ML070740653

License Amendment Request: Changes to Technical Specification Use and Application and Administrative Controls Sections

ML070740653
Person / Time
Site:	Palisades
Issue date:	03/15/2007
From:	Harden P Nuclear Management Co
To:	Document Control Desk, NRC/NRR/ADRO
References
Download: ML070740653 (38)

Text

N Committed to Nuclear Excellent Palisades Nuclear Plant Operated by Nuclear Management Company, LLC March 15, 2007 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 License Amendment Request: Changes to Technical Specification Use and Application and Administrative Controls Sections Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC) requests Nuclear Regulatory Commission (NRC) review and approval of a proposed license amendment for the Palisades Nuclear Plant (PNP).

The proposed amendment affects the following Technical Specifications (TS) sections:

Use and Application, and Administrative Controls.

Proposed changes to TS 1.4 incorporate NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Changes TSTF-284, "Add 'Met vs.

Perform' to Specification 1.4, Frequency," revision 3, TSTF-485-A, "Correct Example 1.4-1," revision 0 and make administrative changes. Proposed changes to TS section 5 incorporate NRC-approved TSTF-258, "Changes to Section 5.0, Administrative Controls," revision 4, and NRC-approved TSTF-273, "[Safety Functions Determination Program] SFDP Clarifications," revision 2, as amended by Westinghouse Owners Group (WOG) editorial change WOG-ED-23 and make administrative changes. provides a detailed description of the proposed changes, background and technical analysis, No Significant Hazards Consideration Determination, and Environmental Review Consideration. Enclosure 2 provides the revised TS pages reflecting the proposed changes. Enclosure 3 provides the annotated TS pages showing the proposed changes.

27780 Blue Star Memorial Highway Covert, Michigan 49043-9530 Telephone: 269.764.2000

Document Control Desk Page 2 NMC requests approval of this proposed license amendment by April 1, 2008, with the amendment being implemented within 60 days.

A copy of this request has been provided to the designated representative of the State of Michigan.

Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 15, 2007.

Paul A. Harden Site Vice-President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosures (3)

CC Regional Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC NRC Resident Inspector, Palisades USNRC

ENCLOSURE l DESCRIPTION OF REQUESTED CHANGES "I0 DESCRIPTION Nuclear Management Company, LLC (NMC) requests to amend Operating License DPR-20 for the Palisades Nuclear Plant (PNP).

The proposed amendment affects the following Technical Specifications (TS) sections: Use and Application and Administrative Controls.

Proposed changes to TS 1.4 incorporate NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Changes TSTF-284, "Add

'Met vs. Perform' to Specification 1.4, Frequency," revision 3, TSTF-485-A, "Correct Example 1.4-1," revision 0 and make administrative changes. Proposed changes to TS section 5 incorporate NRC-approved TSTF-258, "Changes to Section 5.0, Administrative Controls," revision 4, and NRC-approved TSTF-273,

"[Safety Functions Determination Program] SFDP Clarifications," revision 2, as amended by Westinghouse Owners Group (WOG) editorial change WOG-ED-23 and make administrative changes.

2.0 PROPOSED CHANGE

S NMC proposes to revise TS as follows:

TS 1.4, "Frequency" Add text and examples consistent with TSTF-284 to facilitate the use and application of Surveillance Requirement (SR) notes that use the terms "met" and "perform." The examples provide clarification on typical suweillance notes that allow for the SR to "not be required" to be performed.

Revise example 1.4-1 to be consistent with the requirements of SR 3.0.4.

SR 3.0.4 was revised in Amendment 219 to incorporate TSTF-359, "Increased Flexibility in Mode Restraints," revision 9. The proposed changes are consistent with TSTF-485-A, which was developed to reconcile the inconsistencies generically.

Renumber existing pages 1.4-2 through 1.4-4, as 1.4-3 through 1.4-5.

Use the term "plant" instead of "unit" in applying TSTF-284 and TSTF-485 sections; TS at PNP currently make use of the word "plant" instead of "unit" (example: L C 0 3.0.3).

TS Section 5.0 Administrative Controls Incorporate changes in TSTF-258 related to control room staffing and qualifications. Specifically, NMC proposes the following:

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Delete part 5.2.2.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.

Add 5.3.5:

For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed reactor operator (RO) are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m).

NMC also proposes to remove the 10 CFR 50.59 reviewer qualifications from TS. Specifically, NMC proposes the following:

Delete 5.3.4:

The plant staff who perform reviews which ensure compliance 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 considered equivalent to a bachelors degree for the purpose of this specification.

Revise TS 5.5.13, Safety Functions Determination Program (SFDP), to add text consistent with TSTF-273, to clarify the intent of L C 0 3.0.6 if a single inoperable TS support system makes both redundant subsystems of a supported system inoperable (a loss of safety function condition). Add clarification for determining when a loss of safety function condition exists and what L C 0 Actions are required to be taken when a safety function is lost.

Revise TS 5.5.15.b.l to use a generic title for the Quality Assurance (QA) program. Specifically, NMC proposes to replace the reference "CPC-2A" with a reference to the "Quality Assurance Program."

Markups showing the specific changes and the clean TS pages with those changes incorporated are provided in Enclosure 2 and 3, respectively.

3.0 BACKGROUND

NMC has identified several applicable TS improvements. NMC seeks to obtain benefits of approved TSTFs and improve the content and presentation of TS Use and Application, and Administrative Controls sections.

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Amendment 189 to Operating License No. DPR-20, dated November 30, 1999, converted Palisades Nuclear Plant (PNP) TS to Improved Technical Specifications (ITS). The PNP ITS conversion did not include the changes to the TS Use and Application section addressed by TSTF-284.

Amendment 196 to Operating License No. DPR-20, dated May 3, 2001, incorporated the majority of TSTF-258, revision 4, but did not include the proposed changes to TS 5.2.2.b and TS 5.3.5.

4.0 TECHNICAL ANALYSIS

TS 1.4, "Frequencv" The proposed changes to TS 1.4 revise SRs as necessary to appropriately use "met" and "perform" exceptions. The proposed changes insert a discussion and new examples to facilitate the use and application of SR Notes that use the terms "met" and "perform". These changes are considered clarifications that provide explicit direction and alleviate misunderstanding. They do not change requirements or affect operation of the plant. The changes are consistent with NRC-approved TSTF-284, revision 3, with the following deviation. The term "plant" instead of "unit" is used as TS at PNP currently make use of the word "plant" instead of "unit" (example: L C 0 3.0.3).

Amendment 219 to Operating License No. DPR-20, dated November 10,2004, implemented TSTF-359-A, "Increased Flexibility in Mode Restraints," revision 9.

TSTF-359 revised SR 3.0.4 such that certain statements in Example 1.4-1 are inconsistent. The proposed change to modify the discussion of TS 1.4 SR Example 1.4-1 provides clarification to the discussion on surveillances that exceed the interval without being performed while in the applicability. The proposed change is considered administrative in that it modifies the example to demonstrate the proper application of SR 3.0.4 and L C 0 3.0.4. This is necessary as SR 3.0.4 provides that performance of a missed surveillance may have been extended and prior to performance of the missed surveillance, but within the time permitted under SR 3.0.3, a mode change occurs. The proposed change to modify the discussion of TS 1.4 SR Example 1.4-1 is consistent with NRC approved TSTF-485, revision 0.

TS Section 5.0 Administrative Controls The proposed deletion of TS 5.2.2.b removes specific shift staffing requirements from TS. The requirements of 10 CFR 5OS54(m)(2)(iii)and 50.54(k) adequately provide for shift manning. 10 CFR 50.54(m)(2)(iii) requires, "when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a Page 3 of 8

licensed operator or senior operator shall be present at the controls at all times."

Further, 10 CFR 50.54(k) requires, "An operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during the operation of the facility." Current TS 5.2.2.b requirements are already met through compliance with these regulations and are not required to be duplicated in TS. The proposed deletion of TS 5.2.2.b is consistent with NRC approved TSTF-258, revision 4 and NU REG-1432, revision 3.

The proposed addition of TS 5.3.5 provides clarification to specific staffing requirements. Definitions in 10 CFR 55.4 state, "Actively performing the functions of an operator or senior operator means that an individual has a position on the shift crew that requires the individual to be licensed as defined in the facility's technical specifications, and that ...." The proposed addition of TS 5.3.5 clarifies that for the purposes of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m). The proposed addition of TS 5.3.5 is consistent with NRC approved TSTF-258, revision 4 and NUREG-1432, revision 3.

The proposed deletion of TS 5.3.4 removes 10 CFR 50.59 reviewer qualifications from TS. The qualifications to perform reviews in compliance with 10 CFR 50.59 are duplicated in an owner controlled document. The approved Quality Assurance Topical Report (QATR) delegates the Plant Operating Review Committee (PORC) to review written 10 CFR 50.59 evaluations to verify that changes to the facility or procedures, tests or experiments do not involve a change in the Technical Specifications or require prior NRC review. Therefore, the qualifications are not required to be duplicated in TS.

The proposed changes to TS 5.5.1 3 clarify the requirements for the SFDP. The proposed changes clarify that the Actions for a single support system inoperability are addressed by that support system's Actions, without cascading to the supported system The proposed changes clarify the SFDP to be consistent with L C 0 3.0.6. By clarifying the intent of the existing requirements of the SFDP, these changes remove an ambiguity that could lead to a misinterpretation of those requirements. The proposed changes to TS 5.5.13 are administrative. The proposed changes to TS 5.5.13 are consistent with NRC approved TSTF-273, revision 2, as amended by WOG editorial change WOG-ED-23 and NUREG-1432, revision 3.

The proposed change to TS 5.5.15.b.1 removes the title of the former quality program. This change is necessary due to NRC approval of NMC1sQATR by letter dated March 24, 2005. The proposed change to add a generic title facilitates future quality program name changes by eliminating the need to process a license amendment application. Changes to the Quality Assurance Program (QAP) are processed per 10 CFR 50.54(a). NRC approval is required for changes involving a reduction in commitment to the QAP. The proposed Page 4 of 8

change is not a change to the QAP and is not being submitted for review under 10 CFR 50.54(a).

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Si~nificantHazards Consideration Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC) requests to amend Operating License DPR-20 for the Palisades Nuclear Plant.

The proposed change would revise Appendix A, Technical Specifications (TS), to reflect Technical Specification Task Force (TSTF) Standard Technical Specification (STS) changes TSTF-284, "Add 'Met vs. Perform' to Specification 1.4, Frequency," revision 3, TSTF-485, "Correct Example 1.4-1," revision 0, TSTF-258, "Changes to Section 5.0, Administrative Controls," revision 4, and TSTF-273, "SFDP Clarifications," revision 2, as amended by Westinghouse Owners Group (WOG) editorial change WOG-ED-23. NMC also proposes to make administrative changes.

NMC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Response: No.

The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative or provide clarification only.

The proposed changes do not have any impact on the integrity of any plant system, structure, or component that initiates an analyzed event.

The proposed changes will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Thus, the probability of any accident previously evaluated is not significantly increased.

The proposed changes do not affect the ability to mitigate previously evaluated accidents, and do not affect radiological assumptions used in the evaluations. The proposed changes do not change or alter the design criteria for the systems or components used to mitigate the consequences of any design basis accident. The proposed amendment does not involve operation of the required structures, systems, or components (SSCs) in a manner or configuration different from those previously recognized or evaluated. Thus, the radiological consequences of any accident previously evaluated are not increased.

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Therefore, operation of the facility in accordance with the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not involve a physical alteration of any SSC or a change in the way any SSC is operated. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not affect any margin of safety.

The proposed amendment does not involve any physical changes to the plant or manner in which the plant is operated.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

Based on the evaluation above, NMC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

5.2 Apolicable Regulatorv Requirementslcriteria The function of the "Administrative Controls" section of the TS, as stated in 10 CFR 50.36(~)(5),is to provide "provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." The proposed changes continue to meet these objectives.

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The proposed changes are consistent with NUREG-1432 and approved TSTFs.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public..

6.0 ENVIRONMENTAL CONSIDERATION

NMC has determined that the proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(lO). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. TSTF-258, "Changes to Section 5.0, Administrative Controls," Revision 4.

2. TSTF-273, "SFDP Clarifications," Revision 2 as amended by editorial change WOG-ED-23.

3. TSTF-284, "Add 'Met vs. Perform' to Specification 1.4, Frequency,"

Revision 3.

4. TSTF-359, "Increased Flexibility in Mode Restraints," Revision 9.

5. TSTF-485, "Correct Example 1.4-1," Revision 0.

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8.0 PRECEDENT By letter dated January 28,2004 (ADAMS Accession #ML040410564), as supplemented by letter dated November 22, 2004 (ADAMS Accession

1. ML043380203), NMC submitted a license amendment request (LAR) for Duane Arnold Energy Center (DAEC). The LAR requested changes to the DAEC TS, specifically, changes that implement TSTF-273, TSTF-284 and other TSTFs. By letter dated May 12,2005 (ADAMS Accession #ML051110692), the NRC approved the LAR for DAEC. Similar to this submittal, NMC is requesting approval to implement TSTF-273 and the portions of TSTF-284 not previously addressed with the incorporation of Amendment 189 in PNP TS. NMC's PNP submittal differs from NMC's DAEC submittal in that DAEC deviated from TSTF-273 and PNP did not. Also, DAEC completely implemented TSTF-284 and PNP's proposal is for partial implementation of TSTF-284 with deviations.

As previously discussed, NMC's PNP deviations are administrative.

By letter dated January 20,2005 (ADAMS Accession #ML050260253 and

1. ML050250313), as supplemented by letter dated July 5, 2005 (ADAMS Accession #ML052070688), Florida Power and Light Company (FPL) submitted a LAR for Turkey Point Units 3 and 4 (Turkey Point). The LAR requested changes to the Turkey Point TS, specifically, changes that implement TSTF-258 and other TSTFs. By letter dated May 26,2006 (ADAMS Accession

1. ML061080598), the NRC approved the LAR for Turkey Point. Similar to this submittal, NMC is requesting approval to implement the portions of TSTF-258 not previously addressed with the incorporation of Amendment 196 in PNP TS.

NMC's submittal differs from FPL's submittal in that FPL completely implemented TSTF-258 with a deviation. NMC proposes to implement a portion of TSTF-258 with no deviations.

By letter dated September 28,2006 (ADAMS Accession #ML062830057),

Arizona Public Service Company (APS) submitted a LAR for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3. The LAR requested changes to the PVNGS TS, specifically, changes that implement TSTF-485 and another TSTF. By letter dated February 21,2007 (ADAMS Accession #ML070350039),

the NRC approved the LAR for PVNGS. Similar to this submittal, NMC is requesting approval to implement 485. NMC proposes to implement TSTF-485 with no deviations.

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ENCLOSURE 2 LICENSE AMENDMENT REQUEST: CHANGES TO TECHNICAL SPECIFICATION USE AND APPLICATION AND ADMINISTRATIVE CONTROLS SECTIONS REVISED TECHNICAL SPECIFICATION PAGES 1.4-1, 1.4-2, 1.4-3, 1.4-4, 1.4-5, 1.4-6, 1.4-7, 1.4-8 5.0-2, 5.0-4, 5.0-20, 5.0-2 1, and 5.0-23 AND OPERATING LICENSE PAGE CHANGE INSTRUCTIONS 14 Pages Follow

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Remove the following pages of Appendix A Technical Specifications and replace with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT I .4-4 1.4-5, 1.4-6, 1.4-7, and 1.4-8

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.

Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LC0 is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" and "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

Palisades Nuclear Plant 1.4-1 Amendment No. 4-89,

Frequency 1.4 1.4 Freauencv DESCRIPTION Some Surveillances contain notes that modify the Frequency of (continued) performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated L C 0 if any of the following three conditions are satisfied:

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or

b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or

c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.

Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.

EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the L C 0 (LC0 not shown) is MODES 1,2, and 3.

Palisades Nuclear Plant 1.4-2 Amendment No. 4-89, 1

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)

SURVEILLANCE REQUIREMENTS I

SURVEILLANCE I FREQUENCY Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the plant is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the plant is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the L C 0 for which performance of the SR is required, then SR 3.0.4 becomes I applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the L C 0 is considered not met (in accordance with SR 3.0.1) and L C 0 3.0.4 becomes applicable.

Palisades Nuclear Plant 1.4-3 Amendment No. 4-89, I

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 2 25% RTP 1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level

< 25% RTP to 2 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND").

This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to

< 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

Palisades Nuclear Plant 1.4-4 Amendment No. 4-89,

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS 1

SURVEILLANCE I FREQUENCY NOTE............................

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after r 25% RTP.

Perform channel adjustment. 1 7 days The interval continues, whether or not the plant operation is < 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches 2 25% RTP to perform the Surveillance.

The Surveillance is still considered to be performed within the "specified Frequency." The interval continues, whether or not the plant operation is

< 25% RTP between performances. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with power r 25% RTP.

Once the plant reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

Palisades Nuclear Plant 1.4-5 Amendment No. I

Frequency 1.4 1.4 Freauencv EXAMPLES EXAMPLE 1.4-4 (continued)

SURVEILLANCE REQUIREMENTS I

SURVEILLANCE I FREQUENCY /

............................ NOTE............................

Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the plant is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), but the plant was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would require satisfying the SR.

Palisades Nuclear Plant 1.4-6 Amendment No. I

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-5 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

............................ NOTE............................

Only required to be performed in MODE 1.

Perform complete cycle of the valve. 7 days The interval continues, whether or not the plant operation is in MODE 1 , 2, or 3 (the assumed Applicability of the associated LCO) between performances.

As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1.

Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.

Once the plant reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

Palisades Nuclear Plant 1.4-7 Amendment No. I

Frequency 1.4 1.4 Frequencv EXAMPLES EXAMPLE 1.4-6 (continued)

SURVEILLANCE REQUIREMENTS I I SURVEILLANCE FREQUENCY

---

em------- NOTE............................

Not required to be met in MODE 3.

Verify parameter is within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the plant is in MODE 3 (the assumed Applicability of the associated L C 0 is MODES 1,2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), but the plant was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would require satisfying the SR.

Palisades Nuclear Plant 1.4-8 Amendment No. I

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Orqanizations 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.

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 1, 2, 3, or 4.

b. (Deleted) I Palisades Nuclear Plant 5.0-2 Amendment No. 4-89,

Plant Staff Qualifications 5.3 5.0 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 except for the education and experience eligibility requirements for operator license applicants, and changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation dated October 24, 2003.

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: (a) Formal schooling in science or engineering, or (b) operational or technical experience and training in nuclear power.

5.3.3 The individual, required by Specification 5.2.2g1assigned to provide advisory technical support to the plant operations shift crew, shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift (Published in Federal Register 50 FR 43621, October 28, 1985).

5.3.4 (Deleted) 5.3.5 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed reactor operator (RO) are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m).

Palisades Nuclear Plant 5.0-4 Amendment No. 4-89, 4-%,=212-,

Programs and Manuals 5.5 5.5 Proarams and Manuals Technical Specifications (TS) Bases Control Program (continued)

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 Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

Safety Functions Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LC0 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 LC0 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 supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), 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 Palisades Nuclear Plant 5.0-20 Amendment No. 449,

Programs and Manuals 5.5 5.5 Proarams and Manuals 5.5.1 3 Safetv Functions Determination Proqram (SFDP) (continued)

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 L C 0 in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.1 4 Containment Leak Rate Testina Proqram

a. A program shall establish the leakage 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. This program shall be in accordance with the guidelines of Regulatory Guide 1.I 63, "Performance-Based Containment Leakage-Test Program," dated September 1995, as modified by the following exceptions:

1. Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or post-test adjustment of the air lock door seals. However, a seal contact check shall be performed instead.

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.

2. Leakage rate testing at Pa is not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at 210 psig instead.

3. Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on component performance.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 53 psig. The containment design pressure is 55 psig.

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

Palisades Nuclear Plant 5.0-21 Amendment No. 439,494,

Programs and Manuals 5.5 5.5 Proarams and Manuals 5.5.1 5 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 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 Assurance Program. 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.

Palisades Nuclear Plant Amendment No. W , W ,

ENCLOSURE 3 LICENSE AMENDMENT REQUEST: CHANGES TO TECHNICAL SPECIFICATION USE AND APPLICATION AND ADMINISTRATIVE CONTROLS SECTIONS MARK-UP OF TECHNICAL SPECIFICATION PAGES 1.4-1, 1.4-3, 1.4-4, 5.0-2, 5.0-4, 5.0-20, 5.0-21, and 5.0-23 (showing proposed changes)

(additions are highlighted; deletions are strikethrough) 12 Pages Follow

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.

[INSERTc-1 a bz rqskcc! (: ."., 'its bc! p c r ( " ' m n r ( d LCC Tn hn U,

CR "I

' Q A i* I

• 2 I V.V. t

. . w EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LC0 (LC0 not shown) is MODES 1,2, and 3.

Palisades Nuclear Plant 1.4-1 Amendment No. 4-89>

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)

SURVEILLANCE REQUIREMENTS I

SURVEILLANCE I FREQUENCY Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the plant is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the plant is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the plant is not in a MODE or other specified condition in the nce of the SR is required e Surveillance must be MODE- or other s~ecifie

,,C P, 3.0.4. or the L C 0 is considered not met (in accordance with SR 3.0.1) and L C 0 3.0.4 becomes applicable.

Palisades Nuclear Plant 1.4-23 Amendment No. 4-89,

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE ............................

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 2 25% RTP.

Perform channel adjustment. 7 days The interval continues, whether or not the plant operation is < 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is c 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches 2 25% RTP to perform the Surveillance.

The Surveillance is still considered to be performed within the "specified Frequency." The interval continues, whether or not the plant operation is

< 25% RTP between performances. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with power r 25% RTP.

Once the plant reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

[INSERT 21 [INSERT 31[INSERT 41 Palisades Nuclear Plant 1.4-45 Amendment No. 4-89,

INSERT 1 Sometimes special situations dictate when the requirements of a Surveillance are to be met.

They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated L C 0 is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" and "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated L C 0 if any of the following three conditions are satisfied:

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or

b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or

c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.

Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.

INSERT 2 EXAMPLES EXAMPLE 1.4-4 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY

............................ NOTE............................

Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the plant is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), but the plant was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would require satisfying the SR.

INSERT 3 EXAMPLES EXAMPLE 1.4-5 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY

............................ NOTE............................

Only required to be performed in MODE 1.

Perform complete cycle of the valve. 7 days The interval continues, whether or not the plant operation is in MODE 1,2, or 3 (the assumed Applicability of the associated LCO) between performances.

As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1.

Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.

Once the plant reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

INSERT 4 EXAMPLES EXAMPLE 1.4-6 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

............................ NOTE............................

Not required to be met in MODE 3.

Verify parameter is within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the plant is in MODE 3 (the assumed Applicability of the associated LC0 is MODES 1,2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), but the plant was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would require satisfying the SR.

Organization 5.2 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.

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 1, 2, 3, or 4.

Palisades Nuclear Plant 5.0-2 Amendment No. U,

Plant Staff Qualifications 5.3 5.0 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 except for the education and experience eligibility requirements for operator license applicants, and changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation dated October 24,2003.

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: (a) Formal schooling in science or engineering, or (b) operational or technical experience and training in nuclear power.

5.3.3 The individual, required by Specification 5.2.2g, assigned to provide advisory technical support to the plant operations shift crew, shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift (Published in Federal Register 50 FR 43621, October 28, 1985).

5.3.5 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed reactor operator (RO) are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50.54(m).

Palisades Nuclear Plant 5.0-4 Amendment No. W,W,212,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)

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 Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

5.5.13 Safety Functions Determination Proqram (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into L C 0 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 LC0 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 supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

r 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 Palisades Nuclear Plant 5.0-20 Amendment No. 4-89,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 3 Safety Functions Determination Program (SFDP) (continued)

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 5.5.14 Containment Leak Rate test in^ Proaram

a. A program shall establish the leakage 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. This program shall be in accordance with the guidelines of Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program," dated September 1995, as modified by the following exceptions:

1. Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or post-test adjustment of the air lock door seals. However, a seal contact check shall be performed instead.

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.

2. Leakage rate testing at Pais not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at 210 psig instead.

3. Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on component performance.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 53 psig. The containment design pressure is 55 psig.

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

Palisades Nuclear Plant 5.0-21 Amendment No. W , W ,

Programs and Manuals 5.5 5.5 Programs and Manuals 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 State regulations, and other requirements governing the disposal of the radioactive waste.

b. Changes to the Process Control Program:

1. Shall be documented and record erformed shall be retained as required by the Qualit 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.

Palisades Nuclear Plant 5.0-23 Amendment No. ~ , ~