LR-N17-0124, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control
| ML17265A847 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/21/2017 |
| From: | Carr E Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LAR H17-06, LR-N17-0124 | |
| Download: ML17265A847 (124) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 CPSEG Nuclear LLC 10 CFR 50.90 LR-N17-0124 LAR H17-06 SEP 2 1 2011 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:
Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Hope Creek Generating Station.
The proposed change replaces existing Technical Specifications (TS) requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.4.
Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only. contains the camera-ready TS pages.
PSEG requests approval of this license amendment request (LAR)in accordance with standard NRC approval process and schedule. Once approved, the amendment shall be implemented within 180 days. PSEG intends to use the enforcement discretion provided in Enforcement Guidance Memorandum EGM 11-003, "Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel," during the Spring 2018 Hope Creek refueling outage and until this license amendment request is approved and implemented.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.
There are no regulatory commitments contained in this letter.
10 CFR 50.90 LR-N17-0124 Page 2 SEP 2 1 2017 If you have any questions or require additional information, please contact Mr. Lee Marabella at (856) 339-1208.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on 7/'i//;7-(Date)
Sincerely,
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~~~
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Eric Carr Site Vice President Hope Creek Generating Station Attachments:
- 1. Description and Assessment
- 2. Proposed Technical Specification Changes (Mark-up)
- 3. Proposed Technical Specification Bases Changes (Mark-up) (for information only)
- 4. Camera-ready Technical Specification pages cc: Administrator, Region I, NRC Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Corporate Commitment Tracking Coordinator Hope Creek Commitment Tracking Coordinator
LR N17 0124 Attachment 1 Description and Assessment
LR N17 0124 Description and Assessment
1.0 DESCRIPTION
The proposed change replaces existing Technical Specifications (TS) requirements related to "operations which have the potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Hope Creek Safety Limit 2.1.4 (Improved TS based TSTF 542 Safety Limit 2.1.1.3). Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.
2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation PSEG Nuclear (PSEG) has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20, 2016, as well as the information provided in TSTF 542. PSEG has concluded that the justifications presented in TSTF 542 and the safety evaluation prepared by the NRC staff are applicable to Hope Creek Generating Station (Hope Creek) and justify this amendment for the incorporation of the changes to the Hope Creek TS.
The following Hope Creek TS reference or are related to OPDRVs and are affected by the proposed change:
1.0 Definitions 3/4.3.2 Isolation Actuation Instrumentation 3/4.3.3 Emergency Core Cooling System Actuation Instrumentation 3/4.3.7 Radiation Monitoring Instrumentation 3/4.5.2 ECCS Shutdown 3/4.5.3 Suppression Chamber 3/4.6.5.1 Secondary Containment Integrity 3/4.6.5.2 Secondary Containment Automatic Isolation Dampers 3/4.6.5.3.1 Filtration Recirculation and Ventilation System (FRVS) Ventilation Subsystem 3/4.6.5.3.2 FRVS Recirculation Subsystem 3/4.7.2.1 Control Room Emergency Filtration System 3/4.7.2.2 Control Room Air Conditioning (AC) System 3/4.8.1.2 AC Sources Shutdown 3/4.8.2.2 DC Sources - Shutdown 3/4.8.3.2 Electrical Power Systems Distribution Shutdown 2.2 Variations PSEG is proposing the following variations from the TS changes described in the TSTF 542 or the applicable parts of the NRC staffs safety evaluation. These variations do not affect the applicability of TSTF 542 or the NRC staff's safety evaluation to the proposed license amendment.
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LR N17 0124 The following variations are administrative and do not affect the applicability of TSTF 542 to the Hope Creek TS A.1. The Hope Creek TS utilize different numbering and titles than the BWR/4 Standard Technical Specifications (STS) on which TSTF 542 was based. Specifically, the following table shows the differences between the plant specific TS numbering and/or titles and the TSTF 542 numbering and titles. These differences are administrative and do not affect the applicability of TSTF 542 to the Hope Creek TS.
TSTF 542 STS Hope Creek TS Comments 1.1, Definitions 1.0 Definitions DRAIN TIME 1.11.1 - DRAIN TIME For Hope Creek, each definition is individually numbered.
3.3.5.1A, Emergency Core 3.3.3, Emergency Core Cooling System (ECCS) Cooling System Actuation Instrumentation Instrumentation 3.3.5.2A, Reactor 3.3.12, Reactor Pressure Pressure Vessel (RPV) Vessel (RPV) Water Water Inventory Control Inventory Control Instrumentation Instrumentation 3.3.6.1A, Primary 3.3.2, Isolation Actuation Containment Isolation Instrumentation Instrumentation 3.3.6.2A, Secondary 3.3.2, Isolation Actuation Containment Isolation Instrumentation Instrumentation 3.3.7.1A Main Control 3.3.7.1, Radiation Room Environmental Monitoring Instrumentation Control (MCREC) System Instrumentation 3.5, Emergency Core 3.5, Emergency Core Cooling Systems (ECCS), Cooling Systems (ECCS)
RPV Water Inventory and RPV Water Inventory Control, and Reactor Core Control, Isolation Cooling System (RCIC) 3.5.3, RCIC System 3.7.4, Reactor Core Isolation Cooling System 3.6.4.1, Secondary 3.6.5.1, Secondary Containment Containment Integrity 3.6.4.2, Secondary 3.6.5.2, Secondary Containment Isolation Containment Automatic Valves (SCIVs) Isolation Dampers 3.6.4.3, Standby Gas 3.6.5.3, Filtration, Treatment (SGT) System Recirculation and Ventilation System (FRVS) 3.7.4, Main Control Room 3.7.2.1, Control Room Environmental Control Emergency Filtration (MCREC) System System 2
LR N17 0124 TSTF 542 STS Hope Creek TS Comments 3.7.5, Control Room Air 3.7.2.2, Control Room Air Conditioning (AC) System Conditioning (AC) System 3.8.2, AC Sources 3.8.1.2, AC Sources Shutdown Shutdown 3.8.5, DC Sources 3.8.2.2, DC Sources Shutdown Shutdown 3.8.8, Inverters Shutdown N/A Hope Creek does not have an equivalent TS 3.8.10, Distribution 3.8.3.2, Distribution Systems Shutdown Shutdown 5.5.16, Setpoint Control N/A Hope Creek does not Program have an equivalent TS A.2. The Hope Creek TS also differ in format from the Standard Technical Specifications on which TSTF 542 was based. In general, the TS Limiting Condition for Operation (LCO), APPLICABILITY, ACTION and SURVEILLANCE REQUIREMENTS are provided in an outline format. Instead of a single table, the Hope Creek ECCS Actuation Instrumentation TS 3/4.3.3 utilizes three tables to display the following information:
Table 3.3.3 1 - List of trip functions, minimum operable channels, applicable Operational Conditions and Actions Table 3.3.3 2 - Associated trip setpoints and allowable values for the trip functions in Table 3.3.3 1 Table 4.3.3.1 1 - Associated surveillance requirements for the trip functions in Table 3.3.3 1 The proposed TS 3/4.3.12 presents the RPV Water Inventory Control Instrumentation requirements in a similar manner.
A.3. The Hope Creek Technical Specifications contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement Frequencies for Specifications 3/4.3.12 and 3/4.5.2 are "In accordance with the Surveillance Frequency Control Program."
A.4. PSEG has chosen to implement the Reactor Pressure Vessel Water Inventory Control (WIC) Instrumentation specification as TS 3/4.3.12 to avoid renumbering existing TS 3/4.3.3 through 3/4.3.11.
A.5. The STS Table 1.1 1 defines MODES of Operation for STS plants (1 through 5) while Hope Creek TS Table 1.2 defines OPERATIONAL CONDITIONS (1 through 5).
The differences in the definitions of OPERATIONAL CONDITIONS 4 and 5 vs.
MODES 4 and 5 are as follows:
Hope Creek has an Average Reactor Coolant Temperature limit of 140°F for OPERATIONAL CONDITION 5, Refueling, and STS does not.
Hope Creek has notes which describe allowances for repositioning the reactor mode switch and refers to Special Test Exception TS 3.10.1, 3.10.3 and 3.10.8 while STS does not.
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LR N17 0124 These differences have no impact on the TSTF 542 changes or the applicability of the model Safety Evaluation. Therefore, the STS MODES 4 and 5 and the Hope Creek OPERATIONAL CONDITIONS 4 and 5 are considered equivalent.
A.6. Proposed TS Table 3.3.12 1 for RPV Water Inventory Control Instrumentation presents the TS Actions in a manner consistent with the format of the current Hope Creek TS. TS ACTION 83 combines TSTF 542 TS 3.3.5.2 REQUIRED ACTIONs C.1 and E.1 in a single ACTION statement. Similarly, TS ACTION 84 combines TSTF 542 TS 3.3.5.2 REQUIRED ACTIONs D.1 and E.1, and TS ACTION 85 combines TSTF 542 TS 3.3.5.2 REQUIRED ACTIONS B.1 and B.2.
A.7. In Hope Creek TS Table 3.3.3 1 for ECCS Actuation Instrumentation, TS ACTION 32 applies only in OPERATIONAL CONDITIONS 4 and 5. Therefore, consistent with the removal of OPERATIONAL CONDITION 4 and 5 requirements from the ECCS Actuation Instrumentation TS, TS ACTION 32 is being deleted.
A.8. Proposed TS ACTION 3.3.12.a is being added to be consistent with the current Hope Creek instrumentation TS. With an RPV Water Inventory Control instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.12 2, TS ACTION 3.3.12.a would require the channel to be declared inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. The variation provides consistency with the existing Hope Creek instrumentation TS and does not affect the requirements of new TS ACTIONs 83, 84 and 85, which are consistent with TSTF 542.
A.9. Note (c) is being added to TS Table 3.3.12 1 for the Core Spray Reactor Vessel Pressure Low (Permissive) trip function. Note (c) clarifies that the trip function is only required for Divisions 1 and 2. The note is included in the current TS Table 3.3.3 1 for ECCS Actuation Instrumentation.
A.10. Proposed SR 4.5.2.5 would be modified by a note permitting automatic valves capable of automatic return to their ECCS position when an ECCS signal is present to be in position for another mode of operation. The note is included in current SR 4.5.1.a.1.b and is consistent with the STS Bases for SR 3.5.2.5.
A.11 PSEG is proposing to add a Note (d) to TS Table 3.3.12 1 (RPV WIC Instrumentation) function to clarify the intent of allowing credit for an OPERABLE LPCI subsystem when it is aligned and operating in the decay heat removal mode of RHR. This is appropriate since the associated RHR pump minimum flow valve (while operating in the decay heat removal mode) is closed and deactivated to prevent inadvertent vessel drain down events. Because the minimum flow valve is closed and deactivated, the associated TS Table 3.3.12 1 Function 2.b would not be required to be OPERABLE. Without the note, TS 3.3.12 ACTION 84 would require that the associated RHR pump be declared inoperable, which would be contrary to the intent of the existing Note to LCO 3.5.2.b.2 which allows a LPCI subsystem to be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.
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LR N17 0124 A.12 During the development of this LAR to adopt TSTF 542, Rev.2, an administrative error was identified within the TS Index. As part of Hope Creek License Amendment No. 146, TS Definition 1.13, EMERGENCY CORE COOLING SYSTEM (ECCS)
RESPONSE TIME, was relocated to TS page 1 3. The TS Index is being revised to reflect this change, which is administrative in nature and does not affect the applicability of TSTF 542 to the Hope Creek TS.
The following variations describe proposed changes to current Hope Creek TS requirements due to plant specific TS requirements related to OPDRVs, plant specific functions that support automatic and manual initiation of ECCS subsystems, and plant specific systems that provide Secondary Containment, and Standby Gas Treatment system functions. The proposed changes differ from the Standard Technical Specifications on which TSTF 542 was based, but are encompassed in the TSTF 542 justification.
B.1 To align with NUREG 1433, Rev. 4, and consistent with TSTF 542, Rev. 2, PSEG proposes to revise TS 3.5.3, "Suppression Chamber," to remove TS requirements associated with OPERATIONAL CONDITIONS 4 and 5 since they are redundant to the requirements and intent of the newly proposed TS Section 3.5.2, "RPV Water Inventory Control."
Specifically, TS LCO 3.5.3.b is addressed in newly proposed TS LCO 3.5.2 and its associated surveillance requirements 4.5.2.2 and 4.5.2.3.
In OPERATIONAL CONDITIONS 4 and 5, TS LCO 3.5.3.b requires a minimum indicated suppression chamber water level of 5.0" except that the suppression chamber level may be less than the limit, provided that:
- 1. No operations are performed that have a potential for draining the reactor vessel,
- 2. The reactor mode switch is locked in the Shutdown or Refuel position,
- 3. The condensate storage tank contains at least 135,000 available gallons of water, and
- 4. The core spray system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water through the spray sparger to the reactor vessel.
With LCO 3.5.3.b not met, TS Action 3.5.3.b requires CORE ALTERATIONS and operations that have a potential for draining the reactor vessel to be suspended, the reactor mode switch to be locked in the Shutdown position, and SECONDARY CONTAINMENT INTEGRITY to be established within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The minimum required suppression chamber water level in TS LCO 3.5.3.b and SR 4.5.3.2 is made redundant by proposed SRs 4.5.2.2 for LPCI subsystems and 4.5.2.3 for Core Spray subsystems, which are consistent with TSTF 542. Removal of the TS Action 3.5.3.b requirement to suspend operations that have a potential for draining the reactor vessel is consistent with the proposed addition of DRAIN TIME requirements to TS LCO 3.5.2. The TS Action 3.5.3.b requirement to establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is made redundant by proposed TS Actions 3.5.2.c and 3.5.2.d for DRAIN TIMES not meeting LCO 3.5.2, which are consistent with TSTF 542.
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LR N17 0124 TS Action 3.5.3.b is currently modified by a note stating the suppression chamber is not required to be OPERABLE in OPERATIONAL CONDITION 5, provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.
Removal of this note is consistent with the proposed addition of DRAIN TIME requirements to TS LCO 3.5.2, and the removal of the similar allowance from STS 3.5.2 Applicability for MODE 5.
Removal of the TS 3.5.3 requirements related to core alterations and reactor mode switch position are addressed in Item D.2 below.
B.2. Current Hope Creek TS Table 3.3.3 1 includes Trip Functions 1e and 1f, Core Spray Pump Start Delay Time - Normal Power and Core Spray Pump Start Delay Time -
Emergency Power, which are required to be OPERABLE in Operating Conditions 1, 2, 3, 4, and 5. The purpose of the delay times is to stagger the automatic start of the Core Spray pumps, limiting starting transients on their associated 4.16 kV emergency buses. This staggering is unnecessary for manual operation. Therefore, these functions applicable in Operating Conditions 4 and 5 are being removed from Table 3.3.3 1 and are not being included in the proposed TS Table 3.3.12 1 for RPV Water Inventory Control Instrumentation. This is consistent with the intent of TSTF 542 and a similar change made to STS Function 2.f, "Low Pressure Coolant Injection Pump Start - Time Delay Relay."
B.3 The Hope Creek Core Spray Reactor Vessel Pressure Low (Permissive) is initiated from four pressure transmitters each in two divisions. The low pressure permissive for each division is provided in one out of two twice logic. Manual initiation of each Core Spray subsystem requires the low pressure permissive from the associated division.
The Hope Creek LPCI Reactor Vessel Pressure Low (Permissive) is initiated from a pressure switch downstream of each LPCI injection valve. Manual initiation of each LPCI subsystem requires the low pressure permissive from the associated pressure switch.
Note (a) is therefore being added to proposed TS Table 3.3.12 1 for the Core Spray and LPCI Reactor Vessel Pressure Low (Permissive) trip functions. Note (a) states that the Minimum Operable Channels per Trip Function requirement applies to those functions associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, RPV Water Inventory Control.
B.4 The Filtration, Recirculation, and Ventilation System (FRVS) consists of two subsystems that are required to perform post accident, safety related functions. The FRVS Recirculation System recirculates the Reactor Building air through filters for cleanup. This subsystem is the initial cleanup system before discharge is made via the FRVS ventilation subsystem. The FRVS Ventilation System maintains the Reactor Building at a negative pressure with respect to the outdoors. A single FRVS Ventilation Subsystem is capable of maintaining the Reactor Building at a negative pressure with respect to the environment and filter gaseous releases in OPERATIONAL CONDITIONS 4 and 5. FRVS requirements are contained in 6
LR N17 0124 proposed TS Actions 3.5.2.c and 3.5.2.d and are consistent with the requirements for the Standby Gas Treatment system in TSTF 542.
The following variations are TS changes called out in the TSTF 542 document that do not apply to Hope Creek TS.
C.1. TSTF 542 T3.3.6.1 1 Function 6b, Primary Containment Isolation Instrumentation Reactor Vessel Water Level - Low Low, Level 2 removes annotation referring to OPDRVs. The corresponding Hope Creek TS T3.3.2 1 Trip Function 7a is applicable in OPCONs 1, 2 and 3 only and has no existing reference to OPDRVs.
Therefore no corresponding changes are proposed.
C.2. TSTF 542 T3.3.7.1 1, Function 4, Refueling Floor Area Radiation - High, removes annotation referring to OPDRVs. The Hope Creek TS have no corresponding TS Trip Function, therefore no corresponding changes are proposed.
C.3. TSTF 542 TS 3.6.1.3 Condition H and Required Action H1 removes reference to OPDRVs for Primary Containment Isolation Valves (PCIVs). The corresponding Hope Creek TS 3.6.3 is applicable in OPCONs 1, 2 and 3 only and has no existing reference to OPDRVs. Therefore no corresponding changes are proposed.
C.4. TSTF 542 TS 3.8.8, Inverters - Shutdown, Required Action A.2.3 removes reference to OPDRVs. The Hope Creek TS control the corresponding inverters under TS 3.8.3.1, Distribution - Operating which is applicable in OPCONs 1, 2 and 3 only.
Therefore no corresponding changes are proposed.
C.5. Hope Creek TS contain no reference to a Setpoint Control Program as described in the TSTF 542 change to Section 5.5.16. Therefore no corresponding change to the Hope Creek TS is proposed.
The following variations describe proposed changes to additional requirements currently in the Hope Creek TS for ECCS subsystems in OPERATIONAL CONDITIONS 4 and 5 to better align the Hope Creek TS with TSTF 542.
D.1. In alignment with TSTF 542, Rev. 2, Proposed Safety Basis (Section 3.1.2), the existing Hope Creek TS 3.5.2 requirement to suspend core alterations as an action for ECCS inoperability is no longer warranted since there are no postulated events associated with core alterations that are prevented or mitigated by the proposed RPV water inventory control requirements. In addition, loss of RPV inventory events are not initiated by core alteration operations. Refueling Limiting Conditions for Operation (LCOs) 3.9.1, Reactor Mode Switch, 3.9.2, Instrumentation, 3.9.3, Control Rod Position, and 3.9.8, Water Level Reactor Vessel, provide requirements to ensure safe operation during core alterations, including required water level above the RPV flange. Therefore, PSEG proposes to delete TS 3.5.2, Action 'b' in its entirety, including the action relating to core alterations.
D.2. To align with NUREG 1433, Rev. 4, and fully implement TSTF 542, Rev. 2, PSEG proposes to revise TS 3.5.3, "Suppression Chamber," to remove TS requirements associated with OPERATIONAL CONDITIONS 4 and 5. As discussed above in Item B.1, the requirements in TS 3.5.3 related to OPDRVs and Actions to suspend 7
LR N17 0124 OPDRVs are redundant to the requirements and intent of the newly proposed TS Section 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control (WIC)."
LCO 3.5.3.b also requires the reactor mode switch to be locked in the Shutdown or Refuel position when suppression chamber water level is less than TS limits in OPERATIONAL CONDITIONS 4 and 5. By definition, the reactor mode switch is in Shutdown or Refuel in OPERATIONAL CONDITIONS 4 and 5. The requirement to lock the mode switch in either position is an administrative control, rather than an element in the lowest functional capability or performance levels of equipment required for safe operation of the facility required to be included in TS.
The TS Action 3.5.3.b requirement to lock the reactor mode switch in the Shutdown position is not required because it does not provide compensatory measures for suppression chamber water level less than TS limits. The TS Action 3.5.3.b requirement to suspend core alterations is not required because Refueling Operations LCOs 3.9.1, Reactor Mode Switch, 3.9.2, Instrumentation, 3.9.3, Control Rod Position, and 3.9.8, Water Level Reactor Vessel, provide requirements to ensure safe operation during CORE ALTERATIONS.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis PSEG requests adoption of TSTF 542 "Reactor Pressure Vessel Water Inventory Control,"
which is an approved change to the Standard Technical Specifications (STS), into the Hope Creek Technical Specifications (TS). The proposed amendment replaces the existing requirements in the Technical Specifications (TS) related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.
PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards 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 proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. Draining of RPV water inventory in OPCON 4 (i.e., cold shutdown) and OPCON 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in OPCON 4 or OPCON 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.
The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an 8
LR N17 0124 unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.
The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in OPCONs 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in OPCON 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in OPCONs 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.
The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.
The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
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LR N17 0124
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.4. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense in depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, PSEG concludes that the proposed change presents no 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.
4.0 ENVIRONMENTAL EVALUATION The proposed change 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.
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LR N17 0124 Attachment 2 Proposed Technical Specification Changes (Mark up)
The following Technical Specification pages for Renewed Facility Operating License NPF 57 are affected by this change request:
i B3/4 3 8**
x 3/4 3 115*
xi 3/4 5 1 xviii 3/4 5 2 xix 3/4 5 3 12 3/4 5 4 3/4 3 11** 3/4 5 5 3/4 3 16a 3/4 5 6 3/4 3 28** 3/4 5 7 3/4 3 31 3/4 5 8 3/4 3 33 3/4 5 9 3/4 3 34 3/4 6 47 3/4 3 35 3/4 6 49 3/4 3 39 3/4 6 51 3/4 3 40 3/4 6 52a 3/4 3 63** 3/4 7 6 3/4 3 64 3/4 7 6a 3/4 3 66** 3/4 7 8a 3/4 3 67 3/4 8 11 3/4 3 111* 3/4 8 17 3/4 3 112* 3/4 8 21**
3/4 3 113* 3/4 8 22**
3/4 3 114* 3/4 8 23
- New TS Pages
- Information Only TS Pages
1.11.1 DRAIN TIME.......................................................................1-2a 1-2a 1-3
3/4.3.12 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL INSTRUMENTATION.3/4 3-111 Table 3.3.12-1 RPV Water Inventory Control Instrumentation ......................................................3/4 3-112 Table 3.3.12-2 RPV Water Inventory Control Instrumentation Setpoints ......................................3/4 3-114 Table 4.3.12.1-1 RPV Water Inventory Control Instrumentation Surveillance Requirements .......3/4 3-115
(ECCS) AND RPV WATER INVENTORY CONTROL RPV WATER INVENTORY CONTROL
3/4.3.12 RPV WATER INVENTORY CONTROL INSTRUMENTATION...............................................B 3/4 3-13
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- "4&4
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TABLE 3.3.2-1 (Continued)
. When handling recently irradiated fuel in the secondary containment and during operations '.lith a potentiaL for draining the reactor
¥eS-&e-1-,
- When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the Norm position, H Below 20% of RATED THERMAL POWER the ~ain Steamline Radiation Monitor setpoints shall not exceed the values determined using normal full power background radiation levels witb the hydrogen water chemistry (HWC) system shut down. After reaching 20% of RATED THERMAL POWER the normal full power background radiation level and associated trip setpoints may be increased to levels pzeviously measured during full power operation with hydrogen injection. Prior to decreasing below 20%
of RATED THERMAL POWER the background level and associated setpoint shall be returned to the normal full power values. If the Main Steamline Radiation Monitor setpoints have been increased for Ewe operation and a power reduction event occurs so that the reactor power is below 20% of RATED THERMAL POWER without the required setpoint change, control rod motion shall be suspended (except for scram or other emergency actions) until the necessary setpoint adjustment is made.
(a) A channel 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 the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
(b) Also trips and isolates the mechanical vacuum pumps.
(c) Also starts the Filtration, Recirculation and Ventilation System (FRVS) ,
(d) DELETED (e) Sensors arranged per valve group, not pr trip system.
(f) Closes only RWCU system isolation valve(s) HV-FOOl and HV-F004.
(g) Requires system stearn supply pressure~low coincident with drywell pressure-high to close turbine exhaust vacuum breaker valves.
(h) Manual isolation closes HV-FOOB only, and only following manual or automatic initiation of the RCIC systen.
(i) Manual isolation closes HV-F003 and HV-F042 only, and only following manual or automatic initiation of the EFCI system.
(j) Trip functions common to RPS instrumentation.
HOPE CREEK 3/4 3-16a Amendment No. ~
No changes on this page - provided for information only TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (el TEST (el CALIBRATION (el SURVEILLANCE REQUIRED
- 1. PRIMARY CONTAINMENT ISOLATION
- 1) Low Low, Level 2 1,2,3
- 2) Low Low Low, Level 1 1,2,3
- b. Drywell Pressure - High 1,2,3
- c. Reactor Building Exhaust Radiation -
High 1,2,3 (a)
- d. Manual Initiation NA NA 1,2,3
- 2. SECONDARY CONTAINMENT ISOLATION
Low Low, Level 2 1,2,3 and *
- b. Drywell Pressure - High 1,2,3
- c. Refueling Floor Exhaust Radiation - High 1,2,3 and *
- d. Reactor Building Exhaust Radiation -
High 1,2,3 and *
(a)
- e. Manual Initiation NA NA 1,2,3 and *
- 3. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level - 1,2,3 Low Low Low, Level 1
- b. Main Steam Line Radiation - High, High 1,2,3
- c. Main Steam Line Pressure - Low 1
- d. Main Steam Line Flow - High 1,2,3
- e. Condenser Vacuum - Low 1,2**,3**
- f. Main Steam Line Tunnel Temperature-High NA 1,2,3 (a)
- g. Manual Initiation NA NA 1,2,3 HOPE CREEK 3/43-28 Amendment No. 187
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. . TABLE 3.3.3-1 (Cont'd) .
EMfBGENCY CORE COOLING SYSTEM ACTUATION Ir~STRUMENTAIION
. MINIMUM OPERABLE CHANNELS PER , APPll CABLE 5""0 TRIP OPERATIONAL rr1 .TRIP FUNCTION FUNCTION ( a) CONDITIONS ACTION n
- a 4. AUTOMATIC DEPRESSURIZATION SYSTEH#l rt1
- .-:: e. . RHR LPCl Mode Pump Discharge Pressure - High
. (Penl1ss1ve) . , 2/pump 1', 2, 3 31
- f. Reactor Vessel water level - Low, Level 3 (Permissive) 2 1, 2, 3 31
- g. ADS Orywell Pressure Bypass Timer . 4 1, 2, 3 31
- h. ADS Manual Inhibit Switch . 2 1, 2, 3 31
- 1. Manual Initiation' ' 4 1,2, 3, 33 MINIMUM APPLICABLE,
, TOTAL NO. CHANNELS CHANNELS , OPERATIONAL Of CHANNElSlb) TO TRIP(bl OPERABLE< h} CONOIIIONS ACTION
- 5. LOSS OF POWER
- 1. 4.16 kvEBergency Bus Under-voltage (Loss of Voltage) 4/bus 2/bus 3/bus 1, 2, 3, 4**, 5** 36 w 2. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 2/source/ 2/source/ 2/source/ 1, 2, 3, 4**, 5** 36 w bus bus bus
....w (a) Achanne. .ay be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip systl!lli ... the tripped condition provided at least one OPERABLE channel in the *same trip systell.isllOnitoring that parameter. .
(b) Also actuates the associated ellerg~ncy'diesel generators. (c) One trip syst~. Provides signal toHPCI pump suction' valve only. (d) Provides a signal to trip HPCI pump turbine only. , . (el In divisions 1 and 2, the two sensors are, associated with each pump and valve co..,ination. In divisions 3 and 4, the two sensors are associa.ted wUheach PUIIP only. (f) Division 1 and 2 Qnly. (g) In divisions 1 and 2, manual initiat10n is associated with each pump and valve combination; in div,isions 3 and. 4~ Ilanualinitiation ,is associated with eachpUllp only.
~ (h) Each voltage detector is a channel. . '
- t (i) Start t111e delay 15 applicable to LPCIPump C and Donly~
f
- t When thesystell 15 I"e~uil"ed ta he OPERABLE pel" Sp'edficatiaR3.5.~Deleted ;
Required'when ESf equipment is required to be OPERABLE. , . t"+
. I Hot required to be OPERABLE when reactor steam ,donie pressure is less than or equal to 200 psig.* .o Not required to be OPERABLE when reactor steam dome pressure is less than or equal to IOOpsig.
Z II 0'\ N
TABl; 3.3.3-1 (Continued) EMERGENCX kQ8EJ:OQLING SYST~M ACTUATION I~STRUMENIAIIQN
," AtIIQN ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. With one channel inop~rable, place the inoperable channel in the tripped condition within 24 hours or declare the associated system inoperable.
, b. ' With more than one channel inoperable, decl~re the associated system inoperable.
ACTION 31 - With the number of OPERABLE channels less than required by the M1nimumOPERABlEChannels per Trip Function requirement, declare 't the associated ECeS in~perable within 24 hours.
- ACTI ON 32 - '. Wi ttl the Rumber- of OPERA8LE channels 1ess than requ1 red b;y the H1R1mum OPERA8LE Chalm, e:l 5 per Tr1 p ~wnctlon requi rement, pl ace the inoperable chaRnel in the tripped condition within 24 heurs ' ,
I ACTION 33 ~ Wi th the number of OPERABLE charm~ 1s less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore
~ Deleted I the inoperable channel to OPERABLE status within 24 ,hours or declare the associated EeCS inoperable.
ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. For one ,channel inoperable, place the inoperable channel 1n the tripped condition within 24 hours or declare the HPCI system inoperable.
- b. With more than one channel inoperable, declare the HPCl system inoperable.
ACTION 35 - With the number of OPERABLE channel s less'than required by' the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel 1n the tripped condition within "I 24 hours or declare the HPCI system inoperable. ACTION 36 - With the number of OPERABLE channels one less than the Total Number of Channels. place the inoperable channel 1n the tripped condition within 1 hour; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST. ACITON37 - "With the number of 'OPERABLE ,channels less than required by the Minimum OPERABLE channels per Trip Function requirement, open the m1nimumflow bypass valve within one hour. Restore the
.. inoperable channel to OPERAB~E status within 7 days or declare
.. the aS$ociated ECCS inoperable
- HOPE CREEK 3/4~ 3-35 Amendment No. Q2
TABLE 4.3.3.1-1. EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE TRIP FUNCTION CHECK (a) TEST (a) CALl6RAUQr-L(a) REQUIRED
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Water Level- Low Low Low. Level 1 ".
1,2,3~
- b. Drywell Pressure - High 1,2,3
- c. Reactor Vessel Pressure - Low 1,2,3.~
- d. Core Spray Pump Discharge Flow - Low (Bypass) 1,2.3,~
- e. Core Spray Pump StartTime Delay - Normal Power NA 1,2,3,-4~
- f. Core Spray Pump Start Time Delay - Emergency Power NA 1,2,3,-4~
- g. Manual Initiation NA NA 1,2,3,-4~
- 2. LOW PRESSURE COOLANT INJECTlON MODE OF RHR SYSTEM
- a. Reactor Vessel Water Level- Low Low Low, Level 1 1,2,3,~
- b. Orywell Pressure - High 1,2,3
- c. Reactor Vessel Pressure - Low (Permissive) 1,2,3,~
- d. LPCI Pump Discharge Flow - Low (Bypass) 1,2,3,~
- e. LPCI Pump Start Time Delay - Normal Power NA 1,2.3,~
- f. Manual Initiation NA NA 1,2.3,~
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM#
- a. Reactor Vessel Water Level - Low Low, Level 2 1,2,3
- b. Drywell Pressure - High 1,2.3
- c. Condensate Storage Tank Level- Low 1,2.3
- d. Suppression Pool Water Level - High 1,2,3
- e. Reactor Vessel Water Level- High, LevelS 1,2,3
- f. HPCI Pump Discharge Flow - Low (Bypass) 1,2,3
- g. Manual Initiation NA NA 1,2,3 HOPE CREEK 3/43-39 Amendment No. 481
TABLE 4.3.3.1-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE TRIP FUNCTION CHECK (al TEST (a) CALIBRATION (al REQUIRED
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM##
- a. Reactor Vessel Water Level- Low Low Low, Level 1 1,2,3
- b. Drywell Pressure - High 1,2,3
- c. ADS Timer NA 1,2,3
- d. Core Spray Pump Discharge Pressure - High 1,2,3
- e. RHR LPCI Mode Pump Discharge Pressure - 1,2,3 High
- f. Reactor Vessel Water Level - Low, Level 3 1,2,3
- g. ADS Drywell Pressure Bypass Timer NA 1,2,3
- h. ADS Manual Inhibit Switch NA NA 1,2,3
- i. Manual initiation NA NA 1,2,3
- 5. LOSS OF POWER
- a. 4.16 kv Emergency Bus Under-voltage (Loss NA NA of Voltage) 1, 2, 3, 4**, 5**
- b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 1, 2, 3,4**,5**
(a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
* 'Nhcn the systcm is reql:lifed to be OPERABI::E per Specification 3.5.2. Deleted ** Required OPERABLE when ESF equipment is required to be OPERABLE. # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig. ## Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
HOPE CREEK 3/43-40 Amendment No. 481
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TABLE 3.3.7.1-1 (Continued) RADIATION MONITORING INSTRUMENTATION
'TABLE NOT~TION tWhen recen-tJ,y izr.adiated fuel. .is baing handled. .in the sacanda1y conbrlmRent and d.u.rinq nparaUcm.s ~
t:ho peteDt ; aJ f~ draining' the Z'Elaet-.r:m _ssel._
**Activates control roam emergency filtration system.
~~*When the offgas treatment system is operating. fiW1th fuel in the new ~uel-storaqe vault.
~JWith fuel in the spent fuel storage pool.
(a)Alarm only. (b)A1aLm setpoint to be set. in accordance with Speci£ication 3.11.2.7. HOPE CREEK 3/4. 3-64 Amendment No. 159 f
1RFKDQJHVRQWKLVSDJHSURYLGHGIRULQIRUPDWLRQRQO\ TABLE 4.3.7.1.1, (Continued) RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
#With fuel in the new fuel storage vault.
- With fuel in the spent fuel storage pool.
*When recently irradiated fuel is being handled in the secondary containment and during operations 'Nith the potential for draining the reactor vessel.
- When the offgas treatment system is operating.
HOPE CREEK 3/43-67 Amendment No. 48=f
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,16(57 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION (Continued) ACTION:
- c. With DRAIN TIME < 36 hours and 8 hours, within 4 hours:
- 1. Verify secondary containment boundary is capable of being established in less than the DRAIN TIME, AND
- 2. Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME, AND
- 3. Verify one Filtration, Recirculation and Ventilation (FRVS) ventilation unit is capable of being placed in operation in less than the DRAIN TIME.
Otherwise, immediately initiate action to restore DRAIN TIME to 36 hours.
- d. With DRAIN TIME < 8 hours, immediately:
- 1. Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for 36 hours***
AND,
- 2. Initiate action to establish secondary containment boundary, AND
- 3. Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room, AND
- 4. Initiate action to verify one FRVS ventilation unit is capable of being placed in operation.
Otherwise, immediately initiate action to restore DRAIN TIME to 36 hours.
- e. With DRAIN TIME < 1 hour, immediately initiate action to restore DRAIN TIME to 36 hours.
- Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.
EMERGENCY CORE COOLING SYSTEMS (ECCS)AND RPV WATER INVENTORY CONTROL SURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance ~ement 4.5.1. 4.5.2.2 The core spray system shall be determined OPERABLE in accordance ..... ith tAs Sup/eiHance F-requency Control Program by verifying the condensate storage tank reqldired volldme when the condensate storage tank is required to be OPERABLE per Specification 3.5.2.a.2.b. HOPE CREEK 3/4 5-7 Amendment No. m
,16(57 4.5.2.1 Verify DRAIN TIME 36 hours in accordance with the Surveillance Frequency Control Program. 4.5.2.2 Verify, for a required low pressure coolant injection (LPCI) subsystem, the suppression chamber indicated water level is > 5.0 inches in accordance with the Surveillance Frequency Control Program. 4.5.2.3 Verify, for a required Core Spray (CS) subsystem, the Suppression chamber indicated water level is > 5.0 inches or condensate storage tank contains at least 135,000 available gallons of water in accordance with the Surveillance Frequency Control Program. 4.5.2.4 Verify, for the required ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve in accordance with the Surveillance Frequency Control Program. 4.5.2.5 Verify, for the required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position, in accordance with the Surveillance Frequency Control Program.# 4.5.2.6 Operate the required ECCS injection/spray subsystem through the recirculation line for 10 minutes, in accordance with the Surveillance Frequency Control Program. 4.5.2.7 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal, in accordance with the Surveillance Frequency Control Program 4.5.2.8 Verify the required ECCS injection/spray subsystem actuates on a manual initiation signal, in accordance with the Surveillance Frequency Control Program.##
- Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
- Vessel injection/spray may be excluded.
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,EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTRO SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to:
- a. 74.5" in accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITIONS 1, 2, and 3.
IDeleted~", §.O" iR a*** ,gaR"" with tho SY",oiliaRs. F",qYORSY GORtrol ~",g"'m iR OPERATIONAL CONDITIONS 4 and 5*. 4.5.3.2 'Nith the sblppFeSsiElR chamber level less thaR the abEl\'!e limit ElF drained in OPERATIONAL CONDITION 4 ElF 5*, in accElrdance with the Surveillance Frequency Contrel Pregram:
.a, Verify the reqbliFed conditiElns of Specification 3.5.3.b to be satisfied, or
-9: Verify footnote conditions
- to be satisfied.
HOPE CREEK 3/4 5-9 Amendment No . ..:j.g:;z
PLANT SYSTEMS 3/4.7.2 CONTROL ROOM SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2.1 Two control room emergency filtration system subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3
- 1. With one control room emergency filtration subsystem inoperable for reasons other than Condition a.2, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With one or more control room emergency filtration subsystems inoperable due to an inoperable control room envelope (CRE) boundary ,
- a. Immediately, initiate action to implement mitigating actions; and
- b. Within 24 hours, verify mitigating actions ensure CRE occupant exposures to radiological and chemical hazards will not exceed the limits and actions to mitigate exposure to smoke hazards are taken; and
- c. Within 90 days, restore the CRE boundary to operable status; Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION *:
- 1. With one control room emergency filtration subsystem inoperable for reasons other than Condition b.3, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the pressurization/recirculation mode of operation.
- When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.
The main control room envelope (CRE) boundary may be opened intermittently under administrative control. HOPE CREEK 3/4 7-6 Amendment No. 191
PLANT SYSTEMS CONTROL ROOM AIR CONDITIONING (AC) SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2.2 Two control room AC subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3:
- 1. With one control room AC subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With two control room AC subsystems inoperable:
- a. Verify control room air temperature is less than 90°F at least once per 4 hours; and
- b. Restore one control room AC subsystem to OPERABLE status within 72 hours.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION *:
- 1. With one control room AC subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days; or place the OPERABLE control room AC subsystem in operation; or immediately suspend movement of recently irradiated fuel assemblies in the secondary containment and initiate action to suspend operations with a potential for draining the reactor vessel.
- 2. With two control room AC subsystems inoperable, immediately suspend movement of recently irradiated fuel assemblies in the secondary containment and initiate action to suspend operations with a potential for draining the reactor vessel.
- 3. The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
- When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.
HOPE CREEK 3/4 7-8a Amendment No.191
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LR N17 0124 ATTACHMENT 3 Proposed Technical Specifications Bases Changes (Mark ups) (For Information Only) Proposed Technical Specifications Bases Revised Pages B3/4 3 2f B3/4 3 2g B 3/4 3 13 B 3/4 3 14* B 3/4 3 15* B 3/4 3 16* B 3/4 3 17* B3/4 5 1 B3/4 5 1a B3/4 5 2 B3/4 5 2a* B3/4 5 2b* B3/4 5 2c* B3/4 5 2d* B3/4 5 2e* B3/4 5 2f* B3/4 5 3 B 3/4 6 13
- New TS Bases Page
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) actions to ensure that any offsite releases are within the limits calculated in the safety analysis. Reactor Vessel Water Level - Low Low, Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Two channels of Reactor Vessel Water Level - Low Low, Level 2 Function are available and are required to be OPERABLE for each PCIS channel to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level - Low Low, Level 2 Function is required to be OPERABLE in OPERATIONAL CONDITIONS 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In OPERATIONAL CONDITIONS 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these OPERATIONAL CONDITIONS; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs), when handling irradiated fuel in the secondary containment and during CORE ALTERATIONS, because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs. The valve groups actuated by this Function are listed in the Technical Requirements Manual. 2.b Drywell Pressure - High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the FRVS are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Two channels of Drywell Pressure - High Functions are available and are required to be OPERABLE for each PCIS channel to ensure that no single instrument failure can preclude performance of the isolation function. The Drywell Pressure - High Function is required to be OPERABLE in OPERATIONAL CONDITIONS 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in OPERATIONAL CONDITIONS 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these OPERATIONAL CONDITIONS. The valve groups actuated by this Function are listed in the Technical Requirements Manual. Hope Creek B3/4 3-2f Amendment No. 171 (PSEG Issued)
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.c, 2.d. Refueling Floor and Reactor Building Exhaust Radiation - High High Refueling Floor or Reactor Building exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Exhaust Radiation - High is detected, secondary containment isolation and actuation of the FRVS are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 4). The Exhaust Radiation - High signals are initiated from radiation detectors that are located on the ventilation exhaust ducts coming from the reactor building and the refueling floor zones, respectively. Three channels of Reactor Building Exhaust Radiation - High Function and three channels of Refueling Floor Exhaust Radiation - High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Refueling Floor and Reactor Building Exhaust Radiation - High Functions are required to be OPERABLE in OPERATIONAL CONDITIONS 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In OPERATIONAL CONDITIONS 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these OPERATIONAL CONDITIONS; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs, when handling recently irradiated fuel in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. The valve groups actuated by this Function are listed in the Technical Requirements Manual. 2.e. Manual Initiation The Manual Initiation for secondary containment isolation can be performed by manually initiating a primary containment isolation. There is no specific UFSAR safety analysis that takes credit for this Function. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis. There are four push buttons for the logic, one manual initiation push button per PCIS channel. There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on the position of the push buttons. Four channels of Manual Initiation Function are available and are required to be OPERABLE in OPERATIONAL CONDITIONS 1, 2, and 3 and during OPDRVs, when handling recently irradiated fuel in the secondary containment. These are the OPERATIONAL CONDITIONS and other specified conditions in which Hope Creek B3/4 3-2g Amendment No. 171 (PSEG Issued)
INSTRUMENTATION BASES 3/4.3.11 DELETED 3/4.3.12 RPV WATER INVENTORY CONTROL INSTRUMENTATION The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF. If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.4 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures. Technical Specifications are required by 10 CFR 50.36 to include limiting safety system settings (LSSS) for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur. The actual settings for the automatic isolation channels are the same as those established for the same functions in OPERATIONAL CONDITIONS 1, 2, and 3 in TABLE 3.3.2-2, "ISOLATION ACTUATION INSTRUMENTATION SETPOINTS" and TABLE 3.3.3-2, ECCS ACTUATION INSTRUMENTATION SETPOINTS. With the unit in OPERATIONAL CONDITION 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in OPERATIONAL CONDITIONS 4 and 5 to protect Safety Limit 2.1.4 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will be isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The purpose of the RPV Water Inventory Control Instrumentation is to support the requirements of LCO 3.5.2, RPV Water Inventory Control (WIC), and the definition of DRAIN TIME. There are functions that are required for manual initiation or operation of the ECCS injection/spray subsystem required to be OPERABLE by LCO 3.5.2 and other functions that support automatic isolation of Residual Heat Removal (RHR) subsystem and Reactor Water Cleanup (RWCU) system penetration flow path(s) on low RPV water level. The RPV Water Inventory Control Instrumentation supports operation of the Core Spray System (CSS) and the Low Pressure Coolant Injection (LPCI) system. The equipment involved with each of these systems is described in the Bases for LCO 3.5.2. With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.4 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in OPERATIONAL CONDITIONS 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is postulated in which a single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure, (e.g., seismic event, loss of normal power, or single human error). It is assumed, based on HOPE CREEK B 3/4 3-13 Amendment No.
INSTRUMENTATION BASES 3/4.3.12 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION (Continued) engineering judgment, that while in OPERATIONAL CONDITIONS 4 and 5, one low pressure ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level. As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety. Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). Permissive and interlock setpoints are generally considered as nominal values without regard to measurement accuracy. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function-by-Function basis. 1.a, 2.a. Reactor Steam Dome Pressure - Low (Injection Permissive) Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during Modes 4 and 5 that the reactor steam dome pressure will be below the ECCS maximum design pressure, the Reactor Steam Dome Pressure - Low signals are assumed to be OPERABLE and capable of permitting initiation of the ECCS. The Core Spray Reactor Steam Dome Pressure - Low signals are initiated from four pressure transmitters in Divisions 1 and 2 that sense the reactor dome pressure. The LPCI Reactor Vessel Pressure - Low (Permissive) is initiated from a pressure switch downstream of each LPCI injection valve. Four channels of Reactor Steam Dome Pressure - Low Function per associated Division for Core Spray Divisions 1 and 2, and one channel of Reactor Steam Dome Pressure - Low Function per associated LPCI injection valve are required to be OPERABLE in MODES 4 and 5 when ECCS Manual Initiation is required to be OPERABLE, since these channels support the manual initiation Function. In addition, the channels are only required when the associated ECCS subsystem is required to be OPERABLE by LCO 3.5.2. 1.b, 2.b. Core Spray and Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. One flow transmitter per ECCS pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. One channel of the Pump Discharge Flow - Low Function is required to be OPERABLE in MODES 4 and 5 when the associated LPCS or LPCI pump is required to be OPERABLE by LCO 3.5.2 to ensure the pumps are capable of injecting into the Reactor Pressure Vessel when manually initiated. HOPE CREEK B 3/4 3-14 Amendment No xxx
INSTRUMENTATION BASES 3/4.3.12 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION (Continued) 1.c, 2.c Manual Initiation The Manual Initiation push button channels introduce signals into the appropriate ECCS logic to provide manual initiation capability. There is one push button for each Division of low pressure ECCS. In Core Spray Divisions 1 and 2, manual initiation is associated with each pump and valve combination; in divisions 3 and 4, manual inittation is associated with each pump only. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. A channel of the Manual Initiation Function (one channel per Division) is required to be OPERABLE in OPERATIONAL CONDITIONS 4 and 5 when the associated ECCS subsystems are required to be OPERABLE per LCO 3.5.2. 3.a. RHR System Shutdown Cooling Mode Isolation - Reactor Vessel Water Level - Low, Level 3 The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being automatically isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low, Level 3 Function associated with RHR System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR System. Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters (two per valve) that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per valve) of the Reactor Vessel Water Level - Low, Level 3 Function are available, only two channels (all in the same valve group) are required to be OPERABLE. The Reactor Vessel Water Level - Low, Level 3 Allowable Value was chosen to be the same as the Primary Containment Isolation Instrumentation Reactor Vessel Water Level - Low, Level 3 Allowable Value (TABLE 3.3.2-2), since the capability to cool the fuel may be threatened. The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated RHR penetration flow path is credited in calculating DRAIN TIME. 4.a. Reactor Water Cleanup System Isolation - Reactor Vessel Water Level - Low Low, Level 2 The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being automatically isolated by valves that will close automatically without offsite power RPV water level prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low Low, Level 2 Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System. Reactor Vessel Water Level - Low Low, Level 2 is initiated from two channels per valve group that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per valve group) of the Reactor Vessel Water Level - Low Low, Level 2 Function are available, only two channels (all in the same valve group) are required to be OPERABLE. The Reactor Vessel Water Level - Low Low, Level 2 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME. HOPE CREEK B 3/4 3-15 Amendment No xxx
INSTRUMENTATION BASES 3/4.3.12 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION (Continued) ACTIONS ACTION a. provides direction following determination that an instrument channel trip setpoint is less conservative than the value shown in the Allowable Values column of Table 3.3.12-2. ACTION b. directs taking the appropriate ACTION referenced in Table 3.3.12-1. The applicable ACTION referenced in the Table is Function dependent. TABLE 3.3.12-1 ACTION 83 Low reactor vessel pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. If the permissive is inoperable, manual initiation of ECCS is prohibited. Therefore, the permissive must be placed in the trip condition within 1 hour. With the permissive in the trip condition, manual initiation may be performed. Prior to placing the permissive in the tripped condition, the operator can take manual control of the pump and the injection valve to inject water into the RPV. The allowed outage time of 1 hour is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in trip. TABLE 3.3.12-1 ACTION 84 If a Core Spray or Low Pressure Coolant Injection Pump Discharge Flow - Low bypass function is inoperable, there is a risk that the associated low pressure ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat. If a manual initiation function is inoperable, the ECCS subsystem pumps can be started manually and the valves can be opened manually, but this is not the preferred condition. The 24 hour allowed outage time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The allowed outage time is appropriate given the ability to manually start the ECCS pumps and open the injection valves and to manually ensure the pump does not overheat. With the ACTION and associated allowed outage time of ACTION 83 or 84 not met, the associated low pressure ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately. TABLE 3.3.12-1 ACTION 85 RHR System Shutdown Cooling Mode Isolation, Reactor Vessel Water Level - Low, Level 3, and Reactor Water Cleanup System Isolation, Reactor Vessel Water Level - Low Low, Level 2 functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME. If the instrumentation is inoperable, ACTION 85 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation and requires calculation of DRAIN TIME. The calculation cannot credit automatic isolation of the affected penetration flow paths. SURVEILLANCE REQUIREMENTS 4.3.12 states that each RPV WIC actuation instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and LOGIC SYSTEM FUNCTIONAL TEST at the frequencies shown in Table 4.3.12-1. HOPE CREEK B 3/4 3-16 Amendment No xxx
INSTRUMENTATION BASES 3/4.3.12 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION (Continued) REFERENCES
- 1. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup," November 1984.
- 2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.
- 3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," August 1992.
- 4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.
- 5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.
HOPE CREEK B 3/4 3-17 Amendment No xxx
3/4.5 EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN The core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS. The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining. The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment. The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS. As noted, one LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the LPCI mode and is not otherwise inoperable. One LPCI subsystem of RHR is not considered to be OPERABLE while the subsystem is aligned and operating in the Shutdown Cooling Mode during Operational Conditions (OPCONs) 4 and 5 unless realignment of the subsystem can be accomplished from the control room. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment. Verification days that each RHR System cross tie valve on the discharge side of the RHR pumps is closed and power to its operator, if any, is disconnected ensures that each LPCI subsystem remains independent and a failure in the flow path in one subsystem will not affect the flow path of the other LPCI subsystem. Acceptable methods of removing power to the operator include de-energizing breaker control power or racking out or removing the breaker. For the valves in high radiation areas, verification may consist of verifying that no work activity was performed in the area of the valve since the last verification was performed. If one of the RHR System cross tie valves is open or power has not been removed from the valve operator, both associated LPCI subsystems must be considered inoperable. These valves are under strict administrative controls that will ensure that the valves continue to remain closed with either control or motive power removed. HOPE CREEK B 3/4 5-1 Amendment No. 202 (PSEG Issued)
3/4.5 EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING AND SHUTDOWN (Continued) The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling. The capacity of the system is selected to provide the required core cooling. The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor pressures between 1120 and 200 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water. A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem . There is an increased risk associated with entering an OPERATIONAL CONDITION or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance. HOPE CREEK B 3/4 5-1a Amendment No. 202 (PSEG Issued)
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES ECCS-OPERATING and SHUTDOWN (Continued) With the HPCI system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the CSS and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems and the RCIC system. If any one LPCI subsystem or one CSS subsystem is inoperable in addition to an inoperable HPCI system, the inoperable LPCI subsystem/CSS subsystem or the HPCI system must be restored to OPERABLE status within 72 hours. In this condition, adequate core cooling is ensured by the OPERABILITY of the automatic depressurization system (ADS) and the remaining low pressure ECCS subsystems. However, the overall ECCS reliability is reduced because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in reduced ECCS capability to perform its intended safety function. Since both a high pressure system (HPCI) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours is required to restore either the HPCI system or the LPCI/CSS subsystem to OPERABLE status. The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor to be in HOT SHUTDOWN with vessel pressure not less than 200 psig. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment. Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200°F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS. ADS automatically controls five selected safety-relief valves although the safety analysis only takes credit for four valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability. HOPE CREEK B 3/4 5-2 Amendment No. 69 89
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES 3/4 5.2 - REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL
Background:
The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF. If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.4 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures. Applicable Safety Analysis: With the unit in OPERATIONAL CONDITION 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in OPERATIONAL CONDITIONS 4 and 5 to protect Safety Limit 2.1.4 and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur. A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in OPERATIONAL CONDITIONS 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is considered in which single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure, (e.g., seismic event, loss of normal power, or single human error). It is assumed, based on engineering judgement, that while in OPERATIONAL CONDITIONS 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety. Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). Limiting Condition for Operation: The RPV water level must be controlled in OPERATIONAL CONDITIONS 4 and 5 to ensure that if an unexpected draining event should occur, the reactor coolant water level remains above the top of the active irradiated fuel as required by Safety Limit 2.1.4. The Limiting Condition for Operation (LCO) requires the DRAIN TIME of RPV water inventory to the TAF to be 36 hours. A DRAIN TIME of 36 hours is considered reasonable to identify and initiate action to mitigate unexpected draining of reactor coolant. An event that could cause loss of RPV water inventory and result in the RPV water level reaching the TAF in greater than 36 hours does not represent a significant challenge to Safety Limit 2.1.4 and can be managed as part of normal plant operation. One low pressure ECCS injection/spray subsystem is required to be OPERABLE and capable of being manually started to provide defense-in-depth should an unexpected draining event occur. A low pressure ECCS injection/spray subsystem consists of either one Core Spray System (CSS) subsystem or one Low Pressure Coolant Injection (LPCI) subsystem. Each CSS subsystem consists of two motor driven pumps, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV. Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV. HOPE CREEK B 3/4 5-2a Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES RPV WATER INVENTORY CONTROL (Continued) The LCO is modified by a note which allows a required LPCI subsystem to be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the LPCI mode and is not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of the restrictions on DRAIN TIME, sufficient time will be available following an unexpected draining event to manually align and initiate LPCI subsystem operation to maintain RPV water inventory prior to the RPV water level reaching the TAF. Applicability: RPV water inventory control is required in OPERATIONAL CONDITIONS 4 and 5. Requirements on water inventory control are contained in LCO 3.3.12, RPV WATER INVENTORY CONTROL INSTRUMENTATION, and LCO 3.5.2, RPV WATER INVENTORY CONTROL. RPV water inventory control is required to protect Safety Limit 2.1.4 which is applicable whenever irradiated fuel is in the reactor vessel. Actions: Action a. - If the required low pressure ECCS injection/spray subsystem is inoperable, it must be restored to OPERABLE status within 4 hours. In this condition, the LCO controls on DRAIN TIME minimize the possibility that an unexpected draining event could necessitate the use of the ECCS injection/spray subsystem; however, the defense-in-depth provided by the ECCS injection/spray subsystem is lost. The 4-hour allowed outage time for restoring the required low pressure ECCS injection/spray subsystem to OPERABLE status is based on engineering judgment that considers the LCO controls on DRAIN TIME and the low probability of an unexpected draining event that would result in loss of RPV water inventory. If the inoperable ECCS injection/spray subsystem is not restored to OPERABLE status within 4 hours, action must be initiated immediately to establish a method of water injection capable of operating without offsite electrical power. The method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The method of water injection may be manually initiated and may consist of one or more systems or subsystems, and must be able to access water inventory capable of maintaining the RPV water level above the TAF for 36 hours. If recirculation of injected water would occur, it may be credited in determining the necessary water volume. Action b. - Deleted Action c. - With the DRAIN TIME less than 36 hours but greater than or equal to 8 hours, compensatory measures should be taken to ensure the ability to implement mitigating actions should an unexpected draining event occur. Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment. HOPE CREEK B 3/4 5-2b Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES RPV WATER INVENTORY CONTROL (Continued) The secondary containment provides a controlled volume in which fission products can be contained, diluted, and processed prior to release to the environment. Verification that the secondary containment boundary is capable of being established in less than the DRAIN TIME is required. The required verification confirms actions to establish the secondary containment boundary are preplanned and necessary materials are available. The secondary containment boundary is considered established when one Filtration, Recirculation and Ventilation (FRVS) ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment. Verification that secondary containment boundary can be established must be performed within 4 hours. The required verification is an administrative activity and does not require manipulation or testing of equipment. Secondary containment penetration flow paths form a part of the secondary containment boundary. Verification of the capability to isolate each secondary containment penetration flow path in less than the DRAIN TIME is required. The required verification confirms actions to isolate secondary containment penetration flow paths are preplanned and necessary materials are available. Power operated valves are not required to receive automatic isolation signals if they can be closed manually within the required time. Verification that secondary containment penetration flow paths can be isolated must be performed within 4 hours. The required verification is an administrative activity and does not require manipulation or testing of equipment. One FRVS ventilation unit is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Verification of the capability to place one FRVS ventilation unit in operation in less than the DRAIN TIME is required. The required verification confirms actions to place a FRVS ventilation unit in operation are preplanned and necessary materials are available. Verification that a FRVS ventilation unit can be placed in operation must be performed within 4 hours. The required verification is an administrative activity and does not require manipulation or testing of equipment. If any of the above Action c conditions are not met, immediate actions must be initiated to restore DRAIN TIME to 36 hours or greater. Action d. - With the DRAIN TIME less than 8 hours, mitigating actions are implemented in case an unexpected draining event should occur. Immediate action to establish an additional method of water injection augmenting the ECCS injection/spray subsystem required by the LCO is required. The additional method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The Action is modified by a note which states that either the ECCS injection/ spray subsystem or the additional method of water injection must be capable of operating without offsite electrical power. The additional method of water injection may be manually initiated and may consist of one or more systems or subsystems. The additional method of water injection must be able to access water inventory capable of being injected to maintain the RPV water level above the TAF for 36 hours. The additional method of water injection and the ECCS injection/spray subsystem may share all or part of the same water sources. If recirculation of injected water would occur, it may be credited in determining the required water volume. Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment. HOPE CREEK B 3/4 5-2c Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES RPV WATER INVENTORY CONTROL (Continued) The secondary containment provides a control volume into which fission products can be contained, diluted, and processed prior to release to the environment. Actions to immediately establish the secondary containment boundary are required. With secondary containment boundary established, one FRVS ventilation unit is capable of maintaining a negative pressure in the secondary containment with respect to the environment. The secondary containment penetrations form a part of the secondary containment boundary. Actions to immediately verify that each secondary containment penetration flow path is isolated or to verify that it can be manually isolated from the control room are required. One FRVS ventilation unit is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Actions to immediately verify that at least one FRVS ventilation unit is capable of being placed in operation are required. The required verification is an administrative activity and does not require manipulation or testing of equipment. If any of the above Action d conditions are not met, immediate actions must be initiated to restore DRAIN TIME to 36 hours or greater. Action e. - If the DRAIN TIME is less than 1 hour, actions must be initiated immediately to restore the DRAIN TIME to 36 hours. In this condition, there may be insufficient time to respond to an unexpected draining event to prevent the RPV water inventory from reaching the TAF. Surveillance Requirements: Surveillance Requirement (SR) 4.5.2.1 verifies that the DRAIN TIME of RPV water inventory to the TAF is 36 hours. The period of 36 hours is considered reasonable to identify and initiate action to mitigate draining of reactor coolant. Loss of RPV water inventory that would result in the RPV water level reaching the TAF in greater than 36 hours does not represent a significant challenge to Safety Limit 2.1.4 and can be managed as part of normal plant operation. The definition of DRAIN TIME states that realistic cross-sectional areas and drain rates are used in the calculation. A realistic drain rate may be determined using a single, step-wise, or integrated calculation considering the changing RPV water level during a draining event. For a control rod RPV penetration flow path with the control rod drive mechanism removed and not replaced with a blank flange, the realistic cross-sectional area is based on the control rod blade seated in the control rod guide tube. If the control rod blade will be raised from the penetration to adjust or verify seating of the blade, the exposed cross-sectional area of the RPV penetration flow path is used. The definition of DRAIN TIME excludes from the calculation those penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths. A blank flange or other bolted device must be connected with a sufficient number of bolts to prevent draining in the event of an Operating Basis Earthquake. Normal or expected leakage from closed systems or past isolation devices is permitted. Determination that a system is intact and closed or isolated must consider the status of branch lines and ongoing plant maintenance and testing activities. HOPE CREEK B 3/4 5-2d Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES RPV WATER INVENTORY CONTROL (Continued) The Residual Heat Removal (RHR) Shutdown Cooling System is only considered an intact closed system when misalignment issues (Reference 6) have been precluded by functional valve interlocks or by isolation devices, such that redirection of RPV water out of an RHR subsystem is precluded. Further, the RHR Shutdown Cooling System is only considered an intact closed system if its controls have not been transferred to Remote Shutdown, which disables the interlocks and isolation signals. The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential effects of a single operator error or initiating event on items supporting maintenance and testing (rigging, scaffolding, temporary shielding, piping plugs, snubber removal, freeze seals, etc.). If failure of such items could result and would cause a draining event from a closed system or between the RPV and the isolation device, the penetration flow path may not be excluded from the DRAIN TIME calculation. TS 4.0.1 requires SRs to be met between performances. Therefore, any changes in plant conditions that would change the DRAIN TIME requires that a new DRAIN TIME be determined. SRs 4.5.2.2 and 4.5.2.3 - The minimum water level of 5 inches required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the Core Spray System (CSS) subsystem or LPCI subsystem pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, the required ECCS injection/spray subsystem is inoperable unless aligned to an OPERABLE CST. The required CSS is OPERABLE if it can take suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pumps. Therefore, a verification that either the suppression pool water level is greater than or equal to 5 inches or that a CSS subsystem is aligned to take suction from the CST and the CST contains greater than or equal to 135,000 available gallons of water, ensures that the CSS subsystem can supply the required makeup water to the RPV. SR 4.5.2.4 - The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the required ECCS injection/spray subsystems full of water ensures that the ECCS subsystem will perform properly. This may also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. SR 4.5.2.5 - Verifying the correct alignment for manual, power operated, and automatic valves in the required ECCS subsystem flow paths provides assurance that the proper flow path will be available for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR is modified by a note that provides an exception when an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation. SR 4.5.2.6 - Verifying that the required ECCS injection/spray subsystem can be manually started and operate for at least 10 minutes demonstrates that the subsystem is available to mitigate a draining event. Testing the ECCS injection/spray subsystem through the recirculation line is necessary to avoid overfilling the refueling cavity. The minimum operating time of 10 minutes was based on engineering judgement. HOPE CREEK B 3/4 5-2e Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM (ECCS) AND RPV WATER INVENTORY CONTROL BASES RPV WATER INVENTORY CONTROL (Continued) SR 4.5.2.7 - Verifying that each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated RPV water level isolation signal is required to prevent RPV water inventory from dropping below the TAF should an unexpected draining event occur. SR 4.5.2.8 - The required ECCS subsystem is required to actuate on a manual initiation signal. This surveillance verifies that a manual initiation signal will cause the required CSS subsystem or LPCI subsystem to start and operate as designed, including pump startup and actuation of all automatic valves to their required positions. This SR is modified by a note that excludes vessel injection/spray during the surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the surveillance. The Surveillance Frequencies in the above SRs are controlled under the Surveillance Frequency Controlled Program. REFERENCES
- 1. Information Notice 84-81, "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup," November 1984.
- 2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.
- 3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," August 1992.
- 4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.
- 5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.
- 6. General Electric Service Information Letter No. 388, "RHR Valve Misalignment During Shutdown Cooling Operation for BWR 3/4/5/6," February 1983.
HOPE CREEK B 3/4 5-2f Amendment No. xxx
EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCI, CSS and LPCI systems in the event of a LOCA. This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is also required by Specification 3.6.2.1. Repair work might require making the suppression chamber inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5. In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200°F.+ Since pressure suppression is not required below 212°F, the minimum water volume is based on NPSH, recirculation volume and vortex prevention plus a safety margin for conservatism. + See Special Test Exception 3.10.8. HOPE CREEK B 3/4 5-3 Amendment No. 69, 89
CONTAINMENT SYSTEMS BASES Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This ensures that the safety analysis assumptions are valid. Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.25 psid is valid. 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Building and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified. Establishing and maintaining a 0.25 inch water gage vacuum in the reactor building with the filtration recirculation and ventilation system (FRVS) once per 18 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment. In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during movement of recently irradiated fuel assemblies in the secondary containment or during operations with a potential for draining the reactor vessel (OPDRVs). Due to radioactive decay, handling of fuel only requires OPERABILITY of secondary containment when fuel being handled is recently irradiated, i.e., fuel that has occupied part of the critical reactor core within the previous 24 hours. During handling of fuel and CORE ALTERATIONS, secondary containment and FRVS actuation is not required. However, building ventilation will be operating during fuel handling and CORE ALTERATIONS and will be capable of drawing air into the building and exhausting through a monitored pathway. To reduce doses even further below that provided by 24 hours of natural decay, a single normal or contingency method to promptly close secondary containment penetrations is provided in accordance with RG 1.183. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose of the prompt methods (defined as within 30 minutes) is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored. These contingencies are to be utilized after a postulated fuel handling accident has occurred to reduce doses even further below that provided by the natural decay. HOPE CREEK B 3/4 6-13 Amendment No. 187 (PSEG Issued)
LR N17 0124 ATTACHMENT 4 Camera ready Technical Specification Pages
DEFINITIONS SECTION 1.0 DEFINITIONS .......................................................................................................................... PAGE 1.1 ACTION ......................................................................................................................................... 1-1 1.2 DELETED ...................................................................................................................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ......................................................... 1-1 1.4 CHANNEL CALIBRATION ............................................................................................................ 1-1 1.5 CHANNEL CHECK ....................................................................................................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST ................................................................................................... 1-1 1.7 CORE ALTERATION .................................................................................................................... 1-2 1.8 DELETED ...................................................................................................................................... 1-2 1.9 CORE OPERATING LIMITS REPORT ......................................................................................... 1-2 1.10 CRITICAL POWER RATIO ........................................................................................................... 1-2 1.11 DOSE EQUIVALENT I-131 ........................................................................................................... 1-2 1.11.1 DRAIN TIME ............................................................................................................................... 1-2a 1.12 -AVERAGE DISINTEGRATION ENERGY ............................................................................... 1-2a 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ...................................... 1-3 1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ......................... 1-3 1.15 DELETED ...................................................................................................................................... 1-3 1.16 DELETED ...................................................................................................................................... 1-3 1.17 FREQUENCY NOTATION ............................................................................................................ 1-3 1.18 IDENTIFIED LEAKAGE ................................................................................................................ 1-3 1.18.1 INSERVICE TESTING PROGRAM .............................................................................................. 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME .................................................................................... 1-3 1.20 LIMITING CONTROL ROD PATTERN ......................................................................................... 1-3 1.21 LINEAR HEAT GENERATION RATE ........................................................................................... 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST ......................................................................................... 1-4 1.23 DELETED ...................................................................................................................................... 1-4 1.24 MEMBER(S) OF THE PUBLIC ..................................................................................................... 1-4 1.25 MINIMUM CRITICAL POWER RATIO .......................................................................................... 1-4 HOPE CREEK i Amendment No. xxx
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Table 3.3.9-2 Feedwater/Main Turbine Trip System Actuation Instrumentation Setpoints ... 3/4 3-107 Table 4.3.9.1-1 Feedwater/Main Turbine Trip System Actuation Instrumentation Surveillance Requirements .......................... 3/4 3-108 3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION .............. 3/4 3-109 3/4.3.11 Deleted .................................................. 3/4 3-110 3/4.3.12 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL INSTRUMENTATION...........................................3/4 3-111 Table 3.3.12-1 RPV Water Inventory Control Instrumentation .........................................3/4 3-112 Table 3.3.12-2 RPV Water Inventory Control Instrumentation Setpoints ...............................3/4 3-114 Table 4.3.12.1-1 RPV Water Inventory Control Instrumentation Surveillance Requirements ...............3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops ...................................... 3/4 4-1 Figure 3.4.1.1-1 DELETED ............................... 3/4 4-3 Jet Pumps ................................................ 3/4 4-4 Recirculation Loop Flow .................................. 3/4 4-5 Idle Recirculation Loop Startup .......................... 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES Safety/Relief Valves ..................................... 3/4 4-7 Safety/Relief Valves Low-Low Set Function ................ 3/4 4-9 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ................................ 3/4 4-10 Operational Leakage ...................................... 3/4 4-11 Table 3.4.3.2-1 Reactor Coolant System Pressure Isolation Valves ...................... 3/4 4-13 Table 3.4.3.2-2 Reactor Coolant System Interface Valves Leakage Pressure Monitors ...... 3/4 4-14 3/4.4.4 DELETED .................................................. 3/4 4-15 3/4.4.5 SPECIFIC ACTIVITY ........................................ 3/4 4-18 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program .................. 3/4 4-20 HOPE CREEK x Amendment No. XXX
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-21 Figure 3.4.6.1-1 Hydrostatic Pressure and Leak Tests Pressure/Temperature Limits - Curve A 3/4 4-23 Figure 3.4.6.1-2 Non-Nuclear Heatup and Cooldown Pressure/Temperature Limits - Curve B 3/4 4-23a Figure 3.4.6.1-3 Core Critical Heatup and Cooldown Pressure/Temperature Limits - Curve C 3/4 4-23b Table 4.4.6.1.3-1 (Deleted)........................... 3/4 4-24 Reactor Steam Dome...................................... 3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................ 3/4 4-26 3/4.4.8 DELETED................................................. 3/4 4-27 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown............................................ 3/4 4-28 Cold Shutdown........................................... 3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL 3/4.5.1 ECCS - OPERATING........................................ 3/4 5-1 3/4.5.2 RPV WATER INVENTORY CONTROL............................. 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER..................................... 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity........................... 3/4 6-1 Primary Containment Leakage............................. 3/4 6-2 Primary Containment Air Locks........................... 3/4 6-5 Primary Containment Structural Integrity................ 3/4 6-8 Drywell and Suppression Chamber Internal Pressure....... 3/4 6-9 HOPE CREEK xi Amendment No. xxx
INDEX BASES SECTION PAGE INSTRUMENTATION (Continued) Remote Shutdown Monitoring Instrumentation and Controls ........................................... B 3/4 3-5 Accident Monitoring Instrumentation ...................... B 3/4 3-5 Source Range Monitors .................................... B 3/4 3-5 3/4.3.8 DELETED .................................................. B 3/4 3-7 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION .......................................... B 3/4 3-7 Figure B3/4 3-1 Reactor Vessel Water Level ............. B 3/4 3-8 3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION .............. B 3/4 3-9 3/4.3.11 Deleted .................................................. B 3/4 3-13 3/4.3.12 RPV WATER INVENTORY CONTROL INSTRUMENTATION ...............B 3/4 3-13 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM ..................................... B 3/4 4-1 3/4.4.2 SAFETY/RELIEF VALVES ..................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ................................ B 3/4 4-3 Operational Leakage ...................................... B 3/4 4-3 3/4.4.4 CHEMISTRY ................................................ B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY ........................................ B 3/4 4-4 3/4.4.6 PRESSURE/TEMPERATURE LIMITS .............................. B 3/4 4-5 Table B3/4.4.6-1 Reactor Vessel Toughness .............. B 3/4 4-7 Figure B3/4.4.6-1 Fast Neutron Fluence (E>1Mev) at (1/4)T as a Function of Service life .............. B 3/4 4-8 Table B3/4.4.6-2 Numeric Values for Pressure/Temperature Limits ........... B 3/4 4-9 HOPE CREEK xviii Amendment No. XXX
INDEX BASES SECTION PAGE 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........... B 3/4 4-6 3/4.4.8 DELETED.................................... B 3/4 4-6 3/4.4.9 RESIDUAL HEAT REMOVAL...................... B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL 3/4.5.1 ECCS - OPERATING........................... B 3/4 5-1 3/4.5.2 RPV WATER INVENTORY CONTROL................ B 3/4 5-2a 3/4.5.3 SUPPRESSION CHAMBER........................ B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity.............. B 3/4 6-1 Primary Containment Leakage................ B 3/4 6-1 Primary Containment Air Locks.............. B 3/4 6-1 Primary Containment Structural Integrity... B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure B 3/4 6-2 Drywell Average Air Temperature............ B 3/4 6-2 Drywell and Suppression Chamber Purge System B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS................... B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES....... B 3/4 6-5 3/4.6.4 VACUUM RELIEF.............................. B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT...................... B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL..... B 3/4 6-6 HOPE CREEK xix Amendment No. xxx
DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:
- a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement), and
- b. Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. 1.8 DELETED CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant operation within these limits is addressed in individual specifications. CRITICAL POWER RATIO 1.10 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the applicable NRC- approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." HOPE CREEK 1-2 Amendment No. xxx
DEFINITIONS DRAIN TIME 1.11.1 The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming: a) The water inventory above the TAF is divided by the limiting drain rate; b) The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error), for all penetration flow paths below the TAF except:
- 1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;
- 2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or
- 3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.
c) The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d) No additional draining events occur; and e) Realistic cross-sectional areas and drain rates are used. A bounding DRAIN TIME may be used in lieu of a calculated value. -AVERAGE DISINTEGRATION ENERGY 1.12 shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. HOPE CREEK 1-2a Amendment No. xxx
TABLE 3.3.2-1 (Continued) NOTES
- When handling recently irradiated fuel in the secondary containment.
- When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the Norm position.
- Below 20% of RATED THERMAL POWER the Main Steamline Radiation Monitor setpoints shall not exceed the values determined using normal full power background radiation levels with the hydrogen water chemistry (HWC) system shut down. After reaching 20% of RATED THERMAL POWER the normal full power background radiation level and associated trip setpoints may be increased to levels previously measured during full power operation with hydrogen injection. Prior to decreasing below 20% of RATED THERMAL POWER the background level and associated setpoint shall be returned to the normal full power values. If the Main Steamline Radiation Monitor setpoints have been increased for HWC operation and a power reduction event occurs so that the reactor power is below 20% of RATED THERMAL POWER without the required setpoint change, control rod motion shall be suspended (except for scram or other emergency actions) until the necessary setpoint adjustment is made.
(a) A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. (b) Also trips and isolates the mechanical vacuum pumps. (c) Also starts the Filtration, Recirculation and Ventilation System (FRVS). (d) DELETED (e) Sensors arranged per valve group, not per trip system. (f) Closes only RWCU system isolation valve(s) HV-F001 and HV-F004. (g) Requires system steam supply pressure-low coincident with drywell pressure-high to close turbine exhaust vacuum breaker valves. (h) Manual isolation closes HV-F008 only, and only following manual or automatic initiation of the RCIC system. (i) Manual isolation closes HV-F003 and HV-F042 only, and only following manual or automatic initiation of the HPCI system. (j) Trip functions common to RPS instrumentation. HOPE CREEK 3/4 3-16a Amendment No. xxx
TABLE 4.3.2.1-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- When handling recently irradiated fuel in the secondary containment.
- When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the Norm position.
(a) Manual initiation switches shall be tested in accordance with the Surveillance Frequency Control Program. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program as part of circuitry required to be tested for automatic system isolation. (b) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (c) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. HOPE CREEK 3/4 3-31 Amendment No. xxx
TABLE 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUNCTION FUNCTION(a) CONDITIONS ACTION
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Water Level - Low Low Low, Level 1 2(b)(e) 1, 2, 3 30
- b. Drywell Pressure - High 2(b)(e) 1, 2, 3 30
- c. Reactor Vessel Pressure - Low (Permissive) 4/division(f) 1, 2, 3 31
- d. Core Spray Pump Discharge Flow - Low (Bypass) 1/subsystem 1, 2, 3 37
- e. Core Spray Pump Start Time Delay - Normal Power 1/subsystem 1, 2, 3 31
- f. Core Spray Pump Start Time Delay - Emergency Power 1/subsystem 1, 2, 3 31
- g. Manual Initiation 1/division(b)(g) 1, 2, 3 33
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
- a. Reactor Vessel Water Level - Low Low Low, Level 1 2/valve 1, 2, 3 30
- b. Drywell Pressure - High 2/valve 1, 2, 3 30
- c. Reactor Vessel Pressure - Low (Permissive) 1/valve 1, 2, 3 31
- d. LPCI Pump Discharge Flow - Low (Bypass) 1/pump 1, 2, 3 37
- e. LPCI Pump Start Time Delay - Normal Power 1/pump(i) 1, 2, 3 31
- f. Manual Initiation 1/subsystem 1, 2, 3 33
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
- a. Reactor Vessel Water Level - Low Low Level 2 4 1, 2, 3 34
- b. Drywell Pressure - High 4 1, 2, 3 34
- c. Condensate Storage Tank Level - Low 2(c) 1, 2, 3 35
- d. Suppression Pool Water Level - High 2(c) 1, 2, 3 35
- e. Reactor Vessel Water Level - High, Level 8 4(d) 1, 2, 3 31
- f. HPCI Pump Discharge Flow - Low (Bypass) 1 1, 2, 3 37
- g. Manual Initiation 1/system 1, 2, 3 33
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM##
- a. Reactor Vessel Water Level - Low Low Low, Level 1 4 1, 2, 3 30
- b. Drywell Pressure - High 4 1, 2, 3 30
- c. ADS Timer 2 1, 2, 3 31
- d. Core Spray Pump Discharge Pressure - High (Permissive) 1/pump 1, 2, 3 31 HOPE CREEK 3/4 3-33 Amendment No. xxx
TABLE 3.3.3-1 (Cont'd) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUNCTION FUNCTION(a) CONDITIONS ACTION
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM##
- e. RHR LPCI Mode Pump Discharge Pressure - High (Permissive) 2/pump 1, 2, 3 31
- f. Reactor Vessel Water Level - Low, Level 3 (Permissive) 2 1, 2, 3 31
- g. ADS Drywell Pressure Bypass Timer 4 1, 2, 3 31
- h. ADS Manual Inhibit Switch 2 1, 2, 3 31
- i. Manual Initiation 4 1, 2, 3 33 MINIMUM APPLICABLE TOTAL NO. CHANNELS CHANNELS OPERATIONAL OF CHANNELS(h) TO TRIP(h) OPERABLE(h) CONDITIONS ACTION
- 5. LOSS OF POWER
- 1. 4.16 kv Emergency Bus Under-voltage (Loss of Voltage) 4/bus 2/bus 3/bus 1, 2, 3, 4**, 5** 36
- 2. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 2/source/ 2/source/ 2/source/ 1, 2, 3, 4**, 5** 36 bus bus bus (a) A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b) Also actuates the associated emergency diesel generators. (c) One trip system. Provides signal to HPCI pump suction valve only. (d) Provides a signal to trip HPCI pump turbine only. (e) In divisions 1 and 2, the two sensors are associated with each pump and valve combination. In divisions 3 and 4, the two sensors are associated with each pump only. (f) Division 1 and 2 only. (g) In divisions 1 and 2, manual initiation is associated with each pump and valve combination; in divisions 3 and 4, manual initiation is associated with each pump only. (h) Each voltage detector is a channel. (I) Start time delay is applicable to LPCI Pump C and D only.
- Deleted.
- Required when ESF equipment is required to be OPERABLE.
- Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
- Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
HOPE CREEK 3/4 3-34 Amendment No. xxx
TABLE 3.3.3-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours or declare the associated system inoperable.
- b. With more than one channel inoperable, declare the associated system inoperable.
ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ECCS inoperable within 24 hours. ACTION 32 - Deleted ACTION 33 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 24 hours or declare the associated ECCS inoperable. ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours or declare the HPCI system inoperable.
- b. With more than one channel inoperable, declare the HPCI system inoperable.
ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 24 hours or declare the HPCI system inoperable. ACTION 36 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condition within 1 hour; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST. ACTION 37 - With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function requirement, open the minimum flow bypass valve within one hour. Restore the inoperable channel to OPERABLE status within 7 days or declare the associated ECCS inoperable. HOPE CREEK 3/4 3-35 Amendment No. xxx
TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE (a) (a) (a) TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3
- b. Drywell Pressure - High 1, 2, 3
- c. Reactor Vessel Pressure - Low 1, 2, 3
- d. Core Spray Pump Discharge Flow - Low (Bypass) 1, 2, 3
- e. Core Spray Pump Start Time Delay - Normal Power NA 1, 2, 3
- f. Core Spray Pump Start Time Delay - Emergency Power NA 1, 2, 3
- g. Manual Initiation NA NA 1, 2, 3
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
- a. Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3
- b. Drywell Pressure - High 1, 2, 3
- c. Reactor Vessel Pressure - Low (Permissive) 1, 2, 3
- d. LPCI Pump Discharge Flow - Low (Bypass) 1, 2, 3
- e. LPCI Pump Start Time Delay - Normal Power NA 1, 2, 3
- f. Manual Initiation NA NA 1, 2, 3
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
- a. Reactor Vessel Water Level - Low Low, Level 2 1, 2, 3
- b. Drywell Pressure - High 1, 2, 3
- c. Condensate Storage Tank Level - Low 1, 2, 3
- d. Suppression Pool Water Level - High 1, 2, 3
- e. Reactor Vessel Water Level - High, Level 8 1, 2, 3
- f. HPCI Pump Discharge Flow - Low (Bypass) 1, 2, 3
- g. Manual Initiation NA NA 1, 2, 3 HOPE CREEK 3/4 3-39 Amendment No. xxx
TABLE 4.3.3.1-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE (a) (a) (a) TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM
- a. Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3
- b. Drywell Pressure - High 1, 2, 3
- c. ADS Timer NA 1, 2, 3
- d. Core Spray Pump Discharge Pressure - High 1, 2, 3
- e. RHR LPCI Mode Pump Discharge Pressure 1, 2, 3
-High
- f. Reactor Vessel Water Level - Low, Level 3 1, 2, 3
- g. ADS Drywell Pressure Bypass Timer NA 1, 2, 3
- h. ADS Manual Inhibit Switch NA NA 1, 2, 3
- i. Manual initiation NA NA 1, 2, 3
- 5. LOSS OF POWER
- a. 4.16 kv Emergency Bus Under-voltage (Loss NA NA of Voltage) 1, 2, 3, 4**, 5**
- b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 1, 2, 3, 4**, 5**
(a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
- Deleted
** Required OPERABLE when ESF equipment is required to be OPERABLE. # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig. ## Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
HOPE CREEK 3/4 3-40 Amendment No. xxx
TABLE 3.3.7.1-1 (Continued) RADIATION MONITORING INSTRUMENTATION TABLE NOTATION
- When recently irradiated fuel is being handled in the secondary containment.
- Activates control room emergency filtration system.
- When the offgas treatment system is operating.
- Activates control room emergency filtration system.
- With fuel in the new fuel storage vault.
- With fuel in the spent fuel storage pool.
(a) Alarm only. (b) Alarm setpoint to be set in accordance with Specification 3.11.2.7. HOPE CREEK 3/4 3-64 Amendment No. xxx
TABLE 4.3.7.1-1 (Continued) RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
- With fuel in the new fuel storage vault.
- With fuel in the spent fuel storage pool.
- When recently irradiated fuel is being handled in the secondary containment.
- When the offgas treatment system is operating.
HOPE CREEK 3/4 3-67 Amendment No. xxx
INSTRUMENTATION 3/4.3.12 RPV WATER INVENTORY CONTROL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.12 The RPV Water Inventory Control (WIC) actuation instrumentation channels shown in Table 3.3.12-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.12-2. APPLICABILITY: As shown in Table 3.3.12-1 ACTION:
- a. With an RPV WIC actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.12-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With one or more channels inoperable, take the ACTION referenced in Table 3.3.12-1 for the channel immediately.
SURVEILLANCE REQUIREMENTS 4.3.12 Each RPV WIC actuation instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and LOGIC SYSTEM FUNCTIONAL TEST at the frequencies shown in Table 4.3.12.1-1. HOPE CREEK 3/4 3-111 Amendment No. XXX
TABLE 3.3.12-1 RPV WATER INVENTORY CONTROL INSTRUMENTATION MINIMUM OPERABLE CHANNELS APPLICABLE PER TRIP OPERATIONAL TRIP FUNCTION FUNCTION CONDITIONS ACTIONS
- 1. CORE SPRAY SYSTEM (a)(c)
- a. Reactor Vessel Pressure - Low (Permissive) 4/division 4, 5 83 (a)
- b. Core Spray Pump Discharge Flow - Low 1/subsystem 4, 5 84 (Bypass)
(a)
- c. Manual Initiation 1/subsystem 4, 5 84
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM (a)
- a. Reactor Vessel Pressure-Low (Permissive) 1/valve 4, 5 83 (a)(d)
- b. LPCI Pump Discharge Flow - Low (Bypass) 1/pump 4, 5 84 (a)
- c. Manual Initiation 1/subsystem 4, 5 84
- 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
- a. Reactor Vessel Water Level - Low, Level 3 2 (b) 85
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. Reactor Vessel Water Level - Low Low, Level 2 2 (b) 85 (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, RPV Water Inventory Control.
(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME. (c) Division 1 and 2 only. (d) Function not required to be OPERABLE while associated pump is operating in decay heat removal when minimum flow valve is closed and deactivated. HOPE CREEK 3/4 3-112 Amendment No. XXX
TABLE 3.3.12-1 (Continued) RPV WATER INVENTORY CONTROL INSTRUMENTATION ACTION ACTION 83 - Place the channel in trip within 1 hour. Otherwise, immediately declare the associated low pressure ECCS injection/spray subsystem inoperable. ACTION 84 - Restore the channel to OPERABLE status within 24 hours. Otherwise, immediately declare the associated low pressure ECCS injection/spray subsystem inoperable. ACTION 85 - Declare the associated flow path(s) incapable of automatic isolation and calculate DRAIN TIME immediately. HOPE CREEK 3/4 3-113 Amendment No .XXX
TABLE 3.3.12-2 RPV WATER INVENTORY CONTROL INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Pressure - Low (Permissive) 461 psig 481 psig
- b. Core Spray Pump Discharge Flow - Low (Bypass) 775 gpm 650 gpm
- c. Manual Initiation N.A. N.A.
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
- a. Reactor Vessel Pressure-Low (Permissive) 450 psig 460 psig
- b. LPCI Pump Discharge Flow - Low (Bypass) 1250 gpm 1100 gpm
- c. Manual Initiation N.A. N.A.
- 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
- a. Reactor Vessel Water Level - Low, Level 3 12.5 inches* (b)
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. Reactor Vessel Water Level - Low Low, Level 2 -38 inches* -45 inches
- See Bases Figure B 3/4.3-1.
HOPE CREEK 3/4 3-114 Amendment No. xxx
TABLE 4.3.12.1-1 RPV WATER INVENTORY CONTROL INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL LOGIC CONDITIONS CHANNEL SYSTEM FOR WHICH CHANNEL FUNCTIONAL FUNCTIONAL SURVEILLANCE (a) (a) (a) TRIP FUNCTION CHECK TEST TEST REQUIRED
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Pressure - N.A. 4, 5 Low (Permissive)
- b. Core Spray Pump Discharge N.A. 4, 5 Flow - Low (Bypass)
- c. Manual Initiation N.A. N.A. 4, 5
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
- a. Reactor Vessel Pressure-Low N.A. 4, 5 (Permissive)
- b. LPCI Pump Discharge Flow - N.A. 4, 5 Low (Bypass)
- c. Manual Initiation N.A. N.A. 4, 5
- 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
- a. Reactor Vessel Water Level - N.A. (b)
Low, Level 3
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. Reactor Vessel Water Level - N.A. (b)
Low Low, Level 2 (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. (b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME. HOPE CREEK 3/4 3-115 Amendment No. xxx
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:
- a. The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:
- 1. Two OPERABLE core spray pumps, and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
- b. The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:
- 1. One OPERABLE LPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
- c. The high pressure coolant injection (HPCI) system consisting of:
- 1. One OPERABLE HPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
- d. The automatic depressurization system (ADS) with five OPERABLE ADS valves.
APPLICABILITY: OPERATIONAL CONDITION 1, 2*,** #, and 3*,**,##.
- The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
- The ADS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
- See Special Test Exception 3.10.6.
- Two LPCI subsystems of the RHR system may be inoperable in that they are aligned in the shutdown cooling mode when the reactor vessel pressure is less than the RHR shutdown cooling permissive setpoint.
HOPE CREEK 3/4 5-1 Amendment xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION ACTION: NOTE: LCO 3.0.4.b is not applicable to HPCI.
- a. For the Core Spray system:
- 1. With one core spray subsystem inoperable, provided that at least two LPCI subsystem are OPERABLE, restore the inoperable core spray subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With both core spray subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- b. For the LPCI system:
- 1. With one LPCI subsystem inoperable, provided that at least one core spray subsystem is OPERABLE, restore the inoperable LPCI subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With two LPCI subsystems inoperable, provided that at least one core spray subsystem is operable, restore at least one LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 3. With three LPCI subsystems inoperable, provided that both core spray subsystems are OPERABLE, restore at least two LPCI subsystems to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 4. With all four LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.*
- Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
HOPE CREEK 3/4 5-2 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- c. For the HPCI system, provided the Core Spray System, the LPCI system, the ADS and the RCIC system are OPERABLE:
- 1. With the HPCI system inoperable, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig within the following 24 hours.
- 2. With the HPCI system inoperable and either one LPCI subsystem or one CSS subsystem inoperable, restore the HPCI system to operable status within 72 hours or restore the LPCI subsystem/CSS subsystem to operable status within 72 hours.
Otherwise, be in HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 200 psig in the next 24 hours.
- d. For the ADS:
- 1. With one of the above required ADS valves inoperable, provided the HPCI system, the core spray system and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 100 psig within the next 24 hours.
- 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure to 100 psig within the next 24 hours.
- e. With a CSS and/or LPCI header P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 7 days or determine the ECCS header P locally at least once per 12 hours; otherwise, declare the associated ECCS subsystem inoperable.
- f. The discharge line "keep filled" alarm instrumentation associated with a LPCI and/or CSS subsystem(s) may be in an inoperable status for up to 6 hours for required surveillance testing provided that the "keep filled" alarm instrumentation associated with at least one LPCI or CSS subsystem serviced by the affected "keep filled" system remains OPERABLE; otherwise, perform Surveillance Requirement 4.5.1.a.1.a.
- g. In the event an ECCS system is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
- This includes testing of the "Reactor Coolant System Interface Valves Leakage Pressure Monitors" associated with LPCI and CSS in accordance with Surveillance 4.4.3.2.3 HOPE CREEK 3/4 5-3 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL SURVEILLANCE REQUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:
- a. In accordance with the Surveillance Frequency Control Program:
- 1. For the core spray system, the LPCI system, and the HPCI system:
a) Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water. b) Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct* position. c) Verify the RHR System cross tie valves on the discharge side of the pumps are closed and power, if any, is removed from the valve operators.
- 2. For the HPCI system, verifying that the HPCI pump flow controller is in the correct position.
- b. Verifying that, when tested pursuant to the INSERVICE TESTING PROGRAM:
- 1. The two core spray system pumps in each subsystem together develop a flow of at least 6150 gpm against a test line pressure corresponding to a reactor vessel pressure of 105 psi above suppression pool pressure.
- 2. Each LPCI pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of 20 psid.
- 3. The HPCI pump develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of 1000 psig when steam is being supplied to the turbine at 1000, +20, -80 psig.**
- c. In accordance with the Surveillance Frequency Control Program:
- 1. For the core spray system, the LPCI system, and the HPCI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
- Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
- The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
HOPE CREEK 3/4 5-4 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL SURVEILLANCE REQUIREMENTS (Continued)
- 2. For the HPCI system, verifying that:
a) The system develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of 200 psig, when steam is being supplied to the turbine at 200 + 15, -0 psig.** b) The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber - water level high signal.
- 3. Performing a CHANNEL CALIBRATION of the CSS, and LPCI system discharge line "keep filled" alarm instrumentation.
- 4. Performing a CHANNEL CALIBRATION of the CSS header P instrumentation and verifying the setpoint to be the allowable value of 4.4 psid.
- 5. Performing a CHANNEL CALIBRATION of the LPCI header P instrumentation and verifying the setpoint to be the allowable value of 1.0 psid.
- d. For the ADS:
- 1. In accordance with the Surveillance Frequency Control Program, performing a CHANNEL FUNCTIONAL TEST of the Primary Containment Instrument Gas System low-low pressure alarm system.
- 2. In accordance with the Surveillance Frequency Control Program:
a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation. b) Verify that when tested pursuant to the INSERVICE TESTING PROGRAM, that each ADS valve is capable of being opened. c) Performing a CHANNEL CALIBRATION of the Primary Containment Instrument Gas System low-low pressure alarm system and verifying an alarm setpoint of 85 +/- 2 psig on decreasing pressure.
- The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
HOPE CREEK 3/4 5-5 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL 3/4 5.2 RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be 36 hours AND At least one of the following low pressure ECCS subsystems shall be OPERABLE:
- a. Core spray system subsystem comprised of:
- 1. Two OPERABLE core spray pumps, and
- 2. An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:
a) From the suppression chamber, or b) When the suppression chamber water level is less than the limit or is drained, from the condensate storage tank containing at least 135,000 available gallons of water.
- b. Low pressure coolant injection (LPCI) system subsystem` comprised of:
- 1. One OPERABLE LPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.**
APPLICABILITY: OPERATIONAL CONDITION 4 and 5. ACTION:
- a. With none of the above low pressure ECCS subsystems OPERABLE, restore a subsystem to OPERABLE status within 4 hours. Otherwise, immediately initiate action to establish a method of water injection capable of operating without offsite electrical power.
- b. Deleted.
- Deleted.
- A LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.
HOPE CREEK 3/4 5-6 Amendment xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL LIMITING CONDITION FOR OPERATION (Continued) ACTION:
- c. With DRAIN TIME < 36 hours and 8 hours, within 4 hours:
- 1. Verify secondary containment boundary is capable of being established in less than the DRAIN TIME, and
- 2. Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME, and
- 3. Verify one Filtration, Recirculation and Ventilation (FRVS) ventilation unit is capable of being placed in operation in less than the DRAIN TIME.
Otherwise, immediately initiate action to restore DRAIN TIME to 36 hours.
- d. With DRAIN TIME < 8 hours, immediately:
- 1. Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for 36 hours***
and,
- 2. Initiate action to establish secondary containment boundary, and
- 3. Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room, and
- 4. Initiate action to verify one FRVS ventilation unit is capable of being placed in operation.
Otherwise, immediately initiate action to restore DRAIN TIME to 36 hours.
- e. With DRAIN TIME < 1 hour, immediately initiate action to restore DRAIN TIME to 36 hours.
- Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.
HOPE CREEK 3/4 5-6a Amendment xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL SURVEILLANCE REQUIREMENTS 4.5.2.1 Verify DRAIN TIME 36 hours in accordance with the Surveillance Frequency Control Program. 4.5.2.2 Verify, for a required low pressure coolant injection (LPCI) subsystem, the suppression chamber indicated water level is > 5.0 inches in accordance with the Surveillance Frequency Control Program. 4.5.2.3 Verify, for a required Core Spray (CS) subsystem, the Suppression chamber indicated water level is > 5.0 inches or condensate storage tank contains at least 135,000 available gallons of water in accordance with the Surveillance Frequency Control Program. 4.5.2.4 Verify, for the required ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve in accordance with the Surveillance Frequency Control Program. 4.5.2.5 Verify, for the required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position, in accordance with the Surveillance Frequency Control Program.# 4.5.2.6 Operate the required ECCS injection/spray subsystem through the recirculation line for 10 minutes, in accordance with the Surveillance Frequency Control Program. 4.5.2.7 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal, in accordance with the Surveillance Frequency Control Program 4.5.2.8 Verify the required ECCS injection/spray subsystem actuates on a manual initiation signal, in accordance with the Surveillance Frequency Control Program.##
- Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
- Vessel injection/spray may be excluded.
HOPE CREEK 3/4 5-7 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL 3/4.5.3 SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:
- a. In OPERATIONAL CONDITION 1, 2 and 3 with an indicated water level of at least 74.5".
- b. Deleted APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression chamber water level less than the above limit, restore the water level to within the limit within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. Deleted HOPE CREEK 3/4 5-8 Amendment No. xxx
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND RPV WATER INVENTORY CONTROL SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to:
- a. 74.5" in accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITIONS 1, 2, and 3.
4.5.3.2 Deleted HOPE CREEK 3/4 5-9 Amendment No. xxx
CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION: Without SECONDARY CONTAINMENT INTEGRITY:
- a. In OPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINMENT INTEGRITY within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.6.5.1 SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:
- a. Verifying in accordance with the Surveillance Frequency Control Program that the reactor building is at a negative pressure.
- b. Verifying in accordance with the Surveillance Frequency Control Program that:
- 1. All secondary containment equipment hatches and blowout panels are closed and sealed.
- 2. a. For double door arrangements, at least one door in each access to the secondary containment is closed.
- b. For single door arrangements, the door in each access to the secondary containment is closed except for routine entry and exit.
- 3. All secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers/valves secured in position.
- When recently irradiated fuel is being handled in the secondary containment.
HOPE CREEK 3/4 6-47 Amendment No. xxx
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS LIMITING CONDITION FOR OPERATION 3.6.5.2 The secondary containment ventilation system (RBVS) automatic isolation dampers shown in Table 3.6.5.2-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.2-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION: With one or more of the secondary containment ventilation system automatic isolation dampers shown in Table 3.6.5.2-1 inoperable, maintain at least one isolation damper OPERABLE in each affected penetration that is open and within 8 hours either:
- a. Restore the inoperable dampers to OPERABLE status, or
- b. Isolate each affected penetration by use of at least one deactivated damper secured in the isolation position, or
- c. Isolate each affected penetration by use of at least one closed manual valve or blind flange.
Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Otherwise, in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.6.5.2 Each secondary containment ventilation system automatic isolation damper shown in Table 3.6.5.2-1 shall be demonstrated OPERABLE: a Prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control or power circuit by cycling the damper through at least one complete cycle of full travel and verifying the specified isolation time.
- b. In accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each isolation damper actuates to its isolation position.
- c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.
- When recently irradiated fuel is being handled in the secondary containment.
HOPE CREEK 3/4 6-49 Amendment No. xxx
CONTAINMENT SYSTEMS 3.6.5.3 FILTRATION, RECIRCULATION AND VENTILATION SYSTEM (FRVS) FRVS VENTILATION SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3.1 Two FRVS ventilation units shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION:
- a. With one of the above required FRVS ventilation units inoperable, restore the inoperable unit to OPERABLE status within 7 days, or:
- 1. In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. In Operational Condition *, place the OPERABLE FRVS ventilation unit in operation or suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
- b. With both ventilation units inoperable in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3. are not applicable.
SURVEILLANCE REQUIREMENTS 4.6.5.3.1 Each of the two ventilation units shall be demonstrated OPERABLE:
- a. In accordance with the Surveillance Frequency Control Program by verifying that the water seal bucket traps have a water seal and making up any evaporative losses by filling the traps to the overflow.
- b. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 15 minutes.
- When recently irradiated fuel is being handled in the secondary containment.
HOPE CREEK 3/4 6-51 Amendment No. xxx
CONTAINMENT SYSTEMS 3.6.5.3 FILTRATION, RECIRCULATION AND VENTILATION SYSTEM (FRVS) FRVS RECIRCULATION SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3.2 Six FRVS recirculation units shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION:
- a. With one or two of the above required FRVS recirculation units inoperable, restore all the inoperable unit(s) to OPERABLE status within 7 days, or:
- 1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. In Operational Condition*, suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
- b. With three or more of the above required FRVS recirculation units inoperable in Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
- c. With three or more of the above required FRVS recirculation units inoperable in OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.5.3.2 Each of the six FRVS recirculation units shall be demonstrated OPERABLE:
- a. In accordance with the Surveillance Frequency Control Program by verifying that the water seal bucket traps have a water seal and making up any evaporative losses by filling the traps to the overflow.
- b. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and verifying that the subsystem operates for at least 15 minutes.
- When recently irradiated fuel is being handled in the secondary containment.
HOPE CREEK 3/4 6-52a Amendment No. xxx
PLANT SYSTEMS 3/4.7.2 CONTROL ROOM SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2.1 Two control room emergency filtration system subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3
- 1. With one control room emergency filtration subsystem inoperable for reasons other than Condition a.2, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With one or more control room emergency filtration subsystems inoperable due to an inoperable control room envelope (CRE) boundary ,
- a. Immediately, initiate action to implement mitigating actions; and
- b. Within 24 hours, verify mitigating actions ensure CRE occupant exposures to radiological and chemical hazards will not exceed the limits and actions to mitigate exposure to smoke hazards are taken; and
- c. Within 90 days, restore the CRE boundary to operable status; Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION *:
- 1. With one control room emergency filtration subsystem inoperable for reasons other than Condition b.3, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the pressurization/recirculation mode of operation.
- When recently irradiated fuel is being handled in the secondary containment.
The main control room envelope (CRE) boundary may be opened intermittently under administrative control. HOPE CREEK 3/4 7-6 Amendment No. xxx
PLANT SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION (continued)
- 2. With both control room emergency filtration subsystems inoperable for reasons other than Condition b.3, suspend handling of recently irradiated fuel in the secondary containment.
- 3. With one or more control room emergency filtration subsystems inoperable due to an inoperable CRE boundary##, immediately suspend handling of recently irradiated fuel.
- c. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION*.
SURVEILLANCE REQUIREMENTS 4.7.2.1.1 Each control room emergency filtration subsystem shall be demonstrated OPERABLE:
- a. DELETED
- b. In accordance with the Surveillance Frequency Control Program by verifying that the subsystem operates for at least 15 continuous minutes with the heaters on.
- When recently irradiated fuel is being handled in the secondary containment.
The main control room envelope (CRE) boundary may be opened intermittently under administrative control. HOPE CREEK 3/4 7-6a Amendment No. xxx
PLANT SYSTEMS CONTROL ROOM AIR CONDITIONING (AC) SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2.2 Two control room AC subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3:
- 1. With one control room AC subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With two control room AC subsystems inoperable:
- a. Verify control room air temperature is less than 90°F at least once per 4 hours; and
- b. Restore one control room AC subsystem to OPERABLE status within 72 hours.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION *:
- 1. With one control room AC subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days; or place the OPERABLE control room AC subsystem in operation; or immediately suspend movement of recently irradiated fuel assemblies in the secondary containment.
- 2. With two control room AC subsystems inoperable, immediately suspend movement of recently irradiated fuel assemblies in the secondary containment.
- 3. The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
- When recently irradiated fuel is being handled in the secondary containment.
HOPE CREEK 3/4 7-8a Amendment No.xxx
ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and
- b. Two diesel generators, one of which shall be diesel generator A or diesel generator B, each with:
- 1. A separate fuel oil day tank containing a minimum of 360 gallons of fuel.
- 2. A fuel storage system consisting of two storage tanks containing a minimum of 44,800 gallons of fuel.
- 3. A separate fuel transfer pump for each storage tank.
APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *. ACTION:
- a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of recently irradiated fuel in the secondary containment, and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22'-2" above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.
- b. The provisions of Specification 3.0.3 are not applicable.
- c. With one fuel oil transfer pump inoperable, realign the flowpath of the affected tank to the tank with the remaining operable fuel oil transfer pump within 48 hours and restore the inoperable transfer pump to OPERABLE status within 14 days, otherwise declare the affected emergency diesel generator (EDG) inoperable. This variance may be applied to only one EDG at a time.
SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, and 4.8.1.1.3, except for the requirement of 4.8.1.1.2.a.5.
- When handling recently irradiated fuel in the secondary containment.
HOPE CREEK 3/4 8-11 Amendment No. xxx
ELECTRICAL POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four channels of the D.C. electrical power sources, one of which shall be channel A or channel B, shall be OPERABLE with:
- a. Channel A, consisting of:
- 1. 125 volt battery 1AD411
- 2. 125 volt full capacity charger# 1AD413 or 1AD414
- b. Channel B, consisting of:
- 1. 125 volt battery 1BD411
- 2. 125 volt full capacity charger# 1BD413 or 1BD414.
- c. Channel C, consisting of:
- 1. 125 volt battery 1CD411
- 2. 125 volt full capacity charger# 1CD413 or 1CD414
- 3. 125 volt battery 1CD447
- 4. 125 volt full capacity charger 1CD444
- d. Channel D, consisting of:
- 1. 125 volt battery 1DD411
- 2. 125 volt full capacity charger# 1DD413 or 1DD414
- 3. 125 volt battery 1DD447
- 4. 125 volt full capacity charger 1DD444 APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *.
ACTION:
- a. With less than two channels of the above required D.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS and handling of recently irradiated fuel in the secondary containment.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
- When handling recently irradiated fuel in the secondary containment.
- Only one full capacity charger per battery is required for the channel to be OPERABLE.
HOPE CREEK 3/4 8-17 Amendment No.xxx
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *. ACTION:
- a. With less than two channels of the above required A.C. distribution system energized, suspend CORE ALTERATIONS and handling of recently irradiated fuel in the secondary containment.
- b. With less than two channels of the above required D.C. distribution system energized, suspend CORE ALTERATIONS and handling of recently irradiated fuel in the secondary containment.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system channels shall be determined energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker/switch alignment and voltage on the busses/MCCs/panels.
- When handling recently irradiated fuel in the secondary containment.
HOPE CREEK 3/4 8-23 Amendment No. xxx}}