ML20082Q245
| ML20082Q245 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/28/1991 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082Q246 | List: |
| References | |
| NUDOCS 9109120055 | |
| Download: ML20082Q245 (14) | |
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','n UNITED STATES i
I" i
NUCLEAR REGULATORY COMMISSION
- S WASHINGTON, D C. 20W.s o,g ' s f GEORGIA POWER _ COP,PANy OGLETHORPE POWER CORPORATION Mull 1CIPAL ELECTR1C A,U,THO,RITy,0,F, GEORGI A CITY,0FDALT0tl, GEORGIA DOCKET tiO._ 50-32_1 EDWIN 1. HATCH NUCLEAR PLAtlT Uti1T_1 AMEt10MEllT TO FACILITY OPERATillG LICENSE Amendment flo.172 License No. DPR-57 1.
The !!uclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin 1. Hatch fluclear Plant, Unit 1 (the facility) Facility Operating 1.icense No. DPR-57 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (thelicensees)datedJune 13, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable atosurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulati as and all applicable requirements have been satisfied.
l 9109120055 910G20 i
l PDR ADOCK 05000321 P
PDH
, 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-57 is hereby amended to read as follows:
Techn,1, cal Specifications The Technical Specifications contained in Appendices A and B, as revised ttrough Amendmert No.172, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implerented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David 6. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: August 28, 1991
ATTACHMEhT TO LICENSE AMENDMENT NO.172 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with i
the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Pages insert Pages 3.2-23 3.2-23 3.2-48 3.2-48 3.10-2 3.10-2 j
3.10-7 3.10-7 9
i N
1 NORS FOR TABLE 3.2-11
-4 o
I e
The column entit'.e1 *Ref. No."is ordy for converwec<e so that e one-toene re:etsomtup con be establishad between items in Table 3.211 and items in Table 4.2-11.
c Q
b.
tmutmg Condetsons for Operation for the Neutron Morwtoring Spiem are listed in Table.1.2-7.
f W
c.
Weth one or more of the morwtormg chonnels inoperable, either restore the inoperable $.simet(s) to OPER ABLE status witNn 30 days or be in et teest HOT SHUTDOWN withm the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i Contmue<t nperation is permissible for seven deys from and after the date that one of these parameters es j-not indecated in the control room. Suevedlance of local panels wid be substr*uted for indicateon in the cetrol room durmg the seven days.
d.
Drywen and Suppressron Chamb6r Pressure are each recorded on the same recseders. Each output chemal has its own recorder.
3 Drywer and Suppreswon Charrter ser temperature end suppeessson chamber water temperature e.e en recorded 3
on the some recorders. t.ech output charmel has its own secorder. Each recorder takes irput from several i
temperature elements.
1 Hydrogen and Oxygen are indicated on one recorder. The recorder hos two pens. one pen for each parameter.
7 Each channel of the post LOCA redesteon moratoring e srsm includes two detectors one located in the nj drywen end the othat in the suppression chamber. Each detector feeds a segnal te e separate log count rare r..eter. The meter output goes to a twe pen recorder. One Ng5 redie ror-level sierm is provided per channal and entmencretion of eierm is provided in the control e -
High Range DryweR Pressure end High Rengs Drywet Radiation are recordes on the some recorders. Eech output channel has sts own recorder.
53 e.
In tha event that eu indicetions of tNs parameter is desabled and such hidication cormot be restored on g
viu 86) hours, en orderfy shutdown shen be irwtieted and the reector she5 be in a Nt Shutdown condition o
m sin 66) hours and a Coid Shutdown condrtion en the fonowing eighteen (18) hours.
ar+
f.
.f eether the prFnery or secondary indNefeon is inoperable, the forus temperature will be nem.d at o
least once per sNft to oc>sarve any tsnowplemed temperature increase which might be endicative of en open
}
SRV. Weth both the primary end secondary morwtonna chennels of two ce more SRVs inoperable either rasvore suffectent innperable channels such that no rnoes then ene SF.V hos both pnmery and secondary w
N charmels erw.pereHe within 7 days or be in at least hot shutdown witNn the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i l
r 4
6 L
1 4
I L
y Tabs. 4,2-11 m=
Check and Calibration Mrumum Frequency for instrumerdaden WNch Prow, des Survedtence informeroon C
2 l
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Ref.
Instrument enstrument Check Instrument Cahbesteon No.
Mromum Frequency Vrumum Frequency tal (td (c) i r
i t
l 1
Reactor Vossal WW.or Level Each shf
-Once/ocieretmg cycle (f)
}
i 2
Sh<oud Water Level Each shft OnceToperntmg cycle (f) 3 Reactor Pressure Eoch shft Oncoloperstmg cycle (f) i 4
Drywen Pressure Each sf=ft Every 6 mantN 5
DryweN Temperature Eech sNft Every 6 months l
6 Suppress.on Cheneae Air Eech shft Everv 6 months Tempereture i
{
7 7
Surpression Chamber Water Each sNft Every 6 mort.hs g
Temperature j
8 Suerressio., Chamber Water Eech sNft Every 6 months
(
Lewal i
t 9
Suppress.on Chamt>er Each sNft Every 6 months 4
Pressure l
>B
]
10 Red Positum informenort Each sNft NIA l
ct System (RPts) a 5
a (D
l 3
11 64yrfrogen and Oxygen Monthly Every 3 rnenths
]
Analyser j
Z 4'
o 12 Post LOCA Red:stron Each shft Every 18 months
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N' f
i i
(
13 4 Safety /Reisef Velve Poset,on Pri-MontNy Eve y 18 months mary ind.caror G
f bl Sefety/Reisef Valve Posttson MontNy Ewary 18 months Secondary fewfecator F
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,_1 MUM WD1110riS IOR OtiL!dlLri sygdligigyMMWf5 3.10.C.
(ore Monitorina Durina Corg 4.10.C.
(twr,tirdtoriro Durina Core Alterations Altrations
- 1. During nomal core alterations, two Prior to nuking nomal alterations SINS shall te operable; one in the to the core the 9% shall te core quadrant where fuel or control fttctimally tested ard ctalad for ro:fs are teing roved and one in an nwtrun respmse. Tht'reafter, atlacent quadrant, except as specified elle recpired to te cgerable. tie in 2 and 3 telow.
9% will te docked daily for resgme.
for an SIN to te considered operable, it shall te inserted to the ncrmal Use of special novable, dnking operatirg level and shall have a tyre detectors drirg initial ftel minian of 3 cps with all rods capable loadirg and major core alteratims t,f nomal insertion fully inserted.
in place of rmal detectors is pererissible as larg as ite detector
- 2. Prior to spiral unloading the SINS is connected to tre norval 9N shall te preven operable as stated cirtuit.
above, however, during sriral unloading the courit rate may drop Prior to spiral mioa11rg or telow 3 cps, reloalin:i tfe 9h shall te fttrtionally testal. Prior to
- 3. Prior to spiral reload, up to four (4) spiral mluadtry the 96 stxuld fuel assert >lles will te leaded into also te dvrked for neutrtri core positions next to
- trspanse, each of the 4 SINS to obtain the required 3 cps. These assertlies may te any 4tch have teen shown to rneet the criteria for storage in the l
spent fuel pool. Until these assertilies have teen leaded, the 3 cps requirenent is not necessary.
D. Spent fuel Poni Water Level D. RentfuelIbolWatertevel 1
At least 21 fea. M h shall te main-Tre water level in the spent fuel tained over l'w to,. of adiated pool shall te deteminnd to te at fuel assent)h o rW.. s.' the spent least its minian rupired d@th fuel stsrage, W at least once ter 7 days.
E. Control Rod Drir 11. zenann E. Cmtrol Ibd Drive Maintenance
- 1. Eeauirevnts for Withdrawai
- 1. &c.irwrn11_fEMlhdrml of I or 2 Control Rods of I or 2 Cmtml Fhis A maxinn of two control rc.d5 separated by at leas' two control cells in all directions may te withdrawn or renoved from the core for the purpose of perfoming control rod drive maintenance prvvided that:
- a. Tre Hrle $ witch is locked in tie
- a. This surveillance twpirumnt is IffUELposition, The refuelirs the saw as givm in 4.10.A.
f interlock eich prwents trors tMn l
cre contrul red frm teirg withdrui ray te bypassal for one cf the l
cmtrol rods on which naintmance is teirg 1
tiAltri - UNIT 1 3.10-2 Amendment No. 172
~.- - - ~ - _. _ _ _ _
l t
_,,1.A1Lil0R LIM 111NUQMD1110NS FOR OPERA 110N j
i 3.10, A,2. fuel Grapole Hoist load Settino Interlocks Fuel handling is normally conducted with the fuel grapple hoist. The total load on i
this hoist when the interlock is required consist $ of the wel ht of the fuel
[
grapple _and the fuel assembly. This total is appio timately 100 lbs. in comparison j
to the load setting of 485 ; 30 lbs.
j i
3.
An!!iary R11113ltgj! Settina Interlock l
Provisions have als) been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks. The 485.t 30 lb load i
setting of these hoists is adequate to trip the interlock when a fuel bundle is l
being handled.
B.
fuel leadina fi in minimize the possibility of loading fuel into a cell containing ho control rod.
i it is required that all control rois are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks. as designed, will prevent inadvertent criticality.
l C.
Core Monitorino Durina Core Alteratio11 The $RMs are provided to monitor the core during periods of Unit shutdown and to guide the operator during refueling operatians and Unit startup. Requiring two operable SRMs in or adjacent to any core Quadrant where fuel or control rods are L
being moved assures adequate monitoring of that quadrant during such alterations, i
The requirements of 3 counts per second provides assurance thai neutron flux is
[
being monitored.
During spiral unloading. 11 is not necessary to maintain 3 cps because core I
alterations will involve only reactivity removal and will not result in criticality.
[
The loading of up to four fuel bundles around the SRMs before attaining the 3 cps is permissible because these bundles form a subtritical configurati'n.
l 0.
Scent Fuel Pool Water level the minimum water level in the spent fuel pool shall be maintained at least 21 feet I
above the top of-the upper tie plates of the irradiated fuel assemblies seated 1.-
the spent fuel pool racks, lhis minimum level ensures removal of at least 98.A%
of the assumed 10% iodine gas activity released from the rupture of en irradiated fuel assembly. This water depth is consistent with the assumptions for the fuel I
handling accident analysis outlined in Regulatory Guide 1.25 and the requirements
[
in Standard Review Plan l$.7.4 for radiological releases resulting from that accident.
t f.
(.pnfrol Rod Drive Maintenance During certain periods, it is desirable to perform maintenance on two control rod I
drives at the same time.
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3.10-7 Amendment No. 172
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,,q /) t UNITED STATES 6
i NUCLEAR REGULATORY COMMISSION t
[
WASHINGTON, D.C. 20%6 o
GEORGIA POWE,R,, COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECT,RIC, AUTr0RIT,Y 0F,,GE0,RGJ6 CITY OF DALT0!!, GEORGIA DOCKET fl0.50-36C EDW1H 1. HATCH NUCLEAR PLANT, Uti1T 2 AMENDMENT TO FACILITY OPEPATING LICENSE Anendment No. 112 Licence No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin 1. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees) dated June 13, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amer.ded (the Act), and the Consission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this license aniendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendsent is ir sccordance with 10 CFR Part 51 of the Commission's regulations and ill applicable requiremens have been satisfied.
, 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of f acility Operating License No. NPF-5 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised thrcugh Amendment No,112, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
I FOR THE HUCLEAR REGutATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - 1/.
Office of Huclear Reactor Regulatic:,
Attachment:
Technical Specification Changes Date of issuance: August 28, 1991
ATTACHMENT T0 t.! CENSE AMENDMENT NO.112 4
FACitITY OPERATING l.! CENSE NO. NPF-5 DOCKET NO. 50-366 I
Replace the following sages of the Appendix "A" Technical Specifications with the enclosed pcges. T1e revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages 3/4 3-13 3/4 3-13 3/4 3-14*
3/4 3-14 3/4 8-23 3/4 0 23 i
3/4 8-24*
3/4 8-24 i
3/4 9-15 3/4 9-15 l
3/4 9-16*
3/4 9-16 8 3/4 9-2 B 3/4 9-2 4
' Overleaf pages 4
i A.
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,._~...__..m,,m,
1 1
~h' TABLE 3 3 'c 1 tCee,micf1 w
O i
I tSOLATiON ACTUAT*ON EP6 DUVENTAT?cN a
h V ALVE GROW ARNatJ9A NUMBEst
' APPLICABLE
[
OPER ATED BY OPER ADLE CHANNELS OPER %T:ONAL TR'P FUNCT80*4 SIGN Atte!
FEst T3nP sysTEwMk! '
' CON > TON ACTION to 4.
HtGH PRESSURE COOLANT P4)ECTson sv5 TEM tsetATroN a HPCI Steam Lme Flen - ng%
3 1
- 1. 2. 3 26 (2E41 N657 A.8)
In HPC1 Steam surpty Pressure -
Low (2E41-NE58 A 8,C.D1
- 3. 8 2
1, 2. 3 26 c.
HPC1 Turtano E=houst Diephregm Presswee - Mgh (2E41-N655 A.BADI 3
2 1.2.3
- s d.
HPCI Pee Pecetreborn Room 7
Toweerature - Hgh (2E41-N671 A. El 3
1
- 1. 2. 2 26 w
e.
SuppeessMm Poca Aree Amtwent i
Tenmeenture - bgh (2E51-N666 C. D) 3 1
1.2.3 26 I
w t.
Suppeesmon Pool Aree W Te-w - ngh (2E51-N665 C. D; 3
1
- 1. 2. 3 26 2E51-N663 C. D:
2E51 NS64 C. Di g.
Suppression Pool Aree Te versture M
Timee Reseys (2E51 M603 A. ED 3'*
1 1.2.3 26 3
CL 9
h.
Erregeracy A e Coo 6ee Tempersose-Mg5 t2E41-N670 A,88 3
1
- 1. 2. 3 26 r+
Z
- i. Drywen Presswe Hgh
?
I2E15 NSS4 C. D) 8 1
1.23 26
- j. Lope Tower Morstne (2E41-K1)
N A**
.1
- 1. 2. 3 27 1
l j
TABLE 3.3.2-1 (4.~ontinue.8)
ISOI.ATICC ACTUATION INSi5IillEiATICC VAI.VE GHOUPS llINilllRI NtR111).H APPI lCAllli.
OPERATED BY G11HAHl1 CilANNEiS Ot*1_ RAT ION Al.
g THI P_[ItNCTION SIGNAL.(a)
ITH TRIP SYSTEIMI.)(c) CONillTIO88 ACTinH 5.
RI' ACTOR CORE ISOLAl' ION h
{rs01.ING SYSTEll ISOI.ATION a.
RCIC Steam Line Flow-Iligh 4
1 1, 2, 3 26 l
(2E51-N657 A,B) i b.
RCIC Steam Supply Pressure -
le:,w (2E51-N658 A, B, C, D) 4, 9 2
1, 2, 3 26 g
c.
RCIC Turbine Exhaust Diaphragm Pressure - Hi h 4'
2 1, 2, 3 26 t
l (2E51-H685 A, B, C, D) d.
E.crgency Area Cooler Temperature -
g Iligh (2E51-N661 A, B) 4 I
I, 2, 3 26 o
Suppression Pool Area Ambient e.
Temper a ture-lii gh 4
1 1, 2, 3 26 8-l
'(2E51-M666 A, B )
f.
Suppression Pool Area a T-High 4
I I, 2, 3 26 (2E51-N665 A, B; 2E51-N663 A,B; 2E51-N664 A,B) 3 Suppression Pool Area Temperature 4,)
g I
I, 2, 3
26 Timer Relays (2E51-H602 A, B) h.
Iirywell Pressure - Iligh l
y (2 Ell-N694 A, B) 9 I
1, 2, 3 26 E
NA(h)
I I, 2, 3 21
!}
- i. Logic Power lionitor (2E51-KI)
S" 6.
SillrIIK&_'N COOLING SYSTEll ISOLATION Heactor vessel Wter Level-Low (I.evels 6, 10, 11, 2 2
3, 4, 5 26 l
a.
3 3)(21121-N680 A, B, C, II) 12 k
4 Reactcr Steam Dome Prer.sure-Iligh II 1
8, 2, 3 (2831-N679 A, D) w
TABLE 3,8.2.6-l_IContinued]
PRIMARY LQULALNMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTlVE DEVICES DEVICE NUM8ER SYSTEM / COMPONENT AND [0 CATION
- POWERED 13.
600 VAC, MCB, M0 ORYWELL RETURN AIR FAN 2R24-50ll, FR 18C 2T47-C001A 14.
600 VAC, MCB, M0 1RYWELL COOLING UNIT 2R24-5013, FR 3B iT47-8010A 15.
600 VAC, MCB, M0 DhYWELL COOLING UNIT 2R24-5014, FR BA 2T47-8010B
- g. Type 7:
1.
208 VAC, MCB, M0 DRYWELL CHEMICAL DRAIN 2R24-S013, FR llD SUMP PUMP 2Gil-C101 2.
208 VAC, MCB, F0 DRYWELL RETURN AIR FAN 2R24-S012, FR 23C 2T47-C0028 3.
208 VAC, MCB, M0 DRYWELL RETURN AIR FAN 2R24-S0ll, FR 22C 2T47-C002A i
l
- MCB - molded case circuit breaker l
M0 - magnetic only l
TM - thermal magnetic HATCH - UNIT 2 3/4 8-23 Amendment No.112
. ~ - - - - -, -. _ _ _, _.. _ - _ _. - -. _
"ELECTnlCAL POWER SYSTE5t$
3/4.8.7 ONSITE POWER 0157Rl8UT!0ft SYSTEMS TLECTRIC PONER N0fl1TORINM FOR REACMR PROTECTION SYSTEM LIMIT!NG CON 0! TION FOR OPERATION 3."8. 2. 7 The power monitoring system for a RPS MG set or the Altert, ate Source shall be OPERABLE if in service.
APPLICABILITY:
At all times.
ACTION:
With the power monitoring system for a RPS HG set or the Alternate Source inoperable, restore the inoperable power monitoring system to OPERABLE status within 30 minutes or remove the RPS HG set or Alternate Source associated with the inoperable power monitoring system from service.
One channel of a power monitoring system may be inoperable, as necessary for test or maintenance, not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per month.
SURVEILLANCE REOUIREMENTS 4.8.2.7 The above specified RPS power monitoring system instrunentation shall be detemir,ed OPERABLE:
At least once per 6 months by performing a FLHCTIONAL TEST:
a.
and b.
At least once per operating cycle by demonstrating the OPERABILITY of over-voltage, ander voltage and under-frequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints.
1.
Over-volt.cyc s 132 VAC, 2.
Under-voltage > 108 VAC, with time delay relay set to zero*,
and 3.
Under-fretJency 3,57 Hz.
PendinD NRC approval of different value.
1, HATCH - UNIT 2 3/4 8-24 Amen &wnt No. 23
REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - SPENT FUEL STORAGE POOL i
LIMITING CONDITION FOR OPERATION l
3.9.10 At least 21 feet of water shall be maintained over the top of l
irradiated fuel assemblies seated in the spent fuel storage pool racks.
I APPLICABlllTl:
Whenever irradiated fuel assemblies are in the spent fuel storage pool.
ACTION:
[
t With the requirements nf the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area af ter placing the load in a safe condition.
The provisions of Specification 3.0.3 are not applicable.
l t
SVRVEILLANCE REQUIREMENTS i
4.9.10 The water level in the spent fuel storage pool shall be determined to be at least its minimum required depth at least once per 7 days.
I i
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HATCH - UNil 2 3/4 9-15 Amendment No.112 l
REFUELING OPERATIONS 3/4.9.ll_ CONTROL ROD RE m VAL SINGLE CONTROL ROD REH0 VAL LIMITING CONDITION FOR OPERATION 3.9.11.1 One control rod and/or the associated control rod drive mechanism may be removed from the reactor pressure vessel provided that at least the following requirerents are satisfied until the control rod and asso-ciated control rod drive mechanism are reinstalled and the control rod is fully inserted in the core.
a.
The reactor mode switch is OPERABLE and locked in the Refuel position per Specification 3.9.1.
b.
The source range monitors (SRM) are OPERABLE per Specification 3.9.2.
c.
The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed; 1.
May be assumed to be the highest worth control rod required to be assumed to be fully withdrawn by the SHUTDOWN MARGIN test, and 2.
Need not be assumed to be immovable or untrippable, d.
All other control rods in a five-by-five array centered on the control rod being removed are fully inserted and electrically
- disarmed, e.
All other control rods are either fully inserted or have the surrounding four fuel assemblias removed.
APPLICABILITY:
CONDITION S.
ACTION:
With the requirements of the above specification not satisfiedi suspend removal of the control rod and/or associated control rod drive mechanism from the reactor pressure vessel and initiate action to satisfy the above requirements. The provisions of Specification 3.0.3 are not applicable.
HATCH - UNIT 2 3/4 9-16
~ _
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4 4
ELQJJ1NGOPEATIONS EASES i
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i JId 9.6 [0%UNICA110NS a
1he reautrement for communications capability ensures that refueling f
station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel i
within the reactor pressure vessel.
I 3/4.9.7 fRANE AND HDIST OPERABilllf The OPERABillTY requirements of the cranes and hoists used for movement of fuel assemblies ensures that: (1) each has sufficient load capacity to i
lift a fuel element, and (2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently l
engaged during lifting operations.
1 3/4.9.8 CRANE TRAY ((.$PINI FUEL STORAGE P0QL i
The restriction on movement of loads in excess of the nominal weight of a fuel eierent over irradiated fuel assemblies ensures that no more t
than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident anelyses. All fuel loaded into the Edwin 1.
Hatch Nuclear Plant spent fuel pool shall have an uncontrolled lattice ko less than or equal to the limit for high density fuel rach described in the
- General Electric Standard Application for Reactor fuel" (CESTAR 11),
NEDE 240ll-P-A-8.
Alternatively, fuel not described in CESTAR 11 shall have
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been analyzed with another hRC approved methodology to er.sure conformity to the FSAR design basis for fuel in the. spent fuel racks.
I 3/4.9.9WATERLEVEl-REACTORVES$(L The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity
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released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis, i
3/4.9.10 WATER LEVEL-SPENT FUEL STORACE POOL The minimum water level in the spent fuel pool shall be maintained at i
least 21 feet above the top of the upper tie plates of the irradiated fuel assembliet seated in the spent fuel pool racks. This minimum level ensures removal of at least 98.6% of the assumed 10% iodice gap activity released from i
the rupture of an irradiated fuel assembly. This water depth is consistent with the assumptions for the fuel handling accident analysis outlined in Regulatory Guide 1.25 and the requirements in Standard Review Plan 15.7.4 for radiological releases resulting from that accident.
I 3/4.9.11 CONTROL ROD REMOVAL This specification ensures that maintenance or repair of control rods r
or ontrol rod drives will be performed under conditions that limit the
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probAility of inadvertent criticality. The requirements for simultaneous remova' of more than one control rod are more stringent since the SHUTDOWN MARCIN :,pecification provides for the core to remain subcritical with only one control rod fully withdrawn.
HAtta - UNii 2 B 3/4 9-2 Amendment No, 112 I
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