ML20077E197

Proposed Tech Specs,Deleting Ref to Operation of Reactor Recirculation Sys in non-loop Manual (Automatic) Mode of Flow Control

ML20077E197
Person / Time
Site:	Grand Gulf
Issue date:	05/30/1991
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20077E175	List: ML20077E166 ML20077E194 ML20077E197
References
NUDOCS 9106060217
Download: ML20077E197 (11)

Text

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" o U I I I I I OWhlOH1 GRAND GULF-UNIT 1 3/4 2-7b Amendment No. 73j [

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l- WB- 90/o x 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS

_ LIMITING CONDITION FOR OPERATION 3.4.1.1 either: The reactor coolant recirculation system shall be in operation with a.

Two recirculation loops operating with limits and setpoints per Specifications 2.2.1, 3.2.1, and 3.3.6, or b.

A single recirculation loop operating with:

1.

A volumetric loop flow rate less than 44,600 gpm, and 2.

The ic@ recirculetica flew cuni.rtrHn the eenwel acdc', cad 2 4. l Limits and setpoints per Specifications 2.2.1, 3.2.1, and 3.3.6. l Operation is not permissible in Regions A, 8 or C as specified withdrawals for startup.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With no reactor coolant system recirculation loops in operation and the reactor mode switch in the run position, immediately place the reactor mode switch in the shutdown position, b.

With operation in Region A as specified in Figure 3.4.1.1-1, immediately place the reactor mode switch in the shutdown position, c.

With operation in regions B or C as specified in Figure 3.4.1.1-1, observe the indicated APRM, neutron flux noise level. With a sustained APRM neutron flux noise level greater than 10%

peak-to peak of RATED THERMAL POWER, immediately place the reactor mode switch in the shutdown position, a d.

With operation in Region B as specified in Figure 3.4.1.1 '..

immediately initiate action to either reduce THERMAL POWER by inserting control rods or increase core flow if one or more recirculation pumps are on fast speed by opening the flow control valve to w: thin Region 0 of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

e.

With operation in Region C as specified in Figure 3.4.1.1-1, unless operation in this region is for control rod withdrawals during startup, immediately initiate action to either reduce THERMAL POWER orhours.

2 increase core flow to within Region D of Figure 3.4.1.1-1 within f.

During single loop operation, with the volumetric loop flow rate greeter t; than the above limit, immediately initiate corrective action d uce flow to within the above limit within 30 minutes.

• See Special Test Exception 3.10.4.

GRAND GULF-UNIT 1 3/4 4-1 Amendment No. 73, l

l

• WPE-90/o2 REACTOR COOLANT $YSTEM 4

LIMIT 7NG CONDITION FOR OPERATION (Continued)

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Durbg ;inglH::p op;rethn, "ith- the loop- #1ew centubeet tiahe-menwei mede ylece it. i c, Um m.ouab m e 'i+hia 15 abutu.

Ji I. During single loop operation, with temperature differences excee the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5 suspend THERMAL POWER or recirculation loop flow increase,. the l .

f J.

With a change in reactor operating conditions, from two recircula -

tion loops operating to single loop operation, or restoration of l 3.2.1, and 3.3.6 shall be implemented within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> .. ,

and take the ACTIONS required by the referenced spe SURVEILLANCE REQUIREMENTS 4.4.1.1.1 shall be verified to be in operation and not in Regions A, B in Figure control rod 3.4,1.1-1 withdrawalsexcept that operation in Region C is permissible for startup. r ng du i 4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve in by: an operating loop shall be demonstrated OPERABLE at least once per 18 a.

Verifying that the control valve fails "as is" on loss of hydraulic essure at the hydraulic unit, and b,

Verifying that the average rate of control valve movement is:

1.

Less than or equal to 11% of stroke per second opening, and 2.

Less than or equal to 11% of stroke per second closing.

• ' 1.1. 3

-f ka cc W a1Ouring in tha single nicy Upere4e, "aci*y that thc icoy c ec ; ccg uticc,-

operet h; hcp i:

Sheurs, " the - miei =ede :t hc:t erice yer-3 4.4.1.1.g During single loop operation, verify that the volumetric lo op flow rate of the loop in operation is within the limit at least once per 2A hoursI .

GR5ND GULF-UNIT 1 3/4 4-la Amendment No.73, -

WPEA '/olop, POWER DISTR 38UTION LIMITS

. BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO -

The required operating limit MCPRs at steady state operating conditions ding sients. integrity Safety Limit MCPR, and an analysis n-of abn

' For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit

, it MCPR at 2.2.

in Specification any time during the transient assuming instrum To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated tbnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

flow, The type of trusients evaluated were loss of increase temperature in pressure decrease. and power, positive reactivity insertion, and coolant When Specificationadded3.2.3 to the Safety Limit MCPR, the required operating limit is obtained.

defines the analytical basis for generation of the MCFR operating limitsT (References 2 and 3).

MCPR operating limits are defined as functions of exposure (MCPR )

, , flow (MCPR f ), and power (MCPR ).

p The limit to be used at a given operating state is the highest of these three limits.

The purpose of the MCPR, is to define operating lim!ts for all anticipated exposures during the Cycle.

exposure intervals. The MCPR, limits are established for a set of sure for each interval.The limiting transients are analyzed at the limiting expo-The MCPR, operating limits are established based on the largest delta-CPR not intervals.be exceeded during the most limiting transient in each The purpose of the MCPR and MCPR f p is to define operating limits at other than rated core flee and power conditions for all exposures during the cycle .

y The MCPR s are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured. The ref-erence core flow increase event used to establish thefMCPR is a hypothesized slow flowexcee overshoot runout to maximum, that does not result in a scram from neutron fluxl l u , w ..+., dingwo. theu APRM m ., neutron flux-high

.._4a.o m level (Table 2.2.1-1 item 2).

..a_.- .. m 2 , w + , . u- -

- r=:ut anacetien.

Janual trarriaa+ depend: :n .;heth;r th: pl:.nt 1: " L ; ":ne:1 07 Mer Ler; during Loop Manual operation is the runout of one loop because the tw recirculation loops are under independent control. A aeet of both 10:;; ie

-pou4ble-d"*4 ng weaLean Ma n > > = 1 ep:--ti:- 5::evee e ein-1; ;;atrell;r-GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 73j _[

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A/I'E 'lo/o3.  !

POWER DISTRIBUT80N LXMITS:

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BASES MINIMUM CRITIC.AL-POWER RATIO (Continued)

{- egehte: cere '!:e.

With this basis,-the'MCPR f curve /.Ngeneratedfroma series of steady state core thermal hydraulic calculations performed at several

j

t. core power and flow conditions along the steepest flow control line. .In the actual calculations a conservative highly steep generic representation of the ll 105% steam flow rodline flow control line has been used. Assumptions used in the original calculations of this generic flow control line were consistent

, with a slow flow increase transient duration of several' minutes: a the plant 1 heat balance was assumed to be in equilibrium, and (b) core xenon c(on) centration i

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GRAND' GULF-UNIT 1 1 8 3/4.2-4a Amendment No.73, s

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POWER DISTRIBUTION LIMITS-BASE'S MINIMUM CRITICAL POWER RATIO (Continued) was assumed to be constant.

i- several core power thermal-hydraulic /' low evaluations. states at which to perform steady-state core

[jg Loop Manual ead tr Ln; "r=1 mode 4

Consistent with the single failure / single / of operation N analyzed. l runout was pe:teht e fa'-postulated wa e for Loop Manual operation.d;r;;; twe . lee; rar.;;t ;n-ope Leep dea"-! :;;retien. l

' runout was assumed to be 110% of rated flow.The maximum core flow at loop '

Peaking factors were selected decrease below the MCPR Safety Limit.such that the MCPR for the bundle i

The MCPR p is established to protect the core from plant transients other 3

than core flow increase including the localized rod withdrawal error event Core power dependent setpoints are incorporated (incremental contral rod with- .

drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification

, These core setpoints thermal margins allow greater control rod withdrawal at lower core are large.

powers (3.3 where However, the increased rod 'sithdrawal.-requires higher not initial MCPR's to assure the MCPR safety limit Specification violated. ..

(2 1 2) is ments that support the RWL system are presented in Reference 4 For core power below 40% of RATED THERMAL POWEll, where the EOC-RPT and the reacto -

turbine sets separate stopof valve MCPR closure and turbine control valve fast closure are bypassed ,

countforthesignifi8antsensitivitytoinitialcoreflows.liU ts are provided fo above 40%

developed. of RATED THERMAL POWER, bounding power-dependent MCPR li are discussed in Reference 5 and the appropriateNocycle-sp change to the MCPR operating limit is required for single loop operation.

At THERMAL POWER levels. less than or equal to 25% of RATED THERMAL the reactor tor void content will will be beoperating very small. at minimum recirculation pump speed and the mod ,

For all designated control rod patterns which may be employed at this point, operating plant-experience indicates that l

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the resulting MCPR value is in excess of requirements by a considerable m .

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! GRAND GULF-UNIT I l

B 3/4 2-6 Amendment No. 73, l l

l-

/V P i 'lo / o k POWER OISTRIBUTION LIMITS BASES MINIMUM CRIT' CAL POWER RATIO (Continued)

During initval start-up testing of the plant, a MCPR evaluation will be made at 25% of 1ATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR laargin wiii thus be demonstrated such that future MCPR evaluation below this power le ul will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after-the ecmpletion of a

, THERMAL POWER increase of at least 15% of RATED THERMAL POW limits are met after power distribution shifts while still allotting time for i

the power distribution to stabilize. The requirement for calculating MCPR 2

after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod densification is less is than the design linear heat generation even if fuel pellet postaleted.

The LHGR limits of Figure "L.2.4-1 are multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC f ) or the power dependent LHGR factor (LHGRFAC p ) corresponding to the existing core flow and power state to ensure adherence to the fuel mechanical design bases during the limiting transient.

LHGRFAC f 's are generated to protect the core from slow flow runout transients, urve for Eius/ a N provided based on the maximum credible flow runout transient r Loop Manual er N:r t :p " r u:1 operation. The result of a single failure or single operator error during operation in Loop Manual.is the runout i of only oneMano=1

-Non_Looo loop because both recirculation loops are under independent control.

0; riti: . 1 tec=ute : sir,gi : r,t'eller re;;& d ; ellow tee sieei:2?tre:

f4+w, ; renvut of Mth !cep:

LHGRFACp

's are generated to i protect the core from plant transients other than core flow increases.

The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-control shifts bution rod changes. are very slow when there have not been significant power or The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL-POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.

GRAND GULF-UNIT 1 B 3/4 2-7 Amendmant No. 73j f

'. //M- 7dox

. . POWER DISTRIBUTION LfMITS BASES 1

_MININUM-GUHCAt-POWER-RAT 10-(Continued)^ ~

l

References:

l ,

1. f XN-NF-80-19(A), Volume 2 " Exxon Nuclear Methodology for Boiling Water-Reactors:

t EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, l' September 19et.

^

2.

General March 1986. Electric Company, " Maximum Extended Operating Domain Analysis,"

3.

AECM-86/0066, " Final Summary Startup Test Report 12," Letter, 0.0.

Kingsley, MP&L, to J. N. Grace, NRC, February 1986.

4 XN-NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Analyst,;

MCPRo for All Plant Operations Within the Extended Operation Domain "

Exxoh Nuclear Company, October 1986.

5.

GGNS Reactor Performance Improvement Program, Single Loop Operation Analysis, General Electric Final Report, February 1986.

j i

6 GRAND GULF-UNIT 1 B 3/4 2-7a AmendmentNo.73)--.[-

f)/PG 90/OA 3/4.4 REACTOR COOLANT SYSTEM l

BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has vided certain simits and setpoints are modified.been evaluated and fou The "GGNS Single Loop Opera-tion Analysis" identified the applicable fuel thermal limits and APM setpoint modific3tions necessary to maintain the same margin of safety for single loop operation as is available during two loop operation. Aoditionally, loop flow limitations limits. are established to ensure vessel internal vib'ation remains with

^ f b conL el mede restriction h aise Meerper4ted te reduce vel ve-

r as-a--rmult of rutentic 'E centrel et4emn+ e and to nsure valec- n.i ry in_t^ the ceVitatien Tegien uu nui UN.

l An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation accident, loop inoperable, but it does, in case of a design-bu is-the core; thus, the requirement for shutdown of the facility w inoperable.

formance on a prescribed schedule for significant degradation. Jet p During two ECCS LOCA analysis design criteria. loop operation, recirculation loop flow flow coastdown from either recirculation loop following aInLOCA.The cases limits will mitted with one loop in operation.where the mismatch limits cannot be mainta and BThe arepower / flow restricted operating from map is divided into four (4) regions.

operations. Regions A the 80% rod-line and below 40% core flow.They include the operating area above area above the 80% rod-line and between 40% and 45% core flow. Region Operation in C include Region Cfuel required is allowed only for control rod withdrawals during startup for preconditioning.

map.

No core thermal-hydraulic stability related restrictions are applie Region not D since predicted theRegion within potential D. onset of core thermal-hydraulic instabilities is data and required operater actions.The definition of Regions A, B and C is ba bility was observed in Region: A Although a large margin to onset of insta-operating configuration, a conser,vative approach is adopted in the s With no reactor coolant system recirculation loops in operation, and the reactor mode switch in the Run position, an immediate reactor shutdown is required.

Reactor shutdown is not required when recirculation pump motors are de-energized during recirculation pump speed transfers. Upon entry to Region A an immediate reactor shutdown is required.

C, unless operation in Region C is for control rod withdrawals during startup,U either a reduction of THERMAL POWER to below the 80% rod-line by control rod insertion or an recirculation increase loop in core flow to exit the region by opening the FCV is required.

Per the specification, the APRM neutron flux noise level should be observed while in Regions B and C.

In the unlikely event in which a sustained GRAND GULF-UNIT 1 B 3/4 4-1 Amendment No.s 73 - )

Uff-YO/02- l REACTOR COOLANT SYSTEM 3

BASES 3/4.4.1 -

i i

RECIRCULATION SYSTEM (Continued)  :

i APRM neutron flux noise 1gvel excceding 105 peak-to peak

{ POWER is coserved, an immediate reactor shutdown of RATED THERMAL red.

is requi i

j POWER instabilities. is established to ensure early rmal-hydraulic detectinn oMAL peak to peak of RATED THERMAL POWER were during its conditions. first three operating cycles and at different r powe /fl eactor i

stable operations of the Grand Gulf Reactor.This ow operating represents the

! margin to themel limits in the unlikelThe 105 peak-to-peak equate of R i oscillations while in Regions B and C, y event of uncontrolled limit cycle

' regional oscillations. '

i he required operator action of an immedia shutdown upon entry to Region A and upon detection of sustain j

flux noise level greater than the 105 peak-to peak ofLRATED on THEMA times, assures that an adequate margin to thermal limits will POWER be maintain i

i region, the recirevlation loop temIn order to prevent undue stre i

prior to startup of an idle loop. peratures shall be within 50*F of each other

, to the recirculation pump and recirculation .

ermal nozzles shock e

j bottom of the vessel is at a lower temperature than the coolant iSi n the upper

- regions of the core, undue difference was greater than 100*F. stress on the vessel would o emperature result if th

' tion may exist in which the coolant in the bottee head of thDuring sing circulating.

e vessel-is not to power or flow increases from this condition.These differen recirculation loop drive flowsThe recireviation flow control ual valves which in turn will vary the flow rate of and recirculation pump speed. coolant through the reacter n c

The rectreulation flow control system consists i

of the electronic and hydraulic will'generatethe two hydraulically actuated components flow contret valvesnecessary g of for%e+pos l i logic j one of. severs)a flee control valve " motion inhibit" signal in respon i

The " notion inhibit" signal causes hydraulic raulic powe isolation such that the flow control valve This fails "asfeature design is "

insures that the flow control valves do not respond to potentia i .

control signals, us inchvikl 1

GRAND GULF-UNIT 1 i 8 3/4 4-14 Ameneent No. 62j _. l-

-.= - . . . - - . - - _ .