ML20148B248
| ML20148B248 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/15/1980 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20148B243 | List: |
| References | |
| NUDOCS 8001210268 | |
| Download: ML20148B248 (11) | |
Text
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,QAFETY LIff1T L]MITIIIG SAFETY SYGTEM OF'lTIN(*
.1 FLEL CLADDING INTEGRITY 2.1 FUEL CLADDIt!G IllTEGI:lTY or core coolant flow is lesc RB1 (0.66W + k2%)
thr'.n 10% of rated, the core thermal power shall not ex-where:
.ceed 823 Mwt (about 25% of 6
rated thennal power).
33 = Red block setting in perce:it.
of rated thermn] power (3?r)'3 MWL)
W
= Loop recirculation flow rate in perecrit of rated (rnteil loop recircule>6 ion flow rate equaln 314.2 X ~10 lb/hr)
In the event of ooerntion with the core maximum fraction of limiting power density (CMFLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows:
S (0.66U + h2%) FRP H
CMFLPD C.
Whenever the rear: tor ic in C.
Geram and isolit:stion--; 's ;R in, ato,ve the chutdown condition with reactor low enter vennci zero le<"I irr utir.ted fuel in the reac-te* vencel. the water level shall not be lere than 17.7 in. above the top of the D.
Scram--turbi ne stop < 10 riu r"en t.
nom:t?. active fuels zone, valve clocure valve et<>
E.
Serr. --turbine control valve th..i 1.ri p ear 1.
Fan t c l<in* re tin, r:e:t. ir4 '
. s i:::v i e t a l vi 2.
Innu o f cer.t.rol { 'j'>0 p ;.i,.
oil presnu.re F.
Scram--low con-1 P i i ncl.o.
denser vacuum 1i, v scion.
G.
- e ravi--maiti n t e:en < 1it p. c..c.i t.
line iso 3ntion va l v.
"It.i.cc 0
ll, thi n cl.c;un i nnlntion 1 :"i i*a i,-
valve clocure--nuclet: ;y n t e.a i,,w i
precsure 10 9@025033 l-
~j@
8001210
Revised 1-17-79 2.1 M ff
,1.
J. & K.
Peactor low yater level set point f or initiat ion of IIFCI and RCIC, closing main steam isolation valves, and starting LFCI and core spray purope.
These systens maintain adequate coolant inventory and provide core cooling with the objective of preventing escessive clad temperatures.
The desir.n of thenc systems to adequately perform the intended func.
tion is based on the specified lov level scram set point and initia-tion set points. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the systes pressure.
L.
References.
1.
- Linford, P.. E., " Analytical Me thods of Plant Trans tent Eva tustions f or the General Electric Boiling Water Reactor." NIDo-10802, Feb.,1973.
g
- 2. Generic Reload Fuel Application, Licensing Topical Report.
NEDE-240ll-P-A, and Addenda.
5 Amendment No. 35, 47 025034 i
I
Revised 1-17-79 O
1.2 BASES Therefore,.*ollowin; ar.:,' transient pressure monitor higher in the vessel.
i l ted, that is severe enough to cause concern that this se'ety limit was v o a to det.cr.
a calculation vill be perfereed using all available in'erutio::
eine if the safety limit was violated.
REi71.lf!CES_
(E.UPISAitSectionll.'))
1.
Pinnt 59..ety Anslysis
!.5".7. 3 oiler and Pressure Vessel Code Secti n III 2,
us.a.s Pipir.c ode,,cet ten 331.1 c
3.
( 7.7 T'AR stescto. */ :sel and Appurter.an:es !!achenicel.2.s'Ca L.
Sa':secticn L.2) 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-240ll-P-A and Addenda.
O 1
1 0
4 e
4 1r 1
1 O
5 025035 7
[
Amendment No.
33, 47 I
9 1
2.2 BASES REACTOR COOLANT SYSTDI INTEGRITY To meet the safety design basis, thirteen relief valves have: been installed on the unit with a total capacity of 82.6% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steamline isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1375 psig.
To meet the operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open)shows that 12 of the 13 relief valves limit peak system pressure to a va1Ue which is well below the allowable vessel over-pressure of 1375 psig.
30 9d025036
t Revised 1-17-79 r
3.3/4.4 RASFS; D.
Reactivtt1 Anomalics During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.
The magnitude of tbla exrons reactivity may be inferred from the crittral rod conf!Kuration.
As fuel burnup pro-gresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly i n t e r p r e t., b l e data relative to core reactivity.
Furthermore, using power operating base conditions permits frequent reactivity comparisons.
Requirink a reactivity comparison at the specified comparison will be made frequency assures that a before the core reactivity change exceeds 1% d k Deviations in core reactivity greater than I Li k a r e not expected and require thorough evaluation.
One percent reactivity into the core would not lead to transients exceeding design conditions of the reactor O
mystem.
References 1.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.
i 134 Amendment No. 35, 47 9
90025037
/
l
}
LIMITING CONDZTIONS FOR OPERATION SURVEILLANCE REQUIREMENTS I
LHGR 4LHGR
- OP/P)
(t/LT) max max LHGR = Design LHGR = 18.5 kW/f t for 7x7 fuel d
=13.4 kW/ for 8x8 fuel 8x8R and P8x8R fuel (d P/P)
= Maximum power spiking penalty
= 0.026 for 7x7 fuel
= 0.022 for 8x8,8x8R and P8x8R fuel LT = Total core length = 12.0 ft for 7x7 fuel and 8x8
= 12.5 ft for 8x8, 8x8R & P8x8R L = Axial position above bottom of core If at any time during operation it is deter-mined by normal surveillance that the limitin g value for LHGR is being exceeded, action shal l
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 and corresponding action shall continue until reactor operation is within the prescribed limits.
K.
M?.nimum Critical Power Ratio (MCPR)
K.
Minimum Critical Power Ratio From BOC to EOC-2000Mw0/T (MCPR) the MCPR operating limit for EFNP 1 cycle 4 is 1.23 for 7x7 fuel, 1.24 for 8x8 fuel, MCPR shall be determined daily and 1.25 for 8x8R and P8x8R fuel. These during reactor power operation at limits apply to steady state power operation 25% rated thermal power and at rated power and fica.
For core flows following any change in power level other than rated the MCPR shall be greater or distribution that would cause than the above limits times Kg.
Kg is the cperation with a limiting control value shown in Figure 3.5.2.
From EOC rod pattern as described in the
-2000 to EOC the MCPR limits will be 1.23, bases for Specification 3.3.
1.27, and'l.28 for 7x7, 8x8 and 8x8R/P8x8R respectively.
If at any time during oper-ation it is determined by normal surveill-ance 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 MCPR is not returned to within the pre-scribed 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.
Reportine Recuirements If any of the limiting values identified in Specifications 3.5.I, J, or K are g()Q2$038 exceeded and the specified action is taken, the event shall be logged and reported in a 30-day written report.
- 12.5 feet for 8x8R fuel 160 1
REFIEED MAY '.' r 19G 3,3 gyg 1.5.M Maint enanc e,,o,I r i 11,ed n ! = c ha r : e_r,1g LPC I. HPC I S, a nd 8 C IC S a r e no t If the discharte piping of the core array, filled, a water hemmer can develop in this pipinr. vben the pump and/or pocos are started. To minimize da= ate to the discharge pipine and to ensure this Technical Specification added martin in tr.e eperation of these systems, requires the discharp.c lines to ha filled whenever the system is in an i
operable conditjon.
If a discharge p ipe is not filled, the pumps that supply that line must be assueed to be inoperable for Technical Specification pur-
\\
poses.
The core spray and kHR syste= discharge pipina high peine vent is visually checked f or water flow once a sonth prior to testine to ensure that the limos are ftLled.
The' visual checking vill avoid starting the c, ore spray or f
RftR system with a discherte line not filled. In addition to the risual etservation and to ensu.e a filled discharge line other thas prior to testics, a pressure suppressien chamber hesd tank is located appmzir.ately 20 feet abor.
The the discharge line highpcist te supply takeu; vster for these systems.
cocienJt.te head task located approximately 100 feet above the disebarge high point serves as a backup :harging system vhee the pressure suppression cht:ber head tar.k is not in serM ee.
Systes discharge pressure indicators are used to determine the vater level above the dischar6e line high peist. The indicatorn and L5 villrefleet approximately 30 ;sig for a vster level at the high poict pwi6 for a vater level in the pressuresuppression chivabor bead tank and are e itored daily to enture that the discharge lines are f131ed.
8"$
When in their normal standhv candition, the $uction for tne !!PCI and RCIC p.amo g are alleneI to the condensate storsr.e tank. whic% is physic ally et a
h ip.he r el'evattan than the lipCIS and RCICS ptnine.
Thfs asvsres that the terct and ECIC discharme pipina. remains filled.
Further resurance is nrnvLded by nbaarvint water flow frem these systems high points monthly.
Ma xi mu s /.v s r s j e F lana r Lin e a r H e a t, C ene r a tion R.a ta (MP laCX) 3.3.2.
This specification assures that the peak cladding temperature fo1*.rving the postulated design basis less-of-:co1 Ant seCident Vill not SICsed the limit specified in the 10CT150, Appendix 1.
The peak cladding temperature fellowing a postulated Icss-of-coolant scei-dent la prisar11y a function of the averase hcot generation rate of all the rod s of a f;e1 assembly at any avtal location and is only 4:pendrnt second-sr!!y on the red to rod power d f stribution within an assembly. Since er-pected loca: vartations in power di:tribution within a fuel asses.b!y a(fect the calcula cd pesk clad temperature by less than 1 200T relative te,the peak temperJture for a typl:J1 fuel des)gn, the limit on the averar,e f in aa r be s t generation race is suffft!:nt to assure that calculated temperatures are within the 10C/ A$0 Ascend.x X limit. The limiting value for liAPLilGR is shown in Tables 3.5.1-1,-2,-3,-4 -5,-6, and -7 per reference 4.
168 O 'Il i []
9dO25039
Ob $79
.5.J.
Linear Heat Ceneration Rate (f.IICR)
This specification assures that the linear heat generation rate in any rod p,
is less than the desiCn linear bent Ceneration if fuel pellet denaification is postulated. The power spike penalty specified is based on the anal-ysis prreented in Section 3.2.1 of Ref erence 1 as modified in Ref erences 2 and 3, and assua es a linearly increacinC varfstion in exist gaps be-tveen cord bottom and tne, and assures with a 9)% confidence, thbt no note than one fuel rod cu reds the de ign linear heat cencrat f on rate due to pover spikinc. The Ll!CK ac a function of cure beis;ht sh:11 be checked daily dur-ing reactor operation at 1 25% power to deter,31ne if fuel burnup, or con-trol rod movoient has cauced changes in power distribution., For LHCR to be a limiting value belnv 25% rated thermal power, the !!Trr would have to be greater than 10 which is precluded by a considerable r:argin when employing any per'ef s sibic contro'._ red patt ern.,
- 3. 3. r,.
Kimima Cri tic al Pcue r Ra eio (MCPR)
At core thsrmal pcver levels less than or equal to 25:, the reactor vill be operating at e.inimum recirculstten pissp speed and the moderator void content vill b e v e ry esall.
For all designated control rod patterns which tuy be en-ployed at this poin t, operating plant experience and thermal hydraulic anal-ysie ind i c a t ed t ha t the resulting NCPP value te in e.xcess of requirements by a c ern s id e r a b i c it.a:S im.
- 1th this low void' content, any inadvertent c. ore flow locrease would only piece operatien in a more conservative code rels-tive to MCTR.
The daily requirement for esiculating MCPR above 25% rated thermal power is sufficient eince power distribution shif ts are very siev vhen there have not been significent power or control rod changes. The requirement for calculating MCPR vhen a li=iting contrel rod pattern is approached ensures that MCFR vill be kaove folleving a change in pcver or pover shape (regardless of ma gn i tud e ) that could piscs operation at a thermal limit.
b 3.5.L.
Rep r t i.o g Req uir em e n t s The LCO's associated with monitoring the fuel red operating conditions are required to be met at.all t!.mes, i.e., there is na allevable time in which the plant cas knovingly exceed the limiting values f or MAPLPCR, LBCR, end HCTE.
It to a requirement, as etated in Specifications 3 5.1..J. end.1.
t h.a t if at acy time during s teady state pen,er operatien, it is determined that the limiting vs. lues for MAPLHCR, LHCP., or M"PR are exceedef actior, is then initiated to restore operation to within the prescribed limits. This actico is initiated as seen as normal surveillance indicates that an eperatsng lir -
it has been reached. rach event involving stesdy state operation beye.d a specif ied limit shell be legged sad reported quarterly. It must ce reecgnized thu there is always an setien which vould return scy of the pare.cetero (MAP'.H C R,
UICR. er HCTR) to withis prescribed limits, ca=ely power reduction. Under most circumstances, this vill act be the only alternative.
j 1
x, References 1.
"ruel Densificati:t Tif ects on General Electric Ecil.n.; k'see; bacter Puel." Supplecer t s 6, 7, and 8. hD'-10 7 3 5, Augu s t 19M.
.2. S u ppl emen t I to Technical Report on Densi!!cacicos of Genera
- r.l e c t r i c Re a c t o r Tb e l s, D e c e:b e r ll., 19 7I. (USA Raguistory Staff).
3.
Cc==su oi c.a t i o n :
- '. A. Moo r e t o I. 5. Mi t c he ll. "Moc i f ie d CE hac e l f or Puel Densificacic." Docket 50-J21. March 27, 197t,.
O Generic Reload Fuel Application, Licenising Topical Report, NEDE-24011-P-A and addenda.
l 169 m.u se,m 94025040
e REylSED MAY 2 5 1979
/*'s 3. 6/h. 6 cAceS detected rensenably in a matter of feu hours utilizing the available leekoce detcetion sehe es, and if the crigin cannot be determined in a reasonably shor+ time the unit should be shut down to allow further :
investigation and corrective action.
The total leakege rate consists of all leaktge, idet.tified and unidenti-fled, which flows to the drywell floor drain and equipment drain sucps.
The capacity of the dryuell flect sue:p pucp is 50 gpm and the capacity of the dryvell er,uipment su:p pump is also 50 g;n. Renovel of 25 src from either of these cumps can be acco.mplished with considerable narsin.
RETIRETCES 1.
Nuclear System Leakage Rate Limits (BFi? FSAR Subsection h.10) 3.6.D/4.6.D Relief Valves To meet the safety basis thirteen relief valves have been installed on the unit with a total capacity of 82.67.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
&alve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves inoperable, results in adequate margin to the code allowable overpressure limit of 1375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel over-pressure of 1375 psig.
l l
219 90025041
Revised 1-17-79 3.6/4.6 Basts:
O Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. De relief valves are benchtested every second operating cycle to ensure that their set points are within the
+ 1 percent tolerance. D e relief valves are tested in place once per operating cycle to establish that they will open and pass steam, he requirements established above apply when the nuclear tystem can be pressurized above ambient condi tion s. nose requirements are applicable at nuclear system pressures below norinal 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. D e valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurised.
REFERENCES 1.
Nuclear System Pressure Relief System (BFNP TSAR Subsection 4.4) 2.
Amendment 22 in response to AEC Question 4.2 of December 6,1971.
3.
" Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 4.
Browns Ferry Nuclear Plant Design Deficiency Report--Target Ro-:k l
Safety-Relief Valves, transmitted by J. E. G111 eland to F. E. Krussi, August 29, 1973.
5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.
- 3. 6. E/4. 6. E Jet Pumps Failure of a jet pump norrie assembly holddown mechanism, nozzle aarembly and/or riser, would increase the cross-sectional flow area for blev:'ovn following the design basis double-ended line break. Also, failure cf the dif fuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made, ne detection technique is as follows. With the two recirculation Fumps balanced in speed to within + 5 percent, the flow rates in both recircula-
{
tion loops will be veri fied by control room monito
'..ig instzveents. If the evo flow rate values do not differ by more than 10 percent, riser and nozzle sesembly integrity has been verified.
220 p
Amendment No. 47 90025042
,