ML20217A122
| ML20217A122 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 11/14/1990 |
| From: | Pierson R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20217A129 | List: |
| References | |
| NUDOCS 9011200190 | |
| Download: ML20217A122 (22) | |
Text
_
[ o aseg'eg UNITED STATES
)
1'
<f f NUCt. EAR RE@ULATORY COMMISSION i
wasniwoTow. o. c. ro66s
.... *,5 DETROIT EDISON COMPANY DOCKET NO. 50-341 FERMI-2 AMENDMENT _ TO,[A,C))))],pf E,RA,1),N,L, LJyJNH Amendment No. 62 License No. NPF-43 1.
The Nuclear Regulatory Connission (the Connission) has found that:
A.
The application for amendnent by the Detroit Edison Company (the licensee)datedMarch 26, 1990, complies with the standards end requirernents of the Aton.ic Energy Act of 1954, as ernended (the Act),
and the Conanission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conforn.ity with the application, the provisions of the Act, and the rules and regulations of the Conahsion; C.
Thereisreasonableassurance(i)thattheactivitiesauthorizedby this anendrent r.an be conducted without endangering the health an; safety of the public, and (ii) that such activities will be conducted in compliance with the Con.n.ission's regulations; D.
The issuance of this antndnent will not be inimic61 to the coneon dtit.nse and security or to the health and safety of the public; and E.
The issuance v ' this en'codnent is in tccordance with 10 CFR Part 51 of the Conraission's regulations and all applicable requirenents have been satisfied.
[
l 2.
Accordingly, the license is anended by changes to the Technical Specifica-
-tions as indicated in the attachnent to this license amendaent and paragraph 2.C.(2) of Facility Operating License No. NPF-43 is hereby amended to read as follows:
T e c h n i c,a),,Sp e c i f i c a t_ipp s, a nM n yj,rp npe n,t aj,,P,rp,t e,c,tj p n, Pj a n The Technical Specifications contained in Appendix A, as revised through Amendrtent No. 62, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
DECO shall operate the facility in ectordance with the Technical Specifications and the i
Environnental Protection Plan.
l l
9011200190 901114 PDR ADOCK 05000'341 P
PNV
t
.g.
}
3.
Further, paragraph 2.C.(9) of Facility Operating License No. NPF-43 is hereby amended to read as follows:
1 Modifications for Fire Protection (Section 9.5.1. SSER #5 and SSER #6) l DECO shall implement and maintain in effect all provisions of the approved fire protection program as described in its Final Safety Analysis Report for the facility through Amendment 60 and as approved in the SER through Supplement No. 5, subject to the following provision:
(a)
Deco may make changes to the approved fire protection program witnout prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
Deco shall install and make operational, the independent alternate shutdotn system in accordance with the schedule contained in its letter L
dated July 5,1985.
The interim procedures and measures described in Section 9.5.1 and Appendix E of SSER #5 and #6 shall continue to be implemented, including removal of power from the Division I cooling i
i tower bypass valve (No. E1150-F603A) and from either the single series valve (No. E1150-F008) in the reactor heat removal (RHR) system or the two parallel RHR suction valves (Nos. E1150-F608 and E1150-F009) during normal plant operation until the independent alternate system is declared operational.
4.
This license amendment is effective as of its date of issuance with full implementation within 60 days of issuance.
l l
FOR THE NUCLEAR REGULATORY LOMMISSION b
Ah+ T~
l
\\
~
Robert Pierson, Director Project Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications D
l L
Date of Issuance: November 14, 1990 l
l
+
ATTACHMENT TO LICENSE AMENDMENT NO.6?
FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The reviled pages are identified by Amendment number and contain a vertical line indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT v
v viii viii xiii xiii xv xv xxiv xxiv 3/4 3-67 3/4 3-67 3/4 3-68 3/4 3-68 3/4 3-69 3/4 3-69 3/4 3-70*
3/4 3-70*
3/4 7-23*
3/4 7-23*
3/4 7 7-39 3/4 7-40 3/4 7-24 8 3/4 3-5 8 3/4 3-5 B 3/4 3-6*
B 3/4 3-6*
B 3/4 7-3*
B 3/4 7-3*
B 3/4 7-4 0 3/4 7-4 6-1*
6-1*
6-2 6-2 6-9 6-9 6-10*
6-10*
- 0verleaf page provided to maintain document completeness.
No changes contained in these pages.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION i
3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...........
3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................
3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.....................................
3/4 3-23 3/4.3.4 ATWS RECIRCULATION PUMP TRIP SYSTEM ACTUATION INSTRUMENTATION.....................................
3/4 3-32 3/4.3.5 REACTOR CORE ISOLATION COOLING Si3 TEM ACTUATION l'
INSTRUMENTATION.....................................
3/4 3-36 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION...................
3/4 3-41 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................
3/4 3-47 Seismic honitoring Instrumentation..................
3/4 3-51 Meteo o'ogical Monitoring Instrumentation...........
3/4 3-54 Remote Shutdown System Instrumentation and Controls.
3/4 3-57 Accident Monitoring Instrumentation.................
3/4 3-60 Source Range Monitors...............................
3/4 3-64 Traversing In-Core Probe System.....................
3/4 3-65 Chlorine Detection System...........................
3/4 3-66 0eleted.............................................
3/4 3-67 loose-Part Detection System.........................
3/4 3-70 Radioactive Liquid Effluent Monitoring Instrumentation.....................................
3/4 3-71 Radioactive Gaseous Effluent Monitoring Instrumentation.....................................
3/4 3-76 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM..................
3/4 3-85 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION......................................
3/4 3-86 3/4.3.10 RESERVED..............................................
3/4.3.11 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION......
3/4 3-90 FERMI - UNIT 2 v
Amendment No. 53, 59.62
)
inDEx 1)!111).L.~. S.hPJ.119F S. ! P.R. 011pb,1) pp, epp, Sppy;j !!!!qci,,R,tg u i atME NT S SECTION PAGE CON 1AINMQ1,)}$TEk)(Continued) 3/A.6.6 PRIMARY CONTAlhkEh1 ATMOSPHERE CONTROL Drywell and Suppression Chanber Hydrogen Recorr.biner Systems............................................
3/4 6-57 Drywell ar,d Suppression Chanber Oxygen Concentretier..
3/4 6-58 3/.4 2. p{ ppt SYSTEMS
/
3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Rernovel Service Water Systin............
3/4 7-1 En ergency Equipo,ent Coolir.g Water System.............
3/4 7 3 Enit rgency Equirrtent Service Wat er Sy:.ttn..............
3/4 7 4 Ditstl Cinerator Coolir.g Water System................
3/4 7-5 Ul t in.a t e He a t S i n k...................................
3/4 7-6 3/A.7.2 00k1ROL R0011 EPEkGLhtY FIL1kAT10N SYSTEM.............
3/4 7-8 3/4.7.3 SHORE BARkl E R PRO 1 EC110h.............................
3/4 7-11
]
3/4.7.4 REAC10R CORE 150LA110h COOLING SYSTEM................
3/4 7-14 3/4.7.5 S N V B B E R S.............................................
3 /4 7 -16 i
3/4.7.6 SE AL E D SOUkCE CON 1 AMI NATION.......................... 3/4 7-22 3/4.7.7 DELETED 3/4 7-24 through 3/4 7-37 3/A.7.8 DE L ETED..............................................
3 /4 7 3 8
-3/4.7.9 MA1H TURDINE BYPASS SYS1EM AND N0151tlRE SEPARA10R R E HE ATE R............................................. 3 /4 7 2 4 3/4.7.10 PESERVED..............................................-
3/4.7.11 APPENDlX R AL1ELNATIVE SHUTDOKN AllXILIARY SYSTEMS.... 3/4 7-41 FERl11 - tlNIT 2 viii An enanent flo. 77. 77,62
INDEX BASES SECTION PAGE I
_ INSTRUMENTATION (Continued)
MONITORING INSTRUMENTATION (Continued)
Meteorological Monitoring Instrumentation.......
B 3/4 3-4 Remote Shutdown System Instrumentation and Controls........................................
B 3/4 3-4 Accident Monitoring Instrumentation............
B 3/4 3-4 Source Range Monitors...........................
B 3/4 3-4 Traversing In-Core Probe System.................
B 3/4 3-4 Chlorine Detection System.......................
B 3/4 3-5 Deleted.........................................
B 3/4 3-5 Loose-Part Detection System.....................
B 3/4 3-5 Radioactive Liquid Effluent Monitoring Instrumentation.................................
B 3/4 3-5 Radioactive Gaseous Effluent Monitoring Instrumentation.................................
B 3/4 3-6 3/4.3.8 TURBINE L.ERSPEED PROTECTION SYSTEM.............
B 3/4 3-6 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEMS ACTUATION l
INSTRUMENTATION.................................
B 3/4 3-6 1
3/4.3.10 RESERVED
{
3/4.3.11 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION.................................
B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM l
3/4.4.1 RECIRCULATION SYSTEM.................'...........
B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES............................
B 3/4 4-la 3/4.4.3-REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......................
B 3/4 4-2 Operational Leakage.............................
B 3/4 4-2 3/4.4.4 CHEMISTRY.......................................
B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY...............................
B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.....................
B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES................
B 3/4 4-5 3/4.4.8-STRUCTURAL INTEGRITY............................
B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REM 0 VAL...........................
B 3/4 4-5 FERMI - UNIT 2 xiii Amendment No. 6, 53, 57,62
1
)
l INDEX I
BASES SECTION
_PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS...........................
B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM........
B 3/4 7-1 3/4.7.3 SHORE BARRIER PROTECTION........................
B 3/4 7-1 3/4.7.4 REACTOR C0dE ISOLATION COOLING SYSTEM...........
B 3/4 7-la 3/4.7.5
$NUBBERS........................................
B 3/4 7-2 1
3/4.7.6 SEALED SOURCE CONTAMINATION.....................
B 3/4 7-4 3/4.7.7 DELF*iED.........................................
B 3/4 7-4 3/4.7.8 0ELETED.........................................
B 3/4 7-4 3/4.7.9 MAIN TURBINE BYPASS SYSTEM AND MOISTURE l
SEPARATOR REHEATER..............................
B 3/4 7-5 3/4.7.10 RESERVED i-3/4.7.11 APPENDIX R ALTERNATIVE SHUTDOWN AUXILIARY l
SYSTEM5.........................................
B 3/4 7-5 l
3/4.8 ELECTRICAL POWER SYSTEMS 1
3/4.8.1, 3/4.8.2, and 3/4.B.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS............................
B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.........
B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH.............................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION.................................
B 3/4 9-1 1
3/4.9.3 CONTROL R0D P0SITION...........................
B 3/4 9-1 3/4.9.4 DECAY TIME......................................
B 3/4 9-1 l
3/4.9.5 COMMUNICATIONS..................................
B 3/4 9-1 9
l 3/4.9.6 REFUELING PLATF0RM..............................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L............
B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L.......
B 3/4 9-2 3/4.9.10 CONTROL R00 REM 0 VAL.............................
B 3/4 9-2 l
FERMI - UNIT 2 xv Amendment No. 39, BI, 59,62
INDEX LIST OF TABLES (Continued)
TABM PAGE 3.3.7.9-1 DELETED.........................................
3/4 3-68 3.3.7.11-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INST RUMENTATION................................
3/4 3-72 4.3.7.11-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...... 3/4 3-74 3.3.7.12-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION................................
3/4 3-77 4.3.7.12-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...... 3/4 3-81 3.3.9-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION I NST RUMENT AT I ON................................ 3/4 3-87 3.3.9-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS......................
3/4 3-88 4.3.9.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS...... 3/4 3-89 3.3.11-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION.. 3/4 3-91 4.3.11.1-1 APPENDIX R ALTERNATIVE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS....... 3/4 3-92 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLA VALVES.............................. TION 3/4 4-12 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS..............................
3/4 4-12 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS........
3/4 4-15 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...............................
3/4 4-18 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--
WITHDRAWAL SCHEDULE............................
3/4 4-22 4.6.1.1-1 PRIMARY CONTAINMENT ISOLATION VALVES / FLANGES LOCATED IN LOCKED HIGH RADIATION AREAS.........
3/4 6-lb 3.6.3*1 PRIMARY CONTAINMENT ISOLATION VALVES...........
3/4 6-22 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS....................
3/4 6-53 3.7.3-1 SURVEY POINTS FOR SHORE BARRIER................
3/4 7-12 3.7.7.5-1 DELETED.........................................
3/4 7-32 3.7.7.6-1 DELETED.........................................
3/4 7-37 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE.................
3/4 8-8 FERMI - UNIT 2 xxiv Amendment No. 49, 59,62
4
_INSTRUKEttTAJJ{Lh 3/4.3.7.9 DELETED This page has beeri deleted.
l i
EERM1 - UtilT 2 -
3/43-67
/rendirent No. 62
I P
TABLE 3.3.7.9-1 DELETED t
r i
(
i This page has been deleted.
I l
i 1
1 r
l 1l.
l b
i,--
l l
l FERMI - UNIT 2 3/4 3-68 Amendment No. 62
i
?m in
'~
TABLE 3.3.7.9-1 (continued)
DELETED
.'f-4
-Ii
,; 'e
'- l ;' t.
il':
4 L '
-* 4 l
g m
i i
4
.:{"'
.1 3t.
lThispagehas-beendeleted.
J
[.i 4
k )
-i.
r s
.,\\ 1
< w
^ {:'
,a
>- 4 0 ;,
>i;.
1.;;' y s.g
'.;y
. i
.I.i,'
? 1^
k i
..,.s5
- x. I ' n,.,
.v N- -
FERMl=- UNIT'2=
'.3/4 3-69 Amendment:No. 62 -
i
+
.(
,6 1
2y a
INSTRUMENTATION
_ LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.10 The loose part detection system shall be OPERABLE.
APPLICABIL11Y:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With one or more loose part detection system channels inoperable for a.
more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specificelion 6.9.2 within the next 10 days outlining the cause of the malfunction 6nd the plans for restoring the channel (s) to OPERABLE status, b.
The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.10 Each channel of the loose part detection system stall be demonstrated OPERABLE by performance of a:
a.
CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, h
b.
CHANNEL FUNCTI1NAL TEST at least once per 31 days, and l-t c.
CHANNEL CALIBRATION at least once per 18 months.
l l'
FERMI - UNIT 2 3/4 3-70 J
~
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
Stored sources not in use - Each wealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months.
Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use, Startup sources and fission detectors - Each sealed startup source c.
and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
4.7.6.3 Reports - A report shall be prepared and submitted to the Commission on an annual t asis if sealed source or fission detector leakage tests reveal the presence i.f greater than or equal to 0.005 microcurie of removable contamination w-c 1
a FERMI - UNIT 2 3/4 7-23 l
i
~
PLAT:TSLST{k) 3 /479_. FeJ N, J,0j,BJ pf, BJfpS S, SJ S}{ f,[pp, fp),S]pp{, 5((ppp]pp,. p EpEpJ{p LlH1 TING CONDITION.FOR OPERATION
=
3.7.9 The miain turbir.t bytess systen. and Moisturt Separator Reheater shall be OPERABLE.
APPLICABIL11Y: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or T(ual t'o"t W of RATED THERNAL POWER.
ACTION: With the main turbine bypass systeni and/ot Moisture Separator Retienter irioperable, testore the system to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or-tale the ACTION required by Specificatics 3.2.3.
5FFi!).t!!FFf. F19!8)FIFf.NJS,,,,,,...,,;,,,.,,,,,,,,,,,,,,,,;,,,,,;,.;g,,,,,,,.
d.7.9 TLe Inuite turbir.e byrass systerr shall be dtnctistrated OPERABLE at least once-per:
a.
92 deys and during ebch COLD SHilTDOWN, by cycling toch turbine bypast galve through at least one complete cycle of full travel, and I
b.
18 months by:
I 1.
Performing a systeci functional test which it.cludes simulattd
^
autoniatic ectuation arid verifyitig that each autcmatic valve actuates to its correct position.
2.
Denonstrating 1URBIhE BYPASS SYS1EM RESPONSE TIME to be less there or equal to 300 milliseconds, l'
l t
FERM1 - UNIT 2 3/4 7-24 An<endment No, 79,62 l
t i
INSTRUMENTATION BASES T
MONITORING INSTRUMENTATION (Continued) 3/4.3.7.8 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures that an accidental chlorinc release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel.
Upon detection of a high concentration of chlorine, the control room emergency ven-tilation system will automatically be placed in the chlorine mode of opera-tion to provide the required protection.
The detection system required by this specification is consistent with the recommendations of Regulatory Guide 1.95 " Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release", Revision 1, January, 1977.
3/4.3.7.9 DELETED l
3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-cart detection system ensures that sufficient cepability is available to detect loose metallic parts in the primary system t
an3 avoid or mitigate damage to pri. nary system components.
The allowable I
out of-service times and surveille.nce requirements are consistent with the l
recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for j
the Primary System of Light-Water-Cooled Reactors," May 1981.
l l
3/4.3.7.11 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm / trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the'0DCM to ensure that the alarm /' rip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERAB. ITY and use of this instrumentation is consistent with the requirements of Gene <a1 Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
. FERMI - UNIT 2 B 3/4 3-5 Amendment No. 59, 62
t INST 5TPEp1ATjpk
!!SES MONITOR 1hG,Jp5])pkjNTATJ0)(Continued) 3/A.3.7.12 RAD 10AC11VE GASEOUS EFFLUENT S0kl10R1hG-INSTRURENTATION The radioactive geseous effluer.t nonitoring instrun.entation is provided to nonitor and control, as applicable, the releases of radioactive n.ateriels in gestops effluer.ts during actual or potential releases of gaseous effluer.ts.
The alerr.i/ trip setroints for these instrunients shall be calculated and adjusted in eccordance with the nethodology and paraneters in tht: OCCF utilirir.s the systeni design flow rates as specified in the ODCM. This conservative niethod is used because the Fern.12 design does not include flow rate ineasurerent devices. This will (nsure the alarra/ trip will occur prior to exceeding the lin.its cf 10 CFR Part 20. This instiunientation else includes provisiot.s for nonitoring and conticlling the concentrations of poteritially explosivt gas nixtures it. the noin condenser offges treattitr.t systero. Tit OPERABIL11) a r.c' use cf tiit instrunttitation is consistent with the rec;uitenttts of General Desigi Criteria EC, 63, and 64 of Appendix A tt 10 CFR Part 50.
N!. M.. ) PFf) f!. Py!F.Sf!!P. ff M ! N ) PF. sv 5T E M This spe(ificatiot, is previded ie ensure thet t14 turbis e overspeed protection systtro instruntntation and the turbine speed contrcl velves are OPERABLE ar.d will protect the turbine froni e>ctssive tiverspeed.
Protectiot.
fror. turbit.t excessive oversittd is t>ct rec;uit ed to pt otect saftty-relatec' cor.ponents, ec;uipent, or st ructurt 5.
However, it is ittluded in order to irprove overall plant reliability.
? /4 3 9 '. f!!PF.A.T!M:!) F.1FFN f.E, Jp)p, NM R, f,C,1pM J pp,.j p5) ppp{!gj pp The fcedwater/niain turbita trip systeni ectuation instruti.entation is provided to initiate action of the feedwater systen/nain turbine trip systen.
in tho event of e high reactor vesstl water level c've to failure of the ft;tdwater contt oller order riaxintn denand.
2/f..3:33...AIIf fE) ?. f. f.l.l f !fN ) 7.E;,Spp1 ppp,N, J p,5] RppE,NT AT 10 N 1he OPERABIL11) of the alternative shutdown systtoi ensure that a fire will oct preclude echitving safe shutdown. The altt;inative shutdown systen instrunt;ntation is independent of ateas where a fire could don' age systems norn. ally used to shutdown the reactor. Thus, the systeni capbility is consisterit with General Design Criterion 3 at.d Appendix R to 10 CFR 50.
l l
FERMI - UNIT 2 E 3/4 3-6 Antndnient No. 59
PLANT SYSTEMS
(.
BASES SNUBBERS (Continued) to be unprotected and to result in f ailure during an assumed initiating ew Inspections performed before the interval has elapsed may be used as a new,eter-ence point to determine the next inspection.
However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.
Any inspection whose results require a shorter inspection interval will override the previous schedule.
The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers.
For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing.
To provide assurance of snubber functional reliability one of three functional testing methods is used with the stated acceptance criteria:
1.
Functionally test 10% of a type of snubber with an additional 10%
tested for each functional testing failure, or 2.
Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7.5-1, or 3.
Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
Figure 4.7.5-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.
Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subsequent date.
Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
The service life of a snubber is established via manufacturer input and information through consideration of the snubb r service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.) The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a perfo. nance evaluation in view
.of their age and operating conditions.
These records will provide statistical bases for future consideration of snubber service life.
7 FERMI - UNIT 2 B 3/4 7-3
'ht pLANTSYS1EP}
BASES 3/4.7.6 SEALED SOURCE CONTAMINATION testing, including alpha emitters, is based on 10 CFR 70.39(c) quiring leak The limitations on removable contamination for sources re limits for plutonium. This limitation will ensure that leakagt fron byproduct, source, and special nuclear naterial sources will not exceed allowable intake values.
Sealtd sources are classified into three grou)s according to their use, with surveillance requirements corrensurate with tie probability of damage to a source in that group. Those sources which are frequently handled ere required to be tested more often than those which are not.
Sealed sources which are continuously enclosed within a shielded n.echanism, i.e., sealed sources within radiation n;onitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded nechanism.
}/32,7.7__ DELETED 3/4.7.8_ _,Df}f,T{p l
l-i l
l l
l FERMI - UNIT 2 B 3/4 7 4 Amendment No. 62
- 6. 0 ADMINISTRATXVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit safe operation and shall delegate in writing the succession to this responsibility during nis absence.
The Plant Manager shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
6.1.2 The Nuclear Shift Supervisor or, during his absence from the control room, a designated individual shall be responsible for the control room l
command function.
A management directive to this effect, signed by the Vice President Nuclear Operations shall be reissued to all station personnel on an annual basis.
I 6.2 ORGANIZATION 6.2.1 l
0FFSITE AND ONSITE ORGANIZATION l
Onlite and offsite organizations shall be established for unit operation and corporate management, respectively.
The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
Lines of authority, responsibility, and communication shall be a.
l established and defined for the highest management levels through intermediate levels to and including all orerating organization positions.
These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental 9sponsibilities and relationships, and job descriptions for key persa nel positions, or in equivalent forms cf documentation.
These requirements shall be documented in the l
Updated Final Safety Analysis Report, The Senior Vice President shall have corporate responsibility for b.
overall plant nuclear safety and shal.1 take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support of the plant to ensure n'. clear safety, The individuals who train the operating staff and those who carry c.
out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from opereting pressures.
6.2.2 UNIT STAFF Each on duty shift shall be composed of at least the minimum shift a.
crew composition shown in Table 6.2.2-1; b.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Operator shall be in the control room; FERMI - UNIT 2 6-1 Amendment No. II,30, 54
ADMINISTRAllVE CONTROLS UNITSTAJF(Continued) c.
A Health Physics Technician shall be on site when fuel is ir the reactor. The Health Physics Technician position may be unfilled for j
a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to acconeodete 1
l unexpected absence, provided imediate action is taken to fill the j
required positions; t
d.
All CORE ALTERATIONS shall be observed and directly sunrvised by i
either a licensed Senior 0>erator or licensed Senior Operator l
Limited to Fuel Handling w1o has no other concurrent responsibilities
]
l during this operation; j
l l
e.
DELETED l
f.
Administrative procedures shall be developed and inipleniented to limit the workin flitctions (e.g.,g hours of unit staff who perform safety-relatec' i
licensed Senior Operators, licensed Operators, i
health physics personnel, auxiliary operators, and key maintenance 1
personnel).
Ad(quate shif t coverage shall bt naintained without routine heavy J
use of overtime. The objective shall be to have operatin work 6 nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while the unit is operating. 0 perscr.nol
)
- However, in the event that unforeseen problems require substantial anounts of overtime to be used or during extended periods of shutdown for refueling, n.ajor maintenance, or najor unit nodifications, on a teraporary basit the following guidelines shall be followed:
1.
An individual should not be perniitted to work n: ore than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift tuit.over time.
2.
An individual should not be permitted to work n. ore than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period, all exclud-
[
ing shift turnover time.
3.
A break of at least B hours should be allowed between work periods, includit.$ shift tornover time.
4.
Except during extended shutdown periods, the use of overtime should be considered on an individual basis at.d not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the Plant Manager or a Section Superintendent or higher levels of managen,ent, in accordance with established procedures and with docunientation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly _by the Plant Manager or a Section Super-intendent to assure that excessive hcurs have not been assigned.
i Routine deviation from the above guidelines is not authorized.
1 FERH1 - UNIT 2 6-2 Anendment No. 11. M. M, 6'
k ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) 1.
Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the correc-tive action to prevent recurrence to the Vice President-Nuclear Operations and to the Nuclear Safety Review Group; and Review of charges to the PROCESS CONTROL PROGRAM, the 0FFSITE m.
DOSE CALCULATION MANUAL, and major modifications to the Radwaste Treatment Systems.
n.
Review of the Fire Protection Program.
6.5.1.7 The OSRO shall:
Recommend in writing to the Plant Manager approval or disapproval of a.
items considered under Specification 6.5.1.6a. through d. prior to l
their implementation.
b.
Render determinations in writing to the Nuclear Safety Review Group with regard to whether or not each item considered under Specifica-tion 6.5.1.6a. through f. constitutes an unreviewed safety question, Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President-c.
Nuclear Operations and the Nuclear Safety Review Group of disagree-ment between the OSRO and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagree-ments pursuant to Specification 6.1.1.
RECORDS
- 6. 5.1. 8 The OSR0 shall maintain written minutes of each OSRO meeting that, at a minimum, document the results of all OSRO activities performed under the responsibility provisions of these Technical Specifications.
Copies shall be provided to the Vice President-Nuclear Operations and the Nuclear Safety Review Group.
6.5.2 NUCLEAR SAFETY REVIEW GROUP ChSRG)
FUNCTION-I i.
6.5.2.1 The NSRG shall function to provide independent review and audit of designated activities in the areas of:
I a'.
Nuclear power plant operations, b.
Nuclear engineering, i
c.
Chemistry and radiochemistry, d.
Metallurgy, e.
Instrumentation and control, f.
Radiological controls, g.
Mechanical and electrical engineering, and h.
Quality assurance practices.
The NSRG shall report to and advise the Senior Vice President on those areas of responsibility in Specifications 6.5.2.7 and 6.5.2.8.
FERMI - UNIT 2 6-9 Amendment No. II,M,62
ADHlh1STLAllH 00hTROLS..
COMPOS 1Tjg 6.5.2.? The Senior Vice President shall appoint et least nine n,enbers to the NSF:0 and stell designate fron this n:enibership e Cheiro.an and at least one Vice Chaitten. The n.en.bership 51.611 collectively possess experier.ce end conpetence to trovide independent rtview anc' eudit in the arees listed in Section 6.5.F.1.
The Ctairnen er.d Vice Chairn.an shall have nuclear background in engineering or optotions ar'd stell be cepeble of c'etermining when to cell in experts to etsist the NSRG review of conplex problen.s.
All nten.bers shell teve at leest a bache-lor's degree in er.gir.eering or related sciences. The Cteirnen shall teve at least 10 years of profus;or.el level n.enegenent experit.nce in the power field and tech of tie other n.en.bers shell have et itest 5 years of cunpletive profes-sionel itvtl experience iri one or n. ore of the fields listed in Section 6.5.7.1.
A,LTERNATES 6.5.P 3 All alternate rien. tars shell Lt appointed in writing ty the NSFG Chairt.tn to scrve on e tertporer) basis; however, no n. ore then tvt. eltert etts shell per ticipate es voting nenbers in f:SPC activitits at at:y bne tir.te.
,C,0p,SULTANTS 6.5.P.4 Consultants itall be utilirec' es deterniir4d by the NSFC Chairraen to providt opert echice to the NSPC.
PEETIUp,[,y,0Mp,C}
6.5.2.5 The NSR0 shell nut at leest once per 6 n.onths.
QUORUM 6.5.2.6 The quoruni of the USFC r.ecessery for thic peiforr.erict of the E!FG review er d eudit forations of these Technical Specifications shell tensist of the Chaitten or his desigt.ated altert.6te et 1 tost one half of the ren.ainir s USRG neriers of which two ne) Le 61ternetes.
No incre than a niurity of the quorur~ shell tevc lir4 responsibility for operatiori of the unit.
REVIEW 6.5.P.7 The NSRG shall br. responsible for the revitw cf 6.5.P 7.a and shall review 6.5.2.7.b through 1:
a.
Tht safety eyeluations for (1) cher,ges to procedures, equirnent, fetilities or systens ar.d (?) tests or ex:eriroents contpleted under the provisior. of 10 CFR 50.59 to verify taet such actions did not cer stitute an ur.revitktd safety qnstion; b.
Proposed ther.ges to procedures, equippent, or systens which involve en unr(viewtd safety question as defir ed in 10 CFR 50.59; FErJ'I - UMlT 2 6 10 Arrendrnent No. 77. N, 54