ML023380730

Facility Operating License DPR-63, License Amendment Request: Technical Specifications Section 6.0, Administrative Controls - Response to RA Information TAC No. MB2441

ML023380730
Person / Time
Site:	Nine Mile Point
Issue date:	11/22/2002
From:	Conway J Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP1L 1698, TAC MB2441
Download: ML023380730 (119)

Text

P.O. Box 63 Lycoming, New York 13093 0 Constellation Energy Group Nine Mile Point November 22, 2002 Nuclear Station NMP1L 1698 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Nine Mile Point Unit 1 Docket No. 50-220 Facility Operating License DPR-63 License Amendment Request: Technical Specifications Section 6.0, Administrative Controls - Response to Request for Additional Information TAC No. MB2441 Gentlemen:

Nine Mile Point Nuclear Station, LLC (NMPNS) hereby transmits supplemental information requested by the NRC in support of a previously submitted application for amendment to Nine Mile Point Unit 1 (NMP1) Operating License DPR-63. The initial application, dated October 26, 2001, proposed to revise the format and content of Technical Specification (TS) Section 6.0, Administrative Controls, in a manner similar to the Nine Mile Point Unit 2 (NMP2)

Administrative Controls section. Supplemental information was provided by letter dated June 7, 2002.

On August 7, 2002, the NRC staff transmitted by e-mail a list of questions regarding the proposed TS amendment. A subsequent telephone conference was held on August 22, 2002 between the NRC and NMPNS staff representatives to discuss the questions and disposition the corresponding issues and responses. As a result of the telephone conference, the NRC staff issued a Request for Additional Information (RAI) on August 30, 2002. The NMPNS responses to the RAI are provided in Attachment 1.

The following attachments incorporate appropriate changes to the application for amendment to reflect: (1) the RAI responses; (2) additional changes identified by NMPNS (summarized in Attachment 2); and (3) changes to make the submittal consistent with the implementation of License Amendment Nos. 173, 174, and 176 that have been approved by the NRC since the original October 26, 2001 submittal (also summarized in Attachment 2).

Page 2 NMP1L 1698

"* Attachment 3 provides the retyped TS pages. These pages replace in their entirety the retyped pages that were submitted in Attachment A of the October 26, 2001 letter.

" Attachment 4 provides the markups of the NMP1 current TS (CTS) pages and the associated discussion of the changes (DOCs). These replace in their entirety the CTS markup pages and DOCs that were submitted in Attachment B of the October 26, 2001 letter. Note that Attachment 4 includes markups of the TS Bases on CTS pages 115 and 150. These Bases page markups are provided for information only and do not require issuance by the NRC.

" Attachment 5 provides the markups of the Improved Standard Technical Specifications (ISTS) (NUREG-1434, Revision 1) and the associated justifications for deviation (JFDs).

These replace in their entirety the ISTS markup pages and JFDs that were submitted in Attachment B of the June 7, 2002 letter.

" Attachment 6 provides four tables that group the proposed TS changes by change category and provide summary descriptions of the changes. These replace in their entirety the tables that were provided in Attachment C of the June 7, 2002 letter.

For Attachments 4, 5, and 6, the specific changes to the originally-submitted information are identified by a vertical bar and an "A" in the right margin. This letter contains no new commitments.

This supplemental information does not affect the No Significant Hazards Consideration analysis that was provided in the October 26, 2001 submittal. Pursuant to 10 CFR 50.91(b)(1), NMPNS has provided a copy of this supplemental information to the appropriate state representative.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 22, 2002.

Sincerely,

? T. Conway Vice President Nine Mile Point JTC/DEV/jm

Page 3 NMP1L 1698 Attachments:

1. Response to Request for Additional Information

2. Summary of Additional Changes

3. Proposed Technical Specifications Pages (Retyped)

4. Current Technical Specifications (CTS) Markup and Discussion of Changes (DOCs)

5. Improved Standard Technical Specifications (ISTS) (NUREG-1434, Revision 1) Markup and Justifications for Deviation (JFDs)

6. Tables Summarizing the Proposed Changes

7. List of Regulatory Commitments cc: Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)

Mr. J. P. Spath, NYSERDA

ATTACHMENT 1 Response to Request for Additional Information A telephone conference was held on August 22, 2002 between the NRC and Nine Mile Point Nuclear Station, LLC (NMPNS) staff to discuss previously e-mailed questions regarding proposed changes to the Nine Mile Point Unit 1 (NMP1) Technical Specifications, Section 6.0, Administrative Controls. As a result of the telephone conference, the NRC staff issued a formal Request for Additional Information (RAI) on August 30, 2002, which included three questions.

The following information provides the NMPNS responses to the three questions. Each question is repeated verbatim (italics added) from the RAI, followed by the NMPNS response.

6.0-1 See Table 1 In the NMP-I Section 6.0 Current Technical Specifications (CTS) cited in Table 1, the wording is either exactly the same as or similar to the correspondingNMP-1 Section 6.0 proposed Technical Specifications (PTS) cited in Table 1. In converting the CTS to the PTS some of these specificationswere modifiedfrom the Standard Technical Specifications (STS) and the NMP-2 Improved Technical Specifications (ITS) and some justifications(DOC's and JFD's) were provided to justify deviationsfrom the standardand consistency with NMP-2 ITS. Yet for the PTS 6.0 specifications(cited in Table 1), the wording is differentfrom the correspondingNMP-2 ITS 5.0 specification. No justification is provided to describe orjustify these differences/inconsistencies. If the intent of the amendment is to make the Unit 1 TS consistent with the Unit 2 ITS, then the wordingfor similar specificationsin PTS 6.0 should be the same as the correspondingITS 5.0 specification. The same would be truefor those specifications that were not adopted where NMP-2 has specificationsand/orprograms either currently in the CTS or in licensee controlleddocuments. If there is to be consistency between the units, then the specificationsshould be consistent. See Comment Number 6.0-2.

COMMENT: Either make the PTS consistent with the ITS or provide a discussion and justificationfor the deviation/inconsistency. See Comment Numbers 6.0-2.

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Table 1 Cross Reference of Inconsistent Specifications Between Unit 1 and Unit 2 NMP-1 NMP-1 Corresponding Associated Associated CTS PTS NMP-2 NMP-1 NMP-1 Section Section ITS DOCs JFDs Section 6.1.2 6.1.2 5.1.2 M.1 3 6.2.1 Title 6.2.1 Title 5.2.1 Title - 9 6.2.1.a 6.2.1.a 5.2.1.a 2,3 6.2.1.b 6.2.1.b 5.2.1.c - 9 6.2.1.c 6.2.1.c 5.2.1.b - 9 6.2.2 6.2.2 5.2.2 - 9 6.2.2.a; 6.2.2.a 5.2.2.a A.4 4 Table 6.2-1 6.2.2.b; 6.2.2.e; - 5.2.2.b LA.2 TSTF-258 Table 6.2-1 Table 6.2-1 6.2.2.b 5.2.2.c A.5 1, 3, 5, 9 Note 6 6.2.2.h 6.2.2.d 5.2.2.e LA.6, L.1 2, 3, 6, 8, 9, TSTF-258 6.2.2.i 6.2.2.e 5.2.2.f 2, 9 6.3.1 6.3.1 5.3.1 LA. 1 3 6.8.1 6.4.1 5.4.1 3,5

- 6.4.1.a 5.4.1.a A.2 6 6.5 5.5 A.1

- 5.5.3 7

- - 5.5.5 10 6.17 6.17 5.5.6 12

- - 5.5.7 13

-- _ _-_ 5.5.9 15

- 6.5.3.b.2 5.5.10.b.2 A.2, M.1 9, TSTF-364 6.9.1.c 6.6.4 5.6.4 L.I TSTF-258 6.9.1.F.3 6.6.5.c 5.6.5.c 7

-- 5.6.6 -- _ 11 Page 2 of 14

NMPNS Response:

Table 1, Item 1 CTS 6.1.2 PTS 6.1.2 NMP-2 ITS 5.1.2 DOC M.1 JFD 3 Editorial inconsistencies (use of acronyms) were identified between the Nine Mile Point Unit 1 (NMP1) proposed Technical Specifications (PTS) and the corresponding Nine Mile Point Unit 2 (NMP2) Improved Technical Specifications (ITS). NMP1 PTS 6.1.2 has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.1.2. This change affects PTS 6.1.2, retyped TS page 347; the Current Technical Specifications (CTS) markup for Specification 6.1, page 2 of 3 (Insert 6. 1-B); and Improved Standard Technical Specifications (ISTS) markup page 5.0-1.

NMP1 and NMP2 differ in the identification of reactor operating conditions in their respective Technical Specifications (TS). NMP1 uses text descriptions of the operating conditions (e.g.,

power operating condition, hot shutdown condition, etc.) that are defined in Section 1.1 of the NMP1 CTS, whereas NMP2 uses mode numbers (e.g., MODE 1, MODE 2, etc.) that are defined in Table 1.1-1 of the NMP2 ITS. This difference is due to the differing licensing bases for the two units. The NMP1 TS have used the text descriptions of the reactor operating conditions since the NMP1 Provisional Operating License No. DPR-17 was issued in 1969. The NMP2 TS, originally issued as NUREG-1253 in July of 1987, were initially prepared based on NUREG 0123, "Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)," which used OPERATIONAL CONDITION numbers to identify reactor operating conditions. With the conversion of the NMP2 CTS to the ITS in NMP2 License Amendment No.

91 issued on February 15, 2000, the OPERATIONAL CONDITIONS became MODES, consistent with NUREG-1433 and NUREG-1434. Since the submittal properly reflects the NMP1 plant-specific licensing basis, no changes to the submittal are considered necessary.

Table 1, Item 2 CTS 6.2.1 Title PTS 6.2.1 Title NMP-2 ITS 5.2.1 Title JFD 9 An editorial inconsistency (omission of the "s" from "Organizations" in the title) was identified between the NMP1 PTS and the corresponding NMP2 ITS. The NMP1 PTS 6.2.1 title has been revised such that it is the same as NMP2 ITS 5.2.1. This change affects PTS 6.2.1, retyped TS page 347; the CTS markup for Specification 6.2, page 1 of 5; and ISTS markup page 5.0-2.

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Table 1, Item 3 CTS 6.2.1.a PTS 6.2.1.a NMP-2 ITS 5.2.1.a JFD 2, 3 Editorial inconsistencies (use of commas, "chart" versus "charts") were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.1.a has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.2.1.a. This change affects PTS 6.2.1.a, retyped TS page 347; the CTS markup for Specification 6.2, page 1 of 5; and ISTS markup page 5.0-2.

Table 1, Item 4 CTS 6.2.1.b PTS 6.2.1.b NMP-2 ITS 5.2.1.c JFD 9 Editorial inconsistencies (subsection lettering, reversal of items "b" and "c") were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.1.b and 6.2.1.c have been revised such that the subsection lettering is the same as NMP2 ITS 5.2.1. This change affects PTS 6.2.1.b, retyped TS pages 347 and 348; the CTS markup for Specification 6.2, page 1 of 5; and ISTS markup page 5.0-2.

Table 1, Item 5 CTS 6.2.1.c PTS 6.2.1.c NMP-2 ITS 5.2.1.b JFD 9 Editorial inconsistencies (subsection lettering, reversal of items "b" and "c") were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.1.b and 6.2.1.c have been revised such that the subsection lettering is the same as NMP2 ITS 5.2.1. This change affects PTS 6.2.1 .c, retyped TS pages 347 and 348; the CTS markup for Specification 6.2, page 1 of 5; and ISTS markup page 5.0-2.

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Table 1, Item 6 CTS 6.2.2 PTS 6.2.2 NMP-2 ITS 5.2.2 JFD 9 Editorial inconsistencies (lead-in sentence wording differences) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.2 has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.2.2. This change affects PTS 6.2.2, retyped TS page 348; the CTS markup for Specification 6.2, page 2 of 5; and ISTS markup page 5.0-2.

Table 1, Item 7 CTS 6.2.2.a, Table 6.2-1 PTS 6.2.2.a NMP-2 ITS 5.2.2.a DOC A.4 JFD 4 NMP1 and NMP2 differ in the identification of reactor operating conditions in their respective TS. NMP1 uses text descriptions of the operating conditions (e.g., power operating condition, hot shutdown condition, etc.) that are defined in Section 1.1 of the NMP1 CTS, whereas NMP2 uses mode numbers (e.g., MODE 1, MODE 2, etc.) that are defined in Table 1.1-1 of the NMP2 ITS.

This difference is due to the differing licensing bases for the two units. The NMP1 TS have used the text descriptions of the reactor operating conditions since the NMP1 Provisional Operating License No. DPR-17 was issued in 1969. The NMP2 TS, originally issued as NUREG-1253 in July of 1987, were initially prepared based on NUREG-0123, "Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)," which used OPERATIONAL CONDITION numbers to identify reactor operating conditions. With the conversion of the NMP2 CTS to the ITS in NMP2 License Amendment No. 91 issued on February 15, 2000, the OPERATIONAL CONDITIONS became MODES, consistent with NUREG-1433 and NUREG 1434.

The NMP1 and NMP2 TS also differ regarding the number of non-licensed operators that are required during the hot shutdown operating condition (MODE 3 in the NMP2 ITS). NMP1 CTS 6.2.2, Table 6.2-1, requires at least one (1) non-licensed operator in the hot shutdown condition, whereas NMP2 ITS 5.2.2.a requires at least two (2) non-licensed operators when in MODE 3.

This difference is due to the differing licensing bases for the two units. The NMP1 TS have required one non-licensed operator in the hot shutdown condition since Facility Operating License No. DPR-63 was issued in 1974. For NMP2, the requirement to have two non-licensed operators when in MODE 3 (OPERATIONAL CONDITION 3 prior to ITS conversion) has existed since the TS were initially issued as NUREG-1253 in July of 1987.

Based on the above discussions, since the submittal properly reflects the NMP1 plant-specific licensing basis, no changes to the submittal are considered necessary.

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Table 1, Item 8 CTS 6.2.2.b, 6.2.2.e, Table 6.2-1 NMP-2 ITS 5.2.2.b DOC LA.2 JFD TSTF-258 In the markup of ISTS 5.2, ISTS 5.2.2.b has been deleted. Section 5.2.2.b states the following:

"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 in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room."

Deletion of this item is based on Technical Specification Task Force Traveler 258 (TSTF-258),

Revision 4, which was approved by the NRC on June 29, 1999. This change was not incorporated into the NMP2 ITS because NRC approval of TSTF-258 occurred after the NMP2 ITS application was submitted. NMPNS routinely reviews and prioritizes applicable NRC approved TSTF travelers for future incorporation into the NMP2 ITS.

Table 1, Item 9 CTS Table 6.2-1, Note 6 PTS 6.2.2.b NMP-2 ITS 5.2.2.c DOC A.5 JFD 1, 3, 5, 9 Editorial inconsistencies (use of the word "The," capitalization, "two" versus "2" ) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.2.b has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.2.2.c. This change affects PTS 6.2.2.b, retyped TS page 348; the CTS markup for Specification 6.2, page 4 of 5 (Insert 6.2-B) and page 5 of 5; and ISTS markup page 5.0-3.

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Table 1, Item 10 CTS 6.2.2.h PTS 6.2.2.d NMP-2 ITS 5.2.2.e DOC LA.6, L.1 JFD 2, 3, 6, 8, 9, TSTF-258 Editorial inconsistencies (use of acronyms and commas) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.2.d has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.2.2.e. This change affects PTS 6.2.2.d, retyped TS page 348; the CTS markup for Specification 6.2, page 3 of 5; and ISTS markup page 5.0-3.

In the markup of CTS 6.2.2.h regarding working hour limits, certain changes have been incorporated based on TSTF-258, Revision 4, which was approved by the NRC on June 29, 1999. This change was not incorporated into the NMP2 ITS because NRC approval of TSTF-258 occurred after the NMP2 ITS application was submitted. NMPNS routinely reviews and prioritizes applicable NRC-approved TSTF travelers for future incorporation into the NMP2 ITS.

Table 1, Item 11 CTS 6.2.2.i PTS 6.2.2.e NMP-2 ITS 5.2.2.f JFD 2, 9 Editorial inconsistencies (use of acronyms) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.2.2.e has been revised such that acronym use is the same as NMP2 ITS 5.2.2.f. This change affects PTS 6.2.2.e, retyped TS page 349; the CTS markup for Specification 6.2, page 3 of 5; ISTS markup page 5.0-4; and JFD 9 for ISTS 5.2 (deleted).

The NMP1 and NMP2 TS differ regarding the identification of individuals who shall hold a Senior Reactor Operator (SRO) license. NMP1 CTS 6.2.2.i requires that as a minimum, either the Manager Operations or the General Supervisor Operations shall hold an SRO license. NMP2 ITS 5.2.2.f requires that the operations supervisors shall hold an SRO license. This difference is due to the differing licensing bases for the two units. The NMP1 CTS 6.2.2.i requirement was approved by the NRC in NMP1 License Amendment No. 160 issued on February 19, 1998. For NMP2, the ITS 5.2.2.f requirement was included in the conversion of the CTS to the ITS in NMP2 License Amendment No. 91 issued on February 15, 2000. As part of that conversion, NMP2 incorporated the generic title "operations supervisors" in place of the previously identified specific position titles of "General Supervisor Operations" and "Supervisor Operations." JFD 10 for ISTS 5.2 has been added to reference License Amendment No. 160 as the current NMP1 licensing basis, and ISTS markup page 5.0-4 has been revised accordingly. No other changes to the submittal are considered necessary.

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Table 1, Item 12 CTS 6.3.1 PTS 6.3.1 NMP-2 ITS 5.3.1 DOC LA.1 JFD 3 The NMP1 and NMP2 TS reference different industry standards regarding minimum unit staff qualification requirements. NMP 1 CTS 6.3.1 references ANSI N 18.1-1971, whereas NMP2 ITS 5.3.1 references ANSI/ANS 3.1-1978. This difference is due to the differing licensing bases for the two units. The NMP1 TS have referenced ANSI N18.1-1971 since Facility Operating License No. DPR-63 was issued in 1974. For NMP2, the TS have referenced ANSI/ANS 3.1 1978 since the TS were initially issued as NUREG-1253 in July of 1987. NMP1 UFSAR Section XIII-A.4.0 and NMP2 USAR Section 13.1.3 reflect these differing licensing bases.

The NMP1 and NMP2 TS also differ because the NMP1 TS contain certain exceptions to the qualification requirements of ANSI N18.1-1971 for the Operations Manager. These exceptions were approved by the NRC in NMP1 License Amendment No. 160 issued on February 19, 1998.

Similar exceptions have not existed in the NMP2 TS since the TS were initially issued as NUREG-1253 in July of 1987.

Based on the above discussions, since the submittal properly reflects the NMP1 plant-specific licensing basis, no changes to the submittal are considered necessary.

Table 1, Item 13 CTS 6.8.1 PTS 6.4.1 NMP-2 ITS 5.4.1 JFD 3, 5 NMP1 PTS 6.4.1 contains a phrase referencing Sections 5.1 and 5.3 of ANSI N18.7-1972 that does not exist in NMP2 ITS 5.4.1. This difference is due to the differing licensing bases for the two units. The NMP1 TS have included the subject phrase referencing ANSI N18.7-1972 since Facility Operating License No. DPR-63 was issued in 1974. The NMP2 TS have not included any reference to ANSI N18.7 since they were originally issued as NUREG-1253 in July of 1987.

Since the submittal properly reflects the NMP1 plant-specific licensing basis, no changes to the submittal are considered necessary.

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Table 1, Item 14 PTS 6.4.1.a NMP-2 ITS 5.4.1.a DOC A.2 JFD 6 The NMP1 and NMP2 TS reference different revisions of Regulatory Guide (RG) 1.33. NMP1 PTS 6.4.1.a references RG 1.33 dated November 3, 1972, whereas NMP2 ITS 5.4.1.a references RG 1.33, Revision 2 (dated February 1978). This difference is due to the differing licensing bases for the two units. As noted in DOC A.2 for PTS 6.4, the NMP1 RG 1.33 reference is consistent with NMP1 statements of conformance contained in Amendment No. 1 to the Application to Convert Provisional Operating License to Full-Term Operating License, dated November 1973, and in Niagara Mohawk Power Corporation (NMPC) letter to the AEC dated November 16, 1973. For NMP2, the TS have referenced RG 1.33, Revision 2, since the TS were initially issued as NUREG-1253 in July of 1987. NMP1 UFSAR Section XIII-A.4.0 and NMP2 USAR Section 13.1.3 reflect these differing licensing bases. Since the submittal properly reflects the NMP1 plant-specific licensing basis, no changes to the submittal are considered necessary.

Table 1, Item 15 PTS 6.5 NMP-2 ITS 5.5 DOC A.1 Editorial inconsistencies (lead-in sentence omission) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.5 has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.5. This change affects PTS 6.5, retyped TS page 350, and the CTS markup for Specification 6.5, page 1 of 8.

Table 1, Item 16 NMP-2 ITS 5.5.3 JFD 7 The NMP1 Post Accident Sampling System (PASS) has been eliminated, as approved by the NRC in NMP1 License Amendment No. 174 dated August 26, 2002; therefore, the program requirements of ISTS 5.5.3 have not been added. DOC A.3 for PTS 6.5 and JFD 7 for ISTS 5.5 have been revised accordingly.

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Table 1, Item 17 NMP-2 ITS 5.5.5 JFD 10 Component cyclic or transient limits for NMP1 are identified in NMP1 Updated Final Safety Analysis Report (UFSAR) Table V-2. Component cyclic or transient limits for NMP2 are identified in NMP2 Updated Safety Analysis Report (USAR) Table 3.9B-1, Note 5. The nuclear steam supply systems for NMP1 and NMP2 are totally independent with no interconnections.

The NMP2 statement of design conformance with 10 CFR 50, Appendix A, General Design Criterion 5, "Sharing of Structures, Systems, and Components," provided in NMP2 USAR Section 3.1.2.5, states the following:

"Unit 2 is a single unit and does not share structures, systems, and components important to safety."

Therefore, using the same rationale expressed in RAI 6.0-2 below, this submittal does not propose to add component cyclic or transient limit program requirements to the NMP1 TS. No changes to the submittal are considered necessary.

Table 1, Item 18 CTS 6.17 PTS 6.17 NMP-2 ITS 5.5.6 JFD 12 Addition of the Inservice Testing Program specification, consistent with ISTS 5.5.7, has been reviewed and approved by the NRC in NMP1 License Amendment No. 173 dated August 5, 2002. This program is re-numbered as PTS 6.5.4. This change affects PTS 6.5.4, retyped TS page 353; the CTS markup for Specification 6.5, page 4 of 8; ISTS markup pages 5.0-10 and 5.0 11; and JFD 12 for ISTS 5.5.

Table 1, Item 19 NMP-2 ITS 5.5.7 JFD 13 The NMP2 ventilation filter testing program (VFI'P) described in NMP2 ITS 5.5.7 applies to the following engineered safety feature (ESF) ventilation systems:

1. Standby gas treatment system (SGTS), described in NMP2 USAR Section 6.5.1, and

2. Control room envelope filtration (CREF) system, described in NMP2 USAR Section 9.4.1.

For NMP1, ventilation filter testing is performed in accordance with the CTS for the following ventilation systems:

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1. Reactor building emergency ventilation system (RBEVS), described in NMP1 UFSAR Section VII-H, and

2. Control room air treatment system (CRATS), described in NMP1 UFSAR Section III-B.

The NMP1 RBEVS and CRATS are totally independent of, and have no interconnections with, the NMP2 SGTS or CREF system. The NMP2 statement of design conformance with 10 CFR 50, Appendix A, General Design Criterion 5, "Sharing of Structures, Systems, and Components,"

provided in NMP2 USAR Section 3.1.2.5, states the following:

"Unit 2 is a single unit and does not share structures, systems, and components important to safety."

Therefore, for the reason described in JFD 13, and using the same rationale expressed in RAI 6.0-2 below, this submittal does not propose to add ventilation filter testing program requirements to the Administrative Controls portion of the NMP1 TS. No changes to the submittal are considered necessary.

Table 1, Item 20 NMP-2 ITS 5.5.9 JFD 15 See the response to RAI 6.0-2 below.

Table 1, Item 21 PTS 6.5.3.b.2 NMP-2 ITS 5.5.10.b.2 DOC A.2, M.1 JFD 9, TSTF-364 In the markup of ISTS 5.5.11 regarding the Technical Specifications Bases Control Program, changes have been incorporated based on TSTF-364, Revision 0, which was approved by the NRC on June 16, 2000. This change was not incorporated into the NMP2 ITS because NRC approval of TSTF-364 occurred after the NMP2 ITS application was submitted. NMPNS routinely reviews and prioritizes applicable NRC-approved TSTF travelers for future incorporation into the NMP2 ITS.

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Table 1, Item 22 CTS 6.9.1.c PTS 6.6.4 NMP-2 ITS 5.6.4 DOC L.1 JFD TSTF-258 In the markup of CTS 6.9.1 .c regarding monthly operating reports, the requirement to document challenges to safety relief valves or safety valves has been deleted based on TSTF-258, Revision 4, which was approved by the NRC on June 29, 1999. This change was not incorporated into the NMP2 ITS because NRC approval of TSTF-258 occurred after the NMP2 ITS application was submitted. NMPNS routinely reviews and prioritizes applicable NRC-approved TSTF travelers for future incorporation into the NMP2 ITS.

Table 1, Item 23 CTS 6.9.1.f.3 PTS 6.6.5.c NMP-2 ITS 5.6.5.c JFD 7 Editorial inconsistencies (use of acronyms, hyphens, and commas) were identified between the NMP1 PTS and the corresponding NMP2 ITS. NMP1 PTS 6.6.5.c has been revised to eliminate the editorial inconsistencies with NMP2 ITS 5.6.5.c. This change affects PTS 6.6.5.c, retyped TS page 358; the CTS markup for Specification 6.6, page 6 of 7; ISTS markup page 5.0-20; and JFD 7 (deleted) and JFD 14 (added) for ISTS 5.6.

Table 1, Item 24 NMP-2 ITS 5.6.6 JFD 11 As noted in JFD 11 for ISTS 5.6, the requirements for submitting a special report to the NRC in the event that accident monitoring instrumentation is inoperable currently reside in NMP1 CTS 3/4.6.11; therefore, the special report requirements of ISTS 5.6.8, "PAM Report," are not proposed to be added to the Administrative Controls portion of the NMP1 TS as a separate section. As an alternative, NMPNS proposes to add the following item to CTS 6.9.3, "Special Reports" (PTS 6.6.6):

h. Accident Monitoring Instrumentation Report, Specification 3.6.11.a (Table 3.6.11-2, Action 3 or 4) (Within 14 days following the event).

The introductory paragraph of CTS 6.9.3 states the following: "These reports shall be submitted covering activities identified below pursuant to the requirements of the applicable reference specification." Addition of the Accident Monitoring Instrumentation Report item to CTS 6.9.3 (PTS 6.6.6) is consistent with this statement. The special report requirements specified in CTS Page 12 of 14

Table 3.6.11-2, Action 3 or 4, are essentially the same as those described in ISTS 5.6.8 (NMP2 ITS 5.6.6); thus, special reports will continue to be submitted when required. It is not necessary or desirable to repeat these requirements in two separate TS locations.

This change affects PTS 6.6.6, retyped TS page 358; the CTS markup for Specification 6.6, page 7 of 7; DOC A.4 for PTS 6.6 (added); ISTS markup pages 5.0-22 and 5.0-23 (including Insert 6.6.6-A); and JFD 11 for ISTS 5.6.

6.0-2 JFD15 NMP-2 5.5.9 STS 5.5.10 JFD 15 states that the requirementsfor the diesel fuel oil testing program (STS 5.5. I1O/NMP-2 ITS 5.5.4) are not part of the NMP-1 current licensing basis and thus the requirement/program is not being added. However, there are plantprocedures which specify the testing requirements for both new and stored dieselfuel oil at the plant. The justification does not provide sufficient information to determine the interrelationshipbetween the NMP-1 and NMP-2 diesel fuel oil storage and transfer systems. If the systems are the same system, are shared in any manner or interconnected,then the NMP-1 PTS should contain a program on diesel fuel oil testing in PTS 6.5 to be consistent with NMP-2 ITS and the fuel testing program. If the systems are totally independent with no interconnections,then the staff couldfind that the fuel oil testing program does not need to be in the NMP-1 TS. Comment: Provide a descriptionof the dieselfuel oil storage and transfer system at NMP-1and NMP-2 and revise andjustify any changes to PTS 6.5 based on the extent of sharingof the NMP-1 and NMP-2 diesel fuel oil storage and transfer system.

NMPNS Response:

The NMP1 diesel generator fuel oil system is described in NMP1 UFSAR Section IX-B.4.0. The NMP2 diesel generator fuel oil storage and transfer system is described in NMP2 USAR Section 9.5.4. The systems are totally independent with no interconnections. The NMP2 statement of design conformance with 10 CFR 50, Appendix A, General Design Criterion 5, "Sharing of Structures, Systems, and Components," provided in NMP2 USAR Section 3.1.2.5, states the following:

"Unit 2 is a single unit and does not share structures, systems, and components important to safety."

Based on this information and the rationale expressed in RAI 6.0-2, no changes to the submittal are considered necessary.

Page 13 of 14

6.0-3 DOC A.2 JFD4 CTS 6.9.1.f.2 PTS 6.6.5.b CTS 6.9.1.f 2 specifies the analytical methods used to determine the core operating limits that have been reviewed and approved by the stafffor use in developing the Core OperatingLimits Reportfor NMP-1. These methods are specifically describedin the following topical reports specified in CTS 6.4.1.F.2: 1) NEDE-24011-P-A, (latest approved revision), 2) NEDF-30966-P A (latest approved revision), Volumes I and II, 3) NEDO-20556-P-A, (latest approved revision) and 4) NEDO-32465-A (August 1996, revision). The correspondingPTS 6.6 specification is PTS 6.6.5.b. PTS 6.6.5.b only lists NEDE-24011-P-A (latest approved revision), and not the other three topical reports. The justificationfor deleting the other 3 specified topical reports (DOC A.2) states that "NEDE-24011-P-A now contains all the methods reviewed and approved by the NRC for the NMP1 Loss of CoolantAccident Analysis andfor the Stability Analysis. Therefore, the references to the other three report(NEDE-30966-P-A, NEDO-20556-P-A, and NEDO 32465-A) are redundant." While this may be true at this point in time, any one of these documents could be revised and the other associateddocuments not be revised. The approval of the TS changes associatedwith the use of the COLR was based on all four documents, not just the one. Which may or may not have the current analyticalmethods containedin all the topicals. Comment: Revise PTS 6.6.5.b to conform to CTS 6.4.1.f.2.

NMPNS Response:

Based on the August 22, 2002 telephone conference with the NRC staff and subsequent follow up communications, it is NMPNS's understanding that the NRC staff does not require any further information regarding this RAI.

Page 14 of 14

ATTACHMENT 2 Summary of Additional Chan2es This attachment provides a brief summary of additional changes to the original license amendment request that was submitted to the NRC on October 26, 2001, as supplemented by the additional information provided by letter dated June 7, 2002. These additional changes have been identified by Nine Mile Point Nuclear Station, LLC (NMPNS) and do not result from the NRC Request for Additional Information (RAI). The changes are listed in Revised Technical Specification (TS) Section order (as shown on the retyped TS pages in Attachment 3), and are identified on the affected pages by a vertical bar and an "A" in the right margin.

Table of Contents

1. Beginning page numbers for the Administrative Controls sections have been revised to be consistent with the Revised TS pages provided in Attachment 3 of this submittal. These changes affect Revised TS page v, and the CTS markup for pages v and vi.

Revised TS 6.3

1. The acronyms "SRO" (for Senior Reactor Operator) and "RO" (for Reactor Operator) have been added to Revised TS 6.3.2. These editorial changes affect Revised TS 6.3.2, retyped TS page 349; the Current TS (CTS) markup for Specification 6.3, page 2 of 2 (Insert 6.3-A); and Improved Standard Technical Specifications (ISTS) markup page 5.0-5 (Insert 6.3.2-A).

Revised TS 6.5

1. License Amendment No. 176, issued on September 11, 2002, added new descriptions for the following programs:

"* Offsite Dose Calculation Manual (ODCM) (CTS 6.11, Revised TS 6.5.1)

"* Radioactive Effluent Controls Program (CTS 6.18, Revised TS 6.5.3)

"* Explosive Gas and Storage Tank Radioactivity Monitoring Program (CTS 6.19, Revised TS 6.5.5)

These and the other sections included in Specification 6.5 have been renumbered consistent with the ISTS. These editorial changes affect Revised TS 6.5, retyped TS pages 350 through 355; the CTS markup for Specification 6.5, pages 1 of 8 through 6 of 8, and 8 of 8; DOC A.2 and DOC M.1 for Revised TS 6.5; the markup for Current Specification 6.10, page 2 of 2; ISTS markup pages 5.0-7 through 5.0-17 (including Insert 6.5.7-A); and JFDs 5, 8, 14, and 18 (deleted) for ISTS 5.5.

Page 1 of 4

2. A typographical error was noted in CTS 6.1 .a (as revised in License Amendment No. 176).

The word "gases" should be "gaseous." This error has been corrected. This change affects Revised TS 6.5.1, retyped TS page 350, and the CTS markup for Specification 6.5, page 1 of 8.

3. In CTS 6.1 1.b (as revised in License Amendment No. 176), references are made to Specifications 6.9.1.d and 6.9.1 .e. These references have been revised to 6.6.2 and 6.6.3, respectively, consistent with Revised TS 6.6. This change affects Revised TS 6.5.1, retyped TS page 350, and the CTS markup for Specification 6.5, page 1 of 8.

4. License Amendment No. 174 dated August 26, 2002 (Post Accident Sampling System elimination) added the following sentence to CTS 6.14:

"The requirements shall apply to the Post Accident Sampling System (PASS) until such time as administrative controls provide for continuous isolation of the associated penetration(s) or a modification eliminates the potential leakage path(s)."

This sentence has been incorporated into Insert 6.5-A for Revised TS 6.5.2 as a parenthetical phrase, in the same manner as NMP2 ITS 5.5.2 (as revised by License Amendment No. 106 dated August 9, 2002). Minor wording changes have been included to be consistent with the NMP2 ITS (i.e., addition of the word "program" and deletion of the "PASS" acronym). This change affects Revised TS 6.5.2, retyped TS page 351; the CTS markup for Specification 6.5, pages 3 of 8 and 7 of 8 (Insert 6.5-A); DOC A.3 for Revised TS 6.5; the markups for Current Specification 6.13, page 1 of 1, and Current Specification 6.15, Page 1 of 1; ISTS markup page 5.0-8 (Insert 6.5.2-A); and JFD 19 (added) for ISTS 5.5.

5. By letter dated September 11, 2002, the NRC accepted the use of the Offgas Effluent Stack Monitoring System (OGESMS) instead of the Radioactive Gaseous Effluent Monitoring System (RAGEMS) for accident monitoring of noble gases and particulates. To reflect this change, the sentence added to CTS 6.14 by License Amendment No. 174 (see Item 4 above) is modified to include RAGEMS. This change affects Revised TS 6.5.2, retyped TS page 351; the CTS markup for Specification 6.5, pages 3 of 8 and 7 of 8 (Insert 6.5-A); DOC A.5 (added) for Revised TS 6.5; ISTS markup page 5.0-8 (Insert 6.5.2-A); and JFD 19 (added) for ISTS 5.5.

Revised TS 6.6

1. License Amendment No. 176 revised the descriptions for the following reports:

"* Occupational Radiation Exposure Report (CTS 6.9.1.b, Revised TS 6.6.1)

"* Annual Radiological Environmental Operating Report (CTS 6.9.1.d, Revised TS 6.6.2)

"* Radioactive Effluent Release Report (CTS 6.9.1.e, Revised TS 6.6.3)

Page 2 of 4

These revised report descriptions are included in Specification 6.6. These changes affect Revised TS 6.6, retyped TS pages 356 and 357; the CTS markup for Specification 6.6, pages 1 of 7 through 4 of 7; ISTS markup pages 5.0-18 and 5.0-19; and JFD 2 (deleted), 3, 5 (deleted), 6, and 8 for ISTS 5.6.

2. An error was noted in the CTS markup for Specification 6.6. The "s" was omitted from "Reports" in the title for Revised TS 6.6.4. This error has been corrected. This change affects Revised TS 6.6.4, retyped TS page 357, and the CTS markup for Specification 6.6, page 2 of 7.

3. Editorial inconsistencies were identified between NMP1 Revised TS 6.6.5, Items b and d, and the corresponding NMP2 ITS 5.6.5. These included the use of hyphens and commas, and the description of referenced report NEDE-2401 1-P-A. NMP1 Revised TS 6.6.5 has been revised to eliminate these editorial inconsistencies with NMP2 ITS 5.6.5. These changes affect retyped TS page 358; the CTS markup for Specification 6.6, pages 5 of 7 and 6 of 7; ISTS markup page 5.0-20 (including new Insert 6.6.5-B); and JFD 7 (deleted) for ISTS 5.6.

4. License Amendment No.173, issued on August 5, 2002, and License Amendment No. 176 have deleted certain special report requirements from CTS 6.9.3. These changes affect Revised TS 6.6, retyped TS page 358; the CTS markup for Specification 6.6, page 7 of 7; ISTS markup page 5.0-23 (Insert 6.6.6-A); and JFD 13 for ISTS 5.6.

Revised TS 6.7

1. License Amendment No. 176 added a new description for CTS 6.12 (Revised TS 6.7), "High Radiation Area." This section has been renumbered consistent with the ISTS. This editorial change affects Revised TS 6.7, retyped TS pages 359 through 361; the CTS markup for Specification 6.7, pages 1 of 4 through 4 of 4; DOC LA. 1 (deleted) for Revised TS 6.7; ISTS markup pages 5.0-24 and 5.0-25; and JFD 2 for ISTS 5.7.

Miscellaneous Paaes

1. Implementation of License Amendment No. 176 has resulted in the following changes to the submittal:

"* The Amendment No. 176 version of CTS page 8 is marked up to show the TS cross reference revision, replacing the version included in the original submittal.

"* The following CTS markup pages and Revised TS pages are no longer needed and have been removed from the submittal: 296, 301, 302, 304, 306, 315, 324, 331, 332, and 337.

2. Changes to TS cross-references have been made to be consistent with the section numbering for Revised TS 6.5. These editorial changes affect retyped TS pages 108 and 131; TS Bases page 115; and the CTS markup for Miscellaneous Page Changes, pages 3 of 6 through 5 of 6.

Page 3 of 4

3. The reference to Section 6.0, "Administrative Controls," that appears on TS Bases pages 115 and 150 is replaced by a reference to "Quality Assurance Program requirements." This change is consistent with relocation of the requirements of Current Specification 6.5, "Review and Audit," to the Quality Assurance Topical Report (QATR). This change affects the CTS markup for Miscellaneous Page Changes, pages 4 of 6 and 6 of 6.

Page 4 of 4

ATTACHMENT 3 Proposed Technical Specifications Pages (Retyped)

Replace the existing Technical Specifications (TS) pages listed below with the attached revised pages. The revised pages have been retyped on their entirety, with marginal markings (revision bars) to indicate changes to the text.

Remove Insert v v vi vi 8 8 11 11 108 108 131 131 347 through 364 347 through 364 366 366 367 367 368 368 370 370 371 371 371a 371a 371b 371b 372 372 372a 372a 373 373 374 374 375 375 376 376

SECTION DESCRIPTION PAGE 342 5.0 Design Features 342 5.1 Site 342 5.2 Reactor 342 5.3 Reactor Vessel 345 5.4 Containment 346 5.5 Storage of Unirradiated and Spent Fuels 346 5.6 Seismic Design 347 6.0 Administrative Controls 347 6.1 Responsibility 347 6.2 Organization 349 6.3 Unit Staff Qualifications 349 6.4 Procedures 350 6.5 Programs and Manuals 356 6.6 Reporting Requirements 359 6.7 High Radiation Area V

AMENDMENT NO. 442

THIS PAGE INTENTIONALLY BLANK I AMENDMENT NO. 1U2, 159, 173, 176 vi

1.28 (Deleted) 1.29 (Deleted) 1.30 Reactor Coolant Leakage

a. Identified Leakage (1) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2) Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.

b. Unidentified Leakagqe All other leakage of reactor coolant into the primary containment area.

1.31 Core Operatingq Limits Report The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.5. Plant operation within these operating limits is addressed in individual specifications.

AMENDMENT NO. 142, 176 8

SAFETY LIM IT LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING Written procedures will be developed and b. The IRM scram trip setting shall not exceed 12%

followed whenever the reactor water level is of rated neutron flux for IRM range 9 or lower.

lowered below the low-low level set point (5 feet below minimum normal water level). The The IRM scram trip setting shall not exceed procedures will define the valves that will be 38.4% of rated neutron flux for IRM range 10.

used to lower the vessel water level. All other valves that have the potential of lowering the c. The reactor high pressure scram trip setting shall vessel water level will be identified by valve be*_ 1080 psig.

number in the procedures and these valves will be red tagged to preclude their operating during d. The reactor water low level scram trip setting the major maintenance with the water level shall be no lower than -12 inches (53 inches below the low-low level set point. indicator scale) relative to the minimum normal water level (302'9").

In addition to the requirement that at least one licensed Operator be in the control room when fuel is in the e. The reactor water low-low level setting for core reactor, there shall be another control room operator spray initiation shall be no less than -5 feet (5 present in the control room with no other duties than to inches indicator scale) relative to the minimum monitor the reactor vessel water level. normal water level (Elevation 302'9").

f. The reactor low pressure setting for main-steam line isolation valve closure shall be >_850 psig when the reactor mode switch is in the run position or the IRMs are on range 10.

g. The main-steam-line isolation valve closure scram setting shall be _ 10 percent of valve closure (stem position) from full open.

AMENDMENT NO. 142, 143, 153, 168 11

UM ITING CONDITION FOR OPERTION SURVBLLANCE REQUIREM ENT UMITNG ONDTIONFOROPEATIO SUVELANCEREQIRE EN 4.2.7 REACTOR COOLANT SYSTEMI ISOLATION VALVES 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Applicability:

Applicability:

Applies to the periodic testing requirement for the Applies to the operating status of the system of reactor coolant system isolation valves.

isolation valves on lines connected to the reactor coolant system.

Obiective:

Obiective:

To assure the capability of the reactor coolant system To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear event of a rupture of a line connected to the nuclear steam supply system.

steam supply system.

Specification:

Specification:

The reactor coolant system isolation valves

a. During power operating conditions whenever the surveillance shall be performed as indicated below.

reactor head is on, all reactor coolant system isolation valves on lines connected to the reactor

a. At least once per operating cycle the operable coolant system shall be operable except as automatically initiated power-operated isolation specified in "b" below.

valves shall be tested for automatic initiation and closure times.

b. In the event any isolation valve becomes inoperable the system shall be considered b. Additional surveillances shall be performed as operable provided at least one valve in each line required by Specification 6.5.4.

having an inoperable valve is in the mode corresponding to the isolated condition, except as noted in Specification 3.1.1.e.

AMENDMENT NO.142,4145, 173, 108

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT REQUIREMENT LIMITING CONDITION FOR OPERATION SURVEILLANCE 3.3.3 LEAKAGE RATE 4.3.3 LEAKAGE RATE Applicability: Applicability:

Applies to the allowable leakage rate of the primary Applies to the primary containment system leakage containment system. rate.

Objective: Obiective:

To assure the capability of the containment in limiting To verify that the leakage from the primary radiation exposure to the public from exceeding containment system is maintained within specified values specified in 10 CFR 100 in the event of a loss values.

of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a Specification:

metal-water reaction.

a. The primary containment leakage rates shall be To assure that periodic surveillances of reactor demonstrated at test schedules and in containment penetrations and isolation valves are conformance with the criteria specified in the 10 performed so that proper maintenance and repairs are CFR 50 Appendix J Testing Program Plan as made during the service life of the containment, and described in Specification 6.5.7. I systems and components penetrating primary containment. b. The provisions of Specification 4.0.1 are not applicable, and the surveillance interval Specification: extensions are in accordance with the 10 CFR 50 Appendix J Testing Program Plan.

Whenever the reactor coolant system temperature is above 215OF and primary containment integrity is required, the primary containment leakage rate shall be limited to:

AMENDMENT NO. 112, 159, 131

6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility 6.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The plant manager or a designee shall approve, prior to implementation, each proposed test and experiment not addressed in the UFSAR or Technical Specifications, and each modification to systems or equipment that affect nuclear safety.

6.1.2 The Station Shift Supervisor - Nuclear (SSS) shall be responsible for the control room command function. During any absence of the SSS from the control room while the unit is in the power operating or hot shutdown conditions, 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 SSS from the control room while the unit is in the cold shutdown or refueling conditions, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

6.2 Orgqanization 6.2.10Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. The organization chart and the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the UFSAR.

The functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions shall be documented in procedures.

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

AMENDMENT NO. 142, 157, 162 347

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

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

6.2.2 Unit Staff The unit staff organization shall include the following:

a. At least two non-licensed operators shall be assigned when the unit is in the power operating condition; and at least one non-licensed operator shall be assigned when the unit is in the hot shutdown, cold shutdown, or refueling conditions. In addition, if the process computer is out of service for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at least three non-licensed operators shall be assigned when the unit is in the power operating, hot shutdown, cold shutdown, or refueling conditions.

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

c. An individual qualified to implement radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence of on-duty personnel, provided immediate action is taken to fill the required position.

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

The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

AMENDMENT NO. 1442 348

Any deviation from the above guidelines shall be authorized in advance by the plant manager or the plant manager's designee, in accordance with approved administrative procedures, with documentation of the basis for granting the deviation. Routine deviation from the working hour guidelines shall not be authorized.

Controls shall be included in the procedures to require a periodic independent review be conducted to ensure that excessive hours have not been assigned.

e. As a minimum, either the Manager Operations or the General Supervisor Operations shall hold an SRO license.

f. The Shift Technical Advisor (STA) shall provide advisory technical support to the shift supervision in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

6.3 Unit Staff Qualifications 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for; the Manager Operations who, in lieu of meeting the senior reactor operator license requirements of ANSI N18.1-1971, shall 1) hold a senior reactor operator license at the time of appointment, or 2) have held a senior reactor operator license at Nine Mile Point Nuclear Station Unit 1 or at a similar unit, or 3) have been certified for equivalent senior reactor operator knowledge; and the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

6.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 6.3.1, perform the functions described in 10 CFR 50.54(m).

6.4 Procedures 6.4.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and cover the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 3, 1972; AMENDMENT NO. 142, 157, 158, 0, 162 349

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

c. Quality assurance for radioactive effluent and radiological environmental monitoring;

d. Fire Protection Program implementation; and

e. All programs specified in Specification 6.5.

6.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

6.5.1 Offsite Dose Calculation Manual (ODCM)

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

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 6.6.2 and Specification 6.6.3.

c. Licensee initiated changes to the ODCM:

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

(a) Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; AMENDMENT NO. 442 350

2. Shall become effective after the approval of the plant manager or a designee; and

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 in 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 (i.e., month and year) the change was implemented.

6.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include Core Spray, Containment Spray, Emergency Cooling, Shutdown Cooling, Reactor Cleanup, Vacuum Relief, Reactor Water Sampling, Containment Atmosphere Dilution (CAD) H20 2 Monitor, Drywell Containment Atmosphere Monitoring (CAM), Post Accident Sampling, Radioactive Gaseous Effluent Monitoring (RAGEMS) (the program requirements shall apply to the Post Accident Sampling System and RAGEMS until such time as administrative controls provide for continuous isolation of the associated penetration(s) or a modification eliminates the potential leakage path(s)), Offgas Effluent Stack Monitoring (OGESMS), and Post Accident Vent to Reactor Building Emergency Ventilation. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. System leak test requirements for each system at 24 month intervals.

The provisions of Specification 4.0.1 are applicable to the 24 month frequency for performing system leak test activities.

6.5.3 Radioactive Effluent Controls Program This program conforms to 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 shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

AMENDMENT 142,-165 351

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

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402;

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

d. Limitations on the annual and quarterly doses or dose 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. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate *500 mrems/yr to the whole body and a dose rate

• 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days: a dose rate *1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary; conforming to 10 CFR 50, Appendix I; AMENDMENT NO. 142, 157, 1*623 352

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

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190; and

k. Limitations on venting and purging of the primary containment through the Emergency Ventilation System to maintain releases as low as reasonably achievable.

The provisions of Surveillance Requirement 4.0.1 are applicable to the Radioactive Effluent Controls Program surveillance frequencies.

6.5.4 Inservice Testing Program This program provides controls for inservice testing of Quality Group A, B, and C pumps and valves.

a. Inservice testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with requirements for American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components specified in Section XI of the applicable ASME Boiler and Pressure Vessel Code Edition and Addenda, subject to the applicable provisions of I OCFR50.55a;

b. The provisions of Specification 4.0.1 are applicable to the normal and accelerated testing frequencies for performing inservice testing activities;

c. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

6.5.5 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

AMENDMENT NO. 142, 157, 162 353

The program shall include:

a. The limits for concentrations of hydrogen in the Main Condenser Offgas Treatment System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outside temporary liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is <10 Ci, excluding tritium and dissolved or entrained noble gases.

The provisions of Surveillance Requirement 4.0.1 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

6.5.6 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

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

b. Licensees may make changes to the 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 UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

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

d. Proposed changes that meet the criteria of 6.5.6.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

AMENDMENT NO. 112, 157, 162 354

6.5.7 10 CFR 50 Appendix J Testingi Program Plan

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled "Performance-Based Containment Leak-Test Program," dated September 1995 with the following exceptions:

1. Type A tests will be conducted in accordance with ANSI/ANS 56.8-1994 and/or Bechtel Topical Report BN-TOP-1, and

2. The first Type A test following approval of this Specification will be a full pressure test conducted approximately 70, rather than 48, months since the last low pressure Type A test.

b. The peak calculated containment internal pressure (Pac) for the design basis loss of coolant accident is 35 psig.

c. The maximum allowable primary containment leakage rate (La) at Pac shall be 1.5% of primary containment air weight per day.

d. Leakage Rate Surveillance Test acceptance criteria are:

1. The as-found Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than 1.0 La.

2. The as-left Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than or equal to 0.75 La, prior to entering a mode of operation where containment integrity is required.

3. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 La, calculated on a maximum pathway basis, prior to entering a mode of operation where containment integrity is required.

4. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 L,, calculated on a minimum pathway basis, at all times when containment integrity is required.

e. The provisions of Specification 4.0.1 do not apply to the test frequencies specified in the 10 CFR 50 Appendix J Testing Program Plan.

AMENDMENT NO. 142, 157, 1 355

6.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

6.6.1 Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent of > 100 mrems and the associated collective deep dose equivalent (reported in man-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling).

This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ion chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling <20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

6.6.2 Annual Radiological Environmental Operating Report*

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

AMENDMENT NO. 442 356

6.6.3 Radioactive Effluent Release Report*

The Radioactive Effluent Release Report covering the operation of the unit 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 and the waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I,Section IV.B.1.

6.6.4 Monthly Operating Reports no later than the Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis 15th of each month following the calendar month covered by the report.

6.6.5 Core Operating Limits Report (COLR) of a reload

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

and

1. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specifications 3.1.7.a 3.1.7.e.

2. The Kf core flow adjustment factor for Specification 3.1.7.c.

3. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.1.7.c and 3.1.7.e.

4. The LINEAR HEAT GENERATION RATE for Specification 3.1.7.b.

5. The Power/Flow relationship for Specifications 3.1.7.d and 3.1.7.e.

common to all units A single submittal may be made for a multiple unit station. The submittal should combine sections of at the station; however, for units with separate radwaste systems, the submittal shall specify the releases radioactive material from each unit.

357 AMENDMENT NO. 4-42,

previously reviewed and

b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the following documents:

Supplement, (NRC

1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," U.S.

approved version specified in the COLR).

fuel thermal mechanical limits,

c. The core operating limits shall be determined such that all applicable limits (e.g.,

nuclear limits such as shutdown core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, analysis are met.

margin (SDM), transient analysis limits, and accident analysis limits) of the safety upon issuance for each reload

d. The COLR, including any midcycle revisions or supplements, shall be provided cycle to the NRC.

6.6.6 Special Reports These reports shall be submitted Special reports shall be submitted within the time period specified for each report.

reference specification:

covering the activities identified below pursuant to the requirements of the applicable (12 months).

a. Reactor Vessel Material Surveillance Specimen Examination, Specification 4.2.2.(b)

b. (Deleted)

c. (Deleted)

d. (Deleted)

e. (Deleted)

f. (Deleted) 3.6.5.2 (Three months).

g. Sealed Source Leakage In Excess Of Limits, Specification Action 3 or4) (Within 14 days

h. Accident Monitoring Instrumentation Report, Specification 3.6.11.a (Table 3.6.11-2, following the event).

358 AMENDMENT NO. 4,1-,517, 4652

6.7 High Radiation Area areas in As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

6.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area, or

2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 359 AMENDMENT NO. 142,157

(ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

6.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1. All such door and gate keys shall be maintained under the administrative control of the Station Shift Supervisor - Nuclear, radiation protection manager, or his or her designee.

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

1, 157, 162 360 AMENDMENT NO. 142,4*

d. Each individual or group entering such an area shall possess one of the following:

1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

AMENDMENT NO. 142,143 361

THIS PAGE INTENTIONALLY BLANK 32 362 AMENDMENT NO. 14 2,16

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 142 2, 363 33

THIS PAGE INTENTIONALLY BLANK I AMENDMENT NO. 1.42,146 3 364

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 142, 413, 176 366

THIS PAGE INTENTIONALLY BLANK 369 367 AMENDMENT NO. 14.2,6

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 142, 172, 1763 368

THIS PAGE INTENTIONALLY BLANK I AMENDMENT NO. 4-4.27 370

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 14.2,4 371

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 4-..L& 371a

THIS PAGE INTENTIONALLY BLANK 371b AMENDMENT NO. 446

THIS PAGE INTENTIONALLY BLANK 372 AMENDMENT NO. 142.,71.16

THIS PAGE INTENTIONALLY BLANK 372a AMENDMENT NO. 4-;-.6

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 142, 159, 1714 373

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 159, 173, 1763 374

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO...3-7-7 375

THIS PAGE INTENTIONALLY BLANK AMENDMENT NO. 4-77 376

ATTACHMENT 4 Current Technical Specifications (CTS) Markup and Discussion of Changes (DOCs)

Note: Changes to the CTS markup and the DOCs originally submitted by letter dated October 26, 2001 are identified by a vertical bar and an "A" in the right margin.

ATTACHMENT 4.1 Current Technical Specifications Markup and Discussion of Changes Revised Table of Contents

SEC*TION, DESCRIPTION SEC TION DESCRIPTION PAGE 5.0 Design Features 342 5.1 Site 342 5.2 Reactor 342 5.3 Reactor Vessel 342 5.4 Containment 345 5.5 Storage of Unirradiated and Spent Fuels 346 5.6 Seismic Design 346 6.0 Administrative Controls 347 6.1 Responsibility 347 6.2 Oaranization 347 6.3

• Staff Qualifications 349 -*

6.4 Tr ining 351

• eview nd Audit]

R 6.6/ortable Evt Action L6 E Safety Lim' Violatio

(.4 Procedures (0 ,g ero roan" *I' /,A V1 vo,.Is 34 36 Co Reporting Requirements AMENDMENT NO. X#

Ceo,*--*,+s SECTION f6.10 =Record R nntion

,11Off

'e Dose Calb tion Manual (2CM))

nmenrii PAGE G1 - High Radiation Area 373 373 6 16 10 50 Appendix J esting Program 7

6.17 Inservice Testing ogramr 34 6 8 Radioactive luent Controls Pr ram 6.19 Explosi' Gas and Storage ank Radioactivity onitoring Prog*m 37 AMEINDMENT NO 44Z, 464, 4-.*, / vi I&

aoPZ

DISCUSSION OF CHANGES REVISED TS: TABLE OF CONTENTS ADMINISTRATIVE (A)

A.1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" None "Specific" None Page 1 of 1

ATTACHMENT 4.2 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.1 Responsibility

6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility 6.1.1 The lant anager shall be responsible for overall unit operation and shall delegate in writing the succession to thi r o 1t uing his asence.

Uvt~Ser+~6.1-A 5~ .3M 6.1.2 The Station Shift Supervisor - Nuclea I'll, .,, . . . . ;shalle responsible for the contres ronm command function nA mal ee eie irstaVbTo THliesct, sian ighefst by thhe

-6.2 0 r: nzt n* managr

._' eUmen lels re-is nt*ermediat levels * =n.5 n ainclu

-an e- r'"J..

. ans1-sisn relatiOncdianization cat functi 6.2.1 An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the position for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. Those 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 personnel positions or in equivalent forms of documentation. The organization charts shall be documented in the Final Safety Analysis Report, and the functional descriptions of departmental responsibilities and relationships and job descriptions for key personnel positions are documented in procedures.

b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to assure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

c. The Plant Manager shall have responsibility for overall unit operation and shall have control over those resources necessary for safe operation and maintenance of the plant.

AMENDMENT NO. P I nrz 3 347

I Insert6.1-A CA.2LA The lant manager r a desinee hall approve, prior to implementation, each proposed test and experiment not addressed in the UFSAR or Technical Specifications, and each modification to systems or equipment that affect nuclear safety.

Insert 6.1-B (M3 During any absence of the SSS from the control room while the unit is in the power operating or hot shutdown conditions, 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 SSS from the control room while the unit is in the cold shutdown or refueling conditions, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

120g~ 4~

Seet

" /. spec'ýý', G.,

411c a6.5.2.3 modifications to unit structures, systems and components that affect nuclear safety shall be designed by Proposed a qualified individual/organization. Each such modification shall be reviewed by an individual/group other than the indi vi dual /group which designed the modification, but who may be from the same organization as the* indivdu I A".

qru hihdsindth oifction.j(p-e molcanststrtue ytesndom-ponents **

,*,*,,,~~~~~0m I ... ao,*,---

I e-5, e eh LA 1

--. 5.2.4 Individuals responsible for reviews performed in accordance with Specifications 6.5.2.1, 6.5.2.2 and 6.5.2.3 shall be members of the station supervisory staff, previously designated by the Plant Manager to perform such reviews.

Each such review shall include a determination of whether or not additional,,cross-disciplinary, review is necessary.

If deemed necessary such review shall be performed by the appropriate designated station review personnel. A, Proposed tests and experiments which affect station nuclear safety and re..otddressed in the FSAR or Techr Specifications 9h =shall be reviewed by the *lant Manage , or @2-240" 6.5.2.6 The Plant Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President - Nuclear Generation.

6.5.2.7 The facility security program, and implementing procedures, shall be reviewed at least every 1 2 months.

Recommended changes shall be approved by the Plant Manager and transmitted to the Vice President - Nuclear Generation and to the Chairman of the Safety Review and Audit Board. I 6.5.2.8 The facility emergency plan, and implementing procedures shall be reviewed at least every 12 months.

Recommended changes shall be approved by the Plant Manager and transmitted to the Vice President - Nuclear Generation and to the Chairman of the Safety Review and Audit Board.

V AMENDMENT NO. WL Ui, /35 or3 4 -3 354

DISCUSSION OF CHANGES REVISED TS: 6.1 - RESPONSIBILITY ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 The requirements of CTS 6.5.2.3 and CTS 6.5.2.5 regarding Plant Manager reviews and approvals of proposed tests, experiments, and modifications to systems or equipment that affect nuclear safety are proposed to be moved to Revised TS 6.1, "Responsibility,"

except that the phrase "and their safety evaluations" would be deleted rather than relocated. Approval of the safety evaluation is inherent in the approval of the modification, test, or experiment; therefore, a separate requirement to approve the safety evaluation is not necessary. This change is consistent with NUREG-1434, Revision 1.

A.3 The acronym "SSS" has been added for the Station Shift Supervisor-Nuclear position title. This is strictly an editorial change.

A.4 CTS 6.1.2 requires a management directive to be reissued annually to all station personnel stating that the Station Shift Supervisor-Nuclear is responsible for the control room command function. This management directive requirement is being deleted. CTS 6.1.2 and Revised TS 6.1.2 state who is responsible for the control room command function. This requirement appears to serve only as a "reminder" to personnel as to who is in charge. Nowhere else in TS is a management directive required to remind personnel of a TS requirement, and this requirement is not considered to be one of the more important requirements (as it does not directly impact a safety margin). Since the TS responsibility requirement is not being changed, this deletion is considered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. 1 CTS 6.1.2 identifies the Station Shift Supervisor - Nuclear (or during his absence, a designated individual) as responsible for the control room command function. The proposed change would delete the phrase "(or during his absence from the control room, a designated individual)," and add a requirement that an individual with either an active Senior Reactor Operator license or Reactor Operator license (depending on the unit operating condition) shall be designated to assume the control room command function.

This change more clearly specifies the qualifications of the individual designated to assume the control room command function. This is an additional restriction on plant operation and is consistent with NUREG-1434, Revision 1.

Page 1 of 2

DISCUSSION OF CHANGES REVISED TS: 6.1 - RESPONSIBILITY TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA. 1 CTS 6.1.1 uses the title "Plant Manager." This specific title is replaced with the generic title "plant manager." The specific title is proposed to be relocated to UFSAR Section XIII-A, which is where the organizational chart and description of this specific title is currently located. Relocation of specific titles out of the TS is consistent with the NRC letter from C. Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994, as documented in NRC-approved TSTF-65, Revision 1. The various requirements of the individuals are still retained in the Revised TS. In addition, Revised TS 6.2.1 requires the organization chart to be documented in the UFSAR. Therefore, the relocated specific titles are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

"Specific" L. 1 CTS 6.5.2.3 and CTS 6.5.2.5 currently identify the Manager Technical Support as the designated alternate to the Plant Manager for the approval of proposed modifications, tests, and experiments. In Revised TS 6.1.1, the phrase "the Manager Technical Support as previously designated by the Plant Manager" that is currently contained in CTS 6.5.2.3 and CTS 6.5.2.5 is replaced with "a designee." This change provides additional flexibility while maintaining plant manager (changed to the generic title by Discussion of Change LA. 1 above) control over the designation of personnel performing these activities. This is consistent with CTS 6.1.1, which states that the Plant Manager is responsible for overall unit operation, and which allows the Plant Manager to designate an individual to take over this responsibility during the Plant Manager's absence. Since the plant manager is still maintaining this control, the removal of a specific titled individual to whom the plant manager delegates responsibility does not impact plant safety.

Page 2 of 2

ATTACHMENT 4.3 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.2 Organization

'see-TISC-USýiOyl4 4 SeroA-o' 4r Re-59o1Seai'JT 6.0 ADMINISTRATIVE CONTROLS 1

6.1 Responsibility 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Station Shift Supervisor - Nuclear (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Chief Nuclear Officer shall be re-issued to station personnel on an annual basis.

6.2 Organization Onsite an d 0ffsit~e 0rqanizto**'e--*'**

"T6.2.1 ii W insite and Jboffsite organization/*shall be established for unit operation and corporate managementqThe onsite and offsite organizatiorn*shall include the positiortf or activities affe hg.i

. ety of the nuclear owerIplant.

A "" A Jbe nuclear power p,,an

a. Lines of authority, responsibility and communication shall be establishedd -.Ei 15 highest management levels, intermediate levels..Q and all operating organization positions.

relationships shall be documented and updated, as appropriate, in Q organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions,

.or in equivalent forms of documentation, The organization chart shall be documented in the A A&

functional descriptions of departmental res-ponsibilities and relationships.and jo Sde oriotions fo(%key perso~nnel positions n*ted in procedures. A* *-L.i S*.!.aUER . ,shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to: acceptable performance of the staff in operating, maintaining, and providing r technical support

• the-plant*,I nuclear safety.6 a d SThe lant *anage- shall oe r-go t for overall operation s a ave control over those necessary for safe operation and maintenance of the plant.

AMENDMENT NO. WL, a't,3 347 PI et ) _ý 5

. TTc-QOYr kL.ievec) -J r1LArV.. Tk'e.1v Irepevidev -f--o1 er&h? p.ccc+/-.s*J 4syxAk4cLa.nAs 6.2.2 The unit organization shall - the following:

sJt TVn$i-+ C'A *efc n-dutsift sh e corn sed of"least thminimu J'la'/e ift cre ompos 4zciet n s o in Tabl .2-1.

7rsSeri-- G. 2- b. Aast on0 censed "peratorsol be in theeontrol ro when fJI is in theeactor. urin 4o.i re or - A)

(* / -- - *5"eratiory ,hi hl eI -se~doea* ent at tJ c ontrol,s~ h a'tf Sc.*..A~t L/' Ileas)4wo

shuldtown licqpsed Operý*ors and-'during shall/be presen)(

r~e very frorecrrps/'" th -6nrlrg~u-ri-n-grefiitor str-,chd*dece' S.* An individual qualified orlad iatlion protectiong procedures shall be on site when fuel is in the reactor C e A liceo r d .m* or Rea~or 0Ooer a or shall b~e" quir~ i~ro~ i*o11at the R oi d uring'li e oJ . hi'J-'

sh t'ow .....

P" uri w-" e a o"t d* =9=]

L~j~ nd wJT~ nthe em ergn c p an is/ ac tvated . I ii*m ay be t h e :lation S h t d prior ,11:le r orI t *ssistant S dtion Shift S~.rvisor -Nulear or anothe Senior Reacto Oprtr te j;6igpower ~perati~on orl When the *e~rgency pn is activated/during norma operations o ot shutdov* S..shutdow*. the /

IAsni s tagn t ation Shift pervisor - clear become the Shift To fnical Advis *, and the St ion Shift J SSu~pervi/ r - Nuclear, restricted t 'the control rog~until an ad litional licens# Senior Re tor 0peratnr/

akarriv , .-. .-

licens.edeenior S,*/K/

Reactor ~perator orlc e eir to Oprao /Liited to F 1!H a,n dIinn g shal I. drec~*/spervsedllll(a/censd S 0r RactrorO~ratr icndeior ti,'ited'to Operator Retor FuIeha irf*qall be drcynoItoe byt'mebe oftL eco nl/(rop. ..... *x e'u~etf resRditoný foreclliovn in tuofinew" idu rawl'm a bedls th in'mthe inl requndrem. 11 alperao/ftimeshall-b fore o(2 ous nM'ertoaco~mdaettexecKIasec~roiedi dit atjlfi t akety_!! rq ch ifp sitions l be

.AMENDMENT NO. /0 348 rovs; 4.- '- .Ccoe r.-cLr+

CD Pesos)l L

&.2.c Administrative procedures shalt be developed and implemented to limit the working hours of who perform safety~elated function e.g.. licensed ,_lice licnsed Operator~

auxiliary operators~and key mailnenrance personnej. k ro a ll~-o Po+exeho" P'i "IC CA-i+/-.I

/!tended period of shutdown fo refueling, majo maintenance or major plant modi c ations on a te orary

1) A individual shout ot be permitt to work more an 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> stra ht (excluding s t turnover L.A. (

An individua should not be rmitted to work ore than 16 ho in any 24-hour eriod, nor more ani 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />si any 48-hour niod, nor more t n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in an 7 day period (a excluding shift t nover Any dxevit iofrm n thea oed ghutdoinperiodsauthoiedofoethmeby san dnage oqsdre na ni baisnd acordance wnith roeurs wit W metano of thft asA orngrntngth LA. (o deviation. ~ontr sa ncAU1 u - FTU ` ov1 lm-sa erv o h e Pr ident - clear :er~atioLn ~ ign oassEr t exc jehour av no n assi 6,2-DRoutine devia* frorm the E~Rguidelines authorized.

~,,2. ~ As a minimum, either the Manager Operations or the General Supervisor Operations shalt hold AMENDMENT NO. Wigi, HI.i, W / 349 Pa~a3 6P

Insert 6.2-A -41OV *,

a. At least two non-licensed operators shall be assigned when the unit is in the power operating condition; and at least one non licensed operator shall be assigned when the unit is in the hot shutdown, cold shutdown, or refueling conditions. In addition, if the process computer is out of service for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at least three non-licensed operators shall be assigned when the unit is in the power operating, hot shutdown, cold shutdown, or refueling conditions.

Insert 6.2-B E D

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

Insert 6.2-C The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

Insert 6.2-D Controls shall be included in the procedures to require a periodic independent review be conducted to ensure that excessive hours have not been assigned.

Insert 6.2-E CE. A9 su ervs ion in the areas of thermal

f. The Shift Technical Advisor (STA) shall provide advisory technical support to the shift hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.n addition, the STA shall meet) A the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

As t. ation Shift lupervisor (S t TechnicaF Ad or Functio Senior Ope aor License) otes:

(,W At a one ti

• more Ij nsed op'nlicensed erating Wople coo be pres t for mfenanc epairs, re"I outag , etc. A (2) Those operating personnel not holding an "Operator" or "Senior Operator" License. A.4 (3) For operation longer than eight hours without process computer. LA.3 (4*y aot shutd n conditi only. For d shutdow and refuel condition nly one s r operat nd one op or are r ired to be on ift.

(5i) An additio l Senior R ctor Opera or SeniorR^actor Op tor Limite a Fuel Ha ing whRo s no other ocurrentI-, "2 Srespon~ilities shal upervise al ore altera ns.

(6)

• Shift Orew *omposition may be IBtless than the minimum requirements of Table 6.2-1 for a period of time not to exceed )

hours !n order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to, restorethe AMENDMENT NO. 1# p*.e._ 5o" £ 350 350

DISCUSSION OF CHANGES REVISED TS: 6.2 - ORGANIZATION ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 CTS 6.2.2.d uses the phrase "qualified in" as it relates to radiation protection procedures.

In Revised TS 6.2.2.c, this phrase is replaced with "qualified to implement," consistent with the NMP2 ITS. This is an administrative change that does not alter the qualifications of the identified individual.

A.3 CTS 6.2.1.d uses the term "health physics." In Revised TS 6.2.1.d this term is replaced with "radiation protection." CTS 6.2.2.h uses the term "health physicists." In Revised TS 6.2.2.d this term is replaced with "key radiation protection personnel." The change in terminology is considered administrative and is consistent with Revised TS 6.2.2.c and the current organization.

A.4 CTS Table 6.2-1, including Notes (2) and (3), contains requirements for unlicensed operating personnel. These requirements are moved to Revised TS 6.2.2.a and presented in text form rather than the tabular form of CTS Table 6.2-1. For the case where the process computer is out of service for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the specific operating conditions for which three unlicensed operators shall be assigned are listed. Also, the term "unlicensed" is replaced with "non-licensed." These administrative changes do not alter the existing requirements, and are consistent with NUREG-1434, Revision 1.

A.5 CTS Table 6.2-1, Note (6), allows the shift crew composition to be less than the minimum requirements of CTS Table 6.2-1 under certain conditions. This requirement is moved to Revised TS 6.2.2.b. In addition, since CTS Table 6.2-1 is not being retained in the Revised TS, the reference to Table 6.2-1 is replaced with "10 CFR 50.54 (m)(2)(i) and Specification 6.2.2.a." This is consistent with the changes described in DOC LA.2 below and with NUREG-1434, Revision 1. These are administrative changes that do not alter the existing requirements.

Page 1 of 5

DISCUSSION OF CHANGES REVISED TS: 6.2 - ORGANIZATION A.6 Note (6) of CTS Table 6.2-1 does not allow any shift crew position to be unmanned upon shift change because an oncoming crewman scheduled to come on duty is late or absent.

Revised TS 6.2.2.b allows a period of time not to exceed two hours in order to accommodate unexpected absence of "on-duty" shift crew members. The term "on-duty" implies that the absence refers to on-duty shift crew members and not the oncoming crew. If anyone in the oncoming crew is not present, the "on-duty" person may not leave.

Therefore, the requirement of this footnote is covered in Revised TS 6.2.2.b. Since the minimum shift crew requirements continue to be maintained in Revised TS 6.2.2.b, deletion of this portion of the footnote is an administrative change.

A.7 Note (1) of CTS Table 6.2-1, which states that more operators can be assigned if needed, is deleted. The CTS table specifies the requirements of the minimum shift crew composition and thus it is not necessary to specify whether the requirements may be exceeded.

A.8 The specific qualification requirements of the Shift Technical Advisor (STA) contained in CTS 6.3.1 have been moved to Revised TS 6.2.2.f and have been modified to reference the Commission Policy Statement on Engineering Expertise on Shift. This change is consistent with NUREG-1434, Revision 1. Since the policy statement encompasses the current requirements, this change is considered administrative.

A.9 The person to whom the STA provides advisory technical support has incorporated a more generic statement than is indicated in NUREG-1434, Revision 1. In the NUREG, the STA is required to provide advisory technical support to the Shift Supervisor. This term for whom the STA supports was derived from the generic term in NUREG-0737, Item I.A.1.1. At NMP1, both an Assistant Station Shift Supervisor (ASSS) and a Station Shift Supervisor (SSS) are on the operating shift, and both hold senior operator licenses.

As noted in CTS Table 6.2-1, Note (7), normally the ASSS is a combined ASSS/STA; however, there may be instances when a shift may be staffed by two Senior Reactor Operators plus a dedicated STA. This dedicated STA would normally provide support to the ASSS, since the ASSS is normally the control room supervisor. However, when the ASSS is not in the control room, the SSS would assume control room supervisor duties.

Thus, the dedicated STA could provide support to either the SSS or the ASSS at the start of an event. To provide a more generic, but technically accurate, statement as to whom the STA provides technical support, the words "Shift Supervisor" used in NUREG-1434 have been replaced with "shift supervision." This change is consistent with the NMP2 ITS.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. 1 Revised TS 6.2.2.f is added to the TS to describe the duties of the Shift Technical Advisor. This is an additional restriction on plant operation and is consistent with NUREG-1434, Revision 1.

Page 2 of 5

DISCUSSION OF CHANGES REVISED TS: 6.2 - ORGANIZATION TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" LA.1 CTS 6.2.1.c and CTS 6.2.2.h use the title "Plant Manager." In Revised TS 6.2.1.c and Revised TS 6.2.2.d, this specific title is replaced with the generic title "plant manager."

CTS 6.2.1 .b uses the title "Chief Nuclear Officer." This specific title is replaced with the generic term "a specified corporate officer." The specific titles are proposed to be relocated to UFSAR Section XIII-A, which is where the organizational chart and description of these specific titles is currently located. Relocation of specific titles out of the TS is consistent with the NRC letter from C. Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994, as documented in NRC approved TSTF-65, Revision 1. The various requirements of the individuals are still retained in the Revised TS. In addition, Revised TS 6.2.1 requires the organization chart to be documented in the UIFSAR. Therefore, the relocated specific titles are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

LA.2 Details of the minimum shift crew requirements located in CTS Table 6.2-1 are proposed to be relocated to the UFSAR (Section XIII-A). The requirements of 10 CFR 50.54(k),

(1), and (m) adequately provide for shift manning. In 10 CFR 50.54(m)(2)(iii), it is required that "when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be present at the controls at all times." Further, 10 CFR 50.54(k) requires "An operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during operation of the facility." The minimum shift crew requirements for licensed operators and senior operators contained in CTS 6.2.2.a, CTS 6.2.2.b, CTS 6.2.2.e and CTS Table 6.2-1 will be met through compliance with these regulations and do not need to be repeated in the TS. This is consistent with NRC-approved TSTF-258, Revision 4. The minimum shift crew requirements for non-licensed plant equipment operators are transferred from CTS Table 6.2-1 to Revised TS 6.2.2.a. In addition, Revised TS 6.1.2 contains requirements for the control room command function, and Revised TS 6.2.2.f contains requirements for the Shift Technical Advisor (STA). The relocation of the details of the minimum shift crew requirements to the UFSAR is acceptable considering the controls provided by regulations, the remaining requirements in the TS, and the control of changes to procedures governing the conduct of operations under 10 CFR 50 Appendix B programs. Therefore, the relocated requirements are not required to be in the TS to provide adequate protection of the public health and safety.

Page 3 of 5

DISCUSSION OF CHANGES REVISED TS: 6.2 - ORGANIZATION LA.3 CTS 6.2.2.c requires two licensed Operators in the control room during reactor startup, scheduled reactor shutdown, and during recovery from reactor trips. In addition, CTS Table 6.2-1, including Note (4), requires two licensed Operators for the hot shutdown condition. These requirements are proposed to be relocated to the UFSAR. The requirement specifying the minimum number of operators in the control room is adequately controlled by the requirements of 10 CFR 50.54(k), (1), and (m), as discussed in DOC LA.2 above. The requirement for location of these operators is also already specified in current administrative procedures. Therefore, the relocated requirement is not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

LA.4 CTS 6.2.2.e and CTS Table 6.2-1, Note (7) specify staffing requirements when the emergency plan is activated. These requirements are proposed to be relocated to the Site Emergency Plan. Staffing requirements when the emergency plan is activated are documented in the Site Emergency Plan and in administrative procedures. As discussed in DOC LA.2 above, the regulations provide the staffing requirements during the power operating and hot shutdown conditions and are adequate since the personnel required during emergencies are specified. Therefore, the relocated requirement is not required to be in the TS to provide adequate protection of the public health and safety. Changes to the Site Emergency Plan are controlled by the provisions of 10 CFR 50.54(q).

LA.5 Details contained in CTS 6.2.2.f that require all Core Alterations to be supervised by either a licensed Senior Reactor Operator or a licensed Senior Reactor Operator Limited to Fuel Handling are proposed to be relocated to the UFSAR. These CTS requirements are contained in 10 CFR 50.54(m)(2)(iv) and do not need to be repeated in the TS to provide adequate protection of the public health and safety. In addition, CTS 6.2.2.f requires that all fuel moves be directly monitored by a member of the reactor analyst group. This requirement is also proposed to be relocated to the UFSAR. In 10 CFR 50.54(m)(2)(iv), the minimum requirements for moving reactor fuel are specified. It does not require a non-licensed member of the reactor analyst group (or any other type of engineer) to monitor fuel movement. This is an additional administrative requirement that does not need to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

LA.6 CTS 6.2.2.h contains requirements for working hour limits for facility staff who perform safety-related functions. CTS Section 6.2.2.h is proposed to be revised from specific working hour limits to administrative procedures to control working hours, consistent with NRC-approved TSTF-258, Revision 4. The proposed changes will provide reasonable assurance that impaired performance caused by excessive working hours will Page 4 of 5

DISCUSSION OF CHANGES REVISED TS: 6.2 - ORGANIZATION not jeopardize safe plant operation. Specific working hour limits are not otherwise required to be in the TS under 10 CFR 50.36(c)(5). Specific controls for working hours of reactor plant staff are described in procedures that require a deliberate decision-making process to minimize the potential for impaired personnel performance, and established procedure control processes will provide sufficient control of changes to that procedure.

These changes are consistent with the recommendations in the April 9, 1997 letter from C. Grimes to J. Davis, as documented in NRC-approved TSTF-258, Revision 4.

Additionally, the statement "Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Vice President-Nuclear Generation or designee to assure that excessive hours have not been assigned." is being deleted.

There is no guidance in Generic Letter 82-12 that discusses these additional controls.

The additional requirement to have the Plant Manager (or his designee) review individual overtime on a monthly basis is unnecessary since sufficient administrative controls and policies exist, as well as the role of the individuals' supervisors in supervising personnel prevent excessive use or abuse of overtime. Therefore, the working hour limits are not required to be in the TS to provide adequate protection of the public health and safety.

LA.7 Details of the operator license requirements in CTS 6.2.2.i for the specific positions of Station Shift Supervisor Nuclear and Assistant Station Shift Supervisor Nuclear, and the CTS requirement that only licensed individuals may direct licensed activities, are proposed to be relocated to the UFSAR (Section XIII-A). This level of detail is not necessary in the TS to provide adequate protection of the public health and safety. These details are adequately addressed by the requirements contained in 10 CFR 50.54(i), (j),

(k), (1), and (m) and by the qualification requirements in Revised TS 6.3.1. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

"Specific" L. 1 CTS 6.2.2.h currently provides a description of the individuals who can be designated by the Plant Manager to approve modifications to overtime requirements. The proposed change to CTS 6.2.2.h would replace the phrase "higher levels of management" with "the plant manager's designee." This change provides additional flexibility while maintaining plant manager (changed to the generic title by DOC LA. 1 above) control over the designation of personnel who can approve this activity. This is consistent with CTS 6.1.1, which states that the Plant Manager is responsible for overall unit operation, and which allows the Plant Manager to designate an individual to take over this responsibility during the Plant Manager's absence. Since the plant manager is still maintaining this control, the change does not impact plant safety. Therefore, this change is considered acceptable.

Page 5 of 5

ATTACHMENT 4.4 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.3 Unit Staff Qualifications

6.3

• Staff Qualificatio n }pr-4e+* ry A ,5 ,I positions, except for; the Manager Operations who, eet or exceed the in lieu of meeting minimum the seniorofreactor qualifications ANSI N18.1-1971 operator license for comparable 6.3.1 ANSI Each member N18.1-1971, of theshallunit 1) holdshall staff a *enior reactor operator license at the time of appointment, requirements of or 2) have held a senior h Ss e'n io r r e'a c t o r 'o pe rra t o r kk n o w le d g ;ea t hikm, 4- Ir, .. . . . .. ..' ' ý uiR*"

Regulatory Guidelicense reactor operator 1.8, Septemat NineberMile mint Nuclear Station Unit ..1 'or ; at o wahsimilar o s ha ll unit, 197 5. n d te Sift Technical A dvisor w hoshall Have c e e d been m e e toror 3)e xhave n s equivalent ua lific a tiofor th e q certified 9~f. .

/rscience or engineering or a pro fesion al engineer license a bach elo's degree in a physic-a l \

issued by exam ination and shall have received specific training iln )*

• se~~i- cto ;#c'v I.2.

te -

design, and response and analysis of the plant for transients and accidents.

11plant T ra in in- A retraining and

.4 6r4.1 G replacement training program for the facility staff shall be maintained "6,4 Training and shall meet or exceed the recommendations under thee :direectionn of thee Mannager "

10CFR Part 55, and shall include and requirements of Section 5.5 of ANSI N18.1-1971 and of familiarization with relevant industry operational experience.r t i0o Fr l 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager Training and Supervisor-Fire Protection, Nuclear and shall meet or exceed the requirements of Appendix R to 10CFR50.

6.5 Review and Audit S _ . o& ,

-oC TS ( 4v 6.5.1 Station Operations Review Committee (SORC)

Function 6.5.1.1 The Station Operations Review Committee shall function to advise the Plant Manager on all matters related to nuclear safety.

Composition 6.5.1.2 The SORC shall be composed of the:

Chairman: Plant Manager Vice Chairman/Member: Manager Operations Vice Chairman/Member: Manager Technical Support Member: Manager QA Operations Member: Manager Maintenance Member: Manager Chemistry Member: Manager Radiation Protection AMENDMENT No.UC/S,- Se 7 351

5pe_-14" .11 "6"?

Insert 6.3-A 6.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 6.3.1, perform the functions described in 10 CFR I &

50.54(m).

Pa e-

DISCUSSION OF CHANGES REVISED TS: 6.3 - UNIT STAFF QUALIFICATIONS ADMINISTRATIVE (A)

A.1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 The requirements in CTS 6.3.1 regarding the Shift Technical Advisor (STA) are proposed to be moved to Revised TS 6.2, "Organization." Technical changes to these requirements are addressed in the Discussion of Changes for Revised TS 6.2.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. 1 Revised TS 6.3.2 is added to clarify the qualification requirements for licensed Senior Reactor Operators and licensed Reactor Operators. Definitions in 10 CFR 55.4 state:

"Actively performing the functions of an operator or senior operator means that an individual has a position on the shift crew that requires the individual to be licensed as defined in the facility's technical specifications, and that..." Adding TS 6.3.2 ensures that there is no misunderstanding when complying with 10 CFR 55.4 requirements. This change is consistent with the recommendations in the April 9, 1997 letter from C. Grimes (NRC) to J. Davis (NEI), as documented in NRC-approved TSTF-258, Revision 4.

TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" LA.1 CTS 6.3.1 uses the title "Manager Radiation Protection." This specific title is replaced with the generic title "radiation protection manager." The specific title is proposed to be relocated to UFSAR Section XIII-A, which is where the organizational chart and description of this specific title are currently located. Relocation of specific titles out of the TS is consistent with the NRC letter from C. Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994, as documented in NRC approved TSTF-65, Revision 1. The various requirements of the individuals are still retained in the Revised TS. In addition, Revised TS 6.2.1 requires the organization chart to be documented in the UFSAR. Therefore, the relocated specific titles are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the procedures governing the conduct of operations, including the areas of organization, position titles, responsibilities, shift staffing, personnel qualifications and training programs, are controlled under 10 CFR 50 Appendix B programs.

"Specific" None Page 1 of 1

ATTACHMENT 4.5 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.4 Procedures

6.6 Reportable Occurrence Action . .

6.6.1 The following actions shall be taken for REPORTABLE EVENTS: CTSt (6. "Ref PtJ)e. Oc-urr ,ertck ".

a. The Commission shall be notified and a report submitted pursuant to the requirements of Sections 50.72 and 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the SORC and the results of this review submitted to the SRAB and the SVice President - Nuclear Generation.

6.7 Safety Limit Violation - _ dSe C' .4 roe,-,c S ' "L; 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The provisions of 10 CFR 50.36(c)(1)(i) shall be complied with immediately.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Vice President - Nuclear Generation and the SRAB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, within 30 days of the violation, and to the SRAB, and the Vice President - Nuclear Generation within 14 days.

• '* Procedures 4-4'o 1 *] Written proce res and administrative policies shall be established, implemented and maintained that meet or exceed the requirement nd recommendations of Sections 5.1 and 5.3 of ANSI N 18.7-19 72 and ,,* "' " "c",'£ --- " ^

1-fs c.d a. Written procedures shall be established, implemented, and maintained for activities involving the Fire Protec ion Proq ram S 4 Aim lwentation. ".

r j ,D 6.. Each pro dure and adtistrative polic 'of 6.8.1 abovea nd changes th to, shall be re ewed and approvy prior to implem tation by tnranch manag for the functi I area of the Ocedure or high levels of mana ent as governe yJ 7 adm' istrative pro dures. Each p cedure and ad inistrative poli of 6.8.1 abov hall be reviewe periodically as seorth dministrativ procedures.

ein AMENDMENT NO. UL1,I.I 3360k

Insert 6.4-A T+he eegnyoperating procedures required toipent the requirements ofNR 0737 and NUREG077 apliabl pocdures recommended in Regulatory Guide 1.33, Appendix A, Novem7ber ý3,1 97ý2; )ý Ca:Th Supemn

ý 1, as stated in Generic Letter 82-33;

c. Quality assurance for radioactive effluent and radiological environmental monitoring;

". Fire Protection Program implementation; and * '-ro, C-TS I.

8..0 (e. All programs specified in Specification 6.5. - )

pa Z- ~3-

6.8.3 emporary c nges to procedur of 6.8.1 above y be made provide

a. T intent of the origi procedure is not tered.

b. The change is ap oved by two me ers of the plant m gement staff, at lea one of whom holds Senior Reactor Operator's Lic se on the unit aff ted.

c. The chan is documented, iewed and approv within 14 days of i lementation by the anch, manager f the functio I area of the ro ure or higher leve of management as erned by administr ye procedures.

6.9

-- e,-- n- Reouirements ,*LA .I In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted in accordance with 10 CFR 50.4.

6.9.1 Routine Reoorts.

a. S.tartupBReprt. A summary report of plant startup and power escalation testing shall be submitted following. (1) receipt of an operating license, (2) amendment to the license involving a planned increase power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison oftthese values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

aL Uhp IUIjultUo010la U0 UUIIIILLUU off VYILlJ lYI I LUIlIUW1III UW7O IVJUIII .IULIUII UL IIVL IV

• LUO L t00%l IJIU W lWI 0- , 1 1 uaya following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

R-t~e-j ITS! 1-4.

4r AMENDMENT NO. W.* /// P, e- ~3 ~P-3 361

DISCUSSION OF CHANGES REVISED TS: 6.4 - PROCEDURES ADMINISTRATIVE (A)

A.1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 CTS 6.8.1 requires that written procedures and administrative policies be established, implemented, and maintained that meet or exceed the requirements and recommendations of Appendix A of Regulatory Guide 1.33. This requirement is proposed to be moved to a specific sub-item (Item a) within Revised TS 6.4.1. The specific version of the Regulatory Guide 1.33 (i.e., dated November 3, 1972) is also identified, which is consistent with NMPC statements of conformance contained in Amendment No. 1 to Application to Convert Provisional Operating License to Full-Term Operating License (November 1973) and in NMPC letter to the AEC dated November 16, 1973. Since the requirements remain unchanged, this is considered to be a format change only, and therefore is considered an administrative change.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. 1 Revised TS 6.4.1, Items b, c, and e are added to the TS. This change will assure proper procedure control for emergency operating procedures, quality assurance for radioactive effluent and radiological environmental monitoring, and the programs list in Revised TS 6.5. This is an additional restriction on plant operation and is consistent with NUREG 1434, Revision 1.

TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA. 1 CTS 6.8.2 describes details of procedure reviews and approvals, and CTS 6.8.3 describes requirements relating to temporary changes to procedures. The proposed change would relocate these requirements to the Nine Mile Point Nuclear Station Quality Assurance Topical Report (QATR). These changes are consistent with the guidance of AL 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995, and NUREG-1434, Revision 1. The administrative letter concluded that TS administrative quality assurance-related requirements may be relocated to licensee-controlled quality assurance programs. For NMP1, these requirements would be relocated in their entirety to the QATR, with changes only to the format. Requirements for the processes related to review and approval of procedures and changes to procedures are contained in 10 CFR 50 Appendix B, Criterion II and Criterion V; ANSI N18.7-1972; and ANSI/ASME NQA-1-1983, including 1983 Addenda. Relocation of these TS provisions to the QATR will provide adequate controls over procedure review and approval activities for NMP1. Thus, the Page 1 of 2

DISCUSSION OF CHANGES REVISED TS: 6.4 - PROCEDURES relocated details are not necessary to be in the TS to provide adequate protection of the public health and safety. Changes to the QATR are controlled by the provisions of 10 CFR 50.54(a).

"Specific" None Page 2 of 2

ATTACHMENT 4.6 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.5 Programs and Manuals

• 5.S* LJ*SSIoo .* * *** "* . o ,

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the SORC and the SRAB.

S Offsite Dose Calculation Manual (ODCM)

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

b. The ODCM shall also contain the radioactive'effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification and Specification

c. Licensee initiated changes to the ODCM: (o,)

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

(a) Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; AMENDMENT NO. (,o,,,5" PY, ,* MS 0 -371 prrm kc,1 AIJ e4, &

2. Shall become effective after the approval of the plant manager or a designee; and

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 in 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 (i.e., month and year) the change was implemented.

Hiah Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

F6.12 6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.,0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area, or

2. A radiation monitoring device that continuously integrates the radiation dose rates In the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or daSy r 4r.

AMENDMENT NO. I 371aa Rtvi&~ TrS; (.,7 Pa.-2

rs (4.

ih~se-rt &S- R Sf 0 d 6.13,1 An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site licensee personnel or an outside fire protection firm.

613, An inspection and audit by an outside qualified fire consultant shall be performed at intervals no greater than 3 years.

P~varCola'i6Lr45;Ot~Jre C0oA+aiedr~v Tr-er' At- .',-

W.iZ 0§f 5.-=Tereurmnsha ppy to the Post Accident SamplTing Sydstemn (PASS) untilsc time as administrative]

[controls provide for oni u u islto oft ea s cae p ntai ns) ra modification eliminates the potential leakage path(s). J

• 6.15 Iodine Monitorina t ov 0,+ 0Z, Procedures shall be established, implemented and maintained to meet or exceed the requirements and recommeidations of Section C  : 2.1.8.c of NUREG 0578.

6S. 1

• 10 CFR 50 Appendix J Testing Program Plan a.. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and A 10 CFR 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained In Regulatory Guide 1.163, entitled "Performance-Based Containment Leak-Test Program," dated September 1995 with the follo ce

1. Type A tests will be conducted in accordance with ANSI/ANS 56.8-1994 and/or Bechtel 1 bN- I U--1, and

2. The first Type A test following approval of this Specification will be a full pressure test conducted approximately 70, rather than 48, months since the last low pressure Type A test.

6, The peak calculated containment internal pressure (Pac) for the design basis loss of coolant accident is 35 psig.

A

c. The maximum allowable primary containment leakage rate (La) at Pac shall be 1.5% of primary containment air weight per day.

A 4 . Leakage Rate Surveillance Test acceptance criteria are:

A

1. The as-found Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than 1.0 L,.

2. The as-left Primary Containment Integrated Leak Rate Test (Type A Test) acceptance criteria is less than or equal to 0.75 L, prior to entering a mode of operation where containment Integrity is required.

3. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 L8, calculated on a maximum pathway basis, prior to entering a mode of operation where containment integrity is required.

AMENDMENT NO. 442-, 1- / P4.. 3 ©P S 373

4. The combined Local Leak Rate Test (Type B & C Tests including airlocks) acceptance criteria is less than 0.6 L., calculated on a minimum pathway basis, at all times when containment integrity is required.

The provisions of Specification 4.0.1 do not apply to the test frequencies specified in the 10 CFR 50 Appendix J Testing Program Plan.

I Inservice Testing Program This program provides controls for inservice testing of Quality Group A, B, and C pumps and valves. 1

a. Inservice testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with requirements for American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components specified in Section X1 of the applicable ASME Boiler and Pressure Vessel Code Edition and Addenda, subject to the applicable provisions of 10CFR50.55a;

b. The provisions of Specification 4.0.1 are applicable to the normal and accelerated testing frequencies for performing inservice testing activities;

c. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

SRadioactive Effluent Controls Program This program conforms to 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 shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

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

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 20.2402; AMENDMENT NO. 4.", 4 faie 374

6.5,3 (c..XJ) c Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose 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 1;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix 1;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate <500 mrems/yr to the whole body and a dose rate <3000 mrems/yr to the skin, and

2. For iodine-131, iodine-I 33, tritium, and all radionuclides in particulate form with half lives greater than 8 days: a dose rate <1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary; conforming to 10 CFR 50, Appendix I; L Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives >8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190; and AMENDMENT NO.

k. Limitations on venting and purging of the primary containment through the Emergency Ventilation System to maintain releases as low as reasonably achievable.

The provisions of Surveillance Requirement 4.0.1 are applicable to the Radioactive Effluent Controls Program surveillance frequencies.

SExplosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a, The limits for concentrations of hydrogen in the Main Condenser Offgas Treatment System and a surveillance program to ensure ,the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e.,

whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in all outside temporary liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is *10 Ci, excluding tritium and dissolved or entrained noble gases.

The provisions of Surveillance Requirement 4.0.1 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

AMENDMENT NO. #/ 376 tI2 p e- (4cd8 J

Insert 6.5-A This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include Core Spray, Containment Spray, Emergency Cooling, Shutdown Cooling, Reactor Cleanup, Vacuum Relief, Reactor Water Sampling, Containment Atmosphere A.5 Dilution (CAD) H2-0 2 Monitor, Drywell Containment Atmosphere Monitoring (CAM), Post Accident Sampling, Radioactive Gaseous Effluent Monitoring (RAGEMS) (the program requirements shall apply to the Post Accident Sampling System nd RAGEM until/

such time as administrative controls provide for continuous isolation of the associated penetration(s) or a modification eliminates the potential leakage path(s)), Offgas Effluent Stack Monitoring (OGESMS), and Post Accident Vent to Reactor Building Emergency Ventilation. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. System leak test requirements for each system at 24 month intervals. A-4 The provisions of Specification 4.0.1 are applicable to the 24 month frequency for performing system leak test activities.

Insert 6.5-B 6.5.6 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

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

b. Licensees may make changes to the 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 UFSAR or Bases that requires NRC approval pursuant tto10 CFR 50.59.

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

d. Proposed changes that meet the criteria of 6.5.6.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

DISCUSSION OF CHANGES REVISED TS: 6.5 - PROGRAMS AND MANUALS ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 NUREG-1434, Revision 1, and the NMP2 ITS state that licensees may make changes to the TS Bases without prior NRC approval provided the changes do not involve "A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59." The proposed change would revise the quoted phrase to "A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59." This change is consistent with the changes to 10 CFR 50.59 published in the Federal Register (Volume 64, Number 191) dated October 4, 1999, as noted in NRC-approved TSTF-364, Revision 0. The final rule clarifies the specific types of changes, tests, and experiments conducted at a licensed facility that require evaluation, and revises the criteria that licensees must use to determine when NRC approval is needed before such changes, tests, or experiments can be implemented. The final rule also adds definitions of terms that have been subject to differing interpretations, and reorganizes the rule language for clarity. This change to Revised TS 6.5.6 is administrative in nature. IL A.3 CTS 6.14, "Systems Integrity", contains a brief statement indicating that the requirements and recommendations of Section 2.1.6.a of NUREG-0578 will be met or exceeded. In Revised TS 6.5.2, this statement is replaced with a more descriptive paragraph that outlines the elements of the program, and lists the systems to which the program applies.

The revised program description is consistent with Section 2.1.6.a of NUREG-0578 and NUREG-1434. These are administrative changes that do not alter the existing requirements. In addition, License Amendment No. 174 dated August 26, 2002 added the following sentence to CTS 6.14:

"The requirements shall apply to the Post Accident Sampling System (PASS) until such time as administrative controls provide for continuous isolation of the associated penetration(s) or a modification eliminates the potential leakage path(s)."

This sentence has been incorporated into Insert 6.5-A for Revised TS 6.5.2 as a parenthetical phrase, in the same manner as NMP2 ITS 5.5.2 (as revised by License Amendment No. 106 dated August 9, 2002). Minor wording changes have been included to be consistent with the NMP2 ITS (i.e., addition of the word "program" and deletion of the "PASS" acronym).

Page 1 of 2 Revision A

DISCUSSION OF CHANGES REVISED TS: 6.5 - PROGRAMS AND MANUALS A.4 A statement of applicability of Specification 4.0.1 has been added to CTS 6.14 (Revised TS 6.5.2). This statement is needed to maintain allowances for Surveillance Interval extensions contained in the TS, since Specification 4.0.1 is not normally applied to intervals identified in the Administrative Controls section of the TS. Since this change is a clarification required to maintain provisions that would be allowed in the Limiting Conditions for Operation sections of the TS, it is considered administrative in nature.

A.5 By letter dated September 11, 2002, the NRC accepted the use of the Offgas Effluent Stack Monitoring System (OGESMS) instead of the Radioactive Gaseous Effluent Monitoring System (RAGEMS) for accident monitoring of noble gases and particulates.

To reflect this change, the sentence added to CTS 6.14 by License Amendment No. 174 (see Discussion of Changes A.3 above) is modified to include RAGEMS. As with PASS, this change makes clear that CTS 6.14 remains applicable to RAGEMS as long as it is a potential leakage path. Any modifications to eliminate RAGEMS as a potential leakage path would be controlled by the provisions of 10 CFR 50.59. This change is considered administrative in nature since it incorporates the effects of a change previously reviewed and approved by the NRC.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

M. 1 This change proposes to add new Section 6.5.6, Technical Specifications (TS) Bases Control Program, to Revised TS Section 6.0. This program is provided to specifically delineate the appropriate methods and reviews necessary for a change to the TS Bases.

The proposed program is identical to NMP2 TS Section 5.5.10, which was issued by the NRC in NMP2 License Amendment No. 91, except as noted in Discussion of Change A.2 above. This change does not revise any safety limits, limiting conditions for operation or surveillance test requirements for the plant. TS Bases are not considered part of the TS as documented in 10 CFR 50.36.

TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" None "Specific" None Page 2 of 2 Revision A

ATTACHMENT 4.7 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.6 Reporting Requirements

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:* ..

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed and approved within 14 days of implementation by the branch manager for the Sfunctional area of the procedure or higher levels of management as governed by administrative procedures.

jg. Reporting Reauirements D

,4-1U.idi with "-___

10 CFR 50.4. .__-r,__, ^e theO-followingoeports shall be submitted in accordance 6.9.1 i LA.neRoot A ummary rep of plant startup an p~ower escalation te ing shall be subi ed following (1) r *ei~pt ofa prn iese, (2) am endment to the licen involving a planned *crease power is rl,((3) installation o /uel that ha irn ein or ha ~een manufactured/l a different fuel su ~lier, and (4) modi i ations that may h 9e signifi antly altered the clear, thermal, or hy aulic performance othe plant. The re rt shall address ea of the tests ide ifie i heFARt ,dshall.in general in/eud e description of te~measured values of the operatin cditions or aracteristics obtai d during the test prrram and a compari n of these values ith design predicti s and specifications. A corrective actions at were required to tamn satisfactory ration shall also described. Any additional spec* c details required in cense conditions ba d on other commi ants shall be incl ed in this report.

Startup re rts shall be submi d within (1) 90 days ollowing completio of the startup test rogram, (2) 90 days followir resumption or co encement of comm cial power opperatlo , or (3) 9 months f owing initial criticalit whic ver is earliest. If t Startup Report doe not cover all thee ents (i.e., initial cri ality, completion of artup te program, and resu ption or commence nt of commercial er operation), su ementary reports s be ubmitted at least e ry three months unti 11three events hay been completed.

AMENDMENT NO. P 4L@

v --.

• 7361

CAllD 4.(

Occupational Radiation Exposure Report, A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent of

> 100 mrems and the associated collective deep dose equivalent (reported in man-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ion chamber, thermoluminescence dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling

< 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

Monthly Ooerating Repor Routine reports of operating statistics and shutdown experience L S961W- .... ----. be submitted on a monthly basis; shall

?no later than the 15th of each month following the calendar month covered by the report.

AMENDMENT NO.

fage- f- 7 362 a

5 eLcRho (0 0 Annual Radiological Environmental Operating Regort*0 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The'material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

AMENDMENT NO.

Paý-3 4- 7 363 &

(A.I Seo Radioactive Effluent Release Reoort*

The Radioactive Effluent Release Report covering the operation of the unit 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 consistent with the objectives outlined in the ODCM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1, A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

AMENDMENT NO. e 4 f 364 )&

CORE OPERATING LIMITS REPO RT c CL5. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycl$for the following:

AN S

10. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for SpecificationX3.1.7.a and 3.1.7.e.

for Specification 3.1 .7.c.

2*J. The Kf core flow adjustment factor and 3.1.7.e.

RATIO (MCPR) for Specification*3.1.7.c 30), The MINIMUM CRITICAL POWER for Specification 3.1.7.b.

4e)o The LINEAR HEAT GENERATION RATE

.7.d and *{r3,,7.I.

50. The Power/Flow relationship for Specification~(3.1 PAI, I!_ =.* CoL1 and shall be documented in the *R the core operating limits shall beA those previously reviewed and approved k*e. The analytical methods used to determine I in the following documents6*

by the NRC, specifically those described NON~ 4~

AMENDMENT~~~ 6

c(D 44, PPLICA ON METHODOLO Y FOR NON-JET P P PLANTS" fk " ENERAL /ELECTRI COMPANY ANAL ICAL MODEýL I ORDANCE WITH CFR5O APPENDI ". (Latest ove "REACTOR ST ILITY DETECT A SUPPRESS S UTIONE FOR RELOA APPLICATIONS,- ugust 1996.

_ C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermaonechanical limits, core IA thermaoiydraulic limits, CCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the sa ety a'alysis are met.

J*. The 2 0_ 1AN19 I %a - 1,g; WEI., including any miitcycle revisions or supplements.,shall be provide upon issuance for each reload cycleto the NRCo

/*6.9.2 FiProtection Profaram Reports 7; Noncomplia es with the Fire otection Program, Ks described in t Final Safety Anal is Report) that a8ersely affect t ability to hieve and main in safe shutdown i the event of a Cf shall be reported ' accordance wit"he requireme of in fý) .. ,0&I SSvs 2a F0 cl 1

AMENDMENT NO. Pop Po,#ee- ('. £7 367

("

Special Reports (A Special reports shall be submittedWQ .w.ithin the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Reactor Vessel Material Surveillance Specimen Examination, Specification 4.2.2(b) (12 months).

b, (Deleted)

c. (Deleted)

d. (Deleted)

e. (Deleted)

f. (Deleted)A

g. Sealed Source Leakage In Excess Of Limits, Specification 3.6.5.2 (Three months).

Atby c4 4  :ýs+e v.4)

AMENDMENT NO. 44-2-, 1--, #/X 368 1&

Pa"r 7 eP7

DISCUSSION OF CHANGES REVISED TS: 6.6 - REPORTING REQUIREMENTS ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

A.2 CTS 6.9.1.f, Item 2, identifies specific analytical methods used to determine the core operating limits that are documented in the COLR. The proposed change deletes the references to three (3) of the identified reports (NEDE-30966-P-A, NEDO-20556-P-A, and NEDO-32465-A), and retains only the reference to NEDE-2401 1-P-A. NEDE 24011-P-A now contains all of the methods reviewed and approved by the NRC for the NMP1 Loss of Coolant Accident analysis and for the Stability Analysis. Therefore, the references to the other three reports (NEDE-30966-P-A, NEDO-20556-P-A, and NEDO 32465-A) are redundant. This change is administrative in nature. There are no changes to the actual analytical methods being used.

A.3 CTS 6.9.1 .f, Item 4, requires that the Core Operating Limits Report (COLR) shall be provided to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector, CTS 6.9.1.c requires that monthly operating reports be submitted in accordance with 10 CFR 50.4, and CTS 6.9.3 requires that special reports be submitted "in accordance with 10 CFR 50.4 Regional Office." Revised TS 6.6 contains a single statement that requires submittal of reports in accordance with 10 CFR 50.4. The TS do not need to give report submittal details since this material is subject to change and would require a change to the TS. The Revised TS submittal requirements are sufficient without including unnecessary duplication or details.

A.4 CTS 6.9.3 states that special reports shall be submitted covering identified activities, pursuant to the requirements of the applicable referenced specifications. Special reports required by Table 3.6.11-2 of CTS 3.6.11, "Accident Monitoring Instrumentation," are not currently identified in CTS 6.9.3. For completeness, an Accident Monitoring Instrumentation Report item is added to CTS 6.9.3, with the proper reference made to Specification 3.6.11 .a and the applicable time period for the report specified. This change does not alter any existing reporting requirements and, therefore, is considered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None Page 1 of 3 Revision A

DISCUSSION OF CHANGES REVISED TS: 6.6 - REPORTING REQUIREMENTS TECHNICAL CHANGES - LESS RESTRICTIVE (L, LA)

"Generic" LA. 1 CTS 6.9.1 .a requires that a startup report be submitted detailing plant startup and power escalation testing following receipt of an operating license, an increase in licensed power level, installation of nuclear fuel with a different design of manufacturer than the current fuel, and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The proposed change would relocate this requirement to the UFSAR. The startup report required by CTS 6.9.1 .a provides the NRC with a mechanism to review the appropriateness of licensee activities after-the-fact, but there is no requirement for the NRC to approve the report. The quality assurance requirements of 10 CFR 50, Appendix B, and the Startup Test Program provisions contained in the UFSAR provide assurance that the listed activities will be adequately performed and that appropriate corrective actions, if required, are taken. Also, given that the report may be submitted to the NRC up to 90 days following completion of the respective milestone, report completion and submittal is clearly not necessary to assure operation of the unit in a safe manner for the interval between completion of the startup testing and submittal of the report. Thus, the startup report is not required to be in the TS to provide adequate protection of the public health and safety. Changes to the UFSAR are controlled by the provisions of 10 CFR 50.59.

LA.2 The details contained in CTS 6.9.1 .e regarding changes to the Process Control Program (PCP) are proposed to be relocated to the UFSAR. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Compliance with these regulations is required by the NMP1 operating license and, as such, relocation of the requirements regarding changes to the PCP from the TS does not affect the safe operation of the facility. Therefore, the relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the UFSAR are controlled by the provisions of 10 CFR 50.59.

LA.3 The details contained in CTS 6.9.2, "Fire Protection Program Reports," are proposed to be relocated to the UFSAR (Appendix 10A), where the program requirements currently reside. This program is required by an NMP1 commitment to Branch Technical Position APCSB 9.5-1, Appendix A, as stated in the UFSAR, Appendix B. Revised TS 6.4.1 will continue to require that procedures shall be established to implement and maintain the Fire Protection Program. This is consistent with Generic Letter 88-12, which allowed the Fire Protection Program requirements to be relocated to plant controlled documents.

Therefore, the relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Fire Protection Program changes are controlled by the provisions of Paragraph 2.D(7) of the Operating License.

Page 2 of 3 Revision A

DISCUSSION OF CHANGES REVISED TS: 6.6 - REPORTING REQUIREMENTS "Specific" L. 1 The reporting of safety and relief valve failures and challenges is based on the guidance in NUREG-0694, "TMI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states: "Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report." NRC Generic Letter 97-02, "Revised Contents of the Monthly Operating Report," requests the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics.

The generic letter does not specifically identify the need to report challenges to the safety and relief valves. As noted in NRC-approved TSTF-258, Revision 4, an NRC staff member (AEOD) was contacted and he indicated that this information was not required for the Performance Indicator Program and therefore would not need to be reported.

Based on this information, it is acceptable to delete the requirement to provide documentation of all challenges to safety relief valves or safety valves.

Page 3 of 3 Revision A

ATTACHMENT 4.8 Current Technical Specifications Markup and Discussion of Changes Revised TS 6.7 High Radiation Area

.FaA Sef r-Aic'x4i'o v, (o,7 P

Seeý- r4 EDj

2. Shall become effective after the approval of the plant manager or a designee; and

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 in 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 (i.e., month and year) the change was implemented.

I High Radiation Area

':7 As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

SHigh Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area, or

2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or AMENDMENT NO. //7 ea; e- I .c,P 4-371 aI

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose Information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area 3

have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

AMENDMENT NO. 4 PO;L C_ ;ý f 371 b 371 I&

(A.;)ýCAO 40,7

@ High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from anySurface Penetrated by the Radiation. but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, Inaddition:

1. All such door and gate keys shall be maintained under the administrative control of the Station Shift Supervisor - Nuclear, radiation protection manager, or his or her designee.

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b. Access to, and actvities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified In radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work In such areas.

d. Each individual or group entering such an area shall possess one of the following:

1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpolnt, or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual In the area, or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, AMENDMENTNO.442 ,-F 4- 372 AA

(i) Be under the surveillance, as specified In the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance, as specified in the RWP or equivalent, while In the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above, are impractical or determined to be Inconsistent with the "As Low As Is Reasonably Achievable' principle, a radiation monitoring device that continuously displays radiation dose rates In the area.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such Individuals, entry into such areas shall be made only after dose rates Inthe area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

AMENDMENT NO. p 4372a

DISCUSSION OF CHANGES REVISED TS: 6.7 - HIGH RADIATION AREA ADMINISTRATIVE (A)

A. 1 Editorial changes, reformatting, and revised numbering have been adopted to make the Revised TS consistent with the Nine Mile Point Unit 2 Improved Technical Specifications (which are consistent with the BWR Standard Technical Specifications, NUREG-1434, Revision 1).

TECHNICAL CHANGES - MORE RESTRICTIVE (M)

None TECHNICAL CHANGES - LESS RESTRICTIVE (L. LA)

"Generic" None "Specific" None Page I of I Revision A