ML20035E539

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Amend 19 to License NPF-86.Amend Revises TS RCS Heatup & Cooldown Limitations
ML20035E539
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 04/07/1993
From: Stolz J
Office of Nuclear Reactor Regulation
To:
North Atlantic Energy Service Corp
Shared Package
ML20035E541 List:
References
NPF-86-A-019 NUDOCS 9304160152
Download: ML20035E539 (11)
v • d • e


Text

.

[me REGy'%

UNITED STATES j

NUCLEAR REGULATORY COMMISSION g

y wasmucTON. D. C. 20555 s-5 p

....=

NORTH ATLANTIC ENERGY SERVICE CORPORATION. ET AL" DOCKET NO. 50-443 SEABROOK STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.19 License No. NPF-86 1.

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

A.

The application for amendment by North Atlantic Energy Service Corporation, et al. (the licensee), dated August 17, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and l

l regulations set forth in 10 CFR Chapter I;

)

B.

The facility will operate in conformity with the application, I

the provisions of the Act, and the rules and regulations of the l

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 l

conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common i

defense and security or to the health and safety of the public; and 1

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.

  • North Atlantic Energy Service Corporation (NAESCO) is authorized to act as agent for the:

North Atlantic Energy Corporation, Canal Electric Company, The Connecticut Light and Power Company, EUA Power Corporation, Hudson Light and Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Companj, New England Power Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, The United Illuminating Company, and Vermont Electric Generation and Transmission Cooperative, Inc.,

and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

9304160152 930407 PDR ADOCK 05000443 P

PDR

_ _ _ _ _. 2.

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-86 is hereby amended to read as follows:

(2)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised l

through Amendment No.19

, and the Environmental Protection Plan contained in Appendix B are incorporated into Facility License No.

NPF-86.

NAESCO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION I

John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 7,1993 l

)

i 7

ATTACHMENT TO LICENSE AMENDMENT NO 19 FACILITY OPERATING LICENSE NO. NPF-86 l

i DOCKET NO. 50-443 Replace the following pages of Appendix A, Technical Specifications, with the attached pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Overleaf j

pages have been provided Remove Insert i

  • ix
  • ix i

x x

3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 i

B 3/4 4-7 B 3/4 4-7 B 3/4 4-8 B 3/4 4-8

  • B 3/4 4-9
  • B 3/4 4-9 B 3/4 4-10 B 3/4 4-10 4

a 4

l INDEX LIMITING CCNDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTICN PAGE i

3/4.9.4 CCNTAISMENT BUILDING PENETRATIONS..........

3/4 9-4 i

3/4.9.5 CCMMUNICATION5..............................

3/4 9-5 i

3/4.9.6 REFUELING MACHINE........................................

3/4 9-6 i

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS.................

3/4 9-7 l

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Hign Water level.........................................

3/4 9-8 i

Low Water Level...........

3/4 9-9 j

3/4.9.9 CCNTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM...........

3/4 9-10 I

3/4.9.10 WATER LEVEL - REACTOR VE55EL.....................

3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL...............................

3/4 9-12 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM......

3/4 9-13 6

3/4.9.13 5 PENT FUEL ASSEMBLY ST0 RAGE..............................

3/4 9-16 i

FIGURE 3.9-1 FUEL ASSEMBLY BURNUP V5. INITIAL ENRICHMENT.........

i FOR SPENT FUEL' ASSEMBLY STORAGE.....................

3/4 9-17 3/4.9.14 NEW FUEL ASSEMBLY ST0 RAGE................................

3/4 9-18

}

3/4.10 SPECIAL TEST EXCEPTIONS r

3/4.10.1 5HUTDOWN MARGIN..........................................

3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTIGN, AND POWER DISTRIBUTION LIMITS...

3/4 10-2 3/4.10.3 PHYSICS TEST 5............

3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS..........

3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................

3/4 10-5 3/4.10.6 REACTOR COOLANT LOOPS.................................

3/4 10-6 3/4.11 RADI0 ACTIVE EFFLUENTS f

3/4.11.1 LIQUID EFFLUENTS

{

Contentration............................

3/4 11-1 Dose....................................................

3/4 11-2 i

Liouid Racwaste Treatment 5ystem.........................

3/4 11-3 i

Liquid Holcup Tanks.......................................

, 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS A Dose Rate................

3/4 11-5 Dose - Noble Gases.......................................

3/4 11-6 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Materi al i n Pa rti cul ate Fo rm.............................

3/4 11-7.

Gaseous Radwaste Treatment 5ystem........................

3/4 11-8 Explosive Gas Mixture -

5ystem.............

3/4 11-9 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................

3/4 11-10 3/4.11.4 TOTAL 005E.......................-........................

3/4 11-12 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM......................................

3/4 12-1 SEAER00( - UNIT 1 ix Amendment No. 6 Correction letter of 9-12-91

Lcntroiting materiai:

case metai

~

Cop;er con *ent:

0.06 WT%

Ric. initial:

40*F RT?~'after11.1EFPY:

1/4T, 108'F 3/4T, 86*F Curve applicable for heatup rates up to 60*F/hr for the service period up to 11.1 EFPY and contains margins of 10*F and 60 psig for possible instrument errors t

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FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 11.1 EFPY SEABROOK - UNIT 1 3/4 4-31 Amendment No. 19

INDEX l

LIMITING CCNDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS l

SECTICN PAGE 3/4.9.4 CCNTAINMENT SUILDING PENETRATIONS....

3/4 9-4 3/4.9.5 CCMMUNICATIONS..

3/4 9-5 3/a.9.6 REFUELING MACHINE...................................

3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS...............

3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.....

3/4 9-8 Low Water Level.................................

3/4 9-9 3/4.9.9 CCNTAINMENT PURGE AND EAHAUST ISOLATION SYSTEM...........

3/4 9-10 l

3/4.9.10 WATER LEVEL - REACTOR VESSEL............................

3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL...........................

3/4 9-12 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM......

3/4 9-13 3/4.9.13 SPENT FUEL ASSEMBLY STORAGE..

3/4 9-16 FIGURE 3.9-1 FUEL ASSEMBLY BURNUP VS. INITIAL ENRICHMENT.........

FOR SPENT FUEL ASSEMBLY STORAGE.

3/4 9-17 3/4.9.14 NEW FUEL ASSEMBLY STORAGE..

3/4 9-18 l

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN..........................................

3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS...

3/4 10-2 3/4.10.3 PHYSICS TESTS........

3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS.....

3/4 10-4 l

3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN....................

3/4 10-5 3/4.10.6 REACTOR COOLANT LOOPS.

3/4 10-6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration.

3/4 11-1 Dose.

3/4 11-2 Licuic Rac*aste Treatment System........................

3/4 11-3 Liquic Holcup Tanks..

3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS Dose Rate............

3/4 11-5 Dose - Noble Gases........

3/4 11-6 Dose - Iodine-131, Iocine-133, Tritium, and Radioactive Material in Particulate Form.............................

3/4 11-7 Gaseous Radwaste Treatment System.......................

3/4 11-8 i

Explosive Gas Mixture -

5ystem...........................

3/4 11-9 3/4.11.3 SOLID RADI0 ACTIVE WASTES...............................

3/4 11-10 3/4.11.4 TOTAL D05E.............

3/4 11-12 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MCNITORING PROGRAM.....

3/4 12-1 l

SEABROOK - UNIT 1 ix Amendment No. 6 Correction letter of 9-12-91 l

i

l i

l INDEX i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS i

SECTION PAGE l

3/4.12.2 LAND USE CENSUS 3/4 12-3 3/4.12.3 INTERLABORATORY COMPARISON."20 GRAM...........

3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL....................

B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS....................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...............

B 3/4 1-3 i

I 3/4.2 POWER DISTRIBUTION LIMITS B 3/4 2-1 i

I 3/4.2.1 AXIAL FLUX DIFFERENCE B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX H0T CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO B 3/4 2-3 3/4.2.S DNB PARAMETERS.....................

B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION...............

B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION..............

B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 3/4.4.2 SAFETY VALVES B 3/4 4-1 3/4.4.3 PRESSURIZER B 3/4 4-2 3/4.4.4 RELIEF YALVES B 3/4 4-2 3/4.4.5 STEAM GENERATORS....................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.............

B 3/4 4-3 3/4.4.7 CHEMISTRY B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/4 4-7 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE-(E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE................

B 3/4 4-9 FIGURE B 3/4.4-2 (This figure number not used).........

B 3/4 4-10 l

SEABROOK - UNIT 1 x

Amendment No. 19

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

p t.cntroiting materias; case metai Copper centent:

0.05 Wi%

Ric; initial:

40*F Ri ; after 11.1 EFPY:

1/4T, 108'F c

3/4T, 86*F Curve applicable for heatup rates up to 60*F/hr for the service period up to 11.1 EFPY and contains margins of 10*F and 60 psig for possible instrument errors 2S22

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REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 11.1 EFPY SEABROOK - UNIT 1 3/4 4-31 Amendment No.19

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SEAER00K - UNIT 1 3/4 4-32

. Amendment No. 19 l

I

(

REACTOR COOLANT SYSTEM I

l BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and

)

Pressure Vessel Code,Section III, Appendix G:

1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a.

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.

Limit lines for cooldown rates between those presented may be obtained by interpolation; and P

b.

Figures 3.4-2 and 3.4-3 define limits to assure prevention ofnon-ductile failure only.

For normal operation, other inherent plant characteristics, e.g., pump heat addition and l

pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

l 2.

These limit lines shall be calculated periodically using methods provided below,

]

3.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below i

70*F, 4.

The pressurizer heatup and cooldown rates shall not exceed 100*F/h and 200*F/h, respectively.

The spray shall not be used if the temperature difference between the pressurizer and the, spray fluid is greater than 320*F, and 5.

System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Regulatory Guide 1.99, Revision 2, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT,1 EFPY service life periodThe 1 tive full power years (EFPY) of service life.

SEABROOK - UNIT I B 3/4 4-7 Amendment No. 19

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) greater than the RT., of the limiti, at the 1/4T location in the core region is is chosen such that the limiting RT ng unirradiated material. The selection of such a limiting RT., assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT.1; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT.,.

Therefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel content of the materiai in question, can be predicted using Figure B 3/4.4-1 and the value of ART., f computed by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement o Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT., at the end of 11.1 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments.

from the material s, determined in this manner may be used until the results Values of ART urveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10CFR50, Appendix H.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.

Evaluation of surveillance capsule data will be conducted in accordance with NRC Regulatory Guide 1.99, Revision 2.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-79244.

The general method for calculating heatup and cooldown limit curves is l

based upon the principles of the linear elastic fracture mechanics (LEFM)-

)

technology.

In the calculation procedures, a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T 1

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