ML18026B196
| ML18026B196 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/13/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Tennessee Valley Authority |
| Shared Package | |
| ML18026B197 | List: |
| References | |
| DPR-33-A-108, DPR-52-A-102, DPR-68-A-075 NUDOCS 8409040196 | |
| Download: ML18026B196 (50) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 108 License No.
DPR-33 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 9, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No.
DPR-33 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 1O8, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
840904019h 8408i3 PDR ADOCK 05000259 P
3.
This license amendment is effective as of the date of issuance.
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 13, 1984 FOR THE NUCLEAR REGULATORY COMMISSION
~<@&I~
~
~
~
Domenic B. Vassallo, Chief Operating Reactors Branch ¹2 Division of Licensing
ATTACHMENT TO LICENSE AMENDMENT NO.
108 FACILITY OPERATING LICENSE NO.
DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages.
7, 55, 58, 61, 70, 78, 112, 255, 266, 277 2.
The marginal lines on these pages denote the area being changed.
1.0 DE~r,.h'cmTIOHS (Cont 'd) 10 Lactic - A logic is an arrangement of relays,
- concacrs, and ocher componnncs that pzoduces a decision oucpuc.
r'a)
~lniciacln A logic char. receive signals fram channels and produces decision outputs to the actuation logic (b)
Actuation A logic that receives signals (either from initiation logic or channels) and pzoduces decision outputs to accomplish a procective action.
Functional Tescs - A functional test is che manual operation or iniciacion of a systems subsyscem, or components to vez fy.
r e if that it functions within design tolezances (c.g;, the manual start of a core spray pump ro, verify that it zuns and thac it pumps the required volume of water)
~
X.
ghacdoan - Thh raaeror is ia a shccdosn concision ahsn rhs rescror mode sairch is in che shacdean mode posicdoa and no core alterations arc being performed..
Y.
Enrineered Safccuard - An engineered safeguard is a safety 1 to a safery action system cne accions of which are essentia o
required in response to accidi:ts.
Re oi"ab1e Event - A reportable event shall be any of those conditions specified ia section 50.73 to 10'FR Pare 50.
s Surveillance interval - Each Surveillance Requirement shall be performed within the specified time interval with:
1.
A maximum allowable extension not to exceed 25% of the surveillance interval, but:
2.
The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times. the specified surveillance interval.
4 Amendment NO.
108
0 r
e 1
LX.!ITIIIG CO!IDITIONS FOR OP I'.RATION SUIVNI?.LANCE kl OUIREMENTS 3e1 REACTOR PROTECTION SYSTEM 4 ~ 1 REACTOIi PROTECTION SYSTEM A 1icabilitv A oli.cabilitv Applies to the instrumentation and associated devices which initiat a reactor scram.
Applies to the surveillance of the instrumentation and asso-ciated devices which initiate reactor scram.
~nt ective
~Oe ective To assure the operability of the reactor protection system.
To specify the type and frequency of surveillance to be applied to the protection instrumentation.
S ecification S ecification A,
When there is fuel in the vessel, the setpoints, minimum number of tr'p systems, and minimum number of instrument channels that must be operable for each position of the reacto" mode switch shall be as given in Table 3.1.A.
A.
Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4.leA and 4.1.B respec-tively.
B.
Two RPS power monitoring channels for each inservice'PS
."IG sets or alternate sourt r shall h>> oI1erabl e.
- 1. With one RPS electric power monitoring channel for inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel to operabl.e status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the associated RPS MG set or alternate power supply from service.
When it is determined that a channel is failed in the unsafe condition, the other RPS channels that.
monitor the same variably shall be functionally tested immediately before the trip system containing the failure is tripped.
The trip system containing the unsafe failure may be untripped for short periods of time to allow functional testing of the other trip system.
The trip system may be in the
~
untripped position for no more than eight hours per functional test period for this testing Amendment Non. ~108
Minixun So.
Instrunant Channels Operable Level Settin Aetio 1'ABLE 3 2 ~ A PRINART COWTAIlQKPT AND REACTOR BVILOIhG ISOIATION INSTRUME ATIOR Rerarks Instrument Channel 538" above vessel zero Reactor Lov Mater Level (6)
A or (B and r) 1.
Selov trip setting does the foiloving:
a.
Initiates Reactor Building Isolation b
Initiatez Primary Containment Isol~tion ((:roups 2, 3, and 6) c.
Ini.tiates MTS Instrument Channel-Reactor High Pressure Instrument Channel-Reactor Lov Mater Level (LZS 3-56A-Dy SM 6 1) 100 i 15 psig p 470" above vessel zero A
1 Above trip setting isolates the shutdovn coolin suction valves of the RHR system 1.
Belov trip setting initiates Hsin Steam Line Isolation Instrument Channel-Slgh Dryvell Pressure (6)
(PS 6% 56A D) 2.5 pais h or (B and B) 1.
Above trip setting does the follovlngz Initiates Reactor Building Isolation b
Initiates Primar'.t containment Isolation
. c.
Initiates SOTS 2 (3) instrument Ch~nnel-Bigh Radlatlan Main Steam Line Tunnel (6)
Instrument channel-Lov Pressure Main Stea~
Line Instrument channel-Siqh Plov Main Steam Line 3 times normal rated S
full pover background (]3) sl5 psig (a)
S 1e06 of rated steam IlcM B
1.
Above trip setting initiates.Hain Steam Line Isolation 1.
Selov trip setting initiates Mai'n Steam Line Isolation 1
Abov+ trip setting initiates Main Steam Line Isolation
)(inimum No.
Instrument "hannels Operable i'0)>>
Funct io Tri Level Settin Action TABLE 3.2.A PRIHARY COHTQIHHENT AHD REACTOR BUILDIHG ISOLATION IHSTRUHEHTATIOH Remarks 2
Group 2 (Initiatin ) Logic N/A A or (B and E)
Refer to Table 3.7.A for list of valves.
Group 2
(RHR Isolation-Actuation)
Logic Group 8 (Tip-Actuation)
Logic Group 2 (Drywell Sump Drains-Actuation)
Logic H/A H/A Group 2
(Reactor Building H/A 6 Refueling Floor, and Dry-well Vent and Purqe-Actuation) lngic Group 3 (Initiating) Logic ii/A P and G
1.
Part of Group 6 Logic.
1.
Refer to Table 3 7.A for list of valves.
Group 3 (Actuation) Logic H/A Group 6 Logic H/A F and G
1.
Refer to Table 3.7.A for list of valves.
Group 8 (Initiating) Logic H/A 1.
Refer to Table 3 7.A 'or list of valves.
2.
Same as Group 2 initiating logic.
Reactor Building Isolation N/A (refueling floor) Logic Reactor Building Isolation N/A (reactor cone)
Logic BorF B or G
or A
1.
Logic has permissive to refueling floor static pressure regulator.
1.
Logic has permissive to reactor zone static pressure r4.gulator.
6.
Channel shared by RPS and Primary Containment
& Reactor Vessel Isolation Control System.
A channel failure may be a
channel failure in each system.
7.
A train is considered a trip system.
8.
Two out of three SGTS trains required.
A failure of more than one will require action A and F.
9.
There is only one trip system with auto transfer to two power sources.
10.
Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.
11.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the trdpped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be operable for the channel to be operable.
Power operations permitted for up to 30 days with 15 of the 16 temperature'switches operable.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal back-ground at full power.
The allowable setpoints for alarm and rea'ctor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.
61
TAsLE 3.l.l (Contlnucd) tllabaun Ão.
Operable ter
~Trl 5
(ll 1(la) 1(10) l(11) l(11) tunction RHI Area Coolct Faa Loeic Core Spray Area Cooler Fan Losic lnatruaent Channel-Core Spray Hotora A or C Start Iaatruaeat Chaaacl-Corc Spray ltotota
~ or D Start Tri Level Settin W/A
- i/A il/A Action Rcaarka l.
Starta RHRSM punps Al, B3, Cl, and D3 Starta RHRSir pumps Al, B3, Cl, and D3..
1(ll}
1(12) lnatruacnt Chanaal-Corc Spray Loop 1 Accident Slsnal (1S) laatruacnt Chanacl-Corc Spray Loop 2 Accidcat Signal (1$ }
A l.
Starts RMRSM puaps Al, B3, Cl, and D3 A
1.
Starts RNtSM pueps Al< B3, Cl, and D3 l(13)
NgtSu laitlate Logic (14)
RPT logic r
N/A (17) l.
Trips recirculation pumps on turbine control valve fast closure or stop valve closure
> 30j pover.
TABLE 3 2. F SURVEILLA)1CE IHSTRDHEHTATIOH Hinimum I of Operable Instrument Channels Instrument I 46 h LI-3-46 B Instrument Reactor Hater Level Type Indication and Range Indicator - 1$ $"
to
+f0" Hotes tl) t2) t3)
PI-3-54 PI-3-61 Reactor Pressure Indicator 0"1500 psig t1) t2) t))
TI-64-52
".R-64-52 Drywell Temperature Recorder, Indicator tl) t2) t3) 0-4000F TR-64-52 Suppression Chamber hir Temperature Recorder 0-4004F t 1) t2) t3)
H/h PS-64-67 TR-64-52 and
?S-64-58 8 and IS-64-67 LI-84-2h LI-84-13A Control Rod Position Neutron Honitoring Drywall Pressuro Drywell Temperature and Pressure and Timer CAD Tank "A Level CAD Tank B" Lovel 6V Indicating
)
Lights
- SRH, IRH, LPRH 0 to 100% power Alarm at 35 psiq Alarm if temp.
) 2814F and prcssuf'e ) I
$ ~16 after 30 minute delay Indic~tor 0 to 100l Indicator 0 to 1001 t 1) t2) t3) t4) t 1) t2l t)l
3.2 BASES and trips the recirculation oumos.
The low reactor water level instrumentation that is set to trip when r eactor water level is 17;7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray
- Pumps, contributes to ADS initiation, and star ts the diesel generators.
These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.
For large br'eaks up tn the complete circumferential break of a 28-inch recit "ulation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.
The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addit'on 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 instrumentation;
- thus, the results given above are applicable here also.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The pr imary function of the instrumentation is to detect a break in the main steam 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 1000 F, and release of'adioactivity to the environs is well below 10 CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperatur e monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.
Tiips ar e provided on this instrumentation and;. when exceeded, cause closure of isolation'alves.
The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.
For large break~,
the high steam flow instrumentation is a backup to the temperature instrumentation.
High radiation monitors in the main% steam line tonnei have tean provided to detect gross fuel failure as in'the control rod drop I
accident.
Mith the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product, release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.
Reference Section 14.6.2 FSAR.
An alarm with a nominal setpoint of 1.5 x normal full-power background is pr ovided also.
Pressure instrumentation is pt'ovided to close the main steam isolation valves in Run Node when the main steam line pressur e drops belo~
825 psige 112 Amendment No.,
108
Croup 7:
The valvea tn Croup 7 aiba auteaatically actuated by only the follcwtng condition:
1.
Raactor veaaal lov vatar level (47Oee)
Croup 8:
The valvaa in Croup 8 are auteaatically actuated by only the follow tng condition:
1.
High Dryvell praaaura 2.
Reactor vessel low water level (538")
255 Amendment No. pd, 108
Table 3.7.H (Continued)
X-1078 X-108A X-108B X-109 X-110A X-110B X-a~9 Spare (tcstablc)
Power CRD Rod Position Indic.
CRD Rod Position Indic.
Power CRD Rod Position lndic.
Suppression Clsamber
'lacuum Breaker 266 Amendment No. 108
~Cree l - pcoeeas Lines ate isolated by teactot vessel lov vatet level (dso-}
Ln order to nlloby for teeaval of decay heat aubaediuent to a
~creel, ynt isolotr.
Ln cise for ptopet'pct'ation of the cote
~ tandby eorLLng aystesle.
The valves Ln group l are alao closed vhen ptoceaa Lnotruiiicntation detects exccnn Lve aain stean line flovb high tadiation, Lou ptconute, ot main steaIa apace high tenPpetatute,
~Grou I - Isolarion valves are closed by rcacror vessel low warer level (538") or high drywall pressure.
The group 2 isolation signal also "iso-1>>tc:;" thc reactor i>uiidi>>g and starts the standby gas treatment system.
Et is not dcsirabic to actuate the group 2 isolation signal by a transient ot spurious si >>al.
~Grnu 3 - Process lines are normally in use, and ir is cbereiore noc desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes.
To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system.
- Also, since the vessel could potentially be drained through the cleanup
- system, a low level isolation is provided.
Groups 4 and 5
Process lines are designed to remain operable and mitigate thc nnscquc>>ccs of an accident whicn results in the isolation of other process li>>cs.
The signals which initiate isolation of Groups 4 and 5
process lines are therefore indicative of a condition which would render them inoperable.
Gto~>>6 - Lines are connected to the primary containment but not directly to thc reactor vessel.
These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor building ventilation high radiation which would indicate a possible accidl;nt and necessitate primary containment isola cion.
~Gtou 7 - Process lines are closed only on the respective turbine steam
~supply valve no t fully closed.
This assures that the valves are no t open when liPCI or RCIC action is required.
('riis1P 8 - Li>>o (Lt>>v1 i i>>}, in-coro probe) is isolated on high drywcil prcssure or rcncLot low w>>tct level (538").
This is to assure that. this i i>>c docs not ptoviriu a leakage path when containment prcssure or reactor water level i>>dicatcs n possible accident condition.
s The maximum closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
In satisfying this design intent, an additional margin has been included in specifying maximum closure times.
This margin permits identification of degraded valve performance prior to exceeding the design closure times.
277 dasndment No. p9, 108
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROMNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
102 License No.
DPR-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 9, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No.
DPR-52 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. 102, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance.
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 13, 1984 FOR THE NUCLEAR REGULATORY COMMISSION st
)
I
.g~ r~~~ ~
l~
Domenic B. Vassallo, Chief Operating Reactors Branch P2 Division of Licensing
ATTACHMENT TO LICENSE AMENDMENT NO. 102 FACILITY OPERATING LICENSE NO.
DPR-52 DOCKET NO. 50-260 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages.
7, 55, 58, 61, 70, 78, 112, 231, 232, 25la, 255, 266, 277 2.
The marginal lines on these pages denote the area being changed.
1.0 D):.FD)LTlONS (Conr 'd) 10.
Logic A logic is an arrangement of relays,
- contacts, and other components that produces a decision output.
'ta) rnl~aiaain
- h lagla ahaa ranalsa signals iran channels and produces decision o<<tputs to the actuation logic.
(b)
Actuation - A logic that receives signals (either from initiation logic or channels) and'roduces decision outputs to accomplish a protcctivc action.
Functional Tests - A functional test is the manual operation or initiation of a system, subsystem, or componenrs to verify t a if that it functions within design roleranccs (c.gs g the manual start of a core spray pump ter verify that it runs and that it pumps thc rrq<<ircd volume of water).
X.
Slniedcwn - Thc rcacror is in a shutdown condition when the reactor mode switch is in thc shutdown mode position and no core altcxations arc being performed.
En> incered Safe.,uard - An engineered safeguard is a safety system rhe actions of which arc essentia'o a safety ac
'on required in rcsponsc to accidcntsa Za Reportable Event - A reportable event shall be any of those conditions specified in section 50.73 to 10 'CFR 'Part 50.
Surveillance Interval - Each Surveillance Requirement shall be performed within the specified time intc.rvnl. with:
l.
A maximum allowab Lc uxtensioif" not to uxcoud 25K of the surveillance interval, but:
2.
The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
Amendment Ho.
102
Miniz:un Ko.
Zns truczent Channels Operable 1
I'ABLE 3~2 A PRIMARy CoccTAIHMEHI'NO REACTOR BulLOING ISOIATIOH INSTRUMESTATIOM ema rks Instrument Channel Reactor Low Mater Level (6)
? 538" above vessel zero A or
{B and E)
Belov trip setting does the follcwing:
- a.
Initiates Reactor Building Isolat ion b
Initiatez Primary Containment Isolatio" (Groups 2, 3, and 6) c.
Initi'ates SOTS Instrument Channel-Reactor High Pressure Instrument Ch4nnel-Reactor Low 1cater Level
{LZS 3-56A-Og SM l1) 100 i 15 psig
? 470'bove vessel zero A
1.
Above trip setting isolates the shutdown cooling suction valves of the RHR system Belou trip setting initiates Main Steam Line Isolation Instrument Channel-High Oryvell Pressure t6)
(PS-64-56A"D)
Instrument Ch4nnel-High Radiation Main Steam Line Tunnel
{6)
Instrulent Channel-Low Pressure Main Steam Line 2.$ pIIS 3 times normal rated full power background (l3)
?
SpS psig la)
A or (8 and C) l.
Above trip setting does the folIcwingz a,
Znitiates Reactor Building Isolation b
Initiates Primar.r containmnt Zsolation
. c Initiates SGTS t.
Above trip setting initi4teS Main Steam Line Isolation 1.
Below trip setting initiates Mai'n Steam Line Isolation 2 (3)
Instrument Channel-Qigh Plow Main Steam Line S
100'L of raced steam fl~
8 1.
Above trip setting initiates Main Steam Line Isolation
Hinimum No.
Instrument n.
"hannels Operable TABLE 3 ~ 2 A
PRIttAR'C COttTAIltttEttT AND REACTOR BUILDItlG ISOMTJOtt IttSTRUttEttTATIOtt Funct io Tri toivel Settin Action Remarks O
2 Group 2 (Initiatin.) Logic tt/A A or (B and E)
Refer to Table 3.7.A for list o-valves.
Group 2
(RHR Isolation-Actuation)
Logic Croup Q (Tip-Actuation)
Logic Group 2 (Drywell Sump Drains-Actuation)
Logic tl/A D/A Group 2
(Reactor Building tt/A 6 Refueling Floor, and Dry-well Vent, and Purqe-Actuation) Logic P and G
1.
Part of Group 6 Logic.
Group 3 (In)tiatirg) Logic tt/A C
1.
Refer to Table 3.7.A for list of valves.
Group 3 (Actuation) Logic tl/A Group 6 Logic W/A FandG 1.
Refer to Table 3.7.A for list of valves.
Group 8 (Initiatinq) Logic tt/A 1.
Refer to Table 3.7.A for list of valves.
Reactor Building Isolation tt/A (refueling floor) Loqic Reactor Buxldi~ Isolation tt/A (reactor xone)
Logic ttor F DorG or A
2.
Same as Group 2 initiatinq loqic.
1.
Logic has permissive to refueling floor static pressure regulator.
1.
Logic has permissive to reactor xone static pressure regulator.
6.
Channel shared by RPS and Primary Containment 6 Reactor Vessel Esolation Control System.
A channel failure may be a
channel failure in each system.
7.
A train is considered a trip system.
8.
Two out of three SGTS trains required.
A failure of more than one will require action A and F.
9.
There is only one trip system with auto transfer to two power sources.
10.
Refer to Table 3.7.A and its notes for a listing of isolation Valve Groups and their initiating signals.
ll.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the trdpped condition provided at least, one OPERABLE channel in the same trip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be operable for the channel to be operable.
Power operati'ons permitted for up to 30 days with 15 of the 16 temperature'switches operable.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal back-ground wt full power.
The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.
Amendment No. p8, JYc., pA, 102 6l
Ktabaua Ao.
Operablc per T~r>
S
~ (1)
Function TAbLE 3.2.b (Continued)
Tri Level Settin Action Remarks l(10)
RHR Area Cooler Faa Loeic l(10)
Core Spray Area Cooler fan Loafe l(ll) lnatruocnt Channel-Core Spray Hotore A or C Start l(ll) laatruaeat Chaaacl-Corc Spray Notora b or D
Start
) (12)
Inatruacnt Chanaal-Corc Spray Loop 1 AccL4cnt Signal (1$ )
1(12) laatruncnt Chanael-Corc Spray Loop 2 Accf4cnt Signal (1$)
1(13)
RHRSM laltLate ~lc M/A
'l/A il/A W/A R/A 8/A' 1
Starte Rl{RSM pumps Al, B3, Cl, and D3 A
l.
Starte R)(RSM pumps Al, B3, Cl, and D3 A
l.
Scarce RHRSM pumps Al, B3, Cl, and D3 l.
Starts RHRSM pumps Al, B3, Cl, and D3 RPT logic (17) l.
Trips recirculation pumps on turbine control valve fast closure or stop valve closure.
> 30$ pover.
TABLE 3 2 F SURVEILLAHCE IHSTRlJMEHTATIOH Minimum g of Operable Instrument Channels 2
Instrument i U-3-46 A
LI-3-46 B PI-3-54 PI-3-61 PR-64-50 PI-64-67 TI-64-52 TR-64-52 TR-64-52 TI-64-55 TIS-64-55 LI-64-54 A LI 66 PS-64-67 TR-64-52 and PS-64-58 B'and IS-60-67 LI-84-2A LI-84-1 3A Instrument Reactor Mater Level Reactor Pressure Drywell Pressure Drywe1 1 Temperature Suppression Chamber Air Temperatur e Suppression Chamber Mater Temperature Suppression Chamber Mater Level Control Rod Position Neutron Monitoring Drywell Pressure Drysell Temperature and Pressure and Timer CAD Tank "A" Level CAD Tank "B" Le re 1 Type Indication and Range Indicater - l55" to
+6e" Indicator 0-1500 psig Recorder 0-80 psia Indicator 0-80 psia Recorder, Indicator 0-4004F Recorder'-400oF Indicator, 0-400~F Indicator -25" to p 25ll 6V Indicating Lights
)
- SRM, IRH~
LPRN 0 to 100% power
)
Alarm at 35 psig
)
)
Alarm if temp.
)
28'1oF and
)
. pressure
> 2 5 pzfg )
after 30 minute
)
delay
)
Indicator 0 to IOOi Indicator 0 to 100%
Hotes (1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(1)
(2)
(3)
(")
(1)
(21 (3)
(4)
3.2 BASFS and trips the reciv< <<1<<ti<<n n<<mna.
The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B ) initiates the LPCI, Core Spray
- Pumps, contributes to ADS initiation, and starts the d'esel generators.
These trip setting levels wer e chosen to be high enough to prevent spur ious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.
For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.
The high drywell pressur e instrumentation is a diverse signal to the water level instrumentation and, in addit'on 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 instrumentation;
- thus, the results given above are applicable here also.
Venturis are provided in the main steam lines as a means of measuring steam flow ard also limiting the loss of mass inventory from the vessel dur ing a steam line break accident.
The primary function of the instrumentation is to detect a break in the main steam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in con)unction 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
- 1000oF, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperatur e monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.
Trips are provided on this instrumentation
- and, when exceeded, cause closure of isolation valves.
The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it j.s capable of covering the entire spectrum of breaks.
For large
- breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in'the control t od drop acc'dent.
Mith the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.
Reference Section 14.6,2 FSAR.
An alarm with a nominal setpoint of 1.5 x normal full-power backgr ound is provided also.
Pressure instrumentation is provided to close the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psigo 112 Amendment No.
, 102
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 ~ 7 CONTAINMENT SYSTRiS 4 ~ 7 CONTAINMENT SYSTEMS Local Lehk rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation valves, which are not part of a water-sealed
- system, at not less than 49.6 psig (except for the main steam isolation
- valves, see 4.7.A,2.i) and not less than 54.6 psig for water-sealed valves each operating cycle.
Bolted double-gasketed seals shall be tested whenever the seal is closed after being opened and at least once per operating cycle.
Acceptable methods of testing are halide gas detection, soap bubbles, pressure decay, hydrostatically pressurized fluid flow or equivalent.
Amendment No.
102 231 The personnel air lock shall be tested at 6-month intervals at an internal pressure of not less than 49.6 psig.
In addition, ii the personnel air'ock is opened during periods when containment integrity is not
- required, a test at the end of such a period will be conducted at not less than 49.6 psig. If the personnel air lock is opened during a period when containment integrity is required, a test att2.5 psig shall be conducted within 3 days after being opened.
If the air look is opened more frequently than once every 3 days, the air lock shall be tested at least once every 3 days during the period of frequent openings.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
- 3. 7 CONTAINMENT SYSTEMS
- 4. 7 CONTAINMENT SYSTEMS The total leakage from all penetrations and isolation valves shall not exceed 60 percent of L per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Leakage from containment isolation valves that terminate below suppression pool ~ster level may be excluded from the total leakage provided a sufficient fluid inventory is available to ensure the scaling function for at least 30 days at a
prcssure of 54.6 psig.
Leak-age from containment isolatioo valves that are in closed-loop, seismic class I lines that. will be voter sealed during a DBA will be measured but will bc excluded when computing the total leakage.
Penetrations and isolation valves are identified as follows:
(1)
Testable penetrations with double 0-ring seals - Table 3.7,B, (2)
Tastable pcnetrations with testable bellows Table 3.7.C, (3)
Isolation valves with-out fluid seal - Table 3.7.D, (4)
Testable electrical pcnetrations
>> Table 3.7.H, and (5)
Isolation valves scaled with fluid-Tables
- 3. 7. E, and 3. 7. F, 232 Amendment No. pf
~pp
TABLE 3.7.A (Continued)
~Grou Valve Identification Number of Power Operated Valves Inboard Outboard
'.!a ximum Operating Normal Time (Sec.)
Position Action. On Initiating
~Sl nal
.orus Hydrogen Sample Line Valves Analyzer A (FSV-76-55, 56)
NA Note 1
SC
- orus Oxygen Sample Line Valves Analyzer A (FSV-76-53, 54),
Note 1
SC
"".'well Hydrogen Sample Li'ne Valves Analyzer A (FSV-76-49, 50)
KA Note 1
SC
~rn;el 1 Oxygen Sample Line Valves
- =.alyzer A (FSV-76-51, 52)
NA Note 1
SC
==-..pie Return Valves - Analyzer A
.=SV-76-57, 58)
"A 0
~
GC
- orus Hydrogen Sample Line Valves Analyzer B (FSV-76-65, 66)
NA Note 1
SC
.orus Oxygen Sample Line Valves-Analyzer B (FSV 64)
NA Note 1
SC "rywell Hydrogen Sample
'ne Valves-Analyzer B
'.=S'V-76-59, 60)
NA Note 1
SC
- ryvell Oxygen Sample Line
':alves-Analyzer B (FSV 62)
NA Note 1
SC
==-,.pie Return Valves-Analyzer B (FSV-76-67, 6S)
NA 0
GC Note 1:
".alyzers are such that one is sa.-..plinz crywell hydrogen and oxygen (valves from drywell open-
- ."en valves from drywell closed)
Croup 7!
The valvea in Croup 7 are autcnlotically actuated by only. tho follovtng condition:
Reactor veaael lov vatar level (4pp")
Croup 8:
The valves in Croup 8 are autoaatically actuated by only the folloving condition:
1.
High Dryvell praaaure 2.
Reactor vessel low water level (538")
255 Amendment No.
102
X-107r X-1 DGii X-10GH X-109 X-11GA
%-1108 X-200A-SC X-219 Table 3.7.'1
( cntinuod)
~p:r: (testa~~le)
Rows r CHL'od Positivn lnu CRD Rod Position ln~.'io.
Power CRD Hod Position S/RV Test instrumentation (Temporary)
Supnression Chamber Vacuum Breaker
~ 0 266 Amendment No.
102
~
~
~
e
~
5ASf.'S
~crnii l
process Linea are isolated by reactor tpaeceL Lov vater Level (alga")
Ln order to aLLoM fOr reeevaL of decay heat subsequent to a
~cram,, yet isolate Ln tioie d'or proper operation o{ the coro
~ tandby coc Ling ayateltio.
The valves Ln group l are also closed vhcn process Lnotruggcntation detects exccaoive aain stean Line flov, high radiation, Lov prooaure, or iiain ateaa apace high teplperature.
~Crau 3 - isolation valves ara closed by reactor vessel lou uater level (538") or high drywell pressure.
The group 2 isolation signal also "iso-lates" the reactor bui.lding and starts the standby gas treatment system.
Lt is not desirable to actuate the group 2 isoLation signal by a transient or spurious signal.
~Crnu 3
Process lines ara normally 'n use, and it is therefore nor.
desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes.
To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system.
Al o, since the vessel could potentially be drained through the cleanup
- system, a low level isolation i.s provided.
the consequences of an accident which results in the isolation of other process lines.
The signals which initiate isolation of Groups 4 and 5
process lines are therefore. indicative of a condition which should render them inoperablee Cr~ou 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 ventilation high radiation which would indicate a possible accident and necessitate primary con ta inmen t isola tion.
~Crau 7
-. Process lines ara closed only on the respective turbine stcam supply valve no t fully closed.
This assures that the valves are no t open when HPCI or RCIC action is required.
~Gr. ub 8 Lieu (trnvv ling in-core probol ia isolated on high dryuell presstlre or reactor low water level (538").
This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition.
The maximum closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside tne primary containment.
In satisfying t)>is design intent, an additiona.'argin has been included in specifying maximum closure times.
This margin permits 'dentif ication of degraded valve performance prior to exceeding the desi,gn closure times.
Amendment No.,
102
UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
75 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 April 9, '1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
C.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
I 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No.
DPR-68 is hereby amended to read as follows:
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
"~ 7
<rPg
- v~
Domenic B. Vassallo, Chief Operating Reactors Branch
/f2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 13, 1984
I
ATTACHMENT TO LICENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE NO.
DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages.
8, 57, 60, 63, 72, 81,
- 109, 267,
- 284, 294 2.
The marginal lines on these pages denote the area being changed.
AA Surveillance
'Knterval - Each Surveillance Requirement shall be performed within the specified time interval with:
l.
A maximum allowab'e extension not to exceed 25/ of the surveillance interval, but:
2.
The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
Amendment No.
75
0
.'lini.".u=.."o.
Insrr".,DDL Charnels Cperat:le ner Tr'- Svs(1)'1]>
Function R oerks 1.
Belo>> trip setting does the following:
a.
Initiates Reactor Building Isola tf on b.
Initiates Primary Containment Isolation (C'roups 2, 3, and 6) c.
Initiates SGTS k 538~ above vessel zero A or JB and EJ Instrument Channel-Reactor Low Mater Level (6)
TABLE.3. 2 h PRIMARy COHTAIHMEHT AHD REACTOR BUILDIHG ISOlATIOH IHSTRV!lEHTATION'nstrument Channel-Reactor Hfqh Pressure 100
+
15 psig Above trfp setting isolates the shutdown cooling suction valves of the RHR system.
Instrument Channel-Reactor Lo>> Mater Level JLIS 3 56A D, SM
~ 1J 2 470" above vessel zero A
1.
Belo>> trip setting initiates Main Steam Line Isolation Instrument Channel-Hig!i Drywell Pressure (6)
(PS-64-56A D) 2.5 pair A
oz'B and KJ 1.
Above trip setting does the following:
a, Initiates Reactor Building Isol ation b.
Initiates'rimary Containment Isolation c.
Initiates SGTS 2 J3)
Instrument Channel-High Radiation Main Steam Line Tunnel f6)
Instrument Channel Lo<< Pressure Hain Steam Line Instr ument charm~.l Bfqh Flow Hain Steam Line 3 times normal rated full power background (13) 825 Psig la) 6 le01 of rated steam flow B
Above trip setting initiates Hain Steam I.ine Isolation 1.
Belo>> trip setting inftiates Hain Steam Line Isolation l.
Above trip setting initiates Main Steam Line Isolation 2 (]2)
Instrument Channel Main Steam Line Tunnel Biqh Temperatur~.
200o F l.
Above trip setting fnitfates Hain Steam Line Isolation.
}linimum ho.
Instrument Channels Operable TABLE 3.2.A PRIHARY COHTAQQ(ENT AHD REACTGR BUILDING ISOLATION IHSTRUHEHTATIOH Functi
~i'ri ie el Seeti Remarks 2
Group 2 (Initiating) Logic H/A A or (B and E)
Refer to Table 3. 7.A for list of valves.
Group 2
(RHR Isolation-Actuation)
Logic Grip 8 {Tip-Actuation)
Logic Group 2 (Dryvell Swap Drains-Actuation)
Logic N/A N/A.
H/A Group 2 (Reactor Building N/A 6 Refueling Floor, and Dry-vell Vent and Purge-Actuation)
Logic F and G
1.
Part of Group 6 Logic.
1 1
Group 3 (Actuation) Logic N/h Group 6 Logic
,N/A Group 8 (Initiating) Logic N/A Group 3 (Initiating) Logic N/h C
F and G
1.
Refer to Table 3.7 A for list of
-valves.
1.
Refer to Table 3.7.A for list of valves.
1.
Refer to Table 3.7.h for list of calves.
Reactor Bui)ding Isolation H/A (refueling floor) Louie Reactor Building Isolation Nlh (reactor zone)
Logic Nor F BorG
'or A
2.
Same as Group 2 initiating logic.
1.
Logic has permissive to refueling floor static pressure regulator.
1.
Logic has pe+missive to reactor zone static pressure regulator.
There are four channels per steam line of which tvo must. be
- operable, Only required in Run Mode (interlocked with Mode Switch).
5.
Hot required in Run Mode (bypassed:iy mode switch) ~
6.
Channel shared by RPS and Primary Containment 5 Reactor Vessel Zsolation Control System.
A channel failure may be a
channel failure in each system.
7.
A train is considered a trip system.
6, Two out, of three SQTS trains required.
A fai1ure of more than one will require action A and P.
9.
There is only one trip system vich auto t=ansfer to tvo power sources.
10.
Refer to Table S.7.h and ita notes !or a liating of Iaolacion Ualve Gtoupa and their initiatf.ng signals.
h channel may be placed in an inoperable status for up to four hours for required surveillance without. placing the trip system in the tripped condition provided at least one OPERABLE channel in the same tzip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be operable for the channel to be operable.
Power operations permitted foz up to 30 days with 15 of the 16 temperature'switches operable.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 time~
background, respectively) are established based on the normal back-ground at full power.
The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respecti'vely.
6.
Amendment No.
7 75
Table ).).s INST&mUITATIOVTnay INITIATES os COICTAOLS Tilr Conf afis COHTAIHHKHT CW)LIHO
~ YFfrllr N)n)nun Ifo.
opecable Per ytip STS.Q!1
'I II0 I J I 'IO) i I I0)
I )10)
~ 1{10)
Instcusrnt Channel 5100 ~ F Thecansf at Laity'ces coolec ran)
)nstrunrnf chaMrl core spray A o'c C Start lnstruncnt Channel Core Spray 0 or D
Tnetcunrnt Channel S 100ay ther+octet
)Core Spray Area Cooler yan)
ASC Area Coo)ac Fan Log)c N/A Aenar is 1.
Above trip sef fang stacts AH)
~rea cno)ec canst 1.
Starts Core Spcay
~ rea Cooler I ah vh+h Cos e Spf av eotoc st al ts I.
Starts Core Spcay area cooler Can uhen Core Spray notor starts I~
Above trip setting starts Core Spcay acta cooler (ans
'I (IO) 111 I)
I 111) coca spcay Area cooler tan N/A Logic
)nstrunent channel N/A Core Spray Notocs A pr C
St.art Tnstcunent ChaMe l N/A Core Spray Notocs B pr D.
Start Instrunent Channel-lf/A Core Spcay Loop I Ace)dent Signal
)IS)
Instcuncnt ChaMel If/A Coca Spray Loop 1 Accident Signa 1
( I S)
A, l, C3, andD1
'I ~
Starts ANASQ punps A3 B]
A I ~
Starts ABASll punpa A3, Bl, C3, and Dl C3, and Dl A
I ~
Starts AIIAsu p'ps A3 Bl C3 RPT logic H/A (17) 1.
Trips recirculation pumps on turbine control valve fast closure or stop valve closure>
30$ pover.
TABLE 3-2-P SDRVEILLAIICE INSTRUHEh TATIOH Hinimum i of Operablu I>>sr rument Channel.
Instrument a
LI-3->>6 A LI"3-46 B
PI-3-54 PI-3-61 PR SO PI-64-67 TI-64-52 TR-64-52 TR-64-52 Tl-64-55 TIS-64-55 LI-64-54 A
LI 46 H/A PS-64-47 TR 64 52 and PS-64-58 E an3 I--64-67 I.l-84-2A "LI-84-l3A Instrument Reactor Mater Level Reactor Pressure Drywell Pressure Dr~eli Temperature Suppression Chamber Air Teeperature Suppression Chamber Mater Temperature Suppression Chamber water Level Control Rod Posit,ion Heutron Honitoring Dry+all Pressure Or)ve}l Temperature a>>d l'ressure and Timer CAD Tank A
Level
."AD Tank "8" Level Type Inals~tlo>>
and Range Ind(cator - (55" to indscator 0-1500 psig Recorder 0-80 psia Indicator 0-80 psia
- Recorder, Indicator
=-
0-4004 F Recorder 0->>004F Indicator, 0-4004F Ind1cator -25" to i25" 6V Indicating
)
Lights
)
- SRH, IRH, LPRH
)
0 to
)OOS po~er
)
Alarm at 35 psiq
)
Alarm if temp.
)
> 28\\4F and
)
pressure
> 2 5 pa(6 )
after 30 minute
)
d lay
).
Indicator 0 to l004 Indicator 0 to 1006 notes (I)
( )
(3)
{')) l2)
(3)
{I) (2)
(3l
(~)
(2)
(3) l))
(2)
{3)
( ))
(2l (3)
())
{2)
(3)
()) l2l l3) l4)
( l)
('2)
(3)
(4)
~
,s 3.2 BASES and HPCE and trips the recirculation oumos.
The low reactor water level instrumentation that is set to trip when reactor water level is'7.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray
- Pumps, contributes to ADS initiation, and starts the diesel generators.
These trip setting levels were chosen to be high enough to prevent spur ious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.
For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.
The high drywell pressure instruientation is a diverse signal to the water 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 instrumentation;
- thus, the r esults given above are applicable here also.
I Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The pr'mary function of the instrumentation is to detect a break in the main steam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in c'on)unction 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 1000 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas Trips are provided on this instrumentation
- and, when exceeded, cause closure of isolation valves.
The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectr um of breaks.
For large
- breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident.
With the established nominal setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines ar e not exceeded for this accident.
Reference Section 14.6.2 FSAR.
An alarm with a nominal setpoint of 1.5 x normal full-power background is pt ovided also.
109 Amendment No.
75
only the.iollowing condition:
Reactor Vessel Lou Water Level (470".)
Group 8:
The.valves in Group 8 are automatically actuated by only the following condition:
l.
High Drywell Pressure 2.
Reactor vessel low water level (538")
Amendment No.,
75
~
g ~
~
TABLE 3 7
H TESTABLE ELECTRICAL PENETRATIONS X-107B X-108A X-108B X-109 X-110A X-llOB Spare
{testable)
Power CRD Rod Position Indic.
II I~
Power CRD Rod Position Indic.
X-219
'SUDp reeving...
Chambe r Vacuum Breaker Amendment No.
75 284
~ '
3.7.D/4.7.D Primar Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and op'n to the free space of the containment.
Closure of one of the valves in each line would be sufficient 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.
~Gzou 1 - pzooess lines aze isolated hy reactor vessel lcw wats=
level (0~") ir. order to allow for removal of decay heat subsequent to a scram, yet isolate in tae for p"oper operation of the core standby cooling systems.
The valves in q"oup 1 are also closed when process insm~entation de ecm excessive main steam line flow, high radiaticn, low pre sure, or main steam space high Cempezatuxe.
~Grou 2 - Isolation valves are closed by reactor vessel low water level (538") or high drywell pressure.
The group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system.
It is not desirable to actuate the group 2 isolation signal by a transment or spurious signal.
~drou 3 - Process lines are normally in use, and it is thereyore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety-related causes.
To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system.
Also, since the vessel could potentially be drained through the cleanup
- system, a low level isolation is provided.
Groups 4 and 5 - Process lines are designed to r edtain operable and mitigate the 'consequences of an accident which results in the isolation of other process lines.
The signals wh ch initiate isolation of groups 4 and 5
process limes are ther efore indicative of a condition which would render them inoperable.
~Grou 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 ventilation high radiation which would indicate a oossible accident and necessitate primary containment isolation.
~Grou 7 - Process lines are closed only or the r espective turbine steam supply valve not fully closed.
This ensures that the valves are not open when HPCIS or RCICS action is required.
~drou 8 - Line (traveling in-core probe) is isolated on high dr ywell pressure or reactor low water level (538").
This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition.
294 Amendment No.
75