ML20003B354
| ML20003B354 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/12/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003B355 | List: |
| References | |
| NUDOCS 8102100698 | |
| Download: ML20003B354 (27) | |
Text
.
e#p* Re o%'o, UNITED STATES
[ ' q' '( /,[,
NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20656 g,-
d' TENNESSEE VALLEY AUTHORITY DOCKET N0. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Tennessee Valley Authority (the licensee) dated August 27, 1980, (as supplemented by letters dated September 23, 1980 and October 14,1980,) and September 5, 1980 and October 17, 1980, comply 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 applications, 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 i
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. 37, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
9'**050 $
W9 n
. 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
%-]ci-xZsbppolito, Chief ThomasA Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 12, 1981 4
A ATTACHMENT TO LICENSE AMENDMENT NO. 37 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:
1.
Remove the following pages and replace with the identically numbered pages:
17 177 24 178 26 225 27 225a 28 261 29 286A 30 286B (new page) 82 320 136 321 166 325a 167 326 176 328 2.
Marginal lines on each page indicate the revised area.
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4m, e should drop below the top of the fuel during this time, the ability to renove decay heat is reduced. This reduction in coolinq 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 safety limit has been established at 17.7 in. above the top of the irradiated fuel 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 zero).
R EF ER ENC 6 1.
General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10958, and NEDE 10958.
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l Amendment No. 28, 37 l
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posit io n, where protection of the fuel cladding integr ity saf ety limit is provided by the IRM and APRM high neutron Thus, the combination of main steam line low flux scrams.
pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability oi 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.
& K.
Reactor low water level set point for initiation of H PCI and RCIC, closinq main steam isolation valves, and startino LPCI and core sprav pumps These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad tem pe ra tu re s. The design of these systems to adequately perform the intended function is based on the specified low level scram set point and initiation set points.
Transient analyses reported in Section N14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the f uel and the system pressure.
L.
References 1.
- Linford, R.
B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," HEDO-10802, Feb., 1973.
2.
Generic Reload Fuel Application, Licensing Topical Report NEDE-240ll-P-A, and Addenda.
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24 Amendment No. 28,37
LIMITT.NG SMdWiMd 3153WM95ETTING SAILTY 1,I fi t T
- 1. /
HEACTOR COOLANT SYSTEM
- 2. 2 REACTOR COOLANT SYST EM IITEGHITY INTEGRITY A pplica bi li ty Applicability Applies to limits on reactor Applies to trip settings of the Coolant system pressure.
instruments and devices WhiCh are provided to prevent the reactor system safety limts from being exceeded.
Obiective Oblective To establish a limit below To define the level of the wntch 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 Spe ci f i ca tion The limiting safety 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 Setting A.
. Nuclear system 1,250 psig
+ 13 psi saf ety valves open--nuclear (2 valves) system pressure B.
Nuclear system relief valves open--nuclear system pressure f
1,105 psia
+ 11 pst (4 valves) 1,115 psig
+ 11 psi T 4 valves) 26 Anendment No. 28,37
LIMITING SAFETY SYSTEM SETTING SATETY LIMIT
- 1. 2 prACTOR COO _t. ANT SYSTFM 2.2 BEACTOR COOIANT SYSTEM INTEGBITY IftrEGRITY 1,125 psig
+ 11 psi
( 3 valves)
C.
Scram-nuclear i 1,055 pt's system high pressure l
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l 27 Amendment No. 28, 37
1.4
[ A.M H t.ACTt;h COOLAtIT' SY s f LM I ttf EGR ITY Tho ut. t y li mit :. f or the reactor coolant system pressure have tr en salected such that ths/ are below pressures at whten it can be snown that the
.r.egrity of the systen is not e
endangered.
However, the pressure saf ety limits are set high
<nouan such t hat no f oreseeable circumstances can cause the system pressure to rise over these limits.
The' pressure se t -st y l i mi t...il e a con trarily selected to be the lowest traassent ove:3lessures e llowed by the. applicable codes, ASME Doller and P1'eaur e Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The desian pr es:s u ro (1,250 psig) et the reactor vessel is est ablished such t ha t, when the 10 percent allowance (125 pst) allowed by the ASME Boiler and Pressure Vessel Code section III t ur pre:tsure transients is added to the design p r e-s s u r e, a transient pressure limit of 1,375 psig is established.
Correspondinqly, the design pressure (1,148 psag for suction and 1, 3 2 6 pt a q for discharoe) of the reactor recirculation syr. ten pi ping a r e such that, when the 20 percent allowance (230 and 265 psi).silowed by USAS Piping Code, Section B31.1 for nr.essur e transients are added to the design pressures, transicat pressur:.11mits of 1,378 and 1,591 psig are established. Tims, the pressure safety limit applicable to power oleration as estat ished at 1,375 psig (the lowest transient overpressure a towed by the pertinent code s), ASME Boiler and Pressure VesseA Code,Section III, and USAS P.tping Code, Section B31.1.
The current cycle's safety analysis concerning the most severe abner ral operationa.1 transient ressultlng directly in a remeter coolant system pressure increase is given inthe supplemental reload licensing submittal for the current cycle.
The reactor vessel pressure coce linit of 1.3h psig give, in subsection 4.2 of the safety analysis l
report is well-above the Deak 'ressure produced by the overpressure Thti, the pressure safety limit applicable transient described above.
to power operation is well above the peak pressure that can result due to reasonably expected overpressure transients.
I Higher destan pressures have bee.) established for piping within the r ear" or coolant system than for the reactor ve:;:,,1.
These liiereases design pressures c. eate a consistent deston which a9r.ures that, if the pressute within the reactor l
vesnel doan. set "xce e 1 1,375 psig, the pressures within the protein casnot exce d t hair respective transient pressure I
linits due to st.atic and pump heads.
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Amendment No. 5,18, 37
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O 10 The safety limit of 1,375 poig actually applies to any point in the reactor vessel; how ver, because of the static water e
head, the highest pressi,re 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, because 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 thie safety limit was violated, a calculation will be perf ormed 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 Pressure Vessel Code Section III 3.
USAS Piping C3de, Section B31.1 4
Reactor Vessel and Appurtenances Mechanical Design (BFNP FSAR Subsection 4.2) 29 Amendment No. 28, 37
9 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 Coce. 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 r ' lief ',alves shall prevent opening of the safety valves during normal t.iant isolations and load rejections.
The details of the analysis wh.ich shows compliance with the ASME Code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Sumary Technical Report submitted in l
r;esconse to cuestion 4.1 dated December 1,1971.
To meet the safety design basis, thirteen safety-relief valves have been installed on each unit with a total capacity of 84.2". of nuclear boiler rated 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) retdts in a maximum vessel pressure
- which, if a neutron flux scram is as;umed, has adequate margin to t/'a code allowable overpressure limit of 1375 psig.
To meet the op-
-ional design basis, the total safety-relief capacity of 84.2% of nuct r boiler rated has been divided into 707, relief, (11 valves) and 14.27, 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 the supplemental reload licensing submittal for the current cycle.
This analysis shows that the 11 relief valves ifmit pressure at the safety valves to a value which-is below the setting of the saf.ety valves. Therefore, the safety valves will not open. This analysts shows that peak system pressure is limited to a value which is well below the allowed vessel overpressure of 1375 psig.
i 3G Ar,endment No. 28, 37 m.
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TABIE 3.2.F
&arveillance Instrumentation a
Minimum J of Operable Instrument Type Indication Channels Instrument $
Instrument arul Bange Notes 2
It M 94 Drywell and Torus 0.1 - 20%
(1) 2 liydrogen ll M 104 Concentration 2
i O
2 2
PdI-64-137 Drywell to Suppression Indicator (1) (2) (5)
PdI-64-138 Chamber Differential O to 2 paid pressure Amendment No. 12, 19, 37
A In the anh ytical treatment at the transiestts, #10~
milliseconds are allowed between a neutron sensor reaching the scram point and the start of negative reactivity insertion.
This is adequate and conservative when conpared to the typically observed time delay of dbout 270 milliseconds.
Approximately 70 milliseconds df ter neutron flux reaches tne trip point, the pilot scram valve scienoid power supply voltage goes to zero an approximately 200 milliseconds later, control rod motion begins.
The 200 milliseconds are included in the dllowable scram insertion times specified in
- pecification'3.3.C.
In order to perform scram time testing as required by yecificaticn rod Lequence L. 3.C.1, the r elaxation of certain restraints in the control system is required. Individual rod bypass switches :tay be used as described in specification' fe.3.C.l.
The position of any rod byTassed must be known to be in accordance bypassing of rods in the.cnner described with rod withdrawal sequence.
in specificaticn 1. 3.C.1 vill allov the subsequent withdrawn 1 of any rod scra:med in the 100 percent to 50 percent rod density groups; however, it will caintain group notch control over all rods in the 50 percent to o percent rod density groups.
In addition. RSCS will nrevent movement of rods in the 50 percent density to a preset power level range until the scrammed rod has been withdrawn.
D.
Reacti vit y Anomalies Durina each f uel cycle excess operative reactivity varies as ruel 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 tuel burnup proaresses, anomalous behavior in the excess reactivity may be detected by comparison or the critical rod pattern at selected base stdtes to the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly interpretable data relattve to core reactivity.
Fu rthermore, using power operating base conditions permits frequent reactivity conpa ri sons.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% d K.
Deviations in core reactivity greater than 1% d K are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to tran"ients exceeding design conditions or the reactor system.
uterences 1.
Generic Reload Fuel Application, Licensing Topical Report NEDF 4011-P-A, and Addenda.
136 Amendment No. 28, 37
LIMITING CONDITIONS FOR OPERATION CURVEILLANCE REQUIREMENTS 4
- 4. 5 gO_RE__AUD COtTTAIEMEE_GQO_ LEG 3.5 C_ ORE AND CONTAINMENT SYST_ Egg COOLING SYSTEMS J.
Linear Heat ' Generation J.
Linear Heat Generation Rate (LHGR)
. Rate (LHGR)
During steady state power The LHGR shall be checked operation, the linear heat daily during reactor generation rate (LHGR) of operation at >25% rated any rod in any fuel thennal power.
assembly at any axial location shall not exceed 13.4 kW/ft.
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If at any time during operation it is determined by normal surveillance that the limiting value i
for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR 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 166 Antandment No.18, 37
1 LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREMENTS 4.5 co._R_L AND_99t!IMMMgtg COOLING 3.5 CORE AND CONTAINMENT EXEIEEE COOLING SYSTEMS and corresponding action shall continue until reactor operation is within the prescribed limits.
K.
Minimum Critical Power Ratio (MCPR)
The MCPR operating limit is K.
Minimum Critical Power i
1.24 for 8x8 fuel, and 1.25
- C#
for 8x8R f uel, and 1.25 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 be greater than the above power level or limits times K. K is the distribution that would g
g valee shown in Figure 3.5.2.
cause operation with a limiting control rod If at any time during Pattern as described in operation, it is deter-the bases for mined by normal surveillance Specification 3.3.
that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 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 I
reactor operation is within i
the prescribed limits.
L.
Reporting Requirements If any of the limiting values identified in Specifications 3.5.I J,
or K are exceeded and the specifled remedial action is taken, the event shall be logged and reported in a 30-day written report, g
Amendment No. 28, 37
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!O 4&i?S tenting to ensure that the lines are filled.
The visual checkt na will avoid starting the core spray or RHR system sith a discharge line not filled.
In addition to the visual observation and to ensure a filled discharge lir.e other than i
prior to testing. a pressure suppression chamber head tank is located approximately 20 feet a'ove the discharge line highpoint to supply makeup wate; f or these systems.
The cond en sa te nead tank located approximately 100 feet above the disena rge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service.
System discha roe 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 nich point and 45 psig for a water level in the pressure suppression
.. amber head tank and are monitored daily to ensure that the discharge lines are filled.
When in their normal standby condition, the suction for the HPCI and FCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and PCICS piping. This assures that the HPCI and RCIC discha rge piping remains filled, rurther assurance is provided by observing water flow from these systems nigh points monthly.
I.
Maximum Averace Planar Linear Heat Generation Rate (HAPLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will. tot exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature following a postulated loss-of-i ccolant accident is primarily a function of the average heat generation rate of all tne rods of a fuel assembly at any l
axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly a f f ect the calculated peak clad temperature by less than 1 i
200r 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 l
within the 10 CFR 50 Appendix K limit.
The limiting value j
fo r MA FLiiGP is shown in Tables 3 5. I-1, -2, -3 TPe analyses supporting these limiting values is presented in reference 4.
I J.
Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in.ny rod is less than the design linear heat 176 Amendment No. 28, 37
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s.5 is A:;lc; qene ra tion it fuel pellet densification is postulated.
shall be The LHGR checked daily during reactor operation at 2 25% power to or control rod movement has caused determine if fuel burnup, changes in power distribution.
For LHGR to be a limiting value below 25% rated thermal power, the MTPr would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.
K.
Minimum Critical Power Ratio (MCPRI core thermal power levels less than or equsi to 25%, the Atreactor will be operating at minimum recirculation pump speed For all and the moderator void content will be very small.
designated control rcd patterns which may be employed at this point, operating planc experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess With this low void of raquirements by a considerable margin.
content, any inadvertent core flow increase would only place The operation in a more conservative mode relative to MCPR.
daily requirement f or calculating MCPR above 25% rated 4
thermal power is suf ficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR limiting control rod pattern is approached ensures when a that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at thermal limit.
L.
Reportino Requirements The LCC's associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e.,
there allowable time in which the plant can knowingly exceed is na the
.miting values for MAPLHGR, LHGR, and MCPR.
It is a as stated in Specifications 3.5.I,
.J, and
.K.
requir ement, if at any time during steady state power operation, it that LHGR, or is determined that the limiting values for MAPLHGR, MCPR are exceeded action is then initiated to restore This action is operation to within the prescribed limits.
initiated as soon as normal surveillance indicates that an operating limit has been reached.
Each event involving
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steady state operation beyond a specified limit shall be 177 Amendment No. 37
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locaed and reported quarterly.
It must be recognized that i
there is always an action which would return any <Jf the pa rame te rs (MAPLHGR, LHGR, or MCPR) to within prescribed
{
limits, namely pcwer reduction.
Under most circumtances, this will not be the only alternative.
M.
References i
i i
Generic Reload Fuel Application, Licensing Topical Report NEDE 24011-P-A, and Addenda.
4 3
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i 178 Amendment No. 28, 37
1 1
To meet the safety design basis, thirteen safety-relief valves have been i
installed on unit 3 with a total capacity of 84.2% of nu: lear boiler ra ted stean flow. The analysis of the worst overpressure transient.
(3-sec)nd closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assu aed has adequate margin to 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 70% relief (11 valves) and 14.2% safety (2 valves). The analysis of the plaat iso-lation transient (turbine trip with bypass valve f ailure to open) assuming trip scram is presented in the. supplemental reload a turbine l
licensing submittal for the current cycle.
Ihis analysis to shows that the 11 relief valves limit pressure at the safety valves a value which is below the setting of the safety valves. Therefore.
the saf ety valves will not open. This analysis shows that peak sys".em 7
pressure is limited to a value which is well below the allowed vessel overpressure of 1375 psig.
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225 i
Amendment No. 28, 37
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r 3.6/4.6 B AS ES Experience in relief and safety valve operation shows that a
testing of 50 percent of the valves per year is adequate to failures or deteriorations. The relief and safety valves thei r detect are benchtested every second operating cycle to ensure that The relief set points are within the +1 percent tolerance.
valves are tested in place once per operating cycle to establish l
that they will open and pass steam.
The requirements established above apply when the nuclear system These requirements can be pressurized above ambient conditions.
are applicable at nuclear system pressures bel')w normal operating pressures because abnormal operational transi nts could possibly these conditions such that eventual overpressure relief start at would be needed.
However, these transients are much less severe, in terms of pressure, than those starting at rated conditions.
The valves need not be f unctional when the vessel head is removed, since the nuclear system cannot be pressurized.
REFERENC ES 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection
- 4. 4) 225a Amendment No. 28, 37
LIMITING CCNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 C0!TTAISHENT SYSTEMS 4.7 CO!TTAIhHENT SYSTEMS H.
Containment Atmosphere H.
Containment Atmosphere Monitoring (CAM) System -
Monitoring (CAM) System -
H Analyzer H Analyzer 2
4 1.
Whenever the reactor is 1.
Each hydrogen analyzer not in cold shutdown, two system shall be demon-independent gas analyzer strated OPERABLE at systems shall be operable least once per quarter for monitorir.g
.,2 drywell by performing a CHAh3EL and the torus.
CALIBRATION using standard gas samples containing 2.
With one hy..-ogen analyzer a nominal eight volume inoperabic, restore at percent hydrogen balance least twr hydrogen
Each hydrogen analyzer be in at least HOT system shall be demonstrated SHUTDOWN within the next OPERABLE by performing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a CHANNEL FUNCTIONAL TEST scathly.
3.
With no hydrogen analyzer OPERABLE the reactor shall be in HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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' Amendment No. 12, 37 261
Inerting The relatively small containment volume inherent in the GE-BWR pressure suppression containment and the large amount of zirconium in the core are such that the occurrence of a very limited ( a percent or so) reaction of the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a flammable concentration in the containment. If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric quantities, the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain low leakage integrity. The <4% hydrogen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant accident.
The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based.
Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety.
Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant laaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged suf ficient to perform the leak inspection and establish the required oxygen concentration.
To ensure that the hydrogen concentration is maintained less than 4%
following an accident, liquid nitrogen is maintained on-site for contain-ment atmosphere dilution. About 2260 gallons would be suf ficient as a 7-day supply, and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore, a requirement of 2500 gallons is conservative.
Following a loss-of-coolant accident the Containment Air Monitoring (CAM) i l
System continuously monitors the hydrogen concentration of the containment volume. Two independent systems (a system consists of one hydrogen I
sensing circuit) are installed in the drywell and the torus. Each sensor and associated circuit is periodically checked by a calibration gas to verify operation.
Failure of one system does not reduce the ability to monitor system atmosphere as a second independent and redundant system will still be operable.
In terms of separability, redundancy for a f ailure of the torus system is based upon at least one operable drywell system. The drywell hydrogen I
concentration can be used to limit the torus hydrogen concentration during l
post LOCA conditions. Post LOCA calculations show that the CAD system within two hours at a flow rate of 100 scfm will limit the peak drywell 286A Amendment No. 3,12, 37
1 j
Inerting (Cont'd) f and wetweil hydrogen concentration to 3.9% (at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and 3.9% (at 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />),
j respectively. This is based upon purge initiation after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at a flow rate of 100 scfm to maintain containment pressure below 30 psig. Thus, peak torus hydrogen concentration can be controlled below 4.0 percent using J
either the direct torus hydrogen monitoring system or the drywell hydrogen monitoring system with appropriate conservatism (* 3.9%), as a guide for CAD / Purge operations.
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i 2868 Amendment No. 37 E - ----
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LIMITING CONDITICNS FOR OPERATION SURVEILLANCE REQUIRE.'G:NTS i
3,9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILI ARY ELECTRICAL SYSTEM b.
The fourth operable unit 3 diesel generator.
4 Buses and Boards 4.
Undervoltage Relays Available a.
Once every 6 a.
Both start buses to months, the unit 3 are energized.
condition ander which the undervoltage b.
The 4-kV bus tie relays are board and required shall shutdown boards be simulated (3 EA, 3EB, 3EC, with an 3ED) are undervoltage on energized.
start buses 1A and la to c.
The u80-V demonstrate that shutdown boards the diesel associated with generators will the unit are start.
energised, b.
Once every 6 d.
Undervoltage months, the relays operable conditions under on start buses which the 1A or 18 and 4-undervoltage kV shutdown relays are boa rds, 3EA, required shall 3EB, 3 EC, and be simulated 3ED.
with an undervoltage on e.
The 480V diesel each shutdown Aux Boards are board to energized.
demonstrate that tne associated diesel generator f.
The 480V Rx. HOV I
will start.
Boards D & E are energized with M-G Sets 3DN, 3DA, 3EN, and 3EA in service.
320 Amendment No. 35,'37
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l
I 3.9 AUXILI ARY ELEC*rRICAL SYSTEM 4.9 AUXILI ARY ELECTRICAL SYSTEM c.
The undervoltage relays which start the diesel generators from start buses 1A and 1B and the 4-kV shutdown boards, shall be calibrated annually for trip and reset and the measurements logged.
5.
The 250-Volt Shutdown 5.
480-V RMOV boards D and E Board battery and unit batteries and a battery a.
Once per operating cycle, charger for cach battery the automatic transfer and associated battery feature for 480-V RMOV boards are operable.
boards D and E shall be functionally tested to verify auto-transfer capabi.
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, 37
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1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4
1 3.9 AUXILIARY ELECTRICAL 'iYSTEM i
l B.
Undervoltage re'cys on lA or 13 start bus may be inoperable for i
a period of 7 days provided the other start bus and undervoltage relay are operable (within sur-l veillance schedule or 4.9.A.4.a).
9.
Undervoltage relays on a shutdown board may be inoperable 5 days l
provided the other shutdown boards and undervoltage relays are operable (within surveillance schedule of 4.9.A.4.b).
10.
When one 480 volt shutdown board is found to be inoperable, the
+
i reactor will be placed in hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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11.
If one 480-V RMOV board m-g set is inoperable, the reactor' may remain in operation for a period not to exceed seven days, provided the remaining 480-V RMOV board m-g sets and their associated loads remain operable.
i 12.
If any two 480-V RMOV board m-g sets become inoperable, the reactor shall be 31 aced in the cold shutdown ondition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i 13.
If the requirements for operating in the conditions specified by t
3.9.B.1 through 3.9.B.12 cannot be met, an orderly shutdown shall l
be initiated and the reactor shall j
be shutdown and in the cold c.n-dition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Amendment No. 35, 37 325a I..
1,1f'IT:!m CONDITIO'.S l'OR OpF. RATION SURVEILLANCE REQUIREMEh"TS E
3.9 At:X LI ARY ELEC11tICAL SYSTEM 4.9 AlfXILIARY ELECTRICAL SYSTEM C.
theration in Cold Shutdown Coadition Whenever the reactor is in the cold shutdown condition with irradiated fuel in the reactor, the availability of electric power shall be as specified in Section 3.9.A except as specified herein.
1.
At least two unit 3 diesel generators and their associated 4-kV shutdown boards shall be operable.
2.
An additional source of power consisting of one of the following:
a.
One 161-kV transmission line and its accociated cooling tower transformer capable of nupplying power to the unit 3 l
shutdown boards.
A third operable dienc! generator.
J.
At least ene unit 3 48U-V shutdown board nust be operable.
4 A total of two 480-V RMOV board motor-generator (m-g) sets may be inoperable. One loop of the RilR system (LPCI mode) must remain fully operable at all times. "(One m-g set for 480-V RMOV boards D and E must be in service at all times. The two operable m-g sets may not be supplied from the same 480-V shutdown board.)
Amendment No.13, 37 326
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The 250-Volt d-c system is so arranged, and the batteries sized such, that the loss of any one unit bactery will not prevent the safe shutdown and cooldow: of all three units in the event of the loss of offsite power and j
a design basis accident in any one unit. Loss of control power to any engineered safeguard control circuit is annunciated in the main control room of the unit affected.
I The station battery supplies loads that are not essential for safe shutdown and cooldown of the nuclear system. This bactery j
was not considered in the accident load calculations.
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There are two 480-V ac Reactor Motor-Operated Valve (RMOV) Boards that l
contain motor-generator (M-G) sets in their feeder lines. These 480-V ac RMOV boards have an automatic transfer from their normal to alternate power
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source (480-V ac shutdown boards). The M-G sets act as electrical isolators to prevent a fault from propagating between electrical divisions due to an automatic transfer. The 480-V ac RMOV boards involved provide motive power to valves associated with the LPCI mode of the RHR system. Faving an M-G set j
out of service reduces the assurance that full RHR (LPCI) capacity will be l
available when required.
Since suf ficient equipment is available to maintain the minimum complement required for RHR (LPCI) operation, a 7-day servicing period is justified. Having two M-G sets out of service can considerably
{
reduce equipment availability. Therefore, the affected unit shall 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 />.
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328 Amendment No. 35, 37
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