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