ML20212N345

Amend 119 to License DPR-50,revising Tech Specs to Provide DHR Capability in All Modes of Operation

ML20212N345
Person / Time
Site:	Crane
Issue date:	08/14/1986
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp
Shared Package
ML20212N346	List: ML20212N345 ML20212N354
References
DPR-50-A-119 NUDOCS 8608280192
Download: ML20212N345 (12)

Text

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION 5

y WASHINGTON, D. C. 20555 Al METPOPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION D0CKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.119 License No. DPR-50 1.

The Nuclear Regulatory Comission (the Commi'ssion) has found that:

A.

The application for amendment by GPU Nuclear Corporation, et al.

(the licensees) dated March 8, 1985, as supplemented May 14, 1985, and April 24, 1986, 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 (1) that the activities authorized by this amendment can be conducted without endangering the health

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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.

860E280192 860814 PDR ADOCK 05000289 P

PDR

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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. DPR-50 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.119, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGl'LATORY COMMISSION h

F. Stolz, Directorj PW Project Directoratg/#6 D vision of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: August 14, 1986

ATTACHMENT TO LICENSE AMENDMENT 'NO.

119 FACILITY OPERATING LICENSE N0. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appen' dix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

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i Remove Insert ii ii iv iv 3-25 3-25 3-26 3-26 3-26a-3-26a 3-26b 3-26c 4-52 4-52 4-52a 4-52a I

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l Section TABLE OF CONTENTS Page 2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 Safety Limits, Reactor Core 2-1 2.2 Safety Limits, Reactor System Pressure 2-4 2.3 Limiting Safety System Settings, Protection 2-5 Instrumentation 3

LIMITING CONDITIONS FOR OPERATION 3-1 3.0 General Action Requirements 3-1 3.1 Reactor Coolant System 3-la 3.1.1 Operational Components 3-la 3.1.2 Pressurization, Heatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17

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3.1.9

. Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves 3-18b 3.1.12 Pressurizer Power Operated Relief Valve (PORV) 3-18c and Block Valve 3.1.13 Reactor Coolant System Vents 3-18f 3.2 Makeup and Purification and Chemical Addition Systems 3-19 3.3 Emergency Core Cooling, Reactor Building Emergency 3-21 Cooling and Reactor Building Spray Systems 3.4 Decay Heat Removal Capability 3-25 3.4.1 Reactor Coolant System Temperature Greater 3-25 Than 250*F 3.4.2 Reactor Coolant System Temperature 250*F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation

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3.5.2 Control Rod Group and Power Distribution Limits 3-33 l

i 3.5.3 Engineered Safeguards Protection System Actuation 3-37 Setpoints l

3.5.4 Incore Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.6 Reactor Building 3-41 3.7 Unit Electrical Power System 3-42 I

3.8 Fuel Loading and Refueling 3-44 3.9 Radioactive Materials 3-46 l

3.10 Miscellaneous Radioactive Materials Sources 3-46 i

3.11 Handling of Irradiated Fuel 3-55 l

l 3.12 Reactor Building Polar Crane 3-57 i

j 3.13 Secondary System Activity 3-58 i

3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around TMI 3-59 3.14.2 Flood Condition for Placing the Unit in Hot Standby 3-60 i

3.15 Air Treatment Systems 3-61 l,

3.15.1 Emergency Control Room Air Treatment System 3-61 3.15.2 Reactor Buf1 ding Purge Air Treatment System 3-62a 3.15.3 Auxiliary and Fuel Handling Exhaust Air Treatment 3-62c i

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Amendment No. 59, pf, 7), Sf, 9,119 6

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Section TABLE OF CONTENTS Page 4.7 Reactor Control Rod System Tests 4-48 4.7.1 Control Rod Drive System Functional Tests 4-48 4.7.2.

Control Rod Program Verification 4-50 4.8 Main Steam Isolation Valves 4-51 4.9 Decay Heat Removal Capability - Periodic Testing 4-52 4.9.1 Emergency Feedwater System - Periodic Testing 4-52 (Reactor Coolant Temperature Greater Than 250*F) 4.9.2 Decay Heat Removal Capability - Periodic Testing 4-52a (Reactor Coolant Temperature 250'F or Less) 4.10 Reactivity Anomalies 4-53' 1

4.11 Reactor Coolant System Vents 4-54 4.12 Air Treatment Systems 4-55 4.12.1 Emergency Control Room Air Treatment System 4-55 4.12.2 Reactor Building Purge Air Treatment System 4-55b 4.12.3 Auxiliary & Fuel Handling Exhaust Air Treatment System 4-55d 4.13 Radioactive Materials Sources Surveillance 4-56 4.14 Reactor Building Purge Exhaust System 4-57 4.15 Main Steam System Inservice Inspection 4-58 4.16 Reactor Internals Vent Valves Surveillance 4-59 4.17 Shock Suppressors (Snubbers) 4-60 4.18 Fire Protection Systems 4-72 4.18.1 Fire Protection Instruments 4-72 4.18.2 Fire Suppression Water System 4-73 4.18.3 Deluge / Sprinkler System 4-74 4.18.4 CO2 System 4-74 4.18.5 Halon Systems 4-75 4.18.6 Hose Stations 4-76 4.19 OTSG Tube Inservice Inspection 4-77 4.19.1 Steam Generator Sample SeTection & Inspection Methods 4-77 4.19.2 Steam Generator Tube Sanple Selection & Inspection 4-77 4.19.3 Inspection Frequencies 4-79 4.19.4 Acceptance Criteria 4-80 4.19.5 Reports 4-81 4.20 Reactor Buildina Air Temperature 4-86 4.21.1 Radioactive Liquid Effluent Instrumentation 4-87 4.21.2 Radioactive Gaseous Process & Effluent Monitoring 4-90 _

Instrumentation i

4.22.1.1 Liquid Effluents 4-97 4.22.1.2 Dose 4-102 4.22.1.3 Liquid Waste Treatment 4-103 4.22.1.4 Liquid Holdup Tanks 4-104 4.22.2.1 Dose Rate 4-105 4.22.2.2 Dose, Noble Gas 4-110 4.22.2.3 Dose, Radiofodines, Radioactive Material in Particulate 4-111 Form & Radionuclides Other Than Noble Gases 4.22.2.4 Gaseous Radwaste Treatment 4-112 4.22.2.5 Explosive Gas Mixture 4-113 4

4.22.2.6 Gas Storage Tanks 4-114 4.22.3.1 Solid Radioactive Waste 4-115 4.22.4 Total Dose 4-116 4.23.1 Monitoring Program 4-117 4.23.2 Land Use Census 4-121 4.23.3 Interlaboratory Comparison Program 4-122 i

Amendment No.,W, % g, ff, fT, [J6f6',)MI, f,119 fv r

3.4 DECAY HEAT REMOVAL CAPABILITY Applicability Applies to the operating status of. systems and components that function to remove decay heat when one or more fuel bundles are located in the reactor vessel.

Objective To define the conditions necessary to assure continuous capability of decay heat removal.**

Specification 3.4.1 Reactor Coolant System temperature greater than 250'F.

3.4.1.1 With the Reactor Coolant System temperature greater than 250*F, three independent EFW pumps and associated flos paths shall be OPERABLE with:

a. Two EFW pumps, each capable of being powered from an OPERABLE emergency bus, and one EFW pump capable of being powered from an OPERABLE steam supply system.

Specification 3.0.1 applies.

b. With one pump or flow path

• inoperable, restore the inoperable pump or flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With more than one EFW pump or flow path

• inoperable, restore the inoperable pumps or flow paths

• to OPERABLE status or be subcritical within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. Four of six turbine bypass valves OPERABLE.

d. The condensate storage tanks (CST) OPERABLE with a minimum of 150,000 gallons of condensate available in each CST. With a CST inoperable, restore the CST to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. With more than one CST inoperable, restore the inoperable CST to OPERABLE status or be subcritical within I hour, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specification 3.0.1 applies.

• For the purpose of this requirement, an OPERABLE flow path snall mean an unobstructed path from the water source to the pump and from the pump to a steam generator.

• • These requirements supplement the requirements of Sections 3.1.1.1.c.

3.1.1.2, 3.3.1 and 3.8.3.

3-25 Amendment No. g, 7I, J,119 d

3.4.1.2 With the Reactor Coolant System temperature greater than 250*F, all l

eighteen (18) main steam safety valves shall be OPERABLE or, if any are not OPERABLE, the maximum overpower trip setpoint (see Table 2.3-1) shall be reset as follows:

Maximum. Number of Maximum Overpower Safety Valves Disabled en Trip Setpoint Any Steam Generator

(% of Rated Power) 1 92.4 2

79.4 3

66.3 With more than 3 main steam safety valves inoperable, restore at least fifteen (15) main steam safety valves to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.4.2 Reactor Coolant System temperature 250 F or less.

3.4.2.1 With Reactor Coolant temperature 250*F or less, at least two of the following means for maintaining decay heat removal capability shall be OPERABLE and at least one shall be in operation except as allowed by Specifications 3.4.2.2, 3.4.2.3 and 3.4.2.4.

Decay Heat Removal String "A".

a.

b.

Decay Heat Removal String "B".

Reactor Coolant Loop "A", its associated 0TSG, and its associated c.

emergency feedwater flowpath.

d.

Reactor Coolant Loop "B", its associated 0TSG, and its associated emergency feedwater flowpath.

3.4.2.2 Operation of the means for decay hest removal may be suspended provided the core outlet temperature is maintained below saturation temperature.

3.4.2.3 The number of means for decay heat removal required to be operable per 3.4.2.1 may be reduced to one provided that one of the following conditions is satisfied:

The Reactor is in a Refueling Shutdown condition with the Fuel a.

Transfer Canal water level greater than 23 feet above the reactor vessel flange.

b.

Reactor coolant temperature is less than 140*F with BWST level greater than 44 feet and an associated flow path through the RCS OPERABLE such that core outlet temperature can be maintained subcooled for at least 7 days.

3-26 AmendmentNo.g,%,119

Equipment Maintenance on one of the means for decay heat removal c.

specified by 3.4.2.1 is required and the equipment outage does not exceed 7 days.

3.4.2.4 Specification 3.4.2.1 does not apply when either of the following conditions exist:

a.

Decay heat generation is less than 188 KW with the RCS full.

b.

Decay heat generation is less than 100 KW with the RCS drained down for maintenance.

3.4.2.5 With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

2 3-26a Amendment No.119

Bases A reactor shutdown following power operation requires removal of core decay heat. Normal decay heat removal is by the steam generators with the steam dump to the condenser when RCS temperature is above 250*F and by the decay heat removal system below 250*F. Core decay heat can be continuously dissipated up to 15 percent of full power via the steam bypass to the condenser as feedwater in the steam generator is converted to steam by heat absorption. Normally, the capability to return feedwater flow to the steam generators is provided by the main feedwater system.

The main steam safety valves will be able to relieve to atmosphere the total steam flow if necessary.

If Main Steam Safety Valves are inoperable, the power level must be reduced, as stated in Technical Specification 3.4.1.2 such that the remaining safety valves can prevent overpressure on a turbine trip.

In the unlikely event of complete loss of off-site electrical power to the station, decay heat removal is by either the steam-driven emergency feedwater pump, or two half-sized motor-driven pumps. Steam discharge is to the atmosphere via the main steam safety valves and controlled atmospheric relief valves, and in the case of the turbine driven pump, from the turbine exhaust. (1)

Both motor-driven pumps are required initially to remove decay heat with one eventually sufficing. The minimum amount of water in the condensate storage tanks, contained in Technical Specification 3.4.1.1, will allow cooldown to l

250'F with steam being discharged to the atmosphere. After cooling to 250*F, the decay heat removal system is used to achieve further cooling.

When the RCS is below 250'F, a single DHR string, or single OTSG and its associated emergency feedwater flowpath is sufficient to provide removal of decay heat at all times following the cooldown to 250'F. The requirement to maintain two OPERABLE means of decay heat removal ensures that a single failure does not result in a complete loss of decay heat removal capability. The requirement to keep a system in operation as necessary to maintain the system subcooled at the core outlet provides the guidance to ensure that steam conditions which could inhibit core cooling do not occur.

Limited reduction in redundancy is allowed for preventive or corrective maintenance on the primary means for decay heat removal to ensure that a

maintenance necessary to assure the continued reliability of the systems may be accomplished.

As decay heat loads are reduced through decay time or fuel off loading, alternate flow paths will provide adequate cooling for a time sufficient to take compensatory action if the normal means of heat removal is lost.

3-26b Amendment No.119 4

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i With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. The BWST 1

with level at 44 fet:t provides an equivalent reserv.oir available as a heat sink. Operability of the BWST is to be determined using calculations based on actual plant data or through plant testing at the time the system is to be declared operable. At such times that either of these means is detemined to be operable, removal of the redundant or diverse cooling system is permitted.

Following extensive outages or major core off loading, the decay heat generation being removed from the Reactor Vessel is so low that ambient losses are sufficient to maintain core cooling and no other means of heat removal is required.

The system is passive and requires no redundant or 1

diverse backup system. Decay heat generation is calculated in accordance with ANSI 5.1-1979 to determine when this situation exists.

An unlimited emergency feedwater supply is available from the river via either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period of time.

The requirements of Technical Specification 3.4.1.1 assure that before the reactor is heated to above 250*F, adequate auxiliary feedwater capability is available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven pumps (460 gpm each) are specified. However, i

only one half-capacity motor-driven pump is necessary to supply auxiliary feedwater flow to the steam generators in the onset of a small break loss-of-coolant accident.

j REFERENCES (1) FSAR Section 10.2.1.3.

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4 3-26c AmendmentNo.Jd,jjg i

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j 4.9 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING Applicability Applies to the ' periodic testing of systems or co:nponents which function to i

remove decay heat.

Objective To verify that systems / components required for decay heat removal are capable of performing their design function.

Specification i

i 4.9.1 Emergency Feedwater System - Periodic Testing (Reactor Coolant System Temperature greater than 250*F.)

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j 4.9.1.1 Whenever the Reactor Coolant System temperature is greater than 1

250*F, the EFW pumps shall be tested in the recirculation mode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3210.

The test frequency shall be at least every 31 days of plant operation at Reactor Coolant Temperature above 250*F.

4.9.1.2 During testing of the EFW System when the reactor is in STARTUP, HOT STANDBY or POWER OPERATION, if one steam generator flow path

• is made inoperable, a dedicated qualified individual who is in comunication j

with the control room shall be continuously stationed at the EFW local manual valves (See Table 4.9-1).

On instruction from the j

Control Room Operator, the individual shall realign the valves from the test mode to their operational alignment, i

4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be i

verified to be in the statu: specified in Table 4.9-1, when required to be OPERABLE.

4.9.1.4 On a quarterly basis, verify that the manual control (HIC-849/850)

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valve station functions properly.

4.9.1.5 On a quarterly basis, EFV-30A and B shall be checked for proper operation by cycling each. valve over its full stroke.

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4.9.1.6 Prior to start-up, following a refueling shutdown or a cold shutdown greater than 30 days, conduct a test to demonstrate that the motor i

4 driven EFW pumps can pump water from the condensate tanks to the Steam Generators.

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• For the purpose of this requirement, an OPERABLE flow path shall mean an j

unobstructed path from the water source to the pump and from the pump to a l

Steam Generator.

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4-52 Amendment No. M, 119 l

4.9.1.7 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9'.1.1..

4.9.2 Decay Heat Removal Capability - Periodic Testing (Reactor Coolant t

System Temperature 250*F or less).*

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4.9.2.1 On a daily basis, verify operability of the means for decay heat removal required by specification 3.4.2 by observation of console status indication.

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• These requirements supplement the requirements of 4.5.2.2 and 4.5.4.

Bases The 31-day testing frequency will be sufficient to verify that the turbine i

driven and two motor-driven EFW pumps are operable and that the associated valves are in the correct alignment. ASME Section XI, Article IWP-3210

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specifies requirements and acceptance standards for the testing of nuclear safety related pumps. Compliance with the normal acceptance criteria assures

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that the EFW pumps are operating as expected. The test frequency of 31 days i

(nominal) has been demonstrated by the B&W Emergency Feedwater Reliability Study to assure an appropriate level of reliability.

In the case of the EFW System flow, the flow shall be considered acceptable if under %e worst case single pump failure, a minimum of 500 gpm can be delivered when steam i

generator pressure is 1050 psig and one steam generator is isolated. A flow i

of 500 gpm, at 1050 psig head, ensures that sufficient flow can be delivered to either Steam Generator. The surveillance requirements ensure that the overall EFW System functional capability is maintained.

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Daily verification of the operability of the required means for decay heat t

removal ensures that sufficient decay heat removal capability will be maintained.

4-52a Amendment No. 74,119 l

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