ML17227A566
| ML17227A566 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/21/1992 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17227A565 | List: |
| References | |
| NUDOCS 9208260009 | |
| Download: ML17227A566 (62) | |
Text
ATTACHMENT 1 Marked-up St. Lucie Unit 1 Technical Specifications page:
XXI of the Index 3/4 0-3 3/4 1-12 3/4 3-41 3/4 4-1b 3/4 4-8 3/4 4-14 3/4 6-5 3/4 6-20 3/4 6-21 3/4 8-6b 3/4 11-2 3/4 11-10 3/4 12-10 B 3/4 1-4 B 3/4 7-5 B 3/4 9-3 6-24 92082b0009 920821 PDR ADOCK 05000335 P,
il e ~
1 I
~
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMEHTS SECTION PAGE 3 4.0 APPLICABILITY...~......................'.....
3/4 0-1 3.4'.1 REACTHfITY CONTROL SYSTEMS 3/4.1.1 GORATION CONTROL'.
Shutdown Margin - T--
~ 200'F.'.....................
Shutdown. Margin - T
~ 200'F.....................
Boro ren 0$ 1uQOn'ooeoooooooeoeeoeoo
~ oeooooo
~ oeooooooeo nderator-Temperature: Coefficient..................
8fnimum Temperature for CrktfcaTjty........
3/4 1-1 3/4 1-1 3(4 1-3'/4 1-4 3/4 1-5 3'-1-7 3'/4 To2; HORATIO@: SYSTEMS.
Ffoe Pa'ths' Shutdem........
8tof PathS OperaQngeeeeoeeeeeo eo eeoc eoooeo eeoc coo e Cherg7ng P
S Shntdawe.....'..'hergIng Pumpe - OperntIng....'.....................
Borfe Acfd.. Pumps: - Shutdeer...................
Eorfe Aci4 Pumps, - Operating;....................
Berated. Mater Sources - Shutdowr.................
Borated Mater Sources. - Operating...................
3/4 1-8 3/4 1-8 we I-Io A>4 3/4 1-12.
3(4 1-13 3(4 1-14'/4 1-15 3/4 1-16 3/4 1-4 3/4.T.3 MOVABLE CONTROL ASSEMBLIES --.~......
Foll Length CEA Posftfon '..........................
Posftjon; Indicator Channels......................
EA, DrOp Tjttlee e o o ~ e o o ~ ~ e ~ o ~ o o o o o o o o ~ o o o ~ ~ o o e o e o o o e o ~ o C
Shutdee CEA Insertion Limit..;........,..............
Pegulatlng CEA Insertjon Limits;........'.............
3/4 1-20 3/4 1-20 3/4 1-24 3/4 1-26 3(4 1-27 3/4 1-28 ST. LUCIE - UNIT 1 Amendment No37
APPLICABILITY SURVEILLANCE RE UIREHENTS (Continued) 4.0.5 (Continued)
ASNE Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice inspection and testin activities Required frequencies for performing inservice inspection and testing activities Meekly Monthly Quarterly or every 3 months Semiannually or every 6 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 days C ~
The provisions of Specification 4.0e2 are applicable to the above required frequencies for performing inservice inspection and testing (Q activities.
d.
e.
Performance of the above inservice inspection and testing activi ties shall be in addition to other specified Surveillance Requirement
~
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
ST.
LUCIE - UNIT 1
3/4 0~3 Amendment Ho.
90
~ g A
l 4
(
4
~ QI~%
REACTIVITY CONTROL SYSTEMS CHARGING PUM 8 SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or one high pressure safety injection pump* in the boron injection flow path required OPERABLE pursuant to Specifi-cation 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.
APPLICABILITY:
MODES 5 and 6.
ACTION:
With no charging pump or high pressure safety injection pump+OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status.
SURVEILLANCE RE UIREMENTS 4.1.2.3 At least one of the above required pumps shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or.,the high pressure safety injection pump develops a total head of greater than or equal to 2571 ft. when tested pursuant to Specification 4.0.5.
(a) the RCS pressure boundary does not exist, or (b) no charging pumps are operable.
In this case, all charging pumps shall be disabled and heatup and cooldown rates shall be limited in accordance with Fig. 3.1-lb.
HCV-3616 HCV-3626 HCV-3636 HCV-3646 At RCS temperatures below 115'F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:
Hi h Pressure Header AUxiliar Header HCV-361 HCV-3627 HCV-3637 HCV-3647 ST.
LUCIE - UNIT 1
3/4 1-12 Amendment No. 5 O~Nf, gg, 104.
0 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitorin instrumentation channels shown in Table 3.3-11 shall be OPERABLE,
boo APPLICABILITY:
MODES 1, 2, and 3.
ACTION:
a.
Actions per Table 3.3-11.
b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE UIREMEWTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.
-ST LUCIE - UNIT 1
3/4 3-41 Amendment No. 37
0 REACTOR COOLANT SYSTEt)
HOT SHUTDOWN LIMITING CONDITION FOR OPERATION
,3.4.1.3 At 1east two of the loops, listed-'belc'~ shall be: OPERABLE and at least one reactor coolant or shutdown cooling loop shall be. in operation.<<
a Reactor Coo1ant Loop A and its assocfated steam generator and a<
1'east: one associated reactor-coolant pump b
Reactor Coolant'Loop 3 and. fts assocfated steam gener ator Teast one assocfateL reactor coolant pum.
D~W~e.
/g c
Shutdown Coo1fng, Loop A d
Shutdown Cool fng Loop 8;
\\
APPLICABILITY NODE 4 ACTION gD P-E,p&~ QY7H
~/
a..*
Mith. Tess than the-above requfred: re ator coolant. or shutdown coolf~ Toops OPERABLE within. one (1) hour fnftfate correctfve actfoa to return the requfret. loops. to OPERABLE status. If the y'enafnfng OPERABL'F Toop-fs. a, shutdown cooling loop be.fn COLD GitQTIN wfthfa 30'ours.
fifth no reactor coolant or." shutdown: coolfng. Toop fn'peration.
uspenC alT operatfons'nvolvfng, a rsiuctfoa. fn boron concentra-Lfoe of the Reactor Coolant System and wfthfn one (1) hour initiate corrective action.- to.return the r equired coolant loop to'peration.
<<ATT reactor coolant pumps. and shutdown cool fng pumps. may be de-energized for
~
up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1). no: operations are permitted Rat would cause dilu-tion of the Peactor Coolant Systia boron concentratforr.
and (2) core outlet taaperature is maintained at'least. 10.'F below saturatfon temperature.
ST. LUCIE - UNIT 1 3/4 4>>1b Amendment No. 5 6.
of Sse.am Fo MK COmPLCT>>~F EACH lesKpQicE. j45p&qlud SVc THlsE5 DkE NQ~9E~ Q'p ~+~ p~ +graf) j~
%P~
~~~
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y4, ~~
<o~W 5 bN l&5lbN lN 1
<< ~<<~m Puu.sHAhrT V~ 5PMivl~ir~ I',.9.g REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Contfnued 5.
Defect means an imperfection of such severity that it exceeds tepTuggtng Ikmkt.
A tuba containing a da1'act
$ s dai'acttva Any tube which does not permit the passage of the eddy-current inspection probe shall be doomed a defective tube.
6.
PIu fn Limit means the imperfection depth at or beyond w
c t e tu e shall be removed from servfce because ft may become unserviceable prior to the next fnspectfon and is equal to 44%
of the nomfnal tube wall thickness.
7.
Unserviceable describes the condition of a tube ff it leaks
~tat tIra ght title I
t fntegrfty fn the event of an Operatfng Basfs Earthquake, a
loss-of-coolant accident, or a steam lfne or feedwater Ifne break as specfffed fn 4.4.5.3.c, above.
8.
Tube Ins ection means an inspectfon of the steam generator u e rom t e point of entry (hot Ieg sfde) completely around the U-bend to the top support of the cold leg.
b.
The steam generator shall.
be determfned OPERASLK after completing the cor respondfng actfons (plug all tubes exceeding the pIuggfng Ifmft and all tubes contafnfng through-wall cracks) requfred by Table 4.4-2.
4.4.5.5
~Ra orts a.
Following each fnservfce fnspictfon of steer generator tubes, the number of tubes plugged tn each stems generator shall be reported to the Caaafssfon wfthfn IS days.
b.
The complete results of the steam generator tube fnservfce fnspac-tfon shall be included fn the Annual Operatfng Report for the perfod fn whfch thfs fnspectfon was completed.
This report shall fnclude:
Number and extent of tubes fnspected.
2.
Locatfon and percent of wall-thickness penetration for each fndfcatfon of an imperfection.
3.
Identfffcatfon of tubes plugged.
fKlK76.ANQ MPLgCE THE.
4 oggi~LEVE NG5ua'g Of-Yea STa~
C CVMmep. TunE, lNSERuiCE iaSOECTi~d To
'5 PE@l SHAl-L SC Su8ehtTlgQ To TH6 Commi&~50& tM A
'5~H.tWl p EP T Pu 5
OC 4 HA~T Rl p'lcATiow 0 I Q MMLH i
~bg~5 F0LLb44ht46 C,o~pgc'giga/
OF ~5 LQ~P&TAto.
THL5 '5PKCigg REPORT +gag, ST.
LUCIK - UNIT I 3/4 4-8 AmenMnt No 18 108
1I ~
REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITIHG CONDITION FOR OPERATION 3.4.6.2
.a.
b.
C ~
Reactor Coolant System leakage shall be limited to:
Ho PRESSURE BOUHDARY LEAKAGE, GPM UNIDENTIFIED LEAKAGE, 1
GPM total primary-to-secondary leakage through steam generators, d.
10 GPH IDEHTIfI >> ~.YAGE from the Reactor Cool ant System, and e.
Leakage as spCcified in Table 3.4.6-1 for each Reactor Coolant System Pressure Iso.. -~alve identified in Table 3.4.6-1.
APPLICABILITY:
MCOES 1, 2, 3 and 4.
V+
ACTIOI(:
a.
ll b.
c With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With any Reactor Coolant System leakage greater than any,.one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within 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 />.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit in 3.4.6.2.e above reactor operation may continue provided that at least two valves, including check valves, in each high pressure line having a non-functional valve are in and remain in the mode corresponding to the isolated condition.
Motor operated valves shall be placed in the closed position, and power supplies deenergized.
(Note, however, that this may lead to ACTION requirements for systems involved.)
Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTDOWN within the following 30 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a.
Monitoring the containment atmosphere.
gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ST LUCIE - UNIT 1 3/4 4-14 Order dated 4/20/81
~Ssten COtlTAItBENT LEAKAGE PATHS Valve Ta Number Location to Containment Service Test
~Ts Hakeup Hater 8
Station Air PELTATE. PsQ EF Piece wtTH g@Q C~ <- l%"
1 1 "j)
,~(W-v- ><'gq(.g Gate (I-NV-15-1)
Check (I-.V-15328)
I lobe (I-V-18-947)+
Gl obe I-V ++
Check I-V-181118)w++
Globe
( I-SH-18797)+++
Outside Inside Outside Outside Insi e+++
Annulus~++
Station Air Supply Bypass Primary Makeup Pater Bypass 0
~s 10 14 23 25 26 28 Instrument Air Containment Purge Containment Purge Haste lianagement Component Cooling Component Cooling Fuel Transfer Tube CVCS Sampling Ga te
( I-11V-18-1)
Check
( I-V-18195)
Butterfly (I-FCV-25-4)
Outterfly (I-FCV-25-5)
Butterfly (I-FCV-25-3)
Butterfly ( I-FCV-25-2)
Globe (V-6741 Check (V-6779)
Butterfly (I-HCV 14-7)
Butterfly (I-HCV-14-1)
Butterfly (I-HCV-14-6)
Butterfly (I-HCV )
Double Gasket Flange Globe (V-2515)
Globe (V-2516)
Globe (V-5200)
Globe (V-5203)
Globe (I-FCV-03-lE)
Globe
( I-FCV-03-1F)
Outside Inside Inside Outside Inside Outside Outside Outside Outside Outside Outside Outside Inside Inside Inside Outside Outside Outside Outside Containment Purge Exhaust Containment Purge Supply Nitrogen supply to SI Tanks RC Pump CM Supply Type C
1 Type C
Bypass Bypass RC Pump CH Return I
Fuel TranSfer Letdown Line Reactor Coolant Sample SI Tank Sample SI Tank Sample Bypass Bypass Bypass Bypass Bypass Instrument Air Supply Bypass
+To become I-HCV-18-2 upon completion of the modification described in L-B7-123.
++To become I-V-18795 upon completion of the modification described in L-87-123.
t++To become effective upon completion of the modification described in L-87-123.
TASLE 3.6-2 CONTAINMENT ISOLATION VALVES Valve Ta Number A.
CONTAINMENT ISOLATION 1.
I-FCV-25-4,5 2.
I-FCV-25-2,3 3.
I-MV-15-1 4.
I-MV-18-1 5.
V-6741 6.
I-HCV-14-1 8 7
7.
I-HCV-14-6 S 2
8.
V-2515,2516 9.
V-5200,5203 10.
V-5201,5204 ll.
V-5202,5205 12.
V-6554,6555 13.
I-LCV-07-11A,118 Penetration Number 10 ll 7
9 14 23 24 26 28 29 29 31 42 Function Containment purge air exhaust, CIS Containment purge supply, CIS Primary makeup water, CIS Instrument air supply, CIS Nitrogen supply to safety injection
- tanks, CIS Reactor coolant pump cooling water
- supply, SIAS Reactor coolant pump cooling water
- return, SIAS Letdown line, CIS, SIAS Reactor coolant sample, CIS Pressur izer surge line sample, CIS Pressurizer steam space
- sample, CIS Containment vent header, CIS Reactor cavity sump pump discharge, CIS Testable During Plant 0
ration No No Yes No Yes No No No Yes Yes Yes Yes Yes Isolation Time Sec 5
5 19 28 5
5 5
5 5
10 14.
V-6301,6302 15.
V-2505 16.
I-SE-01-1 43 44 Reactor drain tank pump suction, CIS Reactor coolant pump controlled
- bleedoff, CIS Reactor coolant pump controlled
- bleedoff, CIS Yes No 17.+ I-HCV-18-2 8.
Station air supply, CIS
~item 17 to become effective upon completion of the modification described in L-87-173.
Yes
Valve Ta Number B-HANUAL OR REHOTE MANUAL Penetration Number TABLE 3.6-2 Continued Function Testable Ouring Plant 0 ration Isol ation Time Sec DEMT~ Am ERA~
~cd&
X-'9-l%- 1g t 1.
I-V-18-947+
2.
I-V-25-11,12 3.
I-V-25-13,14, 15,16 4.
V-3463 5.
I-V-07009 6.
V-07206, V-07189 7.
V-07170, V-07188 57 8
58 41 46 47 Station air supply, Hanual Hydrogen purge outside air make-up, Hanual (NC)
Hydrogen purge exhaust, Manual (NC)
Safety in)ection tank test line, Hanual (NC)
Safety in)ection tank test line, Hanual (NC)
Refueling cavity purification flow inlet, Hanual (NC)
Refueling cavity purification flow outlet, Hanual (NC)
Yes Yes Yes Yes Yes Yes Yes 1
NA NA*
NA*
NA o
8.
I-FSE-27-1,2,3, 4,8,11 9.
I-FSE-27-5,6,7, 9,10 4&a 5 Hydrogen sampling line, Remote 4&c manual 5la 5 Hydrogen sampling line, Remote 51c manual Yes Yes NA*
NA*
~To become I-V-1&795 upon completion of the modification described in L-&7-123.
TR A
P W
R SYST H
URV AN R
R H NT ontinued QE LET&. A~0 Q.EPLACC Qcc u,MULATc0 g.
At least once per ten years by:
2.
Draining each fuel storage
- tank, removing the accummulated sediment and cleaning the tank using an appropr>a e
cleaning
- compound, and Performing a pressure test of those portions of the diesel fuel oil system designed to USAS B31.7 Class 3 requirements at a test pressure equal to 110% of the system design pressure.
4.8.1.,1.3 ~R~r
- All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2.
Reports of diesel generator failures shall include the information recommended in Regulat'ory Position C.3.b of Regulatory Guide 1. 108, Revision 1, August 1977.
If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide
- 1. 108, Revision 1, August 1977.
/
4.8. 1. 1.4 The Class 1E underground cable system shall be demonstrated OPERABLE within 30 days after the movement of any loads in excess of 80K of the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at l'east 80% of the duct's inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.
ST.
LUCIE - UN?T 1
3/4 8-6b Amendment No. 444 112
~ J.
T.ABLE>>".. 11-1 RAGIOACT:V L:GU:O '~ASTE SAMPL'NG ANO n,'iAL",SIS ?ROGRAM L i qui d
.=,el e as e Type Samoling Frequency Minimum Aralysis Fr quency Tvpe o
Activity Analys i s Lower L imit or Oetection (LLO)
(pCi/ml )
A. Batch Waste Release
- Tanks, P
P Each Batch Each Batch Principal Gamma Sxlo "7 Emittal s P
One Ba ch/M P
b Each Batch Composite I-131 Oissolved and Entrained Gases (Gamma Emitters)
H-3 R
Gross Aloha 1 x10 1 xl0 1 xl 0 1 x10 P
Each Batch Q
Ccrocsit b
S~e9, Sr-90'e-=5 Sxlo 1xlo B. Cantinuopy Releases Gaily Composite
<<7 Principal Gamma 5xlo e
cml tters O '/M Grab Samole Comoosite Gaily I-131 Oissolved and Entrained Gases (Gamma Emi-ers) lxlo 1 xl0 Gaily M
Composite H-3 Gross Alpha lxlo lxl0 C. Set:1'ng Sas in Gaily Grab Sample
'Q Comoos ite Sr-89, Sr" 90 Fe" 55 Principal Gamma e
Emi t"ers I 131
-S 5x10 1xl0 Sx'0
~m'0 "
ST.
LUC IE -
UWIT 1
3/ 2 Amendment 1'Io.
5 9
.I yg F
~t
TABLE 4.11-2 Continued TABLE NOTATION b.
c ~
Sampling and analysis shall also be'performed following shutdown, startup, or' THERMAL POWER change exceeding 15$ of RATED THERMAL POWER within I hour unless (I) analysis shows that the OOSE EQUIVALENT I-131 concentra-tion in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
Samples shall be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler).
Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup or THERMAL. POWER change exceeding 155 of RATED THERMAL POWER in I hour and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing if (1) analysis shows that the OOSE EQUIVALENT I-131 concentration in the primary coolant has incr'eased more than a
factor of 3; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.
When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of '10.
e.
The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.I, 3.11.2.2 and 3.11.2.3.
The principal gamma emitters for which the LLD specification applies exclusively are t'e following radionuclides:
Kr-87, Kr-88, Xe-133, Xe-133m,,Xe-135, and.Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.
This list does not mean that only
'des are to be detected and reported.
Other peaks which ar mea bl nd identifiable, together with the above nuclides, shall also be identified and reported.
PpPP~
- QLTH, We<5+t'~Q(~'T.
LucrE - UN@'
3/4 11-10 Amendment No. 5 3
TABLE 4.12-1 Continued TABLE NOTATION It shooid be recognized that the LLO is defined as an a ~riori (before the fact) limit representing the capability of a measurement system and h
f i li~i<<
ment.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.
Occasionally background fluctuations, unavoidable small sample 'sizes, the presence of interfering nucl ides, or other uncontrollable circumstances may render these LLDs unachievable.
In such cases, the contributing factors shall be identified and described in the Annu iolo ical Environmental Operating Report pursuant to Specification 6.9.1.11.
d LLD for drinking water samples.
If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.
An equilibrium mixture of the parent and daughter isotopes which corresponds to 15 pCi/a of the parent isotope.
OE,LETE Qtdo ~P~~
(A>Thd ST.
LUGI E - UNIT 1
3/4 12-10 Amendment No. 5 3
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Continued The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA.
Conformance with these alignment requirements brings the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints.
However, extended operation with CEAs significantly inserted in the core may lead to pertur-bations in 1) local burnup,
- 2) peaking factors, and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.
Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.
The requi~ement to reduce power in certain time limits, dependinq upon the previous F
, is to eliminate a potential nonconservatism for situations when a
CEA has been declared inoperable.
A worst case analysis has shown that a
DNBR SAFDL violation may occur during the second hour after the CEA misalignment if this requirement is not met.
This potential DNBR SAFDL violation is eliminated by limiting the time operation is permitted at FULL POWER before power reductions are required.
These reductions will be necessary once the deviated CEA has been declared inoperable.
The time allowed to continue operation at a reduced power level can be permitted for the following reasons:
1.
The margin calculations that support the Technical Specific are based on a steady-state radial peak of F 1.70.
Qg4 an 2.
When the actual F < 1.70, significant additional margin exists.
3.
T(is additional margin can be credited to offset the increase in F
with time that can occur following a CEA misalignment.
r 4.
This increase in F
.is caused by xenon redistribution.
r 5.
The present analysis can support allowing a misalignment to e~ist for up to 60 minutes without correction, if the initial F
< 1.67.
Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit.
The CEA "Full In" and "Full Out" limits provide-an addi-tional independent means-for determining the CEA positions wheF'the CEAs are at either their fully inserted or fully withdrawn positions.
Therefore, the ACTION statements applicable to inoperable CEA position indicytors permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.
ST.
LUCIE - UNIT 1
8 3/4 1-4 Amendment No. 18.82~
jf 4
t hf
PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM Continued for operations personnel during and following all credible accident condi-tions.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 ran or less whole body, or its equivalent.
This limi tion is consistent with the r equirements of General Design Criteria 10 of Appendix "A"., 10 CFR 50.
OELETE AMo EEPW~
3 4.7.8 ECCS AREA VENTILATION SYSTEM The OPERABILITY of the ECCS area ventilation sys tern ensur es tha t radio-active materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment.
The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident analyses.
3/4.7. 9 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material.
The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
guantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Parts 30.11-20 and 70.19.
Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.
3 4.7.10 SNUBBERS All safety related snubbers are required to be OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.
Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
The visual inspection fr equency is based upon maintaining a constant level of snubber protection to systems.
Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection.
Inspections performed ST.
LUCIE - UNIT 1
B 3/4 7-5 "Amendment No. AA. g'f
~ 83
~
J
~
~1 fi 4.)i ~
ji V
vE I
t
'lP '
REFUELING OPERATIONS BASES 3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered thr ough the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.
The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assump-tions of the accident analyses.
3/4.9.13 SPENT FUEL CASK CRANE The maximum load which may be handled by the spent fuel cask crane is limited to a loaded single element cask which is equivalent to approxi-
~
mately 25 tons.
This restriction is provided to ensure the structural integrity of the spent fuel pool in the event of opped cask accident.
Structural damage caused by dropping a load i exc of a loaded single lement cask could cause leakage from the spent ue pool in excess of the maximum makeup capability.
P I=PLUM 3/4.9.14
'OECAY TIME -
STORAGE POOL The minimum requirements for decay of the irradiated fuel assemblies in the entire spent fuel storage pool rior to movement of the spent fuel cask into the fuel cask compartmen insur that sufficient time has ela sed to allow radioactive decay of the fission ro ucts.
The decay time of 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br /> is based upon one-third of a core placed in the spent fuel pool each year during refueling until the pool is filled.
The decay time of 1490 hour0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> s is based upon one-third of a core being placed in the spent fuel pool each year during refueling following which an entire core is placed in the pool to fill it.
The cask drop analysis assumes that all of the irradiated fuel in the filled pool (7 2/3 cores) is ruptured and follows Regulatory Guide 1.25 methodology, except that a Radial Peaking Factor of 1.0 is applied to all irradiated assemblies.
ST.
LUCIE - UNIT 1 B 3/4 9-3 Amendment No. 2l, 89, 91
ADMINISTRATI VE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE LI UID, GASEOUS AND SOLID WASTE TREATMENT V
6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):
1.
Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Facility Review Group.
The discussion of each shall contain:
a.
b.
C.
d.
e.
A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; Sufficient detailed information to totally support the reason 0
for the change without benefit of additional or supplemental
~
information;-
ReP A detailed description of the equipment, components and processes involved and the interfaces with other plant syst r
An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; An evaluation of the change which shows the expected maximum exposures to individuals in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; AQi Ll
/y f.
A comparison of the predicted releases of radioactive materials in liquid and gaseous effluents and in solid waste to the actual releases for the period prior to when the changes are to be made; 9.
An estimate of the exposure to plant operating personnel as a
result of the change; and h.
Documentation of the fact that the change was reviewed and found acceptable by the FRG.
2.
Shall become effective upon review and acceptance by the FRG.
Licensees may choose to submit the information called for in this Specifica-tion as part of the annual FSAR update.
ST.
LUCIE - UNIT 1
6-24 Amendment No. 69
1
ATTACHMENT 2 Marked-up St. Lucie Unit 2 Technical Specifications page:
2-4 2-5 3/4 2-9 3/4 3-2 3/4 3-13 3/4 6-2 3/4 6-26 3/4 8-7 3/4 9-9 3/4 11-10 3/4 12-10 6-25
TABLE 2.2"1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIHITS IC:nM m
C FUNCTIONAL UNIT'.
Variable. Power Level - High Four Reactor Coolant Pumps Operating 3.
Pressurizer Pressure - High 4.
Thermal Hargin/Low Pressure (0 Four Reactor Coolant Pumps Operating TRIP SETPOINT Not Applicable
< 9.61'bove THERMAL POWER, with a minimm setpoint of 15K of RATED THERMAL POWER, and a maximum of < 107.0X of RATED THERMAL POQER.
< 2370 psia Trip setpoint adjusted to not exceed the-limit lines of Figures 2.2-3 and 2.2-4.
Minimum value of 1900 psia.
ALLOWABLE VALUES Not Applicable
< 9.61% above THERMAL POWER, aod ~
a minimum setpoint of 15X of RATED THERMAL POWER and a maximm of < 107.0X of RATED THERMAL PMER.
< 2374 psia Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.
Hiniaum value of 1900 psia.
~ 5.
Containaent Pressure - High Steam Generator Pressure - Low Steam Generator Pressure Difference - High (Logic in TH/LP Trip Unit) 6.
f'+
Steam Generator Level - Low 3 0 psig
> 626.0 psia (2)
< 120.0 psid
> 20.5X (3)
< 3.l Psig
> 621.0 psia (2)
< 132.0 psid
> 19.5X (3)
TABID 2.2-1 (CooL~loood ttfACTOlt PROTECTIVE INSTRtHIENIATIDN TRIl'ETPOINT l IHI1S FUNCTlONnt. UNlf 9.
Local l'ower DensiLy - lligl}(5) 10.
Loss of C(apo}}ent Coolin(J Mater tu lteacto} t:ool ant Pui}}l}s-Low ll. Iteactov Protection Sysle}}} l.ogic 12. lteact(}r 'trip ltreakevs 13. ltate ol Cl}all(J>> of l'ower - lligh 14. Ilo iolor Cooloot Flow - l.o (1) t5. l.oss of l.oad (Turl}l(}e) llyd}.aul}c I-l(}td I'}ess(}re - low 1 ltlp SETPOI AT lrip setpoinL adjust.ed Lo nut exceed t,l}e li(}}it"li(}es of F igures
- 2. 2-1 and 2. 2-2.
> 636 gl}}}}"" Not Applicable Not Applicable < 2.49 decades per (}}i(}ute > 95.4X of design ReacLor Coo)a(}L flow witl} four . pu}}}ps operating" > 800 psig ALLOWABLE VALUES Trip set.point adjusted to not. exceed t.he li}}}itlines of Figures 2.2-1 and 2.2-2. > 636 gp(a Not Appl icable. Not. Appl icable < 2.49 decades per a}inute > 94.9X of (lesig}} Reacto} Coolant. flow witl} fo(}r pu((}ps ol)e}'at ing" > 800 psi(J Design reactor cooia}}L flow with fouv pu}t}ps ol}orating is 363,000 (J)}le. 10-}1}i>>}}L(}Li}t}e delay after relay act(}atio>>.
POWER GIS RIBUTION LIMITS TOTAL INTEGRATED RADIAL P<DKING FACTOR - F T LIMITING CONG ITION FOR OPERATiON 3, 2. 3 The calculated value of F, shall be limited to < l. 70. T APPLICABILITY: MOOE l". ACTEON: With F > l.70, within 6 hours either: T r a ~ Be in at least HOT STANDBY, or Reduce THERMAL POWER to bring the comoination of THERMAL POWER ana F to ~ithin the limits or Figure 3.2-3 and withdraw the rull-length CEAs to or oeyond the Long Term Steady State Insertion Limits or Specification
- 3. l.3.6.
The THERMAL POWER limit determined from Figure 3.2-3 shall then be used to establish a revised uoper THERMAL POWER level 'limit on Figure 3.2-4 (truncate Figure 3.2-4 at the 'allowable frac=ion of RATED THERMAL POWER determined by Figure 3.2-3) and subsequent ooeration shall be maintained within the reduced acceptable operation region of Figure 3.2-a. SURVEILLANCE REOUIREHENTS 4.2.3.1 The provisions of Soecification 4.0.4 aro not aooiicable. 4.2.3.2 F. shall be calculated by the expression =.=",i-: ) when F is calculated with a non-iull core power distr'but:on ana'ivs s code and shall be calculated as F' when calculations are oerformed with a r r. .ful 1 cor e "owe. di stribution analysis code. Fshall be determined -o be within its limit a" the Following inter/ais: a. Prior to loadin opera 'bove IG: of RA ED .HERBAL.=rJ~ER =f:ar ~= " =": DELHI, Awg Q.a wc- ~n"q 4 caw~q " b. At least once per 31 days or accumulated ooeration in.'~00K ', and c. Ai hi n '. ilours 1 tile A2 'UTHAL POWER TILT ( ) i s ~ O. 'ee .oec'. al: est Kxceot. on 3. O. 2. ST. LUCIE - JiIIT 2 3 If Amendment.'Io.;
TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION ~ I M foal FUNCTIONAL UNIT 1. Manual Reactor Trip 2. 3. 4. 5. 6. Variable Power Level High Pressurizer Pressure - High Thermal Margin/Low Pressure Containment Pressure - High Steam Generator Pressure - Low TOTAL NO. OF CHANNELS 4 4 4 4 4 4/SG CHANNELS TO TRIP 2 2 2(a)(d)
- nDD, 2 d) 2 2/SG(b)
MINIMUM CHANNELS OPERABLE 4 4 3 3 3 3 3/SG APPLICABLE MODES 1, 2 3A 4A 5A 1, 2 1, 2 1, 2 1, 2 1, 2 ACTION 1 5 2¹ 2¹ 2¹ 2¹ 2¹ 7. Steam Generator Pressure Difference - High 8. Steam Generator Level - Low 9. Local Power Density - High 10. Loss of Component Cooling Water to reactor Coolant Pumps ll. Reactor. Protection System Logic 4/SG 12. Reactor Trip Breakers 13. Wide Range Logarithmic Neutron F lux Monitor a. Star tup and Operating-Rate of Change of Power-High b. Shutdown
- 14. Reactor Coolant Flow Low 4/SG 15.
Loss of Load (Turbine Hydraulic Fluid Pressure - Low) 4 2(a)(d) 2/SG 2(c)(d)' 3/SG 3 3 3 2(f) 2(c) 2(e)(g) hag 0 8 rg 2 2/SG d) 3/SG 1, 2 1, 2 .1 1, 2 1 2 3* 4* 5A 1, 2 3* 4* 5'A 1, 2 3, 4, 5 1, 2 2¹ 2¹ 2¹ 2¹ 2¹ 5 5 2¹ 3 2¹
~ I J ~
TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I C M m C 4/steam generator 4 TOTAL NO. FUNCTIONAL UNIT OF CHANNELS 4. MAIN STEAH LINE ISOLATION (HSIS) a. Hanual (T 2 Button ) AOD b. Steam Generator Pressure - Low c. Containment Pressure-High d. Automatic Actuation 2 Logic 5. CONTAINMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) 2 b. Refueling Mater Storage Tank - Low c. Automatic Actuation Logic 2 CHANNELS TO TRIP 2/steam generator 2 HINIHUH CHANNELS OPERABLE 3/steam generator 3 APPLICABLE MODES ACTION 1,2,3 16 l,2,3 1, 2, 3 13*, 14 12 1,2,3,4 12 1,2,3 1, 2, 3 17 12 1, 2, 3(c) 13", 14
I l Il4t I i
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONOITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: a. An overall integrated leakage rate of: 1. Less than or equal to L , 0.50 percent by weight of the a'ontainment air per 24 hours at P , 41.8 psig, or as b. C. 2. Less than or equal to Lt, 0.35 percent by weight of the containment air per 24 hours at a reduced'ressure of Pt, 20e9 psig. I A combined leakage rate of less than or equal to 0.60 L for all a penetratfons and valves subject to Type B and C tests when pressurized to P DZMT'E. 59 A combined bypass leakage rate of less of than or equal
- 0. 12 L for a
all penetrations identified'n Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P a'PPLICABILITY: HOOES 1, 2, 3, and 4. ACTION: With either (a) the measured overall integrated oontainment leakage rata exceeding 0.75 L or 0.75 Lt, as applicable, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types 8 and C test" exceeding 0.60 L , or (c) with the combined bypass leakage rate exceeding a'.L2 L , restore the overall integrated leakage rate to less than or equal to 0.75 L or less than or equal to 0.75 Lt, as applicable, and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than or equal to 0.60 L , and the bypass leakage rate to less than or equal to O.i2 L, prior to increasing the Reactor Coolant System temperature above 2004F. SURVEILLANCE RE UIREMENTS 4.6. 1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50: a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during ST. LUCIE - UNIT 2 3/4 6-2 Amendment No. 3g, 37
CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.6.5 The primary containme vess 1 OPERABLE with an actuatio set point 0.35 inches water gauge. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: to annulus vacuum relief valves shall be f less than or equal to 9.85 + oBL~76, A~Q C.c pM (G Do I fib 5 c.tpbIA+ With one primary containment vessel to annulus vacuum relief valve inoperable, restore the valve to OPERABLE status within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD 'SHUTDOWN within the following 30 hours. SURVEILLANCE RE UIREMENTS 4.6.5 No additional Surveillance Requirements other than those required by Specification 4.0.5. ST. LUCIE - UNIT 2 3/4 6-26
Ice cc
~ j ELECTR ICAL SURVEILLANCE RE UIRENENTS Continued 7, c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on 'the emergency bus concurrent with a safety injection actuation signal. Verifying the diesel generator operates for at least 24 hours.*+** Our ing the first 2 hours of this test, the diesel gen@.ator shall be loaded within a load band of 3800 to 3985 kW~ and during the remaining 22 hours of this test, the diesel generator shall be loaded within a load band of 3450 to 3685 kW<. The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seonds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24 hour
- test, perform Surveillance Requirement 4.8.1.1.2e.4.b) 8.
9. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3935 kW. Verifying that the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power. b) Transfer its loads to the offsite power source, and DLLC ) G. 10. c) He restored to its standby status. r Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power. Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines. PThis band is meant as guidance to avoid routine 'overloading of the engine. Variations in load in excess of this band due to changing bus loads shall not invalidate this test.
- Thistest may be conducted in accordance with. tFie manufacturer"s recommendations concerning engine pr elu5e period, ST. LUCIE-UNIT 2 3/4 8-7 Amendment No. 39
R FUR IHQ OP( TIOHS LOW MlTER LEV LIHITINQ CONOITION FOR OPERATION
- 3. 9.8. 2 Two-fndependent shutdown coolfng loops shall be OPERABLE and at least one shutdown coo1fng loop shall be fn operatfon.
APPLICABILITY: HOOE 6 when the water level above the top of the reactor pressure vessel flange fs less than 23 feet. ACTEON: a. b. fifth less than the requfred shutdown coolfng loops
- OPERABLE, wfthfn 1 hour inftfate correctfve actfon to return the requfrad loops to OPERABLE status, or to establfsh greater than or equal to 23 feet of water above the reactor pressure vessel
- flange, as soon as possfble.
fifth no shutdown coolfng loop fn operatfon, suspend all operatfons fnvolvfng a reduction fn boron concentratfon of the Reactor Coolant Systii and wfthfn 1 hour fnftfate correctfve actfon to return the requfred shutdown coolfng loop to operatfon. Close all contafnment penetratfons provfdfng dfrect access froa the contafnment atmosphere 'o the outsfde atmosphere wfthfn 4 hours. SURVE LANCE R UIRENEHT 4.9.8.2 At least one shutdown coolfng loop shall be verfffed to be fn oper atfon and cfrculatfng reactor coolant at a flow rate of greater than or equal to 3000 gpi at least once per 12 hours, ST. LUCIE - UNIT 2 3/4 9 9 Amendment No.4,
1 H akf pV 4>> i4 ' 'p V ,I
~ i ~ TABLE 4. 11-2 Continued TABLE NOTATION b." C. d. e. Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15K of RATED THERMAL POWER within 1 hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3. Samples shal 1 be changed at least 4 times a month and analyses shal 1 be completed within 48 hours after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours for at least 7 days following each
- shutdown, startup or THERMAL POWER change exceeding 15X of RATED THERMAL POWER in 1 hour and analyses shall be completed within 48 hours of changing if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.
When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. 'he ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculatiog made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3. The principal gamna emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr 88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that onl the nucl ides are to be detected and reported. Other peaks which ar measureable nd identifiable, together with the above nuclides, shall a so e identified and reported. QE~Tg QNQ kERAQ ~<<H "woes,ut gg4" ST. LUCIE = UNIT 2 3/4 11-10 Amendment No. 25
TABLE 4. 12-1 Continued) TABLE NOTATION it should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a gosterior i (after the fact) limit for a par ticular measure-ment. Analyses shall be performed in such a manner that the stated LLOs will be achieved under routine conditions'ccasionally background fluctuations, unavoidable small sample sizes, the presence of interfering
- nuclides, or other uncontrollable circumstances may render these LLOs unachievable.
In such cases, the contributing factors shall be identified and described in the Annu adiological Environmental Operatin Re pursuant to Specificatio 6.9.1.11. ~+~< ~<+ ~p~ ~qyH i> gq ~ g d LLD for drinking water samples. If no drinking water pathway exists, the LLO of gamma isotopic analysis may be used. An equilibrium mixture of the parent and daughter isotopes which corresponds to 15 pCi/R of the parent isotope. ST. LUCIE " UNIT 2 3/4 12" 10
4 N t lk+ 'I@0 V .I
ADMINIST ATIVE CONTROLS
- 6. 13 PROCESS CONTROL PROGRAM PCP 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP: 1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the chan (s) w made. This submittal shall contain: P g Qhlg ~t3Kl= tA~ ~ 5MPggf1 a. Sufficiently detailed information to totall supor the rat ona e for.the change without benefit of additional r pplemental information; b. A determination that the change did not reduce the overall conformance of the dewatered bead resin to existing criteria for radioactive wastes; and c. Documentation of the fact that the change has been reviewed and found acceptable by the FRG. 2. Shall become effective upon review and acceptance by the FRG.
- 6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain: a. Sufficiently detailed information to totally support the rationale for the change withdut benefit of additional or supple-mental information. Information submitted should consist of a package of those pages of the ODCM to be clianged with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s); b. Co A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and Documentation of the fact that the change has been reviewed and found acceptable by the FRG. -2 Shall become effective. upon review and acceptance by the FRG. ST. LUCIE - UNIT2'-25
ATTACHMENT 3 SAFETY ANALYSIS INTRODUCTION These amendments are submitted to make administrative changes to the St. Lucie Unit 1 and Unit 2 Technical Specifications and achieve consistency throughout the Technical Specifications by removing outdated
- material, making minor text
- changes, and correcting errors.
DISCUSSION The proposed amendments represent an administrative update to the St. Lucie Units 1 and 2 Technical Specifications. Each of the proposed changes are discussed below: St. Lucie Unit 1 1) Item: Page III Action: In Section 3/4.1.2 on page III of the, Index an "s" should be added after Charging Pump, the corrected revision should read Charging Pumps-Shutdown instead of Charging Pump-Shutdown. Discussion: Adding an "s" corrects the index reference and provides accuracy and consistency with the rest of the listings in .Section 3/4.1.2'. 2) Item: Page 3/4 0-3 Action: Correct Surveillance Requirement 4.0.5 part d. by inserting a period (".") at the end of the statement. Discussion: Adding a period at the end of the statement corrects the sentence grammatically. 3) Item Page 3/4 1-12 Action: On Page 3/4 1-12 add an "s" after the word "pump" at the top of the page.
l l P C), r
Discussion: This revision will correct a typographical error. Item: Page 3/4 3-41 Action: Correct Limiting Condition For Operation Section 3.3.3.8 by inserting a period (".") at the end of the statement. Discussion: Adding a period at the end of the statement corrects the sentence grammatically. Item: Page 3/4 4-l,b Action: Correct Limiting Condition For Operation Section 3.4.1.3 part (b) by changing the period (".") at the end of the statement to a comma (","). Discussion: Changing the period to a comma at, the end of the statement in part b provides a clearer transformation from part b to part c. By placing a comma after parts a,b and c, and a period after part d a more consistent format is established. Item: Page 3/4 4-8 Action: Change Surveillance Requirement 4.4.5.5 a to read, "Within 15 days following the completion of each inservice inspection of steam generator
- tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a
Special Report pursuant to Specification 6.9.2. Change Surveillance Requirement 4.4.5.5 b to read, "The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following completion of the inspection. Discussion:. This change will allow the Unit 1 Technical Specj.fications to be consistent with the Unit 2 Technical Specifications. Item Page 3/4-14
a / tp l~ F,c p
Action:, Add the words "Surveillance Requirements" above the double line to accurately label Surveillance Requirement 4.4.6.2. This revision will accurately allow the reader to identify that below the double line is the appropriate Surveillance Requirement. Discussion: This addition willmake the format consistent with the rest of the Surveillance Requirements in the Unit 1 Technical Specifications. 8) Item Pages 3/4 6-5 3/4 6-20 3/4 6-21 Action: Delete the new valve tag numbers, and any references to them, as requested in Florida Power & Light's (FPL) Proposed License Amendment letter L-87-123 and replace them with the existing valve tag numbers. Discussion: FPL is returning to the Technical Specifications that existed prior to the proposed modification described in FPL letter L-87-123. The modification was planned as an enhancement to the Breathing Air System to save time and provide operational flexibility during a refueling outage. The penetration is isolated during normal operation. It has been determined that this modification is economically unjustified. No detrimental effects to the Breathing Air System willresult and the system will remain as originally designed. Minor valve number discrepancies are corrected. 9) Item Page 3/4 8-6b Action: Correct misspelling of the word "accumulated." Discussion: None. 10) Item Page 3/4 11-2 Action: Delete the dash ("-") in the page number.
Discussion: This revision corrects a typographical error. Item Page 3/4 11-10 Action: Correct misspelling of the word "measurable". Discussion: None. Item Page 3/4 12-10 Action: Change the referenced Technical Specification from "6.9.1.11" to "6.9.1.8". This change-.vill accurately identify the Annual Radiological Environmental Operating Report in Chapter 6, the Administrative Controls
- Section, of the Unit 1 Technical Specifications.
Discussion: In Chapter 6 of the Unit 1 Technical Specifications, Section 6.9.1.11 presently does not exist. Specification 6.9.1.8 is the correct Specification. Item Page B 3/4 1-4 Action: -Correct statement (1) in Bases Section 3/4.1.3 by adding an "r" below the "T" in the symbol F g which represents the steady-state radial peak. Discussion: This revision will amend a typographical error and accurately represent the steady-state radial peak. symbol as shown. Item Page B 3/4 7-5 Action: In Bases Section 3/4.7.7 the last sentence should reference General Design Criteria 19 of Appendix A, 10 CFR 50 instead of General Design Criterion 10 of Appendix "A", 10 CFR 50. Discussion:
Pl i,, i ) 4 P 1>> s f t l twW
Criterion 19 of Appendix A, 10 CFR 50 pertains to the Control Room and specifically states
- that, "Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures 'in excess of 5 REMs whole body, or its equivalent to any part of the body, for the duration of the accident."
This correction will accurately reference the proper Design Criteria. 15) Item Page B 3/4 9-3 Action: Change the word "exceed" to "excess". Change the word "insure" to "ensure". Discussion: These changes will correct the appropriate sentence grammatically. 16) Item Page 6-24 Action: In Administrative Controls Section 6.15.1 part (c) change the period (".") at the end of the statement and replace it with a semi-colon (" "). Discussion: By making this revision Sections 6.15.1 parts (a) through (h) will read grammatically correct. The only statement in this section that should end with a period is the last statement, ~ "Section 6.15.1 part (h). ST. LUCIE UNIT 2 1) Item Pages 2-4 2-5 Action: Add the referenced note "(1)" after the word "pressure" in Table 2.2-1 part 4., Thermal Margin/Low Pressure. Add the referenced note "(1)" after the word "Low" in Table 2.2-1 part 14., Reactor Coolant Flow-Low. Discussion: Note (1), in Table 2.2-1, is the reference note for Zero Power Mode Bypass. Note (1) states, "Trip may be manually bypassed below 0.54 of RATED THERMAL POWER during testing pursuant to
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Special Test Exception 3.10.3; bypass shall be automatically removed when the THERMAL POWER is greater than or equal to 0.5> of RATED THERMAL POWER." The Zero Power Mode Bypass allows manual bypassing of Variable Powel Level-High, Thermal Margin/Low Pressure (including ASGT), Steam Generator Pressure Difference High and the Reactor Coolant System (RCS) Low Flow. The addition of these two references will correct Table 2.2-1, to include all of the above trips. 2) Item Page 3/4 2-9 Action:. In Surveillance Requirement 4.2.3.2 part (a) change the period (".") at the end of the statement to a comma (","). Discussion: This revision will correct the statement grammatically. 3) Item Pages 3/4 3-2 Action: Add the referenced note "(a)" after the "2" and before the "(d)" in Table 3.3-1 part 4., Thermal Margin/Low Pressure under the column of Channels to Trip. Add the referenced note "(a)" after'the "2/SG" and before the "(d)" in Table 3.3-1 part 14., Reactor Coolant Flow-Low under the column of Channels to Trip. Discussion: Note (a), in Table 3.3-1, is the reference note for Zero Power Mode. Bypass. ,Note (a) states, "Trip may be manually bypassed below 0.5> of RATED THERMAL POWER in conjunction with (d) below; bypass shall be automatically removed when
- THERMAL, POWER is greater than or equal to 0.54 of RATED THERMAL POWER."
Note (d), as referenced above, states, "Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3." The Zero Power Mode Bypass allows manual bypassing of Variable Power Level-High, Thermal Margin/Low Pressure (including ASGT), Steam Generator Pressure Difference-High and the Reactor Coolant System (RCS) Low Flow. The addition of the these two references will correct Table 3.3-1, to include all of the above trips. 4) Item Page 3/4 3-13 Action: Add a close parenthesis (")") after the word "buttons" in Table 3.3-3 Part 4 (a).
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Discussion: This revision willgrammatically correct the reference to Trip Buttons. 5) Item Page 3/4 6-2 Action: Delete the word "of" from the sentence and add the word "to" to the sentence. Discussion: This revision will correct the sentence grammatically. 6) Item Page 3/4 6-26 Action: Change the words "set point" to "setpoint". Discussion: The word "setpoint" should be one word instead of two. 7) Item Page 3/4 8-7 Action: Delete the close parenthesis (")") at the end of paragraph 7. Discussion: This revision will correct the sentence grammatically. 8) Item Page 3/4 9-9 Action: Delete the comma (",") after Amendment No. 48 at the bottom of the page. Discussion: This revision will eliminate a typographical error. 9) Item Page 3/4 11-10 Action: Correct misspelling of the word "measurable".
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Discussion: None 10) Item Page 3/4 12-10 Action: Change the referenced Technical Specification from "6.9.1.11" to "6.9.1.8". This change willaccur'ately identify the Annual Environmental Operating
- Report, in Chapter 6
(the Administrative Controls Section) of the Unit 2 Technical Specifications. Discussion: Specification 6.9.1.11 presently does not exist in Chapter 6 of the Unit 2 Technical Specification. Specification 6.9.1.8 is the correct Specification and this correction needs to b' <<>>'-made:to accurately reflect the proper reference to the Annual Environmental Operating Report. 11) Item Page 3/4 6-25 Action: Correct misspelling of the word "support". Discussion: None
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ATTACHMENT 4 DETERMINATION OF NO SIGNIFICANT HAZARDS The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in 10 CFR 50.92 which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create tne possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows: (1) Operation of the facility in accordance with the proposed '", amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. Administrative changes to the Technical Specifications do not affect assumptions in plant safety analysis, nor do they affect Technical Specifications that preserve safety analysis assumptions. The corrections and clarifications referenced in this proposed license amendment do not affect the probability or consequences of accidents previously analyzed. (2) Operation of the facility in accordance with this proposed license amendment would not create the possibility of a new or different kind of accident previously evaluated. A new or different kind of accident is not created since Technical Specification LimitingConditions for Operation ,(LCO) and ACTION statement requirements remain unchanged. The administrative changes to the Breathing Air System being proposed by FPL willnot lead to material procedure changes or to physical modifications to the St. Lucie Plant. The Breathing Air System proposed modification was not completed due to budgetary considerations. The Technical Specifications are being returned to those that existed prior to the proposed modification.'inor valve numbering corrections are being made. Therefore, the proposed changes do not create the possibility of a new or different kind of accident. (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
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The changes being proposed do not relate to or modify the safety margins defined in and maintained by the Technical Specifications. Therefore, the proposed changes would not involve any reductions in a margin of safety. Based on the
- above, we have determined that -the amendment request does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety; and therefor does
,not involve a significant hazards consideration.
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