ML20125A564

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Amends 103 & 85 to Licenses DPR-53 & DPR-69,respectively, Changing Tech Specs to Reflect Surveillance Requirements for Snubbers,Clarifying Independence of ECCS & Shutdown Cooling Sys & Deleting Pressurization Curve
ML20125A564
Person / Time
Site: Calvert Cliffs  
Issue date: 05/16/1985
From: John Miller
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20125A568 List:
References
NUDOCS 8506110115
Download: ML20125A564 (62)
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Text

{{#Wiki_filter:! paug g j k UNITED STATES gy-p, NUCLEAR REGULATORY COMMISSION 3 't j WASHINGTON, D. C 20555 k u,/ +.... BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.103 License No. DPR-53 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The applications for amendment by Baltimore Gas & Electric Company (the licensee) dated September 20, 1984 and January 31, 1985, comply 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. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Comission'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 P E. The issuance of this amendment is in accordance with'10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. lDR IM BM8i7 p PDR

'. 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. OPR-53 is hereby amended' to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.103, 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. FOR THE NUCLEAR REGULATORY COMMISSION mm James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 16, 1985

's ATTACHMENT TO LICENSE AMENDMENT NO.103 FACILTIY OPERATING LICENSE NO. OPR-53 DOCKET NO. 50-317 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical.1ines indicating the areas of change. The corresponding overleaf pages are provided to maintain document completeness. Remove Pages Insert Pages 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-41 3/4 3-41 3/4 3-42 3/4 3-42 3/4 6-19 3/4 6-19 3/4 6-20 3/4 6-20 3/4 6-21 3/4 6-21 3/4 6-22 3/4 6-22 3/4 6-23 3/4 6-23 3/4 6-24 3/4 6-24 3/4 6-25 3/4 6-25 3/4 7-26 3/4 7-26 3/4 7-26b 3/4 7-26b 3/4 7-27 3/4 7-27 3/4 7-29 3/4 7-29 3/4 7-30 3/4 7-30 3/4 7-31 3/4 7-31 3/4 7-37 3/4 7-37 3/4 7-46 3/4 7-46 3/4 7-47 3/4 7-47 3/4 7-51 3/4 7-51 3/4 7-52 3/4 7-52 3/4 7-60 3/4 7-60 3/4 9-8a. 3/4 9-8a B 3/4 3-2 B 3/4 3-2 B 3/4 5-1 B 3/4 5-1

's TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT LOCATION OPERABLE 1. WIND SPEED a. Nominal Elev. 10M 1 l b. Nominal Elev 60M 1 [ 2. WIND DIRECTION a. Nominal Elev. 10t1 1 l b. Nominal Elev. 60M 1 l 3. AIR TEMPERATURE - DELTA T (10M-60M) 1 l CALVERT CLIFFS - UNIT 1 3/4 3-35 Amendment No.103

9* TABLE 3.3-10 POST-ACCIDEllT M0flITORIflG IllSTRUMEllTATIOil E MillIMUM S CilAfillELS IriSTRUMEllT OPERABLE 3 1. Deleted w e 2. Containment Pressure 2 C5 3. Wide Range Logarithmic fleutron Flux Monitor 2 -4 ~ 4. Reactor Coolant Outlet Temperature 2 5. Deleted 6. Pressurizer Pressure 2 s-7. Pressurizer Level 2 3 8. . Steam Generator Pressure 2/ steam generator 9. Steam Generator Level (Wide Range) 2/ steam generator 10. Feedwater Flow l 2 11. Auxiliary feedwater Flow Rate 2/ steam generator 12. RCS Subcooled Margin lionitor 1 [ 13. PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve OQ 14. PORY Solenoid Power Indication 1/ valve ow w. 15. Containment Water Level (Wide Range) 1 l i k ot

TABLE 4.3-10 POST-ACCIDEllT M0?lITORING IllSTRUMEllTATI0?l SURVEILLAtlCE REQUIREMEflTS n?M$ CHANilEL CHAfillEL IflSTRUMENT CHECK CALIBRATI0fl n C 7 1. Deleted w 2. Containment Pressure M R Eg 3. Wide Range Logarithmic Neutron Flux Monitor M N.A. 4. Reactor Coolant Outlet Temperature M R 5. Deleted g 6. Pressurizer Pressure M R Y 7. Pressurizer Level M R O 8. Steam Generator Pressure M R 9. Steam Generator Level (Wide Range) M R 10. Feedwater Flow { M R 11. Auxiliary Feedwater Flow Rate M R .N 12. RCS Subcooled flargin Monitor M R R R 13. PORV/ Safety Valve Acoustic Monitor fl. A. R 5 14. PORV Solenoid Power Indication fl. A. fl. A. 2 15. Containment Water Level (Wide Range) H R l I <gu "w

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TABLE 3.6-1 n? M C0ffTAlfattEllT ISOLATI0fl VALVES 5 PE!1ETRATI0ft ISOLATI0ft ISOLATI0ft VALVE ISOLATIO!1 n C

10.

CifAfiflEL IDEflTIFICATI0fl fl0. FUf1CTI0ft TIJ1E (SEC0flDS) lA SIAS A PS-5465-CV R.C. and Pressurizer Sampling <7 SIAS A PS-5466-CV D [7 g SIAS A PS-5467-CV 7 g SIAS B PS-5464-CV ~ 18 SIAS A WGS-2180-CV Contiinment Vent lleader to Waste <7 SIAS B WGS-2181-CV Gas -<7 1C SIAS A CVC-506-CV RCP Seals Controlled Bleedoff .7 SIAS B CVC-505-CV ~7 w2 m 10 flA PS-6529-SV* Post Accident Sampling, g i 1 Liquid Return to RC Drain Tank e 2A SIAS A CVC-515-CV Letdown Line <l3 SIAS B CYC-516-CV ~13 flA CVC-105 ~llA . IIA CVC-103 flA W S 2B flA CVC-517-CV Charging I'.. IIA g flA CVC-518-CV flA !!A CVC-519-CV flA = !!A CVC-435-RV flA o !!A CVC-184 flA 't i ? w

1 TABLE 3.6-1 (Continued) n> h C0flTAlf;MEllT ISOLATI0f1 VALVES PEllETRATI0ft ISOLATI0ft ISOLATIO!I VALVE ISOLATI0fl { f0. CHAfitlELS IDEfiTIFICATI0ft fl0. FUtlCTI0ft TIME (SEC0 fids) 7A flA Blind Flange ILRT flA f;A ILRT-1 f1A E 70 ftA Blind Flange ILRT flA ~ ?!A ILRT-2 flA 8 SIAS A EAD-5462-MOV Containment tiormal Sump <13 SIAS B EAD-5463-MOV <l3 9 FIA SI-340 Containment Spray flA e flA SI-326 f1A E$ 10 HA SI-330 Containment Spray flA f1A SI-316 g flA 13 SIAS A, CRS A CPA-1410-CV(3) Purge Air Inlet <7** F SIAS B, CRS B CPA-1411-CV(3) 77** a aa O w

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TABLE 3.6-1 (Continued) n?M C0!!TAlf;ItEf1T ISOLATIO!3 VALVES 5 PEfiETPATIO!! ISOLATIOt ISOLATI0il VALVE ISOLATI0ft n E FIO. CHAR:fiELS IDE!!TIFICATIO!1 i;0. FUtiCTIOf! TIME (SECO? IDS) ? 6' 14 SIAS A, CRS A CPA-1412-CV (3) Purge Air Outlet <7** c -77** SIAS B. CP.S B CPA-1413-CV (3) 5 15 SIAS A RE-5291-CV Purge Air tionitor <7 SIAS B RE-5292-CV {7 16 CIS A CC-3832-CV Component Cooling Water Inlet <18 18 CIS B CC-3833-CV Component Cooling Water Outlet <18 ~ ? U 19A FIA IA-337 Instrument Air flA CIS A IA-2080-MOV <13 198 NA PA-1040* PikotAir f1A flA PA-1044* flA Ng 20A ftA N -344 flitrogen Supply riA 2 g flA fl -612-CV* riA 2 N -622-CV* !{A g NA 2 N -632-CV* ttA HA 2 = ilA fl2-642-CV* tiA _s ~. 5 w

y TABLE 3.6-1 (Continued) n?g CONTAInftENT ISOLATI0fl VALVES 5: PENETRATION ISOLATION ISOLATION VALVE ISOLATION n NO. CHANNEL IDENTIFICATI0tl NO. FUflCTION TIi1E (SECONDS) 5 203 NA N -384 flitrogen Supply flA 2 'HA N -345 NA 2 E G N -346 Nitrogen Supply im 20C MA 2 N -392 NA NA 2 23 SIAS A RCW-4260-CV R.C. Drain Tank Drains g 24 SIAS B PS-6531-SV 0xygen Sample Line g 37 NA PSW-1019 Plant Water flA NA PSW-1003 NA 38 NA DW-5460-CV* Demineralized Water HA y 39 NA SI-463 . Safety Injection Tank Test Line flA HA SI-455 NA = 5 41 NA SI-652-H0V (2) Shutdown Cooling NA g NA SI-651-MOV (2) NA 4 i ? E w

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h TABLE 3.6-1 (Continued) -c ] C0!ITAlf'f1Ef!T ISOLATI0ft VALVES E PETIETRATIOi! ISOLATIO!! ISOLATI0tl VALVE ISOLATI0ft fl0. CHAi!!IEL IDEllTIFICAT10!! fl0. FUf:CTI9ft TIME (SECOTIDS) 44 f1A FP-141-A Fire Protection flA c I:A FP-141-B flA g !:A FP-6200-MOV* f1A 47A fiA PS-6540A-SV* Ilydrogen Sample Outlet r!A !!A PS-6507A-SV* ?!A 47B TIA PS-6540E-SV* H,drogen Sample Outlet fiA { r:A PS-6507E-SV* tlA T E3 47C !!A PS-6540F-SV* Ilydrogen Sample Outlet fiA ftA PS-6507F-SY* flA Ilyrfrogen Sample Return 47D 1:A PS-6540G-SV* fiA ttA PS-6507G-SV* f1A E' e 48A SIAS-B !!P-6900-MOV Containc:ent Vent Isolation <20** y SIAS-A flP-6901-MOV 120** S st i W l ,q ? wk

[2-TABLE 3.6-1 (Continued) -c ] COITAlfatEllT ISOLATIOil VALVES P PE?iETPAT1071 ISOLAT10:1 ISOLATIOil VALVE ISOLATIOTI E r:0. CIIA*;!iEL IDEf!TIFICATIOil fiO. FUilCTIOil TIME (SECOiDS) 'm 483 ftA HP-104 Hydrogen Purge Inlet flA e r;A HP-6903-MOV t;A ii e 49A f!A PS-65408-SV* liydrogen Sample r;A FIA PS-65078-SV* ftA g 498 ftA PS-6540C-SV* Hydrogen Sample ftA itA PS-6507C-SV* TIA R a e 49C FIA PS-6540D-SV* Hydrogen Sample fiA A rtA PS-6507D-SV* t:A u 50 TIA Blind Flange ILkT fiA Blind Flange tlA flA 59 FIA SFP-170 Refueling Pool Inlet NA g tiA SFP-171 FIA a P. 60 71A ES-144 Steam to Reactor Head Laydown !!A 2 TIA ES-142 riA 1 k u e-9

l 1 o ~ TACLE 3.6-1 (Continued) h C0'iTAI!;'tEf1T ISOLATIO?! VALVES N5 PEriETPJTIC'I ISCLATIC'l ISCLATI0ft YALVE ISCL ATIO?! C TIftE (SEC0!!DS) n f;0. CHAN!;EL IDElITIFICATIO!! f:0. FU?!CTIO?1 4 61 T;A SFP-176 Pefueling Pool Outlet fA flA SFP-174 FIA j i !:A SFP-172 flA E T3 SFP-189 flA w 62 SIAS A PH-6579-M07 Contairnent fleating Outlet 113 64 FIA PH-376 Contairment Heating Inlet fiA ( 7tanual or remote manual valve which is closed during plant operation. w 1 (2) !!ay be oranad telow 300*F to establish shutdown cooling flow. o d> (3) Contaircent purg? an<l containment vent isolation valves will be shut in TiODU.1, 2, 3 and 4 per TS 3/4 6.1.7 and TS 3/4 6.1.8, respectively. f

  • Itay be open on an intermittent basis under aduinistrative control.
    • Contairnent purge isolation valves isolation times will only apply for !!0 DES 5 and 6 during g

which tice these valves may be opened. Isolation time for containment purge and containment g vent isolation valves is FIA for it00ES 1, 2, 3 and 4 per TS 3/4 6.1.7 and TS 3/4 6.1.8, respectively, during which tice these valves cust remain closed. e, Et ms t ?

C0flTAIN'1Etli SYSTEils 3/4.6.5 COMBUSTIBLE GAS CONTROL HYDR 0GEli Af1ALYZERS LIMIT! fig C0flDITION FOR OPERAT!0ft

3. 6. 5.1 Two independent containment hydrogen analyzers shall be OPERABLE.

APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen analyzer inoperable *, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLAftCE REQUIREf1ENTS 4.6.5.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST DASIS by performing a CHANNEL CALIBRATION using sample gases in accordance with manufacturers' recommenda tions.

  • During the period from itay 15 to July 15, 1983, one hydrogen analyzer may be made inoperable, at any given time, for the purpose of replacing system solenoid valves with environmentally qualified valves.

During this time, Specification 3.0.4 is not applicable to this requirement. CALVERT CLIFFS - UNIT 1 3/4 6-26 Amendment No. g 7A, 8 3

4 PLAflT SYSTEris 3/A_.7.8__ StiUBBERS LI_!!!T! fig C0flDIT!0ft FOR OPERAT!0fl 4 ) 3.7.8.1 All snubbers listed in Table 3.7-4 shall be OPERABLE. APPLICABIL!_TY: fiODES 1. 2, 3 and 4. (fl0 DES 5 and 6 for snubbers located on systems required OPERABLE in those MODES ) ACT!0ft: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status, and perform an engineering evaluation

  • per Specification 4.7.8.b. and c. on the supported component or declare the supported system inoperable and follow the appropriate ACT!0fl statement for that system.

i SUPVE!LLAtlCE REQUIREf tEtiTS 4.7.8.1 Each snubber shall be demonstrated OPERABLE by performance of the j following augmented inservice inspection program and the requirements of i Specification 4.0.5. 1 ~ a. V1 wal Inspections ) Visual inspections shall be performed ih accordance with the following schedulet fio. Inoperable Snubbers Subsequent Visual" ger inspection Period Inspection Period # 0 18 months + 25% 1 12 months 7 25% 2 6 months I 25% 3, 4 124 days I25% 5, 6, 7 62 days 1 25". 8 or more 31 days 1 25% The snubbers may be categorized into two groups: Those accessible and those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule, i 1 A documented, visual inspection shall be sufficient to meet the requirements i for an engineering evaluation. Additional analyses, as needed, shall be completed in a reasonabic period of time. The inspection interval shall not be lengthened more than two steps at a time. The provisions of Specification 4.0.2 are not applicable. CALVERT CLIFFS - Ull!T 1 3/4 7-25 Amendment flo. F 4, l l

PLAfiT SYSTEMS SURVE!LLAflCE REQUIRE!!EllTS (Continued) b. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, and (2) that the snubber installation exhibits no visual indications of detachment from foundations or supporting structures. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of estab-lishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and/or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.7.8.d. as applicable. When the fluid port of a hydraulic snubber is found l to be uncovered, the snubber shall be determined inoperable unless it can be determined OPERABLE via functional testing for the purpose of ] establishing the next visual inspection interval. t For the snubber (s) found inoperable, an engineering evaluation shall be performed on the component (s) which are supported by the snubber (s). The scope of this engineering evaluation shall be consistent with the licensee's engineering judgment and may be limited to a visual inspec-tion of the supported component (s). The purpose of this engineering evaluation shall be to determine if thetomponent(s) supported by the snubber (s) were adversely affected by the inoperability of the 3 j snubber (s) in order to ensure that the supported component remains capable of meeting the designed service. 1 c. Functional Tests At least once por 18 months during shutdown, a representative sample of 10% of the snubbers in use in the plant shall be functionally tested either in place or in a bench test.* For each snubber that l j does not meet the functional test acceptance criteria of Specification 4.7.8.d. an additional 5% of the snubbers shall be functionally tested until no more failures are found or until all snubbers have been functionally tested.

  • The Steam Generator snubbers 1-63-13 through 1-63-28 need not be functionally tested until the refueling outage following June 30, 1985.

) CALVERT CLIFFS - Uti!T 1 3/4 7-26 Amendment flo. 55,///,103 1

PLANT SYSTEfG 1 SURVE!LLANCE REQUIREMEfiTS (Continued) Snubbers identified in Table 3.7-4 as "Especially Difficult to Remove" or in "High Exposure Zones" shall also be included in the representative s ampl e.

  • In addition to the regular sample, snubbers which failed the previous l

functional test shall be rotested during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested during the next test j period. Failure of these snubbers shall not entail functional testing of additional snubbers. If any snubber selected for functional testing either fails to lock up i or fails to move, i.e., frozen in place, the cause will be evaluated j and if caused by manufacturer or design deficiency all generically i l susceptible snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated above for snubbers not meeting f 4 l the functional test acceptance criteria, j j For the snubber (s) found inoperable, en engineering evaluation shall be I i performed on the component (s) which are supported by the snubber (s). i The scope of this engineering evaluation shall be consistent with the i licensee's engineering judgment and may be limited to a visual inspec-tion of the supported component (s). The purpose of this engineering i. evaluation shall be to determine if the component (s) supported by the + snubber s were adversely affected by the inoperability of the snubber s in order to ensure that the supported con.ponent remains capable of meeting the designed service. d. Hydraulic Snubbers functional Test Acceptance Criteria i The hydraulic snubber functional test shall verify that: 1. Activation (restraining action) is achieved within the specified range of velocity or accoloration in both tension and compression. ( 2. Snubber bleed, or release rate, where required, is within the specified range in compression or tension. For snubbers specifically required to not displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

  • permanent or other exemptions from functional testing for individual snubbers in these categories may be granted by the Comnission only if a justifiable l

basis for exemption is presented and/or snubber life destructive testing was performed to qualify snubber operability for all design conditions at either the completion of their fabrication or at a subsequent date. CALVERT CLIFFS - UNIT 1 3/4 7-26a Amendment No. (.j 1

PLANT SYSTEMS SURVEILLANCE REQU_!REMENTS (Continued) e. Snubber Service Life Nonitoring* A record of the service life of each snubber, the date at which the designated service life cu nences and the installation and maintenance re' cords on which the designated service life is based shall be main-tained as required by Specification 6.10.2.m. At least once per 18 months, the installation and maintenance records for each snubber listed in Table 3.7-4 shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review." If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement, or reconditioning shall be indicated in the records. ~

  • The Snubber Service Life Program shall be fully implemented by January 1.1983.
    • The provisions of Specification 4.0.2 are applicable.

CALVERT CLIFFS - UNIT 1 3/4 7-26b Amendment No.////,103

.o TACLE 3.7-4 SAFETY RELATED HYDRAlf IC SPFJBBERS* E l -o j n SNUE8ER SYSTEli SIUEBER INSTALLED ACCESSICLE OR HIGli P1,DIATION ESPECIALLY DIFFIEULT l C NO. OTI, LOCATION AND ELEVATION I?LtCCESSIBLE 20'IE** TO Pelt)VE ] (A or I) (Yes or fio) (Yes or No) b 1-11-1 SERVICE HATER FUMP (13 Q SUCTI0rt 5' A 16 no 1-11-2 SEEVICE WATER PtMP #13 SUCTION 5' A Ho no { 1-11-4 SERVICE WATER Pt N #12 SUCTION 5' A No no ? 1-11-5 SERVICE WATER PUMP #12 U SUCTION 5' A No No f }[ 1-11-7 SERVICE WATER PUFF fil { SUCTION S' A No Ho 5 1-11-8 SERVICE WATER Pt w fil SUCTION 5' A No No O 1-11-9 SERVICE WATER PtMP fil A No No SUCTION 5' 1-11-10 SERVICE WATER HEADER FROM i TURBINE BLDG. S' A ho No -~ ,--,_mw.-_.. _,-. _. _.,. _ ~. ,-~ ,.-.,,m

l TABLE 3.7-4 I 9 g SAFETY RELATED HYDRAULIC SNUB 8ERS* E ^ SNUS8ER SYSTEM SNU88ER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT f

2 NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE q (A or I) (Yes or No) (Yes or No) 3 1-11-11 SERVICE ETER FROM CONTAIISENT COOLERS 5* A No Yes l cz U 1-11-11A SERVICE M TER PUpF SUCTION WR 5' A No No ~ i 1-11-13 SERVICE MTER FROM SPENT FUEL R POOL COOLERS 5' .A No Yes ? 1-11-14 SERVICE ETER F30t CONTAIMENT M COOLERS 5' A No No 1-11-16 SERVICE ETER PUpF DISCHARGE MADER 5' 'A No No 1 1-11-18 SERVICE ETER PWF DISCHARGE 2 HEADER 5' A No No a 1-11-18A SERVICE ETER PWF DISCHANGE HEADER S' A No No E I e. 43

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O E~ S R S U Ll EEI ?5 LI BS r C IS o 4 I SE A A A A A A A A1 A L SC A 7 U EC ( A CA 3 R C? I D AI L Y L H 5 E A D K T E i A' T 3? A OT L I E TD R CA 2 '0 UE 1 ? 1 Y DC T SH T EI N R E LT E PG P P P P M E M F LA F? M M M M O T E i A AV O t t L L U R A S I S TE J PL F* P P' P '5 F W Y SL Q' 0 5 5 S l G0 G G; G G G G iE l C' NC

R
R rR i

i G f f i O 5 I I E I I L IE i5 l' i R! R LT LD L LD LD t4 L0 l ? EA ES ON OA 0 OA 0A 0 01 S S DP OE OE 0-Ot 0E 0 0 - S i i S* AH C? CH 05 CH CH CS C E UU EL O P C i I HP I P TE TE TE TE TA Tc O ?ST I ! NG NG iG NG nY N5 R l ! f A G EC t tR tR Ex EL E1 P C Of NC N *J t

A
A nA i1 f

EO I I O C F. OH OH 0H rE O E 7 f TL TL I1 IC PC GC PC JT G: t t S C0 FO FS S t S MS MS C S Y, U0 CR CI 0I GI OI OA UR A f f I S? S0 CF CD 0D CD CU CW CF W 0 R A 1 E 2 3 4 5 6 6 8 9 1 S. S U. 5 5 5 5 5 5 5 5 5 u! 1 1 1 1 1 1 1 1 1 ? S 1 1 I l 1 l 1 1 n?ESbq{U[ R^ ~h >gk3 5.

s b 5W 2 oE u Od

  • de t o

o o o g-g = e u = t0 1 5 C 7 5: = W E C U e o @ug u e 5 3 m ( 9 E 5 wl, d$~ l M $$ E .c .c l 7 8

  • 0 5.

9 S. .b *J c m 6 c-b W" o N bh 9 ~~gn a es s D d 5E t $"pg 5 5 w 3 se = a a ~ -w w 3 5 5 5 wG + ~ h5 ~ s' ! 5 5 ~ u gg h, B d d m-n: wwee 5 m a o o E. iiii h2 M M Z Z s 4 4 4 4 CALVERT CLIFFS - Uff!T 1 3/4 7-31 Amendment flo. 12, 59, 85, 103

TABLE 3.7-4 g SAFETY RELATED HYDRAULIC SNUBBERS

  • G

- SNilBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION ~ INACCESSIBLE ZONE ** TO REMOVE p (A or I) (Yes or No) (Yes or Ho) 1-24-3 EMERGENCY DIESEL #12 EXHAUST 61' A No No 'l-2'4-3A. EMERGENCY DIESEL #12 EXHAUST 61' A .No No ~ EMERGENCY DIESEL '#11 EXHAUST 61' A No No l-24-4 24-4A EMERGENCY DIESEL #11 EXHAUST 61' A No No 1-24-5 DIESEL ~ GENERATOR #21 EXHAUST 92' A No No j-1-24-6 DIESEL GENERATOR #21 EXHAUST 62' A No No 1-24-6A DIESEL GENERATOR #21 EXHAUST 62' A No No 1-28-1 UNIT 1 AFW PUMP ROOM 22' A No ho 1-28-2 UNIT 1 AFW PUMP ROOM 22'- A No No 1-28-3 UNIT 1 AFW PUMP ROOM 22' A No No 6 g l-41-l' SUCTION #13 CHARGING PUMP -10' A No No k~ 1-41-2 AUX.. SPRAY 65' I Yes No z P 1-41 AUX. SPRAY 65'. I Yes No R

4. TABLE 3.7-4 f SAFETY RELATtD HYDRAULIC SNUBBERS

  • 3, H-

-n: ~ SNUBBER . SYSTEM SNUBBER INSTALLED ACCESSIBLE OR IllGil RADIATION ESPECIALLY DIFFICULT k -NO. ON,' LOCATION AND-ELEVATION INACCESSIBLE ZONE ** 10 REMOVE ]n! (A or I) (Yes or No) (Yes or No). ,gc 5- -[ 4,'l-52' SD COOLING UPSTREAM ISOLATION -VALVE 25' 'I- ~ Yes No- ~1-52-25 i SD COOLING-UPSTREAM ISOLATION VALVE 20' I' Yes No- 'l-52-25A~ .50 COOLING UPS1 REAM ISOLATION

g-VALVE;20' I

Yes flo .s N 1-52-25B ~SD COOLING UPSTREAM ISOLATION ]: -VALVE 20' I Yes No .~. 1-52-26 ' . VALVE 31' Yes No 50 COULING UPSTREAM.ISOLAlION -I-52-26A SD COOLING UPSTREAM ISOLAIION VALVE 31' I Yes No 1-52-27 SD COOLING. UPSTREAM ISOLATION VALVE 31'. I Yes No y3 1-52-28 .50 COOLING UPSIREAM ISOLATION h-VALVE 31' I Yes No F m 5-1-b2-28A SD COOLING UPSIREAM ISOLATION g' VALVL 31' f Yes flo Es E w-i'

TABLE 3.7-4 SAFETY RELATED llYDRAULIC SNUBBERS

  • w w-n SNUBBER 1

SYSTEM SNUBBER INSTALLED' ACCESSIBLE OR HIGH RADIATI'ON ESPECIALLY DIFFICULT C NO. ON, LOCAIION AND ELEVATION INACCESSIBLE ZONE **, TO REMOVE (A or I) (Yes or No) (Yes or No)- E 12 2-29. SI TANK lIA DISCHARGE 48' I Yes No 5 y.- l-52-30 HPSI TU LOOP llA 46'~ I Yes No 1-52-31

SI TO LOOP 11 A 48' I

Yes No 1-52 SI TO LOOP 1IA (UPSTREAM MOV .614) 48' I Yes No l-52-32A SI T0-LOOP llA (UPSTREAM M0V 614)48'- I Yes No -l-52-33 -SI TANK,12A DISCHARGE 55' I Yes No oc. I-52-34 SI IANK 12A DISCHARGE 56' I Yes No l-52-34A SI ' TANK 12A DISCHARGE 56' I Yes No 1-52-35 SI TO LOOP 12A~(UPSTREAM MOV 634) 48' I Yes No 1-52-36 SI TO LOOP 12A.48' I Yes No 1-52-37 S1 TO LOOP 12A 48' .I Yes No ~ 5 1-S2-37A SI TO LOOP 12A 48' I Yes No g .I l-52-38 SI TO LOOP 12A 48' 'I Yes No 8 iy' t-4--.

.m, .2. l <{._ TABLE 3.7-4: .,f2 ~ SAFETY RELATED flYDRAULIC SNUBBERS

  • h s.. SNUBBER l I

~ . SYSTEM SNUBBER INSTALLED ' ACCESSIBLE OR. IIIGH RADIATION. - ESPECIALLY-DIFFICULT.- E;X -NO. ON, LOCATION AND ELEVATION -INACCESSIBLE Z0flE** - TO REMOVE- .+q[ (A or 1) (Yes or No) (Yes or No)- 1-54 ERV-404 DISCHARGE 89'- I Yes No 11-54-14. RECIRC. RELIEF VALVE DISCH. .T0 QUENCH ' TANK 28 I Yes No , l-54-15, RECIRC.: RELIEF VALVE DISCH. . TO QUENCH TANK 28' I Yes No I' 1-54 RECIRC. RELIEF VALVE DISCH. y T0 QUENCH I ANK 40' I Yes No - 54-17: - RECIRC.; RELIEF VALVE DISCH. TO QUENCH TANK'49'. I Yes. No .l-54-18 ERV-404-RV:201 DISCH. HEADER 89 t Yes No { 1-54-19 RV 20l!0ISCHARGE 89'- f' ~ Yes No [ l-54-20 RV 200, RVL201 SEAT-4 - DRAIN HEADER 83' .1. Yes No J. p- ~

. E.

-I-54-21' RV.200, RV201 SEAT 8' . DRAIN HEADER-/7'- 1 Yes No ~$ ^[ D i t 23,. ' PRESSURIZER RELIEF 90' 1-Yes No l-r X $. I J ? t ,s .---n

,g TABLE 3.7-4 3 SAFETY RELATED flYDRAULIC SNUBBERS

  • E

-P SNUBBER ' SYSTEM SNUBBER INSTALLED-ACCESSIBLE'OR HIGil RADIATION ESPECIALLY DIFFICULT-E- NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE 'y (A or I) (Yes or No)- (Yes or No) Ew -l m. l., 1-60-3 SERVICE WATER TO CONTAINMENT' COOLER #12, 65' I Yes No 1-60-4 SERVICE WATER TO CONTAINMENT COOLER #14 66' I 'Yes. No M l ^ rs 1-60-6 SERVICE WATER FROM CONTAINMENT ,f COOLER #12, 53' I Yes No 1 .g: l g P- .5- .w

Q-TABLE 3.7-4 SAFETY RELATED llYDRAULIC SilVBBERS* Py SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR llIGli RADIATION ESPECIALLY DIFFICULT g NO. ON, LOCATI0f1 AND ELEVATION INACCESSIBLE Z0flE** TO REMOVE (A or 1) (Yes or flo) (Yes or flo) l-60-8 SERVICE WATER FROM C0flTAINMENT H ' COOLER #13 66' I Yes flo 1-60-9 SERVICE WATER TO CONTAINMENT COOLER #11 44' I Yes No R u N 1. l-60-10 SERVICE WATER TO CONTAINMNET COOLER #11 44' I Yes No 1-60-10A SERVICE WATER T0, CONTAINMENT C00LER #11 43' I Yes No l-60-11 SERVICE WATER FROM CONTAINMENT COOLER #1143' I Yes No 1-60-llA. SERVICE WATER FROM CONTAlflMENT -. 5 COOLER #11 43 I Yes No 'k-1-60-12 SERVICE WATER FROM CONTAINMENT COOLER #11 43' I Yes No ow 'l-60-12A -SERVICE WATER FROM CONTAINMENT j COOLER #11 43' I Yes No

i-TABLE 3.7-4 g 4 . Ci SAFETY RELATED HYDRAULIC SNUBBERS

  • El'

.a EL SNUBBER - SYSTEM SNUBBER INSTALLED ' ACCESSIBLE OR' HIGH RADIATION ESPECIALLY DIFFICULT-G- N0. ON,' LOCATION AND ELEVATION INACCESSIBLE ZONE **' TO REMOVE-- Lvl (A or I). (Yes.or No) (Yes or. No) k ?l-60-18 SPRAY TO CONTAINMENT CHARC0AL. )) FILTER.#11 60' I Yes Ho .1-60-20 SPRAY TO CONTAINMENT CHARC0AL FILTER #11 60' I Yes No he b I e 4 E.. , e~ -g.... .a 'r+ <O '.76 mk. - .l

se.<w kee e e's c',m',

,eeeewe -a

o EBLE 3.7-4 ] SAFETY RELATED HYORAULIC SHUBBERS* P- .M' SNUBBER ~ ' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR-HIGil RADIATI0ft ESPECIALLY DIFFICULT l-s NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE = g' (A or I). ~~~(Yes or No) (Yes or No) 61-17 C0flTAltlMENT SPRAY D/STRM S/D !!/X -15' A-No No 1-63 S.G. #11 BLOWDOWN ORIFICE LINE 70' I Yes No 1-63-10: S.G. #11 BLOWDOWN ORIFICE BYPASS 78' I Yes No 1-63 NITROGEN to S.G. #12 74' I Yes No 1-63-12 NITR0 GEN to S.G. #12 69' I Yes . No 'l-63-13 STEAM GENERATORS 75' I Yes No 50 1-63-14 STEAM GENERATORS 75' I Yes No 1-63-15 STEAM GENERATORS 75' I-Yes No 1-63-16 STEAM GENERATORS 75' I-Yes No 1-63-17 STEAM GENERATORS 75' I Yes No 4- .R-8- n 0, E;Et w-

TABLE 3.7-4 9-G SAFETY RELATED llVDRAULIC SNUBBERS

  • E' w.

P SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT M N0. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE 3 (AorI[-- (Yes or No) (Yes or No) S 'l-63-18 STEAM GENERATORS 75' I Yes No ~ 1-63-19 STEAM GENERATORS 75' I Yes No 1-63-20 STEAM GENERATORS 75' I Yes No 1-63-21 STEAM GENERATORS 75' I Yes No 1-63-22: STEAM GENERATORS 75' I Yes No 1-63-23 STEAM GENERATORS 75'. I Yes No R 1-63-24 STEAM GENERATORS 75' I Yes No e ? 1-63-25 STEAM GENERATORS 7s' I-Yes No-E 1-63-26 STEAM GENERATORS 75' 'I-Yes No 1-63-27 STEAM GENERATORS 75' I Yes No 1-63 STEAM GENERATORS 75' I Yes No [ 1-64 LINE TO PRESS. RELIEF MOV-403 81' I Yes No i.f k 5

i TABLE.3.7-4 g.. SAFETY RELATED HYDRAULIC SNUBBERS *' G. S' UBBER J-SYSTEM SNUBBER INSTALLED-ACCESSIBLE OR-IIIGli RADIATION. ESPECIALLY. DIFFICULT-S N ~] NO. '0N, LOCATION AND ELEVATION INACCESSIBLE: ZONE **- .TO REMOVE (A or I) (Yes or No) ,(Yes or'No)- .1-83 19 MAIN' STEAM LINE 27'~ PENETRATION ]

TUNNEL:

I No Yes. 'l'. 27 l'-83 20 MAIN STEAM LINE 27' PENETRATION 1 TUNNEL ~ I No Yes ] 1-83-21! MAIN STEAM LINE 27' PENETRATION TUNNEL I No Yes. l 1-83-22. MAIN STEAM LINE 27' PENETRATION g TUNNEL I No Yes l ,s. .1-83 MAIN STEAM'LINE 27'. PENETRATION j .. TUNNEL I-No Yes l 1-83 #11 AFPT STEAM SUPPLY HEADER 12' ,'A No No .1-83 MSIV -#12 IlYDRAULIC SUPPLY 5' -A No No t 1-83-31: '. MSIV., #12 HYDRAULIC. SUPPLY 5' A No No-E. ~ 1-83-32 MSIV #12 HYDRAULIC SUPPLY. S' A No No l g:.

1-83 -MSIV.#12 HYDRAULIC SUPPLY 5' A

No No 1-83-34. MSIV #12 IlYDRAULIC SUPPLY 27' A No No 1-83-35

MSIV #12 IlYDRAULIC SUPPLY 27' A

No No ,1-83-36 MSIVc#12 liYDRAULIC SUPPLY 27' A No No 1-83-37 MSIV #12 IlYDRAULIC SUPPLY 27' A No No

TABLE 3.7-4 SAFETY RELATED llYDRAULIC SNUBBERS

  • g G

E -SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT [ NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **- TO REMOVE h '(A or I). (Yes or No) ,(Yes or No) 1-83-38 MS FROM S.G. #12 TO AUX. FEED PUMP 27' A No No z. 4 1-83-40 -MS FROM S.G. #12 TO AUX. FEED PUMP 27' A No No 1-83-40A MS FROM S.G. #12 TO AUX. FEED PUMP 27'- A No No R a STEAM SUPPLY'TO AUX. FEED-7 'l-83-44 PUMP 27' A No No - t 'l-83-47' AUX, FEED PUMP 150. VALVE BYPASS 27' A No No 1-83 AUX. FEED. PUMP 150. VALVE E BYPASS 27' A-No No 1-83-49 MAIN STEAM LINE. ENCAPSULATION 27' A No Yes 1-83-50 MAIN STEAM'LINE ENCAPSULATION 27 .A. No Yes 1 51 MAIN STEAM LINE ENCAPSULATION 27' A No Yes

k J_',

1-83-52 MAIN STEAM LINE ENCAPSULATION 27' A No. Yes 1-83-53 MAIN STEAM LINE ENCAPSULATION 27' A No Yes-1-83-54 MAIN STEAM LINE ENCAPSULATION 27' A No No e

REFUELIf!G OPERATIONS SHUTDOUN COOLING AND COOLANT CIRCULATION ,LIMITIliG CONDITI0t! FOR OPERATION 3.9.8.2 Two (2) shutdown cooling loops shall be OPERABLE *#. APPLICABILhTY: Mode 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet. ACTION: a. With less than the required shutdown cooling loops OPERABLE, initiate corrective action to return loops to OPERABLE status within one hour. b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.8.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

  • Normal or emergency power source may be inoperable for each shutdown cooling loop.
  1. 0ne shutdown cooling loop may be replaced by one spent fuel pool cooling loop when it is lined up to provide cooling flow to the irradiated fuel in the reactor core and the heat generation rate of the core is below the heat removal capacity of the spent fuel pool cooling loop.

CALVERT CLIFFS - UNIT 1 Amendment No. 77,103

3/4.3 INSTRUMENATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combi-nation therof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters. .The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The integrated operation of each of these systems is consistent l with the assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance. tests per-formed at the minimum frequencies are sufficient to demonstrate this capability. The measurement of response time at the specified frequencies pro-vides assurance that the protective and ESF action function associated i with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, over-lapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite. test measurements or 2) utilizing replacement sensors with certified response times.. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1' RADIATION MONITORING INSTRUMENTATION The OPERABILITY-of the radiation monitoring channels ensures that

1) the radiation. levels are continually measured in the areas served L

CALVERT CLIFFS - UNIT l' B 3/4 3-1

INSTRUMENTATION BASES bytheindividual.channelsand2[thealarmorautomaticactionis initiated when the radiation level trip setpoint is exceeded. 3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that suffi-cient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility and is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes", April 1974. 3/4.3.3.4. METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide'1.23, Rev.1 (Proposed)'" Meteorological Programs in Support of Nuclear Power Plants," September 1980. 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. CAL' VERT CLIFFS - UNIT 1 B 3/4 3-2 Amendment No.103

3/4.5 EMERGEllCY CORE C00LIflG SYSTEMS (ECCS) BASES 3/4.5.1 SAFETY INJECTI0ft tat 4KS The OPERABILITY of each of the RCS safety injection tanks ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the safety injection tanks. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe rt ]tures. The limits on safety injection tank volume, boron concentratisn and pressure ensure that the assumptions used for safety injection ta ik injection in the accident analysis are met. l The safety injection tank power operated isolation valves are considered

j. to be " operating bypasses" in the context of IEEE Std. 279-1971, which requires ut bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these safet./ injection tank isolation valves fail to meet single failure criteria, removal cf power to the valves is required. The limits forYperation with a safety injection tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA cvent occurring concurrent wi.th_ failure of an additiona_1 safety injection tank which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one safety injection tank is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two separate ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem th.ough any single failure consideration. Either subsystem operating in conjunction with the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. Portions of the low pressure safety injection (LPSI) system flowpath are common to both subsystems. This includes the low pressure safety injection flow control valve, CV-306, the flow orifice aawnstream of CV-306, and the four low pressure safety injection loop isolation valves. Although the portions of the flowpath are common, the system design is adequate to ensure reliable ECCS operation due to the short period of LPSI system operation following a design basis Loss of Coolant Incident prior to recirculation. The LPSI system design is consistent with the assumptions in the safety analysis. CALVERT CLIFFS - UNIT 1 B 3/4 5-1 Amendment lio.103

EMERGENCY CORE COOLING SYSTEMS BASES The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basesment is provided to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provides thjs protection by dissolving in the sump water and causing its final pH to De raised to > 7.0. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be main-tained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses,.aruL(3) provide an accepfitble level of total' ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The requirement to dissolve a repre-sentative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated post LOCA temperatures. 3/4.5.4 REFUELING WATER TANK (RWT) The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWT minimum volume and baron concentration ensure that 1) sufficient water is available within contain-ment to permit recirculation cooling ficw to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control. rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an' allowance for water not usable because of tank discharge line location or other physical characteristics. CALVERT CLIFFS - UNIT 1 B 3/4 5-2 Amendment No.34

f[pa st%q[k UNITED STATES y*.; g g f/;j j g NUCLEAR REGULATORY COMMISSION WASHINGTON D. C.20555 R.,..N) BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE /cendment flo. 85 License No. DPR-69 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The applications for amendment by Baltimore Gas & Electric Company (the licensee) dated September 20, 1984 and January 31, 1985 comply 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. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 4cility Operating License No. DPR-69 is hereby amended to read as folicas: 2. Technical Scecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 85, 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. FOR THE NUCLEAR REGULATORY COMtlISSION A u bb L / James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 16, 1985

k ATTACHMENT TO LICENSE AMENDMEllT f!0. 85 FACILTIY OPERATING LICEllSE NO. DPR-69 00CKE.' NO. 50-318 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are provided to maintain document completeness. Remove Paoes Insert Paoes 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-41 3/4 3-41 3/4 3-42 3/4 3-42 3/4 4-25 3/4 4-25 3/4 6-19 3/4 6-19 3/4 6-21 3/4 6-21 3/4 6-22 3/4 6-22 3/4 6-23 3/4 6-23 3/4 6-24 3/4 6-24 '3/4 6-25 3/4 6-25 3/4 9-8a 3/4 9-8a B 3/4 3-2 B 3/4 3-2 B 3/4 5-1 B 3/4 5-1 4 w

TABLE 3.3-8 METEOROLOGICAL MONITORING Il4STRUMEflTATION MIfilMUM CHAf!NELS IllSTRU"ENT LOCATION OPERABLE 1. WIfiD SPEED a. flominal Elev.10M 1 { b. Nominal Elev. 60M 1 l 2. WIllD DIRECTION a. Nominal Elev. 10M 1 l b. Nominal Elev. 60M 1 g 3. AIR TEMPERATURE - DELTA T (10M-60M) I I t CALVERT CLIFFS - UNIT 2 3/4 3-35 Amendment No. 85

TABLE 4.3-5 9-METEOROLOGICAL MONITORING INSTRUMEllTATION SURVEILLAtlCE REQUIREMENTS G. g^ CHANilEL CHANNEL INSTRUMENT CHECK CALIBRATI0ft 'M3 1. WIND SPEED b a.. Nominal -Elev. 10M D SA l s m i i .b. Nominal .Eley. 60M D SA I K 2. ' WIND DIRECTION c y a. Nominal Elev. 10M D SA l ] .b. Nominal i Eley. 60M i D SA l g i 3. ~ AIR. TEMPERATURE - DELTA T (10M-60M) D SA I a E. 9 l ~-

r TABLE 3.3-10 POST-ACCIDEtlT MONITORING lilSTRUMErlTATI0rt N MIrllMUM CilANNELS-OPERABLE

p

' INST MENT R 1. ._ Containment Pressure 2 c: 2.- Wide Range Logarithmic Neutron Flux Monitor 2- =* 3. ' Reactor Coolant Outlet Temperature 2 [ m 4. -Pressurizer Pressure 2 5. Pressurizer Level 2 s 6. _ Steam Generator Pressure 2/ steam generator

7..

Steam Generator _ Level (Wide Range) 2/ steam generator 8. Auxiliary Feedwater Flow Rate 2/ steam generator w b -9.

RCS Subcooled Margin-Monitor 1

10. .PORV/ Safety Valve Acoustic F ow Monitoring 1/ valve L. 11. PORV Solenoid Power Indication < 1/ valve W 12. _Feedwater Flow. 2 i a 13. Containment Water Level (Wide Range) 1 l .E 3 ~ " ' ' E, e -. -.c

e. - - - --

e

TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS' 9 e-75 CHANNEL CllANNEL 5 INSTRUMENT' CilECK CALIBRATION O h 1. Containment' Pressure M R 2.. Wide Range Logarithmic Neutron Flux Monitor M N.A. E '3. Reactor Coolant Outlet Temperature M R co - - 4. Pressurizer Pressure M R 5. Pressurizer Level M R 6. Steam Generator Pressure M R '7. Steam Generator Level.(Wide. Range) M R s-8.. Auxiliary Feedwater Flow Rate M R Y0

9..

RCS Subcooled Margin Monitor-M R 10. PORV/ Safety Valve Acoustic Monitor { N.A. R 11. PORV Solenoid Power Indication-N.A. N.A. 'F 12. .Feedwater Flow M R ~ ,$ ;13. Containment Water Level (Wide Range) M R l 5 l g-

L f i DELETED 2 i I i CALVERT CLIFFS - UNIT 2 3/4 4-25 Amendment No.85 t t

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  • FOR SDC OPERATIOa #

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~~ - - + - ~~"~~ 0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T. 'F c FIGURE 3.4-2b Reactor Coolant System Pressure Temperature Limitations for 2 to 10 Years of Full Operation CALVERT CLIFFS-UNIT 2 3/4.4-25a L

r TABLE 3.6-1 g r-M C0flTAlftf1EllT ISOLATI0ft VALVES W PEilETRATI0ft ISOLATI0ft ISOLATI0fl VALVE ISOLATIGft n C fl0. CilAfillE_L_ IDEflTIFICATI0rt fl0. FilflCTIOil IIf1E (SEC0flD_S1 m 1A SIAS A PS-5465-CV R.C. and Pressurizer Sampling -<7 SIAS A PS-5466-CV SIAS A PS-5467-CV Z SIAS B PS-5464-CV ~ IB SIAS A flGS-2180-CV Containment Vent licader to Waste <7 SIAS B WGS-2181-CV Gas ~7 1C SIAS A CVC-506-CV RCP Seals Controlled B1eedoff <7 t' SIAS B CVC-505-CV 77 mL PS-6529SV* Post Accident Sampling 10 flA Liquid Return to RC Drain Tank flA 2A SIAS A CVC-515-CV Lgtdown Line <13 SIAS B CVC-516-CV 1 71 3 flA CVC-105 ~flA flA CVC-103 ilA N R 2B flA CVC-517-CV Charging Line flA flA CVC-518-CV tlA flA CVC-519-CV tlA g flA CVC-435-P,V tlA flA CVC-184 flA \\

i TABLE 3.6-1 (Continued) S CONTAINMENT ISOLATION VALVES 9^ -d PENETRATION ISOLATION ISOLATION VALVE ISOLATION [2 NO.

CHANNELS IDENTIFICATION NO.

FUNCTION , TIME (SECONDS) 7A NA Blind Flange ILRT NA NA ILRT-1 NA -l. EIl-7B NA Blind Flange ILRT NA h) NA ILRT-2 NA .l 8 _SIAS A EAD-5462-M0V Containment Normal Sump <13 SIAS B EAD-5463-MOV {13 9 NA SI-340 Containment Spray NA NA SI-326 NA E .10 NA SI-330 Cbntainment Spray NA NA ..SI-316 NA (- I3 - SIAS A, CRS A- 'CPA-1410-CV(3) Purge Air Inlet <7** gr SIAS B, CRS B CPA-1411-CV(3) 57** st 3e I 9

o TABLE 3.6-1 (Continuedl 9{ C0flTAlfl!!EilT ISOLATI0fl VALVES PErlETRATI0fi ISOLATI0fl ISOLATI0fl VALVE ISOLATI0fl P fl0. CilAfiflELS IDErlTIFICATIOrt fl0. FUllCTIOff~ TIf tE (SEC0!lDS) M ? 14 SIAS A, CRS A CPA-1412-CV (3) Purge Air Outlet 71 ** SIAS B, CRS B CPA-1413-CV (3) 71 ** E 15 SIAS A RE-5291-CV Purge Air f1onitor <7 SIAS B RE-5292-CV [7 1 16 CIS A CC-3832-CV Component Cooling Water Inlet 118 18 CIS B CC-3833-CV Component Cooling Water Outlet 118 as4 19A flA IA-175 Instrument Air flA ~ CIS A IA-2080-MOV 113 19B flA PA-137* flant Air flA flA PA-1044* flA k fl -347 flitrogen Supply flA 20A flA 2 l E flA fl2-612-CV* flA S flA N -622-CV* flA 2 N -632-CV* flA E flA 2 flA fl -642-CV* flA 2 =. P Dy l 4 8:

TABLE 3.6-1 (Continued) CONTAIN!!ENT ISOLATION VALVES !B PENETRATION -ISOLATION. . ISOLATION VALVE-ISOLATION P-NO. CHAtlNEL IDENTIFICATION NO. FIMCTION TIl1E (SECONDS) 'E? .20B NA .N -348 Nitrogen Supply NA 2 N -395 NA .~ NA 2 g 20C NA .N2-349 . Nitrogen Supply NA N -398 NA N NA 2 23 SIAS A RCW-4260-CV. R.C. Drain Tank Drains <7 24 SIAS B PS-6531-SV 0xygen Sample Line <7 Y 37-NA PSW-1020 Plant Water NA NA PSW-1009 NA cn b '38 NA' DW-5460-CV* D anineralized Water NA .g - 39 NA SI-463 Safety Injection Tank Test Line NA NA SI-455 NA h -k(.NA - SI-652-M0V (2) Shutdown Cooling NA NA -SI-651-MOV (2) NA .= . ut ~. L

'./.' 'h. TABLE 3.6-1 (Continued)- <[ CONTAINf1ENT ISOLATION VALVES-P- PENETRATION ISOLATION ' ISOLATION VALVE. ISOLATI0fl.

M

'NO. CllANNEL ' IDENTIFICATION NO. FUNCTI0ft . TIME (SEC0llDS) m-w ~.44 .NA. FP-145-A Fire Protection NA - c=. NA FP-145-B. ilA y

NA-FP-6200-MOV*

NA m.. i47Al _NA. PS-6540A-SV* Hydrogen Sample 0utiet. NA NA PS-6507A-SV* NA ~478 -NA. - .PS-6540E-SV*

Hydrogen Samp1 Outlet flA NA PS-6507E-SV*

NA e 147C fiA - PS-6540F-SV* .: Hydrogen Sampie Outlet NA . cn h-

NA.

PS-6507F-SV* NA 47D 'NA: .' PS-6540G-SV

  • llydrogen :Sampl e. Return flA NA-PS-6507G-SV*

tlA F [48A-SIAS A-HP-6900-MOV Containment Vent Isolation < 20** -{ SIAS B-HP-6901-MOV i20** .S a ? 5:t 1 ~

9 TABLE 3.6-1 (Continued) G CONTAINMENT ISOLATION VALVES ,p

PENETRATION ISOLATION ISOLATION-VALVE ISOLATI0ft i

q NO. CilANNEL IDENTIFICATION NO. FUNCTI0ft TIME (SECONOS) 3: _48B -NA IIP-104 flydrogen Purge Inlet NA e NA HP-6903-MOV NA 5' -4 49A NA PS-6540B-SV* liydrogen Sample NA NA-PS-65078-SV* NA 498- .NA PS-6540C-SV* liydrogen Sample NA NA' PS-6507C-SV* NA R 49C NA PS-6540D-SV* liydrogen Sample NA A, NA PS-65070-SV* NA a 50 NA Blind Flange I RT tlA NA B1ind F1ange flA 59 NA SFP-178 Refueling Pool Inlet NA [ NA SFP-179 NA ig 60 NA ES-144 Steam to Reactor llead Laydown NA NA ES-142 NA 2 t l s - 8:

? TABLE 3.6-1 (Continued) h} CONTAINNENT ISOLATION VALVES h PENETRATION ISOLATION ISOLATION VALVE ISOLATION -d ' il0. CllANNEL IDENTIFICATION N0. FUNCTION TIME (SECONDS) P 0;- 61 NA' SFP-184 Refueling Pool Outlet NA 31 NA SFP-182 NA flA SFP-180-NA c:. NA SFP-186- .NA H. _4 62-SIAS A PH-6579-!10V Containment lleating Outlet <13 64 NA PH-387 Containment IIcating Inlet NA ui (1) Manual or remote manual valve which is. closed during plant operation. 2 (2) May be. opened below 300*F. to establish shutdown cooling flow. o3 E (3) Containment purge and containment-vent isolation valves will be shut in MODES 1, 2, 3 and 4 per TS 3/4 6.1.7 and TS 3/4 6.1.8, respectively. t May be open on an intermittent-basis _ under administrative control. Containment purge isolation valves isolation times will only apply for MODES 5 and 6 during which time these. valves may.be opened. Isolation time for containment purge and containment vent isola-2p tion valves is NA for MODES 1, 2,~3 and 4 per TS 3/4 6.1.7 and TS 3/4 6.1.8, respectively, during j[ which time these valves must remain closed. 5 a EF .g' l q i

~ CONTAINMENT SYSTEMS 3/4.6.5 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION

3. 6. 5.1 Two independent containment hydrogen analyzers shall be OPERABLE.

APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen analyzer inoperable *, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. 1 SURVEILLANCE REQUIREMENTS 4.6.5.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases in ace'ordance with manufacturers' recommendations.

  • During the period from !!ay 15 to July 15, 1983, one hydrogen analyzer may be made inoperable, at any given time, for the purpose of replacing system solenoid valves with environmentally qualified valves.

During this time, Specification 3.0.4 is not applicable to this requirement. CALVERT CLIFFS - UNIT 2 3/4 6-26 Amendment No. AZ, EE, E G 4 j

..~ REFUELING OPERATIONS j. SHUTDOW" COOLING AND COOLANT CIRCULATION-LIMITING CONDITION FOR OPERATION 3.9.8.2' Two (2) shutdown' cooling loops-shall be OPERABLE *#. ~ APPLICABILITY: MODE 6 when the water level-above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet. ACTION: a. With less than the required shutdown cooling loops OPERABLE, initiate corrective action-to return-loops to OPERABLE status within one hour. b. 'The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.8.2 No additional Surveillance Requirements other than those required by Specification 4.0.5. )

  • Normalfor emergency power source may be. inoperable for each shutdown cooling -
loop.

.f0ne shutdown cooling loop may be: replaced by:one spent fuel pool cooling loop'when it is lined up to. provide. cooling flow to.the. irradiated' fuel in-tne reactor core and the heat generation rate of the core is below the heat removal' capacity of the spent-fuel pool ~ cooling;1oop. CALVERTLCLIFFS - UNIT 2 3/4.9-8a Amendment No. 38,85

e s 3/4.3-INSTRUMENATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combi-nation therof exceeds its setpoint, 2) the specified coincidence logic is maintained,'3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements specified-for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability. The measurement of response time at the specified frequencies pro-vides assurance that the protective and ESF action function associated with each channel is1 completed within the time limit assumed in the' accident analyses. No credit was taken in the analyses _for those channels with response times indicated as not appli< sable. Response. time may be demonstrated by any. series of sequential, over-lapping or total channel test measurements provided that'such tests demonstrate the, total channel ~ response time 'as defined. Sensor response-time verification may be demonstrated'by either.1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times. 3/4.3.3 -MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that

1) the radiation levels are continually measured in the areas served L

CALVERT CLIFFS - UNIT 2 B.3/4 3-1 i-L. ag,-

l INSTRUMENTATION j BASES by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that suffi-cient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit _ comparison of the mea _sured response to that used in the design basis for. the facility and is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes", April 1974. 3/4.3.3.4. METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect-the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, Rev.1 (Proposed), " Meteorological Programs in Support of Nuclear Power Plants," September 1980. 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. CALVERT CLIFFS - UNIT 2 B 3/4'3-2 Amendment No. R5 t -w

a I 3/4.5 EMERGENCY CORE C00LIf;G SYSTEMS (ECCS) l l BASES 3/4.5.1 SAFETY INJECTION TANKS The OPERABILITY of each of the RCS safety injection tanks ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure'of the safety injection tanks. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. l The limits on safety injection tank volume, boron concentration and i pressure ansure that the assumptions used for safety injection tank injection in the accioent analysis are met. The safety injection tank power operated isolation valves are considered to be " operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protection function be removed automatically whenever permissive conditions are not met. In addition, as these safety injection tank isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for oepration with a safety injection tank inoperable for any reason except an isolation valve closed minimizes-the time exposure of-the plant to a LOCA event occurring concurrent with failure of an additional safety injection tank which may result in unacceptable peak cladding temper-atures. If a closed isolat. ion valve cannot be immediately opened, the full capability of one safety injection tank is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERxBILITY of two separate ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure. consideration. Either subsystem operating in conjunction with the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. Portions'of the low pressure safety injection (LPSI) system flowpath are common to both subsystems. This includes the low pressure safety injection flow control valve, CV-306, the flow orifice downstream of CV-306, and the four low pressure safety injection loop isolation valves. Although the portions of the flowpath are common, the system design is adequate to ensure reliable ECCS operation due to the short period of LPSI system operation following a design basis Loss of Coolant Incident prior to recirculation. The L. PSI system design is consistent with the assumptions in the safety analysis. CALVERT CLIFFS - UNIT 2 B 3/4 5-1 Amendment No. 85 L_

/* EMERGENCY CORE COOLING SYSTEMS BASES The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basesment is provided to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provides this protection by dissolving in the sump water and causing its final pH to be raised to > 7.0. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be main-tained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses., aE_(3) provide an accep_ table d level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The requirement to dissolve a repre-sentative sample of TSP in a sample of RWT water provides assurance that the stored TSP will dissolve in borated water at the postulated post LOCA temperatures. 3/4.5.4 ' REFUELING WATER TANK (RWT) The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of barated water is available for injaction by the ECCS in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that 1) sufficient water is available'within contain-ment to permit recirculation cooling flow to the core, and.2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods' inserted except for the most reactive control assembly. These assumptions are_ consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 1 CALVERT. CLIFFS - UNIT 2 B 3/4 5-2 Amendment No. 16 I}}

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