ML25134A178

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Enclosure 2, Volume 17: Hope Creek Generating Station - Improved Technical Specifications Conversion ITS Chapter 5.0 Administrative Controls, Revision 1
ML25134A178
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 05/14/2025
From:
Public Service Enterprise Group
To:
Office of Nuclear Reactor Regulation
Shared Package
ML25134A156 List:
References
LR-N25-0040, LAR H24-02
Download: ML25134A178 (1)
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ENCLOSURE 2 VOLUME 17 HOPE CREEK GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision 1

LIST OF ATTACHMENTS

1.

ITS Section 5.1 - Responsibility

2.

ITS Section 5.2 - Organization

3.

ITS Section 5.3 - Unit Staff Qualifications

4.

ITS Section 5.4 - Procedures

5.

ITS Section 5.5 - Programs and Manuals

6.

ITS Section 5.6 - Reporting Requirements

7.

ITS Section 5.7 - High Radiation Area

ATTACHMENT 1 ITS 5.1, Responsibility

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

SECTION 6.0 ADMINISTRATIVE CONTROLS ITS A01 ITS 5.1

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.

6.1.2 The Senior Nuclear Shift Supervisor, 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 senior corporate nuclear officer shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include positions 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. These relationships shall be documented and updated, as appropriate in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the Hope Creek Generating Station Updated Final Safety Analysis Report and updated in accordance with 10 CFR 50.71(e).

b.

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

c.

The senior corporate nuclear 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 and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

UNIT STAFF 6.2.2 The unit organization shall be subject to the following:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1; HOPE CREEK 6-1 Amendment No. 97 ITS A01 ITS 5.1 See ITS 5.2 5

5 LA02 (SS)

LA01 5.0 5.1 5.1.1 5.1.2

TABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION SINGLE UNIT FACILITY POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SNSS*

1 1

NSS*

1 None NCO 2

1 EO 2

1 STA 1

None RPT 1

1 TABLE NOTATION SNSS -

Senior Nuclear Shift Supervisor with a Senior Reactor Operator license on the Unit NSS Nuclear Shift Supervisor with a Senior Reactor Operator license on the Unit NCO -

Nuclear Control Operator with a Reactor Operator license on the Unit EO Equipment Operator STA Shift Technical Advisor RPT Radiation Protection Technician Except for the Senior Nuclear Shift Supervisor, the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 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 of Table 6.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Senior Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 1, 2 or 3, an individual with a valid Senior Reactor Operator license shall be designated to assume the control room command function. During any absence of the Senior Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Reactor Operator license or Operator license shall be designated to assume the control room command function.

In cases where an individual has a Senior Reactor Operator's license on the unit, is a qualified STA, and has a Professional Engineers License by virtue of successful completion of the Professional Engineers examination or a bachelor's degree in a scientific, engineering, or engineering technology discipline from an accredited institution, the individual can serve in a dual role capacity as either the SNSS/STA or NSS/STA. (Note: For those individuals with a bachelor's degree in a scientific or engineering technology discipline, course work must have included physical, mathematical, or engineering science.) Otherwise, there shall be a qualified STA as well as a SNSS and NSS on-shift.

HOPE CREEK 6-5 Amendment No. 52 5.1.2 ITS ITS 5.1 A01 See ITS 5.2 See ITS 5.2 MODE A02 an active (SRO)

A03

DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY Hope Creek Page 1 of 3 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 5.1 uses the term OPERATIONAL CONDITION(S). ITS 5.1 uses the term MODE(S). This changes the CTS by incorporating the ITS MODE definition.

The purpose of CTS 5.1 is to establish the Operational Condition (i.e., ITS MODE) in which the specification is required. This change is acceptable because the CTS definition of an Operational Condition, any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2, and the ITS definition of MODE, corresponds to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel, define similar conditions. This change is designated as an administrative change and is acceptable because it does not result in a technical change to the CTS.

A03 CTS Table 6.2.2-1 Note, states, in part, that during any absence of the Senior Nuclear Shift Supervisor, an individual with a valid Senior Reactor Operator license shall be designated to assume the control room command function. ITS 5.1.2 states, in part, that an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. This changes the CTS from requiring a valid SRO license to requiring an active SRO license.

The purpose of Table 6.2.2-1 Note is to provide a requirement for the qualifications of an individual who is allowed to replace the Senior Nuclear Shift Supervisor. This is an enhanced presentation change because 10 CFR 50.54 and 10 CFR 55 define the qualification for a senior reactor operator allowed to perform the control room command function stating that for an SRO to have a valid license allowing them to actively perform their command function they must maintain an active status, i.e., an active SRO license. This change is designated as an administrative change and is acceptable because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None

DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY Hope Creek Page 2 of 3 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.1.2 uses the title "Senior Nuclear Shift Supervisor." ITS 5.1.2 uses the generic title "shift supervisor." This changes the CTS by moving the specific organizational titles to the Quality Assurance Topical Report (QATR) and replacing them with generic titles.

The purpose of CTS 6.1 is to ensure responsibilities for unit operation are delineated. The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific HCGS organizational titles out of the Technical Specifications is consistent with the NRC letter from C Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994.

The various requirements of the plant manager and shift supervisor are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the QATR. Any changes to the QATR are controlled pursuant to 10 CFR 50.54(a)(3), which ensures changes are properly evaluated and the NRC is updated in accordance with 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA02 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.1.2 requires a management directive identifying the designated individual responsible for the control room command function during the shift supervisor s absence, signed by the senior corporate nuclear officer and reissued to all station personnel on an annual basis. This requirement is being relocated to the UFSAR.

The purpose of CTS 6.1.2 is to identify the designated individual responsible for the control room command function. ITS 5.1.2 retains this requirement but does not include the procedural detail for an annual management directive stating who is a designated individual to act is the absence of the SS. This change is acceptable because the control room command function responsibility requirement is not being changed. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are controlled under 10 CFR 50.59, which ensures that changes are properly evaluated, and the NRC is updated in accordance with 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification requirements are being removed from the Technical Specifications.

DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY Hope Creek Page 3 of 3 LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Responsibility 5.1 General Electric BWR/4 STS 5.1-1 Rev. 5.0 Hope Creek Amendment XXX 4

5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

REVIEWER'S NOTES------------------------------------------------

1.

Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

2.

The ANSI Standard shall be the same ANSI Standard referenced in Section 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title.

Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan.

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

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

During any absence of the [SS] from the control room while the unit is in MODE 4 or 5, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

1 3

3 CTS 6.1.1 Table 6.2.2-1 2

6.1.2

JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY Hope Creek Page 1 of 1

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. Generic titles are used in the ITS consistent with the titles specified in the ISTS and based on ANSI/ANS 3.1-1981, "Selection, Qualification and Training of Personnel for Nuclear Power Plants.

As specified in ITS 5.2.1, the plant Quality Assurance Topical Report will contain the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications.

2. ISTS 5.5.1 requires the plant manager or his designee to approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. This requirement is not included in the ITS consistent with the Hope Creek Generating Station current licensing basis. In Operating License Amendment 97, dated March 21, 1997, (NRC ADAMS Accession No. ML011760502), PSEG relocated the requirements of Technical Specifications Section 6.5 to the Quality Assurance (QA) Program, including requirements for approval or disapproval of proposed tests, experiments, and modifications to plant systems or equipment that affect nuclear safety. In the safety evaluation accompanying License Amendment 97, the NRC concluded that the relocated requirements relating to administrative controls are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. In addition, the NRC found that sufficient regulatory controls exist under 10 CFR 50.54(a), or other applicable regulation to assure continued protection of the public health and safety. Accordingly, the NRC has concluded that these requirements could be relocated from the Technical Specification to the QA Program.
3. The ISTS contains bracketed information and/or values that are generic to General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.
4. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 2 ITS 5.2, Organization

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

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.

6.1.2 The Senior Nuclear Shift Supervisor, 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 senior corporate nuclear officer shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include positions 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. These relationships shall be documented and updated, as appropriate in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the Hope Creek Generating Station Updated Final Safety Analysis Report and updated in accordance with 10 CFR 50.71(e).

b.

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

c.

The senior corporate nuclear 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 and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

UNIT STAFF 6.2.2 The unit organization shall be subject to the following:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1; HOPE CREEK 6-1 Amendment No. 97 ITS A01 ITS 5.2 See ITS 5.1 5

5 5

5.2 5.2.1 5.2.1.a throughout including the generic titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications of the plant specified or perform these individuals 5.2.1.b 5.2.1.c 5.2.1.d include L01 5.2.2 5

Quality Assurance Topical A02

ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

b.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Reactor Operator shall be in the control room;

c.

ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

HOPE CREEK 6-2 Amendment No. 177 ITS A01 ITS 5.2 L03

a.

A non-licensed operator shall be assigned to the unit when the reactor contains fuel and an additional non-licensed operator shall be assigned to the unit when the reactor is operating in MODES 1, 2, and 3.

A03 L02 S2

FIGURE 6.2.1-1 DELETED HOPE CREEK 6-3 Amendment No. 21 A01 ITS 5.2

FIGURE 6.2.2-1 DELETED HOPE CREEK 6-4 Amendment No. 21 A01 ITS 5.2

TABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION SINGLE UNIT FACILITY POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SNSS*

1 1

NSS*

1 None NCO 2

1 EO 2

1 STA 1

None RPT 1

1 TABLE NOTATION SNSS -

Senior Nuclear Shift Supervisor with a Senior Reactor Operator license on the Unit NSS Nuclear Shift Supervisor with a Senior Reactor Operator license on the Unit NCO -

Nuclear Control Operator with a Reactor Operator license on the Unit EO Equipment Operator STA Shift Technical Advisor RPT Radiation Protection Technician Except for the Senior Nuclear Shift Supervisor, the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 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 of Table 6.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Senior Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 1, 2 or 3, an individual with a valid Senior Reactor Operator license shall be designated to assume the control room command function. During any absence of the Senior Nuclear Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Reactor Operator license or Operator license shall be designated to assume the control room command function.

In cases where an individual has a Senior Reactor Operator's license on the unit, is a qualified STA, and has a Professional Engineers License by virtue of successful completion of the Professional Engineers examination or a bachelor's degree in a scientific, engineering, or engineering technology discipline from an accredited institution, the individual can serve in a dual role capacity as either the SNSS/STA or NSS/STA. (Note: For those individuals with a bachelor's degree in a scientific or engineering technology discipline, course work must have included physical, mathematical, or engineering science.) Otherwise, there shall be a qualified STA as well as a SNSS and NSS on-shift.

HOPE CREEK 6-5 Amendment No. 52 ITS A01 ITS 5.2 5.2.2.a 5.2.2.e 5.2.2.c 5.2.2.b 5.2.2.c See ITS 5.1 MODE 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e A05 L01 A04 L04 non-licensed operator an individual radiation protection technician LA01 A06 A06

ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 The Shift Technical Advisor shall provide advisory technical support to the Senior Nuclear Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the PSEG Nuclear Quality Assurance Topical Report.

6.3.2 The Operations Manager or Assistant Operations Manager shall hold a senior reactor operator license. The Senior Nuclear Shift Supervisors, and Nuclear Shift Supervisors, shall hold a senior reactor operator license. The Nuclear Control Operators shall hold a reactor operator license.

6.3.3 The Operations Manager shall meet one of the following:

(1)

Hold a senior reactor operator license, or (2)

Have held a senior reactor operator license for this or a similar unit (BWR), or (3)

Have been certified at an appropriate simulator for equivalent senior operator knowledge.

6.4 TRAINING 6.4.1 DELETED 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Fire Protection Manager and shall meet or exceed the requirements of the SRP (NUREG-0800)

Section 13.2.2.II.6, 10 CFR 50 Appendix R and Branch Technical Position CMEB 9.5.1, Section C.3.d.

6.5 REVIEW AND AUDIT (THIS SECTION DELETED)

HOPE CREEK 6-6 Amendment No. 233 ITS A01 ITS 5.2 5.2.2.d An individual 5.2.2.e 5.2.2.e L05 LA01 5.2.2.d LA01 This individual meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift A07 See ITS 5.3 See ITS 5.3 LA02

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 1 of 8 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.2.1.a requires, in part, that the requirements listed shall be documented in the Hope Creek Generating Station Updated Final Safety Analysis Report and updated in accordance with 10 CFR 50.71(e). ITS 5.2.1.a states that these requirements shall be documented in the Quality Assurance Topical Report (QATR) and does not explicitly require update of the UFSAR in accordance with 10 CFR 50.71(e). Additionally, ITS 5.2.1.a includes a clarifying statement that generic titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the QATR.

This changes the CTS by changing the location where the listed requirements will be located and clarifying that the generic titles delineated in Technical Specifications be referenced in the QATR for the applicable positions.

The purpose of CTS 6.2.1 is to identify requirements for onsite and offsite organizations associated with unit operation and corporate management including positions for activities affecting the safety of the nuclear plant.

CTS 6.2.1.a requires that lines of authority, responsibility, and communication be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. ITS 5.2.1.a similarly requires 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. ITS 5.2.1.a also includes a clarifying statement that, These requirements including the generic titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the Quality Assurance Topical Report. CTS 6.2.1.a requires documentation of these requirements in the Updated Final Safety Analysis Report while ITS 5.2.1.a requires documentation of these requirements in the QATR. Changes to the QATR are controlled pursuant to 10 CFR 50.54(a)(3) and any changes to the plant quality assurance program description that do not reduce the commitments in the program description must be submitted to the NRC in accordance with the requirements of 10 CFR 50.71(e). This change is acceptable because similar onsite and offsite organizational operation and management requirements are being applied

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 2 of 8 consistent with the ISTS allowing for either location to be chosen for documentation because both locations provide similar change control processes and require NRC notification of changes pursuant to 10 CFR 50.71(e). This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS Table 6.2.2-1 provides requirements, in part, for the number of Equipment Operators (EOs) required during Operational Conditions (ITS MODES) 1, 2, 3, 4, or 5. At HCGS the equipment operator is a non-licensed operator. During Operational Condition 1, 2, or 3, two EOs are required as part of the minimum shift crew composition with one EO required during Operational Condition 4 or 5.

ITS 5.2.2.a provides a similar requirement stating that a non-licensed operator shall be assigned when the reactor contains fuel and an additional non-licensed operator shall be assigned when the reactor is operating in MODES 1, 2, and 3.

This changes the CTS by presenting the non-licensed operator requirements in paragraph format rather than table format.

The purpose of the CTS Table 6.2.2-1 is to provide minimum shift crew composition requirements to ensure operations can be performed safety. This change is acceptable because the change is in presentation only, changing the minimum non-licensed operator requirements from a tabular format to a paragraph format, and has no impact on the minimum shift crew composition necessary to ensure operations can be performed safety. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 6.2 uses the term OPERATIONAL CONDITION(S). ITS 5.2 uses the term MODE(S). This changes the CTS by incorporating the ITS MODE definition.

The purpose of CTS 5.1 is to establish the Operational Condition (i.e., ITS MODE) in which the specification is required. This change is acceptable because the CTS definition of an Operational Condition, any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2, and the ITS definition of MODE, corresponds to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel, define similar conditions. This change is designated as an administrative change and is acceptable because it does not result in a technical change to the CTS.

A05 CTS Table 6.2.2-1, Table Notation, states, in part, that except for the Senior Nuclear Shift Supervisor, the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 to accommodate unexpected absence of on-duty shift crew members. ITS 5.2.2 provides the same allowance but does not explicitly make exception to the Senior Nuclear Shift Supervisor. This changes the CTS by not explicitly stating the Senior Nuclear Shift Supervisor is an exception to the unexpected absence allowance.

The purpose of the Table Notation specification associated with shift crew composition absences is to accommodate for unexpected conditions requiring an absence of an on-duty shift crew member. CTS makes an exception for the S2

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 3 of 8 Senior Nuclear Shift Supervisor because this position is assigned responsibility for overall plant operation. It is unnecessary to explicitly state that the Senior Nuclear Shift Supervisor is excepted from the unexpected absence allowance because 10 CFR 50.54(m)(2)(ii) requires that each licensee have at its site a person holding a senior operator license for all fueled units at the site who is assigned responsibility for overall plant operation at all times there is fuel in any unit, thus ensuring no absences are allowed for this position. This change is designated as administrative because it does not result in technical changes to the CTS.

A06 CTS Table 6.2.2-1, footnote

  • includes a description of who specifically may provide advisory technical support to the unit operations shift crew and serve a dual-role SRO/STA position: an individual has a Senior Reactor Operator's license on the unit, is a qualified STA, and has a Professional Engineers License by virtue of successful completion of the Professional Engineers examination or a bachelor's degree in a scientific, engineering, or engineering technology discipline from an accredited institution, the individual can serve in a dual role capacity as either the SNSS/STA or NSS/STA. ITS 5.2.2.e requires an individual to provide advisory technical support to the unit operations shift crew but is revised to eliminate the details of who is qualified to assume the duties. This changes the CTS by eliminating the inference that the STA role is a separate shift crew position instead of a function and eliminating details regarding the individual assigned to provide technical advisory support to the unit operations shift crew and serve a dual-role SRO/STA position.

The purpose of the CTS requirement is to ensure engineering expertise is available on shift to provide advisory technical support. NRC Generic Letter 86-04 "Policy Statement on Engineering Expertise on Shift," promulgated the policy statement regarding engineering expertise on shift and established minimum qualification requirements and the requirements for eliminating the separate STA position. The NRC policy statement offers the licensees two options for meeting the NUREG-0737, "Clarification of TMI Action Plan Requirements," Enclosure 3, Item I.A.1.1 requirement regarding engineering expertise on shift and meeting licensed operator staffing requirements pursuant to 10 CFR 50.54(m)(2). Option 1 provides for elimination of the separate STA position by allowing licensees to combine one of the required SRO positions with the STA position into a dual-role (SRO/STA) function. Option 2 allows a licensee to continue use of an NRC approved STA program while meeting licensed operator staffing requirements. Per the policy statement, the Commission encourages licensees to move toward the dual-role (SRO/STA) function, with the eventual goal of the shift supervisor serving in the dual role. The NRCs policy statement also promulgated the educational requirements in this policy statement (50 FR 43621, October 28, 1985) consistent with the requirements specified in CTS Table 6.2.2-2, footnote *.

ITS 5.2.2.e eliminates the title of "Shift Technical Advisor (STA)," because the on shift advisory technical support function may be fulfilled by one or more of the other on-shift individuals. This change is necessary so that it does not imply that the STA and the shift supervisor must be different individuals. As a result of eliminating the STA "position," it is unnecessary to state when the STA "position" must be manned. This change is designated as administrative because an

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 4 of 8 individual with engineering expertise (i.e., dedicated individual or dual role individual) will continue to be required on shift meeting the same educational requirements as specified in CTS. Therefore, the change does not result in technical changes to the CTS.

A07 CTS 6.2.3.1 states, in part, that the Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room. ITS 5.2.2.e states that this individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. This changes the CTS by not including details specified in the Commission Policy Statement on Engineering Expertise on Shift, but instead references the Commission Policy statement in the ITS.

The purpose of CTS 6.2.3.1 is to ensure the individual assigned as the STA (i.e.,

individual providing technical support) meets minimum qualification requirements.

CTS provides the educational requirements in the specification whereas ITS references the NRCs policy statement that promulgated the educational requirements listing them in this policy statement (50 FR 43621, October 28, 1985). This change is acceptable because the policy statement encompasses the minimum qualification requirements for the STA specified in the CTS. This change is designated as administrative because the qualifications specified for the individual providing advisory technical support to the unit operations shift crew remain the same and does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS Table 6.2.2-1 Table Notation lists minimum shift crew composition by plant specific titles. CTS 6.2.3.1 specifies the plant specific title of Shift Technical Advisor. CTS 6.3.2 specifies the plant specific titles of Operations Manager and Assistant Operations Manager. ITS 5.2 adopts generic titles for these plant specific positions. This changes the CTS by relocating plant specific titles to the QATR.

Plant specific titles are not necessary to be defined in the technical specifications to provide adequate protection of the public health and safety. This change is

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 5 of 8 acceptable because the ITS retains the current requirements, as modified in other Discussion of Changes provided herein, using generic titles and ITS 5.2.1.a requires plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications be documented in the QATR.

The relocation of the details of plant specific titles is acceptable considering the controls provided by the requirements in ITS 5.2.1 and the QATR change control process pursuant to the requirements of 10 CFR 50.54(a)(3). This change is designated as a less restrictive removal of detail change because specific shift crew titles are being removed from the Technical Specifications.

LA02 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.4.2 states that a training program for the Fire Brigade shall be maintained under the direction of the Fire Protection Manager and shall meet or exceed the requirements of the SRP (NUREG-0800)

Section 13.2.2.II.6, 10 CFR 50 Appendix R and Branch Technical Position CMEB 9.5.1, Section C.3.d. ITS 5.2 does not include this requirement. This changes the CTS by relocating the Fire Brigade training program requirements of CTS 6.4.2 to the UFSAR.

The purpose of the CTS requirement is to provide organizational and training requirements for the Fire Brigade. HCGS Renewed Facility Operating License NPF-57 license condition 2.C.(7), Fire Protection, states that PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment No. 15 and as described in its submittal dated May 13, 1986, and as approved in the SER dated October 1984 (and Supplements 1 through 6) subject to a provision on making changes. In PSEG May 13, 1986 submittal (ADAMS Accession No. ML20197G501), PSEG stated that the existing HCGS Fire Protection Program is complete and has the necessary implementing features of the following controlling documents in place. Under the FSAR is listed C. - Section 13.2.2 - Fire Brigade Training Program. UFSAR Section 13.2.4, Fire Brigade Training, states that Fire Brigade training is described in Section 9.5 and the Nuclear Training Manual. UFSAR Section 9.5.1.5.2, Fire Brigade Organization, Training, and Equipment, states that Fire protection training will be conducted in accordance with the guidelines of the SRP (NUREG 0800) Section 13.2.2.II.6, 10CFR50 Appendix R and Branch Technical Position CMEB9.5.1, Section C.3.d. This change is acceptable because a license amendment pursuant to 10 CFR 50.90 is required to alter HCGS Facility Operating License, License Condition 2.(7) and the UFSAR contains the same training requirements as required by CTS 6.4.2. Any changes to the UFSAR are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details of the required Fire Brigade training are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 CTS 6.2.2.a states that each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1. CTS Table 6.2.2-1 lists

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 6 of 8 the minimum shift crew composition by position title. ITS 5.2.2 does not include a table listing minimum shift crew composition positions, thereby deleting minimum licensed and senior licensed operator crew composition requirements, which are duplicative with regulations. This changes the CTS by deleting redundant minimum licensed and senior licensed operator crew composition requirements.

The purpose of the CTS requirement is to specify the minimum required crew composition. The minimum shift crew requirements for licensed and senior licensed operators are contained in 10 CFR 50.54(k), (l), and (m) and are not required to be repeated in the ITS to provide adequate protection of the public health and safety. 10 CFR 50.54(m)(2)(i) requires each licensee meet the minimum licensed operator staffing requirements in the table. The table in 10 CFR 50.54(m)(2)(i) specifies, for one unit operating in a mode other than cold shutdown or refueling with one control room, a minimum of 2 licensed operators and 2 senior licensed operators. This is consistent with the licensed and senior operator requirements of CTS Table 6.2.2-1 when the unit is in Operational Condition 1, 2, or 3. In addition, 10 CFR 50.54(k) requires a licensed operator or senior operator be present at the controls at all times during the operation of the facility, which would include when the reactor is fueled and 10 CFR 50.54(m)(ii) requires each licensee have at its site a person holding a senior operator license at all times there is fuel in any unit. These requirements are consistent with the licensed and senior operator requirements of CTS Table 6.2.2-1 when the unit is in Operational Condition 4 or 5. This change is acceptable because the CTS requirements associated licensed operators and senior licensed operators of Table 6.2.2-1 are contained in 10 CFR 50.54, and are therefore not required to be duplicated. The minimum shift crew requirements for non-licensed plant equipment operators are retained in ITS 5.2.2.a. ITS 5.2.2.e retains shift technical advisor (i.e., individual providing advisory technical support) requirements. This change is designated as less restrictive because specifications which are included in the CTS will not be included in the ITS.

L02 CTS 6.2.2.b states that at least one licensed Reactor Operator (RO) shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Reactor Operator (SRO) shall be in the control room. ITS 5.2.2 does not include this requirement. This changes the CTS by deleting the CTS requirements for when an RO or an SRO must be in the Control Room.

The purpose of the CTS requirement is to specify when an RO or SRO must be in the control room. Title 10 of the Code of Federal Regulation (10 CFR) Section 50.54 (50.54) provides similar requirements. 10 CFR 50.54(k) requires an operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during the operation of the facility.

10 CFR 50.54(m)(2)(iii) requires 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. 10 CFR 50.54(m)(2)(iii) also requires that 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. This change is acceptable because the CTS

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 7 of 8 requirements of 6.2.2.b are contained in 10 CFR 50.54 and are therefore not required to be duplicated. This change is designated as less restrictive because specifications which are included in the CTS will not be included in the ITS.

L03 CTS 6.2.2.c states that all core alterations shall be observed and directly supervised by either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. ITS 5.2.2 does not include this requirement. This changes the CTS by deleting the CTS requirements for all core alterations to be observed and directly supervised by either a licensed SRO or licensed SRO limited to fuel handling who has no other concurrent responsibilities during this operation.

The purpose of the CTS requirement is to ensure a qualified licensed operator observes and provides direction for core alterations. 10 CFR 50.54 provides similar requirements. 10 CFR 50.54(m)(2)(iii) requires an operator or senior operator licensed pursuant to part 55 of this chapter to be present at the controls at all times during the operation of the facility. 10 CFR 50.54(m)(2)(iv) requires that each licensee to have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person. This change is acceptable because the CTS requirements of 6.2.2.c are contained in 10 CFR 50.54, and are therefore not required to be duplicated. This change is designated as less restrictive because specifications which are included in the CTS will not be included in the ITS.

L04 The CTS Table 6.2.2-1 lists the minimum shift crew composition and provides a provision stating that the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 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 of Table 6.2.2-1. The provision further states that this provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crew man being late or absent. ITS 5.2.2.b provides a similar provision but excludes the restriction that eliminates the use of this provision due to an oncoming shift crewman being late or absent. This changes the CTS by allowing the crew composition to be one less than required due to an oncoming shift crew member being late or absent.

ITS 5.2.2.b specifically addresses the unexpected absence of on-duty shift crew members. The ITS requirement does not preclude its application for the allowance for unmanned shift crew positions as a result of oncoming crew members being late or absent if absence of an on-duty shift crew member was unexpected (e.g., an emergency condition necessitated the absence of an on-duty crew member). ITS 5.2.2.b specifically states that the 2-hour allowance to accommodate unexpected absences applies to "the unexpected absence of on-duty shift crew members." Therefore, the 2-hour unexpected absence provision of ITS 5.2.2.b could apply to unmanned shift crew positions due to the absence or tardiness of an oncoming shift crew member if absence of the on-duty shift crew member was unexpected. This change is designated as less restrictive

DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Hope Creek Page 8 of 8 because less stringent requirements are being applied in the ITS than were applied in the CTS.

L05 CTS 6.3.2 provides license requirements for the Senior Nuclear Shift Supervisors, Nuclear Shift Supervisors, and Nuclear Control Operators. ITS 5.2 does not include these requirements. This changes the CTS by eliminating these license requirements from Technical Specifications for the Senior Nuclear Shift Supervisors, Nuclear Shift Supervisors, and Nuclear Control Operators.

The purpose of the CTS 6.3.2 requirement is to define specific positions required to hold a senior reactor operator license or a reactor operator license. The specific positions identified in CTS 6.3.2 include the Senior Nuclear Shift Supervisors, Nuclear Shift Supervisors, and Nuclear Control Operators. The requirements for these positions to hold an operator license to operate the facility are also contained in 10 CFR 50.54(i), (k), (l), and (m). This change is acceptable because the CTS requirements are duplicates of what is contained in 10 CFR 50.54 and essentially no change in actual practice is being made. This change is designated as less restrictive because less stringent requirements are being identified in the ITS than were identified in the CTS.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Organization 5.2 General Electric BWR/4 STS 5.2-1 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite 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 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. These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan].

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.

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

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

a.

A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor is operating in MODES 1, 2, or 3.

REVIEWER'S NOTE----------------------------------------

Two unit sites with both units shutdown or defueled require a total of three non-licensed operators for the two units.

6.2 6.0 6.2.1 6.2.1.a 6.2.1.b.

6.2.1.c 6.2.1.d Quality Assurance Topical Report Table 6.2.2-1 EO 3

5 the unit when the s

to the unit when the 4

generic 2

and S2

Organization 5.2 General Electric BWR/4 STS 5.2-2 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 5.2 Organization 5.2.2 Unit Staff (continued)

b.

Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e 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.

A radiation protection technician shall be on site when fuel is in the reactor.

The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

d.

The operations manager or assistant operations manager shall hold an SRO license.

e.

An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Table 6.2.2-1 Notation Table 6.2.2-1 RPT 6.3.2 6.2.3.1

JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
2. ISTS 5.2.1.a is modified in the ITS to require generic titles of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications to be documented in the Quality Assurance Topical Report. Hope Creek Generating Station (HCGS) does not explicitly state that titles of personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications be documented in any specific location. This deviation from the ISTS clarifies the current technical specification requirement that HCGS include generic title references of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications in the plant Quality Assurance Topical Report consistent with current practice. This ISTS deviation is consistent with a change made to Section 5.2 of the Exelon Generation Company, LLC plants technical specifications (e.g., LaSalle County Station Unit 1 Technical Specifications - ADAMS Accession No. ML052990324) and a change made to Section 5.2 of the Southern Nuclear Operating Company, Inc. plants technical specifications (e.g., Edwin I. Hatch Nuclear Plant Unit 1 Technical Specifications - ADAMS Accession No. ML052930172).
3. The ISTS contains bracketed information and/or values that are generic to General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.
4. ISTS 5.2.2.a is written to include multiple unit sites. Because HCGS is a single unit site, ITS 5.5.2.a has been modified to reflect a single unit site.
5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. Because HCGS is a single unit site, this Reviewers Note does not apply.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 1 of 10 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L01 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of no significant hazards considerations for conversion to NUREG-1433.

CTS require each on duty shift be composed of at least the minimum shift crew composition listed by position title in the minimum shift crew composition table. ITS does not include listing of the minimum required licensed and senior licensed operator crew composition.

The minimum shift crew requirements for licensed and senior licensed operators are contained in 10 CFR 50.54(k), (l), and (m) and are not required to be repeated in the ITS to provide adequate protection of the public health and safety. 10 CFR 50.54(m)(2)(i) requires each licensee meet the minimum licensed operator staffing requirements in the table. The table in 10 CFR 50.54(m)(2)(i) specifies the minimum requirements for one unit in an operating in a mode other than cold shutdown or refueling with one control room, which is consistent with CTS when the unit is in Operational Condition 1, 2, or 3.

In addition, 10 CFR 50.54(k) requires an operator or senior operator licensed be present at the controls at all times during the operation of the facility, which would include when the reactor is fueled and 10 CFR 50.54(m)(ii) requires each licensee have at its site a person holding a senior operator license at all times there is fuel in any unit. These requirements are consistent with the licensed and senior operator requirements of CTS when the unit is in Operational Condition 4 or 5. Since the licensee must comply with the requirements of 10 CFR 50.54, it is unnecessary to duplicate these requirements and, therefore, they are removed from the technical specifications.

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

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

Response: No.

The proposed change has no impact on the probability or consequences of any analyzed accident. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 2 of 10

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

Response: No.

The proposed cannot create or new or different kind of accident because the change only deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change has no impact on the margin of safety. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 3 of 10 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L02 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of significant hazards considerations for conversion to NUREG-1433.

CTS requires at least one licensed Reactor Operator (RO) to be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Reactor Operator (SRO) shall be in the control room. ITS 5.2.2 does not include this requirement.

The purpose of the CTS requirement is to specify when an RO or SRO must be in the control room. Title 10 of the Code of Federal Regulation (10 CFR) Section 50.54 (50.54) provides similar requirements. 10 CFR 50.54(k) requires an operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during the operation of the facility. 10 CFR 50.54(m)(2)(iii) requires 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. 10 CFR 50.54(m)(2)(iii) also requires that 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. Since the licensee must comply with the requirements of 10 CFR 50.54, it is unnecessary to duplicate these requirements and, therefore, they are removed from the technical specifications.

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

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

Response: No.

The proposed change has no impact on the probability or consequences of any analyzed accident. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 4 of 10

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

Response: No.

The proposed cannot create or new or different kind of accident because the change only deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change has no impact on the margin of safety. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 5 of 10 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L03 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of no significant hazards considerations for conversion to NUREG-1433.

CTS requires core alterations be observed and directly supervised by either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. ITS does not include this requirement.

The purpose of the CTS requirement is to ensure a qualified licensed operator observes and provides direction for core alterations. 10 CFR 50.54 provides similar requirements.

10 CFR 50.54(m)(2)(iii) requires an operator or senior operator licensed pursuant to part 55 of this chapter to be present at the controls at all times during the operation of the facility. 10 CFR 50.54(m)(2)(iv) requires that each licensee to have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer) a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person. Since the licensee must comply with the requirements of 10 CFR 50.54, it is unnecessary to duplicate these requirements and, therefore, they are removed from the technical specifications.

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

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

Response: No.

The proposed change has no impact on the probability or consequences of any analyzed accident. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 6 of 10 The change cannot create or new or different kind of accident because the change only deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change has no impact on the margin of safety. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 7 of 10 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L04 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of no significant hazards considerations for conversion to NUREG-1433.

CTS lists the minimum shift crew composition and provides a provision stating that it does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. ITS does not preclude the allowance for unmanned shift crew positions as a result of oncoming crewmembers being late or absent if absence of an on-duty shift crew member was unexpected (e.g., an emergency condition necessitated the absence of an on-duty crew member). ITS specifically states that the 2-hour allowance to accommodate unexpected absences applies to "the unexpected absence of on-duty shift crew members." Therefore, the 2-hour unexpected absence provision could apply to unmanned shift crew positions due to the absence or tardiness of an oncoming shift crew member if absence of the on-duty shift crew member was unexpected.

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

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

Response: No.

The proposed change provides less restrictive administrative controls for operation of the facility. The specific statement proposed for deletion regarding the application of the unexpected absence provision for the minimum shift crew composition is not an assumption in the initiation or mitigation of any analyzed event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change will not introduce new accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 8 of 10

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

Response: No.

The imposition of less restrictive administrative controls does not involve a significant reduction in the margin of safety. Minimum shift composition requirements continue to be regulated by 10 CFR 50.54(m)(2)(i) and the criteria for temporary deviations from the numbers required by the regulation continue to be provided in the ITS.

Deletion of the specific statement that the 2-hour unexpected absence allowance regarding minimum shift crew composition is not applicable to the absence or tardiness of the oncoming shift crew members neither affects the design of any plant structure, system, or component (SSC), nor the manner in which the SSCs are operated and controlled. No changes are proposed to the facility or to any accident analysis assumptions, inputs or expected outcomes and thus, the proposed change cannot adversely affect the likelihood or outcome of any design basis accident.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 9 of 10 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L05 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of no significant hazards considerations for conversion to NUREG-1433.

CTS provides license requirements for the Senior Nuclear Shift Supervisors, Nuclear Shift Supervisors, and Nuclear Control Operators. ITS does not include these requirements. The purpose of the CTS requirement is to define specific positions required to hold a senior reactor operator license or a reactor operator license. The requirements for these positions to hold an operator license to operate the facility are also contained in 10 CFR 50.54(i), (k), (l), and (m). Since the licensee must comply with the requirements of 10 CFR 50.54, it is unnecessary to duplicate these requirements and, therefore, they are removed from the technical specifications.

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

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

Response: No.

The proposed change has no impact on the probability or consequences of any analyzed accident. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The change cannot create or new or different kind of accident because the change only deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION Hope Creek Page 10 of 10

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

Response: No.

The proposed change has no impact on the margin of safety. The change deletes administrative control requirements that are duplicative of requirements specified in the regulations. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

ATTACHMENT 3 ITS 5.3, Unit Staff Qualifications

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 The Shift Technical Advisor shall provide advisory technical support to the Senior Nuclear Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the PSEG Nuclear Quality Assurance Topical Report.

6.3.2 The Operations Manager or Assistant Operations Manager shall hold a senior reactor operator license. The Senior Nuclear Shift Supervisors, and Nuclear Shift Supervisors, shall hold a senior reactor operator license. The Nuclear Control Operators shall hold a reactor operator license.

6.3.3 The Operations Manager shall meet one of the following:

(1)

Hold a senior reactor operator license, or (2)

Have held a senior reactor operator license for this or a similar unit (BWR), or (3)

Have been certified at an appropriate simulator for equivalent senior operator knowledge.

6.4 TRAINING 6.4.1 DELETED 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Fire Protection Manager and shall meet or exceed the requirements of the SRP (NUREG-0800)

Section 13.2.2.II.6, 10 CFR 50 Appendix R and Branch Technical Position CMEB 9.5.1, Section C.3.d.

6.5 REVIEW AND AUDIT (THIS SECTION DELETED)

HOPE CREEK 6-6 Amendment No. 233 See ITS 5.2 See ITS 5.2 See ITS 5.2 ITS A01 ITS 5.3 5.3.2 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 5.3.1, perform the functions described in 10 CFR 50.54(m).

5.3.1 5.3 5

A02 5.3.2 A03 L01

ADMINISTRATIVE CONTROLS (PAGES 6-7 THROUGH 6-12 ARE DELETED)

HOPE CREEK 6-7 through 6-12 Amendment No. 97 ITS A01 ITS 5.3

ADMINISTRATIVE CONTROLS (THIS PAGE INTENTIONALLY LEFT BLANK)

HOPE CREEK 6-13 Amendment No. 97 ITS A01 ITS 5.3

DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS Hope Creek Page 1 of 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.3.1 states that each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the PSEG Nuclear Quality Assurance Topical Report. ITS 5.3.1 states that each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the Quality Assurance Topical Report. This changes the CTS by removing the PSEG Nuclear designation for the Quality Assurance Topical Report.

The purpose of CTS 6.3.1 is to provide the qualification requirements of the unit staff. PSEG Nuclear has a Quality Assurance Topical Report that is applicable to Salem and Hope Creek generating stations. The title of this document is Quality Assurance Topical Report (QATR). This change is acceptable because the title of the document is all that is needed to ensure the user is directed to the proper location containing the required qualifications for the unit staff. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS does not include ISTS Specification 5.3.2, 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 5.3.1, perform the functions described in 10 CFR 50.54(m). This changes the CTS by clarifying unit staff qualification requirements associated with licensed operators.

The purpose of ISTS Specification 5.3.2 is to ensure that there is no misunderstanding when complying with 10 CFR 55.4 requirements. Adding this paragraph is consistent with the recommendations in the April 9, 1997 letter from C. Grimes to J. Davis and was added to the ISTS by Technical Specification Task Force (TSTF) traveler TSTF-258-A, Changes to Section 5.0, Administrative Controls. This change is acceptable because no change is made to the application of HCGS CTS, it only clarifies existing 10 CFR 50.54(m) requirements. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None

DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS Hope Creek Page 2 of 2 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 CTS 6.3.3 requires the Operations Manager to hold or have held a senior reactor operator (SRO) license at HCGS or have held an SRO license for this or a similar unit (BWR), or have been certified at an appropriate simulator for equivalent senior operator knowledge. CTS 6.3.2 also requires either the Assistant Operations Manager or the Operations Manager to hold an SRO license. ITS 5.2.2.d retains the requirement for either the operations manager or the assistant operations manager to hold an SRO license but does not require specific SRO license/certification qualification requirements for the operations manager. This changes the CTS by deleting the specific SRO license/certification qualification requirements for the operations manager.

The purpose of the CTS requirement is to ensure that operations management, specifically the operations manager, has SRO level knowledge of the plant.

However, as long as either the operations manager or assistant operations manager holds an SRO, there is an individual within operations management that is qualified and trained to an SRO level of knowledge at HCGS. This change is consistent with the ISTS and acceptable because eliminating specific SRO license/certification requirements for the operations manager does not adversely affect the operations of the plant. At least one member of the operations management team will continue to be required to hold an SRO license and therefore, SRO level knowledge of the plant. This change is designated as less restrictive because less stringent requirements are being applied in the ITS than were applied in the CTS.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Unit Staff Qualifications 5.3 General Electric BWR/4 STS 5.3-1 Rev. 5.0 Hope Creek Amendment XXX 3

5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications

REVIEWER'S NOTE-------------------------------------------------

Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of

[Regulatory Guide 1.8, Revision 2, 1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. [The staff not covered by Regulatory Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff].

5.3.2 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 5.3.1, perform the functions described in 10 CFR 50.54(m).

referenced for comparable positions as specified in the Quality Assurance Topical Report 1

2 CTS 6.3 6.3.1 DOC A03

JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS Hope Creek Page 1 of 1

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.
2. The ISTS contains bracketed information and/or values that are generic. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis as approved in License Amendment 233, dated March 9, 2023 (ADAMS Accession No. ML23037A971). As stated in the NRC safety evaluation associated with License Amendment 233, the NRC staff concluded that the elimination of specific unit staff qualification requirements in the technical specifications and referencing the Quality Assurance Topical Report for these requirements was acceptable because the licensees overall unit staff qualification requirements were not reduced. The NRC staff further stated that the technical specifications continue to assure operation of the facility in a safe manner in accordance with 10 CFR 50.36(c)(5).
3. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

R1

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS Hope Creek Page 1 of 2 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L01 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of the change and the determination of No Significant Hazards Considerations for conversion to NUREG-1433.

CTS requires the Operations Manager to hold or have held a senior reactor operator (SRO) license at HCGS or have held an SRO license for this or a similar unit (BWR), or have been certified at an appropriate simulator for equivalent senior operator knowledge.

CTS also requires either the Assistant Operations Manager or the Operations Manager to hold an SRO license. ITS retains the requirement for either the operations manager or the assistant operations manager to hold an SRO license but does not require specific SRO license/certification qualification requirements for the operations manager. This changes the CTS by deleting the specific SRO license/certification qualification requirements for the operations manager.

Eliminating specific SRO license/certification requirements for the operations manager does not adversely affect the operations of the plant. At least one member of the operations management team will still continue to be required to hold an SRO license and therefore, SRO level knowledge of the plant.

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

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

Response: No.

The proposed change provides less restrictive administrative controls requirements for operation of the facility. A requirement for a specific individual within the operations management staff to hold a senior reactor operator license is not an assumption in the initiation or mitigation of any analyzed event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The change will not introduce new accident initiators.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS Hope Creek Page 2 of 2 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The imposition of less restrictive administrative controls does not involve a significant reduction in the margin of safety. Eliminating specific SRO license/certification requirements for the operations manager neither affects the design of any plant structure, system, or component (SSC), nor the manner in which the SSCs are operated and controlled. At least one member of the operations management team will continue to be required to hold an SRO license and therefore, SRO level knowledge of the plant. No changes are proposed to the facility or to any accident analysis assumptions, inputs or expected outcomes and thus, the proposed change cannot adversely affect the likelihood or outcome of any design basis accident.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

ATTACHMENT 4 ITS 5.4, Procedures

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a.

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

b.

The applicable procedures required to implement the requirements of NUREG-0737 and supplements thereto.

c.

Refueling operations.

d.

Surveillance and test activities of safety-related equipment.

e.

Security Plan implementation.

f.

Emergency Plan implementation.

g.

Fire Protection Program implementation.

h.

PROCESS CONTROL PROGRAM implementation.

i.

OFFSITE DOSE CALCULATION MANUAL implementation.

j.

Quality Assurance Program for effluent and environment monitoring.

6.8.2 Each procedure and administrative policy of 6.8.1 above, except 6.8.1.e and 6.8.1.f, and changes thereto, shall be reviewed and approved in accordance with requirements in Updated Final Safety Analysis Report (UFSAR) section 17.2 for SORC or for Technical Review and Control, as appropriate, prior to implementation and reviewed periodically as set forth in administrative procedures. Procedures of 6.8.1.e and 6.8.1.f shall be reviewed and approved in accordance with the Facility's Security and Emergency Plans or requirements in Updated Final Safety Analysis Report (UFSAR) section 17.2 for Technical Review and Control, as appropriate, prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 On-the-Spot changes to procedures of Specification 6.8.1 may be made provided:

a.

The intent of the original procedure is not altered;

b.

The change is approved by two members of the unit management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and

c.

The change is documented and receives the same level of review and approval as the original procedure within 14 days of implementation.

HOPE CREEK 6-15 Amendment No. 97 ITS A01 ITS 5.4 5.4 following

e. All programs specified in Specification 5.5.
d.
c.

A05 5.4.1.c 5.4.1.d 5.4.1.a 5.4.1.b 5.4.1.e 5.4.1.e 5.4 5.4.1 A01 A01 A01 emergency operating NUREG-0737, Supplement 1, as stated in Generic Letter 82-33 LA02 A02 A03 LA01 A04 M01

ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

HOPE CREEK 6-23 Amendment No. 159 ITS A01 ITS 5.4 LA03

DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Hope Creek Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.8.1.b requires that written procedures be established, implemented, and maintained covering the applicable procedures required to implement the requirements of NUREG-0737 and supplements thereto. ITS 5.4.1.b requires that written procedures be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. This changes the CTS by clarifying the applicable procedures as emergency operating procedures and supplements thereto as NUREG 0737, Supplement 1, as stated in Generic Letter 82-33.

The purpose of CTS 6.8.1.b is to ensure that written procedures are established, implemented, and maintained covering the applicable procedures required to implement the requirements of NUREG-0737 and supplements thereto.

Supplement 1 to NUREG-0737 enclosed with Generic Letter 82-33 is part of the implementation of the Three Mile Island (TMI) action items of NUREG-0737. The purpose of the generic letter was to provide additional clarification regarding, in part, upgrade of emergency operating procedures associated with TMI Action Item I.C.1. As a result, the change clarifies the existing requirement to comply with NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, for emergency operating procedures, which represents the applicable procedures required to implement the requirements of NUREG-0373 and associated supplement thereto. The change is designated as administrative and is acceptable because it does not result in technical changes to the CTS.

A03 CTS 6.8.1.c and d require written procedures be established, implemented, and maintained covering refueling operations and surveillance and test activities of safety-related equipment, respectively. ITS 5.4.1 does not explicitly require these activities but rather requires written procedures be established, implemented, and maintained to the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 consistent with CTS 6.8.1.a. This changes the CTS by removing the specific wording of CTS 6.8.1.c and CTS 6.8.1.d.

This change is considered administrative because the recommendations of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 require procedures for refueling operations and surveillance tests for safety related activities. Therefore, specifically citing these requirements individually is unnecessary. This change is designated as administrative because it does not result in a technical change to the CTS.

DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Hope Creek Page 2 of 4 A04 CTS 6.8.1.e requires procedures for implementation of the Security Plan and CTS 6.8.1.f requires procedures for implementation of the Emergency Plan. ITS 5.4 does not include these requirements. This changes the CTS by removing the procedure requirements for the Security Plan and the Emergency Plan implementation.

Procedures to implement the Emergency Plan and the Security Plan are required by 10 CFR 50, Appendix E and 10 CFR 50.54(p). Because conformance with 10 CFR Chapter I is a license condition and the Emergency Plan and Security Plan are required to be implemented by 10 CFR Chapter I, specific identification of these plans is unnecessary duplication. This is a change in the presentation of the requirements only and, therefore, is considered an administrative change.

A05 CTS 6.8.1.i requires written procedures for Offsite Dose Calculation Manual (ODCM) implementation. ITS 5.4.1.e requires written procedures for all Programs and Manuals specified in Specification 5.5. This changes the CTS by replacing the specific procedure requirement for ODCM implementation with a generic requirement to have procedures for all programs and manuals specified in Specification 5.5, which includes the ODCM.

The purpose of CTS 6.8.1.i is to ensure written procedures are established, implemented, and maintained for ODCM implementation. ITS 5.4.1.e similarly ensures written procedures are established, implemented, and maintained for ODCM implementation by stating that written procedures are established, implemented, and maintained for all Programs and Manuals specified in Specification 5.5 which includes the ODCM. Because the ODCM procedure requirement remains, this is a change in the presentation of the requirements only and, therefore, is considered an administrative change.

MORE RESTRICTIVE CHANGES M01 CTS 6.8.1 does not include a requirement that written procedures be established, implemented, and maintained covering all programs and manuals specified in Specification 6.0. ITS 5.4.1.e requires written procedures be established, implemented, and maintained covering all programs and manuals specified in ITS 5.5. This changes the CTS by adopting a new requirement for procedures to address all programs and manuals described in ITS 5.5.

This change is necessary to ensure written procedures, including proper procedure control, are established and maintained to address programs required by ITS 5.5. This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications.

RELOCATED SPECIFICATIONS None

DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Hope Creek Page 3 of 4 REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.8.1.g requires that written procedures for the Process Control Program (PCP) be established, implemented, and maintained. The ITS does not include these requirements. This changes the CTS by moving the requirements to the Updated Safety Analysis Report (UFSAR).

The removal of these details, which are related to meeting Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71 and written procedures are necessary to ensure compliance with regulations. Regulations provide an adequate level of control for the affected requirements, and thus, inclusion of this requirement in the Technical Specifications is not necessary.

Also, this change is acceptable because these details will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications.

LA02 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.8.2 requires that each procedure and administrative policy of 6.8.1, except 6.8.1.e and 6.8.1.f, and changes thereto, be reviewed and approved in accordance with requirements in Updated Final Safety Analysis Report (UFSAR) while 6.8.1.e and 6.8.1.f are reviewed and approved in accordance with the Facility's Security and Emergency Plans or requirements in Updated Final Safety Analysis Report (UFSAR). CTS 6.8.3 provides requirements for on-the-spot changes to procedures of Specification 6.8.1. ITS 5.4 does not include these requirements. This changes the CTS by moving the requirements to the UFSAR.

The removal of these procedure approval and change process details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications in order to provide adequate protection of public health and safety. The ITS retains the requirement for written procedures to be established, implemented, and maintained covering the listed activities. In addition the requirements for the establishment, maintenance, and implementation of procedures related to activities affecting quality are contained in 10 CFR 50, Appendix B, Criterion V and Criterion VI.

Also, this change is acceptable because these types of procedural details will be adequately controlled in the UFSAR. The UFSAR is controlled under 10 CFR 50.59 which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting regulatory requirements are being removed from the Technical Specifications.

DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Hope Creek Page 4 of 4 LA03 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.11.1 states that procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. ITS 5.4 does not include this requirement for radiation protection procedures. This changes the CTS by moving the requirements to the UFSAR.

The details contained in CTS 6.11, "Radiation Protection Program," are to be relocated to the UFSAR. This relocated program requires procedures to be prepared for personnel radiation protection consistent with 10 CFR 20. These procedures are for nuclear plant personnel and have no impact on nuclear safety or the health and safety of the public. Requirements to develop, implement and use procedures are contained in 10 CFR 20.1101(b). Periodic review of these procedures is addressed in 10 CFR 20.1101(c). Because the CTS requirements are contained in the regulations, there is no need to repeat them in the ITS to provide adequate protection of the public health and safety. Also, this change is acceptable because these types of procedural details will be adequately controlled in the UFSAR. The UFSAR is controlled under 10 CFR 50.59 which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting regulatory requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Procedures 5.4 General Electric BWR/4 STS 5.4-1 Rev. 5.0 Hope Creek Amendment XXX 2

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

a.

The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978,

b.

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

c.

Quality assurance for effluent and environmental monitoring,

d.

Fire Protection Program implementation, and

e.

All programs specified in Specification 5.5.

CTS 6.8 6.0 6.8.1 6.8.1.a 6.8.1.b 6.8.1.j 6.8.1.g DOC M01 1

3

JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES Hope Creek Page 1 of 1

1. The ISTS contains bracketed information and/or values that are generic to General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.
2. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. Text revised, inserted, or deleted in ITS to correct a typographical or grammatical error.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 5 ITS 5.5, Programs and Manuals

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4 The following programs shall be established, implemented, and maintained:

a.

Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the HPCI, CS, RHR, RCIC, Containment Hydrogen Recombiner, H2/02 analyzer, Post-Accident Sampling, Control Rod Drive Hydraulic (Scram Discharge portion) systems. The program shall include the following:

1.

Preventive maintenance and periodic visual inspection requirements, and

2.

A service pressure leak test for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel,

2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

c.

Deleted HOPE CREEK 6-16 Amendment No.149 ITS A01 ITS 5.5 and Manuals 5.5 5.5.2 5.5.2 This program provides controls to minimize practicable at a frequency in accordance with the Surveillance Frequency Control Program Integrated leak requirements a

b The provisions of SR 3.0.2 are applicable.

A02 LA01 5.5.2.a 5.5.2.b LA02 A01 and S2 S2

ADMINISTRATIVE CONTROLS

d.

Explosive Gas Monitoring This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System. The program shall include the limit for hydrogen concentration in the Main Condenser Offgas Treatment System and a surveillance program to ensure the limit is maintained. This limit shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion).

The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies.

e.

Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

an API gravity or absolute specific gravity within limits for ASTM 2D fuel oil,

2.

a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3.

bulk water and sediment within limits for ASTM 2D fuel oil;

b.

Other properties for new ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and c.

7RWDOSDUWLFXODWHFRQFHQWUDWLRQRIWKHVWRUHGIXHORLOLVPJOZKHQ

tested every 92 days in accordance with ASTM D-2276, modified as follows: The 0.8 micron membrane filters specified in ASTM D-2276 may be replace with membrane filters up to 3.0 microns.

HOPE CREEK 6-16a Amendment No. 100 at a frequency in accordance with the Surveillance Frequency Control Program LA01 5.5.6 Program 5.5.6 5.5.7 5.5.7 oil The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

A clear and bright appearance with proper color or a of the new fuel verify that the content LA03 LA03 L01 SR 3 A03 5.5.7.a 5.5.7.a.1 5.5.7.a.2 5.5.7.a.3 5.5.7.b 5.5.7.c LA04 and Storage Tank Radioactivity 5.5.6.a A01 See CTS 3.11.1.4 for storage tank markups.

ITS A01 ITS 5.5

, other than those addressed in a., above, S2

ADMINISTRATIVE CONTROLS 6.8.4.f Primary Containment Leakage Rate Testing Program A program shall be established, implemented, and maintained to comply with the leakage rate testing of the containment as required by 10CFR50.54(o) and 10CFR50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 50.6 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.5% of primary containment air weight per day.

Leakage Rate Acceptance Criteria are:

a.

Primary containment leakage rate acceptance criterion is less than or equal to 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa,

2)

Door seal leakage rate less than or equal to 5 scf per hour when the gap between the door seals is pressurized to greater than or equal to 10.0 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

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

HOPE CREEK 6-16b Amendment No. 230 5.5.10 5.5.10 5.5.10.a SR 3.0.3

a.
b.
c.
d.
1.
2.

a b

e.
f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.10.b 5.5.10.c 5.5.10.d 5.5.10.d.1 5.5.10.d.2 5.5.10.d.2.a) 5.5.10.d.2.b) 5.5.10.e A04 A04 5.5.3 5.5.3 This program conforms to scfh For each door, A01 5.5.10.f ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

1)

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

2)

Limitations on the concentration of radioactive material released in liquid effluents to unrestricted areas conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,

3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4)

Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50,

5)

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,

6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, HOPE CREEK 6-16c Amendment No. 230 functional capability functional capability from the site at or 5.5.3.g 5.5.3.f 5.5.3.e 5.5.3.d 5.5.3.c 5.5.3.b 5.5.3.a ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4.g Radioactive Effluent Controls Program

8)

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

9)

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

10)

Limitations on venting and purging of the containment through the Reactor Building Ventilation System, Hardened Torus Vent, or the FRVS to maintain releases as low as reasonably achievable, and

11)

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

h.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluents monitoring program and modeling of the environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2)

A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census, and

3)

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

HOPE CREEK 6-16d Amendment No. 230 5.5.3.j

,beyond the site boundary, The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.3k 5.5.3.i 5.5.3.h A05 LA05 Filtration Recirculation and Ventilation System (FRVS)

A01 ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4.i Deleted 6.8.4.j Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

HOPE CREEK 6-16e Amendment No. 205 3

5.5.13 ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3.n. This documentation shall contain:

(1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes(s) and (2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.

Shall become effective after review and acceptance by the SORC and the approval of the Plant General Manager.

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3.n. This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes(s) and
2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
b.

Shall become effective after review and acceptance by the SORC and the approval of the Plant General Manager.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

HOPE CREEK 6-25 Amendment No. 145 5.5.1 5.5.1 5.5.1.a (s) 5.5.1.a.1 5.5.1.a.2 (s)

(s) 5.5.1.b 5.5.1.c NRC i.e., month and year Licensee initiated LA06 LA07 ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS 6.15 TECHNICAL SPECIFICATION (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.

PSEG may make changes to the Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the License, or

2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

Proposed changes to the Bases that require either condition of Specification 6.15.b above shall be reviewed and approved by the NRC prior to implementation.

d.

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).

e.

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

6.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

HOPE CREEK 6-26 Amendment No.173 5.5.8 5.5.8 Licensees d

c 5.5.8 5.5.8.a 5.5.8.b 5.5.8.b.1 5.5.8.b.2 5.5.8.d 5.5.8.d 5.5.8.c 5.5.12 (CRE) 5.5.12.a 5.5.12.b 5.5.12.c U

Add proposed TS 5.5.9, Safety Function Determination Program (SFDP)

M01 5.5.12 ITS A01 ITS 5.5

ADMINISTRATIVE CONTROLS 6.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (Continued)

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the Control Room Emergency Filtration System, operating at the flow rate required by Surveillance Requirement 4.7.2.1.c.1, at a Frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 36 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of Specification 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

HOPE CREEK 6-27 Amendment No.173 LA01 in accordance with the Surveillance Frequency Control Program the VFTP SR 3 5.5.12.d 5.5.12.e 5.5.12.f LA01 ITS A01 ITS 5.5 subsystem S2

A01

DEFINITIONS LIMITING CONTROL ROD PATTERN 1.20 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

1.23 DELETED 1.24 Not Used MINIMUM CRITICAL POWER RATIO 1.25 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFF-GAS RADWASTE TREATMENT SYSTEM 1.26 An OFF-GAS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting reactor coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATIONAL MANUAL (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/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report required by Specifications 6.9.1.6 and 6.9.1.7.

HOPE CREEK 1-4 Amendment No. 230 See ITS 1.0 (ODCM) 5.5.1 5.5.1 5.5.1.a 5.5.1.b activities 5.6.1 and Specification 5.6.2 s

ITS A01 ITS 5.5

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.

ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a.

After each closing of each penetration subject to Type B testing, except the primary containment air locks, if opened following Type A or B test, by leak rate testing in accordance with the Primary Containment Leakage Rate Testing Program.

b.

In accordance with the Surveillance Frequency Control Program by verifying that all primary containment penetrations** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

c.

By verifying each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d.

By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

See Special Test Exception 3.10.1 Except valves, blind flanges, and deactivated automatic valves which are located inside the primary containment, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been de-inerted since the last verification or more often than once per 92 days.

HOPE CREEK 3/4 6-1 Amendment No. 187 See ITS 3.6.1.1 See ITS 3.6.1.3 See ITS 3.6.1.3 See ITS 3.6.1.1 LA08 ITS A01 ITS 5.5

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:

a.

An overall integrated leakage rate (Type A test) in accordance with the Primary Containment Leakage Rate Testing Program.

b.

A combined leakage rate in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which form the boundary for the long-term seal of the feedwater lines, other valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted.

c.
  • Less than or equal to 150 scfh per main steam line and less than or equal to 250 scfh combined through all four main steam lines when tested at 5 psig (leakage rate corrected to 1 Pa, 50.6 psig).
d.

A combined leakage rate of less than or equal to 10 gpm for all containment isolation valves which form the boundary for the long-term seal of the feedwater lines, when tested at 1.10 Pa, 55.7 psig.

e.

A combined leakage rate of less than or equal to 10 gpm for all other penetrations and containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at 1.10 Pa, 55.7 SVLJ¨S

APPLICABILITY:

When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.

ACTION:

With:

a.

The measured overall integrated primary containment leakage rate (Type A test) not in accordance with the Primary Containment Leakage Rate Testing Program, or

b.

The measured combined leakage rate exceeding the leakage rate specified in the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which form the boundary for the long-term seal of the feedwater lines, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, or

c.

The measured leakage rate exceeding 150 scfh per main steam line or exceeding 250 scfh combined through all four main steam lines, or

HOPE CREEK 3/4 6-2 Amendment No. 174 See ITS 3.6.1.3 See ITS 3.6.1.1 See ITS 3.6.1.3 LA08 5.5.10.b 5.5.10.c 5.5.10.d.1 5.5.10.d See ITS 3.6.1.1 See ITS 3.6.1.3 LA08 ITS A01 ITS 5.5 LA08

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

d.

The measured combined leakage rate for all containment isolation valves which form the boundary for the long-term seal of the feedwater lines exceeding 10 gpm, or

e.

The measured combined leakage rate for all other penetrations and containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 10 gpm, restore:

a.

The overall integrated leakage rate(s) (Type A test) to be in accordance with the Primary Containment Leakage Rate Testing Program, and

b.

The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which form the boundary for the long-term seal of the feedwater lines, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and

c.

The leakage rate to less than or equal to 150 scfh per main steam line and less than or equal to 250 scfh combined through all four main steam lines, and

d.

The combined leakage rate for all containment isolation valves which form the boundary for the long-term seal of the feedwater lines to less than or equal to 10 gpm, and

e.

The combined leakage rate for all other penetrations and containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 10 gpm, prior to increasing reactor coolant system temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.1.2.a The primary containment leakage rates shall be demonstrated in accordance with the Primary Containment Leakage Rate Testing Program for the following:

1.

Type A test.

2.

Type B and C tests (including air locks).

b.

DELETED.

c.

DELETED.

Exemption to Appendix "J" of 10 CFR 50.

HOPE CREEK 3/4 6-3 Amendment No. 171 See ITS 3.6.1.3 See ITS 3.6.1.3 See ITS 3.6.1.1 5.5.10.d 5.5.10.b 5.5.10.c 5.5.10.d.1 See ITS 3.6.1.1 See ITS 3.6.1.1 LA08 ITS A01 ITS 5.5 LA08

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each primary containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b.

An overall air lock leakage rate in accordance with the Primary Containment Leakage Rate Testing Program.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2* and 3.

ACTION:

a.

With one primary containment air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed.

2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With the primary containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

See Special Test Exception 3.10.1.

HOPE CREEK 3/4 6-5 Amendment No. 180 See ITS 3.6.1.2 See ITS 3.6.1.2 5.5.10.

5.5.10.d.2 ITS A01 ITS 5.5

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each primary containment power operated or automatic valve shall be determined to be within its limit when tested pursuant to the INSERVICE TESTING PROGRAM.

4.6.3.4 In accordance with the Surveillance Frequency Control Program, verify that a representative sample of reactor instrumentation line excess flow check valves# actuates to the isolation position on a simulated instrument line break signal.

4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE*:

a.

In accordance with the Surveillance Frequency Control Program by verifying the continuity of the explosive charge.

b.

In accordance with the Surveillance Frequency Control Program by removing the explosive squib from at least one explosive valve, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.

Exemption to Appendix J of 10 CFR Part 50.

The reactor vessel head seal leak detection line (penetration J5C) is not required to be tested pursuant to this requirement.

HOPE CREEK 3/4 6-18 Amendment No. 205 See ITS 3.6.1.3 See ITS 3.6.1.3 ITS A01 ITS 5.5 LA08

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or upon determination** that the HEPA filters or charcoal adsorbent could have been damaged by structural maintenance or adversely affected by any chemicals, fumes or foreign materials (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rates are 9,000 cfm +/- 10% for each FRVS ventilation unit.

2.

Verifying within 31 days after removal from the FRVS ventilation units, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity 95%.

3.

Verifying a subsystem flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit during system operation when tested in accordance with ANSI N510-1980.

d.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal from the FRVS ventilation units, that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 95%.

This determination shall consider the maintenance performed and/or the type, quantity, length of contact time, known effects and previous accumulation history for all contaminants which could reduce the system performance to less than that verified by the acceptance criteria in items c.1 through c.3 below.

HOPE CREEK 3/4 6-51a Amendment No. 187 5.5.5.a A program shall be established to implement the following required testing of filter ventilation systems at the frequencies specified in the Surveillance Frequency Control Program and in Regulatory Guide 1.52, Revision 2, and in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1980.

A06 5.5.5.c As specified in Regulatory Guide 1.52, Revision 2 5.5.5.c L02 A06 L02 A07 A08 A07 A08 A07 A08 5.5.5.

A06 A09 5.5 Programs and Manuals 5.5.5 Ventilation Filter Testing Program (VFTP)

ITS A01 ITS 5.5

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge in the ventilation unit while operating the filter train at a flow rate of 9,000 cfm +/-

10% for each FRVS ventilation unit.

2.

Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a.

Manual initiation from the control room, and

b.

Simulated automatic initiation signal.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2 March 1978, while operating the system at a flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit.

g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 9,000 cfm +/- 10% for each FRVS ventilation unit.

HOPE CREEK 3/4 6-52 Amendment No. 187 5.5.5.d As specified in Regulatory Guide 1.52, Revision 2 As specified in Regulatory Guide 1.52, Revision 2 A06 A06 A07 A07 See ITS 3.6.4.3 A08 A08 5.5.5.a 5.5.5.b s

5.5 Programs and Manuals 5.5.5 Ventilation Filter Testing Program (VFTP)

A10 when tested in accordance with ASME N510-1980 s

s The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

A05 ITS A01 ITS 5.5 5

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or upon determination** that the HEPA filters could have been damaged by structural maintenance or adversely affected by any foreign materials (1) after any structural maintenance on the HEPA filters or housings by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rates are 30,000 cfm +/-

10% for each FRVS recirculation unit.

2.

Verifying a subsystem flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit during system operation when tested in accordance with ANSI N510-1980.

d.

not used

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the exhaust duct is less than 8 inches Water Gauge in the recirculation filter train while operating the filter train at a flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit.

2.

Verifying that the filter train starts and isolation dampers open on each of the following test signals:

a.

Manual initiation from the control room, and

b.

Simulated automatic initiation signal.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Position C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2 March 1978, while operating the system at a flow rate of 30,000 cfm +/- 10% for each FRVS recirculation unit.

This determination shall consider the maintenance performed and/or the type, quantity, length of contact time, known effects and previous accumulation history for all contaminants which could reduce the system performance to less than that verified by the acceptance criteria in items c.1 and c.2 below.

HOPE CREEK 3/4 6-53 Amendment No. 187 5.5.5.a A07 5.5.5.d As specified in Regulatory Guide 1.52, Revision 2 A06 See ITS 3.6.4.3 A07 A08 A08 5.5.5.a A06 A06 5.5 Programs and Manuals 5.5.5 Ventilation Filter Testing Program (VFTP) s s

s A09 A10 when tested in accordance with ASME N510-1980 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

A05 ITS A01 ITS 5.5

PLANT SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM SURVEILLANCE REQUIREMENTS (continued)

c.

In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem filter train by:

1.

Verifying that the subsystem satisfies the in-place penetration testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system filter train flow rate is 4000 cfm + 10%.

2.

Verifying within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity 70%.

3.

Verifying a subsystem filter train flow rate of 4000 cfm +10% during subsystem operation when tested in accordance with ANSI N510-1980.

d.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal from the Control Room Emergency Filtration units that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 0.5% when tested in accordance with ATSM D3803 -1989 at a temperature of 30°C and a relative humidity of 70%.

e.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.5 inches Water Gauge while operating the filter train subsystem at a flow rate of 4000 cfm + 10%.

2.

Verifying with the control room hand switch in the recirculation mode that on each of the below recirculation mode actuation test signals, the subsystem automatically switches to the isolation mode of operation and the isolation dampers close within 5 seconds:

a)

High Drywell Pressure b)

Reactor Vessel Water Level Low Low Low, Level 1 c)

Control room ventilation radiation monitors high.

HOPE CREEK 3/4 7-7 Amendment No. 191 See ITS 3.7.3 5.5.5.a L02 A07 A07 A07 A08 A08 A08 5.5.5.c 5.5.5.c 5.5.5.d A06 A06 L02 5.5 Programs and Manuals 5.5.5 Ventilation Filter Testing Program (VFTP)

A09 A10 when tested in accordance with ASME N510-1980 ITS A01 ITS 5.5

PLANT SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM SURVEILLANCE REQUIREMENTS (continued)

3. Verifying with the control room hand switch in the outside air mode that on each of the below pressurization mode actuation test signals, the subsystem automatically switches to the pressurization mode of operation:

a)

High Drywell Pressure b)

Reactor Vessel Water Level Low Low Low, Level 1 c)

Control room ventilation radiation monitors high.

4.

Verifying that the heaters dissipate 13 + 1.3 Kw when tested in accordance with ANSI N510-1980 and verifying humidity is maintained less than or equal to 70% humidity through the carbon adsorbers by performance of a channel calibration of the humidity control instrumentation.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 4000 cfm + 10%.

g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration testing acceptance criteria of less than 0.05% in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 4000 cfm + 10%.

4.7.2.1.2 The control room envelope boundary shall be demonstrated OPERABLE:

a.

At a frequency in accordance with the Control Room Envelope Habitability Program by performance of control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

HOPE CREEK 3/4 7-8 Amendment No. 191 5.5.5.e See ITS 3.7.3 As specified in Regulatory Guide 1.52, Revision 2 A06 As specified in Regulatory Guide 1.52, Revision 2 A06 See ITS 3.7.3 A07 A07 A08 A08 5.5.5.a 5.5.5.b L03 5.5 Programs and Manuals 5.5.5 Ventilation Filter Testing Program (VFTP) 5.5.5.c The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

A05 ITS A01 ITS 5.5

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.1 Each of the above required batteries and chargers shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that:

1.

The parameters in Table 4.8.2.1-1 meet the Category A limits, and

2.

Total battery terminal voltage for each 125-volt battery is greater than or equal to 129 volts on float charge and for each 250-volt battery the terminal voltage is greater than or equal to 258 volts on float charge.

b.

In accordance with the Surveillance Frequency Control Program and within 7 days after a battery discharge with battery terminal voltage below 108 volts for a 125-volt battery or 210 volts for a 250-volt battery, or battery overcharge with battery terminal voltage above 140 volts for a 125-volt /battery or 280 volts for a 250-volt battery, by verifying that:

1.

The parameters in Table 4.8.2.1-1 meet the Category B limits,

2.

There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohms, excluding cable intercell connections, and

3.

The average electrolyte temperature of each sixth cell of connected cells is above 72°F.

c.

In accordance with the Surveillance Frequency Control Program by verifying that:

1.

The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration,

2.

The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material,

3.

The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohms, excluding cable intercell connections, and

4.

The battery charger will supply the current listed below at the voltage listed below for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

CHARGER Minimum Voltage CURRENT (AMPERES) 1AD413, 1AD414 129 200 1BD413, 1BD414 1CD413, 1CD414 1CD444, 1DD414 1DD444, 1DD413 10D423, 10D433 258 50 HOPE CREEK 3/4 8-13 Amendment No. 187 5.5.11.d 5.5.11.d LA09 See ITS 3.8.6 See ITS 3.8.6 See ITS 3.8.4 See ITS 3.8.4 See ITS 3.8.6 LA09 Limits on battery Limits on battery ITS A01 ITS 5.5

TABLE 4.8.2.1-1 BATTERY SURVEILLANCE REQUIREMENTS PARAMETER CATEGORY A: (*)

LIMITS FOR EACH DESIGNATED PILOT CELL CATEGORY B: (*)

LIMITS FOR EACH CONNECTED CELL CATEORY C: (#)

ALLOWABLE VALUE FOR EACH CONNECTED CELL Electrolyte Level

0LQLPXPOHYHO

indication mark and 1/4" above maximum level indication mark(d)

0LQLPXPOHYHO

indication mark and 1/4" above maximum level indication mark(d)

Above top of plates and not overflowing Float Voltage

YROWV

YROWV(c)

> 2.07 volts Specific Gravity(a)

(b)



AND average of all connected cells

> 1.205(b)

Not more than.020 below the average of all connected cells AND Average of all connected cells 1.195(b)

(*)

With parameters of one or more cells in one or more batteries not within limits (i.e.,

Category A, Category B or Category A and B limits not met), the battery may be considered OPERABLE provided that:

1.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, pilot cell electrolyte levels and float voltages are verified to meet Category C Allowable Values, AND

2.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and once per 7 days thereafter, all battery cell parameters meet Category C Allowable Values, AND

3.

Within 31 days, all battery cell parameters are restored to within Category A and Category B limits of this Table.

(#)

Any Category C parameter not within its Allowable Value indicates an inoperable battery.

(a)

Corrected for electrolyte temperature and level.

(b)

OR battery charging current is less than 2 amperes when on float charge.

(c)

May be corrected for average electrolyte temperature.

(d)

Electrolyte level may exceed 1/4" above maximum level indication mark if an equalizing charge is in progress, or an equalizing charge has been completed within the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

HOPE CREEK 3/4 8-15 Amendment No. 87 Add proposed TS 5.5.11, Battery Monitoring and Maintenance Program M02 See ITS 3.8.6 LA10 See ITS 3.8.6 See ITS 3.8.6 5.5.11.b ITS A01 ITS 5.5

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 3.11.1.1 Deleted 3.11.1.2 Deleted 3.11.1.3 Deleted HOPE CREEK 3/4 11-1 Amendment No. 121 A01 ITS 5.5

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTION:

a.

With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.

HOPE CREEK 3/4 11-2 Amendment No. 187 Explosive Gas and Storage Tank Radioactivity Monitoring Program LA11 5.5.6 5.5.6 5.5.6.b LA11 5.5.6.b the amount that would result in a concentration that is 10 times the limits in 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents L04 ITS A01 ITS 5.5

RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS 3.11.2.1 Deleted 3.11.2.2 Deleted 3.11.2.3 Deleted 3.11.2.4 Deleted 3.11.2.5 Deleted 3.11.2.6 Deleted HOPE CREEK 3/4 11-3 Amendment No. 121 ITS A01 ITS 5.5

THIS PAGE INTENTIONALLY BLANK PAGES 3/4 11-5 THROUGH 3/4 11-16 HAVE BEEN DELETED HOPE CREEK 3/4 11-4 Amendment No. 121 ITS A01 ITS 5.5

DESIGN FEATURES 5.4 REACTOR COOLANT SYSTEM (continued)

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 21,970 cubic feet at a nominal steam dome saturation temperature of 547°F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:

a.

A keff equivalent to less than or equal to 0.95 when flooded with unborated water, including all calculational uncertainties and biases as described in Section 9.1.2 of the FSAR.

b.

A nominal 6.308 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The keff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 199' 4".

CAPACITY 5.6.3 The spent fuel storage pool shall be limited to a storage capacity of no more than 4006 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.

HOPE CREEK 5-5 Amendment No. 184 See ITS 4.0 5.5.4 LA12 This program provides controls to track the reactor cyclic and transient occurrences specified in UFSAR Table 5.3-1 to ensure the reactor components are maintained within the design limits.

ITS A01 ITS 5.5

TABLE 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT Reactor 120 heatup and cooldown cycles 70°F to 546°F to 70°F 80 step change cycles Loss of feedwater heaters 180 reactor trip cycles 100% to 0% of RATED THERMAL POWER 130 hydrostatic pressure and Pressurized to 930 and leak tests 1250 psig HOPE CREEK 5-6 LA12 A01 ITS 5.5

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 1 of 14 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.8.4 provides surveillance requirements with specified frequencies.

CTS 4.0.2 (ITS SR 3.0.2) states that each Surveillance Requirement shall be performed within its specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. ITS 5.5.2 includes a clarifying statement that the provisions of SR 3.0.2 are applicable. This changes the CTS by adding specific clarification that the provisions of the SR extension apply to this specification.

A statement of applicability of SR 3.0.2 for ITS Specification 5.5.2 is needed to ensure the frequency extension provision in CTS 4.0.2 is maintained because in the ITS presentation this provision is not applied to frequencies identified in the Administrative Controls Section of the ITS Specifications unless specifically stated. This change is a clarification required to maintain provisions that would be allowed in the LCO sections of the Technical Specifications and is considered administrative because it does not result in a technical change to the CTS.

A03 The Diesel Fuel Oil Testing Program (CTS 6.8.4.e) is a program in the Administrative Controls Chapter 5.0 (ITS 5.5.7) with specified surveillance frequencies. CTS 4.0.2 (ITS SR 3.0.2) states that each Surveillance Requirement shall be performed within its specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. CTS 4.0.3 (ITS SR 3.0.3) allows, in part, if it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever is greater. ITS 5.5.7 includes a clarifying statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable.

This changes the CTS by adding specific clarification that the provisions of ITS SR 3.0.2 and SR 3.0.3 apply to the program.

The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to ensure the provisions of CTS 4.0.2 and 4.0.3 are maintained because in the ITS presentation these provisions are not applied to frequencies identified in the Administrative Controls Section of the ITS Specifications unless specifically stated. The addition of the sentence regarding SR 3.0.2 and SR 3.0.3 being applicable to the program test frequencies will provide consistency with the application of these requirements in CTS 6.8.4.d (ITS 5.5.6), "Explosive Gas Monitoring Program," which already has a statement that Surveillance Requirements 4.0.2 (ITS SR 3.0.2) and 4.0.4 (ITS SR 3.0.3) are applicable to the

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 2 of 14 program surveillance frequencies. Therefore, the lack of applicability of these SRs to the diesel fuel oil testing program introduces confusion. This change is designated as administrative because it is a presentation preference that does not result in technical changes to the CTS.

A04 TS 6.8.4.f states, in part, that the provisions of Specification 4.0.2 do not apply to test frequencies in the Primary Containment Leak Rate Testing Program. ITS 5.5.10.f states, "Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J." This changes the CTS by exchanging the requirement that CTS 4.0.2 does not apply with the statement to clarify that Primary Containment Leakage Rate Testing Program test frequencies required by a regulation cannot be extended by the use of any Technical Specification allowance, including CTS 4.0.2.

The change is designated as an administrative change to clarify that no generic testing frequency allowance shall be construed to allow modifying a testing Frequency required pursuant to a regulation and is acceptable because it does not result in technical change to the CTS programmatic requirement.

A05 The Radioactive Effluent Controls Program (CTS 6.8.4.g) is a program in the Administrative Controls Chapter 5.0 (ITS 5.5.3) and the liquid holdup tank requirements of CTS 3/4.11.4 are included in the Explosive Gas and Storage Tank Radioactivity Monitoring Program (ITS 5.5.6). The Ventilation Filter Testing Program (VFTP) (ITS 5.5.5) incorporates filter testing requirements from CTS 4.6.5.3.1, CTS 4.6.5.3.2, and CTS 4.7.2.1.1. As such, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extension do apply. This changes the CTS by specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the programs.

The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. The addition of the sentence about SR 3.0.2 and 3.0.3 being applicable to the program test frequencies will provide consistency with the application of these requirements moved from CTS 4.6.5.3.1, CTS 4.6.5.3.2, CTS 4.7.2.1.1, and CTS 4.11.1.4, which already allow application of SR 3.0.2 and 3.0.3 to the filter testing frequencies. This change is designated as administrative because it does not result in a technical change to the CTS.

A06 CTS 4.6.5.3.1.c, d, e, f, and g; and CTS 4.6.5.3.2.c, e, and f; and CTS 4.7.2.1.1.c, d, e, f, and g, state that specified tests must be performed in accordance with the surveillance frequency control program or under specific conditions (e.g., 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation). ITS 5.5.5 requires these tests be performed in accordance with the Surveillance Frequency Control Program and as specified in Regulatory Guide 1.52 Revision 2. This changes the CTS by referencing a Regulatory Guide for the specific conditions under which these tests must be performed.

The purpose of the CTS specifications is to provide frequencies and conditions under which the required tests must be performed to ensure the filters are able to perform their functions. In the CTS condition for testing air filtration and

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 3 of 14 adsorption units are specified such as after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, following painting fire or chemical release, etc. This change is acceptable because similar conditions for testing air filtration and adsorption units are provided in U.S. NRC Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."

(RG 1.52 Rev. 2) RG 1.52 Rev. 2, presents methods acceptable to the NRC staff for implementing the Commission's regulations in Appendix A to 10 CFR Part 50 with regard to design, testing, and maintenance criteria for air filtration and adsorption units of engineered-safety-feature (ESF) atmosphere cleanup systems in light-water cooled nuclear power plants. This change is designated as administrative because it does not result in technical changes to the CTS.

A07 CTS 4.6.5.3.1.c.1, c.2, d, f, and g; and CTS 4.6.5.3.2.c.1, and f; and CTS 4.7.2.1.1.c.1, c.2, d, f, and g, state that the tests must be performed in accordance with Regulatory Positions C.5.a, C.5.c, C.5.d, and C.6.b of Regulatory Guide 1.52, Rev 2, as delineated. ITS 5.5.5 states that testing must be performed in accordance with Regulatory Guide 1.52, Rev 2. This changes the CTS by removing specific regulatory positions for the referenced NRC regulatory guide.

The purpose of CTS 4.6.5.3.1.c.1, c.2, d, f, and g; and CTS 4.6.5.3.2.c.1, and f; and CTS 4.7.2.1.1.c.1, c.2, d, f, and g, is to provide direction on the applicable NRC regulatory guidance for testing of the safety related air filtration and adsorption units. This change only removes reference to specific regulatory positions that are already a part of Regulatory Guide 1.52, Revision 2.

Therefore, no changes are being made to the required testing of the air filtration and adsorption units. This change is designated as administrative because it does not result in a technical change to the CTS.

A08 CTS 4.6.5.3.1.c.1, c.2, d, f, and g; and CTS 4.6.5.3.2.c.1, and f; and CTS 4.7.2.1.1.c.1, c.2, d, f, and g, state, in part, that testing is performed in accordance with... Regulatory Guide 1.52, Revision 2, March 1978. ITS 5.5.5 states, in part, that testing is performed in accordance with Regulatory Guide 1.52, Revision 2. This changes the CTS by deleting the date designation of the regulatory guide.

The purpose of CTS 4.6.5.3.1.c.1, c.2, d, f, and g; and CTS 4.6.5.3.2.c.1, and f; and CTS 4.7.2.1.1.c.1, c.2, d, f, and g, is to provide direction on the applicable NRC regulatory guidance for testing of the safety related air filtration and adsorption units. ITS continues to provide this direction with the date designation deleted. This change is acceptable because Regulatory Guide versions include both a revision number and a corresponding date of the revision. By using one or the other, the correct Regulatory Guide version can be determined. Therefore, deleting the date designation of the Regulatory Guide is administrative and does not result in a technical change to the CTS.

A09 CTS 4.6.5.3.1.c.3, CTS 4.6.5.3.2.c.2, and CTS 4.7.2.1.1.c.3 require verifying the specified flow rate within +/-10% of the required flow rate for each FRVS ventilation unit, each FRVS recirculation unit, and CREF subsystems, respectively, during system operation when tested in accordance with ANSI

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 4 of 14 N510-1980. The specified flow rates for these ventilation systems are duplicative of the flow rate requirement specified in CTS 4.6.5.3.1.c.1, CTS 4.6.5.3.2.c.1, and CTS 4.7.2.1.1.c.1, respectively. ITS 5.5.5.a and b require filter testing of the CREF System and the FRVS ventilation units to be performed at a flowrate of 4000 cfm and 9000 cfm, respectively, +/- 10%. Additionally, ITS 5.5.5.a requires filter testing of the FRVS recirculation units to be performed at a flowrate of 30,000 cfm +/- 10%, which accomplishes the same purpose as CTS 4.6.5.3.1.c.3, CTS 4.6.5.3.2.c.2, and CTS 4.7.2.1.1.c.3. Therefore, these specific Surveillances are unnecessary. This changes the CTS by deleting a duplicative requirement to verify FRVS and CREF system flowrates while operating the ventilation system.

This change is designated as administrative because it does not result in a technical change to the CTS.

A10 CTS 4.6.5.3.1.e.1, CTS 4.6.5.3.2.e.1, and CTS 4.7.2.1.1.e.1 require verifying the pressure drop across the combined HEPA filters and charcoal adsorber banks for each FRVS ventilation unit, each FRVS recirculation unit, and CREF subsystems, respectively, during system operation at the specified flow rate.

ITS 5.5.5.d also requires verifying the pressure drop across the combined HEPA filters and charcoal adsorber banks for the FRVS ventilation and recirculation units, and CREF subsystems. This changes the CTS by clarifying when they are tested in accordance with ANSI N510-1980.

This clarification does not result in a technical change because ventilation filter testing, including the pressure drop test, is performed in accordance with ANSI N510-1980. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 The CTS does not include program requirements for the Safety Function Determination Program (SFDP). The ITS includes the SFDP, as required by ITS LCO 3.0.6, to ensure loss of safety function is detected and appropriate actions taken as a result of multiple support system inoperabilities. This changes the CTS by adding the Safety Function Determination Program.

ITS 5.5.9, Safety Function Determination Program (SFDP), is included to support implementation of ITS LCO 3.0.6 and is necessary to determine if loss of safety function exists as a result of the support system inoperability and the LCO 3.0.6 exception to entering supported system Condition and Required Actions. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken. The SFDP shall contain provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected, provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists, provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities, and other appropriate limitations and remedial or compensatory R1 R1

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 5 of 14 actions. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

M02 The CTS does not include a requirement for the Battery Monitoring and Maintenance Program. The ITS includes a requirement for this program. This changes the CTS by adding ITS 5.5.11,"Battery Monitoring and Maintenance Program," consistent with the ISTS. The Battery Monitoring and Maintenance Program is included to provide for battery restoration and maintenance. The specific wording associated with this program is specified in ITS 5.5.11. The Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite-Update to TSTF-360," (76FR54510) references the model application and safety evaluation (SE) for plant-specific adoption of TSTF-500, Revision2 (NRC ADAMS Accession No. ML111751792). PSEG has verified the applicable information specified in Section 2.2 of the TSTF-500 model application, including applicable Updated Final Safety Analysis Report (UFSAR) information. PSEG will update the UFSAR, as necessary, to include any UFSAR information listed in Section 2.2 of the TSTF-500 model application that is not currently reflected in the HCGS UFSAR.

This change is acceptable and necessary because it supports implementation of the less restrictive ITS requirements related to the Class 1E batteries and battery cell parameters specified in ITS Specifications 3.8.4 and 3.8.6. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of SR Frequency to the Surveillance Frequency Control Program) CTS 6.8.4.a contains a program requiring leakage testing of primary coolant sources outside containment with a stated frequency of at refueling cycle intervals. CTS 6.16 requires measurement, at designated locations, of the control room envelope (CRE) pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the Control Room Emergency Filtration System operating at the flow rate required by Surveillance Requirement 4.7.2.1.c.1, at a frequency of 36 months on a Staggered Test Basis and use the results as part of the 36 month assessment of the CRE boundary. CTS 6.8.4.e requires that total particulate concentration of the diesel generator fuel oil is 10 mg/l when tested every 92 days. ITS 5.5.2, 5.5.12, and 5.5.7 contain similar requirements, respectively, but specify the periodic Frequency as "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified periodic Frequency for these tests to the Surveillance Frequency Control Program.

The purpose of these Surveillances is to assure that the necessary quality of systems and components is maintained. The removal of these details related to

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 6 of 14 surveillance frequencies from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing Frequency is removed from Technical Specifications and placed under licensee control pursuant to the methodology described in NEI 04-10. The surveillance test requirements remain in the Technical Specifications. The control of changes to the Surveillance Frequencies is in accordance with the plant Surveillance Frequency Control Program (SFCP), which is retained in ITS Section 5.5. The SFCP provides the necessary administrative controls requiring that surveillance testing, calibration and inspection are conducted at a frequency assuring the necessary quality of systems and components is maintained, facility operation will be within safety limits, and the limiting conditions for operation will be met pursuant to the requirements of 10 CFR 50.36(c)(3). The proposed change to relocate periodic frequencies in the administrative controls section of Technical Specifications has been previously approved for Wolf Creek Generating Station Unit 1 in Amendment 227, dated April 8, 2021 (NRC ADAMS Accession No. ML21053A117), River Bend Station Unit 1 in Amendment 196, dated April 29, 2019 (NRC ADAMS Accession No. ML19066A008), and Grand Gulf Nuclear Station Unit 1 in Amendment 219, dated June 11, 2019 (NRC ADAMS Accession No. ML19094A799). HCGS adopted a Surveillance Frequency Control Program in Amendment No. 187 (ADAMS Accession No. ML103410243) as contained in CTS 6.8.4.j. This change is acceptable because the testing frequencies will be controlled in accordance with the SFCP requirements retained in ITS, which ensure changes are properly evaluated.

This change is designated as a less restrictive removal of detail change, because the Surveillance Frequencies are being removed from the Technical Specifications.

LA02 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.8.4.b, "In-Plant Radiation Monitoring,"

describes a program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. ITS 5.5 does not include this program. This changes the CTS by moving the requirements for the In-Plant Radiation Monitoring Program to the UFSAR.

The removal of this requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The CTS 6.8.4.b program is designed to minimize radiation exposure to plant personnel in vital areas of the plant after an accident and has no impact on nuclear safety or the health and safety of the public. This change is acceptable because the program requirements will be adequately controlled in the UFSAR.

The UFSAR is controlled under 10 CFR 50.59 which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications.

LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.8.4.e.a.1 and CTS 6.8.4.e.a.3 state that new diesel fuel oil requires an API gravity or absolute specific gravity and bulk water

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 7 of 14 and sediment must be within limits for ASTM 2D fuel oil. ITS 5.5.7.a.1 states that new diesel fuel oil API gravity or an absolute specific gravity must be within limits and ITS 5.5.7.a.3 requires that new diesel fuel oil has a clear and bright appearance with proper color or a water and sediment content must be within limits. The ITS Bases for SR 3.8.3.3 contains the test, limits, and applicable ASTM 2D fuel oil standards associated with the new fuel oil parameters. This changes the CTS by removing the procedural details of the applicable fuel oil standard the new fuel oil parameter limits apply to and relocating the type of fuel oil to the Bases.

The purpose of CTS 6.8.4.e.a is to ensure the acceptability of new fuel oil prior to addition to the storage tanks. The removal of these details for the applicable fuel oil standard the new fuel oil parameter limits apply to from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements that the specified fuel oil parameters must be within limits and continues to assure protection of public health and safety. Also, this change is acceptable because this type of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA04 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.8.4.e.c states, in part, that diesel fuel oil total particulate concentration of the stored fuel oil is 10 mg/l when tested periodically in accordance with ASTM D-2276, modified as follows: The 0.8 micron membrane filters specified in ASTM D-2276 may be replaced with membrane filters up to 3.0 microns. ITS 5.5.7.c states that the total particulate concentration of the fuel oil is 10 mg/l when tested in accordance with the Surveillance Frequency Control Program. This changes the CTS by relocating the information associated with testing procedure details and exception to the ITS Bases.

The purpose of CTS 6.8.4.e.c is to ensure the total particulate concentration of the fuel oil is within limits. The removal of these testing procedure details and exception from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirements that total particulate concentration of diesel fuel oil must be within limits, which continues to assure protection of public health and safety. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases (refer to ITS SR 3.8.3.3 Bases). Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

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DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 8 of 14 LA05 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.8.4.h, "Radiological Environmental Monitoring Program," describes a program to monitor the radiation and radionuclides in the environs of the plant. ITS Section 5.5 does not require this program. This changes the CTS by moving the requirements for the Radiological Environmental Monitoring Program to the ODCM.

The purpose of these program requirements is to provide representative measurements of radioactivity in the highest potential exposure pathways, and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The removal of the requirement for this program from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.6.2, "Radiological Effluent Release Report," continues to require an annual report of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ODCM. Changes to the ODCM are controlled by the ODCM change control process specified in ITS 5.5.1, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because the requirements for a program are being removed from the Technical Specifications.

LA06 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.13 describes the process for control of changes to the Process Control Program (PCP). The ITS does not include these requirements. This changes the CTS by moving the requirements of the PCP to the UFSAR.

The removal of these requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Compliance with these regulations is required by the HCGS Operating License, and procedures are the method to ensure compliance with the program. Regulations provide an adequate level of control for the affected requirements and inclusion of this requirement in the Technical Specifications is not necessary. Also, this change is acceptable because these details will be adequately controlled in the UFSAR. Any changes to the UFSAR are controlled under 10 CFR 50.59, which ensures changes are properly evaluated and the NRC is updated in accordance with 10 CFR 50.71(e). This change is designated as a less restrictive removal of requirements because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications.

LA07 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.14.b uses the title "General Plant Manager.

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 9 of 14 ITS 5.1.1.b related to changes to the ODCM, uses the generic title "plant manager." This changes the CTS by moving the specific organizational title to the Quality Assurance Topical Report (QATR) and replacing it with a generic title.

The purpose of CTS 6.12.b is to ensure changes to the ODCM are properly evaluated and approved by plant management. The removal of these details, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific organizational titles out of the Technical Specifications is consistent with the NRC letter from C Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994. The requirement for the plant manager to approve ODCM changes is retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the QATR. Any changes to the QATR are controlled pursuant to 10 CFR 50.54(a)(3), which ensures changes are properly evaluated, and the NRC is updated in accordance with 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA08 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 4.6.1.1.a, 3.6.1.2.b, 3.6.1.2 Action b, 4.6.3.5 footnote *, provide details of primary containment leak rate testing frequencies, exceptions, and exemptions. ITS 5.5.10, Primary Containment Leakage Rate Testing Program, establishes leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. ITS does not include details related to the frequencies, exceptions, and exemptions. This changes the CTS by moving the primary containment leakage testing requirement details to the Primary Containment Leakage Rate Testing Program.

The removal of details related to primary containment leak rate testing frequencies, exceptions, and exemptions from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The requirements associated with primary containment leak rate testing frequencies, exceptions, and exemptions are adequately addressed within 10 CFR 50, Appendix J. Option B, as modified by approved exemptions, and ITS 5.5.10, which requires compliance with the guidelines NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008. Compliance with these regulations is required by the HCGS Operating License, and procedures are the method to ensure compliance with the program. Regulations provide an adequate level of control for the affected requirements. Therefore, since there is no change in the technical requirements, and future revisions must also be made by Technical Specification amendment request or exemption to the regulations, the relocation of these details continues to provide adequate protection of the public health and safety. This change is designated as a less restrictive removal

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 10 of 14 of detail change because details of primary containment leak rate testing are being removed from the Technical Specifications.

LA09 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS 4.8.2.1.b.2 requires verification that there is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohms, excluding cable intercell connections.

CTS SR 4.8.2.1.c requires, in part, verifying: 1. the cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration,

2. the cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material, and 3. the resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohms, excluding cable intercell connections. ITS 5.5.11, "Battery Monitoring and Maintenance Program,"

requires the program be maintained in accordance with IEEE Standard 450-2010 and includes the requirement for the program to provide provisions to limit battery connection resistance (ITS 5.5.11.d). This changes the CTS by removing the information from the specification to the Battery Monitoring and Maintenance Program implementing document.

The removal of details related to battery cell conditions and connection resistance from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Battery Monitoring and Maintenance Program is established in accordance with the requirements contained in IEEE 450-2010. The IEEE standard contains the detail associated with battery preventative maintenance, including visual inspections to detect degradation. The requirements to maintain the station batteries in accordance with the standard remain in the Technical Specifications. Additionally, the program requirement to provide provisions to limit battery connection resistance is also retained in the Technical Specifications. This change is designated as a less restrictive removal of detail change, because the battery visual inspection and cell connection resistance details are being removed from the Technical Specifications.

LA10 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS Table 4.8.2.1-1 footnote (c) states, in part the float voltage of 2.13 volts may be corrected for average electrolyte temperature. ITS 5.5.11.a requires a program with actions to restore battery cells with float voltage < 2.13 V and ITS 5.5.11 b requires a program with actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be < 2.13 V. This changes the CTS by moving information from the specification to the Battery Monitoring and Maintenance Program implementing document.

The purpose of CTS Table 4.8.3.1-1 footnote (c) is to allow correcting connected cell float voltage for variation in the average electrolyte temperature. The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.11 retains the requirement for cell float voltage 2.13 V. Also, this change is acceptable because these types of procedural details will be adequately

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 11 of 14 controlled by the requirements of the Battery Monitoring and Maintenance Program required by ITS Chapter 5. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA11 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 3.11.1.4 specifies Limiting Condition for Operation, Applicability, and Action requirements for Liquid Holdup Tanks.

ITS 5.5.6 does not include these details. This changes the CTS by relocating the details of the methods for implementing the limit requirements to the ODCM.

The purpose of these CTS requirements is to provide limits on the quantity of radioactive material contained in any outside temporary tank. ITS 5.5.6, "Explosive Gas and Storage Tank Radioactivity Monitoring Program," provides regulatory control over the details to be relocated. As a result, the details to be relocated are not required to be in the ITS to provide adequate protection of the public health and safety. Additionally, changes to the ODCM will be controlled in accordance with the ODCM change process described in ITS 5.5.1.

LA12 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 5.7, "Component Cyclic or Transient Limit,"

provides details for maintaining components within cyclic or transient limits and lists the reactor cyclic and transient limits in CTS Table 5.7.1-1. ITS 5.5.4 does not contain these details; rather, ITS 5.5.4 provides controls to track the reactor cyclic and transient occurrences specified in UFSAR Table 5.3-1 to ensure the reactor components are maintained within the design limits. This changes the CTS by relocating the details of the component cyclic and transient limits to the UFSAR.

The requirement to maintain a program to monitor the reactor cyclic and transient occurrences is maintained in ITS. ITS 5.5.4 provides adequate regulatory control over the details to be relocated. As a result, the details to be relocated are not required to be in the ITS to provide adequate protection of the public health and safety. Also, this change is acceptable because this type of procedural detail will be adequately controlled in the UFSAR. Any changes to the UFSAR are controlled under 10 CFR 50.59, which ensures changes are properly evaluated and the NRC is updated in accordance with 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because details of a program are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 6.8.4.e.a.3, requires that new diesel fuel oil bulk water and sediment be within limits. ITS 5.5.7.a.3 states that new fuel oil must be shown acceptable for use prior to addition to storage tanks by determining that the fuel oil has a clear and bright appearance with proper color or a water and sediment content within limits.

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 12 of 14 This changes the CTS by providing an option to ensuring the bulk water and sediment quality of the new fuel oil for the diesel generators is within limits.

The purpose of CTS 6.8.4.e.a.3 is to ensure the acceptability of new diesel fuel prior to addition to the storage tanks. CTS 6.8.4.e.a.3 requires bulk water and sediment to be within limits to establish, in part, the acceptability of new fuel oil for use prior to addition to storage tanks. ITS 5.5.7.a.3 is proposed to allow a clear and bright appearance with proper color test be performed to establish the acceptability of new fuel oil instead of the "bulk water and sediment" test. ASTM D4176-86,"Standard Test Method for Free Water and Particulate Contamination in Distillate Fuels (Clear and Bright Pass/Fail Procedures)," verifies that the new fuel oil has a clear and bright appearance with proper color. The "clear and bright" test is only applicable to fuel oils that meet the ASTM D4176 color rating requirements (i.e., an ASTM D1500, "Test Method for ASTM Color of Petroleum Products (ASTM Color Scale)," color rating of five or less). The clear and bright test is a qualitative test for quickly determining free water and particulate contamination in distillate fuels and is, therefore, subject to human interpretation.

For example, if an attempt is made to use the qualitative "clear and bright" test with darker colored fuels (e.g., for high sulfur fuel oil that has been dyed in accordance with EPA mandated requirements), the presence of free water or particulate could be obscured and missed by the viewer. Therefore, ITS 5.5.11.a.3 has been revised to allow "clear and bright" test to quickly determine if the new diesel fuel oil is acceptable. If unable to ensure the new diesel fuel oil is acceptable to add to the storage tanks by the "clear and bright" test, the water and sediment content test is a quantitative test using centrifuge methods. In ASTM D975-90, ASTM D1796, Standard Method for Water and Sediment in Fuel Oils by the Centrifuge Method (Laboratory Procedure), is an acceptable standard for the water and sediment content test. This change is designated as less restrictive because Surveillance acceptance criteria required in the CTS will have alternative acceptance criteria allowed in ITS.

L02 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.6.5.3.1.c.2 and d, and 4.7.2.1.1.c.2 and d require, in part, that the laboratory test of the charcoal adsorber be completed within 31 days of removal from the ventilation unit. ITS 5.5.5 does not include this requirement. This changes the CTS by removing the surveillance performance requirement of being completed within 31 days.

The purpose of CTS 4.6.5.3.1.c.2 and d, and 4.7.2.1.1.c.2 and d is to verify that a laboratory analysis of a representative carbon sample, is within limits. ITS 5.5.5 continues to ensure a laboratory analysis of a representative carbon sample is within limits. This change is acceptable because it was previously approved by the NRC during the standard technical specification upgrade project of revising NUREG-0123 into NUREG-1433. This change was identified and justified in NEDC-31681. NEDC-31681 stated this was acceptable because plant administrative controls along with the recommendation of the referenced guidelines are sufficient to assure proper testing. The same justification applies to HCGS. This change is designated as less restrictive because less stringent surveillance performance requirements are being applied in the ITS than applied in the CTS.

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 13 of 14 L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 4.7.2.1.1.e.4 requires verification of maintaining less than or equal to 70%

humidity through the carbon adsorbers by performance of a channel calibration of the humidity control instrumentation. ITS 5.5.5 does not include the requirement to verify humidity by performance of a channel calibration of the humidity control instrument. This changes the CTS by eliminating the humidity verification acceptance criteria of a channel calibration.

The purpose of CTS 4.7.2.1.1.e.4 is to verify heater performance and ensure it is adequate to maintain humidity. This change is acceptable because performing a Channel Calibration of humidity control instrumentation does not confirm humidity is or is not maintained within limits, only that the monitoring equipment is functional. ITS continues to verify heater performance and requires humidity through the carbon adsorbers be verified within limits. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than applied in the CTS.

L04 (Category 1 - Relaxation of LCO Requirements) CTS 3.11.1.4 states that the quantity of radioactive material contained in any outside temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. ITS 5.5.6.b provides similar requirements for any outside temporary tank stating, in part, that the quantity of radioactivity contained in any outside temporary tank is less than the amount that would result in a concentration that is 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. This changes the CTS by changing the radioactive material limit from curies to a concentration limit based on dose and increasing the limitation on the concentration of radioactive material contained in any outside temporary tank to ten times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2.

The purpose of CTS 3.11.1.4 is to provide limitations on the concentration of radioactive material contained in any outside temporary tank. ITS 5.5.6.b continues to provide limitations on the concentration of radioactive material released contained in any outside temporary tank updated for the revision of 10 CFR 20 in 1991. As stated in the Introduction to Appendix B of the revised 10 CFR Part 20, the liquid effluent concentration values given in Appendix B, Table 2, Column 2, are concentrations which, if inhaled or ingested continuously for one year, would produce a total effective dose equivalent of 0.05 rem (50 millirem or 0.5 millisieverts). The liquid effluent concentration values given in the old 10 CFR Part 20 are concentrations which, if continuously present, would produce a dose of 500 millirem in a year, therefore the new effluent concentration values are more restrictive by a factor of 10. Therefore, the limit on concentration of radioactive material contained in any outside temporary tank is revised in the ITS based on the new 10 CFR 20 concentration limits, by a factor of 10, to bring the limit on a par with the old 10 CFR 20 concentration limit and will not affect compliance with liquid effluent dose limits. Additionally, changing the concentration units contained in temporary outside storage tanks from a specific curie content to that amount that would result in 10 times the concentration values, allows the radioactivity curie content to change based on

DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 14 of 14 the type of radioactivity contained in the temporary storage tank and the tank location with respect to the nearest unrestricted water source, while ensuring continued compliance with regulations. This change is acceptable because it maintains the same overall level of liquid effluent control to unrestricted areas while retaining the operational flexibility that exists with CTS under the previous 10 CFR 20 requirements. This limitation (i.e., less than 10 times the concentration values...) provides reasonable assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within the Section II.A design objectives of Appendix I to 10 CFR 50 and the restrictions authorized by 10 CFR 20.1301(e). These changes are intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements. This change is designated as less restrictive because less stringent criteria are being applied in the ITS than were applied in the CTS.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-1 Rev. 5.0 Hope Creek Amendment XXX CTS 2

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

5.5.1 Offsite Dose Calculation Manual (ODCM)

a.

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

b.

The ODCM shall also contain 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 [5.6.1] and Specification [5.6.2].

Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1.

Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s) and

2.

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

b.

Shall become effective after the approval of the plant manager, and

c.

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.

6.14 6.14.a 6.14.b 6.14.c 6.14.a.1) 6.14.a.2) 1.27 1.27 1

106 (pre 1994 regulation) 16

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-2 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.2 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

[the Low Pressure Core Spray, High Pressure Coolant Injection, Residual Heat Removal, Reactor Core Isolation Cooling, hydrogen recombiner, process sampling, and Standby Gas Treatment]. The program shall include the following:

a.

Preventive maintenance and periodic visual inspection requirements and

b.

Integrated leak test requirements for each system at least once per

[18] months.

The provisions of SR 3.0.2 are applicable.

[ 5.5.3 Post Accident Sampling

REVIEWERS NOTE ------------------------------------------

This program may be eliminated based on the implementation of NEDO-32991, Revision 0, Regulatory Relaxation For BWR Post Accident Sampling Stations (PASS), and the associated NRC Safety Evaluation dated June 12, 2001, and implementation of the following commitments:

1. [Licensee] has developed contingency plans for obtaining and analyzing highly radioactive samples from the reactor coolant system, suppression pool, and containment atmosphere. The contingency plans will be contained in emergency plan implementing procedures and implemented with the implementation of the License amendment. Establishment of contingency plans is considered a regulatory commitment.
2. The capability for classifying fuel damage events at the Alert level threshold has been established for [Plant] at radioactivity levels of 300 mCi/cc dose equivalent iodine. This capability may use a normal sampling system or correlations of radiation readings to coolant concentrations. This capability will be described in emergency plan implementing procedures and implemented with the implementation of the License amendment. The capability for classifying fuel damage events is considered a regulatory commitment.
3. [Licensee] has established the capability to monitor radioactive iodines that have been released to offsite environs. This capability is described in our emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment.

This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program 6.8.4.a 6.8 Core Spray containment H2/02 analyzer, post-accident sampling and control rod drive hydraulic (scram discharge portion) at a frequency in accordance with the Surveillance Frequency Control Program 4

1 3

DOC LA01 6.8.4.a.1 6.8.4.a.2 S2 S2

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-3 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.3 Post Accident Sampling (continued) shall include the following:

a.

Training of personnel,

b.

Procedures for sampling and analysis, and

c.

Provisions for maintenance of sampling and analysis equipment. ]

5.5.4 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:

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 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.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology 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, 4

3 6.8.4.g 6.8.4.g.1) 6.8.4.g.2) 6.8.4.g.3) 6.8.4.g.4) 6.8.4.g.5) 6.8.4.g.6) 4 14 2

the II 106 (pre 1994 regulation) 16 16 (pre 1994 regulation) 16

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-4 Rev. 5.0 Hope Creek Amendment XXX CTS 2

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

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 mrem/yr to the whole body and a dose rate 3000 mrem/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 mrem/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,

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 Mark II containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable (in BWR/4s with Mark II containments).

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the FSAR Section [ ], cyclic and transient occurrences to ensure that components are maintained within the design limits.

[ 5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.

3 4

Reactor Building Ventilation System, Hardened Torus Vent, or the Filtration Recirculation and Ventilation System (FRVS) 6.8.4.g.7) 6.8.4.g.8) 6.8.4.g.9) 6.8.4.g.11) 6.8.4.g.10) 2 5

4 the the 2

2 13 5.7.1 4

reactor specified in UFSAR Table 5.3-1 the reactor 15 conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1 (pre 1994 regulation),

16

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-5 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance Program (continued)

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ]

5.5.7 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, Revision 2, ASME N510-1989, and AG-1].

a.

Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below

[+/- 10%].

ESF Ventilation System Flowrate

[ ]

[ ]

b.

Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%].

ESF Ventilation System Flowrate

[ ]

[ ]

c.

Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in [Regulatory Guide 1.52, Revision 2], shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity specified below.

ESF Ventilation System Penetration RH Face Velocity (fps)

[ ]

[See Reviewer's

[See

[See Reviewer's Note]

Reviewer's Note]

Note]

5 Control Room Emergency Filtration (CREF) System Filtration Recirculation and Ventilation System (FRVS) - ventilation units FRVS - recirculation units 4000 cfm 9000 cfm 30,000 cfm 4.6.5.3.1.c.1 4.6.5.3.1.f 4.6.5.3.2.c.1 4.6.5.3.2.f 4.7.2.1.1.c.1 4.7.2.1.1.f 4.6.5.3.1.g 4.7.2.1.1.g CREF System FRVS - ventilation units 4000 cfm 9000 cfm CREF System FRVS - ventilation units 4.6.5.3.1.c.2 4.6.5.3.1.d 4.7.2.1.1.c.2 4.7.2.1.1.d 4.7.2.1.1.e.4 0.5%

5%

70%

95%

5 6

1 6

1 6

6 6

6 6

5 1

1 1

Insert 1 S2

Hope Creek Insert Page 5.5-5 INSERT 1 ASME N510-1980, and ASTM D3803-1989, as described herein.

The tests described in Specification 5.5.5.a through 5.5.5 c shall be performed at the frequencies specified in Regulatory Guide 1.52, Revision 2, except for the 18 month periodic frequency. The tests described in Specification 5.5.5.a through 5.5.5 c shall be performed at a periodic frequency in accordance with the Surveillance Frequency Control Program.

The tests described in Specification 5.5.5.d and 5.5.5 e shall be performed at a frequency in accordance with the Surveillance Frequency Control Program.

2 ITS 5.5

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-6 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (continued)

REVIEWER'S NOTE----------------------------------------

The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate.

ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95%

(or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test.

Allowable Penetration = [(100% - Methyl Iodide Efficiently

  • for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor]

When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following:

Safety factor 2 for systems with or without humidity control.

Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions.

If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified.

  • This value should be the efficiency that was incorporated in the licensee's accident analysis which was reviewed and approved by the staff in a safety evaluation.
d.

Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [+/- 10%].

ESF Ventilation System Delta P Flowrate

[ ]

[ ]

[ ]

[ e.

Demonstrate that the heaters for each of the ESF systems dissipate the value specified below [+/- 10%] when tested in accordance with

[ASME N510-1989].

5 4000 cfm 9000 cfm 30,000 cfm CREF System FRVS - ventilation units FRVS - recirculation units 0

7.5 in. w.g.

5 in. w.g.

8 in. w.g.

4.6.5.3.1.e.1 4.6.5.3.2.e.1 4.7.2.1.1.e.1 or the exhaust duct of the FRVS recirculation units 7

4.7.2.1.1.e.4 6

6 1

2 1

1 1

5 ASME N510-1980 the CREF System dissipate 13 kW 4.6.5.3.2.c.2 4.7.2.1.1.c.3 S2

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-7 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (continued)

ESF Ventilation System Wattage ]

[ ]

[ ]

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous radioactivity quantities shall be determined following the methodology in [Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be determined in accordance with [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"].

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the [Waste Gas Holdup 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).

b.

A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed into the offgas treatment system] is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled release of the tanks' contents], and

c.

A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the [Liquid Radwaste Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

6.8.4.d 3.11.1.4 5

6 Main Condenser Offgas Treatment System Main Condenser Offgas Treatment System 8

b a

that is 10 times the concentration values in 6.8.4.d 1

6 1

1 8

2 5

8 1

8 5

any outside temporary tank 3.11.1.4 4.11.1.4 DOC L04 any outside temporary tank 6.8.4.d R1 R1 R1

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-8 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.9 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

An API gravity or an absolute specific gravity within limits,

2.

A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3.

A clear and bright appearance with proper color or a water and sediment content within limits,

b.

Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a.,

above, are within limits for ASTM 2D fuel oil, and

c.

Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days.

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

5.5.10 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 Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license or

2.

A change to the updated FSAR or Bases that 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 FSAR.

7 8

at a frequency in accordance with the Surveillance Frequency Control Program 6.15 6.15.a 6.15.b 6.15.b.1 6.15.b.2 6.15.e 6.8.4.e 6.8.4.e.a 6.8.4.e.a.1 6.8.4.e.a.2 6.8.4.e.a.3 DOC L01 6.8.4.e.b 6.8.4.e.c DOC LA01 5

3 5

U U

2 2

S2

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-9 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.10 Technical Specifications (TS) Bases Control Program (continued)

d.

Proposed changes that meet the criteria of Specification 5.5.10b above shall be reviewed and approved by the NRC prior to implementation.

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

5.5.11 Safety Function Determination Program (SFDP)

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

a.

Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected,

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists,

c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities, and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable,

b.

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

c.

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

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are 8

9

8.

6.15 6.15.c 6.15.d 5

5 5

DOC M01

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-10 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued) required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.12 Primary Containment Leakage Rate Testing Program

[OPTION A]

a.

A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

b.

The maximum allowable containment leakage rate, La, at Pa, shall be [ ]%

of containment air weight per day.

c.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests.

2.

Air lock testing acceptance criteria are:

a)

Overall air lock leakage rate is [0.05 La] when tested at Pa.

b)

For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

d.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

e.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

[OPTION B]

a.

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

9 10 6.8.4.f NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008 5

5 9

1 2

6.8.4.f

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-11 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3.

... ]

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is [45 psig]. The containment design pressure is

[50 psig].

c.

The maximum allowable containment leakage rate, La, at Pa, shall be [ ]%

of containment air weight per day.

d.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and 0.75 La for Type A tests.

2.

Air lock testing acceptance criteria are:

a)

Overall air lock leakage rate is [0.05 La] when tested at Pa.

b)

For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

e.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

f.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

[OPTION A/B Combined]

10 50.6 0.5 5 scfh 6.8.4.f 1

5 2

13 1

2 2

1 9

3.6.1.2.a 3.6.1.2.a 3.6.1.2.a 3.6.1.2.b 6.8.4.f.b 3.6.1.3.b 6.8.4.f 6.8.4.f DOC A04 6.8.4.f 6.8.4.f 6.8.4.f 6.8.4.f

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-12 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

a.

A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A]

test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3.

... ]

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is [45 psig]. The containment design pressure is

[50 psig].

c.

The maximum allowable containment leakage rate, La, at Pa, shall be [ ]%

of containment air weight per day.

d.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests].

2.

Air lock testing acceptance criteria are:

a)

Overall air lock leakage rate is [0.05 La] when tested at Pa.

9

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-13 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued) b)

For each door, leakage rate is [0.01 La] when pressurized to

[10] psig.

e.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

f.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.13 Battery Monitoring and Maintenance Program

REVIEWERS NOTE------------------------------------------

This program and the corresponding requirements in LCO 3.8.4, LCO 3.8.5, and LCO 3.8.6 require providing the information and verifications requested in the Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite - Update to TSTF-360," (76FR54510).

This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below:

a.

The program allows the following RG 1.129, Revision 2 exceptions:

1.

Battery temperature correction may be performed before or after conducting discharge tests.

2.

RG 1.129, Regulatory Position 1, Subsection 2, "References," is not applicable to this program.

3.

In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2, "Inspections," the following shall be used: "Where reference is made to the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery."

4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

11 3

2010 9

10 11 11 5

DOC M02 p

S2 17

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-14 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.13 Battery Monitoring and Maintenance Program (continued)

5.

In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string."

b.

The program shall include the following provisions:

1.

Actions to restore battery cells with float voltage < [2.13] V;

2.

Actions to determine whether the float voltage of the remaining battery cells is [2.13] V when the float voltage of a battery cell has been found to be < [2.13] V;

3.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4.

Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5.

A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

5.5.14 Control Room Envelope (CRE) Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE [Main Control Room Environmental Control (MCREC)] System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of [5 rem whole body or its equivalent to any part of the body] [5 rem total effective dose equivalent (TEDE)] for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power 11 12 6.16 Control Room Emergency Filtration 6.16.a 6.16.b 6.16.c a

b c

d 5

11 1

1 11 11 11 11 11 5

1 1

and

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-15 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.14 Control Room Envelope (CRE) Habitability Program (continued)

Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

[The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. ;and]
d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the [MCREC] System, operating at the flow rate required by the VFTP, at a Frequency of [18] months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the [18] month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

[ 5.5.15 Setpoint Control Program

REVIEWER'S NOTE----------------------------------------

Adoption of a Setpoint Control Program requires changes to other technical specifications. See TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS Functions," Option B, for guidance (Agencywide Documents Access and Management System (ADAMS) Accession Number ML101160026).

This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analyses, provides a means for processing changes to instrumentation setpoints, and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that 12 Control Room Emergency Filtration in accordance with the Surveillance Frequency Control Program 6.16.d DOC LA01 6.16.e 6.16.f 1

5 3

1 2

3 12 6.16 S2

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-16 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.15 Setpoint Control Program (continued) testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verifies that instrumentation will function as required.

a.

The program shall list the Functions in the following specifications to which it applies:

1.

LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation;"

2.

LCO 3.3.1.2, "Source Range Monitor (SRM) Instrumentation;"

3.

LCO 3.3.2.1, "Control Rod Block Instrumentation;"

4.

LCO 3.3.2.2, "Feedwater and Main Turbine High Water Level Trip Instrumentation;"

5.

LCO 3.3.4.1, "End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation;"

6.

LCO 3.3.4.2, "Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation;"

7.

LCO 3.3.5.1, "Emergency Core Cooling System (ECCS)

Instrumentation;"

8.

LCO 3.3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation;"

9.

LCO 3.3.5.3, "Reactor Core Isolation Cooling (RCIC) System Instrumentation;"

10.

LCO 3.3.6.1, "Primary Containment Isolation Instrumentation;"

11.

LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation;"

12.

LCO 3.3.6.3, "Low-Low Set (LLS) Instrumentation;"

13.

LCO 3.3.7.1, "[Main Control Room Environmental Control (MCREC)]

System Instrumentation;"

14.

LCO 3.3.8.1, "Loss of Power (LOP) Instrumentation;"

15.

LCO 3.3.8.2, "Reactor Protection System (RPS) Electric Power Monitoring."

b.

The program shall require the [Limiting Trip Setpoint (LTSP)], [Nominal Trip Setpoint (NTSP)], Allowable Value (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in paragraph a. are calculated using the NRC approved setpoint methodology, as listed below. In addition, the program shall contain the value of the

[LTSP], [NTSP], AV, AFT, and ALT (as applicable) for each Function described in paragraph a. and shall identify the setpoint methodology used to calculate these values.

Reviewer's Note ---------------------------------------

List the NRC safety evaluation report by letter, date, and ADAMS accession number (if available) that approved the setpoint methodologies.

12

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-17 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.15 Setpoint Control Program (continued)

1.

[Insert reference to NRC safety evaluation that approved the setpoint methodology.]

c.

The program shall establish methods to ensure that Functions described in paragraph a. will function as required by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology.

d.

REVIEWERS NOTE -------------------------------------

A license amendment request to implement a Setpoint Control Program must list the instrument functions to which the program requirements of paragraph d. will be applied. Paragraph d. shall apply to all Functions in the Reactor Protection System (RPS) Instrumentation, Control Rod Block Instrumentation, End of Cycle-Recirculation Pump Trip (EOC-RPT)

Instrumentation, Emergency Core Cooling System (ECCS) Instrumentation, and Reactor Core Isolation Cooling (RCIC) Instrumentation specifications unless one or more of the following exclusions apply:

1.

Manual actuation circuits, automatic actuation logic circuits or to instrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded.

2.

Settings associated with safety relief valves are excluded. The performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program.

3.

Functions and Surveillance Requirements which test only digital components are normally excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply.

The program shall identify the Functions described in paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The [LTSP] of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, trip unit calibrations and CHANNEL FUNCTIONAL TESTS that verify the [LTSP or NTSP].

12

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-18 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals 5.5.15 Setpoint Control Program (continued) 1 The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified [LTSP or NTSP].

2.

If the as-found value of the instrument channel trip setting differs from the previous as-left value or the specified [LTSP or NTSP] by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then the instrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program.

3.

If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

4.

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the [LTSP or NTSP] at the completion of the surveillance test; otherwise, the channel is inoperable (setpoints may be more conservative than the [LTSP or NTSP] provided that the as-found and as-left tolerances apply to the actual setpoint used to confirm channel performance).

e.

The program shall be specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference]. ]

[ 5.5.16 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. ]

13 6.8.4.j 12 12

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-19 Rev. 5.0 Hope Creek Amendment XXX CTS 2

5.5 Programs and Manuals

[ 5.5.17 Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

REVIEWER'S NOTE ----------------------------------

The Risk Informed Completion Time is only applicable in MODES supported by the licensee's PRA. Licensees applying the RICT Program to MODES other than MODES 1 and 2 must demonstrate that they have the capability to calculate a RICT in those MODES or that the risk indicated by their MODE 1 and 2 PRA model is bounding with respect to the lower MODE conditions.

b.

A RICT may only be utilized in MODE 1, 2 [, and 3, and MODE 4 while relying on steam generators for heat removal];

c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.
3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.

Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2.

Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

12

Programs and Manuals 5.5 General Electric BWR/4 STS 5.5-20 Rev. 5.0 CTS Hope Creek Amendment XXX 2

5.5 Programs and Manuals 5.5.17 Risk Informed Completion Time Program (continued)

e.

The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval. ]

[ 5.5.18 Spent Fuel Storage Rack Neutron Absorber Monitoring Program This Program provides controls for monitoring the condition of the neutron absorber used in the spent fuel pool storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The program shall be in accordance with NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision 0, May 2017 [, with the following exceptions:

1.

... ].]

12

JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 1 of 3

1. The ISTS contains bracketed information and/or values that are generic to General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.
2. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. ISTS 5.5.2.b requires integrated leak test requirements for each system at least once per 18 months and trended as part of the 18 month assessment of the control room envelope (CRE) boundary. ISTS 5.5.9.c requires total particulate sampling every 31 days. ISTS 5.5.14.d requires CRE pressure measurement testing relative to external adjacent areas to be performed at a frequency of 18 months on a STAGGERED TEST BASIS. Hope Creek Generating Station (HCGS) controls periodic Frequencies for Surveillances in accordance with the Surveillance Frequency Control Program (SFCP) per CTS 6.8.4.j. Therefore, ITS 5.5.2.b, 5.5.7.c, and 5.5.12.d will be performed at a Frequency in accordance with the SFCP. Additionally, the unnecessary 18 month descriptor of the CRE boundary assessment in ISTS 5.5.14.d is not included in ITS 5.5.12.d since the CRE boundary assessment is performed following performance of the CRE boundary test at a frequency in accordance with the SFCP.
4. ISTS 5.5.3, Post Accident Sampling, is not included in the ITS. The NRC Reviewers Note states, "This program may be eliminated based on the implementation of NEDO-32991, Revision 0, Regulatory Relaxation for BWR Post Accident Sampling Stations (PASS), and the associated NRC Safety Evaluation dated June 12, 2001, and implementation of the following commitments. This elimination was approved pursuant to License Amendment 149 (NRC ADAMS Accession No. ML031430416). Section 4.0 of the safety evaluation accompanying License Amendment 149 discusses required verifications and commitments.

Subsequent Specifications are renumbered, as necessary, to support this deviation.

5. The primary containment design does not include pre-stressed concrete and tendons, therefore ISTS 5.5.6, Pre-Stressed Concrete Containment Tendon Surveillance Program, is not included in the ITS. Subsequent Specifications are renumbered, as necessary, to support this deviation.
6. The Filtration Recirculation and Ventilation System (FRVS) is considered an Engineered Safety Feature (ESF) system. However, the Control Room Emergency Filtration (CREF) System is not considered an ESF system. Therefore, reference to ESF in ISTS 5.5.7 is not included in ITS 5.5.5.
7. ISTS 5.5.7, Ventilation Filter Testing Program (VFTP) includes a Reviewer's Note associated with the use of ASTM D3803-1989. The Reviewer's Note is not included in the ITS. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. The subject Reviewer's Note states that the use of ASTM D 3803-1989 is contingent upon the use of a safety factor of 2 in determining the allowable perpetration for charcoal filters. PSEG Nuclear, LLC, has included a safety factor of 2. This is documented in the NRC safety evaluation of the Extended Power Uprate (EPU) License Amendment 174 (NRC ADAMS

JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 2 of 3 Accession No. ML081230640), which lists the charcoal filter efficiencies as: FRVS ventilation units - 90%, FRVS recirculation units - not credited, and CREF subsystems - 99%. These values were also used in a previous analysis approved by the NRC in License Amendment 146, (NRC ADAMS Accession No. ML030760293).

The limits on penetration for the charcoal filters are: FRVS ventilation units - 5% and CREF subsystems - 0.5%. In addition, in the NRC safety evaluation associated with License Amendment 130 addressing NRC Generic Letter 99-02, "Laboratory Testing of Nuclear-Grade Activated Charcoal, (NRC ADAMS Accession No. ML003770220) the NRC stated that the FRVS ventilation mode, FRVS recirculation mode, and the CREF System have actual face velocities below 40 fpm, and therefore specifying testing of actual face velocities in Technical Specifications is not required.

8. The program details of the Explosive Gas and Storage Tank Radioactivity Monitoring Program are described in ISTS 5.5.8.a and 5.5.8.c (ITS 5.5.6.a and 5.5.6.b).

Therefore, the sentence in the introductory paragraph that specifies a method to determine the explosive gas and storage tank radioactivity is not necessary.

Additionally, the requirements specified in ISTS 5.5.8.b are not necessary because there are no radioactive gas storage tanks that feed into the Main Condenser Offgas Treatment System. The requirements of ISTS 5.5.8.c are revised in ITS 5.5.6.b to apply to outside temporary tanks consistent with the CTS.

9. HCGS uses Option B of 10 CFR 50 Appendix J, therefore the Option A and combined Option A and B provisions of ISTS 5.5.12 are not included in the ITS.
10. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. PSEG Nuclear, LLC, has reviewed the model application for plant-specific adoption of Technical Specification Task Force (TSTF) Traveler TSTF-500, (ADAMS Accession No. ML111751792) and will provide the requested Verifications and UFSAR update commitments prior to ITS Implementation as indicated in the Regulatory Commitments enclosure of the ITS Conversion license amendment request.
11. ISTS 5.5.13 is modified in ITS 5.5.11 to reference IEEE 450-2010, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," and Revision 3 of Regulatory Guide (RG) 1.129, "Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Nuclear Power Plants," instead of IEEE 450-2002 and Revision 2 of RG 1.129. RG 1.129, Revision 3 endorses the use of IEEE 450-2010 and eliminates the need for the exceptions specified in ISTS 5.5.13.a; therefore, the exceptions of ISTS 5.5.13.a are not included in ITS 5.5.11. Section 5.4.2 of IEEE 450-2010 states, in part, that specific gravity readings are not recommended to be taken on a regular basis. HCGS batteries are lead-calcium type batteries and therefore, specific gravities do not have to be obtained at each discharge test; therefore, ISTS 5.5.13.b.5 is not included in ITS 5.5.11. Use of IEEE 450-2010 and RG 1.129, Revision 3 in the Battery Monitoring and Maintenance Program has been previously approved in Donald C. Cook Nuclear Plant Amendments 343 and 325 dated February 5, 2019, for Units 1 and 2, respectively (NRC ADAMS Accession No. ML18346A358) and Turkey Point Nuclear Generating Units 3 and 4 Amendments 297 and 290, respectively, dated September 27, 2023 (NRC ADAMS Accession No.

JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 3 of 3 ML23234A192). Subsequent ISTS 5.5.13 (ITS 5.5.11) requirements are renumbered, as necessary, to support this deviation

12. HCGS does not require, and CTS do not include a Setpoint Control Program, Risk Informed Completion Time Program, or a Spent Fuel Storage Rack Neutron Absorber Monitoring Program. Therefore, these ISTS programs (ISTS 5.5.15, ISTS 5.5.17, and ISTS 5.5.18) are not included in the ITS. Subsequent Specifications are renumbered, as necessary, to support this deviation.
13. ISTS 5.5.4.k places limitations on venting and purging of the Mark II containment.

The HCGS containment is of the General Electric Mark I design consisting of a drywell and a torus. Specifying the specific type of containment design is unnecessary when describing the limitations on venting and purging and, therefore, is not included in ITS 5.5.3.k consistent with the CTS. Also, the statement in ISTS 5.5.12.b specifying the containment design pressure is not included in ITS 5.5.10 consistent with the CTS. The primary containment internal design pressure is specified in UFSAR Sections 3.8.2 and 6.2 and it unnecessary to duplicate this design requirement in the Technical Specifications.

14. ISTS 5.5.4.b is revised in ITS 5.5.3.b to reflect that Appendix B is to 10 CFR 20, not specific sections of 10 CFR 20, consistent with CTS and the regulation.
15. The reactor cyclic and transient occurrences are currently addressed in CTS Table 5.7.1-1. This table is proposed to be relocated to the UFSAR. Therefore, ISTS 5.5.5 is revised in ITS 5.5.4 to specify that the Component Cyclic or Transient Limit program provides controls to track reactor cyclic and transient occurrences specified in UFSAR Table 5.3-1 consistent with CTS 5.7.1, considering the relocation to the UFSAR.
16. As described in the NRC safety evaluation (SE) accompanying Operating License Amendment 121, dated September 8, 1999 (NRC ADAMS Accession No. ML20211N544 (Package) and ML20211N553 (SE)), PSEG has retained, in the current Technical Specifications, the level of effluent control pursuant to 10 CFR 20 requirements in effect prior to January 1, 1994 and, pursuant to 10 CFR 20.1008(c),

these requirements were determined to be equivalent to or more restrictive than the applicable requirements specified in 10 CFR 20.1001-20.2402. Therefore, ISTS 5.5.4.b, 5.5.4.c, and 5.5.4.g (ITS 5.5.3.b, 5.5.3.c, and 5.5.3.g) are revised, as necessary, in the ITS consistent with current Technical Specifications requirements referencing 10 CFR 20 regulations in effect prior to January 1, 1994.

17. Text revised, inserted, or deleted in ITS to correct a typographical or grammatical error.

S2

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 1 of 3 10 CFR 50.92 EVALUATION FOR REMOVE DETAIL CHANGE LA01 PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Revision 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the determination of No Significant Hazards Considerations for conversion to NUREG-1433.

CTS contains periodic testing frequencies in administrative program requirements for: leakage testing of primary coolant sources outside containment; total particulate concentration testing of the diesel generator fuel oil; and measurement, at designated locations, of the control room envelope (CRE) pressure relative to external areas adjacent to the CRE boundary during the pressurization mode of operation of the Control Room Emergency Filtration System. ITS specifies the periodic Frequency for these tests as "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified periodic Frequency for these tests to the Surveillance Frequency Control Program (SFCP).

The purpose of these CTS requirements is to assure that the necessary quality of systems and components is maintained. The removal of these details related to surveillance requirement frequencies from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The existing frequency is removed from Technical Specifications and placed under licensee control pursuant to the methodology described in Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1, dated April 2007 (ADAMS Accession No. ML071360456). The surveillance test requirements remain in the Technical Specifications. The control of changes to the test frequencies is in accordance with the SFCP.

The SFCP provides the necessary administrative controls to require that surveillances related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The proposed change to relocate periodic frequencies in the administrative controls section of Technical Specifications has been previously approved for Wolf Creek Generating Station Unit 1 in Amendment 227, dated April 8, 2021 (NRC ADAMS Accession No. ML21053A117), River Bend Station Unit 1 in Amendment 196, dated April 29, 2019 (NRC ADAMS Accession No. ML19066A008), and Grand Gulf Nuclear Station Unit 1 in Amendment 219, dated June 11, 2019 (NRC ADAMS Accession No. ML19094A799). HCGS adopted a SFCP in License Amendment 187 (NRC ADAMS Accession No. ML103410243). This change is acceptable because the testing frequencies will be adequately controlled in accordance with the SFCP requirements retained in ITS, which ensure changes are properly evaluated.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 2 of 3 PSEG has reviewed the proposed no significant hazards consideration (NSHC) determination published in Federal Register 74 FR 32000-32001 dated July 6, 2009. PSEG has concluded that the proposed NSHC presented in the Federal Register notice is applicable to the proposed relocation of the periodic testing frequencies specified herein.

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under the SFCP. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

No new or different accidents result from using the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS Hope Creek Page 3 of 3

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

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the updated final safety analysis report and bases to technical specifications),

since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.

To evaluate a change in the relocated surveillance frequency, PSEG will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Revision 1, in accordance with the Technical Specification SFCP. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications."

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above analysis, PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

ATTACHMENT 6 ITS 5.6, Reporting Requirements

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ITS ITS 5.6 A01 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 24% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 585 psig or core flow less than 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 585 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 With reactor steam dome pressure greater than 585 psig and core flow greater than 10%

of rated flow:

7KH0,1,080&5,7,&$/32:(55$7,2 0&35 VKDOOEH

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With reactor steam dome pressure greater than 585 psig and core flow greater than 10% of rated flow and the MCPR below the value for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

HOPE CREEK 2-1 Amendment No. 229 L01 L01 L01 See ITS 2.0 See ITS 2.0 See ITS 2.0

ITS ITS 5.6 A01 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5.

ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required.

Comply with the requirements of Specification 6.7.1.

HOPE CREEK 2-2 L01 See ITS 2.0

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a.

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

b.

Each REPORTABLE EVENT shall be reviewed by the Station Operations Review Comittee (SORC), and the results of this review shall be submitted to the Nuclear Review Board and the senior corporate nuclear officer.

6.7. SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

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 senior corporate nuclear officer and the senior management position with responsibility for independent nuclear safety assessment activities and quality program oversight shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

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 unit components, systems, or structures, and (3) corrective action taken to prevent recurrence.

c.

The Safety Limit Violation Report shall be submitted to the Commission, the senior management position with responsibility for independent nuclear safety assessment activities and quality program oversight and the senior corporate nuclear officer within 30 days of the violation.

d.

Critical operation of the unit shall not be resumed until authorized by the Commission.

HOPE CREEK 6-14 Amendment No. 97 A02 L01

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the USNRC Administrator, Region 1, unless otherwise noted.

STARTUP REPORT 6.9.1.1 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 in 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 unit.

6.9.1.2 The startup report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these 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.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days 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 operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS*

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

6.9.1.5 Reports required on an annual basis shall include:

a.

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

HOPE CREEK 6-17 Amendment No. 161 in accordance with 10 CFR 50.4 A03 5.6 L02 L03

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS

b.

Documentation of all challenges to main steamline safety/relief valves.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.6 The Annual Radiological Environmental Operating report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

HOPE CREEK 6-18 Amendment No. 161 5

A01 5.6.1 5.6.1 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.

M01 L03 by Offsite Dose Calculation Manual (ODCM)

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.7 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid wastes released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

HOPE CREEK 6-19 Amendment No. 121 5.6.2 in accordance with 10 CFR 50.36a A04 A01 prior to Process Control Program

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.8 Deleted CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the PSEG Nuclear LLC generated CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following Technical Specifications:

2.2 Reactor Protection System Instrumentation Setpoints 3/4.1.4.3 Rod Block Monitor 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.3 Minimum Critical Power Ratio*

3/4.2.4 Linear Heat Generation Rate 3/4.3.1 Reactor Protection System Instrumentation 3/4.3.6 Control Rod Block Instrumentation The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC as applicable in the following document:

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)"

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number title, revision, date, and any supplements).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The COLR, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.10

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Condition for Operation Section 3.4.6, "RCS Pressure/Temperature Limits"
2. Surveillance Requirement Section 4.4.6, "RCS Pressure/Temperature Limits"
  • The MCPR99.9% value, for both Two Recirculation Loop Operation and Single Recirculation Loop Operation, used to calculate the LCO 3.2.3, Minimum Critical Power Ratio, limit shall be specified in the COLR.

HOPE CREEK 6-20 Amendment No. 219 5.6.3 5.6.3.a prior to

, and shall be documented in the COLR A01 5.6.3.a LCO 3.3.2.1 LCO 3.2.1 LCO 3.2.2 LCO 3.2.3 LCO 3.3.1.1

, specifically those described A01 A01 A03 5.6.4 5.6.4 (APLHGR)

(MCPR)

Control Instrumentation (LHGR)

(RPS)

LCO 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RTP) Instrumentation LCO 3.4.1, Recirculation Loops Operating A05 A01 A01 A01 Specification 3.4.10, A01 5.6.3.b 5.6.3.b 5.6.3.c 5.6.3.d 5.6.3.a MCPR limits of LCO 3.2.2

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. BWROG-TP-11-022-A (SIR-05-044), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 1, dated August 2013.
c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the time period specified for each report.

6.9.3 When a report is required by Action 10 of Specification 3/4.3.1, "RPS Instrumentation, a report shall be submitted within 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety.

c.

All REPORTABLE EVENTS submitted to the Commission.

d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e.

Records of changes made to the procedures required by Specification 6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

h.

Records of annual physical inventory of all sealed source material of record.

HOPE CREEK 6-21 Amendment No. 209 A03 LA01 5.6.6 5.6.6 Core Stability Protection Report ACTION I 3.3.1.1 A01 A01 Reactor Protection System (RPS) core Oscillation Power Range Monitoring Instrumentation

ITS ITS 5.6 A01 ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 6.10.3 The following records shall be retained for the duration of the unit Operating License:

a.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.

b.

Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories.

c.

Records of radiation exposure for all individuals entering radiation control areas.

d.

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

e.

Records of transient or operational cycles for those unit components identified in Table 5.7.1-1.

f.

Records of reactor tests and experiments.

g.

Records of training and qualification for current members of the unit staff.

h.

Records of inservice inspections performed.

i.

Records of quality assurance activities required by the Quality Assurance Program.

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 SORC meetings and activities of the Nuclear Review Board (and activities of its predecessor, the Offsite Safety Review (OSR) staff).

l.

DELETED

m.

Records of analyses required by the radiological environmental monitoring program which would permit evaluation of the accuracy of the analyses at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

n.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATIONAL MANUAL and the PROCESS CONTROL PROGRAM.

HOPE CREEK 6-22 Amendment No. 185 LA01

ITS ITS 5.6 A01 Table 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS ACTION 80 -

a.

With the number of OPERABLE channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel to OPERABLE status within 30 days, or immediately initiate actions in accordance with 6.9.2.

b.

With the number of OPERABLE channels less than the Minimum Number of Channels shown in Table 3.3.7.5-1, (except for the Drywell Atmosphere Post Accident Radiation Monitor) restore at least one inoperable channel to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Deleted

d.

With the number of OPERABLE Drywell Atmosphere Post Accident Radiation Monitor channels less than the Minimum Number of Channels requirement shown in Table 3.3.7.5-1, initiate action in accordance with ACTION 81, below.

ACTION 81 -

With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

a.

Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and b.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 82 -

a.

With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, verify the valve(s) position by use of alternate indication methods. If the affected penetration is not isolated by either (i) a closed manual valve, (ii) a blind flange, or (iii) a deactivated automatic valve located outside primary containment, restore the inoperable channel(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

HOPE CREEK 3/4 3-86 Amendment No. 160 A01 5.6.5 When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM)

Instrumentation,"

Instrumentation channels of the Function 5.6.5 Post Accident Monitoring Report See ITS 3.3.3.1 See ITS 3.3.3.1

DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 1 of 4 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.6.1, Reportable Event Action, including CTS 6.6.1.a, specifies, in the case of a Reportable Event, that the Commission be notified and a report be submitted pursuant to the requirements of 10 CFR 50.73. CTS 6.6.1.b specifies that each Reportable Event shall be reviewed by the Station Operations Review Committee (SORC), and the results of this review shall be submitted to the Nuclear Review Board and the senior corporate nuclear officer. The requirements of CTS 6.6.1 are not included in the ITS. This changes the CTS by removing the requirements for Reportable Event Action.

The purpose of CTS 6.6.1 is to specify the appropriate federal regulations requiring a Reportable Event to be submitted and who at the facility shall review and receive the report. This change is acceptable because the requirements of CTS 6.6.1.a are contained in 10 CFR 50.72 and 10 CFR 50.73 and the review requirements are addressed in the plant Quality Assurance Topical Report and associated implementing procedures. Therefore, duplication of these requirements is unnecessary duplication in the Technical Specifications.

Because the HCGS Operating License requires compliance with 10 CFR 50, the change is designated as administrative and does not result in a technical change to the CTS.

A03 CTS 6.9.1 and 6.9.2 specify, in addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the reports shall be submitted to the USNRC Administrator, Region 1, unless otherwise noted. ITS 5.6 requires reports be submitted in accordance with 10 CFR 50.4. This changes the CTS by adding the reference to 10 CFR 50.4.

The purpose of CTS report requirements is to provide direction on where to send the listed reports. ITS 5.6 provides this direction by referencing 10 CFR 50.4.

10 CFR 50.4 provides distribution requirements for written communications to the NRC including reports. This change is acceptable because the 10 CFR 50.4 distribution requirements already apply to each report to be submitted to the NRC. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 6.9.1.7 requires submittal of an Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year to be submitted by May 1 of each year. ITS 5.6.2 similarly requires submittal of a Radioactive Effluent Release Report but includes submitting the report in accordance with 10 CFR 50.36a, Technical specifications on effluents from

DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 2 of 4 nuclear power reactors. This changes the CTS by providing the specific regulation the Radioactive Effluent Release Report must be in accordance with.

The purpose of CTS 6.9.1.7 is to provide requirements for the development and submittal of the Radioactive Effluent Release Report. By adding the regulation covering the submittal of this report does not change the CTS it only adds identification of the specific regulation that the report must be submitted in accordance with, which applies in CTS. This change is acceptable because the 10 CFR 50.36a requirements already apply to each report to be submitted to the NRC. This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 6.9.1.9 requires core operating limits be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle, for the listed technical specifications. ITS 5.6.3.a also requires core operating limits be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and documented in the COLR, for the listed technical specifications, however, the list is updated to provide a complete list of Specifications whose limits are specified in the COLR; including LCO 3.3.4.1, "End of Cycle Recirculation Pump Trip (EOC-RTP)

Instrumentation," and LCO 3.4.1, "Recirculation Loops Operating." This changes the CTS by including additional specifications whose parameters are to be specified in the COLR.

The purpose of the CTS requirement is to ensure technical specification core operating limits that are established on a cycle-by-cycle basis are specified in the COLR. LCO 3.3.4.1 refers to the COLR for MCPR limits when the associated EOC-RPT instrumentation is inoperable. LCO 3.4.1 refers to the COLR for MCPR and APLHGR limits when one recirculation loop is not in operation. This change ensures core operating limits are established and maintained in the COLR for Limiting Conditions for Operations specifying the COLR. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.9.1.6, Annual Radiological Environmental Operating report, does not include a second additional paragraph detailing what the report must include and stating the format of the information provided with an allowance for results not yet available. ITS 5.6.1 includes this paragraph. This changes the CTS by adding additional requirements for the Annual Radiological Environmental Operating report.

The purpose of CTS 6.9.1.6 is to provide requirements for the Annual Radiological Environmental Operating Report. The change is acceptable because the additional detail added will ensure uniform submittal of the report consistent with other licensees submittals making understanding of the information less challenging by the stakeholders. The change is designated as more restrictive because it reduces adds additional and specific reporting information to the Technical Specifications.

DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 3 of 4 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 6.10, includes detail on Record Retention in addition to the applicable record retention requirements of Title 10, Code of Federal Regulations. ITS does not include these details. This changes the CTS by relocated the Record Retention details to the UFSAR.

The requirement for retention of records related to activities affecting quality is contained in 10 CFR 50, Appendix B, Criterion XVII and other sections of 10 CFR 50 that are applicable. These record retention requirements provide a record of certain activities important to plant safety, but the records themselves do not assure safe operation of the facility since review of these records is a post-compliance review. Relocation of these CTS provisions to the UFSAR will provide adequate control over record retention requirements. The UFSAR will be revised to contain adequate detail with respect to these requirements to ensure recordkeeping is implemented in an appropriate manner. As such, the details to be relocated do not need to be duplicated in the ITS to provide adequate protection of the public health and safety. Changes to the UFSAR are controlled in accordance with the provisions of 10 CFR 50.59 to help ensure that appropriate reviews of any changes are performed. This change is designated as a less restrictive removal of detail change because details on record retention are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 8 - Deletion of Reporting Requirements) CTS 2.1.1, 2.1.2, 2.1.3 and 2.1.4 provide direction on complying with the requirements of CTS 6.7.1 when a safety limit is not met. CTS 6.7.1 provides details of reports and notifications to be made after a safety limit violation. ITS 5.6 does not include this detail direction, required reports, and notifications when a safety limit is violated. This changes the CTS by removing reporting and notification details when a safety limit is violated.

The purpose of CTS 2.1.1, 2.1.2, 2.1.3, 2.1.4 and 6.7.1 is to provide direction and details to ensure the NRC is informed when a safety limit is violated. This change is consistent with ISTS, which does not specify detailed reporting actions, and removes procedural reporting details necessary to comply with 10 CFR 50 requirements. 10 CFR 50.36 states, If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be

DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 4 of 4 resumed until authorized by the Commission. In addition, 10 CFR 50.72 requires notification of the NRC within a period stated based on the event identified and the urgency the NRC has determined for notification and states the necessary information. These regulations are adequate to ensure the NRC is informed when a safety limit is violated. Plant procedures provide reporting guidance to comply with 10 CFR 50.36 and 10 CFR 50.72 notification requirements when a safety limit is violated. This change is designated as less restrictive because reports and notifications that would be submitted under the CTS will not be required under the ITS.

L02 (Category 8 - Deletion of Reporting Requirements) CTS 6.9.1.1, CTS 6.9.1.2, and CTS 6.9.1.3 contain requirements for submitting a report of plant startup and power escalation testing following receipt of an operating license; amendments to the license involving planned increase in power level; installation of fuel that has a different design or has been manufactured by a different fuel supplier; and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The ITS does not contain such reporting requirements. This changes the CTS by deleting startup reporting requirements.

The purpose of the startup reporting requirements is to provide a summary report of plant startup and power escalation testing following the four specified conditions as verification that the unit operated as expected. This change is acceptable because the regulations provide adequate reporting requirements. If there were any unit conditions outside the expected parameters during unit startup, they would be reported to the NRC if they met the reporting requirements in the regulations. Otherwise, the reports would document that the unit operated as expected and already approved by the NRC, as required by regulations. This change is designated as less restrictive because reports that would be submitted under the CTS will not be required under the ITS.

L03 (Category 8 - Deletion of Reporting Requirements) CTS 6.9.1.4 and 6.9.1.5 require annual reports covering any instances when the main steamline safety/relief valves are challenged, to be submitted prior to March 1 each year.

ITS 5.6 does not contain any requirements for such a report. This changes the CTS by not including the requirements for the annual reporting of instances when the main steamline safety/relief valves are challenged.

The purpose of the CTS annual reporting requirements is to specify the requirements for certain operational activities. This change is acceptable because the regulations provide adequate details of reporting requirements of challenges to main steamline safety/relief valves. Operations or conditions challenging safe plant operation are required to be reported in accordance with 10 CFR 50.72 and 10 CFR 50.73. Subsequent reports would be provided, if necessary, without requiring a specific annual report. This change is designated as less restrictive because the reports that would be submitted under the CTS will not be required under the ITS.

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Reporting Requirements 5.6 General Electric BWR/4 STS 5.6-1 Rev. 5.0 Hope Creek Amendment XXX 2

5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Annual Radiological Environmental Operating Report

NOTE-------------------------------------------------

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

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.

5.6.2 Radiological Effluent Release Report

NOTE-------------------------------------------------

[ A single submittal may be made for a multiple unit station. The submittal shall 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. ]

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

6.9.1.6 6.9.1.7 1

1 3

DOC M01

Reporting Requirements 5.6 General Electric BWR/4 STS 5.6-2 Rev. 5.0 Hope Creek Amendment XXX 2

5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT

a.

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

[ The individual specifications that address core operating limits must be referenced here. The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR. ]

b.

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

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c.

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

d.

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

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a.

RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

INSERT 1 INSERT 2 Specification 3.4.10, RCS Pressure and Temperature (P/T) Limits.

4 3

2 4

3 6.9.1.9 6.9.1.10 MCPR limits of 6.9.1.9

  • footnote

Hope Creek Insert Page 5.6-2 INSERT 1 LCO 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),

LCO 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR),

LCO 3.2.3 LINEAR HEAT GENERATION RATE (LHGR),

LCO 3.3.1.1 Reactor Protection System (RPS) Instrumentation, LCO 3.3.2.1 Control Rod Block Instrumentation, LCO 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation, and LCO 3.4.1 Recirculation Loops Operating.

INSERT 2 NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II).

The COLR will contain the complete identification for the referenced topical report used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

3 3

4 ITS 5.6 R1 S2

Reporting Requirements 5.6 General Electric BWR/4 STS 5.6-3 Rev. 5.0 Hope Creek Amendment XXX 2

5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the PTLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the NRC staff approval document by date. ]

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

REVIEWER'S NOTE----------------------------------------

The methodology for the calculation of the P-T limits for NRC approval should include the following provisions:

1.

The methodology shall describe how the neutron fluence is calculated (reference new Regulatory Guide when issued).

2.

The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR 50. The reactor vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.

3.

Low Temperature Overpressure Protection (LTOP) System lift setting limits for the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR.

4.

The adjusted reference temperature (ART) for each reactor beltline material shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide 1.99, Revision 2.

5.

The limiting ART shall be incorporated into the calculation of the pressure and temperature limit curves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits.

BWROG-TP-11-022-A (SIR-05-044), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 1, dated August 2013.

5 5

3 6.9.1.10

General Electric BWR/4 STS 5.6-4 Rev. 5.0 Hope Creek Amendment XXX 2

5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

6.

The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

7.

Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2V') specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2V'), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.[3.1], "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

REVIEWER'S NOTE----------------------------------------

These reports may be required covering inspection, test, and maintenance activities. These reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Table 3.3.7.5-1 Action 81.b 5.6.6 Core Stability Protection Report When a report is required by ACTION I of Specification 3.3.1.1, "Reactor Protection System (RPS) Instrumentation, a report shall be submitted within 90 days. The report shall outline the preplanned means to provide backup core stability protection, the cause of the Oscillation Power Range Monitoring Instrumentation inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

5 1

6.9.3 3

2

JUSTIFICATION FOR DEVIATIONS ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 1 of 1

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. The Note is not applicable because Hope Creek Generating Station (HCGS) is a single unit station.
2. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The ISTS contains bracketed information and/or values that are generic to General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is inserted to reflect the current licensing basis.
4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. Consistent with the Reviewer's Note, HCGS has received prior NRC approval to relocate topical report revision numbers and dates to the COLR in License Amendment 154, dated October 20, 2004 (ADAMS Accession No. ML042040164). As such, a statement is added in the ITS, consistent with current Technical Specifications, that the COLR will contain the complete identification for each of the referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).
5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal. HCGS adopted a Pressure Temperature Limits Report (PTLR) in License Amendment 209, dated December 14, 2017 (ADAMS Accession No. ML17324A840) and included a complete identification of the topical report used to prepare the PTLR with report number, title, revision number, and date.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 7 ITS 5.7, High Radiation Area

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

ADMINISTRATIVE CONTROLS 6.12 HIGH 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:

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 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 HOPE CREEK 6-24 Amendment No.142 ITS A01 ITS 5.7 5.7 5.7.1

ADMINISTRATIVE CONTROLS (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.

6.12.2 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 shift supervisor, 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.

d.

Each individual or group entering such an area shall possess:

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 HOPE CREEK 6-24a Amendment No. 142 ITS A01 5.7.2 ITS 5.7

ADMINISTRATIVE CONTROLS

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.

HOPE CREEK 6-24b Amendment No. 142 ITS A01 ITS 5.7

DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA Hope Creek Page 1 of 1 ADMINISTRATIVE CHANGES A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None

Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

High Radiation Area 5.7 General Electric BWR/4 STS 5.7-1 Rev. 5.0 CTS Hope Creek Amendment XXX 1

5.0 ADMINISTRATIVE CONTROLS 5.7 High 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:

5.7.1 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 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 doses rates in the area; who is responsible for controlling personnel exposure within the area, or 6.12 6.12.1

High Radiation Area 5.7 General Electric BWR/4 STS 5.7-2 Rev. 5.0 CTS Hope Creek Amendment XXX 1

5.7 High Radiation Area 5.7.1 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 (continued)

(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.

5.7.2 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 shift supervisor, 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.

6.12.2 6.12.1

High Radiation Area 5.7 General Electric BWR/4 STS 5.7-3 Rev. 5.0 CTS Hope Creek Amendment XXX 1

5.7 High Radiation Area 5.7.2 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 (continued)

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 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 option (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.

6.12.2 1

High Radiation Area 5.7 General Electric BWR/4 STS 5.7-4 Rev. 5.0 CTS Hope Creek Amendment XXX 1

5.7 High Radiation Area 5.7.2 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 (continued)

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.

6.12.2

JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

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