Text

Amends 188 & 165 to Licenses DPR-53 & DPR-69,respectively, Revising TS to Improve Reliability of RCS Porvs,Per GL 90-06
	ML20065P098
Person / Time
Site:	Calvert Cliffs
Issue date:	04/20/1994
From:	Capra R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20065P103	List: ML20065P098; ML20065P164;
References
	REF-GTECI-070, REF-GTECI-094, REF-GTECI-NI, TASK-070, TASK-094, TASK-70, TASK-94, TASK-OR GL-90-06, GL-90-6, NUDOCS 9404280265
	Download: ML20065P098 (111)
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{{#Wiki_filter:- ] 1 \\- UNITED STATES [ j ' . WASHINGTON, D.C. 20606-0001 NUCLEAR REGULATORY COMMISSION h i p p. BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. I j AMENDMENT TO FACILITY OPERATING LICENSE Amendment NoJ188 License No..DPR-53 1. The Nuclear Regulatory Comission (the Conssission) has found that: A. The application for amendment by Baltimore Gas and Electric. Company .(the licensee) dated September 1, 1992, as supplemented. March 17,. 1994, complies with the standards and requirements of the Atomic' Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; m.. 8. The facility will operate in conformity with the application, a the provisions of the Act, and;the rules and regulations of the Commission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and. safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common. defense and security or to the health and safety of the public; and-E. The issuance of this amendment is in accordance with 10 CFR Part. 51 of the Cosmiission's regulations and all applicable requirements-have been satisfied.

2..Accordingly, the license.is amended by changes to the Technical-Specifications as indicated-in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-53 is hereby-amended ~to read as follows:

940420026S 940420 DR ADOCK 0 37 7

(2) Technical Soecifications The' Technical Specifications contained in Appendices A and B, as revised through Amendment No.188, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license-amendment is effective as of the date of its issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION MCLe N Robert A. Capra, Director g. o Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation '

Attachment:

Changes to the Technical Specifications. Date of Issuance: April 20,,1994 s ,mu " 91 =e : ; -- j 1 i

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pnREQy {., UNITED STATES .j { NUCLEAR REGULATORY COMMISSION y WASHINGTON, D.C. 20066-0001 \\...../ BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING-LICENSE Amendment No.165 License No. DPR-69 1. The Nuclear Regulatory Comission (the Commission) has. found.that: A. The application for amendment by.. Baltimore Gas and Electric Company (the licensee) dated September 1, 1992, as supplemented March 17, 1994, complies with the standards and requirements of the' Atomic Energy-Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter.-I; B. Thefacilitywillo$erateinconformitywiththe. application, the provisions of tie Act, and the rules and regulationsLof the-Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and. safety of the public, and (ii) that such ' activities'will be' conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or:to the health and safety.of.the public; and E. The issuance of this amendment;is in.accordance with 10 CFR Part. 51 of the Commission's regulations and all applicable" requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69 is hereby: amended to read as follows: q a( .. 1

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-s. (2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.165, are hereby incorporated in the license. The 4 licensee shall operate the facility in accordance with the Technical Specifications.' 3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION ofukO..{Q Robert A. Capra, Director Project Directorate I-I-Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 20, 1994 4. S to E 1 3

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3-E ATTACHMENT TO LICENSE AMENDMENTS-AMENDMENT NO. 188 FACILITY OPERATING LICENSE NO. DPR-53

-- AMENDMENT No.165 FACILITY OPERATING LICENSE NO. DPR-69 dK'E'TNOS.50-317AND50-318 Revise Appendix A, DPR-53, as follows: Remove Paaes~ Insort Paaes Table of Contents, Page IV . Tab' e of Contents, _ Page. IV - Table ~of Contents, Page V Table of Contents, Page V Table of Contents,-Page X- ' Table of Contents, Page X. Table of Contents, Page XI Table of Contents, Page'XI' 3/4 4-7 3/4 4 3/4 4;8 3/4 4 3/4 4-31* 3/4 4 9.- 3/4 4-32* 3/4 4 3/4 4-34 3/4 4-33.--3/4 4-35' 4:4-36 R 3/4 4-43* - 3/4 4 3/4 4-42* 3/,3/4 4-2 B 3/4 4-2 8 ^ B 3/4 4-3 B 3/4 4-3 B 3/4 4 B 3/4 4-10* B 3/4 4 B 3/4 4-10* + 1 8 3/4 4-11. B 3/4~4-11: ~ j B 3/4 4-12' s Revise Appendix A, DPR-69, as follows:

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.g Remove Paaes Insert Paaes Table of Coritents, Page IV-Table of Contents, Page IV Table of Contents,.Page V Table of. Contents, Page V Table ~ of Contents, Page X i'A - Table of Contents,- Page XI Table of Contents,' Page X Table of Contents, Page XI 3/4 4-7 3/4 4 3/4 4-8 3/4 4 3/4'4-31* 3/4 4-7.'- 3/4'4-32* 3/4 4 3/4 4434-3/4;4-33 m 3/4 4-35-- ) .3/4 4-35.- 3/434-41* 3/4 4-36 --3/4-4-42* B 3/4 4-2 8 3/4 4-2 4 ~B 3/4 4-3 8 3/4 4-3 B 3/4 4 B 3/4'4-9* B 3/4.4 B 3/4 4-9* B 3/4 4-10 B 3/4 4 B 3/4 4-11* B 3/4 4-11*

These pages are text rollover pages with no changes as the. result of this amendment.

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1 TABLE OF CONTENTS - LIMITING CONDITI6NS'FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION O .PJ4GE 3/4.3 INSTRUMENTATION ~

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M ""~i'1ML 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION 3/4:311

.,..,e n.m 3/4.3.2' ENGINEEREFSAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION................... 3/4 3-9 we,, -, .-g,g 1 3/4.3.3 -MONITORING INSTRUMENTATION -4 -Radiation Monitoring' Instrumentation-'P.". W '...' '3/4 3-23 Incore Detectors 3/4 3-27 Seismic Instrumentation............... 3/4 3-30. Meteorological Instrumentation 3/4 3........... Remote Shutdown Instrumentation........... 3/4 3-36 Post-Accident Instrumentation............ 3/4 3-39 Fire Detection Instrumentation 3/4 3-43 Radioactive Gaseous Effluent Monitoring Instrumentation:................... 3/4 3-48 Radioactive Liquid Effluent Monitoring Instrumentation................_..-. 3/4 3-53' 3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION STARTUP And POWER OPERATION.............. .3/4 4-1 H0T STANDRY..................... 3/4 4-2 Shutdown...................... 3/4 4-4 3/4.4.2 SAFETY VALVES..................... -3/4 4-6 3/4.4.3 RELIEF VALVES.................... 3/4 4-7 3/4.4.4 PRESSURIZER..................... 3/4 4-9 l 3/4.4.5 STEAM GENERATORS '3/4.4-10 l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l Leakage ~ Detection Systems....,............: 3/4.4. Reactor Coolant System Leakage 3/4 4-19L 3/4.4.7 CHEMISTRY....................... 3/4 4 lL 3/4.4.8 SPECIFIC ACTIVITY............_...... 3/4.4 l r CALVERT CLIFFS - UNIT 1 .IV Amendment No.:188

l .,A [, j q TABLE OF CONTENTS LIMITIM CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-SECTION Pfgg ... e. 9. ; e. : .v.-..... 3/4:4:9 PRESSURE / TEMPERATURE ~EIMITS ^ s Reactor Coolant S Pressurizer...ystem 3/4 4-28 . w.......-... T......1..ic 3/4~4-32 ..f.i. .z... ,,......,. Overpressure Protection Sys,tems...,..'..y 3/4 4-33 j STRUCTURAL INTEGRITY '[ ~3/4 4-37 l d 3/4.4.10 ASME Code Class 1,.' 2 And 3 Components.;"...,.;..,,.. ;.,. ~ 3/4.4.11 CORE BARREL MOVEMENT'.......'......... 3/4.4-39 l: I 3/4.4.1Fl'TDOWNL$NEEXCESSFLOW E 3/4 4-41 l-3/4.4113 REACTOR COOLANT SYSTEM VENTS 3/4 4-42. l l

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m-3/4.5 EMERGENCY CORE C0OLING SYSTEMS](ECfS)l,},1 g Y.,.,~ j ~ 3/4.5.1 SAFETYINJECTIONTANKS 3/4's'5-1 v. a. v. u 1 3/4.5.2-ECCS SUBSYSTEMS - MODES 1.,2jnd,3 g 7,50 PSIA). 'l 3/4 5-3 m. 4.r 3/4.5.3 ECCSSUBSY$JEMS,,J460'E$'3.(41730., PSIA),and.,,,,.,4.... . 3/4 5-7 3/4 5.4 REFUELING WATER TANK 3/4 5-8 3/,4.6. g A g y TEMS 3 3/4.6.1 PkIMARYCONTAINMENT- ~ ~ CONTAIISIENTi INTEGRITY u. .4.-

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4 3/4 6-1 "" Containment ' Leakage ~...".~.............. 3/4 6-2 Containment Air Locks................ 3/4 6-5 Internal Pressure.................. .3/4 6-7 Air Temperature..,.......'......... 3/4 6-8 Containment-Structural-Integrity 3/4 6-9 Containment Purge System 3/4 6.......--.-........ 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6............... Containment Cooling System.............. '3/4 6-18;. 3/4.6.3 IODINE REMOVAL SYSTEM......-........-.. 3/4 6-20 CALVERT CLIFFS - UNIT .1-- V Amendment No. 188

-. -..~ ~. - g TABLE OF CONTENTS BASES pag SECTION A .,w u g m. m. 3/4.0 "" APPLICABILITY.. n r............ ;... B 3/4 0-1 1 - ;,*x L 3/4.1 REACTIVITY CONTRCt SYSTEMS' q ,y p 3/4.1.1 BORATION CONTROL B 3/4 1-1 ~ - c.. m. e, n.. '. "'.#[.. '."........".~... B'3/4 1-2: 3/4.1.2 BORAT' ION' SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...:.......... B 3/4 1-3 I 3/4.1. <......PNER D.I,,STRbg"n' ION LIMITS wr n 3/4.2.1 LINEAR HEAT RATE ..........~...... B 3/4 2-1

..wtw.a 2 akw.,

s 3/4.2.2, 3/4.2.3, and 3/4.2.4 TOTAL. PLANAR: AN0; ;,,mi . INTEGRATED RADIAL PEAKING. FACTORS - F,AND F, AND AZIMUTNAL POWER TILT - T, B 3/4 2-l' 4 -+ nn www uwt wuw :..:;m m.a 3/4.2.54 DNS PARAMETERF.c.n.x%WW.4 vi.s.r.y.6 c. B 3/4 2-2 + vF .a.ev.c er 3/4.3 INSTRUMENTATION w erm ai.an s o...reee m... w w.:. ::. 8 a 3/4.3.1 and 3/4~.3.2 ' PROTECTIVE AND~ ENGINEEREDTSAFETY FEATURES (ESF) INSTRUMENTATION... ae;.;.1 1 . '.. B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION.............. #,,... B 3/4 3-l' n a,; a ..m. 3/4.4 REACTOR C00UWT< SYSTEM m;. arc .e 3/4.4.1 -COOLANT LOOPS AND COOLANT CIRCULATION........ B 3/4 4 3/4.4.2. SAFETY VALVES.....-............ 0 3/4 4-1 i 3/4.4.3 RELIEF VALVES....................._. B 3/4.4-2 3/4.4.4 PRESSURIZER.........>...... .,r ...,.. B 3/4.4-3' l. 3/4.4.5 STEAM GENERATORS'.................. B 3/4'4-3 i CALVERT CLIFFS - UNIT 1. X Amendment No.-188-l

,h: TABLE OF CONTENTS BASES. 3rp-SECTION gig 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE :........... B 3/4 4-5 l -- 3/4.4.7 CHEMISTRY,......... 4.r... %., 7..... 9. 3.. B.3/4: 4-5 l . ~.. 3/4.4.8 SPECIFIC ACTIVITY..... .......l.._.,.....i B 3/4 4i6 -l b ~ PRESSURE /TEMPERAT RE LIMITS..............B'3/4_4-7 -l 3.4.4.9 3/4.4.10'. STRUCTURAL INTEGRITY.... '..... '.. I..'. ~ 6 3/4 4 l ~ 3/4.4.11 CORE SARREL MOVEMENT.................'a,.. B 3/4 4-11' nr.:,ut, me.r,' 3/4.4.12 LETDOWN LINE EXCESS FLOW.....a.. _....... R;'J/4 4-12 ' l- .y ~ -sw a ~.o,L s. .,. ;..),.J,l,. '&.k.. ws...iB"3/4'4-12 1 3/4.4.13 REACTOR COOLANT SYSTEM VENTS 3/4.5 EMERGENCYCORECOOLINGSYSTEMS(EC$[ 'h S,.,,.,[ ',. ',,, fg;p.mg.,g.,._R.3/,45-1 3/4.5.1 SAFETY INJECTION TANKS 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..'........... B 3/4 5-1 'l'l ue., s o., r 3/4.5.4 REFUELING WATER TANK (RWT) .. m..,4,y.r. 4.s u 8 3/4 5-3' l

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..,:.y%. 3/4.6 CONTAIMENT SYSTEMS i 3/4.6.1 PRIMARY CONTAINMENT... 7/4.'..T.'.f.!!'.".'[.'..TN..B3/46 ' 3/4.6.2 DEPRESSURIZATION AND COOLING"SVST C.#

.:.?.Y/.~B3/46-3 3/4.6.3 10DIKNEM$dtL' $$Ni!I......Q ',7."/.7...$$f. JB' 3/4 6-3 3

3/4.6.4 CONTAIMENT ISOLATION VALVES B 3/4 6-3 3/4.6.5. COMBUSTIBLE GAS CONTROL, ..B 3/4 6-4 3/4.6.6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSTEM.... B 3/4_6 4 CALVERT CLIFFS - UNIT 1 XI Amendment No. 188

'4 3/4.4 REACTORC004NTSYSTEM 3/4.4.3 RELIEF VALVES LIMITIM COM ITION FOR OPERATION Two power-cperated h1Jef Yalvei(PORVs) an(i 0 valves shall be OPERABLE. thef[Esociated block" 3.4.3 .3, m '" " " ~ ' " " " APPLICA8ILITY: MODE 5 1, 2 and 3*. l ACTION: ~] If one or both PORV(s) has excessive seat leak g'e. within 1 hour s. 4 close the associated block valve.(.s).gn(pi,n,t,ain, powerito the l blockvalve(s). .,4. m,3 .y With one PORf inokefable 'due to ca'uN[o'the'r tb ~ b. seat leakage. wit in 1 hour either restore the PORV to OPERABLE I status or close the associated block valve aM.t'emovqtpower from r the block valvei restore the PORV to;.0PERABLE s atus within the following 5 days or be in. NOT STAMBY yithin:the' nay.ill hours and at or below 365*F. within the folloWng.24. houp. [' ; g.,,,, With both ORVs inoperableTduii[o.cM'olfier[thnixcissiire[ N P c. PORV. seat leakage, within 1. hour 0PERABLE status or close the assoettJur restore.oQeJORY cia'ted block valve and remove power from the block valve;. restore one PORV to OPERA 84 status within the followJns 72; hours or he,.is il0T 5thmslf within the next.12' hours.and at or below 365'F"within,thewfoi, low (ps' 24 hours.,,..g a uwnm ~ c. - mm.- With 'one or. both block, valve (s) inoperable. within, associated d. I hour restore the block valve (s)..to OPERABLE status ar placailts PORV(s) in override closeA Restore at least one block valve to i OPERABLE status within the next 72 hours.if both block valves are ) inoperables.mestora any.rentining inoperable block lyalve to OPERABLE status within the following 5 days; otherwise, be in at least NOT STANDtY within the next 12 hours and at or below 365 F within the following 24 hours.4 The provisiini of Specification 3.0.4 are not applicable. ( e. Jt; unnt .r.- 1 i Above 365'F. At or below 365'F, Specification 3/4.4.9.3 applies.. l j CALVERT CLIFFS - UNIT 1 3/4 4-7 Amendment No. 188 1

y 3/4.4 REACTOR C0OLANT SYSTEM $URVEILUWICE REQUIREMDITS 4.4.3.1 Each PO W shall be demonstrated OPERABLE: a. At least once per 31 days by'perfomance of a CHAlglEL FUllCTI0llAL-TEST, in accordance with Table 4.3-1. Item 4. .i ....a. t,..... ....c.m, b. At least once per 18 months by performance of a.CNAaBIEL CALIBRATI0ll. m au i. m :. -. n. u r nr a a a.. 4.4.3.2 Each block valve shall be< demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed to meet the"requirementscof Action.a. b, or c in Specification 3.4.3. m "io s. y.ans

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'f att .o 4 ..s, '1 1 1 I CALVERT CLIFFS - UNIT 1 3/4.4-8 Amendment No. 188

,3 f. 3/4.4 REAF_ TOR COOLANT SYSTEM 3/4.'4.4 PRESSURIZER LIMITING CONDITION FOR OPERATION .; a. 3.4.4 The PressurTzer shall be!0PERABLE withT stea'afbuble'and w'ith at' least 150 kw of pressurizer heater capacity casable of being su plied by ' emergency >ower. The pressurizer level shall ;>e maintained witiin an operating.zand between 133'and 225 inches except'when three charging pumps ~ are operating and letdown flow is l'eii than 25 GPMi If.three charging pumps are operating and letdown flow is less.than 25.6PM press,urizer. level shall be limited to between 133 and 210 inches. APPLICABILITY: MODES l'and 2." ACTION: E T

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With the pressurizer inoperable due to an inoperable emergency a. power supply to the pressurizer heaters either restore the inoperable emergency >ower supp,1y within M ho4rs or be in at least NOT STANDRY wit.11n the next 6' hours and.fii HOT'SuuTD0lAl within the fg110 wing.12,hgurst 7 h.., j, g ,3 ^ With the'piessilrizer ot'h'eN64 (do,N'*rs ope.dt ('jV" fea rk1e b. 1 STANDBYwiththereactortripbrea nwit n,4hoursandin H0T SHUTD0lAl.w{thjn the fo1 Towing,6Jeug.,;,,.,% ..sc ~ .v: mn SURVEILLANCE REQUIRENDITS J 4.4.4 The pressurizer water level shall be detemined to be within the. above band at least once per 12 hours. 1 I CALVERT CLIFFS - UNIT 1 3/4 4-9 Amendment No. 188 -l. ~

I q ) 3/4.4 REACTOR C00LAltf $YSTEM 3/4.4.5 STEAM GENERATORS 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 CPERABLE status prior to increasing T., above 200'F. SURVEILUUICE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by perfomance - of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Stean Generator Samole Se'ection and lisoection - Each steam generator sha'l be deteminet OPERABLE during slutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 ' Steam Gonorator Tube Sample Select' on and lnsouction. The steam generator tube m'n'mun samp a s'ze, inspect'on resu't c'assification, and the corresponding action required.shall be as'specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be perfomed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes: shall be verified acceptable per the acceptance criteria of S acification 4.4.5.4. The tubes selected for each inservice inspection stall 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 randon basis except: 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)ofeachsteamgeneratorshall' include: 1. All nonplugged tubes that previously had detectable wall penetrations.(>20%).and CALVERT-CLIFFS - UNIT 1. 3/4 4-10 Amendment No. 188 T e-m.,. 1,.,.. ,,.m, ,.,ses r3 . vv -p er. ep-wp, y~

-g 3/4.4-REACTOR C00 LAIR $YSTEM SURVEILLANCE REQUIREMENTS (Continued) 2. Tubes in those areas where experience has indicated potential problems., The$econdandthirdinserviceinspectionsmaybeless.thana I c. full tube ins >ection by concentrating (selecting at.least 50% of the tubes to >e 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: 1 Cateaory Jnsoection Results C-1 Less than 5%'of'the'totel. tubes inspected are. degraded tubes and none of the in,spected tubes +4re defective, -d% One or more tibe's[' EiiIot;more than 1% C-2 of the t6taFtutidP1Nigec"fdd are defective, or between 54 and 10% of the total tubesinspected are degraded ,, y.. s, .y.- tubes. 1 C-3 ,.. More them 10$ of the total tubes ~~"""-"*tnspected WVegfsdat!" tubes or more ~- than 1% of the inspected tubes are i defective. Note: In all inspection d_ >reviously degraded tubes must-

i exhibit signiffc n 10%) further wall penetrations to 1

- -] be included in the above. percentage calculations. 1 4.4.5.3 Inspection Frecueic' os - The above. reitdired inservice inspections of steam generator tubes sia' be perfonned; ht the following frequencies: m 4 m.a a. The first inservice inspection shall be perfonned after l 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 at least-20 percent of the tubes were inspected and-the results were in the C-1 Category or if at least 40 percent of the tubes were inspected and were in the C-2 Category during the previous inspection, the next inspection may be extended up to a maximum of 30 months in order to correspond with the next refueling i H d CALVERT CLIFFS - UNIT 1. 3/4 4-11 Amendment No. 188 .l

q i 3/4.4 REACTOR 00LANT SYSTEM SURVEILLANCE REQUIREMENT 5 (Continued) outage if the results of thi two previous inspections were' not in the C-3 Category. However, if the results of either of the . previ.ous two inspections were in the C-2 Category, an engineering "d'? assessdent'fthalPbetptrfonned before operation beyond 24 months and shall provide assurance that all tubes will retain adequate ~ structural margins.against burst,throughout normal operatino. transient, and accident conditions unt'l the end of the fue' cycle or 30 months, whichever occurs first. If two consecutive inspections.sfollowing-service under. AVT conditions.. net including the preservice inspection, result in all inspection results. falling into the C-1 category or if two consecutive inspections - demonstrateethat previously observedudegradation has not continued and no additional degradation has occurred, theinspection intervalreay4be,ontended.<to.a maxings-of once per+ 40 months.-- ..,., m Is -f the inservice inspection of a steam generator conducted in .+1 b. , ' ;.w accordance with Table 4.4-2 at 40-month intervals fall in -~ Category C-3 the inspection frequency shall be increased to at least once pere 20 mont A The,increas4 A inspecties fre shallaapply untti..the. subsequent,inspectionsasatisfy,the.quency-ce,iteria of Spectfication;4.4.5.3.a; thkin6erva34mayethen,be extended to - a maximum of once per 3Gror 40rmonths, amappl.icabile... e. m., r.. ., o.,m..r ... r n ...w... c. Additional.gunscheduled.inserviceJinspections.shalbbe perfonned on each stoomrgenerator insaccordanse with:the first sample ins.pection specified in Table 4.4-2 during the shutdown subsequent to gny of ths followingsconditionsw.cu,a om g, w-,c .. m .s 1. Primary-to-secondary tube leaks (not including leaks originatings,from tube,tottuhe sheet,,, welds).1A excess of the n 11mits4f.Specif.ication 3.4.6.2.m 2. Aseisaic..occurrensegreateQban.tha0perating.B4 sis Earthquake ~ '..m-m m, . L,..u - 3. A los's-obblant accident requiring actuation of the ... engineered safeguards, or - ' ,,,,s., 4. A main steam line or feedwater line break. d. The provisions of Specification 4.0.2 do not apply for extending the frequency for performing inservice inspections as specified. in Specifications 4.4.5.3.a and b. l CALVERT CLIFFS - UNIT 1 3/4 4-12 Amendment No. 188 l .~ .----------O

q = ( 3/4.4 REACTOR 00LANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Accentance Criteria. ~e a. As used..inithis Specifications %. & & n. m.eme c... e. mtt.mf .h.w n uwv.nra n,. 1.wherfection meansaan exception, to' the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below. u 20% of the-nominal tube,wa11tthicknesse 66 detectab % may be. considered ascimperfections, 2.

Dearadation means a service-induced cracking,

wastage, wear.. or general corrosion occurring on either inside or outside of .~ a tube. 3. Dearadeel Tube means a tube containir.g imperfections t 20% of the nom <nal wall thickness caused by degradation. 4. 4 Dearacation means the percentage of the tube wall thickness affectec or removed by degradation. 5. Defecu means an imperfection of such severity.that'it exceeds. the p' ugging Ilmit. A tube containing a defect is defective. Any tube which does not pemit the passage of the eddy-current inspection probe shall be deemed a defective tube. 6. Pluacing Limit means' the-imperfection depth at or beyond which tie tube shall be removed from service because it may. become unserviceable prior to the next ins mctton and is equal to 40% of the nominal tube wall thic(ness. 7. Unserviceable describes ~ the condition of.a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam ' ine or feedwater line break as specified in 4.4.5.3.c. above. 8. Tu m "nspection means an inspection of the steam generator-tum " rom the point of entry (hot leg side) completely around-the U-bend.to the top support of the cold leg. b. The steam generator shall-be detemined OPERABLE after completing the corresponding actions (plug all tubes exceeding the. plugging Itait and all tubes containing through-wall cracks) required by Table 4.,4-2. CALVERT CLIFFS - UNIT 1 3/4 4-13 Amendment'No. 188 l

m _a. (- 3/4.4. REACTOR COOLANT SYSTpt $URVEILUUICE REQUIRDIENTS (Continued), -. 4.4.5.5 Reports ..p,. ',Following each, inservice inspectiori of steam generator tubes. the ?'. rtdabFof thbes plugged in each steam generator shall be reported to the Commission within 15 days pursuant to Specification 6.9.2.. b. The complete results of the steam generator tube inservice inspection.shaniber inoloded in the Annual Operatin the period in which this ins >ection was completed (g Report for pursuant to Specificetten. 6<.9.1.5sh)*. T11s report shall include: 1. Number and extent,ofriuham4nspected.xmt.m. 'a... 4 v.,

e....,9,o. w

,,n non .m., ca.y te 2. Location and percent of wall-thickness penetration for each indication of an imperfection. .,.......... m a ..,,4 mm ,,. c., 3., Identification of tubes plugged. - t..a ;, .. e. c. Results of steam generator tube inspections which fall into! ...y Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hours >rior to resumption of plant operatione. 4heuunttteshiakipuup of ttisemport: shallme provide a description oibinvetetystionanoonducted te,detemine-causeeof the tube degradation and corrective measures taken to prevent recurrence and shall~be submitted within the next 30 days pursuant to.Speciff cation,6.9.2. w,um nwav w,.r r m trar.4ne w e 9 s u m au j t: y .. a. 1 ,. n ;,,in,,ny, . e.m. araumnetm, ~.. i l CALVERT CLIFFS - UNIT 1 -3/4 4-14 ' Amendment No. 188 l~ . s m r-y,_.m,,,

t 9 TABLE 4.4-1 R 4: G g MINIORBI 1585ER OF STEAll GENERATORS TO BE IllSPECTES DURING IllSERHCE INSPECTIOli p-g m 4 7 -Preservice Inspection No Yes ~ g m g No. of Steam Generators per Unit Two Three Four Two-Three - Four S U ' First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections One One One* One' y l l [ i j.~i ~ TABLE NOTATION: The inservice inspection may be limited to one steam generator on a rotating schedule encompasshig. 1 3 N % of the tubes (where N is the number of steam generators in the. plant) if the results of the first or previous inspections indicate that all steam generators are perfoming in a 11_ke manner.. g Note.that under some circumstances, the operating. conditions in one or more steam generators may be ~ l 5 .found to be more severe than'those in other steam generators.- Under such circumstances the sample sequence shall be modifled-to itspect the most severe conditions. ]

{

l. ,2 The other steam generator not inspected during the' first inservice inspection shall be inspected. l g The third and subsequent inspections should follow-the instructions described in 1 above. 3 Each of the other two steam generators not inspected during the first inservice inspections shall be E inspected during the second and third inspections. The fourth and subsequent inspections shall follow the-instructions described in 1 above. r

..g------ ~..: I' p TABLE 4.4-z w r a STEAN GEMBATM TWE M5PECTIM j y I 151 i __ d G;WiXIIENE ZIm L__ d IgortLilUn Jgu i._ d Igo rLLI HNI as 4 h sample 512e Result Actlen Requires Result Action Nequired Result Action llequired S -5 R mininun of 5 luDes per. c-1 Mene - WA IIIA MA - N/A,e I 3

5. G.

E-g rius eefective tunes ans c-1 none n/A wA. 6 3 q iltspect additional'25 c-z ring eefective c-1 uene :; n } tubes in this,5. G. ts&es and inspect y c-Z Flug Strective 8 cz f-additional 45 tubes tubes - 11 C ^ j in thi,s 5. G.. Q J i g c-3 rerform acuen for r C-3 result of i w t i ~ first sample-e c-3 ; rerform action for i: 1 g C-3 result,of first N/A WAj ; g sample c-3 Inspect all tunes in AtI other +y this 5. G., plug-

5. G.s are; None WA N/A,u defectjve tubes and C-1 J

t w 1 inspect:25 tubes la each i l l other 5. S.'s " t s r i-E 24 hour verbal : 4 i some s.s.s -. rerform action for n/A n/A notification to NHC'with C-2 but no' C-2 result' of written follauu 'P< edditionali ' second sample i pursuant to" ' p- ~ nasin enai. saspect alt. tunes

5. G. are C-3 5pecificatt.on 6.9.2.~

( jc ^

5. G. is C-3;s fit each 5. Gc and-2 F. i

~ 3f ta6es. 24 hour . WA; oL .r.

: plus defective ~

N/A ver6al notification c 4i : 9.: t to IRC with written r .i i follemuy pursuant

g y fftcat}em 15 2j s-

-c F S 3 h Where N is the number of steam generators.in the unit [, and n is the number of steam generators inspected during an-inspection -*t' _.j f

- - =.. - yj 7 3/4.4 REACTORC0%ANTSYSTEM l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakaae Detection Systems ma ,q eg LIMITINGCONDITIONFOR'0PERATIONa3iNptrUrmi 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall; beLOPERABLE: a. A Containment Atmosphere Particulate Radioactivity Monitoring

System, b + The Containment Sump Level Alana System 'and A Containment Atmosphere Gaseous Radioactivity Monitoring System.

c. APPLICA8ILITY: ' MODES 1, 2, 3 and 4. ACTION: i a. With only two of the above required Leakage Detection Systems OPERABLE, 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 Particulate Radioactivity Monitoring System is inoperable; othenvise be in at least NOT STAMBY within the next 6 hours and in COLD $NUTD0lAl within the following 30 hours. j l b. With only one of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 7. days provided-that: 1. Grab samples of the containment atmosphere are obtained and analyzed at least once per 12 hours, and .l 2. The Reactor Coolant System water inventory balance of-Surveillance Requirement 4.4.6.2.c is perfonned at least once per 24 hours. Otherwise be in.at least NOT stale 8Y-within the next 6 hours and in COLD $5UTDolel within-the following 30 hours. - i CALVERT CLIFFS - UNIT 1 3/4 4-17 Amendment No. 188 l: 1

4 i 3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: a. Containment. Atmosphere.Gaseouseand' Particula 4Q40nitoring Systems-performancrof:CIUUglEL' CNECK CRAlelEk CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and ,;;g;jf. -,., e, s o,,, b. Containment Sump Level Alam System-performance of CIUUglEL CALIBRATION at least once per 18 months. CALVERT CLIFFS - UNIT 1 3/4 4-18 Amendment No. 188 l

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE \\ Reactor Coolant System Leakaae a, 0,wa ; i,4 4, .m LIMITIM COMITION'FOR OPER TION ~ ~ ~ l 3.4.6.2 ~ Reactor Coolant System leakage;.shall.be Ifmited to: a. No PRESSURE 5005ARY LEAKAGE, b. 1 GPM WID,ENTIFIED LEAKAGE, c. 1 GPM total primary-to-secondary leakage through all steam generators and 100 gallons-per-day through any one staan generator, and e d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System. APPLICABILITY: MODE 5'1/ 2, 3 and 4. ~30'a '~ . c.. n. ACTION: r gant... c a. With any PRES $URE BOUNDARY LEAKAGEL beitn at'least'N00 STANDBY ] within'6' hours'and in COLD SNUTDOW within the following 30 hours. . mev t.a uilit - b. With any R6 actor Coolant: System leakage 1 greater thaniany one of - the above 1imits, excluding ~ PRESSURE B00 M ARY LEAKAGE. reduce the leakage rate to within limits' withins 4'Notts or be in at least-H0T STAM5Y within the next 6 hours ~and in COLD SMTDOW within the following 30 hours. c<. ,c ry. nn, hf ',T h uikf 2 rb SURVEILLANCE REQUIREN GTS 4.4.6.2 Reactor Coolant System leakages shall' be'demonstr&ted to be within each of the above limits'by: a. Either:.C 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' j accordance with the ACTION requirements of Tecinical i Specification 3.4.6.1. 1 CALVERT CLIFFS - UNIT 1 3/4 4-19 Amendment-No. 188 l

5 3/4.4 REACTOR C00LAlli $YSTEM SURVEILLAllCE REQUIREMDITS (Continued) .b. Monitoring the containment sump discharge frequency at least once per 12 hours, when the Containment. Sump Level Alam System is OPERA 8LE, a ~,e . c. .<...,y -.,., g., c. Detemining Reactor Coelant System leakage at least once per-72 hours during steady state operation and at least once per. 24 hours when required by.ACTI0li 3.4.6.1.b. except when operating in the shutdown coolir.g mode, and s .m y;w, d. Monitoring the reactor vessel head closurW: seal Leakage Detection System at least once per 24 hours. .ra: m y ;,.it; w t;;;'.9 m en m c, ( y .hyn 9~' '*t Oe +, + -_-( l.

R *. 3 \\ 1

? f3s uMt ' eutdikkt > w c...

utitt*3tr -@itM90H ......., s 4. s, ; u + m...;TG 43 i... i,,.3 ,..n a : 4 u ...I 1 -i 'k 1 -CALVERT CLIFFS - UNIT 1 3/4.4-20 Amendment No. 188_ l

wi c' 3/4.4 REACTOR C00UWii SYSTEM 3/4.4.7 CHEMISTRY LINITIM 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. ACTION:- MODES 1, 2, 3 and 4 With any one or more chemistry parameter in excess of its Steady a. State Limit but within its Transient Limit, restore the' parameter to within its Steady State Limit within-24 hours or be in at least NOT STANDSY within the next 6 hours.and'in COLD $NTD0lAl within the following 30 hou'rs. gj-r.cc.4.m e b. With any one or more chemistry parameter iMeacess of its Transient Limit, be in at!1 east NOT'5TASSYMthin 6 hours and in o "j COLB $NTDotAl within the' following 30;houg;g _... gg ']" [~~~ T[ MODES and 6: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to 5 500-psia, if applicable, and perfom an engineering evaluation to determine the effects of the out-of-limit condition on-1 the structural integrity of the Reactor Coolant System; detemine l that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia i or prior to proceeding to MDDE 4. ) SURVEILUUICE REQUIREMENTS 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. CALVERT CLIFFS - UNIT 1 3/4 4-21 Amendment No. 188 l

. v'

9

) 3/4.4 REACTOR C0OLANT SYSTEM TABLE 3.4-1 REACTOR COOLANT SYSTEM - CHEMISTRY-LIMITS w .4. 6-A,, STEADY STATE ~ TRANSIENT PARAMUER LIMIT LIMIT DISSOLVED OXYGEN * $ 0.10 ppe ~51'00 ppm CHLORIDE $ 0.15 ppa-5 1.50 ppe FLUORIDE 5 0.15 ppe - -..$ 1.50 ppm ,;. e co,n a 4 G'it'jNe .....,m..w.. 4., J.. ,s mas... .7 + .-.y+ a.. . :_,.-q 7_ . _. r. _.. - _. w. .- 3f .g. f .t .Q s.- o,,. - n ,.4+...n. .r--.e,,,,, 6 e 4 ow. -.w ..4s.... ,4 m...- .J__ -. WeAh ~ J[ u, % + :;'at .. _, g; w**m my n :.x ,_.. :. i... _,_ ... _ _ _,. ~.... + Limit not applicable with Tom 5 250'F. ' CALVERT CLIFFS - UNIT 1 3/4 4-22 Amendment No. 188 l

3 3/4.4 REACTOR C00UWIT SYSTEM TABLE 4.4-3 REACTOR COOLANT SYSTEM . %. a, pg pCllEMISTRY LIMIIS SURVEILLANCE REQUIREMENTS ~ c:: - = =.. PARAMETER ANALYSIS FRE00 ENCOR. DISSOLVED'0XYGEN* , At least once'per 72. hours ~ I^t least de per 72 hours CHLORIDE FLUORIDE At least once per 72 hours u...neo.a.-a.syos _ a sne y s o a .s-. a 9 . :n ~ ut A n e u_:,n.n ,= qr; y._rg . ~o ,v. g m 4 *f .s,-r ..p. ar) m .? we r gie,c + ..m va3 mm. 1 ce.., q. u . a:. .s .e r.,u m ma.,,, a. es : g $ 8 N' Q' '1 - .,,. r v4;Q. o r, #ta

F

..g y r. i +

,4; G.,' ;4 3 ,e,

(pg*

p q .t.'*' .F "e

s.,I 3.,.

.- ( f.. ._ -.m ..r 3 - n.e r u, q 4;if l l Not required with T. .< 250 F. f CALVERT CLIFFS - UNIT 1 3/4 4-23 Amendment No. 188

l

- _ ~ a 3/4.4 REACTOR CMLANT SYSTDt 3/4.4.8-f " SPECIFIC ACTIVITY LIMITINS COMITION FOR OPERATION -. e. 3.4.8' The specific activity of the primah coolant khalk be limited to: a. < 1;0 pCi/ gram D0$E EQUIVALENT I-131; and w,m -v -u .es. 1

b. 1 100/l C1/ gram.

'r> APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*: a. With the specific activity of the primary coolant > 1.0 pCi/ gram 005E EQUIVALENT I-131 but within.tw allowable limit-(below and: to the left of the line) shown on Figure 3.4.8-1, operation'may l continue for up 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 pC1/ gram DOSE EQUIVALENT I-131 for more than 100 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least NT STA38Y with T.,,'< 500'F within 6 hours. c. With the specific activity of the primary coolant - l > 100/l Cf / gram, be in at least NT STANDSY with T,, < 500*F' l within 6 hours. N00E5 1, 2, 3, 4 and 5: i d. With the specific activity of the primary coolant > 1.0 pCf/ gram DOSE EQUIVALENT I 131 or > 100/E pCf/ gram, perform the sampling. and analysis reavirements ef f tra 4 alof Table 4.4-4 until. the-specific acthity of the pi: rah Wolant.is restored to within its limits. Whenever the specific activity of the primary-coolant exceeds' 1.0 pC1/ gram DOSE EQUIVALENT.I-131 for in. excess > of 50 hours for one continuous time interval or 5 percent of the. unit's. total yearly operating time pursuant to ACTION a. above, a Special Report shall:be prepared.and submitted to the Consission ' pursuant to Specification 6.9.2 within the next 30. days. This W1th T.,, >_ 500 F. CALVERT CLIFFS - UNIT 1 3/4 4-24 Amendment-No..188 l~ u.

..q 1 1 i 3/4.4 REACTOR COOLANT. SYSTEM i LINITING CONDITION FOR OPERATION (Continued) 1 report shall contain the results of the specific activity 1 analyses together with the following information: I i 1. Reactor power history starting =48. hours prior:. tog the,first sample:fn whick=thesifini19issF"?~-~'T'E:.f:.si _ ===. m, _ r __ y m.. u =,= 1 ~. ~ x.m. t 2. Fuel bu.sup.bilcoMEN~GEa:s=;=== = WE;=3 ~

3. CleanAffo fijhfr

_ _ mw g _ =a .. m. s =.-. _ - z.y . --- sample in. wh(ickthe:i*itiMiliq[@' ~ ""~f--~'i:MM#dti@first llailt c - ~ Jan ~~ m.TT 'C;.-L~=2:x. :RiW- = i,== & - &===-?

4. -HideryifEAifftNTe~9'[qiUM&14KgjfMBMON:A$)Wed, hours prief to;thej:ffrstssampTeRf1%G -- _.Gai.miDBaab2 L

u andF =^2R#ss=miETt%=t?=w' L --:2 21 - p~ g.< p - qd=vh-. --- =Fs=4 ; - - - - -- _,,2..i 1 '.. " :. :.=.. ;., w.::-~r-- S." That timaJiustfWwheerttheEL[wn :;~ haftliEp, rimary - - bg . ";. :. m -- --.~==a.

u.I""E'"C N 91k 8 dk C

m_. _ W._ 4. 4.. =,, ~ a. =- ---. _ =. ~ - ' ~ ~. ~, ;((;Z"~AL";f CQ%""""Y _ -- '~ .,- t'El-.2 '--'~~~ w =i ", -.[,; . 2 - .. ;: ;;:-- -{w, -

+-- _

m: :,;;_,

- m ~~ ;- -g,_-,..._.,_:. __ 3k-5~ SURVEILLANCCSM--- M. - C ~4 = .. a= =. a.- -,

4.-_n.-.

The specificcactWiti=egf6$jirWINiiff s ~ nt sh~allb,. 1

-**-~*L--

$ Je. ...K m 4.4.8 detennined to ~ s ~ lv*Ai be within the limitsiby~ pes; forum #ct~e@heganH$ngy:d'apa% s program of Tab 1e 4.4-4. = =2n= a =- u=ra =E@;_r===c.=. c=.-g wi ., p:..!]:L

A =

. _= =,;.=.,.:@-- z,.- n s: -..=:. :===- - - - m ' " *;'""'- ="@""t- ;;-.6;2_, ---. *WJ"O I' Q _., ,'_~~g:g%:l;;*a-j .':E % ~i. min ^_W :: M y w s.0..>.w.i.vtg..n.eep ++. - -... -..., un w = =u.... g. = _.g ..a = a _ - =.. .. ge _

r..:v=

v w -.m-- 2=

=y

-- = 1-- iim.A...g6.n-M,39;#- . - - ----==w=ra 5 ~ w ,1;;c:a =m r - s =F.n:= w q _nr u = aa..-:.;=....:.- J.*f='=-'=t=~.=-- w.~~~ ~f=.x* m - - g..:.. - a.g:9.== * $--J.:..:.: ;t.:- ;:=.;i;i=_ ----g=~T;5dL;:]c ..... w... o a D W88 a,... - d r ?.' tr.W., ,.t'.. ..o-a 2 . n. s.p. t. a .e. ~ a 4 3. w/ CALVERT CLIFFS - UNIT 1 3/4 4-25 Amendment No. 188 l

9 TA8LE 4.4-4 w G g PRIMARY C0OLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS P9064AM n 9 TYPE OF MEASUREMENT i $ AMPLE AND ? MODES IN lAIICM SAMPLE AND h 'd AND ANALYSIS AMETSIS FREQUENCY i ANALYSIS REOUIRED .m. n w M/, ~ 2, g, p[4jf!l 1. Gross Activity Determination dy . Y;, 8 f MttlMQ innci per M : 2: W il c 5 Ohkk p [%p,@y;D dnlri;'5yiT 2 i YD.6 ' % d $ f 'l'E -i d kE'.q. 2.'IsotopicAnalysisforDOSEERl"gkag [ 1! g,

iif' i;tk l9 i

}dO! e 4 2f h$ i' 'C 4 ihjliiW@. M ;hj lip Ici 4 lL W i I-131 Concentration .i g C a. ?

hf l

!U! l i o 3;. Radiochemical for E Det4 min j l i 93 g Q[IM,,.'.

4. cIsotopic Analysis for Ifdine ykl 1p 3

, S' ' ' l.gr y q g pgA} .J qg g 1 deiwar,;th i' j iha# b I-131,1-133, and I-135- 'l i j.l J"!L j

q.,ll ;hi.

fl w 2 5 i i _10 gligipJirr"l+ p k h Iswhljh :pd f. l, . gren'! ;! phi d i !!. 44 a i' .[ J. t [l, i i 0' 3 G,i 'l.. ! eb '^- 4 1,4,o 'Vg "E ,l s g:jt["[ ) -'y-.ili i,lj 3; ([ l

j L $.

, 'i F 3 bkMig.(O[pl' k N(di ~i yg NIf S[ I I M 1 h li' ' b 'h i .ll91 I I l i g p r.ikk jj ..tr i! Win l io 4 J si ? ' p 4 j; a 'l Llll .pl$$}. pia i j.4 g p j1 "*n p j unii u Jal chaupit ,qeeda,l-.. ( f - j d f i f g i l' l T ! d 31 i <lmitt411 b o i } ',il l'(l f' j 1 U! if3 fi llllllll5l01,I j'IN 1 l h l [ k

l h 'll llllll Lilil t

t r E [ ( I {w M inll 4 : ' 4 l Ri i i ( .p l rj. A ' }! J - p p = 4 -.i ! L.u. k[.jjl:, I ' T I, l bf E ~ l a I l t r h1 = .s I h isppp 6t 1: t ll@p;p.u. j , !pr sq,tj [11mily. Until the specific activity of: thE Pdm 47 Goolantdi:' fi l tW-vicen y,9! _ <. Am i; SampletobetakenafteramiN1 Sun'of'24FPDand2'% f'POWE(OPERA ONW'e il5 sed since reactor was last subcritical for 48 hours *or longer. me:a

p 3/4.4 REACTOR C0OLANT SYSTEM .-,,..;,.s., o k l \\ e A ' k \\- . t m u,, GM . r i r ' t i. 4.,- -g y h T ' k r l. L r 30 4 i k -,e-c i h ~! UNAOCEPTABIA " :qql i g w OPERATION Jojt: - ( a r o 150g i i 6_ gr, h c mu;. e j c '

m, t!O r

h-e .c 9t ; a7T 0 0n i i c n a t. r y t n' -( s.,e a '~ wino g sn 100 i Ls a h[LL '."

M m _.

3 j C "' W 6 OPERATION 2 - wr -nper.u s ; %g ne - t s -4 .i i i. ~, 4 ,, e 4 I 0 20 30 40 ~50 80 70 80 90 100 P3RCENT OF RATEDTHERMAL POWER FIGURE 3.4.8-1 D05E EQUIVALENT I-131 PRIMARY C0OLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITN THE PRIMARY COOLANT $PECIFIC ACTIVITY

1,0uC1/ HAM DOSE EQUIVALE N I-131 CALVERT CLIFFS - UNIT.1 3/4 4-27 Amendment No. 188-l-

~_ -...-. .,~ \\ c-3/4.4 REACTOR C00UUlf 3Y3 TEM 3/4.4.9-PRESSURE / TEMPERATURE LIMITS Reactor Coolant System" > m? .I LIMITIN COW 57188 FOR SPEMFIN 3.4.9.1 The Reactor Coolant"Tyttee (except'the pr'ssurfter)' temperature ' e and pressure shall be limited in accordance with the limit lines shown on Figures 3.4.9-1 and'3.4.9-2 during heatupp'cooldownferftf tilityPand w o* inservice leak and hydrostatic testing with: A maximum heatup '.of: "," ".re" "N5) *a4 a. l o e n..': + c a2 u9i e Maximum Allowable Heatuo Rate " "' RCS Temperature I 30 F in any one hour period 70'F to 164 F 40 F in any 'one' hour ' period '" """> 164*F W 2567 60 F in any one,hou,r per;i.op g Q 256' Q ?" i. [ *" ' .) b.-Amaximumcooldowyf,. ,g Maximum Allowable Cooldown Rate RCS T = erature I. 100*F.,1rt any one,pr. per1od MQ*170*F7*["i ? E..td 20'F in any one hour pef,1odh ',' 4,{184*F~f 10* Fin,anyone.houtpar,14 c. A maximum temperature change ofn5'E. in anyconthourtperiod during hydrqstatic testig., opera [tions; aboy.e system design, pressure,,o n,., % ,, m, 2 n _,w.g c y,., APPLICABILITY:.Jt all times **RABLE Na ireswee :arety : niectico 4

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ACTION: With any.c m nof the abov.ma limits exceeded.n m.er dx a,..restorg ha gpgature m and/or pressure.to within the limit withi'a 34ssinutess,.per m an.. engineering evaluation to detemine the effects of the out-of-limit condition on the fracture toughnestpropecties-of,the, Reactor,, Coolant Systems determine thatithe Reacton Co'lant System remains acceptable for. o continued operations or be in at le'ast NOT STAWBY within the next 6 hours and reduce the RCS T and pressure to,less than 200*F and 300. psia, + .respectively,within7hefollowing30 hours. , wte a : r. ya um w o,,,.i,..,,. ~.: l.CALVERTCLIFFS-UNIT 1 3/4 4-28 Amendeent No. 188 l

3/4.4 REACTORC0OLANTSYSTM SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be detenmined to be within the limits at least enco per 30 minutes during system heatup, cooldown, and-inservice leak t & hydrostatic testing' operations. a. em. ....v ... c, 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 504 Appendix H. The results of these examinations shall be used to update Figures 3.4.9-1 and 3.4.9-2. i... 3 .. c.,, ..... a 5 w..; C' !h 1 .1 - i . c .:. ug o ,,c w... . WRE #0HV m,. - c . u. .v. 'b 'a r W f g ie fit ,e 3: ..:.. e. nr: ,1 4 .s ~, .nort a. e .w '? tr_ ', T yi s ,e

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== 5 E TUP =_ 4-- = g - * *:* 5885MCE NYDRoSTA11c TEST * ~ = = t, L.f..J t 3i lA, _~ _ - - - i ~ h SGMCE .-g:_ I1500 gTEMPERATURE- -+ _ = - g,154*P - COM CIETICAL?! l1000 h 7 LT. ~ [i RCS Tars e. HAlRATE f 'E 5 70*F 70184*F 530*F/1 HR >184*F TO 29%?f,,540*F/1 HR _f = -- -, >s sear > soo*pr1 HR IAAMURS 301.TUPTERF.,79'F t 54 asAxmsuasFnessuns i Pon soc openAT1oss g o 100 200 30s 40o soo e00 i. -'. r- .. n .n. ": F sacATuo AsAcToR coouWTTEMPERATURE T I C An ~ FIGURE 3,4,9-1 CALVERT CLIFFS UNIT 1 NEATUP CURVE FOR FLUENCE :s 2.61 x 10' n/ca' 1 REACTOR C0OLANT SYSTEM PRESSURE TEMPERATURE LIMIT 5 CALVERT CLIFFS - UNIT 1 3/4 4-30 Amendment No. 188 l

f 3/4,4 REACTOR C0OLANT SYSTDI 1 ) naa 'i 'l i. 1-usemence uvonosrAviciust g 'u m ust E SWMCE 4 l comamm j teo*, teos l ~ .r. l100o. i En 1MW. GSBM l >now stoo*m Ha 27 ear vatseap saeam Ha asumammottwtaur 7 ear i naa mam ma yn g gggg SOC ofWIATIoIG o a soo-aon. - ace: 4es.._, son._ soo secocAvsn nuacron coouwt Taspenatuns T ' . a.. i, c i FIGURE 3.4.9-2 CALVERT CLIFFS UNIT 1 C00LDOWN CURVE, FOR FLUENCE s 2.61 x 10" n/cuf REACTOR C00LMT SYSTDI PRE 5W TDIPERATURE LDETS CALVE.tT CLIFFS - UNIT 1 3/4 4-31 Amendment No. 168 l

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE L{tlUj. Pressurizer LIMITIM CoWETIM FM OPERATim 3.4.9.2 The pressurizer temperature shall be4faited to:- J a. A maximum hea, tup of 100*F in any one hour period.. b. A maximum cooldown of 200*F in any one hour period, and A maximum spray water temperature differentia] of 400*F. c. APPLICA8ILITY: " At' all times. ~ "'"'"."t.*- war + x ACTION: With the pressurizer temperature 'limitI"in excess' of;'an'y of the above limits, restore the temperatureito withth therlimitFwithin-30 minutes; perform an engineering ' valuatioii'f6 detensine'the'eWects of-e the out-of-limit condition on the fracture toughness pro pressurizer; determine'that the pressurizer remaffs-acc%perties of.the ptable for e continued operatton or be'-in at least NOT STAM3Y~withih the iiext.6-hours - and reduce the pressu'rizer pressure.to less than~300[psiswithin the following 30 hours. y,,['"',,,f"""',,',,,,, .., i. m

s no, SURVEILLANCE REQUIREMENTS H

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4.4.9.2 The pressurizar temperaturekshall be'detemined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shal' he detemined to be within the limit at least once per 12 hours.during auxiliary spray operation. t.. oa.t.curnemy 4 .P* b CALVERT CLIFFS - UNIT'1 3/44-32 Amendment'No. 188-l-

(" 1 7 7 3/4.4 REACTOR C00fR T SYSTEN 3/4.4.9 - PRESSURE / TEMPERATURE LIMITS Overoressure Prutection Systems ^ LIMITING CONDITION FOR OPERATION 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. Twopower-operatedreliefvalves(PORVs)wjthatrip. setpoint below the curve in Figure 3.4.9-3 with their associated block valves open, or 2. A single PORV yith a trip setpoint below the curve in Figure 3.4.9-3 with its associated block valve open and a l Reactor Coolant System vent of 11.3 square inches, or 4 3. A Reactor. Coolant System (RCS) vent 3 2.6 square inches,.- t 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 The HPSI loop motor operated valves (MOVs)' shall' be. prevented j c. from automatically aligning HPSI pump flow to the RCS by placing

1 their hand switches in pul' -to-override.-

d a 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 RCS and when used, it must be under manual control and one of the following restrictions shall apply: 1. The total high pressure safety injection flow shall be Ifnited to a 210 sps, or 2. A Reactor Coolant System vent of 1 2.6 square inches shall

exist, e.

When not in use, the above OPERABLE high pressure safety injection pump shall have its handswitch in pull-to-lock. APPLICABILITY: When the RCS temperature is $ 365'F and the RCS is vented-to < 8 square inches. When on shutdown. cooling._ the PORY trip setpoint shall be i429 psia. 'I EXCEPT when required for testing. CALVERT CLIFFS ~- UNIT 1 3/4 4-33 Amendment No.-188-

3/4.4 REACTOR COOLANT SYSTEM i LIMITING CONDITION FOR OPERATION (Continued) ACTION: a. With one PORV inoperable in MODE 3 with the RCS temperature < 365 F or in MODE 4, either restore the inoperable PORV to OPERABLE status within 5 days or depressurize and vent the RCS through a > 1.3 square inch vent (s) within the next 48 hours; maintain tee RCS in a vented condition until both PORVs have been restored to OPERABLE status, b. With one PORV inoperable in MODES 5 or 6, either restore the inoperable PORV to OPERABLE status within 24 hours, or depressurize and vent the RCS through a > 1.3 square inch vent (s) within the next 48 hours; and maintain tee RCS in this vented condition until both PORVs have been restored to OPERABLE status. c. With both PORVs inoperable, depressurize and vent the RCS through l a 1 2.6 square inch vent (s) within 48 hours; maintain the RCS in a vented condition until either one OPERABLE PORY and a vent of 2 1.3 square inches has been establi;hed or both PORVs have been restored to OPERABLE status. d. In the event either the PORVs or the RCS vent (s) are used to l mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Comission 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.- With less than two HPSI pumps' disabled, place at least two HPSI l e. pump handswitches in pull-to-lock within fifteen minutes and disable two HPSI pumps within the next four hours. f. With one or more HPSI loo) MOVs' not prevented from automatically l aligning a HPSI pump to tm RCS, imediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or isolate the affecteo HPSI header flowpath within four hours, and implement the ACTION requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as applicable. g. With HPSI flow exceeding 210 gpm while suction is aligned to the l RWT and an RCS vent of < 2.6 square inches exists, 1. Imediately take action to reduce flow to less than or equal to 210 gpm. EXCEPT when required for testing. CALVERT CLIFFS - UNIT 1 3/4 4-34 Amendment No. 188

1) 3/4.4 REACTOR C00UUIT SYSTEM g LIMITIllG CONDITION FOR OPERATION (Continued) '2. Verify the excessive flow condition did not. raise pressure. above the maximum allowable pressure for the given RCS temperature on~ Figure 3.4.9-1-or Figure 3.4.9-2. 3. If a pressure limit w e exceeded, take action in accordance with Specification 3.4.9.1. a h. The provisions of Specification 3.0.4 are not applicable. l-SURVEILLANCE REQUIREMENT 5 4.4.9.3.1 Each_PORV shall be demonstrated OPERABLE by: a. Perfomance of a CIUUNIEL FullCTIONAL 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 PORY is required OPERABLE. b. Performance of a CIUUNIEL CALIBRATION on the PORY actuation channe1~at least once per 18 months. c. Verifying the PORY block valve-is open at least once per 72 hours -l when the PORV is being used for overpressure protection. d. Testing in accordance with the inservice test requirements pursuant to Specification 4.0.5. j The RCS vent (s) shall' be verified to be open at least once 4.4.9.3.) hen the vent (s). is being used for overpressure protection. per 12 hours w j 4.4.9.3.3 All high pressure safety injection ~ pumps, except.the above OPERABLE pump, shall be demonstrated inoperable :t least once' per 12. hours. by verifying that the motor circuit breakers have been removed from theiri electrical power supply circuits or by verifying their discharge valves are'-- locked shut. The automatic opening feature of the high pressure safety-injection loop MOVs shall be verified disabled at least once per 12 hours. 'l The above OPERABLE pump shall be verified to have its handswitch in pull-to-lock at least once per 12 hours, l Except when the vent pathway is locked, sealed, or otherwise secured in the open position, then verify these vent pathways:open at least i once per 31 days..

i CALVERT CLIFFS - UNIT 1-3/4 4 Amendment No.:188 j

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it ; i i: il l ! 4 l iji' il i.Iii N$3 fil! !!- -lii fli! i srow isee Psaa ifi ih [fil 1['lli !! ! ilii 3dilly !!!: in ?li iii! .m a. 3l1M ii ' 97Ij!! li!{ ii!! !i[ i y 4 4 6!! ith i 4 !il! i - ! t i l Il ii Ill i i i!ii liil ft: !.li !!! n:; I!I-an i l il l I 'il! iil !!i; illi ! i. $iI !!' iiF !!!' ili illi l a i I i i l a $! !I I l 9 -! i ll l ll i!!! 14l ll!- :i i l Il l ! I . I: !0! li! @ iV! iiii !I!! !!! Ili i J i i : 1 l t4 ilE !!!! liil jiii lil- .ll . I i i 0 100 200 300 400 50s 600 ACTUAL. REACTOR COOT. ANT TEndPERATURE T ' C FIGURE 3.4.9-3 CALVERT CLIFFS UNIT 1 FOR FLUENCE s 2.61 x 10" n/cu' MAXIMUM PORV OPENING PRESSURE vs TEMPERATURE CALVERT CLIFFS - UNIT 1 3/4 4-36 Amendment No. 188 l

d 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1. 2 and 3 Components LIMITING CONDITION FOR OPERATION 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 ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations. b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F. c. With the structural integrity of any ASME Code Class 3 component (s) not confonning to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service. d. The provisions of Specification 3.0.4 are not applicabla. SURVEILUUICE REQUIREMENTS 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 4-37 Amendment No. 188 l

I.q'- 3/4.4-REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) in addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the reconnendations"of Regulatory Position C.4.b of Regulatory Guide 1.14. Revision 1 August 1975. w,c 4.4.10.1.2 Auamented Inservice Lnspectier Procram for Main Steam and Main feedwater Pipina - The uneneapsu'ated welds greater taan 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 9djoining 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 aed-Pressure Vessel Code Section XI,1983 Edition and Addenda through-Sumner. 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 s inspection schedule for the repaired welds shall revert back as if a new interval had begun. If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall-also be inspected.- CALVERTCLIFFS-ONIT.1 3/4 4-38 Amendment No. 188 1.

o 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.11 CORE BARREL MOVEMENT LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level. APPLICABILITY: H0DE 1. ACTION: a. With the APD and/or SA exceeding their applicable Alert Levels. POWER OPERATION may proceed provided the following actions are taken: 1. APD shall be measured and processed at least once per 24 hours, 2. SA shall be measured at least once per 24 hours and shall be processed at least once per 7 days, and 3. A Special Report, identifying the cause(s) for exceeding the applicable Alert Level, shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days of detection. b. With the APD and/or SA exceeding their applicable Action Levels, measure and )rocess APD and SA data within 24 hours to determine if the core aarrel motion is exceeding its limits. With the core barrel motion exceeding its limits, reduce-the core barrel motion to within its Action Levels within the next 24 hours or be in HOT STANDBY within the following 6 hours. c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. CALVERT CLIFFS - UNIT 1 3/4 4-39 Amendment No. 188 l [

v. L c e ) 3/4.4' RQiCTOR C00LANT SYSTEM SURVEILLANCE REQUBEMENTS 4.4.11 Routine Monitorino Core barrel movement shall be detemined to be-less than the APD and SA Alert Levels by using the excore neutron detectors. to measure APD and SA at the following frequencies: a. APD data shall be measured and processed at least once per ~ 7 days. b. SA data shall be measured and processed at least once per 31 days. 4 oc .s > J e,. u-4 h CALVERT CLIFFS - UNIT 1 3/4 4-40 Amendment No. 188 l

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.12 LETDOWN LINE EXCESS FLOW LIMITING CONDITION FOR OPERATION 3.4.12 The bypass valve for the excess flow check valve in the letdown-line shall be closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the above bypass valve open, restore the valve to its closed position within 4 hours or be in at least H6T STANDAY within the next 6 hours and in COLD SHUTDOWN within the ~following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.12 The bypass valve for the excess flow check valve in the letdown line shall be determined closed within 4 hours pf or to entering MODE.4 i from MODE 5. . :a e 'k4 i ~! \\ .CALVERT CLIFFS - UNIT 1 3/4 4-41 Amendment No. 188 l

w. i w 3/4.4 REACTOR COOLANT SYS1EM 3/4.4.13 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION.,,, 3.4.13 One Reactor Coolant System vent path consisting of two' solenoid r valves in series shall be OPERABLE and closed at each of the following locations: .. r a. Reactor vessel head -b. Pressurizer vapor space j APPLICABILITY: MODES 1 and 2 ACTION: a. With the reactor vessel head vent path inoperable, maintain the inoperable' vent path closed with power removed from the actuator 1 of the solenoid valves in the inoperable vent l path,' and: 1. If the pressurizer vapor space vent path is also inoperable. restore both inoperable vent paths to OPERA 3LE status within - 1 72 hours or be in'at least NOT STANDBY within 6 hours, or 1 '2. If the pressurizer vapor space vent path.is OPERABLE, restore the inoperable reactor vessel head vent path to. OPERABLE status within 30 days or~be in at least NOT STAN0BY'- 'l within 6 hours. ~ H ~ b.~ With only the pressurizer vapor space vent path inoperable, Itaintain the inoperable vent path closed with >ower removed'from thrvalve actuator of the solenoid valves in tie inoperable vent 1 patM and: ~ H .w w. v.- ~ } 1." Verify at least one PORY and its associated flow path is OPERABLE within 72 hours and rsstore the inoperable pressurizer vapor space vent path ~to OPERABLE status prior to entering MODE 2 following the next NOT SHUT 00WN of sufficient duration, or n 2. Restore the inoperable pressurizer vapor space vent path to-1 OPERABLE status within 30 days, or be in'at least NOTz l STANDBY within 6 hours. ] c. The provisions of Specification 3.0.4 are not applicable. 1 1 CALVERT CLIFFS - UNIT 1 3/4 4-42 ~ Amendment No. 188 l

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.13.1 Each Reactor Coolant System vent path shall be demonstrated OPERABLE by testing each valve in the. vent path per Specification 4.0.5. 4.4.13.2 Each Reactor Coolant System vent' path shall'be demonstrated OPERABLE at least once per REFUELING INTERVAL by: a. Verifying all manual isolation valvss in eacit vent-path are locked in the open position. b. Verifying flow through the Reactor Coolant System vent paths with the vent valves:open. ..hl1 . c .,s.o ..f.. 'f ..y. 3 ,,e q' I ret ,)

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e:. .n r -;g it. t .a 4" 1 -l ] l CALVERT CLIFFS - UNIT 1 _3/4 4-43 AmendmentLNo.'188-l 1

3 3/4.4 REACTOR COOLANT SYSTEM BASES 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 limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power-operated relief valve or steam dump valves. Demonstration of the safety valves' lift 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 capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path. However, the PORVs and their circuitry do not perfonn a safety-related function and, therefore, do not need emergency power.as part of their operability requirements. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORY to OPERABLE status. Power is maintained to the block valve when it is closed to control excessive PORY seat leakage. This allows the PORV and block valve to remain 0?ERABLE should the PORY be needed to control reactor pressure and facilitate decay heat removal during certain accident conditions. The removal of power from a closed block valve for a PORV inoperable due to causes other than excessive PORV seat leakage provides additional assurance that the block valve will not be inadvertently opened when the condition of the PORV is uncertain. RCS temperature, as used in the applicability statement, is detennined as follows: (1) with the RCPs running, the RCS cold leg temperature (Tc) is the appropriate indication, (2) with the Shutdown Cooling System in CALVERT CLIFFS - UNIT 1 B 3/4 4-2 Amendment No. 188

3/4.4 REACTOR COOLANT SYSTEM BASES operation, the shutdown cooling temperature indication is appropriate (3) if neither the RCPs or shutdown cooling'is in operation, the' core' exit thermocouples are the appropriate ' indicators of RCS teinperature. The testing for transferring motive and control power for the PORVs and block valves from the normal to emergency power bus is done under Technical Speci fication 4,8.1.1.2.d.3.,, 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer with the level as programed ensures that the RCS is not a hydraulically solid system and is capable of accomodating pressure surges during operation. The operating band for pressurizer level bounds the programmed level and ensures that RCS pressure remAJns within the bounds of'an-analyzed"c~endition during the'eicessiVe charging event as well as during the limiting dearessurization event Excess Load. The operating band also protects t1e pressurizer code safety valves and power-1 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;pressurizepcode 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 offsite' power condition to maintain natural circulation at NOT STANDBY. 3/4.4.5 STEAM GENERATORS e e The Surveillance Requireme'nts for inspectiontof the steam generator tubes ensure that the structural integrity of this' portion of the.RCS will be maintained. The program for inservice inspection of' steam gamerator tubes is based on a modification of Regulatory Guide 1.83 Revision 1. Inservice inspection of steam generator tubing is essential in order to 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. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. An engineering assessment of steam generator tube integrity.will confirm that no und"e risk is associated with plant operation beyond 24 months of the previous steam generator tube inspection. To provide this confirmation, the assessment would demonstrate that all tubes will retain CALVERT CLIFFS - UNIT 1 B 3/4 4-3 Amendment No. 188

W* x h' l 3/4.4 REACTOR C00UUIT SYSTEM ~ BASES adequate structural margins against' burst during all nomai[ operating, transient, and accident conditions until the:end of.the. fuel, cycle. -This-evaluation would. include the following elements: t + .A ; J 1. An assessment.of the flaws found during the, previous. inspections. 2. An assessment of the-structural margins relative to the criteria oI Regulatory Guide 1.121." Bases for Plugging Degraded PWR Steam Generator Tubes," that can be expected before the end. of,the fuel.. cycle or 30 months,-whichever comes<first. o e s m-3. An update of the assessment model','a[ app M a $,b[s'ed on. comparison of the predicted results of the steam generator' tube integrity assessment with actual inspection:results, from, previous-inspections ~ 'w T y t, 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 h negligible corrosion.of the steam generator tubes. If the secondary-coolant chemistry is not maintained within these limits, localized corrosion.mayilikely result in. stress corrogion cracking.,fhe. extent of: cracking "during plant o'peration would be limited by the:11m' tation of steam: generator tube. leakage between the. Primary. Coolant System, and the Secondary .j Coolant System (primary-to-secondary leakageM.;1 gallon m minute, total).- Cracks having a primary-to-secondarygleakage.less than~tgis limit during operation will have an. adequate margin.of safety to withstand.the loads j imposed during nonna? operation 'and by postulated accidents. OperatingL o plants have. demonstrated that primary-to-secondary leakage of;1 gallon d perminute can readily be ~ detected by radiation monitors of steam generator blowdown... Leakage ir, excess.of.this limit will; require plant shutdown and an unscheduled insp9ction, during which the leaking tubes will.be located and plugged. Wastage-type defects are E likely M prope U.hemistiryItreatment of the f secondary coolant. However, even if a defect should develop' in. service, it will be found dur,ing scheduled inservice. steam generator tube. examinations. Plugging will be required for all tubes with imperfections 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 inspection ' fall. into Category C-3, these resu'ts will'be promptly reported to the - Connission pursuant to Specifications 6.9.2 prior the resumption of plant operation. Such cases will.be considered by the Consission 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. CALVERT CLIFFS - UNIT 1 B 3/4 4-4 Amendment No. 188 l

i P 3/4.4 REACTOR C0OLANT SYSTEM BASES 3/4.4'.6. REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakane Detection Systems w 1 m wnc ' i ) y ,m. 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.>- ^-~~ m.e i e f.y 3/4.4.6.2 Reactor Coolant System Leakane e n =n .m ar m e Industry experience has shown that.while a limited. amount of leakage is expected from the RCS, the unidentified: portion of this leakage can be reduced.to a threshold value of lessc than 1-GPM.a This > threshold-value is sufficiently low to ensure early detection of.' additional-leakage. - w The 10 GPM IDENTIFIED LEAKAGE 1 imitation provides allowance for a 1imited amount of leakage from known sources;whose presence will.not. interfere with the detection of UNIDENTIFIED. LEAKAGE by the Leakage Detection Systems'.. an m c m,..m The total' steam. generator tube leakage limit"of 1 GPM'forrall steam-generators ensures that the dosage contribution from the tube leakage will be limited,to-a smalltfraction of Part 100alimits in the:eventsof either a. steam generator tube rupture or steam line break.+ The 1-GPM: limit is consistent with the assumptions < used in the anal' sis:of these accidents. y n m m The 100 gallon per day leakage: limit per steam generator ensures that steam generator tube ~ integrity is maintained:in+accordance with the-recommendations of Generic Letter 91-04w .,s c i. ,. n. PRESSURE B0UNDARY 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 000NDARY; LEAKAGE 4 requires the unit to be promptly placed in COLD SNUTDOWNw " 6 -r m 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion. of the Reactor Coolant System is minimizediand reduce the4 potential for: Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides: adequate-corrosion protection to ensure the structural' integrity of the ReactorL Coolant System over the life of the plant. The associated effects of exceeding the oxygen,L chloride and' fluoride -limits are time.and temperature dependent. Corrosion studies show that operation maylbe' continued with contaminant concentration levels in excess of the Steady State Limits..up to the Transient-Limits, for the specified limited time intervals without CALVERT CLIFFS - UNIT'1 B 3/4 4-5 Amendment No. 188

3 / 3/4.4 REACTOR C0OLANT SYSTEM RASES having a significant effect on the structural _ integrity of the Reactor Coolant System. The time interval pemitting continued operation within a the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to'within the Steady. State Limits. 4 The surveillance requirements provide adequate assurance that concentrations in excess of the'Ifmits will be detected in sufficient time to take corrective action. l t 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 B0MDARY will not exceed an-appropriately smail: fraction of Part 100 limits following a" steam generator-tuk rupture accident in conjunction with an assumed steady state primary-to econdary 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 upcr. a parametric evaluation by the-NRC of typical site locations. These values are conservative in that specific: site' parameters'of"the Calvert Cliffs site, such as SITE B0MBARY 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 s mcific' activity limits of - 1 this site. This reevaluation may result.in higier. limits.- The ACTION statement pemitting POWER OPERATION to continue' for limited time periods with the primary coolant's specific activity > 1.0~ Ci/ gram. DOSE EQUIVALENT I-131,. but'within the allowable limit shown on Figure 3.4.8-1, acconnodates possible iodine spiking phenomenon which may' 1 occur following changes in THERMAL POWER. L0peration with specific activity levels exceeding 1.0 C1/ gram DOSE EQUIVALENT'I-131 but within the limits shown on Figure 3;4;8-1 must be restricted.to no more than'10~ percent of' the unit's yearly operating time since the activity levels allowed by 1 Figure 3.4.8-1 increase the 2 hour thyroid dose at the SITE BOW RARY by a factor of up to 20 following a postulated steam generator tube rupture. Reducing T to < 500*F prevents the release of. activity should a steam generator Eube rupture since the saturation pressure of'the primary 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 coolant will be detected in sufficient time to take corrective action. Infomation obtained on iodine. spiking will be used to assess the' parameters associated with spiking 1 phenomena. A reduction in frequency of isotopic analyses following power f changes may be pemissible if justified by the data obtained. u l CAi.VERTCLIFFS-UNIT 1 B 3/4 4-6 Amendment No. 188 l

9 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by nomil load transients, reactor trips, and STARTUP and shutdown operation. The various categories of load cycles used for design purposes are provided in Section 4.1.1 of the UFSAR. During STARTUP and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. Operation within the appropriate heatup and cooldown curves assures the integrity of the reactor vessel against fracture induced by combinative thennal 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 3.4.9-1 and 3.4.9-2, are for a p'eak neutron fluence to the inner surface of the reactor vessel of 5 2.61x10 N/cm' (E > 1 MeV). This fluence corresponds to the Pressurized Thermal Shock Screening Criteria defined in 10 CFR 50.61 for weld 2-203 A, B, C. The reactor vessel materials have been tested to determine their initial RT,or; the results of these tests are shown in Section 4.1.5 of the UFSAR. Reactor operation and resultant fast neutron (E > 1 MeV) irradiation will cause an increase in the RT The actual shift in RT,oy of the vessel material will be establishe$r. periodically during operation by removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of 10 CFR Part 50, Appendix H. The shift in the material fracture toughness, as represented by RT,,,, is l calculated using Regulatory Guide 1.99, Revision 2. For a fluence of 2.61x10"N/ca', the adjusted reference temperature (ART) value at the 1/4 T position is 241.4*F. At the 3/4 T position the ART value is 181.0 F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code Section III A)pendix G to calculate heatup and cooldown limits in accordance with tie requirements of 10 CFR Part 50 Appendix G. To develop composite pressure-temperature limits for the heatup transient, the isothennal,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 l CALVERT CLIFFS - UNIT 1 B 3/4 4-7 Amendment No. 188 l

.,s-

x..

S. 3/4.4-REACTOR COOLANT SYSTEM BASE 5 the heatup event. The composite pressure-tem>erature limit for the cooldown transient is -developed similarly. Tie Appendix 6 limits in Figures 3.4.9-1 and 3.4.9-2 assume the following number of RCPs are running: HEATUP Indicated RCS Temperature Maximum Number of RCPs Operatina 70*F to 330 F 2 > 330 F 4 C00LDOWN Indicated RCS Temocrature Maximum Number of R(os Operatinc > 350*F 4 350'F to 150'F 2 < 150'F 0 Both 10 CFR Part 50, Appendix G and ASME, Code Appendix G require the-development of. pressure-temperature limits which are applicable to inservice hydrostatic tests. The minimum' temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the . test pressure (1.1 times nomal operating pressure) and locating the corresponding temperature. This curve is shown for a fluence of < 2.61x10"N/ca' on Figures 3.4.9-1: and 3.4.9-2. 3 Similarly,10 CFR Part 50 specifies that core critical limits be established based on material considerations.= This limit is shown on the heatup curve. Figure 3.4.9-1. Note that this limit does not consider the~ core reactivity safety analyses that actually control the temperature at which the core can be brought critical. The Lowest Service Temperature is the minimum allowable temperature at ressures above 20% of the pre-operational system hydrostatic test pressure. p(625 psta). This temperature is defined as equal to the most limiting RTm. i for the balance of the Reactor Coolant System components plus 100*F per Article N8 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' Boller and Pressure Vessel Code - as 20% of tie pre-operational hydrostatic test. pressure. The change in the-line at 150*F on Figure 3.4.9-2 is due to a cessation of RCP flow' induced pressure deviation, since no RCPs.are permitted to operate during a cooldown below 150 F. CALVERT CLIFFS - UNIT 1 B 3/4 4-8 Amendment No.-188 l

hy 3 i H, n 3/4;4 REACTOR C0OLANT 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 the higher stressed region of the reactor vessel pl for the material of us any effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel Code. The initial-reference temperature of the reactor vessel and closure head flanges was detenmined using the certified material test reports and Branch Technica.1 Position MTEB 5-2. The maximum initial RT, associated with the stressed region of the closure head flange is -10*F. However, in order to comply with the 10 CFR 50, Appendix 6 limits, the minimum allowable reactor vessel temperature with the reactor head attached is 70 F. Hence, the minimum boltup temperature used in. Figures and 3.4.9-1 and 3.4.9-2. The Low-Temperature Overpressure Protection (LTOP) System consists of administrative controls coupled with low-pressure setpoint PORVs. The administrative controls provide the first lir.e of defense'against 1 overpressurization events; the PORVs provide a backup to the administrative: controls. The following section discusses the bases for the PORV setpoint and administrative controls. Low-Temperature Overpressure Protection uses a variable PORV setpoint to take advantage of the increased Appendix G limits at higher RCS temocratures. Reactor Coolant System temperature is measured at the cold leg RTDs. This provides an' accurate temperature indication during' forced circulation, and is also adequate for-natural circulation. However, the T RTDs are not accurate when on shutdown cooling because they are not in tN flow stream. For this reason, the lowest PORV setpoint is maintained whenever on shutdown cooling. This setpoint, which is independent of RCS temperature, is manually set when shutdown cooling is initiated and. maintained until forced circulation is established.after the RCPs are started. I The PORV setpoint is chosen to protect the most limiting of the heatup or cooldown Appendix G limits. Figure 3.4.9-3 shows the maximum PORV opening pressure. This includes corrections for static and dynamic head, and pressure overshoot to account for PORV response time and the maximum-pressurization rate. The actual PORV set >oint is controlled by procedure and accou_nts for device uncertainty -call aration. uncertainty and loop - drift. The design basis events in the low temperature region are: An RCP start with hot steam generators;.and. An inadvertent HPSI actuation with concurrent charging. These' transients are most severe when the RCS is' initially water solid. J Any measures which will prevent or mitigate the design basis events are sufficient for any less-severe incidents. Therefore, this section will l CALVERT CLIFFS - UNIT 1 B 3/4 4-9 Amendment No. 188 1 = a

3/4.4 REACTOR COOLANT SYSTEM BASES \\ 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. i The RCP start transient is a severe LTOP challenge that can quickly exceed the Appendix G limits for a water solid RCS. Therefore, during water solid operations all four RCPs are tagged out of service and their motor circuit breakers are disabled. However, the transient is adequately mitigated by restricting three parameters:

1) the initial water volume in the pressurizer to 170 inches (indicated), thereby providing a volume for the primary coolant to expand into; 2) the indicated secondary water temperature for each steam generator to 30 F above the RCS temperature; and

3) the initial pressure of the pressurizer to 300 psia.

With these restrictions in place, the transient is adequately controlled without the assistance of the PORVs. Failure to maintain one of the initial conditions j could cause the PORVs to open following an RCP start. The mass addition transient from HPSI or multiple charging pumps is a' severe LTOP challenge for a water solid system due to PORY response time. To preclude this event from happening while water solid, all HPSI pumps and two charging pumps are tagged out-of-service during water solid operations. Analyses were )erformed for a HPSI mass addition transient with concurrent j charging and t1e expansion of the RCS water volume following loss of decay heat removal, assuming one PORV available (due to single-failure criteria). This mass addition, determined at the point when the RCS reat E d water solid conditions, must be less than the capability of a single PORY to limit the LTOP event. Suffici d over)ressure protection results when the equilibrium pressure does not aed tie simiting Appendix G curve pressure. Because the equilibrium pressure 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 for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop MOV to limit instrumentation uncertainty, No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition. Three 100% capacity HPSI pumps are installed at Calvert Cliffs. Procedures will require that two of the three HPSI pumps be disabled (breakers racked out) at RCS temperatures less than or equal to 365'F and that the remaining HPSI pump handswitch be placed in pull-to-lock. Additionally, the HPSI pump normally in pull-to-lock shall De throttled to less than or equal to 210 gpm when used to add mass to the RCS. Exceptions are provided for ECCS testing and for response to LOCAs. To provide single failure protection against a HPSI pump mass addition transient when in MPT enable, the HPSI loop MOV handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt CALVERT CLIFFS - UNIT 1 8 3/4 4-10 Amendment No. 188 l

3/4.4 REACTOR COOLANT SYSTEM t BASES of a SIAS signal. Alternative actions, described in the ACTION 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 tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge valves; and (3) disabling all three HPSI pumps. RCS temperature, as used in the applicability statement, is determined as follows: (1) with the RCPs running, the RCS cold leg temperature is the a) pro)riate indication, (2) with the Shutdown Cooling System in operatio1, t1e stutdown cooling temperature indication is appropriate (3) if neither the RCPs or shutdown cooling is in operation, the core exit thennocouples are the appropriate indicators of RCS temperature. l The allowed out-of-service times for degraded low temperature overpressure y protection system in MODES 5 and 6 are based on the guidance provided in j Generic Letter 90-06 and the time required to conduct a controlled, deliberate cooldown, and to depressurize and vent the RCS under the ACTION statement entry conditions. 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for the ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in coinpliance with Section XI of the ASME Boiler and Pressure Vessel Code. 3/4.4.11 CORE BARREL MOVEMENT This speu fication is provided to ensure early detection of excessive core barrel movement if it should occur. Core barrel movement will be detected by using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Spectral Analysis (SA). Baseline core barrel movement Alert Levels and Action Levels will be confirmed during each reactor startup test program following a core reload. Data from these detectors is to be reduced in two forms. Root mean. square (RMS) values are computed from the APD of the signal amplitude. These RMS magnitudes include variations due both to various neutronic effects and internals motion. Consequently, these signals alone can only provide a gross measure of core barrel motion. A more accurate assessment of core barrel motion is obtained from the Auto and Cross Power Spectral Densities (PSD, XPSD), )hase (6) and coherence (COH) of these signals. These data result from tie SA of the excore detector signals. A modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of the affected parts during the plant shutdown at the end of any fuel cycle. CALVERT CLIFFS - UNIT 1 B 3/4 4-11 Amendment No. 188

s-3/4.4 REACTOR C0OLANT SYSTEM BASES 3/4.4.12 LETDOWN LINE EXCESS FLOW This specification is provided to ensure that the bypass valve for the excess flow check valve in the letdown line will be maintained closed during plant operation. This bypass valve is required to be closed to ensure that the effects of a pipe rupture downstream of this valve will not exceed the accident analysis assumptions. 3/4.4.13 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the Primary System that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer vapor space ensures the capability exists to perfonn this function. The valve. redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversilyle actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of THI Action Plan Requirements," #ovember 1980. ..s. J CALVERT CLIFFS - UNIT 1 B 3/4 4-12 Amendment No. 188 l

TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.................. 3/4 3-9 3/4.2.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation 3/4 3-23 Incore Detectors 3/4 3-27 Seismic Instrumentation.............. 3/4 3-30 Meteorological Instrumentation 3/4 3-33 Remote Shutdown Instrumentation.......... 3/4 3-36 Post-Accident Instrumentation........... 3/4 3-39 Fire Detection Instrumentation 3/4 3-43 Radioactive Gaseous Effluent Monitoring Instrumentation.................. 3/4 3-47 Radioactive Liquid Effluent Monitoring Instrumentation.................. 3/4 3-52 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION STARTUP and POWER OPERATION............ 3/4 4 HOT STANDBY.................... 3/4 4-2 Shutdown 3/4 4-4 3/4.4.2 SAFETY VALVES................... 3/4 4-6 3/4.4.3 RELIEF VALVES................... 3/4 4-7 3/4.4.4 PRESSURIZER.................... 3/4 4-9 l 3/4 4-10 l 3/4.4.5 STEAM GENERATORS 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............. 3/4 4 Reactor Coolant System Leakage 3/4 4-19 3/4.4.7 CHEMISTRY..................... 3/4.4-21 l 3/4.4.8 SPECIFIC ACTIVITY................. 3/4 4-24 l CALVERT CLIFFS - UNIT 2 IV Amendment No. 165

1 TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9 PRESSURE /TEMPERATUREilMITS 1 Reactor Coolant System 3/4 4-28 Pres suri zer.................... 3/4 4-32 Overpressure Protection Systems.......... 3/4 4-33 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components....... 3/4 4-36 l 3/4.4.11 CORE BARREL MOVEMENT 3/4 4-38 .l 3/4.4.12 LETDOWN LINE EXCESS FLOW 3/4 4-40 l 3/4.4.13 REACTOR COOLANT SYSTEM VENTS 3/4 4-41 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETYINJECTIUNTANKS '3/[S-1 ~ 3/4.5.2 ECCS SUBSYSTEMS - MODES 1, 2 AND 3 (> 1750 PSIA) 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - MODES 3 (< 1750 PSIA) AN'D 4... 3/4 5-7 3/4 5.4 REFUELING WATER TANK 3/4 5-8 3/4.6 CONTAINNENTrSYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY............... 3/4 6-1 Containmsnt Leakage................ 3/4 6-2 Containment Air Locks............... 3/4 6-5 Internal Pressure................. 3/4 6-7 Air Temperature.................. 3/4 6-8 Containment Structural Integrity 3/4 6-9 Containment Purge System 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6-12 Containment Cooling System 3/4 6-14 3/4.6.3 IODINE REMOVAL SYSTEM............... 3/4 6-16 CALVERT CLIFFS - UNIT 2 V Amendment No. 165

] 9 ] TABLE OF CONTENTS BASES 4 SECTION PAGE 3/4.0 APPLICABILITY................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL B 3/4 1-1 3/4.1.2 BORATION SYSTEMS B 3/4 1-2 I 3/4.1.3 MOVABLE CONTROL AS';EMBLIES B 3/4 1-3 l i 3/4.2 POWER DISTRIBUTION LIM 1TS 3/4.2.1 LINEAR HEAT RATE ..............73;. B 3/4 2-1 l 3/4.2.2, 3/4.2.3, and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS - P AND y F, AND AZIMUTHAL POWER. TILT - T, B 3/4 2-1 3/4.2.5 DNB PARAMETERS......'h. B 3/4 2-2

r 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION B 3/4 3-1 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 PRESSURIZER.................... B 3/4 4-3 l 3/4.4.5 STEAM GENERATORS B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-5 l CALVERT CLIFFS - UNIT 2 X Amendment No. 165

TABLE OF CONTENTS BASES SECTION PAGE 3/4.4.7 CHEMISTRY B 3/4 4-5 l 3/4.4.8 SPECIFIC ACTIVITY................. B 3/4 4-6 l 3.4.4.9 PRESSURE / TEMPERATURE LIMITS............ B 3/4 4-7 l 3/4.4.10 STRUCTURAL INTEGRITY B 3/4 4-10 3/4.4.11 CORE BARREL MOVEMENT B 3/4 4-11 l 3/4.4.12 LETDOWN LINE EXCESS FLOW B 3/4 4-11 3/4.4.13 REACTOR COOLANT SYSTEM VENTS B 3/4 4-11 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY 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 REFUELING WATER TANK (RWT) B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS B 3/4 6-3 3/4.6.3 IODINE REMOVAL SYSTEM............... B 3/4 6-3 3/4.6.4 CONTAINMENT ISOLATION VALVES B 3/4 6-3 3/4.6.5 COMBUSTIBLE GAS CONTROL.............. B 3/4 6-4 3/4.6.6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSTEM B 3/4 6.. CALVERT CLIFFS - UNIT 2 XI Amendment No. 165 .4 i

s 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3 Two power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, and 3*. l ACTION: a. If one or both PORV(s) has excessive' seat leakage, within 1 hour close the associated block valve (s) and maintain power to the blockvalve(s). a b. With one PORV inoperable due to causes other than excessive PORV seat leakage, within 1 hour either restore the PORY to OPERABLE status or close the associated block-valve and remove power from the block valve; restore the PORY to'0PERABLE status within the following 5 days or be in HOT STANDBY within the next 12 hours and at or below 305 F within the fo110 wing 24 hours, c. With both PORVs inoperable due to causes other than excessive PORV seat leakage, within 1 hou'r either restore the PORVs to OPERABLE status or close its associated b1cck valve and remove power from the block valve; restore one PORV-to 0PERABLE status within the following 72 hours or be in H0T STANDBY within the next 12 hours and at or below 305'F within the follow'irfg 24 hours. d. With one or both block valve (s) inoperable, within 1 hour restore the block valve (s) to OPERABLE status or place its associated ~ ~ PORV(s) in override closed. Restore at least one block valve to CPERABLE status within the next 72 hours if both block valves are inoperable; restore any remaining inoperable block ' valve to OPERABLE status within the following 5 days; otherwise, be in at least NOT STANDBY within the next 12 hours and at or below 305 F within the following 24 hours. e. The provisions of Specification 3.0.4 are not applicable. l Above 305 F. At or below 305 F Specification 3/4.4.9.3 applies. l CALVERT CLIFFS - UNIT 2 3/4 4-7 Amendment No. 165

l 3/4.4 REACTOR COOLANT SYSTEM j SURVEILLANCE REQUIREMENTS 4.4.3.1 Each PORY shall be demonstrated 0PERABLE: a. At least 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 once per 92 days by operating the valve through one complete cyclelof full travel unless the block valve is closed to meet the requirements of Action a, b, or e in Specification 3.4.3. i .i l 1 1 CALVERT CLIFFS - UNIT 2 3/4 4-8 Amendment No. 165

4 f 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.4 PRESSURIZER 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 sower. The pressurizer level shall be maintained within an operating sand between 133 and 225 inches except when three charging pumps are operating and letdown flow is less than 25 GPM. If three. charging pumps are operating and letdown flow is less than 25 GPM pressurizer level shall be limited to between 133 and 210 inches. APPLICABILITY: 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 H0T STANDBY within the next 6 hours and in H0T SHUTDOWN within the following 12 hours. b. With the pressurizer otherwise inoperable, be in at least HOT STAND 8Y with the reactor trip breakers open within 6 hours.and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer water level shall be detemined to be within the above band at least once per 12 hours. CALVERT CLIFFS - UNIT 2 3/4 4-9 Amendment No. 165 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS 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.,, above 200 F. SURVEILUUICE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by perfomance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Sample 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. 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 perfomed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. 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 basis except: 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: 1. All nonplugged tubes-that previously had detectable wall penetrations (>20%), and CALVERT CLIFFS - UNIT 2 3/4 4-10 Amendment No. 165 l

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2. Tubes in those areas where experience has indicated potential problems. c. The second and third inservice inspections may be less than a i full tube insaection by concentrating (selecting at least 50% of the tubes to se 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: Catecor_v Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 Oneormoretubes,butnotmorethan1%of the total tubes inspected are defective, or i between 5% and 10% of the total tubes inspected are degraded tubes. j C-3 More 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 i be included in the above percentage calculations. 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be perfonned at the following frequencies: a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.. If at least 20 percent of the tubes wer)e insaected and the results were in the C-1 Category (See Note or tf at least 40 percent of the tubes were inspected and were in the C-2 Category during the previous inspection, the next inspection may be extended up to a maximum of 30 months in order to correspond with the next refueling outage if the results of the two previous inspections

NOTE:

For Cycle 9, an inspection of 15% of the steam generator tubes with inspection results in the C-1 Category shall be acceptable to extend the next inspection up to 30 months to coincide with the next refueling outage. CALVERT CLIFFS - UNIT 2 3/4 4-11 Amendment No. 165 l

h. 3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) were not in the C-3 Category. However, if the results of either of the previous two inspections were in the C-2 Category, an engineering assessment shall be perfonned before operation beyond 24 months and shall provide assurance that all tubes will retain adequate structural margins against burst throughout nonnal o)erating, transient, and accident conditions until the end of tie fu'el cycle or 30 months, whichever occurs first. If two consecutive inssections following service under AVT conditions, not including t1e preservice inspection, result in all inspection results falling into the C-1 category 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. b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The. increase in inspection frequency shall apply until the subsequent inspectio'n 20% of the nominal wall thickness caused by degradation. 4. % Degradation means the percentage of the tube wall thickness affected or removed by degradation. 5. Defect means an imperfection of such severity that it exceeds the pfugging limit. A tube containing a defect is defective. Any tube which does not pennit the passage of the eddy-current inspection probe shall be deemed a defective tube. 6. Pluacina Limit means the imperfection depth at or beyond I which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness. 7. Unserviceable describes the condition of a tube if it leaks or contains a defect 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 the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. b. The steam generator shall be detennined 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. CALVERT CLIFFS - UNIT 2 3/4 4-13 Amendment No. 165 l

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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 Comission within 15 days pursuant to Specification 6.9.2. b. The complete results of the steam generator tube inservice inspection'shall'be' included in the Annual Operating Report for 4 the period in which this inspection was completed (pursuant to Specification 6.9.1.5.b)F - This report shall include 1. Number and extent of tubes < inspected. 2. Location and percent of wall-thickness penetration for each indication of an imperfection. -4 3. Identification of tubes plugged. 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 operatione Theawritten4fe1towup 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. + ,mw _e .v .:v CALVERT CLIFFS - UNIT 2 3/4 4-14 Amendment No. 165 l'

Q TABLE 4.4-1 R G g MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION f2 g M Q n o Preservice Inspection No Yes g No. of Steam Generators per Unit Two Three Four Two Three Four h First Inservice Inspection All One Two Two l l Z 3 Second & Subsequent Inservice Inspections One One One One E R a ti; TABLE NOTATION: 1 The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are perfonning in

~~e-a like manner. Note that under some circumstances, the operating conditions in one or~more steam 5~~
generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions. 5 2 The other steam generator not inspected during the first inservice inspection shall be inspected. g The third and subsequent inspections should follow the instructions described in 1 above. 3 Each of the other two steam generators not inspected during the first inservice inspections shall S be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.
Q TABLE 4.4-2 w r-M STEAN GENERATOR TUSE IBSPECTION 4 15T 5AM LE AD rLGIAUR ZND 5JWLL IDFLLI AUR JHD 5A M LL ID rtLIAUE 30 h 5 ample 512e Result Action Required Result Action Required Result Action Required S Q A alntmum of 5 Iubes per C-1 None M/A N/A N/A N/A O 1 Q
~~5. G.~~
c-z Plug detective tubes and C-1 None N/A N/A inspect additional 25 c-Z Plug detective C-1 None n tubes in this S. G. tubes and inspect Plug detective 8 c 2 additional 45 tubes C-2 tubes C Q in this S. G. Fertonn action g ro C-3 for C-3 result of first sample C-3 Pertonn action for C-3 result of first N/A N/A g sample C-3 Inspect alI tunes in All other this S. G., plug S. G.s are None N/A N/A w defective tubes and C-1 D inspect 25 tubes in each a other S. G. ,8 Some 5. G.s Perfonn action for m C-2 but no C-2 result of N/A N/A additional second sample 24 hour verbal S. G. are C-3 notification to NRC with written followup pursuant to Specification 6.9.2. Additional inspect all tubes S. G. is C in each S. G. and plug defective N/A N/A f tubes. 24 hour
~~3 verbal notification to NRC with written 5~~
followup pursusnt 5 to Specification 6.9.2. 2o S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators ~ ~ g; inspected during an inspection
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems s..
~~ra,i-LIMITING CONDITION FOR OPERATION t 9V Mh 3.4.6.1 The following Reactor Coolant System Leakage Detection 1 Systems shall be OPERABLE~~

~~a. A Containment Atmosphere Particulate Radioactivity Monitoring~~
~~System, b.-~~
The Containment Sump Level Alann 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 required Leakage Detection Systems OPERABLE, 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 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 required Leakage Detection Systems OPERABLE, operation may continue for up to 7 days provided that: 1. Grab samples of the containment atmosphere are obtained and analyzed at least once per 12 hours, and 2. The Reactor Coolant System water inventory balance of Surveillance Requirement 4.4.6.2.c is perfonned at least once per 24 hours. Otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. CALVERT CLIFFS - UNIT 2 3/4 4-17 Amendment No. 165 l
4-J 3/4.4 REACTOR. COOLANT SYSTEN SURVEILLANCE REQUIREMENTS. 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: a. Containment Atmosphere. Gaseous.and Particulate Monitoring Systems-perfonnance'of, CHAISIE1.' CNECK, CM C, ALIBRATION and-CHAINIEL FWICTIONAL' TEST at the~ frequeiiciss -spittfied in Table'4.3-3, and -+elty ur ses , gen.,r: q b. Containment Sump Level Alann System-perfonnanc'e of CHAlHIEL ' ~ ~ CALIBRATI001 at least once per 18 months".' e -c l CALVERT CLIFFS - UNIT 2 3/4'4-18 Amendment No.:165 -l~ 1 ,-c .m.
b 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Reactor Coolant System Leakage en L t.n .. = #, a a LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage,shall,be limited to: a. No PRESSURE B0UNDARY LEAKAGE, b g JPH UNIDENTIFIED LEAKAGE, 1 i c. 1 GPM total primary-to-secondary leakage through all steam 1 generators and 100 gallons-per-day through any one steam generator, and d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System. APPLICABILITT:' MODE $' 14 ' 2, 3 and'4. ~" ACTION: O a a 1. a. With any PRESSURE' B0UNDARY' LEAKAGE. be in at least HOT STANDBY ) within 6 hours and in COLD SHUTDOWN within the followi.ng 30 hours. i< e.. b. With anf Re'at:tbr Coolant System leakage greater than any one of the above limits, excluding PRES $URE B0UNDARY LEAKAGE, reduce the i leakage rate to within limits within'4' hours or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTD0WN within the following 30 hours. o n.
} V. g}{y s
'ri, SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be' demonstrated to be within each of the above 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 Technical Specification 3.4.6.1. CALVERT CLIFFS - UNIT 2 3/4 4-19 Amendment No. 165 l
3/4.4 REACTOR COOLANT SYSTEM' SURVEILLANCE REQUIREMENTS (Continued) b. Monitoring the containment sump discharge frequency at_1 east once per 12 hours, when the Containment Sump Level Alarm System'is
~~OPERABLE,~~
~ c. Determining the Reactor Coolant System water leakage ~ at least once per 72 hours during steady state operation and at least once per 24 hours when required by ACTION 3.4.6.1.b, except when operating in the shutdown cooling mode, and d. Monitoring the reactor vessel head closure'~ seal Leakage' Detection System at least once per 24 hours. .a m as; movAtuar v. ; s.- o ! ~ -a rj t S p-nt T3 1 CALVERT CLIFFS - UNIT 2 3/4 4-20 Amendment No. 165 l
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIM 7 TING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3s4-1. APPLICABILITY: At all times. ACTION: MODES 1, 2, 3 and 4: a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD 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 Steady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to 5 500 psia, if applicable, and perfom an engineering evaluation to detemine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; detemine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to M0DE 4. SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be detemined to be within the limits by analysis of those paraineters at the frequencies specified in Table 4.4-3. CALVERT CLIFFS - UNIT 2 3/4 4-21 Amendment No. 165 l
,~N o. 3/4.4 REACTOR COOLANT SYSTEM TABLE 3.4-1 REACTOR C0OLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN * $ 0.10 ppm $ 1.00 ppe CHLORIDE $ 0.15 ppm 5 1.50 ppe FLUORIDE 5 0.15 ppm 5 1.50 ppe i i L i Limit not applicable with T.,,5 250'F. CALVERT CLIFFS - UNIT 2 3/4 4-22 Amendment No. 165 l
-jl ~ ) 3/4.4 REACTOR C00LANT SYSTEM TABLE 4.4-3 REACTOR C0OLANT SYSTEM CHEMISTRY LIMITS SURVEILUUICE REQUIREMENTS ^ PARAMETER ANALYSIS FREQUENCY DISSOLVED OXYGEN" At least once per 72 hours CHLORIDE At least once per 72 hours FLUORIDE At least once per 72 hours I / Not required with.T., s 250 F. CALVERT' CLIFF 5 - UNIT 2. 3/4 4-23 Amendment No. 165 l
) 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: a. 5 1.0 pCi/ gram DOSE EQUIVALENT I-131, and b. $ 100/E pC1/ gram. APPLICABILITY: MODES 1, 2. 3, 4 and 5. ACTION: MODES 1, 2 and 3*: With the specific activity of the primary coolant > 1.0 pCi/ grani a. DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4.8-1, operation may continue for up 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 pCi/ gram DOSE EQUIVALENT I-131 for more than 100 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least HOT STANDBY with T.,, < 500 F within 6 hours. c. With the specific activity of the primary coolant > 100/E pCi/ gram, be in at least NOT STANDBY with T.,, < 500 F within 6 hours. MODES 1, 2, 3, 4 and 5: d. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/l Ci/ gram, perform the sam)1ing and analysis requirements of item 4 a) of Table 4.4-4 untiL the specific activity of the primary coolant.is restored to within its limits. Whenever the specific activity of the primary coolant exceeds 1.0 pCi/ gram DOSE EQUIVALENT I-131 for in excess of 50 hours for one continuous time interval or 5 percent of the unit's total yearly operating time pursuant to ACTION (a) above, a Special Report shall be prepared and submitted to the With T.,, >_,500 F. CALVERT CLIFFS - UNIT 2 3/4 4-24 Amendment No. 165 l
3 3/4.4 REACTOR C0OLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) Connission sursuant to Specification 6.9.2 within the next 30 days. T11s report shall contain the results of the specific activity analyses together with the following infonnation-1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded, 2. Fuel burnup by core region. 3. Clean-up flow history starting 48 hours prior:to the first sample in which the limit was exceeded. 4. History of de-gassing operation, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and 5. The time duration when the specific activity.of the primary coolant exceeded 1.0 C1/ gram DOSE EQUIVALENT I-131. SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be detennined to be within the limits by performance of the sampling and analysis program of Table 4.4-4. CALVERT CLIFFS - UNIT 2 3/4 4-25 Amendment No. 165 l L-
g TABLE 4.4-4 R G g PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM E2 9 n Q TYPE OF MEASUREMENT SAMPLE AND MODES IN WHICH SAMPLE AND g AND ANALYSIS ANALYSIS FREQUENCY ANALYSIS REQUIRED 9 o c-At least once per 72 1,2,3,4 E 1.
~~ross Activity Determination hcurs "i~~
~~] 2. Isotopic Analysis for DOSE 1 per 14 days 1 E EQUIVALENT I-131 Concentration 3. Radiochemical for E Detennination 1 per 6 months~~
~~1 w~~
4. Isotopic Analysis for Iodine a) Once per 4 hours, l', 2', 3', 4', S' 1 Including I-131, 1-133, and I-135 whenever the DOSE EQUIVALENT I-131 a h exceeds 1.0 pCi/ gram, and b). One sample between 1, 2, 3 2 and 6 hours following a THERMAL l POWER change exceeding 15 percent of the g RATED THERMAL POWER l R within a one hour l 2 period. E I l P Until the specific activity of the Primary Coolant System is restored within its limits. i S Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since l reactor was last subcritical for 48 hours or longer. l
3/4.4 REACTOR C0OLANT SYSTEM g.. L 250 \\ 't T. O \\ T ' L \\ \\ t h i UNACCEPPABLE \\ ' OPERATION L 150 \\ e T L a _]_ \\ L 100 i L w y. ACCEPTABLE 't I50 -- OPERATION \\ 't 1 k=f I O 20 30 40 50 60 70 80 90 100-PERCENT OF RATEDTHERMAL POWER i l-FIGURE 3.4.8-1 DOSE EQUIVALENT I-131 PRIMARY C0OLANT SPECIFIC ACTIVITY LIMIT VERSUS l PERCENT OF RATED THERMAL POWER WITH THE PRIMARY C0OLANT SPECIFIC ACTIVITY >1.0uci/ GRAM DOSE EQUIVALENT I-131 CALVERT CLIFFS - UNIT 2 3/4 4-27 Amendment No 165 l'.
[* 3/4.4 E%CTORCOOLANTSYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS l Reactor Coolant Systen! LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with tS2 limit lines shown on Figures 3.4.9-1 and 3.4.9-2 during heatup, cce Mown, criticality, and-l inservice leak and hydrostatic testing with: a. A maximum heatup of 75 F in any one hour period, l b. A maximum cooldown of: 1 j Maximum Allowable Cooldown Rate RCS Temperature 'l 100 F in any one hour period > 180 F 40 F in any one hour period 180 F to 140 F 15 F in any one hour period < 140 F-j c. A maximum
~~ature change of 5 F in any one hour period, a~~
~~h during hyi 'c testing operations above system design~~
~~pressure, APPLICABILITY: At ai.~~
...e s. ACTION: . With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfonn an engineering evaluation to determine'the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; detennine that the Reactor Coolant System remains acceptable for 1 continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T.,lhe fellowing 30 hours.and pressure to less'than 200 F respectively, within l 1 U i CALVERT CLIFFS - UNIT 2 3/4 4-28 Amendmert No. 165 l 'l
c + ? 3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor' Coolant System temperature and pressure shall be determined to be within the limits at least once r:i 30 minutes during system heatup, cooldown, and inservice leak and h)drostatic 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, as required by 10 CFR Part 50, Appendix H. The.results of these examinations shall be used to update Figures 3.4.9-1 and 3.4.9-2. CALVERT CLIFFS - UNIT 2 3/4 4-29 Amendment No. 165 l. .J
4 0 3/4.4 REACTOR COOLANT SYSTEM i l 2500

~~i-HE ATUP -. _ _,~~

~~l EINSERVICE HYDROSTATIC TEST' [ [ ^1 i- : 2000 ? E i E 1500~~
~~LOWEST~~
' CORE CRITICAL E ] TEMPERATURE
== SERVICE j g g N E 160'F ^ r Q h1000 ~,- E RCS TEMP. H/U RATE a-O ',f ALL TEMPS 575'F/1 HR S 500 MIN. BOLTUP TEMP. 70 *F ' y i MAXIMUM PRESSSURE - FOR SDC OPERATION' O. 100 200-300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE Tc, F FIGURE 3.4.9-1 CALVERT CLIFFS UNIT 2 HEATUP CURVE, for FLUENCE 5 1.92x10" n/cm* REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 2 3/4 4-30 Amendment No. 165 l
3/4.4 REACTOR COOLANT SYSTEM 2500 .=- : EEINSERVICE HYDROSTATIC TESTk NH 2000 'l551 gLOWEST gg SERVICE @ TEMPER ATURE c. Eg160*F ' 1; 1500 m d 'COOLDOWN g 2 5m i 5 l ~ h ./ .RCS TEMP. C/D RATE m >1804 5100V/1 HR O 180V TO 1404 s404/1 HR h <1409 $154/1 HR j ~ S00 - MIN. BOLTUP TEMP 70 'F MAXIMUM PRESS'SUNF.: FOR SDC' OPERATION i 0 100 200 300 400 500 602 INDICATED REACTOR COOLANT TEMPERATURE Tc, F FIGURE 3.4.9 2 CALVERT CLIFFS UNIT 2 C00LDOWN CURVE, for FLUENCE $ 1.92x10" n/ca' REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 2 3/4 4-31 Amendment NO. 165 l
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Pressurizer LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to: a. A maximum heatup of 100 F in any one hour period, b. A maximum cooldown of 200 F in any one hour period, and c. A maximum spray water temperature differential of 400 F. APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the abwe 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 NOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 300 psia within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be detennined 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 2 3/4 4-32 Amendment No. 166 l
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Overpressure Protection Systems LIMITING CONDITION FOR OPERATION 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 of 5 430 psia with their associated block valves open, or l 2. A single PORV with a lift setting of 5 430 psia with its associated block valve open and a Reactor Coolant System vent of 2 1.3 square inches, or 3. A Reactor Coolant System (RCS) vent t 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 handswitches 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 RCS and when used, it must be under manual control and one of the following restrictions shall apply:
1. The total high pressure safety injection flow shall be limited to 5 210 gpm OR 2. A Reactor Coolant System vent of g 2.6 square inches shall exist. e. When not in use, the above OPERABLE HPSI pump shall have its I handswitch in pull-to-lock. APPLICABILITY: When the RCS temperature is 5 305 F and the RCS is vented to < 8 square inches. i Except when required for testing. i CALVERT CLIFFS - UNIT 2 3/4 4-33 Amendment No. 165
~ ^ ^ ~ i 3/4.4 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) ACTION: a. With one PORY inoperable in MODE 3 with RCS temperature 5 305 F or in MODE 4, 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 one PORV inoperable in MODES 5 or 6, either restore the inoperable PORV to OPERABLE status within 24 hours, or-depressurize and vent the RCS through a 2 1.3 square inch vent (s) within the next 48 hours; and maintain the RCS in this vented condition until both PORVs have been restored to OPERABLE status. c. With both PORVs inoperable, depressurize and vent the RCS through? l a 1 2.6 square inch vent (s) within 48 hours; maintain the RCS in-a vented condition until either one OPERABLE PORY and a vent.of 2 1.3 square inches has been established or both PORVs.have been restored to OPERABLE status. d. In.the event either the PORVs or the RCS vent (s) are used to l: mitigate an RCS pressure transient, a Special Report shall be prepared and submitted 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.' With less than two HPSI pumps' disabled, place at least two HPSI l-e. pump handswitches in pull-to-lock within fifteen minutes and disable two HPSI pumps within the next four hours. f. With one or more HPSI loo) MOVs' not prevented from automatically .l aligning a HPSI pump to tie RCS, imediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or iselate the affected HPSI header flowpath within four hours, and implement the action. requirements of Specifications 3.1.2.1 3.1.2.3. and 3.5.3, as ' applicable. g. With HPSI flow exceeding 210 gpm while suction is aligned to' the l RWT and an RCS vent of < 2.6 squire inches exists, 1. Imediately take action to reduce flow to less'than or equal' q to 210 gpm. Except when required for testing. CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No. 165
^' 3/4.4 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) 2. Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the given RCS temperature on Figure 3.4.9-1 or Figure 3.4.9-2. 3. If a pressure limit was exceeded, take action in accordance with Specification 3.4.9.1. h. The provisions of Specification 3.0.4 are not applicable. 1 SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation,.within 31 days prior to entering a condition in which the PORV is required OPERABLE anti at least_once per 31 days thereafter when the PORV is required OPERABLE. b. Performance of a CHANNEL CALIBRATION on the PORV actuation 1 channel at least once per 18 months. c. Verifying the PORV block valve is open at least once per 72 hours l when the PORV is being used for overpressure protection. d. Testing in accordance with the inservice test requirements 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.~~
1 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 discharga valves are locked shut. The automatic opening feature of the high pressure safety injection loop.MOVs shall be verified disabled at least once per 12 hours. The above OPERABLE pump shall be verified to have its handswitch in pull-i to-lock at least once per 12 hours. I 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. CALVERT CLIFFS - UNIT 2 3/4_4-35 Amendment No.,165 f
4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1. 2 and 3 Components LIMITING CONDITION FOR 0PERATION 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 ASME Code Class I component (s) not conforming to the above requirements ' restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations. b. With the structural integrity of any ASME' Code Class 2. component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s). to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F. c. With the structural integrity of any ASME Code Class 3 component (s) not confonning to the above requirements., restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service. d. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 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 2 3/4 4-36 knendment No.165 l
.4 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) In addition to the requirements of Specification 4.0.5, each Reacto-Coolant Pump flywheel shall be inspected per the recomendation of Regulatory P August 1975.psition C.4.b of Regulatory Guide 1.14. Revision 1 4.4.10.1.2 Augmented Inservice Snspection Procram for Main Steam and Main Feedwater Pipina - The unencapsu'ated welds 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 ins sected per the following augmented inservice inspection program using tie 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 detected in the second sampling, the remainder of the welds shall also be inspected. 1 i Reactor coolant pump flywheel inspections for the first inservice inspection interval may be completed during Unit 2 Refueling Outage No. 9 in conjunction with the reactor coolant pump motor overhaul program. j CALVERT CLIFFS - UNIT 2 3/4 4-37 Amendment No. 165 l
4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.11 CORE BARREL MOVEMENT LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level. APPLICABILITY: H0DE 1. ACTION: a. With the APD and/or SA exceeding their applicable Alert Levels. POWER OPERATION, may proceed provided the following actions are taken: 1. APD shall be measured and processed at least once per 24 hours, 2. SA shall be measured at least once per 24 hours and shall be processed at least once per 7 days, and 3. A Special Report, identifying the cause(s) for exceeding the applicable Alert Level, shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days of detection. b. With the APD and/or SA exceeding their applicable Action Levels, measure and 'arocess APD and SA data within 24 hours to determine if the core 'aarrel motion is exceeding its limits. With the core barrel motion exceeding its limits, reduce the core barrel motion to within its Action Levels within the next 24 hours or be in HOT STANDBY within the following 6 hours. c. The provisions of Specifications 3.0.3 and 3.0.4 re not applicable. CALVERT CLIFFS - UNIT 2 3/4 4-38 Amendment No. 165 l o
^
~~l f~~
3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.11 Routine Monitorina Core barrel movement shall be determined to tut. less-than the APD and SA Alert Levels by using the excore neutron detectors to measure APD and SA at the following frequencies: a. APD data shall be measured and-processed at least once per-7 days. b. SA data shall be measured and processed at least once.per 31 days. f I d CALVERT CLIFFS - UNIT 2-3/4 4-39 Amendment No. 165 l. r_ ..,.t.,~
1 . L' : ) 3/4.41 REACTOR COOLANT SYSTEM 3/4.4.12 LETDOWN LINE EXCESS FLOW 'LINITING CONDITION FOR OPERATION 1 3.4.12 The bypass valve for the excess flow check valve in the letdown-line shall be closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the above bypass valve opan, restore the valve to its closed position within 4 hours or be in at least NOT STANDBY within the next-
~~6 hours and in COLD SHUTDOWN within the following 30 hours.~~
SURVEILLANCE REQUIREMENTS 4.4.12 The bypass valve for the excess flow check valve in the letdown line shall be determined closed within 4 hours prior to entering MODE 4 from MODE 5. 'CALVERT. CLIFFS.- UNIT 2 3/4 4-40 knendment No.J165 1:
1.
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.13 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.13 One Reactor Coolant System 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 I and 2. ACTION: a. With the reactor vessel head vent path inoperable, maintain the inoperable vent path closed with power removed from the actuator of the solenoid valves in the inoperable vent path, and: 1. If the pressurizer vapor space vent path is also inoperable, restore both inoperable vent paths to 0PERABLE status within. 72 hours or be in at least H0T STANDBY within 6 hours, or 2. If the pressurizer vapor space vent path is OPERABLE, restore the inoperable reactor vessel head vent path to OPERABLE status within 30 days or'be in at least H0T STAND 8Y 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 PORY and its. associated flow path is: i 0PERABLE within 72 hours and restore the inoperable pressurizer vapor space vent path to GPZRA8LE 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 OPERABLE status within 30 days, or be in at least HOT 'l STAND 8Y within 6 hours. c. The provisions of Specification 3.0.4 are not applicable. I 'CALVERT CLIFFS - UNIT 2
~~3/4 4-41 Amendment No. 165 l~~
j
~ tr 3l* 3/4.4 REACTOR C0OLANT SYSTEN SURVEILUUICE REQUIREMENTS 4.4.13.1.Each Reactor Coolant System vent path shall be demonstrated OPERA 8LE by testing each valve.in.the vent path per Specification 4.0.5. 4.4.13.2 Each Reactor Coolant System ' vent path shall be demonstrated 0PERA8LE at least once per REFUELING INTERVAL by: a. Verifying all manual isolation valves in each vent path are locked in the open position. b. Verifying flow through the Reactor Coolant System vent paths with the vent valves open. P b l 4 i [;. CALVERT CLIFFS - UNIT 2 3/4 4-42 ' Amendment No. 165 l
i
) 1 3/4.4 REACTOR COOLANT SYSTEM BASES 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 limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power-operated relief valve or steam dump valves. Demonstration of the safety valves' lift 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 capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path. However, the PORVs and their circuitry do not perform a safety-related function and, therefore, do not need emergency power as part of their operability requirements. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being perfomed to restore an inoperable PORV co OPERABLE status. Power is maintained to the block valve when it is closed to control excessive PORV seat leakage. This allows the PORV and block valve to remain OPERABLE should the PORV be needed to contain reactor pressure and facilitate decay heat removal during certain accident conditions. The removal of power from a closed block valve for a PORV inoperable due to causes other than excessive PORY seat leakage provides additional assurance that the block valve will not be inadvertently opened when the condition of the PORV is uncertain. RCS temperature, as used in the applicability statement, is determined as follows: (1) with the RCPs running, the RCS cold leg temperature (Tc) is the appropriate indication, (2) with the Shutdown Cooling System in CALVERT CLIFFS - UNIT 2 B 3/4 4-2 Amendment No. 165 u.
3 3/4.4 REACTOR COOLANT SYSTEM BASES operation,theshutdowncoolingtemperatureindicationisappropriate,(3) if neither the RCPs or shutdown cooling is in operation, the core exit thennocouples are the appropriate indicators of RCS temperature. The testing for transferring motive and control aower for the PORVs and block valves from the nomal to emergency power aus is done under Technical Speci fication 4.8.1.1.2.d.3. 3/4.4.4 LRESSURIZER A steam bubble in the pressurizer with the level as programed 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 programed 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 Load. The operating band also protects the pressurizer code 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 offsite power condition to maintain natural circulation at HOT STANDBY. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. 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 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. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. An engineering assessment of steam generator tube integrity will confirm that no undue risk is associated with plant operation beyond 24 months of the previous steam generator tube inspection. To provide this confimation, the assessment would demonstrate that all tubes will retain CALVERT CLIFFS - UNIT 2 B 3/4 4-3 Amendment No. 165
3/4.4 REACTOR COOLANT SYSTEM BASES adequate structural margins against burst during all normal operating, transient, and accident conditions until the end of the fuel cycle. This evaluation would include the following elements: 1. An assessment of the flaws found during the previous inspections. 2. An assessment of the structural margins relative to the criteria of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," that can be expected before the end of. the fuel cycle or 30 months, whichever comes first. 3. An update of the assessment model, as appropriate, based on comparison of the predicted results of the steam generator tube integrity assessment with actual inspection results from previous inspections. 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 steen generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may like~iy 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 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 develo) in service, it will be found during scheduled inservice steam generator tuae examinations. Plugging will be required for all tubes with imperfections 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 inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to S)ecifications 6.9.2 prior to the resumption of plant operation. Suc1 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. CALVERT CLIFFS - UNIT 2 B 3/4 4-4 Amendment No. 165 l
p 3/4.4 JtEACTOR COOLANT SYSTEM BASES 3/4.4.6 R_EACTOR 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 Leakaqe, q Industry experience has shown that while 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. 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. The 100 gallon-per-day leakage limit per steam generator ensures that steam generator tube integrity is maintained in accordance with the reconnendations of Generic Letter 91-04. 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 LEAKAl.c requires the unit to be promptly pl,' iced 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 potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associatedeffects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess ~of the Steady State Limits, up to the Transient Limiti, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor 1 CALVERT CLIFFS - UNIT 2 B 3/4 4-5 Amendment No. 165 l 1 i
I krj J 3/4.4 REACTOR COOLANT SYSTEM BASES Coolant System. The time interval pennitting 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 concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific ^ ivity of the primary coolant ensure that the resulting 2 hour doses at the MTE B0UNDARY will not exceed an ap)ropriately small fraction of Part 100 limits following a steam generator tuae 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 parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters 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 Ci/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4.8-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4.8-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels allowed by Figure 3.4.8-1 increase the 2 hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture. Reducing T, to < 500 F prevents the release of activity should a steam generator t'ube rupture since the saturation pressure of the primary 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 coolant will be detected in sufficient time to take corrective action..Infomation obtained on iodine saiking will be used to assess the parameters _ associated with spiking plenomena. A reduction in frequency of isotopic analyses following power changes may be pennissible if justified by the data obtained. L CALVERT CLIFFS - UNIT 2 B 3/4 4-6 Amendment No. 165 l
3/4.4 REACTOR C0OLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and STARTUP and shutdown operation. The various categories of load cycles used for design purposes are provided in Section 4.1.1 of the UFSAR. During STARTUP and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. Operation within the appropriate heatup and cooldown curves assures the integrity of the reactor vessel against fracture induced by combinative themal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and even more restrictive Pressure / Temperature limits-must be observed. The current limits, Figures 3.4.9-1 and 3.4.9-2, are for up to and including a 8 fluence of 1.92x10 n/cm at the inner surface of the reactor vessel, which corresponds to approximately 13.8 Effective Full Power Years. The reactor vessel materials have been tested to detemine their initial RT,,,; the results of these tests.are shown in Section'4.1.5 of the UFSAR.: Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the'RT The actual shift in RT o' of the vessel-material will be establisheIr. periodically during opera,tlon by removing and' evaluating reactor vessel material irradiation surveillance specimens. installed near the inside wall of the reactor vessel in the core area. 'The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of 10 CFR Part 50 Appendix H. The shift in the material fracture toughness, as represented by RT,or,fis calculated using Regulat e Guide 1.99, Revision 2. For a fluence o l 1.92x10" n/cm', at the 1/s T position, the adjusted reference temperature (ART) value is less than 171 F. At the 3/4 T position the ART value is 125 F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code, Section III, 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 isothemal,1/4 T heatup, and 3/4 i heatup pressure-temperature limits are compared for a given themal 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. 1 CALVERT CLIFFS - UNIT 2 B 3/4 4-7 Amendment No. 165 l
3/4.4 REACTOR COOLANT SYSTEM BASES To develop a composite pressure-temperature limit for the cooldown event, 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 allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline. Both 10 CFR Part 50, Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be detennined by entering the curve at.the test pressure (1.1 times nonnal operating pressure) and locating the corresponding temperature. This curve is shown for a fluence of r 1.92x10" n/cm on Figures 3.4.9-1 and 3.4.9-2. Similarly,10 CFR Part 50 specifies that core critical limits be established based on material considerations. This limit is shown on the heatup curve, Figure 3.4.9-1. Note that this limit does not consider the core reactivity safety analyses that actually control the temperature at which the core can be brought critical. The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia). This temperature is defined as equal to the most limiting RT,or for the balance of the Reactor Coolant System components plus 100 F, per 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 20% of the pre-operational hydrostatic test pressure. The change in the line at 150 F 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 belo s 150 F. 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,o7 for the material of the higher stressed region of the reactor vessel plus any effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was detennined using the certified material test reports and Branch Technical Position MTEB 5-2. The maximum initial RT,,1 associated with the stressed region of the closure head flange is 30 F. The minimum boltup temperature including temperature instrument uncertainty is 30 F + 10 F = 40 F. However, for conservatism, a minimum boltup temperature of 70 F is utilized in the analysis to establish the low temperature PORY lift setpoint. CALVERT CLIFFS - UNIT 2 B 3/4 4-8 Amendment No. 165 l
3/4.4 REACTOR COOLANT SYSTEM BASES The design basis events in the low temperature region assuming a water solid system are: An 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. The RCP start transient is a severe LTOP challenge for a water solid RCS. Therefore, during water solid operations all four RCPs are tagged out of service. Analysis indicates the transient is adequately controlled by placing restrictions on three parameters: initial pressurizer pressure and level, and the secondary-to-primary temperature difference. With these restrictions in place, the transient is adequately controlled without the assistance of the PORVs. The inadvertent actuation of one HPSI pump in conjunction with one charging pump is the most severe mass addition overpressurization event. Analyses were perfonned for a single HPSI pump and one charging pump assuming one PORV available with the existing orifice area of 1.29 in. 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 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 for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop MOV to limit instrumentation uncertainty. No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition. Comparison of the PORV discharge curve with the critical pressurizer pressure of 471.2 psia indicates that adequate protection is provided by 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. The low temperature PORV pressure lift setp(471.2 psia). A PORV setpoint of oint is set to protect the most restrictive Apaendix G pressure limit 430 psia, whic1 includes' instrumentation uncertainties and sufficient margins for PORV response time requirements necess e v for the protection of 471.2 psia, was selected. CALVERT CLIFFS - UNIT 2 B 3/4 4-9 Amendment No. 165 i
3/4.4 RJACTOR COOLANT SYSTEM BASES To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop M0V handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receiptiof a SIAS signal. Alternative actions, described in the ACTION Statement, ar6 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 includes (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge MOVs; and (3) disabling all three HPSI pumps. Three 100% capacity HPSI pumps are installed at Calvert Cliffs. Procedures will require that two of the three HPSI pumps be disabled (breakers racked out) at RCS temperatures less than or equal to 305 F and that the remaining HPSI pump handswitch be placed in pull-to-lock. 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. Exceptions are provided for ECCS testing and for response to LOCAs. A pressurizer steam volume and a single PORY will provide satisfactory control of all mass addition transients with the exception of a spurious actuation of full flow from a HPSI pump. Overpressurization due to this transient will be precluded for temperatures '305 F and less by disabling two HPSI pumps, placing the third in pull-to-lock, and by throttling the third pum) to less than or equal to 210 gpm flow when it is used to add mass to tie RCS. Note that only the design bases events are discussed in detail since the 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 follows: (1) with'the RCPs running, the RCS cold leg temperature is the approariate indication (2) with the Shutdown Cooling System in operation,. the slutdown cooling temperature indication is appropriate, (3) if neither the RCPs or shutdown cooling is in operation, the core exit thennocouples are the appropriate indicators of RCS temperature. The allowed out-of-service times for degraded low temperature nyerpressure protection system in MODES 5 and 6 are based on the guidance provided in-Generic Letter 90-06 and the time required to conduct a controlled, deliberate cooldown, and to depressurize and vent the RCS under the ACTION-statement entry conditions. 3/4.4.10 STRUCTURAL INTEGRITY l The inspection programs for the ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Yessel Code. CALVERT CLIFFS - UNIT 2 B 3/4 4-10 Amendment No. 165 l
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Ai -s hR (f 3/4.4 REACTOR COOLANT SYSTEM /q RASES 3/4.4.11 CORE BARREL MOVEMENT This specification is~ provided to ensure early, detection of-excessive core barrel movement if it should occur. Core barrel movement will be detected . / using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Spectral Analysis (SA). Baseline core barrel movement Alert Levels and Action Levels will be confinned during each reactor STARTUP test program following a core reload. Data from these detectors is to be reduced in two fonns. Root mean square (RMS) values are computed from the APD of the signal amplitude. These RMS magnitudes include variations due both to various neutronic-effects and i internals motion. Consequently, these signals alone can only provide a gross measure of core barrel motion. A more accurate assessment of core barrel motion is contained from the Auto and Cross Power Spectral Densities (PSD. XPSD), phase (6) and coherence (C0H) of these signals. These data result from the SA of the excore detector signals. A modification to the required monitoring program may be justified by an analysis of the data obtained and by an. examination of the affected parts. during the plant shutdown at the end of any fuel cycle. 3/4.4.12: LETDOWN LINE EXCESS FLOW This ' specification is provided to ensure that the bypass valve for the excess flow check valve in the letdown line will be maintained closed during plant operation. This bypass valve is required'to be closed to ensure that the effects of a pip 7 rupture downstream of this valve will not exceed the accident analysis assumptions. ' 3/4.4.13. REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the Primary System that could inhibit natural circulation - core cooling. The OPERASILITY of at least'one Reactor Coolant System vent path from the reactor vessel head and the pressurizer vapor space ensures the capability exists to perfona this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring'that a single failure of a vent valve, power supply or control' system does not prevent isolation of the vent path. ~ The function, capabilities, and tasting requirements of the Reactor Coolant ' System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. a CALVERT CLIFFS - UNIT 2 B 3/4 4-11 . Amendment No,'165 l g w}}

ML20065P098

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Dearadation means a service-induced cracking,

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