I f CALVERT CLIFFS UNIT 1 3/411-17 Amendment No.105
3/4.12 R ADIOLOGIC AL ZNVIRONMENTAL MONITORING 3/4. G.) TYONToTDBbWtbCNG LIMITING CONDITION FOR OPER ATION 3.12.1 The radiolo ical envire. mental monitoring program shall be conducted as spec'. fled in Table 3.1g.1. APPLICABitfTY: At all times. ACTION:
- a. With the radlological environmental monitoring program not being conducted as specified in Table 3.121, prepare and submit to the Commlaslon, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.7, a description of the reuons for not conducting the program as required and the plans for preventing a recurrence.
6 '
- b. With the level of radioactivity as the result of plant effluents in an environmental sample at a specified location exceeding the reporting levels of Table 3.12 2, prepare and submit to the Commission within 30 days after p %
receiving Report thatthe sample identifies analysis,(s)ursuant the cause for exceeding to Specification the limit (s) and6.9.2, definesa 5pecial the l corrective actions to be taken to reduce radioactive affluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC la less than the calendar year limits of Specifications 3.11.1.2, 3.!!.2.2, and 3.11.2.3. When l more than one of the radionuclides in Table 3.12 2 are detected in the sample this report shall be submitted If ,
concentration (1) + con' centration (2) + . .t 1.0 reporting level (1) reportmg level (2) When radionuclides other than those in Table 3.12 2 are detected and are the 4' result of plant effluents, this report shall be submitted if the potential annual dose' to a MEMBER OF THE PUBLIC la equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. 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 shal4 be reported and described in the Annual Radiological Environmental Operating Report. CALVERT CLIFFS UNIT 1 3/4121 Amendment No. JCC.105
Nd R ADIOLOGIC AL ENv!RONMENTAL MON 1TORING - ACTION: (Continued)
- c. With fresh leafy vegetable samples unava11able from one or more of the sample' locations required by Table 3.121, identify kcations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.9.1.7, Identify the cause of the
, l unavailability of samples and identify the new location (s) for obtaining the replacement samples in the next Annual Radiological Environmental Operating Report.
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.121 and the detection capab111tles required by table 4.121. I
'The MEMBE.R methodolobPUBLIC OF TH shall be Indicated in this report.and parameters used to es
=+'.
CALVERT CLIFFS UNIT 1 3/4 12-2 Amendment No. JgD, 105 l
. TABLE 3.12-1 <
d p RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM G O Exposure Patir.vay fiumber of Representative Sampling and Type and Frequency
" Samples and Sample Locations, aml/or Sample Collection Frequency of Analysis P
b h
- 1. DIRECT RADIATION 23 routine monitoring stations At least Quarterly Gamme. dose at least (DRI-DR23) either with two or quarterly.
more dosimeters or with one' E instrument for measuring and q recording dose rate continuously. placed as follows: an inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDART (DRI-DR9); e" an outer ring of stations, one in C* each meteorological sector in the
, 6- to 8-km range from the site (DR10-DR10);
the remaining stations (DR19-DR23) to be placed in special interest areas such as population centers,
$" nearby residences, schools, and in 3 1 area to serve as a coritrol station.
9 3, 2. AIRDORNE a ! # Radiolodine and Samples from 5 locations (Al-AS): Continuous sampler Radiofodine Cannister (ca , [' art icula tes 3 samples (Al-A3) from close operation with sample collection weekly, or I-131 analysis weekly. to the 3 SITE DOUNDART more frequently if Particulate Sampler: locations, in dif ferent required by dust Gross Feta radioactiv-sectors of the highest loading. Ity analysis following
.-___ b
es*
~
j 8 TA!!!E 3.12-1 (Continued)_ r- pyl0t0GICA M NVIRIMMtNTAl_ N0filTORING P, q AM
.1 d Imposure Pathway flundier of Representat Ive Sampilng and Type and frequency p and/or, sample Samples and Sanple Iocations, Collection Frequency, of Analysis k'! calculated annual average fliter change;C ground-level II/Q. Gag Isotopic anal-ysis of romposite Q 1 sample (A4) from the vicinity (by location) i; of a cinnunity having the high- quarterly.
_. est caltulated annual average . ground-Icvel D/Q. I sample (A5) from a control location, es for eraspic 15-30
&m distant and in the least
- prevalent wind direction.
-~
m 3. WATERB0l?NE
- a. Surface 1 sainple at intane area (Wal) Composite sample Gaasna isntopic anal-I sample at dist.harge area over 1-mnnth period" ysisd anonthly. Compos-(Wa2) ite for tritiumeanal-ysis quarterly.
,. h. Sediment fron I sample freen downstream area Semlannually isotopic anal-2 shoreline with existing or potential ysis seislannually.
R recreational value (Wb1).
.. 4. INGESTION z-P a., fish and
~
3 saepics of consacrcially and/or Sample in season, or GaasnaIsgtopic
~~
Invertchrates recreat ionally important species semiantiually if they analysis on cdthie .S (2 fish spet.fes and i Inverte- are not seasonal. portions. brate species) in vicinf ty of I'lant dist harge area (141-I43). 3 samples of same species in . . areas not infleev-nced ley plant discharge (la4-la6).
~
z.
\
TABLE 3.12-1 (Continued) RADIOLOGICAL [NVIRONFENTAL INITORING PH0 GRAM i 3 Exposure Pathway Number of Representative Sampilng and Type and Frequency n and/or Sample Samples and Saarle Lociations, Collection frequency of Analysis
- b. Food Products Samples of 3 different kinds of Monthly during growing Gamme isotopicd and
** broad leaf vegetation grown near season. 1-131 analysis.
* .the site boundary at 2 different t
- g locatloas of highest predicted annual average groundlevel D/Q
_, (Ibl-Ib6). I sample of each of the siellar Monthly during growing Ganna tsotopic ami broad leaf vegetation grown 15- season. 1-131 analysis. 30 km distant in the least '
- prevalent wind direction (Ib7-Ib9). r R. i U *
- F l
! R i l
.N . . ;
o
- o i
I
. t l
i ! *- i L g
i 4-TABLE 3.12-1 (Continued p TABLE P80 TAT 90ff g 1
- i 7 ' The code in perenthesis, e.g. DRI, AI, defines generic semple locations in this specification that can be used to l ,
^ y identify the specific locations in the niep(s) and table in the ODCM. Specific parameters of distance and direction i g sector from the central point bi w the two containment bundIngs and additional descriptien where pertinent, is ;
; provided for each semple location in Table 3.82-1, and in a se*de and figure (s) be the ODCM. Refer to PMMtE
- i. Treparation of Radiological Etfluent Technical Specific 4tlens for Peuclear Power Plants, October 1978", and to L S* Radiological resysired n- Assessment Branch Tectenical Position, Revisten I, IL._.ic.1979. Devlettens are permitted from the j "., schedsle N specimens are unobtaineble due to circumstances such as ioszardens conditions, seasonal unevaHaldHty, and neeMunction of automatic sampling n_'c_..t.
q_ ', _ _ . n. If spechnens are eseobtainalde due to sampling )' period. - AN deviations frene the sa"., schedule shen be elocemented in the Assount R; ' Operating Report pursuant to 5pecification 6.9.8.7 ,kel Envis __ .. ..tal g t i practicable to centbeme to obtain n_,he of the essedia of chatte at the woost desired In location these er l tinee j instances sositalde alternative secedle and locations nomy be dessen for the particular pathway in epsestlese an
- substitutinns made within 30 days be the radiological ;...
- ._. 2..tal monitoring prograne. Fac _ : to Specificatieve j
- w ebtainingidentify the cause of the unevaHeldHty of samples for that pathway and identify the new location (s) for 6.9.I.7, t be the nest Annual Radiological Enviresemental Opating Report and also inchsde be the repegrt a l b' revised f s) and saide for the ODCM reflecting the new location (sp.
l "b 7 m i be essed be piece of, or be addition to, integrating - demi-- .;mOne or sesore instrumesets, For the purposes of this table, a tieermehrn:-m..; desineeter (TLD) is considered to be one 7 _,"m two er more phosphors be a pocket are considered as two er more
- dosimeters. FHen badges shen not be essed as ti._;o, for measuring direct radiation. The freggwesecy of anal readout for TLD systems w111 depend upon the characteristics of the specific system used and should be selected to obtain optineune dose inf=matten with noenineet fading. Due to the geographical Hmitations,9 sectors are monitored ;
{a aroum* the Cafver t CHffs Pdesclear Power Plant. j t i gC aHow for raden and theren ' J.;;. decay. If gross beta activity in air particulat a i 1, if the yearly neeen of control _,k., gamma isotopic =d,,*. shen be performed en the Individual semph; . y L S: ) i I i l m
i e j TAltfE 3.12-1 (Continued)_ M TAlliE NOTAT10f! (ContIntetQ - M,,
~
,,' G:ssna isotopic analysis means the identification and quantification of gaussa-emitting radionuclides that U,,
snay be attributable to the effluents from the facility. S
- A composite sample is one in which tie quantity (aliquot) of liquid sampled is propottional to the e quantity of flowing liquid and in which the mettwxt of sampling employed results in a specimen that is g representative of the liquid flow. In this program, cowosite sample aliquots shall be collected at 3 time intervals that are very short (e.g.. hourly) relative to the compositing period (e.g., sonthly) in order to assure obtaining a representative sample. ,
Em 1, . l l g -
'l 3
1 i; - - l es .
, L
- TAfti E 3.12-2 i.
9tt]'Ott TING 1 EVELS FOR ttfl*] ACTIVITY C(MN ftt ATIONS IN ENVlttONMENTAL SAa4Pt.E5 Mc -
,'.e ItEl'OstTING LEVELS
?:. ;
e, .
~ _
! ', water- Aneborne Partio te Fish a "...s tebrates Milk i
,l' Aeulysis hsCi/I) or Gases (pCi/ne - Food Prtmences (pCi/kg, wes) (pCi/8 (pCi/kg, wet) i 4
Il- 1 '20,000* j (:- ! '~ {' .1 Mn-14 ' .1,000 10,000 ! " 2 Fe-11 400 i 10,000 ! t Co-18 I,000 - 30,000 ; l s. Co 60 300 < i i a: 80,000
- I L; Zn j 300 20,000
. i Zr-Nie-9) 400 f
i I-I 11 2 0.3 f
) 100 !
Cs-I M 30 80 t 1,000 60 1,000 : l' k CS-t17 30 20 2,000 70 7,000 i a j g fla-l a-Ito 200 ;
- .i 300 i j
j . . - - . - - - r l p, 'For enaydriniring fee
- essett. water
.- sasiipics. This is a 40 CFM Part 141 vakse. If em drinking water pathway esists, a value of 30,000 pCill
, t t l f i
, t
- ~ '
e 4* TABLE 4.17-1 n I DETECTION CAPAntLITIES FOR ENVittONMENTAL SAMPLE ANALYSISa,b h I-si 7 LOWER LIMIT OF DETECTION (LLDF
~
Fish & Invertelicates Milk Food Products sediment Water (pCi/kg,vity) (pCi/l) Airborne or GasesParticulj)te (pCi/m (pCl/kg, wet) (pCl/I) (pCifkg, wet) Analysis .fI " Gross Beta 4 0.01 It.) 2000* Mn-34 13 130 Fe-39 10 260 $ Co-33,60 13 130 Zn-63 30 260 o Zr-Nb-93 13 I-131 to 0.07 I 60 130 f3 60 130 Cs-13% I3 0.03 850 t8 30 t30 Cs-t 37 I8 0.06 13 Ba-La-140 83 ?I E - -
*If no drinking water pathway exists, a value et 3000 pC1/1 may be used.
h il
- f .
+
8
j T ABl.E 4.17 1 (Continued) 1 TABLE NOTATION
- This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuc11 des, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
b Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13. C The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that w111 yleid a net count, above system background, that will be detected with 93% probability with only 3% probability of falsely concluding that a blank observation represents a "real" signal. For a particular meuurement system, which may include radiochemical separation: 4.66sb E* V' 2.22
- Y ' exp ( kt t)
Where: I LLD is the "A prior!" lower limit of detection u defined above, as picoeuries per unit mass or volume, sb is the standard deviation of the background counting rate or of the counting rate of a blank sample u appropriate, u counts per minute, W- E is the counting efficiency, u counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picoeurie, Y is the fractional radlochemical yleid, when app!! cable, 1 is the radioactive decay constant for the particular radionuclide, and a t for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typleal values of E, V, Y and At should be used in the calculation. CALVERT CLIFFS UNIT 1 3/4 12-10 Amendment No. 40.105 )
, T ABLE k.121 (Continued TABLE NOTATION lt should be recogic'ed thtt the LLD is defined u an a priori (before the fact) limit representing the capamiisy of a meuvrement system and not u an a pesteriori(af ter the fact) !!mit for a particular measurement. Analyses shall be performeo in such a manner that the stated LLDs will be achieved under routine conditions. Occulonally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. d LLD for drinking water samples, if no drinking water pathway exists, the LLD of gamma isotopic analysis may be used, s
=9*
CALVERT CLIFF 5 UNIT 1 3/4 12-11 Amendment No. 40. 105
i d _3AAl,0pGlCAL ENY!RONMENTAL HON!TORM Y
$-YLAND USE CENSUS L,lMITING CONDITION FOR OPER ATION .
3.12J A land use census shall be ctoducted and shall identify within a distance of 8 km (3 miles) the location in each of the 9 meteorological sectors of ty neareq milk animal, the nearest residence and the nearest garden
- of greater than 50 m (300 f t') producing broad leaf vegetation. (For elevated releases as defined in Regulatory Guide 1.!!!, Revmslon 1, July 1977, the land use census shall also identify within a distance of 3 km (3 miles) the locations in each2 cf the 9 mete rological sectors of au milk aninals and all gardens of greater than 30 m producing broad leaf vegetation).
_APpt1CABILITY At all times. ACTION:
- a. With a land use census identffying a location (s) that yjelds a calculated dose or
- dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location (s) in the next Annual RadiologicaJ Environmental Operating Report, pursuant to Specification 6.9.1.7
- b. With a land use census identifying a location (s) that yleids a calculated dose or dose commitment (vla the sarne exposure pathway) 20 percent greater than at a locathn from which samples are currently being obtained in accordance with }
Specification monitoring program 3.12.1, within add the The 30 days. newsampling location location (s) tos),the radiolog(ical excluding the envlonmen control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from thlt monitoring program af ter October 31 of the year in which this land use census was conducted. Pursuant to Specification 6.9.1.7, identify the naw location (s) in the next Annual Radlological Environmental Operating Report and also include in l a+ - the report a revised figure (s) and table for the ODCM reflecting the new location (s),
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVE!LLANCE REOUTREMENTS 4.12.2 The land census shall be conducted during the growing season at least once per 12 months using that inforrr.ation that will provide the best results, such as by a door todoor survey, serial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annuti Radiological Environmental Operating Report pursuant to Specification 6.9.1.7
- Broad leaf vegeta_ tion sampling of at least three different kinds of vegetation may be performed at theQte boundarylin each of two different direction sectors with the highest predicted D/Qs in lieu f the garden census. Specifications for broad leaf vegetation sampling in Table 3.121. b shall be followed, including analysis of control samples, CALVERT CLIFFS - UNIT 1 hC. 3/4 12 12 Amendment No. HD,105
h7 P,AD10 LOG 2 CAL [NVIRONwENTAL "0N?TORfNG h INTERL ABOR ATORY COMP ARISON PROGR AM hg. LIMITING CONDITION FOR OPERAT!ON 3.12.3. Analyses shad be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.121. APPLICA5fLfTW At all times. ACTION:
- a. With analyses not being ptrformed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Speelfication
. 6.9.1.7. l ., b. The provisions of 5peelfications 3.0.3 and 3.0.4 are not appilcable.
SURVE!LLANCE REOUTREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required interlaboratory l Comparison Program shad be included in the Annual Radiological Environmental Operating Report pursuant to 5pecification 6.9.1.7. 4'. e CALVERT CLIFFS UNIT 1 3/41213 Amendment No. 106.105 1 l
- - - - - - - - - , , - - - - - - - , - - _ - , , - - , , - , - , - - - , _ ._, _ , --a BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION
, AND SURVEILLANCE REQUIREMENTS I
a .-,. ,, , , .~. , -, + - - , . . -
i k 1 l l l l I I NOTE j The summary statements contained in this section i provide the bases for the specifications of . Sections 3.0 and 4.0 and are not considered a part of these technical specifications as provided in 10 CFR 50.36.
3/4.0 APPLIC AM '.lT Y
, FA5!!
The spe:ifications cf this se: tion provide the general requirementt
,! Re:virements applicable towithin eachSe: ef the tion Limiting 3/4 Cenditions for Operation and Surveillen:e l1
!! 3.0.1 This spe:ification defires the a::iicability of ea:t spe:ificatior in ter-s of cefine: OPERATIONAL MDDES cr other spe:ified conditions and is provided to celinente specifically when ea:h spe:ifi:ation is a:Plicable.
l' 3.0.2 Tnis see:ification defines these conditions necessary te corstitute
;ie:-eiten:e with the terms cf an incividual Limiting Conditioe for 0:eretien and
'jesso:iate: Ati10h rc:virement.
k 3.0.3 Inis 5:e:ifi:ation delineates the ACTION to be taken for circum-i stances net cire:tiy Creviced fer in tne ACTION statements and whose occurrence
! wouic vioitte the intent of the spe:ificatier For example, Settification 1:~.5.1 re: vires ea:h Reacter Coolent Syste . safety inje: tion tenk to be OPERABLE
en: crevices extlici; A;i10h re:ercetr.ts if one safety inje: tion tank is
;i. eerable. Under tne terms of See:ification 3.0.3 i' more then one safety
';1rje:tice tenk i s ito:. era bl e . A*.Il0'i ait nin one heue is re:uire: to ir.itiete
;l unit snut: cat at: tt in et leest RT STAC!h witnin the next f hours end in
;!tt lets H ; T S K U I D 0 n"i w t*>in the fcilorin; 6 hours.
i As a further exempie, l '$;e:i fic6 tion 3. 6. 2.1 re:vi re s tw:, C tteinmea,; Spray Systems 10 be OEERL.BLE An: previces expli:it ACTION re:virements if one sorey system is inoperable: I Under the terms cf Spe:ification 3.0.3 if both of the requireo Containment 5 r' e .s Systems are inoperable, a:tien within one hour is recuired to initiate filunit snutcown and be in at least HDI STACSY within the next 5 hours, in at llleast HDT SHUIDOWN witnin the following 6 hours and ir, et least COLD SHUTDOWN llin the nen 2 hcurs, it is assure: that the unit is brought to the required timcoe within the re;uired times ty promotly initiating and carryin; out the
!ap;ro; itte ACT10h STATEMENT.
to
- 1. 3.0.4 This see:ification provides tnet entry into er. OPERATIONAL MDDE or l; :omplement ether spe:ified a;:li:tbility concition must De rede with (a) the fullsystems, ecuipment or ef recuire:
llotherparametersassee:ifie: in the Limiting Conditions fee 0;eration bein; limet witn:ut regare for allowable deviations anc out of service provisions
'!centained in the ACTION statemeras.
!! 3 The intent of this provision is to insure that facility operction is
,lnetinitiatedwitheitne-reaviredequipmentorsystemsinoperableerother lispecifiec limits being exceeded.
Exceptions te snis provision have been proviced for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications. CALVERT ;LIFFS - UNIT 1 E 3/4 0-1 Anenement ho, 52 444, VEL?L1Ffi- m J ,sm e t.2i
AFpL f ?MllfiY
, BHiC -
l 3.0.5 This specification delintates what additional conditiens must be satisfied te permit operation to continue, consisteht with the ACTION state-ments for power sources, when a normal or emergency power source is not OPERABLE. It specifica',1y prohibits operation when one division is inoperable because its normal ei ebergency power source is inoperable and a system. sub-system, train, component or device in another division is inoperable for another retson, i The provisions of this specification permit the ACTION statements essccinted with individual systems, subsysters, trains components, or devices to be censistent with the ACT;0N statements cf the associated electrical power source. !! allows operation to be governec by the time limits of the ACTION statement associated with the Limiting Condition for 0;'eration for the normal or emergency power source, not the individual ACTION' statements foe each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its normal or emergency power source. For example, Specification 3.5.1.1 requires in part that two emergency diesel generate-s be 08ERAELE. The ACT!0N-statement provides for a 72 hour out-of service time wnen one emergency diesel generator is not OPERABLE. If the definitten of OPERABLE were applied without consideration of Specifica. 4 l tien 3.0.5, all systems , subsystems, trains, coroonents and devices supplied by the inoperable emergency power source would also be inoperable. This would cittate invoking the applicable ACTION statement for each of the applicable I Limiting Conditions for Ooeration. However, the provisions of Specification
, 3.0.5 permit the time limits for continued operation to be consistent with the
, ACTION statement for the inoperable emergency diesel generator instead.
l oroviced the other specified conditions are satisfied. In this case, this
' wcule mean that the corresponding normal powe* source must be OPERABLE, and all reduncant systems, subsystems, trains, components, and devices must be OPERABLE,
, er otnerwise satisfy Specification 3.0.5 (i.e., be capable of performing their desipt function and have at least one normal or one emergency power source ~
OFERASLE). If they are not satisfied, action is require in accordance with this $Decification. As a further example, Specification 3.6.1.1 requires in part that two physically independent circuits between the offsite transmission network and the ensite Class !! distribution system be OPERABLE. The ACTION statement provices a 24 hour out.cf-service time when both required offsite circuits are not OPERABLE. If the cefinition 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 inoperable. This would dictate invoking the applicable ACTION statements for each of the applicable LCOs. However, the provisions of Specifi-cation 3.0.5-permit the time limits for continued operation to be consistent with the ACTION statement for the inoperable normal power sources instead, provided the other specified conditions are satisfied. In case, this would
\
i CALVERT CLIFF 5 4 UNIT 1 B 3/4 0-2 Amendment No. 52 1.;9 7 m c0 ,a-- + - m -n
APPLICABillTY BASES mean that for one division the 6mergency power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other division must be OPERABLE or likewise satisfy Specification 3.0.5 (i.e., be capable of performinq their design functions and have an emergencypowersourceOPERABLl). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, trains, components and devices in both divisions must also be OPERABLE. If these conditions are not satisfied, action is required in accordance with this specification. In MODES 5 or 6 Specification 3.0.5 is not applicable, and thus the individual 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 conditions for which the Limiting Conditions for Operation are applicable. 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. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 This specification establishes the limit for which the specified time interval for Surveillance Requirements may be extended, it permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance, e.g., transient conditions or other ongoing surveillance or maintenance activities, it also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling uutage and are specified with a 24 month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages (e.g., surveillances with a FREQUENCY of R (18 months)). The limitation of Specification 4.0.2 is based on engineering judgment and the recognition that the most probable result of any particular surveillance being perfomed is the verification of conformance with the Survelliance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval. 1 CALVERT CLIFFS - UNIT 1 B 3/4 0 3 .enendment No. JUUJ 150 l d
APPLICAB3]Jf BASES 4.0.3 This specification establishes the failure to perform a surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Under the provisions of this specification, systems and components are assumed to be OPERABLE when Surveillance Requirements have been satisfactorily performed within the specified time interval . However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are found or known to be inoperable although still meeting the Surveillance Requirements. This specification also clarifies that the ACTION requirements are applicable when Surveillance Requirements have not been completed within tae allowed surveillance interval and that the time limits of the ACTION requirements apply from the point in time it is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance Requirement within the allowable time limits of the ACTION requirement restores compliance with the requirements of Specification 4.0.3. However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, was a violation of the OPERABILITY requirements of a Limiting Condition for Operation that is subject to enforcement action. Further, the failure to perform a surveillance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requirement and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(1)(B) because it is a condition prohibited by the plant's Technical Specifications. ' If the allowable time limits of the ACTION requirements are less than 24 hours or a shutdown is required to comply with ACTION requirements, e.g., Specification 3.0.3, a 24 hour allowance is provided to permit a delay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have t ' not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown is required to comply with ACTION requirements or before other remedial measures would be required that may preclude completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, and the time required to perform the surveillance. This time limit provision also provides for the completion of Surveillance Requirements that become applicable as a consequence of M00E changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to-the requirements of Specification 4.0.4 is allowed. If a surveillance is not completed
, within the 24 hour allowance, the time limits of the ACTION requirements t .
CALVERT CLIFFS - UNIT 1 B 3/4 0 4 Amendment No M/J/J 150 l 4
~ , , ,. -
----,.--r - - ---,r-m--, , + - - ,- - -
APPLICABILITY BASES are appitcable at that time. When a surveillance is perfomed within the 24 hour allowance and the Surveillance Requirement $t not met, the time - limits of the ACTION requirements are applicable at the time that the surveillance is terminated. Surveillance Requirements do not have to be performed on inoperable equipment because the ACTI requirements define the remedial measures that apply. However, the rveillance Requirements have to be met to 4 , demonstrate that inoperable equipment has been restored to OPERABLE status. ' 4.0.4 This specification establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL MODE or other condition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or sarameter limits are met before entry into a MODE or condition for which t1ese systems and components ensure safe operation of the facility. This provision applies to changes in OPERATIONAL MODES or other specified conditions associated with plant shutdown as well as startup. Under the provisions of this specification, the applicable surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during ini,tial plant startup or following a plant outage. When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation. 4.0.5 This specification ensures that 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 will be performed in accordance with a periodically updated version of Section XI of the ASHE Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55 a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section IX of the ASME Boiler and Pressure Vessel Code and applicable 1 Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical specifications and to
-remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
CALVERT CLIFFS UNIT 1 B 3/4 0 5 Amendment No. U /J/J 450
APPLICABILITY BASES l ' l Under the terms of this specification the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Yessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERA 110KAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested u) to cne week after return te normal operation. And for example, the Tecinical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Yessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. CALVERT CLIFFS UNIT 1 B 3/4 0 6 Amendment No. 52/J/J, 150
3/a.1 REACT 1V7TY C04Tp0t SYSTEMS
$&SES 3 /a .1.1 BORAT10W CCWTR0t 3 /4.1.1.1 a
- d 3 /a .1.1. P SHUTIKNW ERGIN A sufficient SWTDOW MRGIN ensures that 1) the reactor can be made suberitical from all oper;iting conditions, 2) the reactivity transients Associated with postulated accident conditions are controllable within accept-able limits, and 3) the reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition.
The most limiting SHUTDOW MRGIN requirement at beginning of cycle is determined t'y the reauirements of several transients, including Boron Dilution and Steam Line Rupture. The SHUTDOW MRGIN requirements for these transients are relatively small and nearly the same. Howe,er, the most limiting SHUTDOW MRGIN requirement at end of cycle comes from just one transient, the Steam Line Rupture event. The requirement for this transient at end of cycle is significantly larger than that for any other event at that time in cycle and, also, considerably larger than the most limiting requirement at beginning of cycle. The variation in the most limiting requirement with time in cycle has been incorporated into Technical Specification 3.3.1.1, in the form of a specified SHUTDOW MRGIN value which varies linearly from beginning to end of cycle. This variation in specified SETTDOW MRGIN is conservative relative I to the actual variation in the most limiting requirement. Consequently, i adherence to Technic 31 Specification 3.J.1.1 provides assurance that the available SHUTDOW MRGIN at anytime in cycle will exceed the most limiting SHUTDOW MRGIN ttquirement at that time in cycle. In MDDE 5 the reactivity transients resulting from any event are minimal and do not vary significantly during the cycle. Therefore, the specified SHUTDOW MRGIN in MODE 5 via Technical Specification 3.1.1.2 has been set equal to a constant value which is determined by the requirement of the most limiting event at any time during the cycle, i.e., Boron Dilution with the pressurizer level less than 90 inches and the sources of non borated water restricted. Consequently, adherence to Technical Specification 3.1.3.2
. provides assurance that the available SHUTDOW MAGIN will exceed the most limiting SHUTDOW MRGIN requirement at any time in cycle.
lt l
- CALVERT CLIFFS - UNIT 1 B 3/4 1-1 Amendment No. 2!. 32, de,
! 11, l e , IH , 130
3 /t .1 FEACTIVITY CORTROL SYSTEMS l l Pa5ES (h4AA0dTinH ffATR01 3 /a ,1,1. ? ' E0R04 DILUTION 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 System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,601 cubic feet in approximately 24 minutes. The reactivity change rate associated with boren concentration reductions will therefore be within the carability of operator recognition and control. 3 / A .1.1. 4 NODERaTOR TEMPERATURE C0'TF101ENT (MTti The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the HTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured HTC value is within its
. limit provides .a.ssurances that the coefficient .will be saintained within i acceptable values throughout each fuel cycle. I
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CALVERT CLIFFS - UNIT 1 B 3/41 la Amendment No no
REACT!VliY CONTROL SYSTEMS BASES j 3/4.1.1.5 MIN' MUM TEMPERtTUR FOR CRITICALITY i This specificatien ensures that the reactor will not te made critical with the Reactor Coolant System everage temperature less than 5150F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its nomal operating range, 3) the pressurizer is capable of being in an 0FEPASLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RTND7 temperature. 3 / 4.1. 2 BOUTION SYSTEMS The boron injection system ensures that negative reactivity control is available durin; each mode of facility operation. The system also provices coolant flow following an 51A5 (e.g., during a small Break LOCA) to supplement flow from the Safety injection System. The Small Break LOCA analyses assume flow from a single charging pump, accounting for measurement uncertainties and flow mal-distribution effects in calculating a conservative value of charging flow actually delivered to the RCS. The components required to function include 1) berated uter sources, 2) charging pumps, 3)perfom separate this flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) en emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 2000 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional caoability 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 sufficient to provide a SH'JT-DOWN MARBIN from til operating conditions of 3.0% ak/k after xenon decay and cooldown to 2000F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6500 gallons of 7.25% boric acid solution from the beric acid tanks or 55,627 gallons of 2300 ppm borated water from the refueling water tank. However, to be consistent with the ECCS requirements, the RVT is required to have a minimum contained volume of 400,000 gallons during MODES 1, 2, 3 and 4. The maximum boron concentration of the refueling water tank shall be limited to 2700 ppm and the maximum boron concentration of the boric acid storage tanks shall be limited to 8% to preclude the possibility of boron precipitation in tie core during long tern ECCS cooling. With the RC3 temperature below 2000F, 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 ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.
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CALVERT CLIFTS - UNIT 1 B 3/41-2 Amendment No. 27, 4, EE,104
t!!: 1Yl'Y CD': t'. Sv! !V! BASES The boron capability recuired below 200*F isabased ueen oroviding a 3'a Ak/k SHMDOW'i MARSIN af ter xenon decay and cooldown from 200'F to 140'F. Thit. condition recuires either 737 gallons of 7.25% boric acid solution from the beric acid tanks or 9,Bia gallons of 2300 ppm borated water from the refuelin; water tank. The OPERA!ILITY ef one beret inje: tion system during REFUIL1N3 ensures that tnis system is available for reactivity control while in MOO! C. 3 14.1 .? 90Va?tT CD'.*ROL a!? M t1!! Tne specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARSIN is maintained, and (3) the potential effects of a CEA ejection accident' are limited to acceptable levels. The ACT30N statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are cat. The ACTION statements applicable to a stuck or untrippable .CEA and to a large misalignment (1 15 inches) of two or more CEAs, require a prompt shutdown of the teetter since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUT-DOWN MARGIN. For small misalignments (< 15 inches) of the CEAs, there is 1) e small degradation in the peakiiig factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions rela-tive to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARSIN, and 4) a smell effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within its aligtrnent require-ments prior to initiating a reduction in THERMAL POWER. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) I take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution. CALVERT CLIFFS - UNIT 1 B 3/4 1-3 Amendmentho.IA,48yg I 127
~
RE ACTit'ITY CONTRO1 SYSTTMS PASES Overpower margin is presided to protect the core in the event of a tarke misalignment ( t 15 inches) of a CE A. However, this misalignment would cause distortion of the core poser distribution. The reactor protective system would not detect the degradation in radial peaking factors and 'since variations in other system parameters (e.g., pressure a.nd coolant temperature) may not be suffi: lent to cause trips, it is possible that the reactor could be operating with l process variables lets conservative than those usumed its generating LCO and LSSS serpoints. The ACTION statement associated with a large CEA misalignment recaires prompt a:sion to realign the CIA to avoid eacessist margin degradation. If the CEA is not realigned within the given time constraints, a: tion is spe:ified whi:h will preserve margin, in:lu, ling redu:tions in THIRMAL POWER. Fer a single CIA misajisament, the time allowan:e m ie. align the CIA (Figure 3.1-3) is permitted fer the following reasons: J. The margin cal:uiations whi:b support the , power distribution LCOs for DNBR are based on a steady. state F/ as specified in Te:hnica] 5peelfi:ation 30.3.
- 2. When the actual F7 is less than the Te:hni::a3 Spe:ification value, additional margin ex.ists.
- 3. This additional margin can .be credited to offset the increase in F7 with time that mill occur following a CEA mittlignme.nt due to Arnon redistribution.
The requirtment to nduce power level after the time l'r nit of Figure 3.1-3 is reached offseu the continuing increase in F7 that can o: cur due to tenen redistribution. A power reduction is not required below $0% power. Below $0% power there is sufficient conservatism in the DNB power distribution LCOs to completely offset sny, or any additional, senen redistribution effects. The ACTION statemenu applicable to misaligned or inoperable CEAa include requirements to align the OPERAllLE CEA: in a given group with the inoperable .// CEA. Conformance with these alignment isquirtmenn bring $the core, within a shors period of time, to a configuration consistent whh that assumed in generating LCO and LSSS setpoints. However, eatended operation with CIAs significantly inserted in the core may lead to perturbations in 1) local burnup,
- 2) peaking factors, and 3) available abotdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and ISSS setpoinu detarmination. Therefore, time ismin have heem im-d on operation with inoperable CEAs to pre:Inde auch adverse condi'l= from developing.
CA1VIRT CLIFFS - UhTT J .B 3/4 1-4 Amendment No. H, 8/7,127
...2 l nr AcTivrry1:ovTp01 syFTrMs B Asrs Operability of the CEA position indicators b required to determine CEA positions and thereby ensure comphan:e with the CEA alignment and insertier.
limits and ensures proper operation of the rod blo:k cir:uit. The CE A *F ul' In" and
- Full Ou t' limits provide an additional independent i meant for determining the CEA positions when the CIAs art at either their fully inserted or fully withdrawn positions. .
Therefore the OPERABILITY and the ACTION statements appli:able to inoperable CEA position indicators permit continued operations when positions of CEAs with inoperable indi:ators can be verified b.s the
- Full In* or *Fu11 Out* limits.
CEA positions and OPERABILITY of the CEA position indi:ators are re:;uittd to be verified on a nominal basis of once per 10 hours with more frequent verifiettions Te::uired if an automati: monitoring channel is inoperable. These verifi:ation frequen:ies are adequate for assuring that the appli:able LCOs are satished. The surveillan:e reovirements affe: ting CIAs with inoperable position indi:ation channels allow 10 minutes for testing ea:h affected CEA. This time limit was selected so that 1) the time would be long enough for the Tequired testing. and 2) if all position indi:ation were lost during testing, the 1.ime ( enough to allow a power reduction to 70% of maxirnum allowable thermal power yithin ene hour from when the testing was initiated. The time y mui casuTertEA misalignments o: urring during CEA testing are corre:ted d/
-withm the time requirements required by existing specifications.
}
The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses. Heasurement with T > 515'F and with all reactor coolant pumps operating ensures that the NEsiired drop times will be representative of insertion times experienced during a reactor trip at operating conditions. The L555 setpoints and the power distribution LCOs were generated based upon a core burnup which~would be achieved with the core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during M3 DES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion pemitted by the Steady State Insertion 1.imits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control. The Transient Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained. (2) the minimum SHUTDOWN MARG 1H is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long tem operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configuration. CALVERT CLIFFS - UNIT 1 B 3/4 1-5 Amendment No. 32.127
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE ThelimitationonlinearheatrateensuresthatintheevgntofaLOCA, the peak temperature of the fuel cladding will not exceed 2200 F. Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verify-ing that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs '.Ms faction by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant syrrnetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEA is maintained within the allowable limits of Figure 3.2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assump-tions are made: 1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specifica-l tion 3.2.4 are satisfied, and 3) the TOTAL PLANAR RADIAL PEAXING FACTOR does not exceed the limits of Specification 3.2.2. The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1. The setpoints i for these alarms include allowances, set in the conservative directions, for
- 1) a measurement-calculational uncertainty factor of 1.062, 2) an engineering uncertainty factor of 1.03, 3) an allowance of 1.002 for axial fuel densifica-tion and thermal expansion, and 4) a THERMAL F0WER measurement uncertainty factor of 1.02.
3/4.2.2. 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS-Fjy ANDF)ANDAZIMUTHALPOWERTILT-Tq l l The limitations on FT and T are provided to ensure that the assumptions used in the analysis for eltablis0ing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. The limitations on FT and To are provided to ensure that the assumptions used in the analysis establishing the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valigdurngoperationatthevariousallowableCEAgroupinsertionlimits. If Fx,F or Tq exceed their basic limitations, operation may continue under the addit onal restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assump-t Os_msed s in establishinJ the Linear Heat Rate,~ Thermal Margin / Low Pressure and Ltica(lMPower AZI)fuTHA 0WER T LT Deinity
> 010 - Hi LC0i not,exp L3 W set oints r ' in W11d cted,and f it,s ob d oced n
<i-A ]
f subsequent operation wouTd et benest cted to opthose perition 61rehide ify (thecaufjpofthi fxpecte t)lt 7 j dJAALL % 3-{rt D L CALVERT CLIFFS - UNIT 1 B 3/4 2 1 Amendment No. 32, 39, 104
p0MER DISTRIBUTION L1 HITS BASES and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt. The value of T g that must be used in the equation F}y Fxy (1 + Tq) and Ff Fr (1 + Tq ) is the measured tilt. The surveillance requirements for verifying that F}y, Ff and T q are within their limits provide assurance that the actual values of FT,pJ yy and Tq do not exceed the assuaed values. Verifying FT,FT xy 7 after each fuel loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded. 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNB Specified Acceptable Fuel Design Limit (SAFDL) of 1.15 throughout each analyzed transient. In addition to the DNB criterion, there are two other criteria which set the specification in Figure 3.2 4. The second criterion is to ensure that the existing core power distribution at full power is less severe than the power distribution factored into the small break LOCA analysis. This results in a limitation on the allowed negative AX1AL SHAPE INDEX value at full power. The third criterion is to maintain limitations on peak linear heat rate at low power levels resulting from Anticipated Operaticaal Occurrences (A00s). Figure 3.2-4 is used to assure the LHR - criteria for this condition because the lineer heat rate.LCO, for both ex-core and in core monitoring, is set to maintain only the LOCA kw/ft requirements which are limiting at high power levels. At reduced power levels, the kw/ft requirements of certain A00s (e.g., CEA withdrawal), tend to become more limiting than that for LOCA. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored 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. Revised by NRC Letter dated Mav 23. 1000 . CALVERT CLIFFS UNIT 1 B 3/4 2-2 Amendment No. 79//E/EE/77/JE// J9/J9/Ei \
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/
l 3/4.3 -INSTRUMERAT10N
'f %
BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPEPABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combi-nation therof exceeds its setpoint, 2) the spectfied coincidence logic is maintained 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters. The OPERABILITY of these systems is required t'o provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The 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 overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability. The measurement of response time at the specified frequencies pro-vides assurance that the protective and ESF action 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 applicable. Response time may be demonstrated by any series of sequential, over-lapping 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. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that
- 1) the radiation levels are continually measured in the areas served CALVERT CLIFFS - UNIT 1 B 3/4 3-1
IhSTRUMENTATION l BASES I by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3/4.3.3.2 INCORE DETECTORS . The OPERABILITY of the incore detectors with the specified minimum complement of equiptent ensures that the measurgments obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. 3/4.3.3.3. SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the regnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the meas.ured response to that used in the design basis for,the facility and is consistent with the recommendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that
" sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is requit2c 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 Programs," February 1972, as supplemented by Supplement 1 to NUREG-0737, 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to pennit shutdown and maintenance of HOT STANDBY of the facility from lo' c ations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. CALVERT CLIFFS - UNIT 1 B 3/4 3 2 Amendment No. 102,JOS, ddd. 11 l 1
INSTRUMENTATION l BASES 3/4.3.3,6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables 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 Lessons Learned Task Force Status Report and Short-Term Recommendations." The subcooled Margin Monitor (SMM), the Heated Junction Thermocouple (HJTC), and the Core Exit Thermocouples (CET) comprise the Inadequate Core Cooling (ICC) instrumentation required by item II,F.2 NUREG 0737, the Post TMI-2 Action Plan. The function of the ICC instrumentation is to enhance the ability of the plant operator to diagnose the approach to, and recovery from, ICC. Additionally, they aid in tracking reactor coolant inventory. These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37. These instruments are not required by the accident analysis, nor to bring the plant to HOT STANDBY or COLD SHUTDOWN. In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during an extended COLD SHUTDOWN. This is bec'ause the sensors are accessible only after the plant has been cooled down and drained, and the missile shield has been moved. If only one channel is inoperable, it should be restored to OPERABLE status in accordance with the schedule outlined in a Special Report. If both channels are inoperable, the system shall be restored to OPERABLE status in the next refueling outage. 3/4,3.3.7 FIRE DETECTION INSTRUMENTATION 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 their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation'is restored to operability. l 11 CALVERT CLIFFS - UNIT I B 3/4 3-3 Amendment No. 25#3/99#E9,147
INSToVMENTATION l BASES 1/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor : and control, as applicable, the releases of radioactive materials in ' gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCH to ensure that the alarm / trip will occur prior to exceeding the limits of Specification 3.ll.2.1.a based on average annual X/Q. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50. 3/A.3.3.10 RADIOACTIVE L10VID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated and b adjusted in accordang.e vitt the methodology and parameters in the ODCM to MT ensure that the alaryn/thp4 ll occur prior to excecding the limits of 10 CFR Part 20. The OPERA 51LITY and use of this instrumentation is consistent with the requirements of General Design Criteria 6.0, 63, and 64 of Appendix A to 10 CFR 50. l ei CALVERT CLIFFS - UNIT 1 B 3/4 3-4 Amendment No.105
3 /4. 4 PE20TOD COOLANT SYSTEM BASES 3'A.4,1 COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBA above 1.195 during all normal operations and anticipated transients. A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure considerations require plant shutdown if component repairs and/or corrective actions cannot be made within the allowable out of service time. In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient 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 l specification requires two shutdown cooling loops to be OPERABLE. The operation of one Reactor Coolant Pump or one shutdown cooling inp provides adequate flow to ensure mixing, prevent 5 stratification and [ l produce 5 gradual reactivity changes during boron concentration reductions ! in the Reactor Coolant System. The reactivity change rate associated l with boron reductions will, therefore, be within the capability of operator recognition and control. The restrictions on starting a Reactor Coolant Pump during MODES 3, l 4 and 5 with ti.e RCS temperature 5 3270 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 Part 50 (see Bases 3/4.4.9). For operation of the reactor coolant pumps the following criteria apply: (1) restricting the water volume in the l pressurizer (170 inches) and thereby providing a volume for the primary coolant to expand into and (2) by restricting starting of the RCPs to l when the indicated secondary water temperature of each steam generator is less than or equal to 30 F0 above the Reactor Coolant System temperature, (3)_ limit the initial 1.ndicated pressure of the pressurizer to less than or equal to 290 psia. 3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of'2750 psia. Each safety valve l 1s designed to relieve approximately 3 x 105 lbs per hour of saturated
' steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient'to CALVERT CLIFFS - UNIT I B 3/4 4-1 Amendment No. J//JJ/JJ/E2/J/J,146 1
ee tI:: y ::: :'. 5f5 !" j;:!: 9,
,9
i-it the Rea: tor Coolant Syste c essure te w t '- its Safety Litit c' 5:-
e psia foilexing a complete icss of tartine generat:- load while operating a. KAT:.L THEP?AL POWER and assuming no reactor trip until the first Reactor Protecti f System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is' taken for a direct reactor trip on the loss of turbi. e) and also assuming no operation of tne pressurizer power operated relief valve or steam dump valves. Demonstration of the safety valves' lif t settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. _3/4.4.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capeble of being supplied from an emergency power source to ensure the ability to seal this possible RCS leak-age path. 3/4.4.4 PRESSUR12ER A steam bubble in the pressuri:er with the level as programmed ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The operating band for pressurizer level bounds the progranted level and ensures that RCS pressure remains within the bounds of an analyzed condition during the excessive charging event as well as
'during the limiting depressurization event, Excess Loed. The operating band also protects the pressurizer pode safety valves and power operated relief valve against water relief. The power operated relief valves function to relieve RCS pressure during all design transients. Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer--Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
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 off-site power condition to maintain natural circulation at HOT STANDBY. 3/4.4.5 STEA!4 GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will
' W caintained. The program 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 t
. l CALVERT CLIFFS - UNIT I B 3/4 4-2 Amendment No. H , 53,4/),82
REACTOR COOLANT SYSTEM 4 BASES maintain surveillance of the conditions of the tubes in the event that there ' is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. In-service inspection of steam generator tubing also provides a means of i characterizing the nature and cause of any tube degradation so that correc- l tive measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these 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 generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 gallon per minute, total). Cracks having a primary-to-secondary leakage CALVERT CLIFFS - UNIT 1 B 3/4 4-2a Amendment No. 53 g _ _j_ g
I REACTORCOOLANTSYS,TE BASES 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 plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfec-tions exceeding the plugging limit of 40*. of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspec-tion fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specifi'cationsr6.9'.'Eprior the resumption of plant operation. Such cases will be considered by the Comission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS 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", May 1973.
3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE
~
Industry experience has shown that wh le a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GpM. This
- threshold value is sufficiently low to ensure early detection of additional leakage. '.
CALVERT CLIFFS-UNIT 1 B 3/4 4-3
REACTOR COOLANT SYSTEM 1 BASES The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 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 the analysis of these accidents. PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. 3/4.4.7 CHEMISTRY
\
The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the poten-tial for Reactor Coolant System leakage or failure due to stress corro-sion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion orotection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be con-tinued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time inter-vals without having a significant effect on the structural integrity of the Reactor Coolant System. . The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits. The surveillance requirements provide adequate assurance that con-centrations in excess of the limits will be detetted in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY 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 fracticn of Part 100 limits following a CALVERT CLIFFS-UNIT 1 B 3/4 4 4
PFACTOR COOLANT SYSTEM i BASES \ 4 steam generator tube rupture accident in conjunction with an assumed steady state primary to secondary steam generator leakage rate of 1.0 gpm and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent interim limits based upon a p3rametric evaluation by the NRC of typical site locations. These values are conservative in that specific site carameters of the Calvert Cliffs site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity >1.0 uti/ gram DOSE EQUIVALENT l-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 uti/ gram DOSE EQUIVALENT l-131 but within the limits shown on Figure 3.41 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 < 5000 F prevents the release of activity should a steamgenerator,Uberupturesincethesaturationpressureoftheprimary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary cociant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Operation within the appropriate heatup and cooldown curves assures the integrity of the reactor vessel against fracture induced by combinative thermal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and ever more restrictive Pressure / Temperature limits must be observed. The current limits, Figures 3.4-2a and 3.4-2b, are for up to and including 12 Effective Full Power Years (EFPY) of operation. The shift in the material fracture toughness, as represented by N is calculated using Regulatory Guide 1.99, Revision 1. For 12 RT EFPY,nT, a t the 1/4 T position, the adjusted reference temperature (ART) CALVERT CLIFFS - UNIT 1 B 3/4 4 5 Amendment No. 145
REACTOR COOLANT SYSTEM BASES J value is 222 0F. At tne 3/4 T position the ART value is 162.5 0F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code, Section 111, Appendix G to calculate heatup and cooldown limits in accordance with the requirements of 10 CFR Part 50, Appendix G. To develop composite pressure temperature limits for the heatup
/ transient, the isothermal,1/4 T heatup, and 3/4 T heatup pressure-
[ temperature limits are compared for a given thermal rate. Then the most restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event. To develop a composite pressure temperature limit for the cooldown event. 3* the isothermal pressure temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive j allowable pressure temperature limit is chosen resulting in a composite kj limit curve for the reactor vessel beltline. J , 3 -( Both 10 CFR Part 50 Appendix G and ASME, Code Appendix C require the development of pressure temperature limits which are applicable to y;}. D inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the
\
7 I test pressure (1.1 times normal operating pressure) and locating the d corresponding temperature. This curve is shown for 12 EFPY on Figures 3.4-2a and 3.4 2b. i l Similarly,10 CFR Part 50 specifies that core critical limits te established based on material considerations. This limit is shown on the heatup curve, Figure 3.4 2a. Note that this limit does not consider the
, core reactivity safety analyses that actually control the temperature at j which the core can be brought critical.
I The Lowest Service Temperature is the minimum allowable temperature at I pressures above 207 of the pre operational system hydrostatic test i pressure (625 psia). This. temperature is defined as equal to the most limiting RT for the balance of the Reactor Coolant System components plus 1000F,NhTer Article NB 2332 of Section III of the ASME Boiler and Pressure Vessel Code. The horizontal line between the minimum boltup temperature and the Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 207. of the pre operational hydrostatic test pressure. The change in the line at 150 0F on the cooldown curve is due to a cessation of RCP flow induced pressure deviation, since no RCPs are permitted to operate during a cooldown below 1500F. d. CALVERT CLIFFS - UNIT 1 B 3/4 4 6 Amendment No.145
ret.C709 C00tBNi SYSTEM BASES The minimum boltup temperature is the minimum allowable temperature at pressures below 20% of the pre operational system hydrostatic test pressure. The minimum is defined as the initial RT NhT for the material of the higher stressed region of the reactor vessel Elus any effects for irradiation per Article G 2222 of Section 111 of the ASME Boiler and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was determined using the certified material test reports and Branch Technical Position MTEB 5 2. The maximum head flange initial RT@F.TThe is 10 associated with the stressed region of the closure minimum boltup temperature instrument uncertainty is 10,Ftemperature
+ 100 F = 00F.including However, for conservatism, a minimum boltup temperature of 70 0 F is utilized.
The design basis events in the low temperature region assuming a water solid system are: A RCP start with hot steam generators; and. An inadvertent HPSI actuation with concurrent charging. Any measures which will prevent or mitigate the design basis events are sufficient for any less severe incidents. Therefore, this section will
/ discuss the results of the RCP start and mass addition transient
/ analyses. Also discussed is the effectiveness of a pressurizer steam bubble and a single PORV relative to mitigating the design basis events.
S 3 The RCP start transient is a severe LTOP challenge for a water solid RCS. Therefore, during water solid operations all 4 RCPs are tagged out of
} service. Analysis indicates the transient is adequately controlled by placing restrictions on three parameters: initial pressurizer pressure n and level, and the secondary-to-primary temperature difference. With ? these restrictions in place and when decay heat level is low (reactor has g -
been shutdown 8 hours or longer), the transient is- adequately controlled without the assistance of the PORVs. Operating procedures require that J during normal cooldowns, entry into MPT enable (327 0 F and below) will not g . occur until 8 hours af ter reactor shutdown. This restriction is nol Q---- intended to delay cooldown in situations where plant or personnel safety considerations make expeditious cooldown prudent. If RCPs are restored in response to e loss of decay heat removal when decay heat loads are high and cperator actions were either not taken or ineffective, a single PORV will protect the Appendix G limits. The inadvertent actuation of one HPSI pump in conjunction with one charging pump is the most severe mass addition overpressurization event. Analyses were performed for a single HPSI pump and one charging pump assuming one PORV available with the existing orifice area of 1.29 in2 . - For the limiting case, only a single PORV is considered available due to single failure criteria. A figure was developed which shows the calculated RCS pressures versus time that will occur assuming HPSI and
'\ charging pump mass inputs, and the expansion of the RCS following loss of decay heat removal. Sufficient overpressure protection results when the equilibrium pressure does not exceed the limiting Appendix G curve pressure. Because the equilibrium pressure CALVERT CLIFFS - UNIT 1 B 3/4 4-7 Amendment No. Jf) 146 l
1 PEACTOR C00 tant Sv5 TEM PASES exceeds the minimum Appendix G limit for full HPSI flow, HPSI flow is throttled to no more than 210 gpm indicated when the HPSI pump is used l for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS i expansion following loss of decay heat removal. The HPSI flow is i injected through only one HPSI 1000 MOV to limit instrumentation uncertainty, No more than one charging pump (44 gpm) is allowed to I operate during the HPSI mass addition. Comparison of the PORV discharge curve with the critical pressurizer pressureof464.1psiaindicatesthatadgquateprotectionisprovidedby l a single PORV for RCS temperatures of 70 F or above when all mass input is limited to 380 gpm. HPSI discharge is limited to 210 gpm to allow for-one charging pump and system expansion due to loss of decay heat removal. To provide single failure protection against a HPS! pump mass addition l l transient, the MPSI loop MOV handswitches must be placed in l pull to override so tne valves do not automatically actuate upon receipt i of a SIAS signal. Alternative actions, described in the ACTION a STATEMENT, are to disable the affected MOV (by racking out its motor circuit breaker or equivalent), or to isolate the affected HPSI header. Examples of HPSI header isolation actions include; (1) de energizing and ! L (g tagging shut the HPSI header isolation valves; (2) locking shut and tag ing all three HPSI pump discharge MOVs; and (3) disabling all three HPS pumps. i l Three 100% capacity HPSI pumps are installed at Calvert Cliffs. Procedures will require that two of the three HPSI pumps be disabled 0 L (breakers racked out) at RCS temperatures less than or equal to 327 0F and that the remaining HPSI pump handswitch be placed in pull-to lock. l- Additionally, the HPSI pump normally in pull-to lock shall be throttled to less than or equal to 210 gpm when used to add mass to the RCS. n Exceptions are provided for ECCS testing and for response to LOCAs. M A pressurizer steam volume and a single PORV will provide satisfactory control of all mass addition transients with the exception of a spurious l .D * ' actuation of full flow from a HPSI pump. Overpressurization due to this transient will be precluded for temperatures 3270F and less by disabling two HPSI pumps, placing the third in pull-to-lock, and by throttling the third pump to less than or equal to 210 gpm flow when it _is used to add mass to the RCS. I N p Note that only the design bases events are discussed in detail since the l . less severe transients are bounded by the RCP start and inadvertent HPSI actuation analysis. , RCS temperature, as used in the applicability statement, is determined as g follows: (1) with the RCPs running, the RCS cold leg temperature is the appropriate indication (2 with the shutdown cooling system in s operation, the shutdown coo) ling temperature indication is appropria (3) if neither the RCPs or shutdown cooling is in operation, the core exit thermococples are the appropriate indicators of RCS temperature. l. CALVERT CLIFFS - UNIT I B 3/4 4 8 Amendment No. J/E 146 n- _ _ . _ _ - . _ _ - _ - - - _ . - _ . _ _ _ - _ - - - - - - - _ - _ - - - - - - _ . _ _ _ - . _ _ _ , _ _ . - - - _ _ , , _ _ _ _ _ - - - - - . _ - - - - - _ - _ . , _ _ - _ - - _ _ - - _ _ _ _ _ _ _ - . - - _ - - - - - - - _ _ - _ . . _ , . - _ - - _ , _ _ _ . - _ - _ _ - - _ _ _ _ _ _ _ _ . _ _
DELETED e 1 CALVERT CLIFFS - UNIT 1 B 3/4 4 9 Amendment No. 145
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DELETED l
'CALVERT CLIFFS - UNIT 1 B 3/4 4-10 Amendment No.145 l-
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I REACTOR COOLANT SYSTEM MSES DELETED l: t-1 .. l CALVERT CLIFFS - UNIT 1 B 3/4 4-11 Amendment No. #,145 I i-
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II I t ij .::e r t. I-l 3/4.4.'O STR'J:TUE;L ' N~ ECE:TY The inspection programs for ne ASME Code *iass i, 2 gnd 3 comp:nents ensure tha; :he stru:: ural in egri:y Of these components will oe main;aine: a en at:e;; tie levei throughou; the life Of :ne olant. To -56 ex;en; applicacie,
- ne ins:ee-ion ;-ogram for nese : moonen s is in com;1 dance w1:n Section X :f
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Plan; operation. This bypass valve is recuire: to De close: to ensure :na: the effe::s of a pipe ru;;ure downs: ream of tris valve wiii ne er:ee: the a :iden; analyses assum:tiens. l.
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t C ALVEF.T CL:Fr5 . 'JNIT 1 5 2/: :.i 2 Amend en- N: . 3-6
REACTOR COOLANT SYSTEM BASES 3/4.4.13 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncendensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one heactor boolant 3ystem vent T"- path from the reactor vessel head and the pressurizer vapor space ensures the capability exists to perform this function. Thevalveredundancyofthebeactorhoolant ystem vent paths serves to D' minimize the probability of inadvertent or i(rreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does'not prevent isolation of the vent path. Jhe function, capabilities, and testing requirements of the kaactor(coolant ( 1ystem vent systems are consistent with the requirements of Item II.B hUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. P c CALVERT CLIFFS - UNIT 1 B 3/4 4-13 Amendment No.119
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) LAl[S-3/4.5.1 SAFETY INJECTION TANKS The OPERABILITY of each of the RCS safety injection tanks ensure.%
@(_tthat V
icient volume of borated water will be immediately forced into or tore through each of the cold legs in the event the RCS g is below the pressure of the safety injection tanks. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on safety injection tank volume, boron concentration and pressure ensure that the assumptions used for safety injection tank injection in the accident analysis are met. The safety injection tank power operated isolation valves are
. considered to be " operating bypasses" in the context of IEEE Std.
279 1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In
- addition, as these safety injection tank isolation valves fail to meet single failure criteria, removal of power to' the valves is. required.
The limits for operation with a safety injection tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an - additional safety injection tank which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be iminediately opened, the full capability of one safety injecttnu tank is not available and prompt action is required to place the reactor in a n.eds where this capability is not required. ;
,f;,
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two separate ECCS subsystems ensures that I sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem throuah any sin e failure consideration. Either subsystem operating in 4 t :t Fn; wit the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from-the double ended bruak of the largest - 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. Portions of the low pressure safety injection (LPSI) system flowpath are common to both subsystems. This includes the low pressure safety injection flow control valve, CV-306, the flow orifice downstream of CV-306, and the four low pressure safety injection loop isolation valves. Although the portions of the flowpath are common, the system design is adequate .to ensure reliable ECCS operation due to the short period of LPSI system operation following a design loss of Coolant Incident prior to recirculation. The LPSI system design is consistent with the assumptions in the safety analysis. I CALVERT CLIFFS - UNIT 1 B 3/4 5-1 Amendmant No. 103 145
EMERGENCY CORE C00t1NG SYSTEMS l BASES The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basement is provided to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provides this protection by dissolving in the sump water and causing its final pH to be raised to 2 7.0. The requirement to dissolve a representative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated post LOCA temperatures. The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that as a mininium, the assumptions used in the safety analyses are met and the subsystem OPERADILITY is maintained. The
. surveillance requirement for flow balance testing provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure <irop in the piping system to each injection point is necessary to: (1) prevent total pump fiow ' rom exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between' injection points in accordance with the assumptions used in the ECCS LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS LCCA analyses. Minimum HPSI flow requirements for temperatures above 327 0F are based upon small break LOCA calculations which credit charging pump flow following an
{ SIAS. Surveillance testing includes allowances for instrumentation and system leakage uncertainties. The 470 gpm requirement for minimum HPSI flow from the three lowest flow legs includes instruinent uncertainties but not system check valve 1eakage. The OPERABILITY of the charging pumps and the associated flow paths is assured by the-Boration System
- Specification 3/4.1.2. Specification of safety injection pump total developed head ensures pump performance is consistent with safety analysis assumptions.
At temperateres of 327 0 f and less, HPSI injection flow is limited to less than or equal to 210 gpm except in response to excessive reactor coolant leakage. With excessive RCS leakage (LOCA), make up requirements could exceed 210 gpm. Overpressurization is prevented by controlling other parameters, such as RCS pressure and subcooling. This provides overpressure protection in the low temperature region. An analysis has been performed which shows this flow rate is more than adequate to meet core cooling safety analysis assumptions. HPSis are not required to auto start when the RCS is in the HPT enable condition. The Safety
. Injection Tanks provide immediate injection of borated water into the core in the event of an accident, allowing adequate time for an operator to take action to start a HPSI.
Surveillance testing of HPSI pumps is required to ensure pump operability. Some surveillance testing requires that the HPSI pumps deliver flow to the RCS. To allow this testing to be done without increasing the potential for overpressurization of the RCS, either the PNT must be isolated or the HPSI pump flow must be limited to less than or equal to 210 gpm or an RCS vent greater than 2.6 square inches must be provided. CALVERT CLIFFS - UNIT 1 B 3/4 5 2 Amendment No. J//JS//JJ7/J/5,146 l l
EMERGENCY CORE COOLING SYSTEMS BASES-3/4.5.4 REFUELING WATER TANK (RWT) The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS-
.in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure'that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain suberitical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods inserted except-for the most reactive control assembly. These assumptions are consistent-with the LOCA analyses.
The contained water volume limit includes an allowance for water not . usable because of tank discharge line location or other physical l characteristics. 1 1 l l l I I i I F l ! l l l 1 l I CALVERT CLIFFS - UNIT 1 8 3/4 5 2a Amendment No.145 ) i i
f: y 3/4.6 CONTAINMENT SYSTEMS .
; BASES -
i 3/4.6.1 PRIMRY CONTAINMENT _
]
f 3/4.6.1.1 CONTAINMENT INTEGRITY i-3 Primary CONTAINMENT INTEGRITY ensures that the release of radio-active 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 limi-
' tation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
f,- o 3/4.6.1.2 CONTAINMENT LEAXAGE i The limitations on containment leakage rates ensure that the total i i containment leakage volume will not exceed the value assumed in the As an added con-8 accident analyses at the peak accident pressure, P,r. 1 servatism, the measured overall integrated leakage ate is further
- limited to < 0.75 L or < 0 t fof p(as applicable) during perfonnance of theperiodictestsI.oaccoun.75L ossible degradation of the contain-ment leakage barriers between leakage tests.
- The surveillance testing for measuring leakage rates are consistant with the requirements of Appendix "J" of 10 CFR 50.
3/4.6.1.3 C0KTAINMENT AIR LOCKS
- The limitations on closure and leak rate for the containment air
- locks are required to meet the restrictions on CONTAINMEhT INTEGRITY
[ and contairment leak rate. Surveillance tasting 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. 1 i I f I CALVERT CLIFFS - UNIT 1 B 3/4 6-1 L . t- . - . - - . .. ..
CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE T_ The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding it design negative pressure differential with respect to the outside atmosphere of 3.0 psig and 2) the containment peak pressure does not exceed the design pressure of 50 psig during LOCA or steam line break conditions. l The maximum peak pressure expected to be obtained from a LOCA event is 47.6 psig assuming an initial containment pressure of 14.7 psia. The l limit of 1.8 psig for initial positive containment pressure will limit the total pressure to 49.4 psig which is less than the design pressure and is consistent with the accident analyses. The maximum peak pressure expected to be obtained from a steam line break event is 49.2 psig assuming an initial containment pressure of 16.5 psia (1.8 psig). 3/4.6.1.5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 2760 F during LOCA conditions. The containment temperature limit is consistent with the accident analyses. 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY The limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the event of a LOCA. The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchorages and liner and the Type A leakage tests are sufficient to demonstrate this capability. The surveillance requirements for demonstrating the containment's structural integrity are consistent with the intent of the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures", January 1976. The end anchorage concrete exterior surfaces are checked visually for indications of abnormal material behavior during tendon surveillance. Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test. Revised by NRC Letter dated May 23. 1990 . i i CALVERT CLIFFS - UNIT 1 B 3/4 6-2 Amendment No. J M ,
CONTAINMENT SYSTFE ! BASES 3/4.6.1.7 CONTAINMENT PURGE SUPPLY AND EXHAUST ISOLATION VALVES This limitation ensures that containment purge supply and exhaust valves will be maintained shut during MODES where containment pressurization may occur as the result of LOCA or steam line break conditions. The capability of these valves to close during a containment pressurization event and provide isolation of these lines has not been established. l 3/4.6.2 DEPRESSUR12AT10N AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRav SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. 3/4.6.2.2 CONTA!NMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available during post LOCA conditions. 3/4.6.3 IODINE REMOVAL SYSTEM The OPERABILITY of the containment iodine filter trains ensur n that sufficient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity 'are consistent with the assumptions used in the LOCA analyses. # 3/4.6,4 CONTAINMENT ISOLATION VALVES The OPERABILITY cf the containment isolation valves ensure that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent ~with the assumptions used in the analyses for a LOCA. Revised by HRC Letter dated
- v 21 1090 .
t i CALVERT CLIFFS - UNIT 1 B 3/4 6-3
)
e,
,*'O.e
., CONTAINMENT SYSTEMS l
BASES I 1 3/4.6.5 COMBUSTIBLE GAS CONTROL I The OPERABILITY of the equipment and systems required for the detection H and control of hydrogen gas ensures that this equipment will be available to l maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. The detection equipment has been upgraded to meet the requirements of NUREG-0737, which included a detection range of 0 to 10. percent. hydrogen. Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions,
- 2) radiolytic decomposition of water and 3) corrosion of metals within contain-ment.
L 3/4.6.6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSTEM ' The OPERABILITY of the penetration room exhaust system ensures that radio-active materials leaking from the containment atmosphere through containment penetrations following a LOCA are filtered and adsorbed prior to reaching the
- envi ronment. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the LOCA analyses.
s L 3
- CALVERT CLIFFS - UNIT 1 B 3/4 6-4 Amendment No. 6 0
3/4.7 PLANT SYSTEMS l BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1000 psig during the most severe anticipated system operational transient. The total relieving capacity for all valves on all of the steam lines is 12.18 x 106 lbs/hr at 100t RATED THERMAL POWER. The maximum relieving capacity is associated with a turbine trip from 100% RATED THEPXAL POWER coincident with an assumed loss of condenser heat. sink (i.e., no steam bypass to the condenser). The main steam line code safety valves are tested and maintained in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code. The as-left lift settirtgs will be.no less.than EB5 psig to ensure that the lift setpoints wt11 remain within specification during the cycle. . In MODE 3 two main steam safety valves are required OPERABLE per steam : generator. These valves will provide adequate relieving pacity or re. al of both decay heat and reactor coolant pump heat from the actor nolan ystemi via either of the two steam generators. This requirement is prov ed to fact 11 tate the post-overhaul setting and OPEPABILITY testing of the safety valves which can only be conducted when the RCS is at or aDove 5000F. It allows entry into MODE 3 with a minimum number of main steam safety valves OPERABLE so that the set pressure for the remaining valves can be adjusted in the plant. . This is the most accurate means for adjusting safety valve set pressures since the valves will be in thermal equilibrium with the operating environment. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the redt: tion in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are dcrived on the following bases: For two loop operation 3p , (X) - (Y)(V) x 106.5 X F'or single loop operation (two rea tor coolant pumps operating its the same loop) SP = (X) -x(Y)(U) x 46.8 where: .
=
SP reduced reactor trip setpoint in percent of RATED THERMAL F0WER
=
V maximum number of inoperable safety valves per steam line CALVERT CLIFFS - UNIT 1 B 3/4 7-1 Amendment No. $ . 117
PLANT $Y$7 EMS BASES f U = maximum nuncer of inoperable safety valves per operating steam line 106.5 = Power Level - High Trip Setpoint for two loop operation 46.8 = Power Level - High Trip 5etpoint for single loop operation with two reactor coolant pumps operating in the same loop X = Total relieving capacity of all safety valves per steam line in 1bs/ hour Y = Maximum relieving capacity of any one safety valve in 1bs/ hour 3/4.7.1.2 AUXILIARY TEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be coolid down to lest than 300'F from normal operating conditior.s in the event of a total loss of offsite power. A delivered flow of 300 gpm is sufficient to ensure that adequate feedwater flow is availabic to remove decay heat and reduce the Reactor Coolant System temperature to less than 300'F when the shutdown cooling system may be placed into operation. ' g Flow control valves were installed in the system in order to allow auto:..atic flow initiation to a value selected by the Operator. Maximum flow to the steam generators from the motor driven AFW pump powered from the diesel is 300 gpm when feeding both generators (i.e.,150 gpm per leg maximum flow). The flow control valves installed in each leg supplied from . the motor driven AFW pump shall be set at a flow setpoint not to exceed 150 gpm per leg. If the flow is only being directed to one steam generator, it is acceptable to deliver a maximum of 330 gpm because the flow error associated with the non-used loop is eliminated. These motor driven AFW pump capacity limits are imposed to prevent exceeding the emergency diesel generator load limit. If diesel generator loading is not a limiting concern, the delivered flow from the motor driven AFW punp may be increased up to a maximum of 575 gpm (motor HP limit vice diesel loading limit). These upper flow limits do not apply to the steam driven pumps. In the spectrum of events analyzed in wnich automatic initiation of auxiliary feedwater occurs, the following flow conditions are allowed with an operator action time of 10 minutes. CALVERT CLIFFS - UND 1 B 3/4 7-2 Amendment No. 57,57,M ,b .118 i 1
P'. ANT SYS? EMS -
-BASES -P- J a dF Loss of Feedwater: 0 gpm Auxiliary Feedwater Flow Feedline Break: 0 gpm Auxiliary Feedwater Flow :
Main Steam Line Break: 1300 gpm Auxiliary Feedwater Flow (This being the maximum flow through the AFW suction line, with one unit requiring flow. prior to pump cavitation due to L 1owNPSH,) At 10 mihutes af ter an Auxiliary Feedwater Actuation Signal thi operator is assumed to be available to increase or decrease auxiliary feedwater flow to that required by existing plant conditions. 4
. i, e
CALVERT CLIFFS - Uh!T I B 3/4 7-2a Amendment No. 8 g
.u.e--.. ..4 c,, . . .r U. . ..-~..,,-.-.w-... e.,5.-u......,. __.- e . . . . --_%. . .r . .M.-y. ,-..---o,. m_.-,,..m,-,,,-.,m.m_ _ y.,-#,,--., , , , , --
l PL ANT SYSTEMS l BASES 3/4.7.1.3 CONDENStiE STORAGE T ANE The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available te maintain the RCS at HOT STANDBY conditions for 6 hours with steam discharge to atmosphere with concurrent and total loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site 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 effects of a coincident 1.0 GPM primary, to secondary tube lett in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses. 3/4.7.1.5 KAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isola-tion valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. The main steam isolation valves are surveilled to close in less than 5.2 seconds to ensure that under reverse steam flow conditions, the valves will close in less than the 6.0 seconds assumed in the accident analysis. 3/4.7.1.6 SECONDARY WATER CHEMISTEY <@ % d b The sec ary water e istry p ogram i des't edtoprovidMaximum]e pr ection yo b th the ste generato and econde hals. Th mos dama ital re nts enter h systemvi)systemint ondens co6(ingwa/er ingre s. /ging e Accumulation thes impurit in the ste gene tors mby'elept! to
. loss o etallurgical i egrity and/or ev tual.compone ilure. Th s imits in Table 3. 3 are the p scrib by the NSS - upplier as ittited-presen) ion' chemistry rameters a are consi ent wit the opera,t st rece \
indystrys ndards. Bygroutine n toring of t se y rameterg, pla person 1 are able to 4pid y detect and mit e duration of ingress oTs c ically detrimental spec s and t eby ainta n steam g perator tube i egrity. 1 I CALVERT CLIFFS - UNIT 1 B 3/4 7-3 Amendment No. H E7,82,126
PL ANT SYSTEMS , BASES 3/a.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 80'F and 200 psig are based on steam generator secondary side limitations and are sufficient to prevent brittle fracture. 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during nomal and activent conditions. The redundant cooling capacity of this system, assuming a single f a with the assumptions used in the accident analyses. ,ilure, is consistent 3/4.7.4 SERVICE WATER SYSTEM The OPERABillTY of the service water system ensures that sufficient cooling capacity is available for continued operation of equipment during nomal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used I in the accident analyses. 3/4.7.5 SALT WATER SYSTEM The OPERABILITY of the salt water system ensures that sufficient cooling capacity is available for continued operation of equipment during nomal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM . The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowabic temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERA-BILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to S rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 10 of Appendix "A* 10 CFR 50. 3/4.7.7 ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM I The OPERABILITY of the ECCS pump room exhaust air filtration system ensures that radioactive materials leakina from the ECCS equipment within the pumo room followinc a LOCA are filtered prior to reaching the CALVERT CLIFFS - UNIT 1 B 3/4 7-4 Amendment No.126
PLANT SYSTEMS BASES , environment. The operation of this system and the resultant effects on offsite dosage caleviations was assumed in the accident analyses. 3/4.7.8 SNUEBERS .x All safety relateA snubb 5 are auired OPERABLE to ensure that the structural I j integrity of the Yeactor c lent . stem and all other safety related systems is f maintained during and foi owing a seismic or other event initiating dynamic loedt. Snubbers excluded from this inspection program are those installed on non safety related systems and then only if their f ailure or f ailure of the system on which l they are installed would nave no adverse ef fett on any safety related system. The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber f ailures and is determined by the number of inoperable snubbers of each type' found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed lbefore the original required time interval has elapsed (nominal time less 25%) rey not be used to lengthen the required inspection interval. Any inspect.on wh0se results require a snorter inspection interval will override the previous schedule.
;When the cause of the rejection of a snubber is clearly established and remedied for that snubder and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from j being counted as inoperable. Genericall are (1) of a specific make or model, (2)y susceptible snubbers are those which ofthesamedesign,and(3)similarly
/ located or exposed to the same environmental conditions such as temperature, radiation, and vibration. These characteristics of the snubber installation
/ shall be evaluated to determine if further functional testing of similar snubber 3
installations is warranted. Wnen a snubber is found inoperable, an engineering evaluation is performed, in
/ addition to the determination of the snubber mode of failure, in order to deter.
D pd mine if any safety related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degrada-g tion on the supported component or system, gl 4 df - To provide assurance of snubber functional reliability, a representative sample of the installed snubbers of each type
- will be functionally tested during plant shutdowns at 18 month intervals. Observed failures of these sample snubbers
&) ! -, shall require functional testing of additional units.
The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installa-tion and maintenance records (newly installed snubber, seal replaced, spring repla:ed, in high radiation area, in high temperature area , etc. . . . ) . The requirement to monitor the snubber service life is included to ensure that the
<B") anc large bore (>B") hydraulic snubbers are examples of
'5different mall bore ty (ies of snubbers.
B 3/4 7-5 Amendment No. f 4, })f ,125 CALVE:iT CLIFFS - UMT 1
i PLANT SYSTCMS BASES
/,
operating conditions. "The service life program is designed to uniquely reflect
''S"'
the conditions at Calvert Cliffs. The criteria for valuating service life shall be determined, and documented, by the licensee. 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 operation. 3/4.7.9 SEALE0 SOUPtt CONTAMINA110N The limitations on removable contemination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits' for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake va, lues. 3/4.7.10 WATERTIGwT 000R5 This specification is provided to ensure the protection of safety related equip-rent from the effects of water or steam escaping from ruptured pipes or components in adjoining roer.s. 3/4.7.11 FIRE SUPFRE5510N SYSTEMS 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 is located. The fire suppression system consists of the water system, spray and/or sprinklers, g Halon and fire hose stations. The collective capability of the fire suppres-sien systems is adequate to minimize potential damage to safety related equip-g ment 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 affected areas until the inoperable equipment is restored to service.
,} Where a continuous fire watch is required in lieu of fire protection equipment and habitability due to heat or radiation is a concern, the fire watch should be stationed in a habitable area as close as possible to the inoperable equip-
]5 ment.
In the event the fire suppression water system becomes inoperable, immediate corrective men'sures must be taken since this system provides the major fire suppression capability of the plant. The requirement for a twenty-four 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. CALVERT CLIFFS - Uti1T 1 B 3/4 7 6 Amendment tio. 2f, f 7, ' ; 1
. - . ~ . . . . - .
3 PLANT $YST[V$ BA!!$ 3/a 7.12 PEN [*:J"!ON F!t! IAEE!!E! l i The functional integrity of the penetration fire barriers ensures that fires will be confined er adequately retarcac from screading to adjacent portions of the facility. This design feature minimizes tne possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishment. The penetration fire barriers are a passive element in the facility fire protection program and are subject to periccic inspections. During periods of time when the barriers are not functional, a continuous fire watch is recuired to be maintained in the vicinity of the affected barrier until the barrier is rest 0 red to functional status.
. I c
h CALVERT CLIFFS UNIT 1 8 3/4 7 7 113 AmendmentNo.26,ppj n(
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l l 3/4.B ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of the A.C. and D.C. power sources 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 2) the mitigation and control of accident conditions within the facility. The minimum specified indepen-dent and redundant A.C. and 0.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A"
. to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation comensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least one of each of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 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 facility status. 4 CALVERT CLIFFS - UNIT 1 B 3/4 8-1 I
I J 1/A.9 REFUELING :DEU CNS 2ASES i 3/4.9.1 !CRCN *CN~ENTRAT:CN The limitatiens en minimum beren c:ncentration (2300 pom) ensure that: l
- 1) the reacter will remain suberitical during CORE ALTERATIONS, ano 2) a '
unifcrm boron cencentration is maintained for reactivity controi in the water volumes having direct access to the reactor vessel. The limitation on X u$y of:no greater certainties, than 0.95to is sufficient which includes prevent reactor a censervative criticality during allewance for refueling operatiens. 3/4.9.2 -INSTRUMENTA 10N j i- The OPEP.AB:LITY cf the. source range neutren flux monitors ensures that redundant monitoring. capability is available to detect changes in the: react hity ccnditien of the core. 3/A.9;3 DECAY TIME The minimum recuirement for reactor suberiticality prior to mevement t
-of irradiated fuel assemblies in the react:r pressure vessel ensures that sufficient time has' elapsed to alicw.the radioactive decay of the short
-lived fission products. This decay time is cens4 stent with the assumptions-used in tne accident analyses. -
y 3/4.9.4 CONTAINMENT PENETRATIONS .. t The requirements en containment penetration closure and CPERABILITY ensure that a release of radicactive-material within c:ntainment will be '
' restricted from leakage to the environment'. ~ The CPERA!!LITV and clcsure L restricti'ons are sufficient to restrict radicactive material release f-om a fuel' element ructure. based ucon tne lack of centainmer.t oressur-izatien. potential while in the REFUELING MODE. , ,
3/a.9.5 00MPUNICA~ CNS The re:uirement for communications capability ensures inac refueling , -station' personnel can te premetly informed of significant changes in the 7 L -facility status or core reactivity; condition during CORE ALTERATIONS. CALVERT'CLI??S - UN T 1- - S39 91 Amenemenc No 43 y T ; m e-- _ ; -+ _ . - --__, . .. n . e
.,-.cy,~..'... f g.
, y. . . - - _,.h., ,,., , , , .,..w,,.-,,_, , ,,,-,.m .
.w,,, .,,_,%.,_y ,,,#..v.., e . ,,w .rw..- ,-h*c,,,,-aws.d,,.w-s.-,,.
' ' "! FUEL:NG 0*! EAT:0NS \? . s BA!!5 i!3/3.9.6 REFUEL:N3 n:-1NE ODERAB:LITY i: The CDED.A3lL TY re:wifeme".ts for ne refueling a: hine ensure tnat: (1) the refueling ma:nine aill be used f:r move .ent of CEAs anc fuel assemclies, (2) the refueline tachine has sufficient load capacity to lift a CEA or fuel assembly, and (3) :the core internals and pressure vessel are prote:ted frem excessive lifting force in the event they are inadvertently engaged curing lifting 0;eratiens. 3/4.9.7 CRANE uyE; . sp!N ruEL STNAGE ?v:'.0!NG The restri: tion en movement of 1:a s in ex:ess of the no .inal wei;ht of a fuel assemely an: CEA over other fuel assemblies in the stcrage ; col ensures that in the event this lead is drooped (1) the activity release will be limited to that contained in a single fuel assemoly, anc (2) any pessible ditt:rtion of fuel in the st: rage racks will not result in a :ritical array. This assumption is consistent with the activity release assumed in the a::ident
' analyses.
3/4.9.8 CDOLANT CIROULATION
'lhe re:uirement tna* at least ene shutdown cooling loco be in operation enteres that (1) sufficient :ocling capacity is availacle *: regevede:ayheat t'id maintain the water in *he rea:::r :ressure vessel below 140 F as re uired l dering the REFUELING MCDE, and (2) suffi:ient : elant circulatien is mairtained thrtugh the reactor core t minimi:e the effects of a teron cilutien incident '
and prevent beren stratification. i Ihe re:uirtien* 13 have two s*.Y*.:Can 20011tg ICC;s OPERAELE wnen there !
!is less than 23 fee
- Cf water abCve the cre enta"es tha* a single failure of l
lI *ne 0; era'ing shu* 0wn ::CIing ICo; will Act res.it in 4 00mpitte 105 5 Of decay l l{neatremCvalca;atility. With the rea:*0r vessel hee: rem:ve: and 23 feet Of , water acove the* :re, a large heat sink is available -for core c cling, tnus i in the event of a failure of the operating snutcown ecoling loop, adequate i
.. time is Orovided :: initiate emer;ency pr:cecures to c:el the :ere.
l
' 3/J .9.9 CON :NMEN DURSE VALVE ! SOLATION SYS?!v l
t T5e OPERA!!'.*TY Of inis system ensures that tne :entainment : urge vai ves
. I will :e aut:mati:a'if isclate u:en :etection f nign radiation leve'.s witnin 'j:ne contatree t. **e OPEUB. TY of tnis sys em is re:vir e: t: restri:t tne ,
j release Of ra:ica:tive material ft:m the c:ntainment atmos;nere t: tne environment,
,iCALVEET lL*FF5 - L9 l Amen: met: No. 55 $AW~:L :"; "":' : B 3/4 9-2 4enom ^: t. _ . -- ~
1 _ . _ _ _ _ - . . . _ _ _ - - - - _ - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ___ ~
I REFUELING OPERATIONS w. BASES _ 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assyned 10t iodine gap activity released from the rupture of an irradiated fuel assembly.
, The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.12 SPENT FUEL P0OL VENTILATION SYSTEM l. l The limitations on the spent fuel pool ventilation' system ensure that all radioactive material released from an irradiated 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 analyses. 3/4.9.13 SPENT FUEL CASK RANDLING CRANE The restriction on movement of the spent fuel shipping cask within one cask len 1s drop l l and (2) any ped (gth of any fuel assembly ensures will not that in the
- 1) the stored spent fuel assemblies will not be damaged.
possible distortion of fuel in the storage racks i result in a critical array. l 3/4.9.14 CONTAIRMENT VENT ISOLATION VALVES The OPERABILITY and closure restrictions on the containment vent isolation valves are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressuriza-tion potential while in the REFUELING MODE. 1 CALVERT CLIFFS UNIT 1 8 3/4 9-3 AmendmentNo.jpp,108
E ( { 3/4.10 SPECIAL TEST EXCEPTIONS 1 BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA j worth is innediately available for reactivity control when tests are perfonned-for CEAs worth measurement. This special test exception is 3 required to permit the periodic verification of the actual versus I predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations. 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual _ CEAs to be positioned outside of their nornal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure CEA worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions. 3/4.10.3 NO FLOW TESTS 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. 3/4.10.4 CENTER CEA MISALIGNMENT This special test exception permits the center CEA to be misaligned during PHYSICS TESTS required to determine the isothermal temperature coefficient and power coefficient. 3/4.10.5 COOLANT CIRCULATION. This special test exception permits all forced circulation of l reactor coolant to be-suspended during required local leak rate testing of containment penetration number 41 (shutdown cooling) and during maintenance on the common shutdown cooling suction line or on the shut-down cooling flow control valve-(CV-306). I CALVERT CLIFFS - UNIT 1 B 3/4 10-1 l
, R ADIOACTIVE EFFLUENTS BASES 3/h.11.1 L10UID EFFLUENT 5 - 3/4.11.1.1 CONCENTRATION This specitc4 tion is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels speci!!ed in 10 CFR Part 20 Appendix B, Table !!, 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 !!.A design objectives of Appendix 1,10 CFR Part 30, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population. The required detection cap 4,bilities for radioactive materials in Liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the
- LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 Currie, L. A., " Limits for Qualitative Detection and Quantitative Determinanon .
- Application to Radlochemistry," Anal. Chem, 40. 386 93 (1968), and Hartwell, J. K., "
" Detection Limits for Radioanalytical Counung Techniques," Atlantic Richfield Hanford Company Report ARH-5A-215 (June 1973).
3/4.11.1.2 DOSE . This specification is provided to implement the requirements of Sections ILA, !!!.A and IV.A of Appendix 1,10 CFR Part 30. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section !Y.A of Appendix ! to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."
- g. The dose calculation methodology and parameters in the ODCM 1mplement the requirements in Section 111.A of Appendix 1 that conformance with the guides of Appendix
! be shoc by calculational procedures based on models and data, such that the actual exposure of a M.MBER OF THE PUBLIC through appropriate pathways le unilkely to be substantially underestimated. The equations specified in the ODCM for calculating the , doses due to the actual releue rates of radioactive meterials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluenu for the Purpose of Evaluating Compilance with 10 CFR Part 30, Appendix 1," Revision 1 October 1977, Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluenu from Accidental and Routine Reactor P.eleases for the Purpose of implementing Appendix 1," April 1977, and NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants".
CALVERT CLIFFS UNIT 1 B 3/411-1 Amendment No.105 l
l R ADIDACTIVE EFFLUENTS g BASES hzinn 3/4.!! l.) L10VfD RADS'ASTE TREATMENT SYSTEM - The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releues of radioactive materials in licyld effluents will be Wept. "as now as is reuonably achievable". This specificat on implements the requirements of 10 CFR Part X.Ma, General Design Criterion 60 of Appendix A to 10 CFR Part 30 and the design objective given in Section ILD of Appendix ! to 10 CFR Part
- 30. The speelfled limits governing the use of appropriate portions of the 11guld radwaste treatment system were specified as a suitable fraction of the dose design objectives set iorth in Section n.A of Appendix 1,10 CFR Part 30, for 11guld effluents.
3/4.11.2 CASECUS EFFLUENTS 3/4.11.2.1 DO5E R ATE
- This specification is provided to ensure that the dose at any time at and beyond the $1TE BOUNDARY from taseous effluents from all units on the site wl11 be within the annual dose limits of 10 C'R Part 20 to UNRESTRICTED AREAS. The annual dose limits are i the doses associated with the concentrations of to CFR Part 20, Appendix B, Teie n, j Column .1. These limits provide rouonable assurance that radioactive - material i discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE I PUSLlc in an UNRESTRICTED AREA, exceedin Table !! of 10 CFR Part 20 (10 CFR Part 20.106(b)g the limits specified in Appendix B,
). For MEMBERS OF THE PUBLIC who may at times be within the $!TE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compenaste for any increase in the atmospheric diffusion factor above that for the $1TE BOUNDARY.
The required detection capabilities for radioactive materials in gueous waste samples
.p. are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HA5L-300.
Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Appilcation to Radlochemistry," Arnl. Chem. 40. 386-93 (1968), and Hartwell, 3.K.,
" Detection Limits for Radloanalytica,. Counting Techniques," Atlantic Richfield Hanford Company Report ARH-5 A 213 ' June 1973).
CALVERT CLIFFS UNIT I B 3/4112 Amendment No.105 i
,- - - - - , - - - - +,.-w. ,.,-c--+--,+w.-. +.-wm+---+-r---, -
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R AD10 ACTIVE EFFLUENT 5 BASE *h -- - t _ 3/l..!!.2.2 DOSE - NOBLE G ASES This specification is provided to implement the requirements cf Ser. ions.!!.B. III.A and IV.A of Appendix 1,10 CFR Part 30. The Limiting Condition for Operation implements the guides set forth in Section 11.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix ! to usure that the releases of radioactive materialin gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable." The Survelliance Requirements implement the requirements in Section 11LA of Appendix ! that conformance with the guides of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gueous effluenu 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 Part 30, Appendix 1," Revision 1, October 1977 and
. Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efiluents in Routine Releues from Light-Water-Cooled Reactors," Revision 1, .1uly 1977, and NUREG-0133 " Preparation of Radiological Effluent Technical Speelilcations for Nuclear Power Plants".
The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the hhtorical annual average atmospheric conditions.
-f
- CALVERT CLIFFS UNIT 1 B 3/4113 Amendment No.105
I: [ R ADf0 ACTIVE EFFLUENTS 1 ; SA$E$ W . .. 3/4.11.2.3 Dose -10 DINE 131 AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the. requirements of Sections.'!LC, E!.A and
!Y.A of Appendix 1,10 CFR Part 30. The Limiting Canditions for Operation are the guides set forth in Section ILC of Appendix 1. The ACTION statements provide the required o Section !Y.perating fleslbuity and at the same time implement the guldes set forth in A of Appendix ! to assure that the releases of radioactive materials in gaseous
{. effluents to UNRESTRICTED AREA 5 will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveulance Requirements implement the requirements in Section ELA of Appendix ! that conformance with the guides of 3 Appendix 1 be shown by calculational procedures based on models and data, such that the L actual exposure of a MEMBER OF THE PUBLIC throu gh appropriate pathways is unukely i to be nabstantially mderestimated. The ODCk calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.t09,
" Calculation of Annual Doses ta Man from Routine Releases of Reactor Effluents for the
- Pu pre of Evaluating Compliance with to CFR Part 30, Appendix 1," Revision 1, October 1977, Regulatory Guide 1.111, " Methods for ~ Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"
Revision 1, July 1977, and NUREG4133 , " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants". These equations also provide for determinin conditions.g the actual doses based upon the historical annual average atmospheric The release rate specifications for ladine 131 and radionuclides in I particulate form with half lives greater ttan 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the $1TE 50VNDARY. The pathways that were examined in the development of these calculations were: 1) Individual inhalation of airborne radionuclides,2) deposition of radlonuclides onto green 4 leafy vegetation with subsequent consumption by man,3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the muk and meat by man, and 4) deposition on the ground with subsequent exposure of man.- , .. 3/4.!!J 4 GA$EOUS RADWASTE TREATMENT SYSTEM s The requirement that the appropelate portions of these systems be used,'when speelfled, provides reasonable assurance. that the releases of radioactive materials in gaseous effluents wul be kept "as low as is reasonably achievable". i i This speelfication implements the requirements of 10 CFR Part 30.36a, General Design Criterion 60 of Appendix A to 10 CPR Part 30. and the design objectives given in Section
!!.D of ~ Appendix ! to 10 CPR Part 30. The specified limits governing the use of.
appropriate portions of the eyetems were specified as a suitable fraction of the dose design objectives set forth in Sections R.5 and ILC of Appendix 1,10 CFR Part 30, for-gaseous efijuents. [ 1 CALVERT CLIFFS UNIT 1 83/411-4 Amendment No.105 [
R AD10 ACTIVE EFFLUENTS B ASE5 tY6M*"4 J/4.11.2.5 EXPLO5tVE G AS MIXTURE This specification is provided to ensure that the concentration of potentIally explosive gas mixtures contained in the waste gu holdup system is maintained below the Klammability limit of oxygen. Maintaining the concentration of oxygen below its flammability limit provides assurance that the releues of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A ta 10 CFR Part 30. 3/4.11.2.6 G A5 STOR ACE TANKS The tanks incJuded in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technica! Specification to a quantity that is less than the quantity that provides assurance that in
- the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not o exceed 10 CFR 100 limits.
Restricting the quantity of radioactivity contained in each gu storage tank provides r.ssurance that in the event of an uncontrolled releue of the tank's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest $1TE BOUNDARY will not exceed accident guidelines. 3/4.11.3 SOLID RADIOACTIVE WASTE This specification implements de requirements of 10 CFR Part 50.36a and General design Criterion 60 of Appendix A to 10 CFR Part 30. The process parameters are included in the PROCESS CONTROL PROGRAM. 4
- 3/4.!!.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyrold, which nhall be limited to less than or equal to 75 mrems. The Specla! Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible.
CALVERT CLIFFS UNIT 1 B 3/411-5 Amendment No.105 l l
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R AD10 ACTTVE EFFLUENTS BASES hTO If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in vlotation of 40 CFR Part 190 have not already been' corrected), in accordance with the provisions of 40 CFR Part 190.!! and 10 CFR Part6 20.403c, is considered to be a timely request and fulflits the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190. and does not apply in any way to the other requirements for dose ilmitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and 3.!!.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel eyele.
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CALVERT CLIFFS UNIT 1 B 3/4116 Amendment No.105 i
i 3/4.12 RADIOLOGICAL ENV1 0NMENTAt MONITORING BASES 3/4.12.1 MONITORING PROGEAM l The radiological environmental monitoring program recuired by this soeci. l fication provides representative measurements of radiation and of radioactive ' materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures to MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Section IV.E.2 of Appendix ! to 10 CFR Part 50 and thereby supplements the radiological I effluent monitoring program by verifying that the measurable concentrations I' of radioactive materials and levels of radiation are not higner than expected on the basis of the effluent measurements and the modeling of the environmental ,
, exposure pathways. Guidance for this monitoring program is provided by the i Radiological Assessment Branch Technical Position on Environmental Monitoring. l Program changes may be initiated based on operational experience.
The recuired detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs recuired by Table 4.12-1 are considered ootimum for routine environmental reasurements' j in industrial laberatories. It should be recognized that the LLD is defined i as an a criori (before the fact) limit representing the cacability of a i measuriment system and not as an i coste*1eri (after the fact) limit for a particular measurement. ,, Detailed discussion of the LL*D and other detection limits can be found in l l RASL Procedures Manual, HASL 300 (revised annually)' Currie, L. A. , " Limits for
, i Qualitative Detection and Quantitative Detemination Apolication to Racio. i chemistry," anal. Chem. 40, !!6-93 (1960), and Hartwell, J. K., " Detection Limits for Racicanaly 1 cat Counting Techniques," Atlantic Ricnfield Hanfore l Company Report ARH-SA-215 (June 1975).
--:..3/4.12.2 LA'O USE CENSUS This soecificatien is provided to ensure that enanges in the use of areas
~
at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the , results of this census. The best infonnation from the door to-door survey, ' from aerial survey or from consulting with local agricultural authorities sna11 be used. This census satisfies the recuirements of Section IV.B,3 of Aepencix ! , gte 10 CFR Part 50. Restricting the census to garcens of greater than 50 m' : provides assurance that significant exposure pathways via leafy vegetables will j be identified and monitored since a garden of this size is the minimum recuired ' to produce the quantity (25 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garcen size, the following assumptions were made: 1) 20% of the garcen was used for growing l l broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegeta- - l l tien yield of 2 kg/m , ; f . ,g l l CALVERT CLIFF 3 - UN!T 1 B 3/4 12-1 Amencrent No. 100 1
l 3/4.12 RADIOLOG! Cal ENV!RONMErlTAL MONITORING l BASES f 3/4.12.3 INTERLaBOPATORY COMPARISON PROGPJW ' The requirement for participation in an approved Interlsboratory Com:ari. son Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are perforsned as part of the quality assurance program for environ. mental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix ! to 10 CFR Part 50. l
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l i !. l! o ll . 1; CAL Y.E ET ;'. . F r! 8.'.'i:7 *. 5 3/4 12-2 Amencment No.10C I l-
SECTION 5.0 DESIGN FEATURES
5.0 DES!GN FEATURES 5.' SITE MAP DEFIN!NG THE $17! BOUNDARY AND EFFLUENT RELEASE P0INTS 5.1.1 A map of the Calvert Cliffs Nuclear Power Plant site identifying the major plant structures as well as defining the radioactive effluent release points and the $1TE BOUNDARY is shown in Figure 5.1 1. LOW POPULATION 20NE 5.1.2 The low populatien :ene shall be as shown in Figure 5.1 2. 5.2 CONTAINMENT - CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrets building of cylincrical shape, with a come roof and having the following ' design features:
, a. Nominal insice diameter = 130 feet, i
- b. Nominal inside height = 1812/3 feet.
- c. Minimum thickness of concrete walls = 3 3/4 feet,
- d. Minimum thickness of concrete roof = 31/4 feet,
- e. Minimum thickness of concrete floor pad = 10 feet.
- f. Nominal thickness of steel liner = 1/4 inches.
- g. Net free volume = 2 x 106 cubic feet.
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CALVERT CLIFFS UN!T 1 51 Amencment No. 100 I
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DES:GN FEATURES l DESIGN DE'ESSUEE AND ?!Y:EE U ! n ~ ts *ea:::- :aia tr :. :i . i {i.1.1tei e: #:r e ar .::ir.:ernal :-ess re;:'s :es';re: at: stati :e mai
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!* :si; and a :e ; era:ure :n.f 5.3 EE:*TCR ~CRE
. FUEL ASSEMBLIES - '
~ ..- .
.i 5.3.1 assembly containing a maxim.:m of 175 fuel rods clad Each fuel red shall have a n minal active fuel length of 13E. 7 in:hes anc-
- contain a maximum total weight cf 2000 grtes uranium. The ini 1:a:in; shall have a maximum enrichment of 2.9" : weign; cer:e-U.235.
tial ::re Reload fuel snell be similar in pn.vsi:a1 design to the ir.i-iai : ore .. loacin; and shall have a maximum enrichment of 4.1 weign:
. . . . percent U-235. l 5.3.2 under entrol Except ele.forten see:ial test as authori:ed by the NRC, all fuel assemolies previously approved by the NRO. assemblies . snail be sleeved with a sleeve cesign }
/
CONTROL ELEMENT ASSEMBLIES 5.3.3
,::n:r:1 element The rea:ter core shall con:ain 77 full len;;h and no part length assembiits. . .
5.1 REACTOR COOLANT SYS~EM DESIGN PRESSURE AND TEM:E:.ATURE 5 . A .1 The he orholanthstem is designed and shall be maintained: a.
; In.2 ac:ordan:e Of .ne TSAR witn 'Othal',:the coce recuire ents see:ifiec in Set: :n i 1
- ' ne a::ii:atie Iurvetii ncet Re:ivirements,Lence fcr nor ai ce;racation grs t.
F:r a cressure Of 2500 psia, and
?:r a tem: erat *cre 'of 550*F, ex:e:- for the pressuri:er vnich
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' cav!R curFs - um : Amendmen: No. 22,c4 71 51
DESIGN FEATURES YOLUME 5.4.2 Thetotalwaterandsteamvolumeofghe actorbelantbystemis '- 10,614 :t 460 cubic feet at a nominal T,yg of $32 F. . 5.5 METEOROLOGICAL TOWER 30 CAT 10N 5.5.1 The meteorological tower shall be located as shown on Figure 5.!-1. 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum 10 3/32' x 10 3/32" center to center distance between fuel assemblies placed in the storage racks to ensure a k,gf of s 0.95 with the storage pool filled with unborated water. The k f conservativeallowancesforuncertaintiesdescribebofs0.95includesthe in Section 9.7.2 of the FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 5.0 weight percent. CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 inch center to center distance between new fuel assemblies such that k.gf will not exceed 0.95 when fuel having a maximum enrichment of 5.0 weight percent U 235 is in place and var'ious densities of unborated water are assumed including aqueous foam moderation and full flood conditions. The k of 1 0.95 includes the conservative allowance for uncertainties described in,f rSection 9.7.2 of the FSAR. DRAINAGE
- j. 5.6.3 The spent fuel storage pool is designe.1 and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.
CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.71 are designed and shall be maintained within the cyclic or transient limits of Table 5.71. 1 CALVERT CLIFFS UNIT 1 5-5 Amendment No. 27//////j), H,!H 2H, 139 l
I~ TA8tE 5.7-1 Q CORPONENT CYCLIC ON TRAN5IENT LIMITS M n - Component-1 [ e-Cyclic or Transient limit Deslan Cycle or Transient
% Reactor Coolant System 500 heatup and cc)1down cycles 70*r to 532*F-to 70*r Z
g 400 reactor trip cycles 100% to 0% RATED THERMAL POWER Z 10 Frimary Hydrostatic Tests 3125 psla and 60*F > NDTT 320 Primary Leak Tests 2500 psia and 60'r > NOTT Y' t Steam Generator 10 Secondary Hydrostatic Tests 1250 psia Secondary Side and temperature > 100*F 320 Secondary leak Tests 1000 psia Secondary Side With Primary - Secondary Ap of 820 psi and shell side temperature between 100*F and 200'F d k a R E Y 6 6
.._ __ r__m_ ____ _ _ -______ - -_ _ _ - - - _ _ - - -_.-. - . _ . -
SECTION 6.0 ADMINISTRATIVE CONTROLS f l
ADWINICTRATIYE CDKTPol$ __ 6.1 RtspoNsitfLITY 6.1.1 The Manager . Calvert Cliffs Nuclesr Power Plant shall be responsi. l ble for e,verall facility operation and shall delegate in writing the succes. sien to this responsibility during his absence. 6.2 ORfM IEATION 6.2.1 Q!G1IE & OrFSITE ORGAN 11AT10NS Onsite and offsite organizations shall be established for unit operatien and corporate management, respectively. The onsite and offsite organ. itations shall include the positions for activities affecting the safety of the nuclear power plant.
- n. Lines of authority, responsibility and coassunication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization '
positions. These relationships shall be documented and updated, as appropriate, in the fom of organization charts, functional de- " l scriptions of departmental responsibilities and relationships, and Job descriptions for key personnel positions, or in equivalent forms i of documentation. These requirements shall be documented in FSAR Chapter 12, and updated in accordance with 10 CFR 50.7)(e),
- b. The Manager Calvert Cliffs Nuclear Power Plant shall be responsi-ble for overall unit safe operation and shall have control over these onsite activities necessary for safe operation and maintenance l cf the plant.
1 i c. The Vice President huclear Energy shall have corporate responsi-bility for overall plant nuclear safety and shall take any measures needed to ensure acceptable perfomance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety,
- d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence free operating pressures.
6.2.2' UNIT STAFF
- a. Each on 60ty shift shall be composed of at least the minimus shift crew composition shown in Table 6.2 1.
- b. At least eno licensed Operator shall be in the control roos when fuel is in the reactor.
( ll CALVERT CLIFFS - UNIT 1 61 Amendment No. JJ, AJ, E N rtve-*5 " ci " en h uarn, w y m.ll#igg 131 AUG $ 11988 3.w ,
1 ASWINISTP.tTIVE COW'DL 5
- c. At least two licensed Operators shall be present in the control roon c;ri g rea:t:
sta t up, stbeduled reactor shutd wn, tr.d dur n; e recovery fr:9 rea:to- trips.
- d. An individual cualified in radiation protection proceduras shall be on site when fuel is in the reactor, e.
ALL CORE ALTERATIL after the in'itial fuel loading shall be di-rect'y sacerviset by either a liter. sed Ser.ier Rea: tor Operator er Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation,
- f. A site Fire Brigade of at least 5 members shall be maintained onsite at all times. The Fire Brigade shall not include the minimum shift crew necessary for safe shutdown of botn units (4 members) or ary personnel recuired for other essential functions during a fire ererge cy.
- g. Tne General Superviser Nuclear Operations shall hold or have held a senior rea: tor operator license at Calvert Cliffs. The Assistant General Suoervisor Nuclear Operations, Shift Supervisor and CO trc' Roo- Supervise shall hoic a ser.icr reactor operator li-c er.s e ,
licerse. ine Contro' Room Operator sna11 hole a reactor operator t CALVERT CLIFFS UN'.I I E la Amend:,ent No. 2f, 4 , H9, DJ,135
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DELETED l t. CALVERT CLIFFS - UNIT 1 62 Amendment No. JE,f), JE,pE,JJp,131
l ) DELETED I I 63 Amendment No. JJ,2J, CALVERT CLIFFS - UNIT 1 15,5%,95 JJ0,131
r TABLE 6.21 i MINIMUM SHIFT CREW COMPOSITION
- i Condition of Unit 1 - Unit 2 in MODES 1, 2, 3 or 4 LICENSE APPLICABLE MODES C AT EGDP,Y 1, 2. 3 & 4
- 5&6 50L** p p, OL** 3 3 l
hon-Licensed 3 3 Shift Technical Advisor 1pp 1p Condition of Unit 3 . Unit 2 in MODES 5 or 6 APPLICABLE MODES y
- 1. 2, 3 & 4 5&6 SQL** 2 le OL*' 3 2
. Non-Licensed 3 3
. Shift Technical Advisor lo o CALVERT CLIFFS - UNIT 1 Amendment No. NJ, 68
,4 l
l
~ T ABLE 6.21_ (Continued) \
'Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS during fuel reloading.
** Assumes each individual is licensed on each unit.
#Shif t crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accomnodate unexpected absence of on duty shif t crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2.1.
fiThe STA shall be qualified / serve as follows: (1) With one unit in MODE S or 6, and the other unit in MODE 1, 2, 3 or 4, the 50L holder other than the Shift Supervisor shall serve as the STA. (2) With one unit defueled and the other unit in MODE 1, 2, 3 or 4, the STA shall be an 50L holder in addition to the one 50L required. (3) With both units in MODE 1, 2, 3 or 4, the STA shall be an 50L holder in addition to the two 50Ls required. As an alternative to the above, the STA may be an individual with the e following minimum qualifications: a bachelor degree or equivalent in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant for transients and accidents. a 6-5 Amendment No. 53,105 CALVERT CLIFFS - UNIT 1 (
1 l 1 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OVALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N!8.1 1971 for comparable positions, except for (1) the Radiation Safety Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and 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 General Supervisor - Nuclear Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1 1971 and * -R_ af 10 CFR CA as appli-cable. 62.5Hc)j 6.4.2 A training program for the Fire Brigade shall b'e maintained under the direction of the General Supervisor - Quality Control and Support and shall meet or exceed the requirements of Section 27 of the NFPA Code 1975. 6.5 RWIN AND AUDIT 6.5.1 PLANT OPERATIONS AND SAFETY REVIEW C0694fTTEE (POSRC) FUNCTION 6.5.1.1 The POSRC shall function to advise the Manager-Calvert Cliffs Nuclear Power Plant on all matters related to nuclear safety. COMPOSfTION 6.5.1.2 The POSRC shall be composed of the: l Chairman: Manager Calvert Cliffs Nuclear Power Plant Member: General Supervisor - Nuclear Operations Member: General Supervisor - Electrical and Controls Member: General Supervisor - Chemistry l Member: General Supervisor - Mechanical Maintenance l Member: General Supervisor - Technical Services Engineering l Member: General Supervisor - Radiation Safety Member: General Supervisor - Plant and Project Engineering l ALTERNATES ! 6.5.1.3 All alternate' members shall be appointed in writing by the POSRC Chairman to serve on a temporary basis, however, no more than two alternates shall participate as voting members in POSRC activities at any one time. CALVERT CLIFFS - UNIT 1 66 Amendment No. 25,4J, 5),f2,95,119,131 l l
ADMINISTRATIVE CONTROLS I MEETING FRE0VENCY 6.5.1.4 The POSRC shall meet at least once per calendar month and as con-vened by the POSRC Chairman or his designated alternate. 92E!!! 6.5.1.5 A quorum of the POSRC shall consist of the Chairman or his desig-nated alternate and four members including alternates. RESPONSIBillTIES 6.5.1.6 The POSRC shall be responsible 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 Manager - Calvert Cliffs Nuclear Power Plant to affect nuclear safety.
- b. Review of all proposed tests and experiments that affect nuclear safety.
- c. Review of all proposed changes to Appendix 'A" Technical Specifica-tions. l Review of all proposed changes or modifications to plant systems or }
d. equipment that affect nuclear safety.
- e. Investigation of all violations of the Tdchnical Specifications including the preparation and forwarding of reports covering evalu-ation and recomendations to prevent recurrence to the Vice Presi.
dent Nuclear Energy and to the Chairman of the Off Site Safety Review Comittee,
- f. Review of all REPORTABLE EVENTS.
- g. Review of f acility operations to detect potential safety hazards,
- b. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairmen of the Off-Site Safety Review Comittee,
- i. Review of the Plant Security Plan and implementing procedures and shall submit recomended changes to the Off-Site Safety Review Comittee.
1 67 Amendment No. JE,JE, CALVERT CLIFFS - UNIT 1 f),95,195,119,131 l
m . 0 ADMINISTRATIVE C'KTR0ls
- j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Off Site Safety Review Committee.
- k. Review of any accidental, unplanned or uncontrolled radioactive release that exceeds 25% of the limits of Specification 3.11.1.2, 3.11.2.2 or 3.11.2.3, including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and for forwarding of these reports to the Manager Calvert Cliffs Nuclear Power Plant and the Off-Site Safety Review Committee.
- 1. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MNUAL.
CALVERT CLIFFS - UNIT 1 6-7,. Amendment No. JE,JE,
$3,99,195,119,131
ADMINISTRATIVE CONTROLS I AUTHORIT1 6.5.1.7 The Plant Operations and Safety Review Comittee shall:
- a. Recomend to the Manager Calvert Cliffs Nuclear Power Plant written approval or disapproval of items considered under 6.5.1.6(a) through (d) above,
- b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification with 24 hours to the Vice Pres w Tt-Nuclear Energy and the Off Site Safety Review Comittee of dis-agreement between the POSRC and the Manager-Calvert Cliffs Nuclear Power Plant; however, the Manager Calvert Cliffs Nuclear Power Plant shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The POSRC shall maintain written minutes of each meeting ,and copies shall be provided to the Vice President - Nuclear Energy and Chairman of the
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Off-Site Safety Review Comittee. 6.5.2 0FF SITE SAFETY REVIEW ComITTEE fOSSRC) FUHtTION 6.5.2.1 The Off Site Safety Review Comittee shall function to provide independent review and audit of designated activities in the areas of:
- a. nuclear power plant operations
- b. nuclear engineering
- c. chemistry and radiochemistry
- d. metallurgy and non destructive examination
- e. instrumentation and control
- f. radiological safety
.g , mechanical and electrical engineering
- h. quality assurance practices
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Amendment No. /J,JJp,131 CALVERT CLIFFS - UNIT 1 68 f
1 ADMINISTRATIVE CONTROLS ! COMPOSITION 6.5.2.2 The OSSRC shall be composed of at least seven members, including the Chairman. Members of the OSSRC may be from the Nuclear Energy Division I or other BG&E organization or from organizations external to BG&E and shall collectively have expertise in all of the areas of 6.5.2.1. QUALIFICATIONS T 6.5.2.3 The Chairman, members and alternate members of the 055RC shall be appointed in writing by the Vice President - Nuclear Energy and shall have an academic degree in engineering or a physical science, or the equivalent, l 1-and in addition shall have a minimum of five years technical experience in j one or more areas given in 6.5.2.1, No more than two alternates shall participate as voting members in OSSRC activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the OSSRC Chainnan to provide expert advice to the OSSRC.
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MEETING FREQUENCY 6.5.2.5 The OSSRC shall meet at least once per six months. QUORUM 6.5.2.6 The quorum of the OSSRC necessary for the performance of the OSSRC review and audit functions of these Technical Specifications shall consist of more than half the OSSRC membership or at least four members, whichever is greater.* This quorum shall include the Chairman or his appointed alternate and the OSSRC members, including appointed alternates, meeting the require-ments of Specification 6.5.2.3. No more than a minority of the quorum shall have line responsibility for operation of the plant. l CALVERT CLIFFS - UNIT 1 6-9 Amendment No. 43, 110 l
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ADMINISTRATIVE CONTROLS REVIEW 6.5.2.7 The OSSRC shall review:
- a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provisions of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
- b. Proposed changes to procedures, equipment or systems which involve an 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 this Operating License.
- e.
Violations of codes, regulations, orders, Technical Specifications, E license requirements, or of internal procedures or instructions having nuclear safety significance.
)
f. Significant operating abnonnalities or deviations from normal anri expected performance of plant equipment that affect nuclear safety. , l g. All REPORTABLE EVENTS. l
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, syitems or components. s l
- i. Reports and meetings minutes of the POSRC.
l I CALVERT CLIFFS --UNIT 1 6-10 Amendment No. 94
ADMINISTRATIVE CONTROLS AUDITS 6.S.2.8.1 Audits of facility activities shall be perfomed under the cog-nizance of the OSSRC. These audits shall encompass:
- a. The confomance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months. .
- b. The perfomance, training and qualification of the entire facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in
- facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
- d. The perfomance of activities required by the Qu'ality Assurance Program to meet the criteria of Appendix "B",10 CFR Part 50, at least once per 24 months.
- e. Deleted.
- f. The Safeguards Contingency Plan and i'mplementing procedures at least once per 12 months in accordance with 10 CFR 73.40(d).
- g. Any other area of facility operation considered appropriate by the
. OSSRC or the Vice President-Nuclear Energy. l . h. The Facility Fire Protection Pronram end implementing procedures at least once per 24 months.
- 1. An independent fire protection and loss prevention program inspection and audit shall be performed at least ence per 12 months utilizing either qualified offsite licensee persennel or an outside fire protection fim.
- j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.
- k. The radiological environmental monitoring program and the results thereof at least once per 12 months.
- 1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months, m.
The PROCESS CONTROL PROGRAM and implementing procedures for proces-sing and packaging of radioactive wastes at least once per 24 months.
- n. The perfomance of activities required by O..: 4...cy Assurance Program for effluent and environmental monitoring at least once per 12 months.
CALVERT CLIFFS - UNIT l 6-11 Amendment No. 26 82, 205 110
l l l ADMINISTRATIVE CONTROLS 6.5.2,8,2 Review of facility activities shall be performed under the cognizance of the OSSRC. These reviews shall encompass: ,
- a. The Facility Emergency Plan and implementing procedures at least j once per 12 months in accordance with 10 CFR Part 50.54(t).
AUTHORITY l l 6.5.2.9 The OSSRC shall report to and advise the Vice President-Nuclear Energy l 1 on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.
. RECORDS '
1 6.5.2.10 Records of OSSRC activities shall be prepared, approved and distributed as indicated below:
- a. Minutes of each OSSRC meeting shall be prepared, approved and .
forwarded to the Vice President-Nuclear Energy within 14 days - I following each meeting,
- b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President-Nuclear Energy within 14 days following. completion of the review.
c c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice President-Nuclear Energy and to the manage-
' ment positions responsible for the areas audited within 30 days after completion of the aucit.
6.6 ' REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Comission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part .50, and
- b. Each REPORTABLE EVENT shall be reviewed by the POSRC and the results of this review shall be submitted to the OSSRC and the Vice President-Nuclear Energy.
w
-CALVERT CLIFFS - UNIT 1 6-12 Amendment No. 48,82,7A,105,110
ADMINISTRATIVE CONTROLS j.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The facility shall be placed in at least HOT STANDBY within one hour.
- b. The NRC Operations Center shall be notified by telephone as soon as -
possible and in all cases within one hour. The Vice President Nu-clear Energy and the OSSRC shall be notified within 24 hours,
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the POSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective ,
action taken to prevent recurrence,
- d. The Safety limit Violation Report shall be submitted to the Commis-sion, the OSSRC and the Vice President Nuclear Energy within 14 days of the violation,
- f. 8 PROCEDURfJ .
6.8.1 Written procadures shall be established, implemented and maintained covering the activities referenced below:
- a. The applicaDie procedures recomended in Appendix "A" of Regulatory Guide 1.33, Revision 2, Febrtary 1978.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security plan implementation.
- e. Emergency Plan implementation.
- f. Fire Protection Program implementation,
- g. The amount of overtime worked by plant' staff members performing
. safety related functions must be limited in accordance with the NRC PolicyStatwaton}%rkingHours(GenericLetterNo.82-12).
t 1 CALVERT CLIFFS - UNIT 1 6-13 Amendment No. JE,f), 75 S2,195,119,131
l ADNINISTRATIVE CORTROLS _ [
- h. PROCESS CONTROL PROGRAM implementation.
- i. OFFSITE DOSE CALCULATION MANUAL implementation.
6.8.2 Each procedure and administrative policy of 6.8.1 above and changes thereto shall be reviewed by the POSRC and approved by the Manager Calvert Cliffs Nuclear Power Plant prior to implementation and reviewed periodically as set forth in administrative procedures. I
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I CALVERT CLIFFS - UHli 1 6-13a Amendment No. JE,f), 75,87.JEE.JJE 131
l ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures 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 members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected,
- c. The change is documented, reviewed by the POSRC and appproved by the Manager Calvert Cliffs Nuclear Power Plant within 14 days of imple- l mentation.
6.9 REPORTING RE0VIREMENTJ R0llTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted, i STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall 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 different design or has been manufactured by a different fuel supplier, and (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 identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a compari-son of these values with design. predictions and specifications. Any correc-tive actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other comitments shall be included in this report. i CALVERT CLIFFS UNIT 1 6 14 Amendment No. 29,0 , SJ #,JEE,JJE,131
ADMINISTRATIVE CONTROLS 6.9.1.3 Startup reports shall be submitted within (1) 90 days following l completion of the startup test program (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all ]' three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. ANNUAL REPORTS U Annual reports covering the activities of the unit as described ] 6.9.1.4 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. 6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility.
7 '~ and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their apociated man rem exposure according to work and job functions,- e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter. TLO, or film badge measure-ments. Small exposures totalling less than In 20% of the individual the aggregate, at least ) total dose need not be accounted for. 80% of the total whole body dose received from external sources shall be assigned to specific major work functions,
- b. The complete results of steam generator tube inservice inspections performed during the report period (reference Specification 4.4.5.5 b).
- c. Documentation of all failures and challenges to the pressurizer PORVs or safety valves, s If A single submittal may be made for a multiple unit station. The submittal should combine those sections that are comon to all units at the station.
1/ This tabulation supplements the requirements of 120.407 of 10 CFR Part 20. 6-15 Amendment No. 29,6 8 CALVERT CLIFFS - UNIT 1
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1
_ADMINISTR ATIVE CONTROLS MONTHLY OPER ATING REPORT 6.9.1.6. Routine reports of operating statistics and shutdown experience shall be submitted on a montnly basis to the Director, Of fice of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20535, ATTN: Document Control Desk, with a copy to the Regional Administrator and to the NRC Resident inspector, no later than the 13th of each month following the calendar month covered by the report. ANNUAL RADIOLOGICAL ENVIRONMENTAL OPER ATING REPORT
- 6.9.1.7 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.12.2. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period p.irsuant to the locations specified in the Table and Figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explalning the reasons for the missing results. The missing data shat! be submitted as soon as possible in a supplementary report. The reports shall also include the fo!!owing a summary description of the radiological environmental monitoring program; at least two legible maps" covering all sampling locations keyed to a table giving distances and directions from the central point between the two containment buildings; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the sampling schedule of Table 3.12-1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
- A single submittal may be made.
** One map shall cover stations near the site boundary; a second shall include the more distant stations.
I CALVERT CLIFFS UNIT 1 6-16 Amendment No. 29,6B,9#,105 i
ADMINISTR ATIVE CONTROLS The Annual Radiological Environmental Operating Report will include the identification I of the cause of unavailability of samples (if any), and will describe the locations used for the replacement samples. The report will also include any permanent changes in the sample locations which could appear in the monitoring program. It will include a revised figure (s) and table for the ODCM reflecting the new location (s). SEMI ANNUAL R ADIOACTIVE !FFLUENT RELE ASE REPORT
- 6.9.1.8 Routine Radioactive Ef fluent Release Reports covering the operation 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." .
The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseoits effluents and solid waste released from the plant as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes anc' Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-C >oled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. The Radioactive Effluent Release Report to be submitted within 60 days af ter January 1 I of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by- ) hour listing on magnetic tape of wind speeds wind direction, atmospheric stability, and precipitation (if measured), or in the form ofgjnt 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 calendar year. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).
- A single submittal may be made for Calvert Cilffs, since the radwaste systems are common to both units.
" In lieu of submission with the Se ni-Annual Reports, sri 9, Sr90 analyses results may be submitted in a supplementary report within 120 days after January 1 and July 1 of each year.
*** In lieu of submission with the first half year Radioactive Effluent Release Report, this summary of required meteorological data may be retained on site in a file that shall be provided to the NRC upon request.
6-17 Arnendment No. 29,9/,105 CALVERT CLIFF 5 UNIT 1 I l
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ADMINISTRATIVE CONTROLS 2dil1NNW10ACTMFTJFLUENPTMASE JtIDetT7dentiatrstip The radioactive Effluent Release Report to be submitted 60 days after January 1 of each year-shall also include an assessment of radiation ; doses to the likely most exposed MEMBER OF THE PUBLIC.from reactor- 1 releases and other nearby uranium fuel cycle sources, including doses 4 from primary effluent pathways and direct radiation, for the previous i calendar year to show conformance with 40 CFR Part 190 Environmental- : Radiation Protection Standards for Nuclear Power Operation. Acceptable o methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1, October 1977, and , NUREG 0133, " Preparation of Radiological Effluent Technical ! Specifications for Nuclear Power Plants". The Radioactive Effluent Release Reports shall include the following l information-for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during.the report period:
- a. Container volume, l
- b. Total curie quantity (specify whether determined by measurement orestimate),
a
- c. Principal; radionuclides (specify whether determined by: '
measurement or estimate),
- d. Source of waste an,d processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms), !
- e. Solidification agent or absorbent (e.g., cement). !
-The Radioactive Effluent Release Reports shall . include a list and 0 description of unplanned releases from the site to UNRESTRICTED AREAS of ;
radioactive materials in-gaseous and liquid effluents made during the ; reporting period. q
-The- Radioactive Effluent Release Reports shall ' include' any. changes made
- during-the reporting period to the PROCESS CONTROL PROGRAM--(PCP), and-to-the 0FFSITE DOSE CALCULATION MANUAL;(0DCM), as well;as a listing of new U locations' for dose calculations identiified by the-annual land use census pursuant to Specification'3.12.2.
L SPECIAL-REPORTS l- . . J
.6.9.2 Special reports shall be submittsd to the Regional Administrator y i ref the NRC. Regional Office within the time period specified for each '
i ' report. These reports shall- be submitted covering the activities t . identified below pursuant to the requirements of the-applicable reference - specification; y
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-11 CALVERT CLIFFS - UNIT 1 6-18 Amendment No. 29/#f,105
l3 ADMINISTRATIVE CONTROLS
- a. ECCS Actuation Specifications 3.5.2 and 3.5.3 j i
- b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3. 4
- c. Inoperable Meteorological Instrumentation, Specification 3.3.3.4. ,
j
- d. Seismic event analysis, Specification 4.3.3.3.2. j
- e. Core Barrel Movement, Specification 3.4.11. 1
- f. Fire Detection Instrumentation, Specification 3.3.3.7.
- g. Fire Suppression Systems, Specifications 3.7.11.1, 3.7.11.2, 3.7.11.3, 3.7.11.4, and 3.7 ll.5.
- h. Penetration Fire Barriers, Specification 3.7.12.
1
- i. Steam Generator Tube Inspection Results, Specification 4.4.5.5.a and c. ;
- j. Specific Activity of Primary Coolant, Specification 3.4.8.
- k. Containment Structural Integrity, Specification 4.6.16.
- 1. Radioactive Effluents - Calculated Dose and Total Dose, ;
Specifications 3.11.1.2, 3.11.2.2, 3.11.2.3, and 3.11.4.
- m. Radioactive Effluents - Liquid Radwaste, Gaseous Radwaste and Ventilation Exhaust Treatment Systems Discharges, Specifications 3.11.1.3 and 3.11.2.4.
l-
- n. Radiological Environmental Monitoring Program, Specification 3.12.1.
- o. Radiation Monitoring-Instrumentation, Specification 3.3.3.1 1
(Table 3.3-6),
- p. Overpressure Protection Systems, Specification 3.4.9.3.
- q. Hydrogen Analyzers,--Specification 1.6.5.1.
L r. Post-Accident Instrumentation, Specification 3.3.3.6 l l. c l. 1 11 CALVERT CLIFFS - UNIT 1 6-18a Amendment No. JJ/Jf/JJJ/JJ7,147
n. 1 . ADMINISTRATIVE CONTR0'.S 6'.10- RECORD RETENTION 6.10.1. The following records shall be retained for at least five years: ,
- a. Records and logs of facility operation covering time interval l at each power level,
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
- c. All REPORTABLE EVENTS - )
- d. . Records of surveillance activities, inspections and calibrations
. required by these Technical Specifications,
- e. Records of reactor tests and experiments. l f .- Records of changes made to Operating Procedures..
- g. Records of radioactive shipments.
- h. Records of sealed source and. fission detector leak tests and results.
- 1. Records of annual physical inventory of all sealed source material.of record.
6.10.2' The following records shall-be retained for the duration of the Facility Operating License: a.- Records and drawing ch'anges reflecting facility design modifi-tations' made to systems and equipment described in the Final
' Safety Analysis Report.
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
.c. . Records. of facility radiation and contamination surveys.
O L CALVERT CLIFFS . UNIT 1 6-19 Amendment No.-28, 94 1 l^ - . . --. .- . - - -
ADMINISTRATIVE CONTROLS
~
- 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 identified in Table 5.7-1.
- 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 identified in the NRC approved QA Manual as lifetime records.
- j. Records of reviews perfomed for changes made to procedures or equip-nent or reviews of tests and experiments pursuant to 10 CFR 59.59.
- k. Records of meetings of the POSRC and the OSSRC.
CLd / RecorAs'of ronnepaM)ualMitTt%er: cnered-war A- -
,pffvisi of parafraph fr13.
}
j j L-K Records of the service lives of all safety related snubbers including the date at which the service life commences and associated installation and maintenance records. 6.11 RADIATION PROTECTION PROGRAM Procedures fo/ 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.raciation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or "alam signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20:
- a. A high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Special or Radiation Work Pemit and any individual or group of individuals pemitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
I 6-20 Amendment No. Ef,125 1 CALVERT CLIFF 5 - UNIT 1
3 ADMfNISTRATfVE CONTROLS b. A high radiation area in which the intensity of radiation is greater then--1000 mrem /hr shall be subject to the provisions of 6.12.1.a. above, and in addition locked' barricades shall be provided to prevent- unauthorized entry into such areas and the keys 'shall be maintained by the Supervisor-Radiation Control Operations and the Operations-Shift Supervisor on duty under their separate adminis-trative control. 6.13 SYSTEM INTEGRITY The licensee shall ic'ement a program to reduce leakage from systems outsidi serious transient or accident to as low as practical This program levels. co shall include the following:
- ) .
Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.
' Leak test requirements for each system at a frequency not i to exceed refueling cycle intervals.
6.14 IODINE MONITORING The licensee shall implement a program
- which will ensure the capability to accurately detennine the airborne iodine concentration in vital areas under l accident condit1_ons. This program shall include the following:
- 1. Training of personnel.
- 2. Procedures for monitoring, and 3.
L Provisions for maintenance of sampling and analysis equipment. L l 6.15 POSTACCIDENT SAMPLING The licensee shall establish, implement and maintain a-program
- which will .
ensure the capability to obtain and analyze reactor coolant, rediosctive iodines and particulates in plant gaseous-effluents, and containment-atmosphere the following: samples under accident conditions. The program shall include
- 1. Training of personnel, 2.- Procedures for sampling and analysis, and
- 3. -
Provisions for maintenance of sampling and analysis equipment. L L
'Itoperation is. acceptablemanuals if (e-the licensee maintains details of the program in plant -
- c. ..
maintenance procedures, ERPIPs). chemistry procedures, training-instructions, p CALVERT CLIFFS - UNIT 1 6-21 !. 9/ddf 79/24/80, M. 53,78, Am. No.195,198, JJ9,113 d
l ADMINISTRATIVE CONTROLS } 6.16 PROCESS CONTROL PROGRAM (PCPl 6.16.1 The PCP shall be approved by the Comission prior to implementation. 6.16.2 Licensee initiated changes to the PCP:
- 1. Shall be submitted to the Comission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
An evaluation suppor the premise that the change did not a.
>- reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- b. A reference to the date and the POSRC meeting number in which the change (s) were reviewed and found acceptable to the POSRC.
- 2. Shall become effective upon review by the POSRC and approval of the Manager - Calvert Cliffs Nuclear Power Plant.
6.17 0FFSITE DOSE CALCULATION MANUAL (00CN) 6.17.1 The ODCM shall be approved by the Comission prior to implementa-I tion. ' I 6.17.2 Licensee initiated changes to the ODCH:
- 1. Shall be submitted to the Comission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:
- a. Sufficient information to support the rationale for the change.
Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a change number and/or change date together with appropriate analyses or evaluations justifying the change (s);
- b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the POSRC.
- 2. Shall become effective upon review by the POSRC and approval of the Manager Calvert Cliffs Nuclear Power Plant.
AmendmentNo.JSE,JJE, ) CALVERT CLIFFS - UNIT 1 6-22 JJp,JJJ,131 l
ADMINISTRATIVE CONTROLS 6.18 MAJOR CHANGES TO RADICACTIVE LIOUID, GASEOUS AND SOLID WASTE ' TREATMENT SYSTEMS 6.18.1 Licensee initiated major changes to the radioactive waste system.s t (liquic, gaseous and solid) shall be reported to the Commission in the Semi-annual Radioac-ive Effluent Release Report for the period in wnich the modification to 'the waste system is completed. The discussion of each change shall contain: a. A description of the equipment, components and processes involved, b. Documentation of the fact that the change including the safety analysis was reviewed and found acceptable by tne POSRC. e 1 l l I 1 i i CALVERT CLIFFS - UNIT 1 6-23 Amendment NO. 113
4 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT Nos.1 AND 2 1 APPENDIX B PARTI ENVIRONMENTAL TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE Nos. DPR-53 AND DPR-69 DELETED , See Appendix A Technical Specifications 4 Sections 3/4.11 and 3/4.12
-ISSUED BY THE U. S. NUCLEAR REGULATORY COMMIS$10N p' $
CALVERT CLIFFS UNIT 1 Amendment No. 23,76J60,105 - CALVERT CLIFFS UNIT 2 Amendment No. 7. H , 82, 86
_ _ _ .. _. __ . - - _ _ . _ . _ . . . _ . _ . . _ _ _ . _ . ~ . _ _ _ - _ . _ _ _ . _ _ . _ APPENDIX f f" M TOfACILITYOPERATINGLICENSENd(.DPR53-50FF,-C7 CALVERT CLIFFS NUCLEAR POWER PLANT BALTIMORE CAS & ELECTRIC COMPANY DOCKET N0 s. 50-317 a ',0-;^G ENVIRONMENTAL PROTECTION PLAll I e (NON-RADIOLOGICAL) TECHNICAL SPECIFICATIONS i i i- ' CALYEFT CLIrfs t' NIT 1-QEnRT -;i.HTEL '9 TT 7. J,,,;,,.-iUji AMENDhENT ;; , ;; ----NO. 70
100 Ob.iectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during construction and operation of the nuclear facility. The principal objectives of the EPP are as follows:
- 1. Verify that the plant is operated in an environmentally acceptable manner, j as established by *he FES and other NRC environnental impact assessments. ;
1 1
- 2. Coordinate NRC requirements and r.aintain consistency with other Federal, State and local requirements for environmental protection.
- 3. Keep NRC informed of the environmental effects of facility construction .
I and operation and of actions taken to control those effects. Environmental concerns identified in th'e FES which relate to water quality matters are regulated by way of the licensee's NPDES permit. I I i l l 1-1 l l - CA c.LVERT CLI,FFS l' NIT 1 AMENDMENT NO. 70
,ure, r ree im , a Aurynygge 33,_ 3----
l
2.0 Environmental Protection Issues In the FES-OL, the staff considered the environmental impacts associated with the operation of the Calvert Cliffs Plsnt. Certain environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment. The Appendix 8 Environmental Technical Specifications issued with the licenses included discharge restrictions and monitoring programs to resolve the issues. Prior to issuance of this EPP, the requirements remaining in n e ETS were:
- 1. Protection of the aquatic environicent by limiting the discharge of !
dissolved solics and acids and bases and an annual inventory of treatment chemicals added or used in the plant. (ETS2.2.1,2.2.2) j 2. Surveillance programs for fish, crabs and oysters, and water cuality to estehlish impact of plant operation on toe aquatic environment. (ETS 3.11 .
)
1
- 3. Special studies to document levels of intake entrainment and impingement L in relation to the dentities of important species in the plant vicinity.
(ETS 3.1.2.b) , Aquatic issues are now addresseo by the effluent limitations and monitoring requirements continued in the effective NPDES Perinit issued by the State of , Maryland Departa nt of Health and Mental Hygiene. The NRC will rely on this agency for regulation of matters involving water quality and aquatic biota. 2.} .- L
* \
CALVERT CLIFFS tfNit AMENDMENT NO.'70 mu_:a;;5 =q ~_" wun n
, _ , ,, m _.,_ . . , . - - - + ^ ~'
3.0 Consistency Requirements - 3.1 Plant Design and Operation The licensee may make changes in station design or operation or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question. Changes in plant design or operation or performance of tests or experiments which do not affect the environment are not subject to this requirement. Before engaging in unauthorized construction or operational activities which may affect the environment, the licensee shall perfom an environmental evaluation of such activity.* When the evaluation indicates that such activity involves an unreviewed environmer:tal question, the licensee shall provide a written evaluation of such activities and obtain prior approval from the NRC. ,
, A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmentel statement (FES) as modified by staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environ-mental impact appraisals, or in any decisions of the Atomic Safety and
- Activities are excluded from this requirement if all measurable nonradiological effects are confined to the on-site areas previously disturbed during site preparation, plant construction and previous plant operation.
l 3-1 l j' CALVERT CLIFFS UNIT 1 AMENDMENT NO. 70
- j. e x -r-> u : n ;-:' e L JddEN" T. ??
l l l
Licensing Board; or (2) e sigr.ificant change in effluents or power level (in accordance with 10 CFR Part $1.2(b)(2))! cr (3) a matter not previously reviewed ! and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact. The licensee shall maintain records or changes in facllity design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include a written evaluation which provides bases for the determination that the change, test, or experiment does not involve an ' unreviewed environmental cuestion, t Activities governed by Section 3.3 of this EPP are not subject to the requi: vents of this section. - 3.2 Reporting Pelated to the NPDES Pemit and State Certification ' (pursuant to Section 401 of the Clean Water Act) 1. Violations of the NPDES Pemit or the State 401 Certification Conditions shall be reported to the NRC by submittal of copier of the reports required by the NPDES Permit or State 401 Certification. gi 2 The licensee shall provide the NRC with a copy of any 316(a) or (b) studies and/or related documentation at the same time it is submitted to the remitting agency.
- 3. Changes and additions to the NPDES Permit or the State 401 Certification shall be reported to the NRC within 30 days following the date the 32
\
CALYERT CLIFFS UNIT 1 tnu , . __ m ,, uni, - AMENDMENT NO. 70
.. c- m rum m w s1
change is approverd. If a pemit or certification. in part or in its entirety, is appealed and stayed, the fiPC shall be netified ai%hin I 30 days following the date the stay is granted. 4 The NRC shall be notified of changes to the effective NPDES Pemit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the pemitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPDES Pemit at the same time the application is submitted to the pemitting agency. 3.3 Changes Required for Compliance with Other Environmental Regulations 4 Changes in plant design or operation and performance of tests or experiments l which are required to achieve compliance with other federal, State, or , local environmental regulations are not sub,iect to the requirements of Section 3.1. e 3-3 i
- i. CALVERT CLIFFS UNIT 1 AMENDMENT NO. 70 n <**Ec e retu m ,
u s a t e e n h _.
~ . ,, +me.s 9q -.Yv-rgy -- ei V e rp.g---<- ,-g.e me- og-y-Q me ,,w.,"du d'4 me-e n m-N'"=dw-* -* ** 'EP-C~+i--w--~^'"4- --' 'u- + " "h-N *--+M"8 '- '
- l 4.0 Environmental Conditions l
l 4.1 Significant Environmental Events Any occurrener of a significant event that indicates or could result in significant environnental impact causally related to station operation shall be recorded and promptly reported to the WRC within 24 hours followed by a written report within 30 days. No routine monitoring programs are required to implement this condition. The written report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (3) indicate the agencies notified and their preliminary responses. . Events reportable under this subsection which also require reports to other Federal State or local agencies shall be reported in accordance with those reporting requirements in lieu of the reovirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency. The following are examples of significant environmental events: excessive bird impactior events; onsite plant er aninal disease outbreaks; mortality or 4-1 CALVERT CLIFFS UNIT 1 AMENDMENT NO 2 _ID
. =4AWISMdhtffe-tm1T+ ..
O
unusual occurrence of any species protected by the Endangered Species Act , [ of 1973; unusual fish killst and increase in nuisance organisms or cond1Bions. l 1 I t 1 1 i 4-2 ('. AMENDMENT NO. 70-CALVERT eni u CLIFFS,, ,U,NI,T -
;i p -...r.a. g e.- __
*g '
eamuwuvuvuu NUCLEAR REGULATORY COMMISSION
- WASHINGTON, D. C t0585 5
( '*)& n s;f N l'N# 8.. ..' - ENVIRONMENTAL I!@ACT APPRAISAL BY THE OPTICE 07 Nt' CLEAR REACTOR RE::ULATION SUFFORTING AMEhTME*ti NOS. 23 AND 7 TO LICENSE NOS. DPR-53 AND DPR-69 BALTIMORE GAS & ELECTRIC CCtOANY CALVERT CLTTTS STCLEAR PO'1ER PLANT UNIT NCS.1 AND 2 DOCKET Nos. 50-317 AND 50-318 Description of Procesed Action By letter dated January 3, 1977, and supplement thereto dated June 7, 1977, the Baltimore Gas & Electric Company (DC&E) requested an amendment to racility Operating License No. DPR-53 for the Calvert Cliffs Nuciant Power l Plant Unit No. 1. Following discussions with and agreement by BG&E, the reques . was med1fied to include Tacility Operating License No. DPR-69 for the Calvert Cliffs Nuclear Power Plant Unit No. 2. The request, as modified, vould replace the Appendix B (Environmental) Technical Specifications (ETS) of both Unit Nos.1 and 2, in their antirety, with a common set of ITS. The requested content and format of the new common CTS is very closely reisted , s j to the present Unit No. 2 ETS. , 1 l l In replacing the axisting Unit No.1 ETS, BC&E has identified the major changes being proposed as deleting the limiting conditions for Operation (LCO) on e
, maximum' discharge.ta=perature and rate of change of the discharge temperature, modifying the allovaele range of pH for discharge, changing the radioactive afflu-ants section, adding an onsite meteorological monitoring program, modifying or deleting the monitoring requirements for aquatic che:nistry, impingement, the l l benthic feasibility entrainment and intake-velocity studies, modifying the 1
radiological environments 1 nonitoring section, and changing the a 61nistrative controls requirements. Chcnges to the LCO's are necessary to make the ETS
,g, consistent with the NPDES pe mit limitaciens issued by the State of Maryland on June 7,1976. and reconfirmed, following a public hearing, on May 12, 1977.
BG6E requested the deletion of a nunber of monitoring programs following the succe.sful ccepletion of these programs. Environmental Impacts of Procened Action Potential environmental impacts associated with the proposed replace-ment of the ETS vere identified by BC6E in the table entitled "Mc3er Changes in Envircraental Technical Specifications - Csivert Clif f s Unit No. 1." This table was part of the Januar" 3, 1977. sub=ittal and included 12 items. Each of these items is discussed and the enviroi.= ental impact assessed below.
- 1. Deletion _of Maximum Discharte Temp'erature 1.1-1:s J .
Backe,round Inf er :stien The existing limitation on maximum discharge temperature in the ETS is 32.2 C (90 F). The basis f or the limitation was that it was the temperature used for the impact predictions in the Calverc Clif fs Final Environmental Statement (FES) as reflected s in the State of Maryland Department of Water Resources Surface .
. Water Appropriation Pemit # C-70-SAP-1 and the State of Maryland Water Resources Administration Section 401 Certification .
(Permit No. 74-DP-0187) for Unit No.1 and Unit No. 2, respectively. Periods of exception for special studies at higher te=peratures were allowed in the ETS for both units. Though the limit reflected d that set by the Stste of Maryland, the staff further noted a biologiesi basis for a higher limit in that:
a "Most of the indigenous species of fish can tolerate periods of exposure at temperatures of 93*F or higher and due to the mixing characteristics of the submerged jet discharge no significant impact on the aquatic
- ecosystem due to the thermal discharge is expected" .
(ETS, Unit No. 1, Section 2.1.2 Basis). l The State of Maryland issued a combined Statt and NPDES permit I
-(State Discharge Permit No. 74-DP-0187A and NPDES Permit No.
- HD0002399),- effective June 7, 1976, which eliminated the maximum
~
discharge tesperature limite of 32.2*C '(90*7). The thernal limitations sat by the Permit were a daily maximum t.T < 5.56*C 13
- ' (10'F) and a daily maximum heat rejection rata = 1.32 X 10 ,
J
- Joules per hour (1.25 X 1010;3TU per hour). A state adjudicatory i
hearing to review the permit issuance was held on September 15, 16, and 17, 1976. The two. issues subject to adjudication vara
> : challenges to the' deletion of the 90'T maximum temperature and ;
to the provision to allow a: test period ending on November 1,- .c
-)
. -1976..to evaluate temperature . increases across the condensers of greate,r than 10'F.
Expert ~ testimony was presented on behalf of'the Chesapeaka Environmental Protection Association,-Inc., j the County Comissioners for Calvert County, . the State ' ii. of Maryland Water Resources Administration and the licensee. On the basis of the hearing record, the -
. . - _ _ a- n _ _ ___ 2_ 2 _... _ 2 _ .-__._.__ u .. . _ . - ._c.,._. - . _ . _ . - _ _ _ _ _ . . _ . . . _ - _ - .
1-r . 1 l 1 I Hearing Officer recommended l to the Director of the f WaterResourcesAdministrhttenthat (1) the 32.2*C (90*T) maximum limit be reinstated in the State and NTDE3 permit issued on June 7,1976, with a caveat permitting excursions beycnd that limit for a period up to 200 hours per year and (2) no further tests for the purpose of evaluating condenser temperature rise of greater than 10'F be ) l l permitted. However, in his subsequent decision2 , the 3
. 1 1
Director overruled the recommendatiens of the Hearing ] 1 Officer and denied the requested changes to the discharge i l pe rmit . , The salient findings of the Director were that! l s , 1
. l l
(a) the parait in question is designed to minimize and I control. the full range of adverse environmental l 1mpaccc associated with the plant.' The permit is-1 l not and should not be limited to one parameter, , maximum comparature, at the probable expense of incurring other kinds of damage, i.e., mechanical. (b) If temperature enn be naasured at the outfall then measurement of temperature rise across the l l condenser provides similar control because identical measurement techniques are used.
5-(c) With respect to temperatures in excess'of 90*F, the incursions are of very (phemeral nature and their l impacts very limited. (d) . The permit as issued conceins a special condition which provides a supplemental censure of protection should an unforeseen environmental problem arise. (e) The use of a 90*T plus 200 hours per year excursion ' limitation would introduce a regulatory problem that of accepting excursions above maximum limitation. (f) Evidence was presented in the record which indicates that the physical stresses of pumping large volumes of water may cause more environmental damage than allowing short-term exposure of the same organisms to higher temperature differentials. l (g) The Administration (WRA) should not foreclose the possibility of further scientific assessment of environmental impact, should examination of available , data indicate that such testing is warranted. s M i e
- . ,v..,
6-To make the dTS limits on maximum discharge compatible , with State and NPDI:5 Permit requires the deletion of the ' 32.2'c (90'T) limit and* deletion of the limits on duration
.(1 100 hours per year) and temperature maximum (93'T) for the special studies in the ETS for Unit No. 1. Potential impacts, therefore must be evaluated for operation of the :
plant with no limits on the maximum discharge temperature. , The findings of the WRA Director'are consistent with the
.results of our evaluation. '
. Evaluatien of potential Impacts The Come.ission issued Amendment No.16 to the license fer Calvert Cliffs, Unit No. 1, on August 5, 1976, which modified
- the conditions for special studies associated with main-
' condenser. cooling. water discharge temperature3 . The Staff's ,
Environmental Impact Appraisal supporting that amendment . reviewed the licensee's semi-annual monitoring. reports- . (covering the first paar of operation) and concluded that the increased AT,of 12 to 14'F, with. discharge
, temperatura not to exceed-93'T maximus,.would not"
, affact the populations of phytoplankton,.scoplankton and ichthyoplankton in- the Chesapeake-gay by being' entrained '
-through the, plant. -The conclusion'of negligible biological' impact was supported primarily by the short duration.of the. i t
increases in AT above 10'F, the short exposure time (5 minutes), short generation time for phytoplankton and zooplankton, the i e i, -a,,,,,-, e v - ,. -we-,,,me, , , , , , - - , ,w,.-,,--, ,. ew e e ,- w - a- --~ . ,~.- - . . + , - +m~ -~wr-,-ma~,c--wr------~~~- - - - - - - - - -
. 1
~7- o . l low abundance-of ichthyoplankton representing "important" species in the site vicinity, and the insignificant amount of water withdrav.- by Unit 1 comrared to the flow past the site. 'these same factors are pertinent to the current
\
evaluation and support the same conclusion, i.e., negligible i impacts. The only new factor to be evaluated is whether the I biological it. pacts associated with long-term operation of ;
.. l the plant with unlimited discharge tenperature are different. '
from those impacts on the short term which have been evaluated and found negligible. Long term in this context means . operation over the several conths each year when intake temperatures may exceed 32.2*C (90'F). . Operating experience indicates that the intake temperature i exceeded 26.7'C (80*F) for 67 hours during 1975 and for 21 hours during 1976". As shown in Table 1. only one hour
, was,1,ogged with intake temperature in the range ,
, of > 27.8'C (82'F) and 128.3*C (83*F) and none Iogged at > 28.3*C -
(83*F). These two years of operating data indicate that the aaximum discharge temperature would not have exceeded 33.9'c (93'F) within the existing NPDES and ETS limit of AT 15.56*C ' (10*F) and the plant at normal operation *.
, / -
in the ETS limiting condition on AT, the operacion ant less than
" normal" was- recognized as. a possibility ' due to pump outage or
- Normal. operation, as defined in ETS for Unit No. 1, refers to
' operation withLali circulating pumps (6/ unit) in operation.
f
. ~ . . ., ,. . . _ , . , , . . . -.,_.-.__m. . . . - . - - . _ _ . . . . . . . . _ , _ , -, , - . . _ , , , . . . . ,
M g. Table 1 Calvert Cliffs Nuclear Pouer Plant Number of Hours Intake Temperatures T.xceeded 80'T During 1975 and 1976 Months _ , , , , , _ Temeerature Ca t eno rv' (*F) Total Hoiirs 80'F Execeded
>80181 l >61182
>82583 80'T Exci.edeil July 1975 25 5 1 31 August 1975 32 4 0 36 July 1976 1 0 0 1 August 1976 15 5 0 20 Two-Year '
Totals 73 13 1 88 D3ta Source: Letter from A. E. Lur.dvall, Jr. (BC&E) to Office of Nuclear Reactor Regulation (NRC), dated June 7,1977 e S 9 G 4 j
maintanence. To-prsvide ficxibility for such cccurrence, short-term excursions of AT 112*T up to two hours were allowed. Wa-find no biological basis in the TES for the setting of a specific 11.miting condition during such events; the limit
- appears to be based solely on plant operational considerations.
In reviewing the most recent data ,5 mm.h of which were , examined in the State hearing, we feund no results which would negate our previous appraisal. Testimony for the Chesapeake Enviromnental Protection Association indicated that the excursions
-in excess of 90*T would not allow for an adequate saf ety margin6 ,
This conclusion was based on the literature review of Miller and Beck . In cross examination, it was determined that the 90*F limit was based on long exposure time (*. 24 hours) l-0 - whereas the exposure time through the plant is approximately 4 minutss. We conclude that this khort period of exposure provides a " safety margin" for protecting the biota se supported by the results of the Meansee's studies I
- I' '
l Moreover, the withdrawal of wa*.ar, with two unit operation, is only about 2% of the non-tidal surface flow by the plant or about 0.5% of the tidal flow. No significant changes in the biotic consunity outside' of- the immediate- plume are anticipated L: nor have they been' detected,' thus far, by the monitoring program. ( l- \:
. We cor.clude that the limits on AT and anximum discharge temperatures
- y are essentially providing redundant control on thermal' effects l
. and the change in plant operation caused by deletion of_the
, maximum discharge temperature will not crestra winnificant -
adverse impact..
. . , + , ,.%,_- ,. .....m, . . . , , . , . .._,,.,_.m_,._, , _ .
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- 2. Deletian of R.1te of, Chenste af Discharre Temnersturo Limit LTS 2.1.3 timits the rat e of change of the discharge t emperatur -
to not more than 10'r per hour. The licensec proposes to del (*.c this ' spceification as it is nnt in the t' nit 2 CTS nor in the llPDES permit fer the station. The basic for this specificntien indicates ': hat, due to the lov (10'r) te perrture rise of the cooling water and the hich jicchtree ve'locity (9 fps) at the disch:rcejthe rate of 10*F per hour will have a neglicj bic effect en fish in the near-field of the thermal ple cc. Sinee ; fj:n carinot swim for lens in the near-ric11 high velecity area, the 10*r/ hour was chosen as a rest. ens!.le rate assur.in;; fish would t,e attract ed to er re-sidint in the near-field of the pic:c. The licensec and the Marylend Prwar Plant Siting frecrtun have conducted monit orinc of ficih in the ditcherce vicinity. The licensee has ::npled fish b/ otter tr al and gill nc; in the site vicinity including the discharge e.nd has not. detected ritnificent differer.ces in fi:h pcpulations in the disc'harge area compared tn either preoperational sampling or to samples collected at thermally unaffected stations . Fish distribution were niso sturlted by acoustic curveys conducted by the Maryland Power Plant SJting Program . Surveyn made 'n the. plume and in nearby arcas showed that large concentractens of Jnt ce rich do not occur in-the plune. Therefore, thu staff concludes ths.t 6 chance
. in plant operation'due to the clic.ination of the rate of chanfo of dis-charge tc=perature will not 'rcruit in a, significant 1.: pet as fish do not 9
6 m,-r, - ev.- r- .e - - . .-,,,e~.,ev-.~.wr-,, - ..e-- - - --------+-+4, ,e--, - . - .-,,r -e .
concrerste in the near field of the thermal olume where they would be subjceted to a rapid change in tea.perature. , Monitorinc of fish in tt,c dischsrce arca will continue to be conducted secercing to technical specification 4.1.1.5 to ensure thtt crcenisms are not cdver::1y af. fected by the heated ditcharce.
- 3. pH Environmental Pretection Condition Chante The environmental protection condition for pH in the Unit 1 ETS (Section 2.2.1) limits discharges resulting f rom the demineraliter regeneration vaste neutralization tank to a pH range of 6.5 to 8.5.
The similar restriction in the Unit 2 ETS specifies a. pH range of 6.0 to 9.0. Tha proposed change would replace the existing pH range in the Unit 1 ITS with that of the Unit 2 ETS. - The proposed change vould allow the pH of the discharge of the demineralizer regeneration neutralization tank from Unit 1 to vary between 6.0 and 9.0, inclusive, which is less restrictive than the existing ETS range of 6.5 to 8.5, inclusive. The U.S. Environmental Protection Agency, in its publication Quality Criteria for Water recom= ends a water quality standard for pH for the protection of aquatic life in an estuar,ine environm,ent of 6.5 to 9.0, inclusive. Also rec:mmended is the avoidance of rapid fluctuations in pH due to vaste discharges. The EPA has also , published Effluent Limitations and Guidelines for the Steam Electric Generating Point Source Category I0 . These regulations describe minimum standards of performance for the industry for 4
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r , the protection of aqueti.* species in and on the receiving water body. The guideline for pH is the range 6.0 en 9.0, inclusive. The rationale for this effiuent limitation is that, aside from being attainable by the industry, unacceptable harm to the receiving water biota due to differences in discharge and receiving water pH is not likely because of the available buffering capacity of most natural waters. Th'e staff evaluated the discharge of those chemicals likely to cause an alteration in the pH of the discharge (Calvert Cliffs tJnits 1 and 2 TES, Section III.D 3 'and V.C.2.e) . The results of the i evaluation indicated that changes in effluent pH to the extent that effects on the aquatic biota of the receiving waters vould not be expected due to the buffering capacity of the cooling vater.
.This evaluation considered a discharge from the plant water treatment facilities with a pH range of 6.0 to 8.5 at a rate of up to 500 spm prior to mixing with the condenser cooling water with
's pH range of 7.3 to 8.4 at a discharge rate of 600,000 gpm.
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- Evaluation of operational data indicates that discharges frem the plant have not resulted in any discernable horizontal trend s ii s , s a w..l.. t..'.s ....s. a.. ,. I hoe ..s .s s h.s,.. I i n 6.ts ' i s..u.. ...~ t. s i s4 locations cru VJdely separnted and include stat luns which nre directly affected by the dischstr.v plume and stations which are not af fected by the plant discharge.
Within station samples (i.e. , vertical dis tribu tio n) indicates a trend toward elightly higher pH_vajues
at the surf ace than .se the bottom during dsylight hours .15 , 3 However, these dif ferences have not been attributed to plant operational affects and are likely due to depletion of carben dioxide via photosynthesis resulting in slight elevation of the pH in this warser surf ace layer. . 1 Evaluation of the effects that reperted environmental events arising from pH values outside of the currently allowed range has indicated that the resultant pH change of the plant discharge water was ] 0
. immeasurable upon discharge to the Chesapeake Bay .
l-l Ve conclude that a change in the allowable pH range of the discharge from the 'nor=alizar tank of f rom 6.5 to B.5 to 6.0 to 9.0 will not result in' an unacceptable environr: ental impact. 4
- 4. Radioactive Effluents Change .
ETS 2.3 on Radioactive Effluents contains the same discharge limits
- as the previous Unic No.1 Section 2.3 entitled " Radioactive Materials." Amendment No. 20 to the Unit No.1 license dated s February 11, 1977, had moved this specification from Appendix A, l
section 3.9 to Appendix B, Section 2.3. The Unit No. 2 version of this section is formated slightly different and is considered easier to understand. . The staff..with the approval of BG&E, has updated the wording in l~ j accordance with Appendix I of 10 CTR Part - 50 replacing "as-low-as l practicable" with "as low as reasonably achievable".
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e By letter dated June 4, 1976, the licansee providcd addittensi .
'information pursuant to Appendix I to 10 CPR Part 30. After we complete our evaluation of this information we intend to revise the Technical Specifications to reflect the requirements of Append 1:: I.
To redute confusion, the word " site" has been inserted to designate
.which dircharge limits are intended to be for boch units, l The proposed amendment vill not allow the licensee to discharge l concentrations greater than the maximum allowe'd nor to discharge ,
f more activity in a year than the maximum allowed. Compliance with radiological ef fluent Technical Specificati aus vill maintain concentrations of radioactive materials in unrestricted areas to a small fraction of 10 CTR Part 20, St:ndards fer Protection Against l i l Radiation. .
. i 1
. 1 The staf f concludes that no detrimental eff ect on the environment 1
vill reruit from the changes proposed in the radioactive ef fluents i i section. .
- 5. Onsite Heceorological Monitoring Program
' The licensee has proposed that the onsite meteorological monitoring program be considered for both Unit Hos. I and 2. This proposal results in no change to the Unit No. 2 ITS requirement 3.1.1.b.
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, 6. AquatAc Chemistry Monitoring Requirement Change The Environmental Monitoring Requirement in the Unit 1 ETS (Section 4.1.1) requires that four sampling locations near the site be sampled at the surface for a variety of physical and chemical vater quality constituents.
The proposed change would delete the requirement in the Ceneral
. Aquatic Ecological Survey to monitor iron, potassium, chromium, j cobalt, stror.cium, manganese and zine in the Chesapeake Bay in the vicinity of the Calvert Cliffs Plant. .The staff has examined the TES to determine the basis for the inclusion of these parameters in the operational aquatic monitoring program. Preoperational monitoring data available at the time of preparation of the TES )
(for years 1968 and 1969) indicate that, of these parameters, only iron, manganese and potassium were monitored. Additionally, no pre-existing environment.a.1 stresses related to these metals were . identified for the Chesapeake Bay in the TES. The anticipated I chemical wastes from the operatinn of the plant were examined by the staff, and the conclusions reached for these wastes were (1) that they will produce no detectable environmental damsge (TES Section - III.D.3) 'and (2) that the expected concentration of chemicals in the discharge will be well below those believed to cause detrimental effects on ao.uatic life (TES Section V.C.2.e). The l seven metals listed above were not specifically identified as plant l- chemical wastos in the TES. On these bases we conclude that there - 1 was demonstrated to be no causal link between anticipated plant operation and the concentration and effects of these metals in the V waters of the Chesapeake Bay. 1 l vn-rw, ,-- ~ m --,-
'" A summary of the operational monitoring of these seven metals for the yest 1976 and a comparison of the operstional and preoperational values ressured is presented in the latest Semi-Annual Monitoring Report . For calendar year 1976, only potassium measurements in the surface samples exhibited any trend between stations. This trend was an increase in concentration for the louer bay . sampling locations. Slight within station (i.e., vertical) gradients were noted during 1976 for cobalt, strontium and , iron, withibottom samples centaining higher concentrat ons. Overall, in 1976, concentrations of manganese, iron, cobalt and :ine apparently increased. However, a more sensitive detecting system var employed during 1976 than in previous years. This may account for the apparent higher values and.the greater number of values, in
. general, which exceeded the threshold of detection in 1976 than in
.1975. ', ,
These metals are not regarded as being significant toxins in an estuarine environment. For example, of these seve.. metals the - U. S. EPA Quality Criteria for Water identifies a specific u,niform limit only for hexavaient chromium (100 ng/1) and an application factor for =ine of 0.1 of the 96 LC50 for the most sensitive species. At no time during the operations 1 monitoring program did concentrations of these metals reach levels considered harmful to aquatic life. ,
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-s Although not requested by the licensee, the staf f in its reviev of the requirements of this specification noted that sampling is requixed for total bacteria and total colif orms. It was further determined that their inclusion in the specification was not reviewed in the FIS and is not supported by the basis of this specification. As the staf f, at this time, cannot justify a need for sampling of these parameters they can be deleted. On the bases of}}