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)

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I OWhlOH1 GRAND GULF-UNIT 1 3/4 2-7b Amendment No. 73

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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 The reactor coolant recirculation system shall be in operation with either:

Two recirculation loops operating with limits and setpoints per a.

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 l

2 4.

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:

With no reactor coolant system recirculation loops in operation and a.

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, With operation in regions B or C as specified in Figure 3.4.1.1-1, c.

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 />.

With operation in Region C as specified in Figure 3.4.1.1-1, unless e.

operation in this region is for control rod withdrawals during startup, immediately initiate action to either reduce THERMAL POWER or increase core flow to within Region D 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 />.

f.

During single loop operation, with the volumetric loop flow rate greeter than the above limit, immediately initiate corrective action t;

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 LIMIT 7NG CONDITION FOR OPERATION (Continued) 4 Durbg ;inglH::p op;rethn, "ith-the loop- #1ew centubeet ia

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menwei mede ylece it. i c, Um m.ouab m e 'i+hia 15 abutu. t he-Ji During single loop operation, with temperature differences excee I.

the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5 l

THERMAL POWER or recirculation loop flow increase,.

suspend the 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 3.4,1.1-1 except that operation in Region C is permissible du i control rod withdrawals for startup.

r ng 4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve in an operating loop shall be demonstrated OPERABLE at least once per 18 by:

Verifying that the control valve fails "as is" on loss of h d a.

essure at the hydraulic unit, and y raulic 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 Ouring single nicy Upere4e, "aci*y that thc icoy

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operet h; hcp i: " the - miei =ede :t hc:t erice yer-

Sheurs, 3

4.4.1.1.g During single loop operation, verify that the volumetric lo rate of the loop in operation is within the limit at least once per 2A hours op flow I

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

A WPE '/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 integrity Safety Limit MCPR, and an analysis of abn sients.

For any abnormal operating transient analysis evaluation with the n-initial condition of the reactor being at the steady state operating limit MCPR at any time during the transient assuming instrum

, it in Specification 2.2.

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).The type of trusients evaluated were loss of

flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

When added to the Safety Limit MCPR, the required operating limit Specification 3.2.3 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 ), and power (MCPR ).

The limit to be used at a given operating state f

p 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.

The MCPR, limits are established for a set of exposure intervals.

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 be exceeded during the most limiting transient in each intervals.

The purpose of the MCPR and MCPR is to define operating limits at other f

p than rated core flee and power conditions for all exposures during the cycle The MCPR s are established to protect the core from inadvertent core flow y

increases such that the 99.9% MCPR limit requirement can be assured.

erence core flow increase event used to establish the MCPR The ref-f is a hypothesized slow flow runout to maximum, that does not result in a scram from neutron flux l

overshoot excee u, w..+., ding the APRM neutron flux-high level (Table 2.2.1-1 item 2).

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- r=:ut trarriaa+ depend: :n.;heth;r th: pl:.nt 1: " L ; ":ne:1 07 Mer Ler; Janual anacetien.

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 w a e Lean Ma n > > = 1 ep:--ti:- 5::evee e ein-1; ;;atrell;r-GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 73

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POWER DISTRIBUT80N LXMITS:

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BASES MINIMUM CRITIC.AL-POWER RATIO (Continued)

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egehte: cere '!:e.

With this basis,-the'MCPR curve /.Ngeneratedfroma f

series of steady state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line..In the

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actual calculations a conservative highly steep generic representation of the 105% steam flow rodline flow control line has been used.

Assumptions used in ll the original calculations of this generic flow control line were consistent with a slow flow increase transient duration of several' minutes:

heat balance was assumed to be in equilibrium, and (b) core xenon c(on) centration a the plant 1

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t GRAND' GULF-UNIT 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.

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

[jg Loop Manual ead tr Ln; "r=1 mode Consistent with the single failure / single / of operation N analyzed.

l runout was postulated for Loop Manual operation.d;r;;; twe. lee; rar.;;t ;n-ope 4

pe:teht e fa'- wa Leep dea"-! :;;retien.

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

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

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

than core flow increase including the localized rod withdrawal error event 3

Core power dependent setpoints are incorporated (incremental contral rod with-drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification These setpoints allow greater control rod withdrawal at lower core powers (3.3 core thermal margins are large.

However, the increased rod 'sithdrawal.-requires where higher initial MCPR's to assure the MCPR safety limit Specification (2 1 2) is not violated.

ments that support the RWL system are presented in Reference 4 below 40% of RATED THERMAL POWEll, where the EOC-RPT and the reacto For core power turbine stop valve closure and turbine control valve fast closure are bypassed separate sets of MCPR countforthesignifi8antsensitivitytoinitialcoreflows.liU ts are provided fo above 40% of RATED THERMAL POWER, bounding power-dependent MCPR li developed.

are discussed in Reference 5 and the appropriate cycle-sp change to the MCPR operating limit is required for single loop operation.

No At THERMAL POWER levels. less than or equal to 25% of RATED THERMAL the reactor will be operating at minimum recirculation pump speed and the mod tor void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant-experience indicates that the resulting MCPR value is in excess of requirements by a considerable m l

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l GRAND GULF-UNIT I B 3/4 2-6 l

Amendment No. 73, l

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/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 after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that 2

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 is less than the design linear heat generation even if fuel pellet densification is postaleted.

The LHGR limits of Figure "L.2.4-1 are multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC ) or the power dependent LHGR factor f

(LHGRFAC ) corresponding to the existing core flow and power state to ensure p

adherence to the fuel mechanical design bases during the limiting transient.

LHGRFAC 's are generated to protect the core from slow flow runout transients, f

for Eius/ a N provided based on the maximum credible flow runout transient urve 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 one loop because both recirculation loops are under independent control.

-Non_Looo Mano=1 0; riti:. 1 d ; ellow ei:2?tre: ; renvut of Mth !cep:

tec=ute : sir,gi

r,t'eller re;;& tee sie f4+w, LHGRFAC 's are generated to p

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-bution shifts are very slow when there have not been significant power or control rod changes.

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. 73 f

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.. POWER DISTRIBUTION LfMITS BASES 1

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

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References:

f 1.

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

EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, t

September 19et.

^

2.

General Electric Company, " Maximum Extended Operating Domain Analysis,"

March 1986.

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.

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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 are established to ensure vessel internal vib'ation remains with limits.

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

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

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-bu is-

accident, the core; thus, the requirement for shutdown of the facility w inoperable.

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

and B are restricted from 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 C include Region C is allowed only for control rod withdrawals during startup for Operation in required fuel preconditioning.

No core thermal-hydraulic stability related restrictions are applie map.

Region D since the potential onset of core thermal-hydraulic instabilities is not predicted within Region D.

data and required operater actions.The definition of Regions A, B and C is ba Although a large margin to onset of insta-bility was observed in Region: A 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 increase in core flow to exit the region by opening the recirculation loop 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. 73 -

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Uff-YO/02-REACTOR COOLANT SYSTEM 3

BASES 3/4.4.1 i

RECIRCULATION SYSTEM (Continued) i APRM neutron flux noise 1gvel excceding 105 peak-to peak i

POWER is coserved, an immediate reactor shutdown is requi of RATED THERMAL

{

POWER is established to ensure early detectinn o red.

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instabilities.

MAL peak to peak of RATED THERMAL POWER were rmal-hydraulic during its first three operating cycles and at different powe /fl eactor conditions.

stable operations of the Grand Gulf Reactor.This represents the i

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

he required operator action of an immedia regional oscillations.

shutdown upon entry to Region A and upon detection of sustain i

flux noise level greater than the 105 peak-to peak of RATED THEMA on assures that an adequate margin to thermal limits will be maintain j

L POWER

times, 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 i

to the recirculation pump and recirculation nozzles ermal shock bottom of the vessel is at a lower temperature than the coolant iS e

regions of the core, undue stress on the vessel would result if th j

n the upper difference was greater than 100*F.

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

to power or flow increases from this condition.These differen e vessel-is not recirculation loop drive flowsThe recireviation flow control valves which in turn ual and recirculation pump speed. coolant through the reacter c will vary the flow rate of The rectreulation flow control system consists of the electronic and hydraulic components necessary for%e+pos n

will'generatethe two hydraulically actuated flow contret valves i

one of. severs)a flee control valve " motion inhibit" signal in respo logic j l

g of i

The " notion inhibit" signal causes hydraulic powe i

isolation such that the flow control valve fails "as is "

insures that the flow control valves do not respond to potentia raulic control signals, This design feature i

us inchvikl GRAND GULF-UNIT 1 8 3/4 4-14 Ameneent No. 62

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