ML17080A414
| ML17080A414 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 07/03/2017 |
| From: | Jennifer Whitman NRC/NRR/DSS/STSB |
| To: | Technical Specifications Task Force |
| Honcharik M, 415-1774, NRR/DSS/STSB | |
| Shared Package | |
| ML17080A409 | List: |
| References | |
| TAC MF5125 | |
| Download: ML17080A414 (15) | |
Text
July 3, 2017 Technical Specifications Task Force 11921 Rockville Pike, Suite 100 Rockville, MD 20852
SUBJECT:
DRAFT SAFETY EVALUATION OF TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-551, REVISION 3, "REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS" (TAC NO. MF5125)
Dear Members of the Technical Specifications Task Force:
By letter dated October 3, 2016 (Agencywide Documents Access and Management System Accession No. ML16277A226), the Technical Specifications Task Force submitted to the U.S.
Nuclear Regulatory Commission (NRC) for review and approval traveler TSTF-551, Revision 3, Revise Secondary Containment Surveillance Requirements. The NRC staffs draft safety evaluation (SE) of the traveler and a draft model SE are enclosed.
Thirty calendar days are provided to you to comment on any factual errors or clarity concerns contained in the enclosed draft SEs. The final SEs will be issued after making any necessary changes. The NRC staff's disposition of your comments on the draft SEs will be discussed in the final SEs. To facilitate the NRC staff's review of your comments, please provide a marked-up copy of the draft SEs showing proposed changes and provide a summary table of the proposed changes.
If you have any questions, please contact Michelle Honcharik at 301-415-1774 or via e-mail at Michelle.Honcharik@nrc.gov.
Sincerely,
/RA/
Jennifer M. Whitman, Acting Chief Technical Specifications Branch Division of Safety Systems Office of Nuclear Reactor Regulation Project No. 753
Enclosures:
As stated cc: See next page
Package: ML17080A409, Cover letter and Draft traveler SE: ML17080A414, Draft Model SE: ML17080A415; *concurred via e-mail
- from internal memos ML16062A426 and ML16340A067 NRR-106 OFFICE DPR/STSB*
DRA/ARCB*
DSS/SBPB**
DORL/BC*
NAME MHoncharik JDozier for KHsueh RDennig JDanna DATE 3/21/2017 6/5/17 12/6/2016 5/19/2017 OFFICE OGC*
DSS/STSB DSS/STSB*
NAME BHarris MHoncharik JMWhitman DATE 6/29/17 6/29/17 7/2017
Technical Specifications Task Force Project No. 753 cc:
Technical Specifications Task Force c/o EXCEL Services Corporation 11921 Rockville Pike, Suite 100 Rockville, MD 20852 Attention: Brian D. Mann E-mail: brian.mann@excelservices.com James R. Morris Diablo Canyon Power Plant Building 104/5/21A P.O. Box 56 Avila Beach, CA 93424 E-mail: james.morris@pge.com Lisa L. Williams Energy Northwest Columbia Generating Station PO Box 968 Mail Drop PE20 Richland, WA 99352-0968 E-mail: llwilliams@energy-northwest.com Otto W. Gustafson Entergy Nuclear Operations, Inc.
Palisades Nuclear Power Plant 27780 Blue Star Memorial Highway Covert, MI 49043 E-mail: ogustaf@entergy.com Jordan L. Vaughan Duke Energy EC2ZF / P.O. Box 1006 Charlotte, NC 28202 Email: jordan.vaughan@duke-energy.com Jason P. Redd Southern Nuclear Operating Company 42 Inverness Center Parkway Bin B234 Birmingham, AL 35242-4809 E-mail: jpredd@southernco.com
ENCLOSURE 1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 2
TSTF-551, REVISION 3, 3
REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS 4
5
1.0 INTRODUCTION
6 7
By letter dated September 3, 2015 (Agencywide Document Access and Management System 8
(ADAMS) Accession No. ML15246A131), the Technical Specifications (TS) Task Force (TSTF) 9 submitted traveler TSTF-551, Reactor Pressure Vessel Water Inventory Control, Revision 0, 10 for U.S. Nuclear Regulatory Commission (NRC) review and approval. By letter dated 11 January 26, 2016, the TSTF submitted Revision 1 to traveler TSTF-551 (ADAMS Accession 12 No. ML16026A026), and by letter dated May 12, 2016, the TSTF submitted Revision 2 to the 13 traveler (ADAMS Accession No. ML16133A536). By letter dated October 3, 2016 (ADAMS 14 Accession No. ML16277A226), the TSTF submitted Revision 3 of the Traveler TSTF-551.
15 16 Traveler TSTF-551 proposes changes to the Standard Technical Specifications (STS) and 17 Bases for boiling water reactor (BWR) designs BWR/4 and BWR/6.1 The changes would be 18 incorporated into future revisions of NUREG-1433, Volumes 1 and 2 and NUREG-1434, 19 Volumes 1 and 2. NUREG-1433 is based on the BWR/4 plant design, but is also representative 20 of the BWR/2, BWR/3, and, in some cases, BWR/5 designs. NUREG-1434 is based on the 21 BWR/6 plant design, and is representative, in many cases, of the BWR/5 design.
22 23 The proposed changes would allow the [secondary] containment vacuum limit to not be met 24 provided the standby gas treatment (SGT) system remains capable of establishing the required 25
[secondary] containment vacuum and revises NUREG-1433 to permit [secondary] containment 26 access opening to be open to permit ingress and egress similar to the corresponding 27 statements in NUREG-1434.
28 29 Throughout this safety evaluation (SE), items that are enclosed in square brackets signify 30 plant-specific nomenclature or values. Individual licensees would furnish plant-specific 31 nomenclature or values for bracketed items when submitting a license amendment request 32 (LAR) to adopt the changes described in this SE.
33 34 1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/4 Plants, NUREG-1433, Vol. 1, Specifications, Revision 4.0, April 2012, ADAMS Accession No. ML12104A192.
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/4 Plants, NUREG-1433, Vol. 2, Bases, Revision 4.0, April 2012, ADAMS Accession No. ML12104A193.
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/6 Plants, NUREG-1434, Vol. 1, Specifications, Revision 4.0, April 2012, ADAMS Accession No. ML12104A195.
U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/6 Plants, NUREG-1434, Vol. 2, Bases, Revision 4.0, April 2012, ADAMS Accession No. ML12104A196.
2.0 REGULATORY EVALUATION
1 2
2.1 SYSTEM DESCRIPTION 3
4 The [secondary] containment is a structure that encloses the primary containment, including 5
components that may contain primary system fluid. The safety function of the [secondary]
6 containment is to contain, dilute, and hold up fission products that may leak from primary 7
containment following a design basis accident (DBA) to ensure the control room operator and 8
offsite doses are within the regulatory limits. There is no redundant train or system that can 9
perform the [secondary] containment function should the [secondary] containment be 10 inoperable.
11 12 The [secondary] containment boundary is the combination of walls, floor, roof, ducting, doors, 13 hatches, penetrations and equipment that physically form the [secondary] containment. A 14 routinely used [secondary] containment access opening contains at least one inner and one 15 outer door in an airlock configuration. In some cases, [secondary] containment access 16 openings are shared such that there are multiple inner or outer doors. All [secondary]
17 containment access doors are normally kept closed, except when the access opening is being 18 used for entry and exit of personnel, equipment, or material.
19 20
[Secondary] containment operability is based on its ability to contain, dilute, and hold up fission 21 products that may leak from primary containment following a DBA. To prevent ground level 22 exfiltration of radioactive material while allowing the [secondary] containment to be designed as 23 a mostly conventional structure, the [secondary] containment requires support systems to 24 maintain the pressure at less than atmospheric pressure. During normal operation, non-safety 25 related systems are used to maintain the [secondary] containment at a slight negative pressure 26 to ensure any leakage is into the building and that any [secondary] containment atmosphere 27 exiting is via a pathway monitored for radioactive material. However, during normal operation it 28 is possible for the [secondary] containment vacuum to be momentarily less than the required 29 vacuum for a number of reasons, such as during wind gusts or swapping of the normal 30 ventilation subsystems.
31 32 During emergency conditions, the SGT system is designed to be capable of drawing down the 33
[secondary] containment to a required vacuum within a prescribed time and continue to maintain 34 the negative pressure as assumed in the accident analysis. The leak tightness of the 35
[secondary] containment together with the SGT system ensure that radioactive material is either 36 contained in the [secondary] containment or filtered through the SGT system filter trains before 37 being discharged to the outside environment via the elevated release point.
38 39 2.2 CHANGES TO THE STS 40 41 The proposed changes would allow the [secondary] containment vacuum limit to not be met 42 provided the SGT system remains capable of establishing the required [secondary] containment 43 vacuum and revises NUREG-1433 to permit [secondary] containment access opening to be 44 open to permit ingress and egress similar to the corresponding statements in NUREG-1434.
45 46 Corresponding changes are proposed to the STS Bases. A summary of the revised STS Bases 47 and the NRC staffs evaluation of the revised Bases are provided in an attachment to this SE.
48 49 2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 1
2 Surveillance requirement (SR) 3.6.4.1.1 requires verification that [secondary] containment 3
vacuum is [0.25] inch of vacuum water gauge. This SR would be modified by a note that 4
states:
5 6
Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if analysis demonstrates one 7
standby gas treatment (SGT) subsystem is capable of establishing 8
the required [secondary] containment vacuum.
9 10 This change is applicable to NUREG-1433 and -1434.
11 12 2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 13 14 SR 3.6.4.1.3 requires verification that one [secondary] containment access door in each access 15 opening is closed. This SR would be modified by adding the following phrase to the end of the 16 SR statement, except when the access opening is being used for entry and exit.
17 18 This change is applicable to NUREG-1433 only. This provision already exists in NUREG-1434, 19 Revision 4.
20 21 2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 22 23 An editorial change is made to SR 3.6.4.1.4 in which the words standby gas treatment are 24 replaced with the initialism SGT.
25 26 2.3 APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 27 28 Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications 29 Improvements for Nuclear Power Reactors (58 Federal Register 39132), dated July 22, 1993, 30 states in part:
31 32 The purpose of Technical Specifications is to impose those 33 conditions or limitations upon reactor operation necessary to 34 obviate the possibility of an abnormal situation or event giving rise 35 to an immediate threat to the public health and safety by 36 identifying those features that are of controlling importance to 37 safety and establishing on them certain conditions of operation 38 which cannot be changed without prior Commission approval.
39 40
[T]he Commission will also entertain requests to adopt portions of 41 the improved STS [(e.g., TSTF-551)], even if the licensee does 42 not adopt all STS improvements 43 44 The Commission encourages all licensees who submit Technical 45 Specification related submittals based on this Policy Statement to 46 emphasize human factors principles 47 48 In accordance with this Policy Statement, improved STS have 1
been developed and will be maintained for [BWR designs]. The 2
Commission encourages licensees to use the STS as the basis for 3
plant-specific Technical Specifications 4
5
[I]t is the Commission intent that the wording and Bases of the 6
improved STS be used [] to the extent practicable.
7 8
As described in the Commissions Final Policy Statement on Technical Specifications 9
Improvements for Nuclear Power Reactors, recommendations were made by NRC and industry 10 task groups for new STS that include greater emphasis on human factors principles in order to 11 add clarity and understanding to the text of the STS, and provide improvements to the Bases of 12 STS, which provides the purpose for each requirement in the specification. Subsequently, 13 improved vendor-specific STS were developed and issued by the NRC in September 1992.
14 15 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) 16 requires an applicant for an operating license to include in the application proposed TS in 17 accordance with the requirements of 10 CFR 50.36. The applicant must include in the 18 application, a summary statement of the bases or reasons for such specifications, other than 19 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these technical 20 specification bases shall not become part of the technical specifications.
21 22 Additionally, 10 CFR 50.36(b) requires:
23 24 Each license authorizing operation of a utilization facility will 25 include technical specifications. The technical specifications will 26 be derived from the analyses and evaluation included in the safety 27 analysis report, and amendments thereto, submitted pursuant to 28 10 CFR 50.34 [Contents of applications; technical information].
29 The Commission may include such additional technical 30 specifications as the Commission finds appropriate.
31 32 The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required 33 by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are 34 the lowest functional capability or performance levels of equipment required for safe operation 35 of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the 36 licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the 37 condition can be met.
38 39 The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 40 which are requirements relating to test, calibration, or inspection to assure that the necessary 41 quality of systems and components is maintained, that facility operation will be within safety 42 limits, and that the LCOs will be met.
43 44 Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for 45 an amendment must be filed with the Commission, fully describing the changes desired, and 46 following as far as applicable, the form prescribed for original applications.
47 48 Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the 1
applicant, the Commission will be guided by the considerations which govern the issuance of 2
initial licenses to the extent applicable and appropriate.
3 4
The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of 5
NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 6
Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425). As 7
described therein, as part of the regulatory standardization effort, the NRC staff has prepared 8
STS for each of the light-water reactor nuclear designs.
9 10 NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative 11 Source Terms, Revision 0, dated July 2000, provides guidance to the NRC staff for the review 12 of AST amendment requests. SRP 15.0.1 states that the NRC reviewer should evaluate the 13 proposed change against the guidance in RG 1.183.
14 15 Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design 16 Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000, provides acceptable 17 methodology for analyzing the radiological consequences of several design basis accidents to 18 show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable 19 application of alternate source term (AST) (also known as the accident source term) submittals, 20 including acceptable radiological analysis assumptions for use in conjunction with the accepted 21 AST.
22 23 10 CFR 50.67, Accident source term, states that:
24 25 (i)
An individual located at any point on the boundary of the 26 exclusion area for any 2-hour period following the onset of 27 the postulated fission product release, would not receive a 28 radiation dose in excess of 0.25 Sv (25 rem) total effective 29 dose equivalent (TEDE),
30 (ii)
An individual located at any point on the outer boundary of 31 the low population zone, who is exposed to the radioactive 32 cloud resulting from the postulated fission product release 33 (during the entire period of its passage), would not receive 34 a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and 35 (iii)
Adequate radiation protection is provided to permit access 36 to and occupancy of the control room under accident 37 conditions without personnel receiving radiation exposures 38 in excess of 0.05 Sv (5 rem) TEDE for the duration of the 39 accident.
40 41 In the evaluation of plant-specific LARs adopt TSTF-551 changes, the NRC staff will confirm the 42 current licensing basis, which reflects the AST methodology for analyzing the radiological 43 consequences of the design basis accidents using RG 1.183. The NRC staff will also consider 44 relevant information in the updated Final Safety Analysis Report (FSAR), which describes the 45 DBAs and evaluation of their radiological consequences for a specific licensee.
46 47
3.0 TECHNICAL EVALUATION
1 2
3.1 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.1 3
4 A note is being added to SR 3.6.4.1.1. The note allows the SR to not be met for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if 5
an analysis demonstrates that one SGT subsystem is capable of establishing the required 6
[secondary] containment vacuum. During normal operation, conditions may occur that result in 7
SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external 8
pressure or loss of the normal ventilation system that maintains [secondary] containment 9
vacuum may affect [secondary] containment vacuum. These conditions may not be indicative of 10 degradations of the [secondary] containment boundary or of the ability of the SGT system to 11 perform its specified safety function.
12 13 The note provides an allowance for the licensee to confirm [secondary] containment operability 14 by confirming that one SGT subsystem is capable of performing its specified safety function.
15 This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.
16 While the duration of these occurrences is anticipated to be very brief, the allowance is 17 permitted for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which is consistent with the time permitted for [secondary]
18 containment to be inoperable per Condition A of LCO 3.6.4.1.
19 20 The NRC staff intends to evaluate the impact of this note on the licensees design basis 21 radiological consequence dose analyses to ensure that the proposed change will not result in an 22 increase in the dose consequences and that the resulting calculated doses remain within the 23 design criteria specified in 10 CFR 50.67 and the accident specific design criteria outlined in 24 RG 1.183.
25 26 The proposed addition of the note to SR 3.6.4.1.1 does not change the STS requirement to 27 meet SR 3.6.4.1.4 and SR 3.6.4.1.5. SR 3.6.4.1.4 requires verification that the [secondary]
28 containment can be drawn down to [0.25] inch of vacuum water gauge in [120] seconds 29 using one SGT subsystem. SR 3.6.4.1.5 requires verification that the [secondary] containment 30 can be maintained [0.25] inch of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at 31 a flow rate [4000] cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment 32 (SGT) System, must be met; otherwise a licensee shall shut down the reactor or follow any 33 remedial action permitted by STS until the condition can be met.
34 35 As discussed above, [secondary] containment operability is based on its ability to contain, dilute, 36 and hold up fission products that may leak from primary containment following a DBA. To 37 prevent ground level exfiltration of radioactive material the [secondary] containment pressure 38 must be maintained at a pressure that is less than atmospheric pressure. The [secondary]
39 containment requires support systems to maintain the control volume pressure less than 40 atmospheric pressure. Following an accident, the SGT system ensures the [secondary]
41 containment pressure is less than the external atmospheric pressure. During normal operation, 42 non-safety related systems are used to maintain the [secondary] containment at a negative 43 pressure. However, during normal operation it is possible for the [secondary] containment 44 vacuum to be momentarily less than the required vacuum for a number of reasons. These 45 conditions are not indicative of degradations of the [secondary] containment boundary or of the 46 ability of the SGT system to perform its specified safety function. Since the licensee meets the 47 requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the LCO or is following the Actions of TS 48 LCO 3.6.4.3, and the licensees analysis confirms [secondary] containment operability by 49 confirming that one SGT subsystem is capable of performing its specified safety function, then 1
there is reasonable assurance that the [secondary] containment and SGT subsystem will 2
maintain the vacuum requirements during a DBA.
3 4
Therefore, the NRC staff has determined that: if the conditions do not affect (1) the ability to 5
maintain the [secondary] containment pressure during an accident, at a pressure that is less 6
than atmospheric, and (2) the time assumed in the accident analyses to draw down the 7
[secondary] containment pressure, then the [secondary] containment can perform its safety 8
function and may be considered TS operable. This is evident by being able to successfully 9
perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to 10 establish and maintain the required vacuum in the [secondary] containment as assumed in the 11 accident analyses.
12 13 If the specified safety functions of the [secondary] containment and SGT subsystem can be 14 performed in the time assumed in the accident analysis, then the fission products that bypass or 15 leak from primary containment, or are released from the reactor coolant pressure boundary 16 components located in [secondary] containment prior to release to the environment, will be 17 contained and processed as assumed in the design basis radiological consequence dose 18 analyses. If the above statement is true for a plant-specific amendment, then the NRC staff 19 finds that the proposed change does not affect the current radiological consequence analyses.
20 Therefore, the NRC staff concludes this change is acceptable with respect to the radiological 21 consequences of DBAs.
22 23
3.2 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.3 24 25
[NOTE: The proposed change is not applicable if the radiological dose consequence analysis 26 assumes the [secondary] containment pressure is below atmospheric pressure prior to or 27 coincident with the time at which the accident or event occurs. Such an analysis assumption 28 would require a revised radiological dose consequence analysis considering the new release 29 point (the open [secondary] containment doors), with appropriate atmospheric dispersion 30 factors, and any other necessary revisions to the accident or event analysis.]
31 32 The NRC staff review of SR 3.6.4.1.3 was limited to the request to provide an allowance for the 33 brief, inadvertent, simultaneous opening of redundant [secondary] containment access doors 34 during normal entry and exit conditions. Planned activities that could result in the simultaneous 35 opening of redundant [secondary] containment access openings, such as maintenance of a 36
[secondary] containment personnel access door or movement of large equipment through the 37 openings that would take longer than the normal transit time, will be considered outside the 38 scope of the NRC staff's review.
39 40 The NRC staff reviewed the changes to SR 3.6.4.1.3. The NRC staff determined that the SR 41 continues to provide appropriate confirmation that [secondary] containment boundary doors are 42 properly positioned and capable of performing their function in preserving the [secondary]
43 containment boundary. The NRC staff determined that the SRs continue to appropriately verify 44 the operability of the [secondary] containment and provide assurance that the necessary quality 45 of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
46 47 Additionally, the NRC staff evaluated the impact of modifying STS to allow [secondary]
48 containment access openings to be open for entry and exit on the design basis radiological 49 consequence dose analyses to ensure that the modification will not result in an increase in the 1
radiation dose consequences and that the resulting calculated radiation doses will remain within 2
the design criteria specified in 10 CFR 50.67 and the accident specific design criteria outlined in 3
RG 1.183. The NRC staff review of these DBAs determined that there are two DBAs that take 4
credit for the [secondary] containment, and are possibly impacted by the brief, inadvertent, 5
simultaneous opening of both an inner and outer access door during normal entry and exit 6
conditions, the loss-of-coolant accident (LOCA) and the fuel handling accident (FHA) in 7
[secondary] containment.
8 9
3.2.1 LOCA 10 11 Following a LOCA, the [secondary] containment structure is maintained at a negative pressure 12 ensuring that leakage from primary containment to [secondary] containment can be collected 13 and filtered prior to release to the environment. The SGT system performs the function of 14 maintaining a negative pressure within the [secondary] containment, as well as collecting and 15 filtering the leakage from primary containment. The SGT system is credited for mitigation of the 16 radiological releases from the [secondary] containment. In the LOCA analysis, the [secondary]
17 containment draw down analysis assumes that SGT system can draw down the [secondary]
18 containment within [5 minutes]. STS SR 3.6.4.1.4 requires one SGT subsystem to draw down 19 the [secondary] containment, to greater than or equal to [0.25] inches of vacuum water gauge in 20 a maximum allowable time of [120] seconds.
21 22 Conservatively, the DBA LOCA radiological consequence analysis in [UFSAR Chapter 15]
23 assumes that following the start of a DBA LOCA the [secondary] containment pressure of [0.25]
24 inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed that 25 releases into the [secondary] containment prior to the [10]-minute draw down time leak directly 26 to the environment as a ground level release with no filtration. After the assumed [10]-minute 27 draw down these releases are filtered by the SGT system and released via the SGT system 28 exhaust vent.
29 30 Based on this information, the NRC staff concludes that the DBA LOCA analysis has sufficient 31 conservatism by assuming a draw down time of [10] minutes from the start of the DBA LOCA.
32 Margin exists to ensure that the [secondary] containment can be reestablished during a brief, 33 inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 34 assurance that a failure of a safety system needed to control the release of radioactive material 35 to the environment will not result. The brief, inadvertent, simultaneous opening of the 36 secondary containment access doors does not impact the design bases and will not result in an 37 increase in any on-site or off-site dose.
38 39 Based on the above discussion, the NRC staff finds that the proposed change to the STSs does 40 not impact the design basis LOCA radiological consequence analysis, will not result in an 41 increase in any onsite or offsite dose, and is consistent with regulatory requirements and 42 guidance identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the 43 proposed change to the STSs will continue to comply with these criteria and that that the 44 estimates of the dose consequences of the postulated DBAs will comply with the requirements 45 of 10 CFR 50.67 and the accident specific dose guidelines specified in RG 1.183. Therefore, 46 the proposed changes are acceptable with regard to the radiological consequences of the 47 postulated DBAs.
48 49 3.2.2 FHA in [Secondary] Containment 1
2 During normal operation, non-safety related systems are used to maintain the [secondary]
3 containment at [0.25] inches of vacuum water gauge to ensure that any leakage is into the 4
building and that any [secondary] containment atmosphere exiting the building is via a 5
monitored pathway. The refueling floor, which is inside the [secondary] containment, is 6
maintained at a negative [0.25] inches of vacuum water gauge by normal operating ventilation 7
systems. The refueling floor exhaust ductwork in the [secondary] containment is equipped with 8
radiation monitors to detect a fuel handling accident. When a radiological release is sensed by 9
the radiation monitors, a [secondary] containment isolation signal is generated. This initiates 10 the SGT system and the normal ventilation system isolates. The radiation monitor is positioned 11 such that it will detect the release and send a closure signal to the [secondary] containment 12 isolation dampers.
13 14 Following a FHA, the [secondary] containment structure is maintained at a negative pressure by 15 the SGT system ensuring that fission products released from the spent fuel pool to [secondary]
16 containment can be collected and filtered prior to release to the environment. In the FHA 17 analysis, the [secondary] containment draw down analysis demonstrates that SGT system can 18 draw down the [secondary] containment within [5 minutes]. The SGT system is credited for 19 mitigation of the radiological releases from the [secondary] containment. STS SR 3.6.4.1.4 20 requires one SGT subsystem to draw down the [secondary] containment, to greater than or 21 equal to [0.25] inches of vacuum water gauge in a maximum allowable time of [120] seconds.
22 23 Conservatively, the DBA FHA radiological consequence analysis in [UFSAR Chapter 15]
24 assumes that following the start of a DBA FHA the [secondary] containment pressure of 25
[0.25] inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed 26 that releases into the [secondary] containment prior to the [10]-minute draw down time leak 27 directly to the environment as a ground level release with no filtration. After the assumed 28
[10]-minute draw down these releases are filtered by the SGT system and released via the SGT 29 system exhaust vent.
30 31 Based on this information, the NRC staff concludes that the DBA FHA analysis has sufficient 32 conservatism by assuming a draw down time of [10] minutes from the start of the DBA FHA.
33 Margin exists to ensure that the [secondary] containment can be reestablished during brief, 34 inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 35 assurance that a failure of a safety system needed to control the release of radioactive material 36 to the environment will not result. The brief, inadvertent, simultaneous opening of the 37
[secondary] containment access doors does not impact the design bases and will not result in 38 an increase in any on-site or off-site dose.
39 40 Based on the above discussion, the NRC staff finds that the proposed change to the STSs does 41 not impact the design basis FHA radiological consequence analysis, will not result in an increase 42 in any onsite or offsite dose, and is consistent with regulatory requirements and guidance 43 identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the proposed change 44 to the STSs will continue to comply with these criteria and that that the estimates of the dose 45 consequences of the postulated DBAs will comply with the requirements of 10 CFR 50.67 and 46 the accident specific dose guidelines specified in RG 1.183. Therefore, the proposed changes 47 are acceptable with regard to the radiological consequences of the postulated DBAs.
48 49
3.3 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.4 1
2 The changes to SR 3.6.4.1.4 are editorial only and do not change any technical aspects of SR 3
3.6.4.1.4. The NRC staff determined that the change is acceptable.
4 5
4.0 CONCLUSION
6 7
The NRC staff reviewed traveler TSTF-551, Revision 3, which proposed changes to 8
NUREG-1433, Volumes 1 (STS) and 2 (Bases) and NUREG-1434 Volumes 1 (STS) and 2 9
(Bases). The NRC staff determined that the proposed changes to the STS met the standards 10 for TS in 10 CFR 50.36(b). The proposed SRs assure that the necessary quality of systems 11 and components is maintained, that facility operation will be within safety limits, and that the 12 LCOs will be met, and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the STS were 13 reviewed for technical clarity and consistency with customary terminology and format in 14 accordance with SRP Chapter 16.
15 16 The proposed bases, which will be added to future revisions to NUREG-1433, Volume 2, and 17 NUREG-1434, Volume 2, satisfy the Commissions Policy Statement by addressing the 18 questions specified in the policy statement, and cite references to appropriate licensing 19 documentation to support the Bases.
20 21 Additionally, the NRC staff has evaluated the impact of the proposed changes on the design 22 basis radiological consequence analyses against the regulatory requirements and guidance 23 identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the 24 changes to the STSs will continue to comply with the requirements of 10 CFR 50.67 and the 25 guidelines specified in RG 1.183. Therefore, the proposed changes are acceptable with regard 26 to the radiological consequences of the postulated DBAs.
27 28 29 Technical contacts:
Kristy Bucholtz, NRR/DRA/ARCB 30 Nageswara Karipineni, NRR/DSS/SBPB 31 32
Attachment:
Basis for Accepting the Proposed Changes to the Standard Technical 33 Specification Bases, Volume 2 of NUREGs-1433 and -1434 34 35 Date:
36 37 38 39
ATTACHMENT ATTACHMENT 1
2 BASIS FOR ACCEPTING THE PROPOSED CHANGES TO THE STANDARD TECHNICAL 3
SPECIFICATION BASES, VOLUME 2 OF NUREGS-1433 AND -1434 4
5
1.0 INTRODUCTION
6 7
Traveler TSTF-551 proposes changes to Standard Technical Specifications, General Electric 8
BWR/4 Plants, BWR/4 NUREG-1433, Volume 2, Bases, Revision 4.0, April 2012, ADAMS 9
Accession No. ML12104A193 and Standard Technical Specifications, General Electric BWR/6 10 Plants, BWR/6 NUREG-1434, Volume 2, Bases, Revision 4.0, April 2012, ADAMS Accession 11 No. ML12104A196. The changes would be incorporated into future revisions of NUREG-1433, 12 Volume 2, and NUREG-1434, Volume 2. A summary of the changes and the NRC staffs 13 evaluation of those changes are presented in this attachment.
14 15
2.0 REGULATORY EVALUATION
16 17 2.1 APPLICABLE REGULATIONS AND GUIDANCE 18 19 The regulation at 10 CFR 50.36(a)(1) states that each applicant for a license authorizing 20 operation of a production or utilization facility shall include in his application proposed technical 21 specifications in accordance with the requirements of this section. A summary statement of the 22 bases or reasons for such specifications, other than those covering administrative controls, shall 23 also be included in the application, but shall not become part of the technical specifications.
24 25 In its Final Policy Statement on Technical Specifications Improvements for Nuclear Power 26 Reactors, the Commission presented its policy on the scope and purpose of the Technical 27 Specifications. The Commission explained how implementation of the policy statement through 28 implementation of the improved STS is expected to produce an improvement in the safety of 29 nuclear power plants through the use of more operator-oriented TS, improved TS Bases, 30 reduced action-statement-induced plant transients, and more efficient use of NRC and industry 31 resources.
32 33 The Final Policy Statement provides the following description of the scope and the purpose of 34 the Technical Specification Bases:
35 36 Appropriate Surveillance Requirements and Actions should be 37 retained for each LCO which remains or is included in the 38 Technical Specifications. Each LCO, Action, and Surveillance 39 Requirement should have supporting Bases. The Bases should at 40 a minimum address the following questions and cite references to 41 appropriate licensing documentation (e.g., FSAR, Topical Report) 42 to support the Bases.
43 44
- 1. What is the justification for the Technical Specification, i.e., which 45 Policy Statement criterion requires it to be in the Technical 46 Specifications?
47 48
- 2. What are the Bases for each LCO, i.e., why was it determined to 1
be the lowest functional capability or performance level for the 2
system or component in question necessary for safe operation of 3
the facility and, what are the reasons for the Applicability of the 4
LCO?
5 6
- 3. What are the Bases for each Action, i.e., why should this remedial 7
action be taken if the associated LCO cannot be met; how does 8
this Action relate to other Actions associated with the LCO; and 9
what justifies continued operation of the system or component at 10 the reduced state from the state specified in the LCO for the 11 allowed time period?
12 13
- 4. What are the Bases for each Safety Limit?
14 15
- 5. What are the Bases for each Surveillance Requirement and 16 Surveillance Frequency; i.e., what specific functional requirement 17 is the surveillance designed to verify? Why is this surveillance 18 necessary at the specified frequency to assure that the system or 19 component function is maintained, that facility operation will be 20 within the Safety Limits, and that the LCO will be met?
21 22 Note: In answering these questions the Bases for each number 23 (e.g., Allowable Value, Response Time, Completion Time, 24 Surveillance Frequency, etc.), state, condition, and definition (e.g.,
25 operability) should be clearly specified. As an example, a number 26 might be based on engineering judgment, past experience, or 27 PSA insights; but this should be clearly stated.
28 29 The NRC staff used the guidance contained in the Final Policy Statement during its review of 30 the proposed changes to the Bases.
31 32
2.2 DESCRIPTION
OF CHANGES 33 34 Volume 2 of NUREGs-1433 and -1434 contain the Bases for each Safety Limit and each LCO 35 contained in Volume 1. The Bases for each LCO is organized into sections:
36 37
=
Background===
38 Applicable Safety Analyses, LCO, and Applicability 39 Actions 40 Surveillance Requirements 41 References 42 43 The Bases for SR 3.6.4.1.1 in NUREGs-1433 and -1434 is being revised, and the Bases for 44 SR 3.6.4.1.3 in NUREG-1433 is being revised. The following discussion provides a summary of 45 the revised Bases, followed by the NRC staffs evaluation of the revised Bases.
46 47
3.0 TECHNICAL EVALUATION
1 2
3.1 REVISION TO SR 3.6.4.1.1 BASES 3
4 The Bases for SR 3.6.4.1.1 is revised by the addition of a description of the modification to the 5
applicability of the SR acceptance criterion. The revised Bases describe conditions that could 6
lead to the required vacuum not being met and provides a discussion of why these conditions 7
do not indicate a change in the leaktightness of the [secondary] containment boundary. It also 8
provides a description of the analysis needed to determine whether one train of SGT could 9
establish the assumed [secondary] containment vacuum in the unlikely event of an accident 10 occurring.
11 12 The NRC staff reviewed the revised Bases and determined that it adequately provides the basis 13 for the SR, and provides an appropriate description of the note which modifies the SR.
14 15 3.2 REVISION TO SR 3.6.4.1.3 BASES 16 17 The Bases for SR 3.6.4.1.3 are revised in their entirety to describe that the verification of one 18 door being closed is necessary to provide assurance that exfiltration from the [secondary]
19 containment does not occur. The revised bases also provide an explanation that the intent is 20 not to breach the [secondary] containment boundary, but the access openings may be used for 21 entry and exit.
22 23 The NRC staff reviewed the revised Bases and determined that it adequately provides the 24 purpose and the basis for the SR.
25 26
4.0 CONCLUSION
27 28 The NRC staff determined that TS Bases changes are consistent with the proposed TS changes 29 and provide an explanation and supporting information for each of the SRs. Therefore, the NRC 30 staff determined that the revised Bases are consistent with the Commission's Final Policy 31 Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 32 1993 (58 Federal Register 39132).
33 34