Text

Amend 106 to License NPF-29,changing TS to Remove cycle- Specific Reactor Physics Parameters & Incorporate Into New Document Called Core Operating Limits Rept
	ML20127N310
Person / Time
Site:	Grand Gulf
Issue date:	01/21/1993
From:	Hubbard G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20127N314	List: ML20127N310; ML20127N317;
References
	NUDOCS 9301290167
	Download: ML20127N310 (25)
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ENTERGY OPGAT10fil. INC.

SYSTEM ENERGY RESOURCES. INC.

ED EH MISSISSIPPI ELECTRIC POWER ASSOCIATION MISSISSIPPI POWER AND LIGHT COMPANY DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION. UNIT 1 A2ENDMENT TO FAClllTV OPERATING LICENSE Amendment No. 106 License No. NPF-29 1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated October 29, 1992, 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; 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.

9303290167 930121 PDR-ADOCK 05000416 P

PDR

i I' 2.

Accordingly, the license is amended by channes 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) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.106, 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 90 days from its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

)) esp "f'

George T. Hubbard, Acting Director Project Directorate IV-1 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 21, 199?

ATTACHMENT TO LICENSE AMENDMENT NO. 106 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 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. The corresponding overleaf pages are also provided to maintain document completeness.

FEMOVE PAGES INSERT PAGES i

i iv iv xx xx l-2 1-2 1-2a 2-1 2-1 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-2 3/4 2-4 3/4 2-3 3/4 2-5 3/4 2-6 3/4 2-6a 3/4 2-7 3/4 2-4 3/4 2-7a 3/4 2-7b 3/4 2-7c B 3/4 2-1 8 3/4 2-1 B 3/4 2-7 8 3/4 2-7 5-5 5-5 6-19 6-19 6-19a-6-19b l

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_ _ _. - -, _ _ _ _ _,. -.. _...,._ ___,. ~,. _... _.-

I

', u INDEX QEFINITIONS-SECTION 1.0 DEFINITIONS

'PAGE 1.1 ACTION 1-1 1.2 AVERAGE PLANAR EXPOSURE......................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1-1 1.4 CHANNEL CAllBRATION........................ 1 1.5 CHANNEL CHECK..-......................... 1 1.6 CHANNEL FUNCTIONAL TEST 1-1 1.7 CORE ALTERATION...........................

1-2 1.7a CORE OPERATING LIMITS REPORT 1-2' l

1.8 CRITICAL POWER RATIO 1-2 1.9 DOSE EQUIVALENT I-131.......................

1-2 1.10 DRYWELL INTEGRITY......................,,.

1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY..................

1-3 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1-3 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP-SYSTEM RESPONSE' TIME.....

1 1.14 FRACTION OF LIMITING POWER DENSITY

--l-3 1.15 FRACTION OF RATED THERMAL POWER........-......-..

1-3 1.16 FREQUENCY NOTATION 1-3 1.17 DELETED..............................

1-3 1.18 IDENTIFIED LEAKAGE 1-4.

1.19 ISOLATION SYSTEM RESPONSE TIME.

1-4 1.20 LIMITING CONTROL R0D PATTERN 1 1.21 LINEAR HEAT GENERATION RATE....................

1 1.22 LOGIC SYSTEM FORCTIONAt TEST

~

1-4 1.23 MAXIMUM FRACTION OF LIMITING. POWER DENSITY 1-4 1.24 HEMBER(S) 0F THE PUBLIC...............,......- 1-4

-1.25 MINIMUM CRITICAL POVER RATIO 1-5 1.26 0FFSITE DOSE CALCULATION MANDAL (ODCM)

_l-5 1.27-OPERABLE - OPERABILITY 1-5

-GRAND. GULF-UNIT:1,- - - -Amendment No.= W,1106---

]

J

INDEX DEFINITI^NS SECTION DEF]NITIONS (Continued)

PAGE

1. 28 OPE RATIONAL CONDITION - CONDITION....................................

1-5 1.29 PHYSICS TEST5........................................................

1-5 1.30 PRESSURE BOUNDARY LEAKAGE............................................

1-5 1.31 PRIMARY CONTAINMENT INTEGRITY........................................

1-6 1.32 PROCESS CONTROL PROGRAM (PCP)........................................

1-6 1.33 PURGE -

PURGING......................................................

1-6 1.34 RATED THERMAL P0WER...................................

1-6 1.35 REACTOR PROTECTION SYSTEM RESPONSE TIME..............................

1-7 1.36 REPORTABLE EVENT.....................................................

1-7 1.37 ROD DEN51TY..........................................................

1-7 1.38 SECONDARY CONTAINMENT INTEGRITY......................................

1-7 1.39 SHUTDOWN MARGIN......................................................

1-8 1.40 SITE B0VNDARY....................................

18 1.41 DELETED..............................................................

1-0 1.42 DELETED..............................................................

1-8 1.43 STAGGERED TEST BASI 5.................................................

1-8 1.44 THERMAL P0WER........................................................

1-8 3.45 UNIDENTIFIED LEAKAGE.................................................

1-8 1.46 UNRESTRICTED AREA..........................

1-8 1.47 DELETED..............................................................

1-8 1.48 VENTING..............................................................

1-9 TABLE 1.1, SURVEILLANCE FREQUENCY N0TATION................................

1-10 TABLE 1.2, OPERATIONAL CONDIT10NS.........................................

1-11 GRAND GULF-UNIT 1 11 Amendment No. 87

.INDEX 4

(IMITING CONDITIONS FOR OPERATION AND SVRVElllANCE RE0VIREMENTS SECTION EAGI 3/4.0 APPLICABillTY 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.....................

3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod 3/4 1......................

Control Rod Maximum Scram Insertion Times........

3/4 1-6 Control Rod Scram Accumulators 3/4 1-8 Control Rod Drive Coupling 3/4 1-10 Control Rod Position Indication............

3/4 1-12 Control Rod Drive Housing Support...........

3/4 1-14 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control Rod 3/4 1-15 Rod Pattern Control System..............

3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 3/4 2-1 3/4.2.2 DELETED.........................

3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO 3/4 2-3 3/4.2.4 LlhEAR HEAT GENERATION RATE...............

3/4 2-4 i

I l'

l GRAND GULF-UNIT 1 iv Amendment No. 46, 106

..,..,3 A

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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F10w.............................

2-1 THERMAL POWER, High Pressure and High F1ow..........................

2-1 Reactor Coolant System Pressure.....................................

2-1 Reactor Vessel Water Leve1..........................................

2-2

2. 2 LIMITING SAFETY SYSTEM SETT1 HGS Reactor Protection System Instrumentation Setpoints.................

2-3 BASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F1ow.............................

B 2-1 THERMAL POWER, High Pressure and High F10w..........................

B 2-2 Reactor Coolant System Pressure.....................................

B 2-5 Reactor Vessel Water Leve1..........................................

B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.................

B 2-6 GRAND GULF-UNIT 1 iii

INDEX ADMINISTRATIVE CONTROLS SECTION EaQE SAFETY REVIEW COMMITTEE (SRC)

(Continued)

Review 6-10 Audits 6-11 Authority............................

6-12 Records...............

6-12 5.5.3 TECHNICAL REVIEW AND CONTROL Activities 6-12 6.6 REPQBTABLE EVENT _bC1. ION...... -..............

6-13 6.7 S A F ET Y L I KII_yJ OL AT I ON,...................... 6-13 6.8 PROCEDURES AND PROGRAMS 6-14 6.9 REPORTING RE0VIREMENTS Routine Reports.........................

6-15

,ertup Reports.........................

6-15 Anr,ual Reports 6-16 Annual Radiological Environmental Operating Report.......

6-17 Semiannual Radioactive Effluent Release Report.........

6-17 Monthly Operating Reports....................

6-19 Core Operating Limits Report (COLR)...............

6-19 Special Reports.........................

6-19a 6.10 RECORD RETENTIQM 6-19b 5.cl]__BADIAT10N PROTECTION PH0GR68..................

6-21 6.12 H I GH _f (D I AT I ON AR E A........................ 6-21 6.13 PROCLjS PROGRAM CONTROL (PCP)..................

6-22 6,14 0FFSITE DOSE CALCULATION MANUAL (ODCM)

. 6-22 GRAN; GULF-UNIT 1 xx Amendment No. B7,106

..=

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 R E S PO N 51B I L [I(....................................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS..............................

6-1 l 6.2.2 UNIT STAFF............................- 2.....................-

F-1 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function......................................................

6-2 Composition..............,....................................

6-2 Responsibilities..............................................-

6-6 Authority....................................................

6-6 6.2.4 SHIFT TECHNICAL ADVIS0R.......................................

6-6 6, 3 UNIT STAFF QUALIFICATIONS..........................................

6-6 6.4 TRAINING...........................................................

6-6 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

Function......................................................

6-6 Composition...................................................

6-7 Alternates....................................................

6 1 Meeting Frequency.............................................

6-7 Quorum.....................',..................................._

6 Responsibilities..............................................

6-7 Authority............................

6-8 Records.......................................................

6-8_

l 6.5.2 SAFETY REVIEW COMMITTEE (SRC)

]

Function......................................................

6-9 C omp o s i t i o n...................................................

6 Alternates....................................................

'6-9 Consultants...................................................

6 i Meeting Frequency..............................................

5-10 QuoruR........................................................

6-10 1

GRAND GULF-UNIT 1 xix Amendment No. 45 l

l

e l'

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or renetivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of the SRMs IRMs, LPRMs, TIPS, or special movable detectors is not considered to be CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

[0RE OPERATING LIMITS REPORT (COLR) 1.7a The COLR is the Grand Gulf Nuclear Station specific document that provides core operating limits fnr the current reload cycle. These cycle-specific 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.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by applicatien of the ANFB correlation to cause some point in the assembly to experience bulling transition, divided by the actual assembly operating power.

DOSE E0VIVALENT l-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, miciocuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, 1-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of T10-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

l GRAND GULF-UNIT 1 1-2 Amendment No. 35, 73,I02,106 l

,- u OEFINITIONS DRYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when; a.

All drywell penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE drywell automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are opened under administrative controls as permitted by Specification 3.6.4.

b.

The drywell equipment hatch is closed and sealed.

c.

The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.

d.

The drywell leakage rates are within the limits of Specification 3.6.2.2.

e.

The suppression pool is in compliance with the requirements of Specification 3.6.3.1.

f.

The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

GRAND GULF-UNIT 1 1-2a Amendment No.106

a 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

1 With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated

flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

THERMAL POWER. Hiah Pressure and Hiah Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during two loop operation and 1.07 during single loop operation with the l

reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

a.PPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With MCPR less than the above limits and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specifi-cation 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

A_0J1pj:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

GRAND GULF-UNIT 1 2-1 Amendment No. 73, 99, 106 l

~

ILim2 POWER DISTRIBUTION LIMill 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

l APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

During two loop operation or single loop operation, with an APLHGR exceeding the limits, initiate corrective action within 15 minutes and restore APLHGR to I

within the required limits 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 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 RE0VIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required-limits:

a.

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

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at i

least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

l GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 73, 99, 106 t

n POWER DISTRIBUTION LIMITS 3/4.2.3 HINIMUM CRITICAL POWER BillQ LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limits specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With MCPR less than the applicable MCPR limits, initiate corrective action l

within 15 minutes and restore MCPR to within the required limits 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 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 MCPR shall be determined to be equal to or greater than the applicable MCPR limits.

l a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MCPR.

d.

The provisions of Specification 4.0.4-are not applicable.

GRAND GULF-UNIT 1 3/4 2-3 Amendment No. M,106 l

1. ~

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two loop operation are specified in the CORE OPERATING LIMITS REPORT (COLR).

For single-loop operation, a MAPLHGR limit corresponding to the product of the two loop MAPLHGR and a reduction factor specified in the COLR can be l

conservatively used to ensure that the PCT for single loop operation is bounded by the PCT for two loop operation.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No. 73, 99, 106

'; v INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTION 1-1 1.2 AVERAGE PLANAR EXPOSURE,.....................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1............

1.4 CHANNEL CALIBRATION..............._.........

1-1 1.5 CHANNEL CHECK...........................

1-1 1.6 CHANNEL FUNCTIONAL TEST 1-1 1.7 CORE ALTERATION..........................

1-2 1.7a CORE OPERATING LIMITS REPORT 1-2 l

1.8 CRITICAL POWER RATIO 1-2 1.9 DOSE EQUIVALENT I-131.......................

1-2 1.10 DRYWELL INTEGRITY.........................

1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY..................

1-3 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1-3 1.13 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RtSPONSE TIME,....

1-3 1.14 FRACTION OF LIMITING POWER DENSITY 1-3 1.15 FRACTION OF RATED THERMAL POWER,,................

1-3 1.16 FREQUENCY. NOTATION 1-3 1.17 DELETED..............................

1-3 1.18 IDENTIFIED LEAKAGE 1-4 1.19 ISOLATION SYSTEM RESPONSE TIME 1-4 1.2d LIMITING CONTROL R0D PATTERN 1-4 1.21 LINEAR HEAT GENERATION RATE,...................

1-4 l

1.22 LOGIC SYSTEM FUNCTIONAL TEST 1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER DENSITY 1-4 1.24 MEMBER (S) 0F THE PUBLIC...._.................. 4 l-1.25 MINIMUM CRITICAL POWER RATIO 1-5 l-l 1.26 0FFSITE DOSE CALCULATION MANUAL-(ODCM) 1-5 t

1.27 OPERABLE - OPERABILITY 5 GRAND GULF-UNIT 1 i

Amendment No. B7,106 I

a INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

PAGE 1.28 OPERATIONAL CONDITION - CONDITION....................................

1-5 1.29 PHYSICS TESTS........................................................

1-5 1.30 PRESSURE BOUNDARY LEAKAGE............................................

1-5 1.31 PRIMARY CONTAINMENT INTEGRITY........................................

1-6 1.32 PROCESS CONTROL PROGRAM (PCP)........................................

1-6 1.33 PURGE - PURGING......................................................

1-6 1.34 RATED THERMAL P0WER..................................................

1-6 1.35 REACTOR PROTECTION SYSTEM RESPONSE TIME..............................

1-7 1.36 REPORTABLE EVENT.....................................................

1-7 1.37 ROD DENSITY..........................................................

1-7 1.38 SECONDARY CONTAINMENT INTEGRITY......................................

1-7 1.39 SHUT 00VN MARGIN......................................................

1-8 1.40 SITE B0VNDARY........................................................

1-8 1.41 DELETED..............................................................

3-8 1.42 DELETED.............................................................

1-P 1.43 STAGGERED TEST BASIS.................................................

1-8 1.44 T!iERMAL P0WER........................................................

1-8 1.45 UNIDENTIFIED LEAKAGE.................................................

1-8 1.46 UNRESTRICTED AREA....................................................

1-8 1.47 0ELETED..............................................................

1-8 1.48 VENTING..............................................................

1-9 TABLE 1.1, SURVEILLANCE FREQUENCY N0TATION................................

1-10 TABLE 1.2, OPERATIONAL CONDITIONS.........................................

1-11 GRAND GULF-UNIT 1 11 Amendment No. 87

e i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION EASE 114.0 APPLICABILITY 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.....................

3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod 3/4 1-3 Control Rod Maximum Scram Insertion Times........

3/4 1-6 s

Control Rod Scram Accumulators 3/4 1-8 Control Rod Drive Coupling 3/4 1-10 Control Rod Poaition Indication............

3/4 1-12 Control Rod Drive Housing Support............ 3/4 1-14 3/4.1.4 CONTROL R00 PROGRAM CONTROLS Control Rod 3/4 1-15 Rod Pattern Control System..............

3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR liEAT GENERATION RATE 3/4 2-1 3/4.2.2 DELETED.........................

3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE...............

3/4 2-4 l

I GRAND GULF-UNIT 1 iv Amendment No. 46, 106

u INDEX ADMINISTRATIVE CONTROLS IIIIl0B EASE SAFETY REVIEW COMMITTEE (SRC)

(Continued)

Review.............................

6-10 Audits.............................

6-11 Authority............................

6-12 Records.............................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities 6-12 6.6 REPORTABLE EVENT ACTION.....................

6-13 6.7 SAFETY L IMIT VIOL ATION....................... 6-13 6.8 PROCEDVRES AND PROGRAMS 6 6.9 REPORTING RE0VIREMENTS Routine Reports.........................

6-15 Startup Reports..........,..............

6-15 Annual Reports 6-16 Annual Radiological Environmental Operating Report 6-17 Semiannual Radioactive Effluent Release Report 6 Monthly Operating Reports....................

6-19 Core Operating Limits Report (COLR)...............

6-19 Speci al Reports..............

6-19a 6.10 RECORD RETENTION 6-19b l

6_.11 RADIATION PROTECTION PROGRAM...................

6-21 l

6.12 HIGH RADIATION AREA.......................

6-21 l

l 6.13 PROCESS PROGRAM CONTROL (PCP)..................

6-22 l

6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)

.._.._.......... 6 i l

l GRAND GULF-UNIT 1 xx Amendment No. B7, 106 l

N A

R INDEX ADMINISTRATIVE CONTROLS I

1 SECTION t

PAGE i

6.1 RESPONSIBILITY.....................................................

6-1 6.2 ORGANIZATION i

1 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS..............................

6-1 l 6.2.2 UNIT STAFF....................................................

6-1 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function......................................................

6-2 Composition...................................................

6-2 Responsibilities..............................................

6-6 Authority.....................................................

6-6 6.2.4 SHIFT TECHNICAL ADVIS0R.......................................

6-6 6.3 UNIT STAFF QUALIFICATIONS..........................................

6-6 6.4 TRAINING...........................................................

6-6 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

Function......................................................

6-6 Composition...................................................

6-7 Alternates....................................................

6-7 Meeting Frequency.............................................

6-7 Quorum.....................',...................................

6-7 Responsibilities..............................................

6-7 Authority............................

6-8 l

l R e c o rd s.......................................................

6-8 l

6.5.2 SAFETY REVIEW C01911TTEE (SRC)

Function......................................................

6-9 Composition...................................................

6-9 l

Alternates....................................................

6-9 Consultants...................................................

6-10 Heeting Frequency.............................................

6-10 Quorum........................................................

6-10 GRAND GULF-UNIT 1 xix Amendment No. 45 l

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs LPRMs, TIPS, or special movable detectors is not considered to be CORE ALTERATION. Suspar. ion of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT (COLR1 1.7a The COLR is the Grand Gulf Nuclear Station specific document that provides core operating limits for the current reload cycle.

These cycle-specific 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.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the ANFB correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE E0VIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

GRAND GULF-UNIT 1 1-2 Amendment No. 35, 73,I02,106 i

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p a

y f' >

2.0: DEFINITIONS The following terms are defined so that unifere interpretation of these specifications may be achieved. The defined tems appear in capitalized type and shall be applicable throughout these Technical Specifications.-

r ACTION 1.1 ACTION shall be that part of a Specification dich prescribes remedial.

asasures required under.sesignated conditions.

AVERAGE PLANAR EXPOSURE' 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the -

specified bundle at'the specified height divided by the neber of fuel fods in the fuel bundle.

. AVERAGE PLANAR LINEAR NEAT-GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR NEAT GENERATION RATE (APUCR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR NEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the neber of fuel rods in the fuel bundle.

CHANNEL CALIBRATION-1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary,!ef 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 alars and/or trip functions,:and'shall include the CHANNEL FUNCTIONAL TEST. The' CHANNEL CALIBRATION may be performed by any series of sequential, everlapping er total channel steps such that the entire channel is calibrated.

CHANNEL CHECK.

1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where? -

possible, comparison of the channel indication and/or stat;us with other indi-cations and/or. status derived frem independent 1hstrument channels esasuring:

the same parameter.

CHANNEL FUNCTIONAL TEST

-1.6 A CHANNEL FUNCTIONAL TEST shall be:

Analog channels - the injection of a simulated signal into then a.

channel as close to the sensor as practicable to verify OPERABILITY.

including,alars and/or_ trip functions and channel failure trips.

b.-

tutable channels - the inlection of a simulated signal into the sensor to verify OPERA 81LITY including alors and/or trip' functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the antire channel is tested.

GRAND GULF-UNIT 1 1-1

f

,. -,u DEFINITIONS

['

DRYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when:

a.

All drywell penetrations required to be closed during accident-conditions are either:

1.

Capable of being closed by an OPERABLE drywell automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are opened under administrative controls as permitted by Specification 3.6.4.

b.

The drywell equipment hatch is closed and sealed.

c.

The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.

d.

The drywell leakage rates are within the limits of Specification 3.6.2.2.

The suppression pool is in compliance with the requirements of e.

Specification 3.6.3.1.

f.

The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

GRAND GULF-UNIT 1 1-2a Amendment No.106

r

--,, 3 2. 0' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS' i

2.1' SAFETY LIM'ITS THFRMAL POWER. Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% 'of RATED. THERMAL POWER with the s

reactor.

vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER :nd the reactor vessel steam dome pressure less than 785 psig or core-flow less than 10% of rated" fl ow,-

be in at least HOT.SHU130WN within 2 hourr and comply with the requirements'of Specification 6.7.1.

IUERMAL POWER. Hiah Press ire and Hich Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during two loop operation and 1.07 during single loop operation with the-

.l reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

AEELICABILITY: OPERATIONAL' CONDITIONS I and 2.

ACTION:

With MCPR less than the above iimits and the reactor vessel steam dome' pressure greater than 785 psig and core flow greater than 10% of rated flow, be-in at 7

least HOT THUTDOWN within 2. hours and comply with the requirements of Specifi--

cation 6J 1.

REACTOR-C00LANT SYSTEM PRESSURE ~

2.1.3 The' reactor coolant system pressure, as measured lin the reactor vessel-L steam dome, shall-not exceed 1325 psig '

\\

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4; e

ACTION:

'With the reactor coolant system pressure, as measured'in the reactor vessel.

steam dome, above 1325 psig, be in'at least HOT SHUTDOWN with reactor' coolant system pressure.less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

GRAND GULF-UNIT 1 2-1 Amendment No. 73,.99, 106 i

.m,L_

5AFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

~

SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the activa irradiated fuel.

APPLICABILITY:

OPERATIONAL CONDITIONS 3, 4 and 5 ACTION:

\\

With the reactor vessel water'1evel at or below the top of the active.

irradiated fuel, manually initiate the ECCS to restore the water level.

Depressurize the reactor vessel as necessary *for ECCS operation.

Cnaply with the requirements of Specification 6.7.1.

?

I L

GRAND GULF-UNIT 1.

2-2 r

v d

F

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

APPLIr. ABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

During two loop operation or single loop operation, with an APLHGR exceeding the limits, initiate corrective action within 15 minutes and restore APLHGR to I

within the required limits 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 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

5VRVEIllANCE RE0VIREMENTS 4.2.1 All APLHGRs shall be verified to be equal te or less than the required limits:

a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER., and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 73, 99, 106 d

POWER DISTRIBUTION LIMITS 3/4.2.2 IDELETED1 i

P GPMiD GULF-UNIT 1 3/4 2-2 Amendment No. M, 106

a POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limits specified in the CORE OPERATING LIMITS REPORT.

APPllCABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With MCPR less than the applicable MCPR limits, initiate corrective action l

within 15 minutes and restore MCPR to within the required limits 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 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEllLANCE RE0VIREMENTS 4,2.3 MCPR shall be determined to be equal to or greater than the applicable MCPR limits.

l a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when-the reactor is operating with a LIMITING CONTROL R00 PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

l GRAND GULF-UNIT 1 3/4 2-3 Amendment No. 73,106 l

POWER DISTRIBUTION 1IMITS 3/4.2.4 LINEAR HEAT GENERATION RAll-LTHITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater lthan or equal to 25% of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action l

within 15 minutes and restore the LHGR to within the 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 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 RE0VIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than their allowable limits:

a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERHAL POWER, c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR, and d.

The provisions of Specification 4.0.4 are not applicable.

GRAND GULF 1-UNIT 1 3/4 2-4 Amendment No. 73,106 l

p 3/4.2 POWER DISTRIBUTION LIMITS BASFS The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46.

}]_L2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for two loop operation are specified in the CORE OPERATING LIMITS REPORT (COLR).

For single-loop operation, a MAPLHGR limit corresponding to the product of the two loop HAPLHGR and a reduction factor specified in the COLR can be l

conservatively used to ensure that the PCT for single loop operation is bounded by the PCT for two loop operation.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No. M, 99,106

POWER DISTRIBUTION LIMITS BASES HINIMUM 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 changes.

The requirement to calculate MCPR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 MCPR after initially determining a LIMITING CONTROL R0D 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 are multiplied by the smaller of either the flow dependent l

LHGR factor (LHGRFAC,) 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,'s are generated to protect the core from slow flow runout transients. 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 generated to p

protect the core from plant transients other than core flow increases.

The daily requirements 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 Amendment No. M, M,106

l DES!GN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 800 fuel assemblies.

Each fuel assembly shall contain fuel rods and water rods clad with Zircaloy cladding.

Each fuel rod shall have a design nominal active fuel length of 150 inches.

Reload fuel I

shall have mechanical, thermal-hydraulic and neutronic characteristics compatible with the initial core loading.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC Staff-approved codes and methods, shown to comply with all safety design bases, and are identified in the Core Operating Limits Report.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 193 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing a design nominal 143.7 inches of boron carbide, B C, powder surrounded by a cruciform 4

shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pump.

2.

1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3.

1550 psig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575'F.

VOLUME S.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,000 cubic feet at a nominal T,y, of 533*F.

GRAND GULF-UNIT 1 5-5 Amendment No. 57,106 l

l ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to main steam system safety / relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT (COLR) 6.9.1.11 Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT (COLR) for the following:

The Average Planar Linear Heat Generation Rate (APLHGR) for Technical a.

Specification 3.2.1.

b.

The Minimum Critical Power Ratio (MCPR) for Technical Specification 3.2.3.

The Linear Heat Generation Rate (LHGR) for Technical Specification c.

3.2.4.

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

The appropriate revision / supplement number for each document shall be identified in the Core Operating Limits Report.

1)

XN-NF-79-71(P), Exxon Nuclear Plant Transient Methodoloav for Boilina Water Readon, Exxon Nuclear Company, Inc., Richland, WA. Approved by NRC letter dated October 24, 1986.

2)

XN-NF-80-19(P)(A), Volume 1, Exxon Nuclear Methodoloav for Boilina Water Reactors - Neutronic Methods for Desian and Analysis, Exxon Nuclear Company, Inc., Richland, WA.

3)

XN-NF-80-19(P)(A), Volume 1, Advanced Nuclear Fuels Methodoloav for Boilina Water Reactors: Benchmark Results for the CASMO-3G/MICR0 BURN-B Calculation Methodoloav, Advanced Nuclear Fuels Corporation, Richland, WA.

4)

XN-NF-80-19(P)(A), Volume 3, Exxon Nuclear Methodoloav for Boilina Water Reactors THERMEX: Thermal Limits Methodoloav Summary Descriotion," Exxon Nuclear Company, Inc., Richland, WA.

5)

ANF-913(P)(A) Volume 1, COTRANSA2:

A Computer Proaram for Boilina Water Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, Richland, WA.

6)

ANF-1125(P)(A), ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, Richland, WA.

7)

XN-NF-84-105(P)(A), Volume 1, XCOBRA-T: A Comouter Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA.

GRAND GULF-UNIT 1 6-19 Amendment No. 80, 87, 106

~ ' ' '

ADMINISTRATIVE CONTROLS e

' CORE OPERATING LIMITS REPORT-(COLR) (Continued) 8)

XN-NF-573(P),-RAMPEX Pellet-Clad-Interaction Evaluation Code for Power Ramos, Exxon Nuclear Company, Inc., Richland, WA. Approved by NRC letter dated August 28, 1990.

9)

XN-NF-81-58(P)(A),BQQfl2:

Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.

10) XN-NF-85-74(P)(A), RODEX2A (BWR): Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc.,- Richland,. WA.

11)

XN-CC-33(P)(A) HUXY:

A Generalized Multirod Heatuo' Code with 10CFR50 Aooendix K Heatuo Option Exxon Nuclear Company, Inc., Richland, WA.

12)

XN-NF-825(P)(A), BWR/6~ Generic Rod Withdrawal Error Analysis. MCPRA Plant Ooeration Within the Extended Ooeratino Domain," Exxon Nuclear Company, Inc., Richland, WA.

13) XN-NR-81-51(P)(A), LOCA-Seismic Structural Resoonse of an Exxon Nuclear Comoany BWR Jet Pumo Fuel Assembiv Exxon Nuclear Company, Inc.. Richland, WA.

14) XN-NF-84-97(P)(A), LOCA-Seismic Structural Response of-an ENC 9x9 BWR Jet Pumo Fuel Assembly, Advanced-Nuclear Fuels Corporation Richland,. WA.

15) XN-NF-86-37(P), Generic LOCA Break Soectrum Analysis 'for BWR/6 ' Plants, Exxon Nuclear Company, Inc., Richland, WA. Approved by NRC letter dated October 24, 1986.

'16) XN-NF-82-07(P)(A), Exxon Nuclear Company ECCS C1addino -Swellino-and

~

Ruoture Model, Exxon Nuclear Company, Inc., Richland, WA.

17) XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, Exxon Nuclear Methodoloov for-Boilina Water Reactors EXEM BWR ECCS Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.

18) XN-NF-79-59(P)(A), Methodoloov for Calculation of Pressure Droo in BWR Fuel Assemblies, Exxon Nuclear Comp.any,. Inc., Richland, WA. -

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

The COLR, including.any mid cycle revisions or supplements,: shall be provided upon issuance for. each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Nuclear Regulatory Commission pursuant to Section 50.4'of 10 CFR Part 50 within the time period specified for each report.

GRAND GULF-UNIT 1 6-19a Amendment'No.'106

.1.

J*'

c-

' ADMINISTRATIVE CONTROLS-6.10 -RECORD RETENTION:

In addition to the applicable record retention requirements of Title 10, Code of Federal. Regulations, the: following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained forlat least five years:

a.

Records and logs of unit operation covering time interval at each power level, b.

Records and logs of principal maintenance activities, inspections,-

repair and replacement of principal items of equipment related to.

nuclear safety.

c.

All REPORTABLE EVENTS.

i j

i GRAND-GULF-UNIT 1 6-19b Amendment No.106 [

l ADMINISTRATIVE CORTROLS 1

6.10 RECORDRETENTION(Continued) 4 d.

Records of surveillance activities inspections and calibrations requiredbytheseTechnicalSpeciflcations, Records of changes made to the procedures required by Specification e.

6.8.1.

f.

Records of radioactive shipments, g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Unit Operating License:

Records and drawing changes reflecting unit design modifications made a.

to systems and equipment described in the Final Safety Analysis Report.

b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

Records of radiation exposure for all individuals entering radiation c.

control areas.

d.

Records of gaseous and liquid radioactive material released to the environs.

Records of transient or operational cycles for those unit components e.

identified in Table 5.7.1-1.

f.

Records of rea: tor tests and experiments.

g.

Records of training and qualification for curren, members of the unit

staff, h.

Records of in-service inspections performed pursuant to these Technical Specifications, i.

RecordsofQualityAssuranceactivitiesrequiredbytheOperational Quality Assurance Manual not listed in Section 6.10.1.

j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 30 CFR 50.59.

k.

Records of meetings of the PSRC and the SRC.

1.

Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and' associated iitstallation and maintenance records.

Records of analyses required by the radiological environmental m.

monitoring program.

n.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

GRAND GULF-UNIT 1 6-20 Amendment No. 2f, 37

. _ - _ _ _ _ - _ - _ -

ML20127N310

Text

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