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Enclosure 2: Hope Creek Generating Station Improved Technical Specifications Conversion - Volume 14
	ML24142A442
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Site:	Hope Creek
Issue date:	05/20/2024
From:	; Public Service Enterprise Group
To:	; Office of Nuclear Reactor Regulation
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ENCLOSURE 2 VOLUME 14 HOPE CREEK GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 0

LIST OF ATTACHMENTS

1.

ITS 3.9.1, Refueling Equipment Interlocks

2.

ITS 3.9.2, Refuel Position One-Rod-Out Interlock

3.

ITS 3.9.3, Control Rod Position

4.

ITS 3.9.4, Control Rod Position Indication

5.

ITS 3.9.5, Control Rod OPERABILITY - Refueling

6.

ITS 3.9.6, Reactor Pressure Vessel (RPV) Water Level

7.

ITS 3.9.7, Residual Heat Removal (RHR) - High Water Level

8.

ITS 3.9.8, Residual Heat Removal (RHR) - Low Water Level

9.

ISTS Not Adopted

ATTACHMENT 1 ITS 3.9.1, Refueling Equipment Interlocks

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

3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH LIMITING CONDITION FOR OPERATION 3.9.1 The reactor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position. When the reactor mode switch is locked in the Refuel position:

a.

A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.

b.

CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless at least the following associated Refuel position interlocks are OPERABLE for such equipment.

1.

All rods in.

2.

Refuel platform position.

3.

Refuel platform main hoist fuel-loaded.

4.

Service platform hoist fuel-loaded.

APPLICABILITY:

OPERATIONAL CONDITION 5* #

ACTION:

a.

With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position.

b.

With the one-rod-out interlock inoperable, lock the reactor mode switch in the Shutdown position.

c.

With any of the above required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock.

See Special Test Exceptions 3.10.1 and 3.10.3.

The reactor shall be maintained in OPERATIONAL CONDITION 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

HOPE CREEK 3/4 9-1 Amendment No. 31 See ITS 3.9.2 See ITS 3.9.2 See ITS 3.9.2 See ITS 3.9.2 See ITS 3.9.2 A01 ITS 3.9.1 ITS 3.9 3.9.1 Refueling Equipment Interlocks The refueling equipment shall be LCO 3.9.1 Applicability s

the SR 3.9.1.1.a SR 3.9.1.1.b SR 3.9.1.1.c ACTION A One or more ing (s) in-vessel fuel movement Immediately A02 A03 During in-vessel fuel movement with Add proposed Required Actions A.2.1 and A.2.2 L01

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.1.1 The reactor mode switch shall be verified to be locked in the Shutdown or Refuel position as specified:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to:

1.

Beginning CORE ALTERATIONS, and

2.

Resuming CORE ALTERATIONS when the reactor mode switch has been unlocked.

b.

In accordance with the Surveillance Frequency Control Program.

4.9.1.2 Each of the above required reactor mode switch Refuel position interlocks* shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and in accordance with the Surveillance Frequency Control Program during control rod withdrawal or CORE ALTERATIONS, as applicable.

4.9.1.3 Each of the above required reactor mode switch Refuel position interlocks* that is affected shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.

The reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that all control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

HOPE CREEK 3/4 9-2 Amendment No. 187 See ITS 3.9.2 refueling equipment following on SR 3.9.1.1 L03 L02 inputs A01 ITS 3.9.1 ITS See ITS 3.10.2

DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS 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 3.9.1 Applicability states OPERATIONAL CONDITION 5. Specific conditions for Applicability are provided in the CTS 3.9.1 LCO and Actions. CTS 3.9.1.b states, in part, CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless the Refuel position interlocks are OPERABLE. CTS 3.9.1 Action c states, in part, With any of the required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock. ITS 3.9.1 Applicability states During in-vessel fuel movement with equipment associated with the interlocks. This changes the CTS by clarifying the specific refueling condition of the Applicability when the refueling equipment interlocks must be OPERABLE.

The purpose of the CTS 3.9.1 applicability is to define the Conditions that apply when the refueling equipment interlocks must be OPERABLE. This change is consistent with the ISTS presentation and acceptable because the only core alterations that are performed using equipment associated with a refuel position interlock are control rod movement and fuel movement. The ITS 3.9.1 Applicability is consistent with CTS 3.9.1 Action c, which only requires suspension of operations with equipment associated with inoperable interlocks.

The interlock associated with control rod movement is addressed by ITS 3.9.2, Refuel Position One-Rod-Out Interlock. The remaining interlocks (i.e., those addressed by ITS 3.9.1) involve only in-vessel fuel movement. 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 3.9.1.b.4 requires, with the reactor mode switch locked in the refuel position, the refuel position equipment interlocks associated with the service platform hoist (fuel-loaded) to be OPERABLE when the equipment is being used for CORE ALTERATIONS. This changes the CTS by not including CTS 3.9.1.b.4 requirements in ITS 3.9.1 (ITS SR 3.9.1.1) since the refuel service platform is no longer incorporated in the HCGS design.

The current service platform is used for maintenance and in-vessel servicing activities in the reactor vessel. It is not used for core alterations. This change is designated as an administrative change and is acceptable because it does not result in a technical change to the CTS.

DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 2 of 4 MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.9.1 Action c requires suspension of CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock when required Refuel position equipment interlocks are inoperable. ITS 3.9.1 ACTION A provides similar action requirements for the same condition but includes optional actions when one or more required refueling equipment interlocks are inoperable. ITS 3.9.1 Required Actions A.2.1 and A.2.2 require immediately inserting a control rod withdraw block and verifying all control rods are fully inserted. This changes the CTS by adding optional actions to perform when one or more required refueling equipment interlocks are inoperable.

The purpose of CTS 3.9.1 Action is to suspend activities that could increase the probability of a core criticality event during refueling, when required refueling interlocks are inoperable. This change is consistent with the ISTS and acceptable because the optional Required Actions also include remedial measures to minimize risk associated with continuing the refueling activities while providing time to repair inoperable features. Required Action A.2.1 ensures no control rods can be withdrawn, because a block to control rod withdrawal is in place. The withdrawal block utilized must ensure that if rod withdrawal is requested, the rod will not respond (i.e., it will remain inserted). Required Action A.2.2 is performed after placing the rod withdrawal block in effect and provides a verification that all control rods are fully inserted. The optional requirements continue to provide remedial actions, in accordance with the requirements of 10 CFR 50.36(c)(2)(i),

to minimize the probability of a core criticality event during refueling. Therefore, it is acceptable to continue with the in-vessel fuel movements since the proposed alternative Required Actions will ensure all control rods are completely inserted and will remain in the inserted position even if control rod withdrawal is attempted. This change is designated as less restrictive because less stringent Required Actions are being applied In the ITS than were applied in the CTS.

L02 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.9.1.2 requires performance a CHANNEL FUNCTIONAL TEST of the Refuel Position interlocks for refueling equipment within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and in accordance with the Surveillance Frequency Control Program (SFCP). ITS SR 3.9.1.1 continues to require performance of the refueling equipment interlock CHANNEL FUNCTIONAL TEST in accordance with the SFCP. This changes the CTS by

DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 3 of 4 deleting the Frequency requirement, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of, allowing the Surveillance to be performed up to the frequency specified in the SFCP prior to the specified activities (as permitted by ITS SR 3.0.4).

The purpose of Surveillance Frequency is to ensure the LCO is met prior to and during activities that could increase the potential for a core criticality event during refueling. This change is acceptable because the normal periodic Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability and provides adequate assurance of OPERABILITY. As such, the requirement to perform the surveillance requirement within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of the specified activities is unnecessary and deleted. If the Surveillance has been performed within the normal specified interval, reliance on the results is allowed since ITS SR 3.0.4 (CTS 4.0.4) requires only that a Surveillance be performed within the required Frequency prior to entering the applicable MODE or specified condition. The Frequency specified in the SFCP provides adequate assurance that the LCO requirements are satisfied. If the Surveillance is not performed within the normal surveillance interval, in-vessel fuel movement with the equipment associated with the affected refueling interlock cannot commence since ITS SR 3.0.1 (CTS 4.0.1) requires a Surveillance be met within the specified Frequency while in the applicable MODE or specified condition. ITS SR 3.0.1 (CTS 4.0.1) also states that failure to meet the Surveillance constitutes failure to meet the LCO, which would then require the ACTIONS of the LCO to be taken. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3), to ensure refueling equipment interlocks are OPERABLE prior to and during in-vessel fuel movement with equipment associated with the interlocks. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L03 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.1.2 requires performance of a CHANNEL FUNCTIONAL TEST of the Refuel Position interlocks for refueling equipment prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.

ITS 3.9.1 does not include an additional post-maintenance Surveillance Requirement to perform the refueling equipment interlock CHANNEL FUNCTIONAL TEST. This changes the CTS by deleting a specific post-maintenance Surveillance Requirement.

The purpose of Surveillance Requirement is to ensure the reactor mode switch Refuel position interlock function is OPERABLE following repair, maintenance or replacement of any component that could affect Refueling Interlock OPERABILITY. This change is acceptable because the deleted Surveillance Requirement is not necessary to ensure that the reactor mode switch Refuel position interlock is OPERABLE. ITS SR 3.9.1.1 continues to require the CHANNEL FUNCTIONAL TEST be performed on a periodic Frequency specified in the Surveillance Frequency Control Program. Anytime the OPERABILITY of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1 to demonstrate the OPERABILITY

DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 4 of 4 of the affected components. In addition, the requirements of 10 CFR 50, Appendix B, Section XI (Test Control), provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria.

Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3), to ensure refueling equipment interlocks are OPERABLE prior to and during in-vessel fuel movement with equipment associated with the interlocks. This change is designated as less restrictive because Surveillance Requirements which are required in the CTS will not be required in the ITS.

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

Refueling Equipment Interlocks 3.9.1 General Electric BWR/4 STS 3.9.1-1 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LCO 3.9.1 The refueling equipment interlocks shall be OPERABLE.

APPLICABILITY:

During in-vessel fuel movement with equipment associated with the interlocks.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required refueling equipment interlocks inoperable.

A.1 Suspend in-vessel fuel movement with equipment associated with the inoperable interlock(s).

OR A.2.1 Insert a control rod withdrawal block.

AND A.2.2 Verify all control rods are fully inserted.

Immediately Immediately Immediately 3.9.1 3/4.9.1 3/4.9 LCO 3.9.1.b Action c DOC L01

Refueling Equipment Interlocks 3.9.1 General Electric BWR/4 STS 3.9.1-2 Rev. 5.0 Hope Creek Amendment XXX 1

CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:

a. All-rods-in,

b. Refuel platform position,

c.

Refuel platform [fuel grapple], fuel loaded,

[d. Refuel platform fuel grapple fully retracted position,]

[e. Refuel platform frame mounted hoist, fuel loaded,]

[f. Refuel platform monorail mounted hoist, fuel loaded,] and

[g. Service platform hoist, fuel loaded.]

[ 7 days OR In accordance with the Surveillance Frequency Control Program ]

main hoist 4.9.1.2 3.9.1.b.1 3.9.1.b.2 3.9.1.b.3 2

2 2

3 and 1

JUSTIFICATION FOR DEVIATIONS ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The service platform hoist fuel-loaded function is not included in ITS 3.9.1 since the service platform is no longer incorporated in the Hope Creek Generating Station design. The current service platform is used for maintenance and in-vessel servicing activities in the reactor vessel. It is not used for core alterations.

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

Refueling Equipment Interlocks B 3.9.1 General Electric BWR/4 STS B 3.9.1-1 Rev. 5.0 Hope Creek Revision XXX 1

B 3.9 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks BASES BACKGROUND Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures that prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.

GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). The control rods, when fully inserted, serve as the system capable of maintaining the reactor subcritical in cold conditions during all fuel movement activities and accidents.

One channel of instrumentation is provided to sense the position of the refueling platform, the loading of the refueling platform fuel grapple, and the full insertion of all control rods. Additionally, inputs are provided for the loading of the refueling platform frame mounted hoist, the loading of the refueling platform monorail mounted hoist, the full retraction of the fuel grapple, and the loading of the service platform hoist. With the reactor mode switch in the shutdown or refueling position, the indicated conditions are combined in logic circuits to determine if all restrictions on refueling equipment operations and control rod insertion are satisfied.

A control rod not at its full-in position interrupts power to the refueling equipment and prevents operating the equipment over the reactor core when loaded with a fuel assembly. Conversely, the refueling equipment located over the core and loaded with fuel inserts a control rod withdrawal block in the Control Rod Drive System to prevent withdrawing a control rod.

The refueling platform has two mechanical switches that open before the platform or any of its hoists are physically located over the reactor vessel.

All refueling hoists have switches that open when the hoists are loaded with fuel.

1 3

main hoist 1

Refueling Equipment Interlocks B 3.9.1 General Electric BWR/4 STS B 3.9.1-2 Rev. 5.0 Hope Creek Revision XXX 1

BASES BACKGROUND (continued)

The refueling interlocks use these indications to prevent operation of the refueling equipment with fuel loaded over the core whenever any control rod is withdrawn, or to prevent control rod withdrawal whenever fuel loaded refueling equipment is over the core (Ref. 2).

The hoist switches open at a load lighter than the weight of a single fuel assembly in water.

APPLICABLE The refueling interlocks are explicitly assumed in the FSAR analyses for SAFETY the control rod removal error during refueling (Ref. 3) and the fuel ANALYSES assembly insertion error during refueling (Ref. 4). These analyses evaluate the consequences of control rod withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.

Criticality and, therefore, subsequent prompt reactivity excursions are prevented during the insertion of fuel, provided all control rods are fully inserted during the fuel insertion. The refueling interlocks accomplish this by preventing loading of fuel into the core with any control rod withdrawn or by preventing withdrawal of a rod from the core during fuel loading.

The refueling platform location switches activate at a point outside of the reactor core such that, considering switch hysteresis and maximum platform momentum toward the core at the time of power loss with a fuel assembly loaded and a control rod withdrawn, the fuel is not over the core.

Refueling equipment interlocks satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO To prevent criticality during refueling, the refueling interlocks ensure that fuel assemblies are not loaded with any control rod withdrawn.

To prevent these conditions from developing, the all-rods-in, the refueling platform position, the refueling platform fuel grapple fuel loaded, the refueling platform trolley frame mounted hoist fuel loaded, the refueling platform monorail mounted hoist fuel loaded, the refueling platform fuel grapple fully retracted position, and the service platform hoist fuel loaded inputs are required to be OPERABLE. These inputs are combined in logic circuits, which provide refueling equipment or control rod blocks to prevent operations that could result in criticality during refueling operations.

and 1

U 1

main hoist

Refueling Equipment Interlocks B 3.9.1 General Electric BWR/4 STS B 3.9.1-3 Rev. 5.0 Hope Creek Revision XXX 1

BASES APPLICABILITY In MODE 5, a prompt reactivity excursion could cause fuel damage and subsequent release of radioactive material to the environment. The refueling equipment interlocks protect against prompt reactivity excursions during MODE 5. The interlocks are required to be OPERABLE during in-vessel fuel movement with refueling equipment associated with the interlocks.

In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and CORE ALTERATIONS are not possible. Therefore, the refueling interlocks are not required to be OPERABLE in these MODES.

ACTIONS A.1, A.2.1, and A.2.2 With one or more of the required refueling equipment interlocks inoperable, the unit must be placed in a condition in which the LCO does not apply. Therefore, Required Action A.1 requires that in-vessel fuel movement with the affected refueling equipment must be immediately suspended. This action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn).

Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position.

Alternatively, Required Actions A.2.1 and A.2.2 require a control rod withdrawal block to be inserted, and all control rods to be subsequently verified to be fully inserted. Required Action A.2.1 ensures no control rods can be withdrawn, because a block to control rod withdrawal is in place. The withdrawal block utilized must ensure that if rod withdrawal is requested, the rod will not respond (i.e., it will remain inserted). Required Action A.2.2 is performed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3.1.

Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn).

SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the 1

Refueling Equipment Interlocks B 3.9.1 General Electric BWR/4 STS B 3.9.1-4 Rev. 5.0 Hope Creek Revision XXX 1

BASES SURVEILLANCE REQUIREMENTS (continued) other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

[ The 7 day Frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section [7.6.1].

3.

FSAR, Section [15.1.13].

4.

FSAR, Section [15.1.14].

1 2

4 15.4.1 15.4.1.1.2.2 U

7.7.1.4 U

U 2

JUSTIFICATION FOR DEVIATIONS ITS 3.9.1 BASES, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The service platform hoist fuel-loaded function is not included in ITS 3.9.1 since the refuel service platform is no longer incorporated in the HCGS design. The current service platform is used for maintenance and in-vessel servicing activities in the reactor vessel. It is not used for core alterations.

4. The Reviewers 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.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 2 ITS 3.9.2, Refuel Position One-Rod-Out Interlock

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

3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH LIMITING CONDITION FOR OPERATION 3.9.1 The reactor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position. When the reactor mode switch is locked in the Refuel position:

a.

A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.

b.

CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless at least the following associated Refuel position interlocks are OPERABLE for such equipment.

1.

All rods in.

2.

Refuel platform position.

3.

Refuel platform main hoist fuel-loaded.

4.

Service platform hoist fuel-loaded.

APPLICABILITY:

OPERATIONAL CONDITION 5* #

ACTION:

a.

With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position.

b.

With the one-rod-out interlock inoperable, lock the reactor mode switch in the Shutdown position.

c.

With any of the above required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock.

See Special Test Exceptions 3.10.1 and 3.10.3.

The reactor shall be maintained in OPERATIONAL CONDITION 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

HOPE CREEK 3/4 9-1 Amendment No. 31 See ITS 3.9.1 See ITS 3.9.1 A01 ITS 3.9.2 ITS 3.9.2 Refuel Position One-Rod-Out Interlock SR 3.9.2.1 Verify LCO 3.9.2 shall be MODE 5 with and any control rod withdrawn.

Applicability Refuel position Condition A L02 L02 Condition A Refuel position one-rod-out interlock inoperable Add proposed Required Action A.1 and A.2 L02 A03 A04 A03 A02 L01

A01 ITS 3.9.2 ITS REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.1.1 The reactor mode switch shall be verified to be locked in the Shutdown or Refuel position as specified:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to:

1.

Beginning CORE ALTERATIONS, and

2.

Resuming CORE ALTERATIONS when the reactor mode switch has been unlocked.

b.

In accordance with the Surveillance Frequency Control Program.

4.9.1.2 Each of the above required reactor mode switch Refuel position interlocks* shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and in accordance with the Surveillance Frequency Control Program during control rod withdrawal or CORE ALTERATIONS, as applicable.

4.9.1.3 Each of the above required reactor mode switch Refuel position interlocks* that is affected shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.

The reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that all control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

HOPE CREEK 3/4 9-2 Amendment No. 187 Verify SR 3.9.2.1 L03 Add proposed SR 3.9.2.2 Note L05 See ITS 3.10.2 L04 L03 SR 3.9.2.2 L01

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 1 of 6 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 3.9.1 is divided into two separate requirements in ITS. CTS 3.9.1.a provides requirements associated with control rod withdrawal (i.e., requirements for the reactor mode switch to be in the refuel position and the refuel position equipment one-rod-out interlock to be OPERABLE). These requirements are addressed and rewritten in ITS 3.9.2. CTS 3.9.1.b places restrictions on CORE ALTERATIONS associated with fuel movement (the requirement for the refueling equipment interlocks to be OPERABLE). These requirements are addressed and rewritten in ITS 3.9.1, Refueling Equipment Interlocks, (see Discussion of Changes for ITS 3.9.1); where the Applicability addresses the only CORE ALTERATIONS remaining; in-vessel fuel movement. As a result, ITS 3.9.2 requires the refuel position one-rod-out interlock to be OPERABLE and ITS SR 3.9.2.1 requires the reactor mode switch to be locked in the Refuel position. This changes the presentation of the CTS to address the one-rod-out interlock rather than the reactor mode switch position.

The purpose of CTS 3.9.1.a requirement is to ensure the one-rod-out interlock is OPERABLE with any control rod withdrawn so no other control rods can be withdrawn in MODE 5 except as provided by the Special Test Exception Specifications. This change is acceptable because it represents a presentation difference and does not result in technical changes. The ITS presentation continues to ensure the refuel position one-rod-out interlock (which is accomplished by the reactor mode switch in the refuel position) is OPERABLE and the reactor mode switch is locked in the refuel position with any control rod withdrawn in MODE 5.

A03 CTS 3.9.1 Applicability states, OPERATIONAL CONDITION 5#. Footnote #

states The reactor shall be maintained in OPERATIONAL CONDITION 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. CTS 3.9.1.a states, A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE. ITS 3.9.2 Applicability states MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn. The information in the footnote is not included in ITS 3.9.2 because the information is redundant to that provided in the definition of a MODE and ITS Table 1.1-1. This clarifies the CTS Applicability by requiring the refuel position one-rod-out interlock to be OPERABLE when the reactor mode switch is in the refuel position and any control rod withdrawn and eliminates redundant information provided in other Technical Specifications.

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 2 of 6 The purpose of the CTS 3.9.1 applicability is to define the Conditions that apply to the limiting condition for operation (LCO). CTS 3.9.1 is divided into two separate requirements in ITS. ITS 3.9.2 addresses CTS 3.9.1.a which are associated with control rod withdrawal (i.e., requirements for the reactor mode switch to be in the refuel position and the refuel position equipment one-rod-out interlock to be OPERABLE). CTS 3.9.1 requires that the reactor mode switch be locked in the Shutdown or Refuel position when in MODE 5. ITS SR 3.9.2.1 also requires the mode switch to be locked in the refuel position and ITS LCO 3.9.2 requires the Refuel Position one-rod-out interlock be OPERABLE. The requirement that the reactor mode switch be in the shutdown or refuel position is an explicit part of the definition of OPERATIONAL CONDITION 5, as defined in CTS Table 1.2 and the definition of MODE 5, as defined in ITS Table 1.1-1.

Reactor mode switch OPERABILITY is an implicit part of the OPERABILITY of the interlocks required by ITS LCO 3.9.1 and LCO 3.9.2. In addition, CTS 3.9.1 Footnote # is addressed in the definition for a MODE in ITS Section 1.1, which states, in part, A MODE shall correspond to any one inclusive combination specified in Table 1.1-1 with fuel in the reactor vessel, and ITS Table 1.1-1 Footnote (b) modifies MODE 5 stating, One or more reactor vessel head closure bolts less than fully tensioned, which also encompasses the reactor vessel head removed. This change is consistent with the presentation in the ISTS and acceptable because by specifying OPERABLE interlocks and including the position of the reactor mode switch as part of the Applicability, appropriate features of the reactor mode switch are implicitly required to be OPERABLE without an explicit statement. Similarly, CTS 3.9.1 Action a, which establishes requirements with the reactor mode switch not locked in the shutdown or refuel position, becomes "Refuel position one-rod-out interlock inoperable" in ITS 3.9.2, Condition A. This change is designated as administrative because it represents a presentation preference and does not result in technical changes to the CTS.

A04 CTS 3.9.1 Applicability is modified by footnote

which states See Special Test Exceptions 3.10.1 and 3.10.3. ITS LCO 3.9.2 does not contain the footnote or a reference to the equivalent Special Operations Specification(s). This changes the CTS by not including cross references to Special Test Exception Specifications in the ITS.

The purpose of the footnote reference is to alert the user that a Special Test Exception exists that may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 3 of 6 REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.9.1, in part, requires the reactor mode switch to be OPERABLE and locked in the Shutdown or Refuel position while in Operational Condition 5. CTS 3.9.1 Actions include, in part, options of locking the mode switch in the Shutdown position. ITS SR 3.9.2.1 requires the reactor mode switch to be verified locked in the Refuel position. The presentation of ITS 3.9.2 does not address the shutdown position of the mode switch but rather addresses the refuel position one-rod-out interlock.

ITS LCO 3.9.2 is only applicable in MODE 5 with the reactor mode switch in the refuel position and any control rod is withdrawn. This changes the CTS by eliminating the requirement to lock the mode switch in the Shutdown position and eliminating the requirement to lock the mode switch in the Refuel position when all control rods are inserted.

The purpose of CTS requirement and associated actions is to ensure no more than one control rod may be withdrawn during refueling except as allowed by the special test exceptions. Since placing the reactor mode switch in the Shutdown position is an optional requirement, there is no need to explicitly require this feature to be OPERABLE. Reactor mode switch design precludes control rods from being withdrawn with the switch in the shutdown position by inserting a control rod block. Locking the mode switch in this position provides no safety purpose regarding the refuel position one-rod-out interlock. Placing the mode switch in the refuel position activates the one-rod-out interlock. Therefore, it is unnecessary to lock the mode switch in the refuel position unless a control rod is withdrawn. The ITS presentation between ITS 3.9.1, 3.9.2, and 3.9.3 ensure no more than one control rod can be withdrawn during refueling except as allowed by ITS Section 3.10, Special Operations, Specifications. This change is consistent with the ISTS and acceptable because ITS Section 3.9, Refueling Operations, specifications continue to ensure no more than one control rod may be withdrawn during refueling thus providing protection against potential reactivity excursions. This change is designated as less restrictive because less stringent requirements are being applied in the ITS than were applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS LCO 3.9.1, in part, requires the reactor mode switch to be OPERABLE and locked in the Shutdown or Refuel position. CTS 3.9.1 Action a requires suspending Core Alterations and locking the reactor mode switch in the Shutdown or Refuel position if the reactor mode switch is found not locked in the Shutdown or Refuel position. CTS 3.9.1 Action b requires locking the reactor mode switch in the Shutdown position when the one rod out interlock is inoperable. ITS 3.9.2 does not explicitly address the OPERABILITY of the mode switch but rather addresses the refuel position one-rod-out interlock. ITS SR 3.9.2.1 requires the reactor mode switch to be locked in the Refuel position. ITS 3.9.2 Required Actions A.1 and A.2 require suspending

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 4 of 6 control rod withdrawal and initiating action to insert all insertable control rods in core cells containing one or more fuel assemblies if the one-rod-out interlock is inoperable. This changes the CTS by providing actions to ensure all insertable control rods are inserted when the refuel position one-rod-out interlock is inoperable with the reactor mode switch in the Refuel position.

The purpose of the CTS actions is to ensure control rods cannot be withdrawn with the one-rod-out interlock inoperable during refueling to preclude a prompt reactivity excursion. The ITS presentation between ITS 3.9.1, 3.9.2, and 3.9.3 ensure no more than one control rod can be withdrawn during refueling except as allowed by ITS Section 3.10, Special Operations, Specifications. With the one-rod-out interlock inoperable (which would include discovering the reactor mode switch not locked in the Refuel position), ITS 3.9.2 ACTION A requires immediately suspending control rod withdrawal and initiating action to insert all insertable control rods in core cells containing one or more fuel assemblies. This change is consistent with the ISTS and acceptable because ITS 3.9.2 Required Actions A.1 and A.2 compensate for an inoperable one-rod-out interlock and provide adequate protection against potential reactivity excursions. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L03 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.9.1.1.a.1, CTS 4.9.1.1.a.2, and CTS 4.9.1.1.b require performance of the reactor mode switch Refuel position Surveillance within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to beginning CORE ALTERATIONS, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to resuming CORE ALTERATIONS when the reactor mode switch has been unlocked and in accordance with the Surveillance Frequency Control Program. ITS SR 3.9.2.1 requires performance of the reactor mode switch Refuel position Surveillance at the periodic frequency in accordance with the Surveillance Frequency Control Program. CTS 4.9.1.2 requires performance of the refuel position one-rod-out interlock Surveillance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and in accordance with the Surveillance Frequency Control Program. ITS SR 3.9.2.2 requires performance of the refuel position one-rod-out interlock Surveillance at the periodic frequency in accordance with the Surveillance Frequency Control Program. This changes the CTS by deleting the explicit Frequency requirements of prior to, allowing the Surveillance to be performed up to the frequency specified in the Surveillance Frequency Control Program (SFCP) prior to the specified activities (as permitted by ITS SR 3.0.4).

The purpose of the Surveillance Frequency is to ensure the LCO is met prior to and during activities that could increase the potential for a prompt reactivity excursion. This change is acceptable because the normal periodic Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability and provides adequate assurance of OPERABILITY. If the Surveillance has been performed within the normal specified interval, reliance on the results is allowed since ITS SR 3.0.4 (CTS 4.0.4) requires only that a Surveillance be performed within the required Frequency prior to entering the applicable MODE or specified condition. The Frequency specified in the SFCP provides adequate assurance that the LCO requirements are satisfied. If the Surveillance is not performed within the normal surveillance interval, the LCO is not met, and the Action must be entered. ITS SR 3.0.1 (CTS 4.0.1) requires a

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 5 of 6 Surveillance be met within the specified Frequency while in the applicable MODE or specified condition. ITS SR 3.0.1 (CTS 4.0.1) also states that failure to meet the Surveillance constitutes failure to meet the LCO, which would then require the ACTIONS of the LCO to be taken. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3),

to ensure the refuel position one-rod-out interlock is OPERABLE prior to and during refueling conditions with the reactor mode switch is in the Refuel position and a control rod is withdrawn. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L04 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.1.3 requires performance of a CHANNEL FUNCTIONAL TEST of the Refuel Position one-rod-out interlock prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the reactor mode switch Refuel position interlock. ITS 3.9.2 does not include an additional post-maintenance Surveillance Requirement to perform the Refuel position one-rod-out interlock CHANNEL FUNCTIONAL TEST. This changes the CTS by deleting a specific post-maintenance Surveillance Requirement.

The purpose of the Surveillance Requirement is to ensure the reactor mode switch Refuel position one-rod-out interlock function is OPERABLE following repair, maintenance or replacement of any component that could affect Refueling Interlock OPERABILITY. This change is acceptable because the deleted Surveillance Requirement is not necessary to ensure that the reactor mode switch Refuel position interlock is OPERABLE. ITS SR 3.9.2.2 continues to require the CHANNEL FUNCTIONAL TEST be performed on a periodic Frequency specified in the Surveillance Frequency Control Program. Anytime the OPERABILITY of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1 to demonstrate the OPERABILITY of the affected components. In addition, the requirements of 10 CFR 50, Appendix B, Section XI (Test Control), provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3), to ensure the refuel position one-rod-out interlock is OPERABLE prior to and during refueling conditions with the reactor mode switch is in the Refuel position and a control rod is withdrawn.

This change is designated as less restrictive because Surveillance Requirements which are required in the CTS will not be required in the ITS.

L05 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.9.1.2 requires a CHANNEL FUNCTIONAL TEST of the refuel position one-rod-out interlock in accordance with the Surveillance Frequency Control Program. ITS SR 3.9.2.2 also requires a CHANNEL FUNCTIONAL TEST of the refuel position one-rod-out interlock in accordance with the Surveillance Frequency Control Program.

DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 6 of 6 Additionally, ITS SR 3.9.2 2 includes a NOTE that states Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn. This changes the CTS by adding this Surveillance Note.

The purpose of the surveillance is to ensure the refuel position one-rod-out interlock is OPERABLE when a control rod is withdrawn during refueling and the reactor mode switch is in the refuel position. To perform the CHANNEL FUNCTIONAL TEST, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). The Note to ITS SR 3.9.2.2 is consistent with the ISTS and provides time to perform the Surveillance after any control rod is withdrawn. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is considered acceptable because of the procedural controls on control rod withdrawals and indications available in the control room to alert the operator of control rods not fully inserted.

This change is acceptable because the Surveillance Note allowance has been evaluated to ensure that it provides an acceptable level of equipment reliability.

ITS SR 3.9.2.2 continues to provide assurance that the necessary quality of the refuel position one-rod-out interlock will be maintained and the LCO will be met pursuant to the requirements of 10 CFR 50.36(c)(3). This change is designated as less restrictive because a Surveillance allowance is added to the CTS.

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

Refuel Position One-Rod-Out Interlock 3.9.2 General Electric BWR/4 STS 3.9.2-1 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.

APPLICABILITY:

MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out interlock inoperable.

A.1 Suspend control rod withdrawal.

AND A.2 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel position.

[ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR In accordance with the Surveillance Frequency Control Program ]

3.9.1.a 3/4.9 LCO 3.9.1 and Applicability Action a, b DOC L02 3.9.1 4.9.1.1 DOC L01 DOC L03 2

2

Refuel Position One-Rod-Out Interlock 3.9.2 General Electric BWR/4 STS 3.9.2-2 Rev. 5.0 Hope Creek Amendment XXX 1

CTS SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.9.2.2

NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn.

Perform CHANNEL FUNCTIONAL TEST.

[ 7 days OR In accordance with the Surveillance Frequency Control Program ]

2 2

4.9.1.2 DOC L03 DOC L05

JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

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

Refuel Position One-Rod-Out Interlock B 3.9.2 General Electric BWR/4 STS B 3.9.2-1 Rev. 5.0 Hope Creek Revision XXX 1

B 3.9 REFUELING OPERATIONS B 3.9.2 Refuel Position One-Rod-Out Interlock BASES BACKGROUND The refuel position one-rod-out interlock restricts the movement of control rods to reinforce unit procedures that prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod is permitted to be withdrawn.

GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

The refuel position one-rod-out interlock prevents the selection of a second control rod for movement when any other control rod is not fully inserted (Ref. 2). It is a logic circuit that has redundant channels. It uses the all-rods-in signal (from the control rod full-in position indicators discussed in LCO 3.9.4, "Control Rod Position Indication") and a rod selection signal (from the Reactor Manual Control System).

This Specification ensures that the performance of the refuel position one-rod-out interlock in the event of a Design Basis Accident meets the assumptions used in the safety analysis of Reference 3.

APPLICABLE The refueling position one-rod-out interlock is explicitly assumed in the SAFETY FSAR analysis for the control rod withdrawal error during refueling ANALYSES (Ref. 3). This analysis evaluates the consequences of control rod withdrawal during refueling. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.

The refuel position one-rod-out interlock and adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)" prevent criticality by preventing withdrawal of more than one control rod. With one control rod withdrawn, the core will remain subcritical, thereby preventing any prompt critical excursion.

The refuel position one-rod-out interlock satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO To prevent criticality during MODE 5, the refuel position one-rod-out interlock ensures no more than one control rod may be withdrawn. Both channels of the refuel position one-rod-out interlock are required to be OPERABLE and the reactor mode switch must be locked in the refuel position to support the OPERABILITY of these channels.

U 1

Refuel Position One-Rod-Out Interlock B 3.9.2 General Electric BWR/4 STS B 3.9.2-2 Rev. 5.0 Hope Creek Revision XXX 1

BASES APPLICABILITY In MODE 5, with the reactor mode switch in the refuel position, the OPERABLE refuel position one-rod-out interlock provides protection against prompt reactivity excursions.

In MODES 1, 2, 3, and 4, the refuel position one-rod-out interlock is not required to be OPERABLE and is bypassed. In MODES 1 and 2, the Reactor Protection System (LCO 3.3.1.1) and the control rods (LCO 3.1.3) provide mitigation of potential reactivity excursions. In MODES 3 and 4, with the reactor mode switch in the shutdown position, a control rod block (LCO 3.3.2.1) ensures all control rods are inserted, thereby preventing criticality during shutdown conditions.

ACTIONS A.1 and A.2 With one or both channels of the refueling position one-rod-out interlock inoperable, the refueling interlocks may not be capable of preventing more than one control rod from being withdrawn. This condition may lead to criticality.

Control rod withdrawal must be immediately suspended, and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.

SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required.

By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.

[ The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

2

Refuel Position One-Rod-Out Interlock B 3.9.2 General Electric BWR/4 STS B 3.9.2-3 Rev. 5.0 Hope Creek Revision XXX 1

BASES SURVEILLANCE REQUIREMENTS (continued)

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.

This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. [ The 7 day Frequency is considered avdequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not fully inserted.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn.

3 2

3

Refuel Position One-Rod-Out Interlock B 3.9.2 General Electric BWR/4 STS B 3.9.2-4 Rev. 5.0 Hope Creek Revision XXX 1

BASES REFERENCES

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section [7.6.1.1].

3.

FSAR, Section [15.4.1.1].

15.4.1 U

7.7.1.4 U

1 2

JUSTIFICATION FOR DEVIATIONS ITS 3.9.2 BASES, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The Reviewers 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.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 3 ITS 3.9.3, Control Rod Position

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

REFUELING OPERATIONS 3/4.9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be inserted.*

APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.**

ACTION:

With all control rods not inserted, suspend all other CORE ALTERATIONS, except that one control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod-out interlock.

SURVEILLANCE REQUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as above specified:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to:

1.

The start of CORE ALTERATIONS.

2.

The withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out interlock.

b.

In accordance with the Surveillance Frequency Control Program.

Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.

See Special Test Exception 3.10.3.

HOPE CREEK 3/4 9-5 Amendment No. 187 3.9.3 LCO 3.9.3 Applicability When loading fuel assemblies into the core.

A01 ITS 3.9.3 ITS fully fully immediately A03 loading fuel assemblies into the core.

One or more ACTION A SR 3.9.3.1 Verify are fully L01 L02 A02 A02 L01 3.9

REFUELING OPERATIONS 3/4.9.4 DELETED HOPE CREEK 3/4 9-6 Amendment No. 137 A01 ITS 3.9.3

REFUELING OPERATIONS 3/4.9.5 DELETED Hope Creek 3/4 9-7 Amendment No. 137 A01 ITS 3.9.3

REFUELING OPERATIONS 3/4.9.6 DELETED Hope Creek 3/4 9-8 Amendment No. 137 A01 ITS 3.9.3

THIS PAGE INTENTIONALLY DELETED.

HOPE CREEK 3/4 9-9 Amendment No. 31 A01 ITS 3.9.3

REFUELING OPERATIONS 3/4.9.7 DELETED HOPE CREEK 3/4 9-10 Amendment No. 137 A01 ITS 3.9.3

DISCUSSION OF CHANGES 3.9.3, CONTROL ROD POSITION 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 3.9.3 Footnote

states Except control rods removed per Specification 3.9.10.1 or 3.9.10.2 and is a cross-reference to specifications 3.9.10.1, Single Control Rod Removal, and 3.9.10.2, Multiple Control Rod Removal. CTS 3.9.3 Applicability Footnote ** states See Special Test Exception 3.10.3 and is a cross-reference to special test exception 3.10.3, Shutdown Margin Demonstrations. ITS 3.9.3 does not include these cross-references. This changes the CTS by removing cross-references to other specifications.

The purpose of the CTS 3.9.3 footnotes is to alert the user to other specifications that may apply to the specification in use. The ITS presentation does not include cross-references to other specifications. This change is acceptable because the deleted cross-refences are not required to meet the requirements of ITS 3.9.3.

This change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS.

A03 CTS 3.9.3 Action states, in part, With all control rods not inserted, suspend all other CORE ALTERATIONS. ITS 3.9.3 ACTIONS require, with one or more control rods not fully inserted, to immediately suspend loading fuel assemblies into the core. Refer to Discussion of Change L01 for technical change to the action statement. This changes the CTS by explicitly requiring the action to be performed immediately when the LCO is not met.

The purpose of the CTS action is to suspend activities that could result in adverse core reactivity changes when control rods are not fully inserted during refueling operations. This change clarifies the intent of the current action to immediately suspend the stated activities and is acceptable because it does not result in any technical change. All fuel loading operations will continue to be immediately suspended when one or more control rods are withdrawn, except as provided by ITS Section 3.10, Special Operations, Specifications. 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 3.9.3, CONTROL ROD POSITION Hope Creek Page 2 of 3 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS LCO 3.9.3 is applicable in Operational Condition 5, during CORE ALTERATIONS, as modified by footnote

- . Consistent with this CTS Applicability, CTS 3.9.3 Action states With all control rods not inserted, suspend all other core alterations, except that one control rod may be withdrawn under control of the reactor mode switch refuel position one-rod-out interlock. ITS LCO 3.9.3 is applicable when loading fuel assemblies into the core. Consistent with this ITS Applicability, when one or more control rods are not fully inserted, ITS 3.9.3 ACTION A requires immediate suspension of loading fuel assemblies into the core. This changes the CTS by relaxing the Applicability from during all core alterations to only when loading fuel assemblies into the core and changes the corresponding CTS actions to immediately exit the applicability.

The purpose of the current applicability of the Specification is to ensure control rods are fully inserted during refueling conditions that could result in an inadvertent criticality. The intent of the change in the Applicability, and associated Required Action A.1, is to establish the requirement that all control rods are inserted consistent with the assumption in the fuel assembly insertion error analysis. Immediate suspension of fuel loading operations shall not preclude completion of movement of a component to a safe position.

This change is consistent with the ISTS and acceptable because the ITS will continue to ensure conditions are maintained during core alterations to minimize the probability of an inadvertent criticality. The core alterations covered by the CTS 3.9.3 Applicability include: (1) fuel loading, (2) fuel unloading, and (3) control rod movement while fuel is in the associated cell. ITS LCO 3.9.3, along with ITS 3.9.2, Refuel Position One-Rod-Out Interlock, ITS 3.9.4, Control Rod Position Indication, and ITS 3.9.5, Control Rod OPERABILITY - Refueling, preserve the assumptions in the fuel assembly insertion error and control rod withdrawal error during refueling analyses to minimize the probability of an inadvertent criticality. ITS LCO 3.9.3 covers fuel loading during core alterations and ITS LCOs 3.9.2, 3.9.4, and 3.9.5 cover control rod movement during core alterations. ITS Section 3.9, Refueling Operations, does not specifically address position of the control rods while unloading fuel. Eliminating the requirement that all control rods are inserted while unloading fuel is not safety significant because fuel unloading increases overall core SHUTDOWN MARGIN. Special Operations Specifications, ITS 3.10.5, Single Control Rod Drive (CRD) Removal -

Refueling, and ITS 3.10.6, Multiple Control Rod Withdrawal - Refueling, provide additional requirements to minimize the probability of localized criticality

DISCUSSION OF CHANGES 3.9.3, CONTROL ROD POSITION Hope Creek Page 3 of 3 when one or more control rods are withdrawn in MODE 5. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS.

L02 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.9.3 requires verification that control rods are inserted "within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to" the start of core alterations or the withdrawal of one control rod under the one rod out interlock, and at a frequency in accordance with the Surveillance Frequency Control Program (SFCP). ITS SR 3.9.3.1 maintains the requirement to perform the verification in accordance with the SFCP but eliminates the requirement that verifications be performed "within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to" the specified activities. This changes the CTS by allowing the verification to be performed up to the frequency specified in the SFCP prior to the specified activities (as permitted by ITS SR 3.0.4).

The purpose of the surveillance frequencies is to ensure the LCO requirement is met prior to and during performance of the specified activities. This change is acceptable because the requirement that all control rods remain inserted throughout the applicable conditions is unchanged. Elimination of the requirement to perform this verification within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the activity is not significant because the normal periodic Surveillance Frequency is established in the SFCP to provide adequate assurance that requirements are being met. If the Surveillance has been performed within the normal specified interval, reliance on the results is allowed since ITS SR 3.0.4 (CTS 4.0.4) requires only that a Surveillance be performed within the required Frequency prior to entering the applicable MODE or specified condition. The normal Frequency specified in the SFCP provides adequate assurance that the LCO requirements are satisfied. If any Surveillance has not been performed within this interval, control rod withdrawal and fuel loading operations must be immediately suspended. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3), to ensure the control rods are fully inserted prior to and during loading of fuel assemblies into the core. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

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

Control Rod Position 3.9.3 General Electric BWR/4 STS 3.9.3-1 Rev. 5.0 Hope Creek Amendment XXX 1

3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.

APPLICABILITY:

When loading fuel assemblies into the core.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods not fully inserted.

A.1 Suspend loading fuel assemblies into the core.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted.

[ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR In accordance with the Surveillance Frequency Control Program ]

CTS 3.9.3 Applicability Action DOC L01 4.9.3 DOC L02 3./4.9.3 2

2

JUSTIFICATION FOR DEVIATIONS 3.9.3, CONTROL ROD POSITION Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

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

Control Rod Position B 3.9.3 General Electric BWR/4 STS B 3.9.3-1 Rev. 5.0 Hope Creek Revision XXX 1

B 3.9 REFUELING OPERATIONS B 3.9.3 Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the Control Rod Drive System.

During refueling, movement of control rods is limited by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) or the control rod block with the reactor mode switch in the shutdown position (LCO 3.3.2.1).

GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

The refueling interlocks allow a single control rod to be withdrawn at any time unless fuel is being loaded into the core. To preclude loading fuel assemblies into the core with a control rod withdrawn, all control rods must be fully inserted. This prevents the reactor from achieving criticality during refueling operations.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during refueling SAFETY are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the ANALYSES SDM (LCO 3.1.1), the intermediate range monitor neutron flux scram (LCO 3.3.1.1), the average power range monitor neutron flux scram (LCO 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).

The safety analysis for the control rod withdrawal error during refueling in the FSAR (Ref. 2) assumes the functioning of the refueling interlocks and adequate SDM. The analysis for the fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. Thus, prior to fuel reload, all control rods must be fully inserted to minimize the probability of an inadvertent criticality.

Control rod position satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO All control rods must be fully inserted during applicable refueling conditions to minimize the probability of an inadvertent criticality during refueling.

U 1

Control Rod Position B 3.9.3 General Electric BWR/4 STS B 3.9.3-2 Rev. 5.0 Hope Creek Revision XXX 1

BASES APPLICABILITY During MODE 5, loading fuel into core cells with control rods withdrawn may result in inadvertent criticality. Therefore, the control rods must be inserted before loading fuel into a core cell. All control rods must be inserted before loading fuel to ensure that a fuel loading error does not result in loading fuel into a core cell with the control rod withdrawn.

In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, this Specification is not applicable in these MODES.

ACTIONS A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the FSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.

[ The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency takes into consideration the procedural controls on control rod movement during refueling as well as the redundant functions of the refueling interlocks.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section [15.1.13].

3.

FSAR, Section [15.1.14].

2 3

2 1

U 15.4.1.1.2.2 U

15.4.1.1.2.3 U

1

JUSTIFICATION FOR DEVIATIONS 3.9.3 BASES, CONTROL ROD POSITION Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The Reviewers 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.

Determination of No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 3.9.3, CONTROL ROD POSITION Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 4 ITS 3.9.4, Control Rod Position Indication

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

REACTIVITY CONTROL SYSTEMS CONTROL ROD POSITION INDICATION LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod position indication system shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1 or 2 with one or more control rod position indicators inoperable, within 1 hour:

1.

Determine the position of the control rod by using an alternative method, or:

a)

Moving the control rod, by single notch movement, to a position with an OPERABLE position indicator, b)

Returning the control rod, by single notch movement, to its original position, and c)

Verifying no control rod drift alarm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or

2.

Move the control rod to a position with an OPERABLE position indicator, or

3.

When THERMAL POWER is:

a)

Within the preset power level of the RWM, declare the control rod inoperable.

b)

Greater than the preset power level of the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves** either:

1)

Electrically, or

2)

Hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable, move the control rod to a position with an OPERABLE position indicator or insert the control rod.

At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

HOPE CREEK 3/4 1-13 Amendment No. 180 3.9.4 Control Rod Position Indication 3.9 REFUELING OPERATIONS full-in channel for each control rod LCO 3.9.4 See ITS 3.1.3 See ITS 3.1.3 See ITS 3.1.3 MODE One or more required indication channels Initiate action to fully insert associated with the inoperable Required Action A.2.1 Add proposed Required Action A.1.1, A.1.2, A.1.3 M01 A01 ITS 3.9.4 ITS M01 A03 Applicability Add proposed ACTIONS NOTE A04 A02 Add proposed Required Action A.2.2 M01 L01

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:

a.

In accordance with the Surveillance Frequency Control Program that the position of each control rod is indicated,

b.

That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and

c.

That the control rod position indicator corresponds to the control rod position indicated by the "Full Out" position indicator when performing Surveillance Requirement 4.1.3.6.b.

HOPE CREEK 3/4 1-14 Amendment No. 187 See ITS 3.1.3 SR 3.9.4.1 full-in channel for each control rod LCO 3.9.4 Verify required channel has no "full-in" indication on that is not "full-in."

Each time the control rod is withdrawn from the "full-in" position A01 ITS 3.9.4 ITS L01

DISCUSSION OF CHANGES ITS 3.9.4, CONTROL ROD POSITION INDICATION 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 3.1.3.7 Applicability states, in part, OPERATIONAL CONDITION 5. ITS 3.9.4 Applicability states MODE 5. This changes the CTS by incorporating the ITS MODE definition.

The purpose of CTS 3.1.3.7 Applicability is to establish the Operational Condition (i.e., ITS MODE) in which the LCO is required. This change is acceptable because the Applicability of MODE is not changed. 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 3.1.3.7 Applicability is modified by footnote

which states Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. ITS LCO 3.9.4 does not contain the footnote or a reference to the equivalent Special Operations Specification(s). This changes the CTS by not including cross references to other Specifications in the ITS.

The purpose of the footnote reference is to alert the user that other Specifications exist that may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS.

A04 CTS 3.1.3.7 Action b describes Actions to be taken when in OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable. ITS 3.9.4 also describes actions to be taken with one or more required control rod position indication channels inoperable and contains a Note that separate condition entry is allowed for each required channel. This changes the CTS by adding a Note stating that separate condition entry is allowed for each required channel.

The purpose of the CTS Actions is to provide the appropriate compensatory actions with a control rod position indicator inoperable. This proposed change will allow separate condition entry for each inoperable position indication channel.

The Note clarifies that control rod position indication channels are treated as separate entities, each with separate Completion Times. This change is acceptable because it meets the current requirement. The CTS considers each control rod position indication channel to be separate and independent from each other. This change is designated as administrative because it does not result in technical changes to the CTS.

DISCUSSION OF CHANGES ITS 3.9.4, CONTROL ROD POSITION INDICATION Hope Creek Page 2 of 3 MORE RESTRICTIVE CHANGES M01 CTS 3.1.3.7 Applicability, Note

and Action b require position indication be OPERABLE for each withdrawn control rod. ITS 3.9.4 requires a full-in position indication channel be OPERABLE for each control rod, whether the control rod is inserted or withdrawn. This changes the CTS by expanding the Applicability, actions, and surveillances to include full-in position indication channels of each control rod, regardless of whether the control rod is withdrawn.

The purpose of the requirements for full-in position indication channel to be OPERABLE for each control rod when in MODE 5 is to ensure the refueling interlocks and the one-rod-out interlock can preclude more than one control rod being withdrawn while in refueling conditions. Although CTS 3.1.3.7 appears to require OPERABILITY of control rod full-in position indication, CTS 3.1.3.7 Actions do not compensate for an inoperable full-in position indication channel.

It only requires the position of the control rod to be known or the control rod to be inserted. This is corrected by ITS 3.9.4 ACTIONS for inoperable position indication which require that in vessel fuel movement and control rod withdrawal be suspended (Required Actions A.1.1 and A.1.2) and the control rod, if insertable and in a core cell containing one or more fuel assemblies, be fully inserted (Required Action A.1.3), or alternatively, that the control rod with the inoperable position indicator be fully inserted and associated control rod drive disarmed (Required Actions A.2.1 and A.2.2). Required Actions A.1.1 and A.1.2 prevent additional core reactivity changes while actions are being taken to insert all insertable control rods with an inoperable position channel in core cells containing one or more fuel assemblies (Required Action A.1.3). The alternative Required Actions require immediate initiation of action to insert and disarm the control rod associated with the inoperable position channel. These Required Actions ensure that a control rod with an inoperable position channel does not allow violation of the one-rod-out interlock. Finally, the proposed Completion Time has been added to specify that the Required Action be completed "immediately." CTS 3.1.3.7 Action b does not clearly specify a time period to start or complete the Action. This change is acceptable because it provides additional assurance that necessary actions will be performed when one or more required control rod full-in position indication channels are inoperable. This change is designated as more restrictive because it adds specific action requirements with a specific completion time to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None

DISCUSSION OF CHANGES ITS 3.9.4, CONTROL ROD POSITION INDICATION Hope Creek Page 3 of 3 LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.1.3.7 LCO and Action b, and CTS 4.1.3.7 address the control rod position indication system capability to indicate the current position of any withdrawn control rod. ITS 3.9.4 requires only full-in position indication channels to be OPERABLE. This changes the CTS LCO, Actions, and the associated surveillances to require only the full-in position indication channels to be OPERABLE.

The purpose of the CTS requirements is to ensure the position indication of withdrawn control rods are OPERABLE. ITS 3.9.4 deletes the general position indication system requirement and adds a specific requirement for the full-in position indication channel to be OPERABLE for each control rod, regardless of the actual position of the control rod. This added restriction details requirements consistent with the intent of requiring refueling interlocks and the one-rod-out interlock to be OPERABLE. ITS LCO 3.9.4 and LCO 3.9.5 for MODE 5 do not require the specific position of a withdrawn control rod to be indicated. The ITS 3.9.4 requirement only requires that a withdrawn control rod not indicate full-in. Since only one control rod can be withdrawn while in MODE 5 (exceptions to this are addressed in ITS Section 3.10, Special Operations, Specifications),

and the position of the control rod is not a consideration in any accident or transient when in this condition, the precise position of the control rod is insignificant. The critical safety issue, whether the control rod is fully inserted or not to support the refueling interlocks and one-rod-out interlock, is addressed by the ITS 3.9.4 requirement. Additionally, the Surveillance Requirements and ACTIONS are modified to be consistent with the requirement that only the full-in indicator must be OPERABLE. The current requirements to verify the position of each control rod is indicated on a periodic Frequency (CTS 4.1.3.7.a), that the control rod position changes during exercise tests (CTS 4.1.3.7.b), and that the full out indicator functions during control rod coupling checks (CTS 4.1.3.7.c) are deleted because these tests do not support Operability of the full-in position indication channel. ITS SR 3.9.4.1 requires that each time a control rod is withdrawn from the full-in position, the full-in indication is indicating correctly (i.e., it is not indicating full-in when a control rod is withdrawn). Note that failure to indicate full-in when the control rod is not withdrawn results in conservative actuation of the one-rod-out interlock, and therefore, is not explicitly required to be verified by this SR. The full-in position indication channel is considered inoperable even with the control rod fully inserted, if it would continue to indicate full-in with the control rod withdrawn. Performing the SR each time a control rod is withdrawn is considered adequate because of the procedural controls on control rod withdrawals and the visual and audible indications available in the control room to alert the operator to control rods not fully inserted. Therefore, the current requirement to verify the position of the control rod at the specified Frequency is deleted. These less restrictive changes are consistent with the ISTS and are acceptable because they have no impact on safety. 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)

Control Rod Position Indication 3.9.4 General Electric BWR/4 STS 3.9.4-1 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indication LCO 3.9.4 The control rod "full-in" position indication channel for each control rod shall be OPERABLE.

APPLICABILITY:

MODE 5.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each required channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required control rod position indication channels inoperable.

A.1.1 Suspend in vessel fuel movement.

AND A.1.2 Suspend control rod withdrawal.

AND A.1.3 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

OR A.2.1 Initiate action to fully insert the control rod associated with the inoperable position indicator.

AND Immediately Immediately Immediately Immediately 3.1.3.7 DOC L01 Applicability DOC A04 DOC M01 DOC M01 DOC M01 3.1.3.7 Action b DOC M01

Control Rod Position Indication 3.9.4 General Electric BWR/4 STS 3.9.4-2 Rev. 5.0 Hope Creek Amendment XXX 1

CTS ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2.2 Initiate action to disarm the control rod drive associated with the fully inserted control rod.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify the required channel has no "full-in" indication on each control rod that is not "full-in."

Each time the control rod is withdrawn from the "full-in" position DOC M01 4.1.3.7.a DOC L01

JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, CONTROL ROD POSITION INDICATION Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

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

Control Rod Position Indication B 3.9.4 General Electric BWR/4 STS B 3.9.4-1 Rev. 5.0 Hope Creek Revision XXX 1

B 3.9 REFUELING OPERATIONS B 3.9.4 Control Rod Position Indication BASES BACKGROUND The full-in position indication channel for each control rod provides necessary information to the refueling interlocks to prevent inadvertent criticalities during refueling operations. During refueling, the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) use the full-in position indication channel to limit the operation of the refueling equipment and the movement of the control rods. The absence of the full-in position channel signal for any control rod removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this condition causes the refuel position one-rod-out interlock to not allow the withdrawal of any other control rod.

GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during refueling SAFETY are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the ANALYSES SDM (LCO 3.1.1), the intermediate range monitor neutron flux scram (LCO 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).

The safety analysis for the control rod withdrawal error during refueling (Ref. 2) assumes the functioning of the refueling interlocks and adequate SDM. The analysis for the fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. The full-in position indication channel is required to be OPERABLE so that the refueling interlocks can ensure that fuel cannot be loaded with any control rod withdrawn and that no more than one control rod can be withdrawn at a time.

Control rod position indication satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Each control rod full-in position indication channel must be OPERABLE to provide the required input to the refueling interlocks. A channel is OPERABLE if it provides correct position indication to the refueling interlock logic.

Control Rod Position Indication B 3.9.4 General Electric BWR/4 STS B 3.9.4-2 Rev. 5.0 Hope Creek Revision XXX 1

BASES APPLICABILITY During MODE 5, the control rods must have OPERABLE full-in position indication channels to ensure the applicable refueling interlocks will be OPERABLE.

In MODES 1 and 2, requirements for control rod position are specified in LCO 3.1.3, "Control Rod OPERABILITY." In MODES 3 and 4, with the reactor mode switch in the shutdown position, a control rod block (LCO 3.3.2.1) ensures all control rods are inserted, thereby preventing criticality during shutdown conditions.

ACTIONS A Note has been provided to modify the ACTIONS related to control rod position indication channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable control rod position indication channels provide appropriate compensatory measures for separate inoperable channels. As such, this Note has been provided, which allows separate Condition entry for each inoperable required control rod position indication channel.

A.1.1, A.1.2, A.1.3, A.2.1 and A.2.2 With one or more required full-in position indication channels inoperable, compensating actions must be taken to protect against potential reactivity excursions from fuel assembly insertions or control rod withdrawals. This may be accomplished by immediately suspending in-vessel fuel movement and control rod withdrawal, and immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Actions must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

Suspension of in-vessel fuel movements and control rod withdrawal shall not preclude moving a component to a safe position.

Alternatively, actions must be immediately initiated to fully insert the control rod(s) associated with the inoperable full-in position indicator(s) and disarm the drive(s) to ensure that the control rod is not withdrawn.

Actions must continue until all associated control rods are fully inserted and drives are disarmed. Under these conditions (control rod fully inserted and disarmed), an inoperable full-in channel may be bypassed to allow refueling operations to proceed. An alternate method must be used to ensure the control rod is fully inserted (e.g., use the "00" notch position indication).

Control Rod Position Indication B 3.9.4 General Electric BWR/4 STS B 3.9.4-3 Rev. 5.0 Hope Creek Revision XXX 1

BASES SURVEILLANCE SR 3.9.4.1 REQUIREMENTS The full-in position indication channels provide input to the one-rod-out interlock and other refueling interlocks that require an all-rods-in permissive. The interlocks are actuated when the full-in position indication for any control rod is not present, since this indicates that all rods are not fully inserted. Therefore, testing of the full-in position indication channels is performed to ensure that when a control rod is withdrawn, the full-in position indication is not present. Note that failure to indicate full-in when the control rod is not withdrawn results in conservative actuation of the one-rod-out interlock, and therefore, is not explicitly required to be verified by this SR. The full-in position indication channel is considered inoperable even with the control rod fully inserted, if it would continue to indicate full-in with the control rod withdrawn.

Performing the SR each time a control rod is withdrawn is considered adequate because of the procedural controls on control rod withdrawals and the visual and audible indications available in the control room to alert the operator to control rods not fully inserted.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section [15.1.13].

3.

FSAR, Section [15.1.14].

15.4.1 U

7.7.1.4 U

2 2

JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, CONTROL ROD POSITION INDICATION Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.4, CONTROL ROD POSITION INDICATION Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 5 ITS 3.9.5, Control Rod OPERABILITY - Refueling

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

REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 Each control rod scram accumulator shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

NOTE-----------------------------------------------------

Separate condition entry is allowed for each control rod

a.

In OPERATIONAL CONDITIONS 1 or 2:

1.

With one control rod scram accumulator inoperable and reactor pressure 900 psig, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , a)

Restore the inoperable accumulator to OPERABLE status, or b)

Declare the associated control rod scram time slow***, or c)

Insert the associated control rod, declare the associated control rod inoperable and disarm the associated control valves by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With two or more control rod scram accumulators inoperable and reactor pressure 900 psig, a)

Within 20 minutes of discovery of this condition concurrent with charging water pressure < 940 psig, restore charging water header pressure to 940 psig otherwise place the mode switch in the shutdown position**,

and b)

Within one hour, declare the associated control rod scram time slow***,

or c)

Within one hour insert the associated control rods, declare the associated control rods inoperable and disarm the associated control valves by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At least the accumulator associated with each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

Only applicable if the associated control rod scram time was within the limits of Table 3.1.3.3-1 during the last scram time Surveillance. Rods that are already considered slow should be declared inoperable and fully inserted.

HOPE CREEK 3/4 1-9 Amendment No. 193 3.9.5 Control Rod OPERABILITY - Refueling 3.9 REFUELING OPERATIONS A01 ITS 3.9.5 ITS LCO 3.9.5 Applicability See ITS 3.1.5 See ITS 3.1.5 MODE A02 A03 withdrawn A02 See ITS 3.1.5 A02

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

3.

With one or more control rod scram accumulators inoperable and reactor pressure < 900 psig, a)

Immediately upon discovery of charging water header pressure

< 940 psig, verify all control rods associated with inoperable accumulators are fully inserted otherwise place the mode switch in the shutdown position**, and b)

Within one hour insert the associated control rod(s), declare the associated control rod(s) inoperable and disarm the associated control valves either electrically or hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5*:

1.

With one or more withdrawn control rods inoperable, upon discovery immediately initiate action to fully insert inoperable withdrawn control rods.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by verifying that the indicated pressure is greater than or equal to 940 psig unless the control rod is inserted and disarmed or scrammed.

At least the accumulator associated with each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

HOPE CREEK 3/4 1-10 Amendment No. 187 See ITS 3.1.5 See ITS 3.1.5 ACTION A SR 3.9.5.2 withdrawn LCO 3.9.5 each withdrawn control rod scram accumulator A02 A03 Add proposed SR 3.9.5.1 and NOTE M01 A02 A01 ITS 3.9.5 ITS

REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a.

In OPERATIONAL CONDITION 1 and 2 with one control rod not coupled to its associated drive mechanism, within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

1.

If permitted by the RWM, insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod, and:

a)

Observing any indicated response of the nuclear instrumentation, and b)

Demonstrating that the control rod will not go to the overtravel position.

2.

If recoupling is not accomplished on the first attempt or, if not permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves** either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1.

Insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or

2.

If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves** either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

HOPE CREEK 3/4 1-11 Amendment No. 180 L01 ITS 3.9.5 See ITS 3.1.3 See ITS 3.1.3 L01 See ITS 3.1.5 L01

DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING 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 3.1.3.5 requires each control rod scram accumulator to be OPERABLE and the Applicability stats, in part, OPERATIONAL CONDITION 5*. CTS 3.1.3.5 Footnote

states in part, At least the accumulator associated with each withdrawn control rod. The presentation of ITS 3.9.5 includes the requirements of CTS 3.1.3.5 requiring withdrawn control rods to be OPERABLE. The OPERABILITY aspects of the control rods (i.e., OPERABILITY of the control rod scram accumulators) are presented in the associated Surveillances and Bases.

Therefore, ITS 3.9.5 is presented with an Applicability that states, MODE 5 and ITS LCO 3.9.5 states Each withdrawn control rod shall be OPERABLE.

Additionally, CTS 4.1.3.5 requires verifying the OPERABILITY of each control rod scram accumulator by verifying pressure unless the control rod is inserted and disarmed or scrammed. ITS SR 3.9.5.2 provides similar requirements and applies to withdrawn control rods. This changes the CTS presentation by specifying that only withdrawn control rods must be OPERABLE in MODE 5 consistent with the intent of the CTS requirements.

Footnote

of CTS 3.1.3.5 Applicability clarifies that the Specification applies to withdrawn control rods and CTS 3.1.3.5 Action b also applies to withdrawn control rods that are inoperable. Therefore, the intent of CTS limiting condition for operation (LCO) 3.1.3.5 is that the control rod accumulator requirements apply to withdrawn control rods. This is a presentation change consistent with the ISTS and is acceptable because the requirements for the scram accumulators to be OPERABLE are retained by requiring withdrawn control rods to be OPERABLE.

Likewise, ITS SR 3.9.5.1 and SR 3.9.5.2 are required for withdrawn control rods only, consistent with the Applicability, eliminating the need to except control rods that are inserted and disarmed or scrammed. 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 3.1.3.5 Applicability is modified by footnote

which states Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. ITS LCO 3.9.5 does not contain the footnote or a reference to the equivalent Specification(s). This changes the CTS by not including cross references to other Specifications in the ITS.

The purpose of the footnote reference is to alert the user that another Specification exists that may modify the Applicability of the Specification. It is an ITS convention to not include these types of footnotes or cross-references. This

DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING Hope Creek Page 2 of 3 change is designated as administrative as it incorporates an ITS convention with no technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 4.1.3.5 does not contain an explicit requirement to verify that each withdrawn control rod is OPERABLE. ITS SR 3.9.5.1 requires inserting each withdrawn control rod at least one notch. This added requirement ensures that the withdrawn control rod can be inserted. This SR is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4. HCGS controls periodic Frequencies for Surveillances in accordance with the Surveillance Frequency Control Program per CTS 6.8.4.j (ITS 5.5.13).

Therefore, ITS SR 3.9.5.1 will be performed at a Frequency in accordance with the Surveillance Frequency Control Program with an initial Frequency of 7 days consistent with the Frequency established for ISTS SR 3.9.5.1. This changes the CTS by adding a Surveillance Requirement to the ITS.

The purpose of ITS SR 3.9.5.1 is to ensure each withdrawn control rod is not stuck and is free to automatically insert. This change is consistent with the ISTS and necessary because it supports LCO 3.9.5, which requires that each withdrawn control rod be OPERABLE. This change is designated as more restrictive because it adds an SR to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.1.3.6 requires control rods be coupled to their drive mechanisms in OPERATIONAL CONDITION 5 and includes applicable ACTIONS. ITS LCO 3.9.5 requires each withdrawn control rod to be OPERABLE and does not include requirements for control rods to be coupled to their associated control rod drive mechanisms (CRDMs) while in MODE 5. This changes the CTS by eliminating the requirement for control rods to be coupled to their associated CRDMs while in refueling conditions.

The purpose of the CTS requirements is to ensure a withdrawn control rod can insert during refueling conditions and when inoperable, provide actions to ensure the withdrawn control rod is placed in a safe condition. Control rod coupling requirements, as described in CTS 3.1.3.6, are not necessary during refueling

DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING Hope Creek Page 3 of 3 since withdrawal of more than one control rod from a core cell containing fuel assemblies is precluded by refueling interlocks and administrative controls.

Therefore, the probability and consequence of a control rod drop accident (CRDA) are negligible (i.e., reactor will remain subcritical and within the limits of the CRDA assumptions) in MODE 5. Additionally, control rod coupling is not necessary to ensure the control rod can perform its intended safety function of fully inserting. As such, control rod coupling is not required in MODE 5 and the MODE 5 requirements of CTS 3.1.3.6 are deleted. Additionally, control rod coupling requirements are provided in ITS 3.10.8, SHUTDOWN MARGIN (SDM)

Test - Refueling, which allows the option to perform a SDM demonstration in MODE 5 with multiple control rods withdrawn. This change is consistent with the ISTS and acceptable because the ITS 3.9.5 requirements continue to ensure a withdrawn control rod can insert during refueling conditions and when inoperable, appropriate actions are retained to ensure the withdrawn control rod is placed in a safe condition. This change is designated as less restrictive because less stringent requirements are being applied in the ITS than applied in the CTS.

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

Control Rod OPERABILITY - Refueling 3.9.5 General Electric BWR/4 STS

3. 9.5-1 Rev. 5.0 Hope Creek Amendment XXX 1

CTS 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY - Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.

APPLICABILITY:

MODE 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn control rods inoperable.

A.1 Initiate action to fully insert inoperable withdrawn control rods.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1

NOTE------------------------------

Not required to be performed until 7 days after the control rod is withdrawn.

Insert each withdrawn control rod at least one notch.

[ 7 days OR In accordance with the Surveillance Frequency Control Program ]

3.1.3.5 Action b.1 DOC M01 3.1.3.5 Applicability 2

2

Control Rod OPERABILITY - Refueling 3.9.5 General Electric BWR/4 STS

3. 9.5-2 Rev. 5.0 Hope Creek Amendment XXX 1

CTS SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.9.5.2 Verify each withdrawn control rod scram accumulator pressure is [940] psig.

[ 7 days OR In accordance with the Surveillance Frequency Control Program ]

4.1.3.5.a 2

2

JUSTIFICATION FOR DEVIATIONS ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

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

Control Rod OPERABILITY - Refueling B 3.9.5 General Electric BWR/4 STS B 3.9.5-1 Rev. 5.0 Hope Creek Revision XXX 1

B 3.9 REFUELING OPERATIONS B 3.9.5 Control Rod OPERABILITY - Refueling BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD) System, the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes during refueling operation. In addition, the control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. 1). The CRD System is the system capable of maintaining the reactor subcritical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during refueling SAFETY are provided by refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the ANALYSES SDM (LCO 3.1.1), the intermediate range monitor neutron flux scram (LCO 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).

The safety analyses for the control rod withdrawal error during refueling (Ref. 2) and the fuel assembly insertion error (Ref. 3) evaluate the consequences of control rod withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Control rod scram provides protection should a prompt reactivity excursion occur.

Control rod OPERABILITY during refueling satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Each withdrawn control rod must be OPERABLE. The withdrawn control rod is considered OPERABLE if the scram accumulator pressure is

[940] psig and the control rod is capable of being automatically inserted upon receipt of a scram signal. Inserted control rods have already completed their reactivity control function, and therefore are not required to be OPERABLE.

APPLICABILITY During MODE 5, withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and provide the required negative reactivity to maintain the reactor subcritical.

2

Control Rod OPERABILITY - Refueling B 3.9.5 General Electric BWR/4 STS B 3.9.5-2 Rev. 5.0 Hope Creek Revision XXX 1

BASES APPLICABILITY (continued)

For MODES 1 and 2, control rod requirements are found in LCO 3.1.2, "Reactivity Anomalies," LCO 3.1.3, "Control Rod OPERABILITY,"

LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators." During MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions.

ACTIONS A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod(s). Inserting the control rod(s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod(s) is fully inserted.

SURVEILLANCE SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is [940] psig.

[ The 7 day Frequency takes into consideration equipment reliability, procedural controls over the scram accumulators, and control room alarms and indicating lights that indicate low accumulator charge pressures.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

2 3

2

Control Rod OPERABILITY - Refueling B 3.9.5 General Electric BWR/4 STS B 3.9.5-3 Rev. 5.0 Hope Creek Revision XXX 1

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 26.

2.

FSAR, Section [15.1.13].

3.

FSAR, Section [15.1.14].

2 2

15.4.1 U

7.7.1.4 U

1 1

JUSTIFICATION FOR DEVIATIONS ITS 3.9.5 BASES, CONTROL ROD OPERABILITY - REFUELING Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The Reviewers 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.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 6 ITS 3.9.6, Reactor Pressure Vessel (RPV) Water Level

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

REFUELING OPERATIONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.8 At least 22 feet 2 inches of water shall be maintained over the top of the reactor pressure vessel flange.

APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.8 The reactor vessel water level shall be determined to be at least at its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program during handling of fuel assemblies or control rods within the reactor pressure vessel.

HOPE CREEK 3/4 9-11 Amendment No. 193 3.9 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 Applicability ACTION A SR 3.9.6.1 RPV level RPV movement irradiated fuel assemblies within the RPV, During movement of new handling of

RPV, RPV are A02 RPV water level not within limit.

movement and handling of RPV A01 ITS 3.9.6 ITS Verify RPV water level is 22 ft 2 in above the top of the RPV flange.

L01 A01 above LA01

DISCUSSION OF CHANGES ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL 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 3.9.8 Applicability states During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated. ITS 3.9.6 is applicable during movement of irradiated fuel assemblies within the RPV, and during movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV. The ITS Applicability does not include MODE 5. This changes the CTS by deleting OPERATIONAL CONDITION 5 and clarifies the applicability includes both movement of irradiated fuel assemblies, and movement of new fuel assemblies when irradiated fuel assemblies are seated in the RPV.

The purpose of CTS 3.9.8 Applicability is to establish the Operational Condition (i.e., ITS MODE) and other conditions in which the LCO is required.

OPERATIONAL CONDITION 5 is deleted since stating MODE 5 in the Applicability is unnecessary. The only MODE where it is possible to handle fuel assemblies or control rods in the core is MODE 5. In MODES 1, 2, 3 and 4, the reactor vessel head is on, and no fuel movement or control rod handling activities over irradiated fuel assemblies seated in the RPV are possible. This change also more precisely clarifies that the Specification applies during movement of irradiated fuel assemblies, regardless of whether fuel assemblies are seated within the RPV; and the Specification also applies during movement of new fuel assemblies only when there are irradiated fuel assemblies seated within the RPV. This change is consistent with the ISTS and is acceptable because it is unnecessary to state "OPERATIONAL CONDITION 5" (ITS MODE 5). This change is designated as an administrative change because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None

DISCUSSION OF CHANGES ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Hope Creek Page 2 of 3 REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.9.8 Action requires, with the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.

ITS 3.9.6 Action also requires immediate suspension of the specified activities when the LCO requirement is not met but does not include the operational detail after placing all fuel assemblies and control rods in a safe condition. This changes the CTS by relocating operational detail to the ITS Bases.

The removal of the operational detail for performing an action 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 3.9.6 ACTION A retains the action to immediately suspend movement of fuel assemblies and handling of control rods within the RPV when the RPV water level is not within the required limit.

As stated in ITS Section 1.3, when "Immediately" is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner. In this application, the ITS Bases clarifies that suspension of fuel movement and control rod handling shall not preclude completion of movement of a component (i.e., fuel assembly or control rod) to a safe position. Also, this change is acceptable because this type of procedural detail will be adequately controlled in the ITS Bases. 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.

LESS RESTRICTIVE CHANGES L01 (Category 7 - Relaxation of Surveillance Frequency) CTS 4.9.8 requires verification that RPV water level is at least at its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program during handling of fuel assemblies or control rods within the reactor pressure vessel. ITS SR 3.9.6.1 maintains the requirement to perform the verification in accordance with the Surveillance Frequency Control Program (SFCP) but eliminates the requirement that verifications be performed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of the specified activities. This changes the CTS by allowing the verification to be performed up to the frequency specified in the SFCP prior to the specified activities (as permitted by ITS SR 3.0.4).

DISCUSSION OF CHANGES ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Hope Creek Page 3 of 3 The purpose of the surveillance frequencies is to ensure the LCO requirement is met prior to and during performance of the specified activities. This change is acceptable because the requirement that RPV water level is 22 ft 2 in above the top of the RPV flange throughout the applicable conditions is unchanged.

Elimination of the requirement to perform this verification within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the activity is not significant because the normal periodic Surveillance Frequency is established in the SFCP to provide adequate assurance that requirements are being met. If the Surveillance has been performed within the normal specified interval, reliance on the results is allowed since ITS SR 3.0.4 (CTS 4.0.4) requires only that a Surveillance be performed within the required Frequency prior to entering the applicable MODE or specified condition. The Frequency specified in the SFCP provides adequate assurance that the LCO requirements are satisfied. If any Surveillance has not been performed within this interval, handling of fuel assemblies or control rods within the reactor pressure vessel must be immediately suspended. This change is consistent with the ISTS and considered adequate, pursuant to the requirements of 10 CFR 50.36(c)(3), to ensure RPV water level is acceptable prior to and during movement of irradiated fuel assemblies or; during movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

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

[RPV] Water Level -[Irradiated Fuel]

3.9.6 General Electric BWR/4 STS 3.9.6-1 Rev. 5.0 Hope Creek Amendment XXX 1

2 3.9 REFUELING OPERATIONS 3.9.6

[Reactor Pressure Vessel (RPV)] Water Level - [Irradiated Fuel]

LCO 3.9.6

[RPV] water level shall be [23] ft above the top of the [RPV flange].

APPLICABILITY:

During movement of irradiated fuel assemblies within the [RPV],

[ During movement of new fuel assemblies or handling of control rods within the [RPV], when irradiated fuel assemblies are seated within the [RPV]. ]

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [RPV] water level not within limit.

A.1 Suspend movement of fuel assemblies [and handling of control rods] within the

[RPV].

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify [RPV] water level is [23] ft above the top of the [RPV flange].

[ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR In accordance with the Surveillance Frequency Control Program ]

CTS 2

22 ft 2 in 22 ft 2 in 2

2 2

2 3.9.8 3/4.9.8 Applicability Action 4.9.8 DOC L01

JUSTIFICATION FOR DEVIATIONS ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis. The Hope Creek Generating Station technical specifications limit RPV water level to 22 feet 2 inches above the top of the RPV flange during handling of fuel assemblies or control rods in the RPV when the fuel assemblies being handled are irradiated, or the fuel assemblies in the reactor vessel are irradiated. Since the RPV water level limiting condition for operation is the same for during movement of irradiated fuel assemblies, and during movement of new fuel assemblies or handling control rods within the RPV when irradiated fuel assemblies are seated in the RPV, the ISTS 3.9.6 Specification title is revised in the ITS to, Reactor Pressure Vessel (RPV)

Water Level, and ISTS 3.9.7, [ Reactor Pressure Vessel (RPV)] Water Level - [New Fuel or Control Rods ], and associated Bases are not adopted in the ITS.

Subsequent Specifications and associated Bases are renumbered, as applicable.

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

[RPV] Water Level - Irradiated Fuel B 3.9.6 General Electric BWR/4 STS B 3.9.6-1 Rev. 5.0 Hope Creek Revision XXX 1

2 B 3.9 REFUELING OPERATIONS B 3.9.6 [Reactor Pressure Vessel (RPV)] Water Level - [Irradiated Fuel]

BASES BACKGROUND The movement of [irradiated] fuel assemblies [or handling of control rods]

within the [RPV] requires a minimum water level of [23] ft above the top of the [RPV] flange. During refueling, this maintains a sufficient water level in the reactor vessel cavity and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to 25% of 10 CFR 100 limits, as provided by the guidance of Reference 3.

APPLICABLE During movement of [irradiated] fuel assemblies [or handling of control SAFETY rods], the water level in the [RPV] is an initial condition design parameter ANALYSES in the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of 23 ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor of 100 (Regulatory Position C.1.g of Ref. 1) to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine inventory (Ref. 1).

Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained within allowable limits (Ref. 4).

While the worst case assumptions include the dropping of the irradiated fuel assembly being handled onto the reactor core, the possibility exists of the dropped assembly striking the [RPV] flange and releasing fission products. Therefore, the minimum depth for water coverage to ensure acceptable radiological consequences is specified from the [RPV] flange.

Since the worst case event results in failed fuel assemblies seated in the core, as well as the dropped assembly, dropping an assembly on the

[RPV] flange will result in reduced releases of fission gases. [Based on this judgement, and the physical dimensions which preclude normal operation with water level 23 feet above the flange, a slight reduction in this water level is acceptable (Ref. 4).]

[RPV] water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

22 ft 2 in A fuel handling accident is evaluated to ensure that the radiological consequences of the calculated total effective dose equivalent (TEDE) at the exclusion area boundary (EAB), at the low population zone (LPZ) outer boundary, and in the control room, meet the exposure guideline values specified in 10 CFR 50.67 (Ref. 3).

4 2

2 3 and 4 4

1 4

2 1.183 1

s s

> 23 ft above the reactor core (i.e., 22 ft 2 in as measured from the RPV flange) allows an overall effective decontamination factor of 200 99.5 4

4 is not expected in this case due to the lower kinetic energy in the dropped fuel assembly based on the short distance (< 1 ft) between a fully raised fuel assembly and the RPV flange. However

. R 2

2

[RPV] Water Level - Irradiated Fuel B 3.9.6 General Electric BWR/4 STS B 3.9.6-2 Rev. 5.0 Hope Creek Revision XXX 1

2 BASES LCO A minimum water level of [23] ft above the top of the [RPV] flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference 3.

APPLICABILITY LCO 3.9.6 is applicable when moving [irradiated] fuel assemblies [or handling control rods (i.e., movement with other than the normal control rod drive)] within the [RPV]. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. [If irradiated fuel is not present within the [RPV], there can be no significant radioactivity release as a result of a postulated fuel handling accident.] Requirements for handling of new fuel assemblies or control rods (where water depth to the [RPV] flange is not of concern) are covered by LCO 3.9.7, "[RPV] Water Level - New Fuel or Control Rods."

Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8, "Spent Fuel Storage Pool Water Level."

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

LCO 3.9.6 is written to cover new fuel and control rods as well as irradiated fuel. If a plant adopts LCO 3.9.7, however, the second bracketed portion of this Applicability is adopted in lieu of the first bracketed portion, and the LCO name and Required Action A.1 modified appropriately.

ACTIONS A.1 If the water level is < [23] ft above the top of the [RPV] flange, all operations involving movement of [irradiated] fuel assemblies [and handling of control rods] within the [RPV] shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of

[irradiated] fuel movement [and control rod handling] shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of [23] ft above the top of the [RPV]

flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

[ The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

22 ft 2 in 22 ft 2 in 22 ft 2 in 2

5 2

2 2

2 1

7 5

5 5

when moving new fuel assemblies or with irradiated fuel assemblies seated within the RPV 1

[RPV] Water Level - Irradiated Fuel B 3.9.6 General Electric BWR/4 STS B 3.9.6-3 Rev. 5.0 Hope Creek Revision XXX 1

2 BASES SURVEILLANCE REQUIREMENTS (continued)

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES

1.

Regulatory Guide 1.25, March 23, 1972.

2.

FSAR, Section [15.1.41].

3.

NUREG-0800, Section 15.7.4.

4.

10 CFR 100.11.

50.67 3

U 15.7.4 2

3 2

1 1

1 1.183, July 2000 1

JUSTIFICATION FOR DEVIATIONS ITS 3.9.6 BASES, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Hope Creek Page 1 of 1

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. The Reviewers 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.

4. The ISTS Bases is revised in the ITS to reflect the Hope Creek Generating Station current licensing basis. License Amendment 146, Hope Creek Generating Station -

Issuance of Amendment RE: Containment Requirements During Fuel Handling and Removal of Charcoal Filters (TAC NO. MB5548) (NRC ADAMS Accession No. ML030760293), approved changes to the Technical Specifications and included approval to use the alternate source term (AST) pursuant to 10 CFR 50.67 using the guidance provided in Regulatory Guide 1.183. As stated in UFSAR Section 15.7.4, the most severe fuel handling accident from a radiological viewpoint is the drop of an irradiated fuel assembly onto the reactor core. Consequently, the FHA radiological dose consequences assume an overall decontamination factor of 200 for iodine in elemental and particulate because a water level of 22 feet 2 inches above the RPV flange is greater than 23 feet above the reactor core.

5. The Reviewers 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 HCGS current technical specifications, ISTS LCO 3.9.6 is written in the ITS to be applicable during movement of irradiated fuel assemblies within the RPV, and during movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV. As a result, the ISTS 3.9.6 Specification and Bases title is revised in the ITS to, Reactor Pressure Vessel (RPV) Water Level, and ISTS LCO 3.9.7, [ Reactor Pressure Vessel (RPV) Water Level - [New Fuel or Control Rods ], and associated Bases are not adopted in the ITS.

Subsequent Specifications and associated Bases are renumbered, as applicable.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 7 ITS 3.9.7, Residual Heat Removal (RHR) - High Water Level

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

REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.1 At least one shutdown cooling mode loop of the residual heat removal (RHR) system shall be OPERABLE and in operation* with:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet 2 inches above the top of the reactor pressure vessel flange and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5.

ACTION:

a.

With no RHR shutdown cooling mode loop OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal. Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8hour period.

Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained).

HOPE CREEK 3/4 9-17 Amendment No. 187 A01 ITS 3.9.7 ITS 3.9 3.9.7 Residual Heat Removal (RHR) - High Water Level LCO 3.9.7 RHR subsystem LA01 MODE with reactor pressure vessel (RPV)

RPV LCO 3.9.7 NOTE subsystem required RHR Applicability A03 LA02 Required subsystem inoperable ACTION A Add proposed Required Action B.2, B.3, B.4 A04 subsystem Verify an is available loading irradiated fuel assemblies into the RPV Required Action B.1 Required Action B.2, B.3, B.4 Condition C Required Action C.1, C.2 from discovery of no reactor coolant circulation Verify 1

Immediately Verify RHR subsystem is SR 3.9.7.1 A02 immediately 1

A04 M01 Insert Applicability NOTE Applicability Note

DISCUSSION OF CHANGES ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL 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 3.9.11.1 Applicability states, in part, OPERATIONAL CONDITION 5. ITS 3.9.7 Applicability states, in part, MODE 5. This changes the CTS by incorporating the ITS MODE definition.

The purpose of CTS 3.9.11.1 Applicability is to establish the Operational Condition (i.e., ITS MODE) in which the LCO is required. This change is acceptable because the Applicability of MODE is not changed. 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 3.9.11.1 Applicability states, in part, and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5. Footnote ** states Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained). ITS 3.9.7 Applicability Note states, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation. This changes the CTS presentation by adding a Note to the Applicability regarding maintaining Reactor Coolant System (RCS) temperature with no RHR shutdown cooling (i.e., RCS ambient losses are greater than or equal to the RCS heat input).

The purpose of CTS 3.9.11.1 Applicability is to establish the MODE and other applicable conditions when an RHR shutdown cooling subsystem must be OPERABLE and in operation. The presentation of the Applicability Note maintains the intent of the current allowance associated with ambient heat losses, allowing RHR shutdown cooling subsystems to be secured when heat losses to ambient are sufficient to maintain MODE 5. This change is designated as an administrative change and is acceptable because it does not result in a technical change to the CTS.

A04 CTS 3.9.11.1 Action a requires with no required shutdown cooling mode loop (subsystem) OPERABLE, and the requirements cannot be met within the required action times, to suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY. ITS 3.9.7 ACTION B provides similar requirements be performed.

ITS 3.9.7 Required Action B.1 requires immediate suspension of loading irradiated fuel assemblies into the RPV. ITS 3.9.7 Required Actions B.2, B.3, and B.4 provide actions representative of the CTS SECONDARY CONTAINMENT INTEGRITY definition which is deleted in the ITS (See Section 1.0). This includes

DISCUSSION OF CHANGES ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 2 of 4 initiating action to ensure: 1) secondary containment boundary is established; 2) one Filtration Recirculation Ventilation System ventilation unit is OPERABLE; and

3) secondary containment isolation capability in each associated penetration not isolated is restored. This changes the presentation of CTS actions in the ITS to define the remedial actions more precisely.

The purpose of the CTS 3.9.11.1 Actions is to ensure additional decay heat is not added into the core and secondary containment integrity is established when there are no RHR shutdown cooling loops capable of removing decay heat. ITS 3.9.7, Required Action B.1 clarifies more precisely to suspend loading irradiated fuel assemblies into the RPV, which act to suspend any increase in reactor core heat load (fission and decay heat load). Although not explicitly stated, the intended time to suspend the activity is immediate. Therefore, the Completion Time of Required Action B.1 is specified as immediately. ITS 3.9.7, Required Actions B.2, B.3, and B.4 replace the CTS defined term " Secondary Containment Integrity" with multiple requirements that establish all the essential elements of secondary containment integrity. This change is consistent with the ISTS and necessary because secondary containment integrity is not a defined term in ITS. These changes consist of presentation preference changes that meet the same intent and are designated as administrative changes and do not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.9.11.1 Action a requires, in part, with no required shutdown cooling mode loop (subsystem) OPERABLE, and the requirements cannot be met within the required action times, then establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Commensurate with the CTS action establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , ITS 3.9.7 Required Actions B.2, B.3, and B.4 provide actions representative of the CTS SECONDARY CONTAINMENT INTEGRITY definition and require these actions be initiated immediately. This changes the CTS by requiring actions to be initiated immediately.

The purpose of the subject CTS actions is to provide follow-up actions when a Required Action and associated Completion Time is not met. ITS 3.9.7, ACTION B, requires the actions to be initiated immediately to establish the essential elements of secondary containment but does not specify a time to complete the action. CTS establishes completion times for the subject actions that, depending on plant conditions, may not be achievable in the required time. Conversely, plant conditions may be such that the actions could be accomplished in a shorter time than is required. Therefore, CTS actions appear to provide a period of time in which it is not necessary to establish the desired plant conditions even if those conditions can be readily established. Conversely, if plant status is such that the desired plant conditions cannot be established in the required time, the CTS actions result in noncompliance with the Technical Specifications and a requirement for a licensee event report (LER). ITS more appropriately establishes the required actions as " Initiate action" immediately to establish the desired plant conditions. This change eliminates a CTS allowance to avoid establishing the condition for up to the specified action time and replaces it with a

DISCUSSION OF CHANGES ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 3 of 4 requirement to establish the desired condition as quickly as is reasonably achievable. By eliminating the CTS completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , an LER is not required if the best efforts to establish the condition take longer than the time specified in the CTS. This change is designated as more restrictive because it adds specific action requirements with a shorter completion time to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.9.11.1 LCO requires that one shutdown cooling mode loop (subsystem) be OPERABLE, and provides details for each loop consisting of one OPERABLE RHR pump (CTS LCO 3.9.11.1.a) and one OPERABLE RHR heat exchanger (CTS LCO 3.9.11.1.b) to be considered OPERABLE. ITS LCO 3.9.7 does not include these system design details. This changes the CTS by moving these details regarding RHR shutdown cooling subsystem OPERABILITY from Technical Specifications to the ITS Bases.

The removal of these details, which are related to system design and operation, 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 LCO 3.9.7 retains the requirement that one RHR shutdown cooling mode loop (subsystem) be OPERABLE. Also, this change is acceptable because the removed information 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 information relating to system design is being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.9.11.1 Applicability is modified by footnote **

which states, Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained). This changes the CTS by moving these operational details regarding ambient losses from the Technical Specifications to the ITS Bases.

The removal of these operational 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 ITS 3.9.7 Bases clarifies that the LCO is not applicable when decay heat is low enough such that losses to ambient are sufficient to ensure no increase in RCS temperature will occur with RHR shutdown cooling subsystems not in operation. In this condition, a method of reactor coolant circulation must be maintained to provide assurance of continued RCS temperature monitoring

DISCUSSION OF CHANGES ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 4 of 4 capability. ITS LCO 3.9.7 retains the requirement that one RHR shutdown cooling subsystem be OPERABLE, and in operation. With the required RHR shutdown cooling subsystem inoperable, an alternate method of decay heat removal must be provided. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures.

The required cooling capacity of the alternate method should be sufficient to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Also, this change is acceptable because the removed information 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 operational procedure information relating to decay heat ambient losses is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None

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

RHR - High Water Level 3.9.8 General Electric BWR/4 STS 3.9.8-1 Rev. 5.0 7

Hope Creek Amendment XXX 1

3 7

3 CTS 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR) - High Water Level LCO 3.9.8 One RHR shutdown cooling subsystem shall be OPERABLE and in operation.

NOTE--------------------------------------------

The required RHR shutdown cooling subsystem may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

APPLICABILITY:

MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level [23] ft above the top of the [RPV flange].

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown cooling subsystem inoperable.

A.1 Verify an alternate method of decay heat removal is available.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter B. Required Action and associated Completion Time of Condition A not met.

B.1 Suspend loading irradiated fuel assemblies into the RPV.

AND B.2 Initiate action to restore

[secondary] containment to OPERABLE status.

AND Immediately Immediately 7

7 22 ft 2 in 3

3 3.9.11.1 3/4.9.11 Applicability Action a 3.9.11.1 Footnote

Action a DOC M01 2

5

NOTE----------------------------------------------

Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystem in operation.

4 establish boundary

RHR - High Water Level 3.9.8 General Electric BWR/4 STS 3.9.8-2 Rev. 5.0 7

Hope Creek Amendment XXX 1

3 7

3 CTS ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore one standby gas treatment subsystem to OPERABLE status.

AND B.4 Initiate action to restore isolation capability in each required [secondary]

containment penetration flow path not isolated.

Immediately Immediately C. No RHR shutdown cooling subsystem in operation.

C.1 Verify reactor coolant circulation by an alternate method.

AND C.2 Monitor reactor coolant temperature.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of no reactor coolant circulation AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter Once per hour Action b DOC M01 DOC M01 2

Filtration Recirculation and Ventilation System ventilation unit 5

6

RHR - High Water Level 3.9.8 General Electric BWR/4 STS 3.9.8-3 Rev. 5.0 7

Hope Creek Amendment XXX 1

3 7

3 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is operating.

[ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR In accordance with the Surveillance Frequency Control Program ]

SR 3.9.8.2 Verify required RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water.

[ 31 days OR In accordance with the Surveillance Frequency Control Program ]

7 2

2 7

3 4.9.11.1

JUSTIFICATION FOR DEVIATIONS ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 1 of 2

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. Hope Creek Generating Station (HCGS) is not adopting ISTS LCO 3.9.7, Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods. Therefore, subsequent Specifications and associated Bases are renumbered, as applicable, to reflect this ISTS deviation.

4. ISTS 3.9.8 (ITS 3.9.7) Applicability is revised in the ITS to include a note stating, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation. This deviation is consistent with the HCGS current licensing basis established at initial licensing as shown in NUREG-1202, Technical Specification Hope Creek Generating Station, dated July 1986 (NRC ADAMS Accession No. ML20205D512). Decay heat losses to ambient must be such that no increase in reactor vessel water temperature will occur.

Additionally, a method of reactor coolant circulation will be maintained to provide assurance of continued RCS temperature monitoring capability.

5. ISTS 3.9.8 (ITS 3.9.7) Required Action B.2 is modified in the ITS to state, "Initiate action to establish secondary containment boundary," and Required Action B.3 is modified in the ITS from one Standby Gas Treatment (SGT) subsystem to one Filtration Recirculation and Ventilation System (FRVS) ventilation unit. This deviation reflects the HCGS design and licensing basis. OPERABILITY of one FRVS ventilation unit provides the equivalent level of secondary containment protection as one SGT subsystem in MODES 4 and 5. Refer to Section 2.2.6.11 of the NRC safety evaluation issued with HCGS License Amendment 213 (NRC ADAMS Accession No. ML18260A203). Full restoration of the secondary containment to OPERABLE status cannot be accomplished without restoring at least four FRVS recirculation units to OPERABLE status. Since the FRVS recirculation units are not required to provide the filtration function in MODES 4 and 5, it is unnecessary to fully restore the secondary containment to OPERABLE status.

Therefore, ITS 3.9.7 Required Actions B.2 requires initiating action to establish the secondary containment boundary like Required Action D.2 of ISTS 3.5.2, Reactor Pressure Vessel (RPV) Water Inventory Control, instead of restoring secondary containment to OPERABLE status. The ISTS Bases is modified in the ITS to clarify what constitutes establishing the secondary containment boundary. The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

6. The second Completion Time of AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter associated with ISTS 3.9.8, Required Action C.1 is not included in ITS 3.9.7 consistent with HCGS current licensing basis. As stated in the ISTS Bases of Required Actions C.1 and C.2, the reactor coolant temperature must be periodically monitored during the period when the reactor coolant is being circulated by an alternate method (other

JUSTIFICATION FOR DEVIATIONS ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 2 of 2 than by the required RHR Shutdown Cooling System) to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

This frequent monitoring of reactor coolant temperature (Required Action C.2) to confirm the proper operation of alternate method of coolant circulation is considered sufficient in lieu of additional performances of Required Action C.1.

7. ISTS SR 3.9.8.2, associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water, is not adopted in the ITS. NRC Generic Letter (GL) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems" (ADAMS Accession No. ML072910759), required licensees to submit information, in general, regarding concerns that certain safety systems could accumulate gas pockets that may inhibit system flow or damage system pumps/piping during an event which requires system operation. PSEG provided response to GL 2008-01 related to Hope Creek Generating Station (HCGS) management of gas accumulation in letters dated April 10, 2008, October 13, 2008, July 30, 2009, and January 28, 2011 (ADAMS Accession Nos. ML081130672, ML082970219, ML092230347 and ML110400201, respectively). The responses included the results of evaluations performed on systems of concern, summary of procedural controls, and description of support features (such as the Emergency Core Cooling System "keep-fill" system). Based on the review of information provided by PSEG, the NRC found the responses acceptable and subsequently closed the GL 2008-01 request for HCGS in letter to T. Joyce (PSEG) from R.B.

Ennis (NRC), dated June 2, 2011 (ADAMS Accession No. ML111380081), with no further information or action required.

SRs associated with verifying certain systems remain full of water were adopted in the ISTS following NRC approval of Technical Specification Task Force (TSTF) traveler TSTF-523-A, "Generic Letter 2008-01, Managing Gas Accumulation,"

Revision 2, dated January 15, 2014 (79 FR 2700). PSEG subsequently performed an evaluation and determined that adoption of TSTF-523 was not necessary based on established controls and support features previously determined acceptable by the NRC. In addition, current technical specifications do not contain a Surveillance associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water. Based on the considerations provided herein, adoption of ISTS SR 3.9.8.2 is not necessary to ensure associated systems are properly maintained in an OPERABLE condition.

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

RHR - High Water Level B 3.9.8 General Electric BWR/4 STS B 3.9.8-1 Rev. 5.0 7

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B 3.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR) - High Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal.

APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.

The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE and in operation in MODE 5 with irradiated fuel in the RPV and the water level [23] ft above the RPV flange. Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem. Management of gas voids is important to RHR Shutdown Cooling System OPERABILITY.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate 7

3 22 ft 2 in 2

The RHR System consists of four separate subsystems, each with one RHR pump. RHR subsystems C and D are dedicated to low pressure coolant injection (LPCI) mode of operation. RHR subsystems A and B are also used for LPCI and can be aligned for shutdown cooling. Each shutdown cooling subsystem consists of one motor driven pump, a heat exchanger, and associated piping and valves. Both loops have a common suction from recirculation loop B.

The RHR heat exchangers transfer heat to the Safety Auxiliaries Cooling System.

RHR cross tie valves allow the C and D RHR pumps to be used as a means for alternate decay heat removal via their respective RHR loop heat exchanger if the normal shutdown cooling RHR pump (A or B) for that loop is inoperable.

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RHR - High Water Level B 3.9.8 General Electric BWR/4 STS B 3.9.8-2 Rev. 5.0 7

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BASES LCO (continued) core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception for the operating subsystem to be removed from operation every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE and in operation in MODE 5, with irradiated fuel in the reactor pressure vessel and with the water level [23] feet above the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);

Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level

< [23] ft above the RPV flange are given in LCO 3.9.9.

ACTIONS A.1 With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method(s) must be reconfirmed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. This will ensure continued heat removal capability.

Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. The required cooling capacity of the alternate method should be sufficient to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability.

Alternate methods that can be used include (but are not limited to) the Spent Fuel Pool Cooling System, the Reactor Water Cleanup System, or an inoperable but functional RHR shutdown cooling subsystem. The method used to remove the decay heat should be the most prudent choice based on unit conditions.

22 ft 2 in 22 ft 2 in 8, "Residual Heat Removal (RHR) - Low Water Level."

3 2

2 A Note provides exception when decay heat is low enough such that losses to ambient are sufficient to ensure no increase in Reactor Coolant System (RCS) temperature will occur with RHR shutdown cooling subsystems not in operation. In this condition, a method of reactor coolant circulation must be maintained to provide assurance of continued RCS temperature monitoring capability.

5

RHR - High Water Level B 3.9.8 General Electric BWR/4 STS B 3.9.8-3 Rev. 5.0 7

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BASES ACTIONS (continued)

B.1, B.2, B.3, and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending loading of irradiated fuel assemblies into the RPV.

Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE; one standby gas treatment subsystem is OPERABLE; and secondary containment isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, a surveillance may need to be performed to restore the component to OPERABLE status.

Actions must continue until all required components are OPERABLE.

C.1 and C.2 If no RHR Shutdown Cooling System is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Completion Time is modified such that the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that the RHR subsystem is in operation and circulating reactor coolant.

Filtration Recirculation and Ventilation System (FRVS) ventilation unit 3

6 7

boundary is established The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

These Required Actions s

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RHR - High Water Level B 3.9.8 General Electric BWR/4 STS B 3.9.8-4 Rev. 5.0 7

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BASES SURVEILLANCE REQUIREMENTS (continued)

The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. [ The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.9.8.2 RHR Shutdown Cooling System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required RHR shutdown cooling subsystem(s) and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of RHR Shutdown Cooling System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR Shutdown Cooling System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or 4

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RHR - High Water Level B 3.9.8 General Electric BWR/4 STS B 3.9.8-5 Rev. 5.0 7

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BASES SURVEILLANCE REQUIREMENTS (continued) discharge of a pump), the Surveillance is not met. If the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. If it is determined by subsequent evaluation that the RHR Shutdown Cooling System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR Shutdown Cooling System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

[ The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the RHR Shutdown Cooling System piping and the procedural controls governing system operation.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES None.

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JUSTIFICATION FOR DEVIATIONS ITS 3.9.7 BASES, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 1 of 2

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. Hope Creek Generating Station (HCGS) is not adopting ISTS LCO 3.9.7, Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods. Therefore, subsequent Specifications and associated Bases are renumbered, as applicable, to reflect this ISTS deviation and the Specification title is added, where applicable, in accordance with TSTF-GG-05-01, "Writer's Guide for Plant-Specific Improved Technical Specifications."

4. ISTS SR 3.9.8.2, associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water, is not adopted in the ITS. PSEG provided response to NRC Generic Letter (GL) 2008-01 related to Hope Creek Generating Station (HCGS) management of gas accumulation in letters dated April 10, 2008, October 13, 2008, July 30, 2009, and January 28, 2011 (ADAMS Accession Nos. ML081130672, ML082970219, ML092230347 and ML110400201, respectively). The responses included the results of evaluations performed on systems of concern, summary of procedural controls, and description of support features (such as the Emergency Core Cooling System "keep-fill" system).

Based on the review of information provided by PSEG, the NRC found the responses acceptable and subsequently closed the GL 2008-01 request for HCGS in letter to T.

Joyce (PSEG) from R.B. Ennis (NRC), dated June 2, 2011 (ADAMS Accession No. ML111380081), with no further information or action required. Therefore, adoption of ISTS SR 3.9.8.2 is not necessary to ensure associated systems are properly maintained in an OPERABLE condition. Changes to the ISTS Bases are made to support the changes to the Specification.

5. ISTS 3.9.8 (ITS 3.9.7) Applicability is revised in the ITS to include a note stating, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation, consistent with the HCGS current licensing basis. Decay heat losses to ambient must be such that no increase in reactor vessel water temperature will occur. Additionally, a method of reactor coolant circulation will be maintained to provide assurance of continued RCS temperature monitoring capability. Changes to the ITS Bases reflect the change to the Specification.

6. ISTS 3.9.8 (ITS 3.9.7) Required Action B.2 is modified in the ITS to state, "Initiate action to establish secondary containment boundary," and Required Action B.3 is modified from one Standby Gas Treatment (SGT) subsystem to one Filtration Recirculation and Ventilation System (FRVS) ventilation unit. This deviation reflects the HCGS design and licensing basis. As discussed in Section 2.2.6.11 of the NRC safety evaluation issued with HCGS License Amendment 213 (NRC ADAMS Accession No. ML18260A203), FRVS provides the equivalent level of secondary containment protection as the SGT System. One FRVS ventilation unit is capable of maintaining the secondary containment at a negative pressure with respect to the

JUSTIFICATION FOR DEVIATIONS ITS 3.9.7 BASES, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 2 of 2 environment and filter gaseous releases in MODES 4 and 5. The ISTS Bases is modified in the ITS to support the change to the Specification and clarify what constitutes establishing the secondary containment boundary. The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

7. The Reviewers 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.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 8 ITS 3.9.8, Residual Heat Removal (RHR) - Low Water Level

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

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and at least one loop shall be in operation,* with each loop consisting of:

a.

One OPERABLE RHR pump, and

b.

One OPERABLE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet 2 inches above the top of the reactor pressure vessel flange and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5.

ACTION:

a.

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the OPERABILITY of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8hour period.

Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained).

HOPE CREEK 3/4 9-18 Amendment No. 187 A01 ITS 3.9.8 ITS 3.9 3.9.8 Residual Heat Removal (RHR) - Low Water Level RHR subsystems RHR shutdown cooling subsystem MODE with reactor pressure vessel (RPV)

RPV LA01 A03 One or two required subsystems inoperable 1

Verify an is available required subsystem ACTION A ACTION C Add proposed ACTION B M01 subsystem 1

from discovery of no reactor coolant circulation Verify SR 3.9.8.1 Verify RHR subsystem is LCO 3.9.8 NOTE subsystem required operating RHR LA02 A02 LCO 3.9.8 Applicability Insert Applicability NOTE Applicability Note

DISCUSSION OF CHANGES ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL 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 3.9.11.2 Applicability states, in part, OPERATIONAL CONDITION 5. ITS 3.9.8 Applicability states, in part, MODE 5. This changes the CTS by incorporating the ITS MODE definition.

The purpose of CTS 3.9.11.2 Applicability is to establish the Operational Condition (i.e., ITS MODE) in which the LCO is required. This change is acceptable because the Applicability of MODE is not changed. 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 3.9.11.2 Applicability states, in part, and heat losses to ambient** are not sufficient to maintain OPERATIONAL CONDITION 5. Footnote ** states Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained). ITS 3.9.8 Applicability Note states, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation. This changes the CTS presentation by adding a Note to the Applicability regarding maintaining Reactor Coolant System (RCS) temperature with no RHR shutdown cooling (i.e., RCS ambient losses are greater than or equal to the RCS heat input).

The purpose of CTS 3.9.11.2 Applicability is to establish the MODE and other applicable conditions when an RHR shutdown cooling subsystem must be OPERABLE and in operation. The presentation of the Applicability Note maintains the intent of the current allowance associated with ambient heat losses, allowing RHR shutdown cooling subsystems to be secured when heat losses to ambient are sufficient to maintain MODE 5. 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 M01 CTS 3.9.11.2 Action a provides requirements for one or both required shutdown cooling mode loops (subsystems) inoperable but does not provide an action if the requirements cannot be met within the required action times. ITS 3.9.8 ACTION B provides actions for when one or two required shutdown cooling mode loops (subsystems) are inoperable, and the Required Action and associated Completion Time is not met. ITS 3.9.8, ACTION B requires, with the

DISCUSSION OF CHANGES ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 2 of 3 Required Action and associated Completion Time of Condition A not met, to immediately initiate action to: 1) establish the secondary containment boundary;

2) restore one Filtration Recirculation and Ventilation System ventilation unit to OPERABLE status; and 3) restore isolation capability in each required secondary containment penetration flow path not isolated. This changes the CTS by providing actions for when one or two inoperable shutdown cooling mode loops (subsystems) are inoperable, and the Required Action cannot be met within the required Completion Time.

The purpose of ITS 3.9.8 ACTION B is to provide follow-up actions when a Required Action and associated Completion Time is not met. ITS 3.9.8 Required Actions B.1, B.2, and B.3 provide additional actions to minimize any potential fission product release to the environment. These actions may be performed as administrative checks, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons.

The immediate Completion Time reflects the importance of establishing secondary containment. The actions must continue until all required components are OPERABLE. This change is acceptable because it provides additional assurance that any potential fission product release to the environment is minimized when shutdown cooling subsystems are inoperable. This change is designated as more restrictive because it adds additional action requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.9.11.2 LCO requires that two shutdown cooling mode loops (subsystems) be OPERABLE, and provides details for each loop consisting of one OPERABLE RHR pump (CTS LCO 3.9.11.2.a) and one OPERABLE RHR heat exchanger (CTS LCO 3.9.11.2.b) to be considered OPERABLE.

ITS LCO 3.9.8 does not include these system design details. This changes the CTS by moving these details regarding shutdown cooling subsystem OPERABILITY from the Technical Specifications to the ITS Bases.

The removal of these details, which are related to system design and operation, 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 LCO 3.9.8 retains the requirement that two shutdown cooling mode loops (subsystems) be OPERABLE. Also, this change is acceptable because the removed information 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

DISCUSSION OF CHANGES ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 3 of 3 detail change because information relating to system design is being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.9.11.2 Applicability is modified by footnote **

which states, Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained). This changes the CTS by moving these operational details regarding ambient losses from the Technical Specifications to the ITS Bases.

The removal of these operational 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 ITS 3.9.8 Bases clarifies that the LCO is not applicable when decay heat is low enough such that losses to ambient are sufficient to ensure no increase in RCS temperature will occur with RHR shutdown cooling subsystems not in operation. In this condition, a method of reactor coolant circulation must be maintained to provide assurance of continued RCS temperature monitoring capability. ITS LCO 3.9.8 retains the requirement that two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation. With one of the two required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided. With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. The required cooling capacity of the alternate method should be sufficient to maintain or reduce temperature.

Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Also, this change is acceptable because the removed information 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 operational procedure information relating to decay heat ambient losses is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None

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

RHR - Low Water Level 3.9.9 General Electric BWR/4 STS 3.9.9-1 Rev. 5.0 8

Hope Creek Amendment XXX 1

3 8

3 CTS 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR) - Low Water Level LCO 3.9.9 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.

NOTE--------------------------------------------

The required operating shutdown cooling subsystem may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

APPLICABILITY:

MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < [23] ft above the top of the [RPV flange].

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required RHR shutdown cooling subsystems inoperable.

A.1 Verify an alternate method of decay heat removal is available for each inoperable required RHR shutdown cooling subsystem.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter B. Required Action and associated Completion Time of Condition A not met.

B.1 Initiate action to restore

[secondary] containment to OPERABLE status.

AND B.2 Initiate action to restore one standby gas treatment subsystem to OPERABLE status.

AND Immediately Immediately 8

8 22 ft 2 in 3

3 3.9.11.2 Applicability 3.9.11.1 Footnote

Action a DOC M01 Filtration Recirculation and Ventilation System ventilation unit 5

5 DOC M01 2

4 boundary establish

NOTE----------------------------------------------

Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystem in operation.

RHR - Low Water Level 3.9.9 General Electric BWR/4 STS 3.9.9-2 Rev. 5.0 8

Hope Creek Amendment XXX 1

3 8

3 CTS ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore isolation capability in each required [secondary]

containment penetration flow path not isolated.

Immediately C. No RHR shutdown cooling subsystem in operation.

C.1 Verify reactor coolant circulation by an alternate method.

AND C.2 Monitor reactor coolant temperature.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of no reactor coolant circulation AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter Once per hour SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.9.1 Verify one RHR shutdown cooling subsystem is operating.

[ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR In accordance with the Surveillance Frequency Control Program ]

8 2

2 3

DOC M01 Action b 4.9.11.2 2

6

RHR - Low Water Level 3.9.9 General Electric BWR/4 STS 3.9.9-3 Rev. 5.0 8

Hope Creek Amendment XXX 1

3 8

3 CTS SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.9.9.2 Verify RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water.

[ 31 days OR In accordance with the Surveillance Frequency Control Program ]

7

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 1 of 2

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. Hope Creek Generating Station (HCGS) is not adopting ISTS LCO 3.9.7, Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods. Therefore, subsequent Specifications and associated Bases are renumbered, as applicable, to reflect this ISTS deviation.

4. ISTS 3.9.9 (ITS 3.9.8) Applicability is revised in the ITS to include a note stating, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation. This deviation is consistent with the HCGS current licensing basis established at initial licensing as shown in NUREG-1202, Technical Specification Hope Creek Generating Station, dated July 1986 (NRC ADAMS Accession No. ML20205D512). Decay heat losses to ambient must be such that no increase in reactor vessel water temperature will occur.

Additionally, a method of reactor coolant circulation will be maintained to provide assurance of continued RCS temperature monitoring capability.

5. ISTS 3.9.9 (ITS 3.9.8) Required Action B.1 is modified in the ITS to state, "Initiate action to establish secondary containment boundary," and Required Action B.2 is modified in the ITS from one Standby Gas Treatment (SGT) subsystem to one Filtration Recirculation and Ventilation System (FRVS) ventilation unit. This deviation reflects the HCGS design and licensing basis. OPERABILITY of one FRVS ventilation unit provides the equivalent level of secondary containment protection as one SGT subsystem in MODES 4 and 5. Refer to Section 2.2.6.11 of the NRC safety evaluation issued with HCGS License Amendment 213 (NRC ADAMS Accession No. ML18260A203). Full restoration of the secondary containment to OPERABLE status cannot be accomplished without restoring at least four FRVS recirculation units to OPERABLE status. Since the FRVS recirculation units are not required to provide the filtration function in MODES 4 and 5, it is unnecessary to fully restore the secondary containment to OPERABLE status. Therefore, ITS 3.9.8 Required Actions B.1 requires initiating action to establish the secondary containment boundary like Required Action D.2 of ISTS 3.5.2, Reactor Pressure Vessel (RPV) Water Inventory Control, instead of restoring secondary containment to OPERABLE status. The ISTS Bases is modified in the ITS to clarify what constitutes establishing the secondary containment boundary. The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

6. The second Completion Time of AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter associated with ISTS 3.9.9, Required Action C.1 is not included in ITS 3.9.8 consistent with HCGS current licensing basis. As stated in the ISTS Bases of Required Actions C.1 and C.2, the reactor coolant temperature must be periodically monitored during the period when the reactor coolant is being circulated by an alternate method (other

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 2 of 2 than by the required RHR Shutdown Cooling System) to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

This frequent monitoring of reactor coolant temperature (Required Action C.2) to confirm the proper operation of alternate method of coolant circulation is considered sufficient in lieu of additional performances of Required Action C.1.

7. ISTS SR 3.9.9.2, associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water, is not adopted in the ITS. NRC Generic Letter (GL) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems" (ADAMS Accession No. ML072910759), required licensees to submit information, in general, regarding concerns that certain safety systems could accumulate gas pockets that may inhibit system flow or damage system pumps/piping during an event which requires system operation. PSEG provided response to GL 2008-01 related to Hope Creek Generating Station (HCGS) management of gas accumulation in letters dated April 10, 2008, October 13, 2008, July 30, 2009, and January 28, 2011 (ADAMS Accession Nos. ML081130672, ML082970219, ML092230347 and ML110400201, respectively). The responses included the results of evaluations performed on systems of concern, summary of procedural controls, and description of support features (such as the Emergency Core Cooling System "keep-fill" system). Based on the review of information provided by PSEG, the NRC found the responses acceptable and subsequently closed the GL 2008-01 request for HCGS in letter to T. Joyce (PSEG) from R.B.

Ennis (NRC), dated June 2, 2011 (ADAMS Accession No. ML111380081), with no further information or action required.

SRs associated with verifying certain systems remain full of water were adopted in the ISTS following NRC approval of Technical Specification Task Force (TSTF) traveler TSTF-523-A, "Generic Letter 2008-01, Managing Gas Accumulation,"

Revision 2, dated January 15, 2014 (79 FR 2700). PSEG subsequently performed an evaluation and determined that adoption of TSTF-523 was not necessary based on established controls and support features previously determined acceptable by the NRC. In addition, current technical specifications do not contain a Surveillance associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water. Based on the considerations provided herein, adoption of ISTS SR 3.9.9.2 is not necessary to ensure associated systems are properly maintained in an OPERABLE condition.

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

RHR - Low Water Level B 3.9.9 General Electric BWR/4 STS B 3.9.9-1 Rev. 5.0 Hope Creek Revision XXX 1

8 3

B 3.9 REFUELING OPERATIONS B 3.9.9 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.

The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 23 ft above the reactor pressure vessel (RPV) flange both RHR shutdown cooling subsystems must be OPERABLE.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. To meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem.

Management of gas voids is important to RHR Shutdown Cooling System OPERABILITY.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception for the operating subsystem to be removed from operation every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

8 3

22 ft 2 in 2

The RHR System consists of four separate subsystems, each with one RHR pump. RHR subsystems C and D are dedicated to low pressure coolant injection (LPCI) mode of operation. RHR subsystems A and B are also used for LPCI and can be aligned for shutdown cooling. Each shutdown cooling subsystem consists of one motor driven pump, a heat exchanger, and associated piping and valves. Both loops have a common suction from recirculation loop B. The RHR heat exchangers transfer heat to the Safety Auxiliaries Cooling System. RHR cross tie valves allow the C and D RHR pumps to be used as a means for alternate decay heat removal via their respective RHR loop heat exchanger if the normal shutdown cooling RHR pump (A or B) for that loop is inoperable.

1 1

4

RHR - Low Water Level B 3.9.9 General Electric BWR/4 STS B 3.9.9-2 Rev. 5.0 Hope Creek Revision XXX 1

8 3

BASES APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE, and one must be in operation in MODE 5, with irradiated fuel in the RPV and with the water level < [23] ft above the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);

Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the RPV and with the water level [23] ft above the RPV flange are given in LCO 3.9.8, "Residual Heat Removal (RHR) - High Water Level."

ACTIONS A.1 With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced.

Therefore an alternate method of decay heat removal must be provided.

With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of this alternate method(s) must be reconfirmed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. This will ensure continued heat removal capability.

Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. The required cooling capacity of the alternate method should be sufficient to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability.

Alternate methods that can be used include (but are not limited to) the Spent Fuel Pool Cooling System, the Reactor Water Cleanup System, or an inoperable but functional RHR shutdown cooling subsystem. The method used to remove decay heat should be the most prudent choice based on unit conditions.

22 ft 2 in 22 ft 2 in 7

3 2

2 A Note provides exception when decay heat is low enough such that losses to ambient are sufficient to ensure no increase in Reactor Coolant System (RCS) temperature will occur with RHR shutdown cooling subsystems not in operation. In this condition, a method of reactor coolant circulation must be maintained to provide assurance of continued RCS temperature monitoring capability.

5

RHR - Low Water Level B 3.9.9 General Electric BWR/4 STS B 3.9.9-3 Rev. 5.0 Hope Creek Revision XXX 1

8 3

BASES ACTIONS (continued)

B.1, B.2, and B.3 With the required decay heat removal subsystem(s) inoperable and the required alternate method(s) of decay heat removal not available in accordance with Required Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE; one standby gas treatment subsystem is OPERABLE; and secondary containment isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

C.1 and C.2 If no RHR subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Completion Time is modified such that the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

SURVEILLANCE SR 3.9.9.1 REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.

[ The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room.

Filtration Recirculation and Ventilation System (FRVS) ventilation unit 3

8 6

2 boundary is established The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

These Required Actions s

6

RHR - Low Water Level B 3.9.9 General Electric BWR/4 STS B 3.9.9-4 Rev. 5.0 Hope Creek Revision XXX 1

8 3

BASES SURVEILLANCE REQUIREMENTS (continued)

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.9.9.2 RHR Shutdown Cooling System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR shutdown cooling subsystems and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of RHR Shutdown Cooling System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR Shutdown Cooling System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. If it is determined by subsequent evaluation that the RHR Shutdown Cooling 4

2 7

RHR - Low Water Level B 3.9.9 General Electric BWR/4 STS B 3.9.9-5 Rev. 5.0 Hope Creek Revision XXX 1

8 3

BASES SURVEILLANCE REQUIREMENTS (continued)

System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR Shutdown Cooling System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

[ The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the RHR Shutdown Cooling System piping and the procedural controls governing system operation.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES None.

4 7

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8 BASES, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 1 of 2

1. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is changed to reflect the current licensing basis.

3. Hope Creek Generating Station (HCGS) is not adopting ISTS LCO 3.9.7, Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods. Therefore, subsequent Specifications and associated Bases are renumbered, as applicable, to reflect this ISTS deviation.

4. ISTS SR 3.9.9.2, associated with verifying RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water, is not adopted in the ITS. PSEG provided response to NRC Generic Letter (GL) 2008-01 related to Hope Creek Generating Station (HCGS) management of gas accumulation in letters dated April 10, 2008, October 13, 2008, July 30, 2009, and January 28, 2011 (ADAMS Accession Nos. ML081130672, ML082970219, ML092230347 and ML110400201, respectively). The responses included the results of evaluations performed on systems of concern, summary of procedural controls, and description of support features (such as the Emergency Core Cooling System "keep-fill" system).

Based on the review of information provided by PSEG, the NRC found the responses acceptable and subsequently closed the GL 2008-01 request for HCGS in letter to T.

Joyce (PSEG) from R.B. Ennis (NRC), dated June 2, 2011 (ADAMS Accession No. ML111380081), with no further information or action required. Therefore, adoption of ISTS SR 3.9.9.2 is not necessary to ensure associated systems are properly maintained in an OPERABLE condition. Changes to the ISTS Bases are made to support the changes to the Specification.

5. ISTS 3.9.9 (ITS 3.9.8) Applicability is revised in the ITS to include a note stating, Not applicable when Reactor Coolant System temperature can be maintained with no RHR shutdown cooling subsystems in operation, consistent with the HCGS current licensing basis. Decay heat losses to ambient must be such that no increase in reactor vessel water temperature will occur. Additionally, a method of reactor coolant circulation will be maintained to provide assurance of continued RCS temperature monitoring capability. Changes to the ITS Bases reflect the change to the Specification.

6. ISTS 3.9.9 (ITS 3.9.8) Required Action B.1 modified in the ITS to state, "Initiate action to establish secondary containment boundary, and Required Action B.2 is modified in the ITS from one Standby Gas Treatment (SGT) subsystem to one Filtration Recirculation and Ventilation System (FRVS) ventilation unit. This deviation reflects the HCGS design and licensing basis. As discussed in Section 2.2.6.11 of the NRC safety evaluation issued with HCGS License Amendment 213 (NRC ADAMS Accession No. ML18260A203), FRVS provides the equivalent level of secondary containment protection as the SGT System. One FRVS ventilation unit is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases in MODES 4 and 5. The ISTS Bases is modified in the ITS to support the change to the Specification and

JUSTIFICATION FOR DEVIATIONS ITS 3.9.8 BASES, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 2 of 2 clarify what constitutes establishing the secondary containment boundary. The secondary containment boundary is considered established when the secondary containment is sufficiently leak tight such that one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

7. The Reviewers 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.

Specific No Significant Hazards Considerations (NSHCs)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL Hope Creek Page 1 of 1 There are no specific No Significant Hazards Considerations for this Specification.

ATTACHMENT 9 ISTS Not Adopted

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

[RPV] Water Level - [New Fuel or Control Rods]

3.9.7 General Electric BWR/4 STS 3.9.7-1 Rev. 5.0 3.9 REFUELING OPERATIONS 3.9.7

[ Reactor Pressure Vessel (RPV)] Water Level - [New Fuel or Control Rods ]

LCO 3.9.7

[RPV] water level shall be [23] ft above the top of irradiated fuel assemblies seated within the [RPV].

APPLICABILITY:

During movement of new fuel assemblies or handling of control rods within the [RPV], when irradiated fuel assemblies are seated within the [RPV].

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [RPV] water level not within limit.

A.1 Suspend movement of fuel assemblies and handling of control rods within the

[RPV].

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify [RPV] water level is [23] ft above the top of irradiated fuel assemblies seated within the [RPV].

[ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR In accordance with the Surveillance Frequency Control Program ]

1

JUSTIFICATION FOR DEVIATIONS ISTS 3.9.7, [REACTOR PRESSURE VESSEL (RPV)] WATER LEVEL - [NEW FUEL OR CONTROL RODS]

Hope Creek Page 1 of 1

1. ISTS 3.9.7, [Reactor Pressure Vessel (RPV)] Water Level - [New Fuel or Control Rods] and associated Bases are not included in the Hope Creek Generating Station (HCGS) ITS because the HCGS current technical specifications do not include a separate Specification for RPV water level during movement of new fuel assemblies or handling of control rods in the RPV. Subsequent ISTS Specifications and associated Bases are renumbered, as applicable.

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

[RPV] Water Level - [New Fuel or Control Rods]

B 3.9.7 General Electric BWR/4 STS B 3.9.7-1 Rev. 5.0 B 3.9 REFUELING OPERATIONS B 3.9.7 [Reactor Pressure Vessel (RPV)] Water Level - [New Fuel or Control Rods]

BASES BACKGROUND The movement of new fuel assemblies or handling of control rods within the [RPV] when fuel assemblies seated within the reactor vessel are irradiated requires a minimum water level of [23] ft above the top of irradiated fuel assemblies seated within the [RPV]. During refueling, this maintains a sufficient water level above the irradiated fuel. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to 25%

of 10 CFR 100 limits, as provided by the guidance of Reference 3.

APPLICABLE During movement of new fuel assemblies or handling of control rods over SAFETY irradiated fuel assemblies, the water level in the [RPV] is an initial ANALYSES condition design parameter in the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of [23] ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor of 100 (Regulatory Position C.1.g of Ref. 1) to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine inventory (Ref. 1).

Analysis of the fuel handling accident inside containment is described in Reference 2. With a minimum water level of [23] ft and a minimum decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained within allowable limits (Ref. 4).

The related assumptions include the worst case dropping of an irradiated fuel assembly onto the reactor core loaded with irradiated fuel assemblies.

[RPV] water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO A minimum water level of [23] ft above the top of irradiated fuel assemblies seated within the [RPV] flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference 3.

1

[RPV] Water Level - [New Fuel or Control Rods]

B 3.9.7 General Electric BWR/4 STS B 3.9.7-2 Rev. 5.0 BASES APPLICABILITY LCO 3.9.7 is applicable when moving new fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) over irradiated fuel assemblies seated within the [RPV]. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the [RPV], there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8, "Spent Fuel Storage Pool Water Level." Requirements for handling irradiated fuel over the [RPV] are covered by LCO 3.9.6,

"[Reactor Pressure Vessel (RPV)] Water Level - [Irradiated Fuel]."

ACTIONS A.1 If the water level is < [23] ft above the top of irradiated fuel assemblies seated within the [RPV], all operations involving movement of new fuel assemblies and handling of control rods within the [RPV] shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of [23] ft above the top of irradiated fuel assemblies seated within the [RPV] ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

[ The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

1

[RPV] Water Level - [New Fuel or Control Rods]

B 3.9.7 General Electric BWR/4 STS B 3.9.7-3 Rev. 5.0 BASES REFERENCES

1.

Regulatory Guide 1.25, March 23, 1972.

2.

FSAR, Section [15.1.41].

3.

NUREG-0800, Section 15.7.4.

4.

10 CFR 100.11.

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JUSTIFICATION FOR DEVIATIONS ISTS 3.9.7 BASES, [REACTOR PRESSURE VESSEL (RPV)] WATER LEVEL - [NEW FUEL OR CONTROL RODS]

Hope Creek Page 1 of 1

1. ISTS 3.9.7, [Reactor Pressure Vessel (RPV)] Water Level - [New Fuel or Control Rods] and associated Bases are not included in the Hope Creek Generating Station (HCGS) ITS because the HCGS current technical specifications do not include a separate Specification for RPV water level during movement of new fuel assemblies or handling of control rods in the RPV. Subsequent ISTS Specifications and associated Bases are renumbered, as applicable.

ML24142A442

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