ML20100Q486

Amend 93 to License NPF-29, Deleting Refs to Operation of Reactor Recirulation Sys in non-loop Manual Mode of Flow Control

ML20100Q486
Person / Time
Site:	Grand Gulf
Issue date:	03/09/1992
From:	Larkins J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20100Q488	List: ML20100Q486 ML20100Q493
References
NUDOCS 9203170136
Download: ML20100Q486 (14)

Text

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ENTERGY OPERATIONS. INC.

SYSTEM ENERGY RESOURCES. INC2 SOUTH MISS!SS1PPI ELECTRIC POWER ASSOCIATION MISSISSIPP1 POWER AND LIGHT COMPANY DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION. UNIT 1

.AE NDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No. NPF-29 1.

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

A.

The applicatson for amendment by Entergy Operations, Inc. (the licensee) dated May 30, 1991, 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 assurance (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; 4

D.

The issuance of this amendment will not be inimical to the common defense and security 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.

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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. NPF-29 is hereby amended to read as follows:

(2) IgI.hnical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 93, are hereby incorporated into this license.

Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance, FOR TITE NUCLEAR REGULATORY COMMISSION John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects - III/lV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 9, 1992 w

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ATTACHMENT T0 tlCENSE AMQiQt ENT-NO. 93 '

f FACIllTY OPERATING LICENSE NO. NPF-22

DOCKET NO. 50-415 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by amendment number and

- contain vertical lines indicating the area of change.

REMOVE PAGES-INSERT PAGES 3/4 2-5 3/4 2-5 3/4 2-7b 3/4 2-7b

-3/4 4-1 3/4 4-1 3/4 4-la 3/4 4-la B 3/4 2-4 B 3/4 2-4 B 3/4 2-4a B 3/4 2-4a

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B 3-4 2-6 B 3/4 2 B 3-4 2-7 B 3/4 2-7a B 3/4 2-7a B 3/4 4-1 B 3/4 4-1 B 3/4 4-la B 3/4 4-la l-I

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I 3/4.4 REACTOR COOLANT S_Y, STEM

-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:

a.

Two rceirculation 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.

Limits and setpoints per Specifications 2.2.1, 3.2.1, and 3.3.6.

Operation is not permissible in Regions A, B or C as specified in Figure 3.4.1.1-1 except that operation in Region C is permissible during control rod 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 sustailed 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.

d.

With operation in Region B as specified in Figure 3.4.1.1-1, immediately initiate action to either reduce THERMAL POWER by itserting control rods or increase core flow if one or more recirculation pumps are on fast speed by opening the flow control valve to within 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 startop, immediately initiate action to either reduce THERMAL POWER or increase core flow to within 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 />.

f.

During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action to reduce 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, 93

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. LIMITING CONDITION FOR 0PERATION-(Continued) g.

During single loop operation, with temperature differences exceeding theLlimits of SURVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the.

THERMAL POWER-or recirculation Lloop flow increase.

ii.

With a changelin reactor operating' conditions,jfrom two recircula-l tion. loops _operatingLto' single loop operation,Jor restoration of-two loop _ operation,- the -limits and setpoints of Specifications 2.2.1.-

13.2.1, 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 /> or declare the associated equipment inoperable (or the limits to be "not satisfied"),

u and-take the ACTIONS' required by theLreferenced specifications.

SURVEILLANCE REQUIREMENTS

.4.4.1.1.11 At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor. coolant' recirculation system shall be verified to be -in operation and.not in Regions A, B or C as specified zin~ Figure'3.4.1.1-1 except that operation in Region C.is permissible during

' control rod = withdrawals for startup.

L4.4.1.1.2 EEach reactor 1 coolant system recirculation loop flow control valve in an operating. loop'shall be demonstrated OPERABLE at least once per 18 months--

by.

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. Verifying that the control valve fails "as -is" on loss of. hydraulic-

. a.

pressure at

the hydraulic unit, and-b.

Verifying that the average, rate of control valve movement ist 1

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

2.

Less than or equal to IDE of stroke.per second closing.

- 1

.4.4.1.1.3 During single loop operation, verify. that_ the volumetric. loop! flow -

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1 rate of1the loop in operation.is within the limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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i GRAND GULF-UNIT 1 3/4 4-la Amendment No. 73,93

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. POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel clad-ding integrity Safety Limit MCFR, and an analysis of abnormal operational tran-sients.

For any abnormal operating transient analysis evaluation with the initial condition of the redctor being at the steady state operating limit, it is required that the resulting MCPR does not decrease pelow the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal 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 transients evaluai.ed were loss of flow, increase in pressure and power, positive reactivity-insertion, and coolant temperature decrease.

The limiting transient yields the largest delta CPR.

When added to the Safety Limit MCPR, the required operating limit MCPR of Specification 3.2.3 is obtained.

The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits (References 2 and 3).

MCPR operating limits are defined as functions of exposure (MCPR )' IIO*

e (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 limits for all anticipated exposures during the Cycle. The MCPR, limits are established for a set of exposure intervals.

The limiting transients are analyzed at the limiting expo-sure for each interval.

The MCPR operating limits are established based on the largest delta-CPR e

calculated at the limiting exposure and ensure that the MCPR safety limit will not be exceeded during the most limiting transient in each of the exposure intervals.

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

p than rated core flow and power conditions for all exposures during the cycle.

The MCPR s are established to protect the core from inadvertent core flow f

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

The ref-erence core flow increase event used to establish the MCPR is a hypothesized f

slow flow runout to maximum, that does not result in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 item 2).

The result of a single failure or single operator error during Loop Manual I

operation is the runout of one loop because the two recirculation loops are under independent control. With this basis, the MCPR curve was generated from l

f GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 73, 93 1

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

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

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

Assumptions used in the original calculations of this generic flow control line were con-sistent with a slow flow increase transient duration of several minutes:

(a) the plant heat balance was assumed to be in equilibrium, and (b) core xenon concentration GRAND GULF-UNIT I B 3/4 2-4a Amendment No. 72,93

POWER'0ISTRIBUTION LIMITS BASES:

MINIMUM CRITICAL POWER RATIO (Continuedi

-was assumed to be constant.-- The generic flow control line is used to define-

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

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single failure / single operator. error. criterion, one loop: runout:was postulated _

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. The Loop: Manual: mode of operation was analyzed.

Consistent with the for Loop Manual operation.: The maximum core flow at loop runout was assumed to 1

_be 110% of rated flow.. Peaking-factors were selected such that the MCPR for the bundle with the least margin of sa.fety would not decrease below the MCPR Safety-Limit.

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

than core-flow increase including the localized. rod withdrawal-error event.

Core power dependent.setpoints are incorporated (incremental control rod with-drawal limits) in-the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).

These setpoints-allow greater control-rod withdrawal at lower core powers where

' core thermal margins are large.

However, the increased rod withdrawal requires higher' initial MCPR's to assure-the MCPR safety limit Specification (2.1.2) is not violated.

The analyses thatiestablish the power dependent MCPR require-ments that' support the-RWL system are presented in Reference 4.

For core power below 40% of RATED THERMAL POWER, where'the EOC-RPT and the reactor scrams on

-. turbine stop valve. closure and turbine control-valve fast closure are bypassed,

- separate sets'of MCPR limits..are provided_for-high'and low core flows =to ac-countforthesignifiBantsensitivitytoinitialcoreflows.

For core power

above 40% of RATED THERMAL POWER,_ bounding power-dependent-MCPR limits were-developed.- The abnormal operating transients analyzed for single loop operation are discussed in Reference 5 and the appropriate cycle-specific documents.

No

--change-to thelMCPR operating-limit is required for single loop operation.-

At THERMAL POWER levels less than or equal to-25% of RATED THERMAL POWER,

the reactor will be operating'at minimum recirculation pump speed and the modera-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

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

GRAND GULF-UNIT 1 B 3/4 2-6 Amendment No. 73,93 v

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POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

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

The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level 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 charges.

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 completion af a THERMAL POWER increase of et 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 MCPR 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 is less than the design linear heat generation even if fuel pellet densification is postulated.

The LHGR limits of Figure 3.2.4-1 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 A curve is provided based on the maximum credible flow runout transient for Loop Manual operation.

The result of a single failure or single operator error during operation in Loop Manual is the runout of only one loop because both recirculation loops are under independent control.

LHGRFAC 's are I

p generated to 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 centrol 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 R0D 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 Amendment No. 73,93

POWER DISTRIBUTION LIMITS BASES

References:

1.

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

EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982.

2.

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

March 1986.

3.

AECM-86/0066, " Final Summary Startup Test Report 12," Letter, 0.D.

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

4.

XN-NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Analysis; MCPRo for All Plant Operations Within the Extended Operation Domain,"

Exxch Nuclear Company, October 1986.

5.

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

GRAND GULF-UNIT 1 8 3/4 2-7a Amendment No. 73,93

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3/414:-REACTOR COOLANT SYSTEM BASES 3/4.4.1' RECIRCULATION SYSTEM

[

Operation with one reactor core coclant recirculation loop' inoperable has

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been _ evaluated and found to remain within design limits and safety margins pro-vided certain limits and setpoints are modified. The "GGNS Single Loop Opera--

tion Analysis" identified the applicable fuel thermal limits'and APRM setpoint modifications necessary to maintain the same margin of safety for single loop

-operation as.is available during.two loop operation.. Additionally, loop flow limitations are established to ensure vessel internal vibration remains within limits.

l 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-basis-accident, increase the blowdown area and reduce the capability of.reflooding the core;-thus; the requirement for_ shutdown of the facility with a jet pump inoperable.

Jet ' pump failure ~can be detected by mon:toring jet pump per-formance_ on~ a prescribed schedule for significant degradation.

During two loop operation,Jrecirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria.

The limits will ensure-an adequate core flow coastdown from either recirculation loop following a LOCA.

In cases where the misratch limits cannot be maintained, continued operation is per--

-mitted:with one loop-in operation.

The power / flow operating map is divide'd into four (4)' regions.

Regions A and B are= restricted from operations. They include the operating area above the 80% rod-line'and below 40%' core flow.

Region C includes the operating area above the 80% rod-line and between 40% and 45% core flow.

Operation in Region C :is allowed only.for control rod _ withdrawals during startup' for required fuel; preconditioning.

Region D consists of the rest of the operating 3

zmap.

No core thermal-hydraulic stability related restrictions are applied to Region'0 since the potential onset._of core thermal-hydraulic instabilities is not predicted within Region D.

7 The: definition of Regions A, B and-C is based on BWR stability operational

' data and. required operator actions.

Although'a large margin to onset of insta-bility was observed in Regions ~Ac B-and C during GGNS stability tests for typical Loperating configuration, a conservative approach is adopted in the specification.

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 durir.g recirculation pump speed. transfers.

Upon entry to Region A an immediate reactor shutdown is required.

Upon-entry to Region B or Region

~

C, unless; operation _in Region C is for_ control rod withdrawals during startup,

'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 n

. GRAND GULF-UNIT 1 B 3/4 4-1 Amendment-No. 73, 93

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_ REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)

APRM neutron fiex noise icvel exceeding 10% peak-to peak of RATED THERMAL POV" is observed, an immediate reactor shutdun is required.

The APRM neutron flux noise level of 10% peak-to-peak of RATED THERMAL j

POWER is established to ensure early detection of core thermal-hgdraulic instabilities.

APRM neutron flux noisa levels in the range of 2. to 6%

peak to peak of RATED 'iHERMAL POWER were observed for the Grand Culf Reactor during its first three operating cycles and at different power / flow operating conditions.

This represents the typical APRH neutron flux noise level for stable operations of the Grand Gulf Reactor.

The 10% peak-to scak of RATED THERMAL POWER noise level provides adequate margin to thermal l'aits in the unlikely event of uncontrolled limit cycle osc111ations while in Regions B and C, including the even less likely event of regional oscill?.tions.

The required operator action of an immediate reactor shutdewn upon entry to Region A and upon detection of sustained APRM neutron flux nois6 level greater than the 10% peak-to peak of RATED THERNAL POWER assures that an adequate margin to thermal limits will be maintained at all times.

In order to prevent undue stress on the vessel nozzles and bottom head region, the rer:ireviation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.

The loop temperature must also be within 50'F of the reactor pressure vessel ecolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, unoue stress on the vessel would result if the temperature difference was greater than 100'F.

During single loop operation, tha condi-tion may exist in which the coolant in the bottom head of the vessel is not circulating.

Thtse differential temperature criteria are also to be met prior to power or flow Increases from this condition.

The recirculttion fica control valves provide regulation of individual recirculuion loop drive flows; which, in turn, will vary the flow rate of coolnnt through the reactor core over a range consistent with the rod pattern and recirculation pump speed.

The recirculation flow control system consists of the electronic and hydraulic components necessary for individual positioning l

of the two hydraulically actuated flow control valves.

Solid state control logic will rrnerate a flow control valve " motion inhibit" signal in response to any one of several hydraulic power unit or analog control circuit failure

signale, The " motion inhibit" signal causes tiydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is."

This design feature insures that the flow control valves do not respond to potentially erroneous control signals.

GRAND GULF-UNIT 1 B 3/4 4-la Amendment No.

62.93

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