ML20095D586
| ML20095D586 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/08/1995 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20095D582 | List: |
| References | |
| NUDOCS 9512130166 | |
| Download: ML20095D586 (38) | |
Text
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l 1 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY l i LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. ! APPLICABILITY: MODES 1, 2, 3 and 4. l ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT. INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. l SURVEILLANCE REOUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations
- not R16 capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves R207 secured in their positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.
- b. By verifying that each containment air lock is in compliance with the R134 requirements of Specification 3.6.1.3.
- c. Perform required visual examinations and leakage rate testingh__ gigo in accordancen. with
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*Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or containment or the main steam valve vaults and are locked, sealed or othenvise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. June 13, 1995 M 6-1 Amendment No. 12, 130, 176, 191, SEQUOYAH - UNIT 1 9512130166 951208 203 PDR ADOCK 05000327 P PDR _ _ _ . _
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Secondary Containment bypass leakage rates shall be limited to a R207 combined bypass leakage rate of less than or equal to 0.25 L for all penetrationsthataresecondarycontainmentBYPASSLEAKAGEPATHSTOTHE l AUXILIARY BUILDING when pressurized to P,.
- R180 APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION: , With the combined bypass leakage rate exceeding 0.25 L for BYPASS LEAKAGE PATHSTOTHEAUXILIARYBUILDING,restorethecombinedbypassleakageratefrom BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING to less then or equal to 0.25 L, within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. O 13cTsca of LCo 3 lo . l . I , Pr. .c3 d, +..nmant ~ den En te.r t h e. exceed, de,
$cconclary d>,d a.'w enk h p 4ea Eay rc5dh .n oaca\\ ac%n ~ck Ic&g a553 c. ra+c. ecep%c. cc. L.a. .
June 13, 1995 3/4 6-2 Amendment No. 12, 71, 176, 203 SEQUOYMi - UNIT 1
n accorbance toS & DEammenY CONTAINMENT SYSTEMS f7 4 R180 SECONDARY CONTAINMENT BYPASS LEAKAGE l SURVEILLANCE RE0VIREMENTS R180 4.6.1.2 .he secondary co ainment bypass leakage rates shall be demonstrated:
- a. The combined byp ss leakage rate to the auxiliary building shall be determined to be les_s tha_n or eaual to 0.25 L by applicable Type B and C tests . ~1 ast c g g mo6% 3)excepI. for penetrations which are not indivi ua y testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with ,
soap bubbles while the cantainment is pressurized to P, (12 psig) I during each Type A test.W
- b. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, 4 Section III.C.3, when determining the combined leakage rate provided
! the seal system and valves are pressurized to at least 1.10 P (13.2 l psig) and the seal system capacity is adequate to maintain sy, stem l pressure (or fluid head for the containment spray system and RHR l spray system valves at penetrations 48A, 488, 49A and 498) for at least 30 days.
- c. The provisions of Specification 4.0.2 are not applicable.
l - I f*Resitssh 1 be e aluated gainst t accept nce crit ria of pecifi a-l ti n 4.6. 1.c in ccorda e with 10 CFR 50, Appendix J, as m dified y ved xempti s. i l SEQUOYAH - UNIT 1 3/4 6-3 Amendment No. 12,71,101,102, 127,130, 176 February 10, 1994
l l 4 CONTAINMENT SYSTEMS l CONTAINMENT AIR LOCKS d i LIMITING CONDITION FOR OPERATION 1 3.6.1.3 Each containment air lock shall be OPERAB
@ th doors, closed except when the air lock is being used for normal u
transit entry and exit through the containment, then at least one i gr lock door shall be closed,@ _ l
- b. M cver:11 cir 10:k ?::k:;; rate of ler: than er egal t: 0.05 L, :t P,,12 p:%
APPLICABILITY: 'H0 DES l', 2, 3 and 4. i 1 ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed.
- 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days. l l
- 3. Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours. !
l
- 4. The provisions of Specification 3.0.4 are not applicable,
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours.
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a,'r loc k lea hy a. resdts m MAR 251982 1e y e. cafe. aec,phnce er.+cna.. 3/4 6-7 Amendment No. 12 SEQUOYAH - UNIT 1 March 25, 1982 m
4:e \ e e des in a ccordance_. e3 er d N daYaM"'snk Leks. eye Test program CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS CA; I 4.6.1.3 [ ach containme air lock shall be demonstrated OPERABLE:
- a. b/A er acho/ening,e ept whe the air ock is eing us for lti le en ies, the at lea once pe 72 hou s, by v ifying eal eak ge les .than or equal to 0.01 L, i dete ned by recisio flow 52 mea remen s when m sured f r at lea . two m utes wi h the v lume be een.t e door se is at a ressure reater han or qual to 6 psig, M b. B condu ting an vera11 r lock 1 akage st at t less han P, R180 2 psi ) and by verifyi the av all air lock i kage ra e is /
I ithin the limi1 of Spec ficatio 3.6.1.3 b and e resul s evalu ed in ac rdance with 10 C R 50, A endix J as mod fied by approve j exemp ions:# f
- 1. At leas once p six mo ths, an
- 2. Prior o estab ishing NTAINM T INTE ITY if pened w n CONTA MENT I EGRITY as not equire when m ntenanc has been erform d on th air loc that uld aff ct the a r lock
( seal ng cap 111ty.* J b [. At least once per 6 months by verifying that only one door in each air lock can be opened at a time. l
#T provi ions o Speci icatio 4.0.2 e not pplic le.
- xempti n to A pendix "J" of 0 CFR 5 .
SEQUOYAH - UNIT 1 3/4 6-8 Amendment.No. 48, 176 February 10, 1994
I
/ CONTAINMENT SYSTEMS 1
I CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F. SURVEILLANCE RE0VIREMENTS . 4.6.1.6 The structural integrity of the containment vessel shall be determined a180 during 4he shutdown for ::ch Type ^ cent:i tent le:k:g r:t t :t (Sp ific:- tier 0.5.1.1. ) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performe prier t
, the Typ: A : nt: intent le:k:g: r:t t :t to verify no apparent c ges in
? appearance of the surfaces or other abnormal degradation. abnormal j degradation of the containment vessel detected during t ove required inspections shall be reported to the Commission purs nt to Specifica- R40 tion 6.6.1. - l jn ace, educe _ wM N " ** Les 3e &h. Tes+ P rog* ~' l l l I I 1 SEQUOYAH ' UNIT 1 3/4 6-11 Amendment No. 36,176 February 10, 1994 l
. s CONTAINMENT SYSTEMS C'ONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be closed. Operation with purge supply or exhaust isolation valves open for either purging R22 or venting shall be limited to less than or equal to 1000 hours per 365 days. The 365 day cumulative time period will begin every JanJary 1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: R124
- a. With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment purge system lines open, close the isolation valve (s) in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With a containment purge supply and/or exhaust isolation valve R124 having a measured leakage rate in excess of 0.05 L,, restore the inoperable valve to OPERABLE status within 24 hours, otherwise be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.6.1.9.1 The position of the containment purge supply and exhaust isolation valva shall be determined at least once per 31 days. 4.6.1.9.2 The cumulative time that tne purge supply and exhaust isolation valves are open over a 365 day period shall be determined at least once per 7 days. R180 4.6.1.9.3 At least once per 3 months, each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L,.* g fe, L o4cTrora d LCO ,3.int.. I . I , Pr.n a c eG,h.n-ent " iAen overa.Il confa.n ent re sults e x c eecle'n3 furge va lve. l es %3e.
\ ea ug rate a ccephoce. cOer;a_ .
- uit sh 1b ev uatg8agai/sttjeacc/ptan/ecrite/ia Spe fic/-
ion .6. 1.c n cordfncewithIgCFR$0,Ag#endix/,a modif ed l(y apor ved xempion.f 3/4 6-15 Amendment No. 18, 120, 176 SEQUOYAH - UNIT l' February 10, 1994
!., - .l . CONTAINMENT SYSTEMS i 3/4.6.3 CONTAINMENT ISOLATION VALVES 4 LIMITING CONDITION FOR OPERATION . 4 4 . R207 3.6.3 Each containment isolation valve shall, be OPERABLE.* l APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: R207 4
- a. With one or more of the isolation valve (s), except containment vacuum relief isolation valve (s), inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
R201
- 1. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- 2. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
- 3. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- 4. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
R207
- b. With one or more containment vacuum relief isolation valve (s)' inoperable, I the valve (s) must be returned to OPERABLE status within 72 hours, or be in at'least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0'.4 do not apply. Rj07 SURVEILLANCE REQUIREMENTS R207 4.6.3.1 Deleted l
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R207 f.
- enetration flow path (s) may be unisolated intermittently under administrative controls. June 13, 1995 SEQUOYAH - UNIT 1 3/4 6-17 Amendment No. 12 ,
197, 203
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l - 3/4.6 CONTAINMENT SYSTEMS- O f-
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3-i BASES 4 3/4.6.1 PRIMARY CONTAINMENT ! R180 The safety design basis for p imary containment is that the containment must withstand the ?essures and emperatures of the limiting design basis accident (DBA) witho, exceedin the design leakage rates. ! The DBAs that result in a challenge to containment OPERABILITY from high . pressures and temperatures ar a loss of coolant accident (LOCA), a steam line l break, and a rod ejection acc dent (REA). In addition, release of significant ! fission product radioactivit within containment can occur from a LOCA or REA.
; In the DBA analyses, it is sumed that the containment is OPERABLE such that, ; for the DBAs involving rele se of fission product radioactivity, release to the l environment is controlled y the rate of containment leakage. This leakage i rate limitation will limit the site boundary radiation doses to within the I' limits of 10 CFR 100 duri g accident conditions. The containment was designed with an allowable leakage rate of 0.25 percent of containment air weight per day. This leakage rate, sed in the evaluation of offsite doses resulting from accidents, is defined in 10 CF" 50, App;;di J, as L : the maximum allowable containment leakage rate at the calculated peak cont,ainment internal pressure (P resulting from the limiting DBA. The allowable leakage rate represented by,)L, forms the basis for the accontance criteria-i-nae =d on all containment leakage rate testing.f L is ss d to be 0 5p can per ay i the safe y}
ly sa P = 12. p g. s an adde con rvat se, he asur d ov all l e,a ge at is urt r li ited o le s t n or equa to .75 i t at , ur gp fo nce of the eri ic t sts acc unt or p ssib e de ada on f (th con inne t 1 k ge rrie s be ween ests J Primary containment INTEGRITY or operability is maintained by limiting leakagetowithintheacceptancecriteriaofg0CF"50,A;;;;di:J. ndivi ual 1 kage ates spe ified r the ontain nt air ock (L
.6. 3), urge y lves C0 3.6. 9) and second y bypa s leak e (LC0 .6.1. )
re at s cific ly pa t of the accept ce cri eria of 10 CFR 50, App ndix . The efore leaka e rat exceed ng thes indiv ual li its do ot rest t in he pr ary ntain nt be ng inope able ur ess th leaka , when combine with o er Ty.e B an C te leakag s, exceq.ds the ccept ce cri ria of /Appen x l 3/4.6.1.2 SECONDARY CONTAINMENT BYPASS LEAKAGE The safety design basis for containment leakage assumes that 75 percent of the leakage from the primary containment enters the shield building annulus for i filtration by the emergency gas treatment system. The remaining 25 percent of l the primary containment leakage, which is considered to be bypassed to the auxiliary building, is assumed to exhaust directly to the atmosphere without filtration during the first 5 minutes of the accident. After 5 minutes, any bypass leakage to the auxiliary building is filtered by the auxiliary building gas treatment system. A tabulation of potential secondary containment bypass February 10, 1994 SEQUOYAH - UNIT 1 B 3/4 6-1 Amendment No. 102, 127, 176
g & n,,,mb Laa k'a3e. Me Te{ ko rm 3/4.6 CONTAINMENT SYSTEMS BASES leakage paths to the auxiliary building is provided in"-plar.t procedurer. Restricting the leakage through the bypass leakage paths to 0.25 L provides , assurance that the leakage fraction assumptions used in the evaluation of site R180 l boundary radiation doses remain valid. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provide assurance that i the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests'. 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig and 2) the BR containment peak pressure does not exceed the maximum allowable internal pressure of 12 psig during LOCA conditions. 3/4.6.1.5 AIR TEMPERATURE l The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the maximum allowable internal pressure during LOCA conditions and gg
- 2) the ambient air temperature does not exceed that temperature allowable for ,
the continuous duty rating specified for equipment and instrumentation located within containment. The containment pressure transient is sensitive to the initially contained l air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limits of 100*F for the lower compartment, 85'F for the upper compartment, and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper sa temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and ' anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that sa the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA. @ visual inspectior@in confunction with Type-A4eakage-testsr-4s y suffi ient to demonstrate thi capability. arc. Q. acco cloce. O N 0" ' ' wf
~ h Rafe._ Te s June 13, 1995 a SEQUOYAH - UNIT 1 B 3/4 6-2 Amendment No. 102, 127, 176, 203
ADMINISTRATIVE CONTROLS , i1
- ~
- 6) . Limitations on the operability and use of the liquid and gasecus effluent. treatment systems to ensure that the appropriate portions. R152 i .of these systems are used to reduce releases of radioactivity when. ,
the projected doses in a 31-day period would exceed 2 percent of the { guidelines for the annual dose or dose commitment conforming to . Appendix I.to.10 CFR Part 50,
- 7) Limitations.on the dose rate resulting from radioactive material
- released in gaseous effluents from the site to areas at or beyond the. SITE BOUNDARY SHALL BE LIMITED to the following
- R178 L
! 1.: For noble gases: Less than or equal to a dose rate of .i 500 mrem /yr to the total body and less than or. equal to a dose rate of 3000 mrem /yr to the skin, and
- 2. For. Iodine-131, Iodine-133, tritium, and for all radionuclides in . j particulate form with half-lives greater than 8 days: .Less than- 1 or equal to a dose rate of 1500 mrem / year _to any organ. I 8). Limitations on the annual and quarterly air doses resulting from noble gases released in' gaseous effluents from each unit to areas R152-beyond the SITE BOUNDARY conforming to Appendix ! to 10 CFR Part 50,
- 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days -in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and
- 10) Limitations on the annual dose or dose commitment to any MEMBER OF
.THE PUBLIC due-to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
- g. Radioloaical Environmental Monitorina Proaram A program shall be provided to monitor the radiation and radionuclides in the environs of-the plant. The program shall provide (1) repre-sentative measurements of radioactivity in the highest potential expo-sure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling-of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
- 1) Monitoring, sampling, analysis, and reporting of radiation.and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,
- 2) A Land Use Census to ensure t W changes in the use of areas at and beyond the SITE B0UNDARY are 9,1fied and that modifications to the monitoring program are made a r%uired 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 measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
,ZHSERT l A -
SEQUOYAH - UNIT 1 6-18 Amendment Nos. 12, 32, 58, 74, 148, 174 December 9, 1993 l
Insert A
~ h. Containment Leakaae Rate Testina Proaram A progrem shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. Visual examination and testing, including ,
test intervals and extensions, shall be in eccordance with Regulatory Guide l (RG) 1.163, " Performance-Bad Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing instructions. l The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 12.0 psig. , The maximum allowable containment leakage rate, L,, at P., is 0.25% of the , primary containment air weight per day. ) L Leakage rate acceptance criteria are: ) i t a. Containment overall leakage rate acceptance criterion is i 1.0 L,. During j the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 10.60 L, for the combined Type B I and Type C tests, and 10.75 L, for Type A tests; l
- b. Air lock testing acceptance criteria are: l
- 1) Overall air lock leakage rate is 10.05 L, when tested at A P,.
- 2) For each door, leakage rate is 10.01 L, when pressurized to 16 l psig for at least two minutes. '
The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. l The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing ' Program. j i
.. . - . . - - . - - - . ~ . . - - - . . . . . - . - - .--..... ... _ ... . - . - - . _ _ -
3 /4.'6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY' CONTAINMENT i' I CONTAINMENT' INTEGRITY 4 1 f LIMITING CONDITION FOR OPERATION i
- 3.6.1.1 ~ Primary CONTAINMENT INTEGRITY _ shall be maintained.
1
- APPLICABILITY: MODES 1, 2, 3 and 4.
4 ACTION:
-Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within ,
.one hour or be in at least HOT STAND 8Y within the next 6 hours and in COLD
' ' SHUTDOWN within the following 30 hours. l I SURVEILLANCE REQUIREMENTS J 4.6.1.1' Primary CONTAINMENT INTEGRITY shall be demonstrated: .; 1
- a. At least once per 31 days by verifying that all penetrations
- not J capable of being closed by OPERABLE containment automatic isolation i i
valves and required to be closed during accident conditions are
- closed by valves, blind flanges, or deactivated automatic valves gt93 4
secured in their positions, except for valves that are open under - administrative control as permitted by Specification 3.6.3.
- b. By verifying that each containment air. lock is-in compliance with a117 the requirements of Specification 3.6.1.3.
~
- c. Perform required visual examinations and leakage rate testingh:_
in accordance with 10 CF", 50, ".;;;;di: ', :: r dified by : ;rr R167
- ;ti ::. The :I:1- :11=21: 1ede;; r:te, L ,d -
is 0.25% ef
- tti .:::t
- t:tz::t :ir =i;;5t per d:y :t th: ci cul:t-d ;
r...... o, v.,.
. . t fl,e. d.h.need' La e. 'A '
*f* *
*Except valves, blind flanges, and deactivated automatic valves which are R183 located inside the annulus or containment or the main steam valve vaults and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification ;
need not be performed more often than once per 92 days. , June 13, 1995 3/4 6-1 Amendment No. 117, 167, 183, 193 !
.SEQUOYAH - UNIT 2
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Secondary Containment bypass leakage rates shall be limited to a combined bypass leakage rate of less than or equal to 0.25 L for all penetrations that are secondary containmen BYPASSLEAKAGEPATHSTOTHE g3)3 AUXILIARY BUILDING when pressurized to P,. APPLICABILITY: MODES 1, 2, 3 and 4. R167 ACTION: With the combined bypass leakage rate exceeding 0.25 L for BYPASS LEAKAGE PATHSTOTHEAUXILIARYBUILDING,restorethecombinedbypassleakageratefrom BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING to less than or equal to 0.25 L, within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. O 4l,e A C7Tod of LCD 3.6.I. I, " Pr.n ar 9 G4_,,h " wl,,, En+e. r re s k .., aca ad,:,3 8 Pass Le sy sec..,scy ad ..~ak Icr f e rde accef a.,c e cr.$cdo_.
&ba .n.~, e s the o vera ll June 13, 1995 .
SEQUOYAH - UNIT 2 3/4 6-2 Amendment N,:. 63, 167, 193
. .s
** "'" f CONTAINMENT SYSTEMS Lea 6 c. 6Ya 'Ie d f re re wi R167
(]s SECONDARY CONTAINMENT BYPASS LEAKAGE SURVEILLANCE RE0VIREMENTS i R167 4.6.1.2 The secondary containment bypas leakage rates shall be demonstrated:
- a. The combined bypass leakage rate to the auxiliary building shall be l determined to be less than r equal to 0.25 L l and C tests (Jt7TTr'asr ese yer 24 go lis except. by applicable Type Bfor penetratio !
are not individually testa)1e; penetra ions not individually testable
~
shall be determined to have no detectable leakage when tested with soap bubbles while the cgtainment is pressurized to P , (12 psig) during each Type A test.w
- b. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P !
(13.2 psig) and the seal system capacity is adequate to maintitin ! system pressure (or fluid head for the containment . spray system and RHR spray system valves at penetrations 48A, 488, 49A and 49B) for at least 30 days.
- c. The provisions of Specification 4.0.2 are not applicable. I O
l
*
- sults shall be ev uated gainst the ac ptancecriteraofS!e fic -
tion 4 6.1.1 e in a ordan with CFR 5 , Appen ix J, modif[ed by prov d exemp lons. 139, SEQUOYAH - UNIT 2 3/4 6-3 Amendment No. 63,90,104,117,126, 167 February 10, 1994
.- - . . - . - - . - - - _ . - . . . . - ~ . . ~ . - . - - - . . _ - ~ . . . - . - . . - - . -
1-L CONTAINMENT SYSTEMS ,f,- j'* CONTAINMENT AIR LOCKS I l. LIMITING CONDITION FOR OPERATION l
\
l % '
}
- 3. 6.1. 3 Each containmw air lock shall be OPERABLE wp
.@ goth doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one - .
air lock door shall be closed, @ f
- b. AT. L;7;ll d T l^ % 1:0% ^^ 7 t" ^f l^:: th :T ^^"Il t^ C. 05
'.a :t-P,,12 pd;. !
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to CPERABLE status within 24 hours or lock the OPERABLE air-lock door closed. l Cd, ~ 2. Operation may then continue'until performance of the next !
required overall air lock leakage test provided that the OPERABLE i air lock door is verified to be locked closed at least once per ' i 31 days.
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- 4. 'The provisions of Specification 3.0.4 are not applicable,
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
In Va l* =bk <. '$e. pressou s Y 1. An on oferble dr lack N.cor- cIot S n* te, tr e. +,s4 succass &I p< -F.cn..,e a. J aa. ovaedI a.c 1.c K y Aersea I Lcc z. s. u . u . ,
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SEQUOYAH - UNIT 2 3/4 6-7 j t
~
lab raYe 5 in * **r * ^c e-yy g)g k,f ,,$L,b+e_ Le & y. EN E 5 f f CONTAINMENT SYSTEMS SURVEILtANCE REQUIREMENTS {'- 4.6.1.3 ach containmen ir lock shall be demonstrated OPERABLE: a."/ Af er e h openi g, exte when the ir lock is bein used fo ltipi entrie , then a least onc per 72 ours, b verifyi g seal eaka less t an or eq 1 to 0.01 L, as de rmined y precis on flow meas ements hen measu ed for at east tw minute with the volume R40 bet een the oor seals at a pres re great r than r equal t 6 psig f b. conduct ng an ove all air 1 k leakag test a not less han P, R1 12 psig) and by ve fying the overall r lock eakage ra e is within t e limit of Specifica ion 3.6.1.3.b an the resul s evalu ed in acco ance with 10 CFR 50 Appendix J, as m dified by pprove exempt ns:#
- 1. t least o e per six months, a d
- 2. Prior to stablishi CONTAIN NT INTE RITY if ope ed whe CONTAINM T INTEGRI Y was not require when maint nance h s been per ormed on t e air lo that c uld affect he air ock sealing capability /*
- g. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
/
he rovi ions' [ Spe ificat on 4. 2 ar not ppli able.
*Ex.mptio to pend "J" f 10 R 50
~
February 10, 1994 SEQUOYAH - UNIT 2 3/4 6-8 Amendment No. 40,167
CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONUlfiON FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined R167 during-the shutdown for -=ch Type a cantai~aent le= M a -ate test (Specific:- tien '.S.I.1.c) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prier M the Typ: ?. Ocnt:irment le:k:g r:t test to verify no apparent c ges in appearance of the surfaces or other abnormal degradation. Any normal degradation of the containment vessel detected during the ab e required inspections shall be reported to the Commission pursuant to Specifica- R28 tion 6.6.1.
*^
jn etccoede c d. ae L+ Tes+ Poy~ l Amendment No. 28,167 l SEQUOYAH - UNIT 2 3/4 6-11 February 10, 1994
CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be clored. ! i Operation with purge supply or exhaust isolation valves open for either purging 39 l or venting shall be limited to less than or equal to 1000 hours per 365 days. The 365 day cumulative time period will begin every January 1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With a purge supply or exhaust isolation valve open in excess of the giog above cumulative limit, or with more than one pair of containment purge system lines open, close the isolation valve (s) in the purge !
line(s) within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With a containment purge supply and/or exhaust isolation valve gio9 having a measured leakage rate in excess of 0.05 L., restore the inoperable valve to OPERABLE status within 24 hours, otherwise be in :
l at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN gm within the following 30 hours. l 3
') l 4
SURVEILLANCE RE0VIREMENTS 4.6.1.9.1 The position of the containment purge supply and exhaust isolation valves shall be determined at least once per 31 days. l 4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation g9 valves are open over a 365 day period shall be determined at least once per 7 days. , R167 4.6.1.9.3 At least once per 3 months, each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L,.*
- r. ~ a r Cod e.k ~ A e Se. ACTuod E Lco 3. 6. I. I y e o verall cbn$ 4 d
- e. vh.lve. l eah e. reso lf s in exceed.ny fue g %ce. crber.i Iche ra$a. acc.ef O ysu s all be alualed a ains the ccep nce rite ia of Spec fic
( tio 4.6 1.1. in ccorpance with 0 CF 50, ppen ix J, as m ifie by ppro ed ge pti s. Amendment No. 9, 109, 167 SEQUOYAH - UNIT 2 3/4 6-15 February 10, 1994 L
f CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES I i LIMITING CONDITION FOR OPERATION l R193 i 3.6.3 Each containment isolation valve shall be OPERABLE
- l I
i i APPLICABILITY: MODES 1, 2, 3 and 4. i ACTION: 1 R193 . I
- a. With one or more of the isolation valve (s), except containment vacuum !
relief isolation valve (s), inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either: R188 I. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or ! 2. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or ) 3. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or i i
- 4. Be in at least HOT STANDBY within the next 6 hours and in COLD
- SHUTDOWN within the following 30 hours.
R193 ,
- b. With one or more containment vacuum relief isolation valve (s) inoperable, l !
the valve (s) must be returned to OPERABLE status within 72 hours, or be in l i at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within R188 the following 30 hours. R193
- c. The provisions of Specification 3.0.4 do not apply. l SURVEILLANCE RE0VIREMENTS R193 4.6.3.1 Deleted l
Edar fk AC Tiou ef L.co 3.4,.I.I Ch. -ob "
+ aa+..- e Ia Ie. Vl ePr.~<ht+g I2 .,
w 4a, as: ra s 0>e. overalI l%Ya.'nnd~ le< e_ rhe_ accef h nc <
=
ex.cead.b.n cr:tei. _ -- R193 enetration flow path (s) may be unisolated intermittently under administrative l June 13, 1995 SEQUOYAH - UNIT 2 3/4 6-17 Amendment No.193
i
- C E-
- e. -c 3/4.6 CONTAINMENT SYSTEMS *J 4, 7;s[ fro ra m BASES 3/4.6.1 PRIMARY CONTAINMENT R167 The safety design basis for pr mary containment is that the containment must withstand the pressures and eratures of the limiting design basis-accident (DBA) without exceeding he design leakage rates.
The DBAs that result in a hallenge to containment OPERABILITY from high
. pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection acc dent (REA). In' addition, release of significant fission product radioactivit within containment can occur from a LOCA or REA.
In the _ DBA analyses, it is sumed that the containment is OPERABLE such that, for the DBAs involving rele se of fission product radioactivity, release to the environment is controlled y the rate of containment leakage. This leakage rate limitation will limit the site boundary radiation doses to within the limits of 10 CFR 100 duri g accident conditions. The containment was designed with an allowable leakage rate of 0.25 percent of containment air weight per day. This leakage rate, sed in the evaluation of offsite doses resulting from accidents, is defined in .0 CFR 50, ?;;=di J, as L : the maximum allowable containment leakage rate at the calculated peak cont,ainment internal pressure (P resulting from the limiting DBA. The allowable leakage rate represented by,)L, forms the basis for the acceptance criteria imoosed on all containment per day n es ety
~
leakage rate testina. fL is a s to 0 5p ce (Anal sis at = 2.0 ps .A an dded cons rvat se, the as ed vera 1
/in gra 1 ka e r te sf the li ted o? s t an o eq 1t 0.7 L d ing erf a f ep iod to ts t acc unt for ossi le ra ati,o o ec tai nt le ag bar ier bet en ests Primary containment INTEGRITY or operability is maintained by limiting leakage to within the acceptance criteria ofy0 CFR 50, ?;;=di: J.
In ivid al lea age r es s cifi for t cont innen air 1 k (LC l
- 3. .l. , pu ge va es (L 0 3.6. 9) a d seco dary pass eakage (LC0 .6.1. ) i e no spe fical y par of th acce ance e iteri of I CFR 50, Appe dix 0 here ore, eakag rates excee ng th se ind idual limit do no resu in he ;
ri y co ainee t bei inop rable nless he le kage, en c ined with ! othe Type 8 and test eaka s, e eeds t e acc t2nc criter a of l t App dix 3. f 3/4.6.1.2 SECONDARY CONTAINMENT BYPASS LEAKAGE The safety design basis for containment leakage assumes that 75 percent of the leakage from the primary containment enters the shield building annulus for filtration by the emergency gas treatment system. The remaining 25 percent of the primary containment leakage, which is considered to be bypassed to the auxiliary building, is assumed to exhaust directly to the atmosphere without filtration during the first 5 minutes of the accident. After 5 minutes, any bypass leakage to the auxiliary building is filtered by the auxiliary building gas treatment system. A tabulation of potential secondary containment bypass l Amendment No. 91 13g, 167 SEQU0YAH - UNIT 2 8 3/4 6-1 February 10,199d
.. _ f ,,
Q Ghwen$ le L. h*f** 3/4.6 CONTAINMENT SYSTEMS BASES [ 1eakage paths to the auxiliary building is provided inf phr.t precetre . Restricting the leakage through the bypass leakage paths to 0.25 L provides assurance that the leakage fraction assumptions used in the evalualion of site boundary radiation doses remain valid. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. I 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig and 2) the containment peak pressure does not exceed the maximum allowable internal na pressure of.12 psig during LOCA conditions. 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the ( containment air mass is limited to an initial mass sufficiently low to prevent sa A exceeding the maximum allowable internal pressure during LOCA conditions and
- 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.
The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limits of 100*F for the lower compartment, . BR 85'F for the upper compartment, and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and ; anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY . i This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that sa the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA. visual inspectior@in c=j=cti= with-Typc A leakage-tests-is-suff c t to demonstrate thi capability. _ S.a,; manas+ A ey-chd W w - June 13, 1995 SEQUOYAH - UNIT 2 B 3/4 6-2 Amendment No. 91, 139, 167, 193
g ADMINISTRATIVE CONTROLS
- 3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the. precision and accuracy of the measurements of radioactive materials in environmental sample -
matrices are performed as part of the quality assurance program pgggT for environmental monitoring. R134 4
- 6. 9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 lIn addition to the applicable reporting requirements of Title'10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4. R64 STARTUP REPORT
- 6. 9.1.1 A summary report of plant startup and power escalation. testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective 3 actions that were required to obtain satisfactory operation shall also be i 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 completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all 1 three events (i.e., initial criticality, completion of startup test program, j and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. l SEQUOYAH - UNIT 2 6-19 Amendment Nos. 28, 50, 64, 66 134 November $6,1990 h hV 'I F " ' } '
g 4 l ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 l (TVA-SON-TS-95-24) LIST OF AFFECTED PAGES Unit 1 3/4 6-1 3/4 6-2 3/4 6-3 3/4 6-7 3/4 6-8 3/4 6-11 , 3/4 6-15 3/4617 B 3/4 6-1 B 3/4 6-2 6-18 1 Unit 2 3/4 6-1 3/4 6-2 3/4 6-3 3/4 6-7 3/4 6-8 3/4 6-11 3/4 6-15 3/4 6-17 B 3/4 6-1 B 3/4 6-2 l 6-19
<q( .
F Insert A
- h. Containment Leakaos Rate Testina Proaram A program shall be established to implement the leakage rate testing of the
. containment as required by 10 CFR 50.54(c) and 10 CFR 50 Appendix J, Option B,- i~ as modified by approved exemptions. Visual examination and testing, including test intervals and extensions,'shall be in accordance with Regulatory Guide (RG) 1.163, " Performance-Based Containment Leak-Test Program," dated l September 1995 with exceptions provided in the site implementing instructions. 1 The peak calculated containment internal pressure for the design basis loss of coolant accident, P , is 12.0 psig. The maximum allowable containment leakage rate, L., at P , is 0.25% of the primary containment air weight per day. ]. ' Leakage rate acceptance criteria are: I
- a. Containtnent overall leakage rate acceptance criterion is 11.0 L,. During
[ the first unit startup following testing in accordance with this program, the !
- leakage rate acceptance criteria are 10.60 L, for the combined Type B -1
; and Type C tests,'and 10.75 L, for Type A tests; i
- b. Air lock testing acceptance criteria are:
. 1) Overall air lock leakage rate is 10.05 L, when tested at A P,.
j 2) For each door, leakage rate is 10.01 L, when pressurized to A 6 psig for at least two minutes. l' i i The provisions of SR 4.0.2 do not apply to the test frequencies specified in the g , Containment Leakage Rate Testing Program. The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program. 1
):
i I
i ; ENCLOSURE 2 ! i PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE I SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 '
)
(TVA-SON-TS-95-24) , l DESCRIPTION AND JUSTIFICATION FOR INCORPORATING OPTION B TO 10 CFR 50, APPENDIX J l l , l 1
- i l
t l l
- - - . - . - - . - - -- .. .- . - . . - - - . - . . . . ~ . ~ . - . - - - . -
I Description of Chanae
. TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 technical specifications (TSs) to implement the revision to 10 CFR 50, Appendix J. The new Appendix J rule (Option B) provides a voluntary performance based testing option for containment leakage rate testing (CLRT). Option B CLRT requirements are based on system and component performance in lieu of compliance with the current prescriptive requirements. Option B provides flexibility to adopt cost-effective methods, including setting test intervals for implementing the safety cbjectives underlying the requirements of Appendix J. The proposed TS change is as follows:
; General- The proposed change adopts less prescriptive and more performance oriented requirements within TSs. Detailed technical methods for visual examination, containment testing, and test intervals are incorporated into the SON TS by reference to approved industry guidelines (i.e., RG 1.163, Performance-Based Containment Leakage-Test Program).
Tvoe A Test Interval- The proposed chonge implements Option B, which includes an alternative approach to determine surveillance test intervals based on past performance.: An extension of the Type A test interval from three tests in 10 years to one test in 10 years is allowed based on satisfactory performance of two previous tests. However, in accordance with guidance provided in Section C, item 3 of RG 1.163, the visual examination of accessible interior and exterior surfaces of the containment system for structural problems should be conducted prior to initiating a Type A test and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years. Tvoe B and C Test Interval- For Type B and Type C local leakage rate tests, Option B allows licensees to extend the testing frequency on a plant-specific basis based on experience history of each component and established controls to ensure continued performance during the extended testing interval. The Type B test frequency can be extended up to a maximum of once per 120 months. In acc9rdance with guidance provided in Section C, item 2 of RG 1.163, test intervals greater than 60 months for Type C tested components is not presently endorsed by the NRC staff. Further, the interval for Type C tests for containment purge and vent valves can only be extended to once per 30 months. ! Specific changes are described below. I
- 1. TS Surveillance Requirement (SR) 4.6.1.1.c, TS page 3/4 6-1, currently reads:
" Perform required visual examinations and leakage rate testing at P, in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
The maximum allowable leakage rate, L, is 0.25% of containment air weight per day at the calculated peak containment pressure P ,12 psig." I
l 1 - l j The proposed change reads as follows: I
" Perform required visual examinations and leakage rate testing in accordance l with the Containment Leakage Rate Testing Program." !
- 2. A proposed change to TS 3.6.1.2, " Secondary Containment Bypass Leakage," on .l TS page 3/4 6-2 adds the following footnote: .j Enter the ACTION of LCO 3.6.1.1, " Primary Containment," when secondary ,
containment bypass leakage results in exceeding the overe'l containment leakage ; rate acceptance criteria. l
- 3. TS SR 4.6.1.2.c, " Secondary Containment Bypass Leakage," on TS page'3/4 6-3 !
currently contains the following footnote: "Results shall be evaluated against the j acceptance criteria of Specification 4.6.1.1.c in acce' dance with 10 CFR 50, ;
. Appe.edix J, as modified by approved exemptions." The proposed change deletes ;
this footnote. , i 1
- 4. TS LCO 3.6.1.3, " Containment Air Locks," on TS page 3/4 6-7,' currently reads: ,
"Each containment air locks shall be OPERABLE with: !
- a. Both doors closed except when the air lock is being used for normal transit !'
entry and exit through the containment, then at least one air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 L, at P.,12 ,
psig. The proposed change deletes item (b) from the LCO and relocates item (b) to I SON's Containment Leakage Rate Test Program (newly proposed ! Specification 6.8.4.h). Item (a) of the LCO is reformatted to read as follows: ) i "Each containment air lock shall be OPERABLE with both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed." , in addition to the above change, two footnotes are added for clarification of air lock operability. The proposed footnotes read as follows: j i
- a. An inoperable air lock door does not invalidate the previous successful l performance of the overall air lock leakage test.
l
- b. Enter the ACTION of LCO 3.6.1.1, " Primary Containment," when air lock !
leakage results in exceeding the overall containment leakage rate acceptance criteria.
- ..- . . . - . . . . - . - . - - . - --- . . - - . . - . . - - . . . - . - ~ . - - . .
I i
- 5. SR 4.6.1.3, TS page 3/4 6-8, currently reads:
i "Each containment air lock shall be demonstrated OPERABLE:
- a. After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying seal leakage less than or equal to.O.01 L, as determined by precision flow measurement when measured for at least two minutes with the volume between the door seals at a pressure greater than or equal _to 6 psig,
- b. By conducting an overall air lock leakage test at not less than P (12 psig) and by verifying the overall air lock leakage rate is within the limit of Specification 3.6.1.3.b and the results evaluated in accordance with 10 CFR 50, Appendix J, as modified by approved exceptions:#
.1. At least once per six months, and ,
- 2. Prior to establishing CONTAINMENT INTEGRITYif opened when CONTAINMENT INTEGRITY was not required when maintenance has been i performed on the air lock that could affect the air lock sealing capability." i t c. At least onn per 6 months by verifying that only'one door in each air lock can be opene 1 at a time."
#The provisions of Specification 4.0.2 are not applicable.
- Exemption to Appendix "J" of 10 CFR 50.
The proposed change deletes SR 4.6.1.3.(a) and (b) above and relocates the deta:ls associated with these SRs into SON's Containment Leakage Rate Test Program (newly proposed Specification 6.8.4.h). The proposed change also l simplifies SR 4.6.1.3 to read as follows:
"Each containment air lock shall be demonstrated OPERABLE.
I
- a. By verifying leakage rates in accordance with the Containment Leakage Rate i Test Program. ~
- b. At least once per 6 months by verifying that only one door in each air lock I can be operated at a time." l In addition, SR 4.6.1.3(c)is renumbered as Sri 4.6,1.3(b) and the two associated footnotes (# and *) are no longer applicable under ;he proposed change and are deleted. l 9
, ~ - ---
-. ..,.. . ,, m_,.. . ..e. ,_
-- -. .- .- - . - . .- - .. .- . ~ _. . - - - - - - - . -
.4
- 6. SR 4.6.1.6, TS page 3/4 6-11, currently reads:
"4.6.1.6 ' The structural integrity of the containment vessel shall be determined i during the shutdown for each Type A cor'tainment leakage rate test (Specification 4.6.1.1.c) by a visual in',pection of the exposed accessible interior and exterior surfaces of the vessel. - This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.6.1."
The proposed change revises SR 4.6.1.6 to read as follows:
"4.6.1,6 The structuralintegrity of the containment vessel shall be determined during shutdown by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed in accordance .
with the Containment Leakage Rate Test Program to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.6.1."
- 7. TS 3.6.1.9, " Containment Ventilation System," page 3/4 6-15, contains a footnote that reads:
"Results shall be evaluated against the acceptance criteria of Specification 4.6.1.1.c in accordance with 10 CFR 50, Appendix J, as modified '
by approved exemptions." The propose change deletes this footnote and replaces it with the following:
" Enter the ACTION of LCO 3.6.1.1, " Primary Containment," when purge valve leakage results in exceeding the overall containment leakage rate acceptance criteria."
- 8. The proposed change to TS 3.6.3, " Containment Isolation Valves," on TS page 3/4 6-17 adds the following footnote: l I
" Enter the ACTION of LCO 3.6.1.1," Primary Containment," when containment !
isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria." j i
- 9. TVA's proposed TS change includes changes to TS Bases Section 3/4 6.1, i
" Primary Containment," and Bases Section 3/4.6.1.6, " Containment Vessel I StructuralIntegrity." These Bases changes replace current references to 10 CFR 50, Appendix J with the Containment Leakage Rate Test Program. In .
addition, information is deleted to reflect the proposed changes to the TS l sections discussed above. l
l I l I , l
- 10. Administrative Controls, Section 6.8, " Procedures and Programs," TS page 6-18 is evised r under the proposed change to include requirements of a new program entitled Containment Leakage Rate Test Program (Section 6.8.4, item h). The
. programmatic requirements implement 10 CFR 50, Appendix J, Option B, and requires that visual examination and testing in accordance with RG 1.163,
" Performance-Based Containment Leak-Test Program," dated September 1995.
- Information associated with SON's leakage rate acceptance criteria L, and P, is also provided.
Reason for Chance , TVA is revising TSs to implement the recent revision to 10 CFR 50, Appendix J, l Leakage Rate Testing of Containment of Light Water Cooled Nuclear Power Plants. I Currently, CLRT is performed in accordance with the prescriptive requirements of Option A to 10 CFR 50, Appendix J. Option A specifies containment leak-rate test requirements, including the types of tests required. In addition, for each type of test, Appendix J discusses leakage acceptance criteria, test methodology, frequency of testing, and reporting requirements. The Option A details of Appendix J are currently contained in the SON TSs. NRC amended the regulations to provide an Option B to the existing Appendix J. Option B is a performance based approach to Appendix J leakage testing :
- requirements. This option allows licensees with good performance history to reduce the Type A testing frequency from three tests in 10 years to one test in 10 years. For Type B and Type C tests, Option B allows licensees to reduce testing frequency on a plant specific basis based on experience history of each component, and established controls to ensure continued performance during the extended testing interval.
Additionally, Option B allows utilitics to remove the prescriptive details from the TSs. Therefore, TVA is revising the SON TSs to comply with the performance based J approach provided in the revision to 10 CFR 50, Appendix J. The proposed change is a cost beneficiallicensing action. Approval of this TS-amendment will allow an immediate cost savings of approximately $500,000. This 1 cost savings is associated with eliminating performance of a Type A test during the j upcoming Unit 2 Cycle 7 refueling outage. The long-term cost savings for ! implementing Option 8 to 10 CFR 50, Appendix J is estimated at five million dollars ! over the 25-year plant life for both units. ! Justification for Chanaes The function of SON's steel containment vessel is to isolate and contain fission products released from the reactor coolant system following a design basis accident and to confine the postulated release of radioactive material. The safety design basis
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j for. containment is that it must withstand the pressures and temperatures of the i limiting' design basis accident without exceeding the design leakage rate. Periodic l testing of the leak tightness of containment, as well as individual penetrations and J ' valves, is necessary to assure that the assumed release rate in SON's safety analysis is conservative. l 1 in general, TVA's proposed license amendment revises SON TSs to implement the 1 recently promulgated 10 CFR 50, Appendix J, Option B. Prior to this rulemaking, NRC performed a review of current regulatory requirements in an effort to relax or
- eliminate requirements that are marginal to safety and yet impose significant regulatory burden on licensees. Reactor containment leak testing was identified as an area where NRC determined that a change to the regulations was warranted.
As discussed in the final regulatory impact analysis, for the revised rule, the primary consideration in implementing the performance based leakage rate testing requirements of Appendix J, Option B, is that changes will have at most only a marginalimpact on safety. The results of the present analysis confirm the previous observations of insensitivity of population risks from severe reactor accidents to containment leakage rates. This analysis includes comparisons of the predicted reactor accident risks as a function of containment leakage rate with the NRC's safety goals. The calculated risks are well below the safety goals for all the reactors , considered even at assumed containment leakage rates 100-fold above current requirements. The risk to both the general population and the most exposed members of the public were analyzed. Based on a detailed examination of the results of the Probabilistic Risk Assessments (PRAs) for the five plants evaluated in NUREG-1150 (NRC90), the Technical Support Document (TSD) found that leakage rates as high as 100 times those currently permitted by the licensees' TSs would not increase the containment contribution to risk from severe accidents more than approximately one percent. This - increase is marginal to safety. In addition, a change in the allowable leakage rate is estimated to have a negligible impact on occupational radiation exposure. For Type A tests, specific changes in test frequency are recommended based on risk considerations. For Type B and C tests, analyses indicate the viability of reducing the i frequency of testing. Type A Tests - Reducing the frequency of Type A tests (integrated leak rate tests [lLRTs]) from the current three every 10 years to one every 10 years was found to ; lead to an imperceptible increase in risk. The estimatediincrease in risk is very small ;
. because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A L tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage rate tests j is possible with minimalimpact on public risk.
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- Type B Tests - Reducing the frequency of Type B testing of electrical penetrations [
- - should be possible with marginalimpact on risk, based on findings that leakages through these penetrations are both infrequent and small (on the order of one percent i l of the total allowable leakage rate). As the performance history of Type B electrical !
^ penetrations shows no instances where leakage was more than a small fraction of the s - current allowable leakage rate, changing the frequency of testing to coincide with the i schedule for ILRTs is not estimated to result in any change in public radiation
- . exposure.
4 Type C Tests - The considerable majority of leakage paths are identified by local leak , rate tests (LLRTs) of containment isolation valves (Type C tests). Based on the model of component failure with time, it has been found that performance-based alternatives
- to current LLRT requirements are feasible without significant risk impacts. For Type C l tests, the population risk for a performance-based testing schedule would increase 4
overall accident risk by about 2.2 percent per year. This increase is marginal to 4 satety. I- TVA's proposed change reflects a programn atic approach for implementing the - containment leakage rate requirements within SON TSs. The current TS details i associated with visual examination and testing, test intervals, and containment leakage rate acceptance criteria, are relocated to the administrative control section of SON TSs. Implementation of 10 CFR 50, Appendix J, Option B, is provided in a ~ newly proposed TS Section 6.8.4.h, which is entitled, " Containment Leakage Rate Test (CLRT) Program." 4 4 Implementation of SON's CLRT program will be based on Regulatory Guide l (RG) 1.163, " Performance-Based Containment Leak Test Program," dated j September 1995. RG 1.163 endorses NEl 94-01, Industry Guide Line for implementing Performance-Based Option of 10 CFR 50, Appendix J." NEl 94-01 provides methods acceptable to the NRC staff for complying with the provisions of Option B. SON's newly proposed CLRT program recognizes one exception to RG 1.163 (NEl 94-01). The exception is associated with statements in NEl 94-01 that could lead to misinterpretations associated with Type B and Type C test results and . Containment operability. NEl 94-01 (Section 8.0, page 7 and Section 10.2, page 14) states: "The combined as-found leakage rates determined on a MNPLR (minimum path leakage rate) basis for all penetrations shall be less than O.60 L,at all times when containment integrity is required." This statement could be interpreted as a second leakage rate acceptance criteria of 0.60 L, for containment integrity that is in addition to.the existing overall containment leakage limit of 1.0 L,. The creation of a second
' limitation leads to confusion and creates the potential for misinterpretations with regard to containment operability. TVA takes exception to this criteria since current TS leakage criteria for containment operability is based on a 1.0 L, limit. SON's CLRT
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kh ? b i-J l l program maintains a running total of the overall containment leakage (including the
- Type B and Type C leakage). SON's CLRT program contains provisions for evaluating as-found leakage rates in excess of 0.60 L, on a MXPLR basis when containment integrity is required. The evaluation ensures that SON's overall containment leakage - .)
rate does not exceed 1.0 L, TVA willinclude the exception to RG 1.163 in the site ; implementing instructions. T in a'd dition to the single exception from RG 1.163, TVA has included within the SON TS (CLRT program) a definition for P,, L , and a description of the leakage rate acceptance criteria.1The leakage rate acceptance criteria is outlined as follows: i ArAA Acceptance Criteria
- a. Containment overall leakage rate i 1.0 L.
Combined Type B and Type C tests - 10.60 L. Type A tests i O.75 L,
- b. Air Lock
- 1. Overallleakage rate 0.05 L, when tested at A P,
- 2. Door sealleakage rate 0.01 L, when pressurized to 16 psig for at least two minutes Under TVA's proposed change, the leakage rate acceptance criteria listed above are relocated from the individual LCOs or SRs to SON's administrative control section l (Section 6.8.4.h). This is justified based on the fact that the leakage criteria remains j unchanged and is retained in the TS SRs through references to the CLRT program. ]
. It should be noted that two SON TSs (3.6.1.2 - Secondary Containment Bypass Leakage and 3.6.1.9 - Containment Ventilation System) contain SRs that govern
. leakage limits for SON's secondary containment bypass leakage paths (0.25 L.) and purge valves (0.05 L,). Under TVA's proposed change, the leakage limits and surveillance test requirements are being retained in their current form to ensure that these leakage limits continue to be satisfied within their associated SRs. A minor change is proposed for SR 4.6.1.2.a to remove a 24-month frequency for determining j
. that combined bypass leakage is less than or equal to 0.25 L,. The 24-month
' frequency requirement is considered to be prescriptive and is relocated within the Containment Leakage Rate Test Program. !
-l SON's bypass leakage and purge valve specifications contain a footnote that states: l l
"Results shall be evaluated against the acceptance criteria of Specification 4.6.1.1.c l in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions."
l
&i N This footnote provides a means of evaluating leakage test results from bypass leakage paths and purge valves to ensure the overall containment leakage rate limit of 1.0 L,.
is satisfied. TVA's proposed change deletes the current footnote language and replaces it with the following:
- 1. Secondarv Containment Bvoass Leakaae
" Enter the ACTION of LCO 3.6.1.1, " Primary Containment,".wher Secondary Containment Bypass Leakage results in exceeding the overall conteMment leakage
. limit of 1.0 L,."
- 2. Containment Ventilation Svstem (Purae Valve Leakaoe)
" Enter the' ACTION of LCO 3.6.1.1, " Primary Containment", when purge valve leakage results in exceeding the overall containment leakage limit of 1.0 L,."
The newly proposed change to these footnotes accomplishes the same goal by ensuring that leakage test results (bypass and purge valves) are evaluated against the overall containment leakage limit of 1.0 L,. The proposed change to these footnotes is consistent with standard TS language (refer to note 4 from LCO 3.6.3 of Revision 1 to NUREG-1431) and establishes a reference to the ACTION requirements of LCO 3.6.1.1 for containment operability (i.e., integrity). In addition to the change described above, similar footnotes have been added to SON's airlock and containment isolation valve specifications (TS 3.6.1.3 and 3.6.3, respectively). This proposed change is consistent with STS requirements (refer to note 3 from LCO 3.6.2 [Airlocks] and note 4 from LCO 3.6.3 [ Containment isolation Valves]) and ensures that containment leakage results from these pathways are considered for overall containment operability. TVA's incorporation of a second footnote within SON LCO 3.6.1.3, Containment Air Locks, states:
"An inoperable air lock does not invalidate the previous successful performance of the overall air lock leakage test."
This proposed footnote is consistent with SR 3.6.1.2.1 of Revision 1 to the BWR-4 Improved Standard Technical Specifications. This note is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a design basis accident. Failure of the air lock interlock mechanism is an example of a condition that would not affect the leak-tight integrity of the doors. Additionally, seal leakage from a single door would not affect the integrity of the second air lock door or invalidate previous overall air lock leakage test results. L L
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Environmental Imoact Evaluation i The proposed change does not involve an unreviewed environmental question because operation of SON Units 1 and 2 in accordance with this change would not: l
- 1. Result in a significant increase in any adverse environmentalimpact previously i evaluated in the Final Environmental Statement (FES) as modified by NRC's )
testimony to the Atomic Safety and Licensing Board, supplements to the FES, l environmentalimpact appraisals, or decisions of the Atomic Safety and Licensing Board.
- 2. Result in a significant change in effluents or power levels.
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3.' Result in matters not previously reviewed in tae licensing basis for SQN that may ! have a significant environmentalimpact.
O O * (1 1 i ENCLOSURE 3 , PROPOSED TECHNICAL SPECIFICATION CHANGE '
~ SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 ,
k (TVA-SON-TS-95-24) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION i I
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1 Significant Hazards Evaluation. l i TVA has evaluated the proposed technical specification (TS) change and has l
' determined that it does not represent a significant hazards consideration based on !
criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not: ;
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed amendment to SON TSs is in accordance with Option B to ; 10 CFR 50, Appendix J. The proposed amendment adds a voluntary performance l based option for containment leak rate testing. The changes being proposed do ! not affect the precursor for any accident or transient analyzed in Chapter 15 of j SON Updated Final Safety Analysis Report. The proposed change does not l increase the total allowable primary containment leakage rate. The proposed ; change does not reflect a revision to the physical design and/or operation of the i plant. Therefore, operation of the facility, in accordance with the proposed
. change, does not significantly affect the probability or consequences of an accident previously evaluated.
. 2. Create the possibility of a new or different kind of accident from any previously analyzed.
The proposed amendment to SON TSs is in accordance with the new performance-based option (Option B) to 10 CFR 50, Appendix J. The changes ! being proposed will not change the physical plant or the modes of operation defined in the facility license. The proposed changes do not increase the total l allowable primary containment leakage rate. The changes do not involve the j addition or modification of equipment, nor do they alter the design or operation of plant systems. Therefore, operation of the facility in accordance with the i proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Involve a significant reduction in a margin of safety.
The proposed change to SON TSs is in accordance with the new option to 10 CFR 50, Appendix J. The proposed option is formulated to adopt performance-based approaches. This option removes the current prescriptive
' details from the TS. The proposed changes do not affect plant safety analyses or change the physical design or operation of the plant. The proposed change does not increase the total allowable primary containment leakage rate. Therefore, operation of the facility, in accordance with the proposed change, does not involve a significant reduction in the margin of safety.
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