Text

Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control
	ML17045A006
Person / Time
Site:	Dresden
Issue date:	02/10/2017
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-17-029
	Download: ML17045A006 (195)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-17-029 February 10, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 10 CFR 50.90

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (EGC) is submitting a request for an amendment to the Technical Specifications (TS) for Dresden Nuclear Power Station (DNPS),

Units 2 and 3.

The proposed change replaces existing 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 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel (TAF). 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 revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

The proposed change has been reviewed and recommended for approval by the DNPS Plant Operations Review Committee in accordance with the EGC Quality Assurance Program.

Approval of the proposed amendment is requested by August 1, 2017, in order to support the DNPS, Unit 2 refueling outage that is planned for October 2017. Once approved, the amendment shall be implemented within 30 days.

EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official in accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," paragraph (b).

February 10, 2017 U.S. Nuclear Regulatory Commission Page2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10th day of February 2017.

Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments: 1.

2.

3.

4.

Description and Assessment Proposed Technical Specifications Changes (Mark-Up)

Revised Technical Specifications Pages Proposed Technical Specifications Bases Changes (Mark-Up) cc:

NRC Regional Administrator, Region Ill NRG Senior Resident Inspector - Dresden Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

Subject:

Dresden Nuclear Power Station, Units 2 and 3, Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control",

1.0 DESCRIPTION

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration

4.0 ENVIRONMENTAL CONSIDERATION

ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

Exelon Generation Company, LLC (EGC) proposes a change to the Dresden Nuclear Power Station (DNPS), Units 2 and 3 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 2.1.1.3. Safety Limit 2.1.1.3 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 EGC 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. EGC has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the NRC are applicable to DNPS, Units 2 and 3 and justify this amendment for the incorporation of the changes to the DNPS TS.

The following DNPS, Units 2 and 3 TS reference or are related to OPDRVs and are affected by the proposed change:

1.1, Definitions 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.2, Isolation Condenser (IC) System Instrumentation 3.3.6.1, Primary Containment Isolation Instrumentation 3.3.6.2, Secondary Containment Isolation Instrumentation 3.3.7.1, Control Room Emergency Ventilation (CREV) System Instrumentation 3.5.1, ECCS - Operating 3.5.2, ECCS - Shutdown 3.5.3, IC System 3.6.1.3, Primary Containment Isolation Valves (PCIVs) 3.6.4.1, Secondary Containment 3.6.4.2, Secondary Containment Isolation Valves (SCIVs) 3.6.4.3, Standby Gas Treatment (SGT) System

3. 7.4, Control Room Emergency Ventilation (CREV) System 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System 3.8.2, AC Sources - Shutdown 3.8.5, DC Sources - Shutdown 3.8.8, Distribution Systems - Shutdown 2.2 Variations EGC is proposing the following variations from the TS changes described in the TSTF 542 or the applicable parts of the NRC's safety evaluation. These variations do not affect the applicability of TSTF-542 or the NRC's safety evaluation to the proposed license amendment.

Page 1 of 10

ATTACHMENT 1 -DESCRIPTION' AND ASSESSMENT In a few instances, the DNPS TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-542 was based. Specifically, the titles for the following DNPS, Units 2 and 3 TS vary from the STS discussed in TSTF-542:

3.3. 7.1, Control Room Emergency Ventilation (CREV) System Instrumentation 3.3.5.2, Isolation Condenser (IC) System Instrumentation 3.5, Emergency Core Cooling Systems (ECCS) and Isolation Condenser (IC) System 3.5.3, IC System 3.7.4, Control Room Emergency Ventilation (CREV) System 3.7.5, Control Room Emergency Ventilation Air Conditioning (AC) System 3.8.8, Distribution Systems - Shutdown Additionally, the DNPS, Units 2 and 3 TS do not include TS for a Reactor Core Isolation Cooling (RCIC) System or Inverters. These differences are administrative and do not affect the applicability of TSTF-542 to the DNPS TS.

The TSTF-542 Traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). DNPS, Units 2 and 3 were not licensed to the 10 CFR 50, Appendix A, GDC. The DNPS, Units 2 and 3 Updated Final Safety Analysis Report (UFSAR), Section 3.1.1 contains an evaluation of the design basis of DNPS with respect to the first draft of the 70 proposed "General Design Criteria for Nuclear Power Plant Construction Permits" issued by the Atomic Energy Commission in July 1967. The design basis of DNPS, Unit 2 was later evaluated against the final "General Design Criteria for Nuclear Power Plants," published as 10 CFR 50, Appendix A in July 1971.

This evaluation is presented in Section 3.1.2 of the DNPS UFSAR. This difference does not alter the conclusion that the proposed change is applicable to DNPS.

The DNPS TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirement Frequencies for Technical Specifications 3.3.5.2 and 3.5.2 are "In accordance with the Surveillance Frequency Control Program."

The DNPS TS contain the following requirements that differ from the Standard Technical Specifications on which TSTF-542 was based, but are encompassed in the TSTF-542 justification:

There are DNPS specific instrumentation functions that differ from the Standard Technical Specifications (STS). Changes to these instrumentation functions are justified by the discussion in Section 3.4.1 of the TSTF-542 justification. DNPS TS Table 3.3.5.1, Functions 1.e and 2.e describe pump start time delay relays for the core spray (CS) and low pressure coolant injection (LPCI) pumps. The purpose of these time delays is to stagger the automatic start of CS and LPCI pumps that are in each Division 1 and 2, thus limiting the starting transients on the 4.16 kV emergency buses. This staggering is unnecessary for manual operation; therefore, these Functions can be removed from the TS because the required ECCS subsystem is proposed to be started by manual operation.

The DNPS, Units 2 and 3 design features a shutdown cooling (SOC) system that is dedicated to the removal of core decay heat and maintaining the temperature of the reactor coolant. Residual heat removal is not a mode of another system like LPCI; therefore, the Page 2of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT potential for system misalignments are less likely for the DNPS SOC system. The DNPS, Units 2 and 3 TS contain a specific Function that provides automatic isolation of the SOC system penetration flow paths below the TAF on low RPV water level in Modes 4 and 5, which can be credited in Drain Time calculations. Inclusion of this instrumentation Function is justified by the discussion in Section 3.3 of the TSTF-542 justification. Specifically, the definition of Drain Time allows crediting the closing of penetration flow paths that are capable of being automatically isolated by RPV water level isolation instrumentation prior to the RPV water level being equal to the TAF. The Reactor Vessel Water Level - Low Function is only required to be operable when automatic isolation of the associated SOC penetration flow path is credited in calculating Drain Time.

Reactor Vessel Water Level - Low signals are initiated from four level transmitters (i.e., two per trip system) that sense the difference between the pressure due to a constant column of water (i.e., reference leg) and the pressure due to the actual water level (i.e., variable leg) in the vessel. One channel per trip system is required to be operable when automatic isolation of the associated penetration flow path is credited in calculating Drain Time to meet LCO 3.5.2.

This Function was moved from TS 3.3.6.1, Function 6.b to TS 3.3.5.2, Function 3.a. The following changes are proposed:

o The Applicability is changed. The TS 3.3.6.1 Applicability for this Function is Modes 3, 4, and 5. The proposed TS 3.3.5.2 Applicability is "when automatic isolation of the associated penetration flow path(s) is credited in calculating Drain Time." In other words, if the Drain Time calculation assumes the SOC System would be automatically isolated, this Function must be operable to perform that function. This is consistent with the definition of Drain Time and the TS 3.5.2 requirements. The TS 3.3.6.1 Applicability in Mode 3 is retained to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel in Mode 3. In Modes 1 and 2, another isolation (i.e., Recirculation Line Water Temperature-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.

o The TS 3.3.6.1 Required Action 1.1 and 1.2 for an inoperable channel is to immediately initiate action to restore the channel to operable status or to immediately initiate action to isolate the SOC system. The TS 3.3.5.2 Action B requires declaring the associated penetration flow path(s) incapable of automatic isolation. This Action would require Drain Time to be recalculated without crediting automatic isolation of the affected penetration flow paths. The proposed Actions are consistent with the definition of Drain Time and the requirements of LCO 3.5.2. The TS 3.3.6.1 Required Action 1.2 is retained to account for SOC system flowpath isolation in Mode 3 as discussed above.

o A Channel Check and Channel Functional Test are required at the existing Frequency.

A calibration of the trip unit, Channel Calibration, Logic System Functional Test, and Isolation System Response Time tests are no longer required in Modes 4 and 5, as discussed in Section 3.3.3 of the TSTF-542 Justification.

Page 3of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT DNPS, Units 2 and 3 do not currently have the capability to perform Channel Checks for proposed Table 3.3.5.2-1 Functions 1.a, "Reactor Steam Dome Pressure-Low (Permissive}, 1.b, "Core Spray Pump Discharge Flow-Low (Bypass)," 2.a, "Reactor Steam Dome Pressure-Low (Permissive)," and 2.b, "Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)." The current DNPS, Units 2 and 3 TS do not include Channel Checks for these functions; therefore, no Channel Check Surveillance Requirement (SR) was added for these functions.

DNPS, Units 2 and 3 each have two LPCI loops with two pumps in each loop (i.e., four LPCI pumps per unit). The DNPS TS currently require one operable channel for each LPCI loop for the "Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)," Function.

EGC proposes to maintain the one channel per loop for Technical Specification 3.3.5.2, Function 2.b in lieu of one channel per pump as described in the STS, since at DNPS, there is only one flow transmitter for each loop that monitors the flow of both pumps in that loop (i.e., two LPCI loop flow transmitters per unit).

The current DNPS, Units 2 and 3 TS do not include a manual initiation logic function for the CS or LPCI subsystems. This is due to the fact that the design of DNPS, Units 2 and 3 does not include this feature. Therefore, since this function does not exist at DNPS, manual initiation functions for LPCI and CS are not being included in Technical Specification 3.3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," Table 3.3.5.2-1. Additionally, since the manual initiation functions are not included in Table 3.3.5.2-1, the associated Logic System Functional Test would likewise not be required for TS 3.3.5.2; therefore, TS 3.3.5.2 as proposed for DNPS, Units 2 and 3 does not include a Logic System Functional Test SR.

As an alternative, EGC proposes that Technical Specification 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control," include an SR to verify that the DNPS, Units 2 and 3 LPCI and CS subsystem can be manually operated through the manipulation of subsystem components from the Main Control Room (i.e., proposed SR 3.5.2.7 shown in Figure 1 below).

SR 3.5.2.7

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

Vessel injection/spray may be excluded.

Verify the required ECCS injection/spray subsystem can be manually operated.

Figure 1: Proposed SR 3.5.2. 7 In accordance with the Survei 1 lance Frequency Control Program The manual operation of the LPCI and CS subsystems for the control of reactor cavity or RPV inventory are relatively simple evolutions and involve the manipulation of a small Page 4of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT number of components. These subsystem alignments can be performed by licensed operators from the Main Control Room as described below. This alternative is justified by the fact that a draining event is a slow evolution when compared to a design basis loss of coolant accident (LOCA), which is assumed to occur at full power, and thus there is adequate time to take manual actions (i.e., hours versus minutes). Adequate time to take action is assured since the proposed Technical Specification 3.5.2, Condition E, prohibits plant conditions that result in Drain Times that are less than one hour. Therefore, there is sufficient time for the licensed operators to take manual action to stop an unanticipated draining event, and to manually start an ECCS injection/spray subsystem or the additional method of water injection. Consequently, there is no need for manual initiation logic to actuate the required subsystem components.

RPV and Reactor Cavity Inventory Control Using LPCI:

a.

Start one Containment Cooling Service Water (CCSW) Pump in each loop (if available) AND verify 2(3)-1501-3 A/B, "Unit 2(3) Containment Cooling Heat Exchanger A(B) Tube Side Discharge Motor Operated Valves," open.

b.

Start desired LPCI pump.

c.

Close 2(3)-1501-21A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Outboard Isolation Valve."

d.

Open 2(3)-1501-22A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Inboard Isolation Motor Operated Valve."

e.

Throttle open 2(3)-1501-21A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Outboard Isolation Valve," to achieve the desired level in the reactor pressure vessel or reactor cavity.

f.

When LPCI is no longer required for inventory level control then:

(1)

Close 2(3)-1501-21 A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Outboard Isolation Valve."

(2)

Close 2(3)-1501-22A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Inboard Isolation Motor Operated Valve."

(3)

Open 2(3)-1501-21A(B), "Unit 2(3) LPCI Loop 1(11) Coolant Injection Outboard Isolation Valve."

(4)

Stop LPCI pump.

(5)

Stop CCSW pumps started in Step a.

Page 5of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT RPV and Reactor Cavity Inventory Control Using Core Spray:

a.

Start Unit 2(3)A(B) Core Spray Pump.

b.

Throttle open, MO 2(3) 1402 25A(B), "Core Spray Pump Discharge Valve," to achieve the desired level in the reactor pressure vessel or reactor cavity.

c.

When Core Spray is no longer required for inventory control then:

(1)

Close MO 2(3) 1402 25A(B), "Core Spray Pump Discharge Valve."

(2)

Stop Core Spray Pump.

Since the LPCI and CS subsystems can be placed in service using manual means in a short period of time (i.e., within the timeframes assumed in the development of TSTF-542), using controls and indications that are readily available in the Main Control Room, manual operation of the required subsystem would be an equivalent alternative to system initiation via manual initiation logic.

The DNPS design provides for isolation of the reactor water cleanup (RWCU) system at Reactor Vessel Water Level-Low versus Reactor Vessel Water Level-Low Low; therefore, proposed Table 3.3.5.2-1 reflects this RWCU isolation difference.

The automatic isolation on Reactor Vessel Water Level-Low functions for the RWCU and SOC systems at DNPS, Units 2 and 3 varies slightly from the system described in the STS.

These functions receive input from four reactor vessel water level channels. Each channel inputs into one of four trip stings, and two trip strings make up a trip system. The trip systems are aligned in a parallel configuration, so both trip systems must trip in order to cause an isolation of the RWCU or SOC system valves. Any channel will trip its associated trip string and trip system. Therefore, both trip systems with one trip string in each trip system is required to provide for automatic RWCU and SOC system isolation. Proposed Table 3.3.5.2-1 has been revised to reflect the DNPS, Units 2 and 3 requirement for one operable channel in each Reactor Vessel Water Level-Low isolation trip system for RWCU and SOC.

The DNPS TS do not currently contain a Note applicable to LCO 3.5.2 regarding realignment to the Low Pressure Coolant Injection mode. This Note is not applicable to DNPS, Units 2 and 3 because of the separate SOC system that provides for the removal of reactor residual heat. This has no effect on the adoption of the TSTF-542.

At DNPS, verification of Suppression Pool and Contaminated Condensate Storage Tank volumes is contained in a single SR (i.e., SR 3.5.2.2) versus in separate SRs in the STS.

This does not affect the applicability of TSTF-542.

EGC proposes to delete DNPS, Units 2 and 3 TS 3.6.1.3, Condition F and all of its associated Required Actions as shown in Figure 2 below. The Applicability for TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs);" is When associated instrumentation is required to be OPERABLE per LCO 3. 3. 6. 1, "Primary Containment Isolation Page 6 of 10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT Instrumentation." This change is justified since OPDRV requirements have been deleted, and Mode 4 and 5 PCIV requirements have been relocated from TS 3.3.6.1 and 3.6.1.3 to the proposed TS 3.3.5.2. Thus, there are no longer any PC IVs required to be operable by TS 3.6.1.3 during Mode 4 or 5. These requirements are addressed by the proposed TS 3.3.5.2 in their entirety. Following the removal of OPDRV and relocation of Mode 4 and 5 requirements as discussed above, this Condition and associated Actions in TS 3.6.1.3 would never be applicable; therefore, are no longer necessary.

ACTIONS i:: _

Re El 1-1 i re El A ct i o R a R El

..... F ~

- lr------+I ++-A i.....

t.... i.... ii..... t.,..9..... ii..,s....,t..... i..,g A

.......... t-g assosiates Gol!l~letioA sw&~9A9 0~9ratioA&

Til!le of CoAE:litioA A, 1.*itl:1 a ~oteAtial for B, C, or g Rot lllet for E:lraiAiP§ tRe reastor PGIV(s) reEJ1-.1ireE:I to be

'lessel (QPgR'!s).

OPeRABLe Bl-JFiR§ MQge 4

~

i::. 2 lRitiate astioR to restore valve(s) to OP~RABL~ stat1-.1s.

PC I Vs 3.6.1.3 CQl"1PbHIOW Til"1[

lFRFReEliately Iffiflleeiately Figure 2: Proposed Variation That Deletes Technical Specification 3.6.1.3 Condition F The DNPS Control Room Emergency Ventilation (CREV) and the Control Room Emergency Ventilation Air Conditioning (AC) systems (i.e., Technical Specifications 3.7.4 and 3.7.5, respectively) provide Control Room habitability functions. Changes to the TS controls on these systems is justified by the discussion in Section 3.4.3 of the TSTF-542 justification.

Specifically, these DNPS specific systems provide similar Control Room habitability functions as those described in the STS, and changes to these Technical Specifications are similarly justified.

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ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Exelon Generation Company, LLC (EGC) requests adoption of Technical Specifications Task Force Traveler (TSTF)-542 "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (STS), into the Dresden Nuclear Power Station, Units 2 and 3 Technical Specifications (TS). The proposed amendment replaces the existing requirements in the 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 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

EGC 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 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 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 Mode 4 or Mode 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 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 Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 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 Page 8of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT 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 secondary 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 2.1.1.3. 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.

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 2.1.1.3. 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 Page 9of10

ATTACHMENT 1 -DESCRIPTION AND ASSESSMENT 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, EGC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 1 O CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL CONSIDERATION

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 1 O 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 1 O 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.

Page 10of10

Dresden Nuclear Power Station, Units 2 and 3, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 2 - PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)

TOC Page i TOC Page ii 1.1-2 1.1-3 1.1-4 1.1-5 1.1-6 1.1-7 3.3.5.1-2 3.3.5.1-3 3.3.5.1-5 3.3.5.1-10 3.3.5.1-11 3.3.5.1-12 3.3.5.1-13 3.3.5.1-14 3.3.5.2-1 3.3.5.2-2 3.3.5.2-3 3.3.5.3-1 3.3.5.3-2 3.3.6.1 -8 3.3.6.2-5 3.3.7.1-1 3.5.1 -1 3.5.2-1 3.5.2-2 3.5.2-3 3.5.2-4 3.5.2-5 3.5.2-6 3.5.3-1 3.6.1.3-5 3.6.1.3-6 3.6.4.1-1 3.6.4.2-1 3.6.4.2-3 3.6.4.3-1 3.6.4.3-2 3.7.4-1 3.7.4-2 3.7.5-1 3.8.2-3 3.8.2-4 3.8.5-2 3.8.8-2

TABLE OF CONTENTS

1. 0 1.1

1. 2

1. 3 1.4

2. 0

2. 1 2. 2

3. 0 3.0

3. 1 3. 1.1 3.1. 2 3. 1. 3 3. 1. 4 3. 1. 5 3.1. 6 3. 1. 7 3. 1. 8 3.2

3. 2.1

3. 2.2

3. 2.3 3. 3 3.3.1.l

3. 3.1.2

3. 3.1.3 3. 3. 2. 1 3.3. 2.2

3. 3. 3.1

3. 3. 4.1 USE AND APPLICATION Definitions.............................................. 1.1 -1 Logica l Connectors....................................... 1.2-1 Comp l etion Times......................................... 1. 3-1 Frequency................................................ 1. 4-1 SAFETY LIMITS (SLs)

SLs...................................................... 2. 0-1 SL Violations............................................ 2.0-1 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY........ 3. 0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY................. 3. 0-4 REACTI VITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM).................................... 3.1.1 -1 Reactivity Anomalies..................................... 3. 1.2-1 Control Rod OPERABILITY.................................. 3. 1.3-1 Control Rod Scram Times.................................. 3.1.4 -1 Control Rod Scram Accumulators........................... 3. 1. 5-1 Rod Pattern Control...................................... 3. 1. 6-1 Standby Liquid Control (SLC) Sy stem...................... 3.1. 7-1 Scram Discharge Volume (SDV) Vent and Drain Valves....... 3.1. 8-1 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GEN ERATION RATE (APLHGR)...... 3. 2.1 -1 MINIMUM CRITICAL POWER RATIO (MCPR)...................... 3. 2. 2-1 LINEAR HEAT GENERATION RATE (LHGR)...................... 3. 2.3 - 1 INSTRUMENTATION Reactor Protection System (RPS) In strumentation.......... 3.3. 1.1-1 Source Range Monitor (SRM) Instrumentation............... 3. 3. 1. 2-1 Oscillation Power Range Monitor (OPRM) Instrumentation... 3. 3. 1.3-1 Control Rod Block Instrumentation........................ 3. 3.2.1-1 Feedwater System and Main Turbine High Water Level Trip Instrumentation................................... 3.3. 2. 2-1 Post Accident Monitoring (PAM) Instrumentation........... 3. 3.3.1-1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation................... 3.3.4.1-1

3. 3.5.1 Emergency Core Cooling System (ECCS) Instrumentation..... 3.3. 5.1-1 3. 3. 5. 2 I sol ati on Condl'.rnserReactor Pressure Vessel

( ~RPV ) S,ysteR1 InstrYR1entationWater Inventory Control Instrumentation............... 3.3.5. 2-1

3. 3. 5.3 Isolation Condenser (JC) System Instrumentation.......... 3. 3. 5.3-1 3.3. 6. 1 Primary Containment Isolation Instrumentation............ 3.3.6.1-1

3. 3.6. 2 Secondary Containment Isolation Instrumentation.......... 3.3.6. 2-1

3. 3.6. 3 Relief Valve Instrumentation............................. 3.3. 6.3-1

3. 3. 7. 1 Control Room Emergency Ventilation (CR EV ) System Instrumentation........................................ 3.3. 7. 1-1 3. 3. 7. 2 Mechanical Vacuum Pump Trip Instrumentation.............. 3.3. 7.2-1 (continued)

Dresden 2 and 3 Amendment No.

TABLE OF CONTENTS

3. 3 3.3.8.1

3. 3.8.2 3.4 3.4.1 3.4. 2 3.4. 3 3.4.4 3.4.5 3.4.6 3.4. 7 3.4.8 3.4.9 3.4.10

3. 5

3. 5.1

3. 5. 2

3. 5.3 3.6

3. 6.1.1

3. 6.1. 2 3.6.1. 3 3.6. 1.4 3.6. 1.5

3. 6. 1.6

3. 6. 1.7

3. 6.1. 8 3. 6. 2. 1 3.6.2.2

3. 6.2. 3

3. 6. 2.4 3.6. 2.5 3. 6. 3. 1

3. 6.4. 1 3.6.4. 2 3.6.4.3 INSTRUMENTATION (continued)

Loss of Power ( LOP ) Instrumentation...................... 3.3.8.1-1 Reactor Protection System (RPS) Electric Power Monitoring............................................. 3.3.8.2-1 REACTOR COO LANT SYSTEM (RCS)

Recirculation Loops Operating............................ 3.4. 1-1 Jet Pumps................................................ 3.4. 2-1 Safety and Relief Valves................................ 3.4.3-1 RCS Operational LEAKAGE.................................. 3.4.4-1 RCS Leakage Detection Instrumentation.................... 3.4. 5-1 RCS Specif i c Activity.................................... 3.4. 6-1 Shutdown Cooling (SOC) System-Hot Shutdown............... 3.4.7 -1 Shutdown Cooling (SOC) System-Cold Shutdown.............. 3.4.8-1 RCS Pressure and Temperature (P/T) Limits................ 3.4. 9-1 Reactor Steam Dome Pressure.............................. 3.4.10-1 EMERGENCY CORE COOL IN G SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)

WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM EC C S-0 p e r a t i n g........................................... 3. 5. 1 - 1 EGGS Shbltaoi,.rnRPV Water Inventory Control................. 3.5. 2-1 IC Sys tern................................................ 3. 5. 3 - 1 CONTAINMENT SYSTEMS Primary Containment...................................... 3.6.1.1-1 Primary Containment Air Lock............................. 3. 6.1.2-1 Primary Containment Isolation Valves (PCIVs)............. 3. 6.1. 3-1 Drywell Pressure......................................... 3.6. 1.4-1 Drywell Air Temperature.................................. 3.6.1. 5-1 Low Set Relief Valves.................................... 3. 6. 1.6-1 Reactor Building-to-Suppression Chamber Vacuum Breakers............................................... 3. 6. 1. 7 - 1 Suppression Chamber-to-Drywell Vacuum Breakers........... 3.6. 1.8 -1 Suppression Pool Average Temperature..................... 3.6.2. 1-1 Suppress i on Pool Water Level............................. 3.6. 2.2-1 Suppression Pool Cooling................................. 3.6.2.3-1 Suppression Pool Spray................................... 3.6.2.4-1 Drywell-to-Suppression Chamber Differential Pressure..... 3. 6.2.5-1 Primary Containment Oxygen Concentra t ion................. 3. 6. 3.1 -1 Secondary Containment.................................... 3. 6.4.1-1 Secondary Containment Isolation Valves (SCIVs)........... 3. 6.4. 2-1 Standby Gas Treatment (SGT) System....................... 3.6.4. 3-1 (continued)

Dresden 2 and 3 i i Amendment No.

Definitions

1. 1 1.1 Definitions (continued)

CHANNE L FUNCTIONAL TEST CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT I - 131 Dresden 2 and 3 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with t he vessel head removed and fuel in the vessel.

The fo 11 owing 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.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle.

These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5. 6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The dose conversion factors used for this calculation shall be the inhalation committed dose conversion factors in Federal Gui dance Report 11, "Li miting Va l ues of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

(continued) 1.1 -2 Amendment No.

1.1 Definitions DRAIN TINE Dresden 2 and 3 Definitions 1.1 The DRAIN TINE 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 is 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.

(continued)

1. 1-3 Amendment No.

1.1 Definitions DRAIN TIME (continued)

LEAKAGE Dresden 2 and 3 Definitions 1.1 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.

LEAKAGE sha 11 be :

a.

Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank ; or

2. LEAKAGE into the drywel 1 atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b.

Unidentified LEAKAG E All LEAKAGE into the drywell that is not identified LEAKAGE; c.

Tota 1 LEAKAGE Sum of the identified and unidentified LEAKAGE; and

d.

Pressure Boundary LEA KAGE LEAKAGE through a nonisolab l e fau l t i n a Reactor Coolant System CRCS) component body, pipe wall, or vessel wall.

(continued) 1.1 -4 Amendment No.

Definitions

1. 1 1.1 Definitions (continued)

LINEAR HEAT GENERATION RATE (LHGR)

LOGIC SYSTEM FUNCTIONAL TEST MINIMUM CRITICAL POWER RATIO CMCPR )

MODE OPERABLE-OPERABILITY Dresden 2 and 3 The LHGR shall be the heat generation rate per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

A LOGIC SYSTEM FUNCTIONAL TEST shall be a te st of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

The MCPR shall be the smallest critical power ratio (CPR ) that exists in the core for each cla ss of fuel.

The CPR i s that power in the assembly that is calculated by applica t ion of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

A MODE shall correspond to any one inclu sive combination of mode switch position, average reactor coolant temperature, and reactor ve ssel head closure bolt tensioning specified in Table 1.1 -1 with fuel in the reactor ve ssel.

A system, subsy stem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function( s ) are also capable of performing their related support function( s ).

(continued) 1.1-5 Amendment No.

Definitions 1.1 1.1 Definitions (continued)

RATED THERMAL POWER

( RTP)

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN CSDM)

THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TIME Dresden 2 and 3 RTP shal l be a total reactor core heat transfer rate to the reactor coolant of 2957 MWt.

The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact until the opening of the trip actuator.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is~ 68°F, corresponding to the most reactive state; and

c.

All control rods are fully inserted except for the sing le control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER shall be the total reactor core heat tran sfer rate to the reactor coolant.

The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

1. 1-6 Amendment No.

MODE TITLE 1

Power Operation 2

Startup 3

Hot Shutdown '*>

4 Cold Shutdown '*>

5 Refuel i ng Cb>

Table 1. 1-1 (page 1 of 1)

MODES REACTOR MODE SWITCH POSITION Run Refuel ca> or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel Definitions 1.1 AVERAGE REACTOR COOLANT TEMPERATURE

( o F)

NA NA

> 212

~ 212 NA (a)

All reactor vessel head closure bolts fully tensioned.

(b)

One or more reactor vessel head closure bolts less than fully tensioned.

Dresden 2 and 3 1.1-7 Amendment No.

ACTIONS CONDITION B.

As required by Required Action A.1 and referenced in Table 3.3.5.1 -1.

Dresden 2 and 3 B. 1 ECCS Instrumentation

3. 3. 5.1 REQUIRED ACTION

- - - - - - - -NOTE.&- - - - - - - -

1. Only applicable in MODtS 1, 2, and 3.

.i.--Only applicable for Functions 1.a, 1.b, 2.a, 2. b, 2.d, and 2. j.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions B.2

NOTE---------

B.3 Only applicable for Functions 3.a and 3.b.

Declare High Pressure Coolant Injection (HPCI) System inoperable.

Pl ace channel in trip.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued) 3.3.5.1-2 Amendment No.

ACTIONS CONDITION C.

As required by Required Action A.l and referenced in Table 3.3.5. 1-1.

Dresden 2 and 3 c. 1 ECCS Instrumentation 3.3. 5.1 REQUIRED ACTION

- - - - - - - - NOTE&- - - - - - - -

1. Onl,y applicable in ti400ES 1, 2, anti 3.

.&:---Only applicable for Functions l.c, l.e, 2. c, 2. e, 2.g, 2.h, 2.i, and 2.k.

Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divi sions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. 2 Restore channel to OPERABLE status.

(continued) 3.3.5. 1-3 Amendment No.

ACTIONS CONDITION E.

As required by Required Action A. l and referenced in Table 3. 3.5. 1-1.

Dre sden 2 and 3 E. 1 E. 2 ECCS Instrumen t ation 3. 3. 5.1 REQUIRED ACTION

- - - - - - - - NOTE.g.- - - - - - - -

1.

0Rly applicable iR MODES 1, 2, aRd 3.

&.--Only applicable for Functions l.d and 2.f.

Declare supported feature(s) inoperable when its redundant feature ECCS initiati on capability i s inoperable.

Re store channel to OPERABLE status.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for subsystems in both divi si ons 7 day s (continued )

3. 3. 5. 1-5 Amendment No.

FUNCTION

1.

Core Spray System

a.

Reactor Vessel Water Level-Low Low

b.

Orywe l l Pressure-High

c.

Reactor Steam Dome Press ure-Low (Permi ssive)

d.

Core Spray Pump Discharge Flow-Low (Bypa ss )

e.

Core Sp ray Pump Start -Time Delay Relay

2.

Low Pressure Coola nt Injecti on CLPC!l Sys tem

a.

Rea ct or Vesse l Wa ter Level-Low Low

b.

Orywe l l Pressure-High

c.

Reactor Steam Dome Press ure-Low (Permissive)

ECCS In strumentation

3. 3. 5. 1 Table 3.3.5. 1-1 (page 1 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE MOO ES OR OTHER SPECIF! ED CONO!TlONS 1,2, 3, 1,2,3 I,2,3 1,2, 3, 1, 2,3 1,2, 3, l, 2. 3 1,2, 3 REQUI RED CHANNELS PER FUNCTION 4'""'

4'""'

2 1 per pump I per pump 2

CONDITIONS REFERENCED FROM REQUIRED ACTION A. 1 B

B c

c B

B c

SURVEILLANCE REQU I REMENTS SR 3.3.5. l. 1 SR 3.3.5. l.2 SR 3.3.5.l.3 SR 3.3.5.1.5 SR 3.3.5. l.6 SR 3.3.5.l.2 SR 3.3.5.l.4 SR 3.3.5. l.6 SR 3.3.5.l.2 SR 3.3.5.l.4 SR 3.3. 5. l.6

~R J.J.§.1.2

~R J.J.§.1.4 SR 3.3.§.1.e SR 3.3.5.l.2 SR 3.3.5. l. 3 SR 3.3.5.l.5 SR 3.3.5. 1.6 SR 3.3.5. 1.5 SR 3.3.5.1.6 SR 3.3.5.l.1 SR 3.3.5. l. 2 SR 3.3.5. l. 3 SR 3.3.5.l.5 SR 3.3.5.l.6 SR 3.3.5.l.2 SR 3.3.5. 1.4 SR 3.3.5.l.6 SR 3.3.5.l. 2 SR 3.3. 5. l. 4 SR 3.3.5.. 6

~R :LL§.

. 2

~R J.J.e. 1. 4

~R J.J.e.l.e ALLOWABLE VALUE

, -54. 15 inches s 1.81 psig

, 308.5 psig and s 34 l. 7 psi g JQ!l.§ JlSig

-4 341.7 JlSi§

, 802 gpm and s 992 gpm s 11.0 seconds

,. 54. 15 Inches s 1.81 psig

, 308.5 ps ig and s 341. 7 psi g

-4 341.7 ~51§ (continued)

(Da ) Al so required to initiate the associated die se l generator COG).

Dres den 2 and 3 3.3.5.1 -10 Amendment No.

FUN CT JON

2.

LPC I Syst em (continued)

d.

React or St eam Dome Press ure-Low (Brea k Detecti on) e.

Low Press ure Coolant Injection Pump Start-Time Delay Relay Pumps B and D f.

Low Press ure Coolant Inj ec ti on Pump Di scharge Fl ow-Low (Bypass)

g.

Recircul ati on Pump Differenti al Press ure-Hi gh (B reak Detection)

h.

Recircul ation Ri se r Different i a 1 Press ure-Hi gh <Brea k Det ecti on) i.

Reci rcu l ati on Pump Diffe rent i a 1 Pressure Time Del ay-Rel ay

<Break Det ec ti on)

j.

Rea ct or St eam Dome Press ure Time Del ay-Rel ay (Brea k Det ecti on)

k.

Recirculati on Rise r Di ffere nt i a 1 Press ure Time Delay-Relay

<Brea k Detecti on)

Dresden 2 and 3 ECCS Instrumentation 3. 3. 5.1 Tabl e 3.3.5.1 -1 (page 2 of 5)

Emergency Core Cooling Syst em Instrumentati on APPLICABLE MODES OR OTHER SPEC I FI ED CONDITIONS I, 2, 3 1,2, 3, 1,2,3, 1,2,3 1,2,3 1,2,3 1> 2 > 3 1, 2,3 REOUI RED CHANNELS PER FUNCTION 1 per pump 1 per l oop 4 per pump 2

2 2

3.3. 5.1 -1 1 CONDITIONS REFERENCED FROM REQUIRED ACTION A. 1 B

c c

B c

SURVEILLANCE REQUIREMENTS SR 3.3.5. 1.2 SR 3.3.5. l.5 SR 3.3.5. 1.6 SR 3.3. 5. l. 5 SR 3.3. 5. l.6 SR 3.3.5.1. 2 SR 3.3. 5. l. 3 SR 3.3.5. l. 5 SR 3.3.5. 1. 6 SR 3.3. 5. l. 2 SR 3.3.5. l. 5 SR 3.3. 5. 1. 6 SR 3.3.5.1. 2 SR 3.3.5.1.5 SR 3.3. 5. l. 6 SR 3.3.5. 1. 5 SR 3.3.5. 1. 6 SR 3.3. 5. 1. 5 SR 3.3.5.1. 6 SR 3.3.5.1. 5 SR 3.3.5. 1. 6 ALLOWABLE VA LUE

,, 802 psig and s 898 psig s 5.5 seconds

,, 1107 gpm s 5.9 psid s 2. O psi d s 0.53 second s s 2. 12 second s s 0. 53 second s

<continued)

Amendment No.

ECCS In strumentati on

3. 3. 5.1 Table 3.3.5. 1-1 (page 3 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR RE QU I RED REFERE NCED OTHER CHANNELS FROM SPECIFIED PER REQU I RED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REOU I REM EN TS VALUE 3.

Hig h Pre ss ure Coolant Injection (HPCI) System

a.

Reactor Vesse l Wa ter I

B SR 3.3.5.1. 1

., - 54. 15 Leve l - Low Low SR 3.3.5. 1. 2 Inche s 2<.wi l, 3 c.0>

SR 3.3.5.1. 3 SR 3.3.5.1.5 SR 3.3.5. 1. 6

b.

Orywe 11 Pressure-High 1,

B SR 3.3.5.1.2

$ 1.81 psig SR 3.3.5.1.4 2<.0J, 3 <.0l SR 3.3.5.1.6

c.

Reactor Vessel Water

1.

2 c

SR 3.3.5. 1. 1 s 46. 2 inches Leve l - High SR 3.3.5.1.2 2<..ti >, 3<.0J SR 3.3. 5.1. 3 SR 3.3.5. 1.5 SR 3.3.5. 1.6 d.

Contaminated 1,

2 pe r CCST 0

SR 3.3.5. 1. 2

., 11.1158 ft Condensate Storage SR 3.3.5.1.5 for CCST 213 A Tank ( CCSTl Le ve 1-Low 2'"" '

3<< 0J SR 3.3.5.1.6 and

, 7.5637 ft for CCST 213 B

e.

Suppress i on Poo l Water 1,

2 0

SR 3.3.5.1.2

$ I 5 ft Le ve l - High SR 3.3.5.1. 5 5.625 inches 2<Wi l '

3 < ~ti l SR

3. 3.5. 1. 6

f.

High Pressure Coolant 1,

SR 3.3.5. 1. 2

, 616 gpm Injection Pump SR 3.3.5.1.4 Di scharge Flow-Low 2<Wi l ' 3 <e-b )

SR 3.3.5.1. 6

<Bypass l g.

Manual Initiation

1.

c SR 3.3.5.1.6 NA 2<<1>J. 3'*b )

(continued)

( e b ) With reacto r st eam dome pre ss ure > 150 psig.

Dres den 2 and 3 3. 3. 5.1 - 12 Amendment No.

ECCS In strumentati on

3. 3. 5.1 Table 3.3.5.1 - 1 (page 4 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR RE QUI RED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUN CTI ON CONDITIONS FUNCTION ACTION A. 1 REQUIREMENTS VA LUE

4.

Automatic Depress urization System (ADS) Trip System A

a.

Reactor Vessel Water I,

2 SR 3. 3. 5.1. I

., -54. 15 Level - Low Low SR 3.3.5. 1.2 Inches 2<..,. 3<**>

SR 3.3.5. 1.3 SR 3.3.5. 1.5 SR 3.3.5.1.6

b.

Drywel l Pressure-High 1,

2 F

SR 3.3.5.1.2

$ 1.81 psig SR 3.3.5. 1.4 2<<1>>, 3<<*>

SR 3.3.5. 1.6

c.

Automatic

1.

G SR 3.3.5. 1.5 s 113 seconds Depressurization SR 3.3.5.1.6 System Initiation 2<-" >' 3<<1>)

Timer

d.

Core Spray Pump

1.

2 G

SR 3.3.5. 1. 2

., 101. 5 psig Di sc harge SR 3. 3.5.1.4 and Pressure -High 2<..ti >) 3<<*>

SR 3.3.5. 1.6

$ 148.5 psig

e.

Low Press ure Coolant

1.

4 G

SR 3.3.5.1.2

., 101. 5 psig Injecti on Pump SR 3.3.5.1.4 and Di scharge 2 Cb*l,

3 1-i-b )

SR 3.3.5. 1. 6

$ 148. 5 psig Pressure -High

f.

Automatic

1.

G SR 3.3.5. 1.5 s 580 second s Depressur i za tion SR 3.3.5.1.6 System Low Low Wa ter 2<<1>>, 3<<1>>

Leve l Actuation Timer (continued)

( ~b l With reactor steam dome pressure > 150 psig.

Dres den 2 and 3 3.3. 5. 1-13 Amendmen t No.

ECCS Ins trumentati on

3. 3. 5.1 Table 3.3.5. 1-1 (page 5 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR RE QUI RED REFERENCED OTHER CHANNELS FROM SPEC! Fl ED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A. l REQUIREMENTS VA LUE

5.

ADS Trip Sys tem B

a.

Rea ctor Vesse l Wa ter I,

2 F

SR 3.3.5. 1. 1

"' - 54. 15 Level - Low Low SR 3.3. 5. 1.2 inches 2<H>>, 3<< 0>

SR 3.3.5. 1. 3 SR 3.3.5.1.5 SR 3.3.5.1.6

b.

Drywe l l Pressure-Hi gh 1,

2 SR 3.3.5. 1. 2 s 1.81 psig SR 3.3.5. 1.4 2<H> ). 3.. *>

SR 3.3.5.J.6 C.

Automati c 1'

G SR 3.3. 5.1.5 s 11 3 seconds Depress urization SR 3.3.5.1.6 System Initiation 2 CHJ ), 3<<*>

Timer

d.

Core Spray Pump 1'

2 G

SR 3.3.5.1. 2

"' 101.5 psig Discharge SR 3.3.5.1.4 and Press ure-High 21"-' >, 3<

SR 3.3.5.1.6 s 148. 5 psig

e.
Low Pressure Coolant 1,
G SR 3.3.5. 1. 2
"' 101.5 psig Injection Pump SR 3.3.5.1.4 and Di sch a rge 2<"">, 3<<bJ SR 3.3.5. 1. 6 s 148.5 psig Pressure-High
f.
Automati c I,
G SR 3.3.5. J.5 s 580 second s Depress uri zation SR 3. 3.5.1.6 Syst em Low Low Wa t er 2 c ~ 1
, 3<*M Leve l Ac tuat ion Timer (eb ) With reactor steam dome press ure > 150 psig.
Dresden 2 and 3
3. 3. 5. 1-14 Amendment No.
3. 3 INSTRUMENTATION RPV Water Inven t ory Control In strumentation 3.3. 5.2 3. 3. 5. 2 Reactor Pressure Vessel (RPV) Water Inventory Contro l Instrumentation LCO 3. 3. 5. 2 The RPV Water Inventory Control instrumentation for each Function in Table 3. 3. 5. 2-1 shall be OPERABLE.
APPLICABILITY:
According to Table 3. 3. 5.2-1.
ACTIONS
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NO TE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Separate Condition entry is al lowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A. l Enter the Condition Immediately inoperable.
referenced in Table 3. 3. 5. 2-1 for the channel.
B.
As required by B. 1 Declare associated Immediately Required Action A. 1 penetration flow and referenced in path(s) incapable of Table 3.3.5. 2-1.
automatic isolation.
AND B. 2 Calculate DRAIN TIME.
Immediately
c. As required by c.1 Place channel in trip.
1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A. 1 and referenced in Table 3. 3. 5. 2-1.
D. As required by D. 1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A. l OPERABLE status.
and referenced in Table 3. 3. 5. 2-1.
(continued) I Dresden 2 and 3 3.3.5. 2-1 Amendment No.
ACTIONS CONDITION E. Required Action and associated Completion Time of Condition C or D not met.
SURVEILLANCE REQUIREMENTS E. 1 RPV Water Inventory Control Instrumentation
3. 3. 5. 2 REQUIRED ACTION Declare associated low pressure ECCS injection/spray subsystem inoperable.
COMPLETION TIME Immediately
- - --- - - - - -- - - -- -- - - - -- - - - - - -- - - - - -- - - NO TE - - -- - - - - - - - - - - -- - - - - - - - - - - -- - -- - - - --
Refer to Table 3.3. 5.2-1 to determine which SRs apply for each ECCS Function.
SURVEILLANCE SR 3.3.5.2. 1 Perform CHANNEL CHECK.
SR 3.3. 5.2.2 Perform CHANNEL FUNCTIONAL TEST.
Dresden 2 and 3
3. 3.5. 2-2 FREQUENCY In accordance with the Surveillance Frequency Control Program Jn accordance with the Surveillance Frequency Control Program Amendment No.
RPV Water Inventory Control In strumentation 3.3. 5.2 Table 3.3 5.2*1 (Page 1 of 1)
RPV Water Inventory Control Instrumentation APPLICABLE CONDITIONS 110DES REFERENCED OR OTHER REQUIRED FROM SPECIFIED CHANNELS PER RE OU IRED SURVE! L LANCE ALLOWAB, E FUNCTION CONDITIONS FUNCTION ACTION A.1 R£0UIREl1ENTS VALUE
1.
Core Spray System
a.
Reactor Steam Dome 4, 5 2
c SR 3.3.5.2.2
,

308.5 psig Pressure-Low and (Permissive) s341.7 psig
b.
Core Sp1*ay Pump 4, 5 1 per D
SR 3.3. 5.2.2
,

802 gpm Discharge Flow-Low pump (a) and (Bypass) 5 992 gpm
2.
Low Pressure Coolant Injection (LPCIJ System
a.
Reactot' Steam Dome 4, 5 2
c SR 3.3.5.2.2
,

308.5 psig Pressure-Low and (Permissive)
_ 341.7 psig
b.
Low Pressure Coolant 4, 5 1 per' D
SR 3.3.5.2.2
?:llOlgpm Injection Pump loop (a)
Discharge Flow-Low (Bypass J 3
Shutdown Cooling Syster~ I SJCJ fsolat1011
a. Reactor Vesse I Water (b) per trip B
SR 3.3.5.2.1
,

2. 65 incres Level Low system SR 3.3. 5.2.2
4.
Reactor Water Cleanup System Isolation
a.
Reactor Vesse I Water (b) pe1* trip B
SR 3.3.5. 2.1
,

2. 65 inches Level-Low system SR 3.3.5.2.2 (a)
Associated with an ECCS subsystem required to be OPERABLE by cCO 3.5.2, "RPV Wate1' lnvento*'y Control."
(b)
When automatic isolation of the associated penetration flow path(s) is credi~ed in calculating DRAIN Tl!ff Dresden 2 and 3
3. 3.5.2 -3 Amendment No.
3.3 INSTRUMENTATION IC System Instrumentation 3. 3. 5..&3
3. 3.5. -&---3 Isolation Condense r (IC) System Instr umentation LCO
3. 3. 5..&3 Four channel s of Reactor Vesse l Pressure-High instrumentation shall be OPERABLE.
APPLICABILITY :
MODE 1, MODES 2 and 3 with react or steam dome pressure > 150 psig.
ACTIONS
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more Reactor A. l Declare IC Sys t em 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Vessel Pressure-High inoperable.
discovery of channels inoperable.
loss of IC initiation capability AND A.2 Place channel(s) in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.
B.
Required Action and B.l Declare IC System Immediately assoc i ated Completion inoperable.
Time not met.
Dresden 2 and 3
3. 3. 5. 3-1 Amendment No.
IC System Instrumentation 3.3.5.iJ SURVEILLANCE REQUIREMENTS
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Reactor Vessel Pressure-High Function maintains IC initiation capability.
S R 3. 3. 5. iJ. 1 SR 3.3.5.iJ.2 SR 3.3.5.iJ.3 SR 3.3.5.iJ.4 Dresden 2 and 3 SURVEILLANCE Perform CHANNEL FUNCTIONAL TEST.
- - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - - - - -
Not required for the time delay portion of the channel.
Perform CHANNEL CALIBRATION.
The Allowable Value shall be$ 1068 psig.
Perform CHANNEL CALIBRATION for the time delay portion of the channel.
The Allowable Value shall be$ 15 seconds.
Perform LOGIC SYSTEM FUNCTIONAL TEST.
FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program 3.3.5.3-2 Amendment No.
FUNCTION 5 _ Reacto r Wat er Cleanup Sys tem Iso lation
a.
SLC Sys tem Initiati on
b.
Reactor Vesse l Water Leve l-Low
6.
Shutdown Cooling System Isolation
( h) a_
Reactor Vesse 1 Press ure - High
b.
Reactor Vesse l Water Leve l-Low Dresden 2 and 3 Pr imary Containment Isol at i on In strumentat i on
3. 3. 6. 1 Table 3.3.6.1 - 1 (page 3 of 3)
Primary Containment Isolation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3 1,2, 3 1, 2. 3 REQU I RED CHANNELS PER TRIP SYSTEM 2
2
3. 3.6. 1-8 CONDITIONS REFERENCED FROM REOUI RED ACTION C. l H
SURVE ILLANCE REQUIREMENTS SR 3. 3.6. 1. 7 SR 3.3.6. 1.1 SR 3.3. 6.1.2 SR 3. 3.6.1.3 SR 3.3.6.1.6 SR 3.3.6. 1. 7 SR 3.3.6. 1.2 SR 3.3.6.1.6 SR 3.3.6. 1. 7 SR 3. 3.6. 1.1 SR 3.3.6.1. 2 SR 3.3.6. 1. 3 SR 3.3.6. 1.6 SR 3.3. 6. 1. 7 NA ALLOWABLE VA LUE
,, 2.65 inches s 114.1 psi g
<Loop 1, Reactor Wide Range Pressure) s 110.4 psig
<loop 2, Reactor Press ure Feedwa ter Control)
,, 2.65 inches Amendment No.
FUN CTI ON
1.
Reacto r Vesse l Water Level-Low
2.
Drywell - Pressure-High
3.
React or Building Exhaust Radiation-High
4.
Refueling Floor Radiation-High Secondary Containment Isolation Instrumentation
3. 3. 6. 2 Tabl e 3.3.6.2-1 (page 1 of ll Secondary Cont ai nment Isolation Instrumentation APPLICABLE MODES OR OTHER SPEC! FI ED CONDITIONS 1,2,3,
-+;;.+
1' 2,3 1,2,3, Ca l.,+-9+
1, 2 '3' (a l.,+-9+
REQUIRED CHANNELS PER TRIP SYSTEM 2
2 2
2 SURVE ILLANCE REQUJ REMENTS SR 3.3.6.2. 1 SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6 SR 3.3.6.2.2 SR 3.3.6.2. 4 SR 3.3.6.2.6 SR 3.3.6.2. 1 SR 3.3.6.2.2 SR 3.3.6. 2.4 SR 3.3.6.2.6 SR 3.3.6. 2. l SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6. 2.6 ALLOWAB LE VA LUE
<: 2.65 inches
,; 1. 94 psi g
,; 14.9 mR/ hr
,; 100 mR/ hr
( ~a l During movement of recentl y irradiated fuel assemblies in secondary containment.
Dresden 2 and 3
3. 3.6. 2-5 Amendment No.
CREV System Instrumentation 3. 3. 7. 1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation CCREV) System Instrumentation LCO 3.3.7. 1 APPLICABILITY :
ACTIONS Two channels of the Reactor Building Ventilation System-High High Radiation Alarm Function shall be OPERABLE.
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, D1::1ri ng opera ti ens 11i th a potential for draining the reactor vessel ( OPDRVs)..
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Separate Condition entry is allowed for each channel.
CONDITION A.
One or more channels inoperable.
Dresden 2 and 3 A. 1 A.2 REQUIRED ACTION Declare CREV System inoperable.
Restore channe 1 to OPERABLE status.
3.3. 7. 1-1 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of CREV System Instrumentation alarm capability in both trip systems 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (continued)
Amendment No.
ECCS-Operating
3. 5. 1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM
3. 5. 1 ECCS-Operating LCO
3. 5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (AOS) function of five relief valves shall be OPERABLE.
APPLICABILI TY:
MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure ~ 150 psig.
ACTIONS
- - - - - - - - - - - - - - - - - - - - - -- - - -- - - -- - -- -- - NOTE - -- -- - - - - -- - - -- - - - - - - - - - - - - - - - - - - - --
LCO 3. 0.4.b is not applicable to HPCI.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One Low Pressure A. l Restore LPCI pump to 30 days Coolant Injection OPERABLE status.
(LPCI) pump inoperable.
B.
One LPCI subsystem B. 1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition A.
subsystem to OPERABLE status.
.QR One Core Spray subsystem inoperable.
c. One LPCI pump in each C. 1 Restore one LPCI 7 days subsystem i noperab l e.
pump to OPERABLE status.
(continued)
Dresden 2 and 3 3.5. 1-1 Amendment No.
EGGS Sh1:1tdo*1mRPV Water Inventory Contra 7
3. 5. 2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYST EM 3.5. 2
~RPV Water Inventory Control Si:n.it8s1m LCO 3. 5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
4-wG---One low pressure ECCS injection/spray subsystem& shall be OPERABLE.
APP LI CAB I LIT Y :
MODES 4 and 5.,
ACTIONS CONDITION REQUIRED ACT! ON COMPLETION A.
~R equired ECCS A. l Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable.
subsystem to OPERABLE status.
B.
Required Action and B.l Initiate action to Immediately associated Completion establish a method of Time of Condition A not water injection capable met.
of operating without offs i te electrical power.st.ispens spePatisns 1,1i tl:i a petentjal f:ep 8pajnjn13 tl:ie PeaeteP vessel
\\ GPQRlJ&)
c.
+1,19 PeEjt.iiPeS ff CS C.l InHjate aetjen te Immesjately i n;jeeti en.lspra;Y st.ispens GPQRVs.
&t.iesystem& inepePaele.
MW G.2 RestePe ene P9Ejt.iiPeS 4 R8YP5 EGGS i n;j est i en,lsp ra;;*
5YQ5:)1StelR te GPERMnE statt.is.
TIME (continued)
Dresden 2 and 3 3.5. 2-1 Amendment No.
ACTIONS CONDITION C.
DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and;::>: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
Dresden 2 and 3 c. 1 EGGS Shb1tdo 11mRPV Water Inventory Control
3. 5. 2 REQUIRED ACTION Verify secondary containment boundary is capable of being established in less than the DRAIN TIME.
COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. 2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment C. 3 penetration flow path is capable of being isolated in less than the DRAIN TIME.
Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME.
4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued)
3. 5. 2-2 Amendment No.
ACTIONS CONDITION D.
DRAIN TIME < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Required Action C.2 and associated Completion Time not met.
Dresden 2 and 3 0.1 EGGS ShYtd01*mRPV Water Inventory Control 3.5. 2 REQUIRED AC TI ON

NOTE --------

Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.
Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for
~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . restore secondary containment to OPERABLE status.
COMP LETI ON TI ME Immediately D.2 Initiate action to Immediately establish secondary D.3 containment boundaryinitiate action to restore one standby gas treatment subsystem to OPERABLE status..
Initiate action to restore isolation isolate each capability in each reqYired secondary containment penetration flow path
~or verify it can be manually isolated from the control room.
Immediately 3.5. 2-3 Amendment No.
CONDITION 0.4 Dresden 2 and 3 EGGS ShbltdownRPV Water Inventory Contra 7 3.5.2 REQUIRED AC TI ON Initiate action to verify one standby gas treatment system is capable of being placed in operation.
COMPLETION TIME Immediately (cont i nued) 3.5. 2-4 Amendment No.
EGGS Shbltdo*,mRPV Water Inventory Control
3. 5.2 ACTIONS E.
Required Action and E.l associated Completion Time of Condition C or D not met.
Init iate action to restore DRAIN TIME to
~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
DRAIN TIME < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
SURVEILLANCE REQUIR EMENTS SR 3.5.2.1 SR 3.5.2.+2 SR
3. 5.2.iJ SURV EILLANCE Verify DRAIN TIME~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
Verify, for for easRthe required ECCS injection/
spray subsystem, the:
a.
Suppression pool water level is
~ 10 ft 4 inches; or b *
"':'TC:-
Only one reqYired ECC5 injection/spray sblb&y&tem may take sredit for tRis option dblrin§ OPDRVs.
Contaminated condensate storage tanks water volume is~ 140,000 available gallons.
Verify, for ~the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.
Immediately FREQU ENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)
Dresden 2 and 3 3.5. 2-5 Amendment No.
EGGS Shb1tdo 111nRPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SR 3.5.2.J.4 SR
3. 5. 2. 5 SR 3.5.2.6 SR 3.5. 2.7 SURVEILLANCE
- - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -
Not required to be met for system vent flow paths opened under administrative control.
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.
Operate the required ECCS injection/spray subsystem through the recirculation line for ~ 10 minutes.
Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.
- - - - - - - - - - - - - - - - - -NO TE - - - - - - - - - - - - - - - - -
Vessel injection/spray may be excluded.
Verify the required ECCS injection/spray subsystem can be manually operated.
SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.5.2.4 Verify each reqblired EGGS j:Jblmj:J develo13s the s13ecified flow rate against a test line 13ressblre corres13onding to the s13ecified reactor 13ressblre.
TEST LINE PRESSURE NO.
CORRESPONQING OF TO A REACTOR SYSTEM FLO';,' R/\\TE
E!lM:E£ PRESSURE OF FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Inservice Testing Program In accordance with the Inservice Testing Program In accordance with the Survei I lance Frequency Control Program In accordance with the Inservice Testing Program Dresden 2 and 3
3. 5. 2-6 Amendment No.
IC System 3.5.3
3. 5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM 3.5.3 IC System LCO
3. 5.3 The IC System shall be OPERABLE.
APPLICABILITY :
MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
ACTIONS
- - - - - - - - - -- - -- -- -- - - -- - - - - - - -- - - - - -NOTE- - ---- - - - - - - - - - - - - - - - - - -- -- - -- -- - - - - - - -
LCO 3.0.4.b is not applicable to IC.
CONDITION A.
IC System inoperable.
B.
Required Action and associated Completion Time not met.
Dresden 2 and 3 REQUIRED ACTION COMPLETION TIME A. l Verify by Immediately A.2 administrative means High Pressure Coolant Injection System is OPERABLE.
Restore IC System to OP ERABLE status.
- -- - - - - -- -- -NOTE - -- - - - - - - - - - -
LCO 3.0.4.a is not applicable when entering MODE 3.
B.l Be in MODE 3.
14 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.5.3-1 Amendment No.
ACTIONS CONDITION C.
(continued)
D.
MSIV leakage rate not within limit.
E.
Required Action and associated Completion Time of Condition A, B, C, or D not met in MODE 1, 2, or 3.
Dresden 2 and 3 C. 2 D.l E. 1 E.2 REQUIRED ACTION
- - - - - - - -NOTES - - - - - - - -
1. Isolation devices in high radiation areas may be verified by use of administrative means.
2.
Isolation devices that are a locked, sealed, or otherwise secured may be verified by use of administrative means.
PC I Vs
3. 6.1.3 COMPLETION TIME Verify the affected Once per 31 days penetration flow path is isolated.
Restore leakage rate to within limit.
Be in MODE 3.
Be in MODE 4.
8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 12 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued) 3.6.1.3-5 Amendment No. 185/180
/\\CTIONS CONOITION REQUIRED
/l.CTIO~l f
Re q 1:1 i r e d Act i on a n d
-r-f-.-. +l -----:1:-I 1+n +-ittrl-H. a,,..t~e~attC:ritrli-tOl-Hn~t.Y.o associated Completion s1:1spend operations Time of Condition A, with a potential for B, C, or 0 not met for draining the reactor PCIV(s) req1:1ired to be vessel (OPDRVs).
OPERABLE d1:1rin9 MOOE 4
~
GR f.2 Initiate action to restore valve(s) to OPERABLE stat1:1s.
SURVEILLANCE REQUIREMENTS SR 3.6.1.3.1 SURVEILLANCE
- - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - -
Not required to be met when the 18 inch primary containment vent and purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simu ltaneou sly.
Verify each 18 inch primary containment vent and purge valve, except the torus purge valve, is closed.
PC I Vs 3. 6. 1.3 COMPLETION TIME Immediately Immediately FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)
Dresden 2 and 3 3.6.1.3-6 Amendment No.
3.6 CONTAINMENT SYSTEMS Secondary Containment
3. 6.4. 1 3.6.4.l Secondary Containment LCO
3. 6.4. 1 The secondary containment shall be OPE RABL E.
APPLICABILITY:
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary con t ainment, Ol:lring operations 11ith a potential for draining the reactor vessel ( OPDRVs).
ACTIONS CONDITION A.
Secondary containment inoperable in MODE 1, 2, or 3.
B.
Required Action and associated Completion Time of Condition A not met.
C.
Secondary containment i noperable during movement of recently irradiated fuel assemblies in the secondary containment or dl:lring OPDRVs.
A. l REQUIRED ACTION Restore secondary containment to OPERABLE status.
- - - - - - - - - - - -NOTE- - - - - - - - - - - - -
LCO 3.0.4. a is not applicable when entering MODE 3.
B. l C. l Be in MODE 3.
- - - - - - - - NOT E- - - - - - - - -
LCO 3.0. 3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
COMPLETION TIM E 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 12 hours Immediately C.2 Initiate action to Immediately sl:lspend OPORVs.
Dresden 2 and 3 3.6.4. 1-1 Amendment No.
3.6 CONTAINMENT SYSTEMS
3. 6.4. 2 Secondary Containment Isolation Valves (SC IVs)
LCO 3.6.4. 2 Each SCIV shall be OP ERABLE.
APPLICABILITY:
MODES 1, 2, and 3, SCI Vs 3.6.4. 2 During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel ( OPDRVs).
ACTIONS NOTES - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
1.
Penetration flow paths may be unisolated intermittently under administrative control s.
2.
Separate Condition entry is allowed for each penetration flow path.
3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.l Isolate the affected 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s penetration flow path s penetration flow path with one SCIV by use of at least inoperable.
one closed and de-activated automatic valve, closed manual valve, or blind flange.
(continued)
Dresden 2 and 3
3. 6.4.2-1 Amendment No.
ACTIONS CONDITION C.
Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.
D.
Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment or d~rin§ OPDRVs.
C. l C.2 D. l REQUIRED ACTION Be in MODE 3.
Be in MODE 4.

NOTE---------

LCO 3.0.3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
SCI Vs 3.6.4.2 COMPLETION TIME 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours Immediately D. 2 Initiate action to Immediately s~sf)end OPDRVs.
Dresden 2 and 3 3.6.4.2-3 Amendment No.
3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems sha ll be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, SGT System 3.6.4. 3 During movement of recently irradiated fuel assemblies in the secondary containment, D1:1rin§ operations *,1ith a potential for drainin§ the reactor vessel ( OPDRVs).
ACTIONS CONDITION A.
One SGT subsystem inoperable.
B.
Re qui red Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
C.
Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment A. 1 REQUIRED ACTION Restore SGT subsystem to OP ERABLE status.
- - - - - - - - - - - -NOTE - - - - - - - - - - - - -
LCO 3. 0.4.a is not applicable when entering MODE 3.
B.l Be in MODE 3.
- - - - - - - - - - - -NOTE - - - - - - - - - - - -
LCO 3.0.3 is not applicable.
C. 1 Place OPERABLE SGT subsystem in operation.
COMPLETION TIME 7 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Immediately or d1:1rin§ OPDRVs.
Q.R (continued)
Dresden 2 and 3 3.6.4.3-1 Amendment No.
ACTIONS CONDITION C.
(continued)
D.
Two SGT subsystems inoperable in MODE 1, 2, or 3.
E.
Required Action and associated Completion Time of Condition D not met.
F.
Two SGT subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment or dblriRg OPDRVs.
C. 2....+/-
C. 2.2 D. l REQUIRED ACTION Suspend movement of recently irradiated fuel assemblies in secondary containment.
IRitiate actioR to SblspeRd OPDRl/s.
Restore one SGT subsystem to OPERABLE status.
- - - - - - - - - - - - NOTE - - - - - - - - - - - - -
LCO 3. 0.4.a is not applicable when entering MODE 3.
E.1 F. 1 Be in MODE 3.
--- - ----NOTE--------
LCO 3.0.3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in secondary containment.
SGT System 3.6.4.3 COMPLETION TIME Immediately hrniedi atel,y 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 12 hours Immediately F. 2 IRitiate actioR to Immediately sblspeRd OPDRl/s.
Dresden 2 and 3
3. 6.4. 3-2 Amendment No.
3. 7 PLANT SYSTEMS
3. 7.4 Control Room Emergency Ventilation CCREV) System LCO 3.7.4 APPLICABILITY :
The CREV System shall be OPERABLE.
- - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - -
The main control room envelope (CRE) boundary may be opened intermitt ently under administrative control.
MOD ES 1, 2, and 3, CREV System 3.7.4 During movement of recently irradiated fuel assemblies in the secondary containment, Duri n9 operations 'iii th a potential for drai ni n9 the reactor
'lessel ( OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
CREV System inoperable A.1 Restore CREV System 7 days in MODE 1' 2' or 3 for to OPERABLE status.
reasons other than Condition c.
B.
Required Action and
- - - - - - - - - - - -NOTE- - - - - - - - - - - - -
associated Completion LCO 3. 0.4.a is not applicable Time of Condition A when entering MODE 3.
not met in MODE 1' 2' or 3.
B. 1 Be in MODE 3.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)
Dresden 2 and 3
3. 7.4-1 Amendment No.
ACTIONS (continued)
CONDITION C.
CR EV system inoperable C.l due to inoperable CRE boundary in MODE 1, 2, or 3.
D.
Required Action and associated Completion Time of Condition C not met in MODE 1, 2, or 3.
C.2 C.3 D. l D.2 REQUIRED ACTION Initiate action to implement mitigating actions.
Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards wi 11 not exceed limits.
Restore CRE boundary to OPERABLE status.
Be in MODE 3.
Be in MODE 4.
E.
CREV System inoperable

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

during movement of LCO 3.0.3 is not applicable.
recently irradiated fuel assemblies in the secondary containment or dYring OPDRVs.
CREV System inoperable E.1 due to an inoperable M>m CRE boundary during Suspend movement of recently irradiated fuel assemblies in the secondary containment.
movement of recently
~E~.~2~~__._,rn~1r*t~1H*a~trle~a~c~trli~o~n__.,t~o irradiated fuel sYspend OPDRVs.
assemblies in the secondary containment or dYring OPDRVs.
CREV System
3. 7.4 COMPLETION TIME Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours Immediately Immediately Dresden 2 and 3
3. 7.4-2 Amendment No.
3.7 PLANT SYSTEMS Control Room Emergency Ventilation AC System
3. 7.5
3. 7. 5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO
3. 7. 5 The Control Room Emergency Ventilation AC System shall be OPERABLE.
APPLICABILITY :
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, Ol:H'i n§ operati ens 'n'i th a potential for drai ni n§ the reactor vessel ( OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Control Room Emergency A.l Ventilation AC System i noperable in MODE 1, Restore Control Room 30 days Emergency Ventilation AC System to OPERABLE 2, or 3.
B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
C.
Control Room Emergency Ventilation AC System inoperable during movement of recently irradiated fuel assemblies in the secondary containment or d~rin§ OPDRVs.
Dresden 2 and 3 status.
- - - - - - - - - - - -NOTE- - - - - - - - - - - - -
LCO 3.0.4.a is not applicable when entering MODE 3.
B.l Be in MODE 3.
- - - - - - - - - - - -NOTE- - - - - - - - - - - - -
LCO 3.0.3 is not applicable.
C. l Suspend movement of recently irradiated fuel assemblies in the secondary con t ainment.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Immediately C.2 Initiate action to Immediately s~spend OPDRVs.
3. 7. 5-1 Amendment No.
ACTIONS CONDITION A.
(continued)
Dresden 2 and 3 A. 2. 2 A.2.3 REQUIRED ACTION Suspend movement of recently irradiated fuel assemblies in the secondary containment.
IRitiate actioR to SYSpeRd operat iORS
'n'ith a poteRtial for draiRiR§ the reactor vessel ( OPDRVs).
A. 2. 43 In i ti ate act i on to restore required offsite power circuit to OP ERABLE status.
AC Sources-Shutdown 3.8.2 COMPLETION TIME Immediately Immediately Immediately (continued) 3.8. 2-3 Amendment No.
ACTIONS CONDITION B.
One required DG inoperable.
B. 1 B. 2 B.3 REQUIRED ACTION Suspend CORE ALTERATIONS.
Suspend movement of recently irradiated fuel assemblies in secondary containment.
Initiate action to st1spend OPDRVs.
AC Sources-Shutdown 3.8.2 COMPLETION TIME Immediately Immediately Immediately B.43 Initiate action to Immediately restore required DG to OPERABLE status.
Dresden 2 and 3 3.8.2-4 Amendment No.
ACTIONS COND IT ION REQUIRED ACTION A.
(continued)
A.2.3 Initiate action to s~spend operations 11i th a potential for draining the reactor vessel.
A. 2. 43 I n i t i ate a ct i on to restore requ i red DC electrical power subsystems to OPERABLE status.
SURV EI LLANCE REQU I REMENTS SR 3.8.5.1 SURVEILLANCE
- - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - -
The following SRs are not required to be performed for the 250 voe electrical power subsystem:
SR 3.8.4.2 and SR 3.8.4.4.
For DC electrical power subsystems required to be OP ERABLE the following SRs are applicable :
SR 3.8.4.1, SR 3.8.4.2, SR 3.8.4.3, and SR 3.8.4.4.
DC Sources-Shutdown 3.8. 5 COMPL ETION TI ME Immediately Immedi ately FREQUENCY In accordance wi t h applicable SRs Dresden 2 and 3 3.8.5-2 Amendment No.
ACTIONS Distribution Systems-Shutdown 3.8.8 CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A. 2.2 A.2.3 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
Initiate action to sYspend operations
,1 i th a potent i al for draining the reactor vessel.
A. 2. 43 I n i t i a t e a ct i on s to restore required AC and DC electrical power distribution subsystems to OPERABLE status.
Immediately IFRFRediately Immediately A.2.-94 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
SURVEILLANCE REQUIREMENTS SR 3.8.8. 1 Dresden 2 and 3 SURV EILLANCE Verify correct breaker alignments and voltage to required AC and DC electrical power distribution subsystems.
3.8.8-2 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No.
Dresden Nuclear Power Station, Units 2 and 3, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 3 - REVISED TECHNICAL SPECIFICATIONS PAGES TOC Page i TOC Page ii 1.1-2 1.1-3 1.1-4 1.1-5 1.1-6 1.1-7 3.3.5.1-2 3.3.5.1-3 3.3.5.1-4 3.3.5.1-5 3.3.5.1-6 3.3.5.1-7 3.3.5.1-8 3.3.5.1-9 3.3.5.1-10 3.3.5.1-11 3.3.5.1-12 3.3.5.2-1 3.3.5.2-2 3.3.5.2-3 3.3.5.3-1 3.3.5.3-2 3.3.6.1-8 3.3.6.2-5 3.3.7.1-1 3.5.1-1 3.5.2-1 3.5.2-2 3.5.2-3 3.5.2-4 3.5.2-5 3.5.3-1 3.6.1.3-5 3.6.1.3-6 3.6.1.3-7 3.6.1.3-8 3.6.4.1-1 3.6.4.2-1 3.6.4.2-3 3.6.4.3-1 3.6.4.3-2 3.7.4-1 3.7.4-2 3.7.5-1 3.8.2-2 3.8.2-3 3.8.5-1 3.8.5-2 3.8.8-2
TABLE OF CONTENTS
1. 0 1.1
1. 2
1. 3 1.4 2.0 2.1 2.2 3.0 3.0 3.1 3.1.1 3.1. 2 3.1. 3 3.1.4 3.1. 5 3.1. 6 3.1. 7 3.1.8 3.2 3.2.1 3.2.2 3.2.3 3.3 3.3.1.l 3.3.1.2 3.3.1.3
3. 3. 2.1 3.3.2.2 3.3.3.1 3.3.4.1
3. 3. 5. 1 3.3.5.2 3.3.5.3 3.3.6.1 3.3.6.2 3.3.6.3
3. 3. 7.1 3.3.7.2 USE AND APPLICATION Definitions.............................................. 1.1-1 Logical Connectors....................................... 1.2-1 Completion Times......................................... 1. 3-1 Frequency................................................ 1.4-1 SAFETY LIMITS CSLs)
SLs...................................................... 2.0-1 SL Violations............................................ 2.0-1 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY........ 3.0-1 SURVEILLANCE REQUIREMENT CSR) APPLICABILITY................. 3.0-4 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN CSDM).................................... 3.1.1-1 Reactivity Anomalies..................................... 3.1.2-1 Control Rod OPERABILITY.................................. 3.1.3-1 Control Rod Scram Times.................................. 3.1.4-1 Control Rod Scram Accumulators........................... 3.1.5-1 Rod Pattern Control...................................... 3.1.6-1 Standby Liquid Control (SLC) System...................... 3.1.7-1 Scram Discharge Volume (SDV) Vent and Drain Valves....... 3.1.8-1 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE CAPLHGR)...... 3.2.1-1 MINIMUM CRITICAL POWER RATIO (MCPR)...................... 3.2.2-1 LINEAR HEAT GENERATION RATE (LHGR)...................... 3.2.3-1 INSTRUMENTATION Reactor Protection System (RPS) Instrumentation.......... 3.3.1.1-1 Source Range Monitor (SRM) Instrumentation............... 3.3.1.2-1 Dscillation Power Range Monitor (OPRM) Instrumentation... 3.3.1.3-1 Control Rod Block Instrumentation........................ 3.3.2.1-1 Feedwater System and Main Turbine High Water Level Trip Instrumentation................................... 3.3.2.2-1 Post Accident Monitoring (PAM) Instrumentation........... 3.3.3.1-1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation................... 3.3.4.1-1 Emergency Core Cooling System CECCS) Instrumentation..... 3.3.5.1-1 Reactor Pressure Vessel CRPV) Water Inventory Control I n st rumen tat i on........................................ 3. 3. 5. 2 -1 Isolation Condenser (IC) System Instrumentation.......... 3.3.5.3-1 Primary Containment Isolation Instrumentation............ 3.3.6.1-1 Secondary Containment Isolation Instrumentation.......... 3.3.6.2-1 Relief Valve Instrumentation............................. 3.3.6.3-1 Control Room Emergency Ventilation (CREV) System Instrumentation........................................ 3.3.7.1-1 Mechanical Vacuum Pump Trip Instrumentation.............. 3.3.7.2-1 (continued)
Dresden 2 and 3 Amendment No.
TABLE OF CONTENTS 3.3 3.3.8.1 3.3.8.2 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.5 3.5.1 3.5.2 3.5.3 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3.6.1.6 3.6.1.7 3.6.1.8 3.6.2.1 3.6.2.2 3.6.2.3 3.6.2.4 3.6.2.5 3.6.3.1
3. 6.4.1 3.6.4.2 3.6.4.3 INSTRUMENTATION (continued)
Loss of Power (LOP) Instrumentation...................... 3.3.8.1-1 Reactor Protection System (RPS) Electric Power Monitoring............................................. 3.3.8.2-1 REACTOR COOLANT SYSTEM (RCS)
Recirculation Loops Operating............................ 3.4.1-1 Jet Pumps................................................ 3.4.2-1 Safety and Relief Valves................................ 3.4.3-1 RCS Operational LEAKAGE.................................. 3.4.4-1 RCS Leakage Detection Instrumentation.................... 3.4.5-1 RCS Specific Activity.................................... 3.4.6-1 Shutdown Cooling (SOC) System-Hot Shutdown............... 3.4.7-1 Shutdown Cooling (SOC) System-Cold Shutdown.............. 3.4.8-1 RCS Pressure and Temperature CP/T) Limits................ 3.4.9-1 Reactor Steam Dome Pressure.............................. 3.4.10-1 EMERGENCY CORE COOLING SYSTEMS CECCS), REACTOR PRESSURE VESSEL CRPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM ECCS-Operati ng........................................... 3. 5.1-1 RPV Water Inventory Control.............................. 3.5.2-1 IC System................................................ 3.5.3-1 CONTAINMENT SYSTEMS Primary Containment...................................... 3.6.1.1-1 Primary Containment Air Lock............................. 3.6.1.2-1 Primary Containment Isolation Valves (PCIVs)............. 3.6.1.3-1 Drywell Pressure......................................... 3.6.1.4-1 Drywell Air Temperature.................................. 3.6.1.5-1 Low Set Relief Valves.................................... 3.6.1.6-1 Reactor Building-to-Suppression Chamber Vacuum Breakers............................................... 3.6.1. 7-1 Suppression Chamber-to-Drywell Vacuum Breakers........... 3.6.1.8-1 Suppression Pool Average Temperature..................... 3.6.2.1-1 Suppression Pool Water Level............................. 3.6.2.2-1 Suppression Pool Cooling................................. 3.6.2.3-1 Suppression Pool Spray................................... 3.6.2.4-1 Drywell-to-Suppression Chamber Differential Pressure..... 3.6.2.5-1 Primary Containment Oxygen Concentration................. 3.6.3.1-1 Secondary Containment.................................... 3.6.4.1-1 Secondary Containment Isolation Valves (SCIVs)........... 3.6.4.2-1 Standby Gas Treatment (SGT) System....................... 3.6.4.3-1 (continued)
Dresden 2 and 3 i i Amendment No.
Definitions 1.1 1.1 Definitions (continued)
CHANNEL FUNCTIONAL TEST CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT I-131 Dresden 2 and 3 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY.
The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.
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.
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle.
These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.
Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The dose conversion factors used for this calculation shall be the inhalation committed dose conversion factors in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.
(continued) 1.1-2 Amendment No.
1.1 Definitions DRAIN TIME Dresden 2 and 3 Definitions 1.1 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 is 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.
(continued) 1.1-3 Amendment No.
1.1 Definitions DRAIN TIME (continued)
LEAKAGE Dresden 2 and 3 Definitions 1.1 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.
LEAKAGE shall be:
a.
Identified LEAKAGE
1. LEAKAGE into the drywe 11, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or
2. LEAKAGE into the drywel l atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;
b.
Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;
c.
Total LEAKAGE Sum of the identified and unidentified LEAKAGE; and
d.
Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.
(continued) 1.1-4 Amendment No.
Definitions 1.1 1.1 Definitions (continued)
LINEAR HEAT GENERATION RATE (LHGR)
LOGIC SYSTEM FUNCTIONAL TEST MINIMUM CRITICAL POWER RATIO (MCPR)
MODE OPERABLE-OPERABILITY Dresden 2 and 3 The LHGR shall be the heat generation rate per unit length of fuel rod.
It is the integral of the heat flux over the heat transfer area associated with the unit length.
A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.
The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.
The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel.
The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
(continued) 1.1-5 Amendment No.
Definitions 1.1 1.1 Definitions (continued)
RATED THERMAL POWER (RTP)
REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SOM)
THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TIME Dresden 2 and 3 RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2957 MWt.
The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact until the opening of the trip actuator.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
a.
The reactor is xenon free;
b.
The moderator temperature is~ 68°F, corresponding to the most reactive state; and
c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
1.1-6 Amendment No.
MODE TITLE 1
Power Operation 2
Startup 3
Hot Shutdown<al 4
Cold Shutdown<al 5
Refuel i ng<bl
~able 1.1-1 (page 1 of 1)
MODES REACTOR MODE SWITCH POSITION Run Refue1<al or Startup/Hot Standby Shutdown Shutdown Shutdown or Refuel Definitions 1.1 AVERAGE REACTOR COOLANT TEMPERATURE
( o F)
NA NA
> 212
212 NA (a)
All reactor vessel head closure bolts fully tensioned.
(b)
One or more reactor vessel head closure bolts less than fully tensioned.
Dresden 2 and 3 1.1-7 Amendment No.
ACTIONS CONDITION B.
As required by Required Action A.l and referenced in Table 3.3.5.1-1.
Dresden 2 and 3 B.1 ECCS Instrumentation
3. 3. 5.1 REQUIRED ACTION

NOTE--------

Only applicable for Functions 1.a, 1.b, 2.a, 2.b, 2.d, and
2. j.
Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.
COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature(s) in both divisions B.2

NOTE---------

B.3 Only applicable for Functions 3.a and 3.b.
Declare High Pressure Coolant Injection (HPCI) System inoperable.
Place channel in trip.
1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued) 3.3.5.1-2 Amendment No.
ACTIONS CONDITION C.
As required by Required Action A.1 and referenced in Table 3.3.5.1-1.
D.
As required by Required Action A.1 and referenced in Table 3.3.5.1-1.
Dresden 2 and 3 C.1 C.2 D.1 D. 2.1 D.2.2 ECCS Instrumentation 3.3.5.1 REQUIRED ACTION

NOTE--------

Only applicable for Functions 1.c, l.e, 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k.
Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.
Restore channel to OPERABLE status.

NOTE---------

Only applicable if HPCI pump suction is not aligned to the suppression pool.
Declare HPCI System inoperable.
Pl ace channel in trip.
Align the HPCI pump suction to the suppression pool.
COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for feature( s) in both divisions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 hour from discovery of loss of HPCI initiation capability 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours (continued) 3.3.5.1-3 Amendment No.
ACTIONS CONDITION E.
As required by Required Action A.l and referenced in Table 3.3.5.1-1.
F.
As required by Required Action A.1 and referenced in Table 3.3.5.1-1.
Dresden 2 and 3 E.1 E.2 F.1 F.2 ECCS Instrumentation
3. 3. 5.1 REQUIRED ACTION

NOTE--------

Only applicable for Functions 1.d and 2.f.
Declare supported feature(s) inoperable when its redundant feature ECCS initiation capability is inoperable.
Restore channel to OPERABLE status.
Declare Automatic Depressurization System CADS) valves inoperable.
Pl ace channel in trip.
COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of initiation capability for subsystems in both divisions 7 days 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of ADS initiation capability in both trip systems 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from discovery of inoperable channel concurrent with HPCI or isolation condenser CIC) inoperable 8 days (continued) 3.3.5.1-4 Amendment No.
ACTIONS CONDITION G.
As required by Required Action A.l and referenced in Table 3.3.5.1-1.
H.
Required Action and associated Completion Time of Condition B, C, D, E, F, or G not met.
Dresden 2 and 3 G.l G.2 H.l REQUIRED ACTION Declare ADS valves inoperable.
Restore channel to OPERABLE status.
Declare associated supported feature(s) inoperable.
ECCS Instrumentation
3. 3. 5.1 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of ADS initiation capability in both trip systems 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from discovery of inoperable channel concurrent with HPCI or IC inoperable 8 days Immediately 3.3.5.1-5 Amendment No.
SURVEILLANCE REQUIREMENTS ECCS Instrumentation
3. 3. 5.1
- - - - - - - - - -- - - -- -- - - - - - - - - -- - - - - - - - -- NOTES- -- - - -- -- - - - - - -- - - - - - - - -- - - - - - - - - -- -
1.
Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
2.
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c, 3.f, and 3.g; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c, 3.f, and 3.g provided the associated Function or the redundant Function maintains ECCS initiation capability.
SURVEILLANCE SR 3.3.5.1.1 Perform CHANNEL CHECK.
SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST.
SR 3.3.5.1.3 Calibrate the trip unit.
SR 3.3.5.1.4 Perform CHANNEL CALIBRATION.
Dresden 2 and 3 3.3.5.1-6 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Contra l Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Contra l Program (continued)
Amendment No.
SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.5.1.5 Perform CHANNEL CALIBRATION.
SR 3.3.5.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST.
Dresden 2 and 3 3.3.5.1-7 ECCS Instrumentation 3.3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No.
ECCS Instrumentation
3. 3. 5.1 Table 3.3.5.1-1 (page 1 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE L
Core Spray System a_
Reactor Vessel Water 1,2,3 4Ca>
B SR 3.3.5.Ll
,, -54.15 Leve 1-Low Low SR 3.3.5.L2 inches SR 3.3.5.L3 SR 3.3.5.1.5 SR 3.3.5.1.6
b.
Drywell Pressure-High 1,2,3 4<*>
B SR 3.3.5.1.2
$ L81 psig SR 3.3.5.1.4 SR 3.3.5.1.6
c.
Reactor Steam Dome 1,2,3 2
c SR 3.3.5.L2
,, 308.5 psig Pressure-Low SR 3.3.5.1.4 and (Permissive)
SR 3.3.5.1.6
$ 341.7 psig
d.
Core Spray Pump 1,2,3 1 per pump E
SR 3.3.5.1.2
,, 802 gpm Discharge Flow-Low SR 3.3.5.1.3 and (Bypass)
SR 3.3.5.1.5
$ 992 gpm SR 3.3.5.L6
e.
Core Spray Pump 1,2,3 1 per pump c
SR 3.3.5.1.5
$ lL O seconds Start-Time Delay Relay SR 3.3.5.1.6
2.
Low Pressure Coolant Injection (LPCI) System
a.
Reactor Vessel Water 1,2,3 4
B SR 3.3.5.Ll
,, -54.15 Level-Low Low SR 3.3.5.1.2 Inches SR 3.3.5.L3 SR 3.3.5.1.5 SR 3.3.5.1.6
b.
Drywel l Pressure-High 1,2,3 4
B SR 3.3.5.1.2
$ 1.81 psig SR 3.3.5.1.4 SR 3.3.5.1.6
c.
Reactor Steam Dome 1,2,3 2
c SR 3.3.5.1.2
,, 308.5 psig Pressure-Low SR 3.3.5.1.4 and
( Permissive)
SR 3.3.5.1.6
$ 341.7 psig (continued)
(a)
Also required to initiate the associated diesel generator COG).
Dresden 2 and 3 3.3.5.1-8 Amendment No.
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPEC I FI ED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
2.
LPCI System (continued)
d.
Reactor Steam Dome 1,2,3 4
B SR 3.3.5.1-2
~ 802 psig and Pressure-Low (Break SR 3_3_5_ 1-5 5 898 psig Detection)
SR 3.3.5.1-6
e.
Low Pressure Coolant 1,2,3 1 per pump c
SR 3.3.5.1-5 5 5.5 seconds Injection Pump SR 3.3.5.1.6 Start-Time Delay Relay Pumps B and D f_
Low Pressure 1,2,3 1 per loop E
SR 3.3.5.1-2
?; 1107 gpm Coolant Injection Pump SR 3.3.5.1.3 Discharge Flow-Low SR 3.3.5.1.5 (Bypass)
SR 3.3.5.1.6 g_
Recirculation Pump 1,2,3 4 per pump c
SR 3.3.5.1-2
,;; 5.9 psid Differential SR 3.3.5.1-5 Pressure-High (Break SR 3.3.5.1-6 Detection l
h.
Recirculation Riser 1,2,3 4
c SR 3.3.5.1-2
,;; 2.0 psid Differential SR 3.3.5.1-5 Pressure-High (Break SR 3.3.5.1.6 Detection)
i. Recirculation Pump 1,2,3 c
SR 3.3.5.1.5 5 0.53 seconds Differential Pressure SR 3.3.5.1-6 Time Delay-Relay (Break Detection) j -
Reactor Steam Dome 1,2,3 2
B SR 3.3.5.1.5
,;; 2.12 seconds Pressure Time Delay-SR 3.3.5.1.6 Relay (Break Detection)
k.
Recirculation Riser 1,2,3 2
c SR 3.3.5.1.5
,;; 0.53 seconds Differential Pressure SR 3.3.5.1.6 Time Delay-Relay (Break Detection)
(continued)
Dresden 2 and 3 3.3.5.1-9 Amendment No.
ECCS Instrumentation 3.3.5.l Table 3.3.5.1-1 (page 3 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A. l REQUIREMENTS VALUE
3.
High Pressure Coolant Injection (HPCI) System
a.
Reactor Vesse 1 Water 1,
4 B
SR 3.3.5.1.l
,

-54.15 Leve 1-Low Low SR 3.3.5.1.2 Inches 2tbl, 3tb)
SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
b.
Drywel l Pressure-High 1,
4 B
SR 3.3.5.1.2
,;; 1.81 psig SR 3.3.5.1.4 2tbl, 3tb)
SR 3.3.5.1.6
c.
Reactor Vessel Water 1,
2 c
SR 3.3.5.1.1
,;; 46.2 inches Level-High SR 3.3.5.1.2 2(b), 3tb)
SR 3.3.5.1.3 I
SR 3.3.5.1.5 SR 3.3.5.1.6
d.
Contaminated 1,
2 per CCST D
SR 3.3.5.1.2
,

11.1158 ft Condensate Storage SR 3.3.5.1.5 for CCST 2/3 A Tank (CCSTl Level-Low 2(b)' 3tb)
SR 3.3.5.1.6 and ;o, 7.5637 ft for CCST 2/3 B
e.
Suppression Pool llater 1,
2 D
SR 3.3.5.1.2
,;; 15 ft Level-High SR 3.3.5.1.5 5.625 inches 2tb), 3tb)
SR 3.3.5.1.6
f.
High Pressure Coolant 1,
E SR 3.3.5.1.2
o, 616 gpm Injection Pump SR 3.3.5.1.4 Discharge Flow-Low 2tb), 3(b)
SR 3.3.5.1.6 (Bypass)
g.
Manual Initiation 1,
c SR 3.3.5.1.6 NA 2(b)' 3t*.l (continued)
(bl With reactor steam dome pressure > 150 psig.
Dresden 2 and 3 3.3.5.1-10 Amendment No.
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES DR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SU RV EI LLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
4.
Automatic Oepressurization System (ADS) Trip System A
a.
Reactor Vessel Water 1,
2 F
SR 3.3.5.1.1
, -54.15 Leve 1-Low Low SR 3.3.5.1.2 Inches 2(b), 3(b)
SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.l.6
b.
Drywel 1 Pressure-High 1,
2 F
SR 3.3.5.1.2
$ 1.81 psig SR 3.3.5.1.4 2<bJ, 3(b)
SR 3.3.5.l.6
c.
Automatic 1,
G SR 3.3.5.l.5
$ 113 seconds Depressurization SR 3.3.5.l.6 System Initiation 2(b), 3(b)
Timer
d.
Core Spray Pump 1,
2 G
SR 3.3.5.l.2
, 101. 5 psig Discharge SR 3.3.5.l.4 and Pressure-High 2(b), 3U>)
SR 3.3.5.l.6
$ 148.5 psig e_
Low Pressure Coolant 1,
4 G
SR 3_3.5.1.2
, 101. 5 psi g Injection Pump SR 3.3.5.1.4 and Di sch a rge 2(b)' 3(b)
SR 3.3.5.1.6
$ 148.5 psig Pressure-High
f.
Automatic 1,
G SR 3.3.5.l.5
,,; 580 seconds Depressurization SR 3.3.5.1.6 System Low Low Water 2(b)' 3(b)
Level Actuation Timer (continued)
(bl With reactor steam dome pressure > 150 psig.
Dresden 2 and 3 3.3.5.1-11 Amendment No.
ECCS Instrumentation
3. 3. 5.1 Table 3.3.5.1-1 (page 5 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE 5_
ADS Trip System B a -
Reactor Vessel Water 1,
2 F
SR 3.3.5.Ll
" -54.15 Leve 1-Low Low SR 3.3.5.1.2 inches 2, 3<bl SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
b.
Drywell Pressure-High 1,
2 F
SR 3.3.5.1.2
,;; 1.81 psig SR 3.3.5.1.4 2<bl' 3<bl SR 3.3.5.l.6
c.
Automatic 1,
G SR 3.3.5.1.5
,;; 113 seconds Depressurization SR 3.3.5.1.6 System Initiation 2<bl' 3<bl Timer
d.
Core Spray Pump 1,
2 G
SR 3.3.5.l.2
" 101.5 psig Discharge SR 3.3.5.1.4 and Pressure-High 2<b) ~ 3<bl SR 3.3.5.1.6
,;; 148.5 psig e_
Low Pressure Coolant 1,
4 G
SR 3.3.5.1.2
" 101.5 psig Injection Pump SR 3.3.5.1.4 and Discharge 2<bl, 3<bl SR 3.3.5.1.6
,;; 148. 5 psi g Pressure-High
f.
Automatic 1'
G SR 3.3.5.1.5
,;; 580 seconds Depressurization SR
3. 3. 5.1. 6 System Low Low Water 2(b), 3<bl Level Actuation Timer (bl With reactor steam dome pressure > 150 psig.
Dresden 2 and 3 3.3.5.1-12 Amendment No.
3.3 INSTRUMENTATION RPV Water Inventory Control Instrumentation 3.3.5.2 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.2 The RPV Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.
APPLICABILITY:
According to Table 3.3.5.2-1.
ACTIONS
- - - - - - - - -- - -- - --- - - - -- -- - - - - - - - -- -- - -NOTE ---- - - -- - -- - - - - -- - -- - -- - -- - - - - - - - - - - -
Separate Condition entry is allowed for each channel.
, CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A. l Enter the Condition Immediately inoperable.
referenced in Table 3.3.5.2-1 for the channel.
B.
As required by B.l Declare associated Immediately Required Action A. l penetration flow and referenced in path(s) incapable of Table 3.3.5.2-1.
automatic isolation.
AND B.2 Calculate DRAIN TIME.
Immediately
c. As required by C.l Place channel in trip.
1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action A. l and referenced in Table 3.3.5.2-1.
D. As required by D.l Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A. l OPERABLE status.
and referenced in Table 3.3.5.2-1.
(continued) I Dresden 2 and 3 3.3.5.2-1 Amendment No.
ACTIONS CONDITION E. Required Action and associated Completion Time of Condition C or D not met.
SURVEILLANCE REQUIREMENTS E.1 RPV Water Inventory Control Instrumentation 3.3.5.2 REQUIRED ACTION Declare associated low pressure ECCS injection/spray subsystem inoperable.
COMPLETION TIME Immediately
- - -- - - - - - -- - - - - -- - - - - - - - - - - - - -- - - - - -- NOTE - -- - - - -- - - - - - - - - - - - -- - - -- - - - - - - - - - - -
Refer to Table 3.3.5.2-1 to determine which SRs apply for each ECCS Function.
SURVEILLANCE SR 3.3.5.2.1 Perform CHANNEL CHECK.
SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST.
Dresden 2 and 3 3.3.5.2-2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No.
RPV Water Inventory Control Instrumentation 3.3.5.2 Table 3.3.5.2-1 (Page 1 of 1)
RPV Water Inventory Control Instrumentation APPLICABLE CONDITIONS MODES REFERENCED OR OTHER REQUIRED FROM SPEC I Fl ED CHANNELS PER REQUIRED SU RV El LLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE L
Core Spray System
a.
Reactor Steam Dome 4, 5 2
c SR 3_3_5_2_2
,, 308.5 psig Pressure-Low and (Permissive)
,;; 341.7 psig
b.
Core Spray Pump 4, 5 1 per D
SR 3_3_5_2_2
,, 802 gpm Discharge Flow-Low pump (a) and
<Bypass)
,;; 992 gpm
2.
Low Pressure Coolant Injection (LPCll System a_
Reactor Steam Dome 4 > 5 c
SR 3_3_5_2_2
,, 308.5 psig Pressure-Low and (Permissive)
,;; 341-7 psig
b.
Low Pressure Coolant 4, 5 1 per D
SR 3_3_5_2_2
,, 1107 gpm Injection Pump loop (a)
Discharge Fl ow-Low
<Bypass) 3_
Shutdown Cooling System (SDCl Isolation
a. Reactor Vessel Water (b) per trip B
SR 3_3_5_2_1
,, 2.65 inches Level-Low system SR 3_3_5_2_2 4_
Reactor Water Cleanup System Isolation a -
Reactor Vessel Water (b) per trip B
SR 3_3_5_2_1
,, 2.65 inches Level-Low system SR 3_3_5_2_2 (a)
Associated with an ECCS subsystem required to be OPERABLE by LCD 3.5.2, "RPV Water Inventory Control."
(bl When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.
Dresden 2 and 3 3.3.5.2-3 Amendment No.
3.3 INSTRUMENTATION IC System Instrumentation 3.3.5.3 3.3.5.3 Isolation Condenser (IC) System Instrumentation LCO 3.3.5.3 Four channels of Reactor Vessel Pressure-High instrumentation shall be OPERABLE.
APPLICABILITY:
MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.
ACTIONS
- - - - - - - - - - - - - -- -- - - - -- - - -- - - - - - - - - - -- NOTE - --- - -- - - - - - - -- - - - -- - - -- - - - - - - -- - -- - -
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more Reactor A. l Declare IC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Vessel Pressure-High inoperable.
discovery of channels inoperable.
loss of IC initiation capability AND A.2 Place channel (s) in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.
B.
Required Action and B.1 Declare IC System Immediately associated Completion inoperable.
Time not met.
Dresden 2 and 3 3.3.5.3-1 Amendment No.
IC System Instrumentation 3.3.5.3 SURVEILLANCE REQUIREMENTS
- - - -- - - -- -- -- - -- - - - - - - -- - - -- - - - -- -- --NOTE- - -- -- - - - - - - - - -- - - - - - - - - - - - - - - -- - - - - -
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Reactor Vessel Pressure-High Function maintains IC initiation capability.
SR 3.3.5.3.1 SR 3.3.5.3.2 SR 3.3.5.3.3 SR 3.3.5.3.4 Dresden 2 and 3 SURVEILLANCE Perform CHANNEL FUNCTIONAL TEST.
- -- - - - - - - - -- - - ---NOTE- - - --- -- - - - - - - - - - - -
Not required for the time delay portion of the channel.
Perform CHANNEL CALIBRATION.
The Allowable Value shall be~ 1068 psig.
Perform CHANNEL CALIBRATION for the time delay portion of the channel.
The Allowable Value shall be~ 15 seconds.
Perform LOGIC SYSTEM FUNCTIONAL TEST.
FREQUENCY In accordance with the Surveillance Frequency Contra l Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program 3.3.5.3-2 Amendment No.
I FUNCTION
5.
Reactor Water Cleanup System Isolation
a.
SLC System Initiation
b.
Reactor Vessel Water Level-Low
6.
Shutdown Cooling System Isolation
a.
Reactor Vessel Pressure - High
b.
Reactor Vessel Water Level-Low Dresden 2 and 3 Primary Containment Isolation Instrumentation 3.3.6.l Table 3.3.6.1-1 (page 3 of 3)
Primary Containment Isolation Instrumentation APPLICABLE MODES OR OTHER SPECIF! ED CONDITIONS 1,2,3 1,2,3 1,2,3 3
REQUIRED CHANNELS PER TRIP SYSTEM 3.3.6.1-8 CONDITIONS REFERENCED FROM REQUIRED SURVEILLANCE ACTION C.l REQUIREMENTS H
F F
SR 3.3.6.1.7 SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.2 SR 3.3_6_1.6 SR 3.3.6.1.7 SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.l.6 SR 3.3.6.1.7 NA ALLOWABLE VALUE
,,

2.65 inches
,;; 114.1 psig (Loop 1, Reactor Wide Range Pressure)
,;; 110.4 psig (Loop 2, Reactor Pressure Feedwater Control)
,,

2.65 inches Amendment No.
FUNCTION
1.
Reactor Vessel Water Level-Low
2.
Drywell-Pressure-High
3.
Reactor Building Exhaust Radiation-High
4.
Refueling Floor Radiation-High Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)
Secondary Containment Isolation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3 1,2,3 1,2,3, (a) 1,2,3, (a)
REQUIRED CHANNELS PER TRIP SYSTEM 2
2 SURVEILLANCE REQUIREMENTS SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 SR 3.3.6.2.1 SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 ALLOWABLE VALUE
,, 2.65 inches
,; 1.94 psig
,; 14.9 mR/hr
,; 100 mR/hr (a) During movement of recently irradiated fuel assemblies in secondary containment.
Dresden 2 and 3 3.3.6.2-5 Amendment No.
CREV System Instrumentation
3. 3. 7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation LCO 3.3.7.1 APPLICABILITY:
ACTIONS Two channels of the Reactor Building Ventilation System-High High Radiation Alarm Function shall be OPERABLE.
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.
- - -- - - - - - -- - - - -- - -- -- - - - - - - -- - - - -- - --NOTE -- -- -- -- - - - - - - - - -- -- - - - - - - - - - -- - - - - - -
Separate Condition entry is allowed for each channel.
CONDITION A.
One or more channels inoperable.
Dresden 2 and 3 A.1 A.2 REQUIRED ACTION Declare CREV System inoperable.
Restore channel to OPERABLE status.
3.3.7.1-1 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of CREV System Instrumentation alarm capability in both trip systems 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (continued)
Amendment No.
I
ECCS-Operating 3.5.l 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM 3.5.1 ECCS-Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves sha 11 be OPERABLE.
APPLICABILITY:
MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure~ 150 psig.
ACTIONS
- - --- - - -- -- - - - - -- - - - - - - - - - --- -- - -- -- - NOTE -- - - - - -- - - - - - - - - - - - - - - - - - - - - -- - - - - - -
LCO 3.0.4.b is not applicable to HPCI.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One Low Pressure A.1 Restore LPCI pump to 30 days Coolant Injection OPERABLE status.
(LPCI) pump inoperable.
B.
One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition A.
subsystem to OPERABLE status.
OR One Core Spray subsystem inoperable.
c.
One LPCI pump in each C.1 Restore one LPCI 7 days subsystem inoperable.
pump to OPERABLE status.
(continued)
Dresden 2 and 3 3.5.1-1 Amendment No.
RPV Water Inventory Control 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM 3.5.2 RPV Water Inventory Control LCO 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
One low pressure ECCS injection/spray subsystem shall be OPERABLE.
APPLICABILITY:
MODES 4 and 5.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Required ECCS A. l Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection/spray injection/spray subsystem inoperable.
subsystem to OPERABLE status.
B.
Required Action and B.l Initiate action to Immediately associated Completion establish a method of Time of Condition A water injection not met.
capable of operating without offs ite electrical power.
(continued)
Dresden 2 and 3 3.5.2-1 Amendment No.
ACTIONS CONDITION C.
DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and ;::: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
Dresden 2 and 3 C.1 RPV Water Inventory Control 3.5.2 REQUIRED ACTION Verify secondary containment boundary is capable of being established in less than the DRAIN TIME.
COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C.2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment C.3 penetration flow path is capable of being isolated in less than t he D RA I N TI M E
Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME.
4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued) 3.5.2-2 Amendment No.
ACTIONS CONDITION D.
DRAIN TIME< 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
D.1 AND D.2 AND D.3 AND D.4 Dresden 2 and 3 RPV Water Inventory Control 3.5.2 REQUIRED ACTION NOTE --------
Required ECCS injection/spray subsystem or additional method of water injection shall be capable of operating without offsite electrical power.
Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level > TAF for
~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
Initiate action to establish secondary containment boundary.
Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room.
Initiate action to verify one standby gas treatment system is capable of being placed in operation.
COMPLETION TIME Immediately Immediately Immediately Immediately (continued) 3.5.2-3 Amendment No.
RPV Water Inventory Control 3.5.2 ACTIONS E.
Required Action and E.1 Initiate action to Immediately associated Completion restore DRAIN TIME to Time of Condition C or 2 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
D not met.
DRAIN TIME < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SURVEILLANCE Verify DRAIN TIME 2 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
Verify, for the required ECCS injection/
spray subsystem, the:
a.
Suppression pool water level is 2 10 ft 4 inches; or
b.
Contaminated condensate storage tanks water volume is 2 140,000 available gallons.
Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.
FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)
Dresden 2 and 3 3.5.2-4 Amendment No.
RPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS SR 3.5.2.4 SR 3.5.2.5 SR 3.5.2.6 SR 3.5.2.7 Dresden 2 and 3 SURVEILLANCE
- - - - - - - - - - - - - - - - - - -NOTE- - - - - - - - - - - - - - - - - - - -
Not required to be met for system vent flow paths opened under administrative control.
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.
Operate the required ECCS injection/spray subsystem through the recirculation line for ~ 10 minutes.
Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.

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

Vessel injection/spray may be excluded.
Verify the required ECCS injection/spray subsystem can be manually operated.
FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Inservice Testing Program In accordance with the Inservice Testing Program In accordance with the Surveillance Frequency Control Program 3.5.2-5 Amendment *No.
IC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS CECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER CIC) SYSTEM 3.5.3 IC System LCO 3.5.3 The IC System shall be OPERABLE.
APPLICABILITY:
MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
ACTIONS
-- -- - - - - - - - - - - - - - - - - - -- - -- --- -- - -- -NOTE- - -- -- - -- - -- - - --- - - -- - -- - -- - -- - - - - - -- --
LCO 3.0.4.b is not applicable to IC.
CONDITION A.
IC System inoperable.
B.
Required Action and associated Completion Time not met.
Dresden 2 and 3 REQUIRED ACTION COMPLETION TIME A.1 Verify by Immediately A.2 administrative means High Pressure Coolant Injection System is OPERABLE.
Restore IC System to OPERABLE status.
- - - -- - - - - - - - NOTE - - - - -- - - -- - --
LCO 3.0.4.a is not applicable when entering MODE 3.
B.1 Be in MODE 3.
14 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.5.3-1 Amendment No.
ACTIONS CONDITION C.
(continued)
D.
MSIV leakage rate not within limit.
E.
Required Action and associated Completion Time of Condition A, B, C, or D not met in MODE 1, 2, or 3.
Dresden 2 and 3 C.2 D.1 E.1 E.2 REQUIRED ACTION

NOTES--------

1. Isolation devices in high radiation areas may be verified by use of administrative means.
2.
Isolation devices that are a locked, sealed, or otherwise secured may be verified by use of administrative means.
PC I Vs 3.6.1.3 COMPLETION TIME Verify the affected Once per 31 days penetration flow path is isolated.
Restore leakage rate to within limit.
Be in MODE 3.
Be in MODE 4.
8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 12 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.6.1.3-5 Amendment No.
SURVEILLANCE REQUIREMENTS SR 3.6.1.3.1 SR 3.6.1.3.2 SURVEILLANCE
- - - - - - - - - - - - -- -- --NOTE- - -- - - - - - - - - -- - - - --
Not required to be met when the 18 inch primary containment vent and purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided the drywell vent and purge valves and their associated suppression chamber vent and purge valves are not open simultaneously.
Verify each 18 inch primary containment vent and purge valve, except the torus purge valve, is closed.

NOTES------------------

1.
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
2.
Not required to be met for PCIVs that are open under administrative controls.
Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.
PC I Vs 3.6.1.3 FREQUENCY In accordance with the Surveillance Frequency Contra l Program In accordance with the Surveillance Frequency Control Program (continued)
Dresden 2 and 3 3.6.1.3-6 Amendment No.
SURVEILLANCE REQUIREMENTS SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 SR 3.6.1.3.6 SURVEILLANCE
- - - - - - - - - - - - - - - - - -NOTES- - - - - - - - - - - - - - - - - -
1.
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
2.,Not required to be met for PCIVs that are open under administrative controls.
Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked sealed, or otherwise secured and is required to be closed during accident conditions is closed.
Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge.
Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.
Verify the isolation time of each MSIV is 2 3 seconds and~ 5 seconds.
PC I Vs 3.6.1.3 FREQUENCY Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days In accordance with the Surveillance Frequency Control Program In accordance with the Inservice Testing Program In accordance with the Inservice Testing Program (continued)
Dresden 2 and 3 3.6.1.3-7 Amendment No.
SURVEILLANCE REQUIREMENTS SR 3.6.1.3.7 SR 3.6.1.3.8 SR 3.6.1.3.9 SR 3.6.1.3.10 SURVEILLANCE Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.
Remove and test the explosive squib from each shear isolation valve of the TIP System.
Verify the leakage rate through each MSIV leakage path is s 34 scfh when tested at
~ 25 psig, and the combined leakage rate for all MSIV leakage paths is s 86 scfh when tested at ~ 25 psig.
PC I Vs 3.6.1.3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Primary Containment Leakage Rate Testing Program Dresden 2 and 3 3.6.1.3-8 Amendment No.
3.6 CONTAINMENT SYSTEMS Secondary Containment
3. 6. 4.1 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.
ACTIONS CONDITION A.
Secondary containment inoperable in MODE 1, 2, or 3.
B.
Required Action and associated Completion Time of Condition A not met.
C.
Secondary containment inoperable during movement of recently irradiated fuel assemblies in the secondary containment.
Dresden 2 and 3 A.1 REQUIRED ACTION Restore secondary containment to OPERABLE status.
- - - -- - - -- -- -NOTE - -- - - - - - - - - --
LCO 3.0.4.a is not applicable when entering MODE 3.
B.1 C.1 Be in MODE 3.
- - - -- -- -NOTE - - - - - - - - -
LCO 3.0.3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 12 hours Immediately 3.6.4.1-1 Amendment No.
3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 Each SCIV shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, SCI Vs 3.6.4.2 During movement of recently irradiated fuel assemblies in the secondary containment.
ACTIONS NOTES- - - -- - - - - - - -- - -- - - -- - - - -- - - -- - - - - - - -
1.
Penetration flow paths may be unisolated intermittently under administrative controls.
2.
Separate Condition entry is allowed for each penetration flow path.
3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable.
one closed and de-activated automatic valve, closed manual valve, or blind flange.
(continued)
Dresden 2 and 3 3.6.4.2-1 Amendment No.
ACTIONS CONDITION C.
Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, or 3.
D.
Required Action and associated Completion Time of Condition A or B not met during movement of recently irradiated fuel assemblies in the secondary containment.
Dresden 2 and 3 C.1 C.2 D.1 REQUIRED ACTION Be in MODE 3.
Be in MODE 4.
SCI Vs 3.6.4.2 COMPLETION TIME 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours
-- - - - -- -NOTE - - - - - - - --
LCO 3.0.3 is not applicable.
Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.
3.6.4.2-3 Amendment No.
3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, SGT System 3.6.4.3 During movement of recently irradiated fuel assemblies in the secondary containment.
ACTIONS CONDITION A.
One SGT subsystem inoperable.
B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
C.
Required Action and associated Completion Time of Condition A not met during movement of recently irradiated fuel assemblies in the secondary containment.
Dresden 2 and 3 A. l REQUIRED ACTION Restore SGT subsystem to OPERABLE status.
- - - - - - - - -- - -NOTE- -- - - - - - - - - --
LCO 3.0.4.a is not applicable when entering MODE 3.
B.l Be in MODE 3.

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

LCO 3.0.3 is not applicable.
C.l C.2 Place OPERABLE SGT subsystem in operation.
Suspend movement of recently irradiated fuel assemblies in secondary containment.
COMPLETION TIME 7 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Immediately Immediately (continued) 3.6.4.3-1 Amendment No.
CONDITION D.
Two SGT subsystems inoperable in MODE 1, 2, or 3.
E.
Required Action and associated Completion Time of Condition D not met.
F.
Two SGT subsystems inoperable during movement of recently irradiated fuel assemblies in the secondary containment.
Dresden 2 and 3 D.1 REQUIRED ACTION Restore one SGT subsystem to OPERABLE status.
- - - -- - - --- - - NOTE - - - - - - - - - -- - -
LCO 3.0.4.a is not applicable when entering MODE 3.
E.1 F.1 Be in MODE 3.

NOTE--------

LCO 3.0.3 is not applicable.
Suspend movement of recently irradiated fuel assemblies in secondary containment.
SGT System 3.6.4.3 ACTIONS COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 12 hours Immediately 3.6.4.3-2 Amendment No.
3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO
3.7.4 APPLICABILITY
The CREV System shall be OPERABLE.
- - - - - - - - -- -- --- - -NOTE -- - - -- -- - - -- - -- - -
The main control room envelope CCRE) boundary may be opened intermittently under administrative control.
MODES 1, 2, and 3, CREV System 3.7.4 During movement of recently irradiated fuel assemblies in the secondary containment.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3 for to OPERABLE status.
reasons other than Condition c.
B.
Required Action and
- --- - -- - -- - -NOTE - - - - - - - - - - - - -
associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.
not met in MODE 1, 2' or 3.
B.1 Be in MODE 3.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)
Dresden 2 and 3 3.7.4-1 Amendment No.
ACTIONS (continued)
CONDITION C.
CREV system inoperable C.1 due to inoperable CRE boundary in MODE 1, 2, or 3.
D.
Required Action and associated Completion Time of Condition C not met in MODE 1, 2, or 3.
C.2 C.3 D.1 D.2 REQUIRED ACTION Initiate action to implement mitigating actions.
Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.
Restore CRE boundary to OPERABLE status.
Be in MODE 3.
Be in MODE 4.
E.
CREV System inoperable during movement of
-- --- - - - -- - - NOTE - - - - -- - - - - -- -
LCO 3.0.3 is not applicable.
recently irradiated fuel assemblies in the secondary containment.
E.1 CREV System inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
CREV System 3.7.4 COMPLETION TIME Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours Immediately Dresden 2 and 3 3.7.4-2 Amendment No.
3.7 PLANT SYSTEMS Control Room Emergency Ventilation AC System 3.7.5 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Control Room Emergency A.1 Ventilation AC System inoperable in MODE 1, Restore Control Room 30 days 2, or 3.
B.
Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3.
C.
Control Room Emergency Ventilation AC System inoperable during movement of recently irradiated fuel assemblies in the secondary containment.
Dresden 2 and 3 Emergency Ventilation AC System to OPERABLE status.
- -- -- - - --- - -NOTE - - -- - - - - - -- - -
LCO 3.0.4.a is not applicable.
when entering MODE 3.
B.1 Be in MODE 3.
- -- - -- -- - - - - NOTE - - -- - -- - - - - - -
LCO 3.0.3 is not applicable.
C.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Immediately 3.7.5-1 Amendment No.
ACTIONS AC Sources-Shutdown 3.8.2

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

LCO 3.0.3 is not applicable.
CONDITION A.
One required offsite circuit inoperable.
Dresden 2 and 3 REQUIRED ACTION
- - - - - - -- -- - -NOTE - - - - - - - - - - - - -
Enter applicable Condition and Required Actions of LCO 3.8.8, when any required division is de-energized as a result of Condition A.
A.1 OR A.2.1 A.2.2 A.2.3 Declare affected required feature(s),
with no offsite power available, inoperable.
Suspend CORE ALTERATIONS.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
Initiate action to restore required offsite power circuit to OPERABLE status.
COMPLETION TIME Immediately Immediately Immediately Immediately (continued) 3.8.2-2 Amendment No.
ACTIONS CONDITION REQUIRED ACTION B.
One required DG inoperable.
B.1 Suspend CORE AL TERA TI ONS.
AND B.2 Suspend movement of recently irradiated fuel assemblies in secondary containment.
AND B.3 Initiate action to restore required DG to OPERABLE status.
SURVEILLANCE REQUIREMENTS SR 3.8.2.1 SURVEILLANCE
- - - - - - - - - - - - - - - --- -NOTES- - --- - - - - - - - - - - - - - -
1.
The following SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, and SR 3.8.1.14 through SR 3.8.1.19.
2.
SR 3.8.1.13 and SR 3.8.1.19 are not required to be met when associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "ECCS-Shutdown."
For AC sources required to be OPERABLE the SRs of Specification 3.8.1, except SR 3.8.1.9, SR 3.8.1.20, and SR 3.8.1.21 are applicable.
AC Sources-Shutdown 3.8.2 COMPLETION TIME Immediately Immediately Immediately FREQUENCY In accordance with applicable SRs Dresden 2 and 3 3.8.2-3 Amendment No.
DC Sources-Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO
3.8.5 APPLICABILITY
ACTIONS One 250 VDC and one 125 VDC electrical power subsystem shall be OPERABLE to support the 250 VDC and one 125 VDC Class lE electrical power distribution subsystems required by LCO 3.8.8, "Distribution Systems-Shutdown."
MODES 4 and 5, During movement of recently irradiated fuel assemblies in the secondary containment.
- - -- - - - - -- - - - -- - - - - - - - - - - - - --- - -- -- - -NOTE - - -- - - - - - - - - -- -- - - -- - - -- - -- - -- - - - - - --
LCO 3.0.3 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more required A.1 Declare affected required feature(s) inoperable.
Immediately DC electrical power subsystems inoperable.
A.2.1 A.2.2 Suspend CORE ALTERATIONS.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
Immediately Immediately A.2.3 Initiate action to Immediately restore required DC electrical power subsystems to OPERABLE status.
Dresden 2 and 3 3.8.5-1 Amendment No.
SURVEILLANCE REQUIREMENTS SR 3.8.5.1 SURVEILLANCE
- - - - -- - - -- - -- - - -- - -NOTE - - --- - - - - - - - - - - -- - - -
The following SRs are not required to be performed for the 250 voe electrical power subsystem:
SR 3.8.4.2 and SR 3.8.4.4.
For DC electrical power subsystems required to be OPERABLE the following SRs are applicable:
SR 3.8.4.1, SR 3.8.4.2, SR 3.8.4.3, and SR 3.8.4.4.
DC Sources-Shutdown 3.8.5 FREQUENCY In accordance with applicable SRs Dresden 2 and 3 3.8.5-2 Amendment No.
ACTIONS Distribution Systems-Shutdown 3.8.8 CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2.2 A.2.3 A.2.4 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
Initiate actions to restore required AC and DC electrical power distribution subsystems to OPERABLE status.
Immediately Immediately Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
SURVEILLANCE REQUIREMENTS SR 3.8.8.1 Dresden 2 and 3 SURVEILLANCE Verify correct breaker alignments and voltage to required AC and DC electrical power distribution subsystems.
3.8.8-2 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No.
Dresden Nuclear Power Station, Units 2 and 3, Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control" ATTACHMENT 4 -
PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (MARK-UP)
TOC Page i TOC Page ii B 3.3.5.1-10 B 3.3.5.1-12 B 3.3.5.1-14 B 3.3.5.1-15 B 3.3.5.1-30 B 3.3.5.1-32 B 3.3.5.1-35 B 3.3.5.2-1 B 3.3.5.2-2 B 3.3.5.2-3 B 3.3.5.2-4 B 3.3.5.2-5 B 3.3.5.2-6 B 3.3.5.2-7 B 3.3.5.2-8 B 3.3.5.2-9 B 3.3.5.2-10 B 3.3.5.3-1 B 3.3.5.3-2 B 3.3.5.3-3 B 3.3.5.3-4 B 3.3.5.3-5 B 3.3.5.3-6 B 3.3.6.1-19 B 3.3.6.1-21 B 3.3.6.2-4 B 3.3.6.2-5 B 3.3.6.2-7 B 3.3.7.1-3 B 3.5.1-1 B 3.5.2-1 B 3.5.2-2 B 3.5.2-3 B 3.5.2-4 B 3.5.2-5 B 3.5.2-6 B 3.5.2-7 B 3.5.2-8 B 3.5.2-9 B 3.5.2-10 B 3.5.2-11 B 3.5.2-12 B 3.5.2-13 B 3.5.2-14 B 3.5.2-15 B 3.5.2-16 B 3.5.2-17 B 3.5.2-18 B 3.5.2-19 B 3.5.3-1 B 3.5.3-2 B 3.6.1.3-4 B 3.6.1.3-9 B 3.6.1.3-10 B 3.6.2.2-2 B 3.6.4.1-2 B 3.6.4.1-4 B 3.6.4.2-2 B 3.6.4.2-5 B 3.6.4.3-3 B 3.6.4.3-4 B 3.6.4.3-5 B 3.6.4.3-6 B 3.7.4-7 B 3.7.4-10 B 3.7.5-3 B 3.7.5-4 B 3.7.5-5 B 3.8.2-1 B 3.8.2-3 B 3.8.2-4 B 3.8.2-5 B 3.8.2-6 B 3.8.2-7 B 3.8.5-1 B 3.8.5-3 B 3.8.5-4 B 3.8.8-2 B 3.8.8-3 B 3.8.8-4
TABLE OF CONTENTS B 2. 0 B 2.1.1 B 2.1. 2 B 3. 0 B 3. 0 B 3. 1 B 3. 1.1 B 3.1. 2 B 3.1. 3 B 3.1. 4 B 3.1. 5 B 3. 1. 6 B 3.1. 7 B 3. 1.8 B 3.2 B 3. 2.1 B 3.2. 2 B 3. 2. 3 B 3.3 B 3.3.1.1 B 3.3. 1.2 B 3. 3.1. 3 B 3. 3. 2. 1 B 3. 3. 2. 2 B 3. 3. 3. 1 B 3. 3. 4. 1 B 3. 3.5.1 B 3.3.5. 2 B 3.3. 5. 3 B 3.3.6.l B 3.3. 6. 2 B 3. 3. 6. 3 B 3.3.7.1 B 3. 3. 7. 2 B 3. 3. 8. 1 B 3.3.8.2 SAFETY LIMITS (SLs)
Reactor Core SLs.................................... B 2.1.1-1 Reactor Coolant System (RCS) Pressure SL........... B 2. 1. 2-1 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY... B 3.0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............ B 3.0-13 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SOM)............................... B 3.1.1-1 Reactivity Anomalies................................ B 3.1.2-1 Control Rod OPERABILITY............................. B 3.1.3-1 Control Rod Scram Times............................. B 3.1. 4-1 Control Rod Scram Accumulators...................... B 3. 1. 5-1 Rod Pattern Control................................. B 3. 1.6-1 Standby Liquid Control (SLC) System................. B 3.1.7-1 Scram Discharge Volume (SOV) Vent and Drain Valves.. B 3.1. 8-1 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR).......................................... B 3.2. 1-1 MINIMUM CRITICAL POWER RATIO (MCPR)................. B 3.2. 2-1 LIN EAR HEAT GENERATION RATE (LHGR)................. B 3.2. 3-1 INSTRUMENTATION Reactor Protection System (RPS) Instrumentation..... B 3. 3. 1. 1-1 Source Range Monitor (SRM) Instrumentation.......... B 3.3.1.2-1 Oscillation Power Range Monitor (OPRM)
Instrumentation................................... B 3. 3.1.3-1 Control Rod Block Instrumentation................... B 3.3. 2.1-1 Feedwater System and Main Turbine High Water Level Trip Instrumentation.............................. B 3.3. 2.2-1 Post Accident Monitoring (PAM) Instrumentation...... B 3.3. 3.1-1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation.............. B 3.3.4. 1-1 Emergency Core Cooling System (ECCS)
Instrumentation................................... B 3.3. 5.1-1 Isolation Condenser CIC) System Instr~mentationReactor Pressure Vessel (RPVJ Water Inventory Control Instrumentation................................... B 3. 3. 5.2-1 Isolation Condenser (IC) System Instrumentation..... B 3. 3. 5.3-1 Primary Containment Isolation Instrumentation....... B 3.3.6.1-1 Secondary Containment Isolation Instrumentation..... B 3. 3. 6. 2-1 Relief Valve Instrumentation........................ B 3. 3. 6. 3-1 Control Room Emergency Ventilation (CREV)
System Instrumentation............................ B 3. 3. 7. 1-1 Mechanical Vacuum Pump Trip Instrumentation......... B 3. 3. 7. 2-1 Loss of Power (LOP) Instrumentation................. B 3.3.8. 1-1 Reactor Protection System (RPS) Electric Power Monitoring........................................ B 3.3.8.2-1 Dresden 2 and 3 (continued)
Revision
TABLE OF CONTENTS (continued)
B 3.4 B 3. 4. 1 B 3.4. 2 B 3.4.3 B 3.4.4 B 3.4. 5 B 3.4. 6 B 3.4. 7 B 3.4.8 B 3.4.9 B 3.4. 10 B 3.5 COND ENSER B 3. 5. 1 B 3.5. 2 B 3.5.3 B 3.6 B 3.6. 1.1 B 3.6.1. 2 B 3.6. 1.3 B 3.6.1.4 B 3.6.1.5 B 3.6. 1.6 B 3.6. 1.7 B 3.6.1.8 B 3. 6. 2. 1 B 3.6.2. 2 B 3.6.2. 3 B 3. 6.2.4 B 3.6. 2. 5 B 3. 6. 3. 1 B 3.6.4. l B 3.6.4.2 B 3.6.4. 3 B 3.7 B 3. 7. 1 B 3. 7.2 B 3. 7.3 B 3. 7.4 REACTOR COOLANT SYSTEM (RCS)
Recirculation Loops Operating....................... B 3.4. 1-1 Jet Pumps........................................... B 3.4. 2-1 Safety and Relief Valves........................... B 3.4. 3-1 RCS Operational LEAKAGE............................. B 3.4. 4-1 RCS Leakage Detection Instrumentation............... B 3.4. 5-1 RCS Specific Activity............................... B 3.4. 6-1 Shutdown Cooling (SDC) System-Hot Shutdown.......... B 3.4. 7-1 Shutdown Cooling CSDC) System-Cold Shutdown......... B 3.4.8-1 RCS Pressure and Temperature (PIT) Limits........... B 3.4. 9-1 Reactor Steam Dome Pressure......................... B 3.4. 10-1 EMERGENCY CORE COOLING SYSTEMS CECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, - AND ISOLATION CIC) SYSTEM EC C S- 0 p e r a t i n g...................................... B 3. 5. 1 - 1 EGGS ShtltdmmRPV Water Inventory Control............ B 3.5. 2-1 IC System........................................... B 3.5. 3-1 CONTAINMENT SYSTEMS Primary Containment................................. B 3.6.1. 1-1 Primary Containment Air Lock........................ B 3.6. 1. 2-1 Primary Containment Isolation Valves (PCIVs)........ B 3.6. 1. 3-1 Drywell Pressure.................................... B 3.6. 1.4-1 Drywell Air Temperature............................. B 3. 6. 1. 5-1 Low Set Relief Valves............................... B 3.6. 1.6-1 Reactor Building-to-Suppression Chamber Vacuum Breakers.......................................... B 3. 6. 1. 7-1 Suppression Chamber-to-Drywell Vacuum Breakers...... B 3.6.1.8-1 Suppression Pool Average Temperature................ B 3.6.2.1-1 Suppression Pool Water Level........................ B 3.6. 2.2-1 Suppression Pool Cooling............................ B 3.6. 2.3-1 Suppression Pool Spray.............................. B 3.6. 2.4-1 Drywell-to-Suppression Chamber Differential Pressure.......................................... B 3.6.2. 5-1 Primary Containment Oxygen Concentration............ 8 3. 6.3. 1-1 Secondary Containment............................... B 3. 6. 4. 1-1 Secondary Containment Isolation Valves CSCIVs)...... B 3. 6.4.2-1 Standby Gas Treatment (SGT) System.................. B 3.6.4. 3-1 PLANT SYSTEMS Containment Cooling Service Water (CCSW) System..... B 3. 7. 1-1 Diesel Generator Cooling Water ( DGCW) System........ B 3. 7. 2-1 Ultimate Heat Sink (UHS)............................ B 3. 7. 3-1 Control Room Emergency Ventilation CCREV) System.... B 3. 7.4-1 (continued)
Dresden 2 and 3 i i Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Dresden 2 and 3 ECCS Instrumentation B 3. 3. 5.1 Function must have a required number of OPERAB LE channels, with their setpoints within the specified Allowable Values, where appropriate.
The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
The Table 3.3.5. 1-1, footnote~
is added to show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation.
Allowable Values are specified for each ECCS Function specified in the Table.
Nominal trip setpoints are specified in the setpoint calculations.
The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS.
Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.
Trip setpoints are those predetermined values of output at which an action should take place.
The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e. g.,
trip unit) changes state.
The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.
The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors.
These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components.
The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
Some Functions (i.e, Functions l.c, l.d, 2.c, 4.d, 4. e, 5.d, and 5.e) have both an upper and lower analytic limit that must be evaluated.
The Allowable Values and trip setpoints (continued)
B 3.3.5.1-10 Revision
BASES APP LI CABLE SAFETY ANALYSES, LCO, and APPLICABILITY Dresden 2 and 3 ECCS Instrumentation B 3.3.5. 1 l.a. 2.a.
Reactor Vessel Water Level-Low Low (continued)
The Reactor Vessel Water Level-Low Low Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.
Four channels of CS Reactor Vessel Water Level-Low Low Function are only required to be OPERABLE when the CS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude CS and DG initiation.
Also, four channels of the LPCI Reactor Vessel Water Level - Low Low Function are only required to be OP ERABLE when the LPCI System is required to be OPERABLE to ensure no single instrument failure can preclude LPCI initiation.
Refer to LCO 3.5.l and LCO 3.5.2, "EGGS Shb1tdmm," for Applicability Bases for the 10*,1 pressblre EGGS sblbsystems; LCO 3.B.l, "AC Soblrces Operating"; and LCO 3. g. 2, "/\\C Soblrces Shbltdo*,m,"
for Applicability Bases for the DGs.
1. b. 2. b.
Drywe 11 Pressure-High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary CRCPB).
The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel damage.
The Drywell Pressure-High Function, along with the Reactor Water Level-Low Low Function, is directly assumed in the LOCA analysis (Ref. 2).
The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
High drywell pressure signals are initiated from four pressure switches that sense drywel 1 pressure.
The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.
The Drywell Pressure-High Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE.
Thus, four channels of the CS Drywell Pressure-High Function are required to be (continued)
B 3.3.5. 1-12 Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Dresden 2 and 3 ECCS Instrumentation B 3.3.5.l l.c. 2. c.
Reactor Steam Dome Pressure-Low (Permissive)
(continued)
Two channels of Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation.
Refer to LGO 3.5.l and LGO 3.5.2 for Applicability gases for the low press~re EGGS s~bsystems l.d. 2.f.
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 sufficiently 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.
The CS Pump Discharge Flow-Low (Bypass) Function is assumed to be OPERABLE and capable of closing the minimum flow valves to ensure that the CS flow assumed during the transients and accidents analyzed in References 1, 2, and 3 is met.
The LPCI Pump Discharge Flow-Low (Bypass) Function is only required to be OPERABLE for opening since the LPCI minimum flow valves are assumed to remain open during the transients and accidents analyzed in References 1, 2, and 3.
The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limi ts of 10 CFR 50.46.
One flow transmitter per CS pump and one flow transmitter per LPCI subsystem are used to detect the associated subsystems' flow rates.
The logic is arranged such that each transmitter causes its associated minimum flow valve to open when flow is low with the pump running.
The logic will close the minimum flow valve once the closure setpoint is exceeded.
The Pump Discharge Flow-Low (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump.
The Core Spray Discharge Flow-Low (Bypass) Allowable Value is also low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.
For LPCI, the closure of the minimum flow valves is not credited.
(continued)
B 3.3.5. 1-14 Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Dresden 2 and 3 ECCS Instrumentation B 3. 3. 5. 1 l.d. 2. f.
Core Spray and Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)
(continued)
Each channel of Pump Discharge Flow-Low (Bypass) Function (two CS channels and two LPCI channels) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude the ECCS function.
Refer to LGO 3. 5. 1 and LGO 3. 5.2 f or Applicabi l ity Bases for the low pressyre EGGS sybsystems.
1.e. 2.e.
Core Spray and Low Pressure Coolant Injection Pump Start-Time Delay Relay The purpose of this time delay is to stagger the start of CS and LPCI pumps that are in each of Divisions 1 and 2, thus limiting the starting transients on the 4160 V ESS buses.
This Function is only necessary when power is being supplied from the standby power sources (DG).
The CS and LPCI Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident and transien t analyses requiring ECCS initiation.
That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.
There are two CS Pump Start-Time Delay Relays and two LPCI Pump Start-Time Delay Relays, one for each CS pump and one for LPCI pump B and D.
While each time delay relay is dedicated to a single pump start logic, a single failure of a LPCI Pump Start-Time Delay Relay could result in the failure of the three low pressure ECCS pumps, powered from the same ESS bus, to perform their intended function (e.g.,
as in the case where both ECCS pumps on one ESS bus start simultaneously due to an inoperable time delay relay).
This still leaves three of the six low pressure ECCS pumps OPERABLE; thus, the single failure criterion is met (i.e.,
loss of one instrument does not preclude ECCS initiation).
The Allowable Values for the CS and LPCI Pump Start-Time Delay Relays are chosen to be short enough so that ECCS operation is not degraded.
Each CS and LPCI Pump Start-Time Delay Relay Function is required to be OPERABL E only when the associated LPCI subsystem is required to be OPERABLE.
Refer to LGO 3.5. l and LGO 3.5.2 for Applicability Bases for the GS and LPGI sybsystems.
(continued)
B 3.3.5. 1-15 Revision
BASES ACTIONS Dresden 2 and 3 B.l. B.2 and B.3 (continued)
ECCS Instrumentation B 3.3.5.1 systems lose initiation capability. (e) two or more Function 2.d channels are inoperable and untripped such that both trip systems lose initiation capability, or (f) two Function 2.j channels are inoperable and untripped.
For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS and DGs to be declared inoperable.
However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsys tems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and DGs being concurrently declared inoperable.
For Required Action B.2, redundant automatic initiation capability (i.e., loss of automatic start capability for Functions 3. a and 3.b ) is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip sys tem.
In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.3 is not appropriate and the feature(s) associated with the inoperable, untripped channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
As noted (Note 1 to ReqYired Action B.l),
ReqYired Action B.l is only applicable in MODES 1, 2, and 3.
In MODES 4 and 5, the specific initiation time of the 1011 pressYre EGGS is not assYmed and the probability of a LOGA is lower.
ThYs, a total loss of initiation capability for 24 hoYrs (as allowed by ReqYired Action B.3) is allowed dYring MODES 4 and 5.
There is no similar Note provided for ReqYired Action B.2 since HPCI instrYmentation is not reqYired in MODES 4 and 5; thYs, a Note is not necessary.
Notes A note ~i s also provided (the Note-- to Required Action B. 1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable.
This ensures that the proper loss of initiation capability check is performed.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities.
This Completion Time also allows for an exception to the normal (continued)
B 3.3.5.1-30 Revision
BASES ACTIONS Dresden 2 and 3 C.l and C. 2 (continued)
ECCS Instrumentation B 3.3.5. l channels are inoperable in both trip systems, (c) two Function l.e channels are inoperable, (d) two Function 2.e channels are inoperable, (e) two or more Function 2. g channels, associated with a recirculation pump are inoperable such that both trip systems lose initiation capability, (f) two or more Function 2.h channels are inoperable such that both trip systems lose initiation capability, (g) two Function 2. i channels are inoperable, or (h) two Function 2.k channels are inoperable.
Since each inoperable channel would have Required Action C.l applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system to be declared inoperable.
However, since channels for both low pressure ECCS subsystems are inoperable (e.g.,
both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable.
For Functions l.e, and 2. e, the affected portions are the associated low pressure ECCS pumps.
For Functions l.c and 2.c, the affected portions are the associated ECCS pumps and valves.
For Functions 2.g, 2.h, 2. i, and 2.k, the affected portions are the associated LPCI valves.
In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action C. 2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
As noted (Note 1), Required Action C.l is only applicable in MODES 1, 2, and 3.
In MODES 4 and 5, the specific initiation time of the EGGS is not assumed and the probability of a LOGA is lower.
Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action C.2) is allo11ed durin§ MODES 4 and 5.
The Note-states that Required Action C.l is only applicable for Functions l. c, l.e, 2.c, 2.e, 2. g, 2. h, 2.i, and 2.k.
Required Action C.l is not applicable to Function 3.g (which also requires entry into this Condition if a channel in this Function is inoperable), since it is the HPCI Manual Initiation Function which is not assumed in any accident or transient analysis.
Thus, a total loss of HPCI Manual (continued)
B 3.3.5.1-32 Revision
BASES ACTIONS Dresden 2 and 3 D.l. D.2.1 and D. 2.2 (continued)
ECCS Instrumentation B 3.3.5. 1 desired to perform Required Actions D.2.1 and D. 2.2 (e. g.,
as in the case where shifting the suction source could drain down the HPCI suction piping), Condition H must be entered and its Required Action taken.
E. l and E. 2 Required Action E.l is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Core Spray and Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass) Functions result in redundant automatic initiation capability being lost for the feature(s).
For Required Action E.l, the features would be those that are initiated by Functions l.d and 2. f (i.e., low pressure ECCS).
Redundant automatic initiation capability is lost if (a) two Function l.d channels are inoperable or (b) two Function 2. f channels are inoperable.
Since each inoperable channel would have Required Action E.l applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected low pressure ECCS pump to be declared inoperable.
However, since channels for more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable.
In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E. 2 is not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
As noted (Note 1 to ReqYired Action E.1), ReqYired Action E.l is only applicable in MODES 1, 2, and 3.
In MODES 4 and 5, the specific initiation time of the EGGS is not assYmed and the probabi 1 i ty of a LOGA is 1 o*.1er.
ThYs, a total loss of initiation capability for 7 days (as allo*.1ed by ReqYired /\\ction E.2) is allo*.1ed dYrin9 MODES 4 and 5.
A A Note is also provided (the Note~ to Required Action E.l) to delineate that Required Action E.l is only applicable to low pressure ECCS Functions.
Required Action E.l is not applicable to HPCI Function 3. f since the loss of one (continued)
B 3.3.5.1-35 Revision
RPV Water Inventory Control Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3. 5. 2 BASES BACKGROUND Reactor Pressure Vessel (RPVJ Water Inventory Control Instrumentation The RPV contains penetrations below the top of the active fuel (TAFJ 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 2.1.1.3 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 (LSSSJ 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 (SLJ 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 MODES 1, 2, and 3 in LCD 3.3.5.1, "Emergency Core Cooling System (ECCSJ Instrumentation," or LCD 3.3.6.1, "Primary Containment Isolation instrumentation".
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. 1.3 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 (continued)
Dresden 2 and 3 B 3.3.5. 2-1 Revision
BASES RPV Water Inventory Control Instrumentation B 3. 3.5. 2 BACKGROUND (continued)
APPLICABLE SAFETY ANALYSES, LCD, and APPLICABILITY Dresden 2 and 3 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 LCD 3. 5. 2, "Reactor Pressure Vessel (RPV) Water Inventory Control," and the definition of DRAIN TIME.
There are functions that are required for manual operation of the ECCS injection/spray subsystem required to be OPERABLE by LCD 3. 5. 2 and other functions that support automatic isolation of Shutdown Cooling (SDCJ and Reactor Water Cleanup (RWCUJ system penetration flow path(s) on low RPV water level.
The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCJ). The equipment involved with each of these systems is described in the Bases for LCD 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. 1.3 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 MODES 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 a11ows 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, Joss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually operated to maintain adequate reactor vessel water level.
(continued)
B 3. 3. 5. 2-2 Revision
BASES APPLICABLE SAFETY ANALYSES, LCD, and APPLICABILITY (continued)
Dresden 2 and 3 RPV Water Inventory Control Instrumentation B 3.3. 5. 2 As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public heal th and safety. Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2J(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.
Core Sorav and Low Pressure Coolant Infection Systems l.a. 2.a.
Reactor Steam Dome Pressure-Low (Permissive)
Low reactor steam dome pressure signals are used as permi ss i ves 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 Reactor Steam Dome Pressure-Low (Permissive) signals are initiated from two pressure switches that sense the reactor steam dome pressure.
The Allowable Value is low enough to prevent overpressurizing the equipment in the low pressure ECCS.
Two channels of Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE in MODES 4 and 5 when ECCS is required to be OPERABLE by LCD 3.5.2.
(continued)
B 3. 3. 5. 2-3 Rev i sion
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Dresden 2 and 3 RPV Water Inventory Control Instrumentation B 3. 3. 5. 2 1.b. 2.b.
Core Sorav and Low Pressure Coolant Iniection Pump Discharge Fl ow-Low CBvpass)
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 sufficiently 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 fl ow transmitter per CS pump and one fl ow transmitter per LPCI loop are used to detect the associated subsystems' flow rates.
The logic is arranged such that each transmitter causes its associated minimum flow valve to open when flow is low with the pump running.
The logic will close the minimum flow valve once the closure setpoint is exceeded.
The Pump Discharge Fl ow-Low (Bypass) Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump.
The Core Spray Discharge Flow-Low (Bypass) Allowable Value is also low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.
For LPCI, the closure of the minimum flow valves is not credited.
Each channel of Pump Discharge Flow-Low (Bypass) Function is only required to be OPERABLE in MODES 4 and 5 when the associated ECCS subsystem is required to be OPERABLE by LCD 3.5.2 to ensure the pumps are capable of injecting into the RPV when manually operated.
(continued)
B 3. 3. 5. 2-4 Revision
BASES RPV Water Inventory Control Instrumentation B 3. 3. 5.2 APPLICABLE Shutdown Cooling CSDCJ Svstem Isolation SAFETY ANALYSES, LCO, and
3. a - Reactor Vessel Water Level-Low APPLICABILITY (continued)
The definition of Drain Time allows crediting the closing of penetration flow paths that are 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.
The Reactor Vessel Water Level-Low Function associated with SDC System isolation may be credited for automatic isolation of penetration flow paths associated with the SDC System.
Dresden 2 and 3 The Reactor Vessel Water Level-Low Function receives input from four reactor vessel water level channels.
Each channel inputs into one of four trip strings.
Two trip strings make up a trip system and both trip systems must trip to cause an isolation of the SDC suction isolation valves.
Any channel will trip the associated trip string.
Only one trip string must trip to trip the associated trip system.
The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
Therefore, one trip string in each trip system is required to provide for automatic SOC system isolation.
The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the Primary Containment Isolation Instrumentation Reactor Vessel Water Level-Low Allowable Value (LCO 3.3. 6. 1), since the capability to cool the fuel may be threatened.
The Reactor Vessel Water Level-Low Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.
Shutdown Cooling System Isolation Functions isolate some Group 3 valves (SDC isolation valves).
(continued)
B 3. 3.5. 2-5 Revision
BASES RPV Water Inventory Control Instrumentation B 3. 3. 5. 2 APPLICABLE Reactor Water Cleanup CRWCU) System Isolation SAFETY ANALYSES, LCD, and 4. a - Reactor Vessel Water level-Low APPLICABILITY (continued)
The definition of Drain Time allows crediting the closing of penetration flow paths that are 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 Dresden 2 and 3 actuated by RPV water level isolation instrumentation.
The Reactor Vessel Water Level-Low Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.
The Reactor Vessel Water Level-Low Isolation Function receives input from four reactor vessel water level channels.
Each channel inputs into one of four trip stings.
Two trip strings make up a trip system and both trip sy tems must trip to cause an isolation of the RWCU valves.
Any channel will trip the associated trip string.
Only one trip string must trip to trip the associated trip system.
The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
Therefore, one trip string in each trip system is required to provide for automatic RWCU system isolation.
The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Allowable Value (LCD 3. 3. 5.1), since the capability to cool the fuel may be threatened.
The Reactor Vessel Water Level-Low Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.
RWCU Functions isolate some Group 3 valves (RWCU isolation valves).
(continued)
B 3. 3. 5. 2-6 Revision
BASES (continued)
RPV Water Inventory Control Instrumentation B 3. 3.5.2 ACTIONS A Note has been provided to modify the ACTIONS related to RPV Water Inventory Control instrumentation channels.
Section 1. 3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, Dresden 2 and 3 subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition.
Section 1. 3 also specifies that Required Actions continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable RPV Water Inventory Control instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable RPV Water Inventory Control instrumentation channel.
Required Action A. l directs entry into the appropriate Condition referenced in Table 3. 3. 5. 2-1.
The applicable Condition referenced in the table is Function dependent.
Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
8.1 and 8.2 Shutdown cooling (SOC) system Isolation, Reactor Vessel Water Level-Low, and Reactor Water Cleanup System, Reactor Vessel Water Level-Low functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating Drain Time.
If the instrumentation is inoperable, Required Action 8. 1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation.
Required Action 8.2 directs calculation of DRAIN TIME.
The calculation cannot credit automatic isolation of the affected penetration flow paths.
(continued)
B 3.3.5. 2-7 Revision
BASES ACTIONS (continued)
Dresden 2 and 3 RPV Water Inventory Control Instrumentation B 3.3. 5. 2 Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions.
If a required channel of the permissive is inoperable, manual operation of ECCS may be prohibited.
Therefore, the affected channel(s) must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
With the affected channel(s) in the trip condition, manual operation may be performed.
The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in the trip condition.
If a CS or LPCI Pump Discharge Fl ow-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 system to ensure the pump does not overheat.
The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities.
The Completion Time is appropriate given the ability to manually start the ECCS pumps and open the minimum flow valves and to manually ensure the pump does not overheat.
With the Required Action and associated Completion Time of Condition C or 0 not met, the associated low pressure ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately.
(continued)
B 3.3. 5. 2-8 Revision
BASES (continued SURVEILLANCE REQUIREMENTS Dresden 2 and 3 RPV Water Inventory Control Instrumentation B 3.3.5. 2 As noted in the beginning of the SRs, the SRs for each RPV Water Inventory Control instrumentation Function are found in the SRs column of Table 3.3. 5. 2-1.
SR 3. 3. 5. 2. 1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred.
A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.
It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.
A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability.
If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
SR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.
This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.
This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
(continued)
B 3.3.5. 2-9 Revision
BASES SURVEILLANCE REQUIREMENTS REFERENCES Dresden 2 and 3 RPV Water Inventory Control Instrumentation B 3. 3. 5. 2 SR 3. 3.5.2. 2 (continued)
Any setpoint adjustment shall be consistent with the assumptions of the current plant-specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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 Cool ant 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.
B 3.3.5. 2-10 Revision
IC System Instrumentation B 3. 3.5..iJ B 3.3 INSTRUMENTATION B 3
. 3
. 5
. ~3 Isolation Condenser (IC) System Instrumentation BASES BACKGROUND APPLICABLE SAFETY ANALYSES LCD Dresden 2 and 3 The purpose of the IC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser).
A more complete discussion of IC System operation is provided in the Bases of LCO 3.5.3, "IC System. "
The IC System may be initiated by either automatic or manual means.
Automatic initiation occurs for sustained conditions of reactor vessel pressure high.
The variable is monitored by four pressure switches that are connected to four time delay relays.
The outputs of the time delay relays are connected in a one-out-of-two logic to a trip relay.
The output of the trip relays are connected in a two-out-of-two logic arrangement.
Once initiated, the IC logic can be overridden by the operator.
The function of the IC System to provide core cooling to the reactor is used to respond to a main steam line isolation event.
Although the IC System is an Engineered Safety Feature System, no credit is taken in the accident analyses for IC system operation.
Based on its contribution to the reduction of overall plant risk, however, the IC System, and therefore its instrumentation, satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
The OPERABILITY of the IC System instrumentation is dependent upon the OPERABILITY of the four channels of the Reactor Vessel Pressure-High Function.
Each channel must have its setpoint within the Allowable Value specified in SR 3.3.5.2.2.
The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
The Allowable Value for the IC System instrumentation Function is specified in the SR.
Nominal trip setpoints are spec i fied i n the setpoint calculat i ons.
The nomina l setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS.
(continued)
B 3. 3.5. 3-1 Revision
BASES LCO (continued)
APPLICABILITY Dresden 2 and 3 IC System Instrumentation B 3.3.5.iJ Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
A channel i s inoperable if its actual trip setpoint is not within its required Allowable Value.
Trip setpoints are those predetermined values of output at which an action should take place.
The setpoints are compared to the actual process parameter (e.g., reactor vessel pressure), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,
relay) changes state.
The analytic limits (or design limits) are derived from the limiting values of the process parameters obtained from the safety analysis.
The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors.
These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components.
The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the in strumentation.
The Reactor Vessel Pressure-High Allowable Value is set high enough to ensure that a potential event is in process.
The time delay is determined by engineering judgement to avoid spurious unnecessary activations of the IC by allowing time for the pressure spike, caused by a main steam isolation valve or stop valve closure, to decay.
Four channels of Reactor Vessel Pressure-High Function are available and are required to be OPERABLE when IC is required to be OPERABLE to ensure that no single instrument failure can preclude IC initiation.
The Function is required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure > 150 psig since this is when IC is required to be OPERABLE.
(Refer to LCO 3.5.3 for Applicability Bases for the IC System.)
(continued)
B 3.3.5.3-2 Revision
IC System Instrumentation B 3. 3.5.i J BASES (continued)
ACTIONS A Note has been provided to modify the ACTIONS related to Dresden 2 and 3 IC System instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables
. expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition.
Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable IC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels.
As such, a Note has been provided that allows separate Condition entry for each inoperable IC System instrumentation channel.
A.l and A. 2 Required Action A. l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in a complete loss of automatic initiation capability for the IC System.
In this case, automatic initiation capability is lost if two channels associated with the same trip relay are inoperable and untripped.
In this situation (loss of automatic initiation capability),
the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of required Action A.2 is not appropriate, and the IC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of IC initiation capability.
The Completi on Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities.
This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock. "
For Required Action A.1, the Completion Time only begins upon discovery that the IC System cannot be automatically initiated due to two or more inoperable, untripped Reactor Vessel Pressure-High channels.
The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time for discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
(continued)
B 3. 3.5. 3-3 Revision
BASES ACTIONS SURVEILLANCE REQUIREMENTS Dresden 2 and 3 A.l and A. 2 (continued)
IC System Instrumentation B 3.3. 5..iJ Because of the redundancy of sensors available to provide initiation signals and the fact that the IC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status.
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Act ion A. 2.
Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
Alternately, if it is not desired to place the channel in trip (e. g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition B must be entered and its Required Action taken.
With any Required Action and associated Completion Time of Condition A not met, the IC System may be incapable of performing the intended function, and the IC System must be declared inoperable immediately.
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the Reactor Vessel Pressure-High Function maintains initiation capability.
Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
SR 3. 3. 5..i3. 1 A CHANN EL FUNCTIONAL TEST is performed on each required channel to ensure that the channel wi 11 perform the intended function.
A successful test of the required contact(s) of a channel relay may be performed by the verification of the (continued)
B 3. 3.5.3-4 Revision
BASES SURV EIL LANCE REQUIREMENTS Dresden 2 and 3 SR 3
. 3
. 5
. ~3.l (continued)
IC System Instrumentation B 3.3. 5.iJ change of state of a single contact of the relay.
This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.
This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.5.iJ.2 and SR 3.3.5.iJ.3 A CHANN EL CA LIB RATION is a complete check of the instrument loop and the sensor.
This test verifies the channel responds to the measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
A Note to SR 3.3.5. 3.2 states that this SR is not required for the time delay portion of these channels.
This allowance is consistent with the plant specific setpoint methodology.
This portion of the channels must be calibrated in accordance with SR 3.3.5.3.3.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
B 3.3.5. 3-5 Revision
BASES SURVEILLANCE REQUIREMENTS (continued)
REFERENCES Dresden 2 and 3 IC System Instrumentation B 3.3.5. ~3 SR 3.3.5. ~3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel.
The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
1.
GENE-770-06-2-A, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of -Service Times for Selected Instrumentation Technical Specifications," December 1992.
B 3.3.5. 3-6 Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Dresden 2 and 3 Primary Containment Isolation Instrumentation B 3.3.6.1 6.b.
Reactor Vessel Water Level-Low Low RPV water level indicates that the capability to cool the fuel may be threatened.
Should RPV water level decrease too far, fuel damage could result.
Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break.
The Reactor Vessel Water Level-Low Function associated with Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the Shutdown Cooling System is bounded by breaks of the recirculation and MSL.
The Shutdown Cooling System isolation on low RPV water level supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the Shutdown Cooling System.
Reactor Vessel Water Level -
Low signals are initiated from four differential pressure 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.
Four channels (two channels per trip system) of the Reactor Vessel Water Level -
Low Function are available and are required to be OPERABLE to ensure that no single in strument failure can preclude the isolation function.
As noted (footnote (b) to Table 3.3.6.l 1), only one channel per trip system (with an isolation signal available to one shYtdown cooling pYmp sYction isolation valve) of the Reactor Vessel Water Level Low FYnction is re~Yired to be OPERABLE in MODES 4 and 5, provided the ShYtdo11n Cooling System integrity is maintained.
System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel throYgh the system.
The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level -Low Allowable Value (LCO 3.3.1.1 ), since the capability to cool the fuel may be threatened.
The Reactor Vessel Water Level-Low Function is only required to be OPERABLE in MODE& 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel.
In MODES 1 and 2, another isolation (i.e., Recirculation Line Water Temperature-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this B 3. 3.6.1 -19 Revision
BASES APPLICABLE SAFETY /\\NAL YS ES LCO, and
/\\PPLIC/HlILITY ACTIONS Dresden 2 and 3 Primary Containment Isolation Instrumentation B 3. 3. 6. 1
6. b.
React0r Vessel 1Alater Level L0w ( rnnti nbled)
The React0r Vessel 111ater Level L0w Fblncti 0n is 0nly re~blired t0 be OPERABLE in MODES 3, 4, and 5 to prevent this potential fl 01.* path from l o*.ieri ng the reaGtor vessel level to the top of the fblel.
In MODES 1 and 2, another i sol ati on (i.e., Recircbllation Line Water Temperatblre ~igh) and administrative controls ensblre that this flo11 path remains isolated to prevent blnexpected loss of inventory via this fl mi path.
This Fbincti on isolates the Group 3 shbltdo 1.m cooling *1al ves.
A Note has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divi sions, subsystems, components, or variables expressed in the Condition, di scovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also speci f ies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable primary containment isolation instrumentation channel s provide appropriate compen satory measures for separate inoperable channel s.
As such, a Note has been provided that allows separate Condition entry for each inoperable pr imary containment isolation in strumentation channel.
Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s, depending on the Function (12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s for those Functions that have channel components common to RPS instrumentation and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Functions that do not have channel components common to RPS instrumentation), has been shown to be acceptable (Refs. 8 and 9) to permit restoration of any inoperable channel to OPERABLE status.
This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B. 1 Bases).
If the inoperable channel cannot be restored to OPERABLE status within the al l owab l e out of service time, the channel must be placed in the tripped condition per Required Action A.1.
Placing the inoperable channel in trip would conservatively B 3. 3.6.1 -21 Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Dresden 2 and 3 Secondary Containment Isolation Instrumentation B 3.3. 6.2 The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
1.
Reactor Vessel Water Level-Low Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened.
Should RPV water l evel decrease too far, fuel damage could result.
An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release.
The Reactor Vessel Water Level - Low Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals.
The isolation and initiation of systems on Reactor Vessel Water Level-Low support actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Ref. 2).
Reactor Vessel Water Level-Low signals are initiated from differential pressure 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.
Four channels of Reactor Vessel Water Level-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Reactor Vessel Water Level-Low Allowable Value was chosen to be the same as the Reactor Protection System (RPS)
Reactor Vessel Water Level-Low Allowable Value (LCO 3.3. 1.1, "Reactor Protection System (RPS)
Instrumentation"), since this could indicate that the capability to cool the fuel is being threatened.
The Reactor Vessel Water Level-Low Function is required to be OPERABLE in MODES 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 MODES 4 and 5, the probab i lity and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function i s not required.
In addition, the (continued)
B 3. 3. 6. 2-4 Revision
BASES Secondary Containment Isolation Instrumentation B 3.3.6. 2 APPLICABLE
2.
Drywell Pres sure-High 1.
Reactor Vessel \\..'ater Level L011 (con ti n1:1ed)
SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Dresden 2 and 3 High drywell pressure can indicate a break in the reactor F1:1nction is also req1:1ired to be OPERABLE during operations coolant pressure boundary (RCPB).
An isolation of the w-i--t-R a potential for draining the reactor vessel COPDRl/s) to ensure that offsite dose limits are not exceeded if core damage occurs.
secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release.
The isolation and initiating of the systems on Drywell Pressure-High supports action s to ensure that any offsite releases are within the limits calculated in the safety analysis (Ref. 2).
High drywell pressure signals are initiated from pressure switches that sense the pressure in the drywel l.
Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.
The Allowable Value was chosen to be the same as the RPS Drywell Pressure-High Function Allowable Value (LCO 3.3.1.1) since this i s indicative of a loss of coolant accident (LOCA).
The Drywell Pressure-High Function is required to be OPERABLE in MODES 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 MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
(continued)
B 3.3.6. 2-5 Revision
BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY ACTIONS Dresden 2 and 3 Secondary Containment Isolation Instrumentation B 3.3. 6. 2 3. 4.
Reactor Building Exhaust Radiation-High and Refueling Floor Radiation-Hi gh (continued) be OPERABLE during OPDRVsmovement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to f~el ~Rcovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded.
Due to radioactive decay, these Functions are only required to isolate secondary containment during fuel handling accidents involving handl i ng recently irradiated fuel Ci.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1. 3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels.
As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.
Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Functions that have channel components common to RPS instrumentation and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Functions that do not have channel components common to RPS instrumentation), has been shown to be acceptable (Refs. 3 and 4) to permit restoration of any inoperable channel to OPERABLE status.
This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.l Bases).
If the inoperable channel cannot be restored to OPERABLE status (continued)
B 3.3.6. 2-7 Revision
BASES LCO (continued)
APPLICABILITY Dresden 2 and 3 CREV System Isolation In strumentation B 3.3.7.1 CALIBRATIONS.
Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.
Trip setpoints are those predetermined value s of output at which an action should take place.
The setpoints are compared to the actual process parameter (e.g., reactor building ventilation exhaust radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,trip unit) changes state.
The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.
The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, ba sed on the trip setpoint values, by accounting for the calibration based errors.
These calibration based errors are limited to reference accuracy, instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components.
The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrument uncertainties, proces s effects, calibration tolerances, instrument drift, and severe environment errors (for channel s that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriatel y applied for the instrumentation.
The Reactor Building Ventilation System-High High Radiation Alarm Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of recently irradiated fuel assemblies in the secondary containment aRd operatioRs with a poteRtial for draiRiRg the reactor vessel (QPORVs), to ensure that control room personnel can be protected during a LOCA, or fuel handling event, or vessel draiRdowR eveRt.
During MODES 4 and 5, when these specified conditions are not in progress (e. g., OPORVs), the probability of a LOCA or fuel damage is low; thus, the Functions are not required.
Also due to radioactive decay, these Functions are only required to be OPERABLE during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
(continued)
B 3.3.7. 1-3 Revision
ECCS-Operating B 3.5.l B 3.5 EMERGENCY CORE COOLING SYSTEM& (ECCS), REACTOR PRESSURE VESSEL (RPVJ WATER INVENTORY CONTROL, AND ----Af>J..Q-ISOLATION CONDENSER (IC) SYSTEM B 3. 5. 1 ECCS-Operating BASES BACKGROUND Dresden 2 and 3 The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA).
The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA.
The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the Low Pressure Coolant Injection (LPCI) System, and the Automatic Depressurization System (ADS).
The suppression pool provides the required source of water for the ECCS.
Although no credit is taken in the safety analyses for the contaminated condensate storage tank (CCST), it is capable of providing a source of water for the HPCI, LPCI and CS systems.
On receipt of an initiation signal, ECCS pumps automatically start; the system aligns and the pumps inject water, taken either from the CCST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps.
Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed.
The HPCI pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core.
If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized.
If HPCI fails, it is backed up by ADS in combination with LPCI and CS.
In this event, the ADS timed sequence would be allowed to time out and open the relief valves and safety/relief valve (S/RV) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel.
If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.
Water from the break returns to the suppression pool where it is used again and again.
Water in the suppression pool is circulated through a heat exchanger cooled by the Containment Cooling Service Water System.
Depending on the (continued)
B 3. 5. 1-1 Revision
EGGS ShYtdo1<,*nRPV Water Inventory Control B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND ISOLATION CONDENSER (IC) SYSTEM B 3. 5. 2 EGGS ShYtdo*.mRPV Water Inventory Contra l BASES BACKGROUND APPLICABLE SAFETY ANALYSES Dresden 2 and 3 The RPV contains penetrations below the top of the active fuel (TAFJ that have the potential to drain the reactor cool ant 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. 1. 3 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.A description of the Core Spray (GS)
Systeff! and the Lo*,1 Pressyre Coolant Injection (LPG!) Systeffl is provided in the Bases for LGO 3. 5. 1, "EGGS Operatin§. "
With the unit in MODE 4 or 5, RPV water inventory control is The EGGS perforff!ance is evalblated for the entire spectrblffl of not required to mitigate any events or accidents evaluated in the safety analyses.
RPV water inventory contra 7 is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur.break sizes for a postbllated loss of coolant accident (LOCA).
A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems.
Instead, an event is considered 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, single human error).
The long terff! cooling analysis following a design basis LOCA (Ref. 1) deff!onstrates that only one low pressblre EGGS injection/spray Sblbsysteff! is reqYired, post LOCA, to fflaintain adeqYate reactor vessel
'.later level in the event of an inadvertent vessel drai ndo*.m.
- It is reasonable to assumed, based on engineering judgement, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually operated to maintain adequate reactor vessel water level.
To provide redYndancy, a ffli ni lflblffl of t"rn l o*,1 pressblre EGGS B 3. 5. 2-1 Revision
Dresden 2 and 3 EGGS ShbAtdownRPV Water Inventory Control B 3.5.2 injection/spray sblbsystems are reqblired to be OPERAB LE in MODES 4 and 5.
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).
The low pressblre EGGS sblbsystems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
(continued)
B 3.5. 2-2 Revision
BASES (continued)
EGGS Sh1,1tdownRPV Water Inventory Control B 3. 5. 2 LCO The RPV water level must be controlled in MODES 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 2.1.1. 3.
Dresden 2 and 3 The Limiting Condition for Operation (LCD) requires the DRAIN TIME of RPV water inventory to the TAF to be
36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
A DRAIN TIME of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate unexpected draining of reactor coolant.
An event that could cause 1 ass of RPV water inventory and resu 1 t in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1. 1. 3 and can be managed as part of normal plant operation.
+w&---One low pressure ECCS injection/spray subsystem~ -a-f'-e--is required to be OPERABLE and capable of being manually operated to provide defense-in-depth should an unexpected draining event occur.
~A low pressure ECCS injection/
spray subsyste~ consists of.tw.&-either one Core Spray (CS) subsyste ~ ~or one~ Low Pressure Coolant Injection
( LPCI J subsystem£.
~A CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or contaminated condensate storage tan ks ( CCSTs) to the reactor press bl re ves se 1 ( RPV+.
f..a..-A.-A LPCI subsystem consists of one motor driven pump, p1p1ng, and valves to transfer water from the suppression pool or the CCSTs to the RPV.
In MODES 4 and 5, OPERABLE CCSTs can be credited to support the OPERABILITY of the required ECCS subsystem.
A single LPCI pump is required per subsystem because of the similar injection capacity in relation to a CS subsystem.
In addition, in MOD ES 4 and 5, the LPCI System cross-tie valves are not required to be open.
Management of gas voids is important to ECCS injection/spray subsystem OPERABILITY.
(continued)
B 3.5. 2-3 Revision
BASES (continYed)
APPLICABILITY ACTIONS Dresden 2 and 3 EGGS ShYtdownRPV Water Inventory Control B 3. 5.2 RPV water inventory control is required in MODES 4 and 5.
Requirements on water inventory control in other MODES are contained in LCOs in Section 3.3, Instrumentation, and other LCOs in Section 3. 5, ECCS, RPV Water Inventory Control, and IC System.
RPV water inventory control is required to protect Safety Limit 2.1.1.3 which is applicable whenever irradiated fuel is in the reactor vessel.
OPERABILITY of the low pressYre EGGS injection/spray sYbsystems i s reqYired in MODES 4 and 5 to ensYre adeqYate coolant inventory and syfficient heat removal capability for the irradiated fyel in the core in case of an inadvertent drai ndm:n of the 11essel.
ReqYi rements for EGGS OPERABILITY dYring MODES 1, 2, and 3 are discYssed in the Applicability section of the Bases for LCO 3.5.1.
EGGS sybsystems are not reqYired to be OPERABLE dYring MODE 5 11ith the spent fyel storage pool gates removed and the 11ater l e 11el maintained at
> 23 ft above the RPV flange.
This provides syfficient coolant inventory to allow operator action to terminate the inventory loss prior to fyel Yncovery in case of an inadvertent draindown.
The AYtomatic DepressYrization System is not reqYired to be OPERABLE dYring MODES 4 and 5 becaYse the RPV pressYre is
< 150 psig, and the CS System and the LPCI sYbsystems can provide core cooling 11ith0Yt any depressYrization of the primary system.
The ~igh PressYre Coolant Injection System is not reqYired to be OPERABLE dYring MODES 4 and 5 since the lo"' pressyre EGGS injection/spray SYbsystems can provide syfficient flo',1 to the vessel.
A.1 and B. 1 If any onethe required low pressure ECCS injection/spray subsystem is inoperable, the inoperable sybsystemit must be restored to OPERABLE status with in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
In this Condition, the LCD 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 remaining OPERABLE sYbsystem can provide syfficient vessel flooding capability to recover from an inadvertent 11essel drai ndo',:n.
~0'11e
11er, overall system reliability is redYced becaYse a single failYre in the remaining OPERABLE sYbsystem concYrrent with B 3.5.2-4 Revision
Dresden 2 and 3 EGGS Shbltdo*.mRPV Water Inventory Control B 3.5.2 a vessel drai ndo1m cobll d resbll t in the EGGS not being able to perform its intended fblnction.
The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for restoring the required low pressure ECCS injection/spray subsystem to OPERABLE status is based on engineering judgment that considered considers the LCD controls on DRAIN TIMEthe remaining available sblbsystem and the low probability of an unexpected draining vessel draindown event that would result in loss of RPV water inventory.
(continbled)
B 3.5.2 -5 Revision
ACTIOflJS Dresden 2 and 3 EGGS Shbltdo*.mRPV Water Inventory Control B 3.5.2 A.l and B.l (continbled)
.w..i-t-A--If the inoperable ECCS injection/spray subsystem is not restored to OP ERABLE status with in the required Completion Time, action must be initiated immediately initiated 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 operated 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 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
If recirculation of injected water would occur, it may be credited in determining the necessary water volume. sblspend operations Hith a potential for draining the reactor vessel (QPDRVs) to minimize the probability of a vessel draindown and the sblbseqblent potential for fission prodblct release.
Actions mblst continble blntil OPDRVs are sblspended.
(continued) I B 3.5.2-6 Revision
BASES EGGS ShbltcfownRPV Water Inventory Control B 3.5. 2 ACTIONS C.1. C.2 and C.3 (continued)C.l, C.2, D. l, D.2, and D.3 Dresden 2 and 3 With the DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months but greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , compensatory measures should be taken to ensure the ability to implement mitigating actions should an unexpected draining event occur.
Should a draining event 1 ower the reactor cool ant 1 eve l 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.
The secondary containment provides a controlled volume in which fission products can be contained, diluted, and processed prior to release to the environment.
Required Action C.1 requires verification of the capability to establish the secondary containment boundary in less than the DRAIN TIME.
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 Standby Gas Treatment (SGT) subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment.With both of the reqblired EGGS injection/spray sblbsystems inoperable, all coolant inventory makeblp capability may be blnavailable.
Therefore, actions mblst immediately be initiated to sblspend OPDRVs to minimize the probability of a vessel draindmrn and the sblbseqblent potential for fission prodblct release.
Actions mblst continble blntil OPDRVs are Sblspended.
One reqblired EGGS injection/spray Sblbsystem mblst also be restored to OPERABLE statbls within 4 hoblrs.
The 4 hoblr Completion Time to restore at least one low pressblre EGGS injection/spray sblbsystem to OPERABLE statbls ensblres that prompt action will be taken to provide the reqblired coolins capacity or to initiate actions to place the plant in a condition that minimizes any potential fission prodblct release to the en~'i ronmen t.
I f at least one reqbl i red 10*.1 pressblre EGGS injection/spray sblbsystem is not restored to OP ERABLE statbls 1.*i thin the 4 hoblr Completion Time, additional actions are reqblired to minimize any potential fission prodblct release to the environment.
This inclbldes ensblrins secondary containment is OPERABLE; one standby sas treatment Sblbsystem is OPERABLE; and secondary containment isolation capability is B 3.5. 2-7 Revision
Dresden 2 and 3 EGGS Shbl t down RPV Water Inventory Control B 3.5. 2 available i n each associated penet ratio n fl o*..
path not Verification that the secondary containment boundary can be established must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
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.
A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place.
Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCIV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path.
Required Action C. 2 requires verification of the capab i lity to isolate each secondary containment penetration flow path in less than the DRAIN TIME.
The required verification confirms actions to isolate the 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 the secondary containment penetration flow paths can be isolated must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
The required verification is an administrative activity and does not require manipulation or testing of equipment.
(continued)
B 3.5. 2-8 Revision
BASES ACTIONS Dresden 2 and 3 EGGS ShYtdo11nRPV Water Inventory Control B 3. 5.2 C.1. C.2. and C. 3 (continued)
One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases.
Required Action C.3 requires verification of the capability to place one SGT subsystem in operation in less than the DRAIN TIME.
The required verification confirms actions to place a SGT subsystem in operation are preplanned and necessary materials are available.
Verification that a SGT subsystem can be placed in operation must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
The required verification is an administrative activity and does not require manipulation or testing of equipment.
isolated that is assYmed to be isolated to mitigate radioactivity releases (i.e., one secondary containment isolation valve and associated instrymentation are OPERABLE or other acceptable administrative controls to assyre isolation capability.
The administrative controls may consist of stationing a dedicated operator, who is in continYoYs commYnication with the control room, at the controls of the isolation device.
In this way the penetration can be rapidly isolated when a need for (continYed)
B 3.5. 2-9 Revision
Dresden 2 and 3
/\\CTI OM~
EGG~ ~hbltao'.mRPV Water Inventory Control B 3.5.2 CD. l
~D. 2 D
. +/-3
.~
and D
. ~1(continblea)
With the DRAIN TIME less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , mitigating actions are implemented in case an unexpected draining event should occur.
Note that if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action E. l is also applicable.
Required Action D. l requires immediate action to establish an additional method of water injection augmenting the ECCS injection/spray subsystem required by the LCD.
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 Note to Required Action D. l 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 operated and may consist of one or more systems or subsystems.
The additional method of wat er injection must be able to access water inventory capable of being injected to maintain the RPV water level above the TAF for~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
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.
(continued)
B 3.5.2-10 Revision
BASES ACTIONS Dresden 2 and 3 EGGS Shbttcfoi,mRPV Water Inventory Control B 3. 5. 2 D.1. D.2. D.3. and D. 4 (continued)
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.
The secondary containment provides a control volume into which fission products can be contained, diluted, and processed prior to release to the environment.
Required Action 0. 2 requires that actions be immediately initiated to establish the secondary containment boundary.
With the secondary containment boundary established, one SGT subsystem 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.
Required Action D.3 requires that actions be immediately initiated to verify that each secondary containment penetration flow path is isolated or to verify that it can be manually isolated from the control room.
A secondary containment penetration flow path can be considered isolated when one barrier in the flow path is in place.
Examples of suitable barriers include, but are not limited to, a closed secondary containment isolation valve (SCJV), a closed manual valve, a blind flange, or another sealing device that sufficiently seals the penetration flow path.
One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases.
Required Action D.4 requires that actions be immediately initiated to verify that at least one SGT subsystem is capable of being placed in operation.
The required verification is an administrative activity and does not require manipulation or testing of equipment.
secondary containffient is indicated).
OPERABILITY ffiay be verified by an adffiinistrative check, or by exaffiining logs or other inforffiation, to deterffiine whether the coffiponents are obit of service for ffiaintenance or other reasons.
It is not necessary to perforffi the Sbtrveillances needed to deffionstrate the OPERABILITY of the coffiponents.
If, however, any reqbtired coffiponent is inoperable, then it ffibtSt be restored to OPERABLE statbts.
In this case, the Sbtrveillance ffiay need B 3. 5. 2-11 Revision
Dresden 2 and 3 ECG~ ~hbltdm*mRPV Water Inventory Contra 1 B 3. 5. 2 to be performed to restore the component to OPERABLE statbls.
Actions mblst continble blntil all reqblired components are OPERABLE.
Ll If the Required Actions and associated Completion times of Conditions C or D are not met, or if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , actions must be initiated immediately to restore the DRAIN TIME to ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
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.
Note that Required Actions D.l, D.2, D.3, and D.4 are also applicable when DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
(continued)
B 3.5. 2-12 Revision
BASES (continued)
SURVEILLANCE REQUIREMENTS Dresden 2 and 3 SR 3.5.2.1 EGGS Shbltdoi,.rnRPV Water Inventory Control B 3.5. 2 This Surveillance verifies that the DRAIN TIME of RPV water inventory to the TAF i ~ ~ 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
The period of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 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 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.3 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.
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 (continued)
B 3.5.2-13 Revision
BASES SURVEILLANCE REQUIREMENTS Dresden 2 and 3 EGGS Shbitdo*,mRPV Water Inventory Control B 3.5. 2 SR 3.5.2.1 (continued) 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.
Surveillance Requirement 3.0.1 requires SRs to be met between performances.
Therefore, any changes in p 1 ant conditions that would change the DRAIN TIME requires that a new DRAIN TIME be determined.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR
3. 5.2. 2 The minimum water level of 10 ft 4 inches above the bottom of the suppression chamber required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for
-tfl.e-a required CS subsystem ~or LPCI subsys tem pump~.
recirculation volume, and vortex prevention.
With the suppression pool water level less than the required limit,
-a-l-l---the ECCS injection/spray subsystem~ ~is inoperable unless they areit is aligned to OPERABLE CCSTs.
When suppression pool level is< 10 ft 4 inches, the CS oriH+G LPCI subsystem~ ~is considered OPERABLE only if it~ can take suction from the CCSTs, and the CCSTs water volume is sufficient to provide the required NPSH and vortex prevention for the CS pump and LPCI pump.
Therefore, a verification that either the suppression pool water level is
~ 10 ft 4 inches or that the required low pressure ECCS injection/spray subsyste~ iH'-e-is aligned to take suction from the CCSTs and the CCSTs contain~ 140,000 available gallons of water, equivalent to 23 ft in both CCSTs with the CCSTs crosstied, ensures that the required low pressure ECCS injection /s pray (continued)
B 3.5. 2-14 Revision
BASES SURVEILLANCE REQUJ REMENTS Dresden 2 and 3 EGGS ShYtdm.,inRPV Water Inventory Control B 3.5. 2 SR 3.5.2. 2 (continued)
- subsyste m-& can supply at least 140,000 gallons of makeup water to the RPV.
The CS and LPCI suctions are uncovered at the 90,000 gallon level.
l=lo 1i1ever, as noted, only one reqYired lo 1i1 pressblre EGGS injection/spray sYbsystem may take credit for the GGST option dblring OPDRVs.
DYring OPDRVs, the volblme in the GGSTs may not provide adeqYate makeblp if the RPV were (continYed)
B 3.5. 2-15 Revision
SURVU LL/\\NGE REQU I REME~ITS Dresden 2 and 3 EGGS Sh1,1ttleHARPV Water Inventory Control B 3. 5. 2 SR 3.5.2.1 (ceAtiAbled) cempletel,y tlraiAetl.
Therefore, eAly eAe le*.1 press1,1re EGGS i Ajecti eA/spray s1,1bsystem is all e'n*etl te 1,1se the GGSTs.
This
eAsblres the ether req1,1ired EGGS s1,1bsystem has atleq1,1ate makeblp vel1,1me.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.2
. ~3. SR 3.5. 2.4. aAG SR 3.5.2.5 The required low pressure ECCS injection/spray flow path piping and components have the potential to develop voids and pockets of entrained gases.
Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS injection/spray subsystem and may also prevent water hammer and pump cavitation.
Selection of ECCS injection/spray subsystem locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations.
The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.
Maintaining the pump discharge lines of the required ECCS injection/spray subsystem sufficiently full of water ensures that the ECCS subsystem will perform properly.
Acceptance criteria are established for the volume of accumulated gas at susceptible locations.
If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criterion for gas volume at the suction or discharge of a pump), the Surveillance is not met.
If it is determined by subsequent evaluation that the ECCS injection/spray subsystem is not rendered inoperable by the accumulated gas (i. e., the system is sufficiently filled with water), the Surveillance may be declared met.
Accumulated gas should be eliminated or brought within the acceptance criteria limits.
B 3. 5. 2-16 Revision
Dresden 2 and 3 EGGS Shbttdo'llnRPV Water Inventory Control B 3. 5. 2 ECCS injection/spray subsystem locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative subset of susceptible locations.
Monitoring may not be pra cti cal for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety.
For these locations, alternative methods (e. g., operating parameters, remote monitoring) may be used to monitor the susceptible location.
Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not cha/ lenge system OPERABILITY.
The accuracy of the method used for monitoring the susceptible locations and trending of the results shou ld be sufficient to assure system OPERA BILITY during the Survei 7 lance interval.
The gases provided for SR 3.5.1.1, SR 3.5.1. 5, and SR 3. 5.1. 8 are appl icable to SR 3. 5. 2. 2, SR 3.5.2.4, and SR 3. 5. 2. 5, res pectively.
The Surve i 7 7 ance Frequency is contra 7 7 ed under the Surveillance Frequency Control Program.
The Surveillance Frequency may vary by locat ion suscept ible to ga s accumulation.
SR 3.5. 2.J-4 Verifying the correct alignment for manual, power operated, and automatic valves in the required ECCS subsystem flow path~ prov i des assurance that the proper flow path~ will exi st be available for ECCS operati on.
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 potentia l ly being mi spositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controll ed under the Surveillance Frequency Control Program.
The Surveillance is modified by (continued)
B 3.5. 2-17 Revision
BASES SURVEILLANCE REQUIREMENTS Dresden 2 and 3 EGGS Sh1:1tdo*,mRPV Water Inventory Control B 3. 5. 2 SR 3.5.2.4 (continued) a Note which exempts system vent flow paths opened under administrative control.
The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the control room.
This individual will have a method to rapidly close the system vent flow path if directed.
SR
3. 5.2. 5 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 is based on engineering judgement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.2.6 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.
The Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown these components usually pass the Surveillance when performed at the selected Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
B 3. 5. 2-18 Revision
BASES SURVEILLANCE REQUIREMENTS (continued)
REFERENCES Dresden 2 and 3 EGGS Shbltdo*,m RPV Water Inventory Control B 3.5. 2 SR 3.5.2.7 The required ECCS subsystem shall be capable of being manually operated.
This Surveillance verifies that the required CS or LPCI subsystem (including the associated pump and valve(s)) can be manually operated to provide additional RPV Water Inventory, if needed.
The Surveillance Frequency is controlled under the Survei 11 ance Frequency Contra I Program.
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 3141516," February 1983.UFS.A.R, Section 6. 3. 3.4.1.
B 3.5.2-19 Revision
IC System B 3. 5.3 B 3.5 EMERGENCY CORE COOLING SYSTEM& CECCS), REACTOR PRESSURE VESSEL (RPVJ WATER INVENTORY CONTROL, AND ISOLATION CONDENSER CIC) SYSTEM B 3. 5.3 IC System BASES BACKGROUND Dresden 2 and 3 The IC System is not part of the ECCS; however, the IC System is included with the ECCS section because of their similar functions.
The IC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation to provide adequate core cooling.
Under these conditions, the High Pressure Coo l ant Injection (HPCI) and IC systems perform similar functions.
The IC System (Ref.l) is a passive high pressure system comprised of one natural circulation heat exchanger, two AC motor-operated isolation valves, two D. C. motor-operated isolat i on valves, and two tube side high point vent isolation valves to main steam l i ne A".
The IC System functions as a heat sink for decay heat removal from the reactor vessel following reactor scram and isolation from the main conden ser.
This function prevents overheating of the reactor fuel, controls reactor pressure, and limits the loss of reactor coolant through the relief valves.
The IC System is automatically initiated by sustained reactor vessel high pres sure and, once activated, remains in operation until manually removed from service.
The isolation condenser shell contains two tube bundles.
When the IC System is in operation, both tube bundles are in service.
The IC System is designed to provide core cooling for reactor pressure
~ 150 psig.
The shell side of the condenser has a minimum water level of 6 feet which provides an inventory of~ 18,700 gallons.
This minimum level provides~ 11,300 gallons (approximately 3 feet) of water above the top of the tube bundles.
The shell side water temperature must be~ 210°F.
During normal plant operations, when the system is in standby, makeup is from the clean demineralized water storage tank.
Makeup during IC System operation can be provided from the Condensate (continued)
B 3. 5. 3-1 Revision
BASES BACKGROUND (continued)
APPLICABLE SAFETY ANALYSES LCO APPLICABILITY ACTIONS Dresden 2 and 3 IC System B 3.5. 3 Transfer System.
Since during operation of the IC System, water in the shell will boil, the condenser is vented to the atmosphere via one line.
The function of the IC System is to respond to main steam line isolation events by providing core cooling to the reactor.
The IC System is an Eng i neered Safety Feature System, an d credit is taken in the loss of feedwater transient analysis for IC System operation (Ref. 3).
Based on its contr i but i on to t he reductio n of overal l pl ant r i sk,
the system satisfies Criterion 4 of 10 CFR 50.36(c)( 2)(ii).
The OPERABILITY of the IC System provides adequate core cooling such that actuation of any of the low pressure ECCS subsystems is not required in the event of RPV isolation, loss of feedwater, or similar occurrence.
The IC System reduces the loss of RPV inventory during such events.
The IC System is required to be OPERABLE during MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig, since IC is the primary non-ECCS source for core cooling when the reactor is isolated and pressurized.
In MODES 2 and 3 with reactor steam dome pressure~ 150 psig, the low pressure ECCS injection/spray subsystems can provide sufficient core cooling.
....oo-+Jn MODES 4 and 5, IC is not required to be OPERABLE since core cooling is assured by LCD 3.5.2, "RPV Water Inventory Control." the 10\\\\' pressure EGGS injection/spray subsystems can provide sufficient core coolin9'.
A Note prohibits the application of LCO 3.0.4. b to an inoperable IC System.
There is an increased risk associated with entering a MODE or other spec i fied condit i on i n the Applicability with an inoperable IC System and the provisions of LCO 3.0.4. b, which allow entry into a MODE or other specif ied condit i on in the Applicab i lity with the LCO not met aft er performance of a r i sk assessment addressing inoperable systems and components, should not be applied in this circumstance.
(continued)
B 3.5.3-2 Revision
BASES LCO (continued)
APPLICABILITY ACTIONS Dresden 2 and 3 PC I Vs B 3.6.1.3 10 CFR 50 Appendix R requirements) to be de-activated and closed, are considered OPERABLE when the valves are de-activated and closed.
These passive isolation valves and devices are those listed in Reference 1.
MSIVs must meet additional leakage rate requirements.
Other PCIV leakage rates are addressed by LCO 3.6.1.1, "Primary Containment," as Type B or C testing.
This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents.
In MODES 1, 2, and 3, a OBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.
Therefore, !!!&5-t PCIVs are not required to be OPERABLE in MODES 4 and 5.
Certain valves, however, are required to be OPERABLE when the to prevent inadvertent reactor vessel drai ndo*,m.
These valves are those *..ihose associated instrumentation is required to be OPERABLE per LCO 3.3. ~5
. ~2
. "Primary Containment IsolationReactor Pressure Vessel (RPV) Water Inventory Control Instrumentation." (This does not include the valves that isolate the associated instrumentation.)
The ACTIONS are modified by a Note allowing penetration flow path(s) to be unisolated intermittently under administrative controls.
These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room.
In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry i s allowed for each penetration flow path.
This is acceptable, s ince the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV.
Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions.
(continued)
B 3.6.1.3-4 Revision
BASES ACTIONS (continued)
Dresden 2 and 3 PC I Vs B 3.6. 1.3 With the MSIV leakage rate CSR 3.6. 1. 3. 10) not within limit, the assumptions of the safety analysis may not be met.
Therefore, the leakage must be restored to within limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange.
When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device.
If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices.
The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allows a period of time to restore MSIV leakage rate to within limit given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown.
E.l and E. 2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
F. l and F. 2 If any Required Action and associated Completion Time cannot be met for PCIV(s) required OPERABLE in MODE 4 or 5, the unit must be 131 aced in a condition i ri *.ihi ch the LCO does not ap13ly.
Action must be immediately initiated to suspend 013erations *.iith a potential for drainin§ the reactor vessel COPDRVs) to minimize the probability of a vessel draindmm and subsequent potential for fission product release.
Actions must continue until OPDRVs are sus13ended.
If suspendin§ an OPDRV *,1ould result in closin§ the shutdo1m coolin§ isolation valves, an alternative Required Action is provided to immediately initiate action to restore the (continued)
B 3. 6. 1.3-9 Revision
BASES ACTIONS SURVEILLANCE REQUIREMENTS Dresden 2 and 3
~
. l and
~
. 2 (continued)
PCIVs B 3.6. 1.3
1al *1e( s) to OPERABU status.
This all o*.1s shutdo*.in cooling to remain i n service 11hile actions are being taken to restore the valve.
SR 3.6. 1.3.1 This SR ensures that the 18 inch primary containment vent and purge valves are closed as required or, if open, opened for an allowable reason.
If a vent or purge valve is opened in violation of this SR, the valve is considered inoperable.
The torus purge valve, 1601-56, is normally open for pressure control, therefore this valve is excluded from this SR.
However, this is acceptable since this valve is designed to automatically close on LOCA conditions.
The SR is modified by a Note stating that the SR is not required to be met when the vent or purge valves are open for the stated reasons.
The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided the drywell vent and purge valves and their as sociated suppre ss ion chamber vent and purge valves are not open simultaneously.
The 18 inch vent and purge valves are capable of closing in the environment following a LOCA.
Therefore, the se valve s are allowed to be open for limited periods of time.
The Surveillance Frequency i s controlled under the Surveillance Frequency Control Program.
SR 3.6. 1.3. 2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions, i s closed.
The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.
This SR does not require any testing or valve manipulation.
Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
B 3.6. 1.3-10 Revision
BASES (continued)
APPLICABLE SAFETY ANALYSES LCO APPLICABILITY ACTIONS Dresden 2 and 3 Suppression Pool Water Level B 3.6.2.2 Initial suppression pool water level affects suppression pool temperature response calculations, calculated drywell pressure during vent clearing for a OBA, calculated pool swell loads for a OBA LOCA, and calculated loads due to relief valve discharges.
Suppression pool water level must be maintained within the limits specified so that the safety analysis of Reference 1 remains valid.
Suppression pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).
A limit that suppression pool water level be~ 14 ft 6.5 inches and ~ 14 ft 10.5 inches above the bottom of the suppression chamber is required t o ensure that the primary containment conditions assumed for the safety analyses are met.
Either the high or low water level limits were used in the safety analyses, depending upon which is more conservative for a particular calculation.
In MODES 1, 2, and 3, a OBA would cause significant loads on the primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.
The requirements for maintaining suppression pool water level within limits in MODE 4 or 5 is addressed in LCO 3. 5.2,
" EGGS ShbltdownRPV Water Inventory Control."
With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met.
If water level is below the minimum level, the pressure suppression function still exists as long as the downcomers are covered, HPCI turbine exhaust is covered, and relief valve quenchers are covered.
If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and the capability of the Suppression Pool Spray System.
Therefore, continued operation for a limited time is allowed.
The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient to restore suppression pool water level to within limits.
Also, it takes into account the low probability of an event impacting the suppression pool water level occurring during this interval.
(continued)
B 3.6.2.2-2 Revision
BASES APPLICABLE SAFETY ANALYSES (continued)
LCO APPLICABILITY Dresden 2 and 3 Secondary Containment B 3.6.4.l associated leakage rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the SGT System prior to discharge to the environment.
Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment.
For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained, the hatches and blowout panels must be closed and sealed, the sealing mechanisms (e.g., welds, bellows, or 0-rings) associated with each secondary containment penetration must be OPERABLE (such that secondary containment leak tightness can be maintained), and all inner or all outer doors in each secondary containment access opening must be closed.
In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment.
Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.
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 operatioRs with a poteRtial for draiRiRg the reactor vessel (QPDRVsl or d~riRg movement of recently irradiated fuel assemblies in the secondary containment.
Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
(continued)
B 3.6.4. 1-2 Revision
BASES ACTIONS (continued)
SURV EILLANCE REQUIREMENTS Dresden 2 and 3 C.l anEI C.2 Secondary Containment B 3.6.4.1 Movement of recently i rradiated fuel assemblies in the secondary containment anEI OPDRVs can be postulated to cause significant fission product release to the secondary containment.
In such cases, the secondary containment is the only barrier to release of fission products to the environment.
Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperabl e.
Suspension of this activity shall not preclude completing an action that involves moving a component to a safe position.
Also, action ffi~st be iffiffieEliately initiateEI to s~spenEI OPDRVs to ffiiniffiize the probability of a vessel ElrainElm1n anEI s~bseq~ent potential for fission proEl~ct release.
Actions ffi~st contin~e ~ntil OPDRVs are s~spenEleEI.
Required Action C. l has been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recen tl y irradiated fuel assemblies while in MODE 4 or 5, LCO 3. 0.3 would not specify any action.
If moving recently irradiated fuel as semblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SR 3.6.4. 1.1 This SR ensures that the secondary containmen t boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
B 3. 6.4. 1-4 Revision
BASES APPLICABLE SAFETY ANALYSES (continued)
LCO APPLICABILITY Dresden 2 and 3 SCI Vs B 3.6.4.2 leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the environment.
Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.
SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
SCIVs form a part of the secondary containment boundary.
The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.
The power operated, automatic, i solation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic i solation signal.
The valves covered by this LCO, along with their associated stroke times, are listed in the Technical Requirements Manual (Ref. 2).
The normally closed manual SCIVs are considered OPERABLE when the valves are clo sed and blind flange s are in place, or open under admini strative control s.
These pa ssive i solation valves or devi ces are listed in Reference 2.
In MODES 1, 2, and 3, a OBA could lead to a fission product release to the primary containment that leaks to the secondary containment.
Therefore, the OPERABILITY of SCIVs is required.
In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES.
Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during e13erati ens iii th a 13etenti al fer draining the reacter vessel (QPQRVs) er d~ring movement of recently irradiated fuel assemblies in the secondary containment.
Due to radioactive decay, SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
(continued)
B 3.6.4. 2-2 Revision
BASES ACTIONS Dresden 2 and 3
.B....J.
(continued)
SCI Vs B 3. 6.4. 2 The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves.
This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations.
C.l and C. 2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.l and Q.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply.
If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended.
Suspension of this activity shall not preclude completion of movement of a component to a safe position.
Also, if applicable, actions must be immediately initiated to suspend OPQRVs in order to minimize the probability of a vessel draindo*,m and the subsequent potential for fission product release.
Actions must continue until OPQRVs are suspended.
Required Action D. l has been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3. 0. 3 would not specify any action.
If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently i rradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
(continued)
B 3.6.4. 2-5 Revision
BASES APP LI CABILITY (continued)
ACTIONS Dresden 2 and 3 SGT System B 3.6.4.3 In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.
Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during 013erations *,1ith a 13otential for drainin§ the reactor vessel (QPQRVs) or dYrin§ movement of recently irradiated fuel assemblies in the secondary containment.
Due to radioactive decay, the SGT System is only required to be OPERABL E during fuel handling involving handling recently irradiated fuel Ci.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days.
In this condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function.
However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed.
The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System and the low probability of a DBA occurring during this period.
If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the overall plant risk is minimized.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.
(continued)
B 3.6.4.3-3 Revision
BASES ACTIONS Dresden 2 and 3 (continued)
SGT System B 3. 6.4. 3 Required Action B.l is modified by a Note that prohibits the application of LCO 3. 0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
However, where al l owed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met.
The allowed Completion Time is reasonable,
based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.l.and C.2.1. and C.2.2 During movement of recently irradiated fuel assemblies, in the secondary containment or d1::1ring OPDRVs, when Required Action A.l cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation.
This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation will occur, and that any other failure would be readily detected.
An alternative to Required Action C. l i s to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk.
If applicable, movement of recently irradiated fuel assemblies must immediately be suspended.
Suspension of this activity must not preclude completion of movement of a component to a safe position.
Also, if applicable, actions m1::1st immediately be initiated to s1::1spend OPDRVs in order to minimize the probability of a vessel draindo 1i1n and s1::1bseq1::1ent potential for fission prod1::1ct release.
Actions m1::1st contin1::1e 1::1ntil OPDRVs are s1::1spended.
The Required Actions of Condition C have been modified by a Note stat ing that LCO 3. 0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, (continued)
B 3.6.4.3-4 Revision
BASES ACTIONS Dresden 2 and 3 C. l.
and C.2. l. and G.2.2 (continued)
SGT System B 3.6.4.3 LCO 3. 0.3 would not specify any action.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function.
Therefore, one SGT subsystem must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of supporting the required radioactivity release control function in MODES 1, 2, and 3.
This time period also ensures that the probability of an accident (requiring the SGT System) occurring during periods where the required radioactivity release control function may not be maintained is minimal.
Ll If one SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the overall plant risk is minimized.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short.
However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.
Required Action E. l is modified by a Note that prohibits the application of LCO 3.0.4.a.
This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met.
While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
(continued)
B 3.6.4. 3-5 Revision
BASES ACTIONS SURVEILLANCE REQUIREMENTS Dresden 2 and 3 l......l (continued)
SGT System B 3.6.4.3 However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4. b.
Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
F. l. aAd
~
. 2.
When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended.
Suspension of this activity shall not preclude completion of movement of a component to a safe position.
Also, if applicable, actioA must immediately be iAitiated to suspend OPDRVs iA order to miAimize the probability of a vessel draindown and subsequent potential for fission product release.
Actions must continue until OPDRVs are suspended.
Required Action F.l has been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3. 0.3 would not specify any action.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
SR
3. 6.4.3. 1 Operating (from the control room using the manual initiation switch) each SGT subsystem for ~ 15 continuous minutes ensures that both subsystems are OPERABLE and that all associated controls are functioning properly.
It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.
(continued)
B 3.6.4. 3-6 Revision
BASES LCO (continued)
APPLICABILITY ACTIONS Dresden 2 and 3 CREV System B 3.7.4 doors, hatches, floors plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area.
For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CRE.
This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicated.
In MODES 1, 2, and 3, the CREV System must be OPERABLE to ensure that the CRE will remain habitable during and following a OBA, since the OBA could lead to a fission product release.
In MODES 4 and 5, the probability and consequences of a OBA are reduced because of the pressure and temperature limitations in these MODES.
Therefore, maintaining the CREV System OPERABLE is not required in MODE 4 or 5, except ~
the following sitYations Ynder which significant radioactive releases can ae postylated:
a.
Dduring movement of recently irradiated fuel assemblies in the secondary containment-;----a-R4.
a.
DYring operations 'n'ith a potential for draining the reactor vessel COPDRVs).
Due to radioactive decay, the CREV System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
With the CREV System inoperable for reasons other than an inoperable CRE boundary, in MODE 1, 2, or 3, the inoperable CREV System must be restored to OPERABLE status within 7 days.
The 7 day Completion Time is based on the low probability of a OBA occurring during this time period.
(continued)
B 3.7.4-7 Revision
BASES ACTIONS (continued)
Dresden 2 and 3 E. l. aRd t.2 CREV System B 3. 7.4 LCO 3. 0. 3 is not applicable while in MODE 4 or 5.
However, since recently irradiated fuel movement can occur in MODE 1, 2, or 3, the Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies.
The NOTE to the ACTIONS, "LCO 3. 0. 3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.
With the CREV System inoperable or with the CREV System inoperable due to an inoperable CRE boundary, during movement of recently irradiated fuel assemblies in the secondary containment or duriRg OPDRVs, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require the CREV System to be placed in the i solation / pressurization mode of operation.
This places the unit in a condition that minimizes the accident ri sk.
If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately.
Su spension of this activity shall not preclude completion of movement of a component to a safe position.
Also, if applicable, actioR ffiust be iRitiated iffiffiediately to suspeRd OPDRVs to ffiiRiffiize the probability of a vessel draindown and the subsequent potential for fission product release.
Actions ffiust continue until the OPDRVs are suspended.
(continued)
B 3.7.4-10 Revision
BASES APPLICABILITY (continued)
Control Room Emergency Ventilation AC System B 3. 7. 5 In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature l imitations in these MODES.
Therefore, maintaining the Control Room Emergency Ventilation AC System OPERABLE is not required in MODE 4 or 5, except for the foll o*.:i n§ si te1ati ens e1neer which si §ni fi cant raei oacti ve releases can 9e poste1latee:
<!1-7~~~~~~~~~d uring movement of recently irradiated fuel assembl i es in the secondary containment~.
ACTIONS Dresden 2 and 3
9.
Dt1ri n§ operati ens 11i th a potential for erai ni n§ the reactor vessel ( OPDRl/s).
Due to radioactive decay, the Control Room Emergency Ventilation AC System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
With the Control Room Emergency Ventilation AC System inoperable in MODE 1, 2, or 3, the system must be restored to OPERABLE status within 30 days.
The 30 day Completion Time is based on the low probability of an event occurring requiring control room emergency zone isolation and the availability of alternate nonsafety cooling methods.
In MODE 1, 2, or 3, if the inoperable Control Room Emergency Ventilation AC System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes overall plant risk.
To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) and because the time spen t in MODE 3 t o perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.
(continued)
B 3. 7. 5-3 Revision
BASES ACTIONS Dresden 2 and 3
..B........l (continued)
Control Room Emergency Ventilation AC System B 3.7.5 Required Action B.l is modified by a Note that prohibits the application of LCO 3. 0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4. a to enter MODE 3 from MODE 4 with the LCD not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3. 0.4.a.
However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4. b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
C.l and G.2 LCD 3.0.3 i s not applicable while in MODE 4 or 5.
However, since recently irradiated fuel movement can occur in MODE 1, 2, or 3, the Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Entering LCD 3.0. 3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies.
The Note to the ACTIONS, "LCO 3. 0.3 is not applicable," ensures that the actions for immediate suspen sion of recently irradiated fuel assembly movement are not postponed due to entry into LCD 3.0.3.
With the Control Room Emergency Ventilation AC System inoperable during movement of recently irradiated fuel assemblies in the secondary containment or d~rin9 OPDRVs,
action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room.
This places the unit in a condition that minimizes risk.
(continued)
B 3.7.5-4 Revision
BASES ACTIONS SURVEILLANCE REQUIREMENTS REFERENCES Dresden 2 and 3 C. l and C.2 Control Room Emergency Ventilation AC System B 3.7.5 (continued)
If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately.
Suspension of this activity shall not preclude completion of movement of a component to a safe position.
Also, if applicable, action mYst be initiated immediately to sYspend OPDRVs to minimize the probability of a vessel draindown and sYbseqYent potential for fission prodYct release.
Action mYst continYe Yntil the OPDRVs are sYspended.
SR
3. 7. 5.1 This SR verifies that the heat removal capability of the system is sufficient to remove the control room emergency zone heat load assumed in the safety analyses.
The SR consists of a combination of testing and calculation.
The Survei l la nce Frequency is control l ed under the Surveil l ance Frequency Control Program.
1.
UFSAR, Section 6.4.
2.
NEDC-32988-A, Revision 2, Technical Justification to Support Risk - Informed Modification to Selected Required End States for BWR Plants, December 2002.
B 3. 7.5-5 Revision
AC Sources-Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND APPLICABLE SAFETY ANALYSES Dresden 2 and 3 A description of the AC sources is provided in the Bases for LCD 3.8.1, "AC Sources-Operating. "
The OPERABILITY of the m1n1mum AC sources during MODES 4 and 5, and during movement of recently irradiated fuel assemblies in the secondary containment ensures that :
a.
The facility can be maintained in the shutdown or refueling condition for extended periods;
b.
Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and c.
Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as ;2-A-inadvertent draindo*,m of the vessel or a fuel handling accident involving handling recently irradiated fuel.
Due to radioactive decay, AC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )
In general, when the unit is shutdown the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents.
However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required.
The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5.
Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences.
These deviations from OBA analysis assumptions (continued)
B 3.8. 2-1 Revision
BASES LCO (continued)
Dresden 2 and 3 AC Sources-Shutdown B 3.8.2 Systems-Shutdown," ensures that all required loads are powered from offsite power.
An OPERABLE DG, associated with a Distribution System Essential Service System CESS) bus required OPERABLE by LCO 3.8.8, ensures that a diverse power source is available for providing electrical power support assuming a los s of the offsite circuit.
Together, OPERABILITY of the required offsite circuit and DG ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of po stulated events during shutdown (e.g., fuel handling accidents involving handling recently irradiated fuel -a-fl4 reactor vessel drai ndo11n ).
The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective ESS bus(es), and of accepting required loads during an accident.
Qualified offsite circuits are those that are described in the UFSAR and are part of the licensing basis for the unit.
The offsite circuit from the 138 kV or 345 kV switchyard consists of the incoming breakers and disconnects to the 22 or 32 reserve auxiliary transformer (RAT) (or 21 or 31 unit auxiliary transformer CUAT) on backfeed), associated 22 or 32 RAT (or 21 or 31 UAT on backfeed), and the respective circuit path including feeder breakers to 4160 VESS buses required by LCO 3.8.8.
Another qualified circuit is provided by the bus tie between the corresponding ESS buses of the two units.
The required DG must be capable of starting, accelerating to rated speed and voltage, connecting to its respective 4160 V ESS bus on detection of bus undervoltage, and accepting required loads.
This sequence must be accomplished within 13 seconds.
Each DG must also be capable of accepting required loads within the assumed loading sequence interval s, and must continue to operate until offsite power can be restored to the 4160 V ESS bu ses.
These capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions.
Additional DG capabilities must be demonstrated to meet required Surveillances.
Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY.
The necessary portions of the DG Cooling Water and Ultimate Heat Sink System capable of prov iding cooling to the required DG is also required.
(continued)
B 3.8. 2-3 Revision
BASES LCO (continued)
APPLICABILITY ACTIONS Dresden 2 and 3 AC Sources-Shutdown B 3.8.2 It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required divisions.
The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment to provide assurance that :
a.
Systems providin§ adequate coolant inventory makeupthat provide core cooling are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel ;
b.
Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) are available; c.
Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and
d.
Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.
AC power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.1.
LCO 3. 0.3 is not applicable while in MODE 4 or 5.
However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies.
The Note to the ACTIONS, "LCO 3. 0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.
(continued)
B 3.8.2 -4 Revision
BASES ACTIONS (continued)
Dresden 2 and 3 AC Sources-Shutdown B 3.8. 2 An offsite circuit is considered inoperable if it is not available to one required 4160 V ESS bus.
If two or more 4160 V ESS buses are required per LCO 3.8.8, one division with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS-,--- and movement of recently irradiated~
movement, and operations with a potential for draining the reactor vessel. By the allowance of the option to declare required features inoperable that are not powered from offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS.
Required features remaining powered from a qualified offsite circuit, even if that circuit is considered inoperable because it is not powering other required features, are not declared inoperable by this Required Action.
For example, if both Division 1 and 2 ESS buses are required OPERABLE by LCO 3.8.8, and only the Division 1 ESS buses are not capable of being powered from offsite power, then only the required features powered from Division 1 ESS buses are required to be declared inoperable.
A. 2. 1, A.2. 2. A.2.3. A 2 4 B. 1. B.2. and B.3 and B 4 With the required offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable per Required Action A. 1.
Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made.
With the required DG inoperable, the minimum required diversity of AC power sources is not available.
It is, therefore, required to suspend CORE ALTERATIONS~, and movement of recently irradiated fuel assemblies in the secondary containment, and activities that coYld resYlt in inadvertent drainin9 of the reactor vessel.
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.
These actions minimize the probability of the occurrence of postulated events.
It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.
(continued)
B 3.8.2-5 Revision
BASES ACTIONS SURVEILLANCE REQUIREMENTS Dresden 2 and 3 AC Sources-Shutdown B 3.8. 2 A.2.1. A.2.2. A.2.3. A. 2.4. B.l B.2 and B.3. aAd B.4 (continued)
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention.
The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.
Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization.
Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESS bus, ACTIONS for LCO 3.8.8 must be immediately entered.
Thi s Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized.
LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.
SR 3.8.2.1 SR 3.8.2. 1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3 to be applicable.
SR 3.8. 1.9 is not required to be met since only one offsite circuit is required to be OPERABLE.
SR 3.8. 1. 20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE.
SR 3.8. 1.21 is not required to be met because the opposite unit's DG is not required to be OPERABLE in MODES 4 and 5, and during movement of recently irradiated fuel assemblies in secondary containment.
Refer to the corresponding Bases for LCD 3.8.1 for a di scussion of each SR.
This SR is modified by two Notes.
The reason for Note 1 is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and to preclude de-energizing a required 4160 V ESS bus or disconnecting a required offsite circuit during performance of SRs.
With limited AC sources available, a single event (continued)
B 3.8. 2-6 Revision
BASES SURVEILLANCE REQUIREMENTS REFERENCES Dresden 2 and 3 SR 3.8.2.1 (continued)
AC Sources-Shutdown B 3.8.2 could compromise both the required circuit and the DG.
It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG and offsite circuit are required to be OPERABLE.
Note 2 sta t es that SRs 3.8. 1. 13 and 3.8.1. 19 are not required to be met when its associated ECCS subsystem(s) are not required to be OPERABLE.
These SRs demonstrate the DG response t o an ECCS initiation signal (either alone or in conjunction with a loss of offsite power signal).
This is consistent with the ECCS instrumentation requirements that do not require the ECCS initiation signals when the associated ECCS subsystem is not required to be OPERABLE.
per LGO 3.5. 2, "EGGS Sh1:1tdo*,,*n."
None.
B 3.8. 2-7 Revision
DC Sources-Shutdown B 3.8. 5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources-Shutdown BASES BACKGROUND APPLICABLE SAFETY ANALYSES Dresden 2 and 3 A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources-Operating. "
The initial conditions of Design Basis Accident and transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE.
The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation and during movement of recently irradiated fuel assemblies in the secondary containment.
The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.
The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment ensures that :
a.
The facility can be maintained in the shutdown or refueling condition for extended periods;
b.
Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and
c.
Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as e-A-inadvertent draindo*.m of the vessel or a fuel handling accident involving handling recently irradiated fuel.
Due to rad i oactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
(continued)
B 3.8.5-1 Revision
BASES LCO (continued)
APPLICABILITY Dresden 2 and 3 DC Sources-Shutdown B 3.8.5 bus; and b) the required 125 VDC subsystem consisting of one battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus - are required to be OPERABLE to support some of the required DC distribution su bsystems required OPERABLE by LCO 3.8.8, "Distribution Systems-Shutdown. " This requirement ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e. g., fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindo*.m ).
The associated alternate 125 VDC electrical power subsystem may be used to satisfy the requirements of the 125 voe subsystems.
The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that :
a.
Required features to provide adequate coolant inventory make~p core cooling are available for the irradiated f~el assemblies in the core in case of an inadvertent draindown of the reactor vessel ;
b.
Required features needed to mitigate a fuel handling accident involving handling recently irradiated fuel are available; c.
Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and
d.
Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.
Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irra diated fuel (i. e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.
(continued)
B 3.8. 5-3 Revision
BASES (continued)
ACTIONS Dresden 2 and 3 DC Sources-Shutdown B 3.8.5 LCO 3.0.3 is not applicable while in MODE 4 or 5.
However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies.
The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3.0.3.
A. l. A.2.1. A.2.2. and A.2.3 aRd A 2 4 By allowance of the option to declare required features inoperable with associated DC electrical power subsystem(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS.
However, in many instances, this option may involve undesired administrative efforts.
Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the secondary containment, aRd aRy activities that co~ld res~lt iR iRadverteRt draiRiRg of the reactor vessel ).
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.
These actions minimize the probability of the occurrence of postulated events.
It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention.
The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.
(continued)
B 3.8.5 -4 Revision
BASES APP LI CABLE SAFETY ANALYSES (continued)
LCO Dresden 2 and 3 Distribution Systems-Shutdown B 3.8.8 The OPERABILITY of the m1n1mum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5, and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:
a.
The facility can be maintained in the shutdown or refueling condition for extended periods;
b.
Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and c.
Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent drai ndo1m of the vessel or a fuel handling accident involving handling recently irradiated fuel.
Due to radioactive decay, AC and DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).
The AC and DC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition.
Implicit in those requirements is the required OPERABILITY of necessary support features.
This LCO explicitly requires energization of the portions of the electrical distribution system, including the opposite unit Division 2 electrical distribution subsystem, necessary to support OPERABILITY of Technical Specifications required systems, equipment, and components - both specifically addressed by their own LCO, and implicitly required by the definition of OPERABILITY.
Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e. g.,
fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel drai ndmm ).
(continued)
B 3.8.8-2 Revision
BASES (continued)
APPLICABILITY ACTIONS Dresden 2 and 3 Distribution Systems-Shutdown B 3.8. 8 The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that :
a.
Systems to provide adeqyate coolant inventory makeYpthat provide core cooling are available for the irradiated fyel in the core in case of an inadvertent draindown of the reactor vessel ;
b.
Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) are available;
c.
Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and
d.
Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.
The AC and DC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8. 7.
LCO 3.0.3 is not applicable while in MODE 4 or 5.
However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.
If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Entering LCO 3. 0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of recently irradiated fuel assemblies.
The Note to the ACTIONS, "LCO 3.0.3 is not applicable," ensures that the actions for immediate suspension of recently irradiated fuel assembly movement are not postponed due to entry into LCO 3. 0. 3.
(continued)
B 3.8.8-3 Revision
BASES ACTIONS (continued)
Dresden 2 and 3 Di stribution Systems-Shutdown B 3.8.8 A. l. A.2.1. A.2. 2. A. 2. 3. and A. 2.4 and A.2.5 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS,---- and the movement of recently irradiated fuel movement, and operations with a potential for draining the reactor vessel. By allowing the option to declare required features associa t ed with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions.
In many instances this option may involve undesired administrative efforts.
Therefore, the allowance for sufficiently conservative actions is made, Ci.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the secondary containment, and any activities that co~ld result in inadvertent draining of the reactor vessel ).
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.
These actions minimize the probability of the occurrence of postulated events.
It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.
Notwithstanding performance of the above conservative Required Actions, a required shutdown cooling (SOC) subsystem may be inoperable.
In this case, Required Actions A. 2. 1 through A.2.4-3 do not adequately address the concerns relating to coolant circulation and heat removal.
Pursuant to LCO 3.0.6, the SOC ACTIONS would not be entered.
Therefore, Required Action A
. 2. ~ is provided to direct declaring SOC inoperable, which results in taking the appropriate SOC ACTIONS.
The Completion Time of immediately is consistent with the required t imes for act i ons requiring prompt attention.
The restoration of the requ i red di stribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.
(continued)
B 3.8.8-4 Revision

v t e Letter Sequence Request
CAC:MF9295, (Approved, Closed)
Initiation Request, Request Acceptance Supplement, Supplement, Supplement Administration Questions 11 July 2017 Answers Results Approval Other:
MONTHYEAR2017-03-16 [Table View]

RS-17-029, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

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3.8.5 APPLICABILITY

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RS-17-029, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control

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Subject:

Subject:

1.0 DESCRIPTION

2.0 ASSESSMENT

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4.0 ENVIRONMENTAL CONSIDERATION

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

3.7.4 APPLICABILITY

3.8.5 APPLICABILITY

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