Text

Improved Technical Specifications, Volume 20, Chapter 5.0
	ML18344A286
Person / Time
Site:	Palisades
Issue date:	01/20/1998
From:	; Consumers Energy Co, (Formerly Consumers Power Co)
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18344A286 (133)
	v • d • e

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I I

I IMPROVED I

TIECHNICAL

. I

SPECIFICATIONS

I I

ALI SADES UC LEAR LANT I -, I

~------- _J Volume 20 CHAPTER 5.0 t:ons11msrs Energy

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PALISADES NUCLEAR PLANT CONSUMERS ENERGY Docket 50-255

~ Conversion to Improved Technical Specifications License DPR-20 INTRODUCTION: CHAPTER 5.0. ADMINISTRATIVE CONTROLS A. ARRANGEMENl'--AND CONTENT OF THIS CHAPTER OF THE CHANGE REQUEST This Chapter of the Technical Specification Change Request (TSCR) proposes lhanges to those Palisades Technical Specifications addressing Chapter 5.0, "ADMINISTRATIVE CONTROLS." These changes are intended to result in requirements which are appropriate for the Palisades Nuclear Plant, but _closely emulate those of the Standard Technical Specifications, Combustion Engineering Plants, NUREG 1432, Revision 1, Chapter 5.0.

This discussion and its supporting information frequently refer to three sets of Technical Specifications, and to two groups of discussions associated with the proposed changes; the following abbreviations are used for clarity and brevity:

CTS - The Palisades Current Technical Specifications, ITS - The Palisades Improved Technical Specifications, ISTS - NUREG 1432, Revision 1.

DOC - Discussions of Change; these discussions explain and justify the differences between the requirements of CTS and ITS .

JFD - Justifications for Deviation; these discussions explain the differences between the requiremertts of the ITS and the ISTS.

Six attachments are provided to assist the reviewer:

1. Proposed ITS Chapter 5.0 pages

2. This Attachment is not applicable to Chapter 5.0

3. A set of all those CTS pages which contain requirements associated with those in ITS Chapter 5.0, marked up to show the changes from CTS to ITS, and arranged by specification in the order in which the requirements occur in ITS. This attachment also includes a DOC for each change.

Each change from CTS to ITS is classified in the following categories:

ADMINISTRATIVE - A change which is editorial in nature, which only involves movement of requirements within the TS without affecting their technical content, or tlarifies CTS requirements.

MORE RESTRICTIVE - A change which only adds new requirements, or which revised an existing requirement resulting in additional operational restrictions. -

RELOCATED - A change which only moves requirements~ not meeting the

-* 10 CFR 50.36(c)(2)(ii) criteria, from the CTS to the Operating Requirements Manual (which has been included in the FSAR by reference).

1

INTRODUCTION: CHAPTER 5.0. ADMINISTRATIVE CONTROLj ,

A. ARRANGEMENT AND CONTENT OF THIS CHAPTER OF THE CHANGE REQUEST (continued)

LESS RESTRICTIVE - REMOVAL OF DETAIL {LA) - A change in which certain details from otherwise retained Specifications are removed from the ITS and placed in the Bases, FSAR, or other license~ controlled documents.

LESS ~STRICTIVE - A change which deletes any existing requirement, or which revises any existing requirement resulting in reduced operational restrictions.

4. No Significant Hazards Analyses for the changes from CTS to ITS.

An individual No Significant Hazards Analysis is provided for each Less Restrictive change; generic No Significant Hazards Analyses are_ provided for each of the other categories of change.

5. ISTS Chapter 5.0 marked to show the differences between ISTS and ITS.

6. JFDs for the differences between ISTS and ITS.

B. REFERENCE DOCUMENTS This Chapter of the TSCR is based on the following reference documents:

1. CTS as revised through Amendment 178. *

2. The following TSCRs which are currently under review by the NRC:

a. Administrative Controls, initially submitted on December 12, 1995.

b. Electrical, initially submitted on December 27, 1995.

c. Containment, submitted on March 26, 1997.

d.. PCP Flywheel, initially submitted on January 18, 1996.

3. ISTS, as revised by Industry Generic Changes (TSTF) approved as of October 15, 1997.

4. The following changes to ISTS which are currently under review by the NRC:

a. TSTF 52 .

2

- INTRODUCTION: CHAPTER 5.0. ADMINISTRATIVE CONTROLS '

C. THE UNIQUE PALISADES NUCLEAR PLANT FEATURES AFFECTING THIS CHAPT~R Palisades has several physical, analytical,_ and administrative features which*

differ from those newer CE plants upon which the ISTS were based. Palisades_was the first CE plant designed and built. Its design and licensing preceded the issuance of the General Design Criteria so that, in some aspects, its physical systems ar~ not like those of newer plants; its Technical Specifications preceded the issuance of Standard Technical Specifications (STS) so that LCOs, Actions, and Surveillance Requirements are not coordinated as they would be for a STS plant.

Palis~des has purchased all its core reloads from Siemens Power Corporati6n *(or its predecessors), therefore, reload analyses and the associated core physics parameters, as well as certain Safety Analyses are not like those plants using all CE fuel and analyses as were modeled in the ISTS.

D. THE DIFFERENCES BETWEEN CTS "OPERATING CONDITIONS" AND ITS "MODES" The CTS definitions of plant operating conditions have been replaced with the*

operation Mode definitions used in ISTS. In several instances the name for a CTS defined "operating.condition" is the same as that for an ISTS "Mode," but the definition differs~

CTS contain the following definitions for operating conditions:

1.

2.

The POWER OPERATION condition shall be when the reactor is critical and the neutron flux po~er range instrumentation indicates greater than 2% of RATED POWER. .

The HOT STANDBY condition shall be when Tave is greater than 525°F and any of the CONTROL RODS are withdrawn anq the neutron flux power range instrumentation indicates less than 2% of RATED POWER.

3. The HOT SHUTDOWN condition shall be when the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Speci fi ca ti on 3 .10 and Tave is greater than 525°F.

4. The COLD SHUTDOWN condition shall be when the primary coolant is at SHUTDOWN BORON CONCENTRATION and Tave is 1ess than 210°F.

5. The REFUELING SHUTDOWN condition shall be when the primary coolant is at REFUELING BORON CONCENTRATION and Tave is 1ess than 210°F .

3

INTRODUCTION: CHAPTER 5.0, ADMINISTRATIVE CONTROLS'

D.. THE DIFFERENCES BETWEEN CTS "OPERATING CONDITIONS" AND ITS ITS contain the following definition table for Modes:

11 MODES~

1

% RATED AVERAGE PRIMARY REACTIVITY THERMAL COOLANT MOOE TITLE CONDITION POWER (a) TEMPERATURE (keff) ("F) 1 Power Operation ;, 0.99 > 5 NA 2 Startup ;, 0.99 ~ 5 NA 3 Hot Standby < 0.99 NA ;, 300*

4 Hot Shutdown (*l < 0.99 NA 300 > T.,. > 200 5 Cold Shutdown(*) < 0.99 NA ~ 200 6 Refue 1i ng(c) NA NA NA

E.

(a)

(b)

(c)

Excluding decay heat.

All reactor vessel head closure bolts fully tensioned.

One or more reactor vessel head closure bolts less than fully tensioned.

MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES"

1. CTS "Power Operation" is essentially equivalent to ITS "MODE 1." Each represents a condition with the reactor critical and the turbine generator in operation. The only effective difference is the power level which separates that Condition or Mode from the next lower one. During plant startup, the plant must meet all CTS "Power Operation" or ITS "MODE 1" LCOs before the turbine generator is placed on the line; similarly, during plant shutdown, the plant exits CTS "Power Operation" or ITS "MODE 1" when the turbine generator is no longer in service. Therefore, this change in definition will have no operational effect. *

2. CTS "Hot Standby" is similar to ITS "MODE 2." Each represents a condition with the reactor critical, or nearly so, and the turbine generator shut down.

During plant startup, the plant must meet all CTS "Hot Standby" or ITS "MODE 2" LCOs before a reactor startup is started; during plant shutdown, the plant exits CTS "Hot Standby" or ITS "MODE 2" when the reactor is shutdown.

CTS action statements requiring that the plant be placed in "Hot Standby" are effectively equivalent to ITS Actions requiring the plant be placed in "MODE 2." Therefore, this change in definition will have no operational effect.

4

INTRODUCTION: CHAPTER 5.0, ADMINISTRATIVE CONTROL~'

E. MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES" 3.

~ (continued)

CTS "Hot Shutdown" and ITS "MODE 3" are similar at their upper temperature boundary. During plant shutdown, the plant exits CTS "Hot Standby or ITS 11 "MODE 2 when the reactor is shutdown. CTS action statements requiring that 11 the plant be placed in Hot Shutdown" are effectively equivalent to ITS 11 Actions requiring the plant be placed in MODE 3." CTS Hot Shutdown" and 11 11 ITS "MODE 3" are quite different at their lower.temperature boundary; CTS "Hot Shutdown" is exited when Tave drops below 525°F, ITS MODE 3" is not 11 exited until Tave drops below 300°F.

4. CTS does not provide a defined term for the condition when Tave is between 525°F and 210°F (the upper bound for CTS "told Shutdown").

5. CTS "Cold Shutdown" is essentially equivalent to ITS "MOOE 5." Each represents a condition with Tave below boiling. There is no techni ca 1 significance to the difference between the CTS 210°F and the ITS 200°F.

CTS action statements requiring that the plant be placed in "Cold Shutdown 11 are effectively equivalent to ITS Actions requiring the plant be placed in 11 MODE 5." Therefore, this change in definition will have no operational effect. * *

6. CTS "Refueling Shutdown" is essentially equivalent to ITS "MODE 6. Each, 11 when taken with other definitions and LCO requirements, represents a condition with the reactor at least 5% shutdown. Therefore, this change in definition will have no operational effect.

F. THE MAJOR CHANGES FROM CTS (as modified by g~nding TSCRs) TO ITS

1. The Safety Function Determination Program, 5.5.13, has been added to ITS Chapter 5.0, but does not appear in CTS Chapter 6.0. The Safety Function Determination Program is a feature of ISTS which supports LCO 3.0.6 (which has also been added to ITS) in addressing support system operability.

G. THE MAJOR DIFFERENCES BETWEEN ITS AND ISTS

1. The Explosive Gas and Storage Tank Radioactivity Monitoring Program, ISTS 5.5.12, was omitted from the ITS. Palisades CTS contain no equivalent requirements.

2. The Fuel Oil Testing Program, ITS 5.5.11, contains different testing and sampling requirements than its ISTS counterpart. These differences are necessary because the Palisades Fuel Oil Storage Tank serves several components besides the Diesel Generators, and the turn over rate for the fuel oil is quite high. This results in different limiting conditions than a tank where fuel is stored for a long time.

3. The Containment Leak Rate Testing Program, ITS 5.5.14, has been revised to allow use of 10 CFR 50, Appendix J, Option B, performance based testing.

5

ATTACHMENT 1 PALISADES NUCLEAR PLANT

CHAPTER 5.0, ADMINISTRATIVE CONTROLS PROPOSED TECHNICAL SPECIFICATIONS

Responsibility 5.1

5.0 ADMINISTRATIVE CONTROLS

5.1 Responsibility 5.1.1 The plant superintendent shall be responsible for overall plant operation and shall delegate in writing the succession for this responsibility during his absence.

The plant superintendent or his designee shall approve, prior to implementation, each proposed test, experiment or modification to?

systems or equipment that affect nuclear safety.

5.1.2 The Shift Supervisor (SS) shall be responsible for the control room command function. During any absence of the SS from the control room while the plant is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor* operator (SRO) license shall be designated to assume the control room command function.

During any absence of the SS from the control room while the plant is in MODE 5 or 6 an individual with an active SRO license or

Reactor Operator (RO) license shall be designated to assume the control room command function .

Palisades Nuclear Plant 5.0-1 Amendment No. 01/20/98

Organization

.j 5.Z

5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the Palisades plant.

a. Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented, and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key positions, or in equivalent forms of documentation.

These requirements and the plant specific equivalent of those titles referred to in these Technical Specifications shall be documented in the FSAR.

b. The plant superintendent shall be responsible for overall plant safe operation and shall have control over those*

onsite activities necessary for safe operation and maintenance of the plant.

c. A specified corporate executive shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out radiation safety and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures .

Palisades Nuclear Plant 5.0-2 Amendment No. 01/20/98

Organization 5.2 5.2 Organization 5.2.2 Plant Staff

a. A non-licensed operator shall be assigned when fuel is in the reactor and an additional non-licensed operator shall be assigned when the reactor is operating in MODES l, 2, 3,

or 4.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the plant is in MODES 1, 2, 3, or 4, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. *

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i), and 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order

to accommodate unexpected absence of on-duty shift crew.

members provided immediate action is taken to restore the shift crew composition to within the requirements.

d. A radiation safety technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, .

provided immediate action is taken to fi 11 the required position. *

e. Administrative procedures shall be developed and implemented to limit the working houts of. plant staff who perform safety-related functions (e.g., licensed SROs, licensed ROs, radiation safety personnel, auxiliary operators, and key maintenance personnel).

In the event that overtime is used! the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;

2. An individual should not be permitted to work more than 16-hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24.

hours in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift .

Palisades Nuclear Plant 5.0-3 Amendment No. 01/20/98

Organization 5.2

5.2 Organization 5.2.2 Plant Staff (continued)

Any deviations from the overtime guidelines shall be authorized in advance by the plant supertntendent or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the plant superintendent or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

f. The operations manager or an assistant operations manager shall hold an SRO license. The individual holding the SRO license shall be responsible for directing the activities of the licensed operators.

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant

analysis with regard to the safe operation of the plant. If either SRO on shift satisfies the Shift Engineer qualification requirements, then the STA does not need to be stationed. *

Palisades Nuclear Plant 5.0-4 Amendment No. 01/20/98

Plant Staff Qualifications 5.3 5.a ADMINISTRATIVE CONTROLS 5.3 Plant Staff Qualifications 5.3.1 Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.

5.3.2 The radiation safety manager shall meet the qualifications of a Radiation Protection Manager as defined in Regulatory Guide 1.8, September 1975. For the purpose of this section, "Equivalent,". a~

utilized in Regulatory Guide 1.8 for the bachelor's degree requirement, may be met with four years of any one or combination of the following: (a) Formal schooling in science or engineering, or (b) operational or technical experience and training in nuclear power.

5.3.3 The Shift Technical Advisor shall have a bachelor's degree or equivalent and the Shift Engineer shall have a bachelor's degree in a scientific or engineering discipline. Specific.training for both the Shift Technical Advisor and the Shift Engineer shall include plant design, operations, and response and analysis of the.

plant for transients and accidents. The Shift Engineer shall hold a Senior Reactor Operator license.

The plant staff who perform reviews which ensure compliance with 5.3.4 10 CFR 50.59 shall meet or exceed the minimum qualifications of ANS ~.1-1987, Section 4.7.1 and 4.7.2. A Senior Reactor Operator license or certification shall be considered equivalent to a bachelors degree for the purpose of this specification *

Palisades Nuclear Plant 5.0-5 Amendment No. 01/20/98

Procedures 5.4

5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a-. The applicable procedures recommended in of Regulatory Guide 1.33, Revision 2, February 1978.

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Site Fire Protection Program implementation.

d. All programs specified in Specification 5.5 .

Palisades Nuclear Plant 5.0-6 Amendment No. 01/20/98

Programs and Manuals 5.5

5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained:

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from , ,

radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain (1) the radioactive effluent contra 1s a-nd radio 1ogi ca 1 environmental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Release Report required by Specification 5.6.2. and Specification 5.6.3.

c. Changes to ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective after approval by the plant superintendent.

3. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of -

or concurrent with the Radioactive Effluent Release Report for the period of the reprirt in ~hich any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date

Palisades Nuclear Plant (e.g., month/year) the change was implemented .

5.0-7 Amendment No. 01/20/98

Programs and Manuals 5.5

. 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage to the engineered safeguards rooms, from those portions of systems outside containment that could contain highly radioactive fluids Quring a serious transient or accident, to as low as practical.

The systems include the Containment Spray System, the Safety Injection System, the Shutdown Cooling System, and the containment sump suction piping. This program shall include the following:*

a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

c. The portion of the shutdown cooling system that is outside the containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig.

d. Piping from valves CV-3029 and CV-3030 to the discharge of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig. *

e. The maximum allowable leakage from the recirculation heat removal systems* components (which include valve stems, flanges and pump seals) shall not exceed 0.2 gallon per minute under the normal hydrostatic head from the SIRW tank (approximately 44 psig).

5.5.3 Post Accident Sampling Program This program provides controls which will ensure the capability to accurately determine the airborne iodine concentration in vital areas and which will ensure the capability to obtain and analyze reattor coolant, radioactive gases and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:

a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analytic equipment .

Palisades Nuclear Plant 5.0-8 Amendment No. 01/20/98

Programs and Manuals 5.5

-r.

5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial action~ t0 be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10 times the value in 10 CFR 20, Appendix B, Table 2, Column 2.

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with

d.

the methodology and parameters in the ODCM, Limitation on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each plant to unrestricted areas conforming to 10 CFR 50, Appendix I,

e. Limitations ~n the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 times the value listed in 10 CFR 20, Appendix B, Table 2, Column 1.

f. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,

g. Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents rel~ased from each plant to areas beyond the site boundary conforming to 10 CFR 50, Appendix I, *

Palisades Nuclear Plant 5.0-9 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

h. Limitations on the annual doses or dose commitment to any member of the public due to releases of radioactivity arid to radiation from uranium fuel cycle sources conforming to 40 CFR 190. -

5.5.5 Containment Structural Integrity Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Structural Integrity Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE and IWL.

If, as a result of a tendon inspection, corrective retensioning of five percent (8) or more of -the total number of dome tendons is necessary to restore their liftoff forces to within the limits, . a dome delamination inspection shall be performed within 90 days following such corrective retensioning. The results of this

-inspection shall be reported to the NRC in accordance with Specification 5.6.7, "Containment Structural Integrity Surveillance Report."

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Containment Structural Integrity Surveillance Program inspection frequencies.

5.5.6 Primary Coolant Pump Flywheel Surveillance Program

a. Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each _10 years.

b. The provisions of SR 3.0.2 are applicable to the Flywheel Testing Program .

Palisades Nuclear Plant 5.0-10 Amendment No. 01/20/98

Programs and Manuals

.~ 5.5

5.5 Programs and Manuals 5.5.7 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda (B&PV Code) as follows:

B&PV Code terminology Required interval for inservice testing for performing inservice activities testing activities*

Weekly ::: 7 days Monthly ::: 31 days Quarterly or every 3 months ::: 92 days Semiannually or every 6 months ::: 184 days Every 9 months ::: 276 days Yearly or annually ::: 366 days Biennially or every 2 years ::: 731 days The provisions of SR 3.0.2 are applicable ta the above b*

required intervals for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice

testing activities; and

d. Nothing in the B&PV Code shall be construed to supersede the requirements of anj Technical Specification.

5.5.8 Steam Generator Tube Surveillance Program This program provides controls for surveillance testing of the Steam Generator (SG) tubes to ensure that the structural integrity of this portion of the Primary Coolant System (PCS) is maintained.

The program shall contain controls to erisure:

a. Steam Generator Tube Sample Selection and Inspection The inservice inspection may be limited to one SG on a rotating schedule encompassing 6% of the tubes if the results of previous inspections indicate that both SGs are performing in a like manner. If the operating conditions in one SG are found to be more severe than those in the other SG, the sample sequence shall be modified to inspect the most severe conditions .

Palisades Nuclear Plant 5.0-11 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.8 Steam Generator Tube Surveillance Program

a. Steam Generator Tube Sample Selection and Inspection (continued)

The SG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.5.8-1. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all SGs; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

2. The first sample of tubes selected for each inservice inspection of each SG shall include:

a) All nonplugged tubes that previously had detectable wall penetrations greater than 20% .

b) Tubes in those areas where experience has indicated potential problems.

c) A tube inspection shall be performed on each sele~ted tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

3. The tub~s selected as t~e second and third samples (if required by Table 5.5.8-1) during each inservice inspection may be subjected to a partial tube inspection provided:

a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

b) The inspections include those portions of the tubes where imperfections were previously found .

Palisades Nuclear Plant 5.0-12 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.8 Steam Generator Tube Surveillance Program (continued)

4. The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected

tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage

b.

calculations .

Inspection Frequencies The above required inservice inspection of SG tubes shall be performed at the following frequencies:

1. Inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspections results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

2. If the results of the inservice inspection of a SG conducted in accordance with Table 5.5.8-1 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.8.b.1; the interval may

Palisades Nuclear Plant then be extended to a maximum of once per 40 months.

5.0-13 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.8 Steam Generator Tube Surveillance Program (continued)

3. Additional, unscheduled inservice inspections shall be performed on each SG in accordance with the first sample irispection specified in Table 5.5.8-1 during the shutdown subsequent to any of the following conditions:

a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of LCO 3.4.13.

b) A seismic occurrence greater than the Operating Basis Earthquake.

c) . A loss-of-coolant accident resulting in initiation of flow of the engineered safeguards.

d) A main steam line or main feedwater line break.

c. Acceptance Criteria

1. As used in this Specification:

a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

c) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

d) % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

e) Defect means an imperfection of such severity that it exceeds*the plugging limit. A tube containing a defect is defective .

Palisades Nuclear Plant 5.0-14 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.8 Steam Generator Tube Surveillance Program (cont1nued) f) Plugging Limit mea~s the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

g) Unserviceable described the condition of a tube ..

if it leaks or contains a defect large enough* to affect its structural integrity in the event of

an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 5.5.8.b.3, above.

h) Tube Inspection means an inspection of the SG tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

I) Preservice Inspection means an inspection of the full length of each tube in SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This

inspection shall be performed after the shop hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

2. The SG shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 5.5.8-1.

The provisions of SR 3.0.2 are applicable to the Steam Generator Tube Surveillance Program *

Palisades Nuclear Plant 5.0-15 Amendment No. 01/20/98

Programs and Manuals 5.5

lST SAMPLE INSPECTION 1 TABLE 5.5.8-1 STEAM GENERATOR TUBE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Result Action Required Result Action Required Result Action Required C-1 None N/A N/A N/A N/A C-2 Plug defective tubes and inspect additional 2S tubes in this SG. C-1 None N/A N/A C-3 Inspect all tubes C-2 Plug defective C-1 None in this SG, plug tubes and inspect defective tubes and additional 4S tubes C-2 Plug defective inspect 2S tubes in in this SG. tubes each other SG. Perform action for C-3 C-3 result of first Sample C-3 Perform action for N/A N/A C-3 result of first Sample All None N/A N/A

other SGs are C-1 Some SGs Perform action for C-2 but no other C-2 result of second sample N/A N/A SG is C-3 Other SG Inspect all tubes N/A N/A is C-3 each SG and plug defective tubes NOTES: 1 The m1n1mum sample size for the first sample inspection is S tubes per SG where S=(6/n)%, where n is the number of steam generators inspected during an inspection .

Palisades Nuclear Plant 5.0-16 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.9 Secondary Water Chemistry Program A program shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:

a. Identification of a sampling schedule for the ~ritical variables and control points for these variables,

b. Identification of the procedures used to measure the values of the critical variables,

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, ct~ Procedures, for the recording and management of data,

e. Procedures defining corrective actions for all off-control point chemistry conditions, and

f. A procedure identifying (a) the authority responsible for the inte~pretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.

5.5.10 Ventilation Filter Testing Program A program shall be established to implement the following required testing of Control Room V~ntilation (CRV) and Fuel Handling Area Ventilation systems at the frequencies specified in Regulatory Guide 1.52~ Revision 2 (RG 1.52), and in accordance with RG 1.52 and ASME N510-1989, at the system flowrates and tolerances specified below*:

a. Demonstrate for each of the ventilation systems that an inplace test of the High Efficiency Particulate Air (HEPA) filters shows a penetration and system bypass < 0.05% for the CRV and< 1.00% for the Fuel Handling Area Ventilation System when tested in accordance with RG 1.52 and ASME N510-1989:

Ventilation System Flowrate (CFM)

' V-8A or V-88 7300 +/- 20%

V-8A and V-88 10,000 +/- 20%

V-95 or V-96 12,500 +/- 10%

Palisades Nuclear Plant 5.0-17 Amendment No. 01/20/98

Programs and Manuals 5.5

5.. 5 Programs and Manua 1s 5.5.10 Ventilation Filter Testing Program (continued)

b. Demonstrate for each of the ventilation systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% for the CRV and < 1.00% for the Fuel Handling Area Ventilation System when tested in accordance with RG 1.52 and ASME N510-1989.

Ventilation System Flowrate (CFM)

V-8A and V-88 10,000 +/- 20%

V-26A and V-268 3200 +10% -5%

c. Demonstrate for each of the ventilation systems that a laboratory test of a sample of the charcoal adsorber,' when obtained as described in RG 1.52 shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of

~ 30°C and equal to the relative humidity specified as follows:

Ventilation System Penetration Relative Humidity VF-66 6.00% 95%

VFC-26A and VFC-268 0 .157% 70%

d. For each of the ventilation systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with RG 1.52 and ASME N510-1989:

Ventilation System Del ta P (In H2Ql Flowrate (CFM)

V-8A and V-88 6.0 10,000 +/- 20%

VF-26A and VF-268 8.0 3200 +10% -5%

e. Demonstrate that the heaters for the CRV system dissipates the following specified value +/- 20% when tested in accordance with ASME N510-1989:

Ventilation System Wattage VHX~26A and VHX-268 15 kW

Palisades Nuclear Plant 5.0-18 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (continued)

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program frequencies.

Should the 720-hour limitation on charcoal adsorber operation occur during a plant operation requiring the use of the charcoal adsorber - sue~ as refueling - testing may be delayed until the completion of the plant operation or up td 1,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of filter operation~ whichever occurs first.

5.5.11 Fuel Oil Testing Program A fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The

- program shall include sampling requirements, testing requirements, and acceptance criteria, based on the diesel manufacturer's specifications and applicable ASTM Standards. The program shall establish the following:

a. Acceptability of new fuel _oil prior to addition to the Fuel

Oil Storage Tank, and acceptability of fuel oil stored in the Fuel Oil Storage Tank, by determining that the fuel oil has the following properties within limits:

1. API gravity or an absolute specific gravity,

2. Kinematic viscosity~ and

3. Water and sediment content.

b. Other properties of fuel oil stored in the Fuel Oil Storage Tank, specified by the diesel manufacturers or specified for grade 2D fuel oil in ASTM D 975, are within limits.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Fuel Oil Testing Program .

Palisades Nuclear Plant 5.0-19 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.12 Te.chni cal Specifi cations (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 5.5.12.b. above shall be reviewed and approved by the NRC prior to implementation. Changes to the aases implemented

5.5.13 without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.?l(e).

Safety Functions Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actiMs taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provi~ions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; b~ Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and Palisades Nuclear Plant 5.0-20 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.13 Safety Functions Determination Program (SFDP) (continued)

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a

.loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.14 Containment Leak Rate Testing Program Programs shall be established to implement the leak rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The Type A test program shall meet the requirements of 10 CFR 50, Appendix J, Option B and shall be in accordance with the guide 1i nes of Regulatory Gui de 1.163, 11 Performance-Based Containment Leakage-Test Program, dated September 1995. 11 The Type B and Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A, as modified by the exemption from certain requirements of 10 CFR 50 Appendix J which was granted in an NRC letter to Consumers Power Company dated December 6, 1989.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 52.64 psig.

The maximum allowable containment leak rate, La, at Pa, shall be 0.1% of containment air weight per day.

Local leak rate tests, other than Personnel Airlock doors between the seals tests, shall be performed at ~ 55 psig.

Palisades Nuclear Plant 5.0-21 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.14 Containment Leak Rate Testing Program (continued)

Lo~al leak rate tests for checking airlock doors seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed as follows:

a-. A between the sea 1s test sha 11 be performed on the Personne 1 Airlock at ~ 10 psig.

b. A full pressure test shall be performed on the Emergency Escape.Airlock at~ 55 psig. A seal contact check shall be performed on the Emergency Escape Airlock following each full pressure test. Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing.

Leak rate acceptance criteria are:

a. Containment leak rate acceptance criteria is s 1.0 La.

During the first plant startup following testing in accordance with this program, the leak rate acceptance criteria are s 0.60 La for the Type B and Type C tests and s 0.75 La for Type A tests;

b. The leakage for a Personnel Airlock door seal test shall not exceed 0. 023 La.

c. An acceptable Emergency Escape Airlock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces.

Containment OPERABILITY is equivalent to 11 Containment Integrity 11 -

for the purposes of the air lock testing requirements in 10 CFR 50, Appendix J.

The provisions of SR 3.0.2 are not applicable to the Containment

Leak Rate Testing Program requirements.

The provisions *of SR 3.0.3 are applicable to the Containment Leak Rate Testing Program requirements .

Palisades Nuclear Plant 5.0-22 Amendment No. 01/20/98

Programs and Manuals 5.5

5.5 Programs and Manuals 5.5.15 Process Control Program

a. The Process Control Program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and Stat~ regulations, and other requirements governing the disposal of the radioactive waste.

b. Changes to the Process Control Program:

1. Shall be documented and records of reviews performed shall be retained as required by the Quality Program, CPC-2A. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluation justifying the change(s) and b) A determination that the change will maintain

2.

the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

Shall become effective after approval by the plant superintendent .

Palisades Nuclear Plant 5.0-23 Amendment No. 01/20/98

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS

.* . 5.6 - Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occupational Radiation Exposure Report This report shall include a tabulation on an annual basis of the number of stations, utility and other personnel (including contractors), for whom monitoring was performed, receiving an .

annual deep dose equivalent greater than 100 mrem and the associated deep dose equivalent (repo~ted in person-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year .

5.6.2 Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the plant during the previous calendar year shall be submitted before May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results*

of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering operation of the plant in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the .objectives outlined in the Offsite Dose Calculation Manual and Process Control Program, and shall be in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1 .

Palisades Nuclear Plant 5.0-24 Amendment No. 01/20/98

Reporting Requirements 5.6

5.6 Reporting Requirements 5.6.4 Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC no later than the fifteenth of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

3.1.7 Regulating Rod Group Position Limits 3.2.1 Linear Heat Rate Limits 3.2.2 Radial Peaking Factor Limits 3.2.4 AS! Limits

b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the

following documents:

1. XN-75-27(A), "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," and Supplements l(A),

2(A), 3(P)(A), 4(P)(A), and 5(P)(A); Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, &3.2.4)

2. ANF-84-73(P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," and Appehdix B(P)(A) and Supplements l(P)(A), 2(P)(A); Advanced Nuclear Fuels Corporation.

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

3. XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.

(LCOs 3.2.1, 3.2.2, &3.2.4)

4. ANF-84-093(P)(A), "Steamline Break Methodology for PWRs," and Supplement l(P)(A); Advanced Nuclear Fuels Corporation. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

5. XN-75-32(P)(A), "Computational Procedure for

Evaluating Fuel Rod Bowing," and Supplements l(P)(A),

2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

- Palisades Nuclear Plant (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4) 5.0-25 Amendment No. 01/20/98

Reporting Requirements 5.6

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. EXEM PWR Large Break LOCA Model as defined by:

(LCOs 3.1.6, 3.2.1, &3.2.2) a) XN-NF-82-20(A), "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates,"

and Supplements l(P)(A), 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

b) XN-NF-82-07(P)(A), "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company.

c) XN-NF-81-58(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," and Supplements l(P)(A), 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

d) XN-NF-85-16(A), "PWR 17x17 Fuel Cooling Tests Program," Volume 1 and Supplements l(P)(A),

2(P)(A), and 3(P)(A), and Volume 2 and Supplement l(P)(A); Exxon Nuclear Company .

e) XN-NF-85-105(A), "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," and Supplement l(P)(A); Exxon Nuclear Company.

7. XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors,"

.. Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, &3.2.2)

8. ANF-1224(P)(A), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," and Supplement l(P)(A); Advanced Nuclear Fuels Corporation. (LCOs 3.2.1, 3.2.2, & 3.2.4)

9. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

10. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.

(LCOs 3.2.1, 3.2.2, &3.2.4)

Palisades Nuclear Plant 5.0-26 Amendment No. 01/20/98

Reporting Requirements 5.6

5.6 Reporting Requirements 5.6.5 CORE° OPERATING LIMITS REPORT (COLR) (continued)

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

d. The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC.

5.6.6. Post Accident Monitoring Report When a report is required by LCO 3.3.7, "Post Accident Monitoring Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status .

5.6.7 Containment Structural Integrity Surveillance Report Reports shall be submitted to the NRC covering Prestressing, Anchorage, and Dome Delamination tests within 90 days after completion of the tests.

5.6.8 Steam Generator Tube Surveillance Report The following reports shall be submitted to the Commission following each inservice inspection of steam generator tubes:

a. The number of tubes plugged in each steam generator shall be reported to the Commission within 15 days following the completion of each inspection, and

Palisades Nuclear Plant 5.0-27 Amendment No. 01/20/98

Reporting Requirements 5.6

5.6 Reporting Requirements 5.6.8 Steam Generator Tube Surveillance Report (continued)

b. The complete results of the steam generator tube inservice inspection shall be reported to the Commission within 12 months following completion of the inspection~ This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections that fall into Category C-3 shall require 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal notification to the NRC prior to resumption of plant operation. A written followup within the next 30 days shall provide a description of investigations and corrective measures taken to prevent recurrence .

- Palisades Nuclear Plant 5.0-28 Amendment No. 01/20/98

High Radiation Area 5.7

5.0 ADMINISTRATIVE CONTROLS

5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in l~eu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

~ 100 mrem/hr but< 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g.i radiation safety technicians). or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates < 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area .

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who 1s responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Request .

Palisades Nuclear Plant* 5.0-29 Amendment No. 01/20/98

High Radiation Area 5.7

5.7 High Radiation Area 5.7.2 In addition to the requirements of Specification 5.7.1, except as allowed by 5.7.3, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or radiation safety supervision. Doors shall remain lock~d except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas an1 the maximum allowable stay times for individuals in those areas.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be maae by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 For individual high radiation areas with radiation levels of

~ 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a

warning device .

Palisades Nuclear Plant 5.0-30 Amendment No. 01/20/98

ATTACHMENT 2 PALISADES NUCLEAR PLANT

CHAPTER 5.0, ADMINISTRATIVE CONTROLS PROPOSED BASES (NIA for CHAPTER 5.0)

ATTACHMENT 3 PALISADES NUCLEAR PLANT CHAPTER 5.0, ADMINISTRATIVE CONTROLS

CTSMARKUP AND DISCUSSION OF CHANGES

4.4 Deleted 4.5 CONTAINMENT TESTS 4.5.l Integrated Leakage Rate Tests The containment integrated leak rate testing *shall be performed in accordance with the Containment Leak Rate Testing Program.

4.5.2 Local Leak Detection Tests

a. Test (1) Local leak rate tests, other than Personnel Airlock ~o~rs between the seals tests, shall be performed at ~ 55 psig.

(2) Local leak rate tests for checking airlock door seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed as. follows:

5,5,IL{ (a) A between the seals test shall be performed on the Personnel Airlock at ~ 10 psig.

(b) A full pressure test shall be performed on the Emergency Escape Airlock at ~ 55 psig. A seal contact check shall be performed on the Emergency Escape Airlock following each full pressure test. Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing.

soap (4) Th local leak rate shall f llowing components:

Containment pene ations that employ re gaskets, *sealan compounds, or bellows (b) Air lock and quipment door seals.

(c)

(d) Isolatio valve~ on the testable penetra ng the containment.

(e) Other containment components orde to meet the acceptanc lea rate test.

4-19 Amendment No. t, ~. +a-&, ~. 177

4.5 CONTAINMENT TESTS

4.5.2 Local Leak Detection Tests (continued)

b. Acceotance Criteria (1) The total leakage from all penetrations and isolation valves shall not exceed 0.60 La.

(2) The leakage for a Personnel ~irlock door seal test shall not exceed 0.023 La.

(3) An acceptable Emergency Escape Airlock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces.

c. Corrective Action (1) If at any time it is ~etermined that 0.60 La is exceeded, repairs shall be in' iated immediately. If repairs are not completed and conf rmance to the acceptance criterion of 4.5.2.b(l) is not ~emonstrated within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the plant shall be placed

at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in LO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(2) If at any tim it is determined that total containment leakag exceeds L., thin one hour action shall be initiated to plac the plant i at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months d in COLD SH DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(3) If the P sonnel airlock door seal leakage is greater th 0.023 L , or if the Emergency Escape Lock door seal con ct

check ails to meet its acceptance criterion, repairs all be initi ed inmediately to restore the door seal to the acce ance criteria of specification 4.5.2.b(2) or 4 5.2.b(3).

In e event repairs cannot be completed within 7 ys, the pl nt shall be placed in at least HOT SHUTDOWN wi in the next 6 ours and in COLD SHUTDOWN wit~in the followin 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f air lock door seal leakage results in one d r causing total containment leakage to exceed 0.60 La, e door shall be declared inoperable and the remaining OPERAB door shall be i11111ediately locked closed* and tested withi 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . As long as the remaining door is found to be OPERA LE, the provisions of 4.5.2.c(2) do not apply. Repairs shal be initiated

i11111ediately to establish conformance wi specification 4.5.2.b(l). In the event conformance this specification cannot be established within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months he plant shall be placed in at least HOT SHUTDOWN with' the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following hours.

Entry and exit is permissible thro h a "locked" air lock door to perform repairs on the. affected a* lock components.

4-20 CONTAINMENT TSCR G~aR§e 7, REV 2 - Amendment No. H-i, +74, W,

4.5 CONTAINMENT TESTS 4.5.4 Surveillance or Prestressing ystem

s:o a.

b.*

Tendon inspection shall be accomplis a at five-year intervals for the life of the plant. The sche ed inspection* date~ for all subsequent inspections may be ried by not more than plus or mi one year from the base sch The ~urveillance ten e.

s shall be randomly but selected from eac the following groups:

1. of 4 dome tendons including dome ten group.

inimum of 4 vertical tendons.

minimum of 5 hoop tendons.

For each inspection, the tendons s l be selected rin a random basis except that those tendons ose routing has been modified to clear penetrations shall be luded from the sample.

I During each tendon ins ion, the following field testing shal e I performed:

1. Lift-off r. dings shall be taken for each of the tendon The tests shall include the following I10*£ One tendon, randomly selected from e group of tendons during each inspection, shall be s Jected to essentially complete detensioning t~ identi broken or damaged wires .

2.

(b) The simultaneous measur ent of elongation and jacking force during retensi ing shall be made at a minimum of three approximate equally spaced levels of force between the se ng force and zero.

While the ten is in the detensioned state, e checked for continuity.

in the

3. Three res, one from each of a vertical, a op and a dome te n will be removed and identified for 'nspection. At each ccessive surveillance, the wires will e selected from

different tendons. Each of the insp tion wires removed will be visually inspected for evidenc f corrosion or other deleterious eff~cts 'nd samples aken for laboratory testing.

4. The sheathing filler shall e inspected visually for color and coverage and samples sh be obtained for laboratory testi

5. Tendon anchorage h aware such as bearing plates, str~ ng washers, shims buttonheads shall be visually ins* cted for evidence of c rosion or other deleterious effect 4-2la Amendment No. -14, ~. ~. -7+, ~. 174 October 31, 1996

f~l 1 ~q

4.5 CONTAINMENT TESTS 4.5.4 Surveillance for Prestressin (continued)-

d. Following the field testing shall be

1. Three thr inspection wires remo e middle). One additi wire determined by fi 1aboratory 1 be cut from each of the.

(one from each end and one rom specimen shall be cut fro visual inspection to have he greatest amount of rrosion. Each of the wire shall be tested for u mate strength, yield elongation.

2. The sh ing filler samples shall b of eac endon examined. Vertical t on samples shall be taken f m the lower end. Samples s 1 be thoroughly mixed and analyzed for reserve alkali

y, water content, and concentration of water ble chlorides, nitrates, and sulfides. Analyses s 11 be performed in accordance 1th the procedures and wit

the acceptance 1 imits speci 'ed in ASME Code Section XI able IWL-2525-1.

Procedure hall be established to minim' and to assure at the volume of sheathing f' er removed has been rep ed upon completion of the in ection and amounts cumented.

1. The average of all asured tendon forces for each t e of

tendon shall be ual to or greater than the mini m required prestress lev , of 584 kips per tendon for d 615 kips tendon for hoop and vertical measur force in each individual tendo tha % of the predicted force, or tendons and, dons. The all not be less the measured force in not ore than one tendon is 903 and 95% of the pre cted force, and (b) The measured fo s in two tendons located adjac the tendon i a) above are not less than 95%*

rces, and (c) the easured forces in all the remain' tendons e not less than 95% of the predi d f measured force in any tenaon i ess than 903 of its predicted force, the tendon sh be completely detensioned and a determination shall be ade as to the cause of such an occurrence and corrective ction shall be taken. In addition, all such tendons shall ve their forces measured as

additional tendons

the next scheduled inspection period.

The Commission s be notified in accordance with Paragraph 4.5.4f.

4-21b Amendment No. 14, ~. -l-G-9, 174*

October 31, 1996

-* f~ Lf of ?-9

4.5 4.5.4 (continued)

3.

Inspection wires s l indicate no signifi by corrosion or tting.

Tensile test pecimens cut from insp tion wires shall be tested for timate strength. Fai re at less than 11.7 of any o of the test samples r uires the Commission e notifi in accordance with sp ification 4.5.4f.

section

4. Te on anchorage hardware all be free of signi cant rrosion, pitting, crac or other deleteriou effects.

f. I any element of the pre ressing system fails o meet the cceptance criteria of .5.4e., the reporting revisions of 10 CFR 50.73 shall apply.

a. examination shall b performed on the end chorage concrete rface at the surveil nee tindon anchor poi s for signs of crac ng, popouts, spalli , or corrosion. Coner cracks havin widths greater than .010 shall be evaluat and documented.

b. T end anchorage caner. e surveillance inspe e the same as tendo surveillance interval

1. Crack w* ths shall be measured y using' optical compar ors or wi feeler gauge. Movem s shall be measured bY.

dem ntable mechanical ext someters.

oncrete anchorage are are acceptable if no c cracks are wider than 0.010 nches and no signs of n deterioration sine the previous inspection re Concrete surf conditions exceeding t se stated in 4.5.5c.2 above shall evaluated fo~ the effe on tendon and containme structural integrity. e results of evaluation shall b ncluded in the final s eillance report.

Dome Delamination Surveillance If, as a result of a pres tressing system inspection WJ,J;~__,,,.......,.,.........___ __,

_ corrective retensioning of five percent (8) or more of the total number of dome tenggn~ i~ lli!'~~~iCY t.a cestore their liftoff forces to within the limits Gi.f Siec;iJcafinn 4)\ 4} a ome e amination inspection shall

_ be performed within 90 days following such corrective retensioning. The results of this inspection shall be reported to the NRC.,t .

1n o..c.cordo..nc..c. w~ Sfl:cr~ict..-\ion 5.&./, ' Coct+C).*nl'llr.rit Srrtic.+uf"G L Tf\+chr,-#Y 'Sc.1rll,dl o.nc-c. 4-2lc 0 -~***

l'GPon Amendment No. -14, ~, ~. 174 October 31, 1996 5 c{ ~q

P°1"-

rrs S.o S'. \

6,g 6.1 A()MlNISTBATivE CONTRQLS RESPQNSIBILITy SJ,\ 6. 1.1 Tht plant superintendent shall be responsible for overall plant operation and shall delegate in writing the succession for this responsibility during his absence.

The plant superintendent or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect ear safety.

s.1.1- 6.1.2 Tht Shift Supervisor (SS) shall be resp rri~ I. 1., ~ or-I./

t for tht control room r;:;;:.,

\:.:::::,J co11111and function. 0 in n of the SS from the control room while the plant is a WN an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control roa11 c011111nd function. Ourt!IQ~l\Y absence of th* SS from the control roOll while th* phnt is in@JJ!l~iln~ an individual w;th an 1ctht SRO l1c:11nse or Buctor Operator (RO) license sh111 bt designated to 1ssU11e the.control room corr111and function.

~---

S".l... 6.2 ORGANIZATION S.£...\ 6.2.l Onsite and Offsit1 Organizatfgns Onsitt and offsitt organizations shall bt established for plant operation and corporate management, r1sp1ctiv1ly. The onsit1 and offsitt organizations shall include tht positions for activities affecting the safety of the Palisades plant.

a. Lints of authority, responsibility and co1111Unication shall be establishtd and defined for the highest managemtnt levels through inttl'lltdiatt levels to and including all operating organization positions. Thtst relationships shall bt documented, and updated, as appropriate, in the fon11 of organization charts, functional descriptions of dtpartmtntal responsibilities and relationships, and job descriptions for key positions, or in equivalent forms of docu.ntatton. Thtst requirements and the plant specific equivalent of those titles referred to in these Technical Speciftcattons shill bt docwnenttd tn tht FSAR.

b. Tht plant suptrtnttndtnt shall bt responsible for overall plant saft operation and shall havt control over those onsitt activities n1c1ss11"1 for safe operation and 111tnt1n1nc1 of tht plant.

C* A spectf1td corporate 1x1cuttv1 shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff tn operating,

.. tntatntftC) and providing technical support to the plant to ensure nucltar safety.

d. Tht indtvtduals who tratn the operating staff and those who carry out radiation safety and quality assurance functions may report to tht 1pproprt1t1 onsttt manager; however, thty shall have sufficient organtzattonal frttdOll to ensure thttr ind1p1ndtnc1 fro* operating pressures.

6-1 Amtndmtnt No. iii, 1:1, ~. -t;9, 174 October 31, 1996 f~ ~ o-J z_q

6.0 ADMINISTRATIVE CONTROLS 6.2.2 Plant Staff

a. be assigned

\,/141'4 Fut.L. 1s 1t1T'..i"-

e c reactor

~~:..:;:..:.7?i~'!!:!/""-i=F~~::r-~-=a~l~n~o.:.:,n-~l:..,:i-::c.:::.e"l.insed- op er ator sh a11 be I@

(\J_,_lol_1._l'l_SJ-.-,... room

r actor is operating .

- DE IJ 2. J 3, e)lll, * @

b. At least one licensed Reactor Operator (RO) shall be present in the

control room w e Uel is in the reactor. In addition, while the ~

u 0 S W , at least one 1 icensed Senior Reactor A.'2.-

perator ( ) shall be present in the control room. A.~ .

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i}, and 6.2.2.a and 6.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of

on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the requirements.

d. A radiation safety technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Administrative procedures shall be developed and implemented to limit the working hours of plant staff who perform safety-related functions (e.g., licensed SROs, licensed ROs, radiation safety personnel, auxiliary operators, and key maintenance personnel).

-In the event that overtime is used, the following guidelines shall

be f o11 owed:

1.

2.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time; An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work*

periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

6-2 Amendment No. i, 3+; 69, ~. +-S, .w&, 2+, 3-9, ffi, 174

s.o 5.Q ApMINISTBAIIyE CONTROLS S. ?... Z. . 6.2.2.e Plant Staff (Continued)

Any deviations from the overtime guidelines shall be authorized in advance by the plant superintendent or his designee, in accordance with approved administrativ~ procedures, or by higher levels of m1n1ge1111nt, 1n accordance with established procedures and with documentat1on of the basis for granting the d1v11tion.

Controls shall be included in the procedures such that individual overt11111 shall be reviewed monthly by the plant superintendent or h1s d1s1gn1e to ensure that excessive hours have not been assigned.

Rout1n1 deviation from the above gu1d111n1s is not authorized.

f. The optrat1ons manager or an assistant operations manager shall hold an SRO license. The 1nd1v1dual holding the SRO license shall be responsible for d1rect1ng the act1v1t1es of the licensed operators.

g. The Sh1ft Technical Advisor (STA) shall provide advisory technical support to the Sh1ft Supervisor (SS) 1n the areas of thermal hydraulics, reactor eny1~~g, and plant analysis w1th regard to the safe operation of h *. If either SRO on shift satisfies the Shift Engineer qua11f1cat1on r1quir1111nts, then the STA does not need to be stationed *

$", 5 S", 3. \

S.3.L..

6.3 6.3.l 6.3.2 PLANT STAFF QUALIFICATIONS Each mtllblr of the plant staff shall meet or exceed the minimum qualif1cat1ons of ANSI NlS.1*1971 for comparable pos1t1ons.

The radiation safety manager shall meet the qua11ficat1ons of a Radiation Protection Manager as dtf1ntd in Regulatory Gulde 1.8, Stptlllblr 1975. For the purpose of thts section, "Equivalent,* as utilized 1n Regulatory Gutde 1.8 for the bachelor's degree requirement, 11ay be ..t with four years of any one or colllb1nat1on of the following:

(a) For111l schooling in sc1enc1 or eng1ntering, or (b) op1r1t1onal or technical 1xp1r11nc1 and training 1n nuclear power. -

6.3.3 The Shift Technical Advisor shall have a bachelor's degree or equivalent and the Shtft Engineer shall have a bachelor's degr11 1n a sc1ent1fic or tntin11riDCJ discipline. Specific training for both the Shift Technical Advisor and thl Shift Engineer shall tncludt plant design, operations, and response and analysts of the plant for transients and accidents.

The Shift Engin11r shall hold a Sen1or Reactor Operator 11censt.

s."l:i.Y 6.3.4 Th* plant staff who perfor11 rtv1ews which ensure compliance w1th 10 CFR 50.59 shall Dttt or exceed tht *~illUll qualifications of ANS 3.1*1987, Stctton 4.7.1 and 4.7.2. A Senior Reactor Operator ltctnst or certification shall bt cons1dtred equivalent to a bachelors degr11 for tht purpose of th1s sptcificat1on.

6-3 Allltndlllnt No. ~. ~. ii, i-1-, ii, .;.i, WI, W, H-9, 174 October 31, 1996

~ % af?-q

s.O 6.0 ADMINISTRATIVE CONTROLS

s.4 6.4 PROCEDURES s.....\ 6.4.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below: . ~ .

a. The applicable procedures recommended in Ap endix " of Regulatory Guide 1.33, Revision 2, Appendix A, February

b. Refueli g o~erations.

.c. Surv illance and test equi ment .

@)(d). Site Fire Protection Program implementation.

@* All programs specified in Specification 6.5.

f. mentation.

g.

b. ""Th.a_ e.."'"'~j~""c.\ ""f..&.r"-..W"') '?~c~J.~,.e.S ('e..\~:r~J. .\...

i~~\e. ...... e.~+ ~ nt~re.""'e......-\-s .. ~ NLL~(..-c:n11 a.."J @'

t--11..L\2...E(::.-0"111 1

$._fP\e. ....... ~ .....

+I 1 c:.S s+.._~.,t 1'"' G~e.-:c.... L&e...- 8£..-?J, 6-4

Amendment No. i, ~. 6-9, +i, ~. W, +64, +74, P~ 9 of l.9

S". 0 6.0 ADMINISTRATIVE CONTROLS s.s 6.5 PROGRAMS AND MANUALS The following programs shall be established, implemented, and maintained:

s.s.1 6.5.1 Offsjte Dose Calculation Manual CODCMl

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain (1) the radioactive effluent controls and radiological environmental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Rel ea~ Report required by Speci fi cation , 6. 2. and \ @

Specification 6.3. * @)

, s

c. Changes to ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a. . Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective aft~r approval by the plant superintendent.

3. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change.to the ODCM was made. Each change shall be identified by markings in the margin.of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

6-5 Amendment No. 8&, &4, Par lo

~*

aft.'f

s,o s.s: z.

6.9 6.5.2 AQMINIST13ATIYE CQNTBOLS Pr1mary Coglant Sgyrces Oytsjde Contajnment.

Th1s progr1111 prov1des controls to minimize Jeakage to the engineered safeguards rooms, from those portions of systems outs1de containment that could contain highly radioactive fluids during a serious transient or acc1dent, to as low as practical. The systems include the Containment Spray System, the Safety Inject1on System, the Shutdown Cooling System, and the containment sump suction piping. This program shall include the following:

a. Provisions establishing preventive maintenance and periodic visual inspection requ1rernents, and

b. Integrated leak test requirements for each syste* at a frequency not to exc1td refueling cycle intervals. *

c. The portion of the shutdown cooling system that is outside the cont1in111nt shall bt tested either by use in not111al operation or hydrostatically tested at 255 psig.

d. Piping fro* valves CV-3029 and CV-3030 to the discharge of the safety injection PY11$JS and containment spray pumps shall be hydrostattcally tested at no less than 100 psig.

e. The *aximum allowable leakage from the recirculation heat removal syst111s' components {which include valve stems, flanges and pump seals) shall not exc11d 0.2 gallon per minute under th* normal hydrostatic head fro*-th1 SIRW tank {approximately 44 psig).

S".S.3 6.5.3 post Accident Sl!l!l!lfnq program This progr .. provides controls which will ensure the capability to accurately d1t1n1tn1 the airborne iodine concentration in vital areas and which will 1nsur1 tht capability to obtain and analyze reactor coolant radio1cttv1 1 and particulates in plant gaseous ~

1ff u1nts, and contain..n atmosphere samples under accident conditions.

Thts progr.. shall tncludt the following:

a. Traintnt of personnel,

b. Proctdurt1 for s1111>>ltng and analysis, and

c. PT'ovtsion1 for .. 1nt1n1nc1 of sam;>ling and analytic equip1111nt.

6-6 Allltndmlnt No. '1-, ~. 17 4 October 31, 1996 Pa.r- ll Ot2.9

s.o 6.0 ADMINISTRATIVE CONTROLS 6.5.4 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR S0.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: fu",ti'~e..~:l; ~

a. Limitations on the *1* of radioactive liquid and gaseous ~

monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, .

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10 times the value in 10 CFR 20, Appendix B, Table 2, Column 2.

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CF~302 and with the 1\.j\

methodology and parameters in the ODCM, ~ ~

d. Limitation on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid

e.

effluents released from eac~to unrestricted areas conforming to 10 CFR 50, Appendix I, ~

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 times the value listed in 10 CFR 20, Appendix B, Table 2, Column 1.

~

f. Limitations on the annual and quarterly air dose~ resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,

g. Limitations on the annual and quarterly doses to a member of the public from lodine-131, lodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,

h. Limitations on the annual doses or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.

6-7

-~*

Amendment No. -l-S-4, 74, f~ IZ o~2.g

6.0 ADMINISTRATIVE CONTROLS

{.(.( 6.5.S \R~et\iedl C,,1>1T'1J> 1,.J tJf t:.1'tr

< AbD * ~ ~ror---

~rt.Jc:nJn. Al I NuRE.b- 14162.)

,../TtH., ur/ 5'1.1'-v"&.1. .o.N'c' ~o~._.....,

.~ /").<

6.5~6 Primary Coolant Pump Flywheel Syrveillance Program

a. Surveillance of the primary coolant pump flywheels shall conJist of a 100% volumetric inspection of the upper flywheels each 10 years.

b. The provisions of Surveillance Requirement 4.0.2 are applicable to the Flywheel Testing Program.

{'. $'. '1 6.5.7 .. lnservice ll:§iectiOJr' antr::ting Program . .)-"' @

This program provides controls for inserv1ce dnijiectHln an"Q) testing of ASME Code Class 1, 2, and 3 components including applicable supports.

The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel C~de and applicable Addenda (B&PV Code) as follows:

B&PV Code terminology Required interval for inservice testing for performing inservice activities testing activities Weekly s 7 days Monthly s 31 days Quarterly or every 3 months s 92 days Semiannually or every 6 months s 184 days Every 9 months s 276 days Yearly or annually s 366 days Biennially or every 2 years s 731 days ~

b. The provisions -of(S1g;ieilla"* egu:re f

(it?. f'o~ 4.,@are applicable A.!

to the above required intervals for performing inservice testing activities; . ~ lj~~ @

c. The.provi~ions of.CSJ!rve2t#(Re4 (tp?'. <plJ)are applicable to 1nserv1ce testing act1v1t1es; an .

d. Nothing in the B&PV Code shall be construed to supersede the requirements of any Technical Specification.

6-8 Amendment No. -!-74, .}7;,

for- I~ of 2.q

6.0 A[)MINISTBATIYE CONTROLS s.s.S 6.5.8 Steam Generator Tybe Syryeillance Program This progra~ provides controls for surveillance testing of the Steam Generttor (SG) tubes to ensure that the structural integrity of this portion of th* Primary Coolant System (PCS) is maintained. The program shall contain controls to ensure:

a. Steam Generator Tybe Sample Selection and Inspection The ins1rvic1 inspection may bt limited to one SG on a rotating schedule encompassing 6% of the tubes if the results of previous inspections indicate that both SGs are perfonning in a like manne If tht op1rat i ng conditions in ant .SG art found to be more severe than those in the other SG, the sample sequence shall be modified to inspect tht 11ast severe conditions.

The SG tube minimu* sample size, inspection result classification, and t'!!.corrtsponding action required shall bt as specified in 15'Tabl11f)S.8-l. Tht tubes selected for each instrvice inspection

~shall include at least 3' of the total number of tubes in all SGs; the tubes s1l1ct1d for these inspections shall be selected on a randa11 basis except:

1. Where 1xp1r1lnc1 in similar plants with similar water .

ch..1stry indicates critical areas to be inspected, then at least SOS of tht tubes inspected shall bt from these critical areas.

2. Tht first sample of tubes selected for each inservice inspection of each SG shall include:

i) All nonplugged tubes that previously had detectable wall penetrations greater than 20S.

b) Tubes in those areas where 1xper11nc1 has indicated potential problems.

c) A tubt 1nspectton shall be perforllltd on each selected tube. If any selected tubt does not penatt the passage of tht tddy current probt for a tubt inspection, this shall bt recorded and an adjacent tubt shall be selected and subjected to a tubt insptctton.

6-9 H-3, Hi. ~. 174 October 31, 1996 p~ ll./OtC~

~, S.'a 6,Q 6,5,8 AQMINISTBATIYE COHJBOL?

Ste11 Generator Tybt Sycytjllao~Pcogram (continued)

3. Tht tubes selected as he second and third samples (if required by Table©5,8-l) during each ioservice inspection may be subjected to a partial tube inspection provided:

a) Tht tubes selected for these samples include the tubes from thost areas of the tube sheet array where tubes with imptrftctions were previously found.

b) Tht inspections include thost portions of the tubes where imperfections wtrt previously found.

4. Tht results of each sample inspection shall bt classified into ont of the following thrtt categorits:

Cateagry Inspection Results C-1 Less than 51 of the total tubes inspected are degraded tubes and oont of the inspected tubes ace defective, C-2 Ont or 110rt tubes. but not mart than 11 of the total tubes insp1ct1d art defective, or b1tw11n 51 and l°'- of the total tubes insp1ct1d art dtgradtd tubes.

C-3 Nott:

Mort than lei of tht total tubes insp1ct1d are degraded tubes or llOrt than 11 of tht inspecttd tubts art c;11f1ct1v1.

In all inspections. previously degraded tubts must exhibit significant (greater than l°'-) further wall penetrations to bt included 1n the above percentage calculations.

b. Insp1ct1gn fr1qy1nci1s The above requtrtd tnservict tnsp1ct1on of SG tubes shall be p1rf0Cllld at th1 following fr1qu1nc11s:*

1. Ins1rvic1 tnspect1ons shall be p1rf0Clllld at intervals of not 1111 than 12 nor 110r1 than 24 calendar 110nths afttr tht pr1vtou1 tnspect1on. If two cons1cut1v1 inspections following s1rvtc1 under AVT conditions, not including the prtstrvice 1nspec:tton, result tn all inspections results falling into the C*l cat19ory or if two constcut1vt inspections demonstrate that previously obstrvtd degradation has not continued and no add1t1onal degradation has occurred, the inspection interval

- 1 bt extended to a maxilllUll of once per 40 months.

6-10 Allltndllltnt Ho. a.9, 4', H, ~. ~. m. ~. 174*

October 31, 1996

- ~ IS o~ 2<1

6.0 6.5.8 A[)MINJSTBATIYE CONTRS)LS v ill n Pr r (continued) 5

2. If the results of the nservice inspection of a SG conducted I in accordance with Tab li'.15.8-1 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. Tht increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specificat~*o(1'.15.8.b.l; the interval may then be extended to 1 maximum of. ce pee 40 months.

s

3. Additional, unscheduled inservict inspections shall be performed on each SG in accordance with the first sample inspection specified in Table5.8-l during the shutdown subsequent to any of the ~owing conditions: I~

a) Primacy-to-secondary~ubt luks (not including luks originating f~tubt*t~~e sheet welds) l~xcess of the limits of(U?ifkili,2_311.(fc.o !."f.1_

b) A seismic occurrenct greater than the Operating Basis Earthquake.

c) A loss-of-coolant accident resulting in initiation of flow of the enginetrtd safeguards.

c.

1.

d) A main steam lint or main feedwatar line break .

Acceptanc1 Cr1ttria As ustd in this Specification:

a) Imptcfcctton means an exc1ption to the dimensions, finish or contour of a tube from that required fabrication drawings or specifications. Eddy-current testing indications below ZOT. of the nominal tube wall thickness, if detectable, may be considered as imperftctions.

b) Deqradatign 1111ans a s1rvic1-inductd cra'-.king, wastage, wear or general corrosion occurring on either inside or

- outside of a tube.

C) Qtqradtcl Tybt 1111ns a tube containing imperfections greater than or 1qu1l to ZOS of tht nominal wall thtckntss caused by degradation.

d) I Ptqr1d1tign means tht p1rc1nt1g1 of tht tube wall thickness affected or removed by degradation.

6-11 Amendment No. a9, ii, Wi, -Hi, ~. ~. 174 October 31, 1996 P~ lb of Z,9

5.S.'a 6,0 6.5.8 AQMlNJSI13ATJy£ CONTROLS Steam Gen1r1tor Tybt Syryejllance program (continued) e) ~means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

f) plygging Limit means the imperfection depth at or beyond which tht tube shall be removed from service and is equal to 40S of the nominal tube wall thickness.

g) Unseryiceablt described tht condition of 1 tube if it leaks or contains a defect large enough to 1ffect its structural integrity in the 1v1nt of 10 Op1r1ting Basis 1 S.8.b.3, above.

Earthquake, a Joss-of-coolant acc1dtnt, ~ a steill line or fttdwattr 11nt bruk as sptc1fttd 1n@)

h) Tybe Inscect1on means an 1nspactton of the SG tube from tht point of entry (hot 199 s1dt) completely around the U*btnd to tht top support of the cold leg.

1) pres1ryic1 lnspect1on means an inspection of the full length of each tube in SG ptrformtd by eddy current techniques prior to service to establish a baseline condttton of th* tubing. This inspection shall bt ptrfoCllld after th* shop hydrostatic test and prior to initial POWER OPERATION using tht equipment and techniques expected to bt used during subsequent ins1rvic1 inspections.

2. Tht SG shall bt dtttrmined OPERABLE aft1r. completing the corrtspond1ng actions (plug all tubes 1xc11ding tht plugging limit and all tubts containing through-wall cracks) required by Tab~S.B*l.

11.e ~toll ;.s ~O"t\5 o.f 5 it s,o:z.. OAL o.ppl;~~\.e. +o ~ ~ha.w-.

Ge\\e*"-~or {v.\oe, ~v.r.Je',l\o.~ee.. Vro~f't>.M*

6-12 Amendment Mo. a9, H, ** Hi, Hoe, ~. 174 October 31, 1996

TABLE~5.8-I STEAM GENERATOR TUBE INSPECTION 3RD SAMPLE INSPECTION :

lST SNtPLE INSPECTION 2ND SAMPLE INSPECTION ------

Action Required Result Action Required Action Required Result 1 s-.11 Stz9' Result N/A N/A N/A H/A C-1 lone A *tnt- of S Tubes

\ S.6.

pe/ C-1 None H/A N/A r--...._ C-2 Plug defective tubes

-~ - and inspect additional C-1 Mone 2S tubes t* thts S.&. C-2 Plug defective tubes and tnspect additional C-2 Plug defective tubes

f1Jo1'! 1) 4S tubes In this S.G.

C-3 Perfo.-. action for C-3 result of first Snple C-3 Perfo.-. action for C-3 result of first H/A N/A Suiple All other None H/A N/A C-3 Inspect all tubes in this S.G., plug de- S.G.s are fective tubes and C-1 Inspect ZS tubes in N/A H/A Some S.G.s Perform action for 3

each other S.G.

C-2 but no C-2 result of second

© ~

24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal :

notification to HRC up within next

...~o days 1

with written follow

~Ki.rid sample S.G. ~

C-3

~...Ul:

S.G. h

,s ~

Inspect all tubes each S.G. and plug H/A H/A C-l defective tubes. . --- ---

Where n is the nUllber of steill generators inspected during an inspection S

6/n S 6- ll Amendment No. 141, 174 Odobcr 31, 1996

5.9 AC!!INISTSATIY£ COrtJBQLS S.:S:,"\ 6.5.9 Secondary Wtter Chemistry Progrim A pro9rui shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:

a. ld1nt1f1cat1on of a s1111pling schedule for the critical variables and control points for these variabl1s,

b. Identification of tht procedures used to measure the values of the critical variables,

c. ldtnttfication of process sampling points, which shall include 110n1tor1ng tht discharge of tht condensate pumps for evidence of condenser in-l11kag1,

d. Proctdur=s for tht recording and managemtnt of data,

e. Procedures defining corrective actions for 111 off-control point ch111istry conditions, and

f. A procedure identifying (1) tht authority responsible for th*

inttrprttatton of tht data, and (b) tht s1qu1nc1 and timing of 1dllinistrattv1 1v1nts required to tnttiatt corrective actions .

6-14 Amendment Ho. 174 October 31, 1996 f~ 19 c-t z.q

s. s. \0 6.0

6. 5' 10 AQMIHISTSATiyE CONTROLS Ventil1tion filter Testing program A progr1m shill bt est1blished to implement the f llowing required @

testing of Control Room Ventilation (CRV) and Fue Ventilation A.l (FPV) systems 1t the frequencies specified in Regula ory Guide 1.52 Revision 2 (RG 1.52), and in accordance with RG 1.52 and ASME NSlO-i989 1t the system flowr1tes ind tolerances specified below*: '

1. Demonstr1t1 for e1ch of the ventila~ion systems that an inplace test of tht high efficiency particulate air (HEPA) filters shows ~

penetration ind system bypass< 0.05% for the. CRV and< l.00\ for the FPV when tested in accordance with RG l.SZ and ASME NSl0-1989:

Yent1{at1on System ElowrMe l~FMl

r. A or f-88 73 t 0%

V*SA 1nd V-88 10,000 t 20"4 V-95 or V-96 12,500 t 10\

b. Demonstr1t1 for e1ch of the ventil1t1on systems th1t 1n inpl1ce test of the ch1rco1l adsorber shows a penetration and system bypass

< 0.051 for tht CRV and< 1.00"4 for the FPV when tested in 1ccordanc1 with RG 1.52 and ASME N510*1989.

Yentwtlon System

r. and v-88 V-26A and V-268

c. Oet110nstratt for uch of the ventihtion systems that 1 liboratory test of a s1mpl1 of the ch1rcoal adsorber, when obtained as described in RG 1.52 shows the methyl iodide penetration less than the v1lut spec1fied below when tested in accordance with ASTI4 03803-1989 at a temper1ture of s 30'C and equ1l to the relative humidity specified as follows: .

Yentil1t1on System PenX~Eaiion Relatiy~ Hymidjty Vf-66 5,..

VEC*26A and VF~*268 0 .157\ 70,,,

d. For each of tht ventilation systems, demonstr1te the pressure drop 1cross tht combined HEPA filters, the prefilters, and the charco1l 1dsorbtrs Is ltss thin tht v1lut SQtcified below when tested in 1ccord1nct w1th RG 1.52 1nd ASME H510*1989:

vent1Jat1oQ System oelt1 i IIn H2Ql Elowrate ICD!l

{.g and v-88 o.o 10,000 t 'Oi VF-26A and VF-268 8.0 3266 +16'.--s, ~

- Detlonstratt that the hnters for W47J the ~e~@(I d1sslp1testh1 following specified value t 20"4 when tested in

accord1nc1 with ASME NS10*1989: *

~>=c;,...;.....:;..:.,o-:,.:~....:...:...;..:..;~~,,....;-;~;;;..; 1 re app l i c1b 1t to the

Should the 720-hour litTVtation on cNrcoal adsorber oper1tion occur durin; I pllnt OO**tion r9<1u1rin; the uM of th* c:N<coel IClsorb*

sucn 11 retuelin;

tHtin; may bt dlllyed until tne completion of tn* plam oo-1tion °' up to 1 ,!500 nours ot filtlf oper1tion; wnicnever occurs fir St.

6*15 Amendment Ho .. 174 October 31. 1996 e~ /(.a* i 9

{.o 6.0 ADMINISTRATIVE CONTROLS 6.5.11 Fuel Oil Testing Program A fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling requirements, testing requirements, and acceptance criteria, based on the diesel manufacturer's specifications and applicable ASTM Standards. The program shall establish the following:

1. Acceptability ~f new fuel oil prior to addition to the Fuel Oil Storage Tank, and acceptability of fuel oil stored in the Fuel Oil Storage Tank, by determining that the fuel oil has the following properties within limits:

a} API gravity or an absolute specific gravity, b} Kinematic viscosity, and c} Water and sediment content.

2. _Other properties of fuel oil stored in the Fuel Oil Storage Tank, specified by the diesel manufacturers or specified for grade 20 fuel oil in ASTM D 975, are within limits.

~e. -pro.,~sio"-S o~ SR. 3,o:z. o.-1.s£.3.0 ..3 a..ra o.ppf1'~t.le to ~~ Fu.~I ~

Di\ "T"t's+i~ 'l>rogro.rk, \!}_y Technical Specifications CTSl Bases Control Program This program provides a means for processing changes to the Bases of these Technical Speciffcations.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. A change in the TS incorporated in the )icense; or

2. A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

The Bases Control Program shall contain provisions to ensure that

-c.

the Bases are maintained consistent with the FSAR. @ .* @

d. Proposed changes that meet the criteria of Spec i fi cation 5. 12. b.

above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent

- with 10 CFR 50.7l(e}.

6-16 Amendment No. '1-+G, 1-74, f~ '-' of J..Cf

6.9 Ac::t!IHISTBATiyE CONTROLS S.S'. ,3 6.5.13 s.s, \"\ 6.5.14 Containmtnt Loak Batt Testing prgqr111 Progr111s shall be 1st1blish1d to implement th1 11ak r1t1 testing of the contain1111nt as r1quir1d by 10 CFB S0.54(o) and 10 CFR 50, Appendix J, Option B, as llOdifitd by approved exemptions. Thi Typ1 A test program shall ..,t thl requirements of 10 CFB 50, Appendix J, Option B and shall b* in 1ccord1nc1 with th* guidelines of Regulatory Guide 1.163, "P1rfon11anc1*Bastd Containment L11k1g1*Test Progr111, dated September 1995." Thi Type Band Typ1 C test progr111 shall m11t th* requirements of 10 CFR 50, Appendix J, Option A, as modified by the exemption fro~

certain requir...nts of 10 CFR 50 Appendix J which was granted in an NBC letter to Consumers Power Company dated December 6, 1989.

Th* *axillUll 1llow1bl1 contain1111nt leak r1t1,*L., at P., shall be 0.1% of contain1111nt air weight per day.

Leak r1t1 acc1ptanc1 criteria ar1:

a. Contain111nt 111k rate 1ccept1nc1 crit1ri1 is s 1.0 L.. During the first plant startup following testing in 1ccordanc1 with this progr.., the leak rat* acc1ptanc1* crit1ri~ 1r1 s 0.60 L. for the Type I and Type C tests and s 0.75 L. for Type A tests;

b. Air lock leak r1t1 acceptance cr1ttr11 is s 0.923 L. for each door, when pnssuriztd to 1 10 psig. ({s;J©J Thi Surv1illanc1 interval extensions of~~~.2 art not applicable to th* Contain111nt Leak Rate Testing Progr111 requirements.

Th* provisions of~.3 .iJ:l 1pplic1bl1 to th* Containment Leak Rate Testing Progr.. r1qutr...nts.

6-17 Amendment No. 174 October 31, 1996

~ '2l. o~ zq

6.0 ADMINISTRATIVE CONTROLS 5,~151 6.5.15 Process Control Program

a. The Process Control Program shall contain the current formula, sampling, analyses, tests, and determinations to be m*ade to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and State regulations, and other requirements governing the disposal of the radioactive waste.

b. Changes to the Process Control Program:

1. Shall be documented and records of reviews performed shal1 b~

retained as required by the Quality Program, CPC-2A. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluation justifying the change(s) and b) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become effective after approval by the plant superintendent

s.:.1a Amendment No. SS, 54, ~. 174

§,0 A[jfUNI§IMIIyt: CONTROLS REpoBTING BEQUIB£MENTS Tht following reports shall be submitted in accordance with 10 CFR 50.4.

~6.l Occupatjonal Radiation Exposyre Reogrt This report all include a tabul ion on an annual basi of the number of station utility and other pe onnel (including con actors) receivin xposures greater th loo:mrem/year and th ir associated man re* xposur* according to ork and job function (e.g., reactor opera ons and survenlance, inservic1 inspection routine maintenance.

A,Jl spt al maintenance [descr* e maintenance], wast processing and r u1ling). This tabula on supplements th* r uirements of 0 CFR 20.2206. Th* d t assignment to va~io s duty functions may be esti111t1s based on po et dosimeter, electr 1c dosi1111ter, TLD, or f. lm badge 111asur...nts. Small exposures.total ng less than 20f. of the individual total d

need not b* accoun d for. In the aggregat , at least SOI of the otal whole body dose eceived from external s rces shall bt assign to specific major w k functions. The repor. shall be sublli tted by r11 30 of each year. r----------:;._::_____,J

~.2 Bad1olgg1ca1 Eny1rqnmtntal Operating Bepgrt

The B1diol09ical Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 15 of Heh year. The report shall include su11111uies, int1rpr1tat1ons, and analysis of trends of the results of the radiological env1ron1111ntal monitoring program for the reporting period.

The ~attrial provided shall be consistent with the objectives outlined in the Offs1tt Dose Calculation Manual (ODCM) and in 10 CFR 50, Appendix I, Sections lV.B.Z, lV.B.3, and IV.C. ~

Cove.ri~ orer~+I~ o.f

~.l 8adioact1y* Effluent Bolease Btpgrt pla..11\~ *IA\ k rev io\.t.S y~r The Rad1oact1vt Effluent Rtltast Report shall b* subllitt'ed in accordance w1tb 10 CFR 50.361. The report shall include a su11111ary of the .

quantities of radioactive liquid and gaseous effluents and solid waste released froa tbt unit. The matlrhl provided shall be consistent with tbt objtctivts outlined in the Offsitt Dose Calculation Manual (OOCM) and P'roctss Control Progr111, and shall be in conformance with 10 CFR 50.361 and 10 CFR SO, Appendix I, Section IY.B.l.

priot- 1:-.o 1"1t:JL'f .1.. o ea&. I(t!At"'

~5.4 Mgntbly Optrat1nq Bepgrt Bout1nt reports of operating stati,tics and shutdown experience shall be subll1tted on a 110nthly bash to Ult NBCC!i 01't!J no later than the fifteenth of each 110nth following the calendar month covered by the report.

6*19 Amendment No. i:i, ii, H, H, ~. 1-&4, 174

October 31, 1996

SECTION 5.0 INSERT 1 This report shall include a tabulation on an annual basis of the number of stations, utility and other personneL(including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent greater than 100 mrem and the associated deep dose equivalent (reported in person-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year .

6-19

S', t..S 6.Q I~

AQt!INJSTBATiyE CQNTBOLS Core Oceratina Ljmits Becort (COLR) m

a. Core operating lim;ts shall be established prior to each reload cycle, or pr;or to any rema;ning portion of a reload cycle, and shall be documented in the COLR for the following:

3.1..'-\ ~tt.*

] .\do . 'i:!!:lf!ll:.l..JdlW L1mi t s J. "l.. \ .

Linear Heat Rate Lim;ts "3.1.."l... * .2 Radial Peaking Factor L;mits

b. Th1 analytical mtthods used to detennin1 th1 cort operat;ng limits shall bt those approved by the NRC, specifically those described in th* latest approved revision of th1 following documants:

1. XN-75.-27(A), "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors,* and Supplements l(A), 2(A),

3(P)(A), 4(P)~a d~A)~n Nu~Company.

(LCOs~, . . * ~ . , l~)

3.2..'I "J.\.l.o J.l. I 3,1...l..

2. ANF-84-73(P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events,*

and Appendix B(P)(A) and Supplements l(P)(Ali....lJP)~

Advanced N~ Fuels Corporation. (LCOs ~* ~

~' l ~) 3,1,~

3.L4 1.2..\ 'l.'1..1..

3. XN-NF-82-Zl(P)(A), "Application of Exxon Nuclear Company PWR Then11l M&f'9in Methodology to Mixed C~nfig~~s,*

Exxon Nuclear Company. (LCOs ~. ~ l ~

3.l.I..\ 'l.'2.. I 1.'1.. 1..

4. ANF-84-093(P)(A), "Steamlint Break Methodology for PWRs,* and Supple1111nt l(P~Ad~ Nuc~els Corporation.

(LCOs~, ~ . l .

~ J.l.(o 3,1,, :J.2..1...

5. XN-75-32(P)(A), *computational Procedure for Evaluating Fuel Rod Bowing, 9 and Supplements l(P)(A), 2~)._liflLA), and 4(!lllli_ Exxon Nucl 11r Company. (LCOs ~ ~ ~.

l~

1.l... l...

_ J.1..4 1.1.b J."l..I @

6. EXEM ~t~LO~l as defined by:

(LCOs~,

3.\.lo

."l.,\

l

'3.2..

a) XH*NF-82*20(A), "Exxon Nuclear Corapany Evaluation Model EXEM/PWR ECCS Model Updates,* and Supplemtnts l(P)(A),

2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

b) XN-NF-8Z-07(P)(A), "Exxon Nuclear Co111Pany ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company.

c) XN*NF-81-SB(A), "RODEXZ Fuel Rod Thtn111l-M1chanical Response Evaluation Model,* and SuppltMnts l(P)(A),

2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

6-20 Amendment No. -t-ii, 174 October 31, 1996

6,0 AQHrNrSTSATryE CONTROLS s.s.5 6,6.5 ~ (continued) d) XN-NF-85-16(A), "PWR 17xl7 Fuel Cooling Tests Program,"

Volume l and Supplements l(P)(A), Z(P)(A), and J(P)(A),

and Volume 2 and Supplement l(P}(A); Exxon Nuclear Company.

e) XN-NF-85-lOS(A), "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," and Supplement l(P)(A); Exxon Nuclear Company.

7. XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for P~zad Water Reactors," Exxon Nuclur Company. (LCOs * ~ & G:Jl!l)

" \

'?. * *t,, ~- ~,2..1... n-.

8. AHF-1224(P)(A), "Departure from Nucleate Boiling Correlation .

for High Thermal Performance Fuel," and Su'()le~nt.J...iflJ.A):

Advanced Nuclear Fuels Corporation. (LCOs . l. , ~. l ~

Qrltj) ;.i. r.'2... \ ~

J .'1..1.

9. AHF-89-lSl(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA C~ l~ts~anced

~r Fuels Corporation. (LCOs .. , ~. ~* l C7) 3.t.t, ~) ~

3.i..'4 1.l.lo 1.-Z..'

10. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Perform~l," Siemens Power Corporation. (LCOs~*QfR* l ._ _~_,_

3 @

c. The cor1 operating limits shall be datermined such that all applicabl1 li~its (a.g., fual thermal mechanical limits, core ther111l hydraulic limits, Emergency Cori Cooling Systems limits, nuclear limits such as shutdown margin, transient analysis limits, and acctd1nt analysts limits) of th* safety analysis are met.

d. Th1 COLR, tncludtng any mid cycle r1vistons or supplements, shall bl provtdld, upon tssuanc1 for each r1load cycle, to the NRC.

6-21.

Amendment No. 174 October 31. 1996 p°1r- a. 7 c1 a- er

§.0 6.6.~

AQMIHIST8ATIVE CONTROLS Cont1inmtnt Stryctyc1l Integrity Syrvejllanca Reogct

s. lo.8 6. 6--~ Steam Generatgc Tybt Sycye111 ance Beoort The following reports shall be subraitted to the Co11111ission following each insecvice inspection of steam generator tubes:

a. The numbtr of tubes plugged 1n each ste1111 generator shall be reported to the Co11111ssion within 15 days following the completion of each inspection. ancl

b. The complete results of the ste1111 generator tube inservice inspection shall be reported to the Co1111ission within 12 months following completion of the inspection. This report shall include:

1. Number and extent of tubes Inspected.

2. Location and percent of wall-thickness penetration foe each indication of an imperfection.

3. Identification of tubes plugged *

c. Results of stel9 generator tube inspections that fall into Category C*3 shall require 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal notification to tht NRC prior to resumption of plant operation. A written followup within the next 30 days shall pcovidt a description of investigations and corrective 1111suc1s taken to prevent rtcurrenct.

6*22 Amtndmtnt No. ~. ~. 174 October 31, 1996

§.9 6.7 AQMINISTBATIYE CONTROLS HIGH BAPIATION AREA s:1. \ 6.7.1 Pursuant to 10 CFR 20, paragraph 20.l60l(c), in lieu of the reQuirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20. in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit RWP) Individuals ~ualified in radiation pro c ion proce ures e.g., (fnttn ph'tiJ@technicians) or personnel ~

continuously escorted by such individuals may be exempt from the RWP ~

issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates < 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or mort of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation 1110nitoring device that continuously integrates the radiation dost rate in tht area and alarms whtn a preset int19rattd dost is received. Entry into such areas with this monitoring device may bt made after tht dost rate levels in the

c.

area have been established and personnel art aware of them .

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall p1rfor11 periodic radiation surveill~nct at the frequency spt'Mi*d by tht Radiation Work Request. l@

5.1.1,.

6.7.Z In addit~to tht requirements of Sptcification~7.1, except as allowed by~7.3, areas with radiation levels~ 1000 mrem/hr shall be ~I provided *1th locked or continuously guarded doors to prevent ~

unauthorized entry and tht keys shall bt maintained under tht tac:h~.+106 adllin1strat1vt control of the Shift Supervisor on duty or nt sics \-.-,

supervision. Doors shill remain l~cked except durtng perio s o ccess

71 by personnel under 1n approved RWP that shall specify the dost rate 11v1ls 1n the illlldt1t1 work ar11s and the maxilllUll allowable stay times for 1ndtvtduals 1n those areas. In lteu of th* stay t11111 spectfication of th* RWP, dtrtct or remote {such as closed ci-rcuit TV camtras) continuous surv1tllanc1 may be made by p1rsonn1l qualified in radiation prottctton procedures to provide positive exposure control over the act1v1ties being p1rf0Mllld within the 1rea.

s:1.1 5.7.3 For individual high radiation areas with radiation 11v1ls of~ 1000 mrlll,lhr, accessible to personnel, that art located within large areas such as reactor contain..nt, where no enclosure exists for purposes of locking, or that c1nnot bt continuously gu1rded, and where no enclosure can bl reasonably constructed around tht individual area, that individual 1r11 shill bl barricaded and conspicuously posted, and a flashing light shall bl activated as a warning dtv1ct.

6-23 Amendment No. 48, ~. -t-Mo, 174

October 31. 1996

ADMINISTRATIVE CHANGES (A)

ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Tecffirical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involve no technical changes to existing Technical Specifications.

Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development.certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the-design is already approved by the NRC, adding more details does not result in a technical change.

A.2 CTS 6.1.2, 6.2.2a and 6.2.2b use the terminology "above COLD SHUTDOWN." In the proposed ITS, this corresponds to MODES 1, 2, 3, and 4. As discussed in

Chapter 1.0, the CTS COLD SHUTDOWN is essentially equivalent to the ITS MODE 5 (CTS 210 F vs. ITS 200 F). Therefore, "above COLD SHUTDOWN" in the CTS equates to MODES 1, 2, 3, and 4 in the proposed ITS. This change is considered to be an administrative change to adopt the terminology of the ISTS.

A.3 CTS 6.2.2a uses the phrases "assigned to each reactor containing fuel," and "assigned for each control room. " The Palisades Nuclear Plant has only one reactor and one control room. Therefore, the wording in ITS 5.2.2 is being modified to state "assigned when fuel is in the reactor, " and "assigned when the reactor is operating" to more accurately reflect the Palisades plant specific design. This change is considered to be an administrative change since no technical requirements have changed.

A.4 . CTS 6.2.2b, 6.2.2g, and 6.5.4d use the term "unit" when discussing the reactor. The typical term used in the remainder of the CTS is "plant." Therefore, the term "plant" will be used in the proposed ITS 5.2.2. This is an administrative change to reflect tll;e typical Palisades Nuclear Plant terminology .

Palisades Nuclear Plant Page 1of6 01/20/98

A.5 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS CTS 6.4.1 requires that written procedures shall be established, implemented, and maintained for the activities listed. In this list, the CTS contains item b., "Refueling operations, and item c., "Surveillance and test activities of safety-related activities."

These items are included in the procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978 which is referenced in CTS 6.4. la and included in the proposed ITS 5 .4. la~ Therefore, since these procedures are already required by the reference to Regulatory Guide 1.33, Revision 2, February 1978, they are not included in the proposed ITS. This change is an administrative change since no requirements have changed. This change maintains consistency with NUREG-1432.

A.6 CTS 6.4.1 requires that written procedures shall be established, implemented, and maintained for the activities listed. In this list, the CTS contains item f., "Site Security Plan implementation" and item g., "Site Emergency Plan implementation .. " These items were recommended to be removed from the Technical Specifications in NRC Generic Letter 93-07 since they are duplicative of regulations contained in the Code of Federal Regulations part 50 and 73. This change is considered to be an administrative change since these requirements must still be met as required by the Code of Federal Regulations. This change maintains consistency with NUREG-1432.

A.7 CTS 6.5.7 is entitled "Inservice Inspectio~ and Testi~g Program." In the proposed ITS 5.5.7, the title is changed to the "Inservice Testing Program." This change is considered to be an administrative change since the requirements of the program are unchanged. This change maintains

~ ..

consistency with NUREG-1432.

A.8 CTS 6.6.5b.l lists, among referenced LCOs, "3.10.1." That item is unnecessary and has been deleted. Neither CTS 3.10.1, nor its ITS replacement reference the COLR.

Palisades Nuclear Plant Page 2of6 01/20/98

A.9 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS CTS 6.6.8, "Containment Structural Integrity Surveillance Report" requires that a report be submitted to the NRC covering Prestressing, Anchorage, and Liner and Penetration tests. Proposed ITS 5. 6. 7, "Containment Structural Integrity Surveillance

. Report"*--also requires_ that a report be submitted to the NRC but only specifies the Prestressing and Anchorage tests be included. Reference to the Liner and Penetratio~_

tests have been deleted since the requirement for these tests was removed from the technical specifications by Amendment 109 dated October 28, 1987. Initially, the Liner and Penetration tests were included in the CTS since they were relative new designs and a surveillance program was established to assure the affected components would maintain their functional integrity. Based on test data, it was concluded that the liner plates and penetration assemblies were performing as predicted. Therefore, the CTS was amended and the surveillance program terminated. As such, it is no longer necessary to reference these tests in ITS 5. 6. 7.

A.10 CTS 4.5.6, "Dome Delamination Surveillance" has been modified to include reference to ITS 5. 6. 7, "Containment Structural Integrity Surveillance Report. " The intent of this change is to clarify the reporting requirements associated with the dome delamination inspection. As stated in CTS 4.5.6, -a dome delamination inspection

shall be performed within 90 days following corrective retentioning of dome tendons and the results of the inspection reported to the NRC. ITS 5.6.7 requires that a report of the dome delamination test be submitted to the NRC within 90 days after completion of the test. The proposed change is considered administrative in nature since no additional restriction are imposed on plant operation. Inclusion of the dome delamination reporting requirements in the Containment Structural Integrity Surveillance Report is discussed in Discussion of Change M.3 to this Section.

A.11 CTS 6.5.8, "Steam Generator Tube.Surveillance Program," and CTS 6.5.11, "Fuel Oil Testing Program," are revised *to provide statements of applicability for SR 3.0.2 and for SR 3.0.2 and SR 3.0.3, respectively. These statements provide clarity and ensure consistent application of these requirements for the Programs referenced by ITS SRs. This change is consistent with NUREG-1432 as modified by TSTF-118.

A.12 CTS 6.6.1, "Occupational Radiation Exposure Report," and CTS 6.6.3, "Radioactive Effluent Release Report," are revised to incorporate language related to revisions to 10 CFR Part 20, and 10 CFR 50.36a. These changes are administrative since there are not actual changes in the application of the requirements. This change is consistent with NUREG-1432 as modified by TSTF-152 .

Palisades Nuclear Plant Page 3 of 6 01/20/98

A.13 ATTACHMENT 3 DISCUSSION OF,. CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS CTS 6.6.4, "Monthly Operating Report," is revised to omit the words "to arrive" since the Palisades Nuclear Plant has no control of the document once it is mailed.

Further, -~his is inconsistent with typical NRC submittal requirements. This change is considered administrative since it has no effect on plant operations and impacts only the submittal of after-the-fact information. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. l CTS 6.4.1 requires that written procedures be established, implemented, and maintained for the listed activities. Proposed ITS 5.4.1 contains the same wording.

However, proposed ITS 5 .4 .1. b is not in the CTS and is being added. Proposed ITS 5.4.1.b states "The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33." Since this item is not included in the CTS it is considered to be a more restrictive change. This change maintains consistency with NUREG-1432.

M.2 CTS 6.5.3 describes the Post Accident Sampling Program. It states in part" .. and which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, .... " In the proposed ITS, the reference is to "radioactive gases" rather than just radioactive iodines. Because the use of the term "gases" is broader than "iodines" for the sampling and analyzing requirements, this is considered to be a more restrictive change. This change is consistent with NUREG-1432.

M.3 The CTS does not contain a program for Containment Tendon Testing. CTS Sections 4.5.4 and 4.5.5 do address tendon testing and these requirements have been replaced with a program. CTS 4.5.6 contains requirements for containment dome delamination inspection. These dome delamination inspection requirements have been added to the ISTS program requirements. Since the program addresses structural components other than tendons, the program has been titled "Containment Structural Integrity Surveillance Program."

Palisades Nuclear Plant Page 4 of 6 01/20/98

M.4 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS The CTS does not contain a Safety Functions Determination Program. Proposed ITS 5. 5 .13 includes this program. This program is added to work in conjunction with the proposed ITS in identifying any loss of safety function which might exist.

Because-the CTS did not contain this program, and its implementation requires additional evaluations to identify a loss of safety function than what is required in the CTS, this change is considered to be a more restrictive change. This change maintains consistency with NUREG-1432.

M.5 CTS 6.6.7 contains the reporting requirements for specific accident monitoring instrument channels that are not restored to an Operable status within the required

Completion Time. CTS 6.6. 7 requires that a report be submitted within 30 days.

Proposed ITS 5.6.6 also contains reporting requirements for specific ac.cident monitoring instrument channels that are not restored to an Operable status within the required Completion Time. However, the ITS requires that a report be submitted within 14 days. As such, the proposed change imposes an additional restriction on plant operations since the time period allowed to submit the report has been shortened I.

from 30 days to 14 days. This change has been proposed to establish consistency with NUREG-1432 and is deemed acceptable since it only involves a change to administrative requirements and does not alter the way in which the plant is operated.

LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

LA.1 CTS Specification 4.5.4, Surveillance for Prestressing System (page 4-21a) and 4.5.5, End Anchorage Concrete SurVeillance (page 4-21c) were replaced by proposed ITS Specification 5.5.5, the Containment Structural Integrity Surveillance Program. The proposed specification emulates the ISTS treatment of containment structural integrity surveillance requirements. The details associated with containment tendori inspections have been removed from the technical specification and reference has been included in ITS 5.5.5 to ASME Boiler and Pressure Vessel Code,Section XI, Subsections IWE and IWL which establishes the applicable test methods, acceptance criteria and testing frequencies. Removal of these details is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsections IWE and IWL is specified in 10 CFR 50.55a. Thus, this change eliminates duplication of federal regulations and can be made without an impact on public health and safety. Removal of these details from the CTS and the incorporation of a containment tendon surveillance program in Section 5.0 of the ITS is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 5 of 6 01/20/98

ATTACHMENT 3 DISCUSSION OF CHANGES

.I CHAPTER 5.0, ADMINISTRATIVE CONTROLS LESS. RESTRICTIVE CHANGES (L)

There were no "Less Restrictive" changes made to this chapter.

RELOCATED (R)

There were no "Relocated" changes made to this chapter .

Palisades Nuclear Plant Page 6 of 6 01/20/98

ATTACHMENT 4 PALISADES NUCLEAR PLANT

CHAPTER 5.0, ADMINISTRATIVE CONTROLS NO SIGNIFICANT HAZARDS CONSIDERATION

ADMINISTRATIVE CHANGES NO SIGNIFICANT HAZARDS CONSIDERATION ATTAC1'ENT 4 CHAPTER 5.0, ADMINISTRATIVE CONTROLS The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NU:REG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve reformatting; renumbering, and rewording of ....

Technical Specifications. These changes, since they do not involve technical changes to the' Technical Specifications, are administrative.

This type of change is connected with the movement of requirements within the current requirements, or with the modification of wording which does not affect the technical content of the current Technical Specifications. These changes will also include nontechnical modifications of requirements to conform to the Writer's Guide or provide consistency with the Improved Standard Technical Specifications in NUREG-1432. Administrative changes are not intended to add, delete, or relo.cate any technical requirements of the current Technical Specifications.

In accordance with the criteria set forth in 10 CFR 50. 92, Palisades Nuclear Plant staff has

\

evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration:. The following is provided in support of this conclusion.

1. Does the change involve a significant iricrease in the probability or consequences of an accident previously evaluated?

The proposed changes involve reformatting, renumbering, an4 rewording of the existing Technical Specification. These modifications involve no technical changes to the existing Technical Specifications. The majority of changes were done in order to be consistent with NUREG~1432. During the development of NUREG-1432, certain wording preferences or English language conventions were adopted. The changes are administrative in nature and do not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefore, the changes do not involve a significant increase in* the probability or consequences of an accident previously evaluated .

Palisades Nuclear Plant Page lof 5 01/20/98

2.

NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 CHAPTER 5.0, ADMINISTRATIVE CONTROLS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes involve reformatting, renumbering, and* rewording of the existing Technical Specifications. The changes do not involve a physical alteration of __

the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The changes will not impose any new or different requirements or eliminate any existing requirements. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of safety?

The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes are administrative in nature and will not involve. any technical changes. The changes will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Also, since these changes are administrative in nature, no question of safety is involved. Therefore, the changes do not involve a significant reduction in a margin of safety.

MORE RESTRICTIVE CHANGES The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outJined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants. " Some of the proposed changes involve adding more restrictive requirements to the existing Technical Specifications by either making current requirements more stringent or by adding new requirements which currently do not exist.

These changes may include additional requirements that decrease allowed outage time, increase frequency of surveillance, impose additional surveillance, increase the scope of a specification to include additional plant equipment, increase the applicability of a specification, or provide additional actions. These changes are generally made to conform with the NUREG-1432.

In accordance with the criteria set forth in 10 CFR 50.92, the Palisades Nuclear Plant has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.

Palisades Nuclear Plant .Page 2 of 5 01/20/98

1.

NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS ATTACHMENT 4 Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event. If anything, the new requirements may decrease the probability or consequences of an analyzed event by incorporating the more restrictive changes. The changes do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis.

Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. The changes do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing *normal. plant operation. The changes do impose different requirements.

However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of safety?

The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. Adding more restrictive requirements either increases or has no impact on the margin of safety. The changes, by definition, provide additional restrictions to enhance plant safety. The changes maintain requirements within the

safety analyses and licensing basis. As such, no question of safety is involved.

Therefore, the changes do not involve a significant reduction in a margin of safety .

Palisades Nuclear Plant Page 3 of 5 01/20/98

NO SIGNIFICANT HAZARDS CONSIDERATION ATTAC1'ENT 4

CHAPTER 5.0, ADMINISTRATIVE CONTROLS LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering ,

Plants." Some of the proposed changes involve moving details (engineering, procedural, etc.)

out of the Technical Specifications and into a licensee controlled document. This information

may be moved to the ITS Bases, FSAR, plant procedures or other programs controlled by the licensee. The removal of this information is considered to be less restrictive because it is no longer controlled by the Technical Specification change process. Typically, the information moved is descriptive in nature and its removal conforms with NUREG-1432 for format and content.

In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.

- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed changes move details from the Technical Specifications to a licensee controlled document;

The removal of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident I' previously evaluated . I

Palisades Nuclear Plant Page 4 of 5 01/20/98

2.

NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 CHAPTER 5.0, ADMINISTRATIVE CONTROLS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? .

The proposed changes move detail from the Technical Specifications to a licensee controlled document. The changes will not alter the plant configuration (no new or ,

different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Th~re are no design changes or equipment performance

parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes remove details from the Technical Specifications and place them under licensee control. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

LESS RESTRICTIVE CHANGES There were no "Less Restrictive Changes" made in Chapter 5.

Palisades Nuclear Plant Page 5 of 5 01/20/98

ATTACHMENT 5 PALISADES NUCLEAR PLANT

CHAPTER 5.0, ADMINISTRATIVE CONTROLS MARKUP OF NUREG-1432 TECHNICAL SPECIFICATIONS

Responsibility

5. l **

~.o 5.0 ADMINISTRATIVE CONTROLS ls,. I 5.1 Responsibility

&;. I. I 5.1.1 The f{/1ant $uperintenden~shall be responsible for overall

_operation and shall delegate in writing the succession to this

~1) (])

responsibility during his absence.

The ~lant Superintenden~or his designee. shall approve, prior to Ci) implementat1on, each proposed test, experiment.or modification to

.systeins or equipment that affect. nuc r safety.

{,. l. "2.. 5.1.2

-1.. r' .. \Clr\

The {Shift Supervisor (SS) . shall be responsible: fID: t~e control \ CD room conmand function *. Durin any absence of the ';{SSf'from the control room while the u" s in MODE 1, 2, 3; or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the COfltrol room command function.

Curing any absence of the fSSt' fr.om the. c;:on~ ro 1 .room while the CD

~;+/-)~is in HOOE 5 or 6, an i11dividual with an active SRO license

. ~~n~~~r~~0 ~p~~:!~~~ }~~~~~ 0 ~~a l J}e. ~es~gnated to assume the

... (~1>)

CEOG STS . 5. 0-1 Rev 1, 04/07/95

Organization 5.2 lo.D 5. 0 ADMINISTRATIVE CONTROLS

~.L 5.2 Organization l'o.1-.I s.2.1 Onsite and Offsite Organizations ~--...,

~

Onsite and offsite organizations shall be established for~

operation and corporate management, respectively. The onsite anq offsite organizations shall include the positions for activiti~s-affecting sa*fety of the lnu¢iear ~n .Pl a~.

-~l.sc...J~

a. Lines of authority, responsibility, and co11111unication shall be defined and established tnroughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documen.ted and updated,- as appropriate~** in organization *charts~ functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of doc~~nta~ion. These

requi*rements* *sh al 1 be documented in the ,fFSARH (1\-.\SltrLI) -::;.- ~ . -a/

b. The '1{1ant iuperintendent~ shall be responsible for overall safe operat'fon of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

c. ~ t'/ specified corporate executive positieflf shall have (!)~

~porate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

~

fM Staff f.o.7...""L 5.2.2 The ti t staff o I

a. A non-licensed operator shall be

[contQinjng fuell and an additional /C0

CEOG STS 5.0-2 (continued)

Rev 1, 04/07/95

SECTION 5.0 INSERI These requirements and the plant specific equivalent of those titles referred to in these Technical Spe~fications shall be documented in the FSAR.

5.0-2

Organization

. ' 5.2 5.2 Organitation

@~~(i; 5.2.2 ~Staff (continued)

D Two unit sites with both units *shutdown or defueled require a total of three non-licen~ed operators for the

two units. .

]

b. At 1east one 1i censed Reactor Operator (RO} sha 11 be present in the control room when fuel is in the reactor. In 0 add1t1on, while the--.~ is in MODE 1, 2, 3, or 4, at least one 1 i censed Senior Reactor Operator (SRO) shall' be pr.esent (j) in the control room. *

c. Shift crew composition may be less than the minimum requirement of 10-CFR 50.54(m)(2)*(i) and- 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence; of on--duty: shift crew members provided immediate. act]on is taken to, restore the shift crew

- composition:to:withjn~the . minimum.requirements.

d. A ~al tli-.Physics iechnici~nfshal 1.: be-. ~n si te.-wher=r fuel is in the recrctor. The position may be vacant for not more -

th an. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to prov i de,-*f or unexpected absence, provided immediate_action 1s.ta-ken to fill the required position. . * ** - 11 -

e.

. * .-1,

. 9I..~"

Administrative procedures_ shall e developed-and implemented

. 0 to 1 i mi t the working hours *of un staff who perform-** CD safety. rehted ~fun.ct ionsc( e.-g~ . f 1 i censed, SROs, *licensed. ROs,

<IleaYth pbyiic'ijStS)-:auxil iary operators, and key maintenance personne 1)

A-radio+iGI\ ~~ ~f'Sol\11c.I.

- .** .;.-'.* .!.~..:..;.-..:.c.,...*, ..,:.::._. ~. ***. - * *._, * * -* *, .* : .***.. r--. ~--*

Adequ e shift coverage hall be maintain without routine hea use of overtime The objective s 11 be to have o rating personnel ork an [8 or 12] our day, nominal 0 h~ur week whil the unit is aper ing. However, i

the event that unfo seen problems re ire substantial aunts -

of overtime t be used, or duri extended period of shutdown fo refueling, major aintenance, or m or plant modificat' n, ori a temporar asis the followi g guidelines shall b followed: *

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;

(continued)

CEOG STS 5.0-3 Rev 1, 04/07/95

SECTION 5.0 INSERT In the event ihat overtime is used, the following guidelines shall be followed:

. ,**i , . . . . .-. - : :~ * *:*n

-. :.,'(;: j

- .; .~

5.0-3

Organization 5.2 5.2 Organizition

~

5.2.2 ~Staff (continued)

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break cif ~t least S hour~ sho~~d be allowed between work periods, including shift turnover time; 4.. Except during extended shutdown periods, the use of overtime should be considered*on~an-individual basis and not for the entire staff on' a shift.

CD

~ .~

Any deviation from..Jhe ab~ve guidelines Jpall be authorized in advance by the .!ftlant ~uperintendentf'or his designee, in accordance*with-approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting

the deviation. ,_ .. , .

Controls shall .be included in the procedures such that individual overtime shall be reviewed monthly by the ~lant

~upe.rioteo.Q_en~f,:;"or his designee to ensure that excessive hours hive not been assigned. Routine deviation from the

above-- guidel-1 nes,- is- not author'.i zed-*.

[ The am unt of overtime perfo ing safety rela cont lled>in* accorda war ng hours (Gener*

orked by unit ed functions s 11 be limited nd e with the NRC Policy Stateme t on Letter 82-12) f.

g.

.CEOG STS 5.0-4 Rev 1, 04/07 /95

?Ian+

@ Staf( Qualifications 5.3

5.0 ~~~~STRATIVE CONTROLS 5.3 ~Staff Qualifications Reviewer's Noe: Minimum qualificat' ns for members of th unit staff shall be specifi~ by use of an overall q lification statemen referencing an ANS Standard a eptable to the NRC sta f or by specifying i ividual position qualifica ans. Generally, the f'rst method is prefer le; however, the second m hod is adaptable to th se unit staffs requi ng special qualifi ation statements becaus of unique organizat* nal structures.

5.3.1

~tlSJ: rJ18. I-), 11 h...- c. .. -~~"-*I<:.

eos:~o-<

CEOG STS 5.0-5 Rev 1, 04/07/95

SECTION 5.0 INSERT

5. 3. 2 The radiation safety manager shall meet the qualifications of a Radiation Protection Manager as defined in Regulatory Guide 1.8, September 1975. For the purpose of this section, "Equivalent," as utilized in Regulatory Guide 1.8 for the bachelor's degree requirement, may be met with four years of any one or combination of the following: :J' (a) Formal schooling in science or engineering, or (b) ope.rational or technical experience and training in nuclear. power. . . .

5.3.3 The Shift Technical Advisor shall-have a bachelor's degree or equivalent and the Shift Engineer shall have a bachelor's degree in a scientific or engineering discipline.

Specific training for both the~ Shift Technical* Advisor_ and .tbe Shi~* E_ngirie~r shali include plant design, operations, and response and analysis of the plant for transients and accident. The Shift-Engineer shal_! ho~d ~ Senior Reactor Operator license.

5.3.4 The plant staff who perform reviews which ensure coinpliaiice with 10 CFR 50.59 shall meet or exceed the minimum' qualifications of ANS '3.1-1987, Section 4.7.1 and 4.7.2.

A Senior Reactor Operator license or certification-shall be considere4equivalentto a bachelors degree for the purpose of this. specification.

.. /'-. :~._ -. *~: : _. :"'; a ~ * - * ** " -'-*.,*.,t;

. . *- *-~'t'*,5

~ '* ~ . ' -. ' . : ... . *..... :. -!_..'

5.0-5

- - .. *:~

Procedures 5.4 c....-r.s .-*--

(a.O 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures Gi.'1. \ 5.4.1 Written procedures shall be established, implemented, and

_maintained covering the following activities: * *

a. The applicable procedures reconunended in Regulatory Guide 1.33, Revision 2,Q\P"~ndix/A;), February 1978;

b. The emergency operating procedures required to implement the(l) requirements of NUREG-0737 and~ NUREG-0737, Supplement 1, as stated in ~Generic Letter.82-3311 . . ([)

tc* /Quality/assurance for eifluent_and envir9i{mental monitoring;j @

~~ Fire Protection Program implementation; and fl}(j. All programs specified in Specification 5.5.

f. Modtfi ation of core prate ion calculator (CPC addre* *able constants.- *Th se procedures sha 11 nc 1ude provi ions to ensure that sufficient margin is maintained

in C C type I addressabl constants to avoid cessive ope ator interact~on wit CPCs during reactor operation.

Mo 1f1cat1ons to th* C~ software (1ncludin changes of a gori thms and fuel cy le,. spe~.ifi,c data) sh 11 be performed i accordance with**th most-recent version of*ncpc Protection lgorithm Software C nge Procedure,n CEN- 9(A)-P, which has een determined to b applicable to the f cility. Additions or deletions to CPC addressable constant or changes to addressable consta software limit valu s shall not be implemented withou prior NRC approval .

CEOG STS 5.0-6 Rev 1, 04/07/95

Progr~ms and Manuals

. 5.5 5.0 ADMINISTRATIVE CONTROLS

~.s 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.l --Offsite Dose Calculation Manual COPCMl

a. The ODCM shall contain the methodology and parameters used"'"

in the calculation* of offsite doses resulting from radioactive gaseotis and liquid effluents, in the calculation*

of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The OOCM shall also contain the r~dioact~ve effl~ent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Ra~oactive Effluent Release Reports required by Specification t5.6.2.r'and Specification (S.6.3.t'. . *le:>

(L~e~eo/inftiate~ {hanges to the OOCM: (f)

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),

2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR SO, Appendhc I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. ~~all become*effective after the approval of the ~fl ant

~perintendentf; and

c. Shall be submitted to the NRC in the form of.a complete, legible copy of the entire OOCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the OOCM was made.

Each change shall be identified by markings in the margin of

CEOG STS, 5.0-7 (continued)

Rev 1, 04/07/95

Programs and Manuals

' 5. 5

lo. s. \

5. 5 Progra11s ..-nd Manuals 5.5.1 Offsite Dose Calculation Manual (QOCMl (continued) the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Primary Coolant Source} Oytside Containment from those

a. Preveritfve maintenance and periodic visual inspection requirements;~ *

b. Integrated leak test requirements for each system at refueling cycle intervals or less;

~. ~.'3 5.5.3

.(_ J:" I'\ 'i,e.rf) .

This program provides controls that ensure the capability to .t' obtain and analyze reactor coolant, rad*ioactive gases, and

(@

particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.=* The program shall include the fo 11 ow i ng : _'"' i,

a. Training of personnel;

b. Procedures,for sampling and analysis; and

c. Provisions for maintenance of sampling and analysis equipment.

(.,.S.~ 5.5.4 Radioact1ye Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactiv~ effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably

CEOG STS 5.0-8 (continued)

Rev 1, 04/07 /95

SECTION 5.0 INSERT

c. The portion of the shutdown cooling system that is outside the containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig;

d. Piping from valves CV-3029 and CV-3030 to the disc_harge of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig; and

e. The maximum allowable leakage from the recircula.tion heat removal systems' components (which include valve stems, flanges, and pump seals) shall not exceed 0.2 gallon per minute under the normal hydrrn~tatic h~ad from the SIRW tank (approximately 44 psig) .

5.0-8

-- _ I

Progr,~ms and Manuals

5. 5 .*

5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls program (continued) achievable. The program shall be contained in the ODCM, shall be

-i.mplemented by procedures, and shall include remedia.l actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous mon;toring instrumentation including surveillance tests and setpoint detennination. in accordance with the methodology in the ODCH;

b. Limitations on the concentrations of radioactive material re 1eased in 1i quid effluents to urirestri cted areas," /2'\

confor111ing to" O CFR ZO, Appendix B, Table:.Z, Column 2; ldJ I 0 -f\-:<-.S ~.. vc:lvc. l"

c. Monitoring, samp ng, an ana ysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM; *

d. Limitations on the annual and quarterly doses or dose

e.

co11111;tment to a member o,f. the. public from. ra~ioactive materials in liquid"effluents released from each unit to unrestrictea areas,'confonning to 10 CFR 50, Appendix I; Deterntination f cumulat-1-ve~and. p ejected dose contri tions fra11 radioac ve effluents for t e current calendar arter and current alendar year in ac ordance wtth. tne me odology and parame rs in the ODCH at east every 31 _days;

f. son the functiona capability and use of the liquid d gaseous effltient treatm~rit system~ t ensure that approp ate portions of th se system~ are used o reduce relea s of radioactivity when-the proj~cted oses in a pert of 31 days would xceed 2% of the gui elines for the ann 1 dose or dose co itment, conforming o 10 CFR 50, ncltx I; e.(j) Ltllitattons on the dose rate resulting from radioactive utertal rel eased in gaseous effluents to areas beyo.n.:.::d"--=t:.:.:h.:::.e--.. . .

sit* boundary confonning to the dose associated with to*hi.... -\\--..IP.I 10 CFR 20, Appendix B, Table 2, Column l; * .R l-kd 1n 1

- CEOG STS 5.0-9 (continued)

Rev 1, 04/07/95

Programs and Manuals 5.5

5.5 Programs and Manuals

~.S.'-t 5.5.4 Radioactjve Effluent Controls Program (continued)

-F Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

8 days in gaseous effluents released.. from each**unit-to areas beyond the s.ite boundary, conforming.*to 10 *CFR 50, Appendix I; and t1 ~ Limitations on t_he annuafdo.se or-dos.e coriimitm.ent to any member of the public due to ~eleases of radioactivity and to radiation from uranfom fuel cj'cle *sou.rces*, conforming to 40 CFR 190.

5.5.5 This program provide~- ontrolS *fo" t~ra*k' th*e FSAR Sect i,on [

cy~ ic._and tran~ient occu~rence~.to nsure that:c~mp~ents are

~ ntarned withrn t e des1gn ,1im1ts*,.'*;:'. :: .. ** , 'c_ ___ :_ ____ ! . --~

l'\C::.J 5.5.~

This program provides cbntrdls for monitoring any tendon degradation in pre-stressed concrete containments, including Q effectiveness of -its _c9.rro.$:i.9r;i .prote~_tion- :Q'le~i.um, .to ensure containment structural integrity~ . The. progr~m. sh~ll. iriclJ:!!!!_ /

baseline measurements prior-*to initial operations. - The ~do-1')

Surveil hnce Program, i nsp'ectiori'"fr'equeric ;-e*s*; and* acceptance criterh shall be in accordance ~ith l{<Re_g:ijilatory Giiide J.35J x

CRivision 71959)1"" ASrnl DDIW o.rJ H't~vr~ 1/e.Mi...L wd( S<.vt1Cfl 'XI, 1

'/ /'I t:'

,. I ) r < 11--h~e~*> Svbe..c.+1orJ ILVC. a.rd LWL Tlae provisions of SR 3.0.2 and SR 3.0.3 are applicable to the

' 1

{..

r:;-;\.

(ljidOftjSurveillance Program inspection frequencies. ~

~o...rn1'1.JMt1tJf:Snuc-rldAL l~T;~~

~.5.L& s.5.fJ ~ Coolant Pymp Flywhe'el Inspection Program C0 s . . ,vt'.\\, ... 4 ot ...

-\h.L ~r:,..,._c ... , t.<>o\c.~+ fl""""'f' +-l1 .... \.:-.e .. \.5 sl"'.._\\ '-<>~<.*st o~

\ut>1, oo\-~ ..{,:c.. i"'SI>(<--~*""' o.f ~ '-"ft<<r .\:-1-,...,"'ce.\s <!/*'-\...,. 10 Y(.0..r5 *.

T~ pra,;.s:o.-s 0 \'.'. $~ '!.,t),(. c..<'(. c:.n\:co-1-\~ 4-t. ~ f,;,..., .. ,..

1 oo\o.~* ~--'7.Fl"'\v1'-'~e\ :C'n.sot,<Ro ..... f,.. * .,,,.,_,_...

CEOG STS 5.0~10 1, 04/07/95

SECTION 5.0 INSERT If, as a result of a tendon inspection, corrective retensioning of five percent (8) or more of the total number of_dome tendons is necessary to restore their liftoff forces to within the limits, a dome delamination inspection shall be performed within 90 days following such corrective retensioning. The results of this inspection shall be reported to the NRC in accordance with *'

Specification 5. 6. 7, "Containment Structural Integrity Surveillance Report."

5.0-10
Progr.~ms and Manuals
5. 5 .*
5.5 Programs and Manuals (continued)
~- <;.1 5. 5.eG) Inservice Testing Program This program provides controls f6r inservice testing of ASME Code
-Class 1, 2, and 3 components including applicable supports. The program shall include the following:
a. Testing frequencies specified in Section XI of the ASME
0 Boiler and Pressure Vessel Code and applicable Addenda as fa 11 ows : / '() r )
~ 0 1 Pv Co';)lf-.
\--~[
E1~VY Ce~ ASMk!:i1%r Vess Cod and ndt.:Lure app icabl Add nd enn1no ogy or Required Frequencies inservice testing for performing inservice activities testing activities Weekly 7 days Monthly _ 31 days Quarterly or every 3 months 92 days Semiannually or every 6 months Every 9 months .
Yearly or annually Biennially or every 2 years
b. The provisions of SR 3.0;2 are applicable to the above required Frequencies for performing inservice testing activities;
c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and & ~PV *
d. Nothing. in the 1.:..A~s*;_:j construed to superse e
.1::..-=;c~a-=,~~...:..=~~;;.;;;..c..,;;,;..;__;::..;.;:...~ be © lP.s:. B Steam Generator CSGl Tube Surveillance Program
CEOG STS 5.0-11 Rev 1, 04/07/95
SECTION 5.0 INSERT 5 .5. 8 Steam Generator Tube Survemance Pro~am This program provides controls for surveillance testing of the Steam Generator (SG)' "
tubes to ensure that the structural integrity of this portion of the Primary Coolant System (PCS) is mafutained. The program shall contain controls to. ensure:
a. Steam Generator Tube Sample Selection and Inspection The inservice inspection may be limited to one SG on a rotating schedule encompassing 6 % of the tubes if the results of previous inspections indicate that both SGs are performing in a like manner. If the operating conditions in one SG are found to be more severe than those in the other SG, the sample sequence shall be modified to inspect the most severe conditions.
The SG tube minimum sample size, inspection result classification, and the
corresponding action required shall be as specified in Table 5.5.8-1. The tubes selected for each inservice inspection shall include at least 3 % of the total number of tubes in all SGs; the tubes selected for tbese inspe~tiom. shall be selected on a random basis except:
1. Where experience in similar plants with similar water chemistry indicates critical areas to be mspected, then at least 50 % of the tubes inspected shall be from these critical areas;
.. 2. The first sample of tubes selected for each inservice inspection of each SG shall include:
a) All nonplugged tubes that previously had detectable wall penetrations greater than 20 %;
b) Tubes in those areas where experience has indicated potential problems;
5.0-11
SECTION 5.0 INSERT (continued) c) A tube inspection shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
3. The tubes selected as the second and third samples (if required by Table 5.5.8-1) during each inservice inspection may be subjected to a partial tube inspection provided:
a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found; b) The inspections include those portions of the tubes where imperfections were previously found.
4. The results of each sample inspection shall be classified into one of the following three categories:
Inspection Results C-1 Less than 5 % of the to..,tal tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1 % of the total tubes inspected are defective, or between 5 % and 10 % of the total tubes inspected are degraded tubes.
C-3 More than 10 % of the total tubes inspected are degraded tubes or more than 1 % of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations .
5.0-11
SECTION 5.0 INSERT (continued)
b. Inspection Frequencies The above required inservice inspection of SG tubes shall be performed at the following frequencies: ?
1. Inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspections results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
2. If the results of the inservice inspection of a SG conducted in
accordance with Table 5.5.8-1 at 40 month intervals fall into -
Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.8.b.l; the interval may then be extended to a maximum of once per 40 months;
3. Additional, unscheduled inservice inspections shall be performed on each SG in accordance with the first sample inspection specified in Table 5.5.8-1 during the shutdown subsequent to any of the following conditions:
a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of LCO 3.4.13; b) A seismic occurrence greater than the Operating Basis Earthquake; c) A loss-of-coolant accident resulting in initiation of flow of the engineered safeguards;
5.0-11
SECTION 5.0 INSERT (continued) d)
A main steam line or main feedwater line break.
c. Acceptance Criteria
1. As used in this Specification:
a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required fabrication drawings or specifications. Eddy current testing indications below 20%
of the nominal tube wall thickness, if detectable, may be considered as imperfections; b) De~ra<lation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube;
c) d)
De~raded Tube means a tube containing imperfections greater than or equal to 20 % of the nominal wall thickness caused by degradation;
% De~adatjon means the percentage of the tube wall thickness affected or removed by degradation; e) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; t) Plu~~in~ Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40 % of the nominal tube wall thickness; g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 5.5.8.b.3, above;
5.0-11
SECTION 5.0 INSERT (continued) h) Tube Inspection means an inspection of the SG tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; i) Preservice Inspection means an inspection of the full length of each tube in SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed after the shop hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
2. The SG shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by
Table 5.5.8-1.
This provisions of SR 3.0.2 are applicable to the Steam Generator Tube Surveillance Program.
~
r~
5.0-11
SECTION 5.0 INSERT TABLE TABLE 5.5.8-1 STEAM GENERATOR TUBE INSPECTION lST SAMPLE__ INSPECTION 1 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Result Action Required Result Action Required Result Action Required C-1 None N/A N/A N/A N/A C-2 Plug defective tubes and inspect additional ZS tubes in this SG. C-1 None N/A N/A C-3 Inspect all tubes C-2 Plug defective C-1 None in this SG, plug tubes and inspect defective tubes and additional 4S tubes C-2 Plug defective inspect 2S tubes in in this SG. tubes each other SG.
C-3 Perform action for C-3 result of first Sample
C-3 All other Perform action for C-3 result of first Sample None N/A N/A N/A N/A SGs are C-1 Some Perform action for N/A N/A SGs C-2 C-2 result of but no second sample.
other SG is C-3 Other Inspect all tubes N/A N/A SG is each SG and plug C-3 defective tubes NOTES: 1 The minimum sample size for the first sample inspection is S tubes per SG where S=(6/n)%, where n is the number of steam generators inspected during an inspection .
5.0-11
Programs and Manuals 5.5
S.s* Programs and Manuals (continued) l"l5
"' s. 9 s.5.@ Secondary Water Chemistry Program
--- ~--*--:,
,l'l..,..., . r.*"'~ ..--*.:.
This program provides (c;ntrol s for mo~i-;; 5.5.~@ Ventilation fjlter Testing Program~ c~,.,+,.~1 R,oo ...... v.z..-.i:l~;k,":(c~v
~ r. ,.._J != *. J. H-~*<<ili~*'j 1-lrea. V.,1<*1\a.+.o>'l
~.~~~~~*~t sh~l_~n~\:~tm; '~*~ .!~u~~P}£~~)~~~~.~~~~~f~1tf~~;;~
systems it the freqµencies spec1 ie in - Regulatory GlJWe ~\_, and C) in iCCOrdance with {Regulatory Guide 1.52, Revision 2.pASME _ii;Sz. Kt,:r o=2J!'0 NSl0-1989(, ~rid A!fl/t at the system flowrate-r-specified below ho\*
I(i/fOS]j. 8 hkr"~te~ 1* \.':!)
1. O.Onstrate t,pr each of th~;;;~ms that an i npl ace test Ji 4
cj)----i of the .H'igh Afficiency i'articulate i'ir (HEPA) filters shows
/ ____il .. Pt!.n,~.t~~i_on an_cL~tstem bypass ~_;{_Q ._Q_S J%~when tested in CD 0
r .....__ , .
""~< \.oo""lo \,,,*f\...e. *accordance with .{Regulatory Guide 1.52, Revision 2, and ASME Q.
F-~e.1 tl<c..v.!\.i ... ~ Anc.:
V-ev.<t;l.,.-\"icv.. S'is~e~
CEOG STS 5.0-12 Rev 1, 04/07/95
Programs and Manuals 5.5 5.5 Programs and Manuals b,$,10 5.S.'1j)Q§') Ventilation Filter Testing Program {VFTPl (continued) system owrate speci ied as follpws] ~
m Ventilation System Flowrate (c.Ff'I)
~?A LO~
8A oc V-0B o..r..l>. v- gp,
- 'tS o..- v - "\ lo } \) e,-,*h la.-h 0 ~
7lUH
~0,000 :r Zo/..
11., >co 1:: I 0),
CD
b. Demonstrate for each of the system:* that an i npl ace test of the charcoal adsorber shows a penetration and ~ystem 1I ---~r--***---:---*
0
h. ~ ., ________ Q.xP,ass < -~'o. os17_..._~hen tested in accordance with .{"R~g~l atory \f7\.
-\~.c. L-~\. (.. r G_LJ_!_de l.~_Rev ~]On 1._,_and ASME_N5)9_::l~?Jtray the sy~ 0(©
z. \,00~10 -\or hi \:-°v<;\ jtlowratj spedTieo afS foTTOws LJ f0%J1:--------.
1-\c...,._A\;~ A.re.c...
Vewhlj)..i.:10~ s'f~hil'-< @f) Vent i 1at ion System Fl owrate
[ \/~Ft 1/-2..~
o-e.J V-gB }
e>..,J V-L.~B
~o,ooo -rZ.5'.l.
32.00.,..1o?o-S"?. *
~
c. Demonstrate for each of the~ system~ that a laboratory @)
test of a sample of the charcoal adsorber, when obtained as described in fRegulatory Guide 1.52, Revision 2q**, shows the G) methyl iodide penetration less than the value specified below when *tested in accordance with ,fASTM 03803-1989t' at a G) temperature of s f3o*c}-"and greater than or equal to the CI) relative humidity specified as follows:
~Ventilation System Penetration RH
(...oo')o} [qs~ CD
[ o. I '51 i- Lo~
(continued)
CEOG STS 5.0-13 Rev 1, 04/07/95
Programs and Manuals 5.5
c:-1>
r.,,),10 5.5 Programs and Manuals 5.5.~ Ventilation Filter Testing Program <VFTP) (continued)
Note: Allowable p netration .. [1oor/~-~~~~~-l-io_d_i_d/______,
for charcoal cred'ted in staff saf ty evaluation]/
actor). @
Safety actor * [5] for sy terns with heater .
[7] for s stems without he ters.

~------:--___L__ _ _ _ _~_J_-----~----~

l-.J_e~ \,J; 0('\ r d. For each of the'>~ systems, demonstrate the pressure drop @
across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with {Regulatory Guide 1.52, \
Revi si2n __~JD.dJ-SM~-- ~µQ_:-_l9-~~esy$teFfl owrate] G) \G)
[speci~ed as fo~ows [!ll0%]j:
(@Ventilation System Delta P~.1\1..;y Flowrate fj) t;~:;:,:~~-B~1 D*i } ~~1:5. 0
Demonstrate that the heaters for [Uili PT! the@.ysteJ dissipate.sthe following specified value f+/- ~03}-*when tested* .-
in accordance with .fASME N510-1989~: ~
} -,
G The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
5. 5. 12 .
rogram provides c ntrol for potentia y expl.osive gas <le) 1e
mixt res contained in he [Waste Gas Hold System], [the quantity of adioactivity cont ined in gas storag tanks.or fed into the of gas treatment sys em, and the quanti of radioactivity c ntained in unpro cted outdoor liqui storage tanks]. The (continued)
CEOG STS 5.0-14 Rev 1, 04/07/95
Programs and Manuals 5.5
5.5 Progra11s and Manuals 5.5.12
_gaseous radi oact
i ty quantities sha 11 b determined fo 11 owing the methodology in Branch Technical Positi n (BTP} ETSB 11-5, "Postulated R ioactive Release due t Waste Gas System Leak or Failur~*]. e liquid radwaste quan ities shall be determined ih accordance ith [Standard Review Pl n, Section 15.7.3, "Postulated Radioacti ~elease due to Tank F lures"].
The pr ram shall include:
The limits for concentr t1ons of hydrogen and oxygen in th
[Waste Gas Holdup Sys m] and a surveillance program to r:;:;\
ensure the limits ar maintained. Such limits shall be ~
appropriate to the ystem's design criteria (i.e., whe or not the system s designed to withstand a hydroge explosion}; *
b. A surveillanc program to ensure that the quanti of radioactivit contained in [each gas storage t k and fed into the of gas treatment system] is less tha the amount that woul result in a whole body exposure o ~ 0.5 rem to any indi idual in an unrestricted area, in he event of [an uncontr led release of the tanks' conten s]; and
c. A su e11lance program to ensure that e quantity of ra oact1vity contained in all outdoo liquid radwaste tanks t tare not surrounded by liners, kes, or walls, capable f holding the tanks' contents and hat do not have tank overflows and surrounding area dr. ins connected to the
[Liquid Radwaste Treatment Syst ] is less than the amount that would result in concentr ions less than the limits of 10 CFR Part 20, Appendix B, ble 2, Column 2, at the nearest potable water suppl and the nearest surface water supply 1n an unrestricted _rea, in the event of an uncontrolled release of e tanks' contents.
Tbe provisions of SR 3.0.2 nd SR 3.0.3 are applicable to the Explosive Gas and Storage ank Radioactivity Monitoring Program surveillance frequencies
CEOG STS 5.0-15 (continued}
Rev 1, 04/07/95
Proqrams and Manuals 5.5
c:~
&.')ii/
5.5 Progr111s and Manuals 5.5.l@(i) OOti&lfl Fyel (continued)
Oil Testing Program A diesel fuel oil testing program to imple nt required tesytng of both n fuel oil ands red fuel oil sh be establishetj.I.' The prog m shall include ampling and test'ng requirements,./and ac ptance criteria all in accordanc with applicable ASTM andards. The p pose of the prog m is to establi the following:
a. Accepta lity of new fue oil for use prio to addition to ~
stora tanks by dete 'ning that the fu oil has: * ~
An API gravitY. or an absolute s cific gravity wi limits, int and kinemati viscosity withi for fuel oil, and
3. A ear and bright a earance with pro er color;~
b. Othe properties for M20 fuel oil a within limits
c wi in 31 days foll ing sampling and ddition to stora e nks; and Total particul e concentration the fuel oil is ~ 10 mg/l when tested ery 31 days in a ordance with AST D-2276, Method A-2 r A-3.
Technical Soecifications CTSl Bases Control Program This program provides a means for processing changes to t_he Bases of these Technical Specifications.
a. Changes *to the Bases of the TS shall be made under appropriate administrative cbntrols. and reviews.
~. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
A change in the TS incorporated in the license; or A change to the updated FSAR or Bases that involves an unreviewed rafety question as defined in 10 CFR 50.59 .
CEOG STS 5.0-16 (continued)
Rev 1, 04/07 /95
SECTION 5.0 INSERT Fuel OU Testing Program A fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling requirements, testing requirements, and acceptance criteria, based on the diesel manufacturer's specifications and applicable ASTM Standards. The program shall establish the following:
a. Acceptability of new fuel oil prior to addition to the Fuel Oil Storage Tank, and acceptability of fuel oil stored in the* Fuel Oil Storage Tank, by determining* that the fuel oil has the following properties within limits:
1. . API gravity or an absolute specific gravity,
2.
Kinematic viscosity, and
3. Water and sediment content .
b. Other properties of fuel oil stored in the Fuel Oil Storage Tank, specified by the diesel manufacturers or specified for grade 2D fuel oil in ASTM D 975, are within limits.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Fuel Oil Testing Program .
5.0-16
Pro~rams and Manuals 5.5
er~
5.~.17.,
5.5 Progrills and Manuals s.s.1jj Technical Specifications (TSl Bases Control Program (continued}
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
d. Prop~~d changes that meet. the criteria of Specification q:..
5.5.~ above shall be reviewed and approved by the NRC * \..7 prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.7l(e).
5.5.1~ Safety functions Determination Program CSFDP)
This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result.
of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This
program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:
a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed. in the accident analysis does not go undetected;
b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
c. Provisions to ensure that an-inoperable suppol'ted system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
d. Other appropriate limitations and remedial or compe~satory actions.
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or *
CEOG STS - 5.0-17 (continued)
Rev 1, 04/07/95
Programs and Manuals 5.5
5.5 Programs and Manuals Safety Functions Determination Program {continued)
b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to support system(s) for the supported systems (a) and {b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. *
>. s. '!.f
©
CEOG STS 5.0-18 Rev 1, 04/07 /95
SECTION 5.0 INSERT 1 5.5.14 Containment Leak Rate Testin~ Pro~ram Programs shall be established to implement the leak rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The Type A test program shall meet the requirements of 10 CFR 50, Appendix J, Option B and shall be in accordance with the guidelines of Regulatory Guide 1.163, "Performance-Based
Containment Leakage-Test Program, dated September 1995."
The Type B and Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A, as* modified by the exemption from certain requirements of 10 CFR 50 Appendix J which was granted in an NRC letter to Consumers Power Company dated December 6, 1989.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 52.64 psig .
The maximum allowable containment l~ak rate, La, at Pa, shall be 0.13 of containment air weight per day.
Local leak rate tests, other than Persorinel Airlock doors between the seals tests, shall be performed at~ 55 psig.
Local leak rate tests for checking airlock doors seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed as follows: * *
a. A between the seals test shall be performed on the Personnel Airlock ~t
~ 10 psig.
b. A full presslire test shall be performed on the Emergency Escape Airlock at~ 55 psig. A seal contact check shall be performed on the Emergency Escape Airlock following each full pressure test. Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional
. pressure testing .
5.0-18
SECTION 5.0 Leak rate acceptance criteria are:
a. Containment leak rate acceptance criteria is :s:: 1.0 La. During the first plant startup following testing in accordance with this program, the leak rate acceptance criteria are :s:: 0.60 La for the Type Band Type C tests and
s:: 0.75 La for Type A tests;
b. The leakage for a Personnel airlock door seal test shall not exceed 0.023 La.
c. An acceptable Emergency Escape Airlock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces.
Containment OPERABILITY is equivalent to "Containment Integrity" for the purposes of the air lock testing requirements in 10 CFR 50, Appendix J.
The provisions of SR 3.0.2 are not applicable to the Containment Leak Rate
Testing Program requirements .
The provisions of SR 3.0.3
-Program requirements .
~applicable-to the Containment Leak Rate Testing
5.0-18
SECTION 5.0 INSERT 2
5. 5 .15 Process Control Program
a. The Process Control Program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing amt packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and State regulations, and other requirements governing the disposal of the radioactive waste.
b. Changes to the Process Control Program:
1. Shall be documented and records of reviews performed shall be retained as required by the Quality Program, CPC-2A. This documentation shall contain:
a) Sufficient information to support the change together with the
b) appropriate analyses or evaluation justifying the change(s) and A determination that the change wili maintain the overall conformance of the solidified waster product to existing requirements of Federal, State, or other applicable regulations.
2. Shall become effective after approval by the plant superintendent.
5.0-18
Repor~1ng Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1 Occupational Radiation Exoosure Report t-=;;;------r-----~-----__,...--------r-----~- ----*-

---------------------NOTE- ------------ ----------

A single submittal may be made for multiple un't station. The 13' submitt should comb'ne sections mmon to all nits at t ~
station *
(JJ 5.6.2 <[OD?!JDRad1oloqical Environmental Ooerating Recort
- ----------.- --------------------NOTE- -----------------------------*- -
A single subll1ttal may be made for a multiple unit station. The 13' subllittal should combine sections co11111on to all units at the 1.2.J station.
The <li!ahDRad1ological Environmental Operating Report covering ~
tile operation of the.i@!) during the previous calendar year shall 51 subi1tted by Hay 15 of each year. The report shall include su11111aries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual
CEOG STS 5.0-19 (continued)
Rev 1, 04/07 /95
SECTION 5.0 INSERT 1 This report shall include a tabulation on an annual basis of the number of stations, utility and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent greater than 100 mrem and the associated deep dose equivalent (reported in person-rem) according to work and job functions (e.g., reactor operations and.
surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing and refueling). This tabulation supplements the requirements of 10 CPR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent external sources should be assigned to specific major work functions. The report coveritig the previous calendar year shall be submitted by April 30 of each year .

S.0-19
Reporting Requirements 5.6
5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)
(OOCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3,
_ and IV. C.
The Annual Radiolo ical Environmenta Operating Report shal include the resul s of analyses of 11 radiological enviro mental samples and of 1 environmental r diation measurements t ken re::'\
during the per* d pursuant to th locations specified i the table \:::!.)
and figures i the COCH, as wel as summarized and tab ated results of t ese analyses and easurements [in the fo at of the table in th Radiological Ass ssment Branch Techni*c Position, Revision 1 November 1979]. [The report shall iden ify the TLD results t at represent coll cated dosimeters in r~ ation to the NRC TLD rogram and the e osure period associat with each result. In the event t t some individual res ts are not avail le for inclusion ith the report, the r art shall be subm" ted noting and e plaining the reasons f the missing res ts. The missin data shall be submitte in a supplementary re ort as soon as p sible .
~.(,.J 5.6.3 Radioacti~~
Effluent Release Recort
-~--------------- -NOTE-----------
A single s bmittal may be ma e for. a multipl unit station. T e submittal should combine se tions common to 11 units at the r1) station; owever, for unit with separate r dwaste systems, he submitt shall specify t releases of ra ioactive materia from each u t.
Repor+i~g Requirements 5.6
5.6 Reporting Requ;rements 5.6.4 Monthly Operat;nq Reports (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLRl Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload 1..-C...O J.l,°C:, Res.._....k~ ~.J cycle, and shall be documented in the COLR for the r
L::w"'Ow..f' fo s: h'"""" oll owing:
L:-~tt L-~ ,,... ....,,.._,. H ~.J-LC..-=> 1.7.., \ operating 4te L~-:-*s Rd:..Q.
"~ .... \:_:~~
-.=-......~ . . . .
i_.,:_:t-S A-~:I:' L:-:-\-.S L..C...O 7."'t.'-\
©
}
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 S pressure and mperature limits f. r heatup, cooldown, low temperature eration, critical , and hydrostatic
CEOG STS 5.0-21 (continued)
Rev 1, 04/07/95
SECTION 5.0 INSERT
1. XN-75-27(A), "Exxon Nuclear Neutronics Design Methods for Pressurized
\Vater Reactors," and Supplements l(A), 2(A), 3(P)(A), 4(P)(A), and 5(P)(A);
Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
2. ANF-84-73(P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," and Appendix B(P)(A) and Supplements l(P)(A), 2(P)(A); Advanced Nuclear Fuels Corporation.
(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
3. XN-NF-82-2l(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company. (LCOs 3.2.1, 3.2.2, & 3.2.4)
4. ANF-84-093(P)(A), "Steamline Break Methodology for PWRs 11 and Supplement l(P)(A); Advanced Nuclear Fuels Corporation.
(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
5.
6.
XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing,"
and Supplements l(P)(A), 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
EXEM PWR Large Break LOCA Model as defined by:
(LCOs 3.1.6, 3.2.1, & 3.2.2) a) XN-NF-82-20(A), "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates, 11 and Supplements l(P)(A), 2(P)(A),
3(P)(A), and 4(P)(A); Exxon Nuclear Company.
b) XN-NF-82-07(P)(A), "Exxon Nuclear Company ECCS Cladding Swelling and* Rupture Model, 11 Exxon Nuclear Company .
5.0-21
SECTION 5.0.
INSERT (continued) c) XN-NF-81-58(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," and Supplements l(P)(A), 2(P)(A), 3(P)(A), and*
4(P)(A); Exxon Nuclear Company.
d) XN-NF-85-16(A), "PWR 17x17 Fuel Cooling Tests Program," Volume 1 and Supplements l(P)(A), 2(P)(A), and 3(P)(A), and Volume 2 and Supplement l(P)(A); Exxon Nuclear Company.
e) XN-NF-85-105(A), "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs, 11 and Supplement l(P)(A); Exxon Nuclear Company.
7. XN'-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors, 11 Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, & 3.2.2)
8. ANF-1224(P)(A), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," and Supplement l(P)(A); Advanced Nuclear Fuels Corporation. (LCOs 3.2.1, 3.2.2, & 3.2.4)
9. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
10. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation. (LCOs 3.2.1, 3.2.2, & 3.2.4)
5.0-21
Reporti~g Requirements 5.6
5.6 Reporting Requirements 5.6.6 Reactor Coolant S stem RCS LIMITS testing as we as heatup ~nd cooldown r es shall be established nd documented in the PTLR r the following:*
[The lridivi ual specifications that a ress RCS pressure and
~empe~atu limits must be reference here.] *
b. The an ytical methods used to de rmine the RCS pressure and t perature limits shall be ose previously reviewed and pproved by the NRC, specif 'cally those described in the fo owi~g documents: [Identi the NRC staff approval
. cumetlt by date.]
c. The PTLR shall be provide to the'NRC upon issuance for ea reactor vessel fluence ~ riod and* for any revision or supplement thereto.
Reviewers' Notes: Them hodology for the calculation of e P-T *.*,,~~~~pl :
limits for NRC approval hould include the following prov sions: ~
I 1.
2.
the methodology hall de~cribe ~o~ the return flu nee is calculated (re erence new Regulatory Guide when ssued).
The Reactor essel Material Surveillance Pro am shall comply wit Appendix H to CFR 50. The reac r vessel material rradiation surveillance specimen emoval schedule shall b provided, along with how the spe imen examinatitins shall e used to update the PTLR curves
3. Low. emperature Overpressure Protecti n (LTOP) System lift s ting limits for the Power Operat Relief Valves (PORVs),
veloped using NRC-approved metho ologies may be included in the PTLR.
The adjusted reference tempera ure* (ART) for each reactor beltline material shall be c culated, accounting for radiation embrittlement, in ccordance with Regulatory Guide 1.99, Revision 2.
The limiting ART shall incorporated into the calculation of the pressure and te perature limit curves in accordance with NUREG-0800 Stan rd Review Plan 5.3.2, Pressure-Temperature Limits. *
(continued)
CEOG STS 5.0-22 Rev 1, 04/07/95
Report~ ,,g Requirements .
5.6 5.6 Reporting Requirements 5.6.6 TEMPERATURE LIMITS
/
6. Th minimum temperature requir, ments of Appendix G to 10 e~R P t 50 shall be incorporate into the pressure and ;1 emperature limit curves. /
Licensees who have remov two or more capsules shoirld compare for each survei ance material the measure.¢'* increase in reference temperat e (RTNor) to the predicted/increase in RTN 0 r; where the pre cted increase in RTNor is p-'ased on the mean shift in RTNot plus the two standard devi,tion value (2a.) specified i Regulatory Guide 1.99, Re sion 2. If measured value ceeds the predicted value increase in RTNor
+ 2a.), the.li ensee should provide a sup ement to the PTLR to demonstra how the results affect t approved methodology
5.6 . .7 If an ind' idual emergency dies or more lid failures in the st 25 demands, thes any non alid failures experi ced by that EOG in at time period shall reported-within-30 ays~ Reports on EOG fanures shall (EOG) four and inclu the information rec ended in Regulator Guide 1.9, Revi on 3, Regulatory Pas* ion C.5, or existi Regulatory Gui 1.108 reporting req irement.
crs !1:fo,J, '-i !Y1111-.~;*, n ~ I . '* * ..
~.b.1 5.6.~©* ~Reoor~-
Whu a report is required
Post Accident Monitoring
~jo;;/B or G M:>Lco 3.3~
Instrumentation, a report shall 11 Ci) 6)
lie submitted within the fol ow ng 14 days. The report shall
...tltne tba preplanned alternate method of monitoring, the cause ef tbe 1noperab1lity, and the plans and schedule for restoring the 1astrumentat1on channels of the Function to OPERABLE status.
5.6.@6)
CEOG STS L____ '. 5.0-23 Rev 1, 04/07 /95 Do*.!.. 1>0-'-""'*~1,* .. ....,,., t.~~~---:J
Report~ng Requirements c..-n 5 .6.
5.6 Reporting Requirements 5.6.fi]} Tendon Surveillance art (co~tinued) 30 day~ The r~ rt shall include a des-~riptio~ t~* e endon
__ con~;ion, th~-condition .of the concrete (espec-ially a tendon anc rages)y-'the inspection procedures, the tolerance on er eking, ind the corrective act-ion taken. / 1
(... b. 9 5. 6. ~@ Steam Generator Tube In soector Reoort ese report may e require covering inspection, te t, and maintena ce activities. These reports are ff\
determined on n individual b sis for each un t and their l.!J preparation a d submittal are designated in e Technical S ecificatio
CEOG STS 5.0-24 Rev _l, 04/07 /95
SECTION 5e0 INSEiU The following reports shall be submitted to the Commission followiiig each inservice inspection of steam generator tubes:
a. The number of tubes plugged in each steam generator shall be reported t(\
the Commission within 15 days following the completion of each inspection; and
b. The complete results of the steam generator tube inservice inspection shall be reported to the Commission within 12 months following completion of the inspection. This report shall include:
1. Number and extent of tubes inspected;
2. Location and percent of wall-thickness penetration for each indication of an imp~rfection;
3. Identification of tubes plugged .
c. Results of steani generator tube inspections that fall into Category C-3 shall require 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal notification to the NRC. prior to resumption of plant operation. A written followup within the next 30 days shall provide a description of investigations and corrective measures taken to prevent recurrence .
5.0-24
{'High ~Rad'iation Areaf-'CD
. f5.7J-'(D
5.0
=[5.7 ADMINISTRATIVE CONTROLS High Radiation Area?"
5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the
___ requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation. Work Permit (RWP). Individuals qualifM1e in radiation protection procedures (e.g., (~ealth .Physics echniciansf) or personnel continuously escorted by luch ind viduals may be exempt .
from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates L @1000 mrem/hr, provided they are otherwise following plant. .
radiation protection procedures for entry into such high radiation areas *.
An*y individual or group of individuals permitted to enter such areas sha 11 be provided with or accompanied by one or more of the .
following: * -
a. . A radiation monitoring device that continuously indicates the radiation dose rate in the area.
b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received~ Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of
them. *
c. An individual qu~lified in radiation protection procedu~es with a radiati6n dose rate monitoring device, who is responsible for providing positive control over the activities witMn the area and shall perform periodic ..
radhttu surveillance at the frequency specified by the .*
Pf!Rad1ifliri Profect1on Man§9erjfif1 tWel RWP:. , .* * (D
. . . . ~e.J(.:eyt o..s o.l\~ ....e.J. b, ->:ii] . .
5. t. 2 In* addition to the requirements of Specification 5.7.1~ areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shi ft Ge@ on duty or (heajtn pti/s1cs suQitvu1o)t Doors shall remain /1 ocked *except durin ' eri o_ds of access b personne 1 *
Cf2)
CEOG STS 5.0-25
{continued)
Rev 1, 04/07/95 *
'{High Radiation Areaf
{5.7}-
&*7.1..
[5.7 High Radiation Area]
5.7.2 (continued) under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for
-individuals in those areas. In lieu of the stay time specifi~ation of the RWP, direct or remote (such as closed circuit TV cameras) continuous survei 11 ance may be made by personne 1 _
qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. /
~.1.) 5.7.3 For individual high radiation areas with radiation levels of ~
"> (J)lOOO mrem/hr, accessible to personnel, that are located within
' l"'arge areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot- be continuously guarded,.
and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as ~
warning device *
,*I. I;
CEOG STS 5.0-26 Rev 1, 04/07/95
ATTAC1'ENT 6 PALISADES NUCLEAR PLANT
CHAPTER 5.0, ADMINISTRATIVE CONTROLS JUSTIFICATION FOR DEVIATIONS FROM NUREG-1432 .
Change ATTACH1\1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 5.0, ADMINISTRATIVE CONTROLS Discussion a
Note: This attachment provides brief discussion of the deviations from NUREG-1432 that
were made to support the development of the Palisades Nuclear Plant ITS~ The Change Numbers correspond to the respective deviation shown on the "NUREG * *"*
MARKUPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided. * *
2. . Deviations have been made for. clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
3. The requirement/statement has been deleted since it is not applicable to this facility.
The following requirements* have been renumbered, where applicable, to reflect this deletion .
4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification. The Administrative Controls section of the Palisades Technical Specifications has been recently amended to closely emulate NUREG-1432. In some instances the wording and requirements differ from those in NUREG-1432. In order to limit the impact of conversion to improved Technical Specifications, revision of the subject wording has not be proposed.
6.
CTS 6.2.2a uses the phrases "assigned to each reactor containing fuel," and "assigned for each control room." The Palisades Nuclear Plant has only one reactor and one control room. Therefore, the wording in ITS 5.2.2 is being modified to state "assigned when fuel is in the reactor," and "assigned when the reactor is operating" to more accurately reflect the Palisades plant specific design .
Palisades Nuclear Plant Page 1of6 01/20/98
ATTACI'ENT 6
. JUSTIFICATION FOR DEVIATIONS
SPECIFICATION 5.0, ADMINISTRATIVE CONTROLS Change Discilssion
7. CTS 6.2.2f requires the oper~tions manager or the assistant operations manager to hold.*
an SRO license. In addition, it states "The individual holding the SRO license shall be
responsible for directing the activities of the licensed operators. " This statement is ,
added to proposed ITS 5.2.2f to provide clarification on who directs the activities of the licen8ed operators in the situation where the operations manager or the assistant operations. manager does not hold an SRO license. *
8. Proposed ITS 5.2.2g states in the last sentence "In addition, the STA shall meet the qualifications specified by. the Commission Policy Statement on Engineering Expertise on Shift. " This statement is not includec,l in the proposed Palisades ITS. The Administrative Controls Chapter of the Palisades Technical Specifications was recently changed in Amendment 174 dated October 31, 1996to be consistent with the format
and presentation presented in NUREG-1432. The NUREG-1432 statement was not included in the subject amendm~n~ and Palisades f~els that the existing wording in the CTS, including the qualiticatiOiis requirements in proposed ITS 5.3.3 adequately describes the STA's role and qualification requirements. This is a plant specific change based on the Palisades Nuelear Plant CTS. *
, .i '< : ' .' :~ *-:-' ;:
9. In proposed ITS 5. 2. :ig 'tile foilowing statement -is. added: "If either SRO on shift .
satisfies the Shift Engineer qualification requirements, then the STA does not need to be stationed." In CTS 6.3.3 and proposed ITS 5.3.3, the Shift Engineer is required to have a bachelor's degree in a~,scieJ:?.tif.ic, or engineei:µtg dtsciplipe and l}~ve, specific training in plant* design, operations, and response and* analysis of the plant for
transients and accidents. Because the training and qualifications of the Shift Engineer encompasses that of the STA, an SRO on shift who meets the Shift.Engineer qualifications provides the equivalent level of qualifications as *the STA, and therefore the STA does not need to be stationed. This is a plant specific change based on the Palisades Nuclear Plant CTS.
10.
NUREG 1432 Item 5.3.1 in the second sentence discusses "The staff not covered by
..... " The proposed ITS deletes this.second sentence and replaces it with three separate subsections discussing 1) the radiation safety manager qualifications; 2) the Shift
.Tecruiical Advisor and Shift Engineer qualifications; and 3)the qualification requirements for plant staff who perfomi 50.59 reviews. This change is consistent with the "Reviewer's Note" in Section 5.3 of NUREG-1432. This is a plant specific
change based on the Palisades. Nuclear Plant CTS .
Palisades Nuclear Plant Page 2of6 01/20/98
ATTaCHMENT 6
JUSTJI?ICATION FOR DEVIATIONS
. SPECIFICATION.5.0, ADMINISTRATIVE {:'.ONTROLS Change Discussion
11. NUREG-1432 contains Item 5.4.lc which requires that written procedures shall be established, implemented, and maintained for "Quality assurance for effluent and environmental monitormg." The Palisades Nuclear Plant CTS does not contain this requirement. The Administrative* Controls Chapter of the Palisades Nuclear Plant wa;Y recently changed in Amendment 174, dated October 31, 1996 to be consistent with the format and presentation of NUREG-1432. The subject NUREG-1432 statement was not included in this amendment and Palisades feels that the requirements for effluent and environmental monitoring are adequately prescribed in the ODCM.
Therefore, the subject wording is not included in the proposed Palisades ITS. This is .
a plant ~pecific change based on the* Palisades Nudear Plant CTS.
12. The proposed ITS 5.5.2 contains requirements for the program which addresses Primary Coolant Sources Outside Containment. In CTS 6.5.2 which contains the*.
same. program, *three additional aspects *of the program are .listed to address plant specific applications. Part "c" addresses the testing requirements for the shutdown.
cooling system that is outside the containment. Part "d" discusses the testing requirements for piping from valves CV-3029 .and CV-3030 to the discharge of the .
safety injection pumps and containment spray pumps. Part "e" discusses the leakage requirements from the recirculation heat removal systems' components. These items will be included in proposed ITS 5.5.2. These changes are plant specific changes to reflect the CTS and plant specific requirements.
13. The proposed ITS 5.5.3, Post Accident Sampling, is a which provides controls to ensure the capability to obtain and analyze samples under post accident conditions.
An additional requirement is included in the _proposed ITS which is taken from CTS 6.5.3, Post Accident Sampling, to ensure the capability to accurately determine the airborne iodine concentration in vital areas. This iS a plant specific change to retain the requirements contained in the CTS Post Accident Sampling program.
14. NUREG-1432 5.5.4 contains the Radioactive Effluent Controls Program. Items "e" and "f" of this program are not included in the proposed.ITS as they were not part of the Palisades Radioactive Effluent Controls Program contained iri. CTS 6.5.4. This change is a plant specific change to reflect the Palisades CTS. The Palisades
. Administrative Controls Chapter was recently changed in Amendment 174 dated October 31, 1996 to match the format and presentation of NUREG-1432. The CTS
does not contain NUREG-1432 items "e" and "f" and therefore, they are not included in the proposed ITS .
Palisades Nuclear Plant Page.3 of 6 . 01/20/98
ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 5.0, ADMINISTRATIVE CONTROLS
.Change Discussion
23. NUREG-1432 5.5'.6 specifies the requirements for the Tendon Surveillance Program.
The proposed ITS revises this title to the Containment Structural Integrity Surveillance Program and replaces the plant specific requirements which currently exists in CTS 4.5.4 and 4.5.5. This program and the associated report (ITS 5.6.7), include requirements relating to dome delamination in addition to tendon testing. The program and report names were changed accordingly. This change is acceptable because the
NUREG-1432 5.5.6 is in brackets to indicate that the plant specific information is to be provided if the report is applicable. . *
24. NUREG-1432 5.5.9 and 5.5.13 are revised to incorporate TSTF-118 which provides consistent application of SR 3.0.2 and SR 3.0.3 to the Programs referenced by ITS .
SRs. -
25. . NUREG-1432 5,6.l and 5.6.3 are revised to incorporate TSTF-152 which reflects previous revisions to 10 CFR Part 20 and 10 CFR 50.36a.

P~isades Nuclear Plant Page 6of6 01/20/98
ATTACHMENT 6 *
. JUSTIFICATION FOR DEVIATIONS SPECIFICATION 5.0, ADMINISTRATIVE ~CONTROLS Change
Discussion
15. NUREG-1432 Section 5.5.5 contains the program requirements for the Component Cyclic or Transient Limits. The Palisades CTS does not contain an equivalent program and therefore, it is not included in the proposed ITS.
16. NUREG-1432 Section 5.5.7 contains the program requirements for the Reactor Coolant Pump Flywheel Inspection Program. The NUREG-1432 program references .
Regulatory Guide 1.14 .. The proposed ITS 5.5.6, Primary Coolant Punlp Flywheel Inspection Program, is based on CTS 6.5.6. The proposed ITS program requires "Surveillance of the primary coolant pump flywheels shall consist of a 100 %
volumetric inspection of the upper flywheels each 10 years." This wording is a d:ifect transfer of the CTS requirements in 6.5.6, as modified in the Palisades Technical Specification Change Request submitted October 1, 1997 .. These changes are plant .
specific changes to reflect the Palisades CTS requirements.
17. NUREG.:.1432 Section 5.5;11 contains the requirements'for the Ventilation Filter Testing Program (VFTP). In the proposed ITS 5.5.10.a footnote is added to the first
paragraph to modify the testing frequencies. The proposed footnote states:. "Should
the 720-hour limitation on charcoal adsorber operation occur during a plant operation
- requiring the use of the charcoal- such as refueling-testing rriay be delayed until the completion 9f the plant operation or-up to 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months of filter operation, whichever occurs first. " This note is used in CTS 6. 5 .10 which contains the requirements for the VFTP. This note is added to accommodate normal operational occurrences which otherwise might impose an unnecessary burden on plant operations. This change is a plant specific change to reflect the Palisades CTS.
18. NUREG-1432 Section 5.5.12 contains.requirements for the Explosive Gas and Storage Tank Radioactivity Monitoring Program. The Palisades CTS does not contain these requirements and therefore they are not included in the proposed ITS .
Palisades Nuclear Plant. Page 4 of 6 01/20/98
Change
. ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
. SPECIFICATION 5.0, ADMINISTRATIVE* CONTROLS Discussion 19.* Proposed ITS 5.5.14 contains the Containment Leak Rate Testing Program. This program-was added to NUREG-1432 as part of TSTF-52 for the implementation of OPTION B to 10 CFR 50 Appendix J. This program has already been adopted in th~
CTS in sectfon 6. 5 .14. The CTS program described in 6. 5 .14 is included as the proposed ITS 5.5.14, Containment Leak Rate Testing Program, with appropriate information from CTS 4.5.2 concerning test methods and acceptance criteria .
. The proposed ITS Containment Leak Rate :resting Program includes specific acceptance criteria for each airlock. These criteria and the associated exemption to 10 CFR 50 Appendix J were approved as Amendment 177 to the Palisades Technical Specifications on* September 30, 1997.
An additional paragraph was included to assure correct application of those 10 CFR 50 Appendix J testing requirements (e.g., III.D.2.(b)(ii)) which are applicable "when
containment integrity is required by the plant's Technical Specifications."
20.
NUREG-1432 5.6.6 contains the requirements for the Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR). The Palisades Nuclear Plant does not currently utilize a PTLR and does not propose to include one in the proposed ITS.
21. NUREG-1432 5.6.7 contains the requirements for an EDG failures report. The Palisades Nuclear Plant CTS does not have an EDG failures report and does not propose to include one in the proposed ITS. This report is shown in brackets in
NUREG-1432 to indicate. that plants which utilize this report should put in the plant*
specific information. .Therefore, it is not. applicable to the Palisades ITS.
22. NUREG-1432 5.6.9 specifies the requirements for the Tendon Surveillance Report. .
The proposed ITS revises this title to the Containment Structural Integrity Report and includes the plant specific information.which currently exists in CTS 6.6.8. This report and the associated program (ITS 5.5.5), include requirements relating to dome delamination in addition to tendon testing. The program and report names were changed accordingly. This change is acceptable because the NUREG-1432 5.6.9 is in
brackets to indicate that the plant specific information is to be provided if the report is applicable .
Palisades Nuclear Plant Page Sof 6 01/20/98

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