ML19323E367

Proposed Tech Spec Changes Providing Limiting Conditions for Operation & Surveillance Requirements for Degraded & Loss of Voltage Protecting Circuitry.Revised Pages to Tech Specs, App A,Encl

ML19323E367
Person / Time
Site:	Prairie Island
Issue date:	05/16/1980
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML19323E365	List: ML19323E364 ML19323E366 ML19323E367
References
NUDOCS 8005230442
Download: ML19323E367 (42)

Text

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i e f a EXHIBIT A

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Prairie Island Nuclear Generating Plant License Amendment Request dated May 16, 1980 Proposed Changes to the Technical Specifications Appendix A of Operating Licenses DPR-42 and DPR-60 Pursuant to 10 CFR 50.59, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:

1. Safeguards Bus Voltage Protection PROPOSED CHANGE Add a new Table TS.3.5-6 and revise pages TS-iii, TS.3.5-1, TS.3.5-3, TS.3.5-4, Table TS.3.5-1, and Table TS4.1-1 to provide Limiting Condi-tions for Operation and Surveillance Requirements for degraded and loss of voltage protection circuitry. Refer to Exhibit B for the proposed fo rmat and wording of these changes.

REASON FOR CHANGE Modifications to provide automatic degraded voltage protection for the safeguards buses at Prairie Island were made during the 1978 Unit No. 2 and 1979 Unit No. I refueling outages. The modifications substantially conformed to the " Statement of Staf f Positions Relative to Emergency Power Systems for Operating Reactors" provided to Northern States Power Company in a letter dated June 3, 1977 from Mr D K Davis, USNRC. These modifications were described in NSP letters dated May 4, 1978, October 12, 1979, and October 26, 1979.

Proposed Technical Specifications establishing operability and surveillance requirements for the degraded voltage protection circuitry, as well as for the loss of voltage protection circuitry included in the original plant design, are being submitted at the request of the NRC Staff.

SAFETY EVALUATION The proposed changes establish Limiting Conditions for Operation and Surveillance Requirements for safeguards bus voltage protection circuitry.

The proposed setpoints have been justified in the correspondence referenced above. Operability and surveillance requirements conform to guidance provided by the Commission in Enclosure (2) to D K Davis's letter dated June 3, 1977.

The proposed changes add Technical Specifications relating to upgraded bus protection instrumentation. This provides assurance that the protection instrumentation will be operable when required. Existing Technical Specification requirements are not affected by these changes, j l

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i e t e EXHIBIT A

2. ' Emergency charcoal Filter System Test Initiation Signal PROPOSED CHANGE Delete the requirement in Specifications 4.4.B.1 and 4.4.B.2 to require initiation of the shield building ventilation system and the auxiliary building special ventilation system during testing with a simulated safeguards initiation signal. Refer to Exhibit B, page TS.4.4-4.

i REASON FOR CHANGE Initiation of the shield building ventilation system and the auxiliary building special ventilation system using a simulated safety injection signal is verified each refueling outage during the integrated safeguards actuation test. Use of a simulated safety injection test for the more frequent emergency chatcoal filter system testing is unnecessary and is not consistent with the method specified for operability tests of other safeguards equipment.

Providing a simulated safeguards signal for initiation of these two emergency charcoal filter systems requires jumpering terminals in an auxiliary relay cabinet. /_dministrative controls placed on the use of test jumpers greatly complicate the test procedure. The possibility of human error in applying the test jumpers, while remote, can lead to an unnecessary plant trip. We believe that manual initiar#on is a more appropriate method of conducting these tests.

SAFETY EVALUATION This change permits manual initiation of emergency charcoal filter systems for tes ting. This is consistent with the method specified in the Technical Specifications for initiating other safeguards systems for periodic operability tests. Automatic initiation using a safety injection signal is verified each refueling outage.

3. Containment Fan Coolers Design Performance Verification PROPOSED CHANGE Delete the requirement in Specification 4.5.A.3 to verify fan coil unit performance is within design specifications during each refueling.

Refer to Exhibit B, page TS.4.5-2.

REASON FOR CHANGE i.

i Attempts to comply with the original Technical Specification requirement

over the past six years have proven it to be impractical. Air and I

water temperature and flow measurements at nominal containment air

conditions cannot be used to confirm a fan coil unit's performance is acceptable under accident conditions. Performance under accident-conditions has been conservatively predicted using computer codes as described in Section 14 of the Final Safety Analysis Report.

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3. Containment Fan Coolers (continued)

REASON FOR CHANGE (continued)

In lieu of evaluating fan coil unit performance, we propose to add a requirement to monitor fan coil unit "in" and "out" temperatures of air and cooling water as well as cooling water flow during the refueling shutdown operability test. These parameters will be compared to nominal temperature and flow values to confirm unit operability.

SAFETY EVALUATION This change eliminates an impractical Technical Specification requirement to evaluate fan coil unit performance under temperature and humidity conditions which make such an evaluation impractical. Surveillance requirements to verify operability using readily measureable parameters have been proposed in their place. The surveillance requirements for fan coil units, modified in the manner described, provide a high degree of assurance that component failures will not go undetected.

4. RHR System Flow Requirements "rROPOSED CHANGE Revise Specification ' .5.B.3.h.2 to read, "The minimum flow through each RHR Reactor Vessel injection line shall be at least 1800 gpm."

Refer to Exhibit B, page TS.4.5-3A.

REASON FOR CHANGE The existing wording is misleading. Each RHR reactor vessel injection line is supplied by only one pump.

SAFETY EVALUATION This change corrects a misleading Technical Specification requirement.

The change is administrative in nature and does not affect the minimum j low pressure safety injection flow requirement.

5. Diesel Generator Surveillance PROPOSED CHANGE Revise Specification 4.6.A.2.b.3 to incude ground fault, as well as engine overspeed and generator differential _ current, as trips which are not automatically bypassed on receipt of a safety injection signal.

Refer to Exhibit B, page TS.4.6-1.

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5. Diesel Generator Surveillance (continued)

REASON FOR CHANGE This change is needed to correct the wording of the Specification to be consistent with the diesel generator system design.

SAFETY EVALUATION This change corrects a Technical Specification surveillance requirement to conform to plant design features. The change is administrative in nature and does not affect the diesel generator testing requirements.

6 Safety Related Shock Suppressors PROPOSED CHANGE Revise Table TS.3.12-1 to include additional shock suppressors. Refer to Exhibit B, Table TS.3.12-1, page 1 of 8 through page 8 of 8.

REASON FOR CHANGE The method used to categorize shock suppressors as " safety related" for the original Table TS.3.12-1 was reviewed and found to be too restrictive.

A new evaluation of plant shock suppressors was performed which resulted in a greatly expanded list.

The expanded list of safety related shock suppressors now includes all devices which must be operable to provide seismic protection for the reactor coolant systems and systems required to safety shutdown the reactors and maintain them in a safe shutdown condition. The original list of safety related shock suppressors included only those used on systems defined to be " essential" in the FSAR anslysis of the high energy line break event.

SAFETY EVALUATION This change expands the list of shock suppressors categorized as " safety related". Including more shock suppressors in the Technical Specification inspection and maintenance program will increase the protection provided to essential plant systems from seismic events.

7. Miscellaneous Corrections and Clarifications PROPOSED CHANGE Make the following changes in the Technical Specifications to correct errors or provide additional cisrification. These changes are shown in Exhibit B:

a. On Table of Contents page TS-i, correct the spelling of " Power" in )

item 3.10 and the spelling of " Detection" in item 3.14.

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EXHIBIT A

b. On page TS.3.3-4, remove the additional "(1)" in the left margin of Specification 3.3.C.1.a(1).

c. On page 1 of 5 of Table TS.4.1-1, remove the "*" in the " Check" column for item 4, Reactor Coolant Temperature. Amendment No. 25 to DPR-42 (No. 19 to DPR-60) deleted this symbol. The symbol was inadvertently restored with an unrelated change in Amendment No. 39 to DPR-42 (No. 33 to DPR-60),

d. On page TS.4.2-3, last line of Specification 4.2.C, change ". . . in accordance with Specification 6.6. B.10." to ". . . in accordance with Specification 6.6.B.9".

e. On page TS.6.5-2, Specification 6.5.B.1.b, change the wording as shown in Exhibit B to conform to current security practices of doors being locked or attended and the use of keys or key devices for locking doors.

f. On page TS.6.1-1, line one of Specification 6.1.C.1, change "esch on duty shall. . ." to "Each on duty shif t shall. . . ".

REASON FOR CHANGE The changes described above correct typographical errors, correct term-onology, or provide additional clarification of Technical Specification requirements. No change in the substance of any requirement is proposed.

SAFETY EVALUATION This is an administrative change having no affect on existing Technical Specification requirements.

i EXHIBIT A

8. Organizational Changes PROPOSED CHANGE 3 On pages TS.6.2-1, TS.6.2-3, TS.6.2-5, TS.6.2-6, and TS.6.4-1 and on Figures TS.6.1-1 and TS.6.1-2 make the title and minor wording changes as shown in Exhibit B.

REASON POR CHANGE A change in corporate organization involving electric generating plant j responsibilities on January 1,1980 and a change in Power Production I

Department organization involving nuclear plant activities on April 1, j 1980, requires revision of organization charts and titles contained in  ;

sections 6.1, 6.2 and 6.4 of the Appendix A Technical Specifications. l I

! SAFETY EVALUATION The organization changes provide greater headquarters participation in, and technical support for, nuclear plant activities. The Safety Audit Committee retains its independence from line responsibility for plant operation. The proposed changes are of an administrative nature l only.

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9. Definition of Operability PROPOSED CHANGE I

Revise the definition of " operability" in Section 1 of the Technical Specifications to include the additional clarification contained in Enclosure (2) of the letter dated April 10, 1980 from D G Eisenhut, Acting Director, Division of Operating Reactors, USNRC. Refer to Exhibit B, page TS.1-4 and TS.1-5.

REASON FOR CHANGE This change is being made at the request of the NRC Staff.

SAFETY EVALUATION This change provides additional clarification of what is required for equipment to be considered operable. The wording proposed by the NRC

. Staff in their April 10. 1980 letter is used.

This change is administrative in nature. The definition of " operability" proposed by the NRC Staff is concistent with our understanding and use of the term since the beginning of plant operation.

O a 8 e EXHIBIT B License Amendment Request dated May 16, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exhibit B consists of revised pages for the Prairie Island Nuclear Generating Plant Technical Specifica-tions, Appendix A, as listed below showing the proposed changes:

TS-i TS-iii TS.1-4 TS.1-5 TS.3.3-4 TS.3.5-1 TS.3.5-3 TS.3.5-4 Table TS.3.5-1 (Page 1.of 2)

Table TS.3.5-1 (Page 2 of 2) (New Page)

Table TS.3.5-6 (New Page)

Table TS.3.12-1 (Page 1 of 8) (4 New Pages)

Through (Page 9 of 8)

Table TS.4.1-1 (Page 1 of 5)

Table TS.4.1-1 (Page 5 of 5)

TS.4.2-3 TS.4.4-4 TS.4.5-2 TS.4.5-3A TS.4.6-1 TS.6.1-1 Figure TS.6.1-1 Figure TS.6.1-2 TS.6.2-1 TS.6.2-3 TS.6.2-5 TS.6.2-6 TS.6.4-1 TS.6.5-2 I

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TS-i P.EV

< TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 1.0 . Definitions TS.1-1 2.0 Safety Limits and Limiting Safety System Settings TS.2.1-1 1 2.1 Safety Limit, Reactor Core TS.2.1-1

j. 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective Ins trument at ion TS.2.3-1 3.0 Limiting conditions for Operation TS.3.1-1 3.1 Reactor Coolant System Ts.3.1-1 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 3.4 Steam and Power Conversion System TS.3.4-1 3.5 Instrumentation System TS.3.5-1

! 3.6 Containment System TS.3.6-1 3.7 Auxiliary Electrical Systems TS.3.7-1 3.8 Refueliag and Fuel Handling TS.3.8-1 3.9 Radioactive Effluents TS.3.9-1 3.10 Control Rod and Power Distribution Limits TS.3.10-1 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Shock Suppressors (snubbers) TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 3.14 Fire Detection and Protection Systems TS.3.14-1 l j 4.0 Surveillance Requirements TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Primary System Surveillance TF.4.2-1 j 4.3 Reactor Coolant System Integrity Testing TS.4.3-1 4.4 Containment System Tests TS.4.4-1 4.5 Engineered Safety Features TS.4.5-1 4.6 Periodic Testing of Emergency Power System .TS.4.6-1 4.7 Main Steam Stop Valves TS.4.7-1 4.8 Auxiliary Feedwater System TS.4.8-1 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring Program TS.4.10-1 4.11 Radioactive Source Leakage Test TS.4.11-1 4.12 Steam Generator Tube Surveillance TS.4.12-1 4.13 Shock Suppressors (snubbers) TS.4.13-1 4.14 control Room Air Treatment System' TS.4.14-1 4.15. Spent Fuel Pool Special Ventilation System- TS.4.15-1 4.16 Fire Detection and Protection Systems TS.4.16-1

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TS-iii REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data 3.1-1 Unit 2 Reactor Vessel Toughness Data 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Waste Sampling and Analysis 3,9-2 Radioactive Gaseous Waste Sampling and Analysis 3.12 . Safety Related Shock Suppressors (Snubbers) 3.14-1 Safety Related Fire Detection Instruments 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Reactor Coolant System In-Service Inspection Schedule Section 1.0 - Reactor Vessel Section 2.0 - Pressurizer Section 3.0 - Steam Generators and Class A Heat Exchangers Saction 4.0 - Piping Systems Section 5.0 - Reactor Coolant Pumps Section 6.0 - Valves 4.2-2 System Boundaries for Piping Requiring Volumetric Inspection Under Examination Category IS-251 J-l  :

4.2-3 System Boundaries Extending Beyond Those of Table TS.4.2-2 for Piping Requiring Surf ace Inspection Under Examination Category IS-251 J-l 4.2-4 System Boundaries Extending Beyond Those of Tables TS.4.2-2 and -3 for Piping Excluded from Examination under IS-251 but Requiring Visual . Inspection (Which need not Require Removal of Insulation) of all Welds during System Hydrostatic Test ,

4.4-1 Unit I and Unit 2 Penetration Designation for Leakage Tests l 4.10-1 Prairie Island Nuclear Generating Plant- I Radiation Environmental Monitoring Program Sample Collection and Analysis Environmental Monitoring Program 4.12-1 Steam Generator Tube Inspection 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant 1 (Per Unit) l 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Ef fluent From Prairie Island Nuclear Generating Plant (Per Unit) j 6.1-1 Minimum Shif t Crew composition 6.7-1 Special Reports

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TS.1-4 REV G. Limiting Safety System Settings Limiting safety sytem settings. are settings on protective instru-mentation that initiate automatic protective action at a level such that safety limits will not be exceeded.

H. Limiting Conditions for Operation Limiting conditions for operation are those restrictions on unit operation resulting from equipment performance capability that must be met in order to assure safe operation of the unit.

I. Operable A system, subsystem, train, component or device shall be Operable or have Operability when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrica-tion or other auxiliary equipment th at are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support funct ion (s ) .

When a system, subsystem, train, component or device is determined to be inoperable soley because its emergency power source is inoperable, or soley because its normal power source is inoperable, it may be con-sidered opeable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corres-ponding normal or emergency power source is operable; and (2) all of its redundant system (s), subsystem (s), train (s), compo m t(s) and device (s) are OPERABLE, or likewise satis fy the requirement

• u this paragraph.

The operability of a system or component shall be considered to be established when: (1) it satisfies the Limiting Conditions for Operation in Specification 3.0, and (2) it has been tested period-ically in accordance with Specification 4.0 and has met its perfor-nance requirements. Following removal from service, a system or component is considered inoperable until its operability has been reestablished.

J. Power Operation Power operation of a unit is any operating condition that results when the reactor of thec ui:it is critical, and the neutron flux power range instrumentecion it:dicates greater than 2% of rated power.

K. Protection Instrumentation and Logic

1. Protection System The protection system consists of both the reactor trip system and the engineered safety feature system. The protection system encompasses all electrical and mechanical devices and circuitry (from sensors through the actuating devices) which are required to operate in order to produce the required protective function. Tests of protection systems will be considered acceptable when overlapped if run in parts.

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TS.1-5 REV

2. Protection System Channel A protection system channel is an arrangement of components and modules as required to generate a single protective action signal when required by a unit condition. The channel loses its identity where single action signals are combined.

3. Logic Channel A logic channel is a group of relay contact matrices which operate in response to analog channel signals to generate a protective action signal.

L. Quadrant Power Tilt Quadrant power tilt is the ratio of the maximum quadrant power indicated by an upper excore detector to the average reactor power indicated by the upper excore detectors or the ratio of the maximum quadrant power indicated by a lower excore detector to the average reactor power indicated by the lower excore detectors, whichever is greater. Power is proportional to excore detector current times

its calibration factor. Percentage quadrant power tilt is 100 times the amount the quadrant power tilt ratio exceeds one.

l M. Rated Power I Rated power of a unit is the steady state heat output of 1650 megawatts thermal (MWt) from the reactor core of that unit.

N. Reactor Critical A reactor is critical when the neutron chain reaction is self-sustain-ing and k,gg = 1.0.

O. Refueling Operation Refueling operation of a unit is any operation involving movement of those core components that could af fect the reactivity of the core when the reactor vessel head is unbolted or removed.

] P. Shutdown l 1. Hot Shutdown A reactor is in the hot shutdown condition when the reactor is suberitical by an amount greater than or equal to the margin as specified in Figurg TS.3.10-1 and the reactor coolant average temperature is 547 F or greater.

2. Cold Shutdown A reactor is in the cold shutdown condition when the reactor is suberitical by at least 1% Ak/k and the reactor coolant average temperatare'is less than 200 F.

e a TS.3.3-4 REV

c. Any redundant Valve or damper required for functioning of the containment air cooling system and the containment spray system during and following accident conditions may be inoperable provided it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Prior to initiating repairs, all valves in the system that provide redundancy shall be demonstrated to be operable.

C. Component Cooling Water System

1. Single Unit Operation

a. A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200 F, unless the following conditions are satisfied, except as permitted in Specification 3.3 C.1.b. below.

(1) The two component cooling pumps assigned to that unit are operable.

(2) The two component cooling heat exchangers assigned to that unit are operable.

(3) All valves, interlocks, instrumentation and piping associated with the above components, and required for the functioning of the system during accident conditions, are operable,

b. During startup operation or power operation, any one of the following conditions of inoperability may exist provided startup operation is discontinued until operability is restored.

The reactor shall be placed in the hot shutdown condition if during power operation operability is not restored within the time specified, and it shall be placed in the cold shutdown condition if operability is not restored within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(1) One of the assigned component cooling pumps may be out of service for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) One of the assigned component cooling heat exchangers  ;

may be out of service for a period not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. '

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2. Two-Unit Operation  !

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a. A second reactor shall not be made or maintained critical nor shall it be heated or maintained above 200 F, unless the following conditions are satisfied, except as provided by Specification 3.3 C.2.b. below.

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TS.3.5-1 l REV 3.5 INSTRUMENTATION SYSTEM Applicability Applies to protection system instrumentation.

Objectives To provide for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded, and to delineate the conditions of the reactor trip and engineered safety feature instrumentation necessary to ensure reactor safety.

Specification A. Limiting set points for instrumentation which initiates operation

, of the engineered safety features shall be as stated in Table TS.3.5-1.

B. For on-line testing or in the event of failure of a sub-system instrumentation channel, plant operation shall be permitted to continue at rated power in accordance with Tables TS.3.5-2 through TS.3.5-6.

C. If the number of channels of a particular sub-system in service falls below the limits given in the column entitled Minimum Operable Channels, or if the specified Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in the column titled Operator Action of Tables TS.3.5-2 through TS.3.5-6.

D. In the event of sub-system instrumentation channel failure permitted by Specification 3.5.B. the requirements of Tables TS.3.5-2 through TS.3.5-6 need not be observed during the short period of time the operable sub-system channels are tested where the failed channel must be blocked to prevent unnecessary reactor trip. If the test time exceeds four hours, operation shall be limited according to the requirement shown in the column titled Operator Action of Tables TS.3.5-2 through TS.3.5-6.

Basis Instrumentation has been provided to sense accident conditions and to l initiate reactor trip and operation of the Engineered Safety Features (1). I l

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TS.3.5-3 REV Steam Line Isolation In the event of a steam line beak, the steam line stop valve of the af fected line is automatically isolated to prevent continuous, uncon-trolled steam release from more than one steam generator. The s te am lines are isolated on high containment pressure (Hi-Hi) or high steam line flow in coincidence with low T and safety injection or high steam flow (Hi-Hi) in coincidence wile safety injection. Adequate protection is af forded for breaks inside or outside the containment even when it is assumed that the steam line check valves do not function prcperly.

Ventilation System isolation In the event of a high energy line rupture outside of containment, redundant isolation dampers in certain ventilation ducts are closed (4) .

Safeguards Bus Voltage Relays are provided on buses 15, 16, 25, and 26 to detect loss of voltage and degraded voltage (the voltage level at which safety related equip-ment may not operate properly). On loss of voltage, the automatic voltage restoring scheme is initiated immediately. When degraded voltage is sensed, the voltage restoring scheme is initiated if acceptable voltage is not restored within a short t ime pe riod . This time delay prevents initiation of the voltage restoring scheme when large loads are started and bus voltage momentarily dips below the degraded voltage setpoint.

Limiting Instrument Setpoints

1. The high ccatainment pressure limit is set at about 10% of the maximum Initiation of Safety internal pressurI2) or steam line break (39jection loss of coolant protects accidents as against discussed in the safety analysis.

2. The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at about 30% for initiation of steam line isolation. Initiation of Containment Sprayanf2gteamLineIsolationproto;tsaggjgstlargelossof as discussed in coolant or steam line break accidents the safety analysis.

3. The pressurizer low pressure limit is set substantially below system operating pressure limits. However, it is sufficiently high to protect aggjnst a loss of coolant accident as shown in the safety analysis

4. The steam line low pressure signal is lead / lag compensated and its setpoint is set well above the pressure expected in the event of large steam line break accident as shown in the safety analysis.(g) a

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TS.3.5-4 REV

5. The high steam line flow limit is set at approximately 20% of nominal full-load flow at the no-load pressure and the high-high steam line flow limit is set at approximately 120% cf nominal full-load flow at the full load pressure in order to protect against large steam break accidents. The coincident low T setting limit for steam line isolation initiation is set below"$Es hot shutdown value. The safety analysis shows thct th event of a large steam break.ggy settings provide protection in the

6. The degraded voltage protection setpoint is 90,+ 2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the degraded voltage setpoint. The degraded voltage protection time delay of 6+2 seconds has been shown by testing and analysis to be long enough to allow for voltage dips resulting from the starting of large loeds.

This time delay is also consistent with the maximum time delay assumed in the ECCS analysis for starting of a safety injection pump. A maximum limit on the degraded voltage setpoint has been established to prevent unnecessary actuation of the voltage restor-ing scheme.

The loss of voltage protection setpoint is approximately 55% of nominal 4160 V bus voltage. Relays initiate a rapid (less than two seconds) transfer to an alternte source on loss of voltage.

Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design.

This specification outlines limiting conditions for operation necessary to preserve the ef fectiveness of the Reacter Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with suf ficeint redundancy to provide the capability for channel calibration and test at power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The I source and intermediate range nuclear instrumentation system channels l are not intentionally placed in a tripped mode since these are one-out-of-two trips, and the trips are therefore bypassed during testing. ,

Testing does not trip the system unless a trip condition exists in a '

concurrent channel.

Re ferences (1) FSAR - Section 7.5 (2) FSAR --Section 14.3 (3) FSAR - Section 14.2.5 (4) FSAR - Appendix I l

TABLE TS.3.5-1 ENGINEERED SAFETY FEATURES INITIATION INSTRUMENT LIMITING SET POINTS FUNCTIONAL UNIT CHANNEL LIMITING SET POINTS

• 1 High Containment Pressure (Hi) Safety Injection

• f,4 psig

2. High Containment Pressure (Hi-Hi) a. Containment Spray f,23 psig

b. Steam Line Isolation f17 psig of Both Lines

3. Pressurizer Low Pressure Safety Injection * >1815 psig

4. Low Steam Line Pressure Safety Injection * >500 psic Lead Time Constant >l2 seconde Lag Time Constant "[2 seconds

5. High Steam Flow in a Steam Line Steam Line Isolation d/pcorrespgndingto Coincident with Safety Injection of Affected Line f,0.745 x 10 lb/hr and Low T at 1005 psig avg

.540 F 6 High-high Steam Flow in a Steam Line Isolation f,d/pcorresgonding Steam Line Coincident with of Affected Line to 4.5 x 10 lb/hr Safety Injection at 735 psig gg

< to

7. High Pressure Difference Between Containment Vacuum f,0.5 psi p Shield Building and Containment Breakers g m

8 High Temperature in Ventilation Ducts Ventilation Systa= -<120 F h' Isolation Dampers Y e--

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• Initiates also containment isolation, feedwater line isolation and starting of all cont r nment fans.

• d/p means differential pressure. g O

TABLE TS.3.5 * (continued)

ENGINEERED SAFETY INITIATION INSTRUMENTATION LIMITING SET POINTS

_ FUNCTIONAL UNIT ' CHANNEL LIMITING SET POINTS 9 4KV Safeguards Busses a. Degraded Voltage Voltage Restoration Voltage (% nominal) 90 + 2%

Time Delay 6 + 2 sec

b. Loss of Voltage

1. Voltage (% nominal) 55% + 10%

Time Delay 2 + 2 sec

2. Voltage (% nominal) 90 + 2% $

Time Delay 2 + 2 sec 5$.

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TABLE TS.3.5-6 INSTRUMENT OPERATING CONDITIONS FOR AUXILIARY ELECTRICAL SYSTEM 1 ,

2 3 4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN

_ FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE MET

1. Degraded Voltage 1/ Bus 1/ Bus ---

Place inoperable channel in the tripped 4KV Safeguards Busses condition within one hour or het shutdown.***

.2. a. Loss of voltage 1/ Bus 1/ Bus ---

Place inoperable channel in the tripped 4KV Safeguard condition within one hour or hot shutdown ***

Bus (90%)

b'. Loss of voltage 1/ Bus 1/ Bus ---

Place inoperable channel in the tripped 4KV Safeguard condition within one hour of hot shutdown ***

Bus (55%)

• • • If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken to place the unit in cold shutdown E c:nditions. ,5

< TABLE TS.3.12-1 (Page 1 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Area During No. Location Elevation (A or I) to Remove Shutdown UNIT I AFSH-22 A&B Main and Aux- 773'-4-1/4" A AFSH-36 iliary Steam 745'-7-1/4" A AFSH-39 699'-10-1/4" A AFSH-48 699'-6-1/4"" A MSDH-25 736'-6-7/16" A X MSDH-26 756'-7-1/4" I. X MSDH-29 756'-7-1/4" A MSDH-30 736-6-7/16" A MSH-48 739'-1-11/16" A X MSH-62 A&B 735'-6" A MSH-68 A&B 755'-8" A UNIT II AFSH-2 Main and Auxiliary 749'-4" A AFSH-19 Steam 745'-7-1/4" A AFSH-20 745'-7-1/4" A AFSH-24 745'-6" A AFSH-29 A&B 721'-1-9/16" A AFSH-33 707'-5" A AFSH-39 696'-6-1/4" A AFSH-40 696'-6-1/4" A AFSH-44 750'-7-1/2" A AFSH-46 750'-7" A MSDH-17 739'-0" A X MSDH-18 759'-0" A X MSDH-19 739'-0" A MSDH-20 759'-0" A i

TABE TS.3.12-1 (Page 2 of 8)

REV Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No Location Elevation (A or I) to Remove Shutdown UNIT II MSH-23 Main and Auxiliary 739'-1-3/16" A X MSH-54 A&B Steam 756'-0-1/16" I MSH-81 A&B 735'-9" A M3H-82 A&B 755'-8" A MSH-83 761'-13/16"" I UNIT I RHRRH-5 Safety Injection 723'-4-1/4" I RHRRH-41 698'-11" I RHRRH-58 670'-0" A RHRRH-60 670'-0" A RPCH-160 718'-1/2" I RSIH-92 714'-11" I RSIH-93 714'-11" I RSIH-95 711'-2" I RSIH-96 711'-2" I RSIH-98 701'-2" I RSIH-163 717'-9" I RSIH-167 717'-9" I RSIH-413 A&B 722'-8" A RSIH-414 716'-10" I RSIH-442 717'-9-1/2" I RSIH-469 707'-6-1/2" I RSIH-469 707'-6-1/2" I RSIH-476 707'-1-3/4" I SIH-53 737'-1-3/4" A SIRH-9 737'-0" I SIRH-11 718'-6"" I SIRH-17 730'-0" I SIRH-18 730'-0" I SIRH-22 711'-4" I SIRH-23 A&B- 711'-4" I 1

TABLE TS.3.12-1 (Page 3 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbe rs In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT II RH RH-13 Safety Injection 673'-9" A RH RH-14 674'-0" A RHRH-52 670'-6" A RHRH-54 670'-6" A RHRRH-19 700'-11" I RHRRH-23 711'-2" I RHRRH-28 707'-4" I RSIH-265 699'-9" I RSIH-268 713'-9-3/16" I RSIH-343 719'-8-11/16" I RSIH-349 703'-11" I RSIH-350 703'-11" I RSIH-353 A&B 701'-9" I SIH-43 720'-0" A SIH-49 A33 737'-3" A SIH-53 710'-3" A SIRH-4A 711'-6-1/8" I SIRH-4B 711'-3" I SIRH-7 716'-3-1/16" I SIRH-18 722'-6" I l

l

TABLE TS.3.12-1 (Page 4 of 8)

REV-SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT I RCRH-5 A&B Reactor Coolant 732'-6" I RCRH-12 A&B 720'-7" I RCRH-26 762'-8" I RCRH-27 A&B 761'-7" I RCRH-34 764'-7" I RCRH-45 765'-1" I RERH-46 765'-1" I RCRH-47 745'-10" I RHRRH-15 705'-6" I RHRRH-27 705'-6" I RHRRH-29 A&B 705'-6" I UNIT II RCRH-5 Reactor Coolant 731'-6" 1 RCRH-8 717'-6" 1 RCRH-9 712'-0" I RCRH-14 705'-9" I RCRH-20 715'-7" I RC RH-25 732'-2" I RCRH-26 757'-7" I RCRH-31 764'-1" I RCRR-45 724'-6" I RCRH-46 758'-3" I RCRH-47 760'-3" I RCRH-48 765'-1" 1 RCRH-49 765'-1" I RRCH-279 A&B 724'-9" I RRCH-282 723'-2" I RRCH-284 A&B .725'-8" I RHRRH-2 699'-0" I RHRRH-4 705'-11" I RHRRH-9 705'-11" I

! RHRRH-15 699'-0" I

. . . . _ ~ _

TABLE TS.3.12-1 (Page 5 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location ,

Elevation (A or I) to Remove Shutdown UNIT I CWH-359 Cooling Water 705'-8" A CWH-380 706'-11" A CWH-385 709'-0" A CWH-394 731'-0" A CWH-395 746'-6" A CWH-405 707'-10" A CWH-429 722'-11" A CWH-432 722'-11" A CWH-433 735'-11" A CWH-434 735'-11" A CWH-436 737'-11" .A UNIT II CWH-34 Cooling Water 709'-3" A CWH-35 746'-8" A CWH-39 710'-6" A CWH-40 710'-6" A CWH-44 730'-11" A CWH-45 709'-0" A CWH-49 723'-0" A CWH-50 723'-10" A CWH-52 736'-0" A CWH-54 738'-0" A l

l l

1 l

1 l

1

TABLE TS.3.12-1 (Page 6 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbe rs In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT I AFWH-72 Feedwater 752'-0" I AFWH-82 728'-11" A AFWH-84 728'-11" A UNIT II AFWH-72 A&B Feedwater 706'-3/4" A FWH-72 A&B 751'-0" I~

UNIT I 25.12620.003 3 Steam Generator 760'-9-1/2" I X l 25.12620.003 - 4 760'-9-1/2" I X 25.12620.003 - 5 760'-9-1/2" I X 25.12620.003 - 6 760'-9-1/2" I X 25.12620.003 - 7 760'-9-1/2" I X 25.12620.003 - 8 760'-9-1/2" I X 25.12620-003 - 10 760'-9-1/2" I X 25.12620.003 - 15 760'-9-1/2" I X UNIT II  !

25.12620.003 - 1 760'-9-1/2" I X 25.12620.003 - 2 760'-9-1/2" I X 25.12620.003 - 9 760'-9-1/2" I X 25.12620.003 - 11 760'-9-1/2" I X 25.12620.003 - 12 760'-9-1/2" I X 25.12620.003 - 13 760'-9-1/2" I X 25.12620.003 - 14 760'-9-1/2" I X 25.12620.003 - 16 7A0'-9-1/2" I X UNIT I CVCH-182 Chemical & Vol 707'-6" A RCRH-16 A&B Control 705'-2" I RCRH-19 705'-2" I RCRH-21 705'-7" I RCRH-23 A&B 715'-11" I RCVCH-907 A&B 717'-11" I RCVCH-1293 712'-0" I RPCH-22 703'-1" I RPCH-23 703'-1" I RPCH-121 707'-9" I l

RPCH-139 704'-4" I i RPCH-140 707'-7" I i PRCH-146 714'-7" I I RPCH-147 714'-10" I I WDRH-24 707'-9" I l 1

l i

= .- . ,_

TABLE TS.3.12-1 (Pegs 7 of 8)

REV

. SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In digh Accessible or Especially Radiation Snubber- Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown i UNIT II RCVCH-1396 Chemical & Vol 702'-10" I RCVCH-1505 Control 708'-6" I RCVCH-1513 710'-1" I RCVCH-1524 719'-1" I RCVCH-1574 721'-0" 1 RCVCH-1668 705'-5" I RCVCH-1373 722'-11" I-RCVCH-1389 706'-1" I RRCH-253 704'-4" I RRCH-255 704'-8" I

! RRCH-261 707'-2" I RRCH-288 707'-2" I RRCH-291 704'-6" 1 RRCH-292 704'-7" I UNIT I CCH-304 Comp Cooling 717'-7" A CCH-373 712'-4" A CCH-376 A&B 700'-5" A CCH-377 703'-0" A CCH-378 708'-4" A CCH-380 670'-8" A CCH-381 A&B 671'-4" A CCH-397 699'-3" A CCH-398 A&b 671'-4" A i UNIT II 1 CCH-161 Comp Cooling 717'-7" A CCH-166 719'-11" A CCH-167 720'-0" A CCH-172 720'-0" A CCH-173 708'-5" A CCH-176 705'-3" A CCH-179 A&B 671'-4" A-CCH-180 670'-8" A CCH-181 708'-4" A CCH-182 704'-2" A CCH-185 A&B 671'-4" A CCH-186 670'-10" A UNIT I RCSH-81 Containment Spray 76"'-9" I RCSH-82 760'-8" I RSCH-83 A&B 732'-1" I UNIT II CSH-75 A&B Containment Spray 731'-10" I CSH-76 ~752'-7" I CSH-79 751'-9" I CSH-82 A&B 731'-11" I CSH-83 767'-2" I CSH-84 767'-2" I-CSH-210 698'-0" 1 CSH-215 698'-0"' A CSH-224 710'-6" A

l TABLE TS.3.12-1 (Page 8 of 8)

REV SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)

Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Areas During No. Location Elevation (A or I) to Remove Shutdown UNIT I RRHH-20 RHR 704'-3" A i- RRHH-62 705'-10" A UNIT II CVCRH-6 RHR 711'-0" I RRHH-21 70 '+ '-6" A i

,. - . = , _ e

TABLE TS.4.1-1 (Page 1 of 5)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Functional Response Description Check Calibrate Test Test Remarks

1) Once/ shift when in service

1. Nuclear Power S(1) D(2) M(3) R 2) Heat balance Range M(4) Q(4) M(5) 3) Signal to AT; bistable action M(6) (permissive, rod stop, trips),

M(7) with the exception of the items covered in Emsark #7.

4) Upper and lower chambers for axial of f-set using in-core detectors

5) Simulated signal for testing positive and negative rate bistable action

6) Quadrant Power Tilt Monitor

7) P8 and P10 permissives and the 25% High Flux Low Setpoint Trip.

2. Nuclear Inter- *S(1) NA T(2) R 1) Once/ shift when in service mediate Range 2) Log Level; bistable action (permissive, rod stop, trips)

3. Nuclear Source *S(1) NA T(2) R 1) Once/ shift when in service Range 2) Bistable action (alarm, trips)

4. Reactor Coolant S(1,2) R(1,2,3) M(1) R(1) 1) Overtemperature AT l Temperature M(2) R(2) 2) Overpower AT gjg!

T(3) 3) Control Rod Bank Insertion < b8 Limit Monitor N

5. Reactor Coolant S R M NA y Flow p

6. Pressurizer S R M NA g Water Level 4,

7. Pressurizer S R M NA ,,

Pressure jy 3

TABLE TS.4.1-1 (Page 5 of 5)

Channel Functional Response Description Check Calibrate Test Test Remarks

- 35. Post-Accident Monitor- M NA NA NA Includes all those in FSAR ing Instruments Table 7.7-2.that are not itemized in Table TS.4.1-1.

36. Steam Exclusion W R M NA See FSAR Apendix I,

. Actuation System Section I.14.6.

37. Overpressure NA R M NA Mitigation System

38. Degraded Voltage NA R M NA 4KV Safeguard Busses

39. Loss of Voltage NA R M NA 4KV Safeguard Busses S -

Each Shift D -

Daily W -

Weekly M -

Monthly Q -

Quarterly f s; R -

Each refueling shutdown g en P -

Prior to each startup if not done previous week h T -

Prior to each startup following shutdown in excess of 2 days if not done in the previous 30 days 9

NA -

Not applicable 00 u

See Spec 4.1.D 4 4

I TS.4.2-3 REV The following corrective measures shall be applied:

(a) An evaluation of the ef fect of any corroded area upon the structural integrity of the component shall be performed in accordance with the provisions of Article IS-311 of Section XI Code.

(b) Repairs of corroded areas, if necessary, shall be l performed in accordance with the procedures of 1 Article IS-400 of Section XI code. i 1

1

3. The structural integrity of the primary system boundary shall I be maintained at the level required by the original acceptance I standards throughout the life of the plant. Any evidence as a result of the inspections listed in Table TS.4.2-1 that defects have initiated or grown, shall be investigated, including evaluation of comparable areas of the primary system. In the event further unacceptable structural defects are revealed, all remaining system components or piping in this category shall be examined to the extent specified in that examination category.

C. Records Detailed records of each inspection, including the preoperational base line inspection, shall be maintained to allow comparison and evaluation of future inspections in accordance with Specification 6.6.B.9.

TS.4.4-4 REV B. Emergency Charcoal Filter Systems

1. Periodic tests of the shield building ventilation system shall be performed at quarterly intervals to demonstrate operability. Each redundant train shall be determined to be operable at the time of its periodic test if it meets drawdown performance computed for the test conditions with 75% of the shield building inleakage specified in Figure TS 4.4-1 after ini t iat ion.

2. Periodic tests of the auxiliary building special ventilation system shall be performed at approximately quarterly intervals to demonstrate its operability. Each redundant train shall be determined to be operable at the time of periodic test if it isolates the normal ventilation system and produces a measure-able negative pressure in the ABSVZ within 6 minutes af ter initiation.

3. At least once per operating cycle, or once each 18 months, which eve r comes firs t , tests of the filter units in the Shield Building Ventilation System and the Auxiliary Building Special Ventilation System shall be performed as indicated below:

a. The pressure drop across the combined HEPA filters and the charcoal adsorbers shall be demonstrated to be less than 6 inches of water at system design flow rate

(+10%).

b. The inlet heaters and associated controls for each train shall be determined to be operable,

c. Automatic initiation of each train of each ventilation system.

4. a. The tests of Specification 3.6.E.2 shall be performed at least once per operating cycle, or once every 18 months l whichever occurs first, or af ter every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system ;

operation or following painting, fire or chemical release )

in any ventilation zone communicating with the system that could contaminate the HEPA filters or charcoal ads rbers.

TS.4.5-2 REV

3. Containment Fan Coolers Each fan cooler unit shall be tested during each reactor refueling shutdown to verify proper operation of all essential features including low motor. speed, cooling water valves, and normal ventila-tion system dampers. Individual unit performance will be monitored by observing the tenninal temperatures of the fan coil unit and measuring cooling water flow rate.

4. Component Cooling Water System

a. System tests shall be performed during each reactor refueling shutdown. Operation of the system will be initiated by tripping the actuation instrumentation.

b. The test will be considered satisfactory if control board indica-tion and visual observations indicate that all components have operated satisfactorily.

5. Cooling Water System

a. System tests shall be performed at each refueling shutdown. Tests shall consist of an automatic start of each' diesel engine and automatic operation of valves required to mitigate accidents including those valves that isolate non-essential equipment from the system. Operation of the system will be initiated by a simulated accident signal to the actuation instrumentation.

The tests will be considered satisfactory if control board indication and visual observations indicate that all components have operated satisfactorily and if cooling water flow paths required for accident mitigation have been established.

b. Each- diesel engine shall be inspected at each refueling shutdown.

B. Component Tests

1. Pumps

a. The safety injection. pumps, residual heat removal pumps and con-containment spray pumps shall be started and operated at intervals of one month. Acceptable levels of performance shall-be that the pumps start and reach their required developed head on minimum recirculation flow and the control board inlications and visual observations indicate that the pumps are operating properly for at least 15 minutes.

b. A test consisting of a manually-initiated start of each diesel

- engine, and assumption of load within ane minute, shall be con-ducted monthly.

, o a .

TS.4.5-3A REV

h. Following completion of high head safety injection system or RHR system modifications that alter system flow characteristics a flow balance test shall be performed during shutdown to confirm the following injection flow rates are achieved:

1. High Head Safety Injection System:

(a) Flow through all four injection lines plus miniflow shall not exceed 835 gpm with one pump in operation.

(b) The minimum flow ekrough loop A

• B cold legs shall be 670 gpm with one pump in opera ion. The flow rates in each leg shall be within 20 gp of each other with one pump in operation.

(c) Flow orifices and throttling valves will be used to limit and balance flow through the reactor vessel injection lines to a maximum of the total flow limit in Specification 4.5.B.3.h.1.(a) above, with one pump in operation. During this flow test the flow rates in each leg shall be within 50 gpm of each other.

2. ,RHR System:

The minimum flow through each RHR Reactor Vessel Injection line shall be at least 1800 gpm.

Basis The Safety Injection System and the Containment Spray System are principal plant Safety Systems that are normally inoperative during reactor operation.

Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes containment isolation and a Containment Spray System test requires the system to be temporarily disabled.

The method of assuring operability of these systems is therefore to combine systems tes ts to be performed during refueling shutdowns, with more frequent component tests which can be performed during reactor operation.

The systems tests demonstrate proper automatic operation of the Safety Injection and Containment Spray Systems. With the pumps blocked from starting, a test signal is applied to initiate automatic action and verific-tion made that the components receive the safety injection in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.

,e. .

4 TS.4.6-1 REV 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Applicability Applies to periodic testing and surveillance requirements of the emergency power system.

}

Objective i

To verify that the emergency power sources and equipment are operable.

Specification The following tests and surveillance shall be performed:

A. Diesel Generators

, 1. At least once each month, for each diesel generator:

a. Verify the fuel level in the day and engine-mounted tank.

b. Verify the fuel level in the fuel storage tank.

c. Verify that a sample of diesel fuel from the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water, and sediment. .

d. Verify the fuel transfer pump can be started and transfers )

fuel from the storage system to the day tank.

e. Verify the diesel starts from the normal standby condition

f. Verify the generator synchronizes, is loaded to at least 1375 kw, and operated for at least one hour.

2. At least once each 18 months:

a. Subject each diesel generator to a thorough inspection in accordance with procedures prepared in conjunction with the manufacturer's recommendations for this class of standby service.

b. For each unit,- simulate a loss of of fsite power in con-junction with a safety injection signal, and:

1. Verify de-energization of the emergency busses and load shedding from the emergency busses.

2. Verify the diesels start from the normal standby con-

! dition on the auto-start signal and energize the emergency busses in one minute.

3. Verify that. the diesel generator system trips, except those for engine overspeed, ground fault, and generator differential current, are automatically bypassed. l

4. Verify that the auto-connected loads do not exceed 3000 kw.

c. Verify the capability of each generator to operate at least one hour while loaded to 3000 kw.

, d. . Verify the capability of each generator to reject a load of at least 650 kw without tripping.

e. During this test, operation of the emergency lighting system shall be' ascertained.

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

e .

TS.6.1-1 REV 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION A. The Plant Manager has the overall full-time onsite responsibility for safe operation of the facility. During periods when the Plant Manager is unavailable, he may delegate this responsibiliity to other qualified supervisory personnel.

B. The Northern States Power corporate organizational structure relating to the operation of this plant is shown on Figure TS.6.1-1.

C. The functional organization for operation of the plant shall be as shown in Figure TS.6.1-2 and:

1. Each on duty shif t shall be composed of at least the minimum shif t crew composition shown on Table TS.6.1-1,

2. For each reactor that contains fuel: a licensed operator in the control room.

3. At least two licensed operators shall be present in the control room during a reactor startup, a scheduled reactor shutdown, and during recovery from a reactor trip. These operators' are in addition to those required for the other eactor.

4. An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor.

5. All refueling operations shall be directly supe vised by a licensed Senior Reactor Operator of Senior Reactor Operator Limited to Fuel Handling who has no other concurrent re s pons-ibilities during this operation.

6. A fire brigsde of at least five members shall be maintained on site at all times.* The fire brigade shall not include the six members of the minimum shif t crew for safe shutdown of the  :

I reactors.

l

• Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected I absence of Fire Brigade members provided immediation action is taken to l restore the Fire Brigade to within the minimum requirements. i

+ .

FIGURE TS.6.1-1 REV PPESIDENT EXECUTIVE VICE PRESIDENT

• VICE PRESIDENT- VICE PRESIDENT-PLANT ENGINEERING AND CONSTRUCTION

^

GENERAL MANAGERS HEADQUARTERS NUCLEAR PLANTS NUCLEAR SERVICES

^

PLANT MANAGERS NUCLEAR PLANTS l SERVICES I i I l g SAC ADMINISTRATION I

l n

! SAFETY AUDIT GENERAL SUPERINTENDENT ~

I COMMITTEE (SAC)

OPERATIONAL QA l

I I

_ AUDIT & j REVIEW FIGURE 6.1-1 NSP CORPORATE ORGANIZATIONAL RELATIONSHIP TO ON-SITE OPERATING ORGANIZATION

• Has the responsibility for the fire protection program

PLANT m

~

OPERATIONS MANAGER

• COMMITTEE PLANT SUPERINTENDENT SUPT. PLANT SUPERINTENDENT OPERATIONS & QUALITY ENGINEERING &

MAINTENANCE *+ ENGINEERING

• RADIATION PROTECTION *(LS0)

SUPPORT FOR QA SUPERVISOR OF AND QC FUNCTION SECURITY & SERVICES I I I r SUPT. OF SUPT. OF TRAINING SUPT. SUPT. SUPT. SENIOR MAINTENANCE

• OPERATIONS SUPERVISOR RADIATION TECHNICAL OPERATIONS NUCLEAR

• (LS0) , (OPERATOR AND PROTECTION

• ENGINEERING

• ENGINEERING

• ENGINEER

• I FIRE BRIGADE TRAINING)

SHIFT SUPER-VISOR (LS0) l MECHANICAL LEAD PLANT EQUIP -

TECHNICAL TECHNICAL EECHNICAL TECHNICAL

& MENT & REACTOR SUPPORT & SUPPORT & 3UPPORT & SUPPORT &

ELECTRICAL OPERATOR (LO) RADIATION ENGINEERS FOR ENGINEERS FOR ENGINEERS MAINTENANCE PROTECTION INSTRUMENTS, JPERATION, FOR NUCLEAR GROUP SPECIALIST., CONTROLS, & 4AINTENANCE, ENGINEERING gy PLANT EQUIPMENT

& REACTOR COMPUTER; SURVEILLANCE, <g OPERATOR (S)(LO)

INSTRUMENT & & TESTING g CONTROLS SPEC g FIGURE 6.1-2 .

ASSISTANT PLANT ~

~

EQUIPMENT PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATOR (S) h AND FUNCTIONAL ORGANIZATION FOR ON-SITE OPERATING GROUP ANDN

• Key Supervisor FIRE BRIGADE LO Licensed Operator LSO Licensed Senior Operator (AS REQUIRED)

+ Has responsibility for implementation of the fire protection program

TS 6.2-1 REV 6.2 Review and Audit Organizational units for the reviaw and audit of facility operations shall be constituted and have the responsibiities and authorities outlined below:

A. Safety Audit Committee (SAC)

The Safety Audit Committee provides the independent review of plant c;erations from a nuclear safety standpoint. Audits of plant opera-tion are conducted under the cognizance of the SAC.

1. Membership

a. The SAC shall consist of at least five (5) persons.

b. The SAC chairman shall be an NSP representative, not having line responsibility for plant operation, appointed by the Vice President - Power Production. Other SAC members shall be appointed by the Vice President - Power Production or by such other person as he may designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

c. No more than two members of the SAC shall be from groups holding line responsibility for operation of the plant.

d. A SAC member may appoint an alternate to serve in his absence, with concurrence of the Chairman. No more than one alternate shall serve on the SAC at any one time. The alternate member shall have voting rights.

2. Qualifications

a. The SAC members should collectively have the capability required to review activities in the following areas:

nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other. appropriate fields associated with the unique characteristice of the nuclear power plant.

e e TS.6.2-3 REV

f. Investigation of all events which are required by reguia- I tion or technical specifications (Appendix A) to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

g. Revisions to the Facility Emergency Plan, Facility Security Plan, and the Fire Protection Program,

h. Operations Committee minutes to determine if matters con-sidered by that Committee involve unreviewed or unresolved safety questions.

i. Other nuclear safety matters referrd to the SAC by the Operations Committee, plant management or company manage-ment.

1

j. 'All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety' related structures systems, or components.

k. Reports of special inspections and audits conducted in accor-dance with specification 6.3.

6. Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to assure safe facility operation.

a. Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years. The audits shall be performed in accordance with appropriate written instructions and procedures,

b. Periodic review sf the audit program should be performed by the SAC at least twice a year to assure its adequacy,

c. Written reports of the audits shall be reviewed by the Vice President - Power Production, by the SAC at a scheduled meeting, and by members of management having responsibility in the areas audited.

7. Authority The SAC shall be advisory to the Vice President - Power Production.

8. Records

Minutes shall be ' prepared and retained for all scheduled meetings of the Safety Audit Committee. The minutes shall be distributed within one month of the meeting to the Vice President - Power Production, the General Manager Nuclear Plants, each member of the SAC and others designated by the Chairman. There shall be a formal' approval of the minutes.

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TS.6.2-5 REV B. Operations Committee (OC)

1. Membership The Operations Committee shall consist of at least six (6) members drawn from the key supervisors of the onsite staf f.

The Plant Manager shall serve as Chairman of the OC and shall appoint a Vice Chairman from the OC membership to act in his absence.

2. Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3. Quorum A majority of the permanent members, including t!.e Chairman or Vice Chairman

4. Responsibilities - The following subjects shall be reviewed by the Operations Committee:

a. Proposed tests and experiments and their results,

b. Modifications to plant systems or equipment as described in the Final Safety Analysis Repott and having nuclear safety significance or which involve an unreviewed safety question as defined in Paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.

c. Proposals which would effect permanent changes to normal and emergency operating procedures and any other proposed change i r procedures that will affect nuclear safety as detetmit.ed by the Plant Manager,

d. Proposed changes to the Technical Specifications or operating licenses.

e. All reported or suspected violations of Technical Specifica-tions, operating license requirements, administrative procedures, operating procedures. -Results of investigations, including evaluation and recommendations to prevent recurrence will be reported in writing to the General Manager Nuclear Plants and to the Chairman of the Safety Audit Committee.

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TS.6.2-6 REV

f. All events which are required by regulations or Technical Specifications to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

g. Drills on emergency procedures (including plant evacuation) and adequacy of communication with offsite support groups.

h. All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan, shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.

i. Special reviews and investigations. as requested by the Safety Audit Committee.

5. Authority The OC shall be advisory to the Plant Manager. In the event of l a disagreement between the recommendations of the OC and the Plant Manager, the course detrained by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the General Manager Nuclear Plants and the Chairman of the SAC for review.

6. Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Manager 1 Nuclear Plants and others designated by the OC Chairman or l Vice Chairman. ,

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7. Procedures A written charter for the OC shall be prepared that contains: 1 l

a. Repoonsibility and authority of the group

b. Cc ent and method of submission of presentations to the Operations Committee

c. Mechanism for scheduling meetings

d. Provision for meeting agenda i

TS.6.4-1 REV

'6. 4 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED If a Safety Limit is exceeded, the reactor shall be shut down immediately. An immediate report shall be made to the Commission and to the General Manager Nuclear Plants, or his designated alternate in his absence. A complete analysis of the circum-stances leading up to and resulting from the situation, together with recommendations by the Operations Committee, shall also be l prepared. This report shall be submitted to the Commission, to the General Manager Nuclear Plants and the Chairman of the Safety Audit Committee within 14 days of the occurrence.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

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TS.6.5-2 REV

1. a. Paragraph 20.203 " Caution signs, labels, signals and controls". In lieu of the " Control device" or alarm signal required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Pe rmit (or continuous escort by a qualified person for the purpose of making a radiation survey) and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. The above procedure shall also apply to each high radiation area in which the intensity of- radiation is greater than 1000 mrem /hr, except that doors shall be locked or attended to prevent unauthorized entry into these areas and the

> keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager, 1

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