Text

Application for Amend to License DPR-20,revising Administrative Controls Section to Emulate CE STS by Removing Review & Audit Requirements of Section 6.5 & Record Retention Requirements of Section 6.10 from TS
	ML18065A346
Person / Time
Site:	Palisades
Issue date:	12/11/1995
From:	Smedley R; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18065A347	List: ML18065A346; ML18065A348;
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	NUDOCS 9512180017
	Download: ML18065A346 (153)
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co111sumers Power POWERiNii MICHlliAN"S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 December 11, 1 995 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT TECHNICAL SPECIFICATION CHANGE REQUEST - REVISION OF ADMINISTRATIVE CONTROLS A request for a change to the Palisades Technical Specifications is enclosed. This change is desired to allow flexibility in assigning responsibilities and position titles within the Palisades Plant organization. The proposed change revises the Administrative Controls section to emulate the CE Standard Technical Specifications by removing the Review and Audit requirements of Section 6. 5 and the Record Retention requirements of Section 6.10 from the Technical Specifications relocating them in the Consumers Power Company Quality Program Description, Topical Report CPC-2A. It also deletes several requirements which duplicate requirements of 10 CFR 50 and makes numerous administrative changes which result in the Administrative Controls Section being nearly identical to the Standard Technical Specifications, Combustion Engineering Plants, NUREG 1432, both in content and arrangement.

The request to relocate Sections 6.5 and 6.10 from the Technical Specifications to CPC-2A is similar to a Commonwealth Edison change request which relocated the Review, Investigative and Audit Functions, for several Commonwealth Edison plants, from Technical Specifications to the ComEd Quality Assurance Topical Report. That change request was approved on October 20, 1995.

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-9512180017 951211 PDR ADOCK 05000255 P ~P~.- ..

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SUMMARY

OF COMMITMENTS This letter establishes no new commitments and makes no revisions to existing commitments.

Yss-~_..p Richard W Smedley ~

Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Enclosure

~ CONSUMERS POWER COMPANY ~

Docket 50-255 Request for Change to the Technical Specifications License DPR-20 It is requested that the Technical Specifications (TS) contained in the Facility Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on February 21, 1991, for the Palisades Plant be changed as described below. These proposed changes result in an Administrative Controls section which closely emulates the Administrative Controls section of the Standard Technical Specifications for Combustion Engineering Plants, NUREG 1432, Revision 1 (STS).

This emulation of STS involves moving several surveillance requirements from existing Section 4 to Section 6 programs. In each case the existing .

requirements were compared to the STS program to assure that unique requirements were not deleted or replaced and that the proposed TS included the appropriate program references. Each of the STS surveillance requirements (SRs) which references a program was reviewed to assure that appropriate SRs were included in the proposed TS. Attachment 5 provides a compari *son of these STS SRs and the Palisades equivalent.

Five attachments are included with this TS change request:

1. The proposed TS pages

2. Existing pages marked to show the proposed changes

3. A compartson of existing and. proposed Administrative Controls sections

4. A comparison of STS and proposed Administrative Controls sections

5. A listing.of STS surveillance requirements which reference the STS programs, and their equivalent in the proposed Palisades TS.

I. Changes Proposed:

References, in several se~tions, to sub parts of the former 10 CFR 20 were

replaced with the corresponding reference in the revised 10 CFR 20~ These changes in Part 20 references are not discussed individually.

In addition, the specific titles of supervisory positions were replaced by the generic title to allow these titles to be changed without necessitating a change in these TS. Section 6.2.la has been amended to require that the relationship between generic titles *and plant specific titles will be included in the FSAR.

1. Sub-sections 1.1, Qperating Definitions, and 1.2, Miscellaneous Definitions have been combined as 1.0, Definitions.

Definitions for Members of the Public, Process Control Program, Site Boundary, and Unrestricted Area (page 1-5) are deleted from the TS.

Members of the Public, Site Boundary, and Unrestricted Area are terms defined in 10 CFR 20.1003. These definitions need not be restated in the TS. It is proposed in change 44, below, that the Process Control Program no longer be controlled by the TS, so the definition will no longer be needed. These definitions do not appear in STS.

2 Where these defined terms appeared, in the balance of the TS, in capitalized text (to indicate a term defined in Section 1), they have been replaced with lower case text.

The definition for Offsite Dose Calculation Manual (page 1-5) is moved to proposed Section 6.5.1 and reworded slightly to emulate STS.

The only definition remaining in Section 1.2, that for the Core Operating Limits Report (page 1-5), is arranged alphabetically with the definitions from former Section 1.1. The reference to Specification 6.9.1, in the COLR definition, is revised to match the proposed numbering.

2. The requirement to shutdown the reactor following a Safety Limit violation is moved from Section 6.7.1.a (page 6-10) to Sections 2.1.1 and 2.2.1 (page 2-1), which formerly referenced Section 6.7. This change places the operating requirements directly into the Action Statement, as is done in STS, instead of incorporating it by reference. While this TS requirement is redundant to the noted Section of 10 CFR, it is retained in the TS to assure that operators are aware of this immediate action.

3. Section 3.17.4, Accident Monitoring Instrumentation (page 3-70), is revised to reflect the inclusion of an Accident Monitoring Report* into the Administrative Controls section of TS, to emulate STS. Action 3.17.4.7 is revised to reference the reporting requirement of Specification 6.6.7, the proposed reporting requirement. The proposed reporting requirement is unchanged from that of the existing TS.

4. Specification 4.0.5 (page 4-1) has been moved to the Administrative Controls section as Specification 6.5.7, emulating STS Specification 5.5.8. The wording for the proposed specification is taken from STS. There is no change in requirements. The Basis for Section 4 has been reworded accordingly, deleting reference to, and explanation of, Section 4.0.5. References to Specification 4.0.5 in the balance of the TS have been revised to reference Specification 6.5.7.

5. Ventilation filter testing requirements of Table 4.2.3 (page 4-14) have been replaced with a single requirement to "Perform required Control Room Ventilation and Fuel Storage Area filter testing in accordance with the Ventilation Filter Testing Program" emulating STS SR 3.7.11.2 and 3.7.14.2. The wording for the proposed program is taken from STS.

The requirements of Table 4.2.3 which pertain to ventilation system testing other than filter testing have been retained in Table 4.2.3 with the following changes:

Existing Item c.2 (proposed Item 2.a) was editorially revised to more closely agree with current plant terminology.

Existing Item c.3 (proposed Item 2.b) was revised to increase the required differential pressure capability and to require control room pressure to be established with respect to the outside atmosphere and rather than the viewing gallery.

3 The existing requirement to verify control room temperature is below 120°F each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , was revised to require that control room temperature is below 90°F. This change is proposed to assure that the Thermal Margin Monitor in the Reactor Protective System remains within its design temperature range.

An additional requirement was added, proposed Item 4, which requires "Verifying that the Fuel Pool Ventilation System is OPERABLE by initiating flow through the HEPA filter and charcoal adsorbers from the control room." This Item is comparable to the verification that Control Room Ventilation system automatically switches to the emergency mode. The Fuel Pool Ventilation System is manually actuated.

These more restrictive requirements are currently being maintained under administrative controls to provide assurance of ventilation system operability. The revision of Table 4.2.3 made in conjunction with revision of the Administrative Controls section provided a convenient opportunity to request an enhancement of the existing requirements.

6. The remaining text of Specifications 4.2 through 4.6 (pages 4-15a through 4-41) was repaginated to eliminate blank pages. The list of modifying amendments, which appears at the bottom of each page, was rewritten to list only those amendments which modified the subject matter appearing on each proposed page.

Existing pages 4-25 and 4-26 contain a notation at the bottom of the page that they were altered by "Change No. 16" and by "Amendment No. 12". These are two identifiers for the same TS revision; "Change No. 16" to the TS was issued by "Amendment No. 12" to the Facility Operating License on February 25 1975. The references to "Change No. 16" have been omitted.

7. Specification 4.3e, a requirement to reevaluate the lnservice Inspection Program (page 4-16), was deleted; it is redundant to the referenced section of 10 CFR 50 (50.55a (g)(5)), and is included in the Inservice Inspection and Testing Program. Specification 4.3f (page 4-16) and Table 4.3.2 (page 4-23) were deleted.

The Specification 4.3f requirement to inspect the regenerative heat exchanger is redundant to the ASME Boiler and Pressure Vessel Code,Section XI, Category B-A. The frequency in the code is once every 10 years; the existing requirement is once every 5 years. There is no unusual feature of the regenerative heat exchanger to require more frequent testing. Past testing at the 5 year interval has disclosed no problems.

The Specification 4.3f primary coolant pump testing requirements were moved to proposed Specification 6.5.6, Primary Coolant Pump Flywheel Surveillance Program, emulating STS Specification 5.5.7. There is no change in primary coolant pump flywheel testing requirements.

8.

Section 4.5.3, Recirculation Heat Removal Systems (page 4-28a), which contains testing requirements for systems outside the containment which could potentially contain highly radioactive fluids, has been 4

combined with former Section 6.15, Systems Integrity, as the proposed Section 6.5.2, Primary Coolant Sources Outside Containment. This change places all requirements for such testing in one location emulating the STS treatment of these testing requirements. This combining of these requirements was suggested by the NRC in the December 1, 1982, letter which issued Amendment 71 to the Palisades Operating License. There is no change in testing requirements.

The basis discussion (page 4-35) for Section 4.5.3 was deleted.

9. Specifications 4.5.4, Surveillance for Prestressing System (page 4-29); 4.5.5, End Anchorage Concrete Surveillance (page 4-32); and 4.5.8, Dome Delamination Surveillance (page 4-32a) were replaced by proposed Surveillance Requirement 4.5.4, which requires verification of containment structural integrity in accordance with the Containment Structural Integrity Surveillance Program, and Specification 6.5.5, the Containment Structural Integrity Surveillance Program. These proposed specifications emulate the STS treatment of containment structural integrity surveillance requirements. The proposed program requires compliance with Regulatory Guide 1.35, as does the equivalent STS program. That guide contains the details of the testing requirements. The parts of Specification 4.5 Basis and References pertaining to containment structural testing were deleted.

10. Specification 4.5.6, Containment Isolation Valves (page 4-32), was renumbered 4.5.3.

11. Surveillance requirement 4.6.2b, verification that the containment spray nozzles are open (page 4-39), was deleted; Item 4.6.2c was renumbered 4.6.2b. The requirement to inspect the containment spray nozzles is redundant to the ASME Boiler and Pressure Vessel Code,Section XI, Subparagraph ICW-5222(d) and Table ICW-2500-1. The frequency in the code is once every 10 years; the existing requirement is once every 5 years. There is no unusual feature of the Palisades spray nozzles to require more frequent testing. Past testing at the 5 year interval has disclosed no problems.

12. The details of'Specification 4.14, Augmented Inservice Inspection Program for Steam Generators (page 4-66), were moved to the Administrative Controls section of TS, emulating STS. There are no changes in requirements. Parts 4.14.2 through 4.14.5 were moved to proposed Specification 6.5.8, Steam Generator Tube Surveillance Program; part 4.14.6 was moved to proposed Specification 6.6.9, Steam Generator Tube Surveillance Report.

Section 4.14 was retitled "Steam Generator Surveillance" and existing part 4.14.1, which currently requires the steam generators to be demonstrated operable by performance of the specified testing, was reworded to reference the proposed program. The wording used for surveillance requirement 4.14.1 was taken from STS. The reference to the Inservice Inspection and Testing Program (former 4.0.5) was retained.

Table 4.14-1 was eliminated, and its footnote reworded as the first paragraph of the Program.

5 The wording of the existing testing requirements was revised to eliminate redundancies and to remove requirements pertinent only to preservice and initial testing. The wording of the reporting requirements was revised to utilize parallel sentence structure.

The proposed testing program and reporting requirements are equivalent, with exception of preservice and initial testing which has been completed, to the requirements of existing Section 4.14.

13. The record retention requirement of Section 4.16.lf (page 4-74) was revised to eliminate reference to the deleted Section 6.10.

14. A requirement for the plant superintendent to approve tests, experiments, and modifications was added to Section 6.1.1 (page 6-1).

The proposed wording was taken from STS. The existing TS contain a requirement for the Plant Review Committee (PRC) to review and recommend plant superintendent approval of these items. That PRC requirement is deleted from the TS by this proposed change.

15. The wording of Section 6.1.2 (page 6-1) is revised to match STS. The proposed requirements are nearly identical to the existing requirements.

16. Section 6.2.2, Plant Staff (page 6-2), was extensively revised to emulate the STS. The table which provides shift staffing requirements does not appear in STS and was deleted. These changes are administrative, and do not involve any change in requirements.

Item 6.2.2a currently requires each shift on duty to include the staff required by Table 6.2-1. That requirement has been replaced with the STS requirement for assignment of non-licensed operators (Auxiliary Operators or AOs). The table contains requirements for Shift Supervisor (SS), Shift Engineer (SE) or Senior Reactor Operator (SRO),

Reactor Operator (RO), Auxiliary Operator (AO), and Shift Technical Advisor (STA) positions. The SS, SE or SRO, and RO requirements duplicate the requirements of 10 CFR 50.54 (m)(2)(i) and need not be repeated in TS. The AO requirement of that table appears as the proposed 6.2.2a. The STA requirements appear as the proposed 6.2.2g.

The wording of Item 6.2.2b is revised to match STS. The proposed requirements are nearly identical to the existing requirements.

A new Item 6.2.2c has been added. Its wording is taken from STS, and is equivalent to the existing footnote in Table 6.2-1 which allows the shift crew to be less than the required crew during unexpected absences. There is no significant change in requirements.

Existing Item 6.2.2c was renumbered 6.2.2d and reworded to match STS.

The proposed Item 6.2.2d effectively combines the existing Item 6.2.2c and the associated footnote which appears at the bottom of the page.

The footnote is deleted.

Existing Item 6.2.2d, which requires core alterations to be performed under supervision of an SRO, is deleted. There is no equivalent requirement in STS. It is redundant to the requirements of 10 CFR 50.54 (m)(2)(iv).

There is no existing requirement numbered 6.2.2e.

6 Existing requirement 6.2.2f is renumbered as 6.2.2e and reworded, slightly, to more closely match STS. There is no reduction, nor any significant change, in requirements.

A new Item, 6.2.2f, has been added requiring either the operations manager or his assistant to hold an SRO license. This requirement is moved to 6.2.2 from existing 6.3.5 to match the wording of STS.

A new Item, 6.2.2g, describing the requirements for a Shift Technical Advisor has been added. It replaces the STA requirements formerly listed in deleted Table 6.2-1. The proposed wording is a combination of that from STS and the footnote from Table 6.2-1.

17. Reference to formerly deleted Item 6.2.3 was removed from page 6-2a.

Its retention served no function.

18. Section 6.3, Plant Staff Qualifications (page 6-3), was revised. The proposed changes limit the requirements of this section to qualifications, allow assignment of the individuals who perform 50.59 reviews to other departments within the plant staff, and move one Item to Section 6.2.2, emulating its placement in STS.

Item 6.3.1 is unchanged.

Item 6.3.2 has been revised to delete the requirements that the radiation safety manager be designated by the Plant General Manager, and that the radiation safety manager shall have direct access to the plant manager. Neither of these requirements is germane to a section on qualifications. In addition, the associated footnote has been incorporated into the main paragraph.

Item :6.3.3 is unchanged.

Item 6.3.4 was reworded to allow the assignment of the required 50.59 reviews to other portions of the plant staff. The qualification requirement for these individuals is unchanged. The reference to Item 6.5.3 was reworded because this change request proposes deleting that section from TS.

Item 6.3.5 was deleted. The first requirement, for an operations manager to hold an SRO ltcense, appears as proposed 6.2.2f, as it does in STS. The second requirement, for meeting ANSI NIB-I, is redundant to 6.3.1. The third requirement is also moved to 6.2.2f.

19. As discussed under change 16, above, Table 6.2-1 (page 6-4) has been deleted. Those requirements of Table 6.2-1 which do not appear in 10 CFR 50.54 have been added to the text of Section 6.2.2. This change emulates STS.

20. Section 6.4 (page 6-5), formerly deleted, was retitled "Procedures" emulating Section 5.4 of STS. The procedure requirements have been moved to Section 6.4 from Section 6.8.1.

21. The requirements of Section 6.5, Review and Audit, (pages 6-5 through 6-10) were deleted from the TS entirely. These requirements are being relocated to the Quality:Program Description, CPA-2A. There are no compar~ble review and audit requirements in STS.

7 Section 6.5 was retitled "Programs and Manuals", emulating Section 5.5 of STS and replacing existing Section 6.8.4. Programs and Manuals currently required by Section 6, and a newly proposed TS basis control program are gathered in this section.

22. -As mentioned in change 2, above, Item 6.7.la, the Safety Limit violation shutdown requirement, was moved to Sections 2.1.1 and 2.2.1 which formerly referenced Section 6.7. Item 6.7.1.b, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification, and Items 6.7.lc and 6.7.ld, written reporting, (page 6-

10) were deleted. Item 6.7.1.b is redundant to 10 CFR 50.72(b){l){i)(A); Items 6.7.1.c and 6.7.1.d are redundant to 10 CFR 50.36{c)(l)(i)(A) and to 10 CFR 50.73(a)(2)(i)(A), (i)(B), and (ii)(B).

This change would extend the required time for submitting a Safety Limit violation report from 14 to 30 days. Since plant operation may not resume until authorized by the Commission, a slight extension of reporting time would not have any effect on safety.

23. Section 6.8.1 (page 6-11) was retitled "Procedures" and renumbered 6.4, emulating STS Section 5.4. Item 6.8.la was reworded slightly to emulate STS wording. Items 6.8.ld and 6.8.le, which require having written procedures for the security and emergency plans, were deleted in accordance with the recommendations of Generic Letter 93-07. There are no comparable requirements in STS. Parts 50 and 73 of Title 10 of the Code of Federal Regulations include provisions that are sufficient to address these requirements.

24. Sections 6.8.2 and 6.8.3 (page 6-11), which describe requirements for the procedure change process, have been relocated to the Quality Program Description, CPC-2A. This level of detail does not appear in STS.

25. Item 6.8.4a, Radioactive Effluent Controls Program (page 6-12), was renumbered 6.5.4, emulating the STS Item 5.5.4 of the same title.

Editorial revisions were made to make the wording closer to that in STS.

26. Item 6.8.4b, Radiological Environmental Monitoring Program (page 6-13), was deleted. No comparable item exists in STS. The program is contained within the ODCM, as required by proposed Specification 6.5.l. Changes to the ODCM are required to be submitted to the NRC as part of, or concurrent with, the Radioactive Effluent Release Report as required by proposed Specifications 6.5.1 and 6.6.3.

27. Item 6.9, Reporting Requirements (page 6-14) was renumbered 6.6, emulating STS Section 5.6 of the same title. Reporting requirements in existing TS Section 6 were collected in this section.

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28. Item 6.9.la, Startup Report (page 6-14}, wa~ deleted. No comparable report is required by STS. The existing TS require a summary of plant start-up and power escalation testing shall be submitted within 90 days following completion of the start-up test program following "amendment to the license involving a planned increase in power level, installation of fuel that has a different design or has been manufactured by a different fuel supplier and, modifications that may have significantly altered the nuclear~ thermal or hydraulic performance of the plant".

This reporting requirement has been judged as unnecessary for inclusion in the STS. It simply sunvnarizes the complete records which are part of the permanent plant records and are thereby available for NRC review. The 90 days of operation allowed before the report submittal, and the lack of any required approval, imply that the report is not intended to be used for NRC safety decisions.

29. Item 6.9.lb, Annual Report (page 6-14), was renumbered 6.6.1 and retitled "Occupational Radiation Exposure Report" emulating STS Item 5.6.1. The option to base radiation exposure reporting on electronic dosimeters was added. The wording was revised, editorially, to mQre closely match STS~

30 .. Item 6.9.lc, Monthly Operating Report (page 6-15}, was renumbered 6.6.4 emulating STS Item 5.6.4.

31. Item 6.9.ld, Radioactive Effluent Release Report (page 6-15}, was renumbered 6.6.3 emulating STS Item 5.6.3. Reference to the deleted Process Control Program was removed.

32. Item 6.9.le, Radiological Environmental Operating Report (page 6-15),

was renumbered 6.6.2 emulating STS Item 5.6.2.

33. Item 6.9.lf, Core Operating Limits Report (page 6-15), was renumbered 6.6.5 emulating STS Item 5.6.5.

34. Item 6.9.2, Reportable Events" (page 6-17), was deleted. There is no comparable requirement in STS. It is redundant to 10 CFR 50.73. The existing requirement is:

- "The Commission shall be notified of Reportable Events and a report submitted pursuant to the requirements of 10 CFR 50.73. 11

35. Item 6.9.3, Nonroutine Reports (page 6-17), was deleted. There is no comparable requirement in STS. The existing requirement is:

"A report shall be submitted in the event that (a) the Radiological Environmental Monitoring Programs are not substantially conducted as described in the ODCM or (b} an unusual or important event occurs from plant operation that causes a significant environmental impact or affects a potential environmental impact. Reports shall be submitted within 30 days."

Part (a) of this requirement is not redundant to any other requirement, but plant administrative procedures require initiation of a Conditi-0n Report for such an event. Condition Reports are available fQr NRC aud-it and are often reviewed by the NRC resident inspector.

9 Part (b) of the existing requirement is redundant to Item 4.1 of the Palisades Environmental Protection Plan, which is Appendix B to the Facility Operating License. That requirement is:

"Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, telegraph, or facsimile transmissions fo 11 owed by a written report per Subsection 5. 4. 2."

In addition, an event that which could or did result in a significant environmental impact would probably involve a notification of other governmental agencies and thereby are subject to the reporting requirements of 10 CFR 50.72(b)(2)(iv)(A&B).

36. Item 6.9.4a, under Special Reports (page 6-26), was reworded and renumbered 6.6.8 and retitled "Containment Structural Integrity Surveillance Report" emulating STS Item 5.6.9.

37. Item 6.9.4.b (page 6-26) was deleted. There is no comparable requirement in STS. It is redundant to 10 CFR 50.4. The existing requirement is:

"Special reports shall be submitted i~ accordance with 10 CFR 50.4, within the time period specified for each report."

38. The requirements of Section 6.10, Record Retention (page 6-26 through 6-28), were deleted from the TS entirely. These requirements are relocated to the Quality Program Description, CPC-2A. There are no comparable Record Retention requirements in STS.

39. Item 6.11, Radiation Protection Program (page 6-28), has been deleted.

It is redundant to proposed Item 6.4a (existing 6.8.la), in that. it contains requirements for procedures for personnel radiation protection. Procedures for personnel radiation protection are listed in Regulatory Guide 1.33, and maintaining the procedures recommended by RG 1.33 is required by proposed Item 6.4a (existing 6.8.la). The existing requirement is:

"Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20, and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure." *

40. Section 6.12, High Radiation Area (page 6-28), was renumbered 6.7, emulating STS Section 5.7. Items 6.12.1 (page 6-28) and 6.12.2 were renumbered as 6.7.1 and 6.7.2 and clarified with respect to the measurement of radiation dose rate by addition of the 10 CFR 20 phrase "at 30 cm from the radiation source or from any surface which the radiation penetrates."

41. Section 6.15, Systems Integrity (page 6-33), was combined with Section 4.5.3 (see change 8, above) and retitled "Primary Coolant Sources Outside Containment" and renumbered 6.5.2, emulating STS Section 5.5.2.

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42. Sections 6.16, Iodine Monitoring (page 6-33), and 6.17, Post Accident Sampling (page 6~34), were combined, retitled "Post Accident Sampling Program", and renumbered 6.5.3, emulating STS Section 5.5.3.

43. Section 6.18, Offsite Dose Calculation Manual (page 6-35), was combined with its definition from existing Section 1.1 and renumbered 6.5.1, emulating STS Section 5.5.1 of the same title. The wording was revfsed, editorially, to more closely match STS.

44. Section 6.19, Process Control Program (page 6-35), is being relocated to plant procedures. Section 6.19 deals solely with changes to the Process Control Program. No corresponding section appears in STS.

45. Section 6.21, Sealed Source Contamination Program (page 6-37), is being relocated to the ODCM. Section 6.21 was added to the Palisades Technical Specifications by Amendment 98 on October 19, 1986, to emulate Section 3/4.7.10 of the former CE STS, NUREG 0212. The subject requirements were not retained in the current STS.

46. Section 6.22, Secondary Water Chemistry (page 6-38), was retitled "Secondary Water Chemistry Program" and renumbered 6.5.9, emulating STS Section 5.5-10 of that title.

47. A new Section, 6.5.12 "Technical Specification Bases Control Program" was added. That~proposed Section is copied from STS Section 5.5.14.

48. The entire Administrative Controls section wa~ re-paginated to eliminate voids and unused pages. The proposed numbering scheme makes provision for features of the STS which are not yet included in the Palisades TS, a Fuel Oil Testing Program, a safety Functions Determinations Program, and a Pressure - Temperature Limits Report.

The list of modifying amendments, which appears at the bottom of each page, was rewritten to list those amendments which modified the subject matter appearing on each proposed page.

11 I I. Rea,sons for Changes:

These changes are all administrative in nature. They are intended to provide the following benefits:

1. Accomplish various administrative changes to eliminate redundancy and clarify the TS requirements by revising and renumbering the Administrative Controls section to closely emulate the Standard Technical Specifications ..

2. Relocation of Review and Audit requirements to the Quality Program Description to allow these duties to be reassigned to qualified individuals in various departments within the plant staff.

3. Update the TS requirements to reflect the revision to 10 CFR 20.

III. Analysis of No Significant Hazards Consideration Consumers Power Crimpany finds that this proposed Technital Specifications change involves no significant hazards and, accordingly, that a no significant hazards determination per 10CFR50.92(c) is justified.

With the exception of changes number 7 and 11, the proposed changes can all be classified as either "Administrative Changes" which are editorial in nature, only i nvo lv.i ng movement of requirements within the TS with out affecting their technical content, or clarifying existing TS requirements; or as "Relocation Changes" which only move requirements, not meeting the 10 CFR 50.36(c)(2)(ii) criteria, from the TS to documents controlled under 10 CFR 50.54 or 50.59, or which delete requirements which are redundant to NRC regulations or other TS requirements.

"Administrative Changes" and "Relocation thanges" move requirements, either within the TS or to documents controlled under 10 CFR 50.54 or 50.59, or clarify existing TS requirements without affecting their content. Since "Administrative" and "Relocated" changes do not alter the technical content of any requirements, they cannot involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any previously evaluated, or involve a significant reductio'n in a margin of safety.

Change number 7 deletes the explicit requirement in surveillance 4.3.f to inspect the Regenerative Heat Exchanger every five years as redundant to an ASME code requirement for the same inspection every 10 years. Similarly, change number 11 deletes the explicit requirement in surveillance 4.6.4.b to inspect the Containment Spray Nozzles every five years as redundant to an*ASME code requirement for the same inspection every 10 years. Neither the Regenerative Heat Exchang~r nor the Containment Spray Nozzles have any unusual features which would require more frequent testing than required by the ASME code. Testing at the five year interval, throughout the plant operating hi story, has uncove,red no problems with these I terns. The extension of a surveillance interval does no~alter plant equipment or operating conditions, change instrument settings, nor change operating procedures or methods. Therefore, operation of the facility i~ accordance with changes number 7 and number 11 would not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any ~reviously evaluated, or involve a significant reduction in a margin of safety.

ATTACHMENT I CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255

TECHNICAL SPECIFICATION CHANGE REQUEST REVISION OF ADMINiSTRATIVE CONTROLS Proposed Pages

.- 55 Pages

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS

- SECTION 1.0 DEFINITIONS DESCRIPTION PAGE NO 1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1 Table 2.3.l Reactor Protective System Trip Setting Limits 2-2 B2.l Basis - Reactor Core Safety Limit B 2-1 B2.2 Basis - Primary Coolant System Safety Limit B 2-2 B2.3 Basis - Limiting Safety System Settings B 2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 APPLICABILITY 3-1 3.1 PRlMARY COOLANT SYSTEM 3-lb 3 .1.1 Operabl~ Components 3-lb 3.1.2 Heatup and Cooldown Rates 3-4 Figure 3,,.1 Pressure - Temperature Limits for Heatup 3-5 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-6 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 3 .1.6 Maximum PC.S Oxygen* and Halogen Concentration 3-23 3.1. 7 Primary and Secondary Safety Valves 3-24a 3 .1.8 Over Pressure Protection Systems 3-25a Figure 3-4* LTOP Limit Curve 3-25c 3 .1. 9 Shutdown Cooling 3-25h 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT. COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.1 Conta.i.nment Penetrations and Valves 3-40b 3.7 ELECTRICAL SYSTEMS 3-41 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49

i Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS

- SECTION

3.0 DESCRIPTION

LIMITING CONDITIONS FOR OPERATION (continued)

PAGE NO 3-1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 3-50 3.10.l Sh~tdown Margin Requirements 3-50 3.10.2 Deleted 3-51 3.10.3 Part-Length Control Rods 3-51 3.10.4 Misaligned or Inoperable Rod 3-52 3.10.5 Regulating Group Insertion Limits 3-52 3.10.6 Shutdown Rod Limits 3-53 3.10.7 Low Power Physics Testing 3-53 3.11 POWER DISTRIBUTION INSTRUMENTATION 3-56 3.11.1 Incore Detectors' 3-56 3.11.2 Excore Power Distribution Monitoring System 3-57 Figure 3.11-1 Axial Variation ~ounding Condition 3-59 3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-60 3.13 Deleted 3-60 3.14 CONTROL ROOM VENTILATION 3-61 3.15 REACTOR PRIMARY SHIELD COOLING SYSTEM 3-62 3.16 ESF SYSTEM INITIATION INSTRUMENTATION SETTINGS 3-63 Table 3.16.1 ESF System Initiation Instrument Setting Limits 3-63 B3 .16 Basis - ESF System Instrumentation Settings B 3.16-1

3.17 3.17.1 Table 3.17.2 Table 3.17.3 INSTRUMENTATION AND CONTROL SYSTEMS Reactor Protective System Instruments 3.17.1 Instrument Requirements for RPS Engineered Safety Features Instruments 3.17.2 Instrument Requirements for ESF Systems Isolation Functions Instruments 3-64 3-64 3-65 3-66 3-67 3-68 Table 3.17.3 Instrument Requirements Isolation Functions 3-69 3.17.4 Accident Monitoring Jnstruments 3-70 Table 3.17.4 Instrument Requirements for Accident Monitoring 3-71 3.17.5 Alternate Shutdown System Instruments 3-72 Table 3.17.5 Instruments for the Alternate Shutdown System 3-73 3.17.6 Other Safety Feature Instruments 3-74 Table 3.17.6 Instruments for Other Safety Features 3-77 B3.17 Basis - Instrumentation Systems B 3.17-1 3.18 Deleted 3-79 3.19 IODINE REMOVAL SYSTEM 3-79 3.20 SHOCK SUPPRESSORS (Snubbers) 3-80 3.21 CRANE OPERATIONS AND MOVEMENT HEAVY LOADS 3-81 3.22 Deleted 3-84 3.23 POWER DISTRIBUTION LIMITS 3-84 3.23.1 Linear Heat Rate 3-84 3.23.2 Radial Peaking Factors 3-86 3.23.3 Quadrant Power Tilt - Tq 3-87 I Table 3.23-3 Power Distribution Measurement Uncertainty 3-88

ii Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS

SECTION 4.0 4 .1

4.2 DESCRIPTION

SURVEILLANCE REQUIREMENTS

OVER PRESSURE PROTECTION SYSTEM TESTS EQUIPMENT AND SAMPLING TESTS Table 4.2.1 Minimum Frequencies for Sampling Tests PAGE NO 4-1 4-6 4-7 4-9 Table 4.2.2 Minimum Frequencies for Equipment Tests 4-11 Table 4.2.3 Ventilation Systems Tests 4-14 4.3 SYSTEMS SURVEILLANCE 4-16 Table 4.3.1 Primary Coolant System Pressure Isolation Valves 4-18 4.4 Deleted 4-19 4.5 CONTAINMENT TESTS 4-19 4.5.1 Integrated Leakage Rate Tests 4-19 4.5.2 . Local Leak Detection Tests 4-19 4.5.3 Containment Isolation Valves 4-21 4.5.4 Surveillance for Prestressing System 4-21 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 4-24 4.6.1 Safety Injection System

4-24 4.6.2 Containment Spray System 4-24 4.6.3 Pumps 4-24 4.6.4 Valves 4-24 4.6.5 Containment Air Cooling System 4-25 4.7 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-42 4.7.1 Dies~l Generators 4-42 4.7.2 Station Batteries 4-42 4.7.3 Emergency Lighting 4-43 4.8 MAIN STEAM STOP VALVES 4-44 4.9 AUXILIARY FEEDWATER SYSTEM 4-45 4.10 REACTIVITY ANOMALIES 4-46 4.11 Deleted 4-46 4.12 AUGMENTED ISi PROGRAM FOR HIGH ENERGY LINES 4-60 4.13 Deleted 4-65 4.14 AUGMENTED ISi PROGRAM FOR STEAM GENERATORS 4-66 4.15 PRIMARY SYSTEM FLOW MEASUREMENT 4-70 4 .. 16 ISi PROGRAM FOR SHOCK SUPPRESSORS (Snubbers) 4-71 4.17 INSTRUMENTATION SYSTEMS TESTS 4-75 Table 4.17~1 Surveillance for the RPS 4-76 Table 4-17.2 Surveillance for ESF Functions 4-77 Table 4-17.3 Surveillance for Isolation Functions 4-78 Table 4-17.4 Surveillance for ~ccident Monitoring 4-79 Table 4-17.5 Surveillance for Alternate Shutdown 4-80 Table 4-17.6 Surveillance for Other Safety Functions 4-81 84.17 Basis - Instrumentation Systems Surveillance B 4.17-1 4.18 POWER DISTRIBUTION INSTRUMENTATION 4-83 4.18.1 Incore Detectors 4-83 4.18.2 Excore Monitoring System 4-83 4.19 POWER DISTRIBUTION LIMITS 4-84 4.19.1 Linear Heat Rate 4-84 4.19.2 Radial Peaking Factors 4-84 4.20 MODERATOR TEMPERATURE COEFFICIENT (MTC) 4-85 iii Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS

SECTION 5.0 5.1

5. 2 .

DESIGN FEATURES 5.

2.1 DESCRIPTION

SITE CONTAINMENT DESIGN FEATURES

Containment Structures PAGE NO 5-1 5-1 5-1 5-1 5.2.2 Penetrations 5-2 5.2.3 Containment Structure Cooling Systems 5-2 .

5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 5-2 5 .3 .1 Primary Coolant System 5-2 5.3.2 Reactor Core and Control 5-3 5.3.3 Emergency Core Cooling System 5-3 5.4 FUEL STORAGE 5-4 5.4.1 New Fuel Storage 5-4 5.4.2 Spent Fuel Sta.rage 5-4a Figure 5-1 Site Environment TLD Stations 5-5 6.0 ADMINISTRATIVE CONTROLS . 6-1 6.1 . RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 Onsite and Offsite Organizati-0ns 6-1 6.2.2 Plant Staff 6-2 6.3 PLANT STAFF QUALIFICATIONS 6-3 6.4 PROCEDURES 6-4 6.5 PROGRAMS AND MANUALS 6-5 6.5.1 *offsite Dose Calculation Manual 6-5 6.5.2 Primary Coolant Sources Outside Containment 6-6 6.5.3 Post Accident Sampling *Program 6-6 6.5.4 Radioactive Effluent Controls Program 6.,-7 6.5.5

Containment Structural Integrity Surv. Program 6-8 6.5.6 Primary Coolant Pump Flywheel Surv. Program 6-8 6.5.7 Inservice Inspection and Testing Program . 6-8 6.5.8 Steam Generator Tube Surveillance-Program 6-9 6.5.9 Secondary Water Chemistry Program 6-14 6.5.10 Ventilation Filter Testing Program 6-15 6.5.11 Reserved 6-16 6.5.12 Technical Specification Bases Control Program 6-16 6.5.13 Reserved 6-17 6.5.14 Reserved 6-17 6.6 REPORTING REQUIREMENTS 6-18 6.6.1 Occupational Radiation Exposure Report 6-18 6.6.2 Radiological Environmental Operating Report* 6-18 6.6.3 Radioactive Effluent Release Report 6-18 6.6.4 Monthly Operating Report 6-18 6.6.5 Core Operating Limits Report 6-19 6.6.6 Reserved 6-20 6.6.7 Accident Monitoring Instrument Report 6-20 6.6.8 Containment Structural Integrity Surveillance Report 6-21 6.6.9 Steam Generator Tube Surveillance Report 6-21 6.7 HIGH RADIATION AREA 6-22 iv Amendment No.

TECHNICAL SPECIFICATIONS

1.0 DEFINITIONS The fo 11 owi n*g terms are de~i ned for uni form interpretation of these

Technical Specifications.

ASSEMBLY RADIAL PEAKING FACTOR - FrA ASSEMBLY RADIAL PEAKING. FACTOR shall be the maximum ratio of the power 1

generated in any fuel assembly,. t.o the average fuel assembly power.

(Each of these power terms shall be integrated over core height and shall include tilt.)

AVERAGE D'ISINTEGRATION ENERGY - E AVERAGE DISINTEGRATION ENER~Y shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor

.. coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MEV) for isotopes., other than iodines, with half lives greater than 15 minutes, making up at least 953 of the total noniodine activity in the coolant.

AXIAL OFFSET or AXIAL SHAPE INDEX - AO or ASI AXIAL OFFSET or AXIAL SHAPE INDEX shall be the ratio of the power generated in the lower half of the core minus the power generated in

the upper half of the core, to the sum of those powers .

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor, alarm, interlock, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. Neutron detectors may be excluded from CHANNEL CALIBRATIONS.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation .. This determination shall include, where possible, comparison of *the channel indication and status with other indications and status derived from independent instrument channels measuring the same param~ter. A CHANN~L CHECK shall include verificati'o_n that the monitored parameter is within limits imposed by the Technical Specifications. * *

1-1 Amendment No. 3-1-, 43-, .§4, .§+, 68, -H-8, -l-24, 28, -l-a-7-, 1-62-,

1.0 DEFINITIONS .(continued}

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel to verify that it is OPERABLE, including any alarm and trip initiating function.

COLO SHUTDOWN The COLD SHUTDOWN condition shall be* when the primary coolant is at SHUTDOWN BORON CONCENTRATION and Tave is less than 210°F.

CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY is defined to exist when all the following are true:

a. All nonautomatic containment isolation valves and blind flanges are closed (OPERABLE} except as noted in Table 3.6.1.

b. The equipment hatch is properly closed and sealed.

c. At least one door in each personnel air lock is properly closed and sealed.

d. All automatic containment isolation valves are OPERABLE

e.

(as demonstrated by satisfying isolation times specified in Table 3.6.1 and leakage criterion. in Specification 4.5.2) or are locked closed.

The uncontrolled containment leakage satisfies Specification 4.~.

CONTROL ROOS CONTROL ROOS shall be all full-length shutdown and regulating rods.

CORE OPERATING LIMITS REPORT (COLR)

The COLR is the document that provides cycle speclfiG parameter limits for the current reload cycle. These cycle specific p~rameter limits shall be determined for each reload cycle in acco.rdance with Specification*6.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (:µCi/gm}

which alone would produce the same thyroid dose as the quantity and isotopic.mixture nf 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose converston facto~s used for this calculation shall be those listed in Table III of TI0-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

1-2 Amendme.nt No. a-I-, 4a, .§4, .§+, 68, HS, -l-24, HS, H+, ~'

1.0 DEFINITIONS (continued)

HOT SHUTDOWN The HOT SHUTDOWN condition shall be when the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specification 3.10 and Tave is greater than 525°F.

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

LOW POWER PHYSICS TESTING LOW POWER PHYSICS TESTING shall be testing performed under approved written procedures to determine CONTROL ROD worths and other core nuclear properties. Reactor power during these tests shall not exceed 2% of RATED POWER, not including decay heat and PCS Tave and PCS pressure shall be in the range of 371°F to 538°F and 415 psia to 2150 psia, respectively. Certain deviations from normal operating practice which are necessary to enable performing some of these tests are permitted in accordance with the specific provisions in these Technical Specifications.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE, or have OPERABILITY, when it is capable of performing its specified functions, and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication, or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified fQnctions are also capable of performing their related support functions.

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

QUADRANT POWER TILT - Tq QUADRANT POWER TILT shall be the algebraic ratio of quadrant power minus average quadrant power, to average quadrant power.

RATED POWER"H RATED POWER shall be a steady state reactor core output of 2530 MWt .

1-3 Amendment No. 3-1-, 43-, .§4, &7, W, H-8, -l-24, HS, -l-3-7, ~'

1.0 DEFINITIONS (continued)

REACTOR CRITICAL The reactor is considered critical for purposes of administrative control when the neutron flux wide range channel instrumentation indicates greater than 10-43 of RATED POWER.

REFUELING BORON CONCENTRATION.

REFUELING BORON CONCENTRATION shall be a Primary Coolant System boron concentration of at least 1720 ppm AND sufficient to assure the reactor

is subcritical by ~ 5% Ap with all CONTROL RODS withdrawn.

REFUELING OPERATION A REFUELING OPERATION shall be any operation involving movement of core components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel.

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

SHUTDOWN BORON CONCENTRATION SHUTDOWN BORON CONCENTRATION shall be* a Primary Cool ant System boron concentration sufficient to assure the reactor is subcr;tical by ~ 2% Ap with all CONTROL RODS in the core and the highest worth CONTROL ROD fully withdrawn.

SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its .present condition assuming that all CONTROL RODS are fully inserted except for the single highest worth CONTROL ROD which is assumed to be withdrawn.

TOTAL RADIAL PEAKING FACTOR - FrT The TOTAL RADIAL PEAKING FACTOR shall be the maximum product of the ratio of individual assembly power to core average assembly power, times the highest local peaking factor integrated over the total core height, including tilt. Local peaking factor is defined as the maximum ratio of an individual fuel rod power to the assembly average rod power .

1-4 Amendment No. a.I-, 46-, .§4, .§-7, 68, HS, ~' ~' -l-3-7, ~' ~'

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 Safety Limit - Reactor Core The Minimum DNBR of the reactor core shall be maintained greater than or equal to the DNB correlation,safety limit.

Correlation

Safety Limit XNB 1.17 ANFP 1.154 HTP 1.141 Applicability Safety Limit 2.1 is applicable in HOT STANDBY and POWER OPERATION.

Action 2.1.1 If a Safety Limit *is exceeded, the reactor shall be shut down immediately and not restarted until the Commission authorizes resumption of operation in accordance with 10 CFR 50.36(c)(l)(i)(A).

2.2

Safety Limit - Primary Coolant System Pressure (PCS)

The PCS Pressure shall not exceed 2750 psia.

Applicability

2.2.1 Safety Limit 2.2 is applicable when there is fuel in the reactor.

Action If a Safety limit is exceeded, the reactor shall be shut down immediately and not restarted until the Commission authorizes resumption of operation in accordance with 10 C~R 50.36(c)(l)(i}(A).

2.3 Limiting Safety System Settings"- Reactor Protective System (RPS)

The RPS trip setting limits shall be as stated in Table 2.3.1.

Applicability Limiting Safety System Settings of Table 2.3.1 are applicable when the associated RPS channels are required to be OPERABLE by Specification 3.17.1.

Action 2.3.1 If an RPS instrument setting is not within the allowable settings of Table 2.3.1, immediately declare the instrument inoperable and complete corrective action as directed by Specification 3.17.1 .

Amendment No. 3-l, 2-1

~' ~' !4-8, 3-7, ~' .i-GS,

3.17 INSTRUMENTATION SYSTEMS Specification

3.17.4 The Accident Monitoring Instruments listed in Table 3.17.4 shall be OPERABLE. (Specifications 3.0.3, 3.0.4, and 4.0.4 do not apply.)

Applicability Specification 3.17.4 applies when the PCS temperature is > 300°F.

Action 3.17.4.1 With one r~quired channel of functions 1 through 14 inQperable for one or more functions:

a. Restore channel to OPERABLE status within 7 days.

3.17.4.2 With two required channels of functions 1 through 14 inoperable for one or more functions: *

a. Restore one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.3 With position indication inoperable for one or more Containment Isolation Valves:

a. Restore the indication to OPERABLE status or lock the associated valves in the closed position within 7 days.

3.17.4.4 If any action required by 3.17.4.l through 3.17.4.3 is not met AND the associated completion time has expired, *

a.

b.

The reactor shall be plac~d in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and The reactor shall be placed in a condition where the affected equipment is not required, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.5 With one channel of functions 16 through 21 inoperable for one or more functions:

a. Restore the channel to OPERABLE status within 7 days.

3.17.4.6 With two required channels of functions 16 through 21 inoperable for one or more functions:

a. Restore one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.7 If any action required by 3.17.4.5 or 3.17.4.6 is not met AND the associated completion time has expired:

a .. With two CETs in any one quadrant inoperable, complete Action 3.17.4.4 in lieu of Action 3.17.4.7 c),

b. *With two RVWL channels inoperable, initiate alternate monitoring within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

c. Submit a report to the NRC in accordance with Specification 6.6.7.

d. Restore the channels to OPERABLE status prior to startup from the next refueling .

Amendment No. ~' .J-47., ~'

4.0 SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance requirements shall be applicable during the reactor operating conditions associated with individual Limiting Conditions for Operation unless otherwise stated in an individual surveillance requirement.

4.0.2 Unless otherwise specified, each surveillance requirement shall be performed within the specified time interval w*ith:

a. A maximum allowable extension not to exceed 25% of the surveillance interval, and

b. A total maximum combined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute

noncomplian~e with the operability requirements for a Limiting Condition for Operation. The time limits of the action requirements are appl1cable at the time it i~ identified that a Surveillance Requirement has not been performed. The action requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of.the surveillance when the allowable outage time ltmits of the action requirements are less.than 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s~ Surveillance Requirements do not have to be performed on inoperable equipment .

4.0.4 Entry into a reactor operating condition or other specified condition shall not be made unless the Surveillance Requirements associated with a Limiting Condition of Operation has been performed ~ithin .tha stated surveillance interval or as otherwise specified. This provision shall not*prevent passage through or to plant condit~ons as required to comply with action requirements .

4-1 Amendment No. 3-G, .§.l, -l-3-G, ~,

4.0 SURVEILLANCE REQUIREMENT (Continued) 4.0.5 Deleted

4-2 Amendment No. -l-3G, ~,

4.0 BASIS Specifications 4.0.1 through 4.0.4 establish the general reqµirements

applicable to Surveillance Requirements. These requirements are based on the Surveillance requirements stated in the code of Federal Regulattons, 10.CFR 50.36{c){3):

PSurveillance requirements are-requirements relating to test; calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation.will be within safety limits, and that the limiting conditions of operation will be met." .

Specification 4.0.l establishes the requirement that surveillances must be performed during reactor operating .conditions or other conditions for which the requirements of the Limiting Conditions for Operation apply, unless otherwise stated in an individual Surveillance Requirement. The purpose of this speci fi cat.ion is to ensure that survei 11 ances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a reactor operating condition or other specified condition for which the associated Limiting Conditions for Operation are applicable.

Surveillance Requirements do not have to be performed when the facility is in an operational condition for which the requirements of the associated Limiting Condition for Operation do not apply, unless otherwise specified.

The Surveillance Requirements .associated with a Special Test Exception are only applicable ~hen the Special Test Exception is used as an allowable exceptton the requirements of ~ specification.

Specification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requirements may be extended. Item a. permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or'other ongoing surveillance or maintenance activities. Item b.

limits the use of the provisions of item a. to ensure that it is not used repeatedly to extend the surveilla~ce interval beyond that specified. The limits of Spe~ification 4j0.2 are based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed ts the verific~tion of conformance with the Surveillance Requirements. These provisions are sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

Specification 4.0.l establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defi.ned by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the operability requirements for a Limiting Condition for Operation.

Under the provisions of this specification, systems and components are assumed to be operable when Surveillance Requirements have .

4-3 Amendment No. -l-3G, ~'

4.0 BASIS (Continued)

Specification 4.0.4 establishes the requirement that all applicable

surveillances must be met before entry into a reactor operating condition or other condition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component operability requirements or parameter limits are met before entry into an operational condition for which these systems and components ensure safe operation of the facility. This provision applies to changes in reactor operating conaitions or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

When a shutdown i£ required to comply with action requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower operational condition .

4-5 Amendment No. -l-3S, +/-62-,

4.1 OVERPRESSURE PROTECTION SYSTEM TESTS Surveillance Requirements In addition to-the requirements of The Inservice Inspection and Testing Program, Specification 6.5.7, each PORV flow path shall be demonstrated OPERABLE by:

I. Testing the PORVs in accordance with the inservice inspection requirements fcir ASME Boiler and Pressure Vessel Code,Section XI, Section IWV, Category B valves.

2. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 mbnths*.

3. When the PORV flow eath is required to be OPERABLE by Specification 3.1.8.1:

(a. Performing a complete cycle of-the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

(b. Performing a complete cycle of the block valve prior to heatup from COLD SHUTDOWN, if not cycled within 92 days.

4. When the PORV fl ow path is required to be OPERABLE by Speci fi ca ti on 3 .1.8.,2:

(a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excl~ding valve operation, at least once per 31 days.

(b. Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5. Both High Pressure Safety Injection pumps shall be verified incapable of injection.into the PCS at least o~ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , unless the reactor head is removed, when either PCS cold leg temperature is< 300°F, or when both shutdown cooling suction valves, M0-3015 and M0-3016, are open.

Basis With t~e reactor vessel head installed when the PCS cold leg temperature is less than 300°F, or if the shutdown cooling system isolation valves M0-3015 and M0-3016 are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be ~xceeded; therefore, both pumps are rendered inoperable.

For Cycle 11 only, this surveillance need not be performed until prior to startup for Cycle 12 *

Amendment No. -l-3G, -149, GG, 6-2-, -163-, le4, 4-6

4.2 EQUIPMENT SAMPLING AND TESTS

Table 4.2.3 VENTILATION SYSTEM TESTS The Control Room Ventilation and Isolation System*and the *Fuel.Storage Area HEPA/Charcoal Exhaust System shall be demonstrated to be OPERABLE by the fa 11 owing tests:

1. Performing required Control Room Ventilation and Fuel Storage Area filter testing in accordance with the Ventilation Filter Testing Program.

2. At least once per refueling cycle by:

a. Verifying that on a containment high-pressure and high-radiation test signal, the Control Room Ventilation system automatically switches into the emergency mode of operation with flow through the HEPA filter and charcoal adsorber bank.

b. Verifying that the Control Room Ventilation system maintains the Control Room at a positive pressure ~ 1/8 inch WG relative to the outside atmosphere during system emergency mode operation. *

.3. Verifying that the Control Room temperature is ~ *90° F; once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . *

4. V~rifying *that the Fuel Pool Ventilation System is OPERABLE by initiating flow through the HEPA filter and charcoal adsorbers from the control room .

4-14 Amendment N.o. Bl, -l-6-2-,

Basis - Table 4.2.2 Item 12 - Trisodium Phosphate (TSP) Tests Item 12.a - TSP quantity verification Verification of the quantity of TSP in the baskets ensures that neither leakage nor other water sources in the containment reduce the basket content b~low the required minimum. This requirement ensures that there is an adequate quantity of TSP to adjust the pH of the post LOCA sump s.ol uti on to a value between 7.0 and 8.0.

Item 12.b - TSP quality verifi~ation Periodic testing is performed to ensure the solubi.lity and buffering ability of the TSP after exposure to the containment environment. Satisfactory completion of this test assures that the TSP in the baskets is "active" as required by Specification 3.19.

Adequate solubility is verified by submerging a representative sample of TSP from one of the baskets in containment in un-agitated borated water heated to a temperature representing post-LOCA conditions; the TSP must completely dissolve within a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. The test time of 4 .hours is specified to allow time for the dissolved TSP to naturally diffuse through the un-agitated test solution. Agitation of the test solution during the solubility verification is prohibited, since an adequate standard for the agitation intensity {other than no agitation} cannot be specified. The flow and turbulence in the containment sump during recirculation. woul~ signjfi~antly decrease the ttme required for the TSP to dissolve.

Adequate buffering capability is verified by a measured pH of the sample solution, following the solubility verification, between 7 and 8. The sample is cooled and thoroughly mixed prior to measuring pH.

The quantity of the TSP sample, and quantity and boron concentration of the w~ter are chos~n to be representative of post-LQCA conditions.

4-15 Amendment No. ~'

4.3 SYSTEMS SURVEILLANCE APPLICABILITY

Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.

OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.

SPECIFICATIONS a,b,c,d,e,f - Deleted

g. A surveillance program to monitor radiation induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.5.3 of the FSAR. *

h. Periodic leakage testing 1al.lbl on each check valve listed in Table 4.3.1 shall be accomplished prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if such testing has not been accomplished within the previous 9 months, and prior to returning the check

i.

valves to service after maintenance, repair or replacement work is performed on the valves.

Whenever integrity of a pressure isolation valve listed in Table 4.3.l cannot be demonstrated and credit is being taken for compliance with Specification 3.3.3.b, the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily and the position of the other closed valve located in that pressure line shall be recorded daily.

j. Following each use of the LPSI system for shutdown cooling, the reactor shall not be made critical until the LPSI check valves (CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed.

181 To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

~Reduced pressure testing is acceptable (see footnote 5, Table 4.3.1).

Minimum test differential pressure shall not be less than 150 psid .

4-16 Amendment No . .§a., ~' ~' 42-,

4.3 SYSTEMS SURVEILLANCE (Cont'd)

Basis The inspection program specified places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems. 111 In addition, th.at port.ion of the reactor vessel shell welds which will be subj~cted to a fast neutron dose sufficient to change ductility properties will be inspected. The inspections will rely primarily on ultrasonic methods utilizing up-to-date analyzing equipment and trained personnel. To the extent applicable, based upon the existing design and construction of the plant, the requirements of Section XI of the Code shall be complied with. Significant exceptions are detailed in the requests for relief which have received NRC approval and are conta.ined in the Class 1, Class 2 and Class 3 Long-Term Inspection Plans.

Valve Testing To ensure the continued integrity of selected check valves which are relied upon to preclude a potential *LOCA outside containment, special requirements for periodic leak tests are specified~ In addition a valve disk position check for the LPSI check valves is specified following each use of the LP.SI system for shutdown cooling. This position check ensures that the four LPSI check valves have reclosed upon cessation of shutdown cooling flow.

References (1) FSAR, Section 4.5.6 (2)

Deleted (3) Systematic Evaluation Program Topic V-II.A, NRC letter to the licensee transmitting the final topic eval~ation dated November 9, 1981.

4-17 Amendment No. , 14, ~' !42-,

TABLE 4.3.1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum lal System Valve No. Allowable Leakage High Pressure Safety Injection Loop IA, Cold Leg 3101 5.0 gpm 3104 5.0 gpm Loop IB, Cold Leg 3116 5.0 gpm 3119 5.0 gpm Loop 2A, Cold Leg 3131 5.0 gpm 3134 5.0 gpm Loop 2B, Cold Leg 3146 5.0 gpm 3149 5.0 gpm Low Pressure Safety Injection Loop IA, Cold Leg 3103 5.0gpm Loop IB, Cold Leg 3118 5.0gpm Loop 2A, Cold Leg J:l33 5.0gpm

Loop 2B, .Cold Leg Footnote:

(al 3148 5.0gpm I. Leakage rates less than or equal ta 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptabl~ if the l~test measured rate has not exceeded the rate determined by the previous test by an amount that reduce.s the margin between measured leakage rate and the maximum permtssible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than. or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between m~asured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. *Leakage rates greater than 5.0*gpm are considered unacceptable.

5. Measured leakage rates must be adjusted for test pressures less than the maximum potential pressure differential across the valve by assuming leakage to be directly proporti~nal to the pressure differential to the one-half power .

NRC Order Dated April 20, 1981 4-18 Amendment No. ,

4.4 Deleted

4.5 4.5.1 CONTAINMENT TESTS Integrated Leakage Rate Tests A surveillance test program for the containment overall integrated leakage rate shall meet the 10 CFR 50, Appendix J, Type A test requirements or approved exemptions.

4.5.2 Local Leak Detection Tests

a. Test (1) Local leak rate tests shall be performed at a pressure of not less than 55 psig.

(2) Local leak rate tests for checking air lock door seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed at a pressure of not less than 10 psig.

(3) Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent.

(4) The local leak rate shall be measured for each of the following components:

(a} Containment penetrations that employ resilient seal gaskets, sealant compounds, or bellows.

(b} Air lock and equipment door seals.

(c} Fuel transfer tube.

(d} Isolation valves on the testable fl-uid systems' lines penetrating the containment.

(e} Other containment components which require leak repair in order to meet the acceptance criterion for any integrated leak rate test.

b. Acceptance Criteri.a (1) The total leakage from all penetrations and isolation valves shall not exceed 0.60 La.

(2) The leakage for an air lock door seal test shall not exceed

0. 023 La.

4-19 Amendment No. -l-2-, ~' -l-3-5,

4.5 CONTAINMENT TESTS

4.5.2 Local Leak Detection Tests

c. Corrective Action (1)

(continued)

If at any time it is determined that 0.60 La is exceeded, repairs shall be initiated immediately. If repairs are not completed and conformance to the acceptance criterion of 4.5.2.b(l) is not demonstrated with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the Plant shall be placed in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(2) If at any time it is determined that total containment leakage exceeds La, within one hour action shall be initiated to bring

the Plant to hot shutdown within the next six (6) hours and cold shutdown within the following thirty (30) hours.

(3) If air lock door seal leakage is greater than 0.023 La, repairs shall be initiated immediately to restore the door to less than specification 4.5.2.b(2). In the event repairs cannot be completed within 7 days, the Plant shall be brought to a hot shutdown condition within the next six (6) hours and cold shutdown within the following thirty (30) hours.

If air lock door seal leakage results in one .(1) door causing total containment leakage to exceed 0.60 La, the door shall be

declared inoperable and the remaining operable door shall be immediately locked closed and tested within four (4) hours.

As long as the remaining door is found to be operable, the provisions of 4.5.2.c(2) do not apply. Repairs shall be initiated immediately to establish conformance with

~pecification 4.5.2.b(l). In the event conformance to this specification cannot be established within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the Plant

sha 11 be brought to a hot shutdown within the -next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4-20 Amendment No.

ChaAge 7,

~'

4.5 CONTAINMENT TESTS 4.5.2 local leak Detection Tests (continued)

d. Test Frequency (1) Individual penetrations and containment isolation valves shall be leak.rate tested at a frequency of at least every six months prior to the first postoperational integrated leak rate test and at a frequency of at least every refueling thereafter, _not exceeding a two-year interval, except as specified in (a) and (b) below:

(a) The containment equipment hatch and the fuel transfer tube shall be tested at each refueling shutdown or after each time used, if that be sooner.

(b) A full air lock penetration test shall be performed at six-month intervals. During the period between the six-month tests when containment integrity is required, a reduced pressure test for th~ door seals or a full air lock penetration test shall be performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after either each air lock door opening or the first of a series of openings.

(2) Each three months the isolation valves must be stroked to the position required to fulfill their safety function unless it is established that such operation is not practical during plant operation. The latter valves shall be full-stroked during each cold shutdown.

4.5.3 Containment Isolation Valves

a. The isolation valves shall be demonstrated operable by performance of a cycling test and verification of isolation time for auto isolation valves prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit.

-b. Each isolation valve shall be demonstrated operable by verifying that on each containment isolation right channel or left channel test signal, applicable isolation valves actuate to their required position during cold shutdown or at least once per refueling cycle.

c. The isolation time of each power operated or automatic valve shall be determined to be within its limit as specified in Table 3.6.1

. when tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.

4.5.4 Surveillance for Prestressinq System Verify containment structural integrity in accordance with the I Containment Structural Integrity Surveillance Program.

~ 4-21 Amendment No. ~, -J-2.8,

4.5 CONTAINMENT TESTS (continued)

Basis The containment is designed for an accident pressure of 55 psig. 111 While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a temperature of about 104°F. With these initial conditions, following a LOCA, the temperature of the steam-air mixture at the peak accident pressure of 55 psig is 283°F.

Prior to initial operation, the containment was strength-tested at 63 psig and then leak rate tested. The design objective of this preoperational leak rate test was established as 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 55 psig. This leakage rate is consistent with the construction of the containment, 1 ~ which is equipped with independent leak-testable penetrations and contains channels over all unaccessible containment liner welds, which were independently leak-tested during construction.

Accident analyses have been performed on the basis of a leakage rate of 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With this leakage rate and with* a reactor power level of 2530 MWt, the potential public exposure would be below 10 CFR 100 guideline values in the event of the Maximum Hypothetical Accident. rai

. The performance of a periodic integrated leak rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.

In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leak rate test.is to be performed without preliminary leak detection surveys or leak repairs and containment isolation valves are to be closed in the normal manner.

This normal manner is a coincident two-of-four high radiation or two-of-four high containment pressure signals which will close all containment isolation val~es not required for engineered safety features except the component cooling lines' valves which are closed by CHP only. The control system is designed on a two-channel (right and left) concept with redundancy and physical seoaration. Each channel is capable of initiating containment isolation. 141

The Type A test requirements including pretest test methods, test pressure, acceptance criteria, and reporting requirements are in accordance with 10 CFR 50, Appendix J, requirements or approved exemptions.

The frequency of the periodic integrated leak rate test is keyed to the refu~ling schedule for the reactor because these tests can best be performed during refueling shutdowns. The specified frequency is as specified in 10 CFR Part 50, Appendix J which is based on three major considerations.

First is the low probability of leaks in the liner because of (a) the test of the leak tightness of the welds during erection; (b) conformance of the complete containment to a low leak rate at 55 psig during preoperational testing which in consistent with 0.1% leakage at design basis accident (OBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation .

4-22 Amendment No. -l-99, +/-3 4.5 CONTAINMENT TESTS

Basis (continued)

Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isola,tion valves-) .and the low value (0.60L 8 ) of the total leakage that is specified as acceptable from penetrations and isolation valves. Third is the Containment Structural Integrity Surveillance Program which provides assurance that an important part, of the structural integrity of the containment is maintained.

The basis for specification of a total leakage rate of 0.60 La from penetrations and isolation valves is specified to provide assurance that the integrated leak rate would remain within the specified limits during the intervals between integrated leak rate tests. This value allows for possible deterioration in the intervals between tests.

The basis for specification of an airlock door seal leakage rate of -0.023 La is to provide assurance that the failure of a single airlock door will not result.in the total containment leakage exceeding 0.6 La. The seven (7) day LCO specified for exceeding the airlock door leakage limit is acceptable since it requires that the total containment leakage limit is not exceeded.

References (1) Updated FSAR Section 5.8.2.

(2) Updated FSAR Section 5.8.8 (3) Updated FSAR 14.22

( 4). Updated.FSAR Section 8.5.1.2 (5) 10 CFR Part 50, Appendix J.

4-23 Amendment No. -!-2-, -l-99, 2-6, -!a-5,

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS

4.6.1 Surveillance Requirements Safety Injection System

a. System tests shall be performed at each reactor refueling i.nterval.

A test safety injection signal will be applied to initiate operation of the system. The safety tnjection and shutdown cooling system pump motors may be de-energized for thi.s test. The system will be considered, satisfactory 1f control board indication and visual observations inditate that all components have received the

~afety injection stgnal in the proper seq~~nce and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel).

4.6.2 Containment Spray System

a. System test shall be performed at each reactor refueling tnterval.

The test shall be performed with the isolation, valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by tri.ppi-ng the normal a,ctuat ion instrumentation.

b. The test will be considered satisfactory if visual observations.

i.ndicate all components have operated satisfactori.ly .

4.6.3 Pumps

a. The s~fety injection pumps, shutdown cooling pumps, and containment spray pumps shall be started at intervals not to exceed three months. Alternate manual startihg between ccintrol room ~onsole and the local breaker shall be practiced in the test program.

b. Acceptable levels of performance shall be that the pumps start, reach their rated heads on recirculation flow, and operate for at least fifteen minutes.

4.6.4 Valves

a. Each Safety Injection Tank flow path shall be verified OPERABLE withi"n 7 days prior to each reactor startup by verifying each motor operated isolation valve is open by observing valve position indication and valve itself, and locking open the associated circuit breakers.

b. The Low Pressure Safety Injection flow path shall *be verified OPERABLE within 7 days prior to each reactor startup by verifying flow control valve CV-3006 is open, and its air supply is isolated .

4-24 Amendment No. -&!, +3-, %, H-7, -!3-l, 62-,

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS

4.6.4 Surveillance Requirements {continued)

Valves c.

(continued)

The safety injection recirculation path shall be verified OPERABLE within 7 days prior to each reactor startup by verifying valves CV-3027 and 3056 are open and their switches HS-3027A, HS-3027B, HS-3056A, and HS-3056B are open.

d. Each Containment Spray Valve manual control shall be verifi.ed to be OPERABLE at least once each refueling by cycling each valve from the control room while observing valve operation at least each 18 months.

4.6.5 Containment Air Cooling System

a. Emergency mode automatic valve and fan operation will be checked ,

for OPERABILITY during ea~h refueling shutdown.

b. Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months .

I 4-25 Amendment No. W, +!t, +:1-, H-1, 62-,

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS

Basis The safety injection system and the containment spray system are principal plant safety features that are normally inoperative during reactor operation.

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

The method of assuring OPERABILITY of these systems is therefore, to combine systems tests to be performed during annual plant shutdowns, with more frequent component tests, which can be performed during reactor operation.

The refueling interval systems tests demonstrate proper ~utomatic operation of the safety injection and containment spray systems. A test signal is applied to initiate automatic action and verification made that the components receive the Safety Injection Signal in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry. 11 *21

During reactor operation, the instrumentation which is depended on to initiate safety injection and containment spray is generally checked each shift and the initiating circuits are tested monthly. In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judgment that more frequent testing would not significantly increase the reliability (ie, the probability that the component would operate when required), yet more frequent test would result in increased wear over a -long period of time.

Other systems that are also important to the emergency cooling function are the SI tanks, the component cooling system, the service water system and the containment air coolers. The SI tanks are a passive safety feature. In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance.

References (1) FSAR, Section 6.1.3.

(2) FSAR, Section 6.2.3.

(Next Page is 4-42)

4-26 Amendment *No. -i-1+, -l-J..l., -+/--e-2-,

4.14 STEAM GENERATORS SURVEILLANCE

4.14.1 Verify Steam Generator tube integrity is acceptable in accordance with the Inservice Inspection and Testing Program, Specification 6.5.7, and the Steam Generator Tube Surveillance Program, Specification 6.5.8 .

{Next page is 4-70) 4-66 Am~ndment No. H, S, IS, a-3-, a-9, 4§., £2., 9-1-, -I-%, -H-2-, 3-2-, -l-4-1-,

4.16 INSERVICE INSPECTION PROGRAM FOR SHOCK SUPPRESSORS (Snubbersl Applicability Applies to periodic surveillance of safety-related snubbers as described per Specification 3.20.

4.16.1 Specifications Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program in addition to the requirements of Specification-6.5.7. As used in this specification, "type of snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity.

a. Visual Inspection Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these categories (inaccessible and accessible} may be inspected independently according to the following paragraph:

If one or more unacceptable snubbers are found, the next inspection interval shall be 2/3 (-25%) of the previous interval. If no unacceptable snubbers are found, the next interval may be doubled

(-25%), but not to exceed 48 months. The interval extension provisions of Technical Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months .

Inspections performed before the interval has elapsed may be used as a new reference point to determine the next inspection.

However, the results of such early inspections, performed before the original required time interval has elapsed (nominal time less 25%}, may not be used to lengthen the required inspection interval.

Any inspection whose results require a shorter inspection interval will override the previous schedule.

b. Visual Inspection Acceptance Criteria Visual inspection shall verify that (1} the snubber has no visible indications of damage or impaired OPERABILITY, (2} attachments to the foundation or supporting structure are functional, and (3} fasteners for the attachment of the snubb.er to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1} the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers, irrespective of type, that may be generi.cally susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Technical Specification 4.16.ld or 4.16.le, as app l i cable. A11 snubbers found connected to an i nope*rab 1e common hydraulic fluid reservoir sh.all be counted as unacceptable for determining the next inspection interval .

4-71 Amendment No. B, 69, -l-G-7, +/-48, 64,

4.16 INSERVICE INSPECTION PROGRAM FOR SHOCK SUPPRESSORS (Snubbersl I 4.16.1

f. Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each safety related snubber in use in the.plant shall be reviewed to verify that the indicated servi.ce life has not been exceeded or will not be exceeded prior to the next scheduled service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in '

the records .

4-74 Amendment No. 3-, &9, 93-, -l-Q.7., 164

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant superintendent shall be responsible for overall plant operation and shall delegate in writing the succession for this.

responsibility during his absence.

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

. 6.1.2 The Shift Supervisor (SS) shall be responsible for the control room

. command function. During any absence of the SS from the control room while the plant is above.COLD SHUTDOWN, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the SS from the control room while the plant is in COLD SHUTDOWN, an individual with an active SRO license or Reactor Operator (RO) license shall be designated to assume the control room command function.

6.2 ORGANIZATION 6.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the Palisades plant .

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

b. The plant superintendent shail be responsible for overall *plant safe operation and shall have control over those onsit~ activities necessary for safe *operation and maintenance of the plant.

c. A specified corporate executive shall have corporate responsibility I*

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

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

6-1

Amendment No. -!&, +s, 98, -+/--3-9,

6.0 ADMINISTRATIVE CONTROLS

6.2.2 Plant Staff

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating above COLD SHUTDOWN.

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is above COLD SHUTDOWN, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

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

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

e. Administrative procedures shall be developed and implemented to limit the working hours of plant staff who perform safety-related functions (e.g., licensed SROs, licensed ROs, radiation safety personne 1 , auxiliary operators, and key maintenance personne 1) .

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

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

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

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

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

6-2 Amendment No. -l-6, ?rl-, W, fl+, 5-, 98, H-7, +/-3-9, .J-6.2.,

6.0 ADMINISTRATIVE CONTROLS

6.2.2.e Plant Staff (Continued)

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

Conttols shall be included in the procedures such that individual overtime shall be reviewed monthly by the plant superintendent or his designee to ensure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

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

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. If either SRO on shift satisfies the Shift Engineer qualification requirements, then the STA does not need to be stationed .

6.3 6.3.l 6.3.2 PLANT STAFF QUALIFICATIONS Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI Nl8.1-1971 for comparable positions.

The radiation safety manager shall meet the qualifications of a Radiation Protection Manager as defined in Regulatory Guide 1.8, September 1975. For the purpose of this secti-0n, "Equivalent," as utilized in *Regulatory Guide 1.8 for the bachelor's degree requirement, may be met with four years of any one or combination of the following:

(a) Formal schooling in science or engineering, or (b) operati-0nal or technical experience and training in nuclear power.

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

The Shift Engineer shall hold a Senior Reactor Operator license.

6.3.4 The plant staff whQ perform reviews which ensure compliance with 10 CFR 50.59 shall meet or exceed the minimum qualifications of ANS 3.1-1987, Section 4.7.1 and 4.7.2. A Senior Reactor Operator license.or certification shall be considered equivalent to a bachelors degree for the purpose of this specification *

Amendment No. M, 6-3

~, ~, W, fil-, 68, +s, 98, 2+, .J.39,

6.

6.4 PROCEDURES Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory 11 11 Guide 1.33, Revision 2., Appendix A, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety-related equipment.

d. Site Fire Protection Program implementation.

e. All programs specified in Specification 6.5 .

6-4 Amend.ment No. 6, a+, 69, 7&, Ma-, 2+, §4,

6.0 ADMINISTRATIVE CONTROLS

6.5 PROGRAMS AND MANUALS The following programs shall be established, implemented, and maintained:

I.

6.5.1 Offsite Dose Calculation Manual (ODCM)

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

b. The ODCM shall also contain (1) the radioactive effluent controls and radiological environ~ental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Release Report required by Specification 6.6.2. and Specification 6.6.3.

c. Changes to ODCM:

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

a.

b.

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and A determination that the change will maintain t'he level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 5Q.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reli~bility of effluent, dose, or setpoint calculations.

2. Shall become effective after approval by the plant super-i ntendent.

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

6-5 Amendment No. ~' -l-64,

6.0 ADMINISTRATIVE CONTROLS

6.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage to the engineered safeguards rooms, from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident, to as low as practical. The systems include Containment Spray system and Safety Injection system including the containment sump suction piping. This program shall include the following:

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

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

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

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

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

6.5.3 Post Accident Sampling Program This program provides controls which will ensure the capability to accurately determine the airborne iodine concentration in vital areas and which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

This program shall include the following:

a. Training of personnel, I.

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analytic equipm~nt.

Amendment No. 6-7, -l-GG,

6.0 ADMINISTRATIVE CONTROLS

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

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

b. Limitations on the concentrations of radio.active material released in liquid effluents to unrestricted areas ~onforming to 10 CFR 20, Appendix B, Table 2, Column 2.

c. Monitoring, sampling, and analysis of radioa'ctive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

d. Limitation on the annual and quarterly do.ses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to 10 CFR 50, Appendix I,

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

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

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

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

6-7 Amendment No. -l-§4,

6. 0 ADMINISTRATIVE CONTROLS

6.5.5 Containment Structural Integrity Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stres'Sed concrete containments, including effectiveness of Hs corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measu.rements prtor to initial operations. The Containment Structural Integrity S.urveil 1ance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide l.35, Revision 3, 1989.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Containment Structural Integrity Surveillance Program inspection frequencies.

6.5.6 Primary Coolant Pump Flywheel Surveillance Program Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each refueling.

6.5.7 lnservice Inspection and Testi.ng Program This program provides controls for inservice inspection and testing of ASME Code Cl ass 1, 2, and 3 components including app 1i cab 1e supports.

The program shall include the following:

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

B&PV Code terminology Required interval for inservice testing for performing inservice activities testi~g activities Weekly s 7 days Monthly s 31 days Quarterly or every 3 months s 92 days Semi annua 11 y .or every 6 months s 184 days Every 9 months s 276 days Yearly or annually s 366 days Biennially o.r every 2 years s 731 days

b. The provisions of Surveillance Requirement 4.0.2 are applicable to the above required intervals for performing inservice testing activities;

c. The provisi-0ns of Surveillance Requirement 4.-0.3 are applicabl~

to .inservice testing activities; and

d. Nothing in tha B&PV tode shall be construed to supersede the requirements -Of any T~chnical Specification .

6-8 Amendment No.

~------ --

6.0 ADMINISTRATIVE CONTROLS

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

a. Steam Generator Tube Sample Selection and Inspection The inservice inspection may be limited to one SG on a rotating schedule encompassing 6% of the tubes if the results of previous inspections indicate that both SGs areperformi.ng in a like manner.

If the operating*conditions in one SG are found to be more severe than those in the other SG, the sample sequence shall be modified to inspect the most severe conditions.

The SG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 6.5.8-1. The tube$ selected for each inservice inspection shall include at least 3% of the total number of tubes in all SGs; the tubes selected for th~se inspections shall be selected on a random basis except:

I. Where experience in similar plants with similar water

.chemistry indicates critical areas to be inspected; then at least 50% of the tubes inspected sha 11 be from these critical areas.

2. The first sample of tubes s~lected for each tnservice inspection of each SG shall include:

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

b) Tubes in those areas where experi~nce has indicated potential problems.

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

6-9 Amendment No. H, t-S, -16, ~' .a-9, 4§., .§.2., 9-l-, ~' H-2-, ~' !41-

6.0 ADMINISTRATIVE CONTROLS 6.5.8 Steam Generator Tube Surveillance Program (continued)

3. The tubes selected as the second and third .samples (if required by Table 6.5.8-1) during each inservice inspection may be subjected to a partial tube inspection provided:

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

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

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

Category Inspection Results C-1 Less than 53 of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 13 of the total tubes inspected are defective, or between 53 and 103 of the total tubes inspected are degraded tubes .

C-3 Note:

More than 103 of the total tubes inspected are degraded tubes or more than 13 of the inspect~d tubes are defective.

In all inspections, previously degraded tubes must exhibit significant (greater than 103) further wall penetrations to be included in the above percentage calculations.

b. Inspection Frequencies The above required inservice inspection of SG tubes shall be

performed at the following frequencies:

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

6-10 Amendment No . 39, 45, .§.2., -!-% , .f-1.2., +/-3-2-, -141-

6.0 ADMINISTRATIVE CONTROLS

6.5.8 Steam Generator Tube Surveillance Program 2.

{continued)

If the results of the inservice inspection of a SG conducted in accordance with Table 6.5.8-1 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 6.5.8.b.l; the interval may then be extended to a maximum of once per 40 months.

3. Additional, unscheduled inservice inspections shall be performed on each SG in accordance with the first sample inspection specified in Table 6.5.8-1 during the shutdown subsequent to any of the following conditions:

a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.1.5.

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

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

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

c. Acceptance Criteria

1. As used in this Specification:

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

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

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

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

6-11 Amendment No. -39, -&2-, M6-, H-2-, .f-3-2., -14!-

6.0 ADMINISTRATIVE CONTROLS

6.5.8 Steam Generator Tube Surveillance e}

~rogram (continued}

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

f} Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 403 of the nominal tube wall thickness.

g} Unserviceable described the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 6.5.8.b.3, above.

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

i} Preservice Inspection means an inspection of the full length of each tube in SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the shop hydrostatic test and prior to initial POWER OPERATION using the eqµipment and techniques expected to be used during subsequent inservice inspections.

. 2. The SG shall be determined OPERABLE after completing the corresponding action~ (plug all tubes* exceeding.the plugging

.limit and all tubes containing through-wall cracks} required by Table 6.5.8-1 .

6-12 Amendment No. 39, .§-2., G&, H-2-, +/-3-2-,. .J4l.

lST SAMPLE INSPECTION TABLE-.8-1 STEAM GENERATO~BE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A 5 Tubes per 5.G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additional 2S tubes in this S.G. C-2 Plug defective tubes C-1 None and inspect additional 45 tubes in this S.G. C-2 Plug defective tubes C:-3 Perform action for C-3 result of first Sample C-3 Perform action for C~3 result of first N/A N/A Sample C-3 Inspect all tubes tn All other None N/A N/A this ~.G., plug de- S.G.s are fective tubes and C-1 inspect 25 tubes in each other S.G. Some S.G.s *Perform action for N/A N/A

. C-2 but no C-2 result of second additional sample 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal S.G. are notification to NRC C-3 with written follow up within next Addition al inspect all tubes

30 days s ..G. is each S.G. and plug C-3 defective tubes. N/A N/A 5 = 6/n % Where n is the number of steam generators inspected during an inspection 6-13 Amendment No. l4l

6.0 ADMINISTRATIVE CONTROLS

6.5.9 Secondary Water Chemistry P~ogram A program shall be established, implemented and maintained for monitoring of secondary water chemistry*to inhibit steam generator tube degradation and shall *include:

a. lderitification of a sampling schedule for the critical variables and control points for these variables,

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

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,

d. Procedures for the recording and management of data,

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

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

6-14 Amendment No.

6.0 ADMINISTRATIVE CONTROLS 6.5.10 Ventilation Filter Testing Program

A program shall be established to implement the following required testing of Control Room Ventilation {CRV) and Fuel Pool Ventilation (FPV) systems at the frequencies specified in Regulatory*Guide 1.52,

~evision 2 (RG 1.52), and in accordance with RG 1.52 and ASME N510-l989, at the system flowrates and tolerances specified below*:

a. Demonstrate for each* of the *ventilation systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration ana system bypass < 0.05% for the CRV and < 1.00% for the FPV when tested in accordance with RG 1.52 and ASME N510-1989:

Ventilation System Fl owrate * (CFM)

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

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

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

b. Demonstrate for each of *the ventilation systems that an inplace test of the charcoal adsorber shows a penetrat-i on and system bypass

< 0.05% for the CRV and < 1.00% for the FPV when tested in accordance with RG 1.52 and ASME N510-1989.

Ventilation System Flowrate (CFM)

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

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

c. Demonstrate for each of the ventilation systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in RG 1.52 shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of ~ 30°C and equal to the relative humidity specified as follows:

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

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

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

Ventilation System Delta P (In H2.Ql Flowrate (CFM)

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

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

e. De~onstrate that the heaters for each of the ventilation systems dissipate the following specified value +/- 20% when tested in accordance with ASME N510-1989:

Ventilation System Wattage VHX-26A and VHX-268 15 kW The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Ventilation Filter Testing Program frequencies.

Should the 720~hour limitation on charcoal adsorber operation occur during a plant operation requiring

the use of the charcoal adsorber - such as refueling - testing may be delayed until the completion of the plant operation or up to 1,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of filter operation; whichever occurs first.

6-15 Amendment No.

6.0 ADMINISTRATIVE CONTROLS

6.5.11 Reserved 6.5.12 !echnical Specifications (TS) Bases Control Program This program provides a means for processi.ng changes to the Bases of these Technical Specifications.

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

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

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

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

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

d. Proposed changes that meet the criteria of Specification 6.5.12 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). .

6-16 Amendment No. l-1-G,

6.0 ADMINISTRATIVE CONTROLS

6.5.13 6.5.14 Reserved Reserved

- 6-17 Amendment No.

6.0 ADMINISTRATIVE CONTROLS

6.6 6.6.1 REPORTING REQUIREMENTS The following reports shall be submitted tn accordance with 10 CFR 50.4.

Occupational Radiation Exposure Report This report shall include a tabulation on an annual basis of the number of station, utility and other personnel (includ*ing contractors) receiving exposures greater than 100 mrem/year and their assoc.i ated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignment to various duty functions may be estimates based on pocket dosimeter, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions. The report shall be submitted by April 30 of each year.

6.6.2 Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and in 10 CFR 50, Appendix I, Sections IV.B.2t IV.B.3, and IV.C.

6.6.3 Radioactive Effluent Release Report The Radioactive Effluent Re 1ease Report sha 11 be submitted i.n accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive li~uid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and shall be in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section lV.B.l.

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

6-18 Amendment No. 6, 6, 36-, SS, GS, -l-54;

6.0 ADMINISTRATIVE CONTROLS

6.6.5 Core Operating Lim.its Report (COLR)

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

.I 3 .1.1 ASI Limits.

3.10.5' Regulating Group Insertion Limits 3.23.1 Linear Heat Rate (LHR) Limits 3.23.2 Radial Peaking Factor Limits

b. The anal yt i cal methods used *to determ.i ne the core operating l i mi ts shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:

1. .XN-75-27(A), "Exxon Nuclear Neutronics Design Methods for Press.uri zed Water Reactors," and Supplements 1(A), 2(A),

3(P)(A), 4(P)(A), and 5(,P)-(A); Exxon Nuclear Company.

(LCOs 3.1.1, 3.10.1, 3.10.5, 3.23.1, & 3.23.2)

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

and Appendix B(P)(A) and Supplements l(P)(A), 2(P)(A);

Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, &3.23.2)

3.

4.

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

Exxon Nuclear Company. (lCOs 3.1.1, 3.13.1, &3.23.2)

ANF-84-093(P)(A), "Steamline Break Methodology for PWRs," and Supplement.l(P)(A); Advanced Nuclear Fuels Corporation.

(LCOs 3.10.1, 3.10.5, 3.23.1, &3.23.2J 5..

XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing," and Supplements l(P)(A), 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company. (LCOs 3.1.1, 3.10.5, 3.23.1,

&3.23.2)

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

(LCOs 3.10.5, 3.23.1, &3.23.2) a) XN-NF-a2~2o(A), "Exxon Nuclear tompany Evaluation Model EXEM/PWR ECCS Model Updates," and Supplements l(P)(A),

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

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

c} XN.-NF-81-58(A)*, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," and Supplements l(P)(A},

2(P)(A), 3(P){A), and 4(P)(A); Exxon Nu.clear Company .

6-19 Amendment No. 69,

6.0 ADMINISTRATIVE CONTROLS

6.6.5 COLR {continued) d)

XN-NF-85-16{A), "PWR 17x17 Fuel Cooling Tests Program,"

Volume ~ ~nd Supplements l{P){A), 2{P){A), and 3{P){A),

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

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

7. XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company. (LCOs 3.10.5, 3.23.1, & 3.23.2)

8. ANF-1224(P)(A), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," and Supplement l(P)(A);

Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.23.1, &

3.23.2)

9. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, &.

3.23.2)

10. EMF-92-153{P){A), "HTP: Departure from .Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation. (LCOs 3.1.1, 3.23.1, &3.23.2)

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

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

6.6.6 Reserved 6.6.7 Accident Monitoring Instrument Report When a report is required by Condition 3.17.4.7c, "Accident Monitoring Instrumentation," a report shall be submitted within the following 30 days. The report shall outline the .preplanned alternate method of monitoring, the cause, of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status .

6-20 Amendment No.

6.0 ADMINISTRATIVE CONTROLS

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

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

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

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

1. Number and extent of tubes inspected.

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

~ndication of an imperfection.

3. Identification of tubes plugged .

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

6-21 Amendment No. -!GS, -!69,

6.0 ADMINISTRATIVE CONTROLS

6.7 6.7.1 HIGH RADIATION AREA In lieu of the "control device" or "alarm signal" -required by 10 CFR 20.1601, each high radiation area in which the intensity of radiation is greater than 100 mrem/hour but less than 1000 mrem/hour at 30 cm from the radiation source or from any surface which the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a radiation work permit.* Any individual or group of individuals permitted to enter such areas shall be provtded with or accompanied by one or more of the following:

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

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

c. An individu~l qualified in radiation protection procedures who is i equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation work permit.

6.7.2 The above requirements shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hour at 30 cm from the radiation source or from any surface which the radiation penetrates. In addition, locked doors shall be provided to prevent unauthorized entry into such areas (> 1000 mrem/hour) and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or the radiation safety supervisor.

Radiation safety personnel or personnel escorted by radiation safety personnel shall be exempt from the radiation work permit issuance requirement during the performance of their assigned radiation protecti~n duties provided they comply with approved radiation protection procedures for entry into high radiation areas .

6-22 Amendment No. 48, ~, -l-54,

ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255

TECHNICAL SPECIFICATION CHANGE REQUEST REVISION OF ADMINISTRATIVE CONTROLS Existing Pages Marked to Show Changes

79 Pages

1.2 MISCELLANEOUS DEFINITIONS

CORE OPERATING LIMITS .REPORT (COLR}

The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits

sha 11 be determined for each reload eye le in accordance with Specification 6.9.1.flMIWI: .. Plant operation withi.n.these limits is addressed in individu*aT*****-specifications.

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall inchufo all persons *,:ho *are not occ1:Jpationally associated with the p*lant. This category does not incl1:Jde employees of the 1:Jtility, its contractors, or its.vendors. Also

=~~1~:::tf=~mt!h!!1c:a::r~:~rt~~.persons who enter the site to service OFFSITE DOSE CALCULATION MANUAL (ODCMl The OFFSITE DOSE CALCULATION MANUAL shall contain the Cl:!rrent methodology and parameters 1:Jsed in. the calc1:Jlation of offsite doses dl:!e to radioactive gaseo1:Js and liq1:Jid effl1:Jents, in the calc1:Jlation of

~h:e~==d~~~ !fq~~~ R::1=1=:i::~ii=~l~~n:!:~:1aM:nlt!~1=:t~~!;;:~.an#h!n ODCM shall also contain the (1) Radioactive Effll:!eAt CoAtrols aAd

~=~!:1:~lct~)E~:!:~;;~;!:! :;At~:rl:, 0 ~::~r::st:e~:ir:: 1 :~e~Pi:it~:atioA.

Radiological EnviroAmental Operating Report and the Radioactive Effll:!eAt Release Report req1:Jired by Speci.fieatioA 6.9.3. .

PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall eoAtai A the cl:!rreAt formt:il a, sampl iflg, aAalyses, tests, and determiAatioAs to be made to eAsl:!re that the proeessiAg aAd pacl<agiAg of solid radioactive wastes based oA demoAstrated proeessiAg of ~ct1:Jal or siml:Jlated wet solid ~:astes will be accomplished iA s1:Jch a way as to ass1:Jre compliaAce with 10 CFR 20, 10 CFR 71, Federal aAd State re~1:Jlati0As, aAd other req1:JiremeAts goverAiAg the disposal of the radioactive waste. *

SITE BOUNDARY*

The SITE BOUNDARY shall be that liAe beyoAd which the laAd is Reither O'dAed AOr otherwise COAtrolled by the liceAsee.

UNRESTRICTED AREA AA UNRESTRICTED AREA shall be aAy area at or beyoAd the SITE BOUNDARY access to which is not coAtrolled by the'liceAsee for pl:!rposes of protectioA of individl:!als from expos1:Jre t~ radiatioA aAd radioactive materials or, aAy area withiA the SITE BOUNDARY 1:Jsed for resideAtial ql:!arters or for iAdl:Jstrial, commercial, iAstit1:JtioAal, or recreatioAal pl:!rposes .

Amendment No. &§., -!-54, ~, -l69

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 Safety Limit - Reactor Core The Minimum DNBR of the reactor core shall be maintained greater than or equal to the DNB correlation safety limit.

Correlation Safety Limit XNB 1.17 ANFP

1.154 HTP 1.141 Applicability Safety Limit 2.1 is appllcable in HOT STANDBY and POWER OPERATION.

Action

2. I. I 2.2 Safety Li.mit - Primary Coolant System Pressure (PCS)

The PCS Pressure shall not exceed 2750 psia.

Applicability Safety Limit 2.2 is applicable when there is fuel in the reactor.

Action 2.2.1 2.3 Limiting Safety System Settings - Reactor Protective System (RPS)

The RPS trip setting limits shall be as stated in Table 2.3.1.

Applicability Limiting Safety System Settings of Table 2.3.1 are applicable when the associated RPS channels are required to be OPERABLE by Specification 3.17.1.

Action 2.3.l If an RPS instrument setting is not within the allowable settings of Table 2.3.1, immediately declare. the instrument i.noperable and complete corrective action as directed by *Specification 3 .17. I.

Amendment No. 3-1-, 2-5, 43-, -l-18, 3+, W, +/-68 2-1

3.17 INSTRUMENTATION SYSTEMS Specification

3.17.4 The Accident Monitoring Instruments listed in Table 3.17.4 shall be OPERABLE. (Specifications 3.0.3, 3.0.4, and 4.0.4 do not apply.)

Applicability Specification 3.17.4 applies when the PCS temperature is> 300°F.

Action 3.17.4.1 With one required channel of functions 1 through 14 inoperable for one or more functions:

aii Restore channel to OPERABLE status within 7 days.

3.17.4.2 Mith two required channels of functions 1 through 14 inoperable for one or more functions:

a!~!: Restore one channel to OPERABLE status within 48. hours.

3.17.4.3 With position indication inoperable for one or more Containment Isolation Valves:

Restore the indication to OPERABLE status or lock the associated valves in the closed position within 7 days.

3.17.4.4 If any action required by 3.17.4.1 through 3.17.4.3 is not met AND the associated completion time has expired,

The reactor shall be placed in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and .

The reactor shall be placed in a condition where the affected equipment is not required, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.5 With one channel of functions 16 through 21 inoperable for one or more functions:

a:~: Restore the channel to OPERABLE status within 7 days.

3.17.4.6 With two required channels of functions 16 through 21 inoperable for one or more functions:

ail,! Restore one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.7 If any action required by 3.17.4.5 or 3.17.4.6 is not met AND the associated completion time has expired:

With two CETs in any one quadrant inoperable, complete Action 3.17.4.4 in lieu of Action 3.17.4.7 c),

With two RVWL channels inoperable, initiate alternate monito*ring within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , Submit a report to the NRC withiH 30 days after the eveHt, OijtliHiHg the action takeH, the caijse of the iHoperability, aHd the 1

~!~A li1l\1:ili~iili~:lii~lli:iimii,ii:i:i;~:i:l1i!~;l::::::i:~:ii~!i:i: s to Op ERAB LE st ah s ;

Restore the channels to OPERABLE status prior to startup from the next refueling.

Amendment No. 136, 147, 6 4.0 SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance requirements shall be applicable during the reactor operating conditions associated with individual Limiting Conditions for Operation unless otherwise stated in an individual surveillance requirement.

4.0.2 Unless otherwise specified, each surveillance requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval, and

b. A total maximum combined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the operability requirements for~ Limiting Condition for Operation. The time limits of the action requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The action requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits of the action requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not have to be performed on inoperable equipment .

4.0.4 Entry into a reactor operating condition or other specified condition shall not be made unless the Surveillance Requirements associated with a Limiting Condition of Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to plant conditions as required to comply with action requirements.

4.0.5 SurveillaAce RequiremeAts for iAservice iAspectioA aAd testiAg of ASME Code Class 1, 2, aAd 3 compoAeAts shall be applicable as follows:

a. IAservice iAspectioA of ASME Code Class 1, 2, aAd 3 compoAeAts aAd iAservice testiAg of ASME Code Class 1, 2, aAd 3 pumps aAd valves shall be performed iA accordaAce with SectioA XI of the ASME Boiler aAd Pressure Vessel Code aAd applicable AddeAda as required by 10 CFR 50, SectioA 50.55a(g), except where specific writteA relief has beeA graAted by theCommissioA pursuaAt to 10 CFR 50,

SectioA 50.55a(g)(6)(i) .

4-1 Amendment No. 30, 5,1, 130, 62-

4.0 SURVEILLANCE REQUIREMENT (Continued)

- b. SurveillaRce iRtervals specified iR SectioR XI of the ASME Boiler aRd Pressure Vessel Code aRd applicable AddeRda for the iRservice iRspectioR aRd testiRg activities required by the ASME Boiler aRd Pressure Vessel Code aRd applicable AddeRda shall be applicable as follows iR these TechRical SpecificatioAs:

ASME Boiler aRd Pressure Vessel Code aRd applicable AddeRda for perform1Rg 1Rserv1ce termiRology for iRservice iRspectioR aRd testiAg iRspectioR aRd testiRq activities activities Weekly At least oRce per 7 days MoRthly At least oRce per 31 days Quarterly or e*.*ery 3 moRths At least oRce per 92 days SemiaRRually or every 6 moRths At least oRce 13er 184 days Every 9 moRths At least oRce 13er 276 days Yearly or aRRually At least oRce per366 days

c. The provisioRs of SpecificatioR 4.0.2 are a13plicable te the above required frequeRcies for perfermiRg iRservice iAspectieR aRd testiRg activities.

d. PerfermaRce of the above iRservice iRspectioR aRd testiRg activities shall be iR additioR to ether specified SurveillaRce RequiremeRts.

e. NothiRg iR the ASME Boiler aRd Pressure Vessel Code shall be eeRstrued to su13ersede the requiremeRts of aRy TechRieal SpecificatioA .

4-2 Amendment No. ~' 6 4.0 BASIS Specifications 4.0.1 through ~1,i:~:P.MI establish the general requirements applicable to Surveillance Requireili'e'hfs. These requirements are based on the Surveillance requirements stated in the code of Federal Regulations, 10 CFR 50.36(c)(3):

"Surveillance. requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."

Specification 4.0.l establishes the requirement that surveillances must be performed during reactor operating conditions or other conditions for which the requirements of the Limiting Conditions for Operation apply, unless otherwise stated in an individual Surveillance Requirement. The purpose of this specification is to ensure that surveillances are performed to verify the operattonal status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a reactor operating condition or other specified condition for which the associated Limiting Conditions for Operation are applicable.

Surveillance Requirements do not have to be performed when the facility is in an operational condition for which the requirements of the associated Limiting Condition for Operation do not apply, unless otherwise specified.

The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception the requirements of a specification .

Specification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requirements may be extended. Item a. permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. Item b.

limits the use of the provisions of Item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified. The limits of Specification 4.0.2 are based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. These provisions are sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

Specification 4.0.3 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the operability requirements for a Limiting Condition for Operation.

Under the provisions of this specification, systems and components are assumed to be operable when Surveillance Requirements have

4-3 Amendment No . .f-3.{}, ~

4.0 BASIS (Contifluedl Specification 4.0.4 establishes the requ.irement that all applicable surveillances must be met before entry into a reactor operating condition or other condition of operati-0n specified in the Applicability statement. The purpose of this specification is to ensure that system and component operability requirements or parameter limits are met before entry into an operational condition for which these systems and components ensure safe

-operation of the facility. This provision applies to changes in reactor operating conditions or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the appli~able Surveillance Requirements must be performed within the surveilJance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

When a .shutdown is required to comply with action requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower operational condition.

SpecificatioR 4.9.§ establishes the reqlliremeRt that iRservice iRspectioR of ASME Code Class I, 2, aRd 3 compoReRts aRd iRservice testiRg of ASME Code Class I, 2, aRd.3 pllmps aRd valves shall ee performbd. iR accordaRce with a periodically llpdated versioR of SectioR XI of the ASME Boiler aRd Pressllre lfossel Code aRd AddeRda as reqlli red ey IQ CFR §9. §5a. These reqlli remeRts apply, except wheR relief has eeeR provided iR writiRg ey the CommissioR.

This specificatioR iRcllldes clarificatioR of the freqlleRcies for performiRg the iRservice iRspectioR aRd testiRg activities reqllired ey SectioR XI of the ASME Boiler aRd Pressllre Vessel Code aRd applicable AddeRda. This clarificatioR i.s provided to eRsllre coRsisteRcy iR SllrveillaRce iRtervals throllghollt the TechRical SpecificatioRs aRd to remove ameigllities relative to the.f~e,lleRcies for performiRg the reqllired i~service iRspectioR aRd testiRg act1v1trns.

URder the terms ef this specificatieR, the more restriritive reqlliremeRts of the TechRical Specific~ti~Rs take precedeRce ever the ASME Boiler aRd Press lire Vessel Code:*'.a*f!:d

appl i cael e AddeRda. *The reqll i remeRts ef SpecificatioR 4.9.4.t.o<perform SllrveillaRce activities before eRtry iRte a reactor eperatiRg coRdit'i OR or ether specified ceRdi ti eR takes precedeRce ever the ASME Boiler aRd Pressllre Vessel Code previsioR which allows pllmps aRd valves to ee tested lip to eRe week after retllrR te Rormal eperatieR. The TechRical SpecificatioR defiRitioR of eperaele dees Rot allew a grace peried eefere a cempoReRt, that is Rat capable ef performiRg its specified fllRctieR, is declared iRoperaele aRd takes .precedeRce ever the ASME Beiler aRd Pressllre Vessel Cede previsioR which allews a valve to ee iRcapaele ef performiR~ its specified fllRCtioR for lip to 24 hollrS Before eeiRg declared iRoperaele .

4-5 Amendment No. ~, -1&2-

4.1 OVERPRESSURE PROTECTION SYSTEM TESTS Surveillance Requirements In addition to the regui rements of Speei fi eati eFt 4. g. § Mlii!Mli~eiiv!!U$.I .

!!,f!~i!l!!!!!~i!1Jfff!!!'.9.~i~~!l~fli!~'ii111!!i1i.l!iii!il!!iie!i\I, eaar*=*puitv=*=*=*¥ro~t"path

+.Jj. Testi.ng the PORVs in accordance with the inserviae inspection requirements for ASME Boiler and Pressure Vessel Code,Section XI, Section IWV, Category B valves.

-t.I. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months*.

~* When the PORV flow path is required to be OPERABLE by Specification 3.1.8.1:

Performing a complete cycle of the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

Performing a complete cycl~ of the blo¢k valve priDr to heatup f~om COLD SHUTDOWN, if not cycled within 92 days.

4-.lm When the

....... 3.1.8.2:

PORV fl ow path is required to be OPERABLE by Spec ifi cation Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, b~t excluding valve operation, at least once per 31 days.

Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

6.i@ Both High Pressure Safety Injection pumps shall be verified

- - - - incapable of injection into the PCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , unless the reactor head is removed, when either PCS cold leg temperature is <300°F, or when both shutdown cooling suction valves, M0-3015 and M0-3016, are open.

With.the reactor vessel head installed when. the PCS cold leg temperature is less than 300°F, or if the shutdown cooling system isolation valves M0-3015 and M0-3016 are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be exceeded; therefore, both pumps are rendered inoperable.

For Cycle 11 only, this surveillance need not be performed until prior to startup for Cycle 12 .

Amendment No. 3{}, -149, 4-6

~' ~, -!Ga-, le4

4.2 EQUIPMENT SAMPLING AND TESTS T,abl e 4. 2.3 HEPA. FILTER AND CHARCOAL ABSORBER SYSTEMS VENTILATION SYSTEM TESTS lf;~i,:::i:control Room Ventilation and Isolation System (Rated flew: 765 efm) fii,Q aia*:~~~~:A'~~r-The filters iA each ef the abeve systems shall be demeAstrated eperable:

a. At least eAce per 31 days by iAitiatiAg, frem the CeAtrel Reem, flew threugh the HEPA filter aAd charceal adserbers aAd veri fyi Ag that the system eperates fer at least 1§ miAutes.

b. At least eAce per refl:JeliAg cycle er (l) after aAy strl:lctural maiAteAaAce. eA the HEPA filter er charceal adserber hel:lsiflgs, er (2) fellewiAg majer paiAtiAg, fire er chemical release iA aAy veAtilatieA_

zeAe cemmlJAicatiAg with the system wheA the ~EPA Filter er charceal adserbers are iA eperatieA by:

1. \'erifyiAg ~JithiA 31 days after remeval that a laberatery aAalysls ef a represeRtative carbeA sample ebtaiRed iA accerdaAce with Regulatery PesitieR C.. 6.b .. ef Regulatery Guide 1.52, RevisieA 2, March 1978, meets the laberatery testiRg criteria ef Regulatery PesitieA C.6.a ef Regulatery Guide 1.§2, RevisieA 2, March 1978 except that the Fuel Sterage Area shall have a methyl iedide limit ef 94% iAstead ef 99,, er replaciRg. with charceal adserbers meetiAg the specificatieRs ef Reglllatery Guide 1.52, .PesitieR C.6.a, RevisieR 2, March 1978.

2. \'erifyiAg that the HEPA filter baRk remeves greater thaA er eeiual te 991' ef th.e DOP wheR they .are tested i A pl ace. i A accerdaAce with ANSI N51Q .1975.wh.ile eperatiAg the system at its rated flew *nm,.

3. '.'erifyiAg that the charceal adserber remeves greater thaA er eeiual te 991' -~f a hydregeRated hydre.carbeR refrigeraAt test* gas wheA Oley a.re tested i R place i R accerdaRce with ANSI N510 1975 while eperatiRg the system at its rated flew +20%.

el.

At least once per refueling cycle by:

1. * '.'eri fyi Ag that the. pressure* drep acress the cembi Red HEPA filter aAd charceal adserber baRk is less thaR (6) iRches Water Gauge while eperatiRg the system.

~a. Verifying that on . . . ~. . . £.9.nt.. ~tm:l.l~.nt. . . .~..t~h:::..P.r.~. ~.-~:ure (lnd hi gh-radi at ion test s i gnal , the l9in:tr9:l:t:R9!P:1ThlY:!:ni:i:~::~mnnttsy st em au tom at i ca 11 y switches into a r*C'cTretiT'ii'tffi*g*t.ifi!llifjM$.t@j!rj!¢.M mode of operation with flow through the HEPA filter and . . cha*rcoaT . adsorber bank. (CeRtrel Reem veAtilatieR eRly.)

Amendment No. Sl, -l-6 4-14

4.2 EQUIPMENT SAMPLING AND TESTS Table 4.2.3 (continued}

HEPA FILTER AND CHARCOAL ABSORBER SYSTEMS VENTILATION SYSTEM TESTS

~~~i~~I.~,~.Q.~~a~t t~e P!:'~!ii[!'~!'J~~!:l!il\i11!!,!it~~si~:A m:!n!:!~~ i~e 183 4

ii.: ;:1:lllllm~ii~:e.~i ~~ ~ ~ ~~ v:y !~e;him~~:gf:i:~a.;1m&.~i~i1~)p' e~: ~ ~ ~n

'ft'O'fl¥'F8l"ffo'O'ffir::VC'fl"t i l at i eA eAl y * ) ..

,_

,.,,.,:,.;::::::~ :=:~:::::::::::,.,'ji!';i;::::::.,:*:=:<=i=:*:=>::::::::::~ *.

).

4. 'Jeri fyi Ag that 'llith the veAtil ati eA system exha1:1sti Ag thre1:1gh the HEPA/Charceal Filters at its rated fl 01'¥ i:20%, the by13ass flew thre1:1gh damper 1893 is less thaA 1% ef tetal flew. (F1:1el Sterage Area eAly.)

a. After every 729 he1:1rs (see Nate 1) ef charceal adserber e13eratieA hy:

\'erifyiAg withiA 31 days after remeval that a laberatory aAalysis of a re13reseF.1tative carbeA sample ebtaiAed iA accerdaAce with Reg1:1latory PositioA C.6.b. ef Reg1:1latery G1:1ide 1.52, RevisieA 2, March 1978, meets the laberatory testiAg criteria ef Reg1:1latory PesitioA C.6.a ef Reg1:1latery G1:1ide 1.52, RevisieA 2, March 1978 exce13t that the F1:1el Storage Area shall ~ave a methyl iedide limit ef 94% iAstead ef 99%, er re13laciAg with charceal adserbers meetiAg the s13ecificati0As ef Reg1:1latery G1:1idri l..52, 13esitieA C.6., Revisie.A 2, M*arch 1978.

e. After each cempl etc er 13arti al repl acemeAt ef a HEPA filter baAk by:

'Jeri fyi Ag that the HEPA filter baflk removes greater ttrnA or eq1:1al to 99%

of the DOP 'llheA they are.test~d ifl 13lace if.I accordaAce \vith ANSI N510 1975 while 013eratiAg the system at its rated flow +20%.

f. After each complete or 13artial replacemeAt of a charcoal adsorber baAk by+

\'eri fyi Ag that the charcoal adsorber removes greater thaf.I or eq1:1al to 99% of a hydrogeAated hydrocarboA refrigeraAt test gas \.iheA they are tested if.I 13lace if.I accordaAce with ANSI N510 1975 while 013eratiAg the system at its rated flow i:20%.

Verify that the Control Room .temperature is < 120°F~i!~Q:fiif!ii! once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months wheA the tem13erat1:1re i A the CoAtrol Room readies 105 °F.

[Note moved to program 6.5.10]

Note I. Should the 720-hour 1 imitation u1fliHNi)ie'naJi~:~:~~mi~frBefflB.rii:ra!~¥0.n?occur du r i ng a p1 ant opera ti on re qui r'l'h'Q'=*=*='th~:,.:U"s~""cn~,,.,.,,'t'fi'e*=*='R'~'fiR:,.:"FYYiEr' aAd charcoal adsorber - such as d1:1riAg a refueling - testing may be delayed until the completion of the plaAt operation or up to 1,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of filter operation whichever occurs first.

Amendment No. 8+/-, ~

4 15lif:lil

Basis - Table 4.2.2 Item 12 - Trisodium Phosphate (TSP) Tests Item 12a, TSP quantity verifi~ation

Verification *of the quantity of TSP in the baskets ensures that neither leakage nor other water sources in the containment reduce the basket content below the required minimum. This requirement ensures that there is an adequate quantity of TSP to adjust the pH of the post LOCA sump solution to a value between 7.0 and 8.0.

Item 12.b - TSP quality verification Periodic testing is performed to ensure the solubility and buffering ability of the TSP after exposure to the containment environment. Satisfactory completion -0f this test assures that the TSP in the baskets is "active" as required by Specification 3.19 ..

Adequate solubility is verified by submerging a representative sample of TSP from one of the baskets in containment in Un-agitated borated water heated to a temperature representing post-LOCA conditions; the TSP must completely dissolve within a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. The test time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is specified to

allow time for the dissolved TSP to naturally diffuse through the un-agitated test solution. Agitation of the test solution during the solubility verification is prohibited, since an adequate standard for the agitation intensity (other than no agitation) cannot be specified. The flow and turbulence in the containment sump during recirculation would significantly decrease the time required for the TSP to dissolve .

Adequate buffering capability is verified by a measured pH of the sample solution, following the solubility verification, between 7 and 8. The sample is cooled ~nd thoroughly mixed prior to measuring pH.

The quantity of the TSP sample, and quantity and boron concentration of the

water are chosen to be representative of post-LOCA conditions .

Amendment No. 165 May 19, 1995

4.3 SYSTEMS SURVEILLANCE APPLICABILITY Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.

OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.

SPECIFICATIONS a, b, c, dMg!~::m - Deleted

e. The Inservice Inspection program shall be reevaluated as required.

by 10 CFR 50, Section 50.55a(g)(5) ta cansider incarparatian of new inspectiaA techniques that have been praveA practical, and the caAclusians af the evaluatiaA shall be used as apprapriate ta update the iAspectian pragram.

f. Survei 11 ance af the regenerative heat exchanger and primary ca al aAt pump flywheels shall be performed as indicated in Table 4.3.2.

g. A surveillance program to monitor radiation induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.5.3 of the FSAR *

Amendment No . .§3., 3G, ~

4.3 SYSTEMS SURVEILLANCE (Continued)

h. Periodic leakage testing<a>. on each check valve 1isted in Table 4.3.1 shall be accomplished prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if such testing has not been accomplished within the previous 9 months, and prior to returning the check valves to service after maintenance, repair or replacement work is performed on the valves.

i. Whenever integrity of a pressure isolation valve listed in Table 4.3.1 cannot be demonstrated and credit is being taken for compliance with Specification 3.3.3.b, the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily and the position of the other closed valve located in that pressure line shall be recorded daily.
j. Following each use of the LPSI system for shutdown cooling, the reactor shall not be made critical until the LPSI check valves (CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed .
<a>To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
Reduced pressure testing is acceptable (see footnote 5, Table 4.3.1).
Minimum test differential pressure shall not be less than 150 psid .
Amendment No. , =J..!t:, -l3G
4.3 SYSTEMS SURVEILLANCE (Cont'd)
Basis
The inspection program specified places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems.< 1> In addition, that portion of the reactor vessel shell welds which will be subjected to a fast neutron dose sufficient to change ductility properties will be inspected. The inspections will rely primarily on ultrasonic methods utilizing up-to-date analyzing equipment and trained personnel. PreeperatieAal iAspectieAs ~1ill establish base ceAaitieAs by aetermi Ai Ag i Hai cati eAs that might ecc1:1r frem geemetri cal er metal h1rgi cal seijrces aAa frem aisceAtiAijities iA welameAts er plates which might caijse ijASije ceAcerA eA a pestservice iAspectieA. To the extent applicable, based upon the existing design and construction of the plant, the requirements of Section XI of the Code shall be complied with. Significant exceptions are detailed in the requests for relief which have received NRC approval and are contained in the Class 1, Class 2 and Class 3 Long-Term Inspection Plans.
Valve Testing To ensure the continued integrity of selected check valves which are relied upon to preclude a potential LOCA outside containment, special requirements for periodic leak tests are specified. In addition a valve disk position check for the LPSI check valves is specified following each use of th~ LPSI system for shutdown cooling. This position check ensures that the four LPSI
check valves have reclosed upon ce.ssation of shutdown cooling flow.
References (1) FSAR, Section 4.5.6 (2) Deleted (3) Systematic Evaluation Program Tqpic V-Il.A, NRC letter to the licensee transmitting the final topic evaluation dated November 9, 1981.
Amendment No~ -9, ~, -i-a.G, -142-
Footnote:
(a)
I. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
4. Leakage rates greater than 5.0 gpm are considered unacceptable.
5. Measured leakage rates must be adjusted for test pressures less than the maximum potential pressure differential across the valve by assuming leakage to be directly proportional to the pressure differential to the one-half power .
NRC Order Dated April 20, 1981 Amendment No. -93
TABLE 4.3.2
1.
EauipmeAt MiscellaAeous SurveillaAce Items RegeAerative Heat ExchaAger Method FrequeAcv
a. Primary Side Shell to 'lolltmetric § Year Maximum Tube Sheet Welds !Aterval ( 100%)
b. Primary Head Volumetric § Year Maximum IAterval (100%)
2. Primary CoolaAt Pump Volumetric 100% Upper Flywheel Flywheels Each RefueliAg
Amendment No. a-4, ::/-9, ~
4.5 CONTAINMENT TESTS Applicability Applies to coAtaiAmeAt leakage aAd structural iAtegrity.
Objective To verify that poteAtial leakage from the coAtaiAmeAt ai'ld the prestressiAg teAdoA loads are maiAtaiAed ~iithiA specified values.
Speci fi cati OAS 4.5.1 Integrated Leakage Rate Tests A surveillance test program for the containment overall integrated leakage rate shall meet the 10 CFR 50, Appendix J, Type A test requirements or approved exemptions .
ChaAge No. 16 Amendment No. 2-, 3-§.
4.5.2 Local Leak Detection Tests
a. Test Local leak rate tests shall be performed at a pressure of not less than 55 psig.
Local leak rate tests for checking air lock door seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed at a pressure of not less than 10 psig.
Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent.
The local leak rate shall be measured for each of the following components:
a) Containment penetrations that employ resilient seal gaskets, sealant compounds, or bellows.
b) Air lock and equipment door seals .
c) d)
e)
Fuel transfer tube.
Isolation valves on the testable fluid systems' lines penetrating the containment.
Other containment components which require leak repair in order to meet the acceptance criterion for any integrated leak rate test.
b. Acceptance Criteria The total leakage from all penetrations and isolation valves shall not exceed 0.60 La.
The leakage for an air lock door seal test shall not exceed 0.023 La.
c. Corrective Action If at any time it is determined that 0.60 La is exceeded, repairs shall be initiated immediately .
Amendment No . .}2-, ~, l.a-5
4.5 Containment Test (continued) 4.5.2 Local Leak Detection Test (continued)
If repairs are not completed and conformance to the acceptance criterion of 4.5.2ib(l) is not demonstrated with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the Plant shall be placed in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
2. If at any time it is determined that total containment leakage exceeds L8 , ~ithin one hour action shall be initiated to bring the Plant to hot shutdown within the next six (6) hours and cold shutdown within the following thirty (30) hours.
3. If air lock door seal leakage is greater than 0.023 La, repairs shall be initiated immediately to restore the door to less than specification 4.5.2.b(2). In the event repairs cannot be completed within 7 days, the Plant shall be brought to a hot shutdown condition within the next six (6) hours and cold shutdown within the following thirty (30) hours.
If air lock door seal leakage results in one {1) door causing total containment leakage to exceed 0.60 La, the door shall be declared inoperable and the remaining operable door shall be immediately locked closed and tested within four (4) hours.
As long as the remain i.ng door is found to. be operable, the provisions of 4.5.2.c(2) do not apply. Repairs shall be initiated immediately to establish conformance with specification 4.5.2.b(l). In the event conformance to this specification cannot be established within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the Plant shall be brough_t to a hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
d. Test Frequency
1. Individual penetrations and containment isolation valves shall be leak rate tested at a frequency of at least every six months prior to the first postoperational integrated leak rate test and at a frequency of at least every refueling thereafter, not exceeding a two-year interval, except as specified in (a) and (b) below:
a) The-containment equipment hatch and the fuel transfet tube shall be tested at each refueling shutdown or after each time used, if that be sooner *
ChaAge Ne. 7 Amendment No. l-, l2e
4.5 CONTAINMENT TESTS {continued) 4.5.2 Local Leak Detection Tests {continued) b) A full air lock penetration test shall be performed at six-month intervals. During the period between the six-month t_ests when containment integrity is required, a reduced. pressure test for the door sea*l s or a full air lock penetration test shall be performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after either each air l-0ck door ripening or the first of a series of openings.
2. Eath three months th~ isolation valv~s must be stroked to the position required to fulfill their safety function unless it is established that such operation is not practical during plant operation. The latter valves shall be full-stroked during each cold shutdown.
4.§.3 ReeireulatieR Heat Remeval. Systems
a.
Tull (1) The pertieR ef the shutdewR eeeliRg system that is eutside the eeRtaiRmeRt shall be tested either by use iR Rermal eperatieR er hydrestati eally tested at 255 psi g at the iRter.val speeified iR 6.15.
(2) Pi pi Ag frem valves C'/ 3029 aRd CV 3030 te the discharge ef the safety iRjectieR pumps aRd ceRtaiRmeRt spray pumps shall be hydrestatieally tested ~t Re less thaR 100 psig at the i Rterval speeifi ed i A 6 .15 .
Amendment No. +l, l-26
4.5 CONTAINMENT TESTS (Contd) 4.5.3 . RecirculatioR Heat Removal Svstems
i. Visual iRspectioR shall be made for excessive leakage from eompoRents ~f the system at the iRterval specified iR 6.15.
ARy sigRificant leakage shall be measured by collectioR aRd
~11eighiRg Of by aRother equivaleRt method.
b. AcceptaRce GriterioR The maximum allowable leakage from the recirculatioR heat l"emoval systems' compoReRts (which iRclude valve stems, flaRges aRd pump seals) shall not exceed Q. 2 gall Ofl per mi Fl Ute uF1der the F1ormal hydrostatic head from the.SIRW taflk (approximatqly 44 psig).
c. Corrective Action Repairs shall be made as l"equil"ed te maintain leakage withifl the acceptance criterien ef 4.5.3b.
4.5.4 Surveillance for.Prestressinq System lllllilllillilllllllllllllllll.111111111111~111.111111111111.ll:~J:il:::::::l:ii
a. Tendefl iFlspectioA shall be accomplished at five year ifltervals fel" the life of the plaflt. The scheduled iFlspectiofl dates fer all subsequeflt iAspectieAs may be varied by Aet mel"e thaA plus er miAus eAe year frem the base schedule.
b. The surveillaRce teAdeFls shall be rafldemly but represeF1tatively selected frem each ef the fellewiflg greups:
I. A miflimum ef 4 dame teAdeFls iflcludiAg eAe frem each dame teAdeA greup.
2. A miflimum ef 4 vertical teF1deF1s.
3. A miflimum ef 5 heap teF1doAs.
Fer each iAspectieA, the teAdeAs shall be selected efl a rafldem basis except that*these teAdeF1s whese reutiAg has beeA medified te clear peF1etratieAs shall be excluded frem the sample.
c. Duri Fig each teF1deA i Flspecti eA, the fel 1m11iflg field testi Fig shall be performed:
1. Lift eff readiAgs shaJl be takeA fer each ef the surveillaAce teF1deAs. The tests shall iflclude the felleWiAg actiofls:
(a) 0Re teRdeR, raF1domly selected from each group ef teF1deF1s duriAg each iFlspectiofl, shall be subjected to esseAtially complete deteRsiofliAg to ideAtify brokeA er damaged wires .
Amendment No. -14, a+, &6, +l, fG9.
4.5 CONTAINMENT TESTS (CoAtd) 4.5.4 System (CoAtd)
~
S urveillaAee ror ~ PrestressiAg
. - t o£ e,oAnatioA aA d J*aekiA"
( b) fepee du* i Ag !ete~ ~ l ::~all Y sp aeed 1evel s The simultaAeeus me~s uremeA- ~ft§ shallr be1 made ~ at efa miAimum feree of three approx1mai~A~ foree aAd zero.
betweeA the sea . d tate, eaeh wi*e iA the
'*,hi . ld "A the for deteASlOAe eoAtiAuity. s .
2. " le the*,,teAdoAb isle be enee*( d teAdeA 111" -- . 1 heep aAd aeme Th*** 11iPes, eAe f: ... omdeach aAd of ideAtifie~ ~As eetioA.from a vert1eafe*, '.ereeted At each.
3. !.
teAdeA '!ill be**"'i1 ;Aee, the Iii r!s 1111 b: uiPes *emeved '" 11 sueeess1ve surve1 Each of the lAspee io sioA or other be visuallyte~deAs.ted diffe**At iAspec evideAee aAdfe*samples takeAef f""iabe*atery or . testiAg.
delete*ieus effeets ***sually fer eel**.aAd
4. The sheathi* d fi 11 er ss~:g Ag samples ~~ !~::~~::df;~ 1aboratory test1 . Ag.
eeverage aA- . lates, *tress1Ag
5. Te washers, aAc~~::g rosioA Ado A s 1 aAd buttoAheadsd or other shf e 1e !e=~ o~s e hardware sueh as bear1A!i:ually iAspeeted fer effects.
evi de Aee ef eer the fell e11i Ag 1aBeratery
d. . the field testiAg of 4.5.4e, Fell e1fl Ag 1 b deAe' . th
testiAg shale I
. th*ee iAspeet1eA
additieAal spee1meA the middle): ~A~ field visual
'11 be eut frem eaehdefeefrom Th*** teAsile. "\i,;s removed (e~e -*ema~i be eut f*em the t t spee1meAs sna f s~
iAspeft~he
,.;,. dete1"1111::t ef eerPesieA. Ea~~ id streAgth, aAd aeh eAd aA OA A te have the uire samples shall
2. The sheathiAg eaeh teAdeA f1~led . "e*tieal e*am1Ae
les shall teAdeA ~ampughly mi*ed Be tttere Ad aAd t
f~* *::*:::.:*;;i~Ble *;l**i~e~~ :!;:;;J~ft~e w~th,~~~
frem -tthe 1o***er eAd. Samp . *. t . , *at er eoAteAt, a "'d w l l<al l Al-)', "
t tes an-
Ralyzed eeA<!At*at1

Aalyses sul f1 des. HA "it hi A shall the Be pe* e*melimits speeified lff" acceptaAee procedure~
Code Seet10AaA~I**' Table IWL 2525 I.
4-30 Amendment No.J~,l~~~
October 28,
4.5 CONTAINMENT TESTS (Contd) 4.5.4 f r Prestressinq S1*stem (Contd) mi~i~1~e,;meved
d d to
11 aneee * "ei-s an P eedu*es sh~l be estaf :heathing fll e and amounts S**ve1 . bl"shed te has been a::".e that replaced **1*:~.:
uponthecomp e ef the inspeeti**
documented.
shall be as follows:
£or each type o~ ed e.
ncceptance criteria H
f 11 measurede~ t d0 n forCeS h ther minimum req Ulr--
(a) the measur;the d force predicted in no force, and 99% and 951 ef ted adjaeeflt te (b) the measure~
The te*d** '" fr)esb!:et::et~::o~!s!o:~an a ad 95% of the p*edieted fa*ees, a* .. " sample te*dens
. all the rema1n1n~
(c) the measured !~::e~S~nof the predicted force.
a.e **t less .,; af its
p*edi eted fa~ee :~e
. . tendon is less than 99 If measu*ed fa*ee '" ::!de* shall be eamplet:a~se af sueh a*
i: shall be made as t\:h:ake*.
and a dete*mrna ' eet; "e aeti ** shall The Comm1ss1on s a l ,' detensi oned easu*ed as I* additi **
4.5.4f.
2. . µ"1res Inspec t io~ . ha11 indicate no si§nifican t loss of section sitting.
by *****s*a* ** p
shall be testl:~mate Fa1l~*es eei mens eut frem i ~speet~:" 1~*: =e:hafl II. 18 kips 3.
~f ~he *~:~~=:ien Tensile st*e*gth. the Commission be tested f*:
of any on. test samples d nee with spec1 4.5.4f.
etified '" aeeer-a f ef sigflifieaflt 4.
Ten don. anchorage corrosion, hardware shal!t~:r ~:~eterious effects.
J:H'ttin§, cracks or
4-31 8
Amendment No.J~, 1 ~'1~~~
October 2 ,
4.§ CONTAINMENT TESTS (,CoRtd) 4.§.4 SurveillaRce for PrestressiRg Svstem (CoRtd)
f. If aRy elemeRt of the prestressiRg*system fails to meet the acceptaRce criteri a. . . ~-.f. . 4..~J.~..4$,.~ ..'- the reportiRg provi si OR of*
Speci fi cati OR 6.9.2'!'Q'lG'ili'i'i::s.~t\il'
.;.

:.:::::::::::.;-:::*:~:;-;:;.;:;~::~.:::::.;:;:-~:;:;.::~:-:::: shall apply .
4.§.§ ERd ARchorage CoRcrete SurveillaRce
a. ,A VT 1 visual examiRatioR shall be performed OR the eRd aRchorage coRcrete surface at the surveillaRce teRdoR aRchor poiRts for sigRs of crackiRg, popouts, spalliRg, or corrosioR. CoRcrete cracks haviRg widths greater thaR 0.010 shall be evaluated aRd documeAted.
b. The eAd aAchorage coRcrete surveillaRce* iAspectioR iRterval shall be the same as teAdoA surveillaAce iAterval.
c. AcceptaAce criteria I. Gracie widths shall be measured by us i Rg' optical comparators or wire feel er gauge. Mo't'emeRts shall be measured by us i Ag demouAtable mechaRical exteAsometers.
2. CoAcrete aAchorage areas are acceptable if RO coAcrete cracks are wider thaA 0.010 iRches aAd AO sigAs of Rew -0r progressive deterioratioR siRce the previous iRspectioR are fouAd .
3~ CoRcrete surface coRditioRs exceediRg those stated iR 4.§.§c.2 above shall be evaluated for the effect OR teRdoR aAd coRtaiRmeRt structural iRtegrity. The results of evaluatioR shall be iRcluded iR the fiRal surveillaRce report.
Containment Isolation Valv~s
a. The i sol at ion valves shall be demonstrated. operable by performance of a cycling test and verification of isolation time for auto isolation valves prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or powe~ circuit .
Amendment No. H, 6, -l-G9, l-28
4.5 CONTAINMENT TESTS (Continued)
Containment Isolation Valves (Continued)
b. Each isolation valVe.shall be demonstrated operable by verifying that on each containment isolation right channel or left channel test signal, applicable isolation valves actuate to their required position during cold shutdown or at least once per refueling cycle.
c. The isolation time of each. power operated or a~tomatic valve shall be determined to be within its limit as specified in Table 3.6.1 when tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.
4.§.7 Deleted 4.§.8 Dome DelamiAatioA SurveillaAce If, as a result of a prestressiAg system iAspectioA uAder SectioA 4.§.4, corrective reteAsioAiAg of five pereeAt (8) or more of the total AUmber of dome teAdoAs is Aecessary to restore their liftoff forces to ~iithiA the limits of SpecificatioA 4.§.4, a dome ~elami~atioA iAspectioA shall be performed withiA 90 days followiAg such co.rrective reteAsioAiAg. The results of this iAspectioA shall be reported to the NRG. . *
.Amendment No. -14, ~, -!G9, 28
4.5 CONTAINMENT TESTS (Cont'd)
Basis
The containment is designed for an accident pressure of 55 psig. ci>
While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a temperature of about 104°F. With these initial conditions, following a LOCA, the temperature of the steam-air mixture at the peak accident pressure of 55 psig is 283°F.
Prior to initial operation, the containment was strength-tested at 63 psig and then leak rate tested. The design objective of this preoperational leak rate test was established as 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 55 psig. This leakage rate is consistent with the construction of the containment,< 2 >
which is equipped with independent leak-testable penetrations and contains channels over all unaccessible containment liner welds, which were independently leak-tested during construction.
Accident analyses have been performed on the basis of a leakage rate of 0.1%
by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With this leakage rate and with a reactor power level of 2530 MWt, the potential public exposure would be below 10 CFR 100 guideline values in the event of the Maximum Hypothetical Accident. c3 >
The performance of a periodic integrated leak rate test duri.ng plant. 1i fe provides a current assessment of potential leakage from the containment in case of an accident that would\pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leak rate test is to be perfor~ed without preliminary leak detection surveys or leak repairs and containment isolation valves are to be closed in the normal manner.
This normal manner is a coincident two-of-four high radiation or two-of-four high containment pressure signals which will close all containment isolation valves not required for engineered safety features except the component cooling lines' valves which are closed by CHP only. The control system is designed on a two-channel (right *and left) concept with redundancy and physical seoaration. Each channel is capable of initiating containment isolation. t 4 >
*Amendment No. -l-G-9, -l-3--S
4.5 CONTAINMENT TESTS (Cont'd)
The Type A test requfrements including pretest test methods, test pressure,
acceptance criteria, and reporting requirements are in accordance with 10 CFR 50, Appendix J, requirements or approved exempti-0ns.
The frequency of the periodic integrated leak rate test is keyed to the refueling schedule for the reactor because thes~ tests can best be performed during refueling shutdowns. The specified frequency is as specified in 10 CFR Part 50, Appendix J which is based on three major considerations.
First is the low probability of leaks in.the liner because of (a) the test of the leak tightness of the welds durin~ erection; (b) conformance of the complete containment to a low leak rate at 55 psig during preoperational testing which in consistent with 0.1% leakage at design basis accident (OBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, .at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60L~) of the total leakage that is specified as accepta.~J..~.... fr.9.m... P.~.n~t.r..~.tJ.9.ri.~.... ~.r:!.~..J.~..QJ.~t.J. on valves. Third is the teAdoA
!:~ ~ :n~!illf~l!P,!nm:::::::§&ll!illlli:EltlD.liiirmP.¥:ili::s urve i 11 an ce Prag ram wh i ch prov i des
Amendment No. ~' ~' ~
4.5 CONTAINMENT TESTS (Cont'd) an important part, of the structural integrity of the containment is
maintained .
The basis for specificatton of a total leakage rate of 0.60 La from penetrations and isolation valves is specified to provide assurance that the integrated .leak rate would remain within the specified limits during the intervals between integrated leak rate tests. This value allows for possible deterioration iri the intervals between tests.
The basis for specification of an airlock door seal leakage rate of 0.023 La is to provide assurance that the fail-ure of a single airlock door will not result in the total containment leakage exceeding 0.6 La. The seven (7) day LCO specified for exceeding the ~irlock door leakage limit is acceptable since it requires that the total containment leakage limit is not exceeded.
The limitiRg leakage rates frem the shutdewR ceeli~g system are judgmeRt values based primarily eR assuriRg that'the cempeReRts ceuld eperate witheut mechaRical failure fer a peried eR the erder ef 200 days after a DBA. The test pressure (270 psig) achieved either by Rermal system eperatieR er by hydrestatically testiRg gives aR adequate margiR ever the highest pressure withiR the system after a DBA. Similarly, the hydrestatic test pressure fer the returR liRes frem the ceRtaiRmeRt te the shutdewR ceeltRg system (100 psig) gives aR5 adequate margiR ever the highest pressure withiR the liRes after a DBA. < > -
A shutdewR ceeliRg system leakage ef 1/§ gpm will limit eff site expesures due te leakage te. iRsigRificaRt levels relative te these calculated fer leakage directly frem the ceRtaiRmeRt iR the DBA. The eRgiReered safeguards ream veRtilatieR system is equipped with iselatieR valves which clese upeR a high radiatieR sigRal frem a lecal radiatieR detecter. These meRiters shall be set - at 2. 2 x l O~ cpm, which is well be.l ew the expected 1evel , fell e'1'fi Rg a less ef ceelaRt accideRt (LOCA), eveR witheut clad failure. The l/S gpm leak rate is sufficieRtly high te permit prempt detectieR aRd te alle~i fer reaseRable leakage threugh the pump seals aRd valve packiRgs, aRd yet small eReugh te be readily haRdled by th_e sumps aRd radieactive waste system.
Leakage te the eRgiReered safeguards ream sumps will be returRed te the ceRtaiRmeRt cleaR water receiver fellewiRg a LOCA, via the equipmeRt draiR taRk aRd pumps. AdditieRal makeup wateia te the centainment sump inventery caR be readily accemmedated vi a the charging pumps frem either the SIRW tank er the cenceRtrated beric acid sterage tanks.
Amendment No . .J-2., 3-l, -l--99, .J--2.e
4.5 CONTAINMENT TESTS (Cont'd)
In case ef failure te meet the acceptance criteria fer leakage frem the shutdewn ceeling system er the penetratiens, it may be pessible te effect repairs within a shert time. If se, it is censidered unnecessary and unjustified te shut down the reactor. The times allowed for repairs are consistent 1:ith the items developed for other engineered safety feature components.
A reduction in prestr~ssing force and change i~ physical conditions are expected fer the prestressing system. Allowances have been made in the reacter building design for the reduction and changes. The inspection results fer each tendon inspected shall be recorded en the forms provided fer that purpose and comparison will be made with previous test results and the initial ~uality central reeerds.
Foree time records will be established and maintained for each ef the tenden greups, dome, hoop and vertical. If the force measured for a tendon is less than the lower bound curve of the force time graph, two adjacent tendoAs will be tested. If either of the adjacent or more than one of the original sample populatien falls below the lower bound of the fared time graph, an iAvestigatien will be conducted before the next scheduled surveillance. The investigation shall be made to determine 1:hether the rate of force reduction is indeed occurring fer eth6r tendons. If the rate of reduction is confirmed, the investigation shall be extended so as to identify the cause ef the rate ef force reduction. The extension of the investigation shall determine the needed changes in the surveillance inspection schedule and the criteria aAd initial planning for corrective action .
If the force measured for a tendon at any time exceeds the upper bound curve of the band on the force time graph, an investigation shall be made to determine the cause.
If the comparison of corrosion conditions, including chemical tests of the corrosion protection material, inditate a larger than expected change in the conditions from the time of installatioA or last surveillance inspection, and investigation shall be made to detect and correct the causes.< 6 >
Amendment No. !4, -lG9
4.5 CONTAINMENT.TESTS (Cont'd)
The 13resb*essi Ag system is a Aecessary streAgth. el emeflt ef the 131 aflt
safegliards afld it is ceflsidered desirable te ceflfirm that the allewaflces are Aet beiflg exceeded. Th* techAiqije ehesefl fer SlirveillaAce is based lii;>efl the rate ef ehaAge ef feree aAd 13hysical eeflditiefls se that the SijrveillaAce eafl eithef' eeAfirm that the alleuaAees are Sijffieieflt, er reqijire maiAteflaflce
. before miAimijm levels of force or 13hysical ceAditiefls are reached.
The eAd aAcherage ceAcrete is Reeded to maiAtaifl the 13restressiflg forees.
The desigA iAvestiga.tiefls ceAClijded that the desigfl is adeqijate. The 13restl"essiAg seqijeflce has shewfl that the efld aAchorage eeAcrete eaA withstaAd leads iA excess ef these which resijlt whefl the teAdoAs are aAchored. J'.t the time ef iflitial 13ressijre testiflg, thi ceAtaiAmeflt bijildiflg had beeA Sijbjeeted
. to tem13eratlire gradieAts eqijivaleAt to those fer Aermal operatiAg eoAditioAs
'~hile the prestressiflg teAdefl loads are at their maximijm.
However, after the iAitial pressijre test both concrete ereep and prestressing losses increase with the greatest rapidity aAd resijlt in a redistribijtioR of the stresses aAd a redijctieA in eAd aAchor force. Becaijse ef the impertanee of the coAtainmeAt aAd the fact that the design was Aew, it was co.Asidered 13rijdeAt to coAtiflije the SijrveillaRce after the iRitial period. <7> *
Amendment No. !99
4.5 CONTAINMENT TESTS (Cont'd)
Cantainment dame delaminatiaA inspectians performed in 1979 and 1982 have canfirmed that no concrete delamination has occurred. The possibility that delamination might occur in the future is ~emote because dome tendon prestress forces gradually diminish thraugh normal tendon relaxation and concrete strength normally increases over time. To accaunt for this remote possibility, howe*t'er, an additional delamination inspection Hill be performed in the event that §% or more of the installed tendons must be retensioned ta compensate for excessive loss of prestress. This inspection woul~ be to confirm that any systematic excessive prestress loss dfd not result from delamination and that the retensianing process did not result in delamination.
References
1. FSAR, Section §.1.2; Updated FSAR Section 5.8.2.
2. FSAR, Section §.1.8; Updated FSAR Section 5.8.8
3. FSAR and Updated FSAR 14.22
4. FSAR, Section 8.5.4; Updated FSAR Section 8.5.1.2
(§) FSAR and Updated FSAR Section 6i2.3 (6) FSAR, Section §.1.8.4; FSAR, Amendment No. 14, Question §.37; and Updated FSAR Section §.8.8.3 .
(7) Updated FSAR, SectiaA §,8.8.6 s5:~:]:t:::1:2:i1:;1:::::::Riif!I1119lI::leei:n:9:!::1::tg[:i,:
Amendment No. , .f-0-9
4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS
4.6.1 Surveillance Requirements Safety Injection System
a. System tests shall be performed at each reactor refueling interval.
A test safety injection signal will be applied to initiate operation of the system. The safety injection and shutdown cooling system pump motors may be de-energized for this test. The system will be considered satisfactory if control board indication and visual observations indicate that all components have received the safety injection signal in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel).
4.6.2 Containment Spray System
a. System test shall be performed at each reactor refueling interval.
The test shall be performed with the isolation valves.in the spray supply lines at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation.
13. At 1east evel"y five yeal"s the spl"ay rrnzzl es sh al 1 l:>e vel"ifi ed te l:>e
~
e§. The test will be considered satisfactory if visual observations
4.6.3 Pumps indicate all components have operated satisfactorily .
a . . The safety injection pumps, shutdown cooling pumps, and containment spray pumps shall be started at intervals not to exceed three months. Alternate manual starting between control room console and the local breaker shall be practiced in the test program.
b. Acceptable levels of performance shall be that the pumps start, reach their rated heads on recirculation flow, and operate for at least fifteen minutes.
4.6.4 Valves
a. Each Safety Injection Tank flow path shall be verif1ed OPERABLE within 7 days prior to each reactor startup by verifying each motor operated isolation valve is open by observing valve position indication and valve itself, and locking open the associated circuit breakers.
b. The Low Pressure Safety Injection flow path shall be verified OPERABLE within 7 days prior to each reactor startup by verifying flow control valve CV-3006 is open, and its air supply is isolated .
Amendment No . .§+, Flt, %, H-7, +/-3-1-, ~
4. 6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS
Surveillance Requirements Valves c.
(continued)
{continued)
The safety injection recirculation path shall be verified OPERABLE withi*n 7 days prior to each reactor startup by verifying va 1ves CV-3027 and 3056 are open and their switches HS-3027A, HS-3027B, HS~3056A, and HS-3056B are open.
d. Each Containment Spray Valve manual control shall be verified to be OPERABLE at least once each refueling by cycling each valve from the control room while observing valve operation at least each 18 months.
4.6.5 Containment Air Cooling Svstem
a. Emergency mode automatic valve and fan operation will be checked.
for OPERABILITY during each refueling shutdown.
b. Each fan and valve required to function during ace i dent cond-i.t ions will be exercised at intervals not to exceed three months .
Amendment No. &9, =Rt, +:!-, -H+, ~
4.6 SAFETY INJECTION AND CONTAINMENT SPRAY... SYSTEMS TESTS
The safety injection system and the containment spray system are principal plant safety features that are normally inoperative during reactor operation.
Complete systems tests cannot be performed. when the reactor is operating because a safety injection signal causes containment isolation and a containment spray system test requires the system to be temporarily disabled.
The method of assuring OPERABILITY of these systems is therefore, to combine systems tests to be performed during* annual plant shutdowns, with_ more frequent component tests, which can be performed during reactor operation.
The arurnal ii:!ii:J:]::nil::i:Rf~iii:ll systems tests demonstrate proper automatic operation of the safety rnJection and containment spray systems. A test signal is applied to initiate automatic action and verificati-0n made that the components receive the Safety Injection Signal in the proper sequence. The test demonstrates the1 ooeration of the valves 1 pump circuit breakers, and automatic circuitry. <
2>
During reactor operation, the instrumentation which is depended on to initiate safety injection and containment spray is generally checked El-a+l-y jj[ijij!i~!:jill:f:lt: and the initiating circuits are tested monthly. In addition, the
a'Cffv.e""".'C"ofriponents (pumps and valves) are to be tested every three months to check the operation of the starting ci.rcuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judgment that more frequent testing would not significantly increase the reliability (ie, the probability that the component would operate when required), yet more frequent test would result in increased wear over a long period of time. '.'er.ificatioR that the spray pipiRg aRd Rozzles are opeR will be made iRitially by a smoke test or other sliitably seRsitive method, aRd at least *e*1ery fh'e years thereafter. Si Ree the material is all staiRlcss I
.stE;el, Rormally iR a dry coRditioR, ~Rd ~Jith RO plliggiRg mechal'tism available, the retest every five years is coRsidered to be more thaR adeql-Jate.
Other systems that are also important to the emerg~ncy cooling function are the SI tanks, the component cooling system, the service water system and the containment air ~oolers. The SI tahks are a passive safety feature. In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance.
References (1) FSAR, Section 6.1.3.
(2) FSAR, Section 6.2.3. *
Amendment N.o. H-7-, ~, 162-
4.14.1 Each steam.geAerator shall be 9emo~strate9 OPERABLE by performaAce of 4.14.2 The steam {eAerator:$.~ tube mini mum samp 1e size, inspection result
~p:~~ !aa i ~ 0 ~ ab ~d~*~*~~ ~ 0i,:i:~:~:ii:I:~ i nrh:cz ~ ~~r~{~~ i ~~~P~g~ ~ bA b~f as
}f f steam 9eAerator tubes shall ....tfo.... performed at the frequeAci es i;r1~r;~~~~~*~~g~1::~r=:1~~
4 i~;*i!~~~~:;~~;~~!:~~~~~~i!~i~~:i;"h* .
inc 1ude a.t.. ..l east 3% of the tot a1 number of tubes in a11 -s-team geAeratorl§s; the tubes ~elected for these inspections shall be selected on a random basis except:
Where experience in similar plants with similar water chem-istry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.
The first sample of tubes selected for each inservice inspection (sub.~.,equeAt to the preservice iAspectioA) of each steam geAerator~~ shall include:
j[),![ All nonplugged tubes that previously had detectable wall penetrations greater than 20%.
Tubes in those areas where experience has indicated potential problems.
A tube inspection (pursuaAt to S~ecificatioA 4.14.5.a.8) shall be performed on each selec ed tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
The tubes selected as the ....s.e.c.o.nd. and third samples (if
~~~~!~i1o~Ym!~b6~ :u~Je~tla'i~t::~)P~~il~r i~5~ ~g~~~~ir~n provided: *
-I--. i:}:j! The tubes se 1ected for these samp 1es inc 1ude the tubes from those areas of the tube sheet array where tubes with imperfections . ~*m.r~. ~.,.,pr,gy.,.jously found.
Amendment No. H, 4- GG &:smmt~t:P.:iti:u H;"""":fb:.;""""~*;**"""~, 4§, ~' 9+, %, -H-2-, -H-2-, .J.41.
4.14
AUGMENTED INSER'lIGE .INSPECTION PROGRAM FOR STEAM GENERATORS (GeAt' d).
1;~::1:~:~:::;:::~:M:Irn111:11.::::::1~:~=~~i:~:~ttt:::mu~~:J$.~i~i:~:m~::~~~:~::::::~:~~s.~:~m:::::::1i:s~~:1::n~~~:1::
~!!itfi!!!il,1).i The insp.ect i ans inc.l ude those port i.ons of the tubes where imperfections were previously found.
The results of each sample inspection shall be one of the following three categories:
cl~ssified into Category Inspection Results*
C-1 less than 5% of the total tubes inspected are degraded tubes and none of* the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total .tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Ntite: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
4.14.4 :§!~! Inspection Freguenci es
The above required i nservi ce inspection of steam geAerater!S.i tubes sha 11 be performed at the fa 11 owing frequencies:
The first iAservice iAspectieA shall be perfermed after 6 Effective Fl:!ll Pewer MeAths bl:!t withiA 24 caleAdar meAths ef iAal:!gl:!ral criticality fer the steam geAeraters. Sl:!bseql:!eAt inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not-including the preservice inspecti-0n, result in all inspections results falling inta the*
C-.l category or if two consecutive inspect i ans demonstrate that previously observed degradation has not ~ontinued and no additional degradatian*has occurred, the inspection interval may be extended to a maximum of once per 40 months.
If the results of the inservice inspection of a steam geAerater~~ conducted in accordance with Table 4.14 2 ;:~:!fiillif,::1 at 40 montfi i nterva1s fa 11 into Category C-3, the i nsp*e*ctlari"**
frequency shall be increased to at least once per 20 months .
Amendment No. 3-9, %, , m, -H-2-, .J.a-2., -141-
i::iili:~mm:::I::rn::i~i:il.~:~i=~~1~;:~:§~:'1~~~1~::~::~~i~~:;J.::~:~n.~:;~:~:~ii~~~;:;:lt~:;§~;:1::;i;1:;ATO RS (GOH t Id)
The increase in inspection frequency shall apply until the subsequen.t. ..tn.~.R~.f.t.i ons satisfy the criteria of Speci fi cation
!t l:~~~ap!l'i'~:~:!:g:~h!~e interval may then be extended to a maximum Additional, unscheduJ..~d ins*ervice inspections shall be performed on
~ach st~am geHe~a~or~g in accordance w..tJ.h.,. . ,t~,e fi ~st samp 1e rnspect1on spec1f1ed rn Table 4.14 2 §MIMS.Ml: durrng the shutdown subsequent to any of the following cohCffffoHs:
h'.1~11 Primary-to-secondary tube 1eaks (not including 1eaks
..,. . . . originating from tube-to-tube s.heet welds) in excess of the limits of Specification 3.1.5.
.i}! A seismic occurrence greater than the Operating Basis
Earthquake.
a.I~ A loss-of-coolant accident resulting in initiation of flow of
the engineered safeguards.
4-.i:Jj A main steam 1i ne or main feedwater line break.
4.14.§ §;,\ Acceptance Criteria
~ I* As used in this Specificationl
Imperfection means an exception to the dimensions, finish or contour of a tube from that requir~d fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal' tube wall thickness, if detectable, may be considered as imperfections.
Degradation means a service-induted cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused. by degradation.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
Plugging Limit means the imperfecti.on depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness .
Amendment No. 3-9, , Me, -H-2-, 3-2-, t4l
7. Unserviceable described the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line
.Q.r......f.~rn.~Iw..~ter line break as specified in 4.14.4.c sM§@~M::t:i;:, above.
Tube Inspection means an inspection of the steam geAerator§@ tube from the point of entry (hot leg side) compl ete1r***around the U-bend to the top support of the cold leg.
Preservice Inspection means an inspectj..Qn of the full length of each tube in steam geAeratorS.~ performed by eddy current techniques prior to servi"C. e to establish a baseline condition of the tubing. This inspection shall be performed after the shop hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections ..
b!iilj[i[i[iiMiil* The steam geAeratorSG sha 11 be determined OPERABLE after completin~ the corr*e. sponding actions (plug all tubes exceeding
~~~ck!)9~~~~i~!~i~Y aT~bi!l 4~~~e2 it:i:~:!,i:~[~ng through-wall i::~:i:i:1:l:::::::::::::::::::::::::a~1:lil:i~:~=~~~:~:~~:lm~~~:::::::~~~~~:~:1::~=:~ry~:~::::::~~:1~~~
a.
m~~~iH:d~:~m::;mri~~~*~mm~i:=:~~:t::~:~~.;"
b.
m~i:~;~;~;i;~~~m::~~~M~i~**t
-~
I. Number and extent of tubes inspected.
2. Location and percent of wall-thickness penetration for each indication of an imperfection.
3. Identification of tubes plugged
c. Results of steam generator tube inspections that fall into Category C-3 shall require 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal notification to the NRC prior to resumption of plant operation. A written followup within the next 30 days shall provide a description of investigations and corrective measures tak.~.n ... t.9. ... P.r.~.vgnt recurrence.
4----69 :~m~::e::i1::I9:f::1:~:
Amendment No. 3-9, ~' -i-96, 1-2-, 3-2-, -l4l
Table 4.14 I MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSER'JICE INSPECTION Table NotatioA:
I. nie iAsel"vice iAspectioA may be 1 imitee to oAe steam geAeratol" oA a l"otatiAg scheeule eAcompassiAg 6% of the tubes if the results of previous iAspectioAs iAeicate that all steam.
geAerators al"e performiAg iA a like maAAer. Note that uAeer some circumstaAces, the ~perat1Ag coAeitioAs iA oAe or mol"e steam geAerators may be fouAe to be mol"e severe thaA those lA othel" steam geAerators. UAclel" such circumstaAces the sample sequeAce shall be moeifiee to iAspect the most severe coAeitioAs.
4.69a Amendment No. 141
lST SAMPLE INSPECTION TABLE 4.1~ lfllll STEAM GENERATOR TUBI""TNSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G.
C-2 Plug defective tubes C-1 None N/A N/A and inspect additional 2S tubes in this S.G. C-2 Plug defective tubes C-1 None and inspect additional 4S tubes in this S.G. C-2 Plug defective tub~s c:..3 Perform action for C-3 result of first Sample C-3 Perform action for C-3 result of first N/A N/A Sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug de- S.G.s are fective tubes and C-1 inspect 2S tubes in each other S.G. Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second additional sample 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal S.G. are notification to NRC C-3 with written follow up within next Additional Inspect all tubes 30 days S.G. is each S.G. and plug C-3 defective tubes. N/A N/A S - 3 ~-""""~~~--*""JAK>e-ic.reµ.. . .flN. . . . +i.....
s --ft~t:t,p.e--1fl<H-UH!iffi1Hb~er-o""'f~s~tH"e'"aff!.m--1g*e>Rfl,P.er""aHt'.f'lo-ic.r.....
s -'li~fl-t'HA'*e>---H-Ut+fl1Ht....,.,-THaflFH'd~Fl-'1µ's.......+t~t:te-AfffUffflmbAfe"""r"""'-f+of-....s~tp.;ea~mll--H-ge~A'*e~r'""atHO*r-s 0
iAspected duriAg aA iAspectioA S = 6/n % Where n is the number of steam generators inspected during an inspection Amendment No. ~
4.16 INSERVICE INSPECTION PROGRAM FOR SHOCK SUPPRESSORS CSnubbersl Applicability Applies to periodic surveillance of safety-related snubbers as described per Specification 3.20.
4.16.1 Specifications Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspe.~.!..t.9..!J program in addition to the requirements of Specification 4.9.S§@§M* As used in this specification, "type of snubber" shill***fuean snubbers of the same design and manufacturer, irrespective of capacity.
a. Visual Inspection Snubbers are categorized as inaccessible or. accessible during reactor operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the following paragraph:
If one or more unacceptable snubbers are found, the next inspection interval shall be 2/3 (-25%) of the previous interval. If no unacceptable snubbers are found, the next interval may be doubled
(-25%), but not to exceed *4s months. The interval extensi-0n
provisions of Technical Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months .
Inspections performed before the interval has elapsed may be used as a new reference point to determine the next inspection.
However, the results of such early inspections, performed before the original required time interval has elapsed (nominal time less 25%), may not be used to lengthen the required inspection interval.
Any inspection whose reiults require a shorter inspection interval will override the previous sch~dule.
b. Visual Inspection Acceptance Criteria Visual inspection shall verify that (1) the snubber has no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional, and (3) fasteners for the attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of v1sual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers, irrespective of type, that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and
. determined OPERABLE per Technical Specification 4.16.ld or 4.16.le, as applicable. All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable
for determining the next inspection interval .
4-71 Amendment No. 3-, 69, -!-9-7-, HS, -l--64
4.16. INSERVICE INSPECTION PROGRAM FOR SHOCK SUPPRESSORS (Snubbers}
f. Snubber Service Life Monitoring 4.16.1 A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by S~ecificatioA 6.10.2.1.
Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each safety related snubber in use in the plant shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records .
4-74 Amendment No. ~' e9, 93-, -l-G-7,_ 6-4
6.1 RESPONSIBILITY 6.1.1 The Pl aRt GeReral MaRager @J.]~lij~!ili.Ji@i.:f.iUl::M~n~i.:n~ sha 11 be responsible for
overall pl ant operati*on ana*****shaTr***aeTe§.afe******lh. . .*wri ting the succession for this responsibility during his absence.
~
6.1. 2 The Shift Supervisor or iR his abseRce from the coRtrol room, the secoRd liceRsed seRior operator OR duty shall be respoRsible for the shift commaRd fuRcti OR. A di recti\*e to this effeg:t.,.,.~,h..~JJ.. . J?.H.,.,.,.1,§.,§.HR~t.,.,.~.r,\.l'.1.,~.,9.llY 6.2 ORGANIZATION 6.2.1 0 FFS I Tf AND ON SI Tf ORGANIZATION s1::::~~:~:~j:~~j::::::~~g::::::~tf~~i]!:~!!:::::~!=f:R@hm~:~:~:~::§.~j~j Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities
affecting the safety of the Palisades Plant .
a. Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented, and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, .I I
and job descri pt i ens for key posit i.Q.D..~. .,. . . . .QX. . ....i..n. . . gq_y.J.y~J..gnt....forms of
b. --The Pl aRt GeReral MaRager P.tJit~ni:[Jl$.Q.ptgrJ}tutti~p~jgt{J,iEsha 11 be responsible for over a11 pl ant safe operaffCfrl"***and"***shaTr*ffa'ie* control over those onsite activities necessary for safe operation and maintenance of the pl ant.
c.
~~i:P.~i:i:~i p~~: ~ ~ e~ ve ~ ~ ~~ ~ ~= t~P ~= ~~ ~ ~ ~ ~ b'~:::';'~i,j'~'i~i~i~~p;~:fln t hUcTe*a:*r-.. . .fafety and sha 11 take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.
d. The individuals.who,,,,,,t,r,~,Jn,,,,,t,Q,,W:,,,,,,RP.~X:~,ting staf~ and those who car~y*
out heal th physics Jlij~fa::ij=t,::u:ult'$'!li~:ty; and qua 11 ty assurance funct i ens may report to the a"fl"pFO"tl"rTafe. . . . O"ffs. ffe manager; however, they sha 11
have sufficient organizational freedom to ensure their independence from operating pressures.
6-1 Amendment No. 3-2-, ~' W, +§., 98, -l-3-9
ADMINISTRATIVE CONTROLS 6.2.2 PLANT STAFF
a. Each OR duty ~hift shall be composed of at least the miRimum shift
b. ~!Ri~:ftrg~: !ic~~ye~i~g~i~=r~~9c~g=d?~r:~i 0 :t~~~Tih:~ ~~1lhe
~=uf g:~~ :=e ry t~g~ !:~. R!:ct:~i 6ig=af :rtg~s S~~n~r R~~~t:~r Ogg~*:t:~r' shall be preseRt at the coRtrols at all times wheR fuel is iR the e
d* All core alteratioRs after the iRitial fuel loadiRg, shall either
e. (Deleted)
~ ed In the event that overtime is used, the following guidelines shall be followed:
Ail* An individual .shal,l~ljj;g1!;~ r:iot be,Qermitted to \'.'Ork more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straightm f'excludrng shift turnover timet--:;:
An indivi<;lual -sfla+l-~ljp:g)J::~ nqt be permitted to work more.than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months rn any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . . *peri od nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1n,,,jl.P.Y 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months geri od, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seveR day:ti\dfay periodM fall excluding shift turnover timeN:t; =*=*=*=*=*=*=*=*=*=*=*=*=*=*=....*=
A break of at least ~i hours -sfla-1...iJ!hBU:lM be a11 owed between work peri ods;i,: ti ntl udi ng shi n=*=*=*turhhver t ime-H:~::
The RadiatioR Safety TechRiciaR may be abseRt for a period of time Rot to exceed two hours iR order to accommodate unexpected abseRce provided immediate actioR is takeR to restore the miRimum requiremeRts.
6-2 Amendment No. ~, +-&, MS, .f.a.9., ~
ADMINISTRATIVE CONTROLS 6.2.2.fi:::::: PLANT STAFF (Continued)
Gi. Except during extended shutdown periodsl the use of overtime should be considered on an individual basis and not for the entire staff on a shift .
6.2.3 6-2a Amendment No. 39
ADMINISTRATIVE CONTROLS 6.3 PLANT STAFF QUALIFICATIONS
6.3.1 6.3.2 Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI NlB.1-1971 for comparable positions.
~e:::!:i!ciR b;ait!!Y p~:~:g~:11:~i!,iRlf!l~:1~:::::::1f=lil,!i!f=t!!!j!f ~~a!!c:d qualifications of a Radiation Protectjon Manager as defined in the Regulatory Guide 1.8, September 1975. The RadiatioR Safety MaRager shall have direct access to PlaRt GeReral MaRager iR the matters of 6.3.3 The Shift Technical Advisor shall have a bachelor's degree or equivalent and the Shift Engineer shall have a bachelor's degree in a sci~ntific or engineering discipline. Specific training for both the Shift Technical Advisor and the Shift Engineer shall include plant design, operations, and response and analysis of the plant for transients and accidents.
The Shift Engineer shall hold a Senior Reactor Operator License.
!~:c ~}:~: ~ ~f~~~. ~Ri~~:~i:j:ii:l:ilg:iilifmillll~::*1:1;11IJ.:~:~:i:;i:l:i~:i:ii: ~~e~!:~s 6.3.4 or exceed the mini m*U"m***qualTffC.afio"ri"if ..Cif""ANS"***3**;T.::rgar;* .. *sectlc>>n 4. 7.1 and 4.7.2. A Senior Reactor Operator License or certification shall be
6.3.§ considered equivalent to a bachelors degree for the purpose of this specification for RO more thaR two iRdividuals.
Either the PlaRt OperatioRs MaRager or the PlaRt OperatioRs .
SuperiRteRdeRt will hold aR SRO liceRse aRd meet the other requiremeRts of SectioR 6.3.1 of these TechRical SpecificatioRs (as applicable to the OperatioRs MaRager iA ANSI Nl8.l). The iRdividual holdiRg the SRO liceRse shall be respoRsible for directiRg the activities of the liceRsed operators.
For the purpose of this sectioA, "EquivaleAt," as utilized iA Regulatory Guide 1.8 for the bachelor's degree requiremeRt, may be met with four years of aRy oRe or combiAatioR of the followiRg: (a) Formal schooliAg iR scieRce or eRgiReeriRg, or (b) operatioRal or techRical experieRce/traiRiRg iR Ruclear power.
Amendment No. 3+, %, -l-GS, ~
ADMINISTRATIVE CONTROLS
Table 6.2 1 MINIMUM SHIFT CREW COMPOS:IJION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION Po*111er Operati OR, Hot Cold Stn.1tdm:n and StaAdby aAd Hat ShutdewA RefueliAg ShutdewA*
SS Shift Super.vi ser \'d th a Se Ai er Reactor Operators Li ceAse SE Shift EAgi~eer with a SeAier Reactor Operators LiceAse SRO IAdividual with a SeRier Reactor Operators LiceRsc
RO
.~o STA Auxiliary Operator Shift TechAical Ad~isor Except for the Shift Supervisor*, the Shift Crew CompesitieA may be oAe less thaA the miAimum rcquiremcAts of Table 6.2 1 for a period of time Aot to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months iA order to accommodate uAexpected abseAce ef eA duty shift crew members provided immediate actieA is takeA to restore the Shift Crew CompositioA to withiA the miAimum rcquiremeAts of Table 6.2 1. This previsioA dees Aet permit aAy shift crew positioR to be uAmaAAed upoA shift chaAge due ta aA oAcomiAg shift crmJmaA beiRg.late or abseAt.
Docs Rot tRcludc additioRal persoRAel required ~~CR qore altcratioAs are bciRg coRducted. Sec SectieA 6.2.2.d.
There shall be twe iAdivtduals with SeAier Reactor Operator LiceAses OR shift. If either SRO OR shift satisfies the.Shift ERgiReer qualifieatioR requiremeAts, theA the STA dees Aet Reed ta be statioAed.
6-4
Amendment No. -!-&, f:rl-, !GS
ADMINISTRATIVE CONTROLS
6.4 6.4.2 6.§ (Deleted)
(Deleted)
R[VI[W AND AUDIT 6.§.1 PLANT R[VI[W COMMITT(( (PRC)
6.§.1.1 FUNCTION The PlaRt Review Gemmittee (PRC) shall fuRctioR to advise the PlaRt GeAeral MaAager OR all matters related to Auel.ear safety.
6.§.1.2 COMPOSITION The PRC is composed of AiAe regular members. The qualificatioA level fer PRC members shall be at least equivaleAt to those described iR SectioR 4.4 of ANSI Nl8.l 1971. The PRC shall iRclude represeAtatives from the OperatioAs, Radiological Services, MaiAteRaRce aRd [AgiAeeriAg DepartmeAts, The ChairmaR, AlterAate ChairmeA, aAd members shall be desigRated iA admiAistrative procedures by the PlaRt GeReral MaRager~
6.§.1.3 ALT[RNAT[S
,O.lterRate members of the PRC sh al 1 be appoi Rted i A wri tiAg by the PRC
ChairmaA to serve OR a temporary basis. No more thaR two alterAates shall participate as votiAg members at aRy oRe time iR PRC activities.
6-5
Amendment No. -l-6, %, +s, -I-GS, 2-7-, 39, -!-%, §.2., l+G
ADMINISTRATIVE CONTROLS
6.5.1.4
~.5.1.5 MffTING FRfOUfNCY The PRC shall meet at least oAee per ealeAdar moAth with special meetiAgs as required.
QUORUM A quorum of the PRC shall eeAsist of the ChairmaA er alterAate aAd four
.memBers er al terAates.
6.5.1.6 RfSPONSIBILITifS The PRC shall Be respeAsiBle fer Auelear safety review of:
a. Ali procedures aAd programs specified BY SpecificatioA 6.8 aAd.
ehaAges thereto, *aAd aAy other procedures er ehaAges thereto as determiAed BY the PlaAt GeAcral MaAagcr to affect AUelcar safety; all 13ropescd tests er exi:ieri me Ats that affect Auel car safety; all pre13escd ehaAgcs or modifieatioAs to 13laAt systems or cquipmcAt that affect Auelear safety; aAd the Site fmergeAey PlaA.
b. All 13re13osed ehaAges to 013eratiAg LieeAse aAd TcehAieal SpccifieatioAs. *
e. Result's of iAvcstigatioAs of all violatioAs of the TeehAical S13eeifieati0As. (A re13ert shall he 13re13arcd eeveriAg evaluatteA aAd rceommcAdatioAs to 13revcAt reeurrcAec aAd Be forwarded to the
El.
e.
Vice PresidcAt NOD aAd to the Directer, Nuclear PerformaAce AssessmcAt Dc13artmcAt (NPADJ.)
PlaAt opcratioAs to detect 13otcAtial safety hazards.
Rc13orts of speei al rcvi ews aAd i Avcsti gati OAS as .requested BY the Pla~t GcAeral MaAagcr er NPAD. *
f. All re13ortable eveAts as dcfiAcd iA SectioA 6.9.2.
g. All items ideAtified UAder S13eeifieatioA 6.5.3.4 as sigAificaAt to A~clear safety. :
h. MoAthly rc13orts from PlaAt Safety aAd LiccAsiAg.
i. Nuclear iAdustry e13cratiAg cxpcricAce.
PRC rcvi cw of the aBove items may be performed BY routiAg, suBjcet to the requiremcAts of S13ecificatioA 6.5.1.7. PRC may delegate review of Item a. to PlaAt Safety aAd LiceAsiAg as dcscriBcd iA S13ccificatioA 6.5.3..
6-5a
Amendment No. -!-GS, -l-2-7, -146-
ADMINISTRATIVE CONTROLS
6.§.1.7 AUTHORITY The PRC shall:
a. Recom~eRd iR writiRg to the PlaRt GeReral MaRager approval er disapproval of items coRsidered uRder SpecificatioRs 6.§.1.6.a.
through i. above. -
b. ReRder determiRatioRs iR writiRg with regard to whether or Rot each item ceRsidered l:IRder SpeeifieatioRs 6.§.1.6.a, b, e aRd g above eoRstitutes aR uRreviewed safety questioR.
c. Provide writteR RotificatioR withiR 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice PresideRt Nuclear OperatioRs aRd to the Nuclear PerfermaRce Ass~ssmeRt DepartmeRt of aRy disagreemeRts betweeR the PRC aRd the PlaRt GeReral MaRager; however, the PlaRt GeReral MaRager shall have respoRsibility for the resolutioR of such disagreemeRts purs1:1aRt ta SpecificatioR 6.1.1 above.
The PRC Chai rmaR may .recommeRd to the Pl aRt GeReral MaRager appro*1al of these items ideRtified iR SpecificatioR 6.§.1.6 above based OR a routiRg review provided the .follmiiRg coRd.itioRs are met: (1) at least five PRC members iRcludiRg the GhairmaR aRd RO more thaR 2 alterRates, shall .
revi m1 the item, coAcur with detel"miRati OR as to ~:hether or Rot the item ceRstitutes aR l:IRrevieweel safety questieR, aRd provide writteR commeRts OR the item; (2) all commeRts shall be resolved to the satisfactioR of the reviewers provieliRg the commeRts; aRel (3) if the PRC GhairmaR
determiRes that the*commeRts are sigRificaRt, the item (iRclueliRg commeRts aRd resol uti oRs) shall be recirculated to all re*1i ewers for additioRal commeRts.
The item shall be re\'ie*11ed at a PRC meetiRg iR the evel'lt that:
(1) GommeRts are Rot resolved; er (2) the PlaRt GeReral MaAager overrides the recommeRdatioR' of the PRC; or (3) a proposed chaRge to.
the TechRical SpecificatioRs iRvolves a safety limit, a limitiRg safety system set.tiRg or a limitiRg coRditioR for operatioR; or (4) the item was reportable to the NRG.
6.§.1.8 RECORDS The PRC shall maiRtaiR writteR miRutes of each PRC meetiRg afld shall provide copies to the NPAD.
6-6
Amendment No. -24, +£, 8&, 86, -l-GS, 2+, He
ADMINISTRATIVE CONTROLS
6.§.2 6.§.2.l NUCLEAR PERFORMANCE ASSESSMENT DEPARTMENT (NPADl FUNCTION The ~ucl~ar PerformaAce AssessmeAt DepartmeAt (NPAD) shall fuActieA te prevHie iAdepeAd_eAt revie11 ef activities iA the areas ef: .
a. Nuclear pmter pl aAt eperati eA
b. Nuclear eAgiAeeriAg .
c. Chemistry aAd radiechemistry
d. Metallurgy
e. NeAdestructive testiAg
f. IAstrumeAtatieA aAd ceAtrol
g. Radielegical safety
h. MechaAical aAd electrical eAgiAeeriAg
i. AdmiAistrative ceAtrels aAd E}Uality assuraAce practices J. EmergeAcy PlaAAiAg
k. TraiAiAg 6.§.2.2 COMPOSITION The NPAD shall iAclude the Director, who rep~rts to the Vice PresideRt NOD, aAd a full time staff of Nuclear PerformaAce Specialists reportiAg te the Director. The Directo~ 'Ad !he Nuclea~ PerfermaAce Specialists shall meet er exceed the qual1f1cat1eAs described iA SectieA 4.7 ef ANSI/ANS 3.1 1987. The NPAD shall have Fie direct respeRsibility fer.*
activtties subject te its re~iew.
6. §. 2. 3 CONSULTANTS If sufficieAt expertise is Aet available withiA NPAD te review particular issues, the NPAD shall have the autherity te utilize ceAsultaAts er ether qualified ergaRizatieAs fer expert advice.
6.§.2.4 RESPONSIBILITIES 6.§.2.4.l REVIEW The NPAD shall review:
a. The safety evaluatieAs fer: I) chaAges te precedures, equipmeRt er systems, aRd 2) tests er experimeRts cempleted uRder the previsieRs ef IQ CFR §Q.§9 te verify that such actieAs de Aet ceAstitute aA 1:rnreviewed safety f:IUestieR.
b. Prepesed chaAges te precedures, eE}uipmeAt er systems which iAvelve aA UAreviewed safety questieA as drifiAed iA 10 CFR 50.§9.
6-6a
Amendment No. -24, +-&, 8-&, 8e, 08, 2+, -14i
ADMINISTRATIVE CONTROLS
6.§.2.4.1 REVIEW (Centinuea)
c.
d.
PFepe~ef.I tests eF expeFiments which i nvel ve an unFevi ewef.I safety quest1en as ElefiAed in 10 CFR 50.59.
P~epesed changes to Techni.cal Specificaiions er the Operating License.
e. *~o 1at'iens e~
\I' &
ce~es, regulat~ens, eraers, Technical Specificatiens, license requirements, er ef internal precealires er instructiens having nuclear safety signi.ficance.
f. Significant eperating abnermalities er aeviatiens frem nermal ana expected perfermance ef unit eqliipment that affects nliclear safety.
g* All repertable eveAts having nuclear safety significance.
. h. All recegnized indicatiens ef an linanticipated deficiency in same aspect ef aesign er eperatien ef strlietlires, systems, er cempenents that celild affect nlielear safety.
i. Reports and meeting minlites of the PlaAt Review Cemmittee.
J. Fire Protectien Pregram ana ImplementiAg Preeealire Changes.
6.§.2.4.2 AUDITS
Auaits ef eperatioAal nliclear* safety relates facility activities shall be performef.I by the NPAD staff unf.ler the eegAizanee of the Nuclear Perfermance Specialists. These alid.i ts shall eAcompass:
a. The ceAfe~maAce of.P1aAt.eperation te.previsieAs ceF1taiF1ea within the TeehAieal Spee1f1eatieF1s aAd applicable lieeAse eenf.litiens at least eAce per 12 meAths. ..
b. The perfermaAce, traiAiflg aAd qlialificatieAs ef the entire facility staff at least enee per 12 meAths.
e. The perferman~e ~f activitie~ reqliirea by the Qliality Assurance Pregram Descr1~t1e~ fer Operatienal Nliclear Pewer Plants (GPG 2A) to meet the criteria of 10 GFR §0, Appenf.lix B at least ence per 24 menths. *
d. The Site Emergency PlaA and implementiAg precedlires at least eAce per 12 moAths.
e. The Site See1:irity Plan and implementing proeedlires (as requiref.I by the Site Seclirity PlaA) at least once per 12 meAths.
6-7
Amendment No. -24, +.&, ~, -l-2-7-, Me,~ .J.§.2.
ADMINISTRATIVE CONTROLS
6.§.2.4.2 AUDITS (CoRtiRued) f.
g.
ARY o~t-lCr are~ of pl aRt op.erati OR coRsi dered appropriate by NPAD or the Vice PresideAt Nuclear OperatioRs.
least eRce per 24 moRths. .
The plaRt Fire ProtectioR Program aRd implemeRtiRg procedures at
h. AR !RdepeRdeRt fire protectieR aRd loss p~eveRtioR iAspectioR aRd a~dit to be perfermed aAAually utiliziAg either qualified offsite liceRsee persoRRel or aR outside fire protectioR firm.
i. AR iRspectioR aRd audit ef the fire protectioR aRd loss preveRtioR program to be performed by aA outside qualified fire coF1sultaF1t at iRtervals AO greater thaR 3 _years. * -
j. Radiological eAvireRmeRtal meAitoriRg program aRd -the results thereof at least eAce per 12 moRths.
k. The OFFSITE DOSE CALCULATION MANUAL aAd implemeF1tiF1g procedures for precessiRg aRd packagiRg of radieactive wastes at least oRce per 24 mORths.
l. The PROCESS CONTROL PROGRAM aAd implemeAtiRg_procedures for processiflg aRd packagiRg ef radioactive wastes at least oRce per 24 RlORths. _
Audit reports eF1compassed by SpecificatioA 6.5.2.4.2 above shall be forwarded to ~he Di~ec~or N~AD, aRd MaRagemeRt positioRs respeRsible for the areas audited withiA thirty (30) days after completio~ of the audit.
6.5.2.4.3 NPAD re*liew of the s*ubjects iA Spec;ificatioAs 6.5 .. 2.4.1 aAd 6.5.2.4.2 shall be performed by aR assigRed Nuclear PerformaRce Specialist se1ected ?R the basis of his techAical expertise relative_ to the subject be1Rg rev1ewed. If the assigRed Nuclear PerformaRce Specialist determiAes the F1eed for iAterdiscipliRary review~ a committee coF1sistiF1g ef the Director, NPAD, or his desigRate, aRd at least four Nuclear PerformaAce Speciali5ts, shall be assigAed. Such committee shall me~t as.coRditioRs requiriflg iRterdiscipliAary review arise, but FIO less thaA t~fl ce yearly.
6.5.2.5 AUTHORITY The NPAD shall report to aAd advise the Vice PresideF1t NOD of sigRificaRt fiFldiRgs asseciated with these areas ef respeRsibility specified iA SectioAs 6.5.2.4.1 aAd SectieA 6.5.2.4.2.
6-8
Amendment No. S, 8-&, 408, ,£2-7, -!%
ADMINISTRATIVE CONTROLS
6.§.2.6 RECORDS Recerds ef NPAD activities shall be maintained.
prepared and distributed as irtdicated below:
a.
Reperts shall be The results of reviews, performed pursuartt to Sectiort 6.6.2.4.1 artd Sectien 6.6.2.4.2, shall be reported to the Vice President NOD. at least mortthly.
b. A report assessirtg the everall rtuclear safety performartce of Palisades shall be provided to senior Consumers Power Company martagemertt artrtually.
6.§.3 PLANT SAFETY AND LICENSING 6.§.3.l FUNCTION The Plartt Safety and Licertsing staff shall review proposed chartges irt design er eperation and such other matters as the PRC may assign te identify issues significant to nuclear safety artd recommertd rtuclear safety improvements.
6.6.3.2 COMPOSITION The Plant Safety and Licensing staff responsible for the review functien shall be experienced technical staff meeting the qualificatiorts of Sectien 6.3.
6.§.3.3 RESPONSIBILITIES The Plant Safety and Licensing staff may provide nuclear safety review as delegated by PRC for:
a. Procedures, programs and changes thereto identified irt Specification 6.8 and any additional procedures and changes thereto identified by the Plartt Gerteral Martager as significant to rtuclear safety~
b. All preposed tests or experiments.
c. All prepesed changes er modifications to plant systems or equipment.
d. The Site Emergency Plart.
6-9
Amendment No. S, 8-5, G&, 2-7, 146, -l7G
ADMINISTRATIVE CONTROLS
6.§.3.4 AUTHORITY The PlaRt Safety aRd LiceRsiRg staff shall determiRe those issues sigRificaRt to Ruel ear .safety which require review by the PlaRt Review Committee from items coRsidered under Specification 6.§.3.3.a through d.
For those items Rot referred to PRC, PlaRt Safety aRd LiceRsiRg shall Y"ecommeRd in writing to pl ant management approval or di sappro*1al of items considered under 6.5.3.3.
6.5.3.5 RECORDS Reports of PlaRt Safety and LicensiRg activities pursuant to SpecificatioR 6.5.3.3 shall be submitted monthly to PRC.
6.6 (Deleted) 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the eveRt a safety limit is
1i ol ated:
a. The reactor shall be shut down immediately and not restarted until the Commission authorizes resumption of operation (10 CFR 50.36(c)(l)(i)(A)).
b. The safety limit *1iolation shall be reported uithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to the CommissioR iR accordance with 10 CFR 50.36, as well as to the Vice
c.
d.
President Nuclear Operations and to the NPAD .
A report shall be prepared in accordaRce with 10 CFR 50. 36 and 6. 9 of this specification. (The safety limit violation and the report shall be reviewed by the PRC.)
The report shall be submitted withiR 14 days to the Commission (in accordance with the requirements of 10 CFR 50.36), to the Vice President Nuclear Operations and to the NPAD.
6-10
Amendment No. ?J, 5, 93, 2-7, -146
ADMINISTRATIVE CONTROLS 6.8 PROCEDURES ANO PROGRAMS
1;~:1::::::::::::::::::::::1::::::::::::::::::::111111i:1~:~
Written procedures shall be established, implemented!]~] and maintained covering the activities referenced below: ***
i~!~;:!~i~-it:i!:~:*~!~!~~~
a.
Oescriptiofl.
b. Refueling operations.
c. Surveillance and test activities of safety-related equipment.
d. Site Security Plafl implemefltatiofl.
e. Site Emergeflcy Plafl implemeAtatiofl.
.f-.---Site Fire Protection Program implementation.
1@:::::::::::::::::~1:1::1:eiP:iiirn§:::::::1P::is:1::i1::g~:::::::1::9:::::::§:e§2:1:::r:!:s:ii::1:2n:::::::§::~::§::~::
6.8.2 Procedures afld chaflges shall be approved prior to implemefltatiofl by the appropriate* seFtior departmeAt maAager predesignated by the Plaflt GeAeral MaAager subject to the reviews per Specifications 6.5.1.6 afld*
6.5.3 .
6.8.3 Temporary changes to procedures of Speci fi cation 6. 8 .1 abo*1e may be made provided:
a.
b.
The iF1teF1t of the origiAal procedure is Aot altered.
The chaAge is approved by t'.m members (or desigAated alterAates) of the PRC, at least oAe of whom holds a SeAior Reactor Operator LiceAse.
c. The chaAge is documeAted, subsequeAtly reviewed by PlaAt Safety aAd LiceAsiAg withiA 30 days of issuaAce aAd approved by the appropriate* seAior departmeAt maAager predesigAated by the PlaAt GeAeral MaAager.
The determiAatioA of the appropriate seAior departmeAt maAager is based oA the activities addressed by the specific procedure aFtd will be predesigflated ifl \:ritiflg by the Pl aAt GeAeral Manager.
6-11 Amendment No. a.e., ~' 5-, -I-OJ., 2-7, 94
ADMINISTRATIVE CONTROLS 1@g~;::ffi::1::::::1m:1:::::;:;:11§>>.:~~~~::r:~~~~::::::~:~~r~~m~
6-.8.-4 The following programs shall be established, implemented, and maintained:
a-.Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the
~~~~~~~ g~ f~~ i ~~~[ ~ ~emi~&~ii1i~i1.:::riRi.wfuB.1;;ffiE i f;~!n ~~a i ~~~t 1~~e~f}~ uents
~h e ig~~~:~:iFi~=~PP,;~:~}:,~,,,,;,~=i,nJ:ix'.~;gl:i::;,;~;:::;x;~?{*:;:*g*~MO'~?am<
y,:l:J==:S.=::i:=:lif-i::::==~Q$.:~i~b~i~li:t.U.m~=:!ii.:l:PthmdlD.=UAm~===:=.-== it> 2 >C l)s has~a1 l ~eb~ mpco1emen 1
nta i ~e~ ~ n e y o eraffh~f""""foEedTir*e*s';""""'afi"d""""fj")""**5hall'"'"i*nc l ude remedial actions to be taken w~enever t~e program limits are exceeded. The program shall include the following elements:
-lfi.ll Limitations on the operability of radioactive liquid and gaseous
"'"""*******monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, .
2-}IW Limitations on the concentrations of radioactive material released
=*=*=*=*=*=* in l i uid effluents to UNRESTRICTED AREAS inre!iitiUt!t!e(f{~ieias\
con fo ~mi ng to 1o cFR 2o, Appendix B, Tab l e=*=*=*ng*=;=*=*=*=calUffi'lf'*=*t:=:=*=*=*=*=*=*=<*=*=
3-tim Monitoring, sampling, and analysis of radioactive .J...J.q,yid and .
...... gaseous effluents in accordance with 10 CFR 20 .-1-00l.$:0.i and with the methodology and parameters in the ODCM, .............
4-H.iM Limitation on the annual and quarterly dos.es or dose commitment to
""'""'***a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS lllEiii;t1ii~!lllllilir!ii conform i ng to App e As ix I ta 1o cFR 5o, ~iliP:lji!
5)@il~l! Limitations on the dose rate resulting from radioactive material
...... r.~J..~.~.~g9. . .J.o: ...g.aseous effluents to areas beyond the SITE BOUNDARY j.:jj}l'g::::JH:U:m~:j:fy conforming to the doses as.soci ated with 10 CFR 20, At>>tfohd"fx"""ff;*******rabl e H£, Column 1.
etf.\~l Limitations on the annual and quarterly air doses resulting from
. . ...,,, noble gases rel eased in g.~.§.§..9...Y..$.. . . . ~. f.f.1.Y. ~nts from each unit to areas beyond the SITE BOUNDARY :tt.efrb.ld.Uhtt:~fit~ conforming to AppeAai x I te 10 cFR 50 ' lie~i~sliiiil[~M *:*:m*:*:*:*:<<<o:*:*:*:*:M:*:*:*:*:<-:*:<-:*:*:*:*:*:*:*:****:*
+ti@ Limitations on the annual and quarterly doses to a member of the
public from Iodine-131, Iodin_e-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous
~~f:+/-JHi.PJ~,,~,liiJ:.,~,leasedf fr~m etach 11 unit d ~o ~r;as beyCoFnRd t he,lf:~,J,I,,§.]i,§,2,,H~B,ARY lifHilfM9:~mY:!t'.¥ con orm mg o nPP e At:11 x :i t; o 10 5 0 , rlllP.&!!Hm~tW:t SJD ~~:t~~:~mi~~d~~~J~:~mFE~:*::u::::ERS conforming to 40.CFR 190.
6-12 Amendment No. 36, 8§., -l-54
ADMINISTRATIVE CONTROLS.
6.8.4 (centinued)
b. Radieleaical Envirenmental Meniterina Pregram A pregram shall be previded ta me Riter the radi ati eR aRd radieRuclides iR the eRviron~ of the plaRt. The pregram shall previde (1) represeRtative measuremeRts ef radieactivity in the highest poteRtial exposure pathways, aRd (2) verifications of the accuracy of the efflueRt moRitoriRg program aRd modeliRg of eR't'ifoRmeRtal exposure path*,Jays. The program shall ( 1) be
. coRtaiRed iR the ODCM, (2) coRform to the guidaRce of AppeRdix I to IQ CFR §0, aRd (3) iRcludiRg the follo'n'iRg:
1) M6RitoriRg, sampliRg, analysis, aRd reportiRg ~f radiatioR aRd radioRuclides iR the eRviroRmeRt iR accordaRce with the methodology aRd*parameters iR the ODCM.
2) A LaRd Use GeRsus to eRsure *that chaRges iR the use of areas at aRd beyoRd the SITE BOUNDARY are ideRtified aRd that modificatioRs to the moRitoriRg program are made if required by the results of this ceRsus, aRd
3) ParticipatioR iR a lRterlaboratory ComparisoR Program to eRsur~ that iRdependeRt checks OR the precisioR aRd accuracy of the measuremeRts of radioactive materials iR eRviroRmeRtal sample matrices are performed as part of the quality assuraRce
. pregram fer cR¥ireRmeAtal meAiteriRg .
6-13
Amendment No. a-6, -1.§4
ADMINISTRATIVE CONTROLS
. 6.9.1 Reporting Requirements Tneftifb.l:J::&wfi§.Mreports aRd other ~'lritteR commuRi cati oRs sha 11 be
'siihm'fffe'ff'*'*t'Ct"l*he NRG in accordance with the t"equil<*emeRts of 10 CFR 50.4 .
ReHtiRe Reports a.
!caf a~lo

eiesti fig \h~llah~ ::c:nt:: ~!i1!w1::rt1 }p r::~i ~:~~:f afl eperatiRg liceF1se, 1
(2) ameF1dmeF1t to the lieeRse iRvolviRg a plaRRed iflcrease ifl power level, (3) iRstallatiofl ef fuel that has a differeRt desigR or has beeR maRufactured by a different fuel supplier aRd, (4) modificatioRs that may have significantly altered.
the nuclear, thermal or hydraulic performaRce of the plant. The report shall address each of the required tests aRd shall, iR general, include a descriptioR of the measured values of the eperatiRg coRditioRs or characteristics ebtaiRed during the test program aRd a comparisoR of these values with desigR predictions and specificatioRs. ARY corrective actioRs that were required te obtaiR satisfactory operatioR shall also be described. ARY additieRal specific details required iR liceRse coAditioRs based BR ether cemmitmeRts shall be iRcluded iR this report.
Start up reports shall be submitted Mi thi R (i) 90 days foll owi fig cempletien of the start up \est program, (2) 90 days fellewiRg resumptioR or commeRcemcRt of commercial pouer operatioFI or, (3) 9 moAths followiAg iAitial criticality, whichever is earliest. If
the Start Up Report does Aot cover all three eveAts (i.e., iRitial criticality, completioA of start up test program aRd resumptiofl er commeAcemeAt of commercial power operatioA), supplementary reports shall be submitted at .1 east every three moFJths UAtil al 1 three eveRts have beeA completed.
~~~ :!li!~:!!!!:l!!!!!!~! !!!!!!!!:~§!!!fi~:it::P~~~re:; ~=l ~=~a~e~::;
6.6.1 1 te supplement requiremcAts of 10 CFR 20.407 should be submitted prior to March 1 of each year.
This aRRUal report shall iAclude:
This report shall include a tabulation on an annual basis of the number of station, utility and other personne 1 (inc 1udJ.. n.9. . .
contractors} receiving exP.Q.sures greater than 100 mRemfinii.ro/year and their asso.~.i ated man Remj~i.j exposure according to work""""a"fid job
functions le.g., reactor**~~erations and surveillance, jns~rvice inspect ion*;" routine maintenance, speci a1 ma i n.tenan~~ ... ff.:9.~.~.~r..i..~.~....
maintenance73: waste processini and refuel ing:l. J:ffi':§::tt:ib.Ul:atadn
,,,:,~:~il~,:~:f:~i~'=l!*l:!'~~:!J,*~*!~~:~.~!!~;!*5!;~!1!2~1*::-!'!'im~'~'~:H~'~'~':'~'':'~'@
~~~ ~~;e~~~ ~ ~~t eS~a;:~::!!~:~~~:~:~::~p:~:~"i:f:y~:~:nq ~~~] ~~a~ 1 ~ d% b~1g ~he individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external
- n c t i o n s . II
Amendment No. -!&, a.e, 8&, l§4
ADMINISTRATIVE CONTROLS
6.9.1 RoijtiAe Reports e-.
(coAtiAijed)
Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the fifteenth of each month following the calendar month covered by the report.
&.- Radioactive Effluent Release Report The Radioactive Effluent Release Report shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be fl+ consistent with the. . . .9..1.?.J~.f..tt.ves outlined in the ODCM aAd PROCESS CONTROL PROGRAM and ffis:nal:l::tb~ in conformance with 10 CFR 50. 36a
~~~ e!:1:~:;1,::::::,~:i.::!:!p!,~~:1!::~:::~@::]§g~imin:::t!:v:i:~:~:::~::i:::f:IS ect i OA I'.' . B. 1 of e-. Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the unit tj~ring the previous calendar year shall be submitted before May II of each year. The report shall include
summaries, interpretatlons, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the report~ ng ~eri od. The dm~teri a1 protl~,~~'g'f~'.h.MI.1.,.,J?.,d.,.,.,.t.Q.nj.,.i,.if,!:.~,.n.! ~*!.:.tt.~'l it~~~~:~p~~
8
.f. Core Operating Limits Report (COLR)
Core operating limits shall be established prior to each reload cyc)e, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
3 .1.1 ASI Limits.
3.10.5 Regulating Group Insertion Limits 3.23.l Linear Heat Rate (LHR) Limits 3.23.2 Radial Peaking Factor Limits
6-15 Amendment No. 6-, e, 3-6-, 8-§., -MS, §4, -l ADMINISTRATIVE CONTROLS Re1:1tiF1e Reperts
COLR (continued}
The analytical. methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the* latest approved revision of the following documents:
1. XN-75-27(A}[i "Exxon 'Nuclear Neutronics Design Methods for Pressuri zed****"Water Reactors," and Supp 1ements 1(A}, 2(A},
3(P}(A}, 4(P}(A), and 5(P)(A); Exxon Nuclear Company.
(LCOs 3.1.1, 3.10.1, 3.10.5, 3.23.1, &3.23.2}
2. ANF-84-73(P}(A}, "Advanced Nuclear Fuels Methodology for Press1,1rized Water Reactors: Analysis of Chapter 15 Events,"
and Appendix B(P}(A} and Supplements l(P)(A}, 2(P}(A};
Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, &3.23.2)
3. XN-NF-82-21(P}(A}, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"
Exxon Nuclear Company. (LCOs 3.1.1, 3.23.1, &3.23.2)
4. ANF-84-093(P}(A}I "Steamline Break Methodology for PWRs," and Supplement l(P}(A}; Advanced Nuclear Fuels Corporation.
(LCOS 3.10.1, 3.10.5, 3.23.1, &3.23.2}
5. XN-75-32(P}(A}, "Computational Procedure for Evaluating Fuel Rod Bowing," and Supplements l(P}(A}, 2(P}.(A), 3(P)(A), and
6.
4(P)(A); Exxon Nuclear Company. (LCOs 3.1.1, 3.10.5, 3.23.1,
& 3.23.2)
EXEM PWR Large Break LOCA Model as defined by:
(LCOs 3~10.5, 3.23.1, & 3.23.2) a) XN-NF-82-20(A), "Exxon Nuclear Company Evaluation Model EXEM/PWR 'ECCS Model Updates," and Supplements l(P)(A),
2(P)(A}, 3(P}(A), and 4(P)(A}; Exxon Nuclear Company.
b} XN-NF-82-07(P}(A), "Exxon Nuclear Comp~ny ECCS Cladding Swelling and Ru~ture Model," Exxon Nuclear Company.
c) XN-NF-81-58(AJI "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluition Model," and Supplements l(P}(A}i 2(P)(A}, 3(P)(A}, and 4(P)(A); Exxon Nuclear Company.
d) XN-NF-85-16(A), "PWR l7x17 Fuel C.ool ing Tests Program,"
Volume 1 and Supplements l(P)(A), 2'(P)(A), and 3(P)(A},
and Volume 2 and Supplement l(P)(A); Exxon Nuclear Company.
e) XN-NF-85-105(A), "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," and Supplement l(P)(A); Exxon Nuclear Company.
6-16
Amendment No. ~
ADMINISTRATIVE CONTROLS 6.9.l Ro1:1tifle Reports
+.-COLR 7.
(continued)
XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.
(LCOs 3.10.5, 3.23.1, &3.23.2)
8. ANF-1224(P)(AJI "Departure from Nucleate Boiling Correlation for High Therni"ifl Performance* Fµel," and Supplement 1(P) (A);
Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.23.1, &
3.23.2)
9. ANF-89-lSl(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, &
3.23.2)
10. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation~ (LCOs 3.1.1, 3.23.1, &3.23.2) ill~! The core operating limits shall be determined such that all
. applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (EGGS) limits, nuclear limits such as shutdown margin, transient analysis
limits, and accident analysis limits) of the safety analysis are met.
The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC.
6.9.2 Reportable fveAts The Commissiofl shall be notified of Reportable [vents and a report s1:1bmitted p1:1rs1:1ant to the req1:1iremeRts of 10 GFR 50.73.
6.9.3 Nonro1:1tine Reports A report shall be s1:1bmitted iR the eveRt that (a) the Radiological ERviroAmeRtal MoAitoriRg Programs are Rot s1:1bstaRtially c0Rd1:1cted as described iR the ODCM or (b) aR 1:1nus1:1al or importaRt event occ1:1rs from plaAt operatioR that ca1:1ses a sigRificaRt eRviroRmeRtal impact or affects a poteRtial environmental impact. Reports shall be s1:1bmitted withiR 30 days. *
(Next page is 6 26) 6-17 Amendment No. ~
ADMINISTRATIVE CONTROLS
6.9 6~9.4 Re~ertiRq Seecial Reeerts RequiremeRts (ceRtiRued) 1;1;1:i;;1::rMm1:::::;::1.11!~~!:1:?~~~~~:~:~:~!m~!ii~~~:::::::::!:~!:!s~r]i:!~::!!~n'.~¥~~:!~!::!~:~!:!~:~t:!~tl!~
a.SpecialReports shall be submitted to the NRC covering the activities
~
SpecificatieR RepertiRg Area RefereAce PrestressiRg, ARcherage, 4.§.4 90 Days ,O,fter LiRer aRd PeRetratieR 4.§.5 GempletieR ef Tests .the Test*
A test is ceRsidered te be cemplete after all asseciated mechaRical, chemical, etc., tests have beeR cempleted.
b. Special reperts shall be submitted iA accerdaAce with 10 GFR §0.4, withiR the time peried specified fer each repert.
6.10 REGO~D RETENTION
6.10.1 IR additieR te the applicable recerd reteRtieR requiremeRts ef Title IQ, Gede ef Federal RegulatieRs, the fellewiAg recerds shall be retaiRed fer at least the miRimum peried iRdicated:
The fellewiRg recerds shall be retaiRed fer at least five years:
a. Recerds aRd legs ef facility*eperatieR ceveriRg time iRterval at each power level.
b. Recerds aAd legs ef priAcipal maiRteRaRce activities, iAspectieRs, repair aRd replacemeRt ef priAcipal items ef equipmeRt related to Rl:rnlear safety.
e. All reportable e*1eAts as defiRed i R Secti OR 6. 9. 2.
d; Recerds ef surveillaAce activities, iAspectioAs aAd calibratieRs required by these TechAical SpecificatieRs.
6-26
Amendment No. ~' .§4, 85, MS, -146, l54
ADMINISTRATIVE CONTROLS
RECORD RETENTION (Cont'd) c.
f.
Records of changes made to the procedures required by Specification 6.8.1.
Records of radioactive shipments.
g. Records of scaled source leak tests aFld results.
h. Records of aF1F1ual physical iF1veF1tory of all source material of record.
6.IQ.2 The following records sh al 1 be retained for the duration of the Facility Operating LiceF1se:
a. Record aF1d drawiflg chaF1ges ioeflectiflg facility desigfl modificatiofls made to systems afld equipment described ifl the Fiflal Safety Aflalysis Report .
. b. Records of Re'ti and irradiated fuel i F1VeF1tory fuel traflsfers and c.
radiatiofl coRtrol areas.
cL Records of gaseous afld liquid radioactive material released to the eR\'i FORS *
c. Records of tra~iient or ope!a~ional cycles for those facility compoRents des1gned for a 11m1ted Rumber of traF1sients or cycles.
f. Records of iflservice iF1spectioF1s performed pursuaRt to these Technical Specifications.
g. Records of Quality Assurance activities requi~ed by the QA Program Description. .
h. Rec9rds of revie~s per~ormed for chaRges made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR
§0.§9. .
i. Records of meetiF1gs of the PRC and reviews performed by NPAD.
j. Records of monthly facility radiatiofl aF1d coF1tamiflatiofl surveys.
6.:.27
Amendment No. le, &9, -100, l4i Order, dated 10/24/80
ADMINISTRATIVE CONTROLS
k.
l.
Recerds ef secendary ~;ater sampling and quality .
Recerds ef ..the service lives ef all hydraulic and mechanical snubbeins c0*1eined by Specif.i cat.ten 3. 20.
date at which the service life commences and asseciatbd installati en and maintenance inece.inels ..
This shall incl uele the
m. Receinels ef tinai Ai Ag anel quali fi cati ens fain membeins ef the pl ant staff.*
R. Recerds ef reacter tests and experiments.
a. Recel"ds ef reviews performed fer changes made ta the OF'FSITE: DOSE:
CALCULATION MANU.AL and the PROCESS CONTROL PROGRAM.
6.11 RADIATION PROTECTION PROGRAM Pinecedu¥"es fain pel'sennel inadi ati en. pl'etecti en shall b.e pine pained censistent ~11ith the requiroments ef 10 CFR 20, and sh.all be appreveel, maintaineel. anel aelheineel te fain all eperati-0ns invelviflg peinsennel raeli ati en* expesure ..
6.12f:ml HIGH RADIATION AREA In lieu of the "con,t..r.p.1 device" or "alarm signal required by 10 CFR 20.203(c)(2):ii§Qjf., each high radia.t.J..9.r.:t. J!.r.!.. ~. in which the intensity
*''O'sl'e'cf*a's''a*'*hJ9h*F~'d','atToH*'*'a're~
CY requiring issuance of a Radiation Work Permit.* Any individual or group of individuals .permitted to enter such areas shall be provided*
with or accompanied by one or more of the following:
and entrance thereto shall b~ contro~l ed
a. A radiation monitoring device which c6ntinuously indicates the radiation dose rate in the area.
b. A radiation monitoring device which continuously integrates the radiation dose rate i~ the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device .may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
6-28
Amendment No. 6, 48, 69, -l-08, §4
ADMINISTRATIVE CONTROLS
c. An individual qualified in radiation protecti.on procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Permit .
6-2.9
Amendment No. 48
ADMINISTRATIVE CONTROLS (Pages 6 30 Through 6 32 Deleted) 6-30
Amendment No. 48
ADMINISTRATIVE CONTROLS 6 .13 (Deleted) 6.14 (Deleted) 6.15 £¥£H~M£ INlEGRH¥ 6.15 l I,. Provisions establishing preventive maintenance and periodic visual inspection requirements, and 6.1§ 2 :1:. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
4.§.3a(.l) i::i1 The portion of the shutdown cooling system that is outside the containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig at the iRterval specified iR
4:5.3a(2) Iii:
e-.+s .
Piping from valves CV-3029 and CV-3030 to the discharge of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig at the iAterval specified iR 6.15.
4.§.3b The maximum allowable leakage from the recirculation heat removal systems' components (which include valve stems, flanges and pump seals} shall not exceed 0.2 gallon per minute under the normal hydrostatic head from the SIRW tank (approximately 44 psig}.
6.16 IODINE MONITORING The l i ceAsee shall impl emeAt a program which wil 1 eRSlffe the capability to accurately determiRe the airborRe iodiRe coRceRtratioR iR vital areas uRder accideRt coRditioRs. This program shall iRclude the followiRg:
1. TraiRiRg of persoRRel, 2.. Procedures for moRitoriRg, aRd
3. ProvisioRs for maiRteRaRce of sampliRg aRd aRalysis equipmeRt.
6-33
Amendment No. 31-, fH-, +/-GB
ADMINISTRATIVE CONTROLS
e.17§M§l1~ Post Accident Samplinq::~::eii.i~l.J ity particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:
+/-i. Training of personnel,
~
i~.
Procedures for sampling and analysis:f:::mi!Jg Provisions for maintenance of sampling and analytic equipment .
6-34
Amendment No. &&, +/-9G
ADMINISTRATIVE CONTROLS 6-.-l8§1:~::~ml OFFSITE DOSE CALCULATION MANUAL WDCM) i,:~:: Changes to ODCM:
1:.
Shall. be documented and records of reviews performed shall be retai.ned as required by SpecificatioA 6.10.20. This documentation shall contain:
i)
Sufficient information to support the change together*
with the appropriate analyses or evaluations justifying the changes, and A determination that the change will maintain the level of radi oact i vg,J~.ffl uent contra l re qui red by 10 CF R 20 *-1-00l.~A9'?.i, 40 C.f..B.. . .1.~.Q. L ....19.. . . ~.I R 50 .. 36a, and
AppeAd ix I to***10 .. *crn so:;.::tAP,p:g)P.J.l:J.:tx=::::::=i::~: and not adversely impact the accuracy or relfahnny***of effluent, dose, or setpoint calculations.
Sha 11 become effective after the rev i cw aAd acceptaA.QP..J?Y the
,e,,B,§,,,,,,,,a,~,9,'t!'],fta,,,,}~~proval *PM the Pl aRt GeReral MaRager Plllin;I
§:HrMnU:lt:JUlJm:~.
. Shall be submitted to the CommissioRN'itC.~ in the form of a complete, legible copy of the entire'.'tl'btM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was m~de. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.,
month/year) the change was implemented.
6.19 PROC[SS CONTROL PROGRAM Changes to the PROCESS CONTROL PROGRAM:
a. Shall be documented and records of reviews performed shall be retained as required by SpecificatioR 6.10.20. This documeRtatioR shall contain:
1) Sufficient informatioR to s~pport the change together ~~ith the appropriate aRal yses or e*1al uati oR just i fyi Rg the chaRge ( s) and
2) A determination that. the change will maintain the overall conformance of the solidified wa~te product to existiRg requirements of Federal, State, or other applicable regulations .
b. Shall become effective after review and acceptance by the PRC and approval of the Plant General Manager.
6-35 Amendment No. SS., -li4
ADMINISTRATIVE CONTROLS
6.2Q (Deleted) 6-36
Amendment No. a;, 154 January I, 1993
- -------~----
ADMINISTRATIVE CONTROLS
6.21.1 SEALED SOURCE CONTAMINATION Program Each sealed se1:1rce ceRtaiRiRg radieactive material either iR excess ef lQQ m.i crec1:1rtes ef beta aRd/er gamma emi tti Ag material or § mi crec1:1ri es ef alpha emittiRg material.shall be free ef greater thaR er eq1:1al te 9.QQ§ microc1:1ries of removable coRtamiRatioR.
6. 21. 2 With a sealed so1:1rce haviRg removable coRtamiRatioR iR excess of Q.Q9§ micrec1:1ries, immediately withdra*.1 the sealed so1:1rce frem 1:1se aRd either:
1. DeceRtamiRate aRd repair the sealed se1:1rce, er
2. Dispese ef the sealed se1:1rce iR accordaRce with applicable reg1:1lati0Rs.
6. 21.3 Each category of sealed so1:1rces as described iR e.21.1 with a half life greater thaA 3Q days (excl1:1diRg HydrogeR 3), aRd iR aRy ether form thaR gas, shall be tested for leak~ge aRd/or coRtamiRatioR at iRtervals Rot to exceed 6 moRths.
6. 21. 4 The test shall be performed by the liceRsee or by other persoRs 6.21.6 The periodic leak test does Rot apply to sealed sources that are stared aRd Aot beiRg 1:1sed. These se1:1rces ~hall be tested. prior to 1:1se or traRSfer to aRother liceRsee, l:IR]ess tested WithiR the preViOl:IS e moRths. Sealed so1:1rces which are ceF1tiF11:1e1:1sly eRclesed withiR a shielded mechaRism (ie, sealed so1:1rces withiR radiatioR moRitoriRg or boroR mcas1:1riRg devices) are ceRsi.elered ta be stared aRd Reed Rat be tested l:IRless they are removed from the shicl.ded mechaRism.
6.21.7 Scaled so1:1rces ~raRsferred witho1:1t a certificate iRdicatiRg the last test date shall be testeel prier to beiRg placed iR 1:1se.
6.21.8 A report shall be 13re13arcd aRd s1:1bmittcd to the GommissioR OR .aR aRRl:lal basis if sealed so1:1rce leakage tests reveal the preseRce of greater thaR or eq1:1al to Q.90§ microc1:1ries of removable coRtamiRatioR.
6-37
Amendment No. 98
ADMINISTRATIVE CONTROLS
6. 221m1:i:1 SECONDARY WATER CHEMISTRvt::1.1n~iil:w:11tsiii A program shall be establi.shed, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:
~- and Identification of a sampling schedule for the critical variables control points for these variables, Identification of the procedures used to measure the values of the critical variables, Identification of process sampling pointsJ which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, Procedures for the recording and management of data, Procedures defining corrective actions for all off-control point chemistry conditions, and e-1. A procedure. identifying (_a) the authority responsible for the interpretation of the data, and (b). the sequence and timing of administrative events required to initiate corrective actions .
Amendment No. ~
ADMINISTRATIVE CONTROLS (New Page)
ADMINISTRATIVE CONTROLS
ADMINISTRATIVE CONTROLS
IMl:f:::J,:Jliliil[il:ill:i::l!:§.!iiil, [Fuel Oil Testing Program]
1;1;1:f:::11::::::::::::::;:mr:~!~~'~j:~:w.~::::::::~1,~~w~~=:s~~:~::§:~'.~::::::::t:!~::1::::::~~:1~:~~¥!!::~~?~:~§~:E~r~s:~~~
§M:::::rniillillllllllllllllllllllllllilli~ll','i*l~*1:111J!lillllijl.ill~illlllll111
t::~::),]i:Itli]sii,n~1:rn1::n::1::1n:iM:::x~m:~~11;§~ie!!l~P!:::::::~:i:::::::1!i:mm::u~:1n:§~::~:rn1:1 1
~:;:~::::::::::::::11!llllllll~llllllllll,llll~~lill1!11.lili"ll=l; 11111!11*11'1~**11*11111:::::::;:::
lm~m~:;::::::iilililliiB.iii'.r¥.il [Safety Function Determination Program]
lmll[iilililllli::::m:::I~~~~!'.~!~!: [Containment Leak Rate Testing Program]
!ili§l~li!l[lil[l[l[lllililff.~~!~Y~!. [Pressure and Temperature Limits Report]
1:~::1:f::foi:::::::::::rn:::::::m111:1:~~11::::::Ein:m1~ti:n1;:t::r:~~~~~~~~m~:1~::'.::1~Q'.~:~1 (New Page)
ATTACHMENT 3 CONS~MERS POWER C.OMPANY PAL-ISADES PLANT DOCKET 5.0-255 TECHNICAL SPECIFICATION CHANGE REQUEST REVISION OF ADMINISTRATIVE CONTROLS
Comparison.of Existing and Proposed Administrative Controls 1 Page
ATTACHMENT 3 Comparison of Existing and Proposed Administrative Controls Current TS Section Proposed Location 6.0 ADMINISTRATIVE CONTROLS 6.0 (same name) 6.1 RESPONSIBILITY 6.1 (same name) 6.2 ORGANIZATION 6.2 (same name) 6.2.1 Offsite and Onsite Organizations 6.2.1 Onsite and Offsite Organizations 6.2.2 Plant Staff 6.2.2 (same name)
Table 6.2-1 Minimum Shift Crew Composition 50.54 & 6.2.2 Plant Staff
6.3 PLANT STAFF QUALIFICATIONS 6.3 (same name) 6.4 Deleted 6.5 REVIEW AND AUDIT Relocated to CPC-2A 6.5.1 Plant Review Committee Relocated to CPC-2A 6.5.2 Nuclear Performance Assessment Department Relocated to CPC-2A 6.5.3 Plant Safety and Licensing Relocated to CPC-2A 6.6 Deleted 6.7 SAFETY LIMIT VIOLATION 50.72, 50.73, & 2.0 SAFETY LIMITS 6.8 PROCEDURES AND PROGRAMS 6.4 PROCEDURES & 6.5 PROGRAMS & MANUALS 6.8.l Procedures 6.4 PROCEDURES 6.8.2 Procedure Change Approval Relocated to Admin Procedures 6.8.3 Temporary Procedure changes Relocated to Admin Procedures 6.8.4 Programs 6.5 PROGRAMS & MANUALS 6.8.4.a Radioactive Effluent Controls Program 6.5.4 (same name)
6.9 6.8.4.b Radiological Environmental Monitoring Program REPORTING REQUIREMENTS 6.9.1 Routine Reports 6.9.1.a 6.9.1.b Start-up Report
.Annual Report Relocated to ODCM 6.6 (same name) 6.6 REPORTING REQUIREMENTS Deleted 6.6.l Occupational Radiation Exposure Report 6.9.1.c Monthly Operating Report 6.6.4 (same name) 6.9.1.d Radioactive Effluent Release Report 6.6.3 (same name) 6.9.1.e Radiological Environmental Operating Report 6.6.2 (same name) 6.9.l.f Core Operating Limits Report 6.6.5 (same name) 6.9.2 Reportable Events 50.73 6.9.3 Nonroutine Reports (ODCM violation or environmental impact) Deleted 6.9.4 Special Reports 6.6.8 Containment Tendon Surveillance Report 6,10 RECORD RETENTION Relocated to Admin Procedures 6.11 RADIATION PROTECTION PROGRAM 6.4.a (procedures required by RG 1.33) 6.12 HIGH RADIATION AREA 6.7 (same name) 6.13 Deleted 6.14 Deleted 6.15 SYSTEMS INTEGRITY 6.5.2 Primary Coolant Sources Outside Containment 6.16 IODINE MONITORING 6.5.3 Post Accident Sampling Program 6.17 POST ACCIDENT SAMPLING 6.5.3 Post Accident Sampling Program 6.18 OFFSITE DOSE CALCULATION MANUAL 6.5.1 Offsite Dose Calculation Manual 6.19 PROCESS CONTROL PROGRAM Relocated to ODCM 6.20 Deleted 6.21 SEALED SOURCE CONTAMINATION Relocated to ODCM 6.22 SECONDARY WATER CHEMISTRY 6.5.9 Secondary Water Chemistry Program
ATTACHMENT 4 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST REVISION OF ADMINISTRATIVE CONTROLS
Comparison of STS and Proposed Administrative Controls 1 Page
ATTACHMENT 4 Comparison of STS and Proposed Administrative Controls
~
Proposed TS Section (current TS) 5.0 ADMINISTRATIVE CONTROLS 6.0 ADMINISTRATIVE CONTROLS (6.0) 5.1 Responsibility 6.1 RESPONSIBILITY (6.1) 5.2 Organization 6.2 ORGANIZATION (6.2) 5.2.1 Onsite and Offsite Organizations 6.2.1 Onsite and Offsite Organizations (6.2.1) 5.2.2 Unit Staff 6.2.2 Plant Staff (6.2.2) 5.3 Unit Staff Qualifications 6.3 PLANT STAFF QUALIFICATIONS (6.3) 5.4 Procedures 6.4 PROCEDURES (6.B.1) 5.5 Programs and Manuals 6.5 PROGRAMS AND MANUALS (6.8.4) 5.5.1 Offsite Dose Calculation Manual 6.5.1 Offsite Dose Calculation Manual (6.18) 5.5.2 Primary Coolant Sources Outside Containment 6.5.2 Primary Coolant Sources Outside Containment (6.15) 5.5.3 Post Accident Sampling 6.5.3 Post Accident Sampling Program (6.17) 5.5.4 Radioactive Effluent Controls Program 6.5.4 Radioactive Effluent Controls Program (6.8.4.a) 5.5.5 Component Cyclic or Transient Limit Not proposed 5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance 6.5.5 Containment Tendon Surveillance Program (4.5.4) 5.5.7 Reactor Coolant Pump Flywheel Inspection Program 6.5.6 Primary Coolant Pump Flywheel Surveillance Program (4.3f)
5.5.8 Inservice Testing Program 5.5.9 Steam Generator Tube Surveillance Program
5.5.10 Secondary Water Chemistry Program 5.5.ll Ventilation Filter Testing Program 6.5.7 Inservice Inspection and Testing Program (4.0.5) 6.5.8 Steam Generator Tube Surveillance Program (4.14) 6.5.9 Secondary Water Chemistry Control Program (6.22) 6.5.10 Ventilation Filter Testing Program (4.2.3) 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program Not proposed 5.5.13 Diesel Fuel Oil Testing Program 6.5.11 Reserved [Diesel Fuel Program to be included in STS TSCR]
5.5.14 Technical Specifications Bases Control Program 6.5.12 Technical Specifications Bases Control Program (New) 5.5.15 Safety Functions Determination Program 6.5.13 Reserved .[SFDP to be included in STS TSCR] (New) 5.6 Reporting Requirements 6.6 REPORTING REQUIREMENTS (6.9) 5.6.l Occupational Radiation Exposure Report 6.6.1 Occupational Radiation Exposure Report (6.9.1.a) 5.6.2 Annual Radiological Environmental Operating Report 6.6.2 Radiological Environmental Operating Report (6.9.1.e) 5.6.3 Radioactive Effluent Release Report 6.6.3 Radioactive Effluent Release Report (6.9.1.d) 5.6.4 Monthly Operating Reports 6.6.4 Monthly Operating Report (6.9.1.c) 5.6.5 Core Operating Limits Report 6.6.5 Core Operating Limits Report (6.9.l.f) 5.6.6 Pressure and Temperature Limits Report 6.6.6 Reserved [PTLR to be included in STS TSCR] (3.1.2) 5.6.7 EOG Failures Report Not proposed 5.6.8 Post Accident Monitoring Report 6.6.7 Accident Monitoring Instrument Report (3.17.4.7.c))
5.6.9 Tendon Surveillance Report 6.6.8 Containment Tendon Surveillance Report (6.9.4) 5.6.10 Steam Generator Tube Inspector Report 6.6.9 Steam Generator Tube Surveillance Report (4.14.6) 5.7 High Radiation Area 6.7 HIGH RADIATION AREA (6.12)
ATTACHMENT 5 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST REVISION OF ADMINISTRATIVE CONTROLS
STS SRs Which Reference Programs 2 Pages
Attachment 5 STS Surveillance Requirements Which Reference Programs 11::1.111[::1:11::1:i:1i,::1111~1::::11i::::::11m1::111i:1::::::rn1::::::11!i:v?.11;111::W:::
1. . . .
~..R.. .........~. .~. . ~J.9 ~.Jm. . . Verify each pressurizer safety valve is OPERABLE in accordance
[:~:11.J::l.:t:tMIM!Mtt:i: with the shall Inservice Testing+/- 1%.
Program. Following testing, lift
...*.*.*.*.*.*.*.*.*.*.*.*.*.*.*.w....................... settings be within Verify leakage from each RCS PIV is equivalent to ~ 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure ~ [2215] psia and ~ [2255] psia. In accordance with the Inservice Testing Program.
Verify the following ECCS pumps develop the required differential pressure on recirculation flow. In accordance with the Inservice Testing Program.
Verify each charging pump develops a flow of ~ [36] gpm at a discharge pressure of ~ [2200] psig. In accordance with the Inservice ~esting Program.
~R ......~ ~ ~..*.~.. !.~........ Verify the i sol at ion ti me of each power operated and each j@g:f.:l))~iP.~:g$,!I automatic containment isolation valve is within limits. In
accordance with the Inservice Testing Program or 92 days.
SR 3.6.6.5 Verify each containment spray pump [develops ~ [250] psid IM!:;:; differential pressure on recirculation flow]. In accordance with the Inservice Testing Program.
~.R. .*. *.*-~. .~. .§.. .~.z..~ t . . . .*.*. *.*.*******Verify ea*ch spray additive pump develops a differential pressure Nit!!.1:::::1¥P:!Pll1:9.:ill!'-i]!of [100] psid on recirculation flow. In accordance with the
Inservi ce Testing Program.
SR 3.7.1.1 Verify each required MSSV lift setpoint per Table 3.7.1-2 in ffiab.J::e::HII(!?IitII#. accordance with the Inservice Testing Program. Following
,,.,,,,,.,.,.,.,.,,,,,.,,,.,.,,,,,,,,,,,.,.,,~,,.,.,.,,,~,,,.,.,,,,,,,,,.,.,,,,,.,,, testing, lift settings sha 11 be within +/- 1%.
SR 3.7.2.1 Verify closure time of each MSIV is ~ [4.6] seconds on an actual l!l~ll.!Jitllililillllll!lf[:::lllllllllfor simulated actuation
. . . . . . . . . *.* * * * * * * *. .*.*.-.*.*.*.*.*.*.*.*.*.*.*.-.*.*.*.*.*.*. . . . .*.*.*Testing Program o.r [18] signal.
months]. In accordance with the [lnservice SR 3.7.3.1 Verify the closure time of each MFIV [and [MFIV] bypass valve] is i.ij:il:::::g~[§rM~~]:::::::::t~ [7] seconds on an actual or simulated actuation signal. In
=************************************-****~ccordance with the [lnservice Testing Program or [18] months].
SR 3.4.13.2 Verify SG tube integrity is in accordance with the Steam IMJ'.f:[l::[Iffrii;§l!il Generator Tube Survei 11 ance Program .
SR 3.6.1.2 IM!if:llillliii!ii9if[j[f[Conta i nment Tendon Survei 11 ance Program.
Verify containment structural integrity in accordance with the
\ .*
\J' Attachment 5 STS Surveillance Requirements Which Reference Programs
~.!.L. . ~ . . .~. . J9..~..?. ....... Perform required ICS filter testing in accordance with the tUiilllP!P:Ja@i.:JH!i
.*:-..............................*..*.***************************.-:-.-.....*.****** Ventilation Filter Testing Program {VFTP).
SR 3.7.11.2 Perform required CREACS filter testing in accordance with i'''i!t.~'l'f''''''''''''''''''''''''''''':~:
!M~d!f.f@rJtPRB!!:8
[V en t i'l a t.ion Fi'l t er Tes t.i ng Program {VFTP)] *
~ilili:i~:~:~:i.;iiJ,I ~~~f [~~n~~~~i ~~~ Vi~~e~Ri:;~ i ~~ 1~~~g;:~t i~~Ti}] ~ccordance with
.~.B.. ~"~],.,~.1.1.,~,.,?,.,.,.,.,.,. ., Perform required FBACS filter testing in accordance with the l@:~;~:l.:l:llP:i~lil [Vent,il at ion Filter Testing Program (VFTP)].
~R ......~ . *..!.. J.~.~..?. ....... Verify re qui red PREACS filter testing in accordance with the
~9:!¥::::::1,µ:~UH!!l~~ii [Ventilation Filter Testing Program (VFTP)].
SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are tested il.i'lllllilllilllliiP~ e~~~ ~~:~c~i i~~t~~~ ~~~~;:~~ed 0 r 1 within the imits of, the .

ML18065A346

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