ML19242B437
| ML19242B437 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/06/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19242B438 | List: |
| References | |
| NUDOCS 7908080519 | |
| Download: ML19242B437 (44) | |
Text
_
TENNESSEE VALLEY AUTHORITY
~
CH ATTANOOG A. TEN;4 ESSEE 37401 400 Chestnut Street Tower II August 6, 1979 TVA BFNP TS 127 Mr. Harold R. Dentou, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Denton:
In the Matter of the
)
Docket No.30-296 Tennessee Valley Authority
)
In accordance with the provisions of 10 CFR Part 50.59, we are enclosing 40 copies of a requested amendment to license DPR-68 to change the technical specifications of Browns Ferry Nuclear Plant unit 3 (Enclosure 1).
The proposed amendment requests changes in the technical specifications to accoccodate reload 2 cycle 3 operation of unit 3 and as a resu]t of
<'ininating the LPCI loop selection logic approved by Amendment 23 to use DPR-68. Also enclosed are 40 copics of the justification for the.oposed changes as addressed in NIDO--24199 (Enclosure 2).
The LOCA analysis as referenced in NEDO-24199 will be submitted to you in the near future under separate cover.
TVA now plana to shut dovn unit 3 on August 26, 1979, to begin the refueling outage and to restart on November 24, 1979.
In order to avoid impacting the scheduled startup we need your approval of thin proposed change by November 9, 1979.
In accordance with the requirements of 10 CFR Part 170.22, we have determined the proposed amendment to be Class III. Tl'is classification is besed on the fact that the proposed amendment involves a single safety issue which does not involve a significant hazard consideration.
The remittance of
$4000 is being wired to the NRC, Attention: Licensing F w Management Branch.
Very truly yours, TENNESSEE VALLEY AUTHORITY h'\\.
. M. Mills, Manager Nuclear Regulation and Safety Subscribed and sworn to,before day of [/ / /< // / i ff 1979.
me thiu 3, jft l
i,
- g 5G]
}b}
Notary Public My Cornission Expires A /' /-
/'// '[/
7 7008080 Enclosures
/m cc: See page 2 An t %o onnonumtv nnmover M
N 2_
Mr. Harold R. Denton August 6, 1979 cc (EncIcsures)
Mr. Charles R. Christopher Chairman, Limestone Ccunty Comission P.O. Box 188 Athens, Alabama 35611 Dr. Ira L. Myers State Health Officer State Department of Public Health State Of fice Building Montgomery, Alabama 36104 G (J FIG }kb a
1 ENCI.OSURE I r
5 <
(J U 'l
- . 'i l
GUIDE TO PROPOSED CllANGES TO BROWNS FERRY UNIT 3 TECHNICAL SPECIFICATIONS Page 11 Claritication Page 178 -
Reload Page 11 C1arification Page 181 -
Reload Page 17 Reload Page 162 -
Reload Page 18 Reload Page 195 -
Clarification Page 29 Reload Page 196 -
Clarification Page 30 Reload Page 225 -
Reload Page 64 LPCI Mod Page 227 -
Clarification Page 66 LPCI Mod Page 266 -
Clarification Page 67 LPCI !!od Page 267 -
Clarification Page 68 LPCI Mod Page 294 -
C1irifIcation Page 70 LPCI Mod Page 318 -
LPCI Mod Page 75 Reload Page 321 -
LPCI Mod Page 93 LPCI Mod Page 325 -
LPCI Mod Page 94 LPCI Mod Page 327 -
LPCI Mod Page 96 LPCI Mod Page 360 -
Reload Page 97 Reload Page 109 -
LPCI Mod Page 136 -
LPCI Mod Page 149 -
LPCI Mod Page 150 -
LPCI Mod Page 15l -
LPCI Mod Page 154 -
Clarification Page 167 -
Reload Page 169 -
LPC: Mod Page 176 -
Reload 50o
' /r t'
/
1
LI".ITING SAFETY SY STE!1 SETTING
- Al'i.TY
- f. I f
- I T 1.1 FUEL CLADDING INTEGRITY If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within the prescribed limits.
Surveillance requirements for APRt scran set-points are given in Specif ication 4.1. B).
2.
APRM--When the reactor mode switch is in the STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.
3.
IRM--The IRM scram shall be set at less than or equal to 120/125 of full scale.
11 509 i43
SAFETY LIMIT LIMITING SAFETY SYSTEtt SETTING 1.1 PUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY C.
Whenever the reactor is in C.
Scram and isola-2 538 in.
the shutdown condition tion reactor above with i rra dia te<3 fuel in low water vessel the reactor vessel, the level zero water level shall not be less than 17.7 in. above D.
Sc ra m--t u r bine 5 10 per-the top of the normal stop valve cent valve active fuel zone.
closure closure E.
Scram--turbine control valve 1.
Fa st closure--Upon trip of the f ast acting solenoid valves 2.
Loss of con-2 1,100 psig trol oil pressure F.
Scram--low con-2 23 inches denser vacuum Hg vacuun G.
Scram--main 5 10 per-steam line cent valv(
isolation closure H.
Main steam isola-5 850 psig tion valve closure
--nuclear system low pressure I.
Core spray and 2 378 in.
LPCI actuation--
above reactor low water vessel level zero J.
IIPCI and RCIC 2 470 in.
ac tua tion-- reac-above tor low water vessel level zero K.
Main steam isola-2 470 in, tion valve above closure--reactor vessel low water level zero 13 509 144
~
should drop te! -
- he top of the fuel during this time, the e
ability to r'
..mca y heat is reduced.
This reduction in coo!'nq -
-111ty could lead to elevated cladding temperatures a n'.
rforation.
As long as the fuel remains covered with water, sufficient cooling is available to prevent feel clad perforation.
The n.i t et y limit has been established at 17.7 in. above the top of the arradiated fuel to provide a point which can be monitored and also provide adequat e mroin.
This point corr es pond s approximately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378" above vessel zero).
REFEFEtjCE 1.
General Elect ric DWR Thermal Analysi s Basis (GETAB) Da t a,
cor re la ti on and Design Application, NEDO 10958, and NEDE 10950.
,. E.e r a :
ectriu nt'.
u u. r er -.,
- t-it+
or FF:.t mit 7.12, NEDO-2 4199.
m 17 509 145
O posteion, where protectton o '. the fuel cladding integrity sa f ot y limit tu provi' led by the IRM and APRM high neutron tlux scrams.
Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron f lux scram prot c;ction over the entire range of applicability o.
the fuel clariding integrity saf et y limit.
In addition, the i sol ation valve closure scram anticipates the pressure and flux transients that occur durtnq no r rna l or inadvertent isolation valve closure.
With the scrams set at 10 percent of valve closure, neutron flux does not increase.
I.
J.
6 K.
R ea ct or low water lovel not point for init iation of II PC I arul i<CIL closinq main steam isolation valves, and starting LPCI and core spray pumps Trese systems maintain adequate coolant inventory and provide cor e cooling wit h the objective of preventing excessive clad temperatures.
Tho design of these systems to adequately perform tbo intended f un ct ion is baued on the specified low level scram not point and initiation set points.
'"r an s i en t analyss reported in Section N14 or the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.
L.
References 1.
Lintord, ".
H.,
" Analytical Methods of Plant Transient p
Evaluations tor the General Electric loiling Water Reactor," NEDO-10802, Feb.,
1973.
General '. l e c t r i c Supp li nen t a l Reload 1.icensing Submittal for in NI' 1:n i t 1 Re 1oad 2, NEDO-24199.
n 24 509 146
i i
The saf et y limit of 1,375 poig actually applies to any point in the reactor vessel; however, because of the static water head, the highest pressure point will occur at the bottom of the vessel.
Because the pressure is not monitored at this i
point, it cannot be directly determined if this safety limit j
han been violated.
Also, becaute of the potentially varying head level and flow pressure dzapo, an equivalent pressure cannot be a priori cietermined for a pressure monitor higher in the vennel.
Therefore, following any transient that is uevere enough to cause concern that this safety limit was violated, a information to determinecalculation will be perf ormed using all available if the safety limit was violated.
Plant Saf ety Analysis (BFNP FSAR Section N14.0) 2.
ASME Doiler and Pressure Vessel Code Section III 3.
USAS Piping Code, Section B31.1 4
Peactor Vessel and Appurtenances Mechanical Design Subsection 4. 2)
(BFNP FSAR
! 5 General !:l ec t r i c Supplemental Reload Licensing Submittal for PENP Unit 3 Re1oad 2,
FEDO-24199.
509 1M
2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Browns Ferry Nuclear Plant has tieen sized to meet two design bases. First, the total safety / relief valve capacity has been established to meet the over-pressure protection criteria of the ASME Code. Second, the distribution of this required capacity between safety valves and relief valves has been set to meet design basis 4.4.4-1 of sub-section 4.4 which states that the nuclear system relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.
The details of the analysis which shows compliance with the ASME Code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Sumary Technical Report submitted in response to question 4.1 da ted December 1,1971.
To meet the safety design basis, thirteen safety-relief valves have been installed on each unit with a total capacity of 84.2; of nuclear boiler rated stean flow. The analysis of the worst overpressure transient, (3-second closure of all nain steam line isolation valves) neglecting the direct scram (valve position screm) results in a maximum vessel pressure of 1280 psig if a neutron flux scram is assumed. This results in a 95 f
psig margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the tc;al safety-relief capacity of 84.2 of nuclear boiler rated has been <ivided into 70% relief (11 valves) and 14.27 safety (2 valvo.
The analysis of the plant isolation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is presented in Reference 5 on page 29.
This analysis shows that the 11 relief valves limit pressure at the safety valves to 1206psig, well below the setting of the safety Valves. Therefore, the safety valves will not open. This analysis shows that peak systen pressure is limited to 1232 psig which is 143 psig below the allowed vessel overpressure of 1375 psig.
30 509 in:
t Table 3.2.8 INITIATES OR COffrROLS TnE CORI AND CONTAINMUTT COOLItr, SYSTDti INSTFLMENTATION TalAT Minimun No.
TI o level S et t in g m inn Pemirks Orcrable Per l
IIlP Eys (Il ru nctj en 2
In s tr u men t Channel -
2 fuyo*above vessel zero.
A 1.
Delow trip settina initisted H IC 1.
Reactor bow water Level 2
Instrument ch annel -
2 $70*above vessel zero.
A 1.
Reactor Lcw Water Level Multiplier relays initiate FCIC.
2 Instrument Channel -
2 378= atore vessel sero.
A 1.
Below trip setting initiates
- Css, hultiplier relays Peactor Low Water Level initiate LPCI.
(LIS-3-58A-D, SW 81) 2.
Multiplier relay tron Css initiates accident signal (15).
2(16)
Instrument cha nne l -
2 370= alnve vessel aero.
A 1.
Delow trip settings in conjunction with drywell high Reactor Low Water Level pressure, low water level (LIS-3-5 6 A-D, cw # 2) permissive, 120 sec. del tirer and CSi. or PnB pump running,
7:3- - R initiates ADS.
(,}J A
1.
Delow trip setting permissive 1(16)
In s tr eme nt Ch an n e l -
2 54 e = a bove ves sel v ero.
for initiating signals on ADS.
?II J'h) f Reactor Low Water Level
\\d Permissive (LIS-3-189 G f - ~,g i.'
+
185, SW f t) 7" 1
Instrument Channel -
2 312 5/16= above vessel sero.
A 1.
Deinw trip setting prevents s
/
u inalvertent og< ration of r
/
Re a ct or I.ow Water Level (2/3 core height) of conteinment spray during (LITS-3-52 C 62, SW II) accident condition.
f
)-
5-
]
t'
]
C
- 1 Lyn
(
h C:3
.J/
sec)
L J7 E-fi L..
- 2
--a L
T1 42" E,
s;)
h43 a
T s t> 1 "
3.2.6 INSTPUMENTATION T EA! INITIATED 0F CO!?T F C LS THE O M E AND CONTAINMENT COOLI93 SYSTEMS Mintrun NO.
Ope r abl e Per Pemarks
- rip Sys (1)
Function Trip Level Settina Act1 n 1
A 1.
Below trip Setting pe rmis s i ve 2
Instrument Channel -
450 PS14 1 15 f or opening CSS and LPCI admission He act or I/w Pressure valves.
(PS-3-74 A & B, SW 82)
(PS-6 8 '3 5, SW 82)
(PS-68-96, SW 02) 2 Instrument Channel -
230 psig i 15 A
1.
Recirculation discharge C;'
Reactor Low Pressure valve actuation.
(PS-3-74A 6 D, SW # 1)
C'2 (PS-68-95, SW 81)
(FS-68-96, SW $ 1) b 1
Instrument Ch a nn e l -
100 psig i 15 A
1.
Below trip setting in conjunction with contain. ment E-Re act or Low Pressur e i sola t ion signal and both (PS-68-93 6 94, SW 81) suction valves open will close R HF (L PCI) admission valves.
f
\\
L 1.
With diesel power 2
Core Spray Auto 65tS8 secs.
Sequencing Timers (5)
. One per motor
/
8 1.
With diesel power 2
LPCI Auto Sequencing Os ts 1 sec.
k 2.
One per motor Timers (5) g A
1.
With diesel power L.
1 RH RSW A1, 23, C1, and 135t515 sec.
2.
One per pump 1
D3 Timers C.
_ c.. m u,
CD G
m CD
Table 3.2.B IN3Ti<UMENTATION THAT INITIATES OR ComOLS THE COF E AND CONTAINMENT COCLIN's SY M EMS Minimun No.
Operable Per Trip Sys (1) r'u nct ion Trip Level Settina Act2on Femarks 2
Core Spray and LPCI 05ts1 sec.
B 1.
With normsl power Auto Sequencing 65tsB sec.
2.
One por CSS Oct3r Timers (6) 125ts16 sec.
3.
Two p er RIIR mo t o r 185ts24 sec.
1 RHRSW A1, B3, C1, and 275ts29 sec.
A 1.
With noratal power D3 Timers 2.
One ter pump L_
- )
.i
,c-
\\
]
.i
.u 1(16)
ADS Timer 120 sec + 5 A
1.
Above trip s et t i ni in conjunction with low reactor g
water level, high drywell pressure and LPCI or CSS pumps j
running initiates ADS.
1 2
Instrument Channel -
100 + 10 psig A
1.
Below trip setting def ers ADS RHR Discharge Pressure act uat ion.
-)
-]
1 LD L
C'D LJ G
[.. _; f a
u m
T 1Di o 3.i.B INSTFUMENTATION T HAT INITIATts JR OJ PITT O LS THE CORE A NL CGNTAINMENT C00 LING JY K EFE Minimun No.
Ope r abl e Per Tr i p Sys ( 1).
Punct ion Trip level Set
- i nq Acelon P_m;3ay ks 2
In st r um en t Channel 165 + 10 psig A
1.
Pelow t rip setti ng defers ADS a ct ua t ion.
CSS Pump Cischarge Pr es s ur e 1 ( 3)
Core Spray Sparger to 2 psid + 0.4 A
1.
Alarm to detect core spray sparger pipe break.
Peactor Pressure Vessel,3/p 1
RER (LPCI) Trip System N/A C
1.
Monitors availability of power bus power monitor to logic systems.
Q C
C N
Ta D l t' 3.2.b IS3TRUMENTATION TilAT INITIATES % CO W B O LS T HE COF E AN
'JNTAINMENT COOLI'si SY ST EMS Minimun No.
Ope r abl e Per Trip Sys (1)
Pu nct ion Trip Level S +4 t inq A ct ior.
Femaria 2 (2)
In s t r umen t Channel -
5583" a bov e vessel zero.
A 1.
Above trip sett ing t ri ps HPCI t ur bi r.e.
Fe a ct or High Water Level 1
Instrument Cha nne l -
5 30 psi (7)
A 1.
A bove trip settirN isolates KN?I system and trips HPCI tutt ine.
HPCI Turbine Steam Lir.e High Flow 4 (4)
Instrument Cha nn e l -
52000F.
A 1.
Above t rip setting i sola t e s H PCI system and trips HP..
H PCI Steam Line Space turbine.
High Temperatura 1
Core Spray System Logic R/A B
1.
Includes testing auto initiation inhibit to Core Spray Syutems in ot her unit s.
1 BCIC System ' Ini ti ati ng)
N/A B
1.
Includes Group 7 valves.
Refer to Table 3.7.A for Logic list of valves.
o 1
FCIC System (Isolation)
N/A B
1.
Includes Group 5 valves.
Pefer to Table 3.7.A for Logic list of valves.
1(16)
ADS Logic N/A A
1 988 (LPCI) System N/ A B
(Initiation)
L.Tl oo U
10 Only r an" trip sy9t em for each cooler tan.
11 In only two of the tour 4160 V shutdown boards.
See note 13.
1/,
In only oru. of the tour 41t 0 V unutdown boa r ds.
See note 13.
I1 An "m"ryoney 4160 V shutdown board is considered a trip s yst em.
14 Rilksw pump would be inoperable.
Refer to section 4.5.C for the requirements of a RIIRSW pump being inoperable.
I'>
Th" accident signal i: the satisfactory completion of a one-out-o r -two t a ken twic" logic of the drywell high pressure plus low reactor pressure or the vessel low water level (2 178" a t,ov e vessel zero) originating in the core spray system t rip system.
16 The ADS circuitry is capable of accomplishing its protective action with one operable tri p system.
Therefore one trip
- ystem may be taken out of service for functional testing and ca l i br at i on for a periol not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Iwo RPT systems exist, either of which will trip both recirculation
- punp, The s ys tens will be individully +unctionally tested monthly.
- f the test period for one RPT system erceeds 2 consecut ive hours,
the,ystem will be declared inoperable.
If both RPT systems are inoperable or i f 1 RPT systen is inoperable for more than 72 consecutive hours, an orderly power reduction shall be initiated and the reactor power shall be less than 85% within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
509 154
T AB LE 4.2.8 CURVEILLANCE R E QU I f< EM EWS FOR I N3TB UM ENTAT ION TilAT INITI AT E L'R C ON'T ROL T H E C ECS Function F un ct ion i l Test Ca llt r a t ion I n s t ru m en t CP-ck Instru. ment Ch annel (1) once/3 months none fleactor Low Dressure (PS-3-7 4A 6 B)
( PS - 6 8 - 9 5)
(PS-6 8-9 6 )
In st r umen t Ch an nel (1) once/3 months none Reactor Low Pressure (PS-68-93 6 94)
Core Spray Auto Sequencing Timers (4) once/ operating c ycle none (Normal Power)
Core Spray Auto Sequencing Timers (4) once/ operating c ycle none (Diesel Power)
LPCI Auto Saguencing Timer s (4) once/ operating cycle none (Normal Power)
.w LPCI Auto Sequencing Timers (4) once/ operating cycle none (Diesel Power)
RER SW Al, B3, C1, D3 Timers (4) once/ operating cycle none (Normal Power)
R1IR SW A1, B 3, C1, D3 Timers (4) once/ operating cycle none (Diesel Power)
W CD w
(37
TABLE 4. 2. is S UP V E I LLA NC r. REQUIREMENTS FCh I NST P tN ENT AT ION THAT INITI AT E C P CONTROL ' lie CSC Punction Fisict ion :11 Test C a li t r.st i an Inst ru nent Cr Nk ADS Timer (4) once/ operating cycle none Instru.v nt Ch an n el (1) once/3 months none RHR Pump Discharge Pressuro Instrument Channel (1) once/3 nonths none Core Spray Pump Discharge Pre s s ur e Core Spray Sparger to RPV d/p (1) once/3 months once/ day Irip System Bus Power Monitor Once/op(? rating cycle N/A ncne Instrument Ch an nel Cond(nsate Storage Tank Low Level (1) once/ 3 months none LIl CD
TABLE 4.2.9 SUF/EILLANCE BE JOIPEMENT3 FOR IN3TF UMENTATION THAT INITIATE OF CONTh0L THE CSCS Pinct ton Fund iona 1 Test Cal l t r.st a an In st r 2:nont Chock LPCI (Con t a in me n t Spray) Lojic once/6 months (6)
N/ A Core Spray Loop A Discharge N/A once/6 months once/ day P re s s ure (PI-75-20)
Core Spray Loop D Discharge N/A once/6 months once/ day Pres s ur e (PI-75-4 8)
RHR Loop A Discharge P r es s ur e N/A once/6 mur.ths once/ day (PI-74-51)
R HR Loop B Discharle Pressure N/A once/6 months once/ day (PI-74-65)
Instrumant Channel -
Tested during N/A N/A B liR Start f unctional test of RHR gmmo (refer to section 4. 5. 8).
Inst rum + nt Channel -
onc e/ mont h once/6 months N/A m The r mo s ta t (RHP Area Cooler Fan)
Instrument Ch annel -
Tested during N/A N/A Core Spray A or C Start f unct ion a l test of cote spray (refer to section 4.5.A).
Instrument Channel -
Tested during
"/A N/A core Spray B or D start f unct ional test of core spray (refer W
CD W
N
TAB LE 4.2 a S U F V E I L'.A NC : h t /II e EM ENT S FOh IS m UM FSTA T ICN !!GT INITIAT'. CR CC N T FOL T d E C K U.
Pa n c t 1 <>n F unct 1 )na l Tes; c ali t r ation I n s t ruz+nt Check to section 4. 5. A ).
Instrument Channel -
once/ amonth once/f mor.ths N'/ A Thermostat (Core Spray Area Coole r 5'an)
RiiR Area Cooler Fan Loq1c Tested during N/A N/A functional test of ins t rum en t channels, B 101 motor start and t he r"to s t a t (RIIR area cooler f a n). No cther test r eq u ir ed.
Core Spray Area Cooler Fan Loq1c Tested daring I c>T ic N/A N/A system functionsl test of instrumnt ch ann el s, core spray inot or start and t ter:no-s ta t (core spray area cooler f an),
No other test r eq u ir e d.
Instrument Ch annel -
Tested during functional N/A N/A Core Spray Motors A or 0 Start test of core spray pump (ref er to section
- 4. 5. A),
In s t run.en t Channel -
Tested durini f u nct ional N/A N/A Core Spray Motors B or C start test of core
- r ay pump (refer to sect ion 4. 5. A).
.ct
-- t 7 a
.iat.
g~
LD once/ operating cycle N/A N/A I
C l RPT breaker C
U CD
in di" I, and
+ripu t he-r"ci r cu la t ion pumps.
The low reactor water level i n u t r umen + a t s on t hat in not to t il p when teactor water level is 11, 7" (F18" atuve vessel zero) above the top of the active tuel
("able t. ?. li) i ni tiat es the LPCI, Core Spray Pumps, contributes
> Aub in it i a t i on and starts the diesel generators.
These trip
+
s +tinq levels wero chosen to be high enough to prevent spurious actuation Lut low enough to initiate CSCS operation so that post accident cooling cm be accomplished and the guidelines of 10 CPR 100 will not tm violated.
For large breaks up to the complete trcumterential break 01 a 28-inch recirculation line and with tho trip sett ing given above, C3CS initiation is initiated in
+ime to moot the above criteria.
Ph" high drywell preisure instrumentation is a diverse signal to the water level inntrumentation and in addition to initiating
( v 's, tt causes isolation at Groups 2 and 8 isolation valves.
For the breann discussed above, this instr umentation will initiate C ',C ? i operation at about tn e same time as the low water 1ovol innt I umenta tion ; thun the results given above are ipplicalbe here also.
/on+urit are provided in the main steam lines as a means of m tw rina steam tlow and also limiting the loss of mass inventory I r o<n the ve,1el during a steam line treak accident.
The prima ry
'anction of the instrumentation is to detect a break in the main
- toam line.
For tho worst case a;cident, main steam line break outuide the drywell, trip uetting of 140f. of rated steam flow a
con junct ion with the tiow limiters and main steam line valve in closure, limits the mass inventory loss such that fuel is not uncovered, tuel cladding temperatures remain below 10000F and r el +'ase of radioactivity to the environs is well below 10 CFR 100 pilde l i nes.
Peterence Section 14.6.5 FSAR.
'omp"rature mon i t o r i n g instrumentation is provided in the main
,e'am line tunnel to detect leaks in these areas.
Trips are pr< > v ulml on this instrumentation and when exceeded, cause closure
<n isolat ion valvos.
The setting of 2000F for the main steam I in. tunn"1 detector is low enough to detect leaks of the order
<>t 15 unn; thus, it is capable of covering the entire spectrum of htoakn.
For large breaks, the high steam flow instrumentation is hackup to the temperature instrumentation.
a Htqh t a liat ion monitors in the main steam line t unnel have been
,uovided to detect gron> tuol tallure as in the control rod drop wrt<lont.
With th" estaolished setting of 3 times normal hickground, and main steam line isolation valve closure, fission u. m uc t rol ane is limited so that 10 CFR 100 quidelines are not
,,c".-led for this accident.
Reterence Section 14.6.2 FSAR.
An iiarm, wt t h a nominal set point of 1.5 x normal full power hackground, is provided alse.
109 509 15^
In t h-a n i t y t. i ca l t r ea t n."a t 01 the transients, 390
- r. t l l i.; ec u nd ',
at+
allowei f.etween a neutron sensor reachin; the scram point and the st art of negat Ive reactivity insert to n.
This in ade< plat e and conservative wnen comp a r s")
to the typically olmerved t ime delay of about DO m t I li m conds.
Approximately 70 mil 1iseconds af'"I no at r on I'uv reaches t tie trip point, the pilot
.eran va lvi-iol.no.d power lopply volt age g oe s t o zero an appr eix tmatel y /09 milliuecondo later, cont r ol rod mot i on b"qton.
Th.- 200 mi l l i accond u are i nclu led in the allowable r>c r a m insertion times specified in
- p"etfIcarion 1.J.C.
..pi re
..j.
. ; n t i c t:
+
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o percent rod dennits croum In additlon, RSLS will nrevent move ~ nt et rod., in the 50 ;'e r c en t d e n s i t :. to a preset power level range until the cran rd rod has been withdrawn.
D.
le a p t t v i t_y_Anomatioa Durino ' irh t u *1 cycle exceis oper at ive r eact iv ity varies a s tu"1
<le p l < t "s an<i an any burnable poison in supplementary c r in t r < >l 1 :, Lorno 1 Tho magnitule of this excess Ieactivity may in interrel fro 7 t rie critical rod contiquration.
As tuel burnup
",r og r e nn 2 5, anonalous tehavior in the excess r eact t vi t y may b+
detected by comparison of the cr it ica l rod lia t t e r n at Selec' ed f.19e !st a t es to the predicted 10:1 invent or y at tha*
- t a t".
Power operating ba! " conditions pr ovi.t.
the most <"nsitive and directly interpretable data relattv" to cote reactivity.
Furthermore, using power ororatinq hase conditions permits frequent reactivity compari90ns.
Imquiring a reactivity comparican at the specified frequency
- assuro, t-b a t a conparison will be made before the core reactivtty chanje excee is 1%. K.
Deviations in core r "a c t i v t + y qreator than 11
.K are not expected and require t hor o u-lh " valuation.
()ne percent retctivity limit is ci >n si lo t mi la t e 11 rtce an insertion of the r eacti v it y into the core would not Irad to transients exceedinq design conditions or t,
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136
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LIMITING CONDITIONS FOR OPEl% TION SURVEILLANCE REQUIREMENTS Cogg_A1D_Cg[q MtLMELrr
- 4. 5 CORE AND CONTAINMENT COOLING l
l CgoLI NG_;}X STEMS SISTEMS b. '. ii dual llea1 Remova1 B. Residual lleat Removal ') y_s tpm ( EllR S) (LPCI an 1 S_ys t em (RIIRS) (LPCI and Containment Cooling) Containment Cooling) 1. The RilHS shall be 1. a. Simulated Once/ operable: Automatic Operating Actuation Cycle Test (1) prior to a reactor startup from a Cold b. Pump Opera-Once/ Condition; or bility month (2) when there is c. Mator Opera-Once/ irradiated tuel ted valve month in the reactor operability vessel and when the reactor d. Pump Flow Once/ 3 vessel pressure Pate Mon th s is greater than atmospheric, e. Testable Once/ except as check valve operating specified in cycle specifications 3.5.B.2, through Each LPCI pump shall deliver 1.5.B.7 and 9,000 gpm against an indicated 3.9.B.3. system pressure of 125 psig-Two LPCI pumps in the same loep shall 2. With th! reactor vessel pressure less deliver 15,000 gpm against an than 105 psig, the indicated system pressure of Rif fa may be removed 200 psig. from service (except that two RilR pumps-2. An air test on the drywell and torus containment cooling headers and nozzles shall be mode and associated conducted once/5' years. A heat exchangers must water test =ay be performed on remain operable) for the torus header in lieu of the a per ioci not to air test. exceed 24 hours while being drained of 149 509 161
LIMITING CONDITIONS FOR OPEPATION SL'RVEI LLANCE REQUIREMENTS 3. ('O R E ts N D W NTliINMENT 4'S CORE AND CONTAINMEtE_C@ KING coo LI NG :i YSTE M.- SYSTEL4S supprossion chamber quality water and tilled with p r i n.a r y coolant quality water provide 1 that during coo ldown two loops with one pump po r loop or one loop with two pumps, and ausociate.1 diesel generators, in the core spray system are opernble. 3. It one RHH pump ' L PC I 3. When it is determined mode) is inoperable, that one RHR pump the reactor may (LPCI mode) is remain in operation inoperable at a time for a period not to when operability is e xceed ee zen d.: ' required, the provided the reotaining RiiR pum ps remaining RilR pum ps (LPCI mode) and (LPCI mode) and both active components in access paths of the both access paths of RHRS (LPCI mode) and the RHRS (LPCI mode) the CSS and the and the CSS and the diesel generato-s diesel generators remain operable. shall be demonstrated to be operable 4 I f any 2 KliR pumps (LPCI immediately and daily mode) beco e inoperable, thereafter. the reactor shall be placed in the cold shutdown condi-tion within 24 hours. 50% 162
lit 1ITIr3G CONDITIONS FOR OPERATION SURVEILLANCE RFQUIREMENTS 1.5 CORE AND C0!TTAINMENT 4.5 CORE AND CONTAINMENT COOLING COO L I_NG _ S YST EMS SYSTEMS 5 It ona RHP pump 4 No additional surveillance (containment coolin7 required. Mode) or associa;.ed heat exchanger is i.. operable, the reactor may remain in operation for a period not ta exceed 10 days provided the remaining RHR pumps (cont a in me nt cooling mod e) and associated heat exchangers and diesel generators and all access paths of the RilRS (containment cooling mode) are operable. 6. If two RHR pu.n p s (containment cooling mode) or anuociated hoat exchangers a re inoperable, the reactor may remin in 5. When it is determined operation for a that one RHR pump period not to exceed (containment cooling 7 days provided the mode) or associated runaining RHR pumps heat exchanger is (containnent cooling inoperable at a time mod e) an1 associated whei operability is heat exchangers and required, the all access p at hs of remaining RHR pumps th" RHRS (containment (containment cooling cooling mole) are mode), the associated operable. heat exchangers and diesel generators, and all active components in the access paths of the RHRS (containment cooling mode) shall be demonstrated to be operable immediately and waekly thereafter until the inoperable RHR pump (containment cooling Mode) and 151 associated heat /* 509 iM
I. I t11 T I t3G CONDITIONS FOR OPEP1. TION SURVEILLANCE REQUIRE? TENTS 4.5 CORE AND CONTAINMENT COOLING 3.. t'O R E A N D Cottl' AI N 4E NT SYSTEMS (?_O.O_ L L JG.S Y ST E.Mg ( No t o : Beca ni; e adjacent unit is c r on i-connect inoperable at a time capant i tt y in not a when operability is a r t. tm r eq u i r ed, the requir"mont, a remaining BliR pu. p component la not and associated heat conslocred inoper able exchanger on the unit it cron;-connect cross-connection and capabilit.y can be the associated diesel rentored to service generator shall be wit hin ; hou rn.) demonstratei to be operable immediately 12. It one hilk pump or and every 15 days aruociated heat thereafter until the exchanger located on inoperable pump and the un it cross- 'ssociated heat connection in u ni t 2 exchanger are in inoperable for any returned to normal r"ason (including service. valvo inoperability, pip" break, etc.), the reactor may romain in operation 12. No additional for a peria l not to surveillance exceed 30 days f*9"If*d* provide 1 t h'- remaintnq HiiR pump anl af.iociatod diesel 13. No additional surveillance generator are operable. required. 13. It HliR crosu-connectton flow or h'at renoval capahtlity is lost, the unit a ty remain in operation for a per1011 not to exceed 10 days unle,s such capability is restore 1. 154 509 164
I.IMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMENTS m
- 4. 5 CORE ? ND_GQlffM!idElf_f _'IOL 'E
- 1. 5 CORI. AND CONTAISMU;T HElfdE COOLING SYSTEMS and corresponding, action shall contiaue until reactor operation is within the prescribed limits.
K. Minimum Critical Power Ratto (MCPR) The MCPR operat'ng limit is K. Minimum Critical Power 1.28 for 8x8 fuel, and 1.22 Fo (MCPR) for 8x8R fuel, and 1.23 for P8x8R fuel. These limits MCPP shall be determined apply to steady state power daily during reactor power operation at rated power ar.d operation at 2 25% rated tlow For core flows other t her ma l powe r and than rated, the MCPR shall f ollowing any change in b-greater than the above power level or limits times K., K is the distribution that would r value shown in Figure 3.5.2. caupe operation with a limi ti ng c3ntrol rod If at any tir.e during as described in patte operation, it is detar-the bases for mined by nornal surveillanc ~ Specification 3.3. e that the 1in1*.ing value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours. Surveillance and corresponding actio's shall continue until reactor operation is within the prescribed limits. I. Reporting Requirements If any of the limiting values identified in Specifications 3.5.I, J, or K are exceeded and the ^ spectiled recedial action e is taken, the event shall be logged and reported in a '30-day wri t ten report. 167 509 i65
1,', Iw.m .e te y ua te core cooling. With due regard for this margin, the allowable rep a ir time of 7 days was chosen. Should one MR pu. p (LPCI mode ) tecome incrernole, caly ; Fl9 purps (LFCI ode; and the core sprily system ar+' sv111abla Since this.eavas only cr.e fig pw p (LPCI mode ) in reserve, wh;_. al cn.c with +3e re.atning 2 Fyy pu.3 3 (LPCI node) and core spray Oystem., demonst rated to be Opersola 1.mediste;y 'i n d ir. i l y *heraafter, 1 7 by r po.r p-ted nc ifted. Jhcull two ?H3 p e p (LICI trJe) n ' '.e Lopern., tr ' e re-lins
- r. 0 reserve ( r etm bnt ) :Tracit/ w I t' the i. i?3 iLPCI "c 1C 1 -
"herefere, the affecte:1 ' Int anall be placed in col: snutdrwn w ; thin 24 hours. Should one RHR pump (containment cooling mod e) become inoperable, a complerr nt of three full capacity containment heat r emov a l systems still available. Any two of the re ma in ing pumps / heat e enanger combinations would provide more than adequate conta.nment cooling for any abnormal or t pont accident situation. Because of the availability of equipment in access of normal redundance requirements, which is demonstrated to be operable immediately and with specified subsequent performance, a 30-day repair period is justified. Should two RHR pumps (co n ta i nment cooling mode) become inoporable, a tull heat removal systom is still available. The remaining pump / heat exchanger combinations would provide adoquite containment cooling for any abnormal post accident situation. Because of the availability of a f ull complet.ent of heat removal equipment, which is demonstrated to be opocable immediately and with specified performance, a 7-day repair period is iustified. Observat ion of the stated requirements for the containment cooling mode tscures that the suppression pool and the drywell will be sufficiently cooled, following a los s-o f-coolant accident, to prevent primary containment overpressuri7ation. The containment cool in g function of the Hito s 1 o perm'itted only after the core has refloo' led to the two-thirds corc height level. This prevents inadvertently divertiny wat er needed for core 11ooding to the less urgent task of containment cooling. The two-thirds core height le ve l Enterlock may be manually bypassed by a keylock switch. Sinco the hHRS is filled with low quality water during power oporation, it is planned that the system be filled with do mine ra li zed (condensate) water before using the shutdown cooling function of the RHR system. Since it is desirable to 169 / /
1.', HA ;Eb v teutinq to usepa ro that the lines ar e fille l. The visual chm ktnq will avoid starting the c're Spray or RHR system with a discharo" line not 1illed. In addition to the visual obrorvation an1 to ensure a tilled discharge line other thin prtor t o t est inq, a pressure suppression chamber head tank is loca t ed approxima tely 20 feet above the discharge line highpoint to supply makeup water for these systems. The contene.ite head tank located approximately 100 feet above the discharge htqh point serves as a backup charging system when the preauute suppreasion chamber head tank is not in service. Syste-d scharoe pressare indicators are used to determine the water leve l above the discharge line high point..The indicat ors will reflect approximately 30 psig f or a water level at the high point and 45 psig for a water level in the pr es su re suppression chamber head tank and are monitored daily to ensure tnat the di sch a rg e lines are filled. Whon .a their normal utandby condition, the suction for the ll PC I and 1CIC pumps are aligned to t he cor lensa te storage tank, wh ic h is physically at a higher el e,, ion than the HiUIS.uul Rm ICS pipinq. This assures that toe llPCI and RCIC discharge piping remains tilled. Further assurance is providea by observing water flow from these systems high poi n t u monthly. I. Maximum Average ?lanar Linear !!e a t Generation Rate (MAPLilGRJ Th i s apociftaition assures t ha t the peak cladding temperature tollowing the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, A p per.d i x K. Tho pe ak clad ling temperature following a postulated loss-of-coolant accident is primarily a function of the average heat qeneration rate of all tne reds of a f uel assembly at any axial location and is only deperident secondarily on the rod to rol power distr tbotion within an assembly. Since expected lacci variations in power distribution within a fuel assembly attect the calculated peak clad temperature by less than i 20or relative to the peak tempe ra ture for a typical fuel dosign, the limit on th e average lin ea r hea t generation rate is uutficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit. The limiting value for MA FLilG R is shown in Tables 3.5.I-1, -2, -3. The anstlyaec o n ortir then liri'< values 10 Presented in ';EDO-2M 27 and NEDO-24194 .7 Linear lio a t Generation Rate (Lil_GM Th i s specification assures that the linear heat generation rate in any rod is less tnan the design linear heat 176 509 167
.A: E v >: e ! and r eportml jua r terly. It must be recognized that i' aIwtys a :) action which would return any of the ir i,teIh (MAPLW;R, LHG l<, or MC P H) to wittun prescribed i i 'at'., rim"ly power reduction. Under most c irc umta nc e s, will not be the only alternative. ti. i v :"r'ncou 1 "tu"1 Densitication Effects on General Electric Doiling W at et aeactor Fuel," Supplemento 6, 7, and 8, NEDM-1 0 ~/ 15, August 1973. 'upplemont 1 to Technical Report on Densifications of ."neral Electric Reactor Fuels, Decernber 14, 1974 (USA 14m ul a to ry S taf f). ' om:ru n ica ti on : V. A. Moore to I. S. Mitchell, " Modified it ro del for Fuel Densification," Docket 50-321, March i 1,
- 19184, c.'
l.l ect r ic Supplemental Reload Licensing Submittal for i .:1 .e oit i Re1oad 2, SEDO-24199. 178 5<7 168 0m i
TABLE 3.5.I-1 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-3 Fuel Type: Inital Core - Type 2 Average Planar Exposure MAPLHCR (Mwd /t) (kW/ft) 200 11.4 1,000 11.6 5,000 12.0 10,000 12.2 15,000 12.3 20,000 12.1 25,000 11.3 30,0C0 10.2 TABLE 3.5.1-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-3 Fuel Type: Initial Core - Type 1 Average Planar Exposure MAPLHGR (Mwd /t) (kW/ft) 200 11.2 1,000 11.3 5,000 11.8 10,000 12.1 15,000 12.3 20,000 12.1 25,000 11.3 30,C00 10.2 181
TABLE 3.5.I-3 hPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: BF-3 FUEL TYPES: 8DRB265L and P8DRB265L Average Planar Exposure MAPLHGR (Mwd /t) (kW/ft) 200 11.6 1,000 11.6 5,000 12.1 10,000 12.1 15,000 12.1 20,000 11.9 25,000 11.3 30,000 10.7 The values in this table are cor.servative fcr both prepressurized and non-pressurized fuel. 182 .f0 509 il
LIMITING CONDITIONS FOR OPSRATION SURVEILLANCE REQUIREMENTS
- 1. 6 PigfMA RY SYSTEM I4OUNDARY 4.6 PRIMARY SYSTEM BOUN DARY F.
J et P tim p Flow Mis ma tch F. J et Pump Flow Mismatch 1. Recirculation pump speeds shall be checked and loqqed at least once per day. <~ j 1. The reactor shall not be operated with one recirculation loop out of service for more than 24 hours. With the reactor operating, if one rec ircu lttion loai is out of cervice, the plant shall be placed in a hot shutd own condition withi a 24 hours unless t ae loop is sooner returned to service. 3, Following one-pump operation, tl_e discharge valve cf the low speed pump mTy not he opened ^ unless the speed of the f aster pump is less than 50% of its rated speed. 195
LIMI"'ING CONDITIONS FOR OPEMTION SURVEILLANCE REQUIREMENTS 3.6 PhitiA_PJ SYT"!M FOUN O T '4. 6 QIMABY SYSTFM POUNDARY l
- 3. Stoddy state operalien with both G.
Structural Integrity recirculation pr.:~ps out of ser-Vice for up to 12 hrs is per-1. Tabl e 4. 6. A tog ether mi t ted. During sm.h interval with suppl emen t a ry restart of the recirculation notes, specifies the pumps is permitted, provided the inservice inspection 8 loop discharge ter?erature is re u e _n ts o f the nIlthin 750r of the saturation reactor coolant temperature of the reactor system as f ollows: vessel uater as determined by dome pressure. The total a. areas to be elapsed time in natural circula-ircpected tion 3rd cr.e purp c;ieration mus t be no grcater than 24 hrs, b. percent of areas to be inspected during the inspection G. S t ritetural Intwrity in te rval 1. The structural c. inspection integrity of the frequency priciary s ystem shall be maintainad at the d. methods used for level required by the in epection original acceptance standards throughout 2. Evaluation of the lif e of the inservice :sspections pl an t. The reactor will be made to the shall be mais.ta!ned acceptance stan da rds in a cold shutdcen epecified for the condition until each original equipment. Indication of a defect has been 3. The inspection investigated and interval shall be 10 evaluated. years. 4 Addit iona l inspactions shall be perf ormed on ce rtain circumferential pipe welds as listed to provide additional protection against pipe uhlp, which could damage auxiliary and control systems. Feedwater-G W-9, K W-13, GW-12, G W-26 509 172 x m->> c - 29 KW-39, C W-15, y KW - 3 0, and GI 4-3 2
m 3.6/4.6 BASES To meet the safety design basis, thirteen safety-relief valves have been installed on unit 2 with a total capacity of 84.2% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second clo" r" of all main steam line isolation valves) neglecting the direct scran (valve position scran) results in a maximum vessel pressure of 1280 poig it a neutron f l ux scram is assured Thi< reults in an 95 psig margin of the code allowable over-pre-ure 1imit of 117") psir To meet the operational design basis, the total safety-relief capacity of 84.27 of nuclear bo'ler rated has been divided into 70% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant iso-lation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is presented in Reference 5 on page 29. This analysis shows that the 11 relief valves limit pressure at the safety valves to 1206 psig, well below the setting of the safety valves. Therefore, the uafety valves will not open. This analysis shows that peak system pressure is limited to 1232 psig which is 143 psig below the allowed vessel overpressure of 1375 ps!g. Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to d et ect failures or deteriorations. The relief and safety valves benchtested every second operating cycle to ensure that are thei r set points are within the +1 percent tolerance. The relief valves are tested in place once per operating cycle to establish that they will open and pass steam. The requirements established above apply when the nuclear system be can pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pres su r es because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be n eeded, lioweve r, these transients are much les s se ve re, in terms of pressure, than those starting at rated conditions. The valves need not be f unctional when the vessel head is removed, since the nuclear system cannot be pressurized. R EF ERENC ES 1. Nuclea r System Pressure Relief System (BFNP FSAR Sub sec tion
- 4. 4) n O
225 509 173
3.6/4.6 B AS ES A nozzle-riser ayutem t a ilur e could also generate the coincident tailure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-rise r system f ailure.
- 3. 6. F/ 4. 6. F J et Pump Flow Mismatch Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one to two pump ope r ation that excessive vibration of the jet pump risers will not occur.
ECCS perf ormance during reactor operation with one recirculation loop out of service has not been analyzed. Therefore, sustained reactor operation under such conditionais not permitted. 3.6.G/4.6.G Structural Integrity The requi r en en ts for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of f ailure in the system and the need to meet as closely as possible the retiuirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The prog ram reflects the built-in limitations of access to the reactor coolant systems. n 509 I/4 227
NOTES FOR TABLE 3.7.A Key: 0 Open = C = Closed 3C Stays Closed = GC = Goes Closed Note: Isolation qroupinqs are as follows: Group 1: The valves in Group 1 a > actuated by any of the following conditions: 1. Reactor Vessel Low Water Level (4708) 2. Main Steamline liigh Radiation 3. Main Steamline High Flow 4 Main Steamline Space liigh Temperature 5. Main Steamline Low Pressure Group 2: The valves in Group 2 are actuated by any of the tollowing conditions: 1. Reactor Vessel Low Water Level (5 3 8") 2. liigh Drywell Pressure Group 1: The valves in Group 3 are actuated by anv of the f ollowing conditions: 1. Roacto r Low Water Level ( 5 3 8") 2. .loactor Water Cleanup System liigh Temperature 3. iteactor Water Cleanup System liigh Drain '.' em p e r a t u r e Group 4: The /alves in Group 4 are actuated by any of the f ol'.owing conditions: 1. IIPCI Steamline Space liigh Temperature 2. IIPCI Steamline liigh Flow 3. IIPCI Steamline Low Pressure Group 5: The valves in Group 5 are actuated by any of the following conditions: 1. RCIC Steamline Space liigh Temperature 2. RCIC Steamline liigh Flow 3. RCIC Steamline Low Pressurt Group 6: The valves in Group 6 are actuated by any of the f ollowing conditions: 1. Reactor Vessel Low Water Level ( 5 3 8") 2. Illgh Drywell Pressure 3. Reactor Building Ventilation Ifigh Radiation Group 7: The valves in Group 7 are automatically actuated by .n 509 i! 266
only the following condition: 1. Reactor Vessel Low Water Level (470") Group H: Ttle valves in Group 8 are automatically actuated by only the f ollowing condition: 1. Iligh Drywell Pressure f 1 'IfC E10 267 JIJ/ i
3.7.0/4.7.D Primary con ta inment Isolation valves Double isolation valves are provided on lines penetrating the primary co n ta inme nt and open to the f ree space of the containment. Closure of one of the valves in each line would be cufficient to maintain the intaqrity of the pressure suppression system. Automatic initiatier. required to minimi2.e the potential leakage paths frcm a containment in the event of a loss of ccolant accident. Group 1 - process lines are isolated by reactor 'essel low water } 1evel ( 470 ") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time f or proper operation of the core standby cooling systems. The valves in group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiatica, low pressure, or main steam space high temperature. Group 2 - isolation valves are closed by reactor vessel low water level ( 5 3 8 ") or high drywell pressure. The group 2 isolation signal also " isolates" the reactor building and starts the standby ga s treatment sy s tem. It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal. Group 3 - prccess lines are normally in use and it is theref ore not desirable to cause spurious isolation due to high drywell pressure resulting frca non-saf ety related causes. To protect the reactor frcm a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high ficw through the inlet to the cleanup system. Also, since tne vessel could potentially be drained through the cleanup system, a low level isolation is provided. Group 4 and 5 - process lines are designed to remain operable and mitigate the censequences of an accident which results in the isolation of other process lines. The signals which initiate isolation of Group 4 and 5 process lines are therefore indicative of a condition which would render them inoperable. G rouc 6 - lines ere connected to the primary containment but not directly to the reactor vessel. These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilaticn high radiation which would indicate a possible accident and necessitate primary containment isolation. G roup 7 - process lines are closed cnly on reactor low water i j level ( 470"). These close on the same signal that initiates HPCIS and ECICS to ensure that the valves are not open when EPCIS or RCICS action is required. G rcu p 8 - line (traveling in-core probe) is isolated on hich d rywel l pressure. This is to assure that this line does not provid e leakage path when containment pressure indicates a possible accident condi t'. :n. n / 294
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFQUIREMENTS I 3.9 AUXI LI AR Y ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTE_M d. Each diesel generator shall be given an annual inspection in dCCordance with instructions based on the manufacturer's r econimendations. e. Once a month a sample of diesel fuel shall be checked for quality. The quality shall be within the acceptable limits specified in Table 1 of the latest revision to ASTM D975 and logged. 2. Three unit 3 diesel 2 D.C. Power System - Unit Batteries (250-generators shall be operable. Volt) and Diesel Generator Batteries (12 5-Volt) and Shutdown Board Battery (250-Volt) a. Every week the specific gravity and the voltage of the pilot cell, and temperature of an adiacent cell and overall battery voltage shall be measured and logged. 509 178 318
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENI'S
- 3. 9 LUXI LI AR Y ELECTRICAL SYSTEM 4.9 AUXILI ARY EL ECTRICAL SYSTEM c.
The undervoltage relays which start the diesel generators f rom start buses 1A dnd 1B and the 4-kV shutdown boards, shall be calibrated annually for trip and reset and the measurements logged. S. 'l h e 250-Valt Shutdown Board battery and unit batteries and a battery charger for each battery and associated battery boards are operable. 6. Logic Systems ,1. Accident signal logic system is opetable. 7 There shall be a minimuin of 103,300 gallons of diesel fuel in the unit 3 standby diesel generator fuel tanks. 321
LIhITIhG ONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMENTS 1.9 AUXILI ARY ELECTRICAL SYSTDI 4.9 AUXI LI ARY ELECTRICAL SYSTEM 4. From and after the date that the 250-Volt Ehutdown board batterie-or one of the three 250-Volt un't batteries and/or its associat' 4 battery board . found to be inope.able for any reason, continued reactor operation is permissible during the *;ucceedinq seven days. Except for routine surveillance t e. c in g, the NRC shall be notified within 24 hours of the situation, th e precautions to be taken during this period and the plans to return the f ailed component to an operable state. 5. When one division of the Logic System is inoperable, continued reactor operation is permissible under this condition for seven days, provided the CSCS requirements listed in Specif ication 3.9.B. 2 are satisfied. The NRC shall be notified withia 24 hours of the situation, the precautions to be taken during this period and the plans to return the failed component to an operable state. 32s 509 180
3.9 BASES Tim objective of this speci.fication is to assure an adequate sourco ot electrical power to operate facilities to cool the unit <turinq shut lown and to operate the engineered safeguards There are three sources of alternating current fo11oving an accident. the electri"nl anergy available, n uely, the 161-kV transmission system, nuc lear gar.arati ng units, and the diesel generators. 161-kV of f site power supply consists of two lines which are The red f rom dit f erent sect'ons of the TVA 161-kV grid. In the normal mode of operatiu the 161-kV system is operating and four liesel generators are operational. If one diesel generator is out of se rvic e, there normally remain the 161-kV sources, and the other three diesel generators. For a diesel generator to be considered operable its associated 125 V battery must be operable. The mininum t uel oil requirement of 103,300 gallons is srfficient for 7 days of full load operation of 3 diesels and is conservatively based on availability of a replenishment supply. (1fstta auxilinry povar for Frewns Ferry Nuclaar Plant Unit 3 is supplied from two n ur .m 'hn unit station transformers frcm the rain generator or the 161-kV +ransning'on i vate a through t ha cooling tower trancformers. If a cooling tower 'ntinue to operate since the station ' r nn s fe m." is 'ont, the unit can t r ans fo r~ r it in arrvice, t he other cooling tower transfcrmer is available, and four diecel generators are operational. A 4-kV shutdown board is allowed to be out of operation for a brief poriod to allow for maintenance and testing, providing all rema in ing 4-kV shutdown boards and associated diesel generators CS, RilR, (LPCI and Containment Cooling) Systems supplied by the remaining 4-kV shutdown boards, and all emergency 480 V power boards are operable. There are five 250-volt d-c battery systeme each of which consists of a battery, battery charger, and distrioution equipment. Three of these systems provide power f or unit control functions, operative power f or unit motor loads, and alternative drive power f or a 115-volt a-c unit preferred motor-generator set. One 250-volt d-c system provides power for common plant and transmission system control functions, drive power for a 115-volt a-c plant preferred motor-generator set, and emergency drive powe r f or certain unit large motor loads. The fifth battery system delivers control power to a 4-kV shutdown board. The 250-Vo't de system is so arranged, and the batteries sized such, that the los, of any one unit battery will not prevent the safe shutdown and cooldown of all three units in the event of the loss of offsite power and a desien basis accident in,ny one unit. Loss of contrm1 powor to any engineered safeguard control 327 509 181
f ', 9 M A.JO P DE S I G!J FEATO.2'i S.1 LITE F E AT U I4 E S ll r own s Ferry units 1, 2, and 3 are located at Browns Ferry fluclear Plant site on property owned by the United States a'.-l in custo<ly of the TVA. The site shall consist cf approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary co n t a i nme nt building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet. 5.2 REACTOR A. The reactor core raa y ciatain 764 fuel assemblies consisting of 8x8 assemblies having 63 fuel reds each, and 8x8 R (and P8x8R) assemblies having 62 fuel rods each. The number of each type in the core is given i n t he in s t recent reload amendment topical report. B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (n C). compacted to approximately 70 percent of theoretical 4 density. 5.3 R EACTO R VESSEL The r" actor vessel shal' be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR. 5.4 Co tiTA I tim EtiT A. The principal design parameters for the primary containment shall be given in Table 5.2-1 of the FSAR. The applicable design codes shall be as described in Section 5.2 of the FSAR. B. The secondary containment shall be as described in Section 5.3 of the FSAR. C. Penetrations to the primary containment and piping passing through such penetrations shall b.. designed in accordance with the standards set forth in Section
- 5. 2. 3. 4 of the FSAR.
5.5 FUEL STO RAG E A. Th" arrangement of the fuel in the new-fuel storage tacilitity shall be such that k for dry conditions,
- eff, b
360 509 18 1
6 ENCLOSURE 2 1 O' 500 IJ. /}}