ML20141B778

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Proposed Tech Specs Revising TS Sections 3/4.6.1.6,4.6.1.2, 6.8.4 & 6.9.1.11 to Support New Requirements in 10CFR50.55a, Which Requires Utils to Update Existing Containment Vessel Structural Integrity Programs
ML20141B778
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 06/17/1997
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20141B761 List:
References
NUDOCS 9706240140
Download: ML20141B778 (45)
v • d • e


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ATTACHMENT B-1 i

MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37 and NPF-66 BYRON STATION UNITS 1 & 2 REVISED PAGES:.

3/4 6-3 3/46-8 l 3/46-9

< 3/4 6-10 I

B 3/4 6-2 6-20 6-23

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9706240140 970617 PDR ADOCK 05000454 P PDR

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CONTAINMENT SYSTfMS

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l SURVEILLANCE RE001REMENTS (Cor.tinued)

b. The reportino requirements and frequency of Type A tests shall be in accordance with Hegulatory Guide 1.163, September 1995, and 10 CFR 50, Appencix J, Option B.

j c. The accuracy of each Type A test shall be verified by a supplemental i

test conducted in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

1 d. Type B and C tests shall be conducted in accordance with Regulatory i Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

i e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;

f. Purge supply and exhaust isolation valves with resilieat material seals shall be tested and demonstrated OPERABLE by the requirements .

of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and l

. The provisions of Specification 4.0.2'are not applicable. l l

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BYRON - UNITS 1 & 2 3/4 6-3 f.MEMcM:%T NO. 91

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FROMs LICENSING FAM N0.8 639 663 7199 09.ge-e7 12819 P,89

! CONTAlleiDIT SYSTEMS l CONTAlletENT .YE53EL _5TRUCTURAL _ INTEGRITY LIMITING CONDITION FOR OPERATION hN

  • tt,; containment ;;;;;;l shall be l 3.s.1.6 i =i ' h;' :) : 1:7:1 ;-i.;ht;;t ;;it ;"h the e;cepter,ce iiit.,je in The etweetue.1 S t: rity SpeeWiettiene 4.0.1.0.1, t.C.I.O.2, G 4.".4.0.3.

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APPLICABILITY: MODES 1 E, 3, and 4. ,

4 l ACTION:

t hWith more than one predicted lower tendon limit andwith905 an observed of the radicted lift offlower force11betwee r i wit tendon below 905 of the ict lower limit, are the l .tandon(s W required level e integrity within tys and c..

! perform an a ring evaluation of the contal and provide a to Caemission within 30 s in accordance with

, M SERT Spec k1 Repo specification 6.g.2 o in at least ANDBY within the next i 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD th a following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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! b. With any abnotssi doge n of ructural int rity other than

ACT10N a. at a level ow the acceptan iteria o Specifications i 4.6.1.6.1, 4.6. . , and 4.6.1.6.3. restore containment vessel l to the level of integrity within 72 hou perform an engine evaluation of the containment and provi e scial i Re to the consission within 15 days in accordance vi i ification 6.9.2 or be in at least HDT STAMBY within the n

! . 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SWTDOWN within the fellowina 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS N l 4.6.1.6/ L.t2-- .i ": ngl hr.d=; ainment V;;&ei ter.br.e' struc-tural integritw:til 5^ dr =tr:t:d :t the rd f 1, 3, ;. .: : y=r; r,t ve:n1 ;trea s.1 i.,0. . itj test e,.4 e fell;;1Ig 07.; ir.iti:1 ;;r,t:ir * ^-

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J: J: .etretr' t;: 4 m D .rt- @'^ +h 3 ^=p.@pcile5tvuhm  ; r ,a mfubitu t

se vu a m. J Pi .

us - _ T 1

Determining tendons 5 dome (that6avertical random andIKitNfpr'HallGtTViisDpife'YWast a hoop) each have an ob ed

! 11 force within predicted, limits for each. For i subs spection one tendon from each gewup tapt l unchanged to a history and to correl observed data.

If the ebearved 1- force of any one in the original i semple population li the cted lower limit and 905 of
the predicted lower limit, _ . one on each side of this tendon should be checked f ft-off forces. If both of i these adjacent teodo' found to in their predicted

! 11alts, all three s should be resto the irod level of inte rity a single deficiency may be e e unique and i

accept . Unless there is abnormal degradation o containment l ring the first three ins ions, the semple p tion for

ves i

sequent inspections shall inc1 at least 10 tendons (3 l

3 vertical, and 4 hoop);

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was 's ea usy he v s s 's

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A=ve iewh $.IAr y.-AA= NEeNS Iv waYES$ NY O Ivw IvIwI N ~

Vel ums. Tvi- up te 35 t=6dv6 until,0e='eisls si SI M . s f,"r.=::" 7.0. O l BYROF. - UNITS 1 & E. 3/4 6-8 WcEE s2 diGT *N/AM j TF6*d OLP*ON ,

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j CONTAXNMENT SYSTEMS f SURVEILLANCE REQUIREMENTS (Continued) i

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Performing tendon detensioning. inspections, and material tests on i a previously stressed tendon from each group (dome, vertical, and I hoop). i l A randomly selected tendon from each group shall be comple ly m etensioned in order to identify broken or damaged wires and det i ing that over the entire length of the removed wire or stra that:  !

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1) e tendon wires or strands are free of corrosion, cra s, and d a g.e ,

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2) There re no changes in the presence or physical the sh thing filler grease, and -

pearance of  ;

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3) A minimum nsile strength of 240,000 psi (g ranteed ultimate 1 1

strength of he tendon material) for at le t three wire or  !

strand sample (one from each end and one t mid-length) cut from each remo j wire or strand. Fail e of any one of the t

wire or strand s les to meet the mi um tensile strength test i is evidence of abn 1 degradation j structura. the containment vessel }

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4 Performing >.endon retensioni i inspection to their observed 1 oft-th e tendons detensioned for.

of +6%. During et tensioning of f force with a toleranca limit 1 and elongation should be measur ese tendons, the changes in. load I

inultaneously at a minimum of three approximately equally s ced the seating force. If the e ongatio corresponding vels of force between zero and load differs by more than to a specific from that ecorded during installation, an investigation should made to ensur that the difference is not related to wire failure or slip of wires n anchorages;

d. Assuring the observ lift off stresses adjus d to account for elastic losses exc ed the average minimum dest value given below:

i one 143 ksi Vertical 144 ksi .

Hoop 140 ksi e.

Verifyin the OPERA 8ILITY of the sheathing filler grease assuring:

1) o voids in excess of 5% of the net duct volume, i

2 Minimum grease coverage exists for the different parts of No anchorage system, and

3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.

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, e l BYRON - UNITS 1 & 2 3/4 6-9 1

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i 4 integr 6. 2 End Anchorages and Adjacent Concrete Surfaces. The structural i f the end anchorages of all tendons inspected pursuant to cation 4.6. . and the adjacent concrete surfaces shall be d cifi-determining throu rated by visual appearance of inspection that no apparent changes hav curred in the

nd anchorage or the concrete ck patterns adjacent to the end anchorages sinc st inspected. Insp i

be performed during the contai vessel t ons of the concrete shall tion 4.6.1.6.1). n tests (reference Specifica-I 4.6.1.6.3 Containment Vessel aces.

exposed accessible interi The s tural integrity of the including the liner nd exterior surfaces o e containment vessel,

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e, shall be determined during t Type A contai eakage rate test (reference Specificatio utdown for each 4

'i visual ins on of these surfaces. This inspection shall be pe .1.2) by a i to th d prior e A containment leakage rate test to verify no apparent chan in

} arance or other abnormal degradation.

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j BYRON - UNITS 1 & 2 3/4 6-10

l CONTAINMENT SYSTEMS 1 I

BASES 1

3/4.6.1.5 AIR TEMPERATURE l

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The limitations on containment average air temperature ensure that the 1

overall containment average air temperature does not exceed the initial

temperature condition assumed in the accident analysis for a steam line i

break accident. Measurements shall be made at all of the listed running fan j locations, air temperature. whether by fixed or portable instruments, to determine the average

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, 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY

' This limitation ensures that the structural integrity.of the containment will the facility.

be maintained comparable to the original design standards for the life of Structural integrity is required to ensure that the containment will withstand the maximum pressure of 44.4 psig in the event of a cold leg double-ended break accident. S: =::r c;nt f :: 'd= t t::ca Mft ;ff e, the tensile tests of the tendon wires or strands, the visual examina Qg of te s, anchorages and exposed interior and exterior surfaces of t containee and the Type A leakage test are sufficient to demonst 6 leapability. this The Surveillance Re ements for demonstratin containment's l

structural integrity are in 11ance with the ommendations of proposed 1

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Rev. 8 to Regulatory Guide 1.35, ervic rveillan:e of Ungrouted Tendons I in Prestressed Concrete Containment S ures," April 1979 and proposed Regulatory Guide 1.35.1, "Deterni Pres ing Forces for Inspection of Prestressed Concrete Contai s," April 1979.

The required S al Reports from any engineering ev abnormalties s include a description of the tendon condit tion ofthe containment condition of the con e (especially at tendon anchorages), the inspection p edure, the t ances on cracking, the results of the engineering evaluation a he

c. ective actions taken.

3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed (power removed) during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves sealed close.d during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 48-inch contain-ment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator.

The use of the containment purge lines.is re'stricted to the 8 . inch purge '

supply and exhaust isolation valve's since, unlike the 48-inch' valves; the 8-inch valves are capable of closing during a LOCA or steam line break accident.

Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not BYRON - UNITS 1 & 2 B 3/4 6-2

1 ADMINISTRATIVE CONTROLS I

PROCEDURES AND PROGRAMS (Continued)

I 2) 1 1

i A Land Use Census to ensure that changes in the use of areas at j and beyond the SITE B0UNDARY are identified and that modifica-j tions to the monitoring program are made if required by the results of this census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the i measurements of radioactive materials in environmental sample ERT matrices are performed as part of the quality assurance program for environmental monitoring.

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REPORTING REQUIREMENTS 1' ,

j ROUTINE REPORTS J

l 6.9.1 In addition to the applicable reporting requirements of Title 10, Code i of Federal Regulations i

AdministratoroftheNkCRegionalOficeunlessotherwisenoted.thefollowingrep l STARTUP_ REPORT i

6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following:

thelicenseinvolvingap(1)receiptofanOperatingLicense,(2)amendmentto lanned increase in power level, 3)installationof i

fuel that has a different design or has been manufactured (by a different fuel j supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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6.9.1.2 The Start

! Final Safety Analy s Report ort shall address each of the tests identified in the '

FSAR and shall include a description of the measured

values of the operating conditions or characteristics obtained during the test i

program tions. Any and a comparison of these values with design predictions and specifica-corrective actions that were required ".o obtain satisfactory opera-i tion shall also be described. Arty additional specific details required in l license conditions based on other commitments shall be included in this report.

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! 6.9.1.3 Startup Reports shall be submitted within: '

i plation of the Startup Test Program, (2 90 days foll(ow)ing resumption or com j eencement of commercial power operation), or (3) 9011owing months initial

! criticality, whichever is earliest. If the Startup Report does not cover all i

i threetionevents res be s (i.e., initial or commencement criticalityl of commercia operation) completion supplementary of Startup Test Program and rep i

) itted at least every 3 months until all three events have been completed.

! ANNUAL REPORTS

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6.9.1.4 Annual Reports covering the activities of the unit as described below a for.the >revious caleridar year shall'be submitted prior to . March.1 af. each ' '

i year.

Tw initial report shall be submitted prior to March 1 of the year j

following initial criticality.

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j BYRON - UNITS 1 & 2 6-20 fri=CC %. 50

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ADMINISTRATIVE CONTROLS i CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAG 6.9.1.10 Fuel enrichment limits for storage shall be established and documented STORAGE RACKS. in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STA The analytical methods used to detemine the maximum fuel enrichments shall be those previously reviewed and approved by the NRC in

" CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION The FUEL STO fuel enrichment limits for storage shall be detemined so that all applicable limits (e.g., subcriticality) of the safety analysis are met.

i The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOO STATION FUE RACKS report shall be provided upon issuance of any chanms, to the NRC l

j Document Control Desk, with co les to the Regional Admin strator arid the Resident Inspector. g' .

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

i 6.10 RECORD RETENTION

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i In addition to the applicable record retention requirements of Title 10,

Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1 I

The following records shall be retained for at least 5 years:  ;

a.

Records and logs of unit operation covering time interval at each l

! power level;

! b.

( Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to t

nuclear safety;

! c. All REPORTABLE EVENTS;

d.

Records of surveillance activities inspections, and calibrations

' requiredbytheseTechnicalSpecifications; e.

Records of changes made to ^.he procedures required by Specification 6.8;

f. Records of radioactive shipments; g.

Records and of sealed source and fiasion detector leak tests and results;

h.

Records of annual physical inventory of all sealed source material

of record.

6.10,2 The Operating following records shall be retained for the duration of the unit License:

3

, a.

Records and drawing changes reflecting unit design modifications made Report;to systems and equipment described in the Final Safety Analysis; l /> -

b.

Records of'new and irradicted fuel inventory, fuel transfers and

) assembly burnup histories; BYRON - UNITS 1 & 2 6-23 AMENDMENT NO.- 3;

i l INSERT A

g. The structural integrity of the exposed accessible interior and exterior surfaces of the

. containment vessel, including the liner plate, shall be demonstrated during the shutdown for each Type A containment leakage rate test by a visual inspection of these surfaces.

This inspection shall be performed at a frequency in accordance with Regulatory Guide 1.163, September 1995, to verify no apparent changes in appearance or other abnormal degradation.

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i INSERT B If containment is found to be inoperable, restore the containment to OPERA.BLE status within I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1 INSERT C l

Surveillance Requirement 4.6.1.6 ensures that the structural integrity of the containment will be l maintained in accordance with the provisions of the Containment Structural Integrity Program. I Testing and frequency are consistent with the requirements of 10 CFR 50.55a(b)(2)(vi), )

" Effective edition and addenda of Subsection IWE and Subsection IWL,Section XI," and 10 CFR 50.55a(b)(2)(ix), " Examination of concrete containments." Predicted tendon lift-off forces will be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990.

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' INSERT D Containment Vessel Structural Integrity Program g.

This program provides controls for monitoring containment vessel structural integrity including

routine inspections and tests to identify degradation and corrective actions if degradation is found. i The Containment Vessel Structural Integrity Program, incoection frequencies, and acceptance l criteria shall be in accordance with 10 CFR 50.55a(b)(2)(vi)," Effective edition and addenda of  ;
Subsection IWE and Subsection IWL,Section XI," and 10 CFR 50.55a(b)(2)(ix)," Examination (

! of concrete containments," as modified by approved exemptions. Predicted tendon lift-off forces l l

shsil be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990. l l

In addition, Unit 1 may have sheathing filler grease voids in excess of 5% of the net duct volume j for up to 35 tendons until the end of B1R08.

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INSERT E CONT 6INMENT VESSEL STRUCTURAL INTEGRITY REPORT l 6.9.1.11 Any abnormal degradation of the containment structure detected during the tests required by the Containment Vessel Structural Integrity Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

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_ . - -. . _. -. . _ . - - . _ - - . - . _ . - . -._ . _..-. _ . _ . ~ _ - . _ . _ . . . - ..

d l ATTACHMENT B-la MARKED UP PAGES FOR PROPOSED CHANGES TO PROPOSED IMPROVED TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37 and NPF-66 l

BYRON STATION UNITS 1 & 2 l l

REVISED PAGES:

3.6-l

  • 3.6-2 l 5.0-10 l 5.0-44 B 3.6-5 B 3.6-6
  • Page provided for continuity only. No changes are being made to this page.

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{' Contamment 4

3.6.1 j 3.6 CONTAINMENT SYSTEMS j 3 6.1 Containment i-

! LCO 3.6.1 1

Containment shall be OPERABLE.

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l APPLICABILITY: MODES 1. 2, 3, and 4.

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ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4

A. Containment A.1
inoperable. Restore containment I hour to OPERABLE status.

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! B. Required Action and B.1 Be in MODE 3.

i associated Completion 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> i

Time not met. A@

! j B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> I s

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' SURVEILLANCE REQUIREMENTS s

SURVEILLANCE FREQUENCY SR 3.6.1.1 Perform required visual examinations and In accordance leakage rate testing except for containment with the air lock testing, in accordance with the Containment i Containment Leakage Rate Testing Program.
Leakage Rate 4

Testing Program 3

I (continued) l ,

4 BYRON - UNITS 1 & 2 3.6-1 i Revision A v ., m- ~-,--wr.-,y-, v --,_v , --,. . r ,,,,-w -,r- -

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Containment 3.6.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.2 Verify containment structural integrity in in accordance l accordance with the Containment Tcr.dcr. with the  !

Surveillec.ce Program ~

Containment l Vesse l bWbral '

>$[72;'11ar,c3 Program Ante 3r,q 1

l BYRON - UNITS 1 & 2 3.6-2 Revision A

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Programs and Manuals  !

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-m 5.5 Programs and Manuals (continued) ygg gQgl gg VV )

5.5.6 Pre-Stressed Concrete Containment Tendco SNeillence Proaram j T i ram provides controls for monitoring any tendon

( degradatl

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re-stressed concrete containments.

[ NMC effectiveness o .

corrosion protection

. u ing

.. to ensure  !

s

' containment structura rity. T ram shall include i baseline measurements prior t

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1040A Surveillance Program, i 1al operations. The Tendon I ion fre 'es and acceptance criteria shall b eneral conformance w1 osed Regulatory Guide 1.3 1sion 3. April 1979 and proposed Reg r Gu' . 5.1. April 1979.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 Reactor Coolant Pumo Fivwheel InsDection Procram This program shall provide for the inspection of each reactor coolant pump flywheel in general conformance with the recomendations of Regulatory Position c.4.b of Regulatory Guide 1.14. Revision 1. August 1975.

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i (continued) ' i BYRON - UNITS 1 & 2 5.0-10 Revision.A -

I

Reporting Requirements

! 5.6 5.6 Reporting Requirements (continued)

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i 5.6.8 .- ~..._:,,__A r m __ e..._

.m. . ~ peoort onhinment N15d

- 1 n.g;g w Any abnormal degradation of the containmen structure detected

%gsel during the tests required by the Prc strc;;cd Concr;te Containment Stvacdum ' .* Tendon % rvcillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon 74[' condition, the condition of the concrete (especially at tendon anchorages). the inspection procedures the tolerances on cracking, and the corrective action taken.

5.6.9 Steam Generator (SG) Tube insoection Reoorts 4
a. Following each inservice inspection of SG tubes. the number of tubes plugged or repaired in each SG shall be reported to the NRC within 15 days.

j b. The complete results of the SG tube inservice inspection i; shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1. Number and extent of tubes inspected.

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2. Location and percent of wall thickness penetration for j- each indication of an imperfection. and
3. Identification of tubes plugged or repaired.

c.

Results of SG tube insmctions that fall into Category C-3 shall be reported to tie NRC within 30 days and prior to resumption of unit operation. The report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to

prevent recurrence, i

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(continued) i BYRON - UNITS 1 & 2 5.0-44 1

Revision-A-i

Containment

' 5 3.6.1 BASES (continued)

SURVEl(LANCE SR 3 6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visJal examinations and leakage rate test )

requirements of the Containment Leakage Rate Testing Program. Failure to meet air lock and purge valve leakage limits specified in LC0 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A.

B. and C leakage causes the limits to be exceeded. As left leakage prior to the first startup after performing a required leakage test is required to be < 0.6 L, for combined Type B and C leakage following an outage or shutdown that included Type B and C testing only, and

< 0.75 L, for overall Type A leakage following an outage or shutdown that included Type A testing. At all other times between required leakage rate tests. the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program.

These periodic testing requirements verify that the contai sent leakage rate does not exceed the leakage rate assumed in the safety analysis.

SR 3 6.1.2 V4.5 se 1 5tmchnI This SR ensures that the structural integrity of the containment provisions of thewill be maintained Containment TcndoninSurveillancc accordance with A Program. ,

Testing and Frequency are consistent with the g ggnh recc=cnd0tions of propc;cd R; ulator ' Ouide 1.3}.

, pcVi; ion 3 'Ref. O and pre o;e;d R uiator Guide 1.35.1. ,

(Ref.

Af veurnmendn&ns o r ww i lo CFR 50.Ch.db)M%vi) (Aef b On 30 de n (b) C06x) ac S. s) k Am.wbt.d 89 ( Ref. 7), N NRC aproveJ w i provisica b allow Owd I b have shestkn Oller reuse voids

- i y G*l, of N nel dod. vobi. for .og ,le I h.o ou unfil -

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l Nt, e,nd of IIRo%. '

(continued)

BYRON - UNITS 1 & 2 B 3.6 5 Revi si on-A--

l

Containmen:

B 3.6.1 BASES (continued)

REFERE$CES 1. 10 CFR 50. Appendix J. Option B.

2. UFSAR Chapter 15.
3. UFSAR. Section 6.2.

> 4. (

April 1979.

5 PrcosedRcgulatoryCuide1.05.Re,-ision;S uly 199 Propc ed Regulatory Guide 1.35.1. W ..

~C

4. 10CFR 50.Sb(bMe)(vi)
5. io aA so sfo-(bMM d F

/

7. NRt safel Evaloth y

Report for Amendment l

$$l b hed47 opted:a3 Licensu N PF- 3'l and N P F - 66, May 6,l197, ,

4 BYRON - UNITS 1 & 2 B 3.6-6 Revision A.-

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INSERT 5.0-10A i This program provides controls for monitoring containment vessel structural integrity including routine inspections and tests to identify degradation and corrective actions if degradation is found. The Containment Vessel Stmetural Integrity Program, inspection i frequencies, and acceptance criteria shall be in accordance with 10 CFR 50.55a(b)(2)(vi) and 10 CFR 50,55a(b)(2)(ix) as modified by approved exemptions. Predicted tendon lift-

! off forces shall be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990.

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ATTACilMENT B-2 MARKED UP PAGES FOR PROPOSED CHANGES TO

! APPENDIX A, TECHNICAL SPECIFICATIONS, OF l FACILITY OPERATING LICENSES NPF-72 and NPF-77 l

BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES: l I

3/4 6-3 3/4 6-8 3/4 6-9 3/4 6-10 B 3/4 6-2 6 20 ,

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CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

b. The reporting requirements and frequency of Type A tests shall be in accordance with Regulatory Guide.l.163, September *995, and 10 CFR 50, Appendix J, Option 8.
c. The accuracy of each Type A test shall be verified by a suppiamental test conducted in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

! d. Type 8 and C tests shall be conducted in accordance with Regulatory l Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

i e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3; l f. Purge supply and exhaust isolation valves with resilient material i seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and

h. .

The provisions of Specification 4.0.2 are not applicable. l l

SERT b l

'f l BRAIDWOOD - UNITS 1 & 2 3/4 6-3 0;CN^4;ENT NG. 75 l

l .FROMI

_ _LICEHS!HG

________ FM H0.: 630 663 P199 83-09-97 12:11 P.16 j CONTAftME!E SYSTEMS j c0NTAINNENT VESSEL STRUCTURAL INTERRITY , gpgg,

[ LIMITING-CONDITION FOR OPERATION i

3.6.1.6 The strechral ir.t;;rity Of th: containment veseel shall be v

r. int::::d :ti :m 14.::1.:.c.
r,.;ific.  ::?,:ict4.:t .dth the.r4

.c.:, ecce,ter.ce 4.c.L. .s. criterie ir, l

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! APPLICABILITY: MODES 1 2, 3, and 4.

1

! Ellalli .

With more than one tendoe with an observed lift-off force betwo predicted lower limit and 905 of the predicted lower 11 or wit e tendon below 905 of the predicted lower limit. tore the tendon the required level of integrit,y withi days and perform an (s e ring evaluation of the conta t and provide a i

i hNg Special Repo to Commission within 30 s in accordance with

! q Specification 6.9.2 e in at least STANDBY within the next l

u 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD th a following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .,

1 i b. With any abnormal degrad of ructural integrity other than i

ACTION a. at a level aw the acceptanc iteria of $pecifications

! 4.6.1.6.1, 4.6. . and 4.6.1.E.3, restore containment vessel

! to the req 1evelofintegritywithin72 hour perform an

! engin evaluation of the containment and provide >ecial

! Re to he Commission within 15 days in accordance wit N j ecification 6.9.2 or be in at least NOT STANDBY within the nhL

k,_ 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in CED SHUTDOWN within the followina 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. N E RY.EF cANcE REDUIREMENTS A __ __

4.6.1.6 / t--t:* g t ": { { Q . , ainment,v::gi,te-4;.;"struc-g 1"**Ei%nni m.e r r_3 x = 4:u m w : m n T m n-' m m .

rm-m .

= m r yeef mA.=,

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..... . . d by : ( 6 o c w ur=w= don

"'u -T e-uidh fhe. C+ontain "'= eent

.re'essv ei i _wn 2 W ^' MT% 96'*:_ _ .

. potermining snaT. a ranov- ous ..cresentenver . rie or as naass j tendons (5 done, 6 vertical, and 8 hoop) each have an obse ,

11 force within predicted limits for each. For sa j subs spection one tendon from each group i unchanged to 1 a history and to correlate observed data.

i If the observed 1 ff force of any one n in the original '~

i sample population lies the p ad lower limit and 905 of the predicted lower. limit, , one on each side of this

! tendon should be diecked fo -off forces. If both of I these ad,jacent tendou ound to be in their predicted j 4 Timits, all three a should be resto the recutred level 1 of integrity, a siigle deficiency may be con ret unique and i

acceptab . nless thore is abnormal degradation o containment ves uring the first three inspections, the susple pop on for sequent inspections shall include et least 10 tendons (3 d 3 vertical. and 4 hoop): N 1

4

%it 1 u h.se ;h;;thir.G " illa.~;. n 0;id: ^: 0C= " " Of th:

f.41,1=, atil tha ad ;f f.127.

=t drt i ver fer :; t: :: txt:: fr i

l BRAIDWO60 - UNIT 5 1 & 2 3/4 6-8 CM::MCMT X0. 01 l w sT d ate os se eiz Asst 4 >m 1

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i CONTAINMENT SYSTEMS l

j SURVEILLANCE REQUIREMENTS (Continued)

Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (done, vertical, and i hoop).

A randomly selected tendon from each group shall be comple y

!

  • etensioned in order to identify broken or damaged wires and date ning that over the entire.. length of the removed wire or stran that:

~ ~

! 1)

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The teridon wires or strands are free of corrosion, crac's, and j ge,

! 2) The are no changes in the presence or physical earance of

{ the s athing filler grease, and

3) A minimum ensile strength of 240,000 psi (g ranteed ultimate i

j strength o the tendon material) for at lea three wire or strand samp) (one from each and and one t mid-length) cut  !

!~ from each reso d wire or strand. Fail e of any one of the wire or strand s les to meet the sin tensile. strength test

! is evidence of ab real degradation i the containment vessel structure.

c. Performing tendon retension

{ inspection to their observed of ft-th a tendons detensioned for f force with a tolerance limit

of +6%. During retensioning o ese tendons, the changes in load j and elongation should be measur inultaneously at a minimum of i

three approximately equally s ced evels of force between zero and the seating force. If the e ongatio corresponding to a specific load differs by unre than from that ecorded during installation, j an investigation should j

made to ensu that the difference is not related to wire failure or slip of wires n anchorages; I

d. Assuring the observ lift off stresses adjus d to account for i

elastic losses exc ed the average minimum dest value given below:

i i one 143 ksi l Vertical 144 ksi

] Hoop 140 ksi

) e. Verifyi the OPERABILITY of the sheathing filler grease assuring:

1) voids in excess of 5% of the net duct volume, 2 Minimum grease coverage exists for the different parts of

- e anchorage system, and

3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.

BRAIDWOOD - UNITS 1 & 2 3/4 6-9

1 CONTAINMENT SYSTEMS

.SURVEILLANCERE00fREMENTS(Continued) 1 m.

. 6. 2 i

intog End Anchoraces and Ad cation 4. .of the end anchorages o.facent concrete Surfaces. P determining th 1 oand the adjacent that no appar concrete surfacesecifi-- shall(

1 visual appearance inspection strated by to the end anchorages sinend anchorage or the concretent changes ccurred h the in be performed during the conta ast inspected. Ins ack patterns adjacent 1

tion 4.6.1.6.1). t vessel ons of the concrete shall i i

4.6.1.6.3 n tests (reference Specifica- <

exposed acc_ Containment Vessel essible aces. The including the liner inter and exterTor surfacesuctural integrity of the Type A contai visual ins eakage, rate testshall be determined duringthe containm to th reference Specificatshutdown for each I

a e A containment leakageapparent rance or other abnormal degradation.

ratec ormed e testpriorto v 1 s in

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BRAIDWOOD - UNITS 1 & 2 ,

3/4 6-10

1 .

) CONTAINrtENT SYSTEMS l -

j 8ASES k

i 3/4.6.1.5 AIR TEMPERATURE i

The Ifmitations on containment average air temperature ensure that the j overall containment average air temperature does not exceed the initial 4

temper.a.ture condition assumed in the accident analysis for a steam line i break accident. Measurements shall be made at all of the listed running fan t

locations, whether by fixed or portable instruments, to determine the average

{ air temperature.

3/4.6.1.6 CONTADetENT VESSEL STRUCTURAL INTEGRITY l

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This liettation ensures that the structural integrity of the containment 3 will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of _44.4 psig in the event of a cold leg

} double-ended break accident. ^: :==t ;f centei;.;;.;nt tenden Mit efi e, the tensile tests of the tendon wires or strands, the visual examinati

of ns, anchorages and exposed interior and exterior surfaces of the
T containee and the Type A leakage test are sufficient to demonstr is j LrisERT capability.

Y b The Surveillance resents for demonstrating containment's i structural integrity are in liance with the ndations of proposed j k Rev. 3 to Regulatcry Guide 1.35, arvice eillance of Ungrouted Tendons

in Prestressed Concrete Containment S res," April 1979 and proposed a Regulatory Guide 1.35.1, "Determin res ing Forces for Inspection of l } Prestressed Concrete Contai , " April 1979.

l The required S Reports from any engineering av tion of containment

abnonsalties sh nelude a description of the tendon condit the condition of the con (especially at tendon anchorages), the inspection p dure,

! the to ances on cracking, the results of the engineering evaluation a c tive actions taken.

I j 3/4.6.1.7 CONTAIW4ENT PURGE VENTILATION SYSTEM i

The 48-inch containment purge supply and exhaust isolation valves are

required to be sealed closed (power removed) during plant operations since these 4

valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 48-inch contain-

! ment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or preverits power from being supplied to the valve operator, j The use of the containment purge lines is restricted to the 8-inch purge

supply and exhaust isolation valves since, unlike the 48-inch valves, the
8-inch valves are capable of closing during a LOCA or steam line break accident.

j Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not BRAIDWOOD - UNITS 1 & 2 8 3/4 6-2 C i

AUMINISIKAIIVt. LUNIHUL5 PROCEDURES AND PROGRAMS (Continued)

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and r D 3) Participation in a Interlaboratory Comparison Program to

/( s ensure that independent checks on the precision and accuracy f of the measurements of radioactive materials in environmental

_Lt45ERT' I g j sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.9 REPORTING RE0VIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

STARTUP REPORT 6.9.1.1 A sumary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The Startup Report shall address each of the tests identified in the Final Safety Analysis Report FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following com-pletion of the Startup Test Program, (2) 90 days following resumption or com-mencement of comercial power operation, or (3) 9 months following initial

, criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or comencement of comercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual Reports covering the activities of the facility as described l below for the previous calendar year shall be submitted prior to March 1 of each year.

BRAIDWOOD - UNITS 1 & 2 6-20 AMEN 0HENT NO. 59

ADMINTSTRATIVE CONTROLS CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS 6.9.1.10 Fuel enrichment limits for storage shall be established and documented in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS. The analytical methods used to determine t5e maximum fuel enrichments shall be those previously reviewed and approved by the NRC in

" CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE The RACKS."

fuel enrichment limits for stcrage shall be determined so that all applicable limits (e.g., subcriticality) of the safety analysis are met.

The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATIUN FUEL STORAGE RACKS report shall be provided upon issuance of any changes, to the NRC Document Control Desk, with co ies to the Regional Administrator and the Resident Inspector.

I ECIAL REPORTS ~UJsERT 6

\

[ 6.9.2 Special reports shall be submitted to the Regional Administrator oi the NRC Regional Office within the time period *oecified for each report.

6.10 RECORD kETtNTION

' In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at eati power level; b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety;

c. All REPORTABLE EVENTS;
d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;
e. Records of changes made to the procedures required by Specification f.8;
f. Records of radioactive shipments; g.

Records of sealai source and fission detector leak tests and results; and

h. Records of annual pi:ysical inventory of all sealed source material of record.

6.10.2 The following records saall be retair.ed for the duration of the unit Operating License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report;
b. Records of new and irradiated fuel inventory, fuel trans"ers and assembly bJrnup histories; BRAIDWOOD - UNITS 1 & 2 6-23 AMENDMENT NO. %

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INSERT A

)

g. The structural integrity of the exposed accessible interior and exterior surfaces of the i j containment vessel, including the liner plate, shall be demonstrated during the shutdown l 1 for each Type A containment leakage rate test by a visual inspection of these surfaces.  !

I This inspection shall be performed at a frequency in accordance with Regulatory Guide 1.163, September 1995, to verify no apparent changes in appearance or other abnormal degradation.

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. . . - . . . . - - _ - . . - _ = - . . _ . - - - . . . . - . . . _ .

I'NSERT B If containment is found to be inoperable, restore the containment to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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I INSERT C Surveillance Requirement 4.6.1.6 ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Structural Integrity Program.

Testing and frequency are consistent with the requirements of 10 CFR 50.55a(b)(2)(vi),

" Effective edition and addenda of Subsection IWE and Subsection IWL,Section XI," and Section 10 CFR 50.55a(b)(2)(ix), " Examination of concrete containments." Predicted tendon lift-off forces will be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990.

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d INSERT D s

g. Contai;nnent Vessel Structual Integrity Program 5

This program prov,' des controls for monitoring containment vessel structural integrity including routine inspectioas and tests to identify degradation and corrective actions if degradation is fourd The Containment Vessel Stmetural Integrity Program, inspection frequencies, and acceptance criteria shall be in accordance with 10 CFR 50.55a(b)(2)(vi)," Effective edition and addenda of Subsec. ion IWE and Subsection IWL,Section XI," and 10 CFR 50.55a(b)(2)(ix), " Examination of concrete containments," as modified by approved exemptions. Predicted tendon lift-off forces shall be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990.

In addition, Unit 1 may have sheathing filler grease voids in excess of 5% of the net duct volume for up to 35 tendons until the end of AIR 07.

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INSERT E 4

CONTAINMENT VESSEL STRUCTURAL INTEGRITY REPORT i

6.9.1.11 Any abncrmal degradation of the containment structure detected during the tests i required by the Containment Vessel Structural Integrity Program shall be reported to the i NRC within 30 days. The repon shallinclude a description of the tendon condition, the

] condition of the concrete (especially at tendon anchorages), the inspection procedures, j the tolerances on cracking, and the corrective action taken.

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ATTACHMENT B-2a MARKED UP PAGES FOR PROPOSED CHANGES TO PROPOSED IMPROVED TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 and NPF-77 BRAIDWOOD STATION UNITS 1 & 2 1

REVISED PAGES:

3.6-1

  • 3.6-2 5.0-10 j 5.0-44
B 3.6-5 B 3.6-6

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  • Page provided for continuity only. No changes are being made to this page.

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Containment 3.6.1 i 3.6 CONTAINMENT SYSTEMS 3.6.1 Containraent i

j LCO 3.6.1 Containment shall be OPERABLE.

4 6

I APPLICABILITY: MODES 1.-2. 3. and 4.

i ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME

A. Containment A.1 Restore containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
inoperable. to OPERABLE status.

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B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />  !

associated Completion '

Time not met. .AR  ;

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.6.1.1 Perform required visual examinations and In accordance leakage rate testing except for containment with the air lock testing, in accordance with the Containment Containment Leakage Rate Testing Prograra. Leakage Rate Testing Program (continued) 1 4

BRAIDWOOD - UNITS 1 & 2 3.6-1 Revision A

.- .. = . - . .- -, . .- . -- . . ~- - . . _ = . -.

Containment 3.6.1~

SURVEILLANCE REOUIREMENTS (continued)

SURVEILLANCE FREQUENCY 1

SR 3.6.1.2 Verify containment structural integrity in In accordance accordance with the Containment-Tcaden n with the l Surycillcnce Program, n .

Containment Vessel Strachr^l fr$?,1cace i i

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Programs and Manuals 5.5 5.5 Programs and Manuals (continued) -- pef [bbaf._lnYg,'k

-u v -

5.5.6 Pre-Stressed Concrete Containment Tcadon Sur Veillance Proaram ram provides controls for monitoring any tendon degradatl re-stressed concrete containments uding Msep. t- effectiveness o i rrosion protection , to ensure y containment structural 1 't , ogram shall include Sg _to A baseline measurements prior operations. The Tendon Surveillance Program ction freque and acceptance criteria shal n general conformance with p d Regulatory Guide ,

evision 3. April 1979 and proposed Regula

.1.35.1. Aoril 1979 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.7 Reactor Coolant Pumo Flywheel Insoection Proaram This program shall provide for the inspection of each reactor coolant pump flywheel in general conformance with the recommendations of Regulatory Position c.4.b of Regulatory Guide 1.14. Revision 1. August 1975.

4 (continued)

BRAILU000 - UNITS 1 & 2 5.0-10 Revi sion-A--

Reporting Requirements 5.6' L

5.6 Reporting Requirements (continued) t 'ConNnment Osscl ShMA !

5 Tendon Surecilknce Reoort ln ;q l Any abnormal degradation of the containment stru re detected i fene 5 duringthetegsrequiredbythePrc-strc:cdConcret+ Containment 3 c.mn- survci . .ance Program shall be reported to the NRC within 1

1

.$4<#ctwrA( 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon T b anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.9 Steam Generator (SG) Tube Insoection Reoorts

a. Following each inservice inspection of SG tubes the number of tubes plugged or repaired in each SG shall be reported to j the NRC within 15 days.  ;

i

b. The complete results of the SG tul' inservice inspection  ;

shall be submitted to the NRC wiu.in 12 months following the completion of the inspection. The report shall include:

1. Number and extent of tubes inspected, i
2. Location and percent of wall thickness penetration for j each indication of an imperfection, and 3 1
3. Identification of tubes plugged or repaired, j l
c. Results of SG tube insactions that fall into Category C-3  ;

shall be reported to tie NRC within 30 days and prior to  :

resumption of unit operation. The report shall provide a  !

description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

i l

l (continued)

BRAIDWOOD - UNITS 1 & 2 5.0-44 Revision A

Containment B 3.6.1 i

BASES (continued)

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. Failure to meet air lock and purge valve leakage limits specified in LC0 3.6.2 and LC0 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A.

B. and C leakage causes the limits to be exceeded. As left l leakage prior to the first startup after required leakage test is required to be < performing 0.6 L, for a combined Type B and C leakage following an outage or shutdown that included Type B and C testing only, and

< 0.75 L, for overall Type A leakage following an outage or shutdown that included Type A testing. At all other times i between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L the offsite dose consequences are bounded by the I assumpt, ions of the safety analysis. SR Frequencies are as '

required by the Containment Leakage Rate Testing Program.

These periodic testing recuirements verify that the containment leakage rate coes not exceed the leakage rate assumed in the safety analysis.

SR 3.6.1.2 YsS%\ SkcS"O

% ,cq This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tenden Surycilkncc Program.

Testing and Frequency are consistent with the repme4J , _ __. _ . .. of propc:cd Regu htcr" Ouidc 1.30 Revide" 3 (Ref ')andpropc:cdRegulatoryGuide1.35.1.

(Ref. ).

k recometwin%ns d I

lo cM E0 F5- db)lIXVO d'E N ""E h p CFR 50 55 a (b)llRix) LR5f N l , v In Amewkesk sl ( Rst. 7) .he NRc ateroved a provisto s k EUow Onif I lo km.vt ska% filler 3tean voM4 7F7.

of b ari dad unlamt for o, 3 f b do u o d il 4

(

~ Mwd ' '

BRAIDWOOD - UNITS 1 & 2 B 3.6-5 Revision [

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Containment i B 3.6.1 l BASES (continued)

REFERENCES 1. 10 CFR 50 Appendix J. Option B.

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2. UFSAR, Chapter 15. l l
3. UFS/R, Section 6.2. l
4. {rog;ggegulator., Guide 1.05 Rc.isio , 3. l

, , , , ~ , , .

G, 1. orep~ed Regulatory Guide 1.35.1, .^pril 1979. July 1990 J  !

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4. 10 CFR So .55 A (b)(.1)(VO. -

% 1 g, MRc Sakh Evalvdeow Reporf Ar Amendmed) si b facility operdig Llcence N P F-72, and WP E - 17 s M*y 6,199 7.

BRAIDWOOD - UNITS 1 & 2 B 3.6-6 Revision A

. _ . _ . _ . _ . _ _ . . _ _ _. __ _ _ _ ~ . . . _ . _ .. _ ..

INSERT 5.0-10A This program provides controls for monitoring containment vessel structural integrity 1 including routine inspections and tests to identify degradation and corrective actions if degradation is found. The Containment Vessel Structural Integrity Program, inspection frequencies, and acceptance criteria shall be in accordance with 10 CFR 50.55a(b)(2)(vi)

}

and 10 CFR 50.55a(b)(2)(ix) as modified by approved exemptions. Predicted tendon lift-off forces shall be determined consistent with the recommendations of Regulatory Guide 1.35.1, July 1990.

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ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A, l TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES l NPF-37, NPF-66, NPF-72, AND NPF-77 l l

Comed has evaluated this proposed amendment and determined that it involves no significant i hazards considerations. According to Title 10 to the Code of Federal Regulations Part 50 l Section 92 Paragraph c (10 CFR 50.92 (c)), a proposed amendment to an operating license  ;

involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident l

previously evaluated; or 1

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
3. Involve a significant reduction in a margin of safety.

A. INTRODUCTION Commonwealth Edison (Comed) proposes to amend Technical Specification (TS) 3.6.1.6,

" Containment Vessel Structural Integrity," and add new TS 6.8.4.g and TS 6.9.1.11 for Byron Nuclear Power Station, Units 1 & 2 (Byron) and Braidwood Nuclear Power Station, Units 1 & 2 (Braidwood) to incorporate the requirements of 10 CFR 50.55a(b)(2)(vi) and 10 CFR 50.55a(b)(2)(ix), which address the rules for containment reinforced concrete and unbonded post-tensioning systems inservice examinations required by the Staff.

B. NO SIGNIFICANT HAZARDS ANALYSIS

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes revise the surveillance requirements for containment reinforced concrete and unbonded post-tensioning systems inservice examinations as required by 10 CFR 50.55a(b)(2)(vi) and 10 CFR 50.55a(b)(2)(ix). The revised requirements affect the inservice inspection program designed to detect structural degradation of the containment reinfor:ed concrete and unbonded post-tensioning systems program and do not affect the function of the containment reinforced concrete and the unbonded post-

tensioning system components. The reinforced concrete and the unbonded post-4 tensioning rystem are passive components whose failure modes could not act as accident initiators or precursors.

The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involvt a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident initiator or precursor or malfunction mechanism. The l l proposed changes provide an NRC-approved ASME Code inspection / testing methodology to assure age-related degradation of the containment structure will not go undetected. The function of the containment reinforced concrete and the unbonded post-i tensioning system components are not altered by this change. Additionally, there is no I

change in the types or increase in the amounts of any effluent that may be released offsite; l and there is no increase in individual or cumulative occupational radiation exposure.

Therefore, the possibility of a new or different kind of accident from any previously evaluated has not been created.

3. The proposed change does not involve a significant reduction in a margin of safety.

The proposed changes revise the surveillance requirements for containment reinforced concrete and unbonded post-tensioning systems inservice examinations and tests contained in the referenced TS as required by 10 CFR 50.55a(b)(2)(vi) and 10 CFR 50.55a(b)(2)(ix).

The proposed changes do not affect the ability of containment to mitigate design basis accidents, and, therefore, do not result in a reduction in the margin of safety.

Based on the abeve evaluation, Comed has concluded that these changes involve no significant hazards considerations.

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ATTACHMENT D e

i ENVIRONMENTAL ASSESSMENT FOR i PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF ,

FACILITY OPERATING LICENSES I NPF-37, NPF-66, NPF-72, AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed License Amendment Request against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with Title 10 to the Code of Federal Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has determined that this proposed License Amendment Request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based upon the following

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1. The proposed licensing action involves the issuance of an amendment to a '

license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement;

2. This proposed License Amendment Request involves no significant hazards considerations as demonstrated in Attachment C; I
3. There is no significant change in the types or significant increase in the l

amounts of any efiluent that may be released offsite; and l l

4. There is no significant increase in individual or cumulative occupational radiation exposure.

Therefore, pursuant to 10 CFR 51.22(b), neither an environmental impact statement nor an environmental assessment is necessary for this proposed License Amendment Request.

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