RS-20-129, License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition, Part 5 of 5

License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition, Part 5 of 5

ML20303A328
Person / Time
Site:	Byron
Issue date:	10/29/2020
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20304A147	List: ML20303A323 ML20303A325 ML20303A327 RS-20-129, License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition, Part 5 of 5
References
RS-20-129
Download: ML20303A328 (54)

Text

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program CSFDP)

!DELETED ~ This program ensures loss of safety function is detected and

...________, appropriate actions taken . Upon entry into LCD 3.0.6, an evaluation shall be made to determine if loss of safety function exists .

Additionally , other appropriate actions may be taken as a result of the sblpport system inoperability and corresponding exception to entering supported system Condition and Required Actions . -T-l+l-&

program implements the requirements of LCO 3.0.6. The SFDP shall contain the fol l m;i ng:

-a.- Provisions for cross train checks to enSblre a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected ;

ft.. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; t-.- Provisions to ensure that an inoperable supported system ' s Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

.d-.- Other appropriate limitations and remedial or compensatory actions .

A loss of safety function exists *,;hen , assuming no concurrent single failure , a safety function assumed in the accident analysis cannot be performed . For the purpose of this program , a loss of safety function may exist *,.ihen a support system is inoperable , and:

-a.- A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or ft.. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or t-.- A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable .

BYRON - UNITS 1 &2 5. 5 - 20 Amendment ~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5. 15 Safety Function Determination Program CSFDP) (continued)

The SFDP identifies where a loss of safety function exists . -+/--f--a-loss of safety function is determined to exist by this program ,

the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are reqLiired to be entered .

5.5.16 Containment Leakage Rate Testing Program ID_E_L_E_T_E_D_~ A program

.... shall be established to implement the leakage rate testing of the containment as reqLiired by 10 CFR 50 . 54(0) and 10 CFR 50 , Appendix J , Option g, as modified by approved exemptions . This program shall be in accordance with the gLii deli nes contained in ~JLicl ear Energy InstitLite OJEI) Topi cal Report CTR) NEI 94 01 , "IndListry GLii deli ne for Implementing Performance gased Option of 10 CFR 50 , Appendix J ," Revision 3 A, dated JLily 2012 , and the conditions and limitations specified in NEI 94 01 , Revision 2 A, dated October 2008 .

The peak calcLilated containment internal pressLire for the design basis loss of coolant accident , .l2a,- is 42 .8 psig for Unit 1 and 38 . 4 psig for Unit 2 The maximLim allowable contai nrnent leakage rate , ~ -a-t .l2a,- shall be 0. 20% of containment air weight per day .

Leakage Rate acceptance criteria are:

-a.- Containment leakage rate acceptance criterion is ~ -+/---.G- 4=a--.-

DLiri ng the first Linit startLip follm1ing testing in accordance '..'ith this program , the leakage rate acceptance criteria are< 0.60 -bcr fe..r. the Type Band C tests and< 0.75 La for Type A tests; and BYRON - UNITS 1 &2 5. 5 - 21 Amendment B9-

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5. 16 Containment Leakage Rate Testing Program (continued)

-8....- /\ir lock testing acceptance criteria are :

-h- Overall air lock leakage rate is ~ .GB- -6 .:hen tested 1

-a-t ~ -l2a+ -aHtl-b For each door , seal leakage rate is :

+.- < 0. 0024 ,.- ',Jhen pressblri zed to ~ 3 psi g, and

++-. < 0. 01 ,.- \.'hen presSLiri zed to ~ 10 psi g.

The provisions of £R 3.0.2 do not appl y to the test freqblencies specified in the ContainFRent Leakage Rate Testing PrograFR .

The provisions of £R 3.0.3 are applicable to the Cont ainFRent Leakage Rate Testing Program .

5.5.17 gattery Monitoring and Maintenance PrograFR IDELETED ~This program provides for restoration and maintenance , based on the recommendations of IEEE Standard 450 , "IEEE Recommended Practice for Maintenance, Testing , and Repla cement of Vented Lead Ac id gatteri es For Stationary /\ppl i cations ," or of the ba ttery manblfactblrer of the following:

A. Actions to restore battery eel ls 'di th fl oat voltage

< 2.13 V, and

.g. Actions to eqLialize and test battery cells that had been discovered ',Jith electrolyte level bel m,i the mini mLim established design limit .

5.5.18 Control Room Envelope Wabitability PrograFR

!DELETED ~A Control Room Envelope CCRE) Wabitabi l ity Program sha ll be

~---~ established and iFRpl eFRented to ens bl re that CRE habitability is maintained sblch that , ',Ji th an OPERAgLE Control Room Ventilation CVC) Filtration SysteFR , CRE occblpants can control the reactor safely under normal conditions and FRaintain it in a safe condition foll m1i ng a radi ol ogi cal event , haza rdobls chemical rel ease , or a smoke challenge . The program shall ensblre that adeqblate radiation protection is provided to perFRit access and occupancy of the CRE

~:~~fvf::i~:d~:~J:na~:~:~:!e~or:>e::::!t!~n~ ~!!h~~:a~e:~~::~Jve dose equivalent CTEDE) for the dLirati on of the accident. +fie.

program shall incl Lide the foll 0 .Ji ng el eFRents:

1 BYRON - UNITS 1 &2 5. 5 - 22 Amendment ~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Envelope Habitability Program (continued)

-a.- The defi ni ti on of the CRE and the CRE boundary .

.&... Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance .

.f..- Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance ',Jith the testing methods and at the Frequencies specified in Sections C. 1 and C. 2 of Regulatory Gui de 1.197 , "Demonstrating Control Room Envelope Integrity at ~lucl ear Power Reactors ," Revision 0, May 2003 , and (ii) assessing CRE habitability at the Frequencies specified in Sections C. l and C.2 of Regulatory Guide 1. 197 , Revision 0.

4-.- Measurement , at designed locations , of the CRE pressb1re relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the VG Filtration System , operating at the fl ow rate required by the VFTP , at a Frequency of 18 months on a ST/\GGERED TEST gASIS . The results shall be trended and used as part of the 18 month assessment of the CRE boundary .

e-.- The quantitative limits on unfiltered air inleakage into the

-GR-b- These limits shall be stated in a manner to allo'.J direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flo\J rate assumed in the licensing basis anal yses of ogA consequences . Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards ',Jill be ',Jithin the assumptions in the licensing basis .

.f. The provisions of SR 3. 0. 2 are appl i cable to the Frequencies for assessing CRE habitability , determining CRE unfiltered inleakage , and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively .

BYRON - UNITS 1 &2 5. 5 - 23 Amendment ~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall cont ain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program 5.5.20 Risk Informed Completion Time Program IDELETED ~ This program provides controls to cal Cbll at e a Risk Informed Completion Time (RICT) and must be impl emented in accordance with

~J EI 06 09 A, Revision 0, "Risk Managed Technical Specifications (RMTS) Gbli deli nes ." The program shall include the foll m1i ng:

.a-. The RICT may not exceed 30 days;

-9. A RICT may onl y be utilized in MODE 1 and 2; s- ..Jhen a RICT is being used , any change to the pl ant 1

configuration change , as defined in ~JEI 06 09 A, Appendi x A, must be considered for the effect on the RICT .

1. For planned changes , the revised RICT must be determined prior to impl ementation of the change in configuration .

2. For emergent conditions , the revised RICT mblst be determined \Jithin the time l imits of the Required Action Completion Time (i .e., not the RICT) or 12 hou rs afte r the pl ant configuration change , .:hi chever is less .

1

3. Revising the RICT i s not required if the plant configuration change \JObll d l m1er pl ant risk and \JObll d resu l t in a l onger RICT .

BYRON - UNITS 1 &2 5. 5 - 24 Amendment l Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.20 Risk Informed Completion Time Program (continued)

-d--.- For emergent conditions , if the extent of condition evaluation for inoperable structures , systems , or components CSSCs) is not complete prior to exceeding the Completion Time , the RIGT shall account for the increased possibility of common cause failure (CCF) by either :

1. ~lumeri call y accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions CRMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs , and , if practicable , reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs .

.e. The risk assessment approaches and methods shall be acceptable to the ~JRC . The plant PRA shall be based on the as built , as operated , and maintained plant; and reflect the operating experience at the plant , as specified in Regulatory Guide 1.200 , Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program , or other methods approved by the ~JRG for generic use; and any change in the PR;\ methods to assess risk that are outside these approval boundaries require prior MRC approval .

BYRON - UNITS 1 &2 5. 5 - 25 Amendment ~

I NO CHANGES ON THIS PAGE I Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 (Deleted)

BYRON - UNITS 1 &2 5.6 - 1 Amendment 142

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report A single submittel mey be mede for the fecility . The submittel should combine sections common to both units .

The Annual Radiological Environmental Operating Report covering the operetion of the facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of t rends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsi t e Dose Calculation Manual CODCM), and in 10 CFR 50, Appendi x I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Report A single submittel mey be rnede for the fecility . The submittel shell combine sections common to both units .

The Radioactive Effluent Release Report covering the operetion of

.the-facility during the previous year shall be submi t ted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioact ive liquid and gaseous effluents and solid waste released from t he facility.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50 .36a and 10 CFR Part 50, Appendi x I,Section IV.B.l.

5.6.4 (Deleted)

BYRON - UNITS 1 &2 5.6 - 2 Amendment ~

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATHIG LIMITS REPORT CCOLR) ~

-a.- Core operating limits shall be established prior to each reload cycle , or prior to any remaining portion of a reload cycle , and shall be doCblmented in the COLR for the foll m:i ng:

SL 2. 1.1 , " Reactor Core SLs " ;

LCO 3. i.1 , "SHUTDrn*m MARG rn csoM) " ;

LCO 3 . 1. 3, "Moderator Temperatblre Coefficient " ;

LCD 3. 1.5 , "Shbltdmm Bank Insertion Limits " ;

LCO 3. 1.6 , "Control Bank Insertion Limits " ;

LCD 3 . 1. 8, " PHYSICS TESTS Exceptions MODE 2" ;

LCD 3. 2 . 1 , "Heat Fl blX Hot Channel Factor ( FttC Z)) " ;

LCO 3 . 2 . 2 , "Mblclear Enthalpy Rise Hot Channel Factor fft+ ) " ;

LCO 3 . 2 . 3 , "AXI/\L FLUX DIFFERE~lCE (/\FD) " ;

LCO 3 . 2 . 5 , "Departblre from ~lblcleate Boiling Ra t io CIJMBR) " ;

LCO 3 . 4. 1 , " RCS Pressblre , Temperatblre , and Flm1 Departblre from Mblcl eate Boiling ( DMB) Li mi ts " ; and LCO 3. 9 . 1, "Boron Concentration " ;

-&-.- The analytical methods blsed to determine the core operating limits shall be those previoblsly reviewed and approved~,

the MRC , specifically those described in the following doCblments:

h ACAP 9272 P A, "ti.1esti n9hoblse Reload Safety tval blati ans 1

Methodology ," Jblly 1985 .

b \*.'CAP 12472 PA , "BtACml Core Monitoring and Operations Sblpport System ," Ablgblst 1994 .

~ ~JFSR 0016 , "CoFFlmomi'ealth tdi son Company Topi cal Report on BenchFF1ark of P1,.JR ~lbicl ear Desi §n Methods , Jblly 11 1983 .

4-.- ~JFSR 0081 , "CoFF1FF1ommalth tdi son CoFF1pany Topi cal Report on Benchmark of PWR Mblclear Design Methods Using the Phoenix P and MlC CoFFlpblter Codes ," Jbll y 1990 .

.§..... Comtd letter from D. Saccomando to the Office of Mbiclear Reactor Regbllation dated December 21 , 1994 ,

transmitting an attachment that documents applicable sections of WCAP 11992/ 11993 and Comtd application of the UtT methodology addressed in "Additional Information Regarding /\pplication for Amendment to Facility Operating Licenses Reactivity Control Systems . 11 BYRON - UNITS 1 &2 5.6 - 3 Amendment -+/--%

Reporting Requirements 5.6 5.6 Reporting Requirements

5. 6. 5 CORE OPER'\THlG LIMITS REPORT CCOLR) (continued)

-9.- \*JC'\P 16009 P ,'\ , Revision 0 , "Realistic Large Break LOC'\

Evaluation Methodology Using the Automated Statistical Treatment of Uncertai nt:Y Method C'\STRUM) ," January 2005.

+.- \*JCAP 10079 P ,'\ , "NOTRUMP , A Nodal Transient Small Break and General ~l et 1.mrk Code ," August 1985 .

&. \*ICAP 10054 P A, " 1..Jesti nghouse Small Break EGGS Evaluation Model using ~l OTRUM P Code ," August 1985 .

.- \*!C'\P 10216 P A, Revision 1, "Rela xation of Constant Axial Offset Control ~Surveillance Technical Specification ," February 1994 .

-W-. \*!C'\P 8745 P A, "Design Bases for the Thermal Overpm*1er AT and Thermal Overtemperature AT Trip Functions ,"

September 1986 .

1-. WC/\P 14 565 P /\ , "VI PRE 01 Modeling and Qblal ifi cation for Pressurized Water Reactor Non LOCA Thermal Hydr ablli c Sa f ety /\nal ys i s ," Oct ober 1999 .

-+/-2-. \*!C'\P 12610 P ,'\ , "VANTAGE+ Fuel Assembly Reference Core Report ," April 1995 , C'i'iesti nghouse Proprietary) .

-hl-. \*JCAP 12610 P ,'\ & GEM PD 404 P A, Addendum 1 A, "Optimized ZIRLO'," July 2006 , ('i'iestinghouse Proprietary) .

t.- The core operating limits shall be determined such that all applicable limits (e . g., fuel thermal mechanical limits ,

core thermal hydraulic limits , Emergency Core Cooling Systems (EGGS) limits , nuclear limits such as SOM , transient anal ysis limits , and accident anal ysis limits) of the safety analysis are met ; and 4.- The COLR , including any midcycle revisions or supplements ,

shall be provided upon issuance for each reload cycle to the

1. RC-.

BYRON - UNITS 1 &2 5.6 - 4 Amendment -+/-%

Reporting Requirements 5.6 5.6 Reporting Requirements

5. 6. 6 Reactor Cool ant System (RCS) PRESSURE MID TEMPERATURE LIMITS

~ REPORT (PTLR)

(Deleted) .a-. RCS pressblre and temperatblre limits for heat blp , cool dm*m, l 0 11 ternperatblre operation , criticality , and hydrostatic 1

testing as ',,iel l as heatblp and cool dm:n rates , and PmJer Operated Relief Valve CPORV) lift settings shall be established and docl:lrnented in the PTLR for the following :

LCD 3 . ~ . 3 . "RCS Pressblre and TeRiperatl:lre CP/T) LiRiits ," and LCD 3. ~ . 12 , "Lm: Ternperatblre OverpresSblre Protection ( LTOP)

SysteRi" ;

-8..- The analytical rnethods blsed to deterrnine the RCS pressblre and temperatblre limits shall be those previoblsly reviewed and approved b:>' the NRG , specifically those described in the following docblrnents:

-+/---.- NRG letter dated Janblary 21 , 1998 , "Byron Station Units 1 and 2, and Braidwood Station , Units 1 and 2, Acceptance for Referencing of Pressblre Ternperatblre Li rnits Report ,"

b NRG letter dated Ablgblst 8 , 2001 , "Issblance of Exernption frorn the reqblirernents of 10 CFR 50 . 60 and Appendix G, for Byron Station , Units 1 and 2 and Brai d.1ood Sta ti on , Units 1 and 2,"

1 J.- \*lesti nghoblse ,./C/\P 161~3 , "Reactor Vessel Cl osblre 1

~cad / Vessel Flange Reqbli rernents Eval blati on for Byron/ Brai d.1ood Uni ts 1 and 2,"

1 4-.- NRG let ter da t ed Al:lgblst 31 , 2020 , " ~kai di.:ood Station ,

Units 1 and 2, and Byron Station , Unit Nos . 1 and 2, Exemption From the Reqblirements of 10 CFR 50 . 61 and 10 CFR 50 , Appendix G (EPID L 2019 LLE 0022) ," and NRG l etter dated September 18 , 2020 , "Brai dvmod Station ,

Units 1 and 2, and Byron Station Unit Nos . 1 and 2 Issblance of Amendment Nos . 217 , 217 , 221 , and 221 Regarding Reactor Coolant Systern Pressblre and Ternpe ratbl re Limits Repo rt Technical Specifications CEPID L 2019 LLA 0215) ," and

.§.. The PTLR \Ji 11 contain the cornpl ete i dentifi cati on fo r each of the TS referenced Topica l Reports blsed to prepare the PTLR (i .e., report nblFRber , titl e, revision , date , and any Sblpplernents) ; and

-c. The PTLR shall be provided t o t he ~JRC blpon i ssblance fo r each reactor vessel flLience period and for any revision or sblppl ernent thereto .

BYRON - UNITS 1 &2 5.6 - 5 Amendment .m

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6. 7 Post /\ccident Monitoring Report ~

1

..Jhen e report is reqLii red b:Y Condition C or G of LCO 3. 3. 3, Post 11 Accident Monitoring (PAM) Instrw+1enteti on , e report shell be 11 SLibmitted within the following 14 deys . The report shell OLitline the preplenned elternete method of monitoring , the CeLise of the inoperebility , end the plens end schedLile for restoring the i nstrnmenteti on chennel s of the FLincti on to OPER/\BLE stetLis .

5.6.8 Tendon SLirveillence Report ~

An:Y ebnormel degredetion of the conteinment strLictLire detected dLiring the tests reqLiired bj the Pre Stressed Concrete Conteinment 1

Tendon SLirveillence Progrem shell be reported in the Inservice Inspection Sl::lmma ry Report in eccordance with 10 CFR 50 . 55e end ASME Section XI .

5.6.9 St eem Generetor CSG) Tb1be Inspection Report ~

A report shell be sLibmitted .'i thin mo deys after the initial entry 1

into MODE 4 following completion of en inspection performed in eccordence with Specificetion 5.5.9, Steem Generetor CSG) Progrem .

The report shall inclLide:

.a.- The scope of inspections performed on each SG ,

-8....- Degradation mechanisms foLind ,

& ~londestrncti ve exami neti on techni qLies Liti l i zed for each degredetion mechenism ,

-d--.- Location , orientation (if linear) , and measLired sizes (if eveilable) of service indLiced indicetions ,

e.- ~!Limber of tLibes plLigged dLiring the inspection OLitege for eech degredetion mechanism ,

.f-. The nLimber and percentage of tLibes plLigged to date , and the effective plLigging percentege in eech steem generetor ,

.g.- The resLilts of condition monitoring , including the resLilts of tLibe pLills end in sitLi testing ,

BYRON - UNITS 1 &2 5.6 - 6 Amendment -ll+/--

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 St earn Generator CSG) Tblbe Inspection Report Ccontinbled)

-I+. For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individblal steam generator , the entire primary to secondary leakage shoblld be conservatively ass1,1med to be from one steam generator) dblri ng the cycle preceding the inspection *i11hi ch is the sblbject of the report ,

+.- For Unit 2, the calcbllated accident indblced leakage rate from the portion of the tblbes below 14 .01 inches from the top of the tblbesheet for the most limiting accident in the most limiting SG . In addition , if the calcbllated accident indblced leakage rate from the most limiting accident is less than

3. 11 times the maxirnblm operational primary to secondary leakage rate , the report shoblld describe how it was determined , and

.j.-. For Unit 2, the resbllts of monitoring for tblbe axial displacement (slippage) . If slippage is discovered , the implications of the discovery and corrective action shall be provided .

BYRON - UNITS 1 &2 5.6 - 7 Amendment -+/-

ATTACHMENT 4 Markup of Technical Specifications Bases Pages (For Information Only)

Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 MARKED UP TECHNICAL SPECIFICATIONS BASES PAGES B 3.0: Pages 1-24 B 3.7.14: Pages 1-4 B 3.7.15: Pages 1-6 B 3.7.16: Pages 1-6

LCD Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION CLCO) APPLICABILITY BASES LC Os i i LCD 3.0.l throLigh LCD 3.0. establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

LCO 3.0.l LCD 3.0.l establishes the Applicability statement w~ 'thin f Tt each individual Specification as the requirement f r, en acii Y the LCD is required to be met Ci .e., when the tffi.=i--t *s in the MODES or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCD 3.0.2 establishes that upon discovery of a failure to meet an LCD, the associated ACTIONS shall be met. The Completion Time of each Required Act ion for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered, Linless otherwise specified .

The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCD are not met. This Specification establishes that:

-a.- CoFRpl eti on of the ReqLii red /\cti ons *,;i thin the specified CoFRpletion TiFRes constitLites coFRpliance \Jith a Specification ; and

-9.- CoFRpletion of the ReqLiired Actions is not reqLiired

\Jhen an LCD is FRet within the specified CoFRpletion TiFRe , Linless otheR1ise specified .

BYRON - UNITS 1 &2 B 3.0 - 1 Revision -+/--+/-J

LCD Applicability B 3.0 BASES

-hGQ. 3.0.2 (continued)

There are t .10 basic types of Required /\ctions . The first 1

type of Required Action specifies a tiFRe liFRit in \Jhich the LCO FRust be FRet . This tiFRe liFRit is the CoFRpletion TiFRe to restore an inoperable systeFR or coFRponent to OPEPJ\BLE status or to restore variables to within specified liFRits . If this type of Required Action is not coFRpl eted \Ji thin the specified CoFRpletion TiFRe , a shutdrnm FRay be required to place the unit in a MODE or condition in which the Specification is not applicable . (Whether stated as a Required Action or not , correction of the entered Condition i s an action that 1+1ay al '..'ays be conside red bipon entering ACTI OWi3 . ) The second type of Reqbli red /\cti on specifies the re1+1edial 1+1easures that per1+1it continued operation of the blnit that is not fblrther restricted by the CoFRpletion TiFRe .

In this case , co1+1pliance with the Reqblired Actions provides an acceptable level of safety for continued operation .

Completing the Required Actions is not required when an LCD is met or is no longer applicable, blnless othe rwise stated in the individblal Specifications .

The natblre of so1+1e Reqbl i red Acti ons of so1+1e Conditions necessitates that , once the Condition is entered , the Required /\ctions 1+1blst be co1+1pleted even though the associated Condition no longer exists . In this instance ,

the i ndi vi dual LCO ' s /\CTI mis specify the Required Acti ans .

/\n exaFRpl e of this is in LCO 3. 4. 3, "RCS Pressure and TeFRperature ( P/ T) Li FRits . "

BYRON - UNITS 1 &2 B 3.0 - 2 Revision .Q.

LCD Applicability B 3.0 BASES

-hGQ. 3.0.2 (continued)

The Completion Times of the Required Actions are also appl i cable '.then a system or component is removed from service intentionally . The ACTIONS for not meeting a single LCO adequately manage any increase in plant risk , provided any unusual external conditions (e .g., severe weather ,

offsite pm,ier instability) are considered . In addition , the increased risk associated \Ji th simultaneous removal of multiple structures , systems , trains or components from service is assessed and managed in accordance ',Ji th 10 CFR 50 . 65 (a) ( 4) . Indi vidual Speci fi ca ti ans may specify a ti me limit fo r perfo rming an SR .1hen equipment is removed from 1

service or bypassed for testing . In this case , the Completion Times of the Required Actions are applicable when this time limit expires , if the equipment remains removed from service or bypassed .

When a change in MODE or other specified condi t ion is required to comply with Required Actions , the unit may enter a MODE or other specified condition in which another Specification becomes applicable and the new LCD is not met .

In this case , the Completion Times of the nrn1 Required Act i ons woul d appl y from t he point i n t ime t ha t t he new Specification becomes applicable , and the ACTIONS Condition(s) are entered .

BYRON - UNITS 1 &2 B 3.0 - 3 Revision ~

LCD Applicability B 3.0 BASES LCD 3.0.3 LCD 3.D.3 establishes the actions that must be implemented J1 1>1hen 1

an LCO is not met and :

~ .fr. /\n associated Reqbli red Action and Co1+1pl eti on Ti 1+1e is not 1+1et and no other Condition applies ; or

-&. The condition of the blni t is not specifically addressed by the associated /\CTIONS . This means that no single Condition or co1+1bination of Conditions stated in the ACTIONS can be made that corresponds to the actual condition of the unit . So1+1eti1+1es , possible combinations of Conditions are such that entering LCO 3.0.3 is warranted . In sblch cases , the Conditions corresponding to such combinations state that LCO 3.0.3 shall be entered i1+1mediately .

This Specification delineates the time limits for placing the blnit in a safe MODE or other specified condition when operation cannot be 1+1aintained within the limi t s for safe operation as defined by the LCO and its /\CTimlS. Planned entry into LCO 3. 0. 3 shobll d be avoided . If it is not practicable to avoid planned entry into LCO 3.0.3, plant risk shobll d be assessed and managed in accordance \Ji th 10 CFR 50 .65(a)(4) , and the planned entry into LCD 3.0.3 shobll d have less effect on pl ant safet:Y than other practicable alternatives .

Upon entering LCO 3.0.3, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allm:ed to prepare for an orderly shbltdown before initiating a change in blnit operation . This includes time to permit the operator to coordinate the redblction in electrical generation \Jith the load dispatcher to ensbtre the stabi l i t:Y and availability of the electrical grid . The time limits specified to enter lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is ',:ell ',Ji thin the specified 1+1axi 1+1u1+1 cool dmm rate and ',Jithi n the ca pa bi l i ti es of the blnit , assblming that only the minimblm reqblired equip1+1ent is OPERABLE . This reduces thermal st resses on components of the Reactor Coolant System and the potential for a unit upset that could challenge safety systems under conditions to \Jhich this Specification applies . The use and interpretation of specified times to complete t he actions of LCO 3.0.3 are consistent \Jith the discblssion of Section 1.3, Completion Times .

BYRON - UNITS 1 &2 B 3.0 - 4 Revision ~

LCD Applicability B 3.0 BASES

-hGQ. 3.0.3 (continued) f\ unit shutdmm required in accordance \Ji th LCO 3. 0. 3 may be terminated and LCO 3. 0. 3 exited if any of the foll O'vJi ng occurs:

-a.- The LCO is nov1 met .

-8....- The LCO is no longer applicable .

-c.- A Condi ti on exists for \Jhi ch the Required /\cti ans have now been performed .

4-.- ACTimJS exist that do not have expired Completion Times . These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited .

The time limits of LCO 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> from MODE 1, 2, 3, or 4 for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation . If the unit is in a lower MODE of operation when a shLitdown is reqLiired , the time limit for entering the next lm:er MODE applies . If a lm:er MODE i s ente red in less time than allowed , howeve r , the total all m1abl e time to enter MODE 5, or other applicable MODE , is not reduced . For example , if MODE 3 is entered in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> , then the time allmrod for entering MODE 4 is the next 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> , because the total time for entering MODE 4 is not reduced from the all m1abl e limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> .

Therefore , if remedial measures are completed t hat ',mul d permit a return to MODE 1, a penalty is not incurred by having to enter a lmror MODE of operation in less than the total time allmmd .

In MODES 1 , 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications . +fie.

requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficient ly define the remedial measures to be taken .

BYRON - UNITS 1 &2 B 3.0 - 5 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.3 (continued)

Excepti ans to LCO 3. 0. 3 are provided in instances *,Jhere requiring a unit shutdmm , in accordance *,Jith LCO 3.0.3, would not provide remedial measures for the associated condition of the unit . An example of this is in LCO 3.7.14 ,

"Spent Fuel Pool \*later Level . " LCO 3. 7.14 has an Applicability of "During movement of irradiated fuel assemblies in the spent foel pool ." Therefore , this LCO can be applicable in any or all MODES . If the LCO and the Required Acti ans of LCO 3. 7.14 are not met \Jhi le in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdrnm condition . The Required Action of LCO 3. 7.14 of "Sb1spend movement of irradiated fuel assemblies in the spent fuel pool " is the appropriate Required Action to complete in lieu of the actions of LCD 3.D.3. These exceptions are addressed in t he individual Specifications .

LCD 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified condi ti ans in the Appl i ca bi l i ty 1.'hen an LCD 1

is not met . It allows placing the unit in a MODE or other specified condition stated in that Applicabili ty (e .g., the Applicability desired to be entered) v1hen unit condi ti ans are such that the requirements of the LCO would not be met ,

in accordance *,:ith either LCO 3.0.4. a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allm:s entry into a MODE or other specified condition in the Appl i ca bi l i ty *,,ii th the LCO not met \Jhen the associated ACTimJS to be entered foll m1i ng entry into the MODE or other specified condition in the /\ppl i ca bi l i ty *.:i 11 permit continued operation *,:i thin the MODE or other specified condition for an unlimited period of time .

Compliance \Jith ACTIONS that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation . This is *.:ithout regard to the status of the unit before or after the MODE change .

Therefore , in such cases , entry into a MODE or other specified condition in the Applicability may be made and the Required Acti ans foll mJed after entry into the Applicability .

For example , LCO 3. 0. 4. a may be used \Jhen the Required Action to be entered states that an inoperable instrument channel must be pl aced in the trip condition 'n'i thin the Completion Time . Transition into a MODE or other specified BYRON - UNITS 1 &2 B 3.0 - 6 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.4 (continued) condition in the Appl icabil ity may be made in accordance

\Jith LCO 3.0.4 and the channel is subsequently pl aced in the tripped condition within the Completion Time , which begins

',;hen the /\ppl i ca bi l i ty is entered . If t he inst rument channel cannot be pl aced in the tripped condition and the subsequent defaul t ACT ION ( Required Action and associated Compl etion Time not met ") allm:s the OPER/\BLE train to be pl aced i n operation , use of LCD 3.0.4.a i s acceptable because t he subsequent ACTIONS to be entered fo ll m:i ng entry into the MODE i nclude ACTIONS (pl ace the OPERABLE train in operation) that permit safe plant operation for an unlimited period of time in the MODE or other specified condition to be entered .

LCO 3. 0. 4. b allows entry into a MODE or other specified condition in the Appl i ca bi l i t:Y 'xi th the LCO not met after performance of a risk assessment addressing inoperable systems and components , consideration of the resul ts ,

dete rmination of the acceptability of entering the MODE or other specified condition in the Appl icability , and establ ishment of risk management actions , if appropriate .

The risk assessment may use quantitative , qualitative , or bl ended approaches , and the risk assessment wi l l be conducted using the plant program , procedures , and criteria in pl ace to implement 10 C~R 50 . 65 (a)( 4) , ..'hi ch requires 1

t hat risk impacts of maintenance activities to be assessed and managed . The risk assessment , for the purposes of LCD 3.0.4 (b) , must take into account al l inoperabl e Technica l Specifi cati on equipment rega rdless of ',Jhethe r t he equipment is included in the normal 10 C~R 50 .65(a)(4) risk assessment scope . The risk assessments wi l l be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182 , "Assessing and Managing Risk Before Maintenance Acti vi ti es at ~Ju e l ear Pm:er Pl ants ."

Regulatory Guide 1. 182 endorses the guidance in Section 11 of ~J UMARC 93 01 , "Industry Guideline for Mani tori ng the Effectiveness of Maintenance at ~J uel ear Pmmr Pl ants ."

These documents address general guidance for conduct of the risk assessment , quantitative and qualitative guidelines for establ ishing risk management actions , and example risk management actions . These include actions to plan and conduct other activities in a manner t hat controls overa ll risk , increa sed ri sk a'..'a reness by shift and management personnel , actions to reduce the duration of the condition ,

actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures) , and determination that the proposed BYRON - UNITS 1 &2 B 3.0 - 7 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.4 (continued)

MODE change is acceptable . Consideration should also be gi ven to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of /\CTIO~lS Completion Times that ',Joul d require exiting the Applicability .

LCO 3.0.4.b may be used \Jith single , or multiple systems and components unavailable . NUMARC 93 01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components .

The results of the ri sk as sessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions . The LCO 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time . Since this is allowable , and since in general the risk impact in that parti cular MODE bounds the ri sk of trans i t i oni ng i nto and t hrough t he appli ca ble MODES or other specified conditions in the Applicability of the LCO , the use of the LCO 3.0.4. b allowance should be generally acceptable , as long as the risk is assessed and managed as stated above . ~m1ever , there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3. 0. 4. b all m1ance is prohibited . The LCOs governing these system and components contain ~l o tes prohibiting the use of LCD 3.0.4. b b:Y stating that LCD 3.0.4.b is not appli cable .

LCD 3.0.4.c allm1s entry into a MODE or other specified condition in the Appl i cabi l i ty 'tJi th the LCD not met based on a Note in the Speci fi cati on ',1hi ch states LCD 3. 0. 4. c i s appl i cable . These specific all O',.'ances permit entry into MODES or other specified conditions in the Applicability

\1hen the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a ri sk assessment has not been performed . This all o',Jance may apply to all the /\CTIONS or to a specific Required Action of a Specification . The risk assessments performed to justify the use of LCD 3.0.4.b usually only consider systems and components . ~or thi s rea son, LCD 3.0.4.c i s typically applied to Specifi cati an s \lhi ch describe values and parameters (e .g., reactor coolant system specific activity) , and may be applied to other Specifications based on ~lRC pl ant specific approval .

BYRON - UNITS 1 &2 B 3.0 - 8 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.4 (continued)

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systerns or cornponents to OPER/\BLE status before entering an associated MODE or ot her specified condition in the Applicability .

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that a re required to cornpl y *,1ith ACTIO~JS . In addition , the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that resb1lt frorn any blnit shb1tdo 11'n. In this context , a unit 1

shutdown is defined as a change in MODE or other specified condition in the Appl i ca bi l it:Y associated \vi th transitioning frorn MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4, and MODE 4 to MODE 5.

Upon entry into a MODE or other specified condition in the

/\pplicability with the LCO not rnet , LCO 3.0. l and LCO 3.0. 2 require entry into the applicable Conditions and Required Actions until the Condition is resolved , until the LCO is R-iet , or until the unit is not '11 ithi n the /\ppl i cabil it:Y of 1

the Technical Specification .

Surveillances do not have to be perforrned on the associated inoperable equipR-ient (or on variables outside t he specified lirnits) , as perrnitted by SR 3.0.1. Therefore , utilizing LCO 3.0. 4 is not a violation of SR 3.0.1 or SR 3.0. 4 for any Surveillances that have not been perforrned on inoperable equiprnent . Mrnmver , SRs rnust be rnet to ensure OPERABILITY prior to declarin§ the associated equipR-ient OPE RAB LE Cor va ri able *.,iithi n l i rnits) and restoring cornpl i anee \Ji th the affected LCO .

BYRON - UNITS 1 &2 B 3.0 - 9 Revision -W4

LCD Applicability B 3.0 BASES LCD 3.0.5 LCD 3.0.5 establishes the allowance for restoring equipment to service under admi ni strati ve controls .Nhen it has been 1

removed from service or declared inoperable to comply with ACTIONS . The sole purpose of this LCO is to provide an exception to LCO 3.0.2 (e .g., to not comply with the appl i cable Required /\cti on (s)) to all o',J the performance of required testing to demonstrate:

-a.- The OPER~BILITY of the equipment being returned to service ; or

-8.. The OPER~BILITY of other equipment .

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the ti me absolutely necessary to perform the required testing to demonstrate OPER~BILITY .

This Specification does not provide time to perform any other preventive or corrective maintenance . LCO 3.0.5 should not be used in lieu of other practicable alternatives that comply with Required Actions and that do not require changing the MODE or other specified conditions in the Applicability in order to demonstrate equipment is OPEMBLE .

LCD 3.0.5 is not intended to be used repeatedly .

An example of demonstrating the OPER/\BILITY of the equipment being returned to service is reopening a containment i sol ati on valve that has been closed to comply *.1ith Required Actions and must be reopened to perform the required testing .

Another example of demonstrating equipment is OPER~BLE i,,ii th the Required Actions not met is opening a manual valve that was closed to comply with Required Actions to isolate a fl m1path *.. ith excessive Reactor Cool ant 2lystem ( RC2l) 1 Pressure Isolation Valve CPIV) leakage in order to perform testing to demonstrate that RCS PIV leakage is now within limit.

Examples of demonstrating equipment OPER~BILITY include instances in \Jhich it is necessary to ta l~ e an inoperable channel or trip system out of a tripped condition that *,,ias directed b:Y a Requi red 1~cti on , if there is no Required Action ~Jote for this purpose . An example of verifying OPEMBILITY of equipment removed from service is taking a tripped channel out of the tripped condition to permit the logic to function and indicate the appropriate response during performance of required testing on the inoperable channel .

BYRON - UNITS 1 &2 B 3.0 - 10 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.5 (continued)

Examples of demonstrating the OPER~BILITY of other equipment are taking an inoperable channel or trip system out of the tripped condition 1) to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system , or 2) to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system .

The administrative controls in LCO 3.0.5 apply in all cases to systems or components in Chapte r 3 of the Technical Specifi cations , as long as the testing could not be conducted while complying vii th the Required /\ct i ans . +4-i-s-includes the realignment or repositioning of redundant or alternate equipment or trains previously manipulated to comply with ACTIONS , as well as equipment removed from service or declared inoperable to comply with ACTIONS .

LCD 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Specifications (TS) . This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inope rable supported system LCO be entered solely due to the inoperability of the support syst em. Thi s exception i s j usti f i ed because the action s that are requi red to ensu re the unit i s maintained in a safe condition are specified in the support system LCO ' s Required Actions . These Required Actions may include entering the supported system ' s Conditions and Required Actions or may specify other Required Actions .

When a support system is inoperable and there is an LCO specified for it in the TS , the supported system(s) are requi red t o be decla red inope rable if determined to be inoperable as a result of the support system i noperabil i ty .

~mmver , it is not necessary to enter into the supported systems ' Conditions and Requi red Actions unless directed to do so by the suppo rt system ' s Requi red Actions . +fie.

potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems ' Conditions and Required Actions are eliminated by providing all the acti ans that are necessary to ensure the unit is maintained in a safe condition in the support system ' s Required Actions .

BYRON - UNITS 1 &2 B 3.0 - 11 Revision -W4

LCD Applicability B 3.0 BASES

-hGQ. 3.0.6 (continued)

Hm1ever , there are instances ',Jhere a support system ' s Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system . This may occur imediately or after some specified delay to perform some other Required Action . Regardless of *,1hether it is i mediate or after some delay , .Jhen a support system ' s 1

Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system , the applicable Conditions and Req~ired Actions shall be entered in accordance \Jith LCO 3.0. 2.

Specification 5. 5. 15 , "Safety Fblnction Determination Program (SFDP) ," ensblres loss of safety fblnction is detected and appropriate actions are taken. Upon entry into LCD 3.0.6, an evalblation shall be made to determine if loss of safety fblnction exists . Additionally , other limitations , remedial actions , or compensatory actions may be identified as a resbllt of the sblpport system inoperability and corresponding exception to entering sblpported system Conditions and Reqbli red Acti ons . The SFDP implements t he reqbli rements of LCD 3.0.6.

Cross train checks to identify a loss of safety fblnction for those Sblpport systems that support multiple and redundant safety systems are required . The cross train check verifies that the supported systems of the redundant DPERN3LE Sblpport system are DPERN3LE , thereby ensuring safety function is retained . If thi s evaluation determines that a loss of safety fun ction exists , the appropriate Conditions and Required Actions of the LCD in which the loss of safety function exists are required to be entered .

BYRON - UNITS 1 &2 B 3.0 - 12 Revision -W4

LCD Applicability B 3.0 BASES LCD 3.0.7 There are certain specia l tests and operations required to be perforrned at various tirnes over the life of t he unit .

These specia l tests and operations are necessary to dernonstrate sel ect unit perforrnance characteristics , to perforrn special rnaintenance activities , and to perforrn specia l evolutions . Exception LCOs (e .g., LCO 3.1.8, "PHYS ICS TESTS Exceptions MODE 2") all O'A' specifi ed Technica l Specification (TS) requirernents to be changed to perrnit perforrnances of these specia l tests and operations ,

whi ch otherwise coLild not be perforrned if required to cornply Hith the requirernents of these TS . Un l ess othen1ise specified , al l the other TS requirernents rernain unchanged .

This *,Ji 11 ens Lire all appropriate reqLii rernents of the MODE or other specified condition not directly associated with or required to be changed to perforrn the special test or operation will rernain in effect .

The App l icability of an Exception LCO represent s a condition not necessarily in cornpliance with the norrnal requirernents of the TS . Cornpl iance with Exception LCOs is optional . A special operation rnay be perforrned either under the provisions of the appropriate Exception LCO or Linder the other applicable TS reqLiirernents . If it is desired to perform the special operation Linder the provisions of the Exception LCO , the requirernents of the Exception LCO shall be foll mved .

LCD 3.0.8 LCO 3.0.8 establishes the applicability of each Specification to both Unit 1 and Unit 2 operation . Whenever a requirernent applies to only one unit , or is different for each unit , this ',Ji 11 be i denti fi ed in the appropriate section of the Specification (e .g., Appl icabi l ity ,

SLirvei 11 ance , etc . ) *,1i th parenthetical reference , ~l otes , or other appropriate presentation '..'i thin the body of the requirernent .

LCD 3.0.9 LCO 3.0.9 establishes conditions under \Jhich systerns are considered to rernain capabl e of perforrning their intended safety function *,;hen associated snubbers are not capable of providing their associated support function(s) . This LCO states that the supported systern is not considered to be inoperable solely due to one or rnore snubbers not capable of perforrning their associated sLipport function(s) . This is appropriate because a lirnited length of tirne is allm1ed for rnaintenance , testing , or repair of one or rnore snubbers not capable of perforrning their associated support function(s) and appropriate cornpensatory rneasures are specified in the snubber requirernents , \Jhich are located outside of the BYRON - UNITS 1 &2 B 3.0 - 13 Revision -+/--1-J

LCD Applicability B 3.0 BASES

-hGQ. 3.0.9 (continued)

Technica l Specifications (TS) under l icensee control . +fie.

snubber requirements do not meet the criteria in 10 CFR 50 .36(c)(2)(ii) , and , as such , are appropriate for control by the l icensee .

When applying LCO 3.0.9.a, at least one train of /\uxiliary Feed\1ater (/\FvD system must be OPER/\BLE during MODES ',:hen AFl\ i s required to be OPERABLE. h'hen applying LCO 3. 0. 9. a 1

during MODES \/hen /\FW is not required to be DPERAt3LE , a core cooling method (such as the Residual Heat Removal (RHR) syster+1) r+1ust be avail able per applicable site procedures . When applying LCO 3.0.9.b, a r+1eans of core cooling r+1ust rer+1ain avai l able (/\FW, RHR, equipr+1ent necessary for feed and bleed operations , etc . ). Reliance on ;wail ability of a core cooling source during r+1odes A'here 1

/\ Ft~ is not required bj TS provides an equival ent safety 1

r+1argin for plant operations were LCO 3.0.9 not applied and r+1eets the intent of Technical Specifications Task Force Change Travel er TSTF 372 , Revision ~ , "/\ddi ti on of LCD 3.D.8, Inoperability of Snubbers ."

When a sn ubber i s to be rendered i ncapabl e of perforr+1i ng its related support function Ci .e., nonfun ctional) for testing or r+1aintenance or is discovered to not be function al , it r+1ust be deterr+1i ned ',Jhether any syster+1( s) require the affected snubber( s) for s:yster+1 DP ERAB I LIT¥ , and

\Jhether the plant is in a MODE or specified condition in the Applicability that requires the supported syster+1(s) to be OPER/\BLE .

If an analysis deterr+1ines that the supported syster+1(s ) do not require the snubber(s) to be functional in order to support the OPERABILITY of the system(s) , LCO 3.0.9 i s not needed . If the LCO(s) associated with any supported system( s ) are not currentl y appli cable (i .e., the plant i s not in a MODE or other specified condition in the Applicability of the LCO) , LCO 3.0.9 is not needed . If the supported system(s) are inoperable for reasons other than snubbers , LCO 3.0.9 cannot be used . LCO 3.0.9 is an all m:ance , not a requirement . 11i'hen a snubber is nonfunctiona l , any supported system(s) may be declared inoperable instead of using LCD 3.D.9.

Every tir+1e the provisions of LCO 3.0.9 are used , the station ',Ji 11 confi rm that at least one train (or subsyster+1) of systems supported by the inoperabl e snubbers 'A'i 11 remain capable of perforr+1ing their required safety or support functions for postulated design l oads other than seismic loads . A record of the design function of the inoperabl e BYRON - UNITS 1 &2 B 3. 0 - 14 Revision -+/--1-J

LCD Applicability B 3.0 BASES

-hGQ. 3.0.9 (continued) snubber (i .e., seismic vs . non seismic) and the associated pl ant configuration ',Ii 11 be available on a recoverable basis for MRG staff inspection .

LGO 3.0.9 does not apply to non seismic snubbers . +fie.

provisions of LCO 3.0.9 are not to be applied t o supported TS systems unless the supported systems '.Joul d remain capable of performing their required safety or support functions for postulated design loads other than seismic loads . The risk impact of dynamic loadings other than seismic loads ',Jas not assessed as part of the development of LCO 3.0.9. These shock type loads include thrust loads ,

bl owdm*m loads , water hammer loads , steam hammer loads ,

LOCA loads and pipe rupture loads . However , there are some important distinctions between non seismic (shock type) loads and seismic loads which indicate that , in general ,

the risk impact of the out of service snubbers is smaller for non seismic loads than for seismic loads . First , while a seismic load affects the entire plant , the impact of a non seismic load is localized to a certain syst em or area of the plant . Second , although non seismic shock loads may be highe r in t otal forc e and t he impact could be as much or more than seismic loads , generally they are of mb1ch shorter duration than seismic loads . Third , the impact of non seismic loads is more plant specific , and thus harder to analyze generically , than for seismic loads . For these reasons , every time LCO 3.0.9 is applied , at least one train of each system that is supported b:'i' the inoperable snubber(s) should remain capable of performing their requi red safety or support functions for postulated design loads other than seismic loads .

If the allrn1ed time expires and the snubber(s) are unable to perform their associated support function(s) , the affected supported system ' s LGO(s) must be declared not met and the Conditions and Required Actions entered in accordance '.,'ith LCD 3. 0. 2.

LGO 3. 0. 9. a applies '.,'hen one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system . LCO 3.0.9.a allrn1s 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the snubber(s) before declaring the supported system inoperable . The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable based on the low probabi l i t:Y of a seismic event concurrent '.Ji th an event that ',JOul d require opera ti on of the supported system occurring while the snubber(s) are not BYRON - UNITS 1 &2 B 3.0 - 15 Revision -+/--+/-J

LCD Applicability B 3.0 BASES

-hGQ. 3.0.9 (continued) capable of performing their associated support function and due to the availability of the redundant train of the supported system .

LGO 3.0.9.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system . LGO 3.0.9.b allov.'s 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to restore the snubber(s) before declaring the supported system inoperable . The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable ba sed on the l m1 probabil ity of a seismi c event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function .

LGO 3.0.9 requires that risk be assessed and managed .

Industry and MRC guidance on the implementation of 10 GFR 50 .65(a)(4) (the Maintenance Rule) does not address seismic risk . However , use of LCD 3.0.9 should be considered with respect to other plant maintenance activities , and integrated into the existing Maintenance Rule process to t he ext ent possible so t hat mai ntenance on any unaffected train or subsystem is properly controlled ,

and emergent issues are properly addressed . The risk assessment need not be quantified but may be a qualitative a\:a reness of the vul nerabi lit:/ of systems and components

\Jhen one or more snubbers are not able to perform their associated support function .

BYRON - UNITS 1 &2 B 3.0 - 16 Revision -+/--1-J

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT CSR) APPLICABILITY BASES

'}

SRs SR 3.0.1 through SR 3.0. establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. SR 3.0.2 and SR 3.0.3 apply in Chapter 5 only when invoked by a Chapter 5 Specification.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs . This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems end components , end that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance

~with SR 3.0.2, constitutes a failure to meet an L~

Systems end components are assumed t o be OPERABLE when the LCOs are met associated SRs have been met. Nothing in this when the Specification, however, is to be construed as implying that systems or components ere OPERABLE '..'hen:

-e-. The systems or components ere knovm to be i noperebl e, elthough still meeting the SRs ; or

.&. +Re- requirements of the Surveillance(s) are known not to be met between required Surveillance perfor::;~ ltacilityl Surveillances do not have to be performed when the

• is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs essoci eted *.1ith en Exception LCO ere only eppl i cebl e *,:hen the Exception LCO is used es en ellm~ble exception to the requirements of e Specificetion .

Unplenned events mey setisfy the requirements (including eppliceble ecceptence criteria) for e given SR. In this case , the unplanned event may be credited es fulfilling the performance of the SR . This all m,ience includes those SRs

\.'hose performence is normelly precluded in e given MODE or other specified condition .

BYRON - UNITS 1 &2 B 3.0 - 17 Revision ++/-J

SR Applicability B 3.0 BASES

~ 3.0.1 (continued)

Survei ll ances , incl uding Surveil l ances invoked by Required Actions , do not ha ve to be performed on inoperable equipment because the ACT IONS define the remedial measures that apply.

Survei ll ances have to be met and performed in accordance

'A'ith SR 3.0.2, prior to returning equi pment to OPERABLE status .

Upon compl etion of maintenance , appropriate post maintenance I~~i~~~si~n~~~~!;e~p:~i~~g~~r~u~~~J~!~:~e~P;~ia~~t f~d and thei r most recent performance is in accordance \Ji th SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established . In these sitLiations , the eqLiipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possibl e and the eqLiipment is not otherwise bel ieved to be incapabl e of pe rforming its function . This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be compl et ecL SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action \Jith a Compl etion Time that reqLiires the periodi c pe rfo rmance of the Requi red Action on a "once pe r . . . "

interva l . facility SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension faci tates Surveillance scheduling and considers unit operating conditions that may not be suitable for conducting the Surveillance (e.g.,

transient conditions or other ongoing Surveillance or maintenance activities).

BYRON - UNITS 1 &2 B 3.0 - 18 Revision -+/-J

SR Applicability B 3.0 BASES

~ 3.0.2 (continued)

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for

\Jhich the 25% extension of the interval specified in the Frequency does not apply . These exceptions are stated in the individual Specifications . An exaFAple of *,1here SR 3.0.2 does not apply is the ContainFAent Leakage Rate Testing Program . The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations .

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ... " basis . The 25~&

extension applies to each performance after the initial perforFAance . The initial perforFAance of the Required Action , whether it is a particular Surveillance or some other remedial action , is considered a single action \Jith a singl e Completi on Time. One reason for not all owi ng the 25%

extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative FAanner .

The provisions of SR 3.0.2 are not intended to be used repeatedly to extend Surveillance intervals Cothe r than those consistent \Ji th refueling intervals) or periodic CoFApletion TiFAe intervals beyond those specified .

BYRON - UNITS 1 &2 B 3.0 - 19 Revision ++/-J

SR Applicability B 3.0 BASES SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been performed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater, applies from t he point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides adequate time to perform Surveillances that have been missed. This delay period permits the performance of a Surveillance before complying with Required Actions or other remedial measures that might preclude performance of the Surveillance.

~The basis for this delay period includes consideration of

~ conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in complet ing the required Surveillance, and the recognition tha t the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

'A'hen a Survei 11 ance ',1ith a Frequency based not on time intervals , but upon specified unit conditions , operating situations , or requirements of regulations (e .g., prior to entering MODE 1 after each fuel loading , or in accordance

'.,'i th 10 CFR 50 , Appendi x J , as modified b:J' approved exemptions , etc . ) is discovered to not have been performed when specified , SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance .

However , since there is not a time interval specified , the missed Surveillance should be performed at the first reasonable opportunity .

SR 3.0.3 provides a time limit for , and allrnrances for the performance of , Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions .

BYRON - UNITS 1 &2 B 3.0 - 20 Revision ++/-J

SR Applicability B 3.0 BASES

~ 3.0.3 (continued)

SR 3.0.3 is only applicable if there is a reasonable expectation the associated equipment is OPERAgLE or that variables are within limits, and it is expected that the Surveillance will be met when performed. Many factors should be considered, such as the period of time since the Surveillance was last performed, or whether the Surveillance, or a portion thereof, has ever been performed, and any other indications, tests, or activities that might support the expectation that the Surveillance will be met when performed. An example of the use of SR 3.0.3 would be a relay contact that was not tested as required in accordance with a particular SR, but previous successful performances of the SR included the relay contact; the adjacent, physically connected relay contacts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not sufficient to infer the behavior of the associated equipment from the performance of similar equipment. The rigor of determining whether there is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance . If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will be met when performed. For Surveillances that have not been performed for a long period or that have never been performed , a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERAgLE . The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used repeatedly to extend Surveillance intervals.

BYRON - UNITS 1 &2 B 3.0 - 21 Revision -+/-+/-

SR Applicability B 3.0 BASES

~ 3.0.3 (continued)

While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified ~requency is provided to perform the missed Surveillance, it lS expected that the missed Surveillance will be performed at 1.-f-ac-i-lit--.yl the first reasonable opportunity. The determination of the

,____..;;...o first reaso ~le opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting yl the pl ant ~wn to pe~form ~he Su~v~i 11 ance) . and im~a~t on

.-If-ac-i-lit......

. . any analys1s assumpt1ons, ln add1t1on to unit cond1t1ons ,

planning, availability of personnel, and the time required to perform the Surveillance . This risk impact should be managed thro~gh the program in place to implement 10 CFR 50 . 65 (a) ( 4) and its i mpl ementati on guidance , MRC Regulatory Guide 1. 182 , ' Assessing and Managing Risk Before Maintenance f\cti vi ti es at Mucl ear PovJGr Pl ants .' This Regul ator.y Gui de addresses consideration of temporary and aggregate risk impacts , determination of risk management action thresholds ,

and risk management action up to and including plant shutdown . The missed Surveillance should be treated as an emergent condition as disrnssed in the Regulatory Guide .

The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the licensee's Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCD Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable , or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCD Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

BYRON - UNITS 1 &2 B 3.0 - 22 Revision ++/-J

SR Applicability B 3.0 BASES SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

This Specification ensures that system and component OP ERAt3I LIT¥ requirements and va ri able limits are met before variable entry into MODES or other specified conditions in th Applicability for which these systems and components ensure limits safe operation of the unit . The provisions of this the safety of Specification should not be interpreted as endorsing the the facility failure to exercise the good practice of restoring system..- s --------.

or components to OPERABLE status before entering an variables to associated MODE or other specified condition in the within limits Applicability.

A provision is included to allow ent ry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4.

However, in certain circumstances, failing to meet an SR will not res ult in SR 3.0.4 restrict ing a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.D.1, which variables states that surveillances do not have to be performed on outside inoperable equipment . When equipment is inoperable , specified SR 3.D.4 does not apply to the associated SR(s) si limits requirement for the SR(s) to be performed is removed.

Therefore, failing to perform the Surveillance (s) wit in the specified Frequency does not result in an SR 3. D.4 '------Ian LCO restriction to changing MODES or other specified conditions is not met of the Applicability. However, since the LCD is not met in this instance, LCD 3.D.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.D.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.D.3.

The provisions of SR 3.D.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition , the provisions of SR 3. 0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result BYRON - UNITS 1 &2 B 3.D - 23 Revision ++/-J

SR Applicability B 3.0 BASES

~ 3.0. 4 (continued) from any unit shutdmm . In this context , a unit shutdmm is defined as a change in MODE or other specified condition in the /\ppl i ca bi l i ty associated 'IJi th transi ti oni ng from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4, and MODE 4 to MODE 5.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary . The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Sblrvei 11 ance , or both . This a11 ows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCD prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO ' s /\ppl i ca bi l i ty , wobll d have i t s Freqblency specified such that it is not "due " until the specific conditions needed are met . Alternately , the Sblrveillance may be stated in the form of a Note , as not req~ired (to be met or performed) blntil a pa rti cblla r event , conditi on, or time has been reached . Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

SR 3. 0. 5 SR 3. 0. 5 establishes the applicability of each Survei 11 ance to both Unit 1 and Unit 2 opera ti on . i"lhenever a requirement 1

applies to only one unit , or is different for each unit ,

this '.Ji 11 be i denti fi ed '.Ji th parenthetical reference , ~lotes ,

or other appropriate presentation '.Ji thin the SR .

BYRON - UNITS 1 &2 B 3. 0 - 24 Revision -+/--+/-J

Spent Fuel Pool Water Level B 3.7.14

~

B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the st orage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pool design is given in the UFSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the UFSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the UFSAR, Section 15.7.4 (Ref. 3).

APPLICABLE The minimum water level in the spent fuel pool meets the SAFETY ANALYSES assumptions of the fuel handling accident described in Regulatory Guide 1.183 (Ref. 4). The resultant Total Effective Dose Equivalent (TEDE) dose is within 10 CFR 50.67 limits (Ref. 5).

According to Reference 4, there is 23 ft of wa t er between the top of the damaged fuel bundle and the fuel pool water surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly.

In practice, this LCD preserves the assumption for the bulk of the fuel in the storage racks . In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be< 23 ft of water above the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The spent fuel pool water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BYRON - UNITS 1 &2 83 . 7.14-1 Revision Spent Fuel Pool Water Level B 3.7.14 BASES (continued)

LCO The spent fuel pool water level is required to be~ 23 ft over the top of irradiated fuel assemblies seat ed in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3).

As such, it is the minimum required for fuel st orage and movement within the spent fuel pool.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool, since the potential for a release of fission products exists .

ACTIONS TFie ACTimlS Fiave beeA 1t1odi fi ed b:Y a ~late i Adi cati Ag U1at LCD 3. 0. 3 does AOt appl y .

A.1 When the initial conditions assumed in the accident analysis cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pool water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position .

If 1t1ovi Ag i rradiated fu el asse1t1bl i es 'oJA il e i A MODE S or § ,

LCD 3. 0. 3 'oJOul d Aot specify aAy acti OA . If 1t1ovi Ag i rradiated fu el asse1t1bl i es 'vo'Ai l e i A MODES 1, 2, 3 , aAd ~ ,

tFie fue l 1t1ove1t1eAt i s iAdepeAdeAt of rea ctor operatioAs .

TFie refo re , i Aabi l i ty to s uspeAd 1t1ove1t1eAt of i rradiated fuel asse1t1bli es i s Aot s uff i ci eAt reasoA t o requi re a react or sfiutdmm .

BYRON - UNITS 1 &2 B3 .7.14-2 Revision B

Spent Fuel Pool Water Level B 3.7.14 BASES (continued)

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient spent fuel pool wa t er is available in the event of a fuel handling accident. The water level in the spent fuel pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

DuFiA§ FefueliA§ opeFatioAs , tAe level iA tAe speAt fuel pool is i A equi l i bFi Uffi \Ji Hi tAe Fefuel i A§ ca1*i t:Y 'oJAeA tAe,y aFe AydFaulically coupled , aAd tAe level iA tAe FefueliA§ cavit,y is cAeclEed dail:Y iA accoFdaAce 'vv'iH1 SR 3 . 9 . 7 . 1.

REFERENCES 1. UFSAR, Section 9 .1. 2.

2. UFSAR, Section 9.1.3.

3. UFSAR, Section 15.7.4.

4. Regulatory Guide 1.183, July 2000.

5. 10 CFR 50.67 .

BYRON - UNITS 1 &2 B3 .7.14-3 Revision ::;..&

I NO CHANGES ON THIS PAGE I Spent Fuel Pool Water Level B 3.7.14 BASES (continued)

This page intentionally left blank.

BYRON - UNITS 1 &2 B3 .7. 14-4 Revision 0

Spent Fuel Pool Boron Concentration B 3.7.15

~

B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Boron Concentration BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions . (For this discussion, the term OFA is intended to refer to the specific reduced fuel rodlet diameter, and includes all analyzed fuel types with this diameter, such as Vantage 5. ) The 24 spent fuel pool storage racks provide placement locations for a total of 2984 new or used fuel assemblies. Of the 24 spent fuel pool storage racks, four are designated "Region 1" with the remaining 20 racks designated as "Region 2." The analytical methodology used for the criticality analyses is in accordance with established NRC guidelines (Ref. 2).

Region 1 racks contain 396 cells which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes tha t spent fuel assemblies reside in all available cell locations). The stored fuel assemblies may contain an initial nominal enrichment of~ 5.0 weight percent U-235 (with or without IFBAs installed) (Ref. 5).

Region 2 racks contain 2588 cells which are also analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes tha t spent fuel assemblies reside in all available cell locations). For the "All Cells" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of

~ 5.0 weight percent U-235 with credit for burnup.

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under actual operating conditions.

BYRON - UNITS 1 &2 B 3. 7.15 - 1 Revision -68

I NO CHANGES ON THIS PAGE I Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE NRC approved methodologies were used to develop the SAFETY ANALYSES criticality analyses (Ref. 2) for the spent fuel pool storage racks. The fuel handling accident analyses are described in Reference 6. Additional evaluations were performed (Ref. 8) to support placement of the Byron lead test assemblies with higher density pellets in the spent fuel pool.

The criticality analyses for the spent fuel assembly storage racks confirm that keff remains s 0. 95 for the spent fuel pool storage racks (including uncertainties and tolerances) at a 95% probability with a 95% confidence level (95/95 basis),

based on the accident condition of the pool being flooded with unborated water. Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack keff s 0.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies. Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the "double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident (Refs. 9 and 10).

The accident analyses address the following five postulated scenarios:

1) fuel assembly drop on top of rack;

2) fuel assembly drop between rack modules;

3) fuel assembly drop between rack modules and spent fuel pool wall;

4) change in spent fuel pool water temperature; and

5) fuel assembly loaded contrary to placement restrictions.

Of these, only scenarios 2, 3, and 5 have the capacity to increase reactivity for the spent fuel pool storage racks.

Calculations were performed, for the spent fuel pool storage racks, for a spent fuel pool temperature of 4°C (39°F) which is well below the lowest normal operating temperature (50°F).

Because the temperature coefficient of reactivity in the spent fuel pool is negative, temperatures grea t er than 4°C will result in a decrease in reactivity.

BYRON - UNITS 1 &2 B3 . 7.15-2 Revision 68

I NO CHANGES ON THIS PAGE I Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued)

Calculations were also performed to show the largest reactivity increase caused by a Westinghouse 17X17 OFA fuel assembly misplaced into a Region 2 storage cell for which the restrictions on enrichment or burnup are not satisfied.

The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

For the above postulated accident conditions, t he double contingency principle can be applied. Specifically, the presence of soluble boron in the spent fuel pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. For the spent fuel pool storage racks, spent fuel pool soluble boron has been credited in the criticality safety analysis to offset the reactivity caused by postulated accident conditions. Because the Region 1 racks are designed for the storage of fresh fuel assemblies, a fuel assembly misload accident has no consequences from a criticality standpoint (i.e., the acceptance criteria for storage are satisfied by all assemblies in the spent fuel pool).

Should a fuel assembly misload accident occur in the Region 2 storage cells, keff will be maintained~ 0.95 due to the presence of at least 300 ppm of soluble boron in the spent fuel pool water.

The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BYRON - UNITS 1 &2 B3 . 7.15-3 Revision 68

INO CHANGES ON THIS PAGE I Spent Fuel Pool Boron Concentration B 3. 7. 15 BASES LCD The spent fuel pool boron concentration is required to be

~ 300 ppm. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in References 5, 6, and 7. The dissolved boron concentration of 300 ppm bounds the minimum required concentration for accidents occurring during fuel assembly movement within the spent fuel pool.

APPLICABILITY This LCD applies whenever fuel assemblies are stored in the spent fuel pool.

BYRON - UNITS 1 &2 B3 . 7.15-4 Revision 68

Spent Fuel Pool Boron Concentration B 3.7.15 BASES ACTIONS Tfie ACTimlS fiave beeA fflodi fi ed b:Y a ~late i Adi cati Ag tfiat LCD 3.0. 3 does AOt apply .

A.1 and A.2 When the concentration of boron in t he spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude movement of a fuel assembly to a safe position. Immediate actions are also taken to restore spent fuel pool boron concentration.

If fflovi Ag fuel assefflbl i es 'oJAil e i A MODE 5 or § , LCD 3. 0. 3

'oJoul d Ast specify aAy acti GA . If fflovi Ag fuel assefflbl i es

'oJAi le i A MODES 1, 2, 3, aAd 4, tfie fuel fflOVeffleAt is indepeAdeAt of reactor operatioAs . Tfierefore , iAability to suspeAd fflOVeffleAt of fuel assefflblies is Aot sufficieAt reasoA to require a reactor sfiutdO'n'A .

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentrat ion of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program .

BYRON - UNITS 1 &2 B3 .7.15-5 Revision ::;..&

I NO CHANGES ON THIS PAGE I Spent Fuel Pool Boron Concentration B 3.7.15 BASES REFERENCES 1. Deleted

2. NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regula t ory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," dated August 19, 1998.

3. Deleted

4. Deleted

5. Holtec International Report, HI-982094, "Criticality Analysis for t he Byron/ Braidwood Rack Installation Project," Project No. 80944, 1998.

6. UFSAR, Section 15.7.4.

7. "Byron/ Braidwood Spent Fuel Pool Dilution Analysis,"

Rev. 3, dated June 17, 1997.

8 CN-CRIT-141 "Analysis Supporting the LTA Assemblies for Byron/ Braidwood SFP," dated February 4, 1999.

9. Double contingency principle of ANSI Nl6.l - 1975, as specified in t he April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

10. ANSI/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

11. Safety Evaluation Report (SER) dated October 25, 1996, issued by the Office of Nuclear Reactor Regulation for Topical Report WCAP-14416-NP-A "Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

BYRON - UNITS 1 &2 B3 . 7.15-6 Revision 68

Spent Fuel Assembly Storage B 3.7.16

~

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions . (For this discussion, the term OFA is intended to refer to the specific reduced fuel rodlet diameter, and includes all analyzed fuel types with this diameter, such as Vantage 5. ) The 24 spent fuel pool storage racks provide placement locations for a total of 2984 new or used fuel assemblies. Of these 24 spent fuel pool storage racks, four are designated "Region 1" with the remaining 20 racks designated as "Region 2." The analytical methodology used for the criticality analyses is in accordance with established NRC guidelines (Ref. 2).

Region 1 racks contain 396 cells which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes tha t spent fuel assemblies reside in all available cell locations). The stored fuel assemblies may contain an initial nominal enrichment of~ 5.0 weight percent U-235 (with or without IFBAs installed) (Ref. 5).

Region 2 racks contain 2588 cells which are also analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes tha t spent fuel assemblies reside in all available cell locations). For the "All Cells" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of

~ 5.0 weight percent U-235 with credit for burnup.

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under actual operating conditions.

BYRON - UNITS 1 &2 B 3. 7.16 - 1 Revision -68

INO CHANGES ON THIS PAGE I Spent Fuel Assembly Storage B 3. 7. 16 BASES APPLICABLE NRC approved methodologies were used to develop the SAFETY ANALYSES criticality analyses (Ref . 2) for the spent fuel pool storage racks. The fuel handling accident analyses are described in Reference 6. Additional evaluations were performed (Ref. 8) to support placement of the Byron lead test assemblies with higher density pellets in the spent fuel pool .

The criticality analyses for the spent fuel assembly storage racks confirm that keff remains : :; 0. 95 for the spent fuel pool storage racks (including uncertainties and tolerances) at a 95%probability with a 95%confidence level (95/ 95 basi s ),

based on the accident condition of t he pool being flooded with unborated water. Thus, the design of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safety margin to maintain spent fuel assembly storage rack keff : :; 0.95 (also on a 95/ 95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies. Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the "double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident (Refs. 9 and 10).

The accident analyses address the following five postulated scenarios :

1) fuel assembly drop on top of rack;

2) fuel assembly drop between rack modules;

3) fuel assembly drop between rack modules and spent fuel pool wall;

4) change in spent fuel pool water temperature; and

5) fuel assembly loaded contrary to placement restrictions.

Of these, only scenarios 2, 3, and 5 have the capacity to increase reactivity for the spent fuel pool storage racks.

Calculations were also performed for a spent fuel pool temperature of 4°C (39°F) which is well below the lowest normal operating temperature (50°F). Because the temperature coefficient of reactivity in the spent fuel pool is negative, temperatures greater than 4°C will result in a decrease in reactivity.

BYRON - UNITS 1 &2 B 3. 7 . 16 - 2 Revision 68

INO CHANGES ON THIS PAGE I Spent Fuel Assembly Storage B 3.7.16 BASES APPLICABLE SAFETY ANALYSES (continued)

For the fuel assembly misload accident, calculations were performed to show the largest reactivity increase caused by a Westinghouse 17X17 OFA fuel assembly misplaced into a Region 2 storage cell for which the restrictions on enrichment or burnup are not satisfied . The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

For the above postulated accident conditions, t he double contingency principle can be applied. Specifically, the presence of soluble boron in the spent fuel pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. For the spent fuel pool storage racks, spent fuel pool soluble boron has been credited in the criticality safety analysis to offset the reactivity caused by postulated accident conditions. Because the Region 1 racks are designed for the storage of fresh fuel assemblies, a fuel assembly misload accident has no consequences from a criticality standpoint (i.e., the acceptance criteria for storage are satisfied by all assemblies in the spent fuel pool).

Should a fuel assembly misload accident occur in the Region 2 storage ce 11 s, keff wi 11 be maintained ::; 0. 95 due to the presence of at least 300 ppm of soluble boron in the spent fuel pool water.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) .

BYRON - UNITS 1 &2 B 3. 7.16 - 3 Revision 68

I NO CHANGES ON THIS PAGE I Spent Fuel Assembly Storage B 3.7.16 BASES LCO The restrictions on the placement of fuel assemblies within the spent fuel pool in accordance with the requirements in the accompanying LCO ensure that the keff of the spent fuel pool will always remain~ 0.95 assuming the pool is flooded with unborated water for the spent fuel pool st orage racks .

For the spent fuel pool storage racks, in LCD Figure 3.7.16-1, the Acceptable Burnup Domain lies on, above, and to the left of the line.

The use of linear interpolation between minimum burnups is acceptable.

APPLICABILITY This LCD applies whenever fuel assemblies are stored in the spent fuel pool.

BYRON - UNITS 1 &2 B 3. 7 .16 - 4 Revision 68

Spent Fuel Assembly Storage B 3. 7. 16 BASES ACTIONS Tl9e ACTIONS 19aive BCCA ffiOEli fi eEI B)' a ~J o t e i AEli cati A§ tl9at LCO 3.0. 3 Eloes AOt apply .

A.1 When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the requirements of the LCD, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance.

If ffiOVi Ag fuel as seffibl i es 'vJAil e i A MODE 5 or 6' LCO 3. 0. 3

'oJOUl El AOt specify JAY acti OA . If ffiOVi Ag fuel asseffibl i es

'oJAi le i A MODES 1, 2, 3, aAd 4, tfie fuel ffiOVertleAt is inElepeAEleAt of reactor operatioA s. Tfierefore , iAa8ility to suspeAEI ffiOVeffieAt of fuel asseffislies is Aot sufficieAt reasoA to require a reactor sfiutdmm .

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS Item a and item bare performed, as applicable, is performed prior to storing the fuel assembly in the intended spent fuel pool storage location. The frequency is appropriate because compliance with the SR ensures that the relationship between the fuel assembly and its storage loca t ion will meet the requirements of the LCD and preserve the assumptions of the analyses.

This SR verifies by administrative means that t he initial nominal enrichment of the fuel assembly is met to ensure that the assumptions of the safety analyses are preserved.

SR 3.7.16.2 SR 3.7.16 .2 is performed prior to storing the fuel assembly in the intended spent fuel pool storage location. The frequency is appropriate because compliance wi t h the SR ensures that the relationship between the fuel assembly and its storage location will meet the requirements of the LCD and preserve the assumptions of the analyses.

This SR verifies by administrative means that t he combination of initial enrichment, burnup, and decay time, as applicable, of the fuel assembly is within t he Acceptable Burnup Domain of Figure 3.7.16-1 for the intended storage configuration to ensure that the assumptions of the safety analyses are preserved.

BYRON - UNITS 1 &2 B 3. 7 .16 - 5 Revision -68

I NO CHANGES ON THIS PAGE I Spent Fuel Assembly Storage B 3.7.16 BASES REFERENCES 1. Deleted

2. NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," dated August 19, 1998.

3. Deleted

4. Deleted

5. Holtec International Report, HI-982094, "Criticality Analysis for the Byron/Braidwood Rack Installation Project," Project No. 80944, 1998.

6. UFSAR, Section 15.7.4.

7. "Byron/Braidwood Spent Fuel Pool Dilution Analysis,"

Rev. 3, dated June 17, 1997.

8 CN-CRIT-141 "Analysis Supporting the LTA Assemblies for Byron/Braidwood SFP," dated February 4, 1999.

9. Double contingency principle of ANSI N16.1 - 1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

10. ANSI/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

11. Safety Evaluation Report (SER) dated October 25, 1996, issued by the Office of Nuclear Reactor Regulation for Topical Report WCAP-14416-NP-A "Westinghouse Spent Fuel Rack Criticality Analysis Methodology."

BYRON - UNITS 1 &2 B 3. 7.16 - 6 Revision 68