ML17053C461
| ML17053C461 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  (DPR-63-A-041, DPR-63-A-41) |
| Issue date: | 03/19/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17053C462 | List: |
| References | |
| NUDOCS 8103260427 | |
| Download: ML17053C461 (56) | |
Text
gal% RECII 8
UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET HO. 50-220 NINE NILE POINT NUCLEAR STATION UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment Ho.41 License Ho.
DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Hohawk. Power Corporation (the licensee) dated April 21, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assuratice (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and securi ty 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.
Accor'dingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C. (2) of Facility Operating License Ho.
DPR-63 is hereby amended to read as follows:
(2) Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No. 41
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
8108260 A(7
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~i&
Thorn 's A. Ippol ito, Chief Operating Reactors.
Branch 82 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 19, 1981
ATTACHMENT TO LICENSE AHENDYiENT NO. a FACILITY OPERATING LICENSE NO.
DPR-63 DOCKET NO ~ 50-220 Revise Appendix A as follows:
Remove 5
8 10ll 12 16 17 63 64 64a 64c 64e 65 66 67 68 69
~
70 70a 70d 237a 245 248 Insert 5
8 10 11 12 16 17 63 64 64a 64c 64e
~ 65 66 67 68 69, 70.
70a 70d 237a 245 245-1 248 248-1
SAFETY LIHIT LIHITIt)G SAFETY SYSTEH SETTIHG 2.1.1 FUEL CLADOl))G I t)TEGRlTY II~II,tilt Appl les to the interrel a ted. variables assoc la tcd wi th fuel t)dermal behavior.
To establish limits on thc important t)~erma) -)iydraul l c vari ablcs to.assu> e the l>>(curl ty of thc fuel cladd)ng.
~IIII a ~
'r!lien thc reactor prcssure is oreatcr than 000 ps l a and t.hr co *e flow is greater t)ian 10'A, the existence of a
)li>>inurn Cr i tical Po'e'er Ratio
())CPR) less iha>> the Safety Limit Critical Power Ratio (S4CPR)
(Reference 12) sha11 constitute v~olation c.f the fuel c."~ddir integrity safety limit~
b.
l))>en the r cactor pressure is less than or equal to 000 psia or core flow is less than 10'f rated, thc core power shall not exceed 2iX of rated t))crmal power Z.1.2 FUEL CLADO)))G l))TEGR) TY
~ttt lilt Applies to trip settings on automatic p>'otect)vc dcviccs related to variables on which the fuel loading safety limits have bccn placed.
OI~.I cc t i vc.'o prov)~)e automatic corrective action (0 lirevcn t exceeding the fuel cl adding sa fr.ty 1 lml ts.
~Seel fleg (ion'.
.Fuc l cl ac) d ) ng 1 lmi ting sa fe ty sys tem settings shall be as follows:
a.
The flow biased APAH scram trip settings shall be less than or equal to that shown in Figure 2.1.1-b.
The
)Hii scram trip setting shall not exceed 12$ of rated neutron flux.
c.
The reactor high pressure scram trip setting shall bc
< 1000 psig.
Amendment No.
jljTi
I40 ttOTCS I. nnTEpvo;tcnl5I050tr~
2, OESICtt Fl.or/I 0 7.5 g I[f>IIIr
- 3. Oc tCttTOTnl.PExKI CFn Totl -" PP
- 4. Ct)tiC rnC55un(>> ~ 000 p:4 I'20 II.0 Z
VI IJ 4I
)C
'II nC
\\I 00 40 Trr- 'IF-3.00 for a1]
Bx8 fue1
.WIIEIIEt 5~
fI'E tlcVlSCtlnt I htto t(VO UNLOCK t~T.-r- - cnccvinT co wn:<Itluis ToTnl. rEAKIttoFAcTOrl 5
Sctlnht" Itoo Ot ocK stlo'rttI ncovE 10 30 40 50 00 70 TIO Amendment No. g* 41 ncclncvl-nTlotl rcovt. rcllcctlT c'F ocstot I Figure.2.1.1.
Flo:~ Oiased Scran and llPAH Bod 01ock
BASES FOR Z.l.l FUEL CLADOIHG - SAFETY.LIHIT The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient.
Because fuel damage is not directly observable,
000 ps la and core flow 10" of rated tlic margin Lo boll ing Lr<ins i Lion is ca 1 cula Lcd from pl anL opera Ling paramcLcrs sucii as core power, core flow, fccdwa tcr Lcmi)craLurc, and core power dis tribution.
Tlic mirgin for cacli fuel assembly is characLcrizcd by the Cri tice) pounr Ratio (CPR) which is
<lie ratio o.'he handle pouer tth)cll uould produce onset of Lransi Lion boiling divided by Lhc actual bundle power.
Tlic niinimuin value of tliis raLio for any bundle in Liic core is tiic Hinlniu)n CriLical Power RaLlo (HCPlt).
I t is assuncd Lhat the iilant operation is controlled to thc nonilnal proLcctivc sct points via Lhc instrumcntcd variables, by tlic nominal expected flow conLrol linc..
Tiic
-SLCPR Iias sufficient co>>scrvatisni to assure tliat ln Lhc cvcnL of an abnovln<11 operational Lransicnt ini tlaLcd froni a nornral operating condl tlon more tlian 99.9X of tiic fuel rods in Llic core are expected to avoid boiling Lransl tion.
Tiic margin bctwccn HCPft of 1.0 (onsct of transit.ion boiling) and tlic SLCPR is derived from a detailed statistical analysis
'onsidering all of Llic uncertainties in n)onl torlng Lhc conc operating state including uncertainty ln tiic bo 1 1 ing trans l t.l on correl ation as described.in Rc fercnccs 1
and 12.
An)endn)cnt No. 5,
$J, 41
~,
BASES FOR 2.1.1 FUEI CLAOI)IHG - SAFETY LIHIT Oecause the boil ing transition correlation is based on a large quanti ty of full scale data there is.
a very high confidence that operation of a fuel assembly at the condi Lion of the SLCPR
~ would not produce boiling transition.
Thus, al though 1 t is not required Lo establish Lhe safety 1imi t, ad-di tional margin'exists between the safety 1imi t and the actual occurrence of loss of cladding integr I ty.
Iloviever, if boil ing transi tion were to occur, clad perforation would not be expected.
Cladding temperatures w'ovid increase to approximately 1l00 F which is below the perforation temperature of the cladding material.
This has been veri fied by tests in the General Electric Test Reactor (GETR) where similar fuel operated above the critical heat flux for a signi ficant period of time (30 minutes) without clad perforation.
I f reactor pressure should ever exceed 1400 psia during normal power operating (the limit of appli-cabilityy of Lhe boiling trans) tion correlation) it would be assumed that the fuel cladding integ-ri t y sa fe ty 1 imi t has been v I o 1 a Led.
In addition to Lhe boiling transition limit SLCPR operation.
Is constrained to a maximum L}IGR of l3.$ kit/It for OxO fuel and '13.4 kW/ft for OxOR 'fuel.
At 100K power this limit is reached wiLh a tbximum Total Peaking Factor (HTPF) of 3.02 for OxO fuel and 3.00 for OxUR fuel.
For the case of the IITPF exceeding these
- values, operation is permi tted only at less than IOOX of rated thermal Power and only wjth reduced RPRH scram settings as required by Specification 2.1.2.a.
(In cases where for a short period the total peaking factor was above 3.02 for OxO fuel and 3 00 for OxOR 'fuel the equation in =Figure 2.1.1 viill be used to adjust Lhe flow biased scram and APRH rod block set poinLs.
AL pressure equal to or below 000 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi.
At low power and all core flows, this pressure differential is maintained in Lhe bypass region of Lbe core.
Since Lh>> pressur>>
clroP in Lbe bypass region
$ s essentially all elevation
- head, Lhe core pressure drop at low powers and all flows will always be greater Lhaf1 n.5G psi.
Analyses show Lhat with a bundle flow of 20x10 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3'.5 psi.
Therefore, due Lo Lhe 4.5G ps i driving head, the bundle flow will be greater than ZOxIO lb/hr irrespective of total core flow and independent of bundle'ower for the range of bundle powers of concern.
Full scale ATLAS test data take~ at pres-sures from l4.7 psia to 000 psia indicate that the fuel assembly crit.ical power at 20x10 1b/t>r Amendment No. 5, 87, 41
OASES FOR 2.1.1 FUEL CLAOOIllG SAFETY LIHIT is approxlmate1y 3.35 Nit.
1<1th Llic design peaking factor, this corresponds to a core tliermal power of morc than 50X.
T)ius, a core Lliermal power limit of.25'A for reactor pressures below 000 ps I a or core flow less than 10' s
cons e rva t l vc.
Ouring transient operat/on the licat flux (thermal power-to-water) wo~ld lag beliind tlie neutron flux due to the inlicrent heat transfer time constarit of tlic fuel which is 0 to 9 seconds.
Also, tlie limiting safety system scram settings are at valves wliich will not allow the reactor to be operated above tlie safety 1 imi t during nornial operation or dirrlng oLlicr plant operating situations wliicli liavc been analyzed in detail.(3i'1)
In addit.ion, control rod scrams are sucli tliat for normal oli-crating transients tlie neutron flux transient.
is terminated before a significant increase in sur-face heat flux occurs.
Scram tlnies of each control rod are checked periodically to assume adequate
. insertion times.
Exceeding a neutron flux scram setting and a failure of tlie control rods to re-dvce flux to less than the scram setLing witliin 1.5 seconds does not necessarily imply that fuel damaoed; liowevcr, lor Lliis specification a safety limit viclaLion will be aSsunicd any time a
neutron flux scram setting is exceeded for longer Lhan 1.5 seconds.
I f tlie'cram occurs svcli that tlie neutron flux dwell Lirlie above Llie'linilt.lng safety sys tern setting is less Llian 1.7 second's, tlie safety limit will not. bc exceeded for normal turbine or generator trips,'wlilcli are Llie most severe normal operating Lransicnts expected.
Tliese analyses sllow tllat l
even if Lire bypass system fails to operate, Llic deslgii limit'of the SLCPR is not exceeded.
- Thus, use of a 1.5-second limit provides additional margiii.
Tlie process compvter lias a sequcrice annunciat.ion program wliicli will indicate the sequence in whicli sci iinls occur sucli as ncu tron f 1 ux. prcssure, c tc.
Thi s prograin a 1 so I rid i ca Les wlicn Llic scram se t point is cleared, Tliis will provide informat.ion on liow long a scram condition exists and tlius pro-vide sonic nlcasure of Llie energy added during a traris lent.
Tlius, computer inforniatlon normally will be available l'or analyzing sera>>is; liowever, if tlie coniputer information should not be available for any scrain analysis, Specification 2.l.l.c will bc relied on to determine if a safety limit lias been v i ol a Led.
Amendment No.
P, Pf
OASES FOR 2.1.2 FUEL CLAOOIIIG - LS void content are minor, cold water from source." available during startup is not much colder than that already in the'ystem, temperature coef f icinnts.are
- smal1, and contro1 rod patterns are constrained Lo be uniform by operating procedures backed up by tlute rod worth minimizer.
>lorth of individual rods is very low ln a uni form rod paLLern.-- Thus, of a11 possible sources of reacl.ivl ty input, uniform control rod wlthdrawa1 is Lhe most probable cause of significant power rise.
Oecause the flux distrlbuLion associated wilh uniform rod withdrawals does nest in-volve high local peaks, and because several rods must be moved Lo change power by a signi fi-cant percenLage of rated, the rate of power rise is very. slow.
Ceiierally, the heat flux is in near equilibrium with the fission rate.
In an assumed unl fono rod withdrawal approach Lo the scram level, the rate of power rise is no more than 5X of rated per minute, and Live IBf1 system would be more than adequate to assure a scram before Lhe power cou1d exceed Lhe safety limit.
C.
Procedural controls will assure that the IBH scram is maintained up to 20K flow.
TI(is is ac-complished by keeping lhe reactor mode switciv in Lhe startup posit:ion until 20X Flow is ex-ceeded and the Al'BH's ale on scale.
Then Lhe reacto) mode switch may be swlLched to Lhe ru(,
- mode, thereby swl Lchlng scram protection from Lhe IBH Lo the APBH sys Lem.
In order Lo ensure tlvat the IBH provided adequate proLectlon against the single rod withdrawal
- error, a range of rod withdrawal accidnts was lna1yzcd.
This anaiysis included starting the accident at various power 1evels.
The most severe case involves an lnlLlal condi Lion ln which Lhe reactor is.just s(ibcritical and th(. II!Il sysL(m is noL yet on sc>le.
Tl)is condition exists at quarler ro(l densily.
Addi Llonal conservatism was Lak(.n in Llais analysis by a~suml(ag Lhat Lhe IBH channel clusest Co Lhe withdrawn rod ls bypassed.
Tlie resui ts of tliis analysis show that Lhe reactor is sera((vned and p ak power 1 I>'>ILed Lo 1X of rated
- power, thus malntalnlng a
] j'mjt Based on the above analysis, the IBH provides protection against local control rod withdrawal errors and 'continuous>>l thdrawal ol control rods in sequence and provides backup pro l.ec t i on For the AP AH.
As demonstrated in Appendix E-I" and the Technical.
Supplement to PeLition to Increase Power I.evel, the reactor high pressure scram ls a backup to the neutron f1ux scram, turbine stop valve closure
- scram, generator load rejection
- scram, and main steam isolation valve c1osure Amendment No. 9, 8P, 41
ORSES FOR 2. 1.2 FUEL CLRDOlrfG - LS
- scram, for various reactor isolation fncfden[s.
Iiowevcr, rapid isolation at lower power 1eve1s.
genera1ly results in hfgli prcssure scram preceding other scrams because the transients are slower and those trips associated with the turbine generator are bypassed.
The operator wf1.f set Lhe trip setting at 1000 psig or 1owcr.
I(owever, the actua1 set point
'can'be as mucid as 15.0 psf above Lhe 1000 psig indicated sct: point due to Lhe deviations dfs-cusscd above.
A reactor water low level scram trip setting -12 inches (53 fncfics fndfcator sca1e) reIatfve to th~
minimum normai water level (Elevation 302'") will assure that power production wf11 bc termfnate~
with adequate coolant remafnfng fn Lhe. core.
The ana1ysfs of the feedwater pump loss fn the Tecft-nicai Supf~lement to Petition to Increase Power Level, dated April 1970, has demonstrated tIiat approximately 0 feet of water remains above the core following the low level scram.
The opel ator wi11 set the 1ow level irfp sef.ting no lower than -12 fnCfies relative to the lowest.
norma1 operating 1evcl.
Itowcvcr, thc actual sct, poi>>L can bc ar much as Z.G inches'ower due to Lhe deviations discussed above.
A reactor. water iow-low level signal
-5 feet (5 fnclies fndicat.or scale) reIatfve to tive mfnfANm normal water level (E1cvation 302'") wi11 assure Lhat core cool ing wi11 continue even i f level is dropping.
Core spray cooI fng wf11 adequately coof the core, as discussed fn LCO 3.1.0.
- Thc operator will set the 1ow-low level core spray i>>i tiation-point at no less tl>an -5 feet (5 inches fndirator scale) relative to Lhc minimum>>ormat waLcr level (t'fcvation 302'").
ilo'ever, ~
Lhc actual set point. can be as much as 2.6 inches lower duc to Lhe deviations discussed above.
Reactor power ieveI may be varied by moving control rods or by varying the recirculation flow rate.
'ihe APBH system provides a control rod block Lo prevent rod withdrawal beyond a given point at constant recirculation flow rate, a>>d thus to protect. agai>>st, the condition of a HCPB less than the St CptThis rod block trip seCting, which is automatica11y varied with recirculatfon flow rate; prevents an incr ease in thc reactor power level Lo excessive values
.due Lo control rod wi thdl awai.
The f1qw variable trip set ting provides subs tantiai margin from fuel damage
~
assuming a steady-state operation at the trip sct.ting, over the entire recirculation flow range.
Tlie margin to the safety 1imi't increases as Lhe flow decreases for thc specified trip
'setting versus flow relationship; therefore, the worst case HCPR which could occur during Ae>>>>d>>ien<
No. $
> jl>f 17
L IIIITING CONDITION FOR OPERATION SURVE ILLANCE REQUIREMENT FUEL RODS 0.1.7 FUEL. RODS Applicabilit The Limiting Conditions for Operation associated with thr fuel rods apply to those pa: ametcrs which monitor the fuel rod operating conditions.
Objective:
The objective of the Limiting Conditions for Operation is to assure the performance of the fuel rods.
Applicability:
The Surveillance Requirements apply to the pai arne Ler.
wliicI~ moni toi t tie fu~
1 rod operating condit.ions.
Objeci.ive:
The objecLive ot the Surveillance Requirements is to specify the type and frequency of surveillance to be applied t the fuel rods.
a.
Avera e Planar Linear Heat Generation Rate AVLHGR During povier operation, Lhe APLI~GR for eacli type ui'uu',
as a function of average planar exposure shall not exceed the limiting value shown in Figures
- 3. 1.7a, 3.1.7b,
- 3. 1.7c,
- 3. l. 7d and 3. 1.7e.
If at any time durinq power operaLion it is determined by non:<dl surveillance that Llie 1imiLing value for APLIIGR is being exceeded at any node in the
- core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits within two (2) hours, reactor pov~er reductions shall be initiated at a rate not. less than 10% per hour unLil APLIIGR at all nodes is within the prescribeo limits.
Specification:
a.
The API IIGR foi ca<i. type of tuel as a
function of average planar exposure shall be determined daily during reactor operation at 25% rated tliermal power.
Avera e Planar Linear Heat Generation Rate APLHGR 63
0
LIllITIHGCOHDITIOH FOR OVERATIOH SURVEILLNlCE REQUIREIIEHT b.
Linear lleat Generation Rate (LIIGR)
Ourlng power operation, the Linear lleat Generation Rat.e (LIIGR) of any rod in any fuel assembly at any axial location shall.
not exceed 13.4 I;II/FT.
b.
Linear lleat Generation Rate (LIIGR)
The LIIGR as a function of core height shall be checked dally during reactor operation at >25K rated thermal power.
If at any time during power operation I t is determined by normal surveillance that the limiting value for LIIGR is being exceeded at any location, action shall be initiated within 15 minutes to restore operation to within t.he prescribed limits.
If-I.he LIIGR at all locations is not returned to within the prescribed limits wi thin two (2) hours, reactor power reductions shall be initiated at a rate not less than 10K per hour until LIIGtl at all locations is within the prescribed 1 iml ts.
Anendment No. 5, gf, 41
1 IHITING CONDITION I'OR OPERRTION SURVI. II.LRf'>CE RC(JJIREHFNT Hiniimim Cri tical Povrer Ratio (HCI>R)
During power operation, the HCPR for all 8 x 8 fuel at rated power and flovI shall be as shovm in the table below:
LIHITING CONDITION FOR OPERATING HCPR c.
"1inimum Critical Power Ratio HCP+R
",1CPR shall be determined daily during reactor novjer operation at )25% rated thermal Dower.
d.
Power Flow Relationshi Core Averaqe Incremental
GIBED/ST EOC minus 2 GllD/ST to EOC minus 1
GlJD/ST EOC minus 1
G1JD/ST to EOC Limiting HCPR*
> 1.38
) 1.41
- l. 50 Compliance with the power flovI relationship in Section 3.1.7. d shal 1
be determined daily during reactor operation.
Partial Loo Oneration Llnder partial loop operation, surveillance requirements 4.1. 7. a, b, c, and d above are applicable.
If at any t.mme during pov!er operation it is determined by normal surveillance that these limits are no longer met, action shall be*
ini tinted wi thin 15 minutes to restore operation to vIithin the prescribed 1 imits.
If al 1 the operating HCPRs are not returned to within the prescribed limits within tv.'o (2) hours, reactor pov'er reductions shall be initiated at a rate not less than 10K per hnur until HCPR is within the prescribed limits.
For core flows other than rated the HCPR limits shall be the limits identified above times Kf vlhere Kf is as shnvm in Finure 3.1.7-1.
Power Flow Relationshi Durinq Power 0 eration The power/flow relationship shall not exceed the limiting values shown in Fiqure 3.1.7.aa.
- These limits shall be determined to be app'.'cable each operating cycle by analyses performed utilizing the ODYN transient code.
64a
i,irIITINn rnnnr TION rOR OPFRATION Associated pump motor circuit breaker shall bc opened and the breaker removed.
If these condi tions are not met, core power shall be restricted to q0.5 percent of full licensed power.
'lihen operating with three recirculation loops in operation and the two remaining loops isolated.
the reactor may operate at 100 percent of full licensed power in accordance with Figure 3.1
~ 7aa and an APLHGR not to exceed 96 percent of the limiting values shown in Figures 3.1.7a, 3.1.7b and
- 3. 1.7c, provided conditions 1
and 2 above are met for the isolated loops.
If these conditions are not met, core power shall be restricted to 90.5 percent of full licensed Dower.
During 3 loop operation, the limitin. HCPP shall be increased by 0.01.
Power operation is not permitted with less than three recirculation loops in operation.
If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR under one and two isolated loop onera-tion is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLiNR at all nodes in the core is not returned to within the nrescrihed limits for one and two isolated loop operation within two (2) hours, reactor power reduction shall be initiated at a rate not less than 10 percent per hour until APLt!GR at all nodes is within the prescribed 1 imi ts.
100 i<in. H)lc Point Vnft, 1 Limit)ng Power /F1o)v Une Cr taJ (0
G itl C4 C3 100
/'QPid!1lQAt.
>aO gag 7iP q]
i ii'i:Tii,'G PO!i"R f-LOif LjfiE 64e
NINE NILE POINT UNIT 1
10 8.57 8 67 8 58
- 8. 7Q
- 8. 71
- 8. 64 G.GO 0 50 7.72 7.25
- 6. 80 10 15 20 25 30 35 40 45 AVERAGE PLANAR EXPOSuRE (G<<O/ST)
Figure 3.1.7a IIaxinium Allo'i/able Average Planar LIIGR Applicable to 008250 Fuel as described in Reference 8
l4iendmt.nt No. g, 41 65
NINE NILE POINT UNIT 1
10 D.
O uJ 8
)
8 62 8.66 8 58 8.65 8.60 8 56
- 7. 70 7
- 7. 24 6.78 10 15 20 25 30 35 40 AVERAGE PLANAR EXPOSURE (GLJD/ST)
Figure 3.1.7b llaximum Allo>~able Average Plarar LtlGR Applicable to 808274L and 800274fl Fuel as described in Reference 8.
Amendment No. N
NINE NILE POINT UNIT 1
10 LLI (a
8 8.43 8.42
- o. 41 8.37
- 8. 35 7.51 7
7.08 6.66 10 15 20 30 35 40 AVERAGE PLANAR EXPOSURE (Gll0/ST)
Figure 3.1.7c Haximum Allowable Average Plai ir LIIGR Applicable to BDNB277 Fuel as described in Reference 8
Amendment No. jg, 41
NINE MILE POINT UNIT 1
e 10-8.44 8.43 8.42 8.41
- 8. 39 8.37 8
)
- 7. 53 7
7.09 6.56 10 15 20 25 30 35 40 45 AVERAGE PLANAR EXPOSURE (GHO/ST)
Figure 3.1.7d Maxi>num Allowable Average Planners LIIGR Applicable.to P80NB277 and Future Reload Fuel as described in Reference 8.
llmendment
>o. g, 4>
NINE NILE POIHT UNIT 1
10 8.57 8.67 0.73 8.71 8.64 8.60 8.58 0
10 15 I
20 25 30 I
35 40.
AVEPAGE PL".iNAR EXPOSURE (G"'0/ST)
Figure 3.1.7e llAXIHJH ALLOl"ABLE AVERAGE PLAiM>~.;l LIIGR APi"LICABLE'TO 80B252 FUEL AS OESCRIBED IH REFERENCE 8.
I Amendment No.
?7, 4l
R/RES I:OR I.l.i and<
I.7 FUI:L I(OI)9 Averaoe Planar Linear Heat Generation Rate
(/!PLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10CFRSO, Appendix K.
The peak cladding temperature I"ollowing a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent. secondarily on the rod-io-rod power distribution within an assembly.
Since expected local-variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than
- 4. 20 F 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 temp ratures are within the 10CFR~O, Appendix K limit.
The limiting value for APLllGR is shown in Figure 3.1.7.
These curves are based on calculations usina
~he models described in References 1, 2, 3, 5 5 6.
Analysis has been performed (Reference 7) which sho)ls for iso'lation of 1 loop, operation limited to 90'A of t0ie limiting APL)lGR shown in Figure 3.1.7 conservatively assures compliance with lOCl-R~O, Appendix K.
Linear i/eat Generation
.Rate Ll(GR)
Ttlis specification assures that the linear heat generation rate it> any rod is less than the design linear heat qeneration even if fuel pellet densification is postulated (Reference 12).
The LHGR shall be checked daily during reactor operation at
> 25/ power to determine if fuel burnup or control rod movement has caused changes'in power distribution.
Hinimum Critical Power Ratio tiCPR At core thermal power levels less than or equal 'to 25$, the reactor will be operating at a minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employ at this point, operating plant experience and thermal-hydraultc analysis indicated that the resulting HCPR value is in excess of requirements by a considerable margin.
ttith this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative t'o HCPR.
During initial startup testing Amendment No.
7.g, 41
of the plant, a
HCPR cvaluatio.",;;H'.
b made at the 25;l thermal power level Mith minimum rccircvlation pump speed.
The HCPR margin v~ili thus b
demonstrated sue): that future YiCPR evaluations below this po(e'er level will be shown to be unnecessary.
The daily
) cquircmcnt for calcu)at"',ng i~)CPR above 25K rated thermal power is sufficient since power d)stribution shifts are very slow w)>cn there have no't veen significant power or control rod changes.
The requirement for calculating
)iCPR when a limiting control rod pattern is approached ensures Chat llCPR wi" 1 be known fol',owing a'hange in power or power shape
{regardless of magnitude) that could place operation at a
L)~crmal limit.
Figure 3.1.7-1 is use() for calculati>>g HCl'R during operation at other Ulafl rated cond)tions.
For the case of automatic flow control, thc
)l facL'or is drterm)ned such that any automatic
'Increase in power {dvc to flo'v control) will always result in arriving at't)ie nominal required HCPR at 100K power.
For manual flow control, the )!f is detcrniincd such that an inadvertc>>L increase in core flow {i.e., operator error or rccircvlation puinI) speed controller failure) would result in arriving at the 99.9X limit HCPR when core flow reaches thc maxi In'nn possi)>le core flow corresponding to a particular setting of the recirculation pump l'G set scoop tube maximu'n speed cont>ol lin>iti;ig set screws.
These screws are to be calibrated arid set to a particular value a>>d whenever the plant is operating in manual l'low control thc Yf defined. by that setting of the screws is to b used in the deterninatio>> of required HCPR.
T))>s will assure that the reduction in hCPR associated with an inadvertent floki increase always satisfies the 99.9Ã requirement.
Irrespective of thc scoop tube setting tile rcOUiY'e'0 i'IC)'R 'ls ncvc3 allowed to bc less than t))e nominal
)iCPR (4.e.,
Kf is never less than vnity).
Power/F)ow Relations~hi~
The power/flow curve is the locus of critical pow r as a function of'lo < from which. the occurrence of abnormal operating transients will y'lcld results within defined plant safety limits.
Each transient and postulated accigcnt appllcablc
'to operation of the plant.>]as analyzed along the pov'ier/floe] line.
The ana",:.i4" 8~12) justifies thc operating envelope bounded by the power/flow curve as long as other operatin limits are satisfied.
Operati".". under
'he power/flow 'ling'is designed to enable the direct ascension to fulj power wit)>in th des j<:"l l.="si" ~"r t".e plant.
A",endment No.
!l"-FEREil"ES FOR JASES 3.1.7 llllD ~i.l.7 F/JfL AQDS
( 1}
"Ful DensifIcatkon Effects on G"n"!al Electric Dolllng >Jater Reactor Fuel," Supplements 6,
7 and 8, llED'l-10735, Rugust 1973.
(2)
Sup;ilcocnt 1 to Tcchnical Rcpo'. t Dn Densification; of General Electric Reactor Fuels, December 14, 1974 (USA fl"gu) a lory 5 ta ff).
(3)
Cor,.-.unicat.inn'.
V. R. ilaore to 1. S. Hitchel 1, "llcdLflcd GE Hodcl fo>
Fuel Densification," Oocket 50-321, ll~rch 27, 1974.
(4)
"General Electric Ooiling ilatcr Ileactor Generic Reload Rppl /cation for 0 x 0 Fuel," llE00-20360, Supplement 1 to 1;evision 1, Oeccniber 19/4.
(5)
"0"neral Elc tr(c Company Ilnal~'Uic.l i'oriel;or Lo" 5 of Coolant Rnalysis ln Accordance Mith 10CFR50 Appendix K,"
ill.l)0-20566.
(6)
General Electric AefiH lief iood Calculation (Supplcmcnt to SAFE Coile Descript.ion) transmitted to the USAEC by lcLtcr; G. L. Gyorcy (o Victo> Sicl
>o.)r., Ja,'cd l)ccvmi>er 20, 1974.
{7} "/line illle Point flucicar l ones Stal.ion 0nit 1, Load L'inc Limit Analysis," ll'00-24012 ~
'0)
Licensing Top>cal Be>ort G"ner'll E',ec'u lc Uo>ling llater Reactor Generic Reload Fuel Application, llLl)E-B01i-P-A, August,
'9i'G.
(9)
Final Safety Analysis Repor", iodine.Nile Point Nuclear Station, niagara iiohawk Power Corporation, June 1967.
(10) tlRC Safety Evaluation,,"imend;;:-;nt Ho.
24 to DPR-63 contained in a letter from George Lear, NRC, to D. P. Disc dated llay l5, 197Q.
(ll) "Core Flow Oislribulion i:
", 6>> ~",al E]ect: ic Boiling lfater Reactot as Heasured in guad Cities Unit 1,"
JlEl)0-10722A.
(12) Wine Nile Point Jluclear Power Station Unit 1, Extended Load 1
~<c ~..> ni.,
=x;en<ed Load Line Lir>it Analysis, License Amendment Submittal 70d
BASES FOR 3.6.3 At<0 4.6.2 PROTECTIVE INSTRUMENTATION The control rod block functions are provided to prevent excessive control rod withdrawal so that HCPR is maintained greater than the SLCPR.
The trip logic for this function is I oui'f n; e.g.,
any trip on one of the eight APRH's, eight IRH's or four SRH's will result in a rod block.
The minimum instrument channel requirements provide suf'licient instrumentation to assure the single failure criteria is met.
The minimum instrument channel requirements for'he rod block may be reduced by one for a short period of time to allow maintenance, testing, or calibration.
This time period is only MX of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The APRH rod block trip is flow biased and prevents a significant reduction in HCPR especially during operation at reduced flow.
The APRH provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.
The trips are set so that HCPR is maintained gr< <ster than the SLCPR.
The APRH rod block also provides local protection of the core; i.e.,
the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.
The trip point is flow biased.
The worst case single control rod withdrawal error has been analyzed and the resul ts show that with the specified trip settings rod with-drawal is blocked before the HCPR reaches the
- SLCPR, thus allowinq adequate margin.
Below<60'!
power the worst case withdrawal of a single control rod results in a
MCPR SLCPR wi thout rod block action, thus below this level it is not required.
The IRH rod block function provides local as wrl 1 as gross core protection.
The scaling arrange-ment is such that trip setting is less than a factor of 10 above the indicated level.
Analysis of the worst case accident results in rod block action before MCPR approaches the SLCPR.
A downscale indication on an APRH or IRH is an indication the instrument has fai led or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented.
The downscale rod blocks are set at 5 percent of full, scale for IRH and 2 percent of full scale for APRH (APRH signal is generated by averaging the output signals from eight LPRH flux monitors).
Amendment l'lo.
237a
I
6.O ADHI<iisn~nTrVE ConnOLS 6.1 Re~s)ons ibi~lit' 6.1.1 The General Superintendent for Nuclear Generation shall be responsible for overall facility operation and shall delegate in wri ting the succession to th'is responsibility dui ing his absence.
6.2
~0n anization Offsite 6.2.1 The offsite organization for facility management and tcchnical support shall be as shown on Figure 6.2-1.
6.2.2 'he Facility organization sha'll be as shown on Figure 6.2-2 and:
a.
Each on duty shift. shall be composed of't least the minimum shift crew composition shown in Table 6.2-1.
b.
At least one licensed Opera to> shall be in the control room when fuel is in the reactor.
Ouring reacto>
operation this licensed operator shall be present at the controls.of the faci 1 i ty.
c..Rt least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
d, An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
e.
A'enior Licensed Operator shall be responsible for all movement of new and irradiated fuel within the site boundary.
A Licensed Op rator. will be required to manipulat the controls of all fuel droving equipment except the reactor building crane.
All fuel movements by the reactor building crane except new fuel movements from receipt; through d)y storage shall be under the direct supervision of a Licensed Operator.
All/fuel moves>>ithin the core s lail bc directly monitored by a member nl'he reactor analyst group.-
Effective until the end of fuel Cycle 6.
245
6.0 RDt~l NI S 1 IWl'IVE CONTROLS 6.1.1 The General Superintendent for Nuclear Generation shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
62
~66 ll Offsite 6.2.
1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.
Facilit Staff 6.2.2 The facility organization shall be as shown on Figure 6.2-2 and:
a.
Each on-duty shift shall be composed of at least the minimum shift ere~
composition shown in Table 6.2-1.
b.
At least one 1icensed Operator shall be in the control room when fuel is in, the reactor.
During reactor operation, this licensed operator shall be present at the controls of the facility.
c.
At least two licensed Operators shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be on site vhen fuel is in the reactor.
e.
ii licensed Senior keactor Operator sba'll be re:ponsible for a'll n2ove6aent of ne62 and irradiated fuel within the site boundary.
Rll core alterations shall be directly supervised by a licensed senior reactor operator who has no other concurrent responsibilities during this operation.
A Licensed Operator will.be required to manipulate the control: of all fuel handling equipment except movement of new fuel from receipt through dry storage.
All fuel moves within the core shall be directly monitored by a member of the reactor analyst group.
Effective for fuel cycle 7 and all refuelings thereafter.
245-1
License.
Table'.2-1 I
ttIHi."lv'H S!!IFT C!iC'c'Os'ia"OSITION Op."t-- (')
Rc.cto.(")
tiormal Opera tion Shutdown Condi tion H/0 Process Computer Startups Senior Operator Operator Unlicensed
(
Hotes:
(1)
At any one time more licensed or unlicensed operating people could be present for maintenance,
- repairs, fuel outagcs, etc.
(2)
Those operating personnel not holding an "Operator" or "Senior Operator" License.
(3)
For operation longer than eight hours without process computer.
I (0)
For reactor startups, except a scram recovery where the reason for scram is both clearly und r-stood and corrected'ffective until the end of fuel Cycle 6
Table 6.2-1 CREM CO"'iPOSIT'GV<
)
License Senior Operator Normal Operation Shutdown Condition 1(6)
Operation(
!i~/0 Process Computer
'eactor(4)
Startups Operator Unlicensed(2)
1(5) l,'otes:
(1) At any one time, more licensed or unlicensed operating people could be present for maintenance,
- repairs, fuel
- outages, etc.
(2) Those operating personnel not holding an "Operating" or "Senior Operator" License.
(3) For operation longer than eight hours without (4) For reactor startups, except a scram recovery (5) loot shutdown condition only.
process computer.
where the reason for scram is both clearly understood and correcte (6) An additional senior reactor operator who has no other concurrent responsibilities shall supervise all core alterations.
Effective for fuel cycle 7 and all refuelinga thereafter.
248-1