ML20154N290

Proposed Tech Specs Sections 1.0,3.0,4.0 & 5.0,converting to Improved Tech Specs

ML20154N290
Person / Time
Site:	Palisades
Issue date:	10/12/1998
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18068A460	List: ML20154N290
References
NUDOCS 9810210243
Download: ML20154N290 (108)

Text

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CHAPTER 1.0 Vt 1.0-L INSERT (SHUTDOWN MARGIN-CTS)

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[piLaw c***Q j However, with all{ rods verified fully inserted by two independent means, it is notj necessary to account for a stuck rod in the SDM calculation. With anfrods not capable of being fully inserted, the reactivity worth of these rods must be accounted for in the determination of SDM; and b) There is no change in part length rod position.

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i CTS 1-4 9810210243 981012 PDR P

ADOCK 05000255 PDR g'7 ( .F f.

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DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION A.13 The CTS definition of SHUTDOWN MARGIN has the following wording added to it  ;

to from the proposed ITS definition of SHUTDOWN MARGIN (SDM):

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a. " ...

However, with aII(rods verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SDM calculation. With any GA W4 79 p, cept,] rods not capable of being fully inserted, the reactivity worth of these rods must i be accounted for in the determination of SDM; and

b. There is no change in part length rod position." I The first part of the additional wording clarifies that if it can be verified by two independent means that all rods are inserted, no penalty for a stuck rod needs to be incurred. In addition, no credit for part length rods is given in the SDM calculation, which is reflected in the analysis assumption that there is no change in part length rod position. . These changes are considered to be administrative changes as they are providing clarification on the calculation of SHUTDOWN MARGIN without changing the SHUTDOWN MARGIN requirements. This change is consistent with NUREG-1432. IMI f.o - L

% eer)

A.14 The CTS definition of REFUELING OPERATION forms the basis for the proposed ITS definition of CORE ALTERATION. In the CTS, the term " core components" is

. expanded in the ITS to define these components as "any fuel, sources, or control rods."

In addition, the clarifying phrase " CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position." is included in the proposed definition to ensure that there is no confusion over being able to complete movement of a core component if directed to " Suspend CORE ALTERATIONS."

These changes are considered to be administrativd changes since the term " CORE ALTERATIONS" is used to simply replace " REFUELING OPERATION" and provide additional clarification on its application. This change is consistent with NUREG-1432.

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

Insert for DOC A.13:

(RAI1.0-1)

CTS define the term CONTROL R00 as "all full-length shutdown and regulating rods". That definition has not been carried over to ITS (see DOC A.7).

Therefore, changing usage from " CONTROL ROOS" to " full length control rods" constitutes an administrative change.

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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 1.0, USE AND APPLICATION NRC REQUEST:

, 1.0-2 ITS 1.1, Leakage l DOC A.2-JFD 2 The proposed ITS definition for Leakage, item a.2, Identified Leakage, differs l from the STS. The STS markup shows, " Leakage into the containment atmosphere fra sources that are both specifically located and known either not to interfere with the operation of leakage detection systems er and not to be pressure boundary Leakage; or and." JFD 2 states that this is an editorial change for clarity or for consistency with the ITS Writer's Guide, and DOC A.2 states that the inclusion of the Leakage definition is an administrative change. While these statements are partially correct, the proposed ITS definition of Identified Leakage appears to be less restrictive then the STS definition.

Causent: Provide a DOC to address this change and/or revise the JFD to provide additional justification for the deviation from the STS.

Consumers Enerov Resnonse:

The proposed Leakage definition, as marked up from STS states, in part:

LEAKAGE shall be:

a. Identified LEAKAGE

b. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems er KidnottobepressureboundaryLEAKAGE;erahld With the STS definition, leakage from sources that are specifically located is

. classified as " Identified" if it is known to fit into either of the below listed categories. With the ITS definition, such leakage is classified as

" Identified" only if it is known to fit into bsth of those categories.

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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 1.0, USE AND APPLICATION

1) it does not interfere with the operation of leakage detection systems, or

2) it is not pressure boundary leakage.

4 Consider operation with Leakage from a source that is specifically located, which is known not to be Pressure Boundary Leakage, but might interfere with the operation of leakage detection systems:

With the STS definition, such leakage is classified as Identified Leakage. Both ITS and STS LCOs 3.4.13 limit Identified Leakage to 10 gpm. With the ITS wording, such leakage is excluded from Identified Leakage and must thereby be classified as Unidentified Leakage. Both ITS and STS LCOs 3.4.13 limit Unidentified Leakage to 1 gpm.

It is therefore considered that the proposed ITS wording is more restrictive.

It is also considered that the proposed wording implements the intent of the STS definition, and the change is therefore considered to be an administrative clarification as stated in JFD 2 for Section 1.0. Since the changed wording

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does alter the requirement associated with the definition, JFD 15 has been added.

Affected Subetttal Paats:

STS Markup page 1.1-4 Attachment 6 JFD Page 2 of 2 (AllofAttachment6replacedduetorepagination) l l

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Oefinitions 1.1 1.1 Definitions.

ENGINEERED SAFETY function (i.e , the valves travel to their FEATURE (ESF) RESP 0 E required posi ions, pump discharge ressures reach TIME their requir d values, etc.). Tim shall (continued) include die 1 generator starting nd sequence O6 loading del ys, where applicable. The response time may b measured by means of any series of sequentia , overlapping, or tot steps so that the entir response time is me ured.

L, The ma mumallowablecentinmentleakagera/e, L., sh 1 be (0.25]% of co tainment air wet t per day a the calculated pe containment pr sure

( P.) .

LEAKAGE LEAKAGE shall be:

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a. Identified LEAKAGE -

1. LEAKAGE, such as that(from pum seals or valve packing (except

• e olant )

E seal water in_tectior od eakoff)yGmp - , I that is captured and conductePttr--

collection systems or a sump or collecting )

tank;

2. LEAKAGE into the containment atmosphere I from sources that are both specifically located and known 6 not to interfere with the operation of leakage detection

{ h grr n be pressure boundary O'""E 3. M Coolant System ( S) LEAKAGE through a steam generator (SG) to the Secondary System.

b. Unidentified LEAKAGE (excefi M-seg Ca b N All LEAKAGE *that is not identified LEAKAGE; O\t ip.'9 l 5,J. teJr..ED 1. . Pressure Boundary LEAKAGE '

LEAKAGE (except SG LEAKA ) through a Oq l nonisolable fault in an S component body, pipe wall, or vessel wal (continued)

CEOG STS 1.1 4 Rev 1, 04/07/95

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1 ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 1.0, USE AND APPLICATION 8.

Performance of an ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME and REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME test is not part of the current Palisades Technical Specifications. A review during the NRC Systematic Evaluation Program, as stated in the resulting SER, concluded that the addition of

response time testing requirements was not necessary. This will be further discussed in the sections which deal with response time testing

9. The Palisades CTS has a Containment Leak Rate Test Program in Section 6.5.14. La I is defined in this program and it will be retained there in the proposed ITS. Therefore, L does not need to be defined in the Definitions sections. I 1

10. The proposed Palisades ITS does not include a definition for the PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) as Palisades does not propose to have  !

this report. The current pressure and temperature limits for Palisades are valid until l the end of reactor vessel life.

11. The proposed definition for " CORE ALTERATION" does not include the term t

" manipulation" as it is redundant to the discussion of "... movement of fuel or components." This change represents a generic change to NUREG-1432 proposed by the industry owners groups. This change was submitted and approved under TSTF-47.

12. The proposed definition for " Unidentified Leakage," which is found under the defined '

term " LEAKAGE," includes the phrase "(except primary coolant pump seal leakoff)."

This phrase was added to clarify that primary coolant pump seal leakoff should not be part of the amount included as " Unidentified Leakage." This change was presented as a generic change to NUREG-1432 proposed by the industry owners groups. This change was submitted under TSTF-40. The proposed Palisades implementation differs from the proposed only by the fact that the primary coolant pump seal injection portion of the phrase has been deleted since the Palisades pumps do not use seal injection.

13. The NUREG-1432 def'mition of SHUTDOWN MARGIN contains part 'b' which states that the fuel and moderator temperatures are changed to be nominal zero power design level. This statement is not appropriate for the methodology used at Palisades to calculate SHUTDOWN MARGIN. Therefore, part 'b' from NUREG-1432 is not included in the Palisades defimition for SHUTDOWN MARGIN.

14. The Palisades CTS contains the term " Quadrant Power Tilt (Tq)" and this term is also included in the proposed ITS. The Quadrant Power Tilt is defined as "Tq shall be the maximum positive ratio of the power generated in any quadrant minus the average quadrant power, to the average quadrant power."

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Palisades Nuclear Plant Page 2 of 2 01/20/98 3-b

New JFD 15:

(RAI1.0-2) .

1 The wording of the Identified Leakage definition has been altered to clarify that leakage which might affect the operation of leakage detection systems  !

must be classified as unidentified leakage. It is believed that this is the  !

intent of the STS definition. l I

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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION l SECTION 3.0, APPLICABILITY NRC REQUEST:

i 3.0-1 STS LCO 3.0.4 and SR 3.0.4 JFD 6 l

JFD 6 states in part, "Rev. 1 to NUREG-1432 contains the option of limiting the Applicability of LC0 and SR 3.0.4 to entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4. In addition, the Rev. I version of LC0 and SR 3.0.4 would not prevent entry into MODES or other specified conditions in the Applicability that are part of any shutdown of the unit. However, to adopt the Rev.1 version of LCO and SR 3.0.4, an evaluation must be perfomed on the existing technical specifications to detemine where specific restrictions on MODE changes or Required Actions should be included ,

in individual LCOs to justify this change. For the Palisades plant there l appears to be little benefit to adopt the Rev.1 definition. In addition, since the process for detemining where specific restrictions on MODE changes i should occur has not been well defined, Palisades chooses not to adopt the allowances of Rev. 1 LC0 and SR 3.0.4." l Cassent: Provide an explanation of why there appears to be little benefit for Palisades to adopt the NUREG-1432, Rev.1, definition.

Consumers Enerav Retd 0nst:

There is little operational benefit to Palisades in adopting the NUREG-1432, Rev.1 LC0 3.0.4 or SR 3.0.4 wording because there are no instances within the ITS where the Rev. O LCO 3.0.4 or SR 3.0.4 wording proposed in'the ITS would prevent a plant shutdown. Similarly, the restriction of LC0 3.0.4 and SR 3.0.4 to Modes 1, 2, 3, and 4, has little impact.

During preparation of ITS Section 3.0, a review of the ITS LCOs and Actions was perfomed to locate any instances where LC0 3.0.4 would prohibit performance of a plant shutdown due to failure to meet the requirements of another LCO. No such instances *are found. During preparation of this letter, that review was repeated, with the same result.

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I CONVERSION TO IMPROVED TECNNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.0, APPLICABILITY CTS wording for LCO 3.0.4 and SR 4.0.4 are equivalent to STS Rev. 0 wording for LC0 3.0.4 and SR 3.0.4. Since the additional allowances provided by the STS Rev. I wording of these specifications provide little operational benefit, those allowances have not been proposed.

Affacted Paaes:

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L ATTACIBfENT 3

~ DISCUSSION OF CHANGES

, CHAPTER 5.0, ADMINISTRATIVE CONTROLS i

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I RELOCATED (R) l l There were no " Relocated" changes made to this chapter.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS ADMINISTRATIVE CHANGES 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 reformatting, renumbering, and rewording of l

Technical Specificatione 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 l 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 relocate any technical requirements of the current i 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 - I represent a significant hazards consideration. The following is provided in support of this I conclusion.

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

The proposed changes involve reformatting, renumbering, and 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.

l l Palisades Nuclear Plant Page 1 of 7 10/10/95

ATTACHMENT 4

~ NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS

2. 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 convening to the Improved Technical Specifications (ITS) as outlined 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 7 10/10/98

1 ATTACHMENT 4

~ NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS

1. 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 er.isted 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.

i 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 i different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of safety? 1 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 7 10/10/98

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l ATTACHMENT 4 i

~ NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS l

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 even!. 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 previously evaluated.

Palisades Nuclear Plant Page 4 of 7 10/10/98

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l ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAl'rER 5.0, ADMINISTRATIVE CONTROLS l

2. Does the change create the possibility of a new or different kind of accident from l 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. There 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 (L)

The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as l outlined in NUREG-1432, " Standard Technical Specifications, Combustion Engineering l Plants." Changes have been proposed which involve making the requirements in the Current l Technical Specifications (CTS) less restrictive. A description of the less restrictive change l and corresponding No Significant Hazards Consideration are provided on the following pages l for each Specification as applicable. l l

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Palisades Nuclear Plant Page 5 of 7 10/10/98

A'ITACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS LESS RESTRICTIVE CHANGE L.1 l l

The CTS requirements for Type B and C leak rate testing are being revised such that the l .

leakage limit for Type B and C testing is s .60 La only during the first plant startup l l following testing performed in accordance with the Containment Leak Rate Test Program. l l

After this, the new limit will now become s 1.0 La. This means that if the testing is l performed during a refueling outage, the total Type B and C leakage must be "as-left" at s l

.60 La prior resuming power operations. Following this, the leakage limit for the remainder 1 of time until the test is performed again becomes 1.0 La for the total containment leakage. l Overall containment integrity is maintained because the results of Type B and C testing must l be compared against the overall containment leakage limit to ensure that the leakage remains l s 1.0 La. l l

CTS 4.5.2.c(1) states "The total leakage from all penetrations and isolation valves shall not I  ;

exceed 0.60 I,." In the proposed ITS, the acceptance criteria and testing frequency will l 1

only exist in the Containment Leak Rate Testing Program which is found in TS l l Administrative Controls Section 5.5.14. This is similar to the CTS Containment Leak Rate l l Testing Program which is found in CTS 6.5.14. The acceptance criteria in the Containment 'l Leak Rate Testing Program will be that the overall containment leakage will not exceed 1.0 l I2. These changes are acceptable since the overall containment leakage requirements of ITS l 1 3.6.1, which reference the Containment Leak Rate Testing Program for the acceptance l criteria, remain valid at all times. Any increase in Type B and C leakage would have to be l evaluated against this limit as evidenced by NOTE 3 in proposed ITS 3.6.2 and 3.6.3. The l change is considered Less Restrictive since the acceptance criteria of .60 La for Type B and l C tests will now become 1.0 La. This change is consistent with the information in the CTS l 6.5.14, Containment Leak Rate Test Program but is less restrictive than the information I contained in the CTS 4.5.2 section which addresses the local leak rate testing. This change l maintains consistency with NUREG-1432 as modified by the intent of industry owner's group l generic change TSTF-52. l l

l Palisades Nuclear Plant Page 6 of 7 10/10/98

1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS

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

The proposed change increases the acceptance criteria for Type B and C tests from l

.60 La to 1.0 La for the total containment leakage. Previously the .60 La for Type B l and C tests acted as a " trigger point" to ensure actions were taken such that the l overall acceptance criteria of 1.0 la were not violated. In addition the actions to I initiate immediate repairs are not required unless total containment leakage exceeds l 1.0 La. The 1.0 la limit for total containment leakage remains in the proposed ITS. 1 Any leakage from Type B and C tests which would put total containment leakage over l .

1.0 la must still be evaluated. Therefore, there is no significant increase in the l probability or consequences of an accident previously evaluated. l I

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

The proposed change does not involve a physical alteration of the plant. No new or j different type of equipment will be installed or changes made to plant parameters l which govern normal plant operation. The proposed change will continue to ensure l total containment leakage is monitored to ensure that it stays within the bounds of the l analysis. Thus, this change does not create the possibility of a new or different kind l of accident from any accident previously evaluated. l l

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

The proposed change preserves the total containment leakage rate of 1.0 La but does l not require actions to be taken once the Type B and C tests exceed .60 La as long as I the contribution of the Type B and C tests do not make the total containment leakage l to exceed 1.0 La. Therefore, this change does not involve a significant reduction in a l margin of safety. I I

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i Palisades Nuclear Plant Page 7 of 7 10/10/98 i

l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO TNE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.0, APPLICABILITY NRC RE0t/EST:

3.0-2 ITS LC0 and SR 3.0.4 JFD 6 TSTF-103, Rev. 1 (Not Approved as of 08/21/98)-

TSTF-103 ITS LC0 3.0.4 and SR 3.0.4 include TSTF-103 changes. JFD 6 dates in part,  ;

"When Rev. 1 to NUREG-1432 was written, the wording in LC0 and ST 3.0.4 was  !

written to reflect the Rev.1 allowances discussed above [see staff comment 3.0-1 above] without using brackets to indicate acceptable alternative wording if the Rev. I approach to implementing LC0 and SR 3.0.4 was not taken. To correct this, the industry owner's groups wrote TSTF-103 to add brackets to 1 the areas which discussed LCO and SR 3.0.4 being only applicable in MODES 1, 2, 3, and 4, and provided alternative wording where the Rev. 1 version '

discussed "any unit shutdown." The proposed wording in the Palisades ITS for LCO and SR 3.0.4 is modeled after the changes made in TSTF-103 along with some proposed changes to TSTF-103 to correct some consistency errors."

Consent: TSTF-103 Rev. 1, has not been approved (as of August 21,1998). If l this TSTF remains unapproved at the time the Palisades review is completed and the Palisades Safety Evaluation (SE) is being prepared, the ITS will have to l return to the CTS, or new justification for this change will need to be prepared.

Consumers Enerav ReSDOnSt:

When the ITS were submitted in January 1998, it was expected that TSTF-103 would be approved in the near future. As of this writing, it has not yet been approved. The affected pages of our January 29, 1998 conversion request have been revised to eliminate usage of the changes proposed in TSTF-103.

Affected Paats:

ITS pages 3.0-2 and 3.0-5 ITS Bases pages B 3.0-6, B 3.0-15, B 3.0-16 (All bases pages replaced due to repagination)

. Marked up STS pages 3.0-2 and 3.0-5 Marked up STS bases page B 3.0-6, B 3.0-6 insert, B 3.0-16, B 3.0-16 insert

!. Attachment 6. JFD Page 2 of 4

! (All of Attachment 6 replaced due to repagination) -

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LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time.

This Spcification shall not prevent changes in MODES or other specified conditions in the Applicability tha,t are required to comply with ACTIONS torja shutfown performea in p

[resgehse to JAe exp6cted f4ilureAo compfy withACTIONS.

Exceptions to this Specification are stated in the M individual Specifications.

LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LC0 3.0.6 When a supported system LC0 is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Palisades Nuclear Plant 3.0-2 Amendment No. 01/20/98 1

h -0 J

)

SR Applicability 3.0 1

3.0 SR APPLICABILITY t

! SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition (s) must be entered.

1 1

SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LC0's '

Surveillances have been met within their specified '

Frequency. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONSlor snutdow d in ponse he expepted tai re to com y with RAI 3D l 1

i

! Palisades Nuclear Plant 3.0-5 Amendment No. 01/20/98 ts-b t

LCO Applicability B 3.0 BASES LCO 3.0.4 The provisions of LCO 3.0.4 shall not prevent changes in (continued) MODES or other specified conditions in the Applicability that are reauired to comolv with ACTIONS.l In addit on, the

~

provi ions of CO 3.0.4 hall n -prevenf changes 'n MODES or o er spe fied con tions i the Ap ,ju ,-

res t from

$icabilit that shutdow perfo d in rey )onse to e excepted fai ure to omply wi ACTIO . /

Exceptions to LCO 3.0.4 are stated in the individual Nh3I 3'O ' "

Specifications. The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for

, continued operation for an unlimited period of time.

1 Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification.

4 LCO 3. 4 is appl able when en ering all MOD S or other speci ied condit ons, whether nereasing in ODES (e.g.,

MOD 5 to MODE ) or decreas g in MODES ( .g., MODE 4 t MO 5). The equirements ecluding ent into anoth .J2,-

M E or othe specified co ition when t associate

,. TIONS do ot provide fo continued op ration for nlimited eriod of time ensures that he plant m ntains sufficie equipment OP RABILITY and edundancy s assumed in the cident analy s.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing

, MODES or other specified conditions while in an ACTIONS l Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do'not have to be performed due to the associated inoperable equiprent. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

]

i.

a i

4 4

1 Palisades Nuclear Plant B 3.0-6 01/20/98 D ~C-

l l l l SR Applicability B 3.0 l

BASES SR 3.0.4 'SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified Condition in the Applicability. I This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified condi'tions in the l Applicability for which these systems and components ensure l l safe operation of the plant.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that i surveillances do not have to be performed on inoperable I

. equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the i SR(s) to be performed is removed. Therefore, failing to perform the Surveillance (s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.

The provisions of SR 3.0.4 shall not prevent changes in l MODES or other specified conditions in the Acolicability that are required to comply with ACTIONS. I In add ion, tne provisio of LC0 3.0 shall not p vent ' chang in MODES _ r_

or ot specifie onditions in e Applicab ity that resf from a s down perfo in respon to the ex cted faflure to co ly with ACTI R$6 3.0 S l

i;- Palisades Nuclear Plant B 3.0-15 01/20/98 a

. 6-d

- - = . _ - - _ - . - . - . _ _ - . . . - ._ _ . . _ - _ - - -

SR Applicability B 3.0 BASES SR 3.0.4 The precise requirements for performance of SRs are (continued) specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of.Surveillances when the pre. requisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LC0 prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO Applicability, I would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

SR 3.0.4 is a licable when ntering all DES or other specified co itions, whet er increasin in MODES (e.g. '

MODE 5 to DE 4) or de asing in MOD (e.g., MODE to ,,J2 -

MODE 5). he require t precluding ntry into ano er MODE

. or oth specified c ndition when sociated ACTI S do not prov e for contin d operation f an unlimite period of ti ensures that he plant mai ains suffici t equipment 0 RABILITY and edundancy as sumed in th accident nalysis.

RAf 3.o 4 Palisades Nuclear Plant B 3.0-16 , 01/20/98 fL)~(l

LC0 Applicability 3.0 3.0 LCO APPLICA81LITY l l

LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in th.e Applicability that are required to mply with ACTION 5[pr Ithat/are sfrt of/a shutdo6n of the_l EM 3.6-2.

Exceptions to this Specification are stated in the individual _ Specifications._JThese cepuans TEto DES or ther spec led con itions in heiiowentry'\

\

App cabilit when the ssociat ACTIONS o be enter a ow unit peration the M0p or other specified \

O13 onditi_o in the Ao icabilit,f only for a limited riod olj I MODE or other gj LCO 3.0. is only applica e for entry into

/ specift d condition in t Applicability in ODES 1, 2, 3, y and 4. J

~ ~

Reviewers's Note: 0 3.0.4 has been rev sed so that changes in MODES or other specified cond tions in the Applicability tha are part of a shutdo n of the unit shall not be prevented. In addition, LCO 3. 4 has been revised so that it is o y applicable for ent y into a MODE or other spccified condi ion in the Applicabi ity in MODES 1, 2, 3,.

and 4. The M E change restriction in LCO 3.0.4 were previously a licable in all MODES Before this version /of LCO 3.0.4 e be implemented on a plant-specific basis, the licensee m t review the existin technical specifica ns to determi e where specific res rictions on MODE cha es or Required ctions should be inc uded in individual LC to justify his change; such an valuation should be s mmarized

~

in a ma rix of all existing C0s to facilitate NRC staff review of a conversion to t e STS.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under adelnistrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABillTY.

(continued)

CEOG STS 3.0 2 Rev 1, 04/07/95 l

f_[

)

SR Applicability I 3.0 '

3.0 SR APPLICA8ILITY SR 3.0.3 declared not met, and the applicable Condition (s) must be (continued) entered, l

I SR 3.0.4 Entry into a MODE or other specified condition in the l Applicability of an LCO shall not be made unless the LCO's I Surveillances have been met within their specified Frequency. This provision shall not prevent entry into MODES or other specified conditions in the licability _

that are required to comply with ACTIONS t are fart or r )

p 3,g ga sputdowy of the unit./ /

/l140 3.0.4 is nly applicable f entry into a h00E r other 5j Ol, specified c dition in the Ap cability in MODES , 2, 3,

- and 4. ,

J Reviewer's N e: SR 3.0.4 has bee revised so that chan s in MODES or ther specified condit ons in the Applicabi ty .

that are p t of a shutdown of th unit shall not be prevented. In addition, SR 3.0. has been revised so hat it is on1 applicable for entry into a MODE or other le specifie condition in the App icability in MODES 1, 2, 3, and 4. he MODE change restr tions in SR 3.0.4 we e '

previo sly applicable in all DES. Before this v rsion of SR 3. 4 can be implemented n a plant specific b sis, the lice se must review the e sting technical spec fications to termine where specifi restrictions on MOD changes or l Re tred Actions should b included in individ al LCOs to

. ju tify this change; suc an evaluation shoul be sumarized i a matrix of all exis ing LCOs to facilita NRC staff eview of a conversion o the STS. _

\

h CEOG STS 3.0 5 Rev 1, 04/07/95 i

=

g_O -

s k

LCO Applicability l B 3.0 i

'~ l BASES l

I .

LCO 3.0.4 The provisions of LC0 3.0.4 shall not prevent changes in (continued) MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. IIn a 1 tion, ne y provi ons oVLCO 3. .4 shall ot pr nt cha es in ES /

l O6 or her spd'cified onditio in t pplic i ity at l r uit frA any it shut wn. l l

l Exceptions to LCO 3.0.4 are stated in the individual l l Specification J. s Exceptions may apply to all the ACTIONS or l to a specificJRequired Action of a Specification.

13 /7#5E P 7 J /\

- LCO3.0.4isonfapplicablewho entering MODE 4 fr l

\ MODE 5, MODE 3 rom MODE 4, M00 2 from MODE 3, or DE 1  :

from MODE 2. urthemore, LC0 .0.4 is applicabi when

entering an other specified ondition in the Ap icability V l

h only while perating in M00 requireme ts of LCO 3.0.4 1, 2, 3, or 4.

not apply in MOD in other specified condit/ons of the Applica lity (unless T

5 and 6, or l

I l

in M00 1,2,3,or4))ecausetheACTIONS f individual l Speci ications sufficiphtly define the re dial measures to l be ken. l[ h some cases (e.g., ..) these ACTION 5 pr ide a i Note that s tes "While this LC is not met, entry i oa

- MODE or ot r specified condit n in the Applicabil y is not permi ed, unless require to comply with ACTI S.'

[

This Note is a requirement e licitly precluding try into a M00E o other specified c dition of the Appli bility.)

Surveillances do not have to be perfomed on the associated

@l3.04 inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing

~

MODES or other specified conditions while in an ACTIONS Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be perfomed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

t LCO 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with (continued)

CE06 STS B 3.0-6 Rev 1, 04/07/95 l

(3- h _

- CHAPTER 3.0 INRRRT 1 -

l The exceptions allow entry into MODES or other specified conditions in the Applicability when l the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time.

Dr 3.0 - L l INSEIT 2 g LCO 3.0 is applic e when en ring all MO ES or other pecified co itions, whethe l increa g in MO S(e.g.,M E 5 or M E 4) or dec asing in M ES (e.g., MO E 4 to l MO 5). The quirement ecluding en into anoth MODE or o er specified c dition I w n the asso 'ated ACTIO. S do not pr ide for cont' ued operatio for an unlimit period of t' e ensures at the plant laintains su icient equip ent OPERAB ITY and redu9 dancy as assumed i e accident ysis.

/ i 1

l

-e e

B 3.0-6 6 -i

.s

- . _ . - - - -- . . _ - -.--_- - - __-=_.-_.- -

SR Applicability B 3.0 l

BASES SR 3.0.4 that are required to comply with ACTIONS. 1In ddition he l (continued) provi ons of 0 3.0.4 all not revent c nges in DES r or o er sp ied c tions i the Appl ability at Olo re it fro any unit hutdown i The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions l

necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO Applicability, would have its Frequency specified such that it is not "due"

until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in

,Section 1.4, Frequency.

j SR 3.0.4 is only applicable when en ring MODE 4 from - - -/

) MODE 5, M00E 3 rom MODE 4, Mode 2 rom MODE 3, or M00Fy 1 from MODE 2. urthermore, SR 3.0 is applicable when enterin any other specified con ition in the Applic 'ility only wh le perating in MODES 1 2, 3, or 4. The Ob requireme s of SR 3.0.4 do n apply in MODES 5 a 6, or

/c in other specified condition of the Applicabilit (unless in MODE 1, 2, 3, or 4) bec se the ACTIONS of i dividual j Specif cations sufficienti define the remedial measures to {

be t en. _ _ ,

% I 3. 0 - 1.

CE06 STS 8 3.0-16 Rev 1, 04/07/95 l

l

. *J -

l t

A l

SERT

. A--

l S .0.4 is appli e when ent ng all MODE or other specif' d conditions, wh er l ' creasingin ES (e.g., 'DE 5 to MO 4) or decreas' g in MODES (e. , MODE 4 to l MODE 5). e requireme precluding en - into another DE or other spe fled condition l when the sociated AC ONS do not pr ide for contin d operation for;u nlimited period of l

time e es that the ant maintains s ficient equipm t OPERABILITY d redundancy as ass ed in the ace' ent analysis.

l UI 3 0 -Z.

1 l

l B 3.0-16 D'$ '

S

ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY Change ,

niscussion

6. (continued)

When . I to NURE 432 was writte , the wording i CO and SR 3.0 was wri to reflect the v. I allowances tscussed above ithout using bra ets to i ate acceptable ternative wordi if the Rev. I roach to imple nting LCO SR 3.0.4 w not taken. To l

ect this, the i ustry owner's g ps wrote TSTF-103 to d brackets to areas which dis ssed LCO and 3.0.4 being only applicable MODES 1,2, nd 4, and prov' ed alternative w ding where the Rev.

I versio discussed "any t shutdown." e proposed wor ng in the Palisades ITS for L and SR 3.0.4 ' modeled after changes made TSTF-103 along with so e proposed chang s to TSTF-103 t correct some c istency errors.

7. The Bases of SR 3.0.1 has been modified to add the following statement: N M'

" Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case. the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition."

. This change was made as a result of a proposed change by the industry owner's groups ,

as discussed in Technical Specification Task Force (TSTF) 8. Rev. 2. The proposed change acknowledges that credit may be taken for an unphnned event which demonstrates the operability of the system equivalent to the performance of the associated surveillance requirement.

8. The proposed Bases for 3.0.2 is modified to reflect a change by the industry owner's groups as discussed in Technical Specification Task Force (TSTF) change number 52 which was generated to implement Option B of Appendix J. The existing NUREG-1431, Rev.1 Bases for SR 3.0.2 stated at the end of the paragraph "The 25% extension...":

1

, "An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by

. approved exemptions." The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations. Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."

i Palisades Nuclear Plant Page 2 of 4 01/20/98

&-/

t l

l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS ,

RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION  !

SECTION 3.0, APPLICABILITY  ;

l NRC REQt/EST:

-3.0-3 Markup of STS SR 3.0.2 JFD 8 )

TSTF-52 (Not Approved as of 08/21/98)

JFD 8 states, in part, "The proposed Bases for 3.0.2 is modified to reflect a change by the industry's owner's groups as discussed in [TSTF-52], which was generated to' implement Option B of Appendix J. The current NUREG-1431, Rev. 1 Bases for SR 3.0.2 stated at the end of the paragraph "The 25% extension...":

'"An example of where SR13.0.2 does not apply in a Surveillance with a ,

Frequency'of "in accordance with 10 CFR 50,' Appendix J, as modified by l approved txemptions." The requirements of regulation take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations. Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."~

TSTF-52 (which has not been approved as of August 21,1998) changes this sentence to state:

"An example of where SR 3.0.2 does not apply in the Containment Leak Rate Testing Program."

The proposed ITS will no longer reference 10 CFR 50, Appendix J, but will instead reference the Containment Leak Rate Testing Program in the Administrative Controls Section.

Causent: TSTF-52 has not been approved (as of August 21,1998). If this TSTF n

remains unapproved at the time the Palisades review is complete and the Palisades Safety Evaluation (SE) is being prepared, the ITS will have to return to the CTS, or new justification for this change will need to be prepared.

4 e

7

-,L

i l

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION l SECTION 3.0, APPLICABILITY Consumers Enerav Resnonse:

JFD 8 for ITS Section 3.0 has been revised to reference changes made to the CTS by Amendments 174 and 182, rather than TSTF 52, as justification for the proposed differences between ITS and STS.

Affected Paaes:

. Attachment 6 JFD page 2 of 4 )

Attachment 6 JFD page 3 of 4 (All of Attachment 6 replaced due to repagination) l O

d o

8 i

ATTACIBIENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY Chanoe Discussion

6. (continued)

When Rev. I to NUREG-1432 was written, the wording in LCO and SR 3.0.4 was written to reflect the Rev.1 allowances discussed above without using brackets to indicate acceptable alternative wording if the Rev.1 approach to implementing LCO and SR 3.0.4 was not taken. To correct this, the industry owner's groups wrote TSTF-103 to add brackets to the areas which discussed LCO and SR 3.0.4 being only applicable in MODES 1,2,3 and 4, and provided alternative wording where the Rev.

1 version discussed "any unit shutdown." The proposed wording in the Palisades ITS for LCO and SR 3.0.4 is modeled after the changes made in TSTF-103 along with some proposed changes to TSTF-103 to correct some consistency errors.

7. The Bases of SR 3.0.1 has been modified to add the following statement:

" Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition."

This chang: was made as a result of a proposed change by the industry owner's groups as discussed in Technical Specification Task Force (TSTF) 8. Rev. 2. The proposed change acknowledges that credit may be taken for an unplanned event which demonstrates the operability of the system equivalent to the performance of the associated surveillance requirement.

8. The proposed Bases for .0.2 is m 'fied to ref1 t a change by the in ustry ow er's R6 groups as d' cussed

• echnical ification ask Forc (TSTF) e nge num er p.3 1d58"# 52 which as gener ed to impt ent Optio B of Ap di> .T existing ggg 7FO 8. NURE -1431, R .1 Bases r SR 3.0.2 tated at end of the aragraph The 25% xtension .":

example o here SR 3 .2 does n apply is a urveillanc with a F equency o in accorda e with 10 FR 50, Ap ndix J, as odified by approved emptions." e requir ents of re ations take recedence over the TS. e TS cann in and of emselves e end a test i erval specified in the re lations, efore, the is a Note ' e Freque stating, "SR 3.0.2 is applicable."

Palisades Nuclear Plant Page 2 of 4 01/20/98 i hO

ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY l 1

Chance Dienulon

8. (continued) gr

TSTF-52 c es this sente to state

example of re SR 3.0.2 oes not apply s the Conta' nt Leak Rate esting Prog .

p l S eillance Re irement Freq ies in the pr posed ITS will o longer refe nce O CFR 50, pendix J but i tead will refe nce the Contai ent Leak Rat Testin ;

Program w ch is containe n the Admini ative Controls ction. There re, the Bases for R 3.0.2 is revi ed to reflect change,

9. In the proposed ITS LCO 3.0.3b, NUREG 1432 lists a time to reach MODE 4 of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This item is " bracketed" in NUREG-1432 since some plant designs require a different number. In the proposed Palisades ITS, this time is increased to 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />.

Increasing the time allowed to reach MODE 4 allows for more complete degassing of the Primary Coolant System (PCS). The PCS is degassed by venting the pressurizer gas space to the Vacuum Degasifier. The efficiency of this method is maximized by maintaining PCS temperature as high as practical, the subcooling as low as practical, and operating all pressurizer heaters. This results in a net increase in the rate of

, hydrogen removal from the PCS since increased spray flow and lower PCS pressure offse' the lower degas flow rate through the vent path. While the total time to reach MGf. E 4 is increased, the time to reach MODE 5 is the same in the proposed ITS as j specified in NUREG-1432.

10. TSTF-165 is incorporated into the Bases for LCO 3.0.5. The Bases for LCO 3.0.5 is changed to use the word " testing" instead of the acronym "SR." LCO 3.0.5 states,

" Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.

This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY."

While LCO 3.0.5 refers to " testing", the Bases for LCO 3.0.5 inconsistently use the term "SRs" instead of " testing." This change corrects this inconsistency. This caange addresses testing that is required to demonstrate operability that is not a surveillance.  !

For example, post maintenance testing required to demonstrate operability may not be a I Surveillance. This change does not change the intent of the LCO and makes the Bases consistent with the LCO. l Palisades Nuclear Plant Page 3 of 4 01/20/98 8-b '

1 I

New JFD 8:

(RAI3.0-3)

8. The proposed Bases for 3.0.2 is modified to reflect the inclusion of a Containment Leak Rate Testing Program in CTS by Amendment 174. That change was retained in the ITS and is reflected in the wording of the Frequency of the SRs in ITS Section 3.6. The affected ITS SRs do not refer to the requirements of 10 CFR 50 Appendix J, but to the Containment Leak Rate Testing Program instead.

In addition to the reason in the STS bases for SR 3.0 2 not being applicable, CTS Amendment 182 placed the restriction that SR 4.0.2 (ITS SR 3.0.2) could not be applied to the Primary Coolant Pump Flywheel Testing Surveillance Program. Therefore, the Bases for ITS SR 3.0.2 is revised from that of STS to reflect these plant specific differences.

The existing NUREG-1431, Rev.1 Bases for SR 3.0.2 stated at the end of the paragraph "The 25% extension...":

"An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations. Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable." ,

The ITS bases change this sentence to state:

An example of where SR 3.0.2 does not apply is the Containment Leak Rate Testing Program.

8 -c 3

. . - . - - . _ . . .-- - ..- - ,_ - - - - - - _ _ = . . - . _ - . . - -

i CONVERSION TO IMPROVED TECNNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.0, APPLICABILITY NRC REQUEST:

3.0-4 STS LC0 3.0.2 Bases ITS LCO 3.0.2 Bases JFD 12 TSTF-122 TSTF-122 is incorporated to revise the LCO 3.0.2 Bases to remove possible

. confusion. The NUREG-1432 changes states in part, " Additionally, if intentional entry into ACTIONS Alterr,etiv; that would not result in redundant equipment being inoperable, alternatives should be used instead. Doing so limitsthetimebothsubsystems/trainsofasafetyfunctionareinoperableand limits the time other conditions exist which may result in LCO 3.0.3 being entered."

causent: To comply with TSTF-122, the ITS Bases for LC0 3.0.2, 7th paragraph, the word "other" should be deleted. The STS markup should also be revised to reflect this change.

Consumers Enerav Resnonse:

The suggested correction has been made.

Affected acces:

ITS Bases page B 3.0-2 Marked up STS Bases page B 3.0-2 m

9

LCO Applicability B 3.0 BASES LCO 3.0.2 The second type of Required Actic; specifies the remedial (continued) measures that permit continued operation of the plant that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case.

An example of this is in LC0 3.4.3, "PCS Pressure and Temperature (P/T) Limits."

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational

, problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into gg 3,o-H ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing s limits the time both subsystems / trains of a safety functinn -O are inoperable and limits the time which may result in LCO 3.0.3 being conditions

@ entered. exist Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Requirec Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is required to comply with Required Actions, the plant may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are enterec.

Palisades Nuclear Plant B 3.0-2 01/20/98

1.G0 Applicability

. 8 3.0 BASES .

, LCO 3.0.2 ACTIONS.) The second type of Required Action specifies the l

i (continued) tal measures that permit continued operation of the g* j that is not further restricted by the Completion Time.

1 Q(>l t his case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case.

Oq { An example of this is in LCO 3.4.3, " @ Pressure and Temperature (P/T) Limits."

Oc.5 The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance.

Qg;g3' if corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done M.. N in a manner that does not compromise safety. Intentional (ia% 4WonJ5 2 entry into ACTIONS,should not be made for operational

~ convenience. AMwnmvesmir would 1gD result in r & % .4,ses p rodundant equipment beino inocerableashould be used instead.

Dt M 'l O l Doing so limits the the both subsystems / trains of a safeif g conditiens exist whicharesult in LCO 3.0.3 being entered. function are inop N Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassij for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is "

h required to comply with Required Actions, the a MODE or other specified condition in which anot er may enter Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.

(continued)

CE0S ST3 5 3.0-2 Rev 1, 04/07/95 94

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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 4.0, DESIGN FEATURES l NRC RE00ESI:

4.0-1 ITS 4.3.2 Drainage j DOC A.5 ,

I The proposed ITS 4.3.2, Drainage specifies, "The spent fuel storage pool cooling system suction and discharge piping is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 644 ft 5 inches."

The NUREG-1432, 4.3.2 Drainage states, "The spent fuel storage pool is designed and shall be maintained to prevent inadvertent drainage of the pool elevation [23ft].

DOC A.5, states in part, "The discharge piping is at 647' and contains a siphon breaker. The bottom of the suction piping is at elevation 644' 5".

l Since these piping arrangements are pemanent plant features, and no additional operational requirements have been imposed the inclusion of this

~

infomation into the proposed ITS is considered to be an administrative change."

Consent: Revise the DOC or provide an appropriate JFD to address the plant specific spent fuel storage pool elevation.

Consumers Enerav ResD0nst:

Doc A.5 has been revised to better explain the plant specific reason for the proposed ITS fuel pool minimum level due to inadvertent drainage.

Affected naae:

Attachment 3, DOC page 2 of 6 10 h

i ii

l ATTACIBfENT 3 l DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN FEATURES i A.4 CTS 5.1 states "The Palisades reactor shall be located on...." The proposed ITS states l l "The Palisades Nuclear Plant is located in...." This change more appropriately reflects  ;

l the plant name. In addition "shall be" is changed to "is" to improve the sentence l wording and reflect the first that the Palisades Nuclear Plant has already been built, i These changes are considered to be administrative changes since no requirements have l changed.

A.5 The Palisades CTS does not address inadvertent draining of the storage pool. The l suction and discharge piping of the cooling system for the storage pool was designed to j prevent inadvertent draining. The discharge piping is at 647' and contains a tinhon l

^

breaked The bottom of the suction piping is at elevation 644'5."FSince these piping l arrangements are permanent plant features, and no additional operational requirements l have been imposed the inclusion of this information into the proposed ITS is considered l to be an administrative change. This change is consistent with the intent of NUREG-1432.

l A6 CTS 5.4.la contains certain design aspects of the new fuel storage racks and includes a j reference to Siemens Nuclear Power Corporation Report EMF-91-1421 (NP) for the appropriate conservatism used in the calculation of Ke rr. In proposed ITS 4.3.1.3, reference to the Siemens Nuclear Power Corporation Report has been replaced by a reference to FSAR Section 9.11. Section 9.11 of the FSAR documents the design and analysis for the Fuel Handling and Storage Systems. This change is considered -

administrative in nature since it does not alter the design or analysis assumptions of the new fuel storage racks, but merely revises the reference of the document which contains the uncertainties used in the determination of K.tr. This change is consistent with NUREG-1432. g f OT 4.0 -1 TECHNICAL CHANGES - MORE RESTRICTIVE (M)

There were no "More Restrictive" changes added to this chapter.

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l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION I SECTION 4.0, DESIGN FEATURES NRC REQUEST:

4.0-2 ITS 4.3.3 Capacity JFD 4 Proposed ITS 4.3.3, Capacity, is a new TS for Palisades Section 4.0.

ITS 4.3.3 states, "The spent fuel storage pool and north tilt pit are designed and shall be maintained with a storage capacity limited to no more than 892 fuel assemblies."

JFD 4 relates to changes that reflect the facility specific nomenclature, number, reference, system description, or analysis description.

Comment: Provide a DOC for the addition of TS 4.3.3.

Consumers Entrav Resconse:

CTS page 5-4a markup has been revised to show the addition of ITS Section 4.3.3, and DOC A.7 has been provided.

'Affected acces:

CTS markup page 5 of 8 Attachment 3 DOC page 2 of 6

~

i 11 0

Chy.<%.o S

Q Q Fuel storane Irrad ted fuel bundles I be stored, prio a.

in t stainless steel 1 ned spent fuel poo . o off. site snisment (j A

lb. 40eleted)/ l h [6 3Alf,-ll h h@ The spent fuel storage pool and spare (north) tilt pit are divided into two regions identifie as Region ! and Region !! as illustrated in Figure . Region I racks are designed and shall O be maintained with a nominal 10.25' center to center cistance between fuel assemblies with the exception of the single Type E rack which has a nostnal 11.25' center to center distance eetween fuel assemblies. The Region I spent fuel storage racks are designed such that fuel having a maximum assembly planar average M a th, Van enrichment of 4.40 w/o placed in the racks would result in a qq K,,, equivalent to s 0.95 when flooded with unborated water. The K, of $ 0.95 includes a conservative allowance for uncertainties, y 31, Q For enrichments above 3.27 w/o Van, the fuel asseeclies must contain 216 rods which are either 00,, G4,0, UO or solid metal. .

y t 1.1.g. $ Region!!rackshavea9.17inchcentertocenterspacing.!Becausel of this s burnupo heIer spacing',

fuel ass 4mely strict contt4/ols orior itsare emplofed placemenI in to evaluate Reeten If l M -s Eall loc tons.1 Upon detensination that the fusf asseecly Mets .

/f,*j\ the burnup requirements of rTaba 5.e' l'._ placement in a__ReSion il #

p ) q'd' cell is authorized. IIyese positive /controis assure /Ine r oei

4. ! - .,

enrichiptnt Itstts assWined in the sJfety analyses w/11 not to o exceeoed.l N

ie. nei et. i f.

e The inimum spent fuel pool w er boron concentration hall ce '

i (ceIA:/' 17 ppe. Boron concentrati shall be verified at est once nthly.

Ig. The ufent fuel racky'are designed as4 Class I str/cture.l I h. / (Deletid)I h U Uj g Storage in Region !! of the spent fuel pool and spare (north) ttit pit shall be restricted bf burnup and enrichment limits specified

( Sa.a W 3.7 ? in ITahM 1.Mll 7, g , 3 M, , g , Q MQTL; ntil needed for f ~e l storage, one Region !! rack in tn I northeast corner f the spent fuel ol may be removec and I reciated with t cast anti tippin cevice.

Rafaren'cas

.\ SAR Update Cha er 5 FSAR Update Ch ter 9

( @D 9.1,1,b.:m, 4*c pe4e.:eJ'ITT) Ml N'0 M Am No M4, i.H. 140

@ D i'l.3 CMPA(' T"} ) 6 tom January 23, 1992

, h.w 54f

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i ATTACHMENT 3 DISCUSSION OF CHANGES  !

CHAPTER 4.0, DESIGN FEATURES l

A.4 CTS 5.1 states "The Palisades reactor shall be located on...." The proposed ITS  !

states "The Palisades Nuclear Plant is located in...." This change more appropriately l reflects the plant name. In addition "shall be" is changed to "is" to improve the l sentence wording and reflect the first that se Palisades Nuclear Plant has already I been built. These changes are considered to be administrative changes since no requirements have changed.

A.5 The Palisades CTS does not address inadvertent draining of the storage pool. The suction and discharge piping of the cooling system for the storage pool was designed to prevent inadvertent draining. The discharge piping is at 647' and contains a siphon l

breaker. The bottom of the suction piping is at elevation 644'5." Since these piping ,

arrangements are permanent plant features, and no additional operational requirements I have been imposed the inclusion of this information into the proposed ITS is l considered to be an adininistrative change. This change is consistent with the intent I of NUREG-1432.

A.6 CTS 5.4.la contains certain design aspects of the new fuel storage racks and includes a reference to Siemens Nuclear Power Corporation Report EMF-91-1421 (NP) for the appropriate conservatism used in the calculation of K e n. In proposed ITS 4.3.1.3, reference to the Siemens Nuclear Power Corporation Report has been replaced by a reference to FSAR Section 9.11. Section 9.11 of the FSAR documents the design and analysis for the Fuel Handling and Storage Systems. This change is considered administrative in nature since it does not alter the design or analysis assumptions of the new fuel storage racks, but merely revises the reference of the document which contains the uncertainties used in the determination of K,y. This change is consistent with NUREG-1432.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

There were no "More Restrictive" changes added to this chapter.

lNMW AS PMl kD~A Palisades Nuclear Plant Page 2 of 6 01/20/98

//-6

i New DOC A.7:

(RAI4.0-2)

, A new section, 4.3.3, has been added to the Design Features Section of ITS to emulate the contents of STS. This section is descriptive, and contains no ,

requirements. Similar material is contained in the FSAR and other plant '

documents. Since .this change contains no requirements, .the change is classified as Administrative.

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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE OCTOBER 1, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 5.0, ADMINISTRATIVE CONTROLS l

l NRC REQUEST (Telenhoned).: 1 5.5-03 ITS 5.5.14 Containment Leak Rate Testing Program l Omitted DOC 1 Proposed ITS 5.5.14, Containment Leak Rate Testing Program, contains a short paragraph equating " Containment Operability" with " Containment Integrity" which does not appear in CTS. That addition was not identified in the markup of CTS and no Discussion of Change (DOC) was provided.

Consent: Correct the CTS 6.5.14 markup and provide a DOC for th's change.

Consumers Enerav Response:

CTS 6.5.14 markup, on page 6-17, has been revised to show the addition of the subject paragraph and DOC A.14 has been provided.

Affected acces: '

~

CTS markup page 6-17 -

' Attachment 3, DOC page 4 of 6 (All Attachment 3 000 pages replaced due to repagination) l l 12

5.0 RM04a i G 8.0 ADMINISTRATIVf CONTRolf 5.7. G 6.5.13 M ,

. Abb SEr.h Mbs DdkdMen hay.m(.5FD4) u pueM inITS i

i 1

j 5,5,N 4.5.14 Cantainment taak Rate Tattina proaram Progress shall be established to isolement the leak rate testing of the i containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option 8, as modified by approved exemptions. The Type A test program s

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 Lenkage. Test Program, dated September 1995.* The Tyne B and Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A. as modified by the exemption froe certain requirements of 10 CFR 50 Appendix J which was granted in an NRC letter to Consumers Power Company dated December 6, 1989.

d 4 The peak calculated containment internal pressure for the desian basis

' loss of coolant accident, P., is 52.64 psig ((FSAR/rable 14/18.1 a]). A>l l The maximum allowable containment leak rate. L., at P., shall be 0.1% of

- containment air weight per day.

Leak rate acceptance criteria are:

a, Containment leak rate acceptance criteria is s 1.0 L,. During the first plant startup following testing in accordance with this 4

program, the leak rate acceptance criteria are 5 0.60 L, for the

' Type B and Type C tests and s 0.75 L, for Type A tests; ,

I b. Air lock leak rate acceptance criteria is s 0.023 L, for each door, when pressurised to 3 10 psig.

g) ,

The Surveillance interval extensions of .2 Ar.t_ngt applicable to QI IbO3 the Containment Leak Rate Testing Progr w requirements.

The provisions of .3 Ata applicable to the Containment Leak Rate 2 l Testing Program requirements.

j l

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,- l 6 17

^

Amendment No. 174

. October 31, 1996 1

2.7 Ok M I

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A mL m--a, - 5 d ---&---S & A 4 4 A a+A+a 4 - - +e. a .J 4 ,aaAL.J--- L- O Ma' 2 -#u '6 22,mKm.x,~mnu l RAI INSERTS FOR SECTION 5.0 l

1

New Insert for CTS page 6-17

. Containment OPERABILITY is equivalent to " Containment Integrity" for the l l_ purposes of the air lock testing requirements in 10 CFR 50, Appendix J. I 1

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ATTACHMENT 3 i DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS A.13 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, this 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.

ADD 30C. 4,f4 hAl f.f-03)

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 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 4 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 I.etter 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

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RAI INSERTS FOR SECTION 5.0 I

New DOC A.14:

I l'

An additional paragraph was added to ITS 5.5.14 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 )

j Technical Specifications." This change is considered to be Administrative, l since it simply assures that the omission of the CTS tenn " Containment  !

Integrity" does not affect the interpretation of Appendix J. testing j requirements.

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l l l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE OCTOBER 1, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 5.0, ADMINISTRATIVE CONTROLS l

I NRC REQUEST (Telephoned):

5.5-04 ITS 5.5.14 Containment Leak Rate Testing Program Omitted justification l

Proposed ITS 5.5.14, Containment Leak Rate Testing Program, states that

" Containment leak rate acceptance criteria is s 1.0 L,", as does CTS 6.5.14.

However CTS 4.5.2b.(1), which is marked up as being moved to ITS 5.5.14, states "The total leakage from all penetrations and isolation valves shall not exceed 0.60 L,."

causent: Provide justification for this change.

Consumers Enerav Resoonse:

The subject change was addressed in Section 3.6 of our ITS submittal, but not in Section 5.0. The change has been justified by additior of a DOC and a No Significant Hazards Consideration to Section 5.0, Attar' : ants 3 and 4, respectively.

Affected naaes:

CTS markup page 4 20 Attachment 3, DOC page 6 revised, page 7 added (All Attachment 3 000 pages replaced due to change in page numbering)

Attachment 4, page 5 revised, pages 6 and 7 added (All Attachment 4 pages replaced due to change in page numbering)

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57476gW W CNf!AS 4.5 CONTAllMENT TESTS

.4.5.2 Local Leak Detection Tests (continued) l

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

[*{ (2) exceed 0.h23L..The leaka e for a Personnel airlock d QI I- -

An acceptable Emer ency Escape Airlock door seal contact check

((3) consists seals andofthe a veri 1 cation sealing of continuous contact between the surfaces.

l l

c. Corrective Action i '

i (1) If at any time it is etermined that 0.60 L, is exceeded, l repairs shall be in lated immediately. If repairs are not completed and conf rmance to the acceptance criterion of l

f is not emonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the plant fM 4.5.2.b(1) shall be p lacedat least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in LD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />'.

3.lo j (2) If at any ti 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> d in COLD SH DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(3) If the P sonnel airlock door seal leakage is greater th l l

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 imediately to restore the door seal to the acce ance criteria of specification 4.5.2.b(2) or 4 S.2.b(3).

In e event repairs cannot be completed within 7 ys, the p1 t shall be placed in at least HOT SHUTDOWN wi 'in the next

. 6 ours and in COLD SHUTDOWN within the followin 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l (4) f air lock door seal leakage results in one d r causing total containment leakage to exceed 0.60 L., e door shall be declared inoperable and the remaining OPERAB door shall be immediately locked closed

• and tested withi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As long as the remaining door is found to be OPE LE, the provisions of 4.5.2.c(2) do not apply. Repairs shal be initiated immediately to establish conformance wi specification 4.5.2.b(1). In the event conformance this specification cannot be established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> he plant shall be placed in at least HOT SHUTDOWN with the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following - hours.

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

4-20 CONTAINNENT TSCR REV 2 Chg 7, Amendment No. 46, U4, W4, he.2ef29 l

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• l ATTACIBIENT 3 DISCUSSION OF CHANGES l CHAMER 5.0, ADMINISTRATIVE CONTROLS LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) l LA.2 CTS 6.5.7, Inservice Inspection and Testing Program has been revised to delete the l phrase " including applicable supports." Requirements for inservice inspections of l ASME Code Class 1,2, and 3 components are specificed in 10 CFR 50.55a(g). As I used in CTS 6.5.7, " applicable supports" is intended to apply to the inspection of I snubbers. Adaptation of this phrase in the CTS was consistent with the NRC's I approach to address concerns related to the relocation of the Snubber LCO from the l l ISTS NUREGs. Subsequently, the ISTS NUREGs have been modified to delete this l l phrase in recognition that it duplicates requirements specifed in the CFRs (See Section l l 5.0, JFD 26 addressing TSTF-279). As such, the deletion of this phrase from the CTS 1 can be made without a significant impact on safety since the inspection of applicable l supports continues to be required by 10 CFR 50.55a(g). l l

LESS RESTRICTIVE CHANGES (L) 1

.n. _ __ . . . _ __ _ _ o r _ _ ._.-.....

n ._: _.: . . . _ t _ _ _ _ _ __ _ m . .u _t__.__ '

- .,.- .._- . - . . . , - . . 1 Id%Ea r t. . t RELOCATED (R) ' '

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

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. l Palisades Nuclear Plant Page 6 of 6 09/04/98 e 8

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RAI INSERTS FOR SECTION 5.0 New "L" LESS RESTRICTIVE CHANGES (L)

L.1 The CTS requirements for Type B and C leak rate testing are being revised such that the leakage limit for Type B and C testing is s .60 La only during the first plant startup following testing perfomed in l accordance with the Containment Leak Rate Test Program. After this, the J new limit will now become s 1.0 La. This means that if the testing is performed during a refueling outage, the total Type B and C leakage must be "as-left" at s .60 La prior resuming power operations. Following this, the leakage limit for the remainder of time until the test is perfomed again becomes 1.0 La for the total containment leakage.

Overall containment integrity is maintained because the results of Type B and C testing must be compared against the overall containment leakage limit to ensure that the leakage remains s 1.0 La. CTS 4.5.2.c(1) specifies actions to be taken if .60 La is exceeded for Local Leak Detection Tests. The actions in 4.5.2.c(1) to initiate repairs immediately and shut down if the acceptance criteria of 4.5.2.b(1) is not met (ensuring total leakage from all penetrations and isolation valves shall not exceed .60 La) will no longer apply. In the proposed ITS, the acceptance criteria and testing frequency will only exist in the Containment Leak Rate Testing Program which is found in TS Administrative Controls Section 5.5.14. This is similar to the CTS Containment Leak Rate Testing Program which is found in CTS 6.5.14. The acceptance criteria in the Containment Leak Rate Testing Program will be that the overall containment leakage will not exceed 1.0 La. These changes are acceptable since the overall containment leakage requirements of ITS 3.6.1, which reference the Centainment Leak Rate Testing Program for the acceptance criteria, remain valid at all times.

Any increase in Type B and C leakage would have to be evaluated against this limit as evidenced by NOTE 3 in proposed ITS 3.6.2 and 3.6.3. The 4

change is considered Less Restrictive since the acceptance criteria of

.60 La for Type B and C tests will now become 1.0 La. This change is consistent with the infomation in the CTS 6.5.14, Containment Leak Rate Test Program but is less restrictive than the infomation contained in the CTS 4.5.2 section which addresses the local leak rate testing. This change maintains consistency with NUREG-1432 as modified by the intent of industry owner's group generic change TSTF-52.

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ATTACIBiENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 5.0, ADMINISTRATIVE CONTROLS

2. 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 i assumptions made in the safety analysis and licensing basis. Therefore, the changes l 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 l mitigative actions are initiated. There 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 TMic we uv L.. Remictivo C: .p. made m Obpr 57

%gg h t:.d k$ b Palisades Nuclear Plant Page 5 of 5 01/20/98

/3c)

1 RAI INSERTS FOR SECTION 5.0 l New NSHC for new "L.1" (2 pages)

LESS RESTRICTIVE CHANGES (L)

.The Palisades Nuclear Plant is converting to the Improved Technical Specifications '(ITS) as outlined in NUREG-1432, " Standard Technical l

Specifications,. Combustion Engineering Plants." Changes have been proposed j which involve making the requirements in the Current Technical Specifications (CTS) less restrictive. A description of the less restrictive change and corresponding No Significant Hazards Consideration are provided on the  ;

following pages for each Specification as applicable.

l LESS RESTRICTIVE CHANGE L.1 i The CTS requirements for Type B and C 1.eak rate testing are being revised such that the leakage limit for Type B and C testing is s .60 La only during the first plant startup following testing perfomed in accordance with the  ;

Containment Leak Rate Test Program. After this, the new limit will now become 1

- s 1.0 La. This means that if the testing is perfomed during a refueling i outage, the total Type B and C leakage must be "as-left" at s .60 La prior resuming power oprations. Following this, the leakage limit for the  !

remainder of time until the test is perfomed again becomes 1.0 La for the  !

total containment h akage. Overall containment integrity is maintained -

i

'because the results of Type B and C testing must be compared against the .

overall containment hakage limit to ensure that the leakage remains j

-s 1.0 La. '

i CTS 4.5.2b.(1) states "Ne total leakage from all penetrations and isolation ,

valves shall not exceed 0.60 L,." In the proposed ITS, the acceptance i criteria and testing frequency will only exist in the Containment Leak Rate Testing Program which is found in TS Administrative Controls Section 5.5.14. 1 This is similar to the CTS Containment Leak Rate Testing Program which is i found in CTS 6.5.14. The acceptance criteria in the Containment Leak Rate  !

Testing Program will be that the overall containment leakage will not exceed 1.0 La. These changes are acceptable since the overall containment leakage requirements of ITS 3.6.1, which reference the Containment Leak Rate Testing Program for the acceptance criteria, remain valid at all times. Any increase i in Type B and C leakage would have to be evaluated against this limit as l

' evidenced by NOTE 3 in proposed Ils 3.6.2 and 3.6.3. The change is considered Less Restrictive since the acceptance criteria of .60 La for Type B and C l tests will now become 1.0 La. This change is consistent with the information l

in the CTS 6.5.14, Containment Leak Rate Test Program but is less restrictive 1

/.3 e r

1 m

RAI INSERTS FOR SECTION 5.0 than the information contained in the CTS 4.5.2 section which addresses the local leak rate testing. This change mcintains consistency with NUREG-1432 as modified by the intent of industry owner's group generic change TSTF-52.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? I l

The proposed change increases the acceptance criteria for Type B and C tests from .60 La to 1.0 La for the total containment leakage.

Previously the .60 La for Type B and C tests acted as a " trigger point" to ensure ac ions were taken such that the overall acceptance criteria of 1.0 La were not violated. In addition the actions to initiate ,

I immediate repairs are not required unless total containment leakage l exceeds 1.0 La. The 1.0 La limit for total containment leakage remains in the proposed ITS. Any leakage from Type B and C tests which would ,

put total containment leakage over 1.0 La must still be evaluated. '

Therefore, there is no 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 change does not involve a physical alteration of the plant.

. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure total containment leakage is monitored to ensure that it stays within the bounds of the analysis. Thus, this change does 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?

The proposed change preserves the total containment leakage rate of 1.0 La but does not require actions to be taken once the Type B and C tests exceed .60 La as long as the contribution of the Type B and C tests do not make the total containment leakage to exceed 1.0 La. Therefore, this change does not involve a significant reduction in a margin of safety.

/3 - f

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l l

l l ATTACHMENT 2

)

l

CONSUMERS ENERGY COMPANY l l PALISADES PLANT .

DOCKET 50-255 l

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION l

l REVISED PAGES FOR SECTION 1.0 i

=

t i

(

l

l l

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 1.0 1

Pace Chanae Instructions Revise the Palisades submittal for conversion to Improved Technical l Specifications by removing the pages identified below and inserting the i attached pages. The revised pages are identified by date and contain vertical l

lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REY DATE NRC COMMENT #

ATTACHMENT 1 No page change ATTACHMENT 2 No page change ATTACHMENT 3' CTS 1-4 page 6 of 6 CTS 1-4 page 6 of 6 10/10/98 RAI 1.0-1 ~

DOC 1.0 page 6 of 12 000 1.0 page 6 of 12 10/10/98 RAI 1.0-1 ATTACHMENT 4 No page change ATTACHMENT 5 NUREG 1.1-4 NUREG 1.1-4 10/10/98 RAI 1.0-2 ATTACHMENT 6 JFD page 1 of 2 JFD page 1 of 3 10/10/98 RAI 1.0-2 JFD page 2 of 2 JFD page 2 of 3 10/10/98 RAI 1.0-2 JFD page 3 of 3 10/10/98 RAI 1.0-2 d

r

i CHAPTER 1.0 INSERT (SHUTDOWN MARGIN-CTS)

However, with all full length control rods verified fully inserted by two independent l means, it is not necessary to account for a stuck rod in the SDM calculation. With any full length control rods not capable of being fully inserted, the reactivity worth of these I rods must be accounted for in the determination of SDM; and b) There is no change in part length rod position.

e 4

CTS 1-4 Pacse G & &

mmsm 1W10/DS

i ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION l

l A.13 The CTS definition of SHUTDOWN MARGIN has the following wording added to it.

to from the proposed ITS definition of SHUTDOWN MARGIN (SDM):

a. "...However, with all full length control rods verified fully inserted by two l indept;ndent means, it is not necessary to account for a stuck rod in the SDM calculatien. With any full length control rods not capable of being fully l l inserted, the reactivity worth of these rods must be accounted for in the determination of SDM; and

b. There is no change in part length rod position."

The first part of the additional wording clarifies that if it can be verified by two independent means that all rods are inserted, no penalty for a stuck rod needs to be incurred. In addition, no credit for part length rods is given in the SDM calculation, which is reflected in the analysis assumption that there is no change in part length rod position. These changes are considered to be administrative changes as they are providing clarification on the calculation of SHUTDOWN MARGIN without changing -

the SHUTDOWN MARGIN requirements. CTS define the term CONTROL ROD as l "all full-length shutdown and regulating rods." That definition has not been carried l over to ITS (see DOC A.7). Therefore, changing usage from " CONTROL RODS" l to " full length control rods" constitutes an administrative change. This change is l consistent with NUREG-1432.

A.14 The CTS definition of REFUELING OPERATION forms the basis for the proposed ITS definition of CORE ALTERATION. In the CTS, the term " core components" is expanded in the ITS to define these components as "any fuel, sources, or control rods." In addition, the clarifying phrase " CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position." is included in the proposed definition to ensure that there is no confusion over being able to complete movement of a core component if directed to " Suspend CORE ALTERATIONS."

These changes are considered to be administrative changes since the term " CORE ALTERATIONS" is used to simply replace " REFUELING OPERATION" and provide additional clarification on its application. This change is consistent with NUREG-1432.

l i

Palisades Nuclear Plant Page 6 of 12 10/10/98

1 1

Definitions 1.1 11 1.1 Definitions.

ENGINEERED SAFETY function (i.e , the valves travel to their FEATURE (ESF) RESP 0 E required posi ions, pump discharge essures reach TIME their requir d values, etc.). Tim shall

@ (continued) include die 1 generator starting nd sequence loading del ys, where applicable. The response time may b measured by means of any series of sequentia , overlapping, or tot steps so that the entir response time is me ured.

L, The c4 mum allowable cont inment leakage ra,t'e, L , sh 1 be (0.25]% of co tainment air wei ht per day a the calculated pe containment pr sure

( P.) .

LEAXAGE LEAXAGE shall be: '

a. Identified LEAKAGE

1. LEAXAGE, such as that (from pum seals or valve 01antfamp-

@ packing (except seal water *ee (n.ie<tior Til eakoff),

that is captured and conducte'h -3 collection systems or a sump or collecting tank;

2. LEAXAGE into the containment atmosphere from sources that are both specifically located and known m +ter; not to interfere lf with the operation of leakage detection I

h systems (Er not o be pressure boundary LEAXAGE; @ em

^ " 1}-

3. M Coolant Y System ( S) LEAXAGE through a steam generator (SG) to the Secondary System.

b. Unidentified LEAKAGE eneri b:-*q C..d All LEAXAGE*that is not identified LEAKAGE:

1 {(p% 9sedledah [.. Pressure Boundary LEAKAGE LEAXAGE (except SG LEAKAGE) through a hl nonisolable fault in an 5 component body, pipe wall, or vessel wal g (continued)

CEOG STS 1.1 4 Rev 1, 04/07/95 REMSED 1ename e

ATTACHMENT 6

- JUSTIFICATIONS FOR DEVIATIONS i CHAlrfER 1.0, USE AND APPLICATION

{

TECHNICAL SPECIFICATIONS 1 1

l NOTE: The first five justifications for these changes from NUREG-1432 were '

generically used throughout the individual LCO section markups. Not all l generic justifications are used in each section.

1. The brackets have been removed and the proper plant specific information or value has been provided.

1

2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

3. The requirement / statement has been deleted since it is not applicable to this facility.

The foHowing 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 specifications.

6. Palisades does not use a methodology which utilizes an AZIMUTHAL POWER TILT Instead, the Palisades methodology utilizes an ASSEMBLY RADIAL PEAKING FACTOR, Fa^, and a TOTAL RADIAL PEAKING FACTOR, FaT.

Therefore, the NUREG-1432 AZIMUTHAL POWER TILT is not included in the proposed Palisades ITS and the ASSEMBLY RADIAL PEAKING FACTOR and the TOTAL RADIAL PEAKING FACTOR definitions have been included.

7. The Palisades CTS has a defined term for both an AXIAL OFFSET, (which is determined by using the incore monitoring system,) and an AXIAL SHAPE INDEX (ASI) (which is determined by using the excore monitoring system). The proposed Palisades ITS adds the defined term AXIAL OFFSET to those included in NUREG-1432 and also adds the clarification ht the ASI is detennined using the excore monitoring system. In additbu, the equation used to explain ASI is not used as it is not contained in the CTS and is omitted because it adds no real clarification to the definition itself.

Palisades Nuclear Plant Page 1 of 3 10/10/98

4 i ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAITER 1.0, USE AND APPLICATION

8. Performance of an ENGINEERED SAFETY FEATURE (ESP) RESPONSE TIME and REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME test is not part of the current Palisades Technical Specifications. A review during the NRC Systematic Evaluation Program, as stated in the resulting SER, concluded that the addition of response time testing requirements was not necessary. This will be further discussed in the sections which deal with response time testing

9. The Palisades CTS has a Containment I2ak Rate Test Program in Section 6.5.14.

L, is defined in this program and it will be retained there in the proposed ITS.

Therefore, L, does not need to be defined in the Definitions sections.

10. The proposed Palisades ITS does not include a definition for the PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) as Palisades does not propose to have this report. The current pressure and temperature limits for Palisades are valid until the end of reactor vessel life.

11. The proposed definition for " CORE ALTERATION" does not include the term -

" manipulation" as it is redundant to the discussion of "... movement of fuel or components." This change represents a generic change to NUREG-1432 proposed by the industry owners groups. This change was submitted and approved under TSTF-47.

12. The proposed definition for " Unidentified leakage," which is found under the defined term " LEAKAGE," includes the phrase "(except primary coolant pump seal leakoff)." This phrase was added to clarify that primary coolant pump seal leakoff should not be part of the amount included as " Unidentified Ixakage." This change was presented as a generic change to NUREG-1432 proposed by the industry owners groups. This change was submitted under TSTF-40. The proposed Palisades implementation differs from the proposed only by the fact that the primary coolant pump seal injection portion of the phrase has been deleted since the Palisades pumps do not use seal injection.

13. The NUREG-1432 definition of SHUTDOWN MARGIN contains part 'b' which states that the fuel and moderator temperatures are changed to be nominal zero power design level. This statement is not appropriate for the methodology used at Palisades to calculate SHUTDOWN MARGIN. Therefore, part 'b' from NUREG-1432 is not included in the Palisades definition for SHUTDOWN MARGIN.

d Palisades Nuclear Plant Page 2 of 3 10/10/98

N i

l ATTACIBIENT 6 l , JUSTIFICATIONS FOR DEVIATIONS I CHAPTER 1.0, USE AND APPLICATION l

14. The Palisades CTS contains the term " Quadrant Power Tilt (Tq)" and this term is

' l also included in the proposed ITS. The Quadrant Power Tilt is defined as "Tq shall l

, be the maximum positive ratio of the power generated in any quadra'It minus the ' l l average quadrant power, to the average quadrant power."

l

15. The wording of the Identified Leakage definition has been altered to clarify that l l leakage which might affect the operation of leakage detection systems must be [

classified as unidentified leakage. It is believed that this is the intent of the STS l definition. l l

l l

l l

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l Palisades Nuclear Plant Page 3 of 3 10/10/98 l

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t 5

ATTACHMENT 3 CONSUMERS ENERGY COMPANY -

PALISADES PLANT DOCKET 50-255 4

CONVERSION TO IMPROVED' TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.0 1

4 4

4

l 1 l

CONVERSION TO IMPROVED TECHNICAL SPECIFICATI0HS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION i

REVISED PAGES FOR SECTION 3.0

, Paae Chanae Instructions l

Revise the Palindes submittal for conversion to Improved Technical

, Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT #

ATTACHENT 1 ITS 3.0-2 ITS 3.0-2 10/10/98 RAI 3.0-2 ITS 3.0-5 ITS 3.0-5 10/10/98 RAI 3.0-2 ATTACHE NT 2 ITS B 3.0-1 ITS B 3.0-1 through through ITS B 3.0-16 ITS B 3.0-16 10/10/98 RAI 3.0-2 -

RAI 3.0-4 ATTACHENT 3 No page change ATTACHENT 4 No page change ATTACHENT 5 NUREG 3.0-2 NUREG 3.0-2 10/10/98 RAI 3.0-2 NUREG 3.0-5 NUREG 3.0-5 10/10/98 RAI 3.0-2 NUREG B 3.0-2 NUREG B 3.0-2 10/10/98 RAI 3.0-4 NUREG B 3.0-6 NUREG B 3.0-6 10/10/98 RAI 3.0-2 NUREG B 3.0-6 insert NUREG B 3.0-6 insert 10/10/98 RAI 3.0-2 NUREG B 3.0-16 .

NUREG B 3.0-16 10/10/98 RAI 3.0-2 NUREG B 3.0-16 insert ----------

10/10/98 RAI 3.0-2 ATTACHENT 6 JFD page 1 of 4 JFD page 1 of 4 through through JFD page 4 of 4 JFD page 4 of 4 10/10/98 RAI 3.0-2 RAI 3.0-3 l

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when  !

the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the ')

Applicability for an unlimited period of time.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are i required to comply with ACTIONS. l j Exceptions to this Specification are stated in the individual Specifications.

1 LC0' 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to l demonstrate its OPERABILITY or the OPERABILITY of other equipment. -This is an exception to LC0 3.0.2 for the system returned to service under administrative control to perfonn -

the testing required to demonstrate OPERABILITY.

~

LC0 3.0.6 When a supported system LC0 is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are i not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13. " Safety Function Determination Program (SFDP)." 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. When a support system's Required Action directs a supported system to be

. declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.

Palisades Nuclear Plant 3.0-2 Amendment No. 10/10/98

- . . . - . .. . - . - . - . - .- . -. ...- .= ._

I SR Applicability

, 3.0 3.0 SR APPLICABILITY SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LC0 must ininediately be declared not met, and the applicable Condition (s) must be entered.

SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be made unless the LC0's Surveillances have been met within their specified Frequency. This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. l l

Palisades Nuclear Plant 3.0-5 Amendment No. 10/10/98

. - _. - - =- -. - . _ . - - . . . . - .

I i LC0 Applicability l B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY I

BASES LC0 LC0 3.0.1 through LC0 3.0.7 establish the general requirements applicable to all Specifications and apply at l all times unless otherwise stated. '

LCC 3.0.1 LC0 3.0.1 establishes the Applicability statement within each individual Specification as the cequirement for when the LC0 is required to be met (i.e., when the plant is in the MODES or other specified conditions of the Applicability statement of each Specification).

LC0 3.0.2 LC0 3.0.2 establishes that upon discovery of a failure to

meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS ,

Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LC0 are not met. This Specification establishes that:

a. Completion of Required Actions within the specified Completion Tin,e .onstitutes compliance with a Specification; and

b. Completion of the Required Actions is not required when an LC0 is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LC0 must be met. This time limit is the Completion Time to

. restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits.

If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the plant in a MODE or condition in which the Specification is not applicable. (Whetherstatedasa Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.)

i l

l Palisades Nuclear Plant B 3.0-1 10/10/98

I i.

LCO Applicability B 3.0 BASES LC0 3.0.2 The second' type of Required Action specifies the remedial .

(continued) measures that permit continued operation of the plant that l 1s not' further restricted by the Completion Time. In this .

case, compliance with the Required Actions provides an

acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LC0 is met or is no longer applicable, unless otnerwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LC0's ACTIONS specify the Required Actions where this is the case.

An example of this is in LC0 3.4.3, "PCS Pressure and Temperature (P/T) Limits."

The Completion Times of the Required Actions are also applicable-when a system or component is removed from -

service intentionally. The reasons for intentionally relying on the. ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing so limitsthetimebothsubsystems/trainsofasafetyfunction are inoperable and limits the time conditions exist which may result in LC0 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required

, Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE.or other specified condition is required to comply with Required Actions, the plant may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.

Palisades Nuclear Plant B 3.0-2 10/10/98 L - - , -

LCO Applicability l B 3.0 l l

BASES LC0 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:

a. An associated Required Action and Completion Time is not met and no other Condition applies; or

b. The condition of the plant is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions' stated in the ACTIONS can be made that exactly corresponds to the actual condition of the plant. Sometimes, possible combinations of Conditions are such that entering l LC0 3.0.3 is warranted; in such cases, the ACTIONS l specifically state a Condition corresponding to such

! combinations and also that LC0 3.0.3 be entered

! immediately.

This Specification delineates the time limits for placing the plant in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe -

operation as defined by the LC0 and its ACTIONS. It~is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

f Upon entering LC0 3.0.3,1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach

! lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the

! specified maximum cooldown rate and within the capabilities l

,of the plant, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Primary Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3,

Completion Times.

Palisades Nuclear Plant B 3.0-3 10/10/98

LCO Applicability B 3.0 BASES

, LCO 3.0.3 A plant shutdown required in accordance with LCO 3.0.3 may (continued) be terminated and LC0 3.0.3 exited if any of the following occurs: .

a. The LCO is now met.

b. A Condition exists for which the Required Actions have now been performed.

c. ACTIONS exist that do not have expired Completion

~ Times. These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LC0 3.0.3 is exited.

4~

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for the plant to be in MODE 5 when a shutdown is required during MODE 1 operation. If the plant is in a lower MODE of- -

operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, however, the total -

allowable time to reach MODE 5, or other applicable MODE, is not reduced. For example, if MODE 3 is reached in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then the time allowed for reaching MODE 4 is the next 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />, because the total time for reaching MODE 4 is not reduced from the allowable limit of 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />. Therefore, if

remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4 LC0 3.0.3 provides actions for

, Conditions not covered in other Specifications. The t

requirements of LC0 3.0.3 do not apply in MODES 5 and 6 because the plant is already in the most restrictive

-Condition required by LCO 3.0.3.

. The requirements of LC0 3.0.3 do not apply in other

,specified conditions of the Applicability (unless in MODE 1, 2, 3 or 4) because the ACTIONS of individual Specifications

, sufficiently define the remedial measures to be taken.

Exceptions to LC0 3.0.3 are provided in instances where requiring a plant shutdown, in accordance with LC0 3.0.3, would not provide appropriate remedial measures for the associated condition of the plant. An example of this is in LC0 3.7.14. " Spent Fuel Pool Water Level."

1 a

Palisades Nuclear Plant B 3.0 4 10/10/98 L , , , . . _ . . , .. , . c. w -_

t 1 LCO Applicability l B 3.0 BASES LC0 3.0.3 LCO 3.7.14 has an Applicability of "During movement of (continued) irradiated fuel assemblies in the spent fuel pool."

Therefore, this LCO can be applicable in any or all MODES. .

l If the LC0 and the Required Actions of LCO 3.7.14 are not met while in MODE 1, 2, or 3, there is no safety benefit to t

be gained by placing the plant in a shutdown condition. The Required Action of LCO 3.7.14 of " Suspend movement of irradiated fuel assemblies in spent fuel pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications.

1 LCO 3.0.4 LC0 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It precludes placing the plant in a MODE or other specified condition stated in that Applicability (e.g., Applicability desired to be entered) when the i

following exist: -

a. Plant conditions are such that the requirements of the LC0 would not be met in the Applicability desired to be entered; and

b. Continued noncompliance with the LC0 requirements, if the Applicability were entered, would result in the

! plant being required to exit the Applicability desired to be entered to comply with the Required Actions.

Compliance with Required Actions that permit continued operation of the plant for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the plant before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made

'in accordance with the provisions of the Required Actions.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

Palisades Nuclear Plant B 3.0-5 10/10/98

. . _ . _ . . 7. .

l LCO Applicability B 3.0 i i- l i

BASES 1

LCO 3.0.4 The provisions of LC0 3.0.4 shall not prevent changes in (continued) MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. l Exceptions to LC0 3.0.4 are stated in the individual Specifications. The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated' ACTIONS to be entered do not provide for continued operation for an unlimited period of time.

Exceptions may apply to all the ACTIONS or to a specific

, Required Action of a Specification.

l- l l

Surveillances do not have to be performed on the associated i inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS

Condition, in compliance with LC0 3.0.4 or where an

exception to LC0 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY.

prior to declaring the associated equipment OPERABLE (or

, variable within limits) and restoring compliance with the affected LCO.

3

' LC0 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been

. removed from service or declared inoperable to comply with I ~ ACTIONS. The sole purpose of this Specification is to provide an exception to LC0 3.0.2

- the applicable Required Action (s))(e.g., to allow.the to'performance not comply with of required testing to demonstrate:

a. The 0PERABILITY of the equipment being returned to service; or

b. The OPERABILITY of other equipment.

1 The administrative controls ensure the time the equipment is 4

returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to

. perform ~the required testing to demonstrate OPERABILITY.

l This Specification does not provide time to perform any j other preventive or corrective maintenance.

4 Y

i

?-

1 Palisades Nuclear Plant B 3.0-6 10/10/98

- .yl,- ...a, - - ,w-. ,- -.-.n..-y- ,..,,m ,,m,,.

i LC0 Applicability l B 3.0

~

BASES LCO 3.0.5 An example of demonstrating the OPERABILITY of the equipment (continued) being returned to service is reopening.a containment isolation valve that has been closed to comply with Required -

Actions and must be reopened to perform the required testing.

An example of demonstrating the OPERABILITY of other l equipment is taking an inoperable channel or trip system out l of the tripped condition to prevent the trip function from i occurring during the performance of required testing on I another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is i taking an inoperable channel or trip system out of the l tripped condition to permit the logic to function and l indicate the appropriate response during the performance of required testing on another channel in the same trip system.

LCO 3.0.6 LC0 3.0.6 establishes an exception to LCO 3.0.2 for support

• i systems that have an LC0 specified in the Technical t

Specifications (TS). This exception is provided because LC0 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LC0 be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the plant is maintained in a l safe condition are specified in the support system LC0's L Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.

! When a support system is inoperable and there is an LC0 l specified for it in the TS, the supported system (s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability, j .However, it is not necessary to-enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCO's Conditions and Required Actions are

! eliminated by providing all the actions that are necessary

! to ensure the plant is maintained in a safe condition in the

support system's Required Actions.

4 l

Palisades Nuclear Plant B 3.0-7 10/10/98 l.

LCO Applicability I B 3.0 BASES LCO 3.0.6 However, there are instances where a support system's (continued) Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and .l Required Actions 'for the supported system. This may occur inmediately or after some specified delay to perform some other Required Action. Regardless of whether it is innediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required l Actions fce a supported system, the applicable Conditions l and Required Actions shall be entered in accordance with l LCO 3.0.2.

l Specification 5.5.13 " Safety Functions Determination Program (SFDP)," ensures loss of safety function is detected l and appropriate actions are taken. Upon entry into LC0 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and

corresponding exception to entering supported system l Conditions and Required Actions. The SFDP implements the

l. requirements of LC0 3.0.6.

Cross train cheers to identify a loss of safety function for thor support systems that support multiple and redundant i safety systems are required. The cross train check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.

If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.

i-(

i Palisades Nuclear Plant B 3.0-8 10/10/98

l LCO Applicability B 3.0 BASES i LC0 3.0.7 Special tests and operations are required at various times over the plant's life to demonstrate performance characteristics, to. perform maintenance activities, and to perform special evaluations. Because TS normally preclude i these tests and operations, Special Test Exceptions (STEs) l allow specified requirements to be changed or suspended j under controlled conditions. STEs are included in i applicable sections of the Specifications. Unless otherwise l specified, all other TS requirements remain unchanged and in '

l effect as applicable. This will ensure.that all appropriate

[ requirements of the MODE or other specified condition not directly associated with or required to be changed or suspended to perform the special-test or operation will remain in effect.

The Applicability of an STE LCO represents a condition not necessarily in compliance with the normal requirements of ^

the TS. Compliance with STE LC0 is optional.

A special test may be performed under either the provisions ~

of the appropriate STE LC0 or the other applicable TS requi rements. If it is desired to perform the special test l under the provisions of the STE LCO, the requirements of the STE LCO shall be followed. This includes the SRs specified j in the STE LCO.

l- Some of the STE LCO require that one or more of the LC0 for l normal operation be met (i.e., meeting the STE LCO requires

meeting the specified normal LCO). The Applicability.

ACTIONS, and SRs of the specified normal LCO, however, are not required to be met in order to meet the STE LCO when it is in effect. This means that, upon failure to meet a specified normal LCO, the associated ACTIONS of the STE LC0

, apply, in lieu of the ACTIONS of the normal LCO. Exceptions

! to the above do exist.

.There are instances when the Applicability of6 the specified normal LC0 must be met, where its ACTIONS must be taken, where certain of its Surveillances must be performed, or where all of these requirements must be met concurrently with the requirements of the STE LCO.

Unless the SRs of the specified normal LC0 are suspended or changed by the special test, those SRs that are necessary to i meet the specified normal LC0 must be met prior to performing the special test. During the conduct of the special test, those Surveillances need not be performed

unless specified by the ACTIONS or SRs of the STE LCO.

l Palisades Nuclear Plant B 3.0-9 10/10/98

j LCO Applicability c

8 3.0 BASES l

! LC0 3.0.7 ACTIONS for STE LCO provide appropriate remedial measures i (continued) upon failure to meet the STE LCO. Upon failure to meet l

these ACTIONS, suspend the performance of the special test -

i and enter the ACTIONS for all LCOs that are then not met. l l Entry into LC0 3.0.3 may possibly be required, but this l l

determination should not be made by considering only the '

failure to meet the ACTIONS of the STE LCO, l

l i

i Palisades Nuclear Plant B 3.0-10 10/10/98

l f

l 1

! SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY l BASES l

SR 3.0.1 through SR 3.0.4 establish the general requirements

~

SRs applicable to all Specifications and apply at all times,

j. unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LC0 apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed

, to verify the OPERABILITY of systems and components, and l that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance l with SR 3.0.2, constitutes a failure to meet an LCO.

l Systems and components are assumed to be OPERABLE when the .

associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

l

a. The systems or components are known to be inoperable, although still meeting the SRs; or 1

b. The requirements of the Surveillance (s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the plant is in a MODE or other specified condition for which the requirements of the associated LC0 are not applicable, unless otherwise specified. The SRs associated with a Special Test Exception (STE) are only applicable when the STE is used as an allowable exception to the requirements of a Specification.

! Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

Palisades Nuclear Plant B 3.0-11 10/10/98

7 SR Applicability l B 3.0 BASES SR 3.0.1 Surveillances, including Surveillances invoked by Required (continued) Actions, do not have to be performed on inoperable equipment .

because the ACTIONS define the remedial measures that apply.

Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the l

Applicability due to the necessary plant parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to -

, proceed.to a MODE or other specified condition where other

necessary post maintenance tests can be completed.

An example of this process is:

a. High Pressure Safety Injection (HPSI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPSI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenante testing.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required

' Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."

interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g.,

transient conditions or other ongoing Surveillance or maintenance activities).

Palisades Nuclear Plant B 3.0 12 10/10/98 l

l

I SR Applicability B 3.0 BASES SR 3.0.2 The 25% extension does not significantly degrade the (continued) reliability that results from performing the Surveillance at ,

l its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is the Containment Leak Rate Testing Program.

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per..." basis. The 25%

extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% -

extension to this Completion Time is that such an action i usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used

, repeatedly merely as an operational convenience to extend l

Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not

'been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is less, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

Palisades Nuclear Plant B 3.0-13 10/10/98 l

l

1 SR Applicability i B 3.0 l

BASES l

SR 3.0.3 This delay period provides an adequate time to complete (continued) Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of plant conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillar.ce, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a Frequency based not on time intervals, but upon specified plant conditions or operational situations, is discovered not to have been performed when specified, SR 3.0.3 allows the full delay -

period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform the Surveillance.

SR 3.0.3 also provides a time limit for completion of Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

If a Surveillance is not completed within the allowed delay period, then the equiprnent is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LC0 Conditions begin imediately upon expiration of the

' delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LC0 Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

l l

Palisades Nuclear Plant B 3.0-14 10/10/98

i SR Applicability i

, B 3.0 i 4

BASES SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or.other specified .

Condition in the Applicability.

l This Specification ensures that system and component OPERABILITY requirements and variable limits are met'before 1 entry into MODES or other specified conditions in the  !

Applicability for which these systems and ceroonents ensure

. safe operation of the plant.

i The provisions of this Specification should not be interpreted as endorsing the failure to exercise the: good 1

practice of restoring systems or components to CFERABLE status before entering an associated MODE or other specified condition in the Applicability. I i

However, in certain circumstances, failing to meet m SR

l. will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, sutsystem, -

division, component, device, or variable is inoperable or '

outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances der not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance (s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LC0 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.

The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability

,that are required to comply with ACTIONS. l l

[

Palisades Nuclear Plant B 3.0-15 10/10/58

SR Applicability B 3.0 BASES SR 3.0.4 The precise requirements for performance of SRs are l (continued) specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition (s)'specified in a Surveillance procedure require entry into the MODE or other specified condition in the ,

Applicability of the associated LC0 prior to the performance -l or completion of a Surveillance. A Surveillance that could not be performed until after entering the LC0 Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form s. a Note as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

-l Palisades Nuclear Plant B 3.0-16 10/10/98

I i

1 l

LCO Applicability 3.0 4

3.0 LCO APPLICA8ILITY LC0 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in tt Applicability that are required

, 4 to emply with ACTION ,'at/are sfrt of/a shutdo6n of thel

~

1 4

Exceptions to this Specification are stated in the 1 individual _ Specifications. jThese cepu ans iiow entry l finto DES or ther spec ted con tions in he App cabilit when the ssociat ACTIONS o be enter I I a ow unit peration the M0 or other specific >d i

, onditio -_in the Ao icabilityp only for a limited riod a ti LCO 3.0. is only appl.er' a for entry into MODE or other

/ specifi d con.11 tion in t Applicability in ODES 1, 2, 3, q and 4. -

Reviewers's Note: 0 3.0.4 has been rev sed so that changes in MODES or other specified cond tions in the Applicability tha are part of a shutdo n of the unit shall

• l not be prevented. In addition, LCO 3. 4 has been revised

so that it is o y applicable for ent y into a MODE or other i

. specified condi ion in the Applicabi ity in MODES 1, 2, 3,< ~

and 4. The M E change restriction in LCO 3.0.4 were i previously ap licable in all MODES- Before this versionfof LCO 3.0.4 c be implemented on a plant specific basis,j'the licensee m t review the existin technical specificat ons  !

to determi e where specific res rictions on MODE chan es or j Required etions should be inc uded in individual LC s to ,

justify his change; such an valuation should be s mmarized in a ma rix of all existing C0s to facilitate NRC staff

. review of a conversion to t e STS. _

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under  !

administrative control solely to perform testing requireo to l demonstrate its ODERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform  !

, the testing required to demonstrate OPERABILITY. l i

l l

l (continued) i l

CEOG STS 3.0 2 Rev 1, 04/07/95  !

/g i O === ,

1stierte e

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

SR Applicability

3.0 3.0 SR APPLICABILI1Y SR 3.0.3 declared not met, and the applicatle Condition (s) must be (continued) entered.

SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LCO's i Surveillances have been met within their specified Frequency. This provision shall not prevent entry into MODES or other specified conditions in the Applicability _

h j that are required sNutdowf or the to unit./ comply withrACTIONS [or t are art or g l 140 3.0.4 is nly appllcable f entry into a MODE r other 5j O[, f specified c dition in the Ap cability in MODE , 2, 3, L and 4. .

Reviewer's N e: SR 3.0.4 has bee revised so that chan s in MODES or ther specified condit ons in the Applicabi ty .

that are p t of a shutdown of th unit shall not be prevented. In addition, SR 3.0. has been revised so hat it is on1 applicable for entry into a MODE or other

6 specifie condition in the App icability in MODES 1, 2, 3 .

and 4. he MODE change restr tions in SR 3.0.4 we e i- previo ly applicable in all DES. Before this y rsion of SR 3. 4 can be implemented n a plant-specific b sis, the lice se must review the e sting technical spec fications to termine where specifi restrictions on M00 changes or i Re tred Actions should b included in individ al LCOs to ju tify this change; suc an evaluation shoul be sumarized 1 a matrix of all exis'ing LCOs to facilita NRC staff eview of a conversion /;o the STS. _

t i

I CE06 STS 3.0 5 Rev 1, 04/07/95 RERASED 1stwee e .

i LCO Applicability B 3.0

. 1 i BASES ,,

1l

'LC0 3.0.4 ACTIONS.) The second type of Required Action specifies the

(continued). r ial measures that permit continued operation of the i g that is not further restricted.by the Completion Time.

' L his case, compliance with the Required Actions provides

, an acceptable level of safety for continued operation.

I Completing the Required Actions is not required when an LCO

~

is met or is no longer applicable, unless otherwise stated j in the individual Specifications.

l The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the i- Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case.

l L{ } An example of this is in LCO 3.4.3, " @ Pressure and Temperature (P/T) Limits." gc5 The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to,'

performance of Surveillances, preventive maintenance. -

corrective maintenance, or investigation of operational lQy(3'@j(

Meu,d e $

. problems. Entering ACTIONS for these reasons must be done -

in a manner that does not compromise safety. Intentional Mo Ac 05 ;- entry into ACTIONS should not be made for operational convenience.Wuternatwavmt would @ result in C2--gg;gep> redundant equipment being inoperable 6should be used instead.

Doing so limits the time both subsystems / trains of a safety function are inoperable and limits the time 6 conditions exist whicharesult in LCO 3.0.3 being entered. Individual O#3 Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is "

h required to comply with Required Actions, thei a MODE or other specified condition in which anot er may enter Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.

(continued)

CE0G STS B 3.0-2 Rev 1, 04/07/95 e nsen 1eNeme 4

l g .

LCO Applicability B 3.0

' ' ~

BASES LC0' 3.0.4 The provisions of LCO 3.0.4 shall not prevent changes in (continued) MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. IIn a 1tlon, he ,/

provi ons LCO 3. 4 shall ot pr nt cha es in ES ' /

O6 or her s ified onditto in t pplic i ity at r uit f any it shut .

Exceptions to LC0 3.0.4 are stated in the individual SpecificationsJ. Exceptions may apply to all the ACTIONS or to a specificfRequired Action of a Specification.

Og /7#$fET

\

U/ LCO 3.0.4 is onT applicable who entering MODE 4 fr MODE 5, MODE 3 ' rom MODE 4, M00 2 from MODE 3, or E1 from MODE 2. urthermore, LCO .0.4 is applicabl when {

entering an other specified ondition in the Ap icability only while perating in MOD 1, 2, 3, or 4.

F.

O(o requirene s of LCO 3.0.4 not apply in M00 T

5 and 6, or in other specified condit ons of the Applica lity (unless in MOD 1,2,3,or4) cause the ACTIONS f individual Speci cations suffici tly define the re dial measures to

, be ken. l [ some cases (e.g., ..) these ACTION 5 pr Ide a Note that s tes "While this LC is not met, entry i oa MODE or ot r specified condit n in the Applicabil y is; not permi ed, unless require to comply with ACT! S."

[

This Note is a requirement e licitly precluding try into a MODE o other specified c dition of the Appli bility.)

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, in compliance with LC0 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LC0 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with (continued)

CEOG STS B 3.0-6 Rev 1, 04/07/95 REVISED 1effeme N

. CHAPTER 3.0 INSERT 1 The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited

! period of time.

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B 3.0 l

! BASES i

SR 3.0.4' that are require.d to comply with ACTIONS. lin ddition he (continued) provi ons of 0 3.0.4 all not revent c nges in DES r I

or o er sp ied co tions i the Appli ability at re it fro any unit hutdown The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance l or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, l the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of l the specific formats of SRs' annotation is found in

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Section 1.4, Frequency. ~

• m

) SR 3.0.4 is on1 applicable when en ring MODE 4 from

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) MODE 5, MODE 3 rom MODE 4, Mode 2 rom MODE 3, or M00 1 from MODE 2. - urthermore, SR 3.0 is applicable wher l entering any other specified con ition in the Applic4b)ility b only while perating in MODES 1 2, 3, or 4. The /c requireme s of SR 3.0.4 do n apply in MODES S a 6, or in other specified condition of the Applicabilit (unless in MODE 1, 2, 3, or 4) bec se the ACTIONS of i dividual Specif cations sufficienti define the remedial measures to be t en.

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! CEOG STS B 3.0-16 Rev 1, 04/07/95 4

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ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS 1 CHAPTER 3.0, LIM.ITING CONDITION FOR OPERATION (LCO) APPLICABILITY 'l l

Channe  : Discussion NOTE: The first five justifications for these changes from NUREG-1432 were

{

generically used throughout the individual LCO section markups. Not all generic justifications are used in each section.

1. The brackets have been removed and the proper plant specific information or value has been provided. I

2. Editorial change for clarity or for consistency with the Improved Technical i Specifications (ITS) Writer's Guide.

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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 I reflect the facility specific nomenclature, number, reference, system description, or analysis description. ,

5. This change reflects the current licensing basis / technical specifications.

6. Rev. I to NUREG-1432 contains the option of limiting the Applicability of LCO and SR 3.0.4 to entry into a MODE or other specified condition in the Applicability in MODES 1,2,3 and 4. In addition, the Rev.1 version of LCO and SR 3.0.4 would not prevent entry into MODES or other specified conditions in the Applicability that are part of any shutdown of the unit. However, to adopt the Rev.1 version of LCO and SR 3.0.4, an evaluation must be performed on the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions l should be included in individual LCOs to justify this change. For the Palisades plant l there appears to be little benefit to adopt the Rev. I definition. In addition, since the l process for determining where specific restrictions on MODE changes should occur has not been well defined, Palisades chooses not to adopt the allowances of the Rev.1 LCO and SR 3.0.4.

I Palisades Nuclear Plant Page I of 4 10/10/98

yi ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 3.0, LIMITING CONDITION FOR OPERA

• ION (LCO) APPLICABILITY Change Discussion i

7. The Bases of SR 3.0.1 has been modified to add the following statement: i

" Unplanned events may satisfy the requirements (including applicable -

acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those i SRs whose performance is normally precluded in a given MODE or other J

.specified condition." l This change was made as a result of a proposed change by the industry owner's '

groups as discussed in Technical Specification Task Force (TSTF) 8, Rev. 2. The l proposed change acknowledges that credit may be taken for an unplanned event which demonstrates the operability of the system equivalent to the performance of the i associated surveillance requirement. l 1 l

8. The proposed Bases for 3.0.2 is modified to reflect the inclusion of a Containment. l 12ak Rate Testing Program in CTS by Amendment 174. That change was retained in l the ITS and is reflected in the wording of the Frequency of the SRs in ITS Section ,l 3.6. The affected ITS SRs do not refer to the requirements of 10 CFR 50 Appendix l J, but to the Containment Leak Rate Testing Program instead. l l

In addition to the reason in the STS bases for SR 3.0.2 not being applicable, CTS l Amendment 182 placed the restriction that SR 4.0.2 (ITS SR 3.0.2) could not be l applied to the Primary Coolant Pump Flywheel Testing Surveillance Program. l Therefore, the Bases for ITS SR 3.0.2 is revised from that of STS to reflect these l plant specific differences. l l

The existing NUREG-1431, Rev.1 Bases for SR 3.0.2 stated at the end of the l paragraph "The 25% extension...": l l

, "An example of where SR 3.0.2 does not apply is a Surveillance with a l Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by I approved exemptions." The requirements of regulations take precedence over j the TS'. The TS cannot in and of themselves extend a test interval specified in l the regulations. Therefore, there is a Note in the Frequency stating, "SR l 3.0.2 is not applicable." l l

The ITS bases change this sentence to state: l l

An example of where SR 3.0.2 does not apply is the Containment Leak Rate l Testing Program. l I

l Palisades Nuclear Plant Page 2 of 4 10/10/98

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ATTACHMENT 6

{ JUSTIFICATIONS FOR DEVIATIONS CHAPTER 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY Change -

Discussion

9. In the proposed ITS LCO 3.0.3b, NUREG 1432 lists a time to reach MODE 4 of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This item is " bracketed" in NUREG-1432 since some plant designs require a different number. In the proposed Palisades ITS, this time is increased to 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />. .

Increasing the time allowed to reach MODE 4 allows for more complete degassing of the Primary Coolant System (PCS). The PCS is degassed by venting the pressurizer gas space to the Vacuum Degasifier. The efficiency of this method is maximized by i maintaining PCS temperature as high as practical, the subcooling as low as practical, and operating all pressurizer heaters. This results in a net increase in the rate of hydrogen removal from the PCS since increased spray flow and lower PCS pressure

[ offset the lower degas flow rate through the vent path. While the total time to reach MODE 4 is increased, the time to reach MODE 5 is the same in the proposed ITS as specified in NUREG-1432.

10. TSTF-165 is incorporated into the Bases for LCO 3.0.5. The Bases for LCO 3.0.5 is changed to use the word " testing" instead of the acronym "SR." LCO 3.0.5 states,-

" Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing ,

required tc, demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY," While LCO 3.0.5 refers to " testing", the Bases for LCO 3.0.5 inconsistently use the term "SRs" instead of " testing." This change corrects this inconsistency. This change addresses testing that is required to demonstrate operability that is not a surveillance. For example, post maintenance testing required to demonstrate operability may not be a Surveillance. This change does not change the intent of the LCO and makes the Bases consistent with the LCO.

11. TSTF-166 is incorporated to revise LCO 3.0.6 to explicitly require an evaluation per the Safety Function Determination Program, and delete the statement " additional . . .

limitations may be required" from LCO 3.0.6. There is an inconsistency between LCO 3.0.6, the Safety Function Determination Program (SFDP), and the LCO 3.0.6 Bases. As currently written, LCO 3.0.6 does not explicitly require an evaluation in accordance with the SFDP, rather it states that additional evaluations may be required. Both the SFDP and the LCO 3.0.6 Bases state that upon entry into LCO 3.0.6, an evaluation shall be made to determine if a loss of safety function exists. In addition, because LCO 3.0.6 states that the evaluation be done in accordance with the SFDP and the SFDP states that other appropriate actions may be taken, there is no need for the statement " additional . . . limitations may be required" in LCO 3.0.6.

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Palisades Nuclear Plant Page 3 of 4 10/10/98

1 II ATTACIBIENT 6 l JUSTIFICATIONS FOR DEVIATIONS l l CHAPTER 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY l l

Change -

Discussion

12. TSTF-122 is incorporated to revise the LCO 3.0.2 Bases to remove possible confusion. This change revises the following two sentences, " Alternatives that would' l not result in redundant equipment being inoperable should be used instead. Doing so l limits the time other conditions exist which result in LCO 3.0.3 being entered." to

(

read, " Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing so limits the time conditions exist which may result in LCO 3.0.3 being entered." The original I wording is confusing in that it begins to discuss inoperability of redundant equipment without introducing the topic. This topic of inoperable redundant equipment seems to l be more appropriate for the Bases of LCO 3.0.3, but an appropriate discussion is i already provided there. The proposed wording retains the intent while presenting the I material in the appropriate context of LCO 3.0.2.

1 i 13. TSTF-104 is incorporated to relocate a discussion of exceptions from LCO 3.0.4 to l

the Bases. This change removes the additional discussion provided in LCO 3.0.4 with respect to the use of exceptions and provides the necessary discussion in the Bases. This change provides consistency with LCO 3.0.3 by moving the discussion of exceptions from the LCO to the Bases. In addition, this change reduces the potential for confusion by revising the discussion to eliminate the repeated use of the phrase " Modes or other specified conditions in the Applicability" to increase clarity.

14. TSTF-71, Rev.1, is not incorporated into the Bases of LCO 3.0.6. This change does not affect the Specification but would only add an example of SFDP application to the LCO 3.0.6 Bases. This brief example is not added in the Bases for LCO 3.0.6 since

it considered to be unnecessary, insufficient, and potentially the cause of further confusion. Although the application of LCO 3.0.6 is the cause of a great deal of confusion, the application can be better explained in the implementation documents using other more detailed examples. This change to the Bases does not alter the technical content of LCO 3.0.6.

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Palisades Nuclear Plant Page 4 of 4 10/10/98

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ATTACHMENT 4

, , CONSUMERS ENERGY COMPANY ,

PALISADES PLANT DOCKET 50-255

! CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION l

REVISED PAGES FOR SECTION 4.0 l

4 l

i CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE AUGUST 21, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 4.0 Paae chanae Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REY DATE NRC COMMENT #

ATTACHMENT 1 No page change ATTACHMENT 2 No page change ATTACHMENT 3 CTS page 5 of 8 CTS page 5 of 8 10/10/98 RAI 4.0-2 -

DOC page 2 of 6 000 page 2 of 6 10/10/98 RAI 4.0-1 RAI 4.0-2 ATTACHMENT 4 No page change ATTACHMENT 5 No page change ATTACHMENT 6 No page change l

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.-- . . . . . . -- . - . ~ . .. --

s-Clufu4.o

@ Q Fuil Storaae I

a. Irradiftedfuelbundlesw)(1bestored,priordooffsiteshipment (')

in the stainless steel lined spent fuel pooV. " V- ,

l lb. 40eletedV l h g g 3,3,37, ,gg g -

h@ The spent fuel storage pool into two regions identifie

'ano spare (north) tilt pit are divided as Region I and Region !! as l

l illustrated in Figure . Region I racks are designed and shall g be maintained with a nominal 10.25' center to center distance between fuel assemblies with the exception of the single Type E rack which has a nominal 11.25' center to center distance between fuel assemblies. The Region I spent fuel storage racks are designed such that fuel having a maximum assembly planar average 4 3.1.% U3s enrichment of 4.40 w/o placed in the racks would result in a 9 y,4 K ,, equivalent to s 0.95 when flooded with unborated water. The K,., of s 0.95 includes a conservative allowance for uncertainties. '

q al,Q For enrichments above 3.27 w/o U: 3., the fuel assemblies must i

, contain 216 rods which are either U0, 2 Gdr0 3 DO: or solid metal. I y$

y Region !! racks have a 9.17 inch center to center spacing.lBecause f this s Ier spaciny, strict cent.cols are empi d to evaluate l urnup o he fuel asambly orior t4 itLphtme in_Regico_IL _. /8' , V) ell loc ions.1 Upon determination that the fue assemely meets O Q*,*\ the burnup requirements of ITable L F U._phcemenLin a Region !! }7;p. tim fp I

p ) qp' 'g cell is authorized. ITyese positive / controls assure / Int f uer enrich 9Ipnt limits assVmed in the s4fety analyses wfil not be 4 exceedd.l '

le. / (DeletedQI h I

f. The inimum spent fuel pool w er boron concentration hall be ce 3 I ppe. Boron concentrati shall be verified at ast once nthly.

19 The sgent fuel rackf are designed as/a Class I str/cture.l l h. / (Delet/d)l h i3 Uj $ Storage in Region !! of the spent fuel pool and spare (north) tilt pit shall be restricted by burnup and enrichment limits specified

( Sa aba M 7 in ITabM 5.Mll4g ,g, g NOTE: ntil needed for f '

e l storage, one R ion !! rack in th northeast corner the spent fuel 1 may be removed and 4. l '

replaced with t cask anti tippin device.

Ra faradcas

.\ SAR Update Cha er 5 l

FSAR Update Ch ter 9

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( ,@D 411., b n wy, b rn e ased '1T5) 5 4a g A01) '63 3) Cv$cmf, Fe,8 fWbmso N dment No. W6, H4, 140 January 23, 1992 h.y 5 .C f

)

ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN FEATURES A.4 CTS 5.1 states "The Palisades reactor shall be located on...." The proposed ITS states "The Palisades Nuclear Plant is located in...." This change more appropriately reflects the plant name. In addition "shall be" is changed to "is" to improve the .

sentence wording and reflect the first that the Palisades Nuclear Plant has already been built. These changes are considered to be administrative changes since no requirements have changed.

A.5 The Palisades CTS does not address inadvertent draining of the storage pool. The suction and discharge piping of the cooling system for the storage pool was designed to prevent inadvertent draining. The discharge piping is at 647' and contains a siphon .

l breaker designed to prevent siphoning fuel pool level below about 648'. The l l discharge pipe itself is cut off a Elevation 645'. The bottom of the suction piping is l at elevation 644'5." Failure # a fuel pool cooling pipe cannot drain the fuel pool l below 644'5". Since these piping arrangements are permanent plant features, and no additional operational requirements have been imposed the inclusion of this -

information into the proposed ITS is considered to be an administrative change. This change is consistent with the intent of NUREG-1432. ,

A.6 CTS 5.4.la contains certain design aspects of the new fuel storage racks and includes a reference to Siemens Nuclear Power Corporation Report EMF-91-1421 (NP) for the appropriate conservatism used in the calculation of K,y. In proposed ITS 4.3.1.3, reference to the Siemens Nuclear Power Corporation Report has been replaced by a reference to FSAR Section 9.11. Section 9.11 of the FSAR documents the design and analysis for the Fuel Handling and Storage Systems. This change is considered i

administrative in nature since it does not alter the design or analysis assumptions of the new fuel storage racks, but merely revises the reference of the document which contains the uncertainties used in the determination of K,y. This change is consistent with NUREG-1432.

l- A.7 A new section,4.3.3, has been added to the Design Features Section of ITS to l emulate the contents of STS. This section is descriptive, and contains no l requirements. Similar material is contained in the FSAR and other plant documents.

l Since this change contains no requirements, the change is classified as Administrative.

l TECHNICAL CHANGES - MORE RESTRICTIVE (M)

There were no "More Restrictive" changes added to this chapter.

Palisades Nuclear Plant Page 2 of 6 10/10/98

ATTACHMENT 5 i

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CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255

'I CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE OCTOBER 1, 1998 TELEPHONE QUESTIONS REVISED PAGES FOR SECTION 5.0 A

j.

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

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE OCTOBER 1, 1998 TELEPHONE QUESTIONS REVISED PAGES FOR SECTION 5.0 Pace Chanac Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.

REMOVE PAGES INSERT PAGES REV DATE NRC COMME 11_f ATTACHMENT 1 No page change ATTACHMENT 2 No page change ATTACHMENT 3

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CTS markup 4-20 CTS markup 4-20 10/10/98 RAI 5.5-04 CTS markup 6-17 CTS markup 6-17 10/10/98 RAI 5.5-04 DOC pages 1 through 6 Pages 1 through 7 10/10/98 RAls 5.5-03 & 04 ATTACHMENT 4 Pages 1 through 5 Pages 1 through 7 10/10/98 RAI 5.5-04 ATTACHMENT 5 No page change ATTACHMENT 6 No page change

t n 50 ADtv>M '

dpWT M 4.5 CONTAIWlENT TESTS .

4.5.2 Local taak Detection Tests (continued)

b. Accentar.ca Criteria (1) The total leakage from all penetrations and isolation valves

g. shall not exceed 0.60 L,.

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{*{ (2) The leaka exceed 0.be 23 L.,for a Personnel airlock door seal test shall not An acceptable Emergency Escape Airlock door seal contact check

((3) consists seals andofthe a verification of continuouc contact between the sealing surfaces,

c. Corrective Action (1) If at any time it is etermined that 0.60 L, is exceeded, repairs shall be in lated immediately. If repairs are not '

completed and conf ance to the accepthnce criterion of f 3 is not emonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the plant (p~ 4.5.2.b(1) shall be p laced at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in LD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Noj (2) If at any ti 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> d in COLD SH DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(3) If the P sonnel airlock door seal leakage is greater th . I 0.023 L , or if the Emergency Escape Lock door seal con et check ails to meet its acceptance criterion, repairs all be initi ed immediately to restore tne 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 t shall be placed in at least HOT SHUTDOWN wi in the next 6 ours and in COLD SHUTDOWN within the followin 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l (4) f air lock door seal leakage results in one d r causing total containment leakage to exceed 0.60 L., e door shall be declared inoperable and the remaining OpERAB door shall be innediately locked closed

• and tested withi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As long as the remaining door is found to be OPE LE, the provisions of 4.5.2.c(2) do not apply. Repairs shal be initiated immediatel specification 4.5.2.b(1)y toInestablish

. the eventconformance conformance wi this specification cannot be established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> he plant shall be placed in at least HOT SHUTDOWN with the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 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 ru__ ,

i REV 2 Amendment No. W6, W4, W

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I 5.0 % C4,a ,

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8.0 AfMINISTRATIVE CONTROLS 55.13 6.5.13 W

. Ae s.n+pbs 06~e,Jem hoy..,ov% 3,

.5 p se4e4 to rrs 5.5, N 4.5.14 cantai-at taak anta Testine proaram Programs shall be estabitshed to implement the leak rate testing of the i containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, L Option B, as medtrted by approved exemptions. The Type A test program shall meet the requirements of 10 CFR 50 Appendix J. Option I and shall be in accordance with the guidelines of Regulatory Guide 1.163,

' Performance. lased Containment Lankage Test Program, dated September .

1995.* The Type I and Type C test program shall meet the requirements-of 10 CFR 50 Appendix J, Option A, as modified by the exemption froe c 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 desion basis loss of coolant accident, P., is 52.64 psig ((FSARAable 14/18.1 4 D. bl The maximum allowable containment leak rate, L., at P., shall be 0.17. of containment air weight per day.

Leak rate acceptance criteria are:

a. Containment leak rate acceptance criteria is 5 1.0 L . During the first plant startup following testing in accordance with this program, the leak rate acceptance criteria are 5 0.60 L, for the Type 5 and Type C tests and 5 0.75 L, for Type A tests; ~

AN b. Air lock leak rate acceptance criteria is 5 0.023 L, for each door, when pressurized to 2 10 pstg.

g) 3 4.0.2 ut.nnt applicable to The Surveillance interval extensions of

- the Centainment Leak Rate Testing Program requirements.

The previsions of .3 Ar,g applicable to the Containment Leak Rate Testing Progran requirements.

6 17 Amendment No. 174

' October 31, 1996 i

27 Ok N

\

I annsen 1stietse T .

1 RAI INSERTS FOR SECTION 5.0 NewInsertforUTSpage6-17:

Containment OPERABILITY is equivalent to " Containment Integrity" for the purposes of the air lock testing requirements in 10 CFR 50, Appendix J. ,

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' ATTACHMENT 3

~ DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS ADMINISTRATIVE CHANGES (A)

A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical 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 u. sed 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 the typical Palisades Nuclear Plant terminology.

l Palisades Nuclear Plant Page 1 of 7 10/10/98

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ATTACHMENT 3

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

l A.7 CTS 6.5.7 is entitled " Inservice Inspection and Testing 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.1 lists, among referenced'LCOs, "3.10.1." That item is unnecessary and l has been deleted. Neither CTS 3.10.1, nor its ITS replacement reference the COLR. l l CTS 6.6.5 a. lists the core operating limits that are established and documented in the l COLR prior to each core reload. Specifically, these limits are: ASI Limits l (CTS 3.1.1), Regulating Group Insertions Limits (CTS 3.10,5), Linear Heat Rate l Limits (CTS 3.23.1), and Radial Peaking Factor Limits (CTS 3.23.2). CTS 6.6.5 b. l list the documents approved by the NRC that describe the analytical methods used to I determine the core operating limits. As part of this listing, cross references are made l to the LCOs pertaining to the affected limit (e.g., ASI Limits, Regalating Group l Insertion Limits, etc...). In error, CTS 6.6.5 b.1. lists CTS 3.10.1 (Shutdown l Margin Requirements) as an LCO related to a document that describes analytical l methods used to determine the core operating limits. Since Shutdown Margin is not a j cycle dependent limit (the limit is contained in the technical specifications and not in l j the COLR), referencing CTS 3.10.1 in CTS 6.6.5 b.1 is inappropriate and has been l l deleted. This change has been characterized as administrative in nature since it does l l not alter any requirement of the CTS, but simply corrects an administrative oversight. l l Palisades Nuclear Plant Page 2 of 7 10/10/98

ATTACHMENT 3

_ DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS A.9 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 Stmetural 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 Penetration tests have been deleted since the requirement for these tests was removed from the i 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 l

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 -

l delamination inspection. As stated in CTS 4.5.6, a dome delamination inspection l 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 I 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 l consistent with NUREG-1432 as modified by TSTF-152.

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ATTACHMENT 3

_ DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS  ;

A.13 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, this is inconsistent with typical NRC submittal requirements. This change is l i

considered administrative since it has no effect on plant operations and impacts only j the submittal of after-the-fact information. This change is consistent with l

NUREG-1432.

]

A.14 An additional paragraph was added to ITS 5.5.14 to assure correct application of l  !

those 10 CFR 50 Appendix J testing requirements (e.g., III.D.2.(b)(ii)) which are l ,

applicable "when containment integrity is required by the plant's Technical l l

Specifications." This change is considered to be Administrative, since it simply l l assures that the omission of the CTS term " Containment Integrity" does not affect the l l interpretation of Appendix J testing requirements. l l l

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M.1 CTS 6.4.1 requires that written prccedures be established, implemented, and f maintained for the listed activities. Proposed ITS 5.4.1 contains the same wordmg. l However, proposed ITS 5.4.1.b is not in the CTS and is being added. Proposed l

4 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 I which will ensure the capability to obtain and analyze reactor coolant, radioactive I iodines and particula *.es in plant gaseous effluents,...." In the proposed ITS, the reference is to " radioactive gases" rather thanjust radioactive iodines. Because the I 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 wi:A 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 Stmetural Integrity Surveillance Program."

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1 ATTACHMENT 3

. DISCUSSION OF CHANGES CHAl'I'ER 5.0, ADMINISTRATIVE CONTROLS M.4 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 i additional evaluations to identify a loss of safety function than what is required in the l

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 accident

, 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 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 4

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 i proposed specification emulates the ISTS treatment of containment structural integrity surveillance requirements. The details associated with containment tendon 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 i public health and safety. Removal of these details from tile CTS and the incorporation of a containment tendon surveillance program in Section 5.0 of the ITS j is consistent with NUREG-1432.

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ATTACHMENT 3

~ DISCUSSION OF CHANGES CHAPTER 5.0, ADMINISTRATIVE CONTROLS LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) i LA.2 CTS 6.5.7, Inservice Inspection and Testing Program has been revised to delete the l .

phrase " including applicable supports." Requirements for inservice inspections of l I ASME Code Class 1,2, and 3 components are specificed in 10 CFR 50.55a(g). As ( )

used in CTS 6.5.7, " applicable supports" is intended to apply to the inspection of l l snubbers. Adaptation of this phrase in the CTS was consistent with the NRC's l approach to address concerns related to the relocation of the Snubber LCO from the l ISTS NUREGs. Sub;equently, the ISTS NUREGs have been modified to delete this l phrase in recognition that it duplicates requirements specifed in the CFRs (See Section l 5.0, JFD 26 addressing TSTF-279). As such, the deletion of this phrase from the l CTS can be made without a significant impact on safety since the inspection of l applicable supports continues to be required by 10 CFR 50.55a(g). l l

l LESS RESTRICTIVE CHANGES (L) -

L.1 The CTS requirements for Type B and C leak rate testing are being revised such that l the leakage limit for Type B and C testing is s .60 La only during the first plant l startup following testing performed in accordance with the Containment Leak Rate l t Test Program. After this, the new limit will now become s 1.0 La. This means l l that if the testing is performed during a refueling outage, the total Type B and C l 1 l leakage must be "as-left" at s .60 La prior resuming power operations. Following l ,

this, the leakage limit for the remainder of time until the test is performed again l l becomes 1.0 La for the total containment leakage. Overall containment integrity is l l

maintained because the results of Type B and C testing must be compared against the l overall containment leakage limit to ensure that the leakage remains s 1.0 La. CTS l 4.5.2.c(1) specifies actions to be taken if .60 La is exceeded for Local Leak Detection l Tests. The actions in 4.5.2.c(1) to initiate repairs immediately and shut down if the l acceptance criteria of 4.5.2.b(1) is not met (ensuring total leakage from all l penetrations and isolation valves shall not exceed .60 L2) will no longer apply. In the l

proposed ITS, the acceptance criteria and testing frequency will only exist in the l E

Containment 1.eak Rate Testing Program which is found in TS Administrative l l Controls Section 5.5.14. This is similar to the CTS Containment Leak Rate Testing l

! Program which is found in CTS 6.5.14. The acceptance criteria in the Containment l Irak Rate Testing Program will be that the overall containment leakage will not l

exceed 1.0 La. These changes are acceptable since the overall containment leakage I i requirements of ITS 3.6.1, which reference the Containment Leak Rate Testing l j Program for the acceptance criteria, remain valid at all times. l l

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