LR-N12-0107, License Amendment Request - APRM Operability Requirements in Opcon 5

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License Amendment Request - APRM Operability Requirements in Opcon 5
ML12158A532
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 06/06/2012
From: Jamila Perry
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Material Safety and Safeguards, NRC/RGN-II
References
LAR H12-02, LR-N12-0107
Download: ML12158A532 (19)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 OPSEG NuclearLLC

'JUN 0'6 2012 10 CFR 50.90 LR-N12-0107 LAR H12-02 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

LICENSE AMENDMENT REQUEST - APRM OPERABILITY REQUIREMENTS IN OPCON 5 In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment to Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS). In accordance with 10 CFR 50.91(b)(1), a copy of this request for amendment has been sent to the State of New Jersey.

The proposed amendment would revise Technical Specification (TS) 3/4.3.1, "Reactor Protection System Instrumentation," and TS 3/4.3.6, "Control Rod Block Instrumentation" by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The proposed amendment would eliminate the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable in the "Refueling" Operational Condition (OPCON) 5.

No new regulatory commitments are established by this submittal. to this letter provides an evaluation supporting the proposed changes. The marked-up TS pages, with the proposed changes indicated, are provided in Attachment 2 to this letter.

PSEG requests approval of the proposed change by July 1, 2013, with the amendment being implemented within 60 days of issuance.

If you have any questions or require additional information, please do not hesitate to contact Mr.

Brian Thomas at (856) 339-2022.

Document Control Desk Page 2 LR-N12-0107

~UN 0"6 2012 I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, f~e:~

Site Vice President Hope Creek Generating Station Attachments (2)

w. Dean, Regional Administrator - NRC Region I J. Hughey, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek P. Mulligan, Manager IV, NJBNE Commitment Coordinator - Hope Creek PSEG Commitment Coordinator - Corporate LAR H12-02 LR-N12-0107 LICENSE AMENDMENT REQUEST (LAR) H12 APRM OPERABILITY REQUIREMENTS IN OPCON 5 Table of Contents
1. DESCRiPTION ................................................................................................................ 2
2. PROPOSED CHANGE .................................................................................................... 2
3. BACKGROUND .............................................................................................................. 3
4. TECHNICAL ANAL YSIS .................................................................................................. 4
5. REGULATORY ANALYSIS ............................................................................................. 8
6. ENVIRONMENTAL CONSIDERATION ......................................................................... 11
7. REFERENCES .............................................................................................................. 11 1 of 11 LAR H12-02 LR-N12-0107

1.0 DESCRIPTION

This license amendment request (LAR H12-02) proposes changes to Technical Specification (TS) 3/4.3.1, "Reactor Protection System Instrumentation," and TS 3/4.3.6, "Control Rod Block Instrumentation" by modifying the operability requirements for the average power range monitoring (APRM) instrumentation system. The proposed amendment would eliminate the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable in the "Refueling" Operational Condition (OPCON) 5, as described below.

TS Table 3.3.1-1, "Reactor Protection System Instrumentation" and TS Table 4.3.1.1-1, "Reactor Protection System Instrumentation Surveillance Requirements" would be revised by eliminating the APRM "Neutron Flux - Upscale, Setdown" and "Inoperative" function requirements in OPCON 5 (functional units 2.a and 2.d, respectively). Additionally, Note (d) of Table 3.3.1-1 would be revised to reflect the change to the APRM operability requirements. TS Table 3.3.6-1, "Control Rod Block Instrumentation" and TS Table 4.3.6-1, "Control Rod Block Instrumentation Surveillance Requirements" would be revised by eliminating the APRM "Inoperative" and "Neutron Flux - Upscale, Startup" function requirements in OPCON 5 (trip functions 2.b and 2.d, respectively).

In the Refueling operational condition these APRM reactor scram and control rod withdrawal block functions do not provide any meaningful core protection. The APRM system is designed to monitor reactor neutron flux in the power operating range and is only one part of the in-core Neutron Monitoring System (NMS). Other TS required neutron monitoring instrumentation, refueling interlocks and plant controls provide adequate monitoring and protection for reactivity events that could potentially occur in OPCON 5.

2.0 PROPOSED CHANGE

The proposed TS changes are described below and are indicated on the marked up TS pages provided in Attachment 2 of this submittal. No changes to the TS Bases are required.

2.1 TS Table 3.3.1-1, "Reactor Protection System Instrumentation" For Functional Unit 2, "Average Power Range Monitor," Table 3.3.1-1 currently requires that Functional Unit 2.a, "Neutron Flux - Upscale, Setdown" and Functional Unit 2.d, "Inoperative" be operable in Operational Condition 5. The proposed amendment would eliminate the requirements that these two APRM scram functions be operable in OPCON 5. Additionally, Note (d) of Table 3.3.1-1 would be revised to reflect the APRMs are not required to be operable.

2.2 TS Table 4.3.1.1-1, "Reactor Protection System Instrumentation Surveillance Requirements" For Functional Unit 2, "Average Power Range Monitor," Table 4.3.1.1-1 currently requires surveillances to be performed on Functional Unit 2.a, "Neutron Flux - Upscale, Setdown" and function 2.d, "Inoperative" in Operational Condition 5. The proposed amendment would eliminate the surveillance requirements in OPCON 5 for these two APRM scram functions.

2 of 11 LAR H12-02 LR-N12-0107 2.3 TS Table 3.3.6-1, "Control Rod Block Instrumentation" For Trip Function 2, "APRM," Table 3.3.6-1 currently requires that Trip Function 2.b, "Inoperative" and Trip Function 2.d "Neutron Flux - Upscale, Startup" be operable in Operational Condition 5. The proposed amendment would eliminate the requirements that these two control rod block functions be operable in OPCON 5.

2.4 TS Table 4.3.6-1, "Control Rod Block Instrumentation Surveillance Requirements" For Trip Function 2, "APRM," Table 4.3.6-1 currently requires surveillances to be performed on Trip Function 2.b, "Inoperative" and Trip Function 2.d "Neutron Flux - Upscale, Startup" in Operational Condition 5. The proposed amendment would eliminate the surveillance requirements in OPCON 5 for these two APRM control rod block functions.

3.0 BACKGROUND

The Neutron Monitoring System (NMS) monitors the neutron flux level in the reactor in three separate, overlapping ranges, all using in-core instrumentation systems (refer to Hope Creek Updated Final Safety Analysis Report (UFSAR) Figure 7.6-1). The system provides automatic core protection signals in the event of power transients. The NMS includes the Source Range Monitor (SRM) system, Intermediate Range Monitor (lRM) system, and power range monitoring 1. The power range monitoring is accomplished by the APRM system, which receives core flux level signals from the Local Power Range Monitors (LPRM). These systems are described below. Additional information on the safety related elements of the NMS are provided in Section 7.6 of the Hope Creek UFSAR.

SRM System The SRM subsystem is composed of four detectors that are inserted into the core during shutdown conditions to monitor neutron flux levels during shutdown, refueling, plant startup and low power operation. The SRM system monitors neutron flux from the fully shutdown condition (source level) through criticality to a neutron flux of approximately 5 x 108 n/cm2/sec. Per UFSAR Figure 7.6-1, the SRMs are withdrawn after overlap with the IRMs during startup. Although the SRMs are not safety-related, they are important to plant safety. The SRMs are required by TS 3/4.9.2 to be operational in OPCON 5. During refueling operations, the plant operators use the SRMs to ensure that neutron flux remains within an acceptable range. In OPCON 5 (and OPCON 2, Startup) the SRMs generate a control rod block signal for upscale, downscale, inoperative and detector not full in conditions (Refer to TS Table 3.3.6-1, Trip Function 3).

An Oscillation Power Range Monitor (OPRM) subsystem is also provided. This system detects power oscillations which can result from thermal-hydraulic reactor core instabilities, and provides alarms which alert the Control Room operator to their occurrence. The system is not required to be operable in OPCON 5.

3 of 11 LAR H12-02 LR-N12-0107 IRM System The IRM subsystem monitors neutron flux from the upper portion of the SRM range to the lower portion of the power range as shown on UFSAR Figure 7.6-1. The IRMs are designed to monitor neutron flux levels at eight local core locations and provide protection against local criticality events caused by control rod withdrawal errors. The IRMs monitor neutron flux levels from the upper portion of the SRM range (about 1 x 108 n/cm 2/sec, approximately 10-4 percent power) to the lower portion of the APRM range (> 15 percent of rated thermal power). They are normally fully inserted during startup and are withdrawn after the reactor mode selector switch is placed in "Run" (OPCON 1) The mode switch is placed in Run when the APRMs are on scale (4 to 12 percent rated thermal power), ensuring IRM/APRM overlap and continuity of neutron flux monitoring. The IRMs provide control rod block and scram functions at 108 and 120, respectively, of a 125 division scale.

The IRMs are required to be operable in OPCON 5 providing scram signals (refer to TS Table 3.3.1-1, Functional Unit 1) and control rod block signals (refer to TS Table 3.3.6-1, Trip Function 4).

APRM System The APRMs do not have in-core detectors of their own but receive input from the Local Power Range Monitor (LPRM) detectors which are located at various levels throughout the core. The APRMs monitor core power from about 1% of rated thermal power to greater than 100% of rated thermal power. The APRMs represent a core average power level while the IRMs and SRMs indicate a local power level. When the mode switch is in Run (OPCON 1), the APRM trip reference signal is provided by a signal that varies with recirculation flow. This provides a power following reactor scram setpoint. As power increases, the reactor scram setpoint also increases up to a fixed setpoint above 100 percent.

In OPCON 5, the APRMs operate in the setdown mode to provide a control rod block and scram function at 11 % and 14% core average power, respectively.

4.0 TECHNICAL ANALYSIS

The APRM system monitors neutron flux in the power operating range from approximately one percent to greater than rated thermal power. The system generates a scram signal at or below 120 percent of the rated thermal power during bulk neutron flux level transients. The system also generates a control rod withdrawal block signal to mitigate postulated single control rod withdrawal error events. Both the scram and rod block setpoints vary as a function of reactor recirculation flow when the reactor mode switch is in Run (OPCON 1). These APRM automatic protective functions prevent damage to the fuel for postulated reactivity insertion events during power operating conditions, such as the Control Rod Withdrawal Error event and the Control Rod Drop Accident.

The proposed amendment does not have any effect on the UFSAR analyses for these postulated at-power reactivity insertion events since the TS will continue to require that the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions remain operable when the reactor mode switch is in the Startup and Run positions.

4 of 11 LAR H12-02 LR-N12-0107 In the refueling operational condition (OPCON 5), the reactor mode switch is in the Shutdown or Refuel position 2 , the reactor coolant system temperature is less than 140°F, and the vessel head closure bolts are less than fully tensioned or the head is removed (refer to TS 1.29 and TS Table 1.2). In addition, all control rods are inserted (TS 3/4.9.3). TS, refueling interlocks and plant procedures allow only one control rod to be withdrawn at a time while the plant is in OPCON 5 and the mode switch is in Refuel. No control rods can be withdrawn when the mode switch is in "Shutdown,,3 4. The core loading pattern is designed to ensure that the core is subcritical by a specified margin with the most reactive control rod at the full out position (refer to TS 3/4.1.1). Withdrawal of one control rod would not cause criticality and the event would not be detected by the APRMs.

Since reactor neutron flux levels during refueling are below the APRM indicating range (1% of rated thermal power), the APRM system does not provide any meaningful core monitoring in OPCON 5. Since the APRMs operate in setdown mode in OPCON 5 (providing a control rod block and scram function at 11 percent and 14 percent rated thermal power, respectively), they also do not provide any meaningful core protection; the SRMs and IRMs are designed and calibrated to measure neutron flux at significantly lower levels than the APRMs as shown below:

NMS System RPS Scram Setpoint - Upscale Rod Block Setpoint - Upscale SRM NA (unless shorting links  ::;; 1.0 X 105 cps (less than removed, then at less than 0.01 % Rated Thermal power) 0.01 % Rated Thermal Power)

IRM (on lowest (most  ::;; 120/125 divisions of full scale  ::;; 108/125 divisions of full scale sensitive) range) (less than 0.01 % Rated Thermal (less than 0.01 % Rated Power) Thermal Power)

APRM (setdown mode)  ::;; 14% of Rated Thermal Power  ::;; 11 % of Rated Thermal Power The SRM system and the IRM system provide adequate neutron flux monitoring during refueling and automatically initiate protective actions (scram or control rod withdrawal block) when required in OPCON 5. Operability of the SRM and IRM systems is TS required in OPCON 5, as discussed in Section 3.0.

The possibility of inadvertent criticality due to a control rod withdrawal or fuel assembly insertion into the core is addressed in U FSAR Section 15.4.1.1. This possibility is minimized by the Refueling Interlocks (discussed later in this evaluation). If a core alteration event were to occur it would result in a change in local flux readily detected by the IRMs (and/or SRMs).

The IRMs are a safety-related subsystem of the NMS and are designed to indicate and respond to neutron flux increases at local core locations (the APRMs are designed to monitor and respond to a core average neutron flux level). The IRMs will generate a RPS scram or control rod block if neutron flux increases to the applicable setpoint. The IRM subsystem is designed and calibrated to respond to a neutron flux level that is significantly less than the flux level 2

With the exception of TS 3/4 10.3, "Shutdown Margin Demonstrations," which is discussed later in Section 4.0 3

In OPCON 5, a single control rod or control rod drive mechanism may be removed from the core in accordance with TS 3/4.9.10.1.

4 In OPCON 5 multiple control rods or control rod drive mechanisms may be removed from the core in accordance with TS 3/4.9.10.2. Multiple control rod removal requires that the four fuel bundles around each rod are removed prior to control rod removal, therefore there is no possibility of inadvertent criticality.

5 of 11 LAR H12-02 LR*N12-0107 monitored by the APRMs. During refueling, when the IRMs are on their most sensitive range, the IRMs will generate a scram signal at less than 0.01% core average power while the APRMs will generate a scram signal at 14% core average power. With the reactor mode switch in Refuel or Shutdown (or Startup) an IRM upscale or inoperative trip signal actuates a NMS trip of the RPS. Only one of the IRM channels must trip to initiate a NMS trip of the associated RPS trip channel.

With the mode switch in the Refuel (or Startup) position 5 , any of the following conditions initiate a control rod block (via Reactor Manual Control System (RMCS)6):

(1 ) Any SRM detector not fully inserted into the core when the SRM count level is below the retract permit level and any IRM range switch on either of the two lowest ranges. This ensures that no control rod is withdrawn unless all SRM detectors are correctly inserted when they must be relied on to provide the operator with neutron flux level information.

(2) Any SRM upscale level alarm - This ensures that no control rod is withdrawn unless the SRM detectors are correctly retracted during a reactor startup. The rod block setting is selected at the upper end of the range over which the SRM is designed to detect and measure neutron flux.

(3) Any SRM downscale alarm - This ensures that no control rod is withdrawn unless the SRM count rate is above the minimum prescribed for low neutron flux level monitoring.

(4) Any SRM inoperative alarm - This ensures that no control rod is withdrawn during low neutron flux level operations unless neutron monitoring capability is available.

(5) Any IRM detector not fully inserted into the core - This ensures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capability is available and correctly located.

(6) Any IRM upscale alarm - This ensures that no control rod is withdrawn unless the intermediate range neutron monitoring equipment is correctly upranged during a reactor startup. This rod block also provides a means to stop rod withdrawal in time to avoid conditions requiring RPS action (trip) in the event that a rod withdrawal error is made during low neutron flux level operations.

(7) Any IRM downscale alarm except when range switch is on the lowest range - This ensures that no control rod is withdrawn during low neutron flux level operations unless the neutron flux is being correctly monitored. This rod block prevents the continuation of a reactor startup if the operator upranges the IRM too far for the existing flux level.

Thus, the rod block ensures that the IRM is on scale if control rods are to be withdrawn.

(8) Any IRM inoperative alarm - This ensures that no control rod is withdrawn during low neutron flux level operations unless neutron monitoring capability is available.

5 With the mode switch in the "shutdown" position, no control rod can be withdrawn. This enforces compliance with the intent of the shutdown mode (UFSAR 7.7.1.1.2.2).

6 The RMCS provides the operator with the means to make changes in nuclear reactivity via the manipulation of control rods so that reactor power level and core power distribution can be controlled.

The RMCS includes the interlocks that inhibit rod movement (rod block) (UFSAR 7.7.1.1).

6 of 11 LAR H12-02 LR-N12-0107 In OPCON 5 the IRMs and SRMs act as a backup protection system to the Refueling Interlocks (Rls). Rls are required to be operational during refueling operations in OPCON 5 (TS 3/4.9.1).

The purpose of the Rls are to restrict the movement of the control rods and the operation of the refueling equipment to reinforce operational procedures that prevent inadvertent criticality during refueling operations. With the reactor mode switch in the Refuel position, the mode switch refueling interlock is required to be operable (TS 3/4.9.1). To prevent the possibility of loading fuel into a cell containing no control rod, it is required that all control rods be fully inserted when fuel is being loaded into the core. Rls will prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on the hoist. Similarly, if the refueling platform is over the core and fuel is on the hoist, control rod motion is blocked by the interlocks. When the refueling platform is not over the core, or fuel is not on the hoist, and the mode switch is in the Refuel position, only one control rod can be withdrawn. Any attempt to withdraw a second rod results in a rod block by the refueling interlocks. Since the core is designed to meet shutdown requirements with the highest worth rod withdrawn, the core remains subcritical even with one rod withdrawn. The design of the control rod does not physically permit the upward removal of the' control rod without the simultaneous or prior removal of the four adjacent fuel bundles. These rod block interlocks and refueling platform interlocks provide two independent levels of interlock action. The interlocks that restrict operation of the platform hoist and grapple provide a third level of interlock action.

A review of UFSAR Section 15.9, "Plant Nuclear Safety Operational Analysis (a System Level/Plant Nuclear Safety Operational Analysis)," was performed. The intent of the review was to identify any conditions when the APRMs were required to be operable to mitigate unacceptable consequences of inadvertent operational or transient conditions in OPCON 5.

The review concluded that if assumed operator errors occur, followed by postulated equipment malfunctions, there were adequate systems and interlocks without the APRMs to preclude potential inadvertent criticality or violation of a safety limit.

TS 3/4.10.3, Shutdown Margin Demonstrations TS 3/4.10.3 is an infrequently invoked special test performed during OPCON 5 if Shutdown Margin Demonstration (SDM) is required? During this test the mode switch is temporarily moved to Startup and the Rls are no longer available. However, performance of shutdown margin demonstrations in OPCON 5 requires additional restrictions in order to ensure that criticality is properly monitored and controlled. These additional restrictions are specified in the TS LCO:

a. The source range monitors are OPERABLE with the RPS circuitry "shorting links" removed per TS 3.9.2.
b. The rod worth minimizer (RWM) is OPERABLE per TS 3.1.4.1 and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.
c. The "rod-out-notch-override" control shall not be used during out-of-sequence movement of the control rods.

7 An example of when this T8 would be invoked would be if there was a fuel design change or methods change.

7 of 11 LAR H12*02 LR*N12*0107

d. No other CORE ALTERATIONS are in progress.

The controls and restrictions in place during this test are sufficiently robust even without the Rls.

Besides the same SRM and IRM protective actions, the additional SRM RPS trip is operable, the RWM is operable and programmed for the shutdown margin demonstration, use of the "rod-out-notch-override" control is prohibited, and no other core alterations are allowed. Therefore, during this infrequent operation, operability of the APRMs are also not required; they would not provide any meaningful core monitoring or protection.

Conclusion The APRMs are not necessary for safe operation of the plant while in OPCON 5. The APRM "Upscale" and "Inoperative" reactor scram and control rod withdrawal block functions need not be operable. Existing TS requirements, design features and procedural controls minimize the possibility of an inadvertent criticality event due to a control rod withdrawal error during refueling. The SRM and IRM systems provide adequate monitoring and core protection if such an event were to occur.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria In 10 CFR 50.36 establishes the requirements related to the content of TSs. TSs are required to include items in the following five categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires an LCO to be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The proposed changes to the APRM system operability requirements are consistent with 10 CFR 50.36 since the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions are not part of the primary success path to mitigate a design basis accident or transient when the plant is in the Refuel operating mode (OPCON 5).

The NRC has granted two previous amendments that are precedents for the proposed change:

1. Nine Mile Point Nuclear Station, Unit No.1-Issuance of Amendment Regarding Revisions to Average Power Range Monitor Instrumentation System Operability Requirements Technical Specification 3.6.2, Protective Instrumentation (TAC NO.

ME5010), dated October 31, 2011 [ADAMS ML112870012]

2. APRM Operability During OPCON 5 (TSCR NO. 90-02-0), Limerick Generating Station, Units 1 and 2 (TAC NOS. 76958/76959), dated July30, 1990 [ADAMS ML011550016]

5.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," PSEG Nuclear LLC (PSEG) requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS). The proposed amendment would change Technical Specification (TS) Section 3.3.1, "Reactor Protection System," and TS Section 3.3.6, "Control Rod Block Instrumentation" by modifying the operability 8 of 11 LAR H12-02 LR-N12-0107 requirements for the average power range monitoring (APRM) instrumentation system. The proposed amendment would eliminate the requirements that the APRM "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable in the "Refueling" Operational Condition (OPCON) 5.

PSEG has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three conditions set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The APRM system is not an initiator of or a precursor to any accident or transient. The APRM system monitors the neutron flux level in the power operating range from approximately one percent to greater than rated thermal power and initiates automatic protective actions for postulated at-power reactivity insertion events. Thus, the proposed changes to the TS operability requirements for the APRM system will not impact the probability of any previously evaluated accident.

The design of plant equipment is not being modified by the proposed amendment. The TSs will continue to require operability of the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions when the reactor is in the Startup and Run modes (OPCON 2 and OPCON 1) to provide core protection for postulated reactivity insertion events occurring during power operating conditions. Thus, the consequences of previously evaluated at-power reactivity insertion events are not affected by the proposed amendment.

The proposed elimination of the TS requirements that the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable when the reactor is in the Refueling mode (OPCON 5) also does not increase the consequences of an accident previously evaluated. The possibility of inadvertent criticality due to a control rod withdrawal error during refueling is minimized by design features and procedural controls that are not affected by the proposed amendment. Since the core is designed to meet shutdown requirements with the highest worth rod withdrawn, the core remains subcritical even with one rod withdrawn. Any attempt to withdraw a second rod results in a rod block by the Refueling Interlocks (RI). In addition, since reactor neutron flux levels during refueling are below the APRM indicating range, the APRM system does not provide any meaningful core monitoring or protection in the refueling operating condition (OPCON 5).

The source range (SRM) and intermediate range (IRM) neutron monitoring systems provide adequate neutron flux monitoring during refueling and automatically initiate protective actions (scram or control rod withdrawal block) when required during refueling.

Additionally, if the infrequently performed TS 3/4.10.3, "Shutdown Margin Demonstrations,"

is performed in OPCON 5, the additional controls and restrictions in place during this test are sufficiently robust even without the Rls when the mode switch is temporarily placed in Startup. In addition to the OPCON 5 SRM and IRM protective actions, the SRM RPS trip is made operable, the RWM is operable and programmed for the shutdown margin demonstration, use of the "rod-out-notch-override" control is prohibited, and no other core alterations are allowed. Therefore, during this infrequent operation, operability of the 9 of 11 LAR H12-02 LR-N12-0107 APRMs is not required as they would not provide any meaningful core monitoring or protection.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes to the TS operability requirements for the APRM system do not introduce any new accident precursors and do not involve any physical plant alterations or changes in the methods governing normal plant operation that could initiate a new or different kind of accident. The proposed amendment does not alter the intended function of the APRM system and does not affect the ability of the system to provide core protection for at-power reactivity insertion events. The other existing TS-required neutron monitoring systems (SRM and IRM) provide for core monitoring and protection in the refueling mode (OPCON 5). Additionally, if the infrequently performed TS 3/4.10.3, "Shutdown Margin Demonstrations" is performed in OPCON 5, the additional controls and restrictions in place during this test are sufficiently robust even without the Rls when the mode switch is temporarily placed in "Startup".

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

Margin of safety is related to the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed amendment does not alter setpoints or limits established or assumed by the accident analyses. The proposed TS changes to eliminate the requirements that the APRM system "Upscale" and "Inoperative" scram and control rod withdrawal block functions be operable when in OPCON 5 have no impact on the performance of the fission product barriers. These APRM functions do not provide any meaningful core monitoring or protection in the Refueling operating condition, including the infrequently performed special test TS 3/4.10.3. The other existing TS required neutron monitoring systems (SRM and IRM) provide for core monitoring and protection in the refueling mode (OPCON 5). In the Startup and Run modes the TSs will continue to require operability of these APRM functions to provide core protection for postulated reactivity insertion events occurring during power operating conditions, consistent with the plant safety analyses.

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

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

10 of 11 LAR H12-02 LR-N12-0107 5.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

None 11 of 11 LAR H12-02 LR-N12-0107 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License NPF-57 are affected by this change request Technical Specification 3/4.3.1, "Reactor Protection System Instrumentation" 3/4 3-2, 5 and 7 3/4.3.6, "Control Rod Block Instrumentation" 3/4 3-57 and 60

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2. Average Power Range MoRitor(e):

w

..... a. Neutron flux - Ups£AJe. S.t~R I

b. fl~ aiased Si.ulated lhe~'

Powe:- Upscale 1 Z

c. fixed Neutron flux - upscale 1 2
d. Jaoperat.hf *

~.~ .......-y""'""'~~~-~

3. Reactor Vessel Sh_ 0 -
1. Z(f)

~ "'-'"'"

2

-~~

I Pressure - fl~ SJh 1

):.

  • Reactor Vessel Wilter Lev. I .. low,

!II 4.

t
1. '2 2 1 I~ u\e' 3

.z 0

S. Main St... line Isolation Valve -

Closure l(g) .. ..

~

Ul

~.

  • TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYS~EM INSTRUMENTATION TABLE NOTATIONS (a) A ohannel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in th~ tripped condition provided at least one OPERAB~E channel in the same trip system is monitoring that parameter.

(b) This funotion shall be automatically bypassed wnen the reactor mode switch is in the Run position.

(c) Unless adequate shutdown margin has been demonstrated per Specifioation 3.1.1, the "shorting link.s" shall be removed from the Rl?S circuitry prior to and during the time any control rod is withdrawn*.

(d) The non-ooincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OP~RABLE Channels Fer the Trip System ar~

~6 IRMS and 2 SRMS.

(e) An APRM channel is inoperable i f there are less than :2 LPRM inputs per level O~ less than 14 LPRM inputs to an APRM channel.

(f) This function is not required to be OPERABLE When the reactor pressure vessel bead is removed per Specification 3.10.1.

(g) This function shall be aut.omatically bypassed when the reaotor mode switch is not in the Run position.

(b) This function is ~ot required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not r.equired.

(i} With any cont:tol rod withdrawn. Not applicable to control rods :removed per Specification 3.9.10.1 or 3.9.10.2.

(j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to THERMAL POWER lass than 24% of RATmO THERMAL POWER.

(k) Also actuates the EOC-RPT system.

  • Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
  • HOPE CREEK 3/4 3-5 Amendment No. 174

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDtTlONS FOR WHICH*

FUNCTIONAL UNII CHECK(m} TEST (In) CALIBRATION (a) (m) SURVEILLANCE REQUIRED

1. Intermediate Range Monitors:
a. Neutron Flux - High (b)  ::r, 2 3,4.5

~

b. Inoperative NA NA 2. 3, 4. 5
2. Average Power Range MonitorW:
a. Neutron Flux - Upscale, (b) (I}

Setdown 3.~

b. Flow Biased Simulated Thennal (g) {d} (e)QI)

Power-Upscale 1

c. Fixed Neutron Flux .. Upscale (d) 1
d. Inoperative NA NA 1,2.3*0
3. Reactor Vessel Steam Dome (k)

Pressure - High 1,2

4. Reactor Vessel Water Level- Low.

Level 3 (k) 1~ 2

5. Main steam Une isolation Valve -

Closure NA ~. 1

6. This item intentionally blank
7. Drywell Pressure - High (k}

1.2 HOPE CREEK 3f43-7 Amendment No. 187

TMLE 3.3.6"'1

% CONJ1OL ROO BLOCK INSTlUlENTAlIOII

.0 III n

IIHIIUI OPfWlf CtwUlELS mUCABlE::

OPfRATIOAAL

" .;10 III JII TI.'" fUNCTION PER TRIPRINCnON £ONOJTJONS . ACTION

~

1. ROD BlOCK tIIIlToa(a)
a. upsula 2 1- 60
b. Inoperative 2 1- 60
c. Downscale 2 1* 60
2. APRIl

\.

a.-fJow Biased Neutron flux -

k@

upscale 61

b. laoper..tive 61
c. OowRscaJe
d. Neutron flux - Upscale, Startup .

4 61 61

3. SOURCE BMGE fDtlTOftS
a. Detector not fun tA(b) 3 2 61 I ...... 2 5 61 tAt
b. Up5c~le(c) 3 2 61 w- 2 5 61 CA
c. lnoperative(c) 3 2 61 2 5 61
d. Oownsc:ale(d} 3 2 61 2 5 61
4. INTfHDIAT£ RAllGE JIlHITORS

. a. tiitecto.- not fUn 1'- ,; 2. 5 61

b. Upscale 2, 5 61 "
c. I~r.tirl) 6 2. 5 61
d. IJowlsQl. 6 2. 5 61
5. SCJtNI DISCfWIGf VOLUME
a. Water level-High {float SWltdl) 2 1. 2. 5-* 62 6.

a

  • upscale
b. Itqterative REACTOR COOLMT SYSTEM IECIICtilATION flOW 2

2 1

1 62 62 i c. Cellparater 2 1 62

'I

,,) . ' ' 7. RL\C....tllt~l)~ PqSIIlOH 2 J. 4 61

.i ".'""... .

~WITtH ~HtJTOOWN

TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK{f) TEST(f} CALIBRATION(a) (f) SURVEILLANCE REQUIRED

1. ROD BLOCK MONITOR
a. Upscale NA (e) 1+
b. Inoperative NA (e)

NA 1--

c. Downscale NA (e) 1*
2. APRM a Flow Biased Neutron Flux - Upscale NA
b. Inoperative
c. Downscale*

NA NA 1'~

V NA

d. Neutron Flux - Upscale, Startup NA
3. SOURCE RANGE MONlTORS
a. Detector not full in NA NA 2,5
b. Upscale NA 2,5
c. Inoperative NA NA 2,5-
d. Downscale NA 2,5
4. INTERMEDIATE RANGE MONITORS
a. Detector not full in NA NA 2,5
b. Upscale NA 2,5
c. Inoperative NA NA 2,5
d. Downscale NA 2,5
5. SCRAM DISCHARGE VOLUME
a. Water Level-High (Float Switch) NA 1,2,5**
6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW
a. Upscale NA 1
b. Inoperative NA NA 1
c. Comparator NA 1
7. REACTOR MODE SWITCH SHUTDOWN POSITrQN NA (el NA 3,4 HOPE CREEK 3/43-60 Amendment No. 187
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