LR-N18-0098, License Amendment Request: Remote Shutdown System

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License Amendment Request: Remote Shutdown System
ML18304A191
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 10/30/2018
From: Fleming J
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H18-05, LR-N18-0098
Download: ML18304A191 (29)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 PSEG Nuclea-r LLC 10 CFR 50.90 LR-N18-0098 LAR H18-05 OCT 3 0 LiJI8 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

License Amendment Request: Remote Shutdown System In accordance with 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment to Renewed Facility Operating License No. NPF-57 for Hope Creek Generating Station (HCGS).

This license amendment request proposes changes to Technical Specification (TS) 3.3.7.4, REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS. The proposed change would make the HCGS requirements consistent with Improved Standard Technical Specification (ISTS) 3.3.3.2, Remote Shutdown System.

The proposed change will increase the allowed outage time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance. It will also delete Tables 3.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION, 3.3.7.4-2, REMOTE SHUTDOWN SYSTEMS CONTROLS, and 4.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS, to be relocated to the Technical Requirements Manual (TRM) where they will be directly controlled by HCGS.

PSEG's technical and regulatory evaluation of this LAR and the TS change are provided in an enclosure to this letter.

The proposed change has been evaluated *in accordance with 10 CFR 50.91 (a)(1), using the criteria in 10 CFR 50.92(c), and it has been determined that this request involves no significant hazards considerations.

There are no regulatory commitments contained in this letter.

PSEG requests NRC approval of the proposed License Amendment within one year of submittal acceptance, to be implemented within 60 days of issuance.

95-2168 REV. 7/99

GCT . Q i01.E 10 CFR 50.90 Page 2 LR-N18-0098 In accordance with 10 CFR 50.91(b)(1), a copy of this request for amendment has been sent to the State of New Jersey.

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 __ 1:.o"-/'-12,_J ..1'-.., /=9-, .;;.o ..._..55!

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Respectfully,

!f/)p----7{

(

Jean Fleming L.l Director- Site Regulatory Compliance

Enclosure:

Evaluation of the Proposed Change C. Administrator, Region I, NRC Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. MacEwen, Hope Creek Commitment Tracking Coordinator

LR-N18-0098 LAR H18-05 Enclosure Evaluation of the Proposed Change

LR-N18-0098 LAR H18-05 Enclosure HOPE CREEK NUCLEAR GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 License Amendment Request: Remote Shutdown System Table of Contents 1.0

SUMMARY

DESCRIPTION ............................................................................................ 3 2.0 DETAILED DESCRIPTION............................................................................................. 3 2.1 SYSTEM DESIGN AND OPERATION ................................................................ 3 2.2 CURRENT TECHNICAL SPECIFICATION REQUIREMENTS ............................ 3 2.3 REASON FOR THE PROPOSED CHANGE ....................................................... 4

2.4 DESCRIPTION

OF THE PROPOSED CHANGE ................................................ 4

3.0 TECHNICAL EVALUATION

............................................................................................ 5

4.0 REGULATORY EVALUATION

....................................................................................... 7 4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA ..................... 7 4.2 PRECEDENT ...................................................................................................... 8 4.2.1 License Amendments ......................................................................... 8 4.3 NO SIGNIFICANT HAZARDS CONSIDERATION............................................... 8

4.4 CONCLUSION

...................................................................................................10

5.0 ENVIRONMENTAL CONSIDERATION

.........................................................................10

6.0 REFERENCES

..............................................................................................................10 ATTACHMENTS:

1. Technical Specification Page Markups
2. Technical Specification Bases Page Markups (for information only) 2

LR-N18-0098 LAR H18-05 Enclosure 1.0

SUMMARY

DESCRIPTION This license amendment request proposes a change which would revise Hope Creek Technical Specification (TS) Limiting Condition for Operation (LCO) 3.3.7.4 and associated ACTIONS and SURVEILLANCE REQUIREMENTS concerning operability of the Remote Shutdown System.

The proposed change would increase the allowed outage time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance. It will also delete Tables 3.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION, 3.3.7.4-2, REMOTE SHUTDOWN SYSTEMS CONTROLS, and 4.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS, to be relocated to the Technical Requirements Manual (TRM) where they will be directly controlled by HCGS.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Remote Shutdown System (RSS) provides the means for achieving and maintaining safe shutdown conditions from outside the main control room in the unlikely event the main control room becomes uninhabitable.

The primary control station for the RSS is the remote shutdown panel (RSP). In the event of a failure at the RSP, sufficient redundant safety grade instrumentation and controls are available remote from both the main control room and the RSP to ensure that safe shutdown of the reactor can be achieved in accordance with the requirements of 10 CFR 50 Appendix A General Design Criteria (GDC) 19. The systems for which the RSP provides remote instrumentation and controls to accomplish this function are as follows:

2.1.1. Reactor Core Isolation Cooling (RCIC) - to maintain reactor water level.

2.1.2. Residual Heat Removal (RHR) system (loop B) - for suppression pool cooling and shutdown cooling.

2.1.3. Safety Auxiliaries Cooling System (SACS) (loop B) - to supply cooling water to the RHR (B) heat exchanger, RCIC and RHR pump room coolers, RHR (B) motor oil and seal coolers, the Standby Diesel Generator (SDG) cooling loads, and other necessary loads.

2.1.4. Station Service Water System (SSWS) (loop B) - to supply cooling water to the SACS loop B heat exchangers.

2.1.5. Control Area Chilled Water System (CACWS) (loop B) - for cooling various ESF equipment rooms and the main control room.

2.1.6. Nuclear boiler instrumentation - to monitor reactor vessel pressure and level.

2.1.7. Portions of the Fuel Pool Cooling System (FPCS).

2.1.8. Main steam line safety/relief valves (manual actuation) - to lower reactor vessel pressure.

2.2 Current Technical Specification Requirements The current TS 3.3.7.4, Remote Shutdown System Instrumentation and Controls, Limiting Condition for Operability (LCO) states, The remote shutdown system instrumentation and controls shown in Table 3.3.7.4-1 and Table 3.3.7.4-2 shall be OPERABLE.

3

LR-N18-0098 LAR H18-05 Enclosure TS 3.3.4 ACTIONS a. and b. state,

a. With the number of OPERABLE remote shutdown monitoring instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-2, restore the inoperable control(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENT 4.3.7.4.1 states, Each of the above required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted by Table 4.3.7.4-1.

SURVEILLANCE REQUIREMENT 4.3.7.4.2 states, At least one of each of the above remote shutdown control switch(es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) in accordance with the Surveillance Frequency Control Program.

2.3 Reason for the Proposed Change The proposed changes, as shown in Section 2.4 and described in Section 3.0 below, revise TS 3.3.7.4 to be consistent with Improved Standard Technical Specifications (ISTS) 3.3.3.2. In doing so it removes unnecessary information and relocates it to a licensee controlled document preventing needless expenditure of licensee and NRC resources processing license amendments to revise tables when the licensee can adequately control the information.

The change also increases the Allowed Outage Time for inoperable Remote Shutdown System components from 7 to 30 days which is more consistent with their safety significance and is consistent with the ISTS 3.3.3.2 REQUIRED ACTION A.1 COMPLETION TIME.

2.4 Description of the Proposed Change TS 3.3.7.4 is being revised as shown below:

LCO:

The remote shutdown system instrumentation and controls shown in Table 3.3.7.4.1 and Table 3.3.7.4-2 functions shall be OPERABLE.

ACTION:

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

Separate ACTION entry is allowed for each Function

a. With one or more of the number of OPERABLE required remote shutdown monitoring functions inoperable instrumentation channels less than required by Table 3.3.7.4-1, 4

LR-N18-0098 LAR H18-05 Enclosure restore the inoperable function(s) to OPERABLE status within 7 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-2, restore the inoperable control(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS:

4.3.7.4.1 Each of the above normally energized required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the a CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequenciesfrequency specified in the Surveillance Frequency Control Program unless otherwise noted by Table 4.3.7.4-1.

4.3.7.4.2 At least one of each of the aboveEach required control circuit and transfer switch remote shutdown control switch(es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) in accordance with the Surveillance Frequency Control Program.

4.3.7.4.3 Each required instrumentation channel* shall be demonstrated OPERABLE by performance of a CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.

Safety Relief Valve Position, Standby Diesel Generator Breaker Indication, and Switchgear Room Cooler Status Indication are excluded from CHANNEL CALIBRATION.

Tables 3.3.7.4-1, 3.3.7.4-2 and 4.3.7.4-1 are being deleted and relocated to the Technical Requirements Manual (TRM) where they will be directly controlled by HCGS. Revisions to the TRM are reviewed pursuant to 10 CFR 50.59. The APPLICABILITY for LCO 3.3.7.4 is not being revised.

3.0 TECHNICAL EVALUATION

The technical basis for each of the proposed changes listed in Section 2.4 is provided in the following Table:

Proposed Change Evaluation

1. Revise TS 3.3.7.4 Limiting Condition for The definition of OPERABLE in the TS Operation (LCO) to simply state that the provides adequate guidance for determining Remote Shutdown System functions shall what instrumentation and controls are be OPERABLE. References to necessary for a particular Remote Shutdown instrumentation as shown in the associated System function. It is unnecessary to list Tables 3.3.7.4-1 and 3.3.7.4-2 are deleted. specific instruments and controls in the TS to provide adequate assurance that the functions Current TS 3.3.7.4 Tables 3.3.7.4-1, can be performed. GDC 19 requires that 3.3.7.4-2 and 4.3.7.4-1 list the specific remote shutdown capability be provided. The instruments and controls for the remote 5

LR-N18-0098 LAR H18-05 Enclosure shutdown system. The proposed change functions will be described in the TRM, which will relocate this information to licensee is sufficient to assure that the system will be controlled documents. OPERABLE. Listing specific instrumentation and controls is unnecessary and may lead to needless expenditure of licensee and NRC resources processing license amendments to revise the table when the licensee can adequately control the information. These details are not necessary to adequately describe the actual regulatory requirement.

Therefore, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in controlled documents provides adequate assurance that they will be maintained.

Revisions to the TRM are reviewed pursuant to 10 CFR 50.59 as described in UFSAR Section 13.5.4. Changes to the TRM will not affect the requirements of the LCO for the Remote Shutdown System.

2. Revise TS 3.3.7.4 ACTION a. to delete the Table 3.3.7.4-1 is being deleted as discussed reference to monitoring instrumentation in 1 above. This proposed change to TS channels as shown in the associated Table 3.3.7.4 ACTION a. is consistent with the ISTS 3.3.7.4-1. The ACTION will be simplified to CONDITION A. wording.

state the entry condition for the action as The proposed note is also consistent with "one or more of the required functions ISTS. Without the proposed note, the inoperable".

proposed new wording could be interpreted as Add a note to permit separate condition not allowing separate entry for each function.

entry for each function. Consequently, the proposed change is considered to be an administrative change.

3. Revise TS 3.3.7.4 ACTION a. to change the Extending the allowed outage time to 30 days required action time from 7 days to 30 days. is reasonable based on operating experience and the low probability of an event occurring that would require the control room to be evacuated. This proposed change is consistent with the ISTS CONDITION A.

COMPLETION TIME.

4. Delete TS 3.3.7.4 ACTION b. Table 3.3.7.4-2 is being deleted and relocated to a licensee controlled document so reference to it is no longer applicable. In addition, the action is encompassed by ACTION a. through rewording to now say One or more required functions.
5. Delete Tables 3.3.7.4-1, 3.3.7.4-2 and Deletion of the tables is discussed in 1 above.

4.3.7.4-1 and relocate the Remote In addition, the Tables were removed from Shutdown System equipment information to NUREG-1433 by the adoption of TSTF-266.

the TRM.

6

LR-N18-0098 LAR H18-05 Enclosure

6. Revise SURVEILLANCE REQUIREMENT a) Table 4.3.7.4-1 deletion is described in Item (SR) 4.3.7.4.1 and add SR 4.3.7.4.3 to 1 above.

a) delete reference to Table 4.3.7.4-1 which b) Splitting out the CHANNEL CALIBRATION is being deleted into a new SR and rewording the SR makes them consistent with ISTS.

b) relocate CHANNEL CALIBRATION to a new SR 4.3.7.4.3 c) Performing a channel check of normally de-energized instrumentation is not practical or c) require the channel check only for feasible during power operation. This change normally energized Remote Shutdown is consistent with ISTS SR 3.3.3.2.1.

System instrumentation d) The note makes the Remote Shutdown d) Add a note to SR 4.3.7.4.3 that the System consistent with the requirements Safety Relief Valve Position, Standby already established for Safety Relief Valve Diesel Generator Breaker Indication, and Position, Standby Diesel Generator Breaker Switchgear Room Cooler Status Indication Indication, and Switchgear Room Cooler are excluded from CHANNEL Status Indication in current TS Table 4.3.7.4-1.

CALIBRATION and e) Administrative wording change e) replace the with a,

7. Revise SR 4.3.7.4.2 replacing, At least These changes make the SR wording one of each of the above remote shutdown consistent with ISTS SR 3.3.3.2 and reflect the switch(es) and control circuits with each removal of TS Table 3.3.7.4-2.

required control circuit and transfer switch.

8. Revise TS Index pages viii and ix to reflect Administrative change.

TS Section 3.3.7.4 title and deletion of the tables.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria 10 CFR 50.36, Technical Specifications, identifies the requirements for the Technical Specification categories for operating power plants: (1) Safety limits, limiting safety system settings, and limiting control settings, (2) Limiting conditions for operation, (3) Surveillance requirements, (4) Design features, (5) Administrative controls, (6) Decommissioning and (7)

Initial notification, and (8) Written Reports. For Limiting conditions for operation, 10 CFR 50.36 states: Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. For Surveillance Requirements, 10 CFR 50.36 states: Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

7

LR-N18-0098 LAR H18-05 Enclosure The Remote Shutdown System is required to provide equipment at appropriate locations outside the control room with a capability to promptly shut down and maintain the unit in a safe condition thus satisfying Criterion 4 of 10 CFR 50.36(c)(2)(ii). HCGS UFSAR Section 7.4.1.4 (Reference 6.1) describes the design of the Remote Shutdown System and its design bases.

The proposed changes to TS 3/4.3.7.4 do not affect the UFSAR description of the HCGS Remote Shutdown System, its design bases, or performance.

10 CFR 50 Appendix A, GDC 19, "Control Room," states that equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures. Since no physical changes are being made, and current design bases are not being affected, there is no impact on compliance with GDC 19.

4.2 Precedent 4.2.1 License Amendments The changes proposed herein to the Remote Shutdown System are similar to those previously approved by the NRC for the South Texas Project Electric Generating Station. This previous approval is discussed below.

South Texas Project Electric Generating Station By letter dated November 4, 2003 (ADAMS accession ML033140308), as supplemented by letter dated June 29, 2004 (ML041890388), STP Nuclear Operating Company requested NRC approval of a South Texas Project Generating Station TS change to the Remote Shutdown System to be consistent with the requirements of NUREG-1431. The NRC approved the change in License Amendment Nos. 163/152 for STP Units 1 and 2, issued August 20, 2004 (ML042370841). The amendment issued for STP was substantively equivalent to the amendment requested herein for the HCGS, in that it revised the TS associated with the Remote Shutdown System to be consistent with the requirements of NUREG-1431 which is substantively equivalent to the NUREG-1433 TS for the Remote Shutdown System.

4.3 No Significant Hazards Consideration PSEG Nuclear LLC (PSEG) requests approval of a change to the Hope Creek Generating Station (HCGS) Technical Specifications (TS) concerning operability of the Remote Shutdown System. The proposed change would increase the allowed outage time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and with the requirements of NUREG-1433, Standard Technical Specifications - General Electric BWR/4 Plants. It will also delete Tables 3.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION, 3.3.7.4-2, REMOTE SHUTDOWN SYSTEMS CONTROLS, and 4.3.7.4-1, REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS, to be relocated to the Technical Requirements Manual (TRM). Although the Remote Shutdown System component list is relocated from the TS to the TRM, the information being relocated will be controlled and further revisions to the TRM Table will be subject to 10 CFR 50.59.

8

LR-N18-0098 LAR H18-05 Enclosure PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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 TS amendment does not involve potential accident initiators; therefore, there is no significant increase in the probability of an accident previously evaluated.

There is no proposed change to the design basis or configuration of the plant and the extension of the allowed outage time of the Remote Shutdown System functions is consistent with the low probability of an event requiring control room evacuation during the allowed outage time and does not have a significant effect on safety.

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

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

Response: No.

The proposed amendment does not involve physical alteration of the HCGS. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which the HCGS is operated. There are no setpoints at which protective or mitigating actions are initiated that are affected by this proposed action. The change does not alter assumptions made in the safety analysis. This proposed action will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alteration is proposed to the procedures that ensure the HCGS remains within analyzed limits, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.

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

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

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change, which makes the HCGS TS for Remote Shutdown System consistent with the requirements of NUREG-1433, does not exceed or alter a setpoint, design basis or safety limit.

9

LR-N18-0098 LAR H18-05 Enclosure Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based upon the above, PSEG Nuclear concludes that the proposed amendment 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.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

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

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

6.0 REFERENCES

6.1 Hope Creek Generating Station Updated Final Safety Analysis Report 10

LR-N18-0098 Attachment 1 Technical Specification Pages with Proposed Changes

LR-N18-0098 LAR H18-05 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Pages Index viii, ix 3/4.3.7.4 3/4 3-74 thru 3/4 3-83 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumentation.. . . . . . . . . . . ..... ..3/4 3-52 Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints. .. .....3/4 3-54 . .

Table 4.3.5.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements . . . . . . . ...... . . . . . . . . ... .. ...3/4 3-55 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION. . . .. ...

. . . . . . . 3/4 3-56 Table 3.3.6-1 Control Rod Block Instrumentation. . . . . . . . . . .3/4 3-57 Table 3.3.6-2 Control Rod Block Instrumentation Setpoints.. . . . . .

. . . .. .. . . . . . . . . . . . . . . . . . . . . . 3/4 3-59 Table 4.3.6-1 Control Rod Block Instrumentation Surveillance Requirements.... .. . . . . . . . . . . . . . . 3/4 3-60 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation . . . . .. . . . .

. . . . . . 3/4 3-62 Table 3.3.7.1-1 Radiation Monitoring Instrumentation . . . . . . . 3/4 3-63 Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements . . . . . . . . . . . ... . . . .3/4 3-66 Remote 3-74 HOPE CREEK viii Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Ta:Sle ..7.4 1 ReH!ete SatilteS'sJB HeBiteriB§ IBstrtilHISBtatieB. . /4 79 Ta:Sle ..7.4 2 ReH!ete Sb:uteown SysteH!s CoBtrols . . . . . . . /4 77 Ta:Sle 4..7.4 1 ReH!ote Sb:uteo'sffi 4oBitoriB§ IBstrtilHISBtatioB SYFJeillaBoe ReqyireH!eBts Accident Monitoring Instrumentation 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumentation . . . . . 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumentation Surveillance Requirements 3/4 3-87 Source Range Monitors . . 3/4 3-88 3/4.3.8 DELETED . . . . . . . . . . 3/4 3-10 3 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION ..........................3/4 3-10 5 Table 3.3.9-1 Feedwater/Main Turbine Trip System Actuation Instrumentation .. . . . 3/4 3-10 6 HOPE CREEK ix Amendment No.

()li E: --------------------------------------------------------

Separate AClll() entry is allowed for each Function INSTRUMENTATIO 3.3. 7.4 The r mote shutdown system instrumentation and controls shotvn in Table 3.3.7.4 1 shall be OPERABLE.

tli functions I OPERATIONAL CONDITIONS 1 and 2.

r one or more of 1

a. With the number of OPERA,QlE remote shutdown monitoring instrumentation required functions --7' channels less than re'luired by Table 3.3.7.4 1, restore the inoperable s}

inoperable, to OPERABLE status within 7- ay be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

- '---@_Qj b... \"lith the number of OPERAQlE remote shutdo*Nn system controls less than re'luirsd in Table 3.3.7.4 2, restore the inoperable contml(s) to OPERABlE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

normally energized

.f.:

4.3.7.4.1 Each of the above required remote shutdo1Nn *nstrumentation channels shall be demonstrated OPERABLE by performance of HANNEL CHECK and CHANNEl s c ed in the Surveillance Frequency Control I

frequency E:ach required control circuit and transfer switch I 4.3.7.4.2 At l&ast on& of each of the above remote shwtdown control sv.Jitch(es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) in accordance with the Surveillance Frequency Control Program.

4.3.7.4.3 Each required instrumentation channel* shall be demonstrated OPERABLE by performance of a CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.

Safety Rl?lief Valve Position, Standby Diesel Generator Breaker Indication, and Switchgear Room Cooler Status Indication are excluded from CHANNEL CALIBRATION.

HOPE CREEK 3/4 3-74 Amendment No. 487-

TABLE 3.3.7.4-1 REMOTE SHY199WN MONITORING INSTRUMENTATION MINIMUM Pages 3/4 3-75 th rough 3/4 3-84 are deleted INSTRI:JMENTS INSTRI.-.IHENT OPERABLE*

lo Reacter Vessel Pressure 2-: Reaeter 'lessel Water level Safety/Relief Valve Pesitien, (3) valves  !/valve

4. Suppression Chamber water Level

§. Suppression Chamber Water Temperature

6. RHR System Flew 7; Safety Auxiliaries Ceeling System Flew l fh Safety Auxiliaries Ceeling System Temperature l 9: ACIC System Flew RCIC Turbine Speed lh R£1£ Turbine Bearing O.il Pressure Lew Indieatien RCIC High Pressure/lew Pressure Tttfbine Bearing Temperature High Indieatien Cendensate Sterage Tank Leve1 lew-Lew Indicatien Standby Diesel Generator 1AG400 Breaker Indieatien
  • Either primaty loeatien (Remete Shutdewn Panel, ,19£399) eP altePnate lecatien.

HOPE CREEK 3/4 3-75 through 3/4 3-83 Amendment No. XXX I

TABLE 3.3.7.4 1 (Continued)

REMOTE SHUTDOWN MONITORING INSTRUMENTATION MINIMUM INSTRUMENTS INSTRUMENT (Continued OPERABLE*

Stand Diesel Generator 1BG400 Breaker Indication

16. Standby Diese l Generator 1CG400 Breaker Indication 117 Standby Diesel Generator 1DG400 Breaker Indication l Switchgear Room Cooler IAVH401 Status Indication Switchgear Room Cooler 1BVH401 Status Indication l Switchgear Room Cooler IC\401 StatHs Indication l-Switchgear Room Cooler 1DVH401 Status Indication
  • Either primary location (Remote Shutdown Panel 10C399) or alternate location.

TABLE 3.3.7.4 2 E SHUTDOWN SYSTEMS CONTROLS 15 HSS 4410A 15:-HSS-44108 lSV liSS 4410G lSu HSS 44100 ls\-HSS 4410N RGIC SYSTEM RSP RCIC Tt:rbifle Trlpr . /Throttle Valve RCIC TI:II"Bif'le f Val **e St'u::Jto6u .

RCIC Steam S:ppl tbeard Isolat,O ii Va4-ve lFC HV F007 CeAtrel RCIC Steam Sl:lply In boa r d Isoletlon Va+¥e lBO HV F031 Cef'ltrel StipressieA Peel te RCIC Pump Suetlon Va+ve 180 HV FOlO CeMdensate Storage Tank to RCIC Pmp Suction Va l e l8D-SV F019 RCIC P ump D15ehare Mirrimt:m Flew 11* l'"e Water :pply ! alve S

180 II\' F046 180 HV F013 RCIC T:rbit'l RCIC Pump D1seha* ge Cee l flgto Feedwater L1ne 11 "e Iselat1e RCIC Steam '8:* LHte lflbeard Isolai 10n '

lFC HV F076 ..

  • uahe ual "e (2) 180 HV F012(3} RCIC Pump D i ee h r lBDHV*F622 Tes t Returt'l ralveg;e ' ce ae s at e:

Sterae Ta (2) .

lFC Hi' F059 IndieaMofl RCIC Turbine Exhaus t to Suppress1o Peel Valve (2) "ae ttum Pump Discharge lFC H\' F060 IndieatieA CIC Condenser (2)

I ruH eat ion IC Turbine:- Exhau t Outbeard Vacuum lFC HV F062

.. , I *1 *
!
.;:1i :boo ViH-ve lFC HV F084 (
2) IMdieation Breaker l

RCIC Tt!rb ne Ise en '1al"e rd VaEtitim a

(3)

  • t
  • lFC H'i F025 IndicatioA RCIC Condensate Pet Draif'l te Maln
  • lFC H'i' F004 (3) I ndi eati on Condenser Valve RCIC Vacuum Tank Condensate to Clean Rad Waste Pump 180 BP228 (4) If'ldieatief'l h{c} Joekey P::.::P 2 8Vah'e

..P 9P229 lFC OP220 Cef'ltrol RCIC 'laeuum lFC OP219 CoAtrel Tank RCIC Gland Seal Cefl CeeAsel" VaeutJm Pump lFC FIG 4158 GP2+/-9 RCIC System InJee tieR Flew miR SYSTEM

  • RHR Pump BP202 Suet1en F rem Recipe bine lBC*HV F004B Va+Ye RIIR Pump.BP2 2 Suetiefl Frem Suppres51on eel Valve HOPE CREEK

TABLE 3.3.7.4 2 {Continued)

REMOTE SHUTDOI.'"iN SYSTEMS CONTROLS BHR SYSTEM RSP (Cont.l 18G HV F0078 Control RHR Pump 8P202 Minimum Flo'N ¥alva te Suppression Peel 18G HV F0488 Control RHR Loop 8 l=lcat ElEahan§er By13ess V-aWe

+BG I-IV F0168 Control RHR loop 8 Shutdown GooliRg Retum

¥aWe 18G HV F009 Gont:rol RHR Sl1utdown Cooling S:oion Freffi Reotrs line lnboarellselaaen Valve 18C HV F008 Control RHR Shutdovm Cooling S:etion Frorn Reoiro Uno Outboard lsolatien Val*.'e 1BC HV F1228 Control RHR beep B Shutdown Cooling Inj ection Cheek Valve Bypass Valve 1BC HV 4439 Control RHR Discharge to Liquid Radweste Reactor Building Isolation Valve 18C HV F0248 Control RR P1::1rnp BP202 Test Return Valva to SuppressioA Pool 18G HV F0478 Centro! RHR beep B Heat E*ohanger Sl=lell Side lnletVal¥e 18C HV F0038 Control RHR loop B l=leat Exchanger Shell Side Outlet Valve 18C HV F04Q Co ntrol RHR Discharge to liquid Radwaste Inboard lsolatiefl Val'<'e 18G HV F040 Control RHR Discharge to liquid Rad\Vaste Outboard Isolation Valve 1 DC-IIV-F0063> Indication RI=IR Pump AP202 Suction From Reeiro Line Valve 18C-IIV*F010B(31 lndicatien RHR Pl::lffip [)P2G2 Test Return Val\'e to Suppression Pool 1 DC=II\t..F016BC31 Indication RHR loop B Containment Spray Ol:ltboard Isolation Valve 1 BG-IIV-F0278(31 lndieation RHR Loop 8 Suppression Pool Spray Line Isolation Valve 1 OC*IIV*F017B13> lndieation RI=IR Low Pressure Coolant Injection Leop-B Injection Vai\'O 180-1 IV-F004D£21 lndioation RHR Pump DP202 Suction From 81::1ppression Pool Vah*e 1 BC-IIV*F021A'31 Indication RI=IR Loe13 A Containment Spray Inboard Isolation Valve 1 BC*IIl.f*F921813l IRdiaation RHR Loop B Containment Spray Inboard Isolation Vallis 1 BC 8P2Q2 Control RHR Purnp BP202 1BC H88 44168 Control Transfer Switch For RHR Pump 8P202 1 BC-DP228141 Indication EGGS (RHR 8) Joekoy Pump DP228 3/4 a 78 Amendment No.

t-92

Local Control tocal Control ocal Contlol lBC 11'1 F947A lBC II\' F993A Local Control local Contlol Indication A RettJrn Flom TACS Inboard lES

  • 2496£ lES II'*' z512A Local Eontrol RIIR loop 6otlet alwe

TABLE 3.3.7.4 2 (Cotinued)

RBOTE SHUTDOWN SYSTEMS CONTROLS SACS REDUNDANT CONTROLS (Cont.)

lEG AP210 Leeal CMtrel SACS Leo A Pm AP210 lEG CP210 Leeal Cef'ltre1 SACS Loop A Pump CP210 STATION SERVICE WATER SYSTB (SSWS) RSP lEA HV 2204 Control Reactor Auxiliaries Coolin§ S st m (RAGS) Heat Exchanger Supply valte l

(From SACS Loop B) lEA HV 23558 CeRtrol SACS Lee B Heat ExehaRger B2E201 Ot:Jtl et Val lie lEA IIV 23718 CoFitl"ol " SACS Leap 8 lleat Exchanger B1E201 Outlet Valve lEA H'/*23578 CeFitrol SACS leap B ta CeolH\Q Tower ,,va 1 "e lEA HV 21988 Cef'ltrel SSWS Pump BP502 Discharge Vale lEA HV*21980 CeRtrel SSWS Pump OP502 D1Se

  • h a{Je 11al"e * .".

lEA HlJ 21978 Control SSJS Strai Mer BF509 Maul Backwastl lEA*HV 21979 CeFitrel ViH-ve Strainer DF509 Mai f'l Baek'w*asl'l lEA BP502 Cef'ltrel s&,.'S Pump BP502 lEA lEA HSS.2219B OPS02.

Cef'ltrol Centl"el TraFisfer SSWS Pump DPS02 Swite Fer SSWS Pump BPS02 lEA HSS 2219D Cef'ltrol TraFisfel" Switch Fer SSWS Pump DP502 SSWS - REDUNDANT CONTROLS lEA HV 2203 I-.Lo¥eLA-a+l -fC;e.owf'ltt:"l'r"Oof-1 R1RACS-IHeat-t-EE xelehMatnn ger--&Supf3l:)ir' -¥V-a-a1weve (From SACS Loop A) lEA AP502 Leeal CORtrel SSWS Pump AP502 lEA CP502 Local Cef'ltrel ss.*s Pijmp CPS02.

CONTROL AREA CHILLED WATER S¥STE1 (CACWS) RSP lGJ BK400 CeFitrol

  • CeMtrol Area Chil4er BK400 lGJ HSS 96528 Cefltrel
  • TraMsfer Switch Fer Control Area Sffi Her BK400 lGJ BK403 Control Safety Related Panel Room Ciller lGJ HSS 966684 CeFitl"el fer Swite Fer Safety-Related P a ne l Room Ch i ll r BK403 e lGJ BP400 " Cel"'trel Control Area Chilled Water Cirelating Pijmp BP400 lGJ BP414 Central Safety Related Panel Room Chllled Wate r Cireulatif'lg Pump BP414 HOPE CREEK

I li

LR-N18-0098 Attachment 2 Technical Specification Bases Page with Proposed Changes (For information only)

LR-N18-0098 LAR H18-05 TECHNICAL SPECIFICATION BASES PAGE WITH PROPOSED CHANGES The following Technical Specifications Bases for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Page 3/4 3.7.4 B 3/4 3-5

INSTRUMENTATION BASES 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63 and 64.

3/4.3.7.2 DELETED 314.3.7.3 DELETED ISYST :J I 3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION AND CONTROL$

F The OPERABILITY of the remote shutdown mgr:itgriRQ iRstr:meRtatigR aRd sgr:tmls ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1980 and NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. For a discussion of SPIRAL RELOAD and SPIRAL UNLOAD and the associated flux monitoring requirements, see Technical Specification Bases Section 3/4.9.2. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

3/4.3.7.7 DELETED HOPE CREEK B 3/4 3-5 Amendment No. xxx (PSEG Issued)

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