ML19290B748
| ML19290B748 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/30/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19290B749 | List: |
| References | |
| NUDOCS 7912130706 | |
| Download: ML19290B748 (49) | |
Text
i w
UNITED STATES o
f g
NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D C. 20555 5
- y Y
p TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 28 License No. DPR-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated August 6, 1979, as supplemented by two letters dated September 26, 1979 and additional letters dated October 10, 1979 and October 25, 1979, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
These is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No. DPR-68 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 28, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
1553 060 7912130 70 6
i 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l' 0 s
ief Operating Reactors Branch (3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 30, 1979 1553 061
ATTACHMENT TO LICENSE AMENDMENT NO. 28 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET N0. 50-296 Revise Appendix A as follows:
1.
Remove the following pages and replace with the identically numbered pages:
11 153 13 154 17 167 24 169 26 176 27 178 29 181 30 182 64 195 66 196 67 225 68 225a (new page) 70 227 75 266 93 267 94 276 96 281 97 294 109 31 8 136 321 149 325 150 327 151 360 2.
The marginal lines on each page indicates the revised area.
1553 062
4 LIMITING SAFETY SYSTEM SETTING SAVI;TY f.IMIT 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 APMi scram set-points are given in Specification 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.
1553 063 Amendment No.
28 n
i SAFETY LIMIT LIMITING mal'ETV SY STEPt S t:TT I Nt; 1.1 FUEL 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 irradiated f uel in low water vessel the reactor vessel, the level zero water level shall not be less than 17.7 in. above D.
Scram--turbine 5 10 per-the top of the normal stop valve cent valve active fuel zone.
closure closure E.
Sc ra m--turbine control valve 1.
Fast closure--Upon trip of the fast acting solenoid valves 2.
Loss of con-2 1,100 paiq 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 actua tion--
above reactor low water vessel level zero J.
actuation--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 1553 064 Amendment No. 28
mi should drop below the top of the f uel during this time, the ability to remove decay heat is reduced. This reduction in i
cooling capability could lead to elevated cladding temperatures and clad perforation.
As long as the fuel remains covered with water, sufficient cooling is available to prevent fuel clad perforation.
The natety limit has been established at 17.7 in. above the top of the a rradiat ed f ueri to provide a point which can be monitored and also provide adequate margin.
This point corresponds approximately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378" above vessel sero).
REFERENCE 1.
General Electric BWR Thermal Analysis Basis (GETAB) Data, correlation and Design Application, NEDO 10958, and NEDE 10958.
?-
General Electric Supplemental Relt ad Licensing Submittal for EFHP unit 3 Helord 2, NEDO-24199.
1553 065 i
17 Amendment No. 28
D**D *D'T]
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ponteion, where protection of the tuel cladding instergrit y Sd f ot y limit iS provided by the IRM and APRM high neutron flux scrams.
Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability o. the fuel cladding integrity safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.
With the scrams set at 10 percent of valve closure, neutron flux does not increase.
I.
J.
5 K.
R ea ctor low water level set point for initiation of HPCI and RCIC, closing main steam isolation valves, and starting LPCI and core sprav pumps These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad t emperatures. Tho deqign of these systems to adequately perform the intended function is based on the specified low level scram ne t. point and initiation set points.
.inalyses reported in Section N14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.
L.
References 1.
- Linford, R.
B.,
" Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO-10802, Feb., 1973.
2.
General Electric Supplemental Reload Licensing Submittal for BFNP Unit 3 Reload 2, NEDO-24199.
1553 066 Amendrient No. 28 24
D
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2 LIMITING SAFETY SYSTEM SETTING SAlCTY LI"IT 1.A khACTOR COOLANT SYSTEM 2.2 REACTok COOLANT SYST EM INTEGRITY INT EGH ITY Applicability Applicability Applies to limits on reactor Applies to trip settings of the instruments and devices which coolant system pressure.
are provided to prevent the reactor system safety limts f rom being exceeded.
obiective Ob iective To establish a limit below To define the level of the which the integrity of the process variables at which reactor coolant. system is not automatic protective action is threatened due to an initiated to prevent the overpressure condition.
pressure safety limit from being exceeded.
Specification Specification The 'imiting saf ety system settings shall be as specified A.
The pressure at the lowest below:
point of the reactor vessel shall not exceed Limiting 1,375 psig whenever Safety irradiated fuel is in the Protective System reactor vessel.
Action Settina A.
Nuclear system 1,250 psig safety valves 1 13 psi open--nuclear (2 valves) system pressure B.
Nuclear system relief valves open--nuclear system pre s s ure Target - Rocks 1,105 psia 2 11 psi
( 4 valves) 1,115 psig
+ 11 psi I 4 valves) 26 Amendment No. 28 1553 067
D l'}
+]D 9
D W
Ju c 3
LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT 2.2 REACTOR COOLANT SYSTQj
- 1. 2 pgAg OR_ COOLANT $YSTEM INTEGRITY INTEGPITY 1,125 psig
+ 11 psi (1 valve i Crosbys**
1,150 peig
+ 0 psi
- 22 psi (2 valves) j OR Target-Rock **
I,125 psig 1
11 psi (2 valves)
C.
Scram-nuclear i 1,055 psig system high pressure
'* Analyses have been run which allow operation with either 9 Target-Rocks and 2 Crosby's or 11 Target-Rocks as indicated in the above specification.
The results of these analyses are presented in the. Bases.
1553 068 Amendment No. 28 27
The saf ety limit of 1,375 peig 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 point, it cannot be directly determined if this safety limit has been violated.
Also, becauLe of the potentially varying head level and flow pressure draps, an equivalent pressure cannot be a priori determined for a pressure monitor higher in the vessel. Therefore, following any transient that is severe enough to cause concern that this safety limit was violated, a calculation will be performed using all available information to determine if the safety limit was violated.
Plant Saf ety Analysis (BFNP FSAR Section N14.0) 2.
ASME Boiler and l'reneure Vessel Code Section III 3.
IJSAS Piping Code, Section DJ1.1 4
Reactor VEsuel and Appurtenances Mechanical Design (BFNP FSAR Subsection 4. 2) 5.
General Electric Supplemental Reload Licensing Submittal for BFNP Unit 3 Reload 2,
NED0-24199.
1553 069 29 Amendment No. 28
2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Browns Ferry Nuclear Plant has been 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 Sunmary Technical Report submitted in resoonse to cuestion 4.1 dated December 1,1971.
9 Target Rock And 2 Crosby Valves To meet the safety design basis, thirteen safety-relief valves have been instal}ed on each unit with a total capacity of 81.08%of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1293 asig if a neutron flux scram is assumed. This results in a 82 psig margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 81.08%of nuclear boiler rated has been divided into 66.88% relief (11 valves) and 14.2% safety (2 valves). 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 1218 psig, well below the setting of the safety valves. Therefore, the safety valves will not open. This analysis shows that peak system pressure is limited to 1243 psig which is 132 psig below the allowed vessel overpressure of 1375 psig.
11 Target Rock Valves Only To meet the safety design basis, thirtean safety-relief valves have been installed on each unit with a total capa city of 84.2% of nuclear boiler rated steam flow. The analysis of the gorst overpressure transient, (3-second closure of all main steam line it olation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1280 psig if a neutron flux scram is assumed. This results in a 95 psig margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated has been divided into 70% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plant isolation transient (turbine trip with bypass valve failure to open) assuming a turbine trio 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 system pressure is limited to 1232 psig which is 143 psig below the allowed vessel overpressure of 1375 psig.
30 i553 070 Amendment No.
28
Table 3.2.3 MI2r* COOLItG SYSTEPS INSTPDENTATIoM THAT INITIATES O2 CONTROLS TnE CoAE AsiD CLN*AI?
Minimun No.
Re ma rks Operable Per Trio Level dettina Action Trin Sys (1L Pu nct ion Below trip setting initiated A
1.
2 Instrument Channel.
3 4uyo*above vessel sero.
H PCI.
Reactor Low Water Level A
1.
2 Instrumeat channel -
2 MO*above vessel zero.
Reactor Low Water Level Multiplier relays initiate RCIC.
Below trip setting initiates A
1.
Instrument channel -
2 378* above vessel sero.
CSS.
Multiplier relays 2
Reactor Low Water Level initiate LPCI.
(LIS-3-58A-D, SW t il Multiplier relay from CSS 2.
initiates accident signal (151 A
1.
Below trip settings in 2(16)
Instrument chanael -
2 378* above vessel sero.
conjunction with drywell hign Reactor Low Water Level pressure, low water level (LIS-3-584-D, SW 02) permissive,120 sec. del timer and CSS or RSA pump running, initiates ADS.
30 A
1.
Below trip setting permissive 1(16)
Instrument channet -
2 544' above vessel sero.
for initiating signals on ADS.
Reactor Low Water Level Permissive (LIS-3-184 5 18 5, SW 81)
Instrument channel -
2 312 5/16* above vessel sero.
A 1.
Below trip setting prevents inadvertent operation of Reactor Low Water Level (2/3 core height) of containment spray during 1
ILITS-3-52 & 62. SW t il accident condition.
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m nt No.
28
Table 3.2.B N
INSTRUMENTATION TBAT INITIATES OR COffrROLS TliE Coke AG CONTAINMENT CCCLI'3G S h
S 25 m
Mini: Pun No.
Remarks Action Operable Per Trio Level Settina
,E Trip sys til Function A
1.
Below trip setting permissive f or opening CSS and LPCI admission Instrument Channel -
450 psig. 15 2
Reactor Iow Pressure valves.
(PS-3-74 A & B, SW 82)
(PS-68-95, SW 42)
(PS-68-96, SW 82)
A 1-Recirculation discharge 2
Instrument Charu el -
230 psig i 15 valve actuation.
Reactor Low Pressure (PS-3-74A & B, SW 81)
(PS-68-9 5, SW 81)
(PS-6 8-96, SW 81)
A 1.
Below trip setting in Instrument channel -
100 psig
- 15 conjunction with containment 1
Reactor Low Pressure isolation signal and both (PS-68-93 & 94, SW 81) suction valves open will close RER (LPCI) admission valves.
S B
1.
With diesel power 65ts8 secs.
2.
One per motor 2
Core Spray Auto Sequencing Timers (5)
B 1.
With diesel power LPCI Auto Sequencing 05ts1 sec.
2.
One per motor 2
Timers (5)
A 1.
With diesel power 1
RHRSW A1, B3, C1, and 135ts15 sec.
2.
One per pump D3 Timers
---s Ln U
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g Table 3.2.8 g
INSTRtMENTATION TBAT INITIATES OR CONTROIE THE CORE AND CONTAINMENT COOLING SYSTEMS 3
CL 23 (D
3 Minieran No.
Operable Per Trip SYS (1)
PJnct ion Trip Level Settino Action Re: marks
,2 Core Spray and LPCI 05t51 sec.
E 1.
With normal power to Auto Sequencino 65tS8 sec.
2.
one per CSS motor W
Timers (6) 125ts16 sec.
3.
ho per RHR motor 185ts24 sec.
1 RUSW A1, B3, C1, and 275tS29 sec.
A 1.
With normal power D3 Timers 2.
One per pump O
1(16)
ADS Timer 120 sec
- 5 A
1.
Above trip settirq in conjunction with low reactor water level, high drywell pressure and LPCI or CSS pumps running initiates ADS.
2 Instrument channel -
100 + 10 psig A
1.
Below trip setting defers ADS RHR Discharge Pressure actuation.
Ln U
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Table 3.2.B INSTRINENTATION IBAT INITIATES OR COtfrROLS THE CORE AND CO*r*AINMENT COOLING SYSTEMS G-5s r'
Miniaun No.
=
OperaDie Per O
Trip Sys (1) runct ion Trio Level Setting Action Remar ks 2
Instrument Channel 185 1 10 psig A
1.
Below trip setting defers ADS a ct uat ion.
ng CSS Pump Discharge C3 Pressure S
1 ( 3)
Core Spray sparger to 2 paid + 0.4 A
1.
Alarm to detect core spray sparger pipe break.
Reactor Pressure vessel d/p 1
1.
Monitors availability of power to logic systems.
bus power monitor LT1 LN CD N
4
Table 3.2.B INSTRUMEtrTATION THAT INITIATES OR CO!frROLS THE CORE AND CONTAINME!rr COOLIPR3 SYSTEMS
]
&5=
ct Minipun N 3.
o Trip Sys (1)
Punct ion Trio Level Sett ing A ct ion Remarks Operable Per 2
2 (2)
Instrument Channel -
$583* above vessel zero.
A 1.
Above trip setting trips HPCI turbine.
na Reactor High Water Level CD A
1.
Above trip setting isolates NPCI 1
Instrument channel -
5 90 psi (7) system and trips HPCI turbine.
HPCI Turbine Steam Line High Flow A
1.
Above trip setting isolates e (4)
Instrument Channel -
5200*F.
HPCI system and trips HPCI BPCI Steam Line Space t urbine.
High Temperature B
1.
Includes testing auto 1
Core Spray System Logic N/A initiation inhibit to Core Spray Systems in other units.
B 1.
Includes Group 7 valves.
1 RCIC System (Initiating)
N/A Refer to Table 3.7.A for Logic list of valves.
B 1.
Includes Group 5 valves.
O 1
RCIC System (Isolation)
N/A Refer to Table 3.7.A for Logic list of valves.
N/A A
1(16)
ADS Logic M/A B
1 RER (LPCI) System (Initiation)
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- 10. Only sne-t ri p system for each cooler fan.
- 11. In only two of the four 4160 V shutdown boards.
See note 13.
- 12. In <>nly on<* of the f our 4160 V shutdown boards.
See note 13.
- 11. An emorgency 4160 V shutdown board is considered a trip synt.em.
- 14. HilR:;W pump would be inoperable.
Refer to section 4.5.C for the requirements of a RHRSW pump being inoperable.
- 15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water level (2 378" above vessel zero) originating in the core spray system trip system.
- 16. The ADS circuitry is capable of accomplishing its protective dCtion with one operable trip system. Therefore one trip system may be taken out of service for f unctional testing and calibration for a period not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly.
If the test period for one RPT system exceeds 2 consecutive hours, the system will be declared inoperable. If both RPT systems are inoperable or if 1 RPT system 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 />.
1553 076 Amendment No.
28 75
TABLE 4.2.B SURVEILLANCE REQUIREMENTS FCR INSTRtMENTATION THAT INITIATE OR CONTROL THE CSCS g
m C'
Function Funct ion al Test Calibrat_ ton In s t ru ment Check s
2 P
Instrument Channel (1) once/3 months none Reactor Low Pressure (PS-3-74A & B) hj (PS-68-95)
(PS-6 8-96)
Instrument Channel (1) once/3 months none Reactor Low Pressure (PS-68-93 6 94)
Core Spray Auto Sequencing Tipers (4) once/ operating cycle none (Normal Power) core Spray Auto Sequencing Timers (4) once/ operating cycle none (Diesel Power)
LPCI Auto Sequencing Timers (4) once/ operating cycle none (Normal Power)
LPCI Auto Sequencing Timers (4) once/ operating cycle none w
(Diesel Power)
RHRSW A1, 83, Cl, D3 Timers (4) once/ operating cycle none (Mormal Power)
RERSW Al, B3, Cl, D3 Timers (4) once/ operating cycle none (Diesel Power)
LT1 u
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F to TABLE 4.2.B 3
Q SURVEILLANCE REQUIREMENTS TCR INST *UMDff ATION TliAT INITIATE CR CONTROL THE CSCS ros Punction Fmetional Test Calibration Inst ru men t Check I"
O ro CD ADS Tisar (4) once/ operating cycle none Instru:nent Channel (1) once/3 months none RRR Pump Discharge Pressure Instrument Channel (1) once/3 months none Core Spray Pump Discharge Pressure Core Spray Sparger to RPV d/p (1) once/3 months once/ day Trip System Bus Power Monitor once/ operating cycle N/A none Instrument Channel Condensate Stcrage Tank Low level (1) once/3 months none LT1 U
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m TABLE 4.2.B
{
SUFvEIGN0Z REWIEEMENI5 EGA IS3TPUMENTATICS T;IAT INITIATE OF CO*7 TROL THE CSCS s
m
- 3
- unction Fuictional Test Calibration Instrument Check et O.
N LPCI (Containment Spray) Ingic once/6 months (6)
N/A Co Core Spray Ioop A Discharge N/A once/6 months once/ day Pressure (PI-75-20)
Core Spray Loop B Discharge N/A once/6 months once/ day Pressure (PI-75-4 8)
RER Inop A Discharge Pressure N/A once/6 months once/ day (PI-74-51)
RER Ioop B Discharge Pressure N/A once/6 months once/ day Q
(PI-74-65)
Instrument Channel -
Tested during N/A N/A RER Start functional test of RER pump (refer to section 4.5.B).
Instrument channel -
once/ month once/6 months N/A
$ Thermostat (RHR Area Cooler Fan)
Instrument Osannel -
Tested during N/A N/A core Spray A or C Start functional test of mre spray (refer to section 4.5.A).
Instrument Channel -
Tested during N/A N/A M
Core Spray B or D start functional test of core spray (refer (n
LeJ CD NW
9 F
to TABLF 4.2.B E
SOFVEILLANCI PUJIBEMENTS FOR INSTRUMDrTATIGN InAT INITI ATE OR CONTROL THE CSCS ct (D
3 Calibration Instrusent Chec4 et Funct iona 1 Test Punction 2o to section 4.5.A).
N/A once/ month once/6 months N
Instrument Channel -
Thermostat (Core Spray Area Cooler Fan)
N/A N/A Tested during RHR Area Cooler Fan *mic f unctional test c'f instrument channels, RER motor start and the rtnosta t (RER area cooler fan).
No other test required.
N/A Tested during logic N/A Core Spray Area Cooler Fan Logic system functional test of instrument channels, oore spray motor start and thermo-stat (core spray area cooler f an). No other test required.
F,A Tested during functional N/A e
m Instrument Channel -
test of core spray pump Core Spray Motors A or D Start (refer to section 4.5. A).
N/A Tested during functional N/A Instrument channel -
test of core spray pump Core Spray Motors a or C Start (ref er to section 4.5. A).
(b
!!/A M
!/A once/ month RPT initiate : i:
N/A b
once/operat.ing cycle N/A t= a l RPT breaker c--
~
E W
2EED W
P u
CD CD CD
a]
~h3A1
}
=
s an1 ilPe t,
and tr ips t h" rec 2rculation pumps.
The low reactor water level inst rumen ta tion that in set to trip when reactor water level is 17.7" (118" above vessel zero) above the top of the active fuel (Ta ble 1. 2. B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation and starts the diesel generators.
These trip satting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that post accident cooling can be accomplished and the guidelines of 10 CPR 100 will not be violated.
For large breaks up to the complete circumfen ential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to ceet the above criteria.
The high drywell pressure instrumentation is a diverse signal to the wate level instrumentation and in addition to initiating CSCs, it causes isolation of Groups 2 and 8 isolation valves.
For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level inttrumentation; thus the results given above are applicalbe here also.
Venturin.a r e provided in the main steam lines as a means of mea suring steam flow and also limiting the loss of mass inventory trom the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a break in the main utoam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 10000F and release of radioactivity to the environs is well below 10 CFR 100 quidelines.
Reference section 14.6.5 FSAR.
Tempo rat un e monitoring in9trumentation is provided in the main stoam lino tunnel to dotect leaks in these areas.
Trips are provided on t h i t, inntrumentation and when exceeded, cause closure 01 inolat ion valven.
The setting of 2000F for the main steam 1ino tunnel detector is low enough to detect leaks of the order ut 15 ypn; thus, it is capable of covering the entire spectrum of breaks.
For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
tiiyh radiation monitors in the main steam line tunnel have been provided to detect gross fuel f ailure as in the control rod drop accident.
With the established setting of 3 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 quidelines are not exceeded for this accident.
Reference Section 14. 6. 2 FSAR.
An ala rm, with a nominal uet point of 1.5 x normal full power background, i s provided also.
Amendment No.
28 109
D9*
- D
~ 9 3[
o In tbe asuiyt scal treatment ut the t t anutentn, 19 0 milliseconds at.
.illowed te'. ween a neutron sensor reaching the scram point and the start of n egat ive reactivity insertion.
Thin is adequate and conservative compared to the typically observed time delay of when about 270 milliseconds.
Approximately 70 milliseconds af ter neutron tlux reaches the trip point, the pilot scram valve solenoid power supply voltage goes to zero an approximately 200 milliseconds later, control rod motion begins.
The 200 milliueconds are included in the allowable scram insertion times specified in Spacification 1.J.C.
In crder te perform sernm time testing as requires! by ulecifichticn in the rui sequence l. 3.C.1, the relaxutiori of certain restraints Individual rod bypass switches any be i
control system ir. required.
used no described in specification h. 3.C.I.
The ponition or nni rod bypanned must l e known t o be in accordance wi th rod wi t tnienwn ! ceque me.
liypunning of ri-h in t he nonner deteribed in s pec i rlen t. b n h. i.t'.1 w i t ! nilow the nubnequent wi t hdenwn! of nny r..d 100 1,e rc en t to 50 percent rod denuity groupu, however, nernmrrd in the to it vil1 n.aintain rroup notch control crer all rods in the 50 percent In addition. RSCS will nrevent movement O percent rod density groups.
of rods in the 50 percent density to a preset power level range until the scrammed rod has been withdrawn.
D.
Reactivity Anomalies Durino each f uel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.
The magnitude of this excess reactivity may be inferred from the critical rod configuration.
As fuel burnup progresses, anomalous behavior in the excess r eact ivit y may be detected by comparison of the critical rod pa t t o r n a t selected base states to the predicted tod invent or y at that state.
Power operating bane conditions pr ovi ele t h" mos t nennitive and ditectly interpretable data relative to cor e react ivity.
Furthermore, using power operatinq bane condittons permits trequent reactivity compa r i fionn.
reactivity comparison at the specified frequency Requiring a assures that a comparison will be made before the core reactivity change exceeds 1% AK.
Deviations in core reactivit y grea ter t han 1% AK are not expected and require thorough evaluation.
One percent reactivity limit is considered safe since an insetrion of the reactivity into the core would not lead to transients exceeding design conditions 01 the reactor system.
References General Electrie "upplemental Reload Licensing Submittal for 1.
!WUP unit 3 Helond 2, NEDO-24199, July 1979.
136 Amendment No. 28 1553 082
LJ JJ
_3 LIMITING CONDITIONS FOR OPEP.ATION SURVEILLANCE RFQUIREMENTS
- 3. 5 CQRE AND COLGAIUMEE 4.5 CORE AND CONTAINMENT COOLING COOLING SYMS SYSTEMS b.
I:"sidual Heal Removal B.
Residual Heat Removal System (RHRS) (LPCI and
']y st em iRHRS) (LPCI and Containment Cooling)
Containment Cooling) 1.
The RHRS shall be 1.
- a. Simulated once/
Automatic Operating operable:
Actuation Cycle Test (1) prior to a reactor startup
- b. Pump Opera-Once/
from a Cold bility month Condition; or (2) when th=re is
- c. Motor Opera-Once/
ted valve month irradi.it ed t uel operability in the reactor vessel and when the reactor d.
Pump Flow Once/3 vessel pressure Rate Months is greater than a t.nos pheric,
- e. Testable Once/
except as check valve operating specified in cycle specifications 3.5.B.2, through Each LPCI pu=p shall deliver 3.5.B.7 and 9,000 gpm against an indicated 3.9.B.3.
system pressure of 125 psig.
'Ivo LPCI pumps in the same loop shall 2.
With the reactor deliver 15,000 gpm against an vessel pressure less indicated system pressure of than 105 psiq, the 200 psig.
RHR may be removed trom nervice (except 2.
An air test on the drywell and torus that two HHR pumps-containment cooling headers and nozzles shall be 1
mode and associated conducted once/5 years. A heat exchangers must water test may be performed on remain operable) for the torus header in lieu of the period not to air test.
a exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while being drained of 15{ O8s3 i
v Amencimeiit fic.
23 149
LIMITING CONDITIONS FOR OPERATION CURVEILLANCE REQUIRE!!ENTS 4.5 ggRE_MD_GQMblgEtfr COOLIUg 3. 's CORE AND CONTAINMENT COOLING SYSTEMS SYSTEd@
suppression chamber quality wat er and filled with primary coolant quality water provided that during cooldown two loops with one pump per loop or one loop with two pumps, and associated diesel g en e ra to r s, in the core spray system are operable.
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 exceed seven days required, the provided the remaining RHR pumps remaining RHR pumps (LPCI mode) and (LPCI mode) and both active components in access paths of the both access paths of RilRS (LPCI mode) and the RHRS (LPCI mode) the CSS and the and the CSS and the diesel generators diesel generators remain operable.
shall be demonstrated to be operable 4
If any 2 RilR pumps (LPCI immediately and daily mode) become inoperable, thereafter.
the reactor shall be placed in the cold shutdown condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1553 084 ISO AmendmenL No. 28
I I,I M I T I Nf. CONDITIONS FOR OPERATION SURVEILLANCE RFQt'I REMENTS
- 4. 5 CORE AND CONTAINMEtTr COOLING 3.5 CORE AND CorTTAINMENT SYSTEMS (UOLING_ SYSTEMS S.
If one R"3 pump 4
No additional surveillance (containment cooling required.
Mode) or associated heat exchanger is inoperable, the reactor may remain in operation for a period not'to exceed 30 days provided the remaining RHR pumps (containment cooling mode) and annociated heat exchangorn and dienel qenet.ators and all accenn paths of t he RilRS (containment cooling mode) are D
I]
@' g ~ p
- }r operable.
G'b v[ u
_ h
,,i( "A I
6.
If two RilR pumps (containment cooling mode) or associated heat exchangers are iaoperable, the reactor may remain 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 remaining RilR pumps heat exchanger is (containment cooling inoperable at a time mode) and associated when operability is heat exchangers and required, the all access paths of remaining RHR pumps the RHRS (containment (containment cooling cooling mo le) are mode), the associated operable.
heat exchangers and diesel generators, and all active components in the access paths of the RRRS (containment cooling mode) shall be demonstrated to be operable immediately and weekly thereafter until the inoperable RHR pump (containment c
ling Mode) and Amendment No. 28 151 associated heat 1553 085
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.5 CORE AND CONTAINMENT COOLING 3.5 COh%_AND CONTAINMENT SYSTEMS COOLING SYSTEMS second operable 8.
If specifications access path for the 3.5.B.1 through same phase of the 3.5.B.7 are not met, mode (drywell sprays, an orderly shutdown snall be initiated suppression chammer sprays and and the reactor shall suppression pool be shutdown and cooling) shall be placed in the cold demonstrated to be condition within 24 operable daily hours.
thereafter until the second path is 9.
When the reactor returned to normal vessel pressure is service.
atmospheric and irradiated fuel is in 8.
No additional the reactor vessel at surveillance least one RHR loop requir ed, with two pumps or two loops with one pump per loop shall be 9.
When the reactor e
operable. The pumps vessel pressure is associated diesel atmospheric, the RHR generators must also be operable, pumps and valves that are required to be 10.
If the conditions of operable shall be demon-specification 3.5. A.5 strated to be operable are met, LPCI and monthly.
containment cooling are not r eq uired.
10.
No additional Snrva111pne,
- 9"II'U*
11 When there is irradiated fuel in 11.
The B and D RHR pumps the reactor and the on unit 2 which reactor vessel supply cross-connect pressure is greater than atmospheric, capability shall be unit 2 RHR pumps B operable monthly when and D with associated the cross-connect heat exchangers and capability is required.
valves must be operable and capable 12.
When it is determined of supplying cross-that one RRR pump or connect capability associated heat except as specified exchanger located on in specification the unit cross-connection 3.5.B.12 below.
in the 153 Amendment No. 28 1553 086
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREHENTS 4.5 CORE AND CCNTAINMENT i:OOLIrv; l.5 (SR E MJD COfff AINAENT SYSTEMS
[OOLl.3G SJSTEMS (Note:
Because adjacent usit is cross-connect inoperabl-at a time capability is not a when operability is required, the r equir ement, a remaining RHR pump component is not and associated heat considered inoperable exchanger on the unit if cross-connect cross-connection and capability can be the associated diesel
~
restored to service generator shall be within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)
demonstrated to be operable immediately 12.
If one RHR pump or and every 15 days associated heat thereafter until the exchanger located on inoperable pump and the unit cross.
associated heat connection in unit 2 exchanger are is inoperable for any returned to normal reason (including service.
valve inoperability, pipe break, etc.),
the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pump and associated diesei generator are required.
13.
If RHR cross-14.
All recirculation pump connection flow or discharge valves shall heat removal be tested for operability capability is lost, during any period of the unit may remain in operation for a reactor cold shutdown period not to exceed exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
10 days unless such if operability tests capability is have not been performed rest red, during the preceeding 31 days.
14.
All rec!roulation pump discharge valves shall be operable prior to reactor startup (or closed if permitted 154 elsewhere in these Specifications).
1553 087 Amendment No. 28
~
if " D f f Elf 4
A d
[CW j
( f' a
u_
1,IMITING NNDITIONS FOR OPERATION SURVEILLANCE RFQUIREMENTS 4.5 G9BE_A._UD_G M bib 5 M _@ llU9 3.5 CORI.AND CONTAINMENT EIEIldE C001.1 ha' SYSTEMS and cort esponding.ic t lott shali cont ini:e unt il reactor operation is within the prescribed limits.
K.
Minimum Critical Power Ratio (MCPR)
The MCPR operating limit is K.
Minimum Critical Power 1.28 for 8x8 fuel, and 1.25 Ratio (MCPRI for 8x8R fuel, and 1.26 for P8x8R fuel. These limits MCPR shall be determined apply to steady state power daily during reactor power operation at rated power and operation at 2 251 rated flow.
For core flows other thermal power and than rated, the MCPR shall following any change in power level or be greater than the above distribution that would limits times K. K is the 7
7 cause operation with a value shown in Figure 3.5.2.
limiting control rod 11 at any time duriny, pattern as described in operation, it in deter-the bases for mined by normal surveillance Specification 3.3.
that the limiting value for MCPR is being exceeded,
. action chall 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 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limitn.
1,
_ Reporting Requirementn 11' any of the Iimiting values identified in Specifications 3.5.I, J, or K are exceeded and the specified remedial action is taken, the event shall 1CC7 nOO 1JJJ UOO be logged and reported in a 30-day written report.
167 Amendment No.
28
D
- 0
'D
' 3
~
_oo o
.5 s, ', P A:m.
.viquate core cooling.
With due regard for this margin, the allowable repair time of 7 days was chosen.
Should one RHR pump (LPCI mode) becomelinoperable, caly 3 RHR pt:.a.rps (LPCI mode) and the core spray system ere available. Since this leaves only or.e IUG pump (LFCI mode) in reserve, which along with the remaining 2 RHR pu:nps immedistely (LPCI mode) and core spray system is demonstrated to be operable and daily thereafter, a 7 day repeir period is jus *1fied.
Should two RHR pumps (LPCI mode) Lecome inopera' ole, tnere remains no reserve (redundant) capacity within the PRES (LPCI moce).
Therefore, the af fected unit.snall be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Shoul<1 one RHR pump (containment cooling mode) become inoperable, a complement of three full capacity containment removal systems is still available.
Any two of the heat remaining pumps / heat exchanger combinations would provide than adequate containment cooling for any abnormal or more accident situation.
Because of the availability of post in access of normal redundance requirements, which equipment is demonstrated to be operable immediately and with specified subsequent performance, a 30-day repair period is justified.
Shodld two RHR pumps (containment cooling mode) become inoperable, a f ull heat removal system is still available.
The remaining pump / heat exchanger combinations would provide adequate Containment Cooling for any abnormal post accident situation.
Because of the availability of a full complement of heat removal equipment, which is demonstrated to be operable immediately and with specified performance, a 7-day repair period is iustified.
Observation of the stated requirements for the containment cooling mode.nasures that the suppression pool and the drywell will be nufficiently cooled, following a loss-of-coolant accident, to prevent primary containment The containment cooling function of the overpressurization.
is permitted only af ter t he core has reflooded to the NHRS two-thirds core height level.
This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment cooling.
The two-thirds core height level interlock may be manually bypassed by a keylock switch.
Since the RHRS is filled with low quality water during power operation, it is planned that the system be filled with demine ralized (condensate) water before using the shutdown cooling function of the RHR system.
Since it is desirable to I553 089 169 Amendment No.
26
s.'
P A:.f N tent 2nq t o ensu re-that the lines ar e filled.
The visual checkinq will avoid starting the core spray or RHR system with a discharge Line not filled.
In addition to the visual observat ion and to ensure a filled discharge line other than pressure suppression chamber head tank is to testing, a prior located approximately 20 feet above the discharge lineThe highpoint to supply makeup water for these systems.
conden sa te head tank located approximately 100 f eet above the discha rge high point serves as a backup charging system when suppression chamber head tank is not in service.
the pressure System discharge pressure indicators are used to determine the water level above the discharge line high point..The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.
When in their normal standby condition, the suction for the llPC I and FCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCI'i and PCICS piping. This assures that the HPCI and RCIC di*icharge pipinq temains 1111ed.
Further assurance is provided by observing water flow from these systems high poin tu mont hl y.
I.
Maximum Averaqe Planar Linear Heat Generation Rate (MAPLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all tne rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variationn in power distribution within a fuel assembly attect the calculated peak clad temperature by less than i 200F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFH 50 Appendix K limit.
The limiting value in Tables 3. 5. I-1. -2 3.
The analyses for MAFLHGH in nhown n up po r t. I ng t.hm c Iimit Ing values in prenented in NITO-Old 27 and NilHb24194.
.r.
Linear iteat Generation Pate_(MLQF1 This specif ication assures that the linear heat generation rate in any rod is less than the design linear heat 176 Anendment No. 28 1553 090
t
It must be recognized that there is always an action which would return any of the pa rame te rs (MA PLHG R, LHGR, or MCPR) to within prescribed limits, namely power reduction.
Under most circumtances, this will not be the only alternative.
M.
Heferences
" Fuel Di nsit ication Et f ects on General Electric Boiling 1.
water Reactor Fuel," Supplemente 6, 7, and 8, NEDM-10735, August 1973.
2.
Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staf f).
3.
Communica tion:
V.
A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, P.s rch 27, 1974.
4.
G:neral Electric Supplemental Reload Licensing Submittal for BFNP Unit 3 Reload 2, NEDO-24199.
1553 091 178 Anendment. No. 28
TABLE 3.5.I-l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: SF-3 Fuel Type: Inital Core - Type 2 Average Planar Exposure MAPLHGR (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 30,000 10.2 TABLE 3.5.I-2 MAPLHGR VERS 1'S 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
'i,000 11.8 10,000 12.1 15,000 12 3 20,000 12.1 25,000 11 3 30,000 10.2 181 Amendment NO. 28
TABLE 3.5.I-3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FUEL TYPES: 8DRB265L Plant: BF-3 and P8DRB265L Average Planar MAPLHGR Exposure (Hwd/t)
(kW/ft) 11.6 200 11.6 1,000 5.000 12.1 10,000 12.1 12.1 15,000 20,000 11.9 25,000 11.3 10.7 30,000 The values in this table are conservative fer both prepressurized and non-pressurized fuel.
Amendment No.
28 182
()hjh
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
- l. 6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTF.M DOUNDARY F.
J et Pump Flow Mismatch F.
Jet Pump Flow Mismatch 1.
Recirculation pump speeds shall be checked and loqqed at least once per day.
1.
The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With the reactor operating, if one recirculation loop is out of service, the plant shall be placed in a hot shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to service.
2.
Following one-pump operation, the discharge valve of the low speed pump rmy not ha opened unlenn the npoed of the f aster pump is less than 50% of its rated speed.
195 Amendment No. 28
O D
D 3
on5 J
AX'n LIMITING CONDITIOND FOR OPERATICN SURVEILLANCE RTOUIRCMINTS L'EltkPI_lLP.inM_lY)'J:T"X
".6 Enit>aELgnTrti_touNDARY 36 c.
s t ruc t o rit n t enrity l
- 3. Ste.idy state cperati n utth huth recirculation pu ps out of ser-vice for up to 12 hrs is per-1.
Tabl e
- 4. 6. A together with cupplemen t a ry mitted.
During so.h interval notes, specifies the restart of the retirculatica inservice inspection pump'; is permitted, provided the
""fV? illa"C' loop disc.harge tteperature is requiremento f the nlithin 75 F of the saturatica reactor coolant 0
terrperature of the reactor cystem as f ollows:
vessel water as deter'nined by dome pressure. The total a.
areas to be clapsed tue in n3tural circula-inopected tion 3r.d cr.e pu p cperciticn
,2 L
b.
percent of areas be ne greater than 24 hrs.
~ '
to be inspected during the inspection in te rval G.
Stru m r_at_ Int *qEltY 1.
The structural c.
inspection integrity of th-frequency primary s ystert shall be maintain **d at th e d.
methods used for level required by th i innpection original acceptance standards throughout 2.
Evaluation of the life of th e inservice inspections plant. T:,e reactor will be made to the shall be maintained acceptance standards in a cold shutdown opecified for the condition until each original equipment, indication of a defect has Leen 3.
The inspection invectigated and interval chall be 10 evaluated.
Years.
4.
Addi t ional inspactions shall be perf ormed on certain circtmf erential pipe welds as linted to provido additional protection against pipo whip, which could damage auxiliary and control systens.
Teodwater-GrW-9, 1(IN-13, Grd-12, Giv-26 KIV-31, G7V-2 9, Krd-39, C19-15, Amendment No. 28 l'J 6 Kiv-Jo, and GIv-32 1553 095
a e 3.6/4.6 B AS ES 9 Target Rock And 2 Crosby Valves To meet the saf e ty design basis, thir teen safety-relief valves h tve been installed on, unit 3 with a total capacity of 81.08% of nuclear botier rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scran (valve position scram) results in a maximum vessel pressure of 1293 psig if a neutron fl ux scram is assumed This results in an 82 psig margin of the code allowable over-pressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 81.08% of nuclear boiler rated has been divided into 66.88% relief (11 valves) and 14.2~ safety (2 valves). The analysis of the plant ino-lation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is prescated in Reference 5 on page 29.
This analysis shows that the 11 relief valves limit pressure at the safety valves to 1218 psig, well below the setting of the safety valves. There f o re,
the saf ety valves will not open.
This analysis shows that peak system pressure is limited to 1243 psig which is 132 psig below the allowed vessel overpressure of 1375 psig.
11 Target Rock Valves Only To meet the safety design basis, thirteen safety-relief valves have been installed on unit 3 with a total capacity of 84.2% of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient.
(3-second clonure of all main steam line isolation valves) neglecting the direct scram (valve position scran) results in a maximum vessel pressure of 1280 psig if a neutron fl ux scram is assumed This results in an 95 psig margin of the code allowable over-pressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated has been divided into 707. 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 ectam 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 settirg of the safety valves. Therefore, the saf ety 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 psig.
225 Amendment No.
28 096
3.6/4.6 UAsrs in relief and safety valve operation shows that a Expe rienceof 50 percent of the valves per year is adequate to testing The relief and safety valves failures or deteriorations.
detect their are benchtested overy second operating cycle to ensure that percent tolerance.* The relief set points are within the 11in place once per operating cycle to establish valves are tested that they will open and pass steam.
The requirenents established above apply when the nuclear system These requirements can be pressurized above ambient conditions.
applicabic at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly are start at these conditions such that eventual overpressure relief would be n eeded.
However, these transients are much less severe, than those starting at rated conditions.
in terr.s of pressure, The valvee need not be f unctional when the vesnel head is since the nuclear system cannot be pressurized.
- removed, REFERENC ES 1.
Nuclear Systen Pressure Relief System 03FNP FSAR Subsection
- 4. 4)
- This in plus zero (+ 0 psi), minus 2% (- 22 pai) for Crosby valves 225a 155J (107 V//
Amendment No. 28
D**D D 'T l v dl o c.
3.h/4.6 sq1p A nozzle-ricer nyntem f a ilur e could also generat e the coinc ident t ailut e of a jet pump diffuser body; however, the converse is not t ue.
The lack of any substantial stress in the jet pump diffuser body makes f ailure impossible without an initial nozzle-riser system f ailure.
3.6.F/4.6.F Jet Pump Flow Mismatch Hequiring 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 f rom one to two pump operation that excessive vibration of the jet pump risers will not occur.
ECCS performance during reactor operation with one recirculation loop out of service has not been analyzed.
Therefore, sustained reactor operation under such conditionais not permitted.
i.6.G/4.6.G Structural Integrit_y The requiremento for the reactor coolant systems inservice i n spec t ion program have been icientified by evaluating the need f or a sampling examination of areas of high stress and highest protability of tailure in the e.ystem and the need to meet as closely as possible the requirements of Section XI, of the ASME iloiler and Pressure vessel Code.
The program reflects the built-in limitations of access to the reactor coolant systems.
Amendment No.
28 227 l a, 090 y
fR>TE S FOR TABLE 3.7.A Key:
0 = Open C = Closed 3C = Stays Closed GC = Goes Closed Note:
Isolation groupings are as follows:
Group 1:
Ti!e valven in Group 1 are actuated by any of the 101 lowing conditions:
1.
Reacto r Vennel Low Water Level (4708) 2.
Main Steamline liigh Radiation 3.
Main Steamline Iligh Flow 4
Main Steamline Space High Temperature 5.
Main Steamline Low Pressure Group /:
The valves in Group 2 are actuated by any of the f ollowi ng conditions:
1.
Reactor vessel Low Water Level (5 3 8")
2.
High Drywell Pressure Group 3:
The valves in Group 3 are actuated by any of the following conditions:
1.
Reactor Low Water Level ( 5 3 8")
2.
Reactor Water Cleanup System !!igh Temperature 3.
Reactor Water Cleanup System High Drain T em pera ture Gr.)up 4:
The valves in Group 4 are actuated by any of the i01 lowing conditLons:
1.
II PC I Steami t ne Space liigh Temperature 2.
HPCI Steamline High Flow 3.
HPCI Steamline Low Pressure Group 5:
The valves in Group 5 are actuated by any of the following conditions:
1.
RCIC Steamline Space High Temperature 2.
RCIC Steamline High Flow 3.
RCIC Steamline Low Pressure Group 6:
The valves in Group 6 are actuated by any of the f ollowing conditions:
1.
Reactor Vessel Low Water Level ( 53 8")
2.
liigh Drywell Pressure 1.
R ea cto r Huilding ventilation High Radiation Group 7:
The valves in Group 7 are automatically actuated by 1553 099 2"
Amendment No.
28
only the followirbi condition:
1.
Reactor Vennel Low Water Level (4 70" )
Group H:
Tho valves in Group 8 are automatically actuated by only the following condition:
1.
liigh Drywell Pressure 1553 100 267 Amendment No. 28
d
%.2,
-,...,,,.-.,n_.s. 3,,.
. O g C On :- l'a t ;,
,,... c.....II., D frw. sui-vs w.J 1
- W r.
we frJ.s Valve Val ms Test Identification Medium 84-83 Ccntci=ent i.tzce-re:-i 1.ilution Air N
8k-19 Centainment Atcosp' heric Eilution S
76-L9 Contai=unt At Osph:ric Monitor Air Air c.
e 70-50 Contairment Atzo pherie Menitcr Air s
76-51 Coc.si=ent Atc::pheric "enitor Air
[
76-52 Centc.incent Atr.ospheric Onitor Air 6
76-53 Ccatainuent Atec cheric i'anitor Air 76-54 Centair=cnt Atzchheric Mcaiter Air g
76-55 Con..i =ent At capheri: ::onitor Air 7 -50 Con ai:eent Atr;ccpheric Menitor Air 76-57 Contai=ent Atr_os,theric Monit:r Air 76-53 contairment Ataccpheric ?:oni;cr Air 76-59 Containment t.tnospheric Monitor Air 76-60 Contairecnt At:acspheric Monitor Air 76-61 Contairment Atacepheric Mcnitor Air N
70-62 Contai.nent Atnostheric Monitcr Air n
76-63 Contai=catAtmohhericMonitor Air 76-64 Conteirnent Atnospheric Monitor Air 76-65 centat=ent Atrocpheric Maniter Air
%g 70-67 Contnincent At cspherie ;:onitor Air 76-08 Containment Atmospheric Monitor Air 76-215 Contai=ent Attorp:: erie 21onitor Air b
76-217 Containnant Atmospheric Iionitor Air 75-220 Containi..ent Attc pheric Monitor Air 76-222 Contairment Atn.ospheric Moniter Air 76-225 Con minnent At cepheric Monitor Air 76-226 Ccntainment Atmospheric ?' nitor Air g
o 70-229 Ccntainment Atnorpharie Monitor Air pg 76-230 Centai=ent Attecrherie ::onitor Air c =n tn 76-237 Containment Atnospheric l'onitor Air en u
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TABLE 3.7.G CHECK VAL'.*ES ON IRWELL IliTLiKJC LINES Valve Te u t, Valves Identification Medium 3-554 Feedvater Air 3-558 Fee &mter Air 3-568 Feedwater Air 3-572 Feetater Air 63-525 stanc y Liquid Control Air Discharge 63-526 standby Liquid Control Air Discharge 69-579 ECU Return (Feedwater Line B)
Air l69-624 E'.iCU Return (Feedwater Line A)
Air 71-40 RCIC A7 Discharge Air 73-45 HPCI A p Discharge Air 85-576 CRD Hydraulic Return Air 281 1553 102 Amendment ko. 28
3.7.0/4.7.D Primary Conta inment Isolation valvan Double isolation valves are provided on lines penetrating the primary containment 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 integrity of the pressure suppression system.
Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.
process lines are isolated by reactor vessel low water Group 1 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 radiaticn, low pressure, or main steam space high temperature.
Group 2 - isolation valves are closed by reactor vessel low water level ( 53 8 ")
or high drywell pressure.
The group 2 isolation signal also " isolates" the reactor building and starts the standby ga s treatment system.
Ic is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.
Group 3 - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-saf ety related causes.
To protect the reactor from a possible pipe break in the system, i solaticn is provided by high temperature in the cleanup system area or high flow thro ugh the inlet to the cleanup system. Also, since the 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 consequences 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.
Grouc 6 - lines are 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 ventilatien high radiation which would indicate a possible accident and necessitate primary containment isolation.
G roup 7 - process lines are closed only on reactor low water j
level ( 470"). These close on the same signal that initiates HPCIS and BCICS to ensure that the valves are not open when HPCIS or RCTCS a ction Ln r equired.
Group 6-line (traveling in-core probe) is isolated on high drywell pres sure.
This is to assure that this line does not provide a leakage path when containment pressure indicates a possible accident condition.
1553 103 Amendment No. 28
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3. ')
AUXILIARY ELECTRICAL SYSTEM 14. 9 AUXILI ARY ELECTRICAL SYSTMj d.
Each diesel generator shall be given an annual inspection in accordance with instructions based on the manufacturer's recommendations.
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 AS E D975 and logged.
2.
Three unit 3 diesel 2.
D.C.
Power system -
Unit Batteries (250-generators shall be Volt) and Diesel operable.
Generator Batteries (125-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.
1553 104 Amendment No. 28 310
I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.9 AUXILI ARY ELECTRICAL SYSTEM 3.9 AUXILIARY ELECTRICAL SYSTEM The undervoltage c.
relays which start the diesel generators f rom start buses 1A and 1B and the 4-kV shutdown boards, shall be calibrated annually for trip and reset and the measurements loqqed.
5.
The 250-Volt Shutdown Board battery and unit batteries and a battery charger for each battery and associated battery boards are operable.
6.
Logic Systems a.
Accident signal logic system is operable.
7.
There shall be a minimum of 103,300 gallons of diesel fuel in the unit 3 standby diesel generator fuel tanks.
321 Amendment No.
28
SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.9 AUXILIARY ELECTRICAL SYSTEM 1.9 AUXll.I ARY ELECTRICAL SYSTI11 4.
From and after flie dat e t hat the 2SO-Volt Slio t ihiwn board ont t er ica or
]'E -D '
3,,}
one of the thice M
D 2so-volt unit D
batteries and/or its b6 associated battery board is found to be inoperable for any reason, continued reactor operation is permissible during the succeeding seven days.
Except for routine surveillance testing, the NRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the situation, the precautions to be taken durinq this period anil the plans to return t he 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 C3cs requirements listed in Specification 3.9.B.2 are satisfied.
The NRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the situation, the precautions to be t aken ilus ing this period anal the plans to ret urn the failed 1553 106 compon ent to an operable state.
Amendment No. 28 325
DT M
j
). 9 BASES The ob lective of this specification is to assure an adequateelectrical power to ope source ot during shutdown and to operate the engineered safeguards There are three sources of alternating current following an accident.
namely, the 161-kV transmission system, the eleotrical encryy avsflable, nue ! car gar;arati ng unita, nnil the 11enel generators.
The 161-kV ot tait e power supply consista of two lines which are fed f rom dif f erent sections of the TVA 161-kV grid.
In the 161-kV system is operating and f our normal mo le of operation, the If one diesel generator is diesel qenerators are operational.there normally remain the 161-kV sources, and the out of service, For a diesel generator to be other three diesel generators.
considered operable its associated 125 V battery must be operable.
The minimum tuel oil requirement of 103,300 gallons is sufficient for 7 days of full load operation of 3 diesels and is supply.
conservatively based on availability of a replenishment Of feite auxiliary power for Prowns Ferry Nuclear Plant Unit 3 is supplied from unit station transformers from the main generator or the 161-kV two nourren: the t ransninolon nystem through the cooling tower transformers. If a cooling tower trans femor is Iont, the unit enn continue to operate since the station other cooling tower transformer is available, t rann fo rw r is in nervice, tho and four.tienet generntorn are operational.
A 4-kV shutdown boarti is allowe6 to be out of operation for a brief period to allow for maintenance and testing, providing all 4-kV shutdown boards and associated diesel generators remain ing CS, RHR, (LPCI and Containment Cooling) Systems supplied by the remaining 4-kV shutdown boards, and all emergency 480 V power boards are operable.
250-volt d-c battery systems each of which There are fivea battery, battery charger, and distribution consists of Three of these systems provide power for unit control equipment.
and alternative tunctions, operative power f or unit motor loads, 115-volt a-c unit preferred motor-generator drive power for a 250-volt d-c system provides power for common plant and act.
One tranumisulon nystem control functiono, drive power for a 115-volt and emergency drive a-c plant preferred motor-generator net, for certain unit large motor loads. The fifth battery system power delivers control power to a 4-kV shutdown board.
ic system is so arranged, and the batteries sized such, that The 250-Volt the loss of any one unit battery will not prevent the safe shutdown and in the event of the loss of offsite power and cooldown of all three units a desten basis accident in any one unit.
Insa of contral power to any engineered safeguard control 327 Amendment No. 28 1553 107 l
'. 's M A.JO P DESIGff F EAT f1P E3 5.1 SITE FFATUEES Il r own r. Ferry units 1,
2, and 3 are located at Browns Ferry tiuclear Plant site on property owned by the United States and in cuatody of the TVA.
The site shall consist of 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 con 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 rna y contain 764 fuel assemblies consisting of 8x8 assemblies having 63 fuel rods each, and 8x8 R (and P8x8R) assemblies having 62 fuel rods each.
The number of each type in the core is given in the roost recent reload amendment topical report.
B.
The reactor core shall contain 185 cruciform-shaped control rods.
The control material shall be boron carbide powder (B C). compacted to approximately 70 percent of theoretical 4
density.
5.3 R EACTOR VESSEL The reactor vessel shall 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 CONTAIllMENT A.
The principal design parameters for the primary containment shall be given in Table 5.2-1 of the FSAR.
The applicable denign codes shall be as described in Section 5.2 of the FSAR.
D.
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 be designed in accordance with the standards set forth in Section
- 5. 2. 3. 4 of the FSAR.
5.5
_ FUEL STORAGE A.
The arrangement of the fuel in the new-fuel storage tacilitity shall be such that k for dry conditions,
- eff, 360 Amendment No. 28 1553 108