Text

Proposed Tech Specs Re Reactivity Control Sys & Removal of cycle-specified Parameter Limits
	ML20248H313
Person / Time
Site:	Hatch
Issue date:	10/04/1989
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20248H309	List: ML20248H306; ML20248H313;
References
	NUDOCS 8910110263
	Download: ML20248H313 (21)
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ENCLOSURE 1

' PLANT. HATCH -- UNIT.1 NRC DOCKET:50-321 OPERATING ~ LICENSE DPR-57 REOUEST TO REVISE. TECHNICAL SPECIFICATIONS-REACTIVITY CONTROL SYSTEMS AND-

' REMOVAL OF CYCLE-SPECIFIED PARAMETER LIMITS.

p;s i

PAGE CHANGE INSTRUCTIONS 4

Removed Page Insert Page i

1~.0-7 1.0 <

3.11-2 3.11-2 3.11-4a-

'3.11-4a 6-15d 6-15d (Note 2)-

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Minimum Critical Power Ratio (MCPR) - Minimum Critical Power Ratio

. (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of.that power in a fuel assembly, which is calculated to cause.some point in the assembly to experience boiling transition, to.

the' actual assembly operation power.

NN. Trio System.- A trip system means an arrangement of instrument channel-

. trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of-protective action may require the tripping of a single trip s'ystem or the coincident tripping of two trip systems.

00.

(Deleted)

PP. Core Operatina Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11.

Plant operation within these operating limits is addressed in individual specifications.

0Q.. Channel Calibration - A Channel Calibration is the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Functional Test. The Channel Calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

RR.

Channel Functional Test - A Channel Functional Test shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as. practicable to verify operability including alann and/or trip functions.

b.

Bistable Channels - the injection of a simulated signal into the channel sensor to verify operability including alarm and/or trip i

functions.

SS. Fraction of Limitina Power Density (FLPD) - the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for the bundle type.

TT. Core Maximum Fraction of Limitina Power Density (CMFLPD) - the CMFLPD l

is the highest value existing in the core of the FLPD.

HATCH - UNIT 1 1.0-7 Proposed 15/02994/229-133 L

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LIM 111NG COND1110NS FOR OPERATION

$ SURVEILLANCE REQUIREMENTS 3.11. B. Linear Heat Generation Rate (LHGR)'

l

'(Continued)

L:

f operation it is determined by normal 1

surveillance that the limiting value j

for LHGR is being exceeded, action

.Shall be initiated within 15 minutes to restore operation to.within the prescribed limits. If the LHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,

'then reduce reactor power to less than 25% of rated thermal l

power within,the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the limiting condition for operation is restored prior to

. expiration of the specified time interval, then further progression to less than 25% of rated thermal l

power is not required.

C.

Minimum Critical tower Ratio (MCPR) 4.11 C.1.

Minimum Critical Power Ratio (MCPR1 The minimum critical power ratio MCPR shall be determined to be (MCPR) shall be equal to or greater equal to or greater than the than the operating limit MCPR applicable limit, daily during (OLMCPR) provided in the CORE reactor power operation at t 25%

l OPERATING LIM 115 REPORT.

rated thermal power and following any change'in power level or If at any time during operation it distribution that would cause is determined by normal surveillance operation with a limiting control that the limiting value for MCPR rod pattern as described in the is being exceeded, action shall be bases for Specification 3.3.F.

initiated within 15 minutes to restore operation to within the prescribed limits. If the steady 4.11.C.2.

Minimum Critical Power Ratio limit

. state MCPR is not returned to within the prescribed limits within two (2)

The MCPR limit at rated flow and hours, then reduce reactor power to rated power shall be determined, less than 25% of rated thermal power

as provided in the CORE OPERAT-within the next four(4) hours. If ING LIMITS REPORT, using:

the Limiting Condition for Operation is restored prior to expiration of a.

t=1.0 prior to initial scram the specified time interval, then time measurements for the further progression to less than cycle, performed in accordance 25% of rated thermdl power is not with Specification 4.3.C.2.a.

required, or b.

t is determined f rom scram time measurements performed in accordance with Specifica-tion 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.3.C.2.

HATCH - UNIT 1 3.11 -?

Proposed TS/0300g/229-157 E-_____m________

.P,,'

g, BASE 5 FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

'4' a

3.11.C. ' Minimum Critical Power Ratio (MCPR) (Continued).

. According to the CORE OPERATING LIMITS REPORT the '1005 power,1005

}

. flow operating limit MCPR (OLMCPR) depends on the average scram time, e, of the control rods, where:

t = 0 or ' ave

B, whichever is greater

'A ~'B 3.,

i

.where:

'A = 1.096 sec (Specification 3.3.C.2.a. scram time limit to notch 36)

'B = y + 1.65 "1

I#2

[ Reference 7) a nINi

i. =1 0.B22 sec (mean scram time used in the transient where: y =

analysis)

o. =.018 sec (standard deviation.of y) n INvI

ave =

i=1 n

IN

. i=1 wherei n'= number of surveillance tests performed to date in the cycle Ni = number of active control rods measured in the ith surveillance test ti - average scram time to notch 36 of all rods in the ith survelliance test N = total number of active rods measureo in 4.3.C.2.a 3

I HATCH - UNIT 1 3.ll-4a Proposed TS/03000/229-157

_-_L__--_-___-

g7.

g, ADMINISTRATIVE CONTROLS e.

Type of container, e.g., LSA, type A, type B, large quantity.

ll f.

Solidification agent, e.g., cement.

The Radioactive Effluent Release Report shall include (on a quarterly basis)

L unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquia effluents that were in excess of 1 C1, excluding dissolved and entrained gases and tritium for liquid effluents, or those in excess of 150 Ci of noble gases or 0.02 Ci of radiciodines for gaseous releases.

The Radioactive Effluent Release Report shall include any changes to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MAkJAL made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10.

Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Of fice cf Management and Program Analysis, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the Regional Office of Inspection and Enforcement no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.11.a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cyc.le or any remaining part of a reload cycle for the following:

(1) Operation with a Limiting Control Rod Pattern (for Rod l

Withdrawal Error,-RWE) for Specification 3.3 F.

(2) The Average Planar Linear Heat Generation Rate (APLHGR) for Specification 3.11. A.

(3) The Linear Heat Generation Rate (LHGR) for Specification 3.11.8, and (4) The Mini.num Critical Power Ratio (MCPR) for Specifications 3.3.F and 3.11.C and Surveillance Requirement 4.ll.C.

b,. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents.

(1) NEDE-240ll-P-A, ' General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

(2) " Safety Evaluation by the Of fice of Nuclear Reactor Regulation Supporting Amendment No.157 to Facility Operating License DPR-57," dated September 12, 1988.

c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) i of the safety analysis are met, j

i d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon j

issuance, for each reload cycle. to the NRC Document Control Desk with copies to the Regional Administrator and Resident l

Inspector.

HATCH - UNIT 1 6-15d Proposed TS/03010/229-149 i

L ll y

1;.

ENCLOSURE 1 PLANT HATCH - UNIT 2 l

NRC DOCKET 50-366 OPERATING LICENSE NPF-5 RE00EST TO REVISE TECHNICAL SPECIFICATION REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS l

PAGE CHANGE INSTRUCTIONS Removed Pagg Insert Paga p

3/4 2-7 3/4 2-7 l

B 3/4 2-4 B 3/4 2-4 6-14d 6-14d i

l l

7 l

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Page 1 of 1

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' ' ~

.1 3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED)

ACTION:

With MCPR less than the applicable operating limit provided in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

_ SURVEILLANCE REQUIREMENTS 4.2.3.1 The MCPR operating limits shall be determined, as provided in the CORE OPERATING LIMITS REPORT. using:

a. 't = 1.0 prior to the initial scram time measurements for the cycle performed in accordance with Specification 4.1.3.2.a, or b.

t is determined from scram time measurements performed in accordance with Specification 4.1.3.2.

The determination of-the limit must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of.each -scram time surveillance test required by Specification 4.1.3.2.

4.2.3.2 All-MCPRs shall be determined to be equal to or greater than the applicable limits:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Whenever THERMAL POWER has been increased by at least 15% of i

RATED THERMAL POWER and steady state operating conditions have been established, and c.

Initially and at least once per 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months when the reactor is

. operating with a LIMITING CONTROL ROD PATTERN for MCPR.

l HATCH - UNIT 2 3/4 2-7 Proposed TS/0302q/229-89

i7 m g

7,.

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

As depicted in the CORE OPERATING LIMITS REPORT the.100% power, 100% flow l

operating limit MCPR (OLMCPR) depends 'on the-average scram time, t', of the

' control. rods, where:

t = 0 or

ave TB, whichever is greater

'tATB where:

A = 1.096 sec-(Specification 3.1.3.3, scram time limit to notch 36)

B = y + 1'.65 N

1/2, 3

n E N, i=1

where:

'u = 0.822 sec (mean scram time used in the transient analysis) o = 018 sec (standard deviation of y) n E N t,

ave =

i=1 n

E IN

.iel g I

where:

n = number of surveillance tests performed to date in the cycle "i = number of active control rods measured in the ith surveillance test Ti = average scram time to notch 36 of all rods in the ith surveillance test N1 = total number of active rods measured in 4.1.3.2.a.

HATCH - UNIT'2' B 3/4 2-4 Proposed TS/0304 /229-89 4

L___-__

L.

ADMANISTRATfvE CONTROLS I

Type of container, e.g., LSA, type A, type B, large quantity e.

f.

Solidification agent, e.g., cement.

The Radioactive Effluent Release Report shall include (on a guarterly basis)

{

unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents that were in excess of 1 C1, excluding dissolved and entrained gases and tritium for liquid effluents, or those in excess of 150 C1 of noble gases or 0.02 C1 of radiciodines for gaseous releases.

The Radioactive Effluent Release Report shall include any changes to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL made during the reporting period.

4 MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the Regional Office of Inspection and Enforcement no later than the 15th of.each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.11.a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

(1) Control Rod Program Controls - Rod Block Monitor for Specification 3.1.4.3, (2) The Average Planar Linear Heat Generation Rate for Specification 3.2.1 and Surveillance Requirement 4.2.1, (3) The Minimum Critical Power Ratio for Specifications 3.1.4.3 and 3.2.3 and Surveillance Requirement 4.2.3, and (4) The Linear Heat Generation Rate for Specification 3.2.4 and Surveillance Requirement 4.2.4.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents.

(1) NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

(2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment Nos. 151 and 89 to Facility Operating Licenses DPR-57 and NPF-5 " dated January 22, 1988.

c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle to the NRC Occument Control Desk with copies to the Regional Administrator and Resident Inspector.

HATCH - UNIT 2 6-14d Proposed TS/03034/229-86 1

?'

DRAFT ENCLOSURE 2 GEORGIA POWER COMPANY

~

EDWIN 1. HATCH HUCLEAR PLANT UNIT 2 FUEL CYCLE 8 (RELOAD 7)

CORE OPERATING LIMITS REPORT REVISION O I

(date)

SOUTHERN COMPANY SERVICES, INC.

P. O. BOX 1295 BIRMINGHAM, AL 35201 I

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o

' ~

1.

INTRODUCTION

' This CORE OPERATING LIMITS REPORT for Hatch Unit 2 Cycle 8 is prepared in accordance with the requirements of Hatch Technical Specification 6.9.1.11.

- The core operating limits p7asented here were developed using NRC-approved methods (References 1 and 2). Results from the reload analyses for the

~ General Electric fuel in Hatch Unit-2 Cycle 8 are documented in Reference 3.

i The following cycle-specific core operating limits are included in this

report, a.

Control Rod Program Controls - Rod Block Monitor

.)

(Technical Specification 3/4.1.4.3)

)

J

b.. Average' Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specification 3/4.2.1) c.

Minimum Critical Power Ratio -(MCPR) Operating Limit (Technical Specification 3/4.2.3) d.

Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3/4.2.4) 2.

ROD; BLOCK'HONITOR (TECHNICAL SPECIFICATION 3/4.1.4.31

)

Both Rod Block Monitor (RBM) channels shall be operable as specified in 1

Technical Specification 3/4.1.4.3 and when.

a.

THERMAL POWER is < 90% of RATED THERMAL POWER and the MCPR is less l

than 1.70, or b.

THERMAL POWER is 1 90% of RATED THERMAL POWER and the MCPR is less than 1.40.

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3..

APLHGR LIMIT (TECHNICAL SPECIFICATION 3/4.2.1)

~

The APLHGR limit is given by the applicable rated-power, rated-flow limit taken from Figures 3-3 through 3-5, multiplied by the smaller of either:

u a.

The MAPFACF factor given by Figure 3-1, or

b..The MAPFACp factor given by Figure 3-2.

For the: fuel types whose APLHGR limits are shewn. in Figures. 3-3 through 3-5, the APLHGR limit shown shall be applied to each axial location in the fuel assembly.

4.

MCPR OPERATING LIMIT (TECHNICAL SPECIFICATION 3/4.2.3)

'.The MCPR' operating-limit (0LMCPR) is a function of fuel design, average scram time,. core flow, number of operating recirculation loops, and core power.

4.1 Two Recirculation loop Operation For 25% 1 Power < 30%, the OLMCPR is given in Figure 4-1.

For Power ;t 30%,

I the OLMCPR is the greater of either:

a.

The applicable limit determined from Figure 4-2, or t

b.

The appropriate Kp given by Figure 4-1, multiplied by the appropriate limit from Figure 4-3 where T is determined in

.accordance with the Technical Specifications Bases section.

j 4.2 Sinale Recirculation loop Operation

)

1 For single-loop operation, the MCPR operating limit shall be 0.01 greater than the two-loop value which is determined as specified in Section 4.1. __

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5..

LHGR LIMIT (TECHNICAL SPECIFICATION 3/4.2.41 l

The LHGR limit shall be 13.4 kw/ft for all 8x8 fuel. The LHGR limit for-the ANF.9x9 LFAs is conservatively accounted for in the exposure-dependent f

APLHGR limits for that fuel type; therefore, there is no explicit LHGR limit.for 9x9 fue1 ~.

6.

REFERENCES 1.

" General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A-9 (September 1988).

2.

Letter, L. P. Crocker (NRC) to J. P. O'Reilly (GPC), "Issuan'ce of Amendment Nos. 151 and 89 to Facility Operating Licenses DPR-57 and NPF Edwin 1. Hatch Nuclear Plant Units 1 and 2 (TACS 66524/66525)," January 22, 1988.

3.

" Supplemental Reload Licensing Submittal fc* Edwin I. Hatch.

E' Nuclear Plant Unit 2, Reload 7, Cycle 8,' GE Document 23A5884 (January 1988).

4.

"Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31376-P, December 1986, as amended.

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