ML20235N571
| ML20235N571 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/24/1989 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20235N575 | List: |
| References | |
| TAC-69408, NUDOCS 8903010290 | |
| Download: ML20235N571 (13) | |
Text
3 UNITED STATES 1
l g.
NUCLEAR REGULATORY COMMISSION 5
ij WASHINGTON, D. C. 20S55
\\, +... + /
GEORGIA POWER COMPANY l
OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.160 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-57 filed by. Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Munici)al Electric Authority of Georgia, and City of Dalton, Georgia, (tle licensee) dated September 6,1988, 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 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; l
D.
The issuance of this amendment will not be inimical to the common l
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 regulations and all applicable requirements have l
been satisfied.
l 8903010290 890224 "
ADOCK0500g DR 1
. 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations 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:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.160, 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 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 24, 1989
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby i
amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 160, 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 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/g David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects-1/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 24, 1989 ayg y
e.
0FFICIAL RECORD C0 j
f LA:PQLI-3 PM:PDII-3 bdSPLB 0G -vF,
H -3 MROWI LCrocker:1s IJCraig 15Matthews 01/ 6 /89 01/9/89 I 01/')9 89 Of/ 3/89 (f)./M/89
i j
l l
4 ATTACHMENT TO LICENSE AMENDMENT NO.160 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET N0. 50-321 l
I Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised page is identified by amendment number and contains a vertical line indicating the area of change.
Remove Page Insert Page 1.0-1 1.0-1 1.0-2 1.0-2 3.2-4 3.2-4 3.5-2 3.5-2 3.5-3 3.5-3 4
3.5-4 3.5-4 3.5-5 3.5-5 3.6-7 3.6-7 3.7-12 3.7-12 L__--_______---______________
1.0 Definitions The following terms are defined so that a uniform interpretation of these specifications may be achieved.
A.
(Dewted)
B.
Cold Shutdown Condition - Cold shutdown condition means reactor operation with the Mode Switch in the SHUTDOWN position, coolant temperature
<212*F, and with no core alterations permitted.*
l
- During the performance of inservice hydrostatic or leakage testing with all control rods fully inserted and reactor coolant temperature >212*F, and/or reactor vessel pressurized, the reactor may be considered to be in the Cold Shutdown Condition for the purpose of determining Limiting Condition for Operation applicability. Note that the Cold Shutdown Condition may be referred to in different ways throughout the Technical Specifications. For example, " reactor subcritical and reactor coolant teinperature < 212 F," " irradiated fuel in the reactor vessel and the reactor is depressurized," " reactor water temperature < 212"F and reactor coolant system vented," or " reactor is not pressurized (s.e.,
212*F)" should be interpreted as COLD S!!UTDOWN. However, compliance with an ACTION requiring COLD SHUTDOWN shall require a reactor coolant temperature < 212 F.
In addition, compliance with the following Specificatioiis is required when performing the hydrostatic or leakage testing under the identified conditions:
3.5.B.1.b, 3.5.C.1.c, 3.6.F.2.d, 3.7.C.1.a(7), 3.9.c, and opplicable notes in Table 3.2-1.
HATCH - UNIT 1 1.0-1 Amendment No. 160
C.
Core Alteration - Core alteration shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Suspension of core alterations shall not preclude completion of the movement of a component to a safe conservative position.
D.
Design Power - Design power refers to the power level at which the reactor is producing 105 percent of reactor vessel rated steam flow. Design power does not necessarily correspond to 105 percent of rated reactor power. The stated design power in megawatts' thermal (MWt) is the result of a heat balance for a particular plant design.
For }'Otch Nuclear Plant Unit I the design power is approximate 1537 MWt.
l E.
Engineered htety Features - Engineered safety features are those features provided for mitigating the consequences of postulated accidents, including for example containment, emergency core i
cooling, and standby gas treatment system.
F.
Hot Shutdown Condition - Hot shutdown condition means reactor operation with the Mode Switch in the SHUTDOWN position, coolant temperature greater than 212 F, and no core alterations are permitted.*
l G.
Hot Standby Condition - Hot standby condition means reactor operation with the Mode Switch in the START & HOT STANDBY position, coolant j
temperature greater than 212 F, reactor pressure less than 1045 psig, cri ti cal.
H.
Immediate - Immediate means that the required action shall be i
initiated as soon as practicable, considering the safe operation of the Unit and the importance of the required action.
I.
Instrument Calibration - An instrument calibration means the adjustment of an instrument output signal so that it corresponds, within acceatable range and accuracy, to a known value(s) of the l
parameter w1ich the instrument monitors.
1 J.
Instrument Channel - An instrument channel means an arrangement of a sensor and auxilibry equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.
l l
l I
l
- During the performance of inservice hydrostatic or leakage testing with I
all control rods fully inserted and reactor coolant temperature > 212 F, and/or reactor vessel pressurized, the reactor may be considered to be in the Cold Shutdown Condition for the purpose of determining Limiting Condition for Operation applicability.
However, compliance with an ACTION requiring COLD SHUTDOWN shall require a reactor coolant temperature
< 212 F.
HATCH - UNIT 1 1.0-2 Amendment No.160
l' l
r e
d h
e t
h o
s e
i he r
t e e l
tvce o
n b gh b
oig f
a nt a
rbra l
li t
oaek s
o l rt s
f ha m
o aua n
e spe e
c hdh o
sisl t
s t
rdg e
d o
s rw s
een m i t
b oemr y
oo mrd wtigo e
rrto s
tl euee oarnr n
ua c
todl pi u i f l
p a
l t c
p awt shib cdr pm.
c oati i
eoo y va ro u gi o r rrat r
rl n s2or osddnrce p
te a
t re tse it w
h omts d
hhi ipopp catdt ro m.
i npst ees
,o a
celdep t
edetuet s
ceso e
tt r
n ter p
p tpm) rnpsstfd o
g td i
hhi uee axueaae p
ttr t
tr e
ejri t
nr y
i esh r
iit arsa hl d
csa t
iwtc t
ww hoy(
tel asol r m
a aail e
tfs ve l se l
rwse r
ldr s
heveesv%
e n
itope l
sas mbrm iwdlebvae0 r
d ys o
eeu
, e pi tl hrhs sts sli e
l2 e
e yld t
v eie eate nna nnw o
wcr b
vcr tr t
tuomnaoo aee a
op seds 1
soi o i o
oewc r
rs ypey -
ertsidtel t
troi e
osl sots tgatteab 7
pv p
tal na kinani e
re r
o c e pii p 3 ncdiiids o
iohwe anw ireor e i n
ci oaaodmal tos o
eoy brparre 1
r l n w
rir tbst l
t tre v a
pn i t
td s
b cnee tl e eyat a eohsl eal 2
eg ea t
shnn e
hted ha a
T jit adm 3
a ewia b
tnhe tmeh nt pu rr r
h m m eT r ht n
e t
i rrg l
nm i fmt i iafiteoe i_
i orcbnw o
l toon l
ou.u oe n nd tal gp a
r q
t gi n eae b
o l
,f i a
s rm h
gtse esnh e grudfl n er s
nsl r n yi t v o l eoair T
ll r
c1 c
f e
tint pw i rrat hho e.
a a po ine o s pi g devassst o
n-tier r
s no r pi yw i t ic f
e2 atl e e
el an oirr e hodcnll a i
iip h
t eh frte r peoeebe s
n4 drh t
vcs t
t a e tsmvar e
e ecwr ee i t
te hl aseet gl d eaum n tll arl sg n
o t
ve i st d
a ey m oa o uu u en o
nl N
ob aip,
e krpt fhr i ffcdsn i l naoeh ca t
eL r
a e pi eota t
ol aaeirs r
i oxM i
etir b
ip l
tac T
nrcW a
u l arg l w on ate q
wte te e o rm l mt i afl fi mc 5 e
r t
ongm s a r aeriob al r
ol rn nnna to vod a i
lys eeed eoi nais snne ast s
cs aht p
ol srrs m brse i
iset ccst u de n ue ne l eec tel n ea o ehdnodo se ot lhvx y
avbe sl th r nteoinho i
s at e
t t
am a
t G
situ t
n,atao4 id h
r i
srrn lene a rbal r2r n snoo r
ooei ngcg h l
ett g
rt pa ninn c pe i ug o
".a yh c e
dcot asai a
wddckni N-itwat t
ya n h l r e eoeal cir o1 n
heoo c pl o h pgr a ah p
,e o i
rn rwc iit h
t n
cbt r
2 e
erei t
dad id c
,t y l
gl tvn oehnrewn f3 eaes t
e tchl n
iFsr b
ro iw ttcueh a Re nite e
v*ea aeum a
ohTt l
it v
m r2rm rns d
rrert en b
ine n
e1ii eo
)
e o p
t i% ke.so b gi da tril i
s2ur o,d s i
e sr t
iee rrl i r0yc net eT nc a
n qp e(
a p2 a a n o b c ltn m gi e n
vre feie c
mboi e nn w o
go s ovun o sn tlj ii nb pu eqn s rasedcl n o
c.
t iil ns aoep n i aia rwea s uhtheeai ee tgu yo m rc eorh a otnTsjh n n ft ri rete bI c
h i ansn l
ro.bi e
ni of a at or b m
f s
sg 4ep n.og u
e ml cod mc fu) n v 2ttttnto u
ne in rt2d aessenr l n year ru ea e dsi l
n e s e e gi d pf prl r eedn a oe rs' u.
s eei h poi v
ir ttooy ce aol w mp st nh npvu t pi a hgp r ph et ir2es ec emeq irm e
t ieedt he ru1 re ha hete free h itrhhyef Tb P p2 pt We Wt(r I tpr T
Wattthso a
b c
d e
".r1
>5mDOBe3 Z. ~aO s
o, 3
T3HnI ' :c2El "-
1l'l1 l!
l ll
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.5.A.2.
Operation with Inocerable 4.5.A.2.
Surveillance with Inocerable Components Components If one CS system loop is inoper-When it is determined that one able, the reactor may remain in core spray loop is inoperable operation for a period not to at a time when operability is exceed seven (7) days providing required, the other core spray
{
all active components in the loop and the RHR system LPCI J
other CS system loop, the RHR mode shall be demonstrated to
{
system LPCI mode and the diesel be operable immediately. The generators (per Specification operable core spray loop shall i
4.9. A.2.a) are operable.
be demonstrated to be operable When performing an inservice daily until both loops are hydrostatic or leakage test returned to normal operation, with the reactor coolant temperature above or below 212'F the CS system is not required to be operable.
3.
Shutdown Requirements If Specification 3.5.A.l.a. or 3.5.A.2. cannot be met the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
B.
System (LPCI and Containment System (LPCI and Containment Coolina Mode)
Coolina Mode) l 1.
Normal System Availability 1.
Normal Operational Tests RHR system testing shall be performed as follows:
Item Freauency a.
The RHR System shall be operable:
- a. Air test on Once/5 years l
drywell head-(1) Prior to reactor startup ers and nozzles from a cold condition, or and air or water test on (2) When irradiated fuel is in torus headers the reactor vessel and the and nozzles reactor pressure is greater l
than atmospheric except as stated in Specification 3.5.B.2.
l HATCH - UNIT 1 3.5-2 Amendment No. 160 1
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i
3.5.B.l.
Normal System Availability (Cont.)
4.5.B.1.
Normal Operational Tests b.
One RHR loop with two pumps or two l Item Freauency loops with one pump per loop shall be operable in the shutdown cool-
'b. Simulated Once/ Operating ing mode when irradiated fuel is Automatic Cycle in the reactor vessel and the reac-Actuation l
I tor pressure is atmospheric except Test prior to a reactor startup as stated in Specification 3.5.B.l.a.
During an inservice hydrostatic or leak-age test, one RHR loop with two pumps or two loops with one pump per loop shall also be operable c.
System flow Once/3 months in the LPCI mode, rate: Each RHR pump c.
The reactor shall not be started up shall deliver with the RHR system supplying at least 7700 cooling to the fuel pool, gpm against a system head of d.
During reactor power operation, the at least 20 psig.
LPCI system discharge cross-tie valve, Ell-F010, shall be in the d.
Pump Opera-Once/ month closed position and the associated bility valve motor starter circuit breaker shall be locked in the e.
Motor Oper-Once/ month off position. In addition, an ated valve annunciator which indicates that operability the cross-tie valve is not in the fully closed position shall be f.
Both recirculation pump discharge available in the control room.
valves shall be tested for oper-ability during any outage exceeding e.
Both recirculation pump discharge 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have valves shall be operable prior to not been performed during the reactor startup (or closed if per-preceding month, mitted elsewhere in these speci-fications).
2.
Doeration with Inoperable 2.
Surveillance with Inoperable Components Components a.
One LPCI Pumo Inoperable a.
One LPCI Pump Inocerable 4
4 If one LPCI pump is inoperable,the When one LPCI pump is inoperable, reactor may remain in operation for the remaining LPCI pumps and asso-a period not to exceed seven (7) ciated flow paths and the Core days provided that the remaining Spray system shall be demonstrated LPCI pumps, both LPCI subsystem to be operable immediately and flow paths, the Core Spray system, daily.thereafter, until the inoper-and the associated diesel genera-able LPCI pump is restored to tors are operable (per Specifica-normal service, tion 4.9.A.2.a).
b.
One LPCI Subsystem inoperable b.
One LPCI Subsystem Inoperable
)
A LPCI subsystem is considered to When one LPCI subsystem is inoper-be inoperable if (1) both of the able, all active components of the LPCI pumps within that system are remaining LPCI subsystem and the inoperable or (2) the active Core Spray system shall be demon-valves in the subsystem flow path strated to be operable, immediately are inoperable.
HATCH - UNIT 1 3.5-3 Amendment No. 160
f 1
e LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B.2. Operation with Inocerable 4.5.B.2. Surveillance with Inocerable Components (Continued) foiiioonents (Continued)
- 1 i
b.
If one LPCI subsystem is inoper -
and daily thereafter, until i
able, the reactor may. remain in the inoperable LPCI subsystem l
operation for a period not to is restored to normal service.
i exceed seven (7) days provided that all active components of the remaining LPCI subsystem, the Core Spray system, and the associated diesel generators are operable (per Specification 4.9.A.2.a).
c.
When performing an inservice hydrostatic or leakage test with the reactor coolant temperature above or below 212'F., comply with Specification 3.5.B.1.b.
HATCH - UNIT 1 3.5-4 Amendment No. 160
{
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B.3.
Shutdown Requirements If Specification 3.5.B.1.a. or 3.5.B.2. cannot be met, the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
RHR Service Water System 4.5.C.
RHR Service Water System 1.
Normal System Availability 1.
Normal Operational Tests The RHR service water system RHR service water system testing shall be operable:
shall be performed as follows:
Item Freauency a.
Prior to reactor startup
- a. Pump & Valve Once/3 months from a Cold Shutdown Operability Condition, or
- b. Pump Capacity After pump b.
When irradiated fuel is in Test:
maintenance the reactor vessel and the Each RHR ser-and once/3 reactor vessel pressure is vice water months greater than atmospheric pump shall pressure except as stated in deliver at Specification 3.5.C.2.
least 4000 gpm at a system head of at least 841 feet, c.
When irradiated fuel is in the reactor vessel and the reactor is depressurized at least one RHR service water loop shall be operable.
2.
One Pumo Inocerahle 2.
One PumD inoDerable If one RHR service water When one RHR service water pump is inoperable the pump is inoperable.the remaining reactor may remain in operation active components of both RHR for a period not to exceed service water subsystems shall be )
30 days provided all demonstrated to be operable other active components of both immediately. An operable RHR subsystems are operable.
service water pump shall be When performing an inservice demonstrated to be operable hydrostatic or leakage test, daily thereafter until the i
comply with Specification inoperable pump is returned to I
3.5.C.l.c.
normal service.
i HATCH - UNIT 1 3.5-5 Amendment No. 160 l
J
~ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.F 2.c.
When the time limits or maxi-4.6.F.2.c.3.
Primary coolant pH shall mum conductivity or chloride be measured at least once concentration limits are er-every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> whenever ceeded, an orderly shutdown reactor coolant conduc-shall be initiated and the tivity is > 2.0pmho/cm reactor shall be in the Cold at 25'C.
Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
Whenever the reactor is not pressurized, a sample of the reactor coolant sha'il be analyzed at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> for chloride ion content and pH.
l G.
Reactor Coolant Leakaae G.
Reactor Coolant Leakaae*
1.
Unidentified and Total Unidentified sources of reactor Any time irradiated fuel is in coolant system leakage shall be the reactor vessel and reactor checked by the drywell floor coolant temperature is above drain sump system and 212*F:
recorded at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Identified sources of a.
reactor coolant system leak-reactor coolant system age into the primary contain-leakage shall be checked by ment from unidentified sources the equipment drain sump shall not exceed 5 gpm when system and recorded at least averaged over a 24-hour period; once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The readings provided by the primary b.
reactor coolant system leakage containment atmosphere into the primary containment particulate radioactivity f rom unidentified sources monitoring system, the shall not increase more than primary containment radio-2 gpm when averaged over a iodine monitoring system.
24-hour period; and and the primary containment gaseous radioactivity c.
the total reactor coolant sys-monitoring system shall tem leakage into the primary also be recorded at least containment shall not exceed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
25 gpm when averaged over a 24-hour period; when checked in accordance with 4.6.G.
2.
Leakaae Detection Systems a.
At least one of the leakage measurement instruments assoc-iated with each sump shall be operable and two of the other three leakage detection systems identified in Tabir 3.2-10, note c shall be operable when irradiated f uel is l
}
l t
- Not required during performance of an inservice hydrostatic or leakage test even if reactor coolant temperature is above 212*F.
HATCH - UNIT 1 3.6-7 Amendment No. 160 I
LIMITING CONDITIGHS FOR OPERATION SURVEILLANCE REQUIREMENTS C.
- C.
Secondarv Containment
- 1. Secondary Containment Intearity 1.
Surveillance While Inteority Maintained a.
Integrity of the secondary con-tainment shall be maintained Secondary containment surveillance during all modes of Unit 1 plant shall be performed as indicated operation except when all of the below:
following conditions are met:
(1) The reactor is suberitical and Specification 3.3.A. is met.
a.
A preopertional secondary contain-ment capability test shall be (2) The reactor water temperature is conducted af ter isolating the below 212*F and the reactor secondary containment and placing coolant system is vented.
the standby gas treatment system filter trains in operation. Such (3) No activity is being performed tests shall demonstrate the capa-which can reduce the shutdown bility to maintain a minimum margin below that stated in 1/4-inch of water vacuum under Specification 3.3.A.
calm wind (< 5 mph) conditions with each filter train flow rate not (4) The fuel cask or irradiated fuel more than 4000 cfm.
is not being moved in the re-actor building, b.
Secondary containment capability to maintain a minimum 1/4-inch (5) All hatches between Unit 1 of water vacuum under calm wind secondary containment and
(< 5 mph) conditions with each Unit 2 secondary containment filter train flow rate not more are closed and sealed.
than 4000 cfm shall be demonstrated at each refueling outage, prior to (6) At least one door in each refueling, access path between Unit I secondary containment and Unit 2 secondary containment is closed.
(7) Inservice hydrostatic or leakage test of reactor vessel is not in progress, b.
Integrity of the Unit 1 secondary containment shall be maintained during all modes of Unit 2 plant operations except Operational Condition 4 as defined in the Unit 2 Technical Specifications.
l
- For secondary containment during 1982 ref ueling outage, see page 3.7-12a.
l HATCH - UNIT 1 3.7-12 Amendment No. 160 l