ML19276G031
| ML19276G031 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 11/12/2019 |
| From: | Victor Cusumano NRC/NRR/DSS/STSB |
| To: | Technical Specifications Task Force |
| Honcharik M, NRR/DSS, 301-415-1774 | |
| Shared Package | |
| ML19262E115 | List: |
| References | |
| EPID L-2017-PMP-0024 | |
| Download: ML19276G031 (13) | |
Text
Enclosure 2 General Directions: This Model SE provides the format and content to be used when preparing 1
the plant-specific SE of an LAR to adopt traveler TSTF-568, Revision 2. The bolded bracketed 2
information shows text that should be filled in for the specific amendment; individual licensees 3
would furnish site-specific nomenclature or values for these bracketed items. The italicized 4
wording provides guidance on what should be included in each section. The italicized wording 5
should not be included in the SE.
6 7
DRAFT MODEL SAFETY EVALUATION 8
BY THE OFFICE OF NUCLEAR REACTOR REGULATION 9
TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 10 TSTF-568, REVISION 2 11 REVISE APPLICABILITY OF BWR/4 TS 3.6.2.5 AND TS 3.6.3.2 12 USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 13 (EPID L-2017-PMP-0024) 14 15
1.0 INTRODUCTION
16 17 By application dated [enter date] (Agencywide Documents Access and Management System 18 (ADAMS) Accession No. [MLXXXXXXXXX]), [as supplemented by letters dated [enter 19 date(s))), [name of licensee] (the licensee) submitted a license amendment request (LAR) for 20
[name of facility or facilities (abbreviated name(s)), applicable unit(s)].
21 22 The proposed changes would revise [TS 3.6.2.5, Drywell-to-Suppression Chamber 23 Differential Pressure, and] TS 3.6.3.2, Primary Containment Oxygen Concentration. The 24 proposed changes simplify and clarify the applicability statements, which if misapplied, could 25 conflict with the corresponding required actions. The proposed changes also remove the 26 undefined term "scheduled plant shutdown" and provide adequate terminal actions.
27 28 The proposed amendment is based on Technical Specifications Task Force (TSTF) traveler 29 TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2 (ADAMS 30 Accession No. ML19141A122). The U.S. Nuclear Regulatory Commission (NRC or the 31 Commission) approved TSTF-568, Revision 2, by letter dated [DATE] (ADAMS Accession 32 No. [ML19XXXXXXX]). The NRC staffs safety evaluation (SE) of the traveler was enclosed 33 with the NRC staffs approval letter.
34 35
((The licensee has proposed variations from the TS changes described in traveler 36 TSTF-568, Revision 2. The variations are described in Section [2.2.3] of this SE and 37 evaluated in Section [3.3].] OR [The licensee is not proposing any variations from the TS 38 changes described in TSTF-568, Revision 2, or the applicable parts of the NRC staffs SE 39 of TSTF-568, Revision 2.))
40 41
[The supplemental letter(s) dated [enter date(s)], provided additional information that 42 clarified the application, did not expand the scope of the application as originally 43
noticed, and did not change the NRC staffs original proposed no significant hazards 1
consideration determination as published in the Federal Register on [enter date] (cite FR 2
reference).]
3 4
2.0 REGULATORY EVALUATION
5 6
2.1 Description of Structures, Systems, Components and TS Sections 7
8
{NOTE: Section 2.1.1 is only applicable for plants with Mark I containments.}
9 10 2.1.1 Current Drywell-to-Suppression Chamber Differential Pressure Control 11 12 The drywell-to-suppression chamber differential pressure control is a safety-related operational 13 feature of Mark I containment designs. The TS 3.6.2.5 requires a minimum differential pressure 14 of [1.5] pounds per square inch differential (psid) to reduce the loss-of-coolant accident (LOCA) 15 hydrodynamic loads during the Mark I containment load definition short-and long-term 16 programs. The LOCA pool swell loads are significantly reduced because the differential 17 pressure control reduces the length of water leg in the downcomer. The LOCA vent clearing 18 and pool swell due to bubble formation would occur earlier (i.e., at a lower drywell pressure 19 resulting in lesser forces on the suppression chamber thereby increasing the safety margin for 20 containment integrity, containment internal structures, and pressure boundary). Decreasing the 21 allowable suppression chamber water level has a similar effect.
22 23 It is difficult to control the differential pressure during startup and shutdown transients. This is 24 because of the variation of the drywell heat loads from the primary and auxiliary systems and 25 because the inerting (during startup) or the de-inerting (during shutdown) of containment.
26 Inerting the containment during startup involves the addition of large volumes of nitrogen.
27 De-inerting containment during shutdown involves the addition of large volumes of air. In order 28 to allow operation during the time differential pressure control is difficult, the current TS 3.6.2.5 29 is applicable from [24] hours following startup after the reactor thermal power exceeds 30
[15] percent to [24] hours prior to reducing thermal power less than [15] percent reactor thermal 31 power (RTP) during a scheduled shutdown.
32 33 2.1.[2] Current Containment Oxygen Concentration Requirement 34 35 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, 36 Combustible gas control for nuclear power reactors, states that for a plant with an inerted 37 containment atmosphere, the oxygen concentration in the primary containment is required to be 38 maintained below 4 percent by volume during normal plant operation. This requirement ensures 39 that an accident that produces hydrogen does not result in a combustible mixture inside the 40 primary containment. The current TS 3.6.3.2 requires primary containment oxygen 41 concentration to be less than 4 percent by volume when in Mode 1 during the period from 42
[24] hours after the thermal power exceeds [15] percent RTP following startup, and to 43
[24] hours prior to reducing the RTP to less than [15] percent RTP during next scheduled 44 shutdown. TSTF-568, Revision 2, stated that the [24]-hour allowance above [15] percent RTP 45 is provided in the primary containment oxygen concentration specification to delay inerting the 46 primary containment in a plant startup and to accelerate de-inerting for a plant shutdown. This 47 allowance is provided so that plant personnel can safely enter the primary containment without 48 breathing apparatus to perform the needed inspections and maintenance adjustments.
49 50
{NOTE: Use this paragraph for Mark I containments.}
1 2
The containment consists of a drywell (in the shape of an inverted light bulb), a suppression 3
chamber (in the shape of a toroid), and a network of vents which extend radially outward and 4
downward from the drywell to the suppression chamber. The containment atmosphere is 5
inerted with nitrogen gas during normal operation to prevent a combustible mixture of hydrogen 6
and oxygen from forming during accident conditions. Long-term control of post-LOCA hydrogen 7
gas concentration is accomplished by adding additional nitrogen gas and then venting the 8
primary containment through the standby gas treatment system.
9 10
{NOTE: Use this paragraph for Mark II containments.}
11 12 The containment consists of a drywell (in the shape of a truncated cone), a suppression 13 chamber directly below the drywell (in the shape of a right circular cylinder), and a network of 14 vertical vents extending downward from the drywell to the suppression chamber. The 15 containment atmosphere is inerted with nitrogen gas during normal operation to prevent a 16 combustible mixture of hydrogen and oxygen from forming during accident conditions.
17 Long-term control of post-LOCA hydrogen gas concentration is accomplished by adding 18 additional nitrogen gas and then venting the primary containment through the standby gas 19 treatment system.
20 21 2.1.[3] Pressure Suppression Following a LOCA 22 23 The drywell is immediately pressurized when a postulated line break occurs within the primary 24 containment. As drywell pressure increases, drywell atmosphere (primarily nitrogen gas) and 25 steam are blown down through the vents into the suppression pool via the downcomers. The 26 steam condenses in the suppression pool which suppresses the peak pressure in the drywell.
27 Non-condensable gases discharged into the suppression pool collect in the free air volume of 28 the suppression chamber, increasing the suppression chamber pressure. As steam is 29 condensed in the suppression pool and on the structures in the drywell, the pressure decreases 30 until the suppression chamber pressure exceeds the drywell pressure and the suppression 31 chamber-drywell vacuum breakers open and vent non-condensable gases back into the drywell.
32 33
{NOTE: Section 2.1.4 is only applicable for plants with Mark I containments.}
34 35 2.1.[4] TS 3.6.2.5, Drywell-to-Suppression Chamber Differential Pressure 36 37 A drywell-to-suppression chamber differential pressure limit is required to ensure the 38 containment conditions assumed in the safety analyses are met. Failure to maintain the 39 required differential pressure could result in excessive forces on the suppression chamber due 40 to higher water clearing loads from downcomer vents and higher-pressure buildup in the drywell 41 during a LOCA. Drywell-to-suppression chamber differential pressure must be controlled when 42 the primary containment is inert. The TS requires that the drywell pressure be maintained 43
[1.5] psid above the pressure of the suppression chamber.
44 45 2.1.[5] TS 3.6.3.2, Primary Containment Oxygen Concentration 46 47 The primary containment oxygen concentration is maintained to ensure that a LOCA, a 48 postulated event that produces hydrogen, does not result in a combustible mixture inside 49 primary containment. The TS requires that the primary containment oxygen concentration be 50
maintained below 4 volume percent. Below this concentration, the primary containment is 1
inerted and no combustion can occur.
2 3
2.2 Description of Proposed Technical Specification Changes 4
5
{NOTE: Section 2.2.1 is only applicable for plants with Mark I containments.}
6 7
2.2.1 Proposed Changes to TS 3.6.2.5, Drywell-to-Suppression Chamber 8
Differential Pressure 9
10 The Applicability of TS 3.6.2.5, Drywell-to-Suppression Chamber Differential Pressure, is 11 revised as shown below.
12 13 Current TS Applicability Proposed TS Applicability MODE 1 during the time period:
- a. From [24] hours after THERMAL POWER is > [15]% RTP following startup, to
- b. [24] hours prior to reducing THERMAL POWER to < [15]% RTP prior to the next scheduled reactor shutdown.
MODE 1 with THERMAL POWER > [15]%
RTP.
14 Required Action A.1 and the completion time (CT) are revised as shown below.
15 16 17 18 The NRC staff understands the overall purpose of the proposed changes is to simplify the 19 applicability statement by adding a new note and revising the CT. This change provides similar 20 operational flexibility but more closely follows established TS conventions.
21 22
[2.2.2] Proposed Changes to TS 3.6.3.2, Primary Containment Oxygen 1
Concentration 2
3 The Applicability of TS 3.6.3.2, "Primary Containment Oxygen Concentration," would be revised 4
as shown below.
5 6
Current TS Applicability Proposed TS Applicability MODE 1 during the time period:
- c. From [24] hours after THERMAL POWER is > [15]% RTP following startup, to
- d. [24] hours prior to reducing THERMAL POWER to < [15]% RTP prior to the next scheduled reactor shutdown.
MODES 1 and 2.
7 Required Actions A.1 and B.1 and their associated CTs are revised as shown below.
8 9
10 11 The NRC staff understands the overall purpose of the proposed changes is to simplify the 12 applicability statement by adding a new note and revising the CT. This change provides 13 operational flexibility but more closely follows established TS conventions and requires that the 14 plant be in Mode 3 if oxygen concentration cannot be restored to within limits.
15 16
[2.2.3 Variations]
17 18
{Note: If the licensee identifies variations in the LAR, other than differences in the numbering of 19 the TS and nomenclature, they should be described in this section.}
20 21 2.3 Applicable Regulatory Requirements and Guidance 22 23 Section 50.90 of 10 CFR, Application for amendment of license, construction permit, or early 24 site permit, requires that whenever a licensee desires to amend the license, application for an 25 amendment must be filed with the Commission fully describing the changes desired, and 26 following as far as applicable, the form prescribed for original applications.
27
1 Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment 2
are to be guided by the considerations that govern the issuance of initial licenses or construction 3
permits to the extent applicable and appropriate. Both the common standards for licenses and 4
construction permits in 10 CFR 50.40(a), and those specifically for issuance of operating 5
licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the 6
activities at issue will not endanger the health and safety of the public.
7 8
The regulation, 10 CFR 50.36, Technical specifications, establishes the regulatory 9
requirements related to the content of TSs. Section 50.36(a)(1) requires an application for an 10 operating license to include proposed TSs. A summary statement of the bases or reasons for 11 such specifications, other than those covering administrative controls, must also be included in 12 the application, but shall not become part of the TSs.
13 14 The regulation, 10 CFR 50.36(b), requires:
15 16 Each license authorizing operation of a utilization facility will include 17 technical specifications. The technical specifications will be derived from the 18 analyses and evaluation included in the safety analysis report, and amendments 19 thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; 20 technical information]. The Commission may include such additional technical 21 specifications as the Commission finds appropriate.
22 23 The categories of items required to be in the TSs are listed in 10 CFR 50.36(c).
24 25 In accordance with 10 CFR 50.36(c)(2), limiting conditions for operation (LCOs) are the lowest 26 functional capability or performance levels of equipment required for safe operation of the 27 facility. When LCOs are not met, the licensee must shut down the reactor or follow any 28 remedial action permitted by the TSs until the condition can be met. In addition, 10 CFR 29 50.36(c)(2)(ii)(B) requires a TS LCO of a nuclear reactor must be established for a process 30 variable, design feature, or operating restriction that is an initial condition of a design basis 31 accident or transient analysis that either assumes the failure of or presents a challenge to the 32 integrity of a fission product barrier.
33 34 The regulation, 10 CFR 50.44(b)(2)(i), states that All boiling water reactors with Mark I or 35 Mark II type containments must have an inerted atmosphere. Section 50.44(a)(1) defines 36
[i]nerted atmosphere as a containment atmosphere with less than 4 percent of oxygen by 37 volume.
38 39
{NOTE: Use this paragraph for Mark I containments.}
40 41 Chapter 6.2.1.1.C, Revision 7, Pressure-Suppression Type BWR Containments of 42 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 43 Power Plants: LWR [Light-Water Reactor] Edition (SRP), March 2007 (ADAMS Accession 44 No. ML063600403) states: The acceptability of LOCA pool dynamic loads for plants with 45 Mark I containments is based on conformance with NRC acceptance criteria found in 46 NUREG-0661[, Safety Evaluation Report Mark I Containment Long-term Program Resolution of 47 Generic Technical Activity A-7, July 1980 (ADAMS Accession No. ML072710452).]
48 49 The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical 50 Specifications, of the SRP, March 2010 (ADAMS Accession No. ML100351425). As described 51
therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard 1
Technical Specifications (STSs) for each of the LWR nuclear designs. Accordingly, the NRC 2
staffs review includes consideration of whether the proposed changes are consistent with the 3
applicable reference STSs (i.e., the current STSs), as modified by NRC-approved travelers.
4 The STS applicable to [abbreviated name of facility] is NUREG-1433, Revision 4.0, Standard 5
Technical Specifications, General Electric Plants BWR/4, Volume 1, Specifications, and 6
Volume 2, Bases, April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193, 7
respectively).
8 9
3.0 TECHNICAL EVALUATION
10 11 The proposed amendments are based on the NRC-approved TSTF-568, Revision 2. The NRC 12 staff also considered the regulations and guidance discussed in Section 2.3 of this SE in its 13 review.
14 15
{NOTE: The changes to TS 3.6.2.5 discussed in Section 3.1 are only applicable to plants with 16 Mark I containments.}
17 18
3.1 PROPOSED CHANGE
S TO TS 3.6.2.5 19 20 3.1.1 Proposed Changes in the Applicability 21 22 The licensee proposed to delete the time periods, dependent on startup and shutdown times, 23 from the applicability section and to replace them with a thermal power value. These time 24 periods are a. From [24] hours after THERMAL POWER is > [15] percent RTP following 25 startup, to b. [24] hours prior to reducing THERMAL POWER to < [15] percent RTP prior to the 26 next scheduled reactor shutdown. These time periods would be replaced by flexibilities and 27 requirements in the revised completion times and the inserted note referencing LCO 3.0.4.c.
28 This would result in requiring the drywell pressure during Mode 1 to be maintained above the 29 specified limit whenever the thermal power is above [15] percent. The current limitations of 30 applicability, dependent on startup and shutdown, were established to allow licensees 31 operational flexibilities, such as containment entry to perform maintenance and surveillances 32 while at power.
33 34 In TSTF-568, Revision 2, Attachment, General Electric (GE) Safety Communication (SC) 02-10, 35 page 4, under the heading Corrective/Preventive Actions, item 2, it is recommended that 36 Mark I plants that use TS 3.6.2.5 should confirm that their containment is structurally designed 37 for pool swell loads with a zero drywell-to-suppression chamber differential pressure. For these 38 plants, the Mark I containment load definition program has defined the pool swell loads 39 associated with zero drywell-to-suppression chamber differential pressure. NUREG-0661, 40 Appendix A, Section 2.3, states that each plant with a differential pressure control (i.e.,
41 TS 3.6.2.5) perform a structural assessment to demonstrate that the containment can maintain 42 its functional capability when the differential pressure control is out-of-service (i.e., the 43 differential pressure is zero).
44 45
[Browns Ferry, Units 1, 2, and 3/Dresden Units 2 and 3/Quad Cities, Units 1 46 and 2/FitzPatrick is/are] applying the drywell-to-suppression chamber differential pressure 47 control TS 3.6.2.5. The licensees plant-specific analysis report called PUAR [Plant Unique 48
Analysis Report] was approved by the NRC.1 As stated in SC02-10, page 3, structural 1
assessment based on zero drywell-to-suppression chamber differential pressure pool swell load 2
definition was used to confirm the functional capability of the suppression chamber against the 3
Service Level D limit. The SC02-10 also identifies the following two major conservatisms in the 4
pool swell load definitions based on the Mark I Quarter Scale tests:
5 6
(a) The drywell pressurization test transient was based on the predicted drywell pressure 7
from the NRC approved conservative GE code M3CPT. This code predicts about 8
50 percent higher drywell pressurization than a realistic analysis using the GE-Hitachi 9
code TRACG.
10 11 (b) The break was simulated by air to pressurize the drywell, which produces a more severe 12 pool swell response than a realistic nitrogen/steam mixture and enhances the bubble 13 growth.
14 15 The NRC approval confirmed that the licensee met the acceptance criteria specified in 16 NUREG-0661, Appendix A, and reviewed and approved any exceptions the licensee took from 17 the acceptance criteria. Therefore, the NRC staff approval of the PUARs confirmed that with 18 the drywell-to-suppression chamber differential pressure out-of-service, the containment is 19 structurally designed for the pool swell loads during a large-break LOCA.
20 21 Based on the PUARs, the NRC staff finds it acceptable for the reactor to not be depressurized 22 when the differential pressure is out-of-service at [15] percent RTP. Further NUREG-0661, 23 Section 3.12.7, concluded that if the differential pressure is out-of-service, the probability of 24 occurrence of a large-break LOCA, is less than 10E-7 per reactor-year, which is sufficiently 25 small. This minimal probability of occurrence paired with the short period during which plants 26 are in the transition state of less than [15] percent rated thermal power, support the adequacy of 27 this change because the LOCA dynamic loads are not adversely affected. The NRC staff 28 determined the proposed deletion of the time periods is acceptable because they are now 29 included in the note insertion (discussed in Section 3.1.2 of this SE) and change in the CT 30 (discussed in Section 3.1.3 of this SE). In addition, the proposed change is acceptable since it 31 simplifies and clarifies the applicability statement and continues to provide the lowest functional 32 capability of equipment required for safe operation of the facility as required by 10 CFR 33 50.36(c)(2) by protecting containment integrity.
34 35 3.1.2 Proposed Changes in Required Action A.1 36 37 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to add the 38 following note to Required Action A.1: LCO 3.0.4.c is applicable. LCO 3.0.4 states:
39 40 When an LCO is not met, entry into a MODE or other specified condition in the 41 Applicability shall only be made:
42 43
- a. When the associated ACTIONS to be entered permit continued operation in 44 the MODE or other specified condition in the Applicability for an unlimited 45 period of time; 46 47 1 Cut and paste appropriate references here: Browns Ferry use Reference 1; Dresden use References 2, 3, and 4; Quad Cities use Reference 5; FitzPatrick use Reference 6.
- b. After performance of a risk assessment addressing inoperable systems and 1
components, consideration of the results, determination of the acceptability of 2
entering the MODE or other specified condition in the Applicability, and 3
establishment of risk management actions, if appropriate; exceptions to this 4
Specification are stated in the individual Specifications, or 5
6
- c. When an allowance is stated in the individual value, parameter, or other 7
Specification.
8 9
This Specification shall not prevent changes in MODES or other specified 10 conditions in the Applicability that are required to comply with ACTIONS or that 11 are part of a shutdown of the unit.
12 13 The criteria applicable to TS LCO 3.6.2.5 is LCO 3.0.4.c since this LCO establishes an 14 individual value or parameter (i.e., drywell pressure maintained above a certain value). The 15 new note will allow entry into the mode of applicability of TS LCO 3.6.2.5 with the drywell 16 pressure outside of the required limit. This note allows the licensee operational flexibility as it 17 permits entry into Mode 1 at greater than [15] percent RTP when drywell pressure is outside of 18 the required limit during startup configurations. The NRC staff concludes that the addition of the 19 note is acceptable because it clarifies and simplifies the intent of the current TS LCO 3.6.2.5 20 applicability statement a. of allowing startup operation with the LCO not met.
21 22 3.1.3 Proposed Changes in the CT of Condition A 23 24 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to change 25 the CT for Required Action A.1 from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. TSTF-568, Revision 2, stated the 26 proposed change will permit safe entry of personnel into the containment in Modes 1 and 2.
27 The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provides: [24] hours to de-inert the containment to permit safe personnel access, 28
[24] hours to perform the required work, and [24] hours to re-inert containment. The NRC staff 29 finds that the extended CT incorporates the time currently allowed through the applicability 30 statement in Section 3.1.1 of this SE. The NRC staff finds that 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable to 31 conduct these activities based on operating experience and the requested completion time does 32 not present a significant change in risk given the low probability that a large line break would 33 occur during this period. Therefore, NRC staff finds this change acceptable.
34 35 3.1.4 Conclusion for Proposed Changes to TS 3.6.2.5 36 37 The NRC staff finds the changes proposed in TS 3.6.2.5 acceptable and continue to meet 38 10 CFR 50.36(c)(2) since the revised LCO provides the lowest functional capability of 39 equipment required for safe operation of the facility by protecting containment integrity.
40 41
3.2 PROPOSED CHANGE
S TO TS 3.6.3.2 42 43 3.2.1 Proposed Changes in the Applicability 44 45 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to expand 46 the applicability of this LCO to Modes 1 and 2 without exception. The NRC staff finds the 47 proposed change acceptable because it is more restrictive since an unlikely LOCA event 48 leading to a degraded core that could produce hydrogen has the highest probability of 49 occurrence during Modes 1 and 2 conditions.
50 51
3.2.2 Proposed Changes in Required Action A.1 1
2 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to add the 3
following note to Required Action A.1: LCO 3.0.4.c is applicable. As stated in Section 3.1.2 of 4
this safety evaluation, TS LCO 3.0.4.c allows entering the mode of applicability of 5
TS LCO 3.6.3.2 with the LCO not met. Therefore, the proposed change would permit entry into 6
Modes 1 and 2 with primary containment oxygen concentration higher than the required limit.
7 The NRC staff concludes the addition of the note is acceptable because it clarifies and simplifies 8
the intent of the current TS LCO 3.6.3.2 applicability statement a. of allowing startup operation 9
with the LCO not met.
10 11 3.2.3 Proposed Changes in the CT of Condition A 12 13 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed changing 14 the CT from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 based on the following sequence of operations: allow [24] hours to 15 de-inert the containment to permit safe personnel access, allow [24] hours to perform the 16 required maintenance or repair work, and allow [24] hours to inert the containment. The NRC 17 staff determined that the presence of a higher oxygen concentration for the 72-hour CT is 18 appropriate considering the low safety significance of the change for potential accidents and the 19 additional restrictions and conservatisms in the revised applicability.
20 21 3.2.4 Proposed Changes in Required Action B.1 22 23 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to change 24 the applicability statement of TS LCO 3.6.3.2 to Modes 1 and 2. If the oxygen concentration 25 cannot be restored within the required limit and CT of Required Action A.1, the reactor should 26 be brought to Mode 3. In this mode, the reactor would be in a hot shutdown condition (control 27 rods fully inserted) with all reactor vessel head bolts fully tensioned.
28 29 The NRC staff recognizes that on entering Mode 3, the decay heat is rapidly decreasing. Steam 30 is initially available for operating the reactor core isolation cooling/high pressure coolant 31 injection steam turbine-driven pumps until the reactor pressure and thus water temperature is 32 substantially reduced. As the decay heat continues to decrease, operators have increased time 33 and options for achieving adequate water injection using the low-pressure emergency core 34 cooling system to avoid core damage and associated generation of combustible gas.
35 Therefore, the occurrence of a LOCA leading to degraded core is highly unlikely in Mode 3.
36 37 The NRC staff finds the proposed change in Required Action B.1 acceptable because it 38 provides a more appropriate terminal action since it requires the plant to be placed in a mode in 39 which the LCO does not apply and the oxygen concentration limit is no longer required. The 40 previous terminal action allowed an indefinite period of operation at [15] percent RTP.
41 42 Due to the low potential for hydrogen generation when the reactor is in Mode 3, inerting of 43 containment in Mode 3 is not needed. Therefore, the NRC staff concluded the proposed change 44 is acceptable because it continues to protect containment integrity and meets 10 CFR 45 50.36(c)(2) by providing the lowest functional capability of equipment required for safe operation 46 of the plant.
47 48
3.2.5 Proposed Changes in the CT of Condition B 1
2 In accordance with approved traveler TSTF-568, Revision 2, the licensee proposed to change 3
the Condition B CT from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, stating that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable time to 4
reduce reactor power from full power conditions to Mode 3 in an orderly manner and without 5
challenging plant systems. The proposed change from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for bringing the 6
reactor to a hot shutdown condition from full power is acceptable to NRC staff because it is not 7
a significant change and is based on industry operating experience.
8 9
3.2.6 Conclusion for Proposed Changes to TS 3.6.3.2 10 11 The NRC staff concludes the proposed changes in the applicability statement for TS 3.6.3.2 are 12 acceptable since they are more restrictive as the applicability now extends to Modes 1 and 2 13 without exception. In addition, the occurrence of a LOCA that could lead to degraded core 14 conditions with containment de-inerted, while in Mode 3, is unlikely. Therefore, the changes 15 proposed in TS 3.6.3.2 are acceptable and continue to meet 10 CFR 50.36(c)(2).
16 17
{Note: If the plants TS differs from the standard TS, include the bracketed language below, 18 replacing the discussion of the differences as necessary. Differences should be identified by 19 the licensee in the LAR. For differences beyond TS numbering and nomenclature, justification 20 should be provided by the licensee and a more thorough evaluation of the applicability of 21 TSTF-568 should be included in Section 3.3. More extensive differences may be considered 22 exceptions to the approved traveler (i.e., may exceed the scope of what is allowable in CLIIP 23 applications).}
24 25 3.3 ADDITIONAL CHANGES 26 27
[The licensee identified differences between the TSs for [abbreviated name of facility]
28 and NUREG-1433, upon which TSTF-568 is based. These differences included [the TS 29 numbering and nomenclature]. The NRC staff determined that these differences do not 30 affect the applicability of TSTF-568 for [plant name].]
31 32
{Note: If the licensee identifies variations other than differences in the numbering of the TS and 33 nomenclature, they should be evaluated in this section.}
34 35
4.0 STATE CONSULTATION
36 37 In accordance with the Commission's regulations, the [Name of State] State official was notified 38 of the proposed issuance of the amendment on [enter date]. The State official had [no]
39 comments. [If comments were provided, they should be addressed here].
40 41
5.0 ENVIRONMENTAL CONSIDERATION
42 43
{NOTE: This section is to be prepared by the PM. As needed, the PM should coordinate with 44 NRRs Environmental Review and NEPA Branch (MENB) to determine the need for an EA.
45 Specific guidance on preparing EAs and considering environmental issues is contained in NRR 46 Office Instruction LIC-203, Procedural Guidance for Preparing Categorical Exclusions, 47 Environmental Assessments, and Considering Environmental Issues.}
48 49 The amendment changes requirements with respect to the installation or use of facility 50 components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has 51
determined that the amendment involves no significant increase in the amounts, and no 1
significant change in the types, of any effluents that may be released offsite, and that there is no 2
significant increase in individual or cumulative occupational radiation exposure. The 3
Commission has previously issued a proposed finding that the amendment involves no 4
significant hazards consideration, which was published in the Federal Register on [DATE 5
(XX FR XXX)], and there has been no public comment on such finding. Accordingly, the 6
amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
7 Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment 8
need be prepared in connection with the issuance of the amendment.
9 10
6.0 CONCLUSION
11 12
{NOTE: This section is to be prepared by the PM.}
13 14 The Commission has concluded, based on the considerations discussed above, that: (1) there 15 is reasonable assurance that the health and safety of the public will not be endangered by 16 operation in the proposed manner (2) there is reasonable assurance that such activities will be 17 conducted in compliance with the Commissions regulations, and (3) the issuance of the 18 amendment will not be inimical to the common defense and security or to the health and safety 19 of the public.
20 21
7.0 REFERENCES
22 23
{NOTE: The references below are provided so the PM can cut and paste the appropriate 24 references into the footnote in Section 3.1.1 of this SE, then References 1-6 can be deleted.
25 However, if the PM prefers to include a reference section, then delete only those not applicable 26 and add all other references here. The full citation for other documents referenced throughout 27 this model SE can be copied from the traveler SE (ADAMS Accession No. [ML19XXXXXX]).
28 The DORL PM would also need to add any plant-specific references (i.e., incoming LAR, RAI 29 responses, etc.).}
30 31
- 1.
Vassallo, Domenic B., U.S. Nuclear Regulatory Commission, letter to Hugh G. Parris, 32 Tennessee Valley Authority, Mark I Containment Long-Term Program, Re: Browns 33 Ferry Nuclear Plant, Units 1, 2, and 3, dated May 6, 1985 (ADAMS Package Accession 34 No. ML18029A537).
35 36
- 2.
Zwolinski, John A., U.S. Nuclear Regulatory Commission, letter to Dennis L. Farrer, 37 Commonwealth Edison Company, Mark I Containment Long-Term Program, Re:
38 Dresden Nuclear Power Station, Unit Nos. 2, and 3, dated September 18, 1985 39 (ADAMS Accession No. ML17195A950).
40 41
- 3.
U.S. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear 42 Reactor Regulation Related to Mark I Containment Long-Term Program Pool Dynamic 43 Loads Review, Commonwealth Edison Company, Docket Nos. 50-237/249, dated 44 September 18, 1985 (ADAMS Accession No. ML17195A952).
45 46
- 4.
U.S. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear 47 Reactor Regulation Related to Mark I Containment Long-Term Program Structural 48 Review, Commonwealth Edison Company, Docket Nos. 50-237/249, dated 49 September 18, 1985 (ADAMS Accession No. ML17195A953).
50 51
- 5.
Zwolinski, John A., U.S. Nuclear Regulatory Commission, letter to Dennis L. Farrar, 1
Commonwealth Edison Company, Mark I Containment Long-Term Program, Re: Quad 2
Cities Nuclear Power Station, Units 1 and 2, dated February 15, 1986 (ADAMS 3
Accession No. ML19199A123).
4 5
- 6.
Vassallo, Domenic B., U.S. Nuclear Regulatory Commission, letter to C. A. McNeill, Jr.,
6 Power Authority of the State of New York, Mark I Containment Long-Term Program Re:
7 James A. Fitzpatrick Nuclear Power Plant, dated December 12, 1984 (ADAMS 8
Accession No. ML19203A093).
9 10
{NOTE: These are the principal contributors for the model SE of the traveler. Replace these 11 names with those who prepared the plant-specific SE. Since this is a CLIIP Traveler, the only 12 reviewer necessary is DSS/STSB (unless there were significant variations.}
13 14 Principal Contributors: A. Sallman, NRR/DSS/SXRB 15 C. Tilton, NRR/DSS/STSB 16 17 Date:
18