ML20128E989

Amend 84 to License DPR-22,permiting Implementation of Expanded Operating Domain Resulting from Max Extended Load Line Limit Analysis & Increased Core Flow

ML20128E989
Person / Time
Site:	Monticello
Issue date:	01/27/1993
From:	Marsh L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20128E993	List: ML20128E989 ML20128E999
References
NUDOCS 9302110199
Download: ML20128E989 (10)

Text

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UNITED STATES J"

NUCLEAR REGULATORY COMMISSION n

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NORTHERN STATES POWER COMPANY QQCYLT NO. 50-2Q tiONTICELLQ NUCLEAR GENER/tIJRG. PLANT NiWDliWI_l0_JjiCILITY OPERATING LICENSE Amendment No. 84 License No. DPR-22 1.

The Nuclear Regulatory Connission (the Commission) has found that:

A.

The application.or amendment by Northern States Power Company (the l

licensee) dated September 16, 1992, complies with the standards.and l

requirements of the Atomic Energy Act of 1954, as amended (the Act),

I and the Corraission's rules and regulations set forth in 10 CFR Chapter 1;

1.

B.

The facility will operate in conformity with the application, the 1

provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activitius authorized by l

l this. amendment can be conducted without endangering the health and l

safety of the public, and (ii) that such activities will be conducted l

in compliance with the Commission's regulations; U.

The issuance of this amendmont will not be inimical to the common defense and security or to che health and safety of the public; and l

E.

The issuance of this amendtent is in accordance with 10 CFR part-51 of the Commission's regulatior.s and all applicable requirements have been-satisfied.

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 Operating License No. OPR-22 is hereby

-amended to read as follows:

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i 9302110199 930127 PDR. ADOCK 05000263,

-P PDR b

It.chnitLLSrftif_tratius The Technical Specifications contained in Appendix A, as revised through Amendment flo.

84, are hereby incorporated in the license.

The licensee shall operate the facility in accordence with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance.

FOR THE flVCLEAR REGULATORY COMMISSI0ld

['W Ledyard B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects - III/IV/V Office of fluclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

January 27, 1993

AUEEMEELIQ_LICIllSE AMEllDMEfG_EQ. 84 fM1L1TY OPERATItLG_LiffliSLtta, EEE-22 QQCKET f40. 50-763 Revise Appendix A '. ethnical Specifications by removing the pages identified below and inserting who attac1ed pages.

The revised pages are identified by amendment number t.nd contain vertical lines indicating the areas of change.

ELM 0E JE1EBI 6

6 15 15 16 16 28 28 56 56 a

58 58 249b 249b i

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2.0 SAFETY LIMITS' LIMITING SAFETY SYSTEM SETTINGS n

-2.1 FUEL CIADOING INTEGRITY 2.3 FUEL CIA) DING INTEGRITY 4

Applicability Applicability 7-Applies to the' interrelated variables associated Applies to trip settings of the instruments and with fuel thermal behavior devices which are provided to prevent the reactor system safety limits from being exceeded.

Oblectivel Obiectivet To define the level of the process variables at T

~~

To establish. limits below which the integrity which e-stomatic protective action is initiated to i

of - the fuel cladding is preserved.

prevent the safety limits from being exceewd.

Specification:

i:

Specification:

The Limiting safety system settings shall be at specified below:

A.

Core Thermal Power Limit (Reactor Pressure >800 psia and Core Flow is >10% of Rated)

A.

Neutron Flux Scram When the reactor pressure is.>800 psia and core 1.

APRM - The 'APRM flux s.: ram trip setting flow is >10% of rated,L the existence of a shall be; minimum crirical power ratio'(MCPR) less than a.

For two recirculation loop operation 1.07, for two recirculation loop operation, or (TLO):

-less than 1.08 for single loop operation, shall S 5 0.66V + 70s

where, constitut.e violation'of the fuel. cladding S - Setting in percent of rated integrity safety limit.

thermal power, rated power being 1570 MWT V - Percent of the drive flo.e required to produce a rated 6

i

^

core flow of 57.6 x 10 lb/hr b.

For single recirculation loop operation (SIA):

S $.0.58(W - 5.4) + 624

c. No greater than 1204.

2.1/2.3 6

Amendment No. 29, 47, 84

1 I

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Bases Continued:

1 i

l For analyses of the thermal' consequences of the transients, the Operating MCPR Limit (T.S.3.11.C) is i

conservatively assumed to exist prior to initiation of the transients.

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This choice of using conservative values of controlling parameters and initiating transients at the design

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j; power level, produces mere pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

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Deviations from as-left settings of setpoines are expected due to inherent instrument error, operator setting l

l error, drift.of'the setpoint, etc.

Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to

.j F

that of the conservatisms discussed above. Although the operator will set the setpoints within the trip l

settings specified, the actual values of the various setpoints can vary from the specified trip setting by cl.e

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allowable deviation.

.j b

A violation of. this specification is assmed to occur only when a device is knowingly set outside I

of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the i

reactor modes in which the functions listed above are required.

A.

Neutron Flux Scram The average power range monitoring (AFRri) system, which is calibrated using heat I

balance data taken during steady state conditions, reads in percent of rated thermal power (1670

[

MWt).

Because.. fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel I

(reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the l

c l-fuel. Therefore, during abnormal operstional transients, the thermal power of. the fuel will be less than that. Indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 1204 i

scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety

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Limit and there is a substantial margin from fuel damage. The re fore, the use of flow teferenced i

scram trip-provides even additional' margin.

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t 2.3 BASES 15 l

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Amendment No. 84

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Bases Continued-Maximum Extended Load Line Limit Analyses have been perforced to allow operation at higher powers at flows below 87t.

The flow referenced scram (and rod block line) have increased (higher slope and y-intercept)

{ ;o forftwo loop' operation (See Core Operating Limits Report). These analyses have not changed the allowed operation for single. loop operation. The supporting. analyses are discussed in CE NEDC-31849P report (Reference : Letter from NSP to NRC dated September 16, 1992.).

s

!o Increased Core Flow analyses have been performed to ' allow operating at flows above 100% for powers equal to 2

. or less than 1001 (See Core Operating. Limit Report). The supporting analyses are discussed in General

. Electric NEDC-31778P report (Raference-Letter from NSF to NRC dated September 16, 1992).

L For operation in tha startup mode while' the reactor! is at low pressure, the IRM scram setting of 20%

of

. rated power. provides adequate thermal margin between the setpoint and the safety limit, 25% of rated. The margin 'is adequate to accommodate anticipated maneuvers associated vitis power plant startup.

Effects of increasing pressure at zero or low void content are minor, cold water' from sources available during startup is not much colder than that already in the system, teeperature coefficients are small, and control red patterns are constrained to be uniform by operating procedures.

Worth of' individual rods is.very low in a uniform rod pattern.

Thus, of all possible sources of reactivity input, uniform control' rod withdrawal is the mest probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to chenge 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 to the scram level, the rate of power t-rise-is no more than 5% of rated power per minute, and the IRM system would.be more than adequate to l-assure a scram before the power could exceed the safety limit. The IRM scram remains active until the mode switch'is placed in the run position and the associated APRM is not downscale. This switch

}'

occurs when reactor pressure is greater than 850 psig.

The operator will set the APRM neutron flux trip setting no greater than that stated in Specifica-tion 2.3.A.l.

However. the actual setpoint can be as much as 34 greater than that stated in Specification 2.3.A.1 for recirculation d*iving flows less than 50% of design and 24 greater l'

than that shown. for recirculation driving flows greater than 50% of design due to the deviations discussed on page 39.

0

'B.

Deleted 11 2;3 BASES-16 Next Page.is 18 i--

1J Amendment No. 29, 60, 63, 85

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1 2

TABLE 3.1.1 i

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS

-r i

- Modes in which func-Total No. ot Min. No. ot Operable tion must be Oper-Instrument or Operatinr. Instru-Limiting able or Operating **

Channels per ment Channels Per Required Trip Function Trip Settin&s Refuel (3) Startup Run Trip System Trip System (1)

Condition

• 1.

Mode Switch in Shutdown I

X X

1 1

a

- 2.

Manual Scram X

X X

1 1

A-t

-3.

Neutron Flux 1RM s 120/125 l-(See Note 2) of full scale X

X 4

3 A

a. High-High

b. Inoperative i

i-l 4

Flow Referenced See Specifi--

Neutron Flux APRM cations (See Note 5)-

2.3A.1 X

3 2

A or E i

a. High-High

b. Inoperative

.c. High Flow Clacp. g 120 %-

l 5.

High Reactor Pressure s 1075 psig X

X(f)

X(f) 2 2

A

.(See Note 9) j-i:.

6.

High Drywel1 l'

Pressure-s 2 psig

'X.

~ X(e f) X(e.f) 2 2

A-(See Note 4) l 7.

Reactor low Water level

= 7 in.(6)

X-X(f)

X(f) 2 2

A I

8.

Scram Discharge

. Volume High level a.-East s 56 gal.(8)

X(a)

X(f)

X(f) 2 2

A

b. West s'56 gal.(8)

X(a)

X(f)

X(f) 2 2

A

~ Turbine Coluknser

- 9.

I.ow Vacuu:n.

= 23 in._Hg.

X(b)

X(b.f) X(f) 2 2

A or C 3;1/4.1-28 I

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.unen ment No. II. 50, 63, 84i a

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IR I IE I IABLE, 3. Z. 3 Instrumentation That Initiates Rod Block Reactor Modes Wilich Function Must be Operable Total No. of Mi.n. No. of Oper-or Operating and Allow-Instrument able or Operating able Bypass Conditions **

Channels per Instrument Channels Required Function Trip Settint'4 Pefuel Startup Run Trip Sys tern per Trip System Conditions *

1. m a.

Upscale s5x105 X

X(d) 2 1(Note 1, 3, 6)

A cr 3 or C cps b.

Detector y(a)

X(a) 2 1(Note 1, 3, 6)

A or B or C not fully inserted

2. E a.

Downscale 23/125 X(b)

X(b) 4 2(Note 1, 4, 6)

A or B or C full scale b.

Upscale s108/125 X

X 4

2(Note 1, 4, 6)

A or B or C full scale

3. 8EBM X

3 1(Note 1, 6, 7)

D or E a.

Upscale

.(1) TIA Biased ~< O.66V + 58%

Flow (Note 2)

(2) SID Flow f 0.58(V - 5.4) + 50%

Stased (Note 2)

(3) High $ 108%

Flow Ciamp 3/125 full scale X

3 1(Note 1, 6, 7)

D er E b.

Downscale 2

56

'!.2/4.2 Amendment No. 29, 47

1 Table 3.2.^.,- continued Instrumentetion that Initiates Red Elort i

30l!21 -

1

' (1) There shall be two' operebte er operatino trip systems for each fmetton. If the mininass rumbar of operable or operJting instrument charewits camot be met for cro of the two trip systems, this condit:m may exist up to sevm days orovidM that dJring this time the egnfabte system is ftrctionelty tested lessm$lately and daily thereof ter.

s (2) *se is the percent of drive flow regsired to prodxe a rated core f toi. cf 5/.6 m to tb/hr (3) only one of the four SRM chamels any be htessed.

(6) There sust be at least one :iperable or og.erating IRM channet monitoring eact. core gJadrant.

(5) An ass channet ;,ilt te considered inoperebte if there are tess tte half the toint ramber of norest ircuts..

'(6) Upon discovery that minieum regJireaents for the rumber of cperable er coerating trip systems or instrument chamels are riet satisfied actions shalt be initiated to:

(a) Satisfy the regJirements by placies appropriate chamets o system in the triseed condit on or i

t (b) Place the plant trder the specified required cenditicris using normat operating procedLres.

(7) There must be a total of at least 4 operable or cperating APRSt chamels j

(8) There e-e 3 upscete trip tevels. Only one is applied over a sgecified everating core therac s :mer range. Att asst trips

- are automaticetty bypassed below 3C% thernet power. Trip settings are provided in the Core Cperating Limits Report.

- 3.2/6.2 58 i

Amend:nent No. 29. H. 70, 84

7.

Core Operatine Limits Report Core operating limits shall be established and documented in the Core Operating Limits Report before each reload cycle or any remaining part of a reloed cycle for the following:

a.

Rod Bicek Monitor Operability Requirements (Specification 3.2.C.2a)

Rod Block Monitor Upscale Trip Settings (Table 3.2.3, Item 4.a)

Maxiaum Average Planar Linear Heat Generation Rate Limits (Specification 3.ll.A)

Linear Heat Generation Ratio Limits (Specification 3.11.B)

Minimum Critical Power Ratio Limits (Specification 3.11.C)

Power to Flow Map (Bases 2.3.A)

I The analytical methods used to datermine the core operating limits shall be those previously reviewed b.

and approved by the NRC, specifically those described in the following documents NEDE-240ll-P-A, *Ceneral Electric Standard Application for Reactor Fuel" (latest approved version)

NSPNAD-8608-A, " Reload Safety Evaluation Methods for Application to the Monticello l

Nuclear Generating Plant" (latest approved version) l NSPNAD-8609-A, "Qutlification of Reactor Physics Methods for Applic8 tion to Monticello" (latest approved version) limits (e.g., fuel thermal-The core operating limits shall be determined such that all applicablecore thermal-hydraulic limits, ECCS limit c.

wechanical limits, transient analysis limits and accident analysis limits) of the safety analysis are met.

including any mid-cycle revisions or supplements shall be supplied d.

The Core Operating Limits Report, to the NRC Document Control Desk with copes to the Regional upon issuance, for each reload cycle, Administrator and Resident Insp(

'ar.

249b 6.7 O

Amendment No. 70. E4

.