ML20066A029

Amend 181 to License DPR-52,revising Tech Specs to Allow for Expanded Reactor Operation in Region of Higher Core Power Vs Core Flow

ML20066A029
Person / Time
Site:	Browns Ferry
Issue date:	12/18/1990
From:	Hebdon F Office of Nuclear Reactor Regulation
To:
Shared Package
ML20066A032	List: ML20066A029 ML20066A035
References
NUDOCS 9101020392
Download: ML20066A029 (21)

Text

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UNITED STATES :

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g NUCLEAR REGULATORY COMMISSION -

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WASHINGTON, D. C. 20665 '

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l TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.181 License No. DPR-52 I

1.

The Nuclear Regulatory Comission (the Comission)'has found that:

- A.

The application for amendment!by Tennessee Valley Authority (the licensee) dated June 8, 1990, complies with the standards-and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter 1; t

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of.the Comission; i

C.

There is reasonable assurance-(1) 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' s

conducted in compliance with the Comission's regulations; L

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and' safety 1of the public; and' E.-

The issuance of'this amendment is in-accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have-been satisfied.

9101020392 901210

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PDR ADOCK 050002604 P.

PDR

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b 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility.0perating License No. DPR-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specificatiens contained in Appendices A and B, as revised through Amendment No.

are hereby incorporated in the-license. The licensee shall opera,te.the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

1. Eb:.

0 rtN-Frederick J. H& don, Director Project Directorate II-4 Division of Reactor _ Projects - I/II Office of Nuclear Reactor Regulation Attachrent:

Changes to the Technical-Specifications-Date of Issuance: December 18,-1990 l

l-l l

l

ATTACHMENT'TO LICENSE AMENDMENT NO 181c FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Appendix 'A Technical Specifications by removing-the pages-identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Overleaf pages* are'provided to reintain document completeness, j

REMOVE INSERT-1.1/2.1-1 1.1/2.1-1*

1.1/2.1-2 1.1/2.1-2 1.1/2.1-3 1.1/2.1-3 1.1/2.1-4 1.1/2.1-4*

1.1/2.1-6 1.1/2.1-6 1.1/2.1-6a.

1.1/2.1-7 1.1/2.1 1.1/2.1-7a-1.1/2.1-12 1.1.2.1 1.1/2.1-13 1.1/2.1-13*

1.1/2.1-14

'1.1/2.1-14 1.1/2.1-15 1.1/2.1-15 1.1/2.1-16 1.1/2.1-16 l

1.1/2.1-16a l

3.2/4.2-25 3.2/4.2-25 3.2/4.2-25a

'3.5/4.5-20

-3.5/4.5-20 3.5/4.5-20a 3.5/4.5-20a*

l

• Denotes overleaf or spillover.page O

y a,_.

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1.1/2d FUEL CilDDING' INTEGRITY SAIITY. LIMIT LIMITING SAFETY SYSTEM SETTING

.,1,1 M L Ct. ADDING INTEGR m 2.1 - WEL CLADDING IRTEGR11Y Apoll'embility Aeolicability Applies to the interrelated Applies to trip settings of variables associated with fuel the instruments and devices thermal behavior.

which are provided to prevent the. reactor system safety limits from being exceeded.

Obiective Obiective To entablish limits which To define the level of the ensure the integrity of the process variables at which fuel cladding.

automatic protective action is initiated to prevent the fuel cladding integrity safety limit from being exceeded.

Specifications Specifications The limiting safety system settings shall-be as specified below:

A.

Thermal Power Limits A.

Neutron Flux Trio Settinns-1.

Reactor Pressure >800-1.

APRM Flux Scram pais and Core Flow Trip Setting

> 10% of Rated.

18 (RUN Mode) (Flow Biased)

When the reactor pressure is greater a.

When the Mode than 800 psia, the Switch is in existence of a minimtsa the RUN.

critical power ratio (MCPR) less than 1.07 position, the APRM flux shall constitute violation of the fuel.

scram trip setting cladding integrity shall bet safety limit.

BFN 1.1/2.1-1 Amendment 181 Unit 2

1.1/2.1 FUEL CLADDING INTEGRITY

^

SATETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flur Trio' Settings 2.1.A.I.a (Cont'd)-

S1(0.58W + 62%)

where:

S = Setting in percent of rated.

thermal power-

'(3293 MWt)

W = Loop.

recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2x106 lb/hr) b.- For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120%

-of rated thermal power.

I l

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BrN 1.1/2.1 Amendment 181

(

Unit 2 l

I

1.1/2,1 FUEL CLAhDING INTEGRITY-SAFETY LIMIT-LINITING SAFETY SYSTEM SETTING O

-2.1.A Neutron Flur Trin# Settinas 2.1.A.1.b. (Cont'd)

HQII : These settings assume operation within the basic thermal hydraulic design criteria.

These criteria are e

LHGR 113.4 kW/ft and MCPR

.within limits of Specification 3.5.K.-

It' it is determined that either'of these design criteria is being violated during operation, action shall be initiated within 15 minutes to. restore operation within-prescribed limits.

Surveillance' requirements for APRM: scram setpoint are given in= Specification 4.5.L.-

c.

The APRM Rod Block trip setting shall be:

[

S g1-(0.58W + 50%)

R L

. where:

- Syg==

Rod Block setting in

. percent of-rated F

thermal power

-(3293 MWt)

W

= Loop recirculation s

flow rate in

-percent'of rated-(rated loop recirculation

. flow rate equals

'34.~2 x 106 lb/hr)

(

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>>c w x

.u :c BFN.

1.1/2.1 Amendment 181 Unit 2 4

1.1/2.1 rtTEL CLADDING INTEGRITY SATETY LIMIT LIMITING SATETY SYSTEM SETTING j

~

1.1.A Thermal Power Limits 2.1.A Neutron Flur Trio Settints (Cont'd)

d. Tixed High Neutron Flux Scram Trip j

Setting

'"4en the mode switch is in the RUN-position, the APRM fixed high flux scram trip setting shall be S1120% power.

2.

Reactor Pressure 1800 2.

APRM and IRM Trip Settings psia or Core Flow 110%

(Startup and Hot Standby of rated.

Modes).

When the reactor pressure a.

APRM -When the is 1800 psia or core flow reactor mode switch is 110% of rated, the core

~is in the STARTUP thermal power shall r.ot position, the APRM-exceed 823 MWt (25% of scram shall be set at rated thermal power).

less than or equal to 15% of rated power.

b.

IRM-The IRM scram shall be set at less than or equal to 120/125 of full scale.

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AMENDMENT NO.14 3,181 BTN 1.1/2.1-4 l

l Unit 2

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1 130 12 0 -

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110 -

100 -

APRM Flow Biased Scram

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-20 40 00

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-12 0

' Recirculation Flow (% of Design)

APRM Flow Reference Scram and APRM Rod Block Settings Fig. 2;1-1-.

BFN 1.1/2.1-6 Amendment 181 Unit 2

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1.1/2.1-7 Amendment 181 Unit 2

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BFN 1.1/2.1-7a

/dendment 181 Unit 2

l 1

J 2.1 SiSES (C:nt'd)

In aummary

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l

?1.

The licensed maximum power level is 3,293 W t.

1 i

f i

2.

Analynes of transients employ adequately conservative values of the controlling reactor parameters.

j 3.

The abnormal operational transients were analyzed to a power i

]

level of 3,440 W t.

l 4

The analyttral procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for the parameters.

i The bases for individual setpoints are discussed belowt l

A.

Neutron Flur scram i

1.

APRM Flot & lased Blah Flur Scrag Trio Settina (RUN Mode)

The average power range monitoring (APRM) system, which is-calibrated using. heat balance data taken during 1

steady-state conditions, reads in percent of rated power (3,293 W t).

Because. fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

During power increase transients, the instantaneous fuel-l surface heat flux-is less than the instantaneous neutron flux by an amount depending upon the duration of the i

transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the-fuel-time constant. As a result of this filterint, APRM flow-biased scram will occur only if l

the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel sortace heat flux-which is equivalent to the neutron flux trip'setpoint. This setpcint is variable up to 120 percent of rated power based

'on recirculation drive flow according to the equatione given in section 2.1.A.1 and the graph in Figure 2.1-2 For the purpose of licensing transient analysis, neutron

- 1 i

flux scram-is assumed to occur at 120 percent of rated I

power. _Therefore,-the flow biased scram provides l

additional margin =to the thermal limite for slow transientsi such as loss of feedvater heating. No safety credit is l

taken for flow-blased scrams.

I 4

BFN 1.1/2.1-12 Amendment-181 Unit 2

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2. -.- ~.. -. - - - - - - -. - - -. -.

i 1

s 2.1 MAE1 (Cont'd) v l

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.07 when the transient 4

is initiated from McPR limits speelfled in Specification 3.5.k.

p h

2.

APRM Flur Scram Trin settina (Refuel or Start & Mot Standhv Mode) i For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the j

safety limit, 25 percent of rated. The margin is adeguate to i

accommodate anticipated maneuvers sesociated with power plant startup. Effects of increasing pressure at sero or low void s

content are minor, cold water from sources available during.

j startup is not much colder than that already in.the system, i

temperature coefficients are small, and control rod patterns are l

constrained to be uniform by operating procedures backed up by l

the rod worth minimizer and the Rod Sequence Control System.

l Thus, of all possible sources of reactivity input, uniform.

control rod withdrawal is the most probable cause of significant j

power rise. Because the flux distribution associated with j

uniform rod withdrawals does not involve high local peaks, and J

because several rods must be moved to change power by a significant-percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is-in near equilibrium with the fission rate.

In-an assumed uniform rod withdrawal approach 1

j to the scram level, the rate of power rise is no more than five l

Percent of rated power _per minute, and the APRM system would be more than adequate-to assure a-scram before the power could exceed the safety limit.. The 15 percent APRM serair remaine active until.the mode switch is placed in the RtTN position.

This switch occurs when reactor pressure is greater than 850 l

pais.

i i

e 3.

IRM Flum Seram Trin Rettina-l N

n

.The IRM System consists of eight chambers, four in each of the-reactor protection system logic channels. The IRM is a l

five-decade instrument:which covers the range of power level l

2 between that covered by the SRM and the APRM. The five decades-

{

are covered by the IRN by means'of a range switch and the five decades.are broken down into 10 ranges, each being one-half of a decade in sise.- The IRM scram setting of 120 divisions is 1

i active in-each range of-the IRM. ' For example if the:instriment -

l l

were on range 1,~the scram setting would be at-120 divisions for

)

i that rangeg likewise if-the. instrument.was on range 8, the scram setting would be 120 divisions on that range., e:.

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- Unit 2.

i _

1.1/2.1-13

- Amendment 181.

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2.1 R& ERA (Cze.t'd)

IaN Flur Beran Trin Battina (CentigggQ Thus, as the IBM is 'rsnsed up to accommodate the increase in I

power level, the scram setting is also rensed up. A scram at 120 divisions on the IRM instruments remains in effect as lens as the reactor is in the startup mode.

In addition, the APRM 15 percent scram prevents higher power operation without being in the RUN mode.

The IRM scram provides protection for changes which occur both locally and over the entire core. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control l

rods that heat flux is in equilibrium with the neutron flux. An IRM scram would result in a reactor shutdown well before any safety limit is exceeded.

For the case'of a single control tod withdrawal error, a ranse of rod withdrawal accidents was analyzed. This analysis included starting th: accident at various power levels. The most severe case _ involves an initial I

condition in which the reactor is just suberitical-and the IRM-system is not yet on scale. - This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the FsAR. Additional conservatism was taken in this analysis by assumins that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis.show that the reactor is scrammed and peak power limited to one percent of rated power,'thus maintaining MCPR above 1.07.

Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal:of control rode in sequence.

4. Fired Riah Neutron Flur Ecram Trig The averase power rense monitoring (APRM) system which is calibrated usins heat balance data taken durins etcady-state conditions,-reads in percent of rated power (3,293 MWt). The 1

APRM system responde directly to neutron flux.*Licensins-analyses.have demonstrated that with a neutron flux'oeram of 120 percent of rated power, none of the abn'ernal operational transients analysed violate the fuel Safety limit and there is a substantial margin from fuel damake.

B.

APRM Control kod Block s

Reactor power level may be varied by moving control rods or by varying the rectrealation flow rate.- The APRM system provides a control rod block to prevent rod withdrawal'beyond a siven point at constant recirculation flow rate and-thus to protect against the condition' of a.MCPR less than-1~.07.- This rod block trip settins,':

which is automatically varied with recirculation loop flow' rate,'

prevents an' increase in the reactor power level to excess values due i

to control rod withdrawal.: The flow variable' trip settins~provides substantial' margin from fuel damage, asesoins a steady-state operation at the trip setting over the entire power / flow domain, l

v N

SF1t 1.1/2.1-14'

' Amendment-181 tinit 2

' * ' ~ "

1

3.1 &&111 (Cont'd) including above the rated rod line (Reference 3).

The margin to the r

safety Limit increases 'as the flow decreases for the specified trip setting versus flow relationship; therefore, the wolst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because ot' the APRM rod bicek trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system.

C.

Reactor Water Low Level Scram and Isolation (Except Main Steam linea)

The setpoint _ for the low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in-F8AR subsection 14.5 show that scram and isoletion of all process j

lines (except main steam) at this level adequately protects the fuel 1

and the presstire barrier, because MCPR la greater _ than 1.07 in all cases, and system pressure does not reach the safety valve 1

settings.__The scram setting is sufficiently below normal operating range.to avoid spurious scrams.

D.

Turbine 8tto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure,-

neutron flux and heat flux increases _that would result from closure of the stop valves. With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed. '(Reference 2)

R.

Turbine Control Valve Fast Closure or Turbine Trin Scram Turbine control valve fast closure or turbine trip scram anticipates 8

the pressure, neutron flux. and heat flux increase that could result from control valve fast c1 ante due to load rejection or control valve closure due to turbine tripg each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds af ter the start of control-valve fast closure -

due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic ~ control oil pressure at the main turbine control valve actuator disc dump valves.- This Icos of pressure is sensed by preneure switches Whose contacts form the one-out-of-two-twice logic input to the reactor protection system.~ This trip setting, a nominally 50 percent greater closure _ time and a different_ valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for_the stop valve. No significant change in MCPR occure. -Relevant transient analyses are discussed in References 2 and 3 of the _ Final Safety Analysis Report. jrhis scram is bypassed when turbine steam flow is below 30- percent of rated,, as measured by

-turbine first state pressure.,

4, BrR 1.1/2.1-15L Amendment 181.

' Unit 2_

4

2.1 BAlli (Cont'd)

T.

(Deleted)

G. & H.

tigjn Steam line_Igglit,lon on Low Pressure and Main Steam Ling Isolation Scram The low pressure isolation of the main steam lines at 825 peig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.

The scram feature that occurs when the main stermline isolation valses close shute down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 pais requires that the reactor modo switch be in the STARTUP pocition, where protection of the fuel cladding integrity asfety limit is provided by the IRM and APRM high neutron flux scrams.

Thus, the combination of main steamline low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of 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. Rtactor_ tow water Levei setcoint for initiation or utcl_and Reic C1211BL_ Main Steam isolation valves. ang_Startina LPCI and core 3 Dray Pusog2 These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the TSAR demonstrate that these conditions result in adequate v

safety margins for both the fuel and the system pressure.

L.

ReferenIAA 1.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

2.

Generic Reload ruel Application, Licensing Topical Report NEDE-20411-P-A, and Addenda.

3.

Browns Ferry Nuclear Plant Unit 2, Cycis 6, Licensing Report, Extended Load Line Limit Analysis, TVA-BTE-052, April, 1990.

l l

BTN 1.1/2.1-16 Amendment 181 Unit 2 "w

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BrN 1.1/2.1-16a Amendment 181 Unit 2

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TABLE 3.2.C hE INSTRUMENTATION THAT IleITIATES ROO 8 LOCKS

~

t Minimum Operable t

Channels Per i-Tria Function (5)

Function l

Trie towel Setties

?

4(1)

Ares Upscale (Flow Blas) 10.58W + SOE (2) 4(1):

17m Upscale (Startup Rede) (8) 1121 f

1 i

4(1)

APRM Demnscale (9) 131

[

4(1)

APIBt Insperative (10b) 2(7) 84R Upscale (Flow 81as) 10.66W + 40E (2)(13) 2(7)

RRr: Dowr: scale (9) 131 2(7).

ROM Ineperative (10c) 6(1)

Ilet Upscale (8) 1108/125 of f=11 scale I

6(1)

IRR Downseale (3)(8) 15/125 of fell scale i

6(1)

Ilet Detector not in Startup Pest tien (8)

(11) 6(1)

IRR Ineperative (8)

(104)'

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P 3(1) (6) 5 set Upscale (8) i 1X105

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• Z 3(1) (6) 3RM Dounseale (4)(8) 23 counts /sec.

7 3(1) (6)

SIWI Detec*.or not in startup Positten (4)(8)

(11)

+

i

- 7 3(1) (6)

$m Ietive (8)

(10a) j U

2(1)

Flow Slas Camperator 119E difference in recirculation flows i

2(1)

Flow Slas Upscale 11151 recircolatiew flew I

1 Red 81 set Leyte N/A 2(1)

RCSC Restraint (P585-41A,8) 147 poly terbine first stage pressere 1(I2)

Migh Water Level in West 125 gal.

Scram Discherge Tank j

(N)

^

1(12)

Nigh Water Level in East

- 125 gal.

y Scras Discharge Tank g

(L5 45-45R) i I

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t BrN 3.2/4.2-25a

_ Amendment 181 Unit 2

1 3.5/4.5 CORE AND CONTAIRME!'T C00LIMG SYSTEMS LIMITING CONDITIONS FOR OPERATION

• SURVEILLANCE REQUIREMENTS 3.5 krt_and_.Grntainment Coolint Systems 4.5 G.gre and Containment Coolint Systems I

L.

APRM Setooints L.

APRM Setooints 1.

Whenever the core thermal TRP/CMTLPD shall be power is 1 25% of rated, the determined daily when ratio of TRP/CMrLPD shall the reactor la 1 25% of be 1 1.0, or the APRM scram rated thermal power.

and rod block setpoint equations listed in Section 2.1.A shall be mu.!tiplied by TRP/CMPLPD as follows:

SI (0.58W + 62%) (TR- -)

CMTLPD SRBI (0.58W 4 50%) (FRPCMTLPD) 2.

When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3.

If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to 125% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

M.

Core Thermal-Hydraulic Stability

-M. @Ip_Ihtlagl-Hydraulic Stability i

1.

The reactor shall not be 1.

Verify that the-reactor is operated at a thermal power outside of Region I and 11 and core flow inside of of Figure 3.5.M-1:

Regions I and II of Figure 3.5.M-1.

a.

Followins any increase of more than 5% rated 2.

If Region I of Figure 3.5.M-1 thermal power while' is entered, immediately initial core flow is less initiate a manual scram.

than 45% of rated, and 3.

If Region II of Figure 3.5.M-1 b.

Following any decrease is entered:

of'more than 10% rated core flow while initial thermal' power is greater than 40% of rated.

I BFN 3.5/4.5 Amendment 181 Unit 2

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l 1

3.5/4.5 CORE AMD CONTAINMENT. COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 Core and Containment Coolint_Sy.tl.tma 4.5 Core and Contafra.tAL i

Coolint Svetem9 l

3.5.M.3. (Cont'd) a.

Immediately initiate action and exit the region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by inserting control rods or by increasing core flow (starting a recircu-lation pump to exit the region is h21 an appropriate action), and i

i b.

While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by APRM escilla-tions which exceed 10 percent peak-to-peak of rated or LPRM oscillations which exceed 30 percent peak-to-peak of scale.

If periodic LPRM upacale or downscale alarm 6 occur, immediately check the APRH's and individual LPRM's for evidence of thermal-hydraulic instability.

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1 AMENDMDU ND. 174'181 BTN.

3.5/4.5-20a Unit 2 4

4