ML19344E653
| ML19344E653 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/27/1980 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18025B055 | List: |
| References | |
| NUDOCS 8009020408 | |
| Download: ML19344E653 (14) | |
Text
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ENCLOSURE 1 PROPOSED TECliNICAL SPECIFICATION CllANCES BROWNS FERRY NUCLEAR PLANT IP41T 3 8009020408 J
shculd drop below the top of the f uel during this time, the ability to remove decay heat ia reduce.d. This reduction in cooling capability could lead to ele';ated 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 aero).
REFERENCE 1.
General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10958, and NEDE 10958.
17 k
position, where protocilon of the f uel cladding integrity saf ety limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability o. the fuel cladding integrity safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.
I. J.
& K.
Peactor low water level set poin t for initiation of HPCI and RCIC, closinq main steam isolation valves, and starting LPCI and core spray pumps These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad tem pe ratu 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 fuel and the system pressure.
L.
References 1.
Linford, R.
B.,
" Analytical Methcds of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO-10802, Feb., 1973.
2.
Generic Reload Fuel Application, Licensing Topical Report NEDE-20411-P-A, and Addenda.
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sal'CTY LIftIT LIMITING FadWiMd 31535ef9CETTING k
- 2. 2 REACTCh MLANT SYSTEM
- 1. 2 JAJTOR COOLANT SYSTEM INTEGH ITY INTEGRITY A pplica bi li t y Applicability Applies to limits on reactor Applies to trip settings of the coola n t system pressure.
instruments and devices which
'are provided to prevent the reactor system safety limts from being exceeded.
Obiective Obiective To establish a limit below To define the level of the which the integrity of the process variables at which reactor coolant system is not automatic protective action is threatened due to an initiated to prevent the overpressure condition.
pressure safety limit from being exceeded.
Specification speci f i ca tion The limiting saf ety system settings shall be as specified A.
The pressure at the lowest below:
point of the reactor vessel shall not exceed Limiting 1,375 psig whenever Safety irradiated fuel is in the Protective System reactor vessel.
Action Settino A.
. Nuclear system 1,250 psig safety valves
+ 13 psi open--nuclear (2 valves) system pressure B.
Nuclear system relief valves open--nuclear system pressure
,1,103 psto
+ 11 pst (4 valves) 1,115 psig
+ 11 psi T 4 valves) 26 r
c SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- 1. 2 py gIOBJOOLA?JT SYSTFM 2.2 REACTOR COOIANT SYSTEM IJ.1,TEGR ITY ItTTEGRITY 1,125 psig
?,11 psi
( 3 valves) r C.
Scram-nuclear i 1,055 psig system high pressure o
27 6-
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- 1. J P A..h!j H E At* Toft COO!Aftr SYS f D1 Ittf EGR ITY The-bat ety limit u for the reactor coolant system pressure have teen salected such that ths/ are below pressores at which it can be shown that the. tit.egrity of the systen is not endingered. However, the pressure safety limits are set high enougn such t hat no foreseeable circumstances can cause the system pressure to rise over these limits.
The' pressure sat et y lirru ta a t., acnitrarily selected to be the lowest t r a.is t ent oven gelessures allowed by the. applicable codes, ASME Doller and Pr essure vessel Code,Section III, and USAS Piping Code, Section 831.1.
The design presuura (1,250 psig) or the reactor vessel is established such that, when the 10 percent allowance (125 pst) allowed by the ASME Boiler and Pressure vessel Code Section III t or pressure transients is added to the design pressure, a tratistent pressure limit of 1,375 psig is r
established.
Co r r e';pund i ng l y,' the design pressure (1,1f48 psig for suction and 1,326 pt.,19 f or discharge) of the reactor recirculation synten piping are such that, when the 20 percent allowa nce (230 and 26 *, ps i) allowed by USAS Piping Code, Section B31.1 f or nr.essum e transients are added to the design pressures, t r asisi en t pressure listiits of 1,378 and 1,591 psig are established.
Thits, the pressurt safety limit applicable to power operation is established at 1,375 psig (the lowest transient overpressure allowed by the pertinent codes), ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
De current cycle's safety analysis concerning the most severe abnerral operationa.1 transient ressu.lting directly in a reacter coolant system pressure increase is Elven inthe supplemental reload licensing submittal for the current cycle.
The reactor vessel pressure coce limit of 1.375 psig given in subsection 4.2 of the safety analysis report is well above the peak pressure produced by the overpressure transient described above. Thus, the pressure safety limit applicable to power operation is well above the peak pressure that can result due to reasonably expected overp,ressure transients.
Higher design pressures have beea established for piping witntn the r aar" n r coolant system than for the reactor vessel.
These increaseis design pressures c eate a consistent design which asnures that, if the pressure within the reactor ve..ncl doan *iet "xceed 1,375 psig, the pressures within the pintain cannot extre *d their reapective transient pressure linits due to st.atic and pump heads.
28 e
i The saf ety limit of in the reactor vessel 1,375 psig actually applies to any point however, because of the static water head, the highest pressure point will occur at the bottom of the vessel.
Because the pressure is not monitoted at this point, it cannot has been violated.be directly determined if this safety limit head level and flow pressure draps, an equivalentAlso, because of the pot pressure in the vessel.cannot be a priori determined for a pressure monitor higher Therefore, following any transient that is severe enoagh to cause concern that this safety limit was violated, a calculation will be performed using all available information to determine if the saf ety limit was violated.
REPER ENCES 1.
Plant Saf ety Analysis (jBFNP FSAR Section N14.0) 2.
ASME Boiler and Pressure Vessel Code Section III 3.
USAS Piping Code, Section B31.1 4
Peactor Vessel and Appurtenances Mechanical Design (BFNP FSAR Subsection 4.2) 29
2.2. BASES REACTOR COOLANT SYSTEM INTEGRITY ihe pressure relief system for each unit at the Browns Ferry Nuclear
~
Plant has been sized to meet two design bases. First, the total safety / relief valve capacity has been established to meet the over-pressure protection criteria of the ASME Code. Second, the distribution of this required capacity between safety valves and relief valves has been set to meet design basis 4.4.4-1 of sub-section 4.4 which states that the nuclear system relief valves shall prevent opening of the safety valves during nornal plant isolations and load rejections.
The details of the analysis which shows compliance with the ASME Code requirements is presented in subsection 4.4 of the FSAR and the Reactor
' Vessel Overpressure Protection Summary Technical Report submitted in r,esconse to ouestion 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) results in a maximum vessel pressure
- which, if a neutron flux scram-is assumed, has adequate margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated has been divided into 70% reliet (11 valves) and 14.2% safety (2 valves). The analysis of the plant isolation transient (turbine trip with bypass valve failure to open) assuming a turbine trip scram is presented in the supplemental
' reload licensing submittal for the current cycle.
This analysis shows that the 11 relief valves limit pressure at the safety valves to a value which is below the setting of the safety valves. Therefore, the safety valves will not open. This analysis shows that peak system pressure is limited to a value which is well below the allowed vessel overpressure of 1375 ps19 4
30 j
mc
e In the ann ytical treatment ot t.he t ransierits, f)dr milliseconds are allowed between a neutron sensor reachino the scram point and the start of negative reactivity insertion.
This is adequate and conservative when compared to the typically observed time delay of dbout 270 milliseconds.
Approximately 70 milliseconds af ter neutron flux reaches the trip point, the pilot scram valve solenoid power supply voltage goes to zero an approximately 200 milliseconds later, control rod motion begins.
The 200 milliseconds are included in the dllowable scram insertion times specified in Specification 3.3.C.
In order to perform scram time testing as required by c;ccifiesticn L.3.C.1, the relaxatien of certain restraints 12. the rvd sequence control systers is required. Individual rod typass switches ;:ay be used as described in specificatiori k.3.C.1.
The position of any rod bypassed must be known to be in accordance vith rod withdrawal sequence. Bypassin6 of rede in the 1.unner deceribed i
in specification L.3.C.1 vill allow the subsequent withdravnl of any rod l
scra=ned in the 100 percent to 50 percent rod density groups; howcVer, it will caintain group notch control over all rods in the 50 percent to o percent rod density arouos. In addition. RSCS will nrevent movennt of rods in the 50 percent density to a preset power level range until the scrammed rod has been withdrawn.
D.
Reactivit y Anomalies During each f uel cycle excess operative reactivity varies as j
fuel depletes and as any burnable poison in supplementary 1
control ts burned.
The magnitude of this excess reactivity
]
may be inferred from the critical rod configuration.
As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod i
pa tte rn a t selected base states to the predicted rod inventory at that state.
Power operating base conditions provide the nost sensitive and directly interpretable data relative to core reactivity.
Fu rthermore, using power operating base conditions permits frequent reactivity compa risons.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% dK.
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 of the reactor system.
References 1.
Generic Reload Fuel Application, Licensing Topical Report NEDE-20411-P-A,~and Addenda.
136 e
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT 4.5 903 ANp ggtr_rMtiggt!T_@LE9 COOLING SYSTEMS SIIEIIME 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 1.24 for 8x8 fuel, and 1.25
- C#
for 8x8R fuel, 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 25% 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 value shown in Figure 3.5.2.
cause operation with a If at any time during limiting control rod operation, it is deter-pattern as described in t
en or mined by normal surveillance g
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 reactor operation is within 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 specified remedial action is taken, the even*: shall be logged and repe,rted in a 30-day written report.
Y.
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testing to ensure that the lines are filled.
The visual checkt nq will avoid starting the core spray or RHR system with a discharge line not filled.
In addition to the Jisual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line highpoint to supply makeup water for these systems.
The conden sate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service.
System discharge pressure indicators are used to determine the water level above the discharge line high point..The indicators will reflect approximately 30 psig for a water level at the nigh point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.
When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled.
Further assurance is provided by observing water flow from these systems high points monthly.
I.
gaximum Averace Planar Linear Heat Generation Rate (MAPLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all tne rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly af f ect the calculated peak clad temperature by less than i 200F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.
The limiting value for MA PLHGR is shown in Tables 3 5.I-1,
-2, -3.
The analyces supporting these limiting values is presented in reference 4.
J.
Linear Heat Generation Rate (LIIGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat 176 e
~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
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It must be recognized that there is always an action which would return any of the pa rame te rs (MA PLdGR, LHGR, or MCPR) to within prescribed limits, namely power reduction.
Under most circumtances, this will not be the only alternative.
M.-
References 1.
" Fuel Densification Effects on General Electric Boiling Water Reactor Fuel,a supplements 6, 7, and 8, NEDM-10735, August 1973.
2.
Supplement 1 to Technic:al Report on Densifications 02 General Electric Reactor Fuels, December 14, 1974 (USA Regulatory S taf f).
3.
Communica tion:
V.
A. Moore to I.
S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.
4' Generic Reload Fuci Application, Licensing Topical Report NEDE-20411-P-A, and Addenda.
178 m--
= -,
e 3.6/4.6 E&EEE To meet the safety design basis, thirteen safety-relief valves have been installed on unit 3 with a total capacity of 84.2% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second 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 assumed has adequate margin to the code allowable over-pressure limit of 1375 psig.
To meet the operational design basis, the total safety-rclief capacity of 84.2% of nuclear boiler rated has been. divided into 70% relief
+
(11 valves) and 14.2% safety (2 valves). The analysis of the plant iso-lation 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.
Ihis analysis shows that the 11 relief valves limit pressure at the safety valves to a value which is below the setting of the safety valves. Therefore, the safety valves will not open. This analysis shows that peak system pressure is limited to a value which is well below the allowed vessel overpressure of 1375 psig.
225
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 f ailures or deteriorations.. The relief and safety valves detect thei r are benchtested every second operating cycle to. ensure that The relief points are within the +1 percent tolerance.
l set valves are tested in place once per operating cycle to establish that they will open and pass steam.
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief 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)
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