ML20247A735
| ML20247A735 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/21/1989 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247A741 | List: |
| References | |
| GL-83-43, NUDOCS 8903290179 | |
| Download: ML20247A735 (42) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
r i.#-
{I e ni
-[k
'o,,
UNITED STATES 0, v.
g:
NUCLEAR REGULATORY COMMISSION E
- j WASHINGTON, D. C. 20555 ARKANSAS POWER AND LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.118 License No. DPR-51 1.
The Nuclear Regulatory.Comission (the Comission) has found that:
A.
The application for amendment by Arkansas' Power and Light Company (the licensee) dated November 17, 1986 and a supplement dated January 13, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance:
(1)thattheactivitiesauthorized 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 Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
l 8903290179 890321 DR ADOCK 0500 3
= = _ _ _ - _ _
' 2.
Accordingly, the license. is amended by 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-51 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 118, are hereby incorporated in the license. The licensee shall pperate the facility in accordance with the Technical Specif~ications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION G. &
Jose A. Calvo, Director Project Directorate - IV.
Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Techt.ical Specifications Date of Issuance: March 21, 1989 l
l
[
t 4.
ATTACHMENT TO LICENSE _ AM_EljDtTf!T N0.118 FACILITY.0PERATING LICENSE.NO. DPR-51 DOCKET NO. 50-313 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by An.Endrent number and contain verticc1 lines indicating the area of change.
I REMOVE.PAGES INSERT PAGES iia iia
- f 27 27.
45f 45f 53d 53d i
53f 53f 531 531 57 57 57a 57a 66b 66b 66m 66m.
66n 66n 660 660 66v 66v 66w 66w 66x 66x 662 66z 66aa 6666 66bb 66bb 66cc 66cc 66ff 66ff 85a 85a 86 86 110n 110n 11001 11001 110ss 110ss 110tt 110tt 110zz 110zz 110bbb 110bbb 121 121 124 124 126 126 128 128 142 142 143 143 144 144 145 145 4
146 146 146a 146a 146b 146b i
e..
J.,
J' l
'e SECTION TITLE PAGE 4.16 SHOCK SUPPRESSORS (SNUBBERS) 110e 4.17 FUEL HANDLING AREA VENTILATION SYSTEM SURVEILLANCE 110h 4.18 STEAM GENERATOR TUBING SURVEILLANCE 4.19 FIRE DETECTION INSTRUMENTATION 110j 4.20 FIRE SUPPRESSION WATER-SYSTEM 110p 4.21 SPRINKLER SYSTEMS 110q 110t
.4.22 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 1 4.23 FIRE HOSE STATIONS 110v 4.74 PENETRATION FIRE BARRIERS 110w 4.25 REACTOR BUILDING PURGE FILTRATION SYSTEM 110x 4.25.
REACTOR BUILDING PURGE VALVES 1102 4.27 DECAY HEAT REMOVAL 110aa 4.28 EXPLOSIVE GAS MIXTURE 110bb 4.29 RADICACTI.VE EFFLUENTS.
110cc.
4.29.1 Radioactive Liouid Effluents 110cc-4.29.1.1 Concentration 110cc 4.29.1.2 Liquid Holdup' Tanks 110gg
'4.29.1.3 Liquid Radioactive Effluent Instrumentation 110hh 4.29.2 Radioactive Gaseous Effluents 4.29.2.1 Dose Rate-110jj 110jj 4.29.2.2 Gas Storage Tanks 110mm 4.29.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation 110nn 4.29.3 Dose Calculations for Radioactive Effluents 110rr 4.29.4 Solid Radioactive Waste 110rra 4.30 RADIOLOGICAL ENVIRONMENTAL MONITORING.
110ss 4.30.1 Radiological Environmental Monitorina Procram Description
~~
110ss 4.30.2 Land Use Census 11022 4.30.3 Interlaboratory Comparison Procram 110bbb 5.
DESIGN FEATURES 111 5.1 SITE 111 5.2 REACTOR BUILDING 112 5.3 REACTOR 114 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 116 6.
- ADMINISTRATIVE CONTROLS,
117 6.1 RESPONSIBILITY 117 6.2 ORGANIZATION 117 6.3 FACILITY STAFF QUALIFICATIONS 117 6.4 TRAINING 117 6.5 REVIEW AND AUDIT 117
- 6. 6 REPORTABLE EVENT ACTION 126 1
6.7 SAFETY LIMIT VIOLATION 126 6.8 PROCEDURES 127 6.9 RECORD RETENTION 128 6.10 RADIATION PROTECTION PROGRAM 129 6.11 HIGN RADIATION AREA 129 6.12 REPORTING REQUIREMENTS 140 6.13 ENVIRONMENTAL QUALIFICATI0N' 147 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 148 its Amendment No. 88,33,118
3.1.6 Leakage Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.2
)
If unidentified reactor coolant leakage (exceeding normal evaporative losses) exceeds 1 gpm or if any reactor coolanf leakage is evaluated as unsafe, the reactor shall be shut-down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.3 If it is determined that any reactor coolant leakage exists through a non-isolable fault in a reactor coolant system strength boundary (such as the reactor vessel, piping, valve body, etc,.
except steam generator. tubes), the reactor shall be shutdown and a cooldown to the cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.
3.1.6.4 Deleted 3.1.6.5 Action to evaluate the safety implication of reactor coolant leakage shall be ' initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection.
The nature, as well as the magnitude of the leak, shall be considered in this ev'aluation. The safety evaluation shall assure that the exposure of offsite personnel to radiation is within the guidelines of 10CFR20.
3.1.6.6 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, or 3.1.6.3 the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected.
3.1.6.7 When the reactor is at power operation, three reactor cool-ant leak detection systems of different operating principles shall be in operation.
One of these systems is sensitive to radioactivity and consists of a radioactive gas detector and an air particulate activity deb": tor.
Both of these instru-ments may be out-of-servfce simuhaneously for a period of no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided two other means are available to detect leakage and reactor building air samples are taken and analyzed in the laboratory at least once per shift 1 other-wise, be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.1.6.8 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which l
Amendment No. 57,118 27
- 5. d 4t t
s t
7 ne b
2.
e u
a n
n 2al e
e fb t
l o
e n1
,b ll z I
a u
ri h
i.
a b ai l
t rc ot w
h6yr ang l
se or a
t t e ror
.e h
i sl d
inip ei e sh et
,c yi e
wol on pt n t s l
se at v
i ini d
oie r
bn ur dn o
stbi d -
an ai t
e m
uaa s
0 od e t o rh ae 7v e
tcrsr 3
t ad si et th r
aiel u t
pi st ny tf peo i c y
si o n h i
rn pc ow oas me en h n l
cnn h
t r
ut on li t e l
eeia2 t
anu po t w b
i g a
l p
.h7 i
cio r
o am wr c
bSec ihh tp sd re e
i a
h n
dt n
l t et sm t
rothi ni4 ae eu ps ue a
et th iw2 l g nh oy t
m p
f ot t
oa ns s
am o
otobi edl ot a
o t o t
n w
t h ea cl hd tn so u
oaeh t vn o
cl o
r a
tustl oo rv o
si e
s ui e nmi or ec l t ll s
l raWv l
ret t e l
ea b o i
euc e
b uris cd bn nl ar np l
t d r an ai ni rt s
n e
erd u eu r
at en s
anh sn reao r
ed h n po a
h otuo w
h e
pn ce oc p
ci
,ti roe gh oa v
y s
at eph2 nt n
e oe B
esnt a h
t1 i
is r
r l y th li esi t
td r,
r e
e bc t
bmk d
)
i nng ue eu w
w an l
amaea l
d eain dl h o o
o re ef e
rotl r hi b
th p
p eg no r
eC b
t
,d t w sa pr n
u p.nat n
d eio d r e6 r
r oe an s
oeorn o
eswl ne r
o o
nm h o s
nhi ee h
C ro l
op ot t
t i e ci e
i ttpm
(
il do co t x c
c t
r con se a
a em ea p
eoa i
ucef en 1
q s
sis en e
e ho l r ht oa l
eeoe e
r r
r t o b e G
t ett b
1 rvl h 0 si e
r ap S
th n
l ct 3il
,h e
ro T
ert mo O
ro ec 5
na p
dt 0
5 rl e
avd n nt p e
1 4
oo pnn opgt 3
h ei aia rn t r oio g
t ense tkih h u ii o
o st nai i
s R i yh E
cft t c4 uh t
t en itt s
e nst i
L soii r1 qt ro na p
rli B
sl rw ri ei p
p c
eir al eg A
ebe ece rw u
u r
hae
'0 rith ne h
T l
vy t
t ue t mp 5 ecutiv t
amo nn t
t oh o
7 heo ro e
l sh ep erN aw a
a ht ed t p gob l tb p ra h o rnf w
i S
,nt a td d
d 1f oao o
e tii e
u gl n yt s ae t
e e
o t
l anrnd t
nf o t
s su s
s n
see e
)
eooe ainr osi sh s
s in etd b
1t vt mb h rne nt w es a
a h o rao (i es i
ceaw ir l
p p
ti im d
m ef r vco r
e t
y y i t
,t e
,b eroc e
p orh so b
b wa ein s
euh s
l re f ot l h r
l no s.
l st rse l
bove o
e e
e
,e bii ag b odd b
v dt nn b
b ep a
t pi a d gf o a
,lal ei,
ni l o rsa ys rnn hs resva sur a
y y
b erl bp eaiet n pr v
scu h e a
a an puu p
wl eo ouk aro cb m
m ri ooc e0 oeoumi ea nh r b5 nrld t
ocf pih fh oe yc f rs n
n pt i
i 7
ial eec i
o li b
1 oor o
o onn 6c y
poht a 4bl s
u i
i io l
e as l ef ca l
r2 e
esn rso t
t l ai etr md er sne e
er bai erh c
c emt nx e
nps rh b nh ph b u n
n n
a nee ne n
yd et mit c yyt mo0 u
u ndr anh oc a) ant uh i
abi uh3 f
f ane h
t ix h2dal o h
ntft m
w n
h ap ce t e c(
at iI a c
2g p
p c
o hn c
0s ew m
sid e7 n i
i e
eti ne er3 nen h
o l t e h
i r
r otf n
ur no at o t s
.r ear t nw t
t nao onm f u o
nl ai i o i
ie s
,i pi t ust nmo ht rc not hhl s
s ht e hht ss hsh ti h
t aue at s ttl i
i tid tt s ie t yt eid t
y h r i aih nd i
it ol h ue ii o h
h ino ii Wsh e C ar Wwf T
T Wim Wws T p Wdwti a W
2 3
4 5
6 7
8 9
0 1
1 1
1 1
1 1
1 1
2 2
>@a25 m*.
a*.
. k ';. G 3*
ll il i
3.5.5 Fire Detection Instrumentation Applicability This specification applies to fire detection instrumentation utilized within fire areas containing safety related equipment or circuitry for the purposes of protecting that safety related equipment or circuitry.
Objective To provide immediate notification of fires in areas where there exists a potential for a fire to disable safety related systems.
Specification 3.5.5.1 A minimum of 50% of the heat / smoke detectors in the locations specified in Table 3.5-5 shall be operable.
3.5.5.2 If less than 50% of the fire detectors in any of the locations designated in Table 3.5-5 are operable, within one hour esta-blish a fire watch patrol to inspect the zone (s) with the in-operable instrument (s) at least once per hour and restore the equipment to operable status within 14 days, or prepare and submit a Special Report to the Com' mission pursuant to Specification 6.12.5 l
within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to operable status. The provisions of
-Specifications 3.0.3 and 3.0.4 are not applicable.
Bases The various detectors provide alarms that notify the operators of the existence of a fire in its early stages thus providing early initiation of fire protection.
The detectors in the main and auxiliary control rooms also provide automatic fire protection initiation.
The detectors required to be operable in the various areas represent one half of those' installed.
Operability of the fire detection instrumentation ensures that operable warning capability is available for the prompt detection of fires.
This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment.
In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas (s) is required to provide detection capability until the inoperable instrumentation is restored to operability.
Amendment No. 25, 62, 57,118 53d 1
(-
~
3.5.6 Radioactive Liquid Effluent Instrumentation Applicability: During releases via this pathway.
Ob.iect0 =2 To provide instrumentation for radioactive liquid releases.
Specification:
3.5.6.1 The radioactive liquid effluent monitoring instrumentation shown in Table 3.5.6-1 shall be operable with their alarm / trip setpoints set to ensure that the limits of specification 3.25.1.1 are not exceeded.
3.5.6.2 With alarm / trip setpoints less conservative than required by the above specification, immediately suspend the release of radioactiya liquid effluents monitored by the affected channel, until the setpoint is changed to an acceptably conservative value.
3.5.6.3 With less than the minimum number of channels operable, take the action shown in Table 3.5.6-1.
Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Release Report why th'e inoperability was not corrected.
3.5.6.4 Specifications 3.0.3 and 3.0.4 are not applicable.
l Bases:
The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases.
The alarm / trip setpoints for these instruments shall be calculated in accordance with the methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
53f Amendment No. 88.118
3.5.7
' Radioactive Gaseous Effluent Instrumentation Applicability: As shown in Table 3.5.7-1.
Objective:
To provide instrumentation for radioactive gaseous releases.
Specification:
3.5.7.1 The radioactive gaseous effluent monitoring instrumentation shown in Table 3.5.7-1 shall be operable with their alarm / trip setpoints set to ensure that the limits of
- Specification 3.25.2.1 are not exceeded.
3.5.7.2 With a channel alarm / trip setpoint less conservative than required, declare the channel inoperable.
3.5.7.3 With less'-than the minimum number of channels operable, take the action shown in Table 3.5.7-1.
Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected.
3.5.7.4 Specifications 3.0.3 and 3.0.4 are not applicable.
Rases:
The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with methods in the ODCH to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.105.
l 531 Amendment No. 88,118 u________________---__
[
ez
'f
~
3.7.2 A.
The specifications in 3.7.1.may be modified to allow one of the following conditions to exist after the reactor has.been heated above 200F.
Except as indicated in the following conditions, if any of these. conditions are.not met, a hot shutdown shall be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the condition is not cleared with-in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be brought to cold shutdown within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.-
B.
In the event that one of the offsite power sources specified in 1
3.7.1.A (1 or 2) is inoperable, reactor operation may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the availability of the diesel generators is immediately verified.
C.
Either one of the two diesel generators may be inoperable for up to 7 days in any month provided that -during such 7 days the operability.of the remaining diesel generator is demonstrated immediately and daily thereafter, there are no inoperable ESF components associated with the operable diesel generator, and provided that the two sources of off-site power specified in 3.7.1.A(1) or 3.7.1.A(2) are available.
D.
Any 4160V, 480V, or 120V switchgear, load center, motor control center, or distribution panel in one of the two ESF distribution systems may be inoperable for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, provided that the operability of the diesel generator associated with the operable ESF distribution system is demonstrated immediately and all of the components of the operable d'stribution system are operable.
If the ESF distribution system is not returned to service at the end of the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, Specification 3.7.2.A shall apply.
E.
Two station battery chargers may be inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, after which Specification 3.7.2.A shall apply.
F.
One of the two station batteries and the associated distribution system may be. inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided that there are no inoperable safety related components associated with the remaining station battery which are redundant to the inoperable station battery and the operability of the diesel generator is verified
~
immediately.
If the battery is not returned to service at the end of the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, Specification 3.7.2.A shall apply.
G.
Two control power sources from the plant to the switchyard and the attendant distribution system may be inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, after whichSpecification3.7.J.Ashallapply.
H.
If the requirements of Specification 3.7.1.G cannot be met.
either:
(1) place all Startup Transformer No. 2 feeder breakers in
" pull-to-lock" within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, restore the inoperable interlocks to operable status within 30 days, or submit within 30 days a Special Report pursuant to Specification 6.12.5 outlining the cause of the failure, proposed l
corrective action and schedule for implementation; or (NOTE: The provisions of Specification 3.0.4 are not applicable.)
(2) apply the action requirements of Table 3.5.1-1, Note 14.
Amendment No. 55,118.
3.7.3 Deleted Bases The electrical system is de'igned to be electrically self-sufficient and s
provide adequate, reliable power sources for all electrical equipment during startup, normal operation, safe shutdown and handling of all emergency situations.
To prevent the concurrent loss of all auxiliary power, the various sources of power are independent of and isolated from each other.
In the ' event that the offsite power sources specified in 3.7.1. A (l'or 2) l are inoperable, the required capacity of one emergency storage tank plus one day tank (20,160 gallons) will be sufficient for not less than three and one-half days operation for one diesel generator loaded to full capacity.
(ANO-1 FSAR 8.2.2.3) The underground emergency storage tanks are gravity fed from the bulk storage tank and are normally full, while the day tanks are fed from transfer pumps khich are capable of being cross connected at their suction and discharges and automatically receive fuel oil when their inventory is less than 180 gallons.
Thus, at least a seven day total diesel oil inventory is available onsite for emergency diesel generator operation during complete loss of electric power conditions.
Technical Specification 3.7.2 allows for the temporary modification of the specifications in 3.7.1 provided that backup system (s) are operable with safe reactor operation and combined availability of the engineered safety features ensured.
Technical Specifications 3.7.1.F and 3.7.1.G provide assurance that the Startup Transformer No. 2 loads will not contribute to a sustained degraded grid voltage situation. This will protect ESF equipment from damage caused
'by sustained undervoltage.
i Amendment No. 75, 55,118
-57a-
1 1
3.12 MISCELLANEOUS RADI0 ACTIVE MATERIALS SOURCES Applicability Applies to byproduct.-source, and special nuclear radioactive material sources.
Objective To assure that leakage from byproduct, source and special nuclear radio-active material sources does not exceed allowable limits.
Specification 3.12.1 The source leakage test performed pursuant to Specification 4.14 shall be capable of detecting the presence of 0.005 pCi of radio-active material'on the test sample.
If the test reveals the pre--
sence of 0.005 pCi or more of removable contamination, it shall immediately be withdrawn from use, decontaminated and repaired, or be disposed of in accordance with Commission regulations. Sealed sources are exempt from such leak tests when the source contains 100 pCi or less of beta and/or gamma emitting material or 5 pCi or less of alpha emitting material.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
3.12.2 A Special Report shall be prepared and submitted to the Commis-sion pursuant to Specification 6.12.5 within 90 days if source l
I leakage tests reveal the presence of 20.005 microcuries of removable contamination.
3.12.3 A complete inventory of licensed radioactive materials in posses-sion shall be maintained current at all times.
4 9
Amendment No. 13, 38, 57,118 66b
t,
3.17' fire Suppression Water System Applicability This specification applies to the portions of the fire suppression water system necessary to provide fire protection to safety related equipment.
Objective To assure that fire suppression is available to safety related equipment.
Specification 3.17.1 The fire suppression water system shall be operable at all times with two high pressure pumps, each with a capacity of at least 2500 gpm, with their discharges aligned to the fire suppression header, and with an operable flow path capable of transferring water through distribution piping with operable sectionalizing control valves to the shutoff valve ahead of each hose standpipe and the water flow alarm device on each sprinkler system.
3.17.2 With one pump inoperable, restore the inoperable equipment to operable status within seven days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.12.5 l
.within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system.
3.17.3 With the fire suppression water system otherwise inoperable:
Establish a backup fire suppression water system within 24 a.
hours; and b.
If "a." above cannot be fulfilled, place the reactor in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The provisions of Specification 3.0.4 are not applicable.
Bases o
The fire pumps supply the only source of water for all fire suppression systems utilizing water.
full flow required for proper fire suppression water system operation The pumps start automatically on low system pressure.
In the event that the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.
1 Amendment No. JB, 57,118 66m i
l '.
3.18 FIRE SUPPRESSION SPRINKLER SYSTEM Applicability This specification applies to the following fire suppression sprinkler systems protecting safety related areas:
. Each of the four reactor building cable penetration areas.
a.
b.
Each of the four cable penetration rooms'.
Each of the two emergency diesel generator rooms.
c.
d.
Cable spreading room.
e.
Each of the two diesel generator fuel vaults.
f.
Hallway El 372 (Zone 98-J).
g.
Condensate demineralized area.*
Objective To assure that fire suppression is a,vailable to safety related equipment located in the above-listed areas.
Specification 3.18.1 The above-listed sprinkler systems shall be operable at all times.
3.18.2 hfith one or more of the above-listed sprinkler systems inoperable, establish a continuous fire watch (or operable smoke and/or heat detection equipment with control room alarm) with backup fire sup-pression equipment for the applicable area (s) within one hour.
Restore the system (s) to operable status within 14 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.12.5 within the next 30 days outlining the action l
taken, the cause of the inoperability and the plans and schedule for restoring the system (s) to operable status. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
Bases Safety related equipment located.in various areas is protected by sprinkler systems. The operability of these systems ensures that adequate fire sup-pression capability is available to confine and extinguish a fire occurring in the applicable areas.
In the event a system is inoperable, alternate backup fire fighting equipment or operable detection equipment is required to be made available untti the. inoperable equipment is restored to service.
- To be implemented no later than July 30, 1979.
)
Amendment No. 35, #3, 57, 118 66n lN
r__-__-
3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON. SYSTEMS Applicability
.)
\\
This specification applies to the Halon systems utilized as the fire j
suppression system for the control room and auxiliary control room.
1 Objective
{
To assure that fire suppression is available to the safety related equipment in the control room and auxiliary control room.
Specification' j
3.19.1 The three control room and auxiliary control room Halon systems shall be operable at all times with the storage tanks having at least 95% of full charge weight and 90% of full charge pressure.
3.19.2 With any of the control room and auxiliary control room Halon systems inoperable, establish backup fire suppression equipment for the affected area within one hour.
Restore the system (s) to operable status within 14 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.12.5 within l
30 days oulining the actions taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
Bases Safety related circuitry located in portions of the control room and auxiliary control room is protected by the Halon systems.
The operability of these systems ensures that adequate fire suppression capability is available to confine and extinguish a fire occurring in the control room or the auxiliary control room.
In the event that the system (s) is inoperable, alternate backup fire fighting equipment is required to be made available in the affected area until the inoperable equipment is restored to service.
Amendment No. 30, 57,118 660
i.
=.
1
'3.25 RADI0 ACTIVE EFFLUENTS 3.25.1-Radioactive Liouid Effluents 3.25.1.1 Concentration l
Applicability: At all times i
Objective:
To ensure that the limits of 10 CFR 20 are met.
Specifications:
l 3 25.1.1 A.
The concentration of radioactive material released to the discharge canal shall be limited to the concentration specified in 10 CFR Part 20, Appendix 8. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.
For dissolved or entrained noble gases, the total concentration released shall be limited to 2 x 10-4 pCi/ml.
B.
With the concentration of radioactive material released exceeding the above limits, immediately initiate action to restore concentration to within limits and provide notification to the Commission within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In lieu cf any other report, prepare and submit a Special Report within 30 days pursuant to Specification 6.12.5 C.
Specifications 3.0.3 and 3.0.4 are not applicable.
I Bases:
This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II.
This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II A design obfectives of Appendix I,10 CFR Part 50, to a member of the public.
The concentration Ifmit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP)
Publication 2.
66v Amendment No. 88118
Radioactive Liquid Effluents 3.25.1.2 Dose Applicability: At all times Objective:
To ensure that the dose limits of 10 CFR 50, Appendix I, Section IV A, are met.
Specifications:
3.25.1.2 A The dose commitment to a member of the public from radio-active material in liquid effluents released from AND-1 to the discharge canal shall be:
1)
During any calendar quarter less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and 2)
During any calendar year less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ.
B.
With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days, pursuant to Specification 6.12.5.
C.
The provisions of Specifications 3.0.3 and 3.0.4 are not l
applicable.
Bases:
Specification 3.25.1.2 provides assurance that releases of liquid effluents will result in concentrations far below the limits of 10CFR20. The specification provides the required operating flexibility and at the same time assures that the release of radioactive material in liquid effluents will be kept "as low as reasonably achtevable".
66w Amendment No. 88,118
l' Radioactive Liquid Effluents 3.25.1.3 Waste Treatment
)
_ Applicability: At all times J
Objective:
To assure that the amount of radioactive material in liquid effluents will be "as low as reasonably achievable."
Specifications.
b.25.1.3 A.
The appropriate parts of the liquid radwaste treatment system shall-be used to reduce the radioactive materials in liquid waste prior to their discharge when it is projected that the cumulative dose during a calendar quarter due to liquid I
effluent releases would exceed 0.18 arem to the total body or 0.625 mrem'to any organ.
B.
The provisions of this specification do not apply to the laundry tanks due to their incompatibility with the radwasie system.
C.
With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report, prepare' and submit a Special Report to ths Commission within 30 days per Specification 6.12.5.
D.
The provisions of Specifications 3.0.3 and 3.0.4 are not l
applicable.
Bases:
The requirements that the appropriate portions of this system de used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Seccion II A of Appendix I,10 CFR Part 50, for liquid effluents.
The values of 0.18 mrem and,0.625 mrem are approximately 25% of the yearly design objectives on a quarterly basis. The yearly design objectives are given in 10 CFR 50, Appendix I, Section II.
l 66x Amendment No. 88,118
3.25.2 Radioactive Gaseous Effluents 3.25.2.1 Dose Rate Applicability: At all times Obiective:
To ensure that the dose rate in unrestricted areas from gaseous effluents will be within the limits of 10 CFR 20.
l Specifications:
3.25.2.1 A.
The dose rate in unrestricted areas (see Figure 5.1-1) due to radioactive materials released in gaseous affluents from the i
site shall be:
1)
For noble gases:
Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.
2)
For iodine-131, for tritium and for all radionuclides in particulate form with half lives greater than 8 days:
Less than or equal to 1500 mrem /yr to any organ.
During periods of reactor building purging the dose rate may be averaged over a one, hour interval.
B.
With the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s)
C.
Specifications 3.0.3 and 3.0.4 are not applicable.
I Bases:
This specification is provided to ensure that, at any time, the dose rate due to gaseous effluents from all units on the site will be within the limits of 10 CFR 20 for unrestricted areas.
This specification apolies to the release of gaseous effluents from all reactcrs et the site.
o I
66z Amendment No. 33,118
4 Radioactive Gaseous Effluents 3.25.2.2 Dose - Noble Gases Applicability: At all times Objective:
To ensure that the design objective doses of 10 CFR 50, Appendix I, Section IV A, are not exceeded.
Specifications:
3.25.2.2. A.
The dose due to noble gases released in gaseous effluents from ANO-1 to unrestricted areas (see Figure 5.1-1) shall be:
1)
During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads' for beta radiation, and 2)
During any calendar year, less than or equal to 10 meads for gamma radiation and less than or equal to 20 mrads for beta radiation.
B.
With the calculated dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days, pursuant to Specification 6.12.5.
C.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
I Bases:
Specification 3.25.2.2 implements the design guides specified in 10 CFR 50, Appendix I, Section II, and the limiting condition for operation as set forth in Section IV A of Appendix I.
The specifications provide the required operating flexibility and at the same time implement the guides set forth in Section IV A. Appendix I, to assure that the releases of radioactive material ir. gaseous effluents will be kept "t.s 12 as is reasonably achievable."
These limits provide reasonable assurance that radioactive material dis-charged in gaseous effluents will not result in the exposure of an indi-vidual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFR Part 20 [10 CFR Part 20.106(b)].
For individuals who may at times be within the exclusion area boundary, the occupancy of the individual will be sufficiently low to i
compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.
i 66aa Amendment No. 3g,118 I
o
Radioactive Gaseous Effluents 3.25.2.3 Dose - Iodine-131, Tritium, and Radionuclides in Part,1culate Form Applicability: At all times Obiective:
To ensure thst the dose limits of 10 CFR 50, Appendix I, Section IV A, are met.
Specifications:
3.25.2.3 A.
The dose to a member of the public from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from AND-1 to unrestricted areas (see Figure 5.1-1) shall be:
1)
During any calendar quarter, less than or equal to 7.5 mrems to any organ, and 2)
During any calendar year, less than or equal to 15 mrems to any organ.
B.
With the calculated dose from the release of iodine-131, tritium and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special. Report to the Commission within 30 days, pursuant to Specification 6.12.5.
C.
The provisions of Specifications 3.0.3 and 3.0.4 are not I
applicable.
Bases:
Specification 3.25.2.3 implements the design guides set forth in 10 CFR 50, l
Appendix I, Section 11 C, and the limiting conditions for operation as set forth in Appendix I,Section IV A.
The specifications provide the required ' operating flexibility and at the same time implement the guides set forth in Section IV A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable".
66bb Amendment No. 88,118
Radioactive Gaseous Effluents 3.25.2.4 Gaseous Radwaste Treatment Applicability: At all times Objective-To assure that the amount of radioactive material in gaseous effluents is "as low as reasonably achievable."
Specifications:
3.25.2.4 A.
Ventilation exhaust treatment systems shall be used to reduce radioactive materials in gaseous waste prior to discharge when the projected doses due to gaseous effluent releases from AND-1 to unrestricted areas (see Figure 5.1-1) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation over a calendar quarter; or when the projected doses due to iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days would exceed 1.0 stem to any organ over a calendar quarter.
B.
When degasifying the reactor coolant system, the gaseous radwaste treatment system shall be utilized to process the degassing effluent to reduce the concentration of radioactive materials prior to discharge when the projected doses due to gaseous effluent releases from AND-1 to unrestricted areas (see Figure 5.1-1) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation over a calendar quarter.
C.
With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report, prepare and submit to the Commission within 30 days a Special Report, per Specification 6.12.5.
D.
The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
l Bases:
The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable."
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II B and II C of Appendix I,10 CFR Part 50, for gaseous effluents.
The values 0.625 mrad,1.25 mrad, and 1.0 arem are approximately 25% of the yearly design objectives on a quarterly basis. The yearly design objectives are given in Specifications 3.25.2.2 and 3.25.2.3.
66cc Amendment No. 88.118
3.25.4 Solid Radioactive Waste Applicability: At all times Objective:
To ensure solid radwaste is processed in accordance with the Process Control Program to meet shipping and burial ground requirements.
Specifications:
3.25.4.1 With the provisions of the Process Control Program not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive waste from the site.
3.25 4.2 The provisions of Specifications 3.0.3 and 3.0.4 are not l
applicable.
Bases:
This specification implements the requirements of 10CFR50.36a and General Design Criterion 60 of Appendix A to 10CFR50.
l l
66ff Amendment No. 88,118
l #
Should the inspection of one of the wires reveal any significant physical change (pitting or loss of area), additional wires shall be removed from the applicable surveillance tendons and inspected to determine the extent and cause change. The sheathing filler-will be sampled and inspected for changes in physical appearance.
(See Applicable Acceptance Criteria in Section 4.4.2.1.3) 4.4.2.1.3 Acceptance Criteria-The Reactor Building Post Tensioning System shall be considered acceptable if the following acceptance criteria are met.
1.
Each surveillance tendon has a normalized left-off force equaling or exceeding its expected prestress force for the time of the test.
See Figures 4.4.2-1,
-2, and -3.
If the normalized lift-off force of any one tendon in a group lies between the expected prestress force and the lower bound prestress force, an adjacent tendon on each side shall be checked for lift-off force.
If both of these tendons are found acceptable, the surveillance program may proceed considering the single deficiency as unique and acceptable.
If either of the adjacent tendons is found unacceptable,.it shall be considered evidence of possible abnormal degradation of the containment structure.
(See TS 6.12.5) l If the normalized lift-off force of any single tendon lies below the lower bound prestress force, the occurrence should be considered evidence of possible abnormal degradation of the containment structure.
(See TS 6.12.5) l 2.
The wires removed from three detensioned surveillance tendons (one dome, one vertical and one hoop) shall be inspected for corrosion, cracks, or other damage over-the entire length of the wire. The presence of abnormal corrosion, cracks, or other danage shall be considered evidence of possible abnormal degradation of the containment structure.
(See TS , 6.12.5)
A minimum of three (3) wire samples cut from each removed wire (one from each end and one at mid length) shall be subjected to a tensile test.
Failure of any one of these wire samples to meet a minimum ultimate tensile strength of 240 ksi shall be considered evidence of possible abnormal degradation of the containment struct.ure.
(See TS 6.12.5) l 3.
Sheathing Filler material samples from each surveillance shall be considered acceptable provided the results of the tests performed on the sanples fall within the following limits.
1.
Water Soluble Chlorides less than 10 ppm 2.
Water Soluble Nitrates less than 10 ppm 3.
Water Soluble Sulfides less than 10 ppm 4.
Water Cantent less than 10% Dry Weight Amendment No. 86,118 85a
1 f
4.4.2.2 ~ Inspection Intervals and Reports i
The inspection intervals, measured from the date of the. initial structural test, shall be one year, three years,-five year's, and every five years thereafter.
Tendon surveillance may be conducted during reactor operation provided design conditions regarding loss j
of adjacent tendons are satisfied at all times.
A quantitative, analytical Special Report covering results of each inspection shall be submitted within 90 days after completion of the tests and inspections, pursuant to Technical Specification 6.12.5, and shall especially address the following conditions, should they develop:
(1) Broken wires.
(2) The force-time trend line for any tendon, when extrapolated, that extends beyond either the upper or lower bounds of the predicted design band.
(3) Unexpected changes in tendon conditions or sheathing filler properties.
l 4.4.2.3 End Anchorage Concrete Surveillance The end anchor /ges of the surveillance tendons and adjacent a.
concrete surface will be inspected.
b.
The inspection interval will be one-half year and one year after the structural integrity test.
The selected inspection location shall include:
c.
(1) Four (4) locations on one buttress (hoop tendon anchorage).
~
(2) Two (2) locations on the top of the ring girder (vertical tendon anchorage).
(3) One (1) location on#the ring girder (dome tendon anchorage).
Amendment No. S. BB,118 86
cr?1 -
4.18.6 Reports Following each inservice inspection of steam generator tubes, the complete results of the inspection shall be reported to the.NRC. This report, to be submitted within 45 days of inspection completion, shall include:
a.
Number and extent of tubes inspected; b.
Location and percent of wall-thickness penetration for each indication of an imperfection; and c.
Identification of tubes plugged and tubes sleeved.
This report shall be in addition to a Special Report (per Specification 6.12.5) required for the results of steam generator tube inspections which fall into Category C-3 as denoted in Table 4.18-2.
The Commission shall be notified of the results of steam generator tube inspections which fall into Category C-3 prior to resumption of plant operation.
Bases The surveillance requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the 'RCS will be maintained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to malatain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
1 Amendment No. 24, 41, 86, 205, 118 110n
I i
vb of usd T
u euioe s
i C
q ett l
eh v E
e l
ct p rt o P
R seal m A
A A
A or S
A A
v ua
/
/
/
/
2bp N
n
/
/
rimss N
N N
N I
o N
N ot re Crd i
cort m
on E
t egef s
a r
a L
c nuf r3r e
oe P
A ol eei t
Fnn M
MPdPCf s
oo A
i f
nt Sll lIlllillllll lIl lll llllllliII o
sic D
t t
a R
l r
l s 3
u e
utl s
b sl a e
m eui R
A A12 3
A A
A A
u rsd
/
/- -
/
/
/
/
n ee N
NCC C
N N
N N
1rm e
e h
C3r ll lllll lllllll lll lll lIllllll1l ll llllIllllllI lll lll lIlIllIi1l t
rCf s
o o
5 C
i fr 4
R ol d
Nf n
dFa r
r o
e v
3 nl a a G. fo o
s o
d r.or n
f ed
.tl n
i N
d 2
O e
sos bn s
an a
usp uaeetvo qep i N
I r
eei n
nf O
T i
vbt sof ooet vbroi eba I
C u
euiiioe i
l
.euort t
ru T
E q
etdht l
tspl G. l et ppc i
tC C
P e
l dt ct p e
ct ml epa n
s R
E S
R e
sea al m n
al aaSseR a u
i eN P
N n
v n
ua o
us v
l v
S I
n o
riti mss N
ms th ril C a e
nit 2
N o
N ot c re red ccotaRi h
ot s I
E i
cesort ornea ciNd t
i ce 8
L t
gepef s
f opegec e
t eu 1
E P
c A
uf sb r3 r r2 cs uf ed m n
cf q B
M A
/
l enue i
e-ennl epne i
aee 4
U A
N PditPCf PCsIi pdS ar dr T
S sno E
Il ll lill[l l lll lil lIl lllllllI1l roNed L
R D.
oi vn B
O N
tt ea A
T 2
t s
s s
ac: e T
A l
i i
i re3l s R
u r
r r
ep se n s a. r u b
E s
A l
2 3
e e.
e h G.1 h G.2 h G.3 en N
e
/
E R
N C
C C
t
- t
- t gi 3 o t 0SC0SCoSC G
mn8ge ll ll llllllllll lll lll lllllllllI aa1 uv M
ll ll llllllllll lil lll lllllllIlI A
d e
.li E
d nl t g4 pt T
e aa sg-
- n sn c
S r
ns eut ni iose i
soi bl ci t
f rt rf u
eeih upe s
t n ou oe q
vbtt t
fd e ra dttd e
eui
.enb ou r
na G.d a u ps edare R
etd n l
er5 b euev l
di l
t R u4 mt sne G.
n sea aS es p
ucree o
v s
ve5 N
i rit e tseb2S l
2 n e u gl O
t e
ot cb ci e u aC1 pp s
I c
n cau ehl tt r iR es m
pt s ce cN6 hndar T
A o
gept s
eeh e
ti eeo G.
C M
uf s n n r v pt poo tt E
l en5 P
Pdi25 I i oiso Stt sscsg S
i re u
o phl tt Nl illllIIllllll ll1lllllllllllll I
t Nt stpr l
ano o
E u
erib
,p L
s re seh P
e 1
2 3
ensrrRt M
R h eeoo A
C C
C W gb tl e ueaar Sl ill lIilllllllllllllllllllllII ht nrio oecm T
S e
r 3
r ne I
z f e
=
onepo op S
Fi gSN i
S 1
2 ms e
ue l
mbt S
p i u E
m n T G.
i T
a S
m5S O
N A
Il lllll lllll lll ll ll llll llllllIl
>2 E% ?. zP ~*. w". m*. w@L5 g P_
e 4.30
' RADIOLOGICAL ENVIRONMENTAL MONITORING 4.30.1-Radiological Environmental Monitoring Procram Description Applicability: Applies at all times.
Objective:
To provide information on.the radiological affects of station operation on the environment.
t b ecifications:
l 4.30.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 4.30-1 and shall be analyzed pursuant to the requirements of Tables 4.30-1 and 4.30.2.
The sample locations shall be shown in table 4-1 in the ODCH.
4.30.1.2 a.
With the radiological environmental monitoring program not being conducted as ispecified in Table 4.30-1, prepare and submit to the Commission in the Annual Radiological Environmental Report a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
(Deviations are permitted from the required sampling schedule if specimens are not obtainable due to hazardous conditions, seasonal unavailability, or to malfunction of sampling equipment.
If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.)
With the level of radioactivity as h.he result' of plan b.
effluents in an environmental sampling medium at one or more of the locations specified in Table 4.30-1 exceeding the limits of Table.4.30-3 when averaged over any calendar 3
i quarter, prepare and submit to the Commission, within 30 days from the end of the affected quarter, a Special Report which l
includes an evaluation of any release conditions, environmental factors or other aspects which caused the
, limits of Table 4.30-3 to be exceeded, and defines the actions taken to reduce radioactive effluents so that the potential annual dose to's member of the public is less than the calendar year limits of Specifications 3.25.1.2 and 3.25.2.2.
When more than one of the radionuclides in Table 4.30-3 are detected in the sampling medium, this Special l
Report shall be submitted if:
1 Concentration (1) reporting level (1)
- Concentration (2) reporting level (2) * *" 11.0 When radionuclides other than those in Table 4.30-3 are detected and are the result of plant effluents, this Special l
Report shall be submitted if the potential annual dose to a member of the public is equal to or greater than the calendar
)
110ss Amendment No. 88.118 j
O,.
-year limits of Specifications 3.25.1.2 and 3.25.2.2. 'This Special Report is not required if the measured level of radioactivity was not the result of plant affluents, however, l
l-in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.
With allk or fresh leafy vegetable samples unavailable from c.
any of the-sample locations required by Table 4.29-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.
The specific locations from which samples were unavailable may then be deleted from the monitoring program.
Identify the causes of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised table for the ODCM reflecting the new location (s).
d.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
4.30.1.3 The results of analyses performed on the radiological environmental monitoring samples shall be summarized-in the Annual Radiological Environmental Report.
Bases:
-The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for.those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation.
This monitoring program thereby supplements the radiological effluents monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected
- on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The initially specified monitoring program will be effective for at least the first three years of ::ommercial operation.-
Following this period, program changes may be initiated based on operational experience.
)
I i
The detection capabilities required by Table 4.30-2 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.
1 1
110tt Amendment No. 58,118 l
L_____-____-__-_____-----
r=-
=
~
o J
-J Radiological Environmental Monitoring j
4.30.2 Land Use census i
_ Applicability: Applies at all times
.l Objectives:
This specification will identify changes in use of the unrestricted areas.
p Specifications:
4.30.2.1 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles from the ANO-1 reactor building.
4.30.2.2 The land use census shall be conducted at least once per 12 months.
between the dates of June 1 and October 1, by door-to-door survey, aerial survey, or by consulting local agricultural authorities.
4.30.2.3 a.
With a land use census identifying a location (s) which yields a calculated dose commitment due to I-131, tritium, and radionuclides in particulate form greater than the values currently being calculated in Unit 1 Specification 4.29.3 and Unit 2 Specification 4.11.2.3 submit location description in the Semiannual Radioactive Effluent Release Report per Specification 6.12.2.6.
b.
With a land use census identifying a location (s) which yields a calculated dose commitment (via the same exposure pathway) greater than at a location from which samples are currently being obtained in accordance with Specification 4.30.1.1, identify the new location in the Semiannual Radioactive Effluent Release Report per Specification 6.12.2.6.
The new^
location shall be added to the radiological environmental
, monitoring program within 30 days, if possible.. The sampling location having the lowest calculated dose commitment (via the same exposure pathway) may be deleted from this monitoring program after' October 31 of the year in which this 1and use census was conducted.
4.30.2.4 The results of the land use census shall be included in the Annual Radiological Environmental Report.
l t
4.30.2.5 The provisions of Specifications 3.0.3 and 3.0.4 are not I
applicable.
- Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.
1 110Zz Amendment No. 88,118 o
4 Radiological Environmental Monitoring 4.30.3 Interlaboratorv Comparison Program Applicability: Applies to the off-site radiochemistry laboratory.
Objective:
To provide independent checks on the accuracy of the measurements of radioactive material in environmental samples.
Specifications:
4.30.3.1 Analyses shall be performed on radioactive materials supplied as part of Interlaboratory Comparison Program which has been approved by NRC.
4.30.3.2 With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report.
4.30.3.3 The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report pursuant to Specification 6.12.2.5.
4.30.3.4 The provisions of Specifications 3.0.3 and 3.0.4 are not I
applicable.
Bases:
The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental I
monitoring in order to demonstrate that the results are reasonably valid.
I i
{
l 110bbb Amendment No. 88,118
4 9 -
ALTERNATES 6.5.1.3 All alternate membe'rs shall be appointed in writing by the PSC i
Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PSC activities at any one time.
MEETING FREQUENCY 6.5.1.4 The PSC shall meet at least once per calendar month and as convened by the PSC Chairman or his designated alternate.
QUORUM
~
6.5.1.5 The minimum quorum of the PSC necessary for the performance of the PSC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES
- 6. 5.1. 6 The Plant Safety Committee shall be responsible for; a.
Review of 1) all procedures required by Specification 6.8 and revisions thereto, 2) any other proposed procedures or revisions thereto as determined by the Executive Director, ANO Site Operations to affect nuclear safety.
b.
Review of all proposed tests and experiments that affect nuclear safety.
c.
Review of all proposed changes to the Appendix "A" Technical Specifications.
I d.
Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e.
Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Executive Director, ANO Site Operations and to the Chairman of the Safety Review Committee.
f.
Review of REPORTABLE EVENTS.
Amendment No. 25, 99,109, IZ2,118 121
REVIEW 6.5.2.7
'The SRC shall review:
The safety evaluations for 1) changes to procedures, -
a.
equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which
~
involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
Proposed tests or experiments which involve an unreviewed c.
safety question as defined in Section' 50.59, 10 CFR.
d.
Proposed changes in Technical Specifications or licenses.
Violations of applicable statutes, codes, regulations, e.
orders, Technical. specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviation from normal and expected performance of unit equipment that affect nuclear safety.
g.
All REPORTABLE EVENTS pursuant to Specification 6.6.2.
l h.
All recognized indications of an unanticipated deficiency in some aspects of design or operation of structures, systems, or components that could af_fect nuclear safety.
1.
Reports and meeting minutes of the Plant Safety Committee.
j.
Changes to Offsite Dose Calculation Manual and Process Control Program.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the SRC. These audits shall encompass:
i The conformance of unit operation to provisions contained a.
within the Technical Specifications and applicable license conditions at least once per 12 months.
b.
The performance, training, and qualifications of the entire unit staff at least once per 12 months.
The result of actions taken to correct deficiencies occurring c.
in unit equipment, structures, systems or methods of operation that affect nuclear safety at least once per 6 i
months.
Amendment No. If, 39, #7, 88, 124 39, 118 u--______________-_-_----._._--
- 6. 6 -
REPORTABLE EVENT ACTION 6.6.1-A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
6.6.2 The following actfons shall be taken for REPORTABLE EVENTS:
a.
A report shall be submitted to the Commission pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and-b.
Each REPORTABLE EVENT shall be reviewed by the PSC, and the results of this review shall be submitted to the SRC and the Vice President, Nuclear Operations by the Director, Site Nuclear Operations.
6.7 SAFETY LIMIT VIOLATION 6.7.1-The following actions shall be taken in the event a Safety Limit is violated:
The facility shall be placed in at least hot shutdown within one a.
hour.
b.
The Nuclear Regulatory Commission shall be notified pursuant to 10 CFR 50.72 and a report submitted pursuant to the requirements of 10 CFR 50.36 and Specification 6.6.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below; The applicable procedures recommended in Appendix "A" of a.
Regulatory Guide 1.33, November,1972.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementation.
~
f.
Fire Protection Program implementation.
g.
New and spent fuel storage, h.
Offsite Dose Calculation Manual and Process Control Program implementation at the site.
i.
Post accident sampling (includes sampling of reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and the containment atmosphere).
Amendment No. I6, 30, 34, 37, #2, 126
- 7, 58, 82, if,118
d'
,l' 6.9 RECORD RETENTION 6.9.1 The following records shall be retained for at least five years:
Records and logs of facility operation covering time interval a.
at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
REPORTABLE EVENTS.
d.
Records of surveillance activities, inspections and calibrations required by the Appendix A Technical Specifications.
e.
Records of reactor tests and experiments.
f.
Records of changes made to procedures required by Specification 6.8.1.
g.
Records of radioactive shipments.
h.
Test results, in units of microcuries, for leak tests performed on licensed sealed sources.
i.
Results of annual physical inventory, verifying accountability of licensed sources on record.
6.9.2 The following records shall be retained for the duration of the Facility Operating License:
Records and drawing changes reflecting facility design a.
modifications made to systems and equipment described in the' Final Safety Analysis Report.
- b.
- Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories, Records of facility radiation and contamination surveys.
c.
d.
Records of radiation exposure for all individuals entering radiation control areas.
e.
Records of gaseous and liquid radioactive material released to the environs.
f.
Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.
g.
Records of training and qualification for current members of the plant staff.
Amendment No.16, Jg,118 128 L______-_-_______-_____
1 LI..
.e 6.12.3 Delited.
6.12.3.1 Deleted
\\_
6.12.3.2 Deleted l
\\-
1.
(Page Intentionally left blank)
Amendment No. f#, 29, 88,118 142
i 4
1
\\
l I
I l
(page intentionally left blank)
- 143 -
Amendment No. 5,118
-_-__- -,--=;
- f. 4
..p.
s g_
.-n.
E 1
,,s e;..,
J
. f.
k s.
.l 4 ;
- l..
(page intentionally'left blank) f.
Amendment No. 5, 47, 32,118
-144
._-___--___.--_______-__a
J.h.J i j ';'1'.)
r yg.
j., 9 (page intentionally left blank)
)
i i
Amendment No. 5, 88, 118 145 a
___u______________
i g:
4:-
4',
s s
6.12.4-Deleted l
L i, (page intentionally left blank).
Amendment No. 5, 52, N,118
-146-
______._______-_-_____.-------__m-
a;
.+
6.12.5 Special Reports Special reports shall.be submitted to the Administrator of the appropriate Regional Office within the time period specified for each report.
These reports 'shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.
Tendon Surveillance. Specification 4.4.2.2 ~
a.
.b.
Inoperable Containment Radiation Monitors, Specification 3.5.1 Table 3.5.1-1.
Radioactive Effluents Specifications 3.25.1, 3.25.2, 3.25.3, and c.
3.25.4.
This report shall include the following:
1)
Description of occurrence.
2)
Identify the cause(s) for exceeding the limit (s) 3)
Explain corrective action (s) taken to mitigate occurrence.
4)
Define action (s) taken to prevent recurrence.
-5)
Summary of consequence (s) of occurrence.
6)
Describe levels exceeding 40CFR190 in accordanc.e with 10CFR20.405(c), as applicable.
d.
Steam Generator Tubing Surveillance. Category C-3 Results, Specification 4.18.
Miscellaneous Radioactive Materials Source Leakage Tests, Specification e.
3.12.2.
f.
Radiological Environmental Monitoring Sample Analysis, Specification 4.30.1.
An ur. planned offsite release during any one hour period of 1) more than g.
1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall be submitted within 30 days of the occurrence and shall include the following information:
1.
A description of the event and equipment involved.
2.
Cause(s) for the unplanned release.
3.
Actions taken to prevent recurrence 4.
Consequences of the unplanned release.
146a Amendment No. 55,118
1
_.e.,
e I
h; I'noperable Fire Detection Instrumentation, Specification 3.5.5.2.
t 1.
Inoperable Fire Suppression Systems.-Specifications 3.17.2, 3.18.2, and 3.19.2.
j DegradedAuxiliaryElectricalSystems, Specification 3.7.2.H.
4 146b Amendment No.118
- - _ _ - - _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _. _ _ _ _