ML20066D286
| ML20066D286 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/02/1991 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20066D288 | List: |
| References | |
| NUDOCS 9101140301 | |
| Download: ML20066D286 (21) | |
Text
.-.
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UNITED STATES
~,
["
',g NUCLEAR REGULATORY COMMISSION 1 E WASHINGTON, D. C. 20655 o
E
%,....*/
TEhiESSE[ YALLEY AUTHORITY E
DOCKET NO. 50-260 BROWNS FERRY NUCLE _AR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.183 License No. DPR-52 i
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the
~
licensee) dated August 6,1990, as supplemented October 9,1990, complies with the standards and requirements of the Atomic Energy Act l
of 1954, as amended (the Act), and the Comission's rules and
(
regulations set forth in.10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 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 Comission's regulations and all applicable ~ requirements have been satisfied.
9101140301 910102 ADOCK 05 % gO DR
- r 2-2.
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 Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.183, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
i 3.
This licenre amendment is effective as of its date of issuance and shall be implemer.ted within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A
,f Frederick J. HeMon, Director Project Directorate II-4:
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 2,1991 i
+
ATTACHMENT TO LICENSE AMENDMENT NO. 183 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness, i
REMOVE INSERT 1.1/2.1-5 1.1/2.1-5 1.1/2.1-5a 1.1/2.1-10 1.1/2.1-10 1.1/2.1-11 1.1/2.1-11*
3.2/4.2-7 3.2/4.2-7 3.2/4.2-8 3.2/4.2-8*
3.2/4.2-14 3.2/4.2-14 3.2/4.2-15 3.2/4>2-15*
3.2/4.2-23 3.2/4.2-23*
3.2/4.2-24 3.2/4.2-24' 3.2/4.2-65 3.2/4.2-65 3.2/4.2-66 3.2/4.2-66*
3.7/4.7-25 3.7/4.7-25 3.7/4.7-26 3.7/4.7-26*
3.7/4.7-30 3.7/4.7-30 3.7/4.7-31 3.7/4.7-31*
3.7/4.7-49 3.7/4.7-49 3.7/4.7-50 3.7/4.7-50*
- Denotes overleaf or spillover page.
1
~.
FU'L' CLADDING INTEGRITY 1.1/2.1 E
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING-1.1.B.
Egwgr Transient.
2.1.B.
Power Transient Trio Settinas To ensure that the Safety 1.
Scram and isola-1 538.in.
Limits established in tion (PCIS groups above.
Specification 1.1.A are 2,3,6) reactor vessel not exceeded, each required low water leve1L zero scram shall be initiated by-its expected scram signal.
2.
Scram--turbine 4 10 per-The Safety Limit shall be stop valve cent valve assumed to be exceeded closure closure when scram is accomplished l
by means other than the 3.
. Scram--turbine 1 550 psig expected scram signal, control valve fast closure or-turbine trip 4.
(Deleted) 5.
Scram--main 1 10 percent steam _line
. valve isolation closure 6.
Main steam 1 825 psig isolation valve closure
--nuclear system low pressure C.
Reactor V,gggel Water Level C.
Water Level Trio Settinas Whenever there is irradiated 1.
Core spray and 1 398 in.
fuel in the reactor vessel, LPCI actuation-- above-the water level shall be
-reactor low vessel-greater than or equal to water level zero l
372.5 inches above vessel zero.
2.
actuation--
above reactor low vessel water level zero 3.
Main steam 1.398 in.
-l isolation
-above valve closure--
vessel reactor low zero water level i
l.
l' BrH 1.1/2.1-5 Amendment 183 Unit 2 l
e n
THIS PAGE INTENTIONALLY LEFT BLANK l
I i
f 4
BFN 1.1/2.1-Sa Unit 2
1.1-BASES'(Cont'd).
The safety limit has been established-at 372.5 inches above_ vessel l
zero to provide a point which can be monitored'and also provide adequate margin to assure sufficient cooling.
j REFERENCE-1.
General Electric BWR Therraal Analysis Basis (GETAB) Data, i
Correlation and Design Application, NEDO 10958 and NEDE 10938.
2.
General Electric Document No. EAS-65-0687, Setpoint Determination for Browns Ferry Nuclear Plant, Revision 2.
t l
l l
BFN 1.1/2.1-10 Amendment 183 Unit 2
3.1 BASES
LIMfT2NG SAFETY SYSTEM SETTINGS RELATED TO TUEL CLADDING IttfEGRIII The abnormal operational transients applicable to operation of the Browns Ferry Nuclear plant have been analyzed throughout the spectrum of planned ope' rating conditions up to the design thermal power condition of 3,440 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR.
In addition, 3,293 MWt is the licensed maximum power level of Browns Ferry Ruclear plant, and this represents the maximum steady-state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance.
Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model. The comparisons end results are summarized in Reference 1.
The void reactivity coefficient and the scram worth are described in-detail in Reference 1.
The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications as further described in Reference 1.
The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The' rapid insertion of negative reactivity is assured by the time requirements for 5 percent and 20 percent insertion. By the time the rods are 60 percent inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect.
The times for 50 percent and 90 percent insertion are given to assure proper completion of the expected performe.nce in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.
For analyses of the thermal consequences of the transients a MCPR > limits specified in Specification 3.5.k is conservatively assumed to exist prior to initiation of the transients. This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by.using expected values of control parameters and analyzins at higher power levels.
BFN 1.1/2.1-11 Unit 2 l
~
.l TABLE 3.2.A PRIMLRY CONTAINMENT AND REACTOR SUILDING ISOLATION INSTRUMENTATION i
c: en 2 M Z%
Minimum No.
i Instrument Channels Operable Per Trio Sys(11(11)
Func tion Trio Level Settino Action (11 Remarks 2
Instrument Channel -
1 538a above vessel zero A or 1.
Selow trip setting does l
Reactor Low Water level (6)
(9 and E) the following:
(LIS-3-203 A-0) a.
Initiates Reactor Building Isolation b.
Initiates Primary Containment Isolation c.
Initiates SGTS 1
Instrument Channel -
100 2 15 psig D
1.
Above trip setting isolates Reactor High Pressure the shutdown cooling section (PS-68-93 and -94)
. valves of the RHR system.
j 2
Instrument Channel -
1 398* above vessel zero A
1.
Below trip setting l
Reactor low Water initiates Main Steam Level (LIS-3-56A-0)
Line Isolation t
2 Instrument Channel -
5,2.5 psig A or 1.
Above trip setting does the w
High Drywell Pressure (6)
(8 and E) following:
N*
(PIS-64-56A-0) a.
Initiates Reactor N
1 Sv11 ding Isolation b.
Initiates Primary Containment Isolation c.
Initiates SGTS 11 8-a W
.{
4 N
1 l
- v. a.
TABLE 3.2.A (Continued)
PRIMARY CONTAlfrENT AMD REACTOR 9tJILDING ISOLATION INSTRUMENTATION ee hE Minisuo No.
Instrument M
. Channels Operable Per Trio Sys(11f11)
Function Trio tevel Settino Action (1)
Remarks 2
Instrument Channel -
1 3 times normal rated 8
1.
Above trip setting High Radiation Main Steam full power background initiates Main Steam Line Line Tunnel (6)
Isolation 2
Instrument Chaenel -
1825 psig (4) 8 1.
Selow trip setting Low Pressure Main Steam line initiates Main Steam Line Isolation (PIS-1-72, 76, 82. 86) 2(3)
Instrument Channel -
i 140f of rated steam flow 8
1.
Above trip setting High Flow Main Steam Line initiates Main Steam
?
(PdIS-1-13A-0, 25A-0 Line Isolation 36A-0, 50A-0) 2(12)
Instrument Channel -
1 200*F B
1.
Above trip setting P
Main Steam Line Tunnel initiates Main Steam w
High Temperature Line Isolation.
I 2(14)
Instrument Channel -
160 - 180*F C
1.
Above trip setting y
Reactor Water Cleanup initiates Isolation tz System Floor Drain of React 7r Water High Tamperature Cleanup Line from Reactor and Reactor Water Return Line.
2 Instrussent Channel -
160 - 180*F C
1.
Same as above Reactor Water Cleanup System Space High Temperature
- n-2 17strument Channel -
1 150*F C
1.
Same as ihove E
Reacter Water Cleanup E
System ripe Trench e:::E 1
Instrument Channel -
1100 er/hr or downscale G
1.
I upscale or 2 downscale will h
Reactor But! ding a.
Initiate SGTS,
- Ventilatien High b.
Isolate reactor zone and
- 8 Aadiation - Reactor Zone refueling floor.
>4 c.
C1oss abnesphere GM control system.
r cp.
^
L TABLE 3.2.8 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS c: e,
- s n Z*
Minisiuns No.
Operable Per I
Trio Sys(1)
Function Trio Level Setting Action Remarks 2
Instrisment Channel -
1470* above vessel zero.
A 1.
Below trip setting initt.ated Reactor low Water Level HPCI.
(LIS-3-584-0) 2 Instrtament Chantel -
1470" above vessel aero.
A 1.
Multiplier relays initiate r
Reactor low Watir level RCIC.
(LIS-3-58A-0) 2 Instrument Channel -
1398" above vessel aero.
A 1.
Selow trip setting initiates Reactor Low Water Level CSS.
(LS-3-58A-0)
Multiplier relays initiate LPCI.
2.
Multiplier relay from CSS initiates accident signal (15).
2(16)
Instrisnent Channel -
1398" above vessel aero.
A 1.
Below trip settings, in u
Reactor low Water level conjunction with drywell (LS-3-58A-0)
Z high pressure, low water 1evel permissive, 105 sec.
6 delay timer and CSS or RHR pump running, initiates A35.
e
~e 2.
Selow trip settings, in conjunction with low reactor water level permissive, 105 see. delay timer.
E3 121/2 min. delay timer.
CSS or RHR pump running, initiates A05.
m 1(16)
Instrument Channel -
1 544" above vessel zero.
A 1.
Below trip setting pennissive 3
Reactor Low Water level for initiating signals on ADS.
Permissive (LIS-3-184
-^
185) m w
1 Instrument Channel -
1 312 5/16" above vessel aero.
A 1.
Below trip setting prevents Reactor low Wa:er Level (2/3 core height) inadvertent operation of (LIS-3-52 and LIS-3-62A) containment spray during accident condition.
I i
.. ~. -
c tn TABLE 3.2.8 (Continued) 52 Minimus No.
rt Operable Per w
Trio Svsill Function Trio tevel Settino Action Remarks 2
Instrument Cha w el -
11 p12.5 psig A
1.
Below trip setting prevents Drywell High Pressure (PIS-64-58 E-H) inadvertent operatloc of i
containment spray during accident conditions.
2
' Instrument Channel -
1 2.5 psi.
A 1.
Above trip setting in con-t Drywell High Pressure (PIS-64-58 A-D) junction with low reactor l
pressure initiates CSS.
Multiplier relays initiate HPCI.
2.
Multiplier relay from CSS initiates accident signal. (15) 2 Instrument Channel -
1 2.5 psig A
1.
Above trip setting in Drywelf High Pressure (PIS-64-58A-0) conjunction with low reactor pressere initiates t.PCI.
2(16)
Instrument Channel -
i 2.5 psig A
1.
Above trip setting, in Drywell High Pressure u
(PIS-64-57A-0) conjunction with low reactor water level low reactor water level permissive, 105 sec. delay timer and
initiates ADS.
- MN aE g
P M-C3 4i 1
/
.~.
NOTES *FOR TABLE 3.2.B 1.
Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two OPERABLE trip systems except as noted.
If a requirement of the first column is reduced by one, the indicated action shall be taken.
If the same function is inoperable in more than one trip system or the first col'umn reduced by more than one, action B shall be taken.
Action:
A.
Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If the function is not OPERABLE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take action B.
B.
Declare the system or component inoperable.
C.
Immediately take action B until power is verified on the trip system.
D.
No action required; indicators are considered redundant.
2.
In only one trip system.
3.
Not considered in a trip system.
4.
Requires one channel from each physical location (there are 4 locations) in the steam line space.
5.
With diesel power, each RHRS pump is scheduled to start immediately and each CSS pump is sequenced to start about 7 sec. later.
6.
With normal power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about 7 sec. with similar pumps starting after about 14 sec. and 21 sec.,
at which time the full complement of CSS and RHRS pumps would be operating.
7.
The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure. The RCICS setting of 450" of water corresponds to et least 150 percent above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.
Similarly, the HPCIS setting of 90 psi corresponds to at least 150 percent above maximum steady state flow while also ensuring the initiation of isolation following a postulated break.
8.
Note 1 does not apply to this item.
9.
-The head tank is designed to assure that the discharge piping from the CS and RHR pumps are full. The pressure shall be maintained at or above the values listed in 3.5.H, which ensures water in the discharge piping and up to the head tank.
BFN 3.2/4.2-23 Unit 2 l
a
R2IES..f0R TABLE 3.2.B (Cont'd) 10.
Only one trip system for each cooler fan.
11.
In only two of the four;4160 V shutdown boards.
See note 13.
12.
In only one of the four 4160 V shutdown boards.
See note 13.
13.
An emergency 4160 V shutdown board is considered a trip system.
14.
RRRSW pump would be inoperable.
Refer to Section 4.5.C for the requirements of a RHRSW pump oeing 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 (1 398" above vessel zero) originating in-the core spray system trip system.
16.
The ADS circuitry is capable of accor.plishing its protective action with one OPERABLE trip system. Therefore one trip system may be taken out of service for functional testing and calibration for a period not to exceed-eight hours.
17.
Two RPT systems exist, either of which will trip bo'th recirculation pumps. The systems will be individually functionally tested monthly.
If-the test period for one RPT system exceeds two 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power shall be less than 85 percent within four hours.
BFN 3.2/4.2-24 Amendment 183 Unit 2
3.8 hASES In addition to reactor protection instrumentation which initiates a
+
reactor scram, protecti.ve instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequenceo. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby _ gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods whet.
portions of such systems are out of service for maintenance, and (li) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. -Such instrumentation must be available whenever primary containment integrity is required.
i The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at 538 inches above vessel zero closes isolation valves in the RER System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves). The low reactor water level instrumentation that is set to trip when reactor water level is 470 inches above vessel zero (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems.
The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).
The low water level instrumentation set to trip at 1 398 inches above vessel zero (Table 3.2.A) closes the Main Steam Isolation Valves, the Main Steam-Line Drain Valves, and the Reactor Water Sample Valves (Group 1).
Details of valve grouping and required closing times are given in Specification 3.7.
These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.
The low reactor water level instrumentation that is -set to trip when reactor water level is 1 398 inches above vessel zero (Table 3.2.B) l BFN 3.2/4.2-65 Amendment 183 Unit 2
1 3.2 BASIS (Cent'd) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.
These trip setting levels vere chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplish.ed 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 instrumentation 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 alli initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also.
ADS provides for automatic nuclear steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool.
ADS initiates when the following conditions exists low reactor water level permissive (level 3), low reactor water level (level 1), high dryvell pressure or the high drywell pressure bypass timer timed out (12 1/2 min.), and a 105 second tima delay.
In addition, at least one RHR pump or two core spray pumps must be running.
The high pressure bypass timer is added to meet the requirements of NUREG 0737, Item II.K.3.18.
This timer will bypass the high drywell pressure permissive af ter a sustained low vnter level.
The worst case condition is a main steam line break outside primary containment with HPCI inbperable. With the bypass timer set at 15 minutes, a Peak Cladding Temperature (PCT) of 1424' F is reached for the worst case event. This temperature is well below the limiting PCT of 2200' F.
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 daring 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 percent of rated steam flow in conjunction with the flow limiters and main steam line valve cicoure 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 valven.
The setting of 200'T for the main steam line tunnel detector is low enough to detect leaks of the order of 15 spm; thus, it is capable of covering the entire spectrum of breaks.
For large breaks, the high steam AMENDMENT RD.16h BFN 3.2/4.2-66 Unit 2
1 TABLE 3.7.A c tzs 52 PRIMARY CONTAIPMENT IS0tATION VALVES n
Muster of Power Maximum Action on u
Operated Valves Operating Normal Initiating fe Ryg Valve Identification inboard Qutboard Time (sec.)
position Signal r
1 Main steamline isolation valves
( FCV-1-14, -26, ~37, & -51; 1-15, t
-27, -38, & -52) 4 4
3<T<S O
GC 1
Main steamline drain isolation valves (FCV-1-55 & 1-56)
I 1
15 0
GC 1*
Reactor. water sample lir.e -
isolation valves 1
1 5
C SC 2
RHRS shutdown cooling supply isolation valves (FCV-74-48 & -47) i 1
40 C
SC 2
RHRS - LeCI to reactor (FCV-74-53 & -67) '
2 40 C
SC 2
RHR$ flush and drain vent to F
suppression chamber (FCV-74-102, y
-103,'-119, & -120) 4 20 C
SC o.
2 Suppression chamber drain 7
(FCV-75-57 & -58) 2 15 0**
GC 2
Drywn11 equipment drain discharge isolation valves (FCV-77-15A & -158) 2 15 0
GC i
2 Drywell floor drain discharge isolation g
valves (FCV-77-2A & -28) -
2 15 0
GC 5
"These valves isolate only on reactor vessel low low low water level (1398") and stin steam line high radiation of Group 1 isolations.
r+
- These valves are normally open when the pressure suppression head tank is aligned to serve the RHR and **,
CD discharge piping and closed when the condensate head tank is used to serve the RrtR and CS discharge piping.
(See Specification 3.5.H)
TABLE 3.7.A (Continued)
Number of Power Maximum Action on-E$
Operated valves Operating Normal Initiating
- z Grevo Valve Identification inboard Outboard Time (stL 1 Posi tion Signal t
N 3
Reactor water cleanup system supply isolation valves (FCV-69-1 & -2) 1 1
30 0
GC l
4 FCV-73-81 (Eypass around FCV-73-3) 1 10 0
GC 4
HPCIS steamline isolation valves (FCV-73-2 & -3) 1 1
20 0
GC 5
RCICS steamilne isolation valves (FCV-71-2 & -3) 1 1
15 0
GC 6
Drywell nitrogen purge inlet i
isolation valves TFCV-76-13) 1 5
C SC 6
Suppression chambsr nitrogen purge Inlet isolation valves (FCV-76-19) 1 5
C SC 6
Drywell main enhaust isolation 4
valves (FCV-64-29 & -30) 2 2.5 C
SC ta 6
Suppression cho6er main exhaust Z
tsolation valves (FEV-64-32 & -33) 2 2.5 C
SC b.
L 6
Drywell/ suppression chamber purge 4
intet (FCV-64-17) 1 2.5 C
SC 6
Drywell atmasphere purge in1st (FCV-64-18) 1 2.5 C
SC d
I i
G 4
O I
1 NOTES FOR TABLE 3.7.A Key:
0 = Open C = Closed SC = Stays Closed GC = Goes, Closed
.l Note:
Isolation groupings are as follows:
Group 1: The valves in Group 1 are actuated by any one of the following conditions:
1.
Reactor Vessel Low Low Low Water Level (1 398")
2.
Main Steemline High kadiation 3.
Main Steamline High Flow 4.
Main Steamline Space High Temperature 5.
Main Steamline Low Pressure Group 2: The valves in Group 2 are actuated by any of the following conditions:
1.
Reactor Vessel Lov Water Level (538")
2.
High Drywell Pressure..
Group 3: The valves in Group 3 are actuated by any of the.following conditions:
1.
Reactor Low Water Level (538")
2.
Reactor Water Cleanup System High Temperature 3.
Reactor Water Cleanup System High Drain Temperature Group 4: The valves in Group 4 are actuated by any of the following conditions:
1.
HPCI Steamline Space High Temperature 2.
HPCZ Steamline High Flow 3.
HPCI Steam 11ne Low Pressure 4.
HPCI Turbine Exhaust Diaphragm high Pressure Group 5: The valves in Group 5 are actuated by any of the foilowing l
conditions:
l 1.
RCIC Steamline Space High Temperature 2.
RCIC Steamline High Flow 3.
RCIC Steamline Low Pressure 4.
RCIC Turbine Exhaust Diaphragm High Preseure Group 6: The valves in Group 6 are actinated by any of the following conditionst 1.
Reactor Vessel-Low Water Level (538")
2.
High Drywell Pressure 3.
Reactor Building-Ventilation High Radiation BFN 3.7I4.7-30 Amendment 183 l
Unit 2
RQIT1 JfR TABLE 3.7.A_Ggntinggil' Group 7: The valves in Group 7 are automatically actuated by only the' following condition:
1.
The respective turbine steam supp W valve not fully closed.
Group 8: The valves in Group 8 are automatically actuated by only the
- l following conditions:
1.
High Dryvell Pressure 2.. Reactor vessel Low Water Level (538")
s BFN 3,7/4.7-31 Unit 2 1
c-yy r
=-.
r
- r r +=
~ =
v 1-wme e
-~e' m
3.7/4,7 IlnSIS (Cont'd) follow ASTM D3803. The charcoal-adsorber efficiency test procedures should i
allow for the removal of one..adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples.
Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.-
If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Guide 1.52.
The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerocol shall be performed in accordance to ANSI M510-1975. Any HEPA filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.
All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of each filter train for a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> each month will prevent moisture build.Jp in the filters and adsorber system.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
If significant painting, fire or chemical release occurs such that the NEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.
The determination of significance shall be made by the operator on duty at the time of the incident.
Knowledgeable staff members should be consulted prior to making this determination.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
If one standby gas treatment system is inoperable, the other systems must be tested daily.
This substantiates the availability of the operable systems and thus reactor-operation and refueling operation can continue for a limited period of time.
3.7.D/4.7.D Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open 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 LOCA.
fttgyp_1 - Process lines are isolated by reactor vessel low water level (1 398") in order to allow for removal of decay heat subsequent to a scram, l
yet isolate in time for proper operation of the core standby cooling systems.
The valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature. The
(-
reactor water sample line valves isolate only on reactor low water level at
- 2. 398" or main steam line high radiation.
l
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BFN 3.7/4.7-49 Amendment 183 Unit 2
3.7/4.7 BASES (Cont'd)
GInup_2 - Isolation valves are closed by reactor vessel low water level ($38")
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 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 nonsafetyrelatedcauses.);Toprotectthereactorfromapossiblepipebreak in the system, isolatia. 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 remain operable and mitigate 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 would render them inoperable.
Group.6 - Lines are connected to the primary containment but not directly te the reacter 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 necussitate primary containment isolation.
Groue 7 - Process lines are closed only on the respective turbine steam supply valve not fully closed. This assures that the valves are not open when HPCI l
or RCIC action is required.
Groun 8 - Line (traveling in-core probe) is isolated on high drywell 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.
l The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intt... 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 includeduin specifying 7 closure times. This margin permits identification of degraded va.
ance prior to exceeding the design closure times.
In order tr nat the doses that may result from a steam line break do not exceed
~?R 100 guidelines, it is necessary that no fuel rod perforation y; from the accident occur prior to closure of the main steam line iso 4am an valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds, l
BFN 3.7/4.7-50 NOMM NO. I 77 Unit 2