TSTF-17-10, TSTF Comments on Draft Safety Evaluation for Traveler TSTF-551, Revision 3, Revise Secondary Containment Surveillance Requirements.
| ML17214A804 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 08/02/2017 |
| From: | Gustafson O, Morris J, Redd J, Vaughan J, Linda Williams BWR Owners Group, Babcock & Wilcox, Combustion Engineering, PWR Owners Group, Technical Specifications Task Force, Westinghouse |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TSTF-17-10 | |
| Download: ML17214A804 (32) | |
Text
TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY August 2, 2017 TSTF-17-10 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
TSTF Comments on Draft Safety Evaluation for Traveler TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements"
REFERENCE:
Letter from Jennifer Whitman (NRC) to the TSTF, "Draft Safety Evaluation of Technical Specifications Task Force Traveler TSTF-551, Revision 3,
'Revise Secondary Containment Surveillance Requirements'," dated July 3, 2017 (ADAMS Accession No. ML17080A409).
On October 3, 2016, the TSTF submitted traveler TSTF-551, Revision 3, " Revise Secondary Containment Surveillance Requirements," to the Nuclear Regulatory Commission (NRC) for review (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16277A226). In the referenced letter, the NRC provided the draft Safety Evaluations for TSTF-551 for comment.
Attachment 1 contains a summary table providing the TSTF's comments on the draft Safety Evaluations. Attachment 2 contains a mark-up reflecting the TSTF's comments.
Should you have any questions, please contact us.
James R. Morris (PWROG/W) Lisa L. Williams (BWROG)
Otto W. Gustafson (PWROG/CE) Jordan L. Vaughan (PWROG/B&W)
Jason P. Redd (APOG) 11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation TSTF Comments on the TSTF-551 Draft Safety Evaluations TSTF Markup of Draft Safety Evaluations cc: Michelle Honcharik, Technical Specifications Branch, NRC Robert Tjader, Technical Specifications Branch, NRC Jennifer Whitman, Technical Specifications Branch, NRC
Attachment 1 TSTF Comments on the TSTF-551 Draft Safety Evaluations Page(s) Line(s)1 Comment Traveler Draft Safety Evaluation 1 9, 13 The ADAMS Accession numbers for the Revision 0 and Revision 1 submittals of TSTF-551 are incorrect. The correct numbers are ML14304A034 and ML15246A131, respectively.
1 26 Recommend adding a footnote explaining why "secondary" is bracketed.
1 27 In some locations, the SE refers to the SR 3.6.4.1.3 note as pertaining to 2 45 "ingress and egress" instead of "entry and exit." This is an unnecessary departure from the TS wording and contrary to plain language.
Recommend using the TSTF-551 and TS wording of "entry and exit."
5 11-46 The SE discusses only radiological consequence analysis using alternative 6 23-25 source term (AST), 10 CFR 50.67, and Regulatory Guide 1.183.
8 3-4, 46 TSTF-551 is applicable to all BWR plants regardless of whether the Loss 9 46-47 of Coolant Accident and Fuel Handling Accident analyses are based on 10 25-26 AST or traditional source terms. The SE should be revised to encompass any plant's source term licensing basis. The TSTF proposes eliminating references to AST and using the phrase from page 7, line 20, "the current radiological consequence analyses."
Draft Model Safety Evaluation for Plant-Specific Adoption
-- -- The model SE is written as a plant-specific SE instead of evaluating the licensee's adoption of TSTF-551. The model SE repeats all of the generic evaluation of TSTF-551 and does not discuss variations from the approved traveler. The TSTF believes this approach undermines the efficiency of the traveler process. The TSTF has not provided a rewrite, but we recommend the NRC revise the technical evaluation to confirm that the traveler SE assumptions and conclusions are applicable.
1 6 Recommend adding note explaining the use of brackets.
1 15 The bolded phrase "Month, Day, 2017 (ADAMS Accession No. MLXXXX)" should be placed in brackets, consistent with similar information in the paragraph.
1 27 In some locations, the model SE refers to the SR 3.6.4.1.3 note as 2 45 pertaining to "ingress and egress" instead of "entry and exit." This is an unnecessary departure from the TS wording and contrary to plain language. Recommend using the TSTF-551 and TS wording of "entry and exit."
1 Line numbers correspond to the documents provided by the NRC and not to the attached proposed revision.
Page 3
1 19 In discussing the SR 3.6.4.1.1 note, the model SE states, "provided that the standby gas treatment (SGT) system remains capable of establishing the required [secondary] containment vacuum within the [specified time]." The SR note does not refer to the time required to establish containment vacuum. Reference to time should be eliminated to be consistent with the TS.
2 29 In the model SE, the TSTF recommends not referring to the Standard 2 41 Technical Specifications NUREGs, but to instead refer to the applicable BWR types.
3 39-49 The model SE discusses only radiological consequence analysis using 4 1-20 alternative source term (AST), 10 CFR 50.67, and Regulatory Guide 5 12 1.183. TSTF-551 is applicable to all BWR plants regardless of whether 6 39-40 the Loss of Coolant Accident and Fuel Handling Accident analyses are 9 4-5 based on AST or traditional source terms. The SE should be revised to 9 25-26 encompass any plant's source term licensing basis. The TSTF proposes eliminating references to AST and using the phrase from page 7, line 20, "the current radiological consequence analyses."
5 5 Remove the unmatched bracket before the word "which."
9 13 Recommend adding a section "Variations from the Approved Traveler,"
similar to other recent model SEs. The section should include example discussions of the variations included in the TSTF-551 model application.
Page 4
Attachment 2 TSTF Markup of Draft Safety Evaluations
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. ML14304A034 ML15246A131), the Technical Specifications (TS) Task 10 Force (TSTF) submitted traveler TSTF-551, Reactor Pressure Vessel Water Inventory Control, 11 Revision 0, for U.S. Nuclear Regulatory Commission (NRC) review and approval. By letter 12 dated January 26, 2016, the TSTF submitted Revision 1 to traveler TSTF-551 (ADAMS 13 Accession No. ML15246A131ML16026A026), and by letter dated May 12, 2016, the TSTF 14 submitted Revision 2 to the traveler (ADAMS Accession No. ML16133A536). By letter dated 15 October 3, 2016 (ADAMS Accession No. ML16277A226), the TSTF submitted Revision 3 of the 16 Traveler TSTF-551.
17 18 Traveler TSTF-551 proposes changes to the Standard Technical Specifications (STS) and 19 Bases for boiling water reactor (BWR) designs BWR/4 and BWR/6.1 The changes would be 20 incorporated into future revisions of NUREG-1433, Volumes 1 and 2 and NUREG-1434, 21 Volumes 1 and 2. NUREG-1433 is based on the BWR/4 plant design, but is also representative 22 of the BWR/2, BWR/3, and, in some cases, BWR/5 designs. NUREG-1434 is based on the 23 BWR/6 plant design, and is representative, in many cases, of the BWR/5 design.
24 25 The proposed changes would allow the [secondary]2 containment vacuum limit to not be met 26 provided the standby gas treatment (SGT) system remains capable of establishing the required 27 [secondary] containment vacuum and revises NUREG-1433 to permit [secondary] containment 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 Plants of BWR/4 design have differing names for the secondary containment. As a result, the BWR/4 ISTS uses the convention, "[secondary] containment." In the ISTS, brackets indicate plant-specific information. BWR/6 plants also have differing names for secondary containment, or the primary containment serves a similar function. The BWR/6 ISTS uses the convention "[secondary containment]."
In this discussion, the phrase "[secondary] containment" applies to both BWR/4 and BWR/6 plants.
ENCLOSURE 1
1 access opening to be open to permit entry and exit ingress and egress similar to the 2 corresponding statements in NUREG-1434.
3 4 Throughout this safety evaluation (SE), items that are enclosed in square brackets signify 5 plant-specific nomenclature or values. Individual licensees would furnish plant-specific 6 nomenclature or values for bracketed items when submitting a license amendment request 7 (LAR) to adopt the changes described in this SE.
8 9
2.0 REGULATORY EVALUATION
10 11 2.1 SYSTEM DESCRIPTION 12 13 The [secondary] containment is a structure that encloses the primary containment, including 14 components that may contain primary system fluid. The safety function of the [secondary]
15 containment is to contain, dilute, and hold up fission products that may leak from primary 16 containment following a design basis accident (DBA) to ensure the control room operator and 17 offsite doses are within the regulatory limits. There is no redundant train or system that can 18 perform the [secondary] containment function should the [secondary] containment be 19 inoperable.
20 21 The [secondary] containment boundary is the combination of walls, floor, roof, ducting, doors, 22 hatches, penetrations and equipment that physically form the [secondary] containment. A 23 routinely used [secondary] containment access opening contains at least one inner and one 24 outer door in an airlock configuration. In some cases, [secondary] containment access 25 openings are shared such that there are multiple inner or outer doors. All [secondary]
26 containment access doors are normally kept closed, except when the access opening is being 27 used for entry and exit of personnel, equipment, or material.
28 29 [Secondary] containment operability is based on its ability to contain, dilute, and hold up fission 30 products that may leak from primary containment following a DBA. To prevent ground level 31 exfiltration of radioactive material while allowing the [secondary] containment to be designed as 32 a mostly conventional structure, the [secondary] containment requires support systems to 33 maintain the pressure at less than atmospheric pressure. During normal operation, non-safety 34 related systems are used to maintain the [secondary] containment at a slight negative pressure 35 to ensure any leakage is into the building and that any [secondary] containment atmosphere 36 exiting is via a pathway monitored for radioactive material. However, during normal operation it 37 is possible for the [secondary] containment vacuum to be momentarily less than the required 38 vacuum for a number of reasons, such as during wind gusts or swapping of the normal 39 ventilation subsystems.
40 41 During emergency conditions, the SGT system is designed to be capable of drawing down the 42 [secondary] containment to a required vacuum within a prescribed time and continue to maintain 43 the negative pressure as assumed in the accident analysis. The leak tightness of the 44 [secondary] containment together with the SGT system ensure that radioactive material is either 45 contained in the [secondary] containment or filtered through the SGT system filter trains before 46 being discharged to the outside environment via the elevated release point.
47 48 2.2 CHANGES TO THE STS 49
1 The proposed changes would allow the [secondary] containment vacuum limit to not be met 2 provided the SGT system remains capable of establishing the required [secondary] containment 3 vacuum and revises NUREG-1433 to permit [secondary] containment access opening to be 4 open to permit entry and exit ingress and egress similar to the corresponding statements in 5 NUREG-1434.
6 7 Corresponding changes are proposed to the STS Bases. A summary of the revised STS Bases 8 and the NRC staffs evaluation of the revised Bases are provided in an attachment to this SE.
9 10 2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 11 12 Surveillance requirement (SR) 3.6.4.1.1 requires verification that [secondary] containment 13 vacuum is [0.25] inch of vacuum water gauge. This SR would be modified by a note that 14 states:
15 16 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 17 standby gas treatment (SGT) subsystem is capable of establishing 18 the required [secondary] containment vacuum.
19 20 This change is applicable to NUREG-1433 and -1434.
21 22 2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 23 24 SR 3.6.4.1.3 requires verification that one [secondary] containment access door in each access 25 opening is closed. This SR would be modified by adding the following phrase to the end of the 26 SR statement, except when the access opening is being used for entry and exit.
27 28 This change is applicable to NUREG-1433 only. This provision already exists in NUREG-1434, 29 Revision 4.
30 31 2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 32 33 An editorial change is made to SR 3.6.4.1.4 in which the words standby gas treatment are 34 replaced with the initialism SGT.
35 36 2.3 APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 37 38 Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications 39 Improvements for Nuclear Power Reactors (58 Federal Register 39132), dated July 22, 1993, 40 states in part:
41 42 The purpose of Technical Specifications is to impose those 43 conditions or limitations upon reactor operation necessary to 44 obviate the possibility of an abnormal situation or event giving rise 45 to an immediate threat to the public health and safety by 46 identifying those features that are of controlling importance to 47 safety and establishing on them certain conditions of operation 48 which cannot be changed without prior Commission approval.
49
1 [T]he Commission will also entertain requests to adopt portions of 2 the improved STS [(e.g., TSTF-551)], even if the licensee does 3 not adopt all STS improvements 4
5 The Commission encourages all licensees who submit Technical 6 Specification related submittals based on this Policy Statement to 7 emphasize human factors principles 8
9 In accordance with this Policy Statement, improved STS have 10 been developed and will be maintained for [BWR designs]. The 11 Commission encourages licensees to use the STS as the basis for 12 plant-specific Technical Specifications 13 14 [I]t is the Commission intent that the wording and Bases of the 15 improved STS be used [] to the extent practicable.
16 17 As described in the Commissions Final Policy Statement on Technical Specifications 18 Improvements for Nuclear Power Reactors, recommendations were made by NRC and industry 19 task groups for new STS that include greater emphasis on human factors principles in order to 20 add clarity and understanding to the text of the STS, and provide improvements to the Bases of 21 STS, which provides the purpose for each requirement in the specification. Subsequently, 22 improved vendor-specific STS were developed and issued by the NRC in September 1992.
23 24 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) 25 requires an applicant for an operating license to include in the application proposed TS in 26 accordance with the requirements of 10 CFR 50.36. The applicant must include in the 27 application, a summary statement of the bases or reasons for such specifications, other than 28 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these technical 29 specification bases shall not become part of the technical specifications.
30 31 Additionally, 10 CFR 50.36(b) requires:
32 33 Each license authorizing operation of a utilization facility will 34 include technical specifications. The technical specifications will 35 be derived from the analyses and evaluation included in the safety 36 analysis report, and amendments thereto, submitted pursuant to 37 10 CFR 50.34 [Contents of applications; technical information].
38 The Commission may include such additional technical 39 specifications as the Commission finds appropriate.
40 41 The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required 42 by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are 43 the lowest functional capability or performance levels of equipment required for safe operation 44 of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the 45 licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the 46 condition can be met.
47 48 The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 49 which are requirements relating to test, calibration, or inspection to assure that the necessary
1 quality of systems and components is maintained, that facility operation will be within safety 2 limits, and that the LCOs will be met.
3 4 Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for 5 an amendment must be filed with the Commission, fully describing the changes desired, and 6 following as far as applicable, the form prescribed for original applications.
7 8 Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the 9 applicant, the Commission will be guided by the considerations which govern the issuance of 10 initial licenses to the extent applicable and appropriate.
11 12 The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of 13 NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 14 Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425). As 15 described therein, as part of the regulatory standardization effort, the NRC staff has prepared 16 STS for each of the light-water reactor nuclear designs.
17 18 NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative 19 Source Terms, Revision 0, dated July 2000, provides guidance to the NRC staff for the review 20 of AST amendment requests. SRP 15.0.1 states that the NRC reviewer should evaluate the 21 proposed change against the guidance in RG 1.183.
22 23 Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design 24 Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000, provides acceptable 25 methodology for analyzing the radiological consequences of several design basis accidents to 26 show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable 27 application of alternate source term (AST) (also known as the accident source term) submittals, 28 including acceptable radiological analysis assumptions for use in conjunction with the accepted 29 AST.
30 31 10 CFR 50.67, Accident source term, states that:
32 33 (i) An individual located at any point on the boundary of the 34 exclusion area for any 2-hour period following the onset of 35 the postulated fission product release, would not receive a 36 radiation dose in excess of 0.25 Sv (25 rem) total effective 37 dose equivalent (TEDE),
38 (ii) An individual located at any point on the outer boundary of 39 the low population zone, who is exposed to the radioactive 40 cloud resulting from the postulated fission product release 41 (during the entire period of its passage), would not receive 42 a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and 43 (iii) Adequate radiation protection is provided to permit access 44 to and occupancy of the control room under accident 45 conditions without personnel receiving radiation exposures 46 in excess of 0.05 Sv (5 rem) TEDE for the duration of the 47 accident.
48
1 In the evaluation of plant-specific LARs adopt TSTF-551 changes, the NRC staff will confirm the 2 current licensing basis, which reflects the AST methodology for analyzing the radiological 3 consequences of the design basis accidents using RG 1.183. The NRC staff will also consider 4 relevant information in the updated Final Safety Analysis Report (FSAR), which describes the 5 DBAs and evaluation of their radiological consequences for a specific licensee.
6 7
3.0 TECHNICAL EVALUATION
8 9
3.1 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.1 10 11 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 12 an analysis demonstrates that one SGT subsystem is capable of establishing the required 13 [secondary] containment vacuum. During normal operation, conditions may occur that result in 14 SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external 15 pressure or loss of the normal ventilation system that maintains [secondary] containment 16 vacuum may affect [secondary] containment vacuum. These conditions may not be indicative of 17 degradations of the [secondary] containment boundary or of the ability of the SGT system to 18 perform its specified safety function.
19 20 The note provides an allowance for the licensee to confirm [secondary] containment operability 21 by confirming that one SGT subsystem is capable of performing its specified safety function.
22 This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.
23 While the duration of these occurrences is anticipated to be very brief, the allowance is 24 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]
25 containment to be inoperable per Condition A of LCO 3.6.4.1.
26 27 The NRC staff intends to evaluate the impact of this note on the licensees design basis 28 radiological consequence dose analyses to ensure that the proposed change will not result in an 29 increase in the dose consequences and that the resulting calculated doses remain within the 30 current radiological consequence analysesthe design criteria specified in 10 CFR 50.67 and the 31 accident specific design criteria outlined in RG 1.183.
32 33 The proposed addition of the note to SR 3.6.4.1.1 does not change the STS requirement to 34 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]
35 containment can be drawn down to [0.25] inch of vacuum water gauge in [120] seconds 36 using one SGT subsystem. SR 3.6.4.1.5 requires verification that the [secondary] containment 37 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 38 a flow rate [4000] cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment 39 (SGT) System, must be met; otherwise a licensee shall shut down the reactor or follow any 40 remedial action permitted by STS until the condition can be met.
41 42 As discussed above, [secondary] containment operability is based on its ability to contain, dilute, 43 and hold up fission products that may leak from primary containment following a DBA. To 44 prevent ground level exfiltration of radioactive material the [secondary] containment pressure 45 must be maintained at a pressure that is less than atmospheric pressure. The [secondary]
46 containment requires support systems to maintain the control volume pressure less than 47 atmospheric pressure. Following an accident, the SGT system ensures the [secondary]
48 containment pressure is less than the external atmospheric pressure. During normal operation, 49 non-safety related systems are used to maintain the [secondary] containment at a negative
1 pressure. However, during normal operation it is possible for the [secondary] containment 2 vacuum to be momentarily less than the required vacuum for a number of reasons. These 3 conditions are not indicative of degradations of the [secondary] containment boundary or of the 4 ability of the SGT system to perform its specified safety function. Since the licensee meets the 5 requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the LCO or is following the Actions of TS 6 LCO 3.6.4.3, and the licensees analysis confirms [secondary] containment operability by 7 confirming that one SGT subsystem is capable of performing its specified safety function, then 8 there is reasonable assurance that the [secondary] containment and SGT subsystem will 9 maintain the vacuum requirements during a DBA.
10 11 Therefore, the NRC staff has determined that: if the conditions do not affect (1) the ability to 12 maintain the [secondary] containment pressure during an accident, at a pressure that is less 13 than atmospheric, and (2) the time assumed in the accident analyses to draw down the 14 [secondary] containment pressure, then the [secondary] containment can perform its safety 15 function and may be considered TS operable. This is evident by being able to successfully 16 perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to 17 establish and maintain the required vacuum in the [secondary] containment as assumed in the 18 accident analyses.
19 20 If the specified safety functions of the [secondary] containment and SGT subsystem can be 21 performed in the time assumed in the accident analysis, then the fission products that bypass or 22 leak from primary containment, or are released from the reactor coolant pressure boundary 23 components located in [secondary] containment prior to release to the environment, will be 24 contained and processed as assumed in the design basis radiological consequence dose 25 analyses. If the above statement is true for a plant-specific amendment, then the NRC staff 26 finds that the proposed change does not affect the current radiological consequence analyses.
27 Therefore, the NRC staff concludes this change is acceptable with respect to the radiological 28 consequences of DBAs.
29 30
3.2 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.3 31 32 [NOTE: The proposed change is not applicable if the radiological dose consequence analysis 33 assumes the [secondary] containment pressure is below atmospheric pressure prior to or 34 coincident with the time at which the accident or event occurs. Such an analysis assumption 35 would require a revised radiological dose consequence analysis considering the new release 36 point (the open [secondary] containment doors), with appropriate atmospheric dispersion 37 factors, and any other necessary revisions to the accident or event analysis.]
38 39 The NRC staff review of SR 3.6.4.1.3 was limited to the request to provide an allowance for the 40 brief, inadvertent, simultaneous opening of redundant [secondary] containment access doors 41 during normal entry and exit conditions. Planned activities that could result in the simultaneous 42 opening of redundant [secondary] containment access openings, such as maintenance of a 43 [secondary] containment personnel access door or movement of large equipment through the 44 openings that would take longer than the normal transit time, will be considered outside the 45 scope of the NRC staff's review.
46 47 The NRC staff reviewed the changes to SR 3.6.4.1.3. The NRC staff determined that the SR 48 continues to provide appropriate confirmation that [secondary] containment boundary doors are 49 properly positioned and capable of performing their function in preserving the [secondary]
1 containment boundary. The NRC staff determined that the SRs continue to appropriately verify 2 the operability of the [secondary] containment and provide assurance that the necessary quality 3 of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
4 5 Additionally, the NRC staff evaluated the impact of modifying STS to allow [secondary]
6 containment access openings to be open for entry and exit on the design basis radiological 7 consequence dose analyses to ensure that the modification will not result in an increase in the 8 radiation dose consequences and that the resulting calculated radiation doses will remain within 9 the current radiological consequence analysesthe design criteria specified in 10 CFR 50.67 and 10 the accident specific design criteria outlined in RG 1.183. The NRC staff review of these DBAs 11 determined that there are two DBAs that take credit for the [secondary] containment, and are 12 possibly impacted by the brief, inadvertent, simultaneous opening of both an inner and outer 13 access door during normal entry and exit conditions, the loss-of-coolant accident (LOCA) and 14 the fuel handling accident (FHA) in [secondary] containment.
15 16 3.2.1 LOCA 17 18 Following a LOCA, the [secondary] containment structure is maintained at a negative pressure 19 ensuring that leakage from primary containment to [secondary] containment can be collected 20 and filtered prior to release to the environment. The SGT system performs the function of 21 maintaining a negative pressure within the [secondary] containment, as well as collecting and 22 filtering the leakage from primary containment. The SGT system is credited for mitigation of the 23 radiological releases from the [secondary] containment. In the LOCA analysis, the [secondary]
24 containment draw down analysis assumes that SGT system can draw down the [secondary]
25 containment within [5 minutes]. STS SR 3.6.4.1.4 requires one SGT subsystem to draw down 26 the [secondary] containment, to greater than or equal to [0.25] inches of vacuum water gauge in 27 a maximum allowable time of [120] seconds.
28 29 Conservatively, the DBA LOCA radiological consequence analysis in [UFSAR Chapter 15]
30 assumes that following the start of a DBA LOCA the [secondary] containment pressure of [0.25]
31 inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed that 32 releases into the [secondary] containment prior to the [10]-minute draw down time leak directly 33 to the environment as a ground level release with no filtration. After the assumed [10]-minute 34 draw down these releases are filtered by the SGT system and released via the SGT system 35 exhaust vent.
36 37 Based on this information, the NRC staff concludes that the DBA LOCA analysis has sufficient 38 conservatism by assuming a draw down time of [10] minutes from the start of the DBA LOCA.
39 Margin exists to ensure that the [secondary] containment can be reestablished during a brief, 40 inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 41 assurance that a failure of a safety system needed to control the release of radioactive material 42 to the environment will not result. The brief, inadvertent, simultaneous opening of the 43 secondary containment access doors does not impact the design bases and will not result in an 44 increase in any on-site or off-site dose.
45 46 Based on the above discussion, the NRC staff finds that the proposed change to the STSs does 47 not impact the design basis LOCA radiological consequence analysis, will not result in an 48 increase in any onsite or offsite dose, and is consistent with regulatory requirements and 49 guidance identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the
1 proposed change to the STSs will continue to comply with these criteria and that that the 2 estimates of the dose consequences of the postulated DBAs will comply with the current 3 radiological consequence analysesthe requirements of 10 CFR 50.67 and the accident specific 4 dose guidelines specified in RG 1.183. Therefore, the proposed changes are acceptable with 5 regard to the radiological consequences of the postulated DBAs.
6 7 3.2.2 FHA in [Secondary] Containment 8
9 During normal operation, non-safety related systems are used to maintain the [secondary]
10 containment at [0.25] inches of vacuum water gauge to ensure that any leakage is into the 11 building and that any [secondary] containment atmosphere exiting the building is via a 12 monitored pathway. The refueling floor, which is inside the [secondary] containment, is 13 maintained at a negative [0.25] inches of vacuum water gauge by normal operating ventilation 14 systems. The refueling floor exhaust ductwork in the [secondary] containment is equipped with 15 radiation monitors to detect a fuel handling accident. When a radiological release is sensed by 16 the radiation monitors, a [secondary] containment isolation signal is generated. This initiates 17 the SGT system and the normal ventilation system isolates. The radiation monitor is positioned 18 such that it will detect the release and send a closure signal to the [secondary] containment 19 isolation dampers.
20 21 Following a FHA, the [secondary] containment structure is maintained at a negative pressure by 22 the SGT system ensuring that fission products released from the spent fuel pool to [secondary]
23 containment can be collected and filtered prior to release to the environment. In the FHA 24 analysis, the [secondary] containment draw down analysis demonstrates that SGT system can 25 draw down the [secondary] containment within [5 minutes]. The SGT system is credited for 26 mitigation of the radiological releases from the [secondary] containment. STS SR 3.6.4.1.4 27 requires one SGT subsystem to draw down the [secondary] containment, to greater than or 28 equal to [0.25] inches of vacuum water gauge in a maximum allowable time of [120] seconds.
29 30 Conservatively, the DBA FHA radiological consequence analysis in [UFSAR Chapter 15]
31 assumes that following the start of a DBA FHA the [secondary] containment pressure of 32 [0.25] inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed 33 that releases into the [secondary] containment prior to the [10]-minute draw down time leak 34 directly to the environment as a ground level release with no filtration. After the assumed 35 [10]-minute draw down these releases are filtered by the SGT system and released via the SGT 36 system exhaust vent.
37 38 Based on this information, the NRC staff concludes that the DBA FHA analysis has sufficient 39 conservatism by assuming a draw down time of [10] minutes from the start of the DBA FHA.
40 Margin exists to ensure that the [secondary] containment can be reestablished during brief, 41 inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 42 assurance that a failure of a safety system needed to control the release of radioactive material 43 to the environment will not result. The brief, inadvertent, simultaneous opening of the 44 [secondary] containment access doors does not impact the design bases and will not result in 45 an increase in any on-site or off-site dose.
46 47 Based on the above discussion, the NRC staff finds that the proposed change to the STSs does 48 not impact the design basis FHA radiological consequence analysis, will not result in an increase 49 in any onsite or offsite dose, and is consistent with regulatory requirements and guidance
1 identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the proposed change 2 to the STSs will continue to comply with these criteria and that that the estimates of the dose 3 consequences of the postulated DBAs will comply with the current radiological consequence 4 analysesthe requirements of 10 CFR 50.67 and the accident specific dose guidelines specified 5 in RG 1.183. Therefore, the proposed changes are acceptable with regard to the radiological 6 consequences of the postulated DBAs.
7 8
3.3 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.4 9
10 The changes to SR 3.6.4.1.4 are editorial only and do not change any technical aspects of SR 11 3.6.4.1.4. The NRC staff determined that the change is acceptable.
12 13
4.0 CONCLUSION
14 15 The NRC staff reviewed traveler TSTF-551, Revision 3, which proposed changes to 16 NUREG-1433, Volumes 1 (STS) and 2 (Bases) and NUREG-1434 Volumes 1 (STS) and 2 17 (Bases). The NRC staff determined that the proposed changes to the STS met the standards 18 for TS in 10 CFR 50.36(b). The proposed SRs assure that the necessary quality of systems 19 and components is maintained, that facility operation will be within safety limits, and that the 20 LCOs will be met, and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the STS were 21 reviewed for technical clarity and consistency with customary terminology and format in 22 accordance with SRP Chapter 16.
23 24 The proposed bases, which will be added to future revisions to NUREG-1433, Volume 2, and 25 NUREG-1434, Volume 2, satisfy the Commissions Policy Statement by addressing the 26 questions specified in the policy statement, and cite references to appropriate licensing 27 documentation to support the Bases.
28 29 Additionally, the NRC staff has evaluated the impact of the proposed changes on the design 30 basis radiological consequence analyses against the regulatory requirements and guidance 31 identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the 32 changes to the STSs will continue to comply with the current radiological consequence 33 analysesrequirements of 10 CFR 50.67 and the guidelines specified in RG 1.183. Therefore, 34 the proposed changes are acceptable with regard to the radiological consequences of the 35 postulated DBAs.
36 37 38 Technical contacts: Kristy Bucholtz, NRR/DRA/ARCB 39 Nageswara Karipineni, NRR/DSS/SBPB 40 41
Attachment:
Basis for Accepting the Proposed Changes to the Standard Technical 42 Specification Bases, Volume 2 of NUREGs-1433 and -1434 43 44 Date:
45 46 47
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
1 DRAFT MODEL SAFETY EVALUATION 2 BY THE OFFICE OF NUCLEAR REACTOR REGULATION 3 TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 4 TSTF-551, REVISION 3, 5 REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS 6 [NOTE: Throughout this safety evaluation (SE), items that are enclosed in square brackets 7 signify plant-specific nomenclature or values to be taken from the licensee's submittal.]
8 9
1.0 INTRODUCTION
10 11 By application dated [enter date], (Agencywide Documents Access and Management System 12 (ADAMS) Accession No. [MLXXXXXXXXX], [name of licensee] (the licensee) requested 13 changes to the technical specifications (TS) for [name of facility]. Specifically, the licensee 14 requested changes to the TSs to adopt Technical Specifications Task Force (TSTF) traveler, 15 TSTF-551, Revision 3, Revise Secondary Containment Surveillance Requirements, dated 16 October 3, 2016 (ADAMS Accession No. ML16277A226). The NRC approved the traveler on 17 [Month, Day, 2017 (ADAMS Accession No. MLXXXX]).
18 19 The proposed changes would allow the [secondary] containment vacuum limit to not be met 20 provided that the standby gas treatment (SGT) system remains capable of establishing the 21 required [secondary] containment vacuum within the [specified time] and revises the TS to 22 permit [secondary] containment access opening to be open to permit ingress and egress.
23 24
2.0 REGULATORY EVALUATION
25 26 2.1 SYSTEM DESCRIPTION 27 28 The [secondary] containment is a structure that encloses the primary containment, including 29 components that may contain primary system fluid. The safety function of the [secondary]
30 containment is to contain, dilute, and hold up fission products that may leak from primary 31 containment following a design basis accident (DBA) to ensure the control room operator and 32 offsite doses are within the regulatory limits. There is no redundant train or system that can 33 perform the [secondary] containment function should the [secondary] containment be 34 inoperable.
35 36 The [secondary] containment boundary is the combination of walls, floor, roof, ducting, doors, 37 hatches, penetrations and equipment that physically form the [secondary] containment.
38 Routinely used [secondary] containment access openings contain at least one inner and one 39 outer door in an airlock configuration. In some cases, [secondary] containment access 40 openings are shared such that there are multiple inner or outer doors. All [secondary]
41 containment access doors are normally kept closed, except when the access opening is being 42 used for entry and exit of personnel, equipment, or material.
43 ENCLOSURE 2
1 [Secondary] containment operability is based on its ability to contain, dilute, and hold up fission 2 products that may leak from primary containment following a DBA. To prevent ground level 3 exfiltration of radioactive material while allowing the [secondary] containment to be designed 4 as a mostly conventional structure, the [secondary] containment requires support systems to 5 maintain the pressure at less than atmospheric pressure. During normal operation, non-safety 6 related systems are used to maintain the [secondary] containment at a slight negative pressure 7 to ensure any leakage is into the building and that any [secondary] containment atmosphere 8 exiting is via a pathway monitored for radioactive material. However, during normal operation it 9 is possible for the [secondary] containment vacuum to be momentarily less than the required 10 vacuum for a number of reasons, such as during wind gusts or swapping of the normal 11 ventilation subsystems.
12 13 During emergency conditions, the SGT system is designed to be capable of drawing down the 14 [secondary] containment to a required vacuum within a prescribed time and continue to 15 maintain the negative pressure as assumed in the accident analysis. For [name of facility], the 16 SGT must be able to establish the required vacuum within [insert time requirement]. The leak 17 tightness of the [secondary] containment together with the SGT system ensure that radioactive 18 material is either contained in the [secondary] containment or filtered through the SGT system 19 filter trains before being discharged to the outside environment via the elevated release point.
20 21 2.2 PROPOSED TECHNICAL SPECIFICATION CHANGES 22 23 The proposed changes would allow the [secondary] containment vacuum limit to not be met 24 provided the SGT system remains capable of establishing the required [secondary]
25 containment vacuum. The proposed changes would also allow for the temporary opening of the 26 inner and outer doors of [secondary] containment for the purpose of ingress and egress (i.e.,
27 normal opening and prompt closure of a door for transit).
28 29 2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 30 31 [NOTE: This change is applicable to all BWR types.NUREGs-1433 and -1434.]
32 33 Surveillance requirement (SR) 3.6.4.1.1 requires verification that [secondary] containment 34 vacuum is [0.25] inch of vacuum water gauge. This SR would be modified by a note that 35 states:
36 37 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 38 standby gas treatment (SGT) subsystem is capable of establishing 39 the required [secondary] containment vacuum.
40 41 2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 42 43 [NOTE: This change is applicable to BWR/2, BWR/3, BWR/4, and BWR/5 plants NUREG-1433 44 only.]
45 46 SR 3.6.4.1.3 requires verification that one [secondary] containment access door in each 47 access opening is closed. This SR would be modified by adding the following phrase to the end 48 of the SR statement, except when the access opening is being used for entry and exit.
49
1 2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 2
3 An editorial change is made to SR 3.6.4.1.4 in which the words standby gas treatment are 4 replaced with the initialism SGT.
5 6 2.3 REGULATORY REQUIREMENTS AND GUIDANCE 7
8 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) 9 requires an applicant for an operating license to include in the application proposed TS in 10 accordance with the requirements of 10 CFR 50.36. The applicant must include in the 11 application, a summary statement of the bases or reasons for such specifications, other than 12 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these technical 13 specification bases shall not become part of the technical specifications.
14 15 Additionally, 10 CFR 50.36(b) requires:
16 17 Each license authorizing operation of a utilization facility will 18 include technical specifications. The technical specifications will 19 be derived from the analyses and evaluation included in the safety 20 analysis report, and amendments thereto, submitted pursuant to 21 10 CFR 50.34 [Contents of applications; technical information].
22 The Commission may include such additional technical 23 specifications as the Commission finds appropriate.
24 25 The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required 26 by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are 27 the lowest functional capability or performance levels of equipment required for safe operation 28 of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the 29 licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the 30 condition can be met.
31 32 The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 33 which are requirements relating to test, calibration, or inspection to assure that the necessary 34 quality of systems and components is maintained, that facility operation will be within safety 35 limits, and that the LCOs will be met.
36 37 The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of 38 NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 39 Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425).
40 41 NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative 42 Source Terms, Revision 0, dated July 2000 (ADAMS Accession No. ML003734190), provides 43 guidance to the NRC staff for the review of alternate source term (AST) amendment requests.
44 SRP 15.0.1 states that the NRC reviewer should evaluate the proposed change against the 45 guidance in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for 46 Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000 47 (ADAMS Accession No. ML003716792).
48
1 RG 1.183 provides acceptable methodology for analyzing the radiological consequences of 2 several design basis accidents to show compliance with 10 CFR 50.67. RG 1.183 provides 3 guidance to licensees on acceptable application of AST (also known as the accident source 4 term) submittals, including acceptable radiological analysis assumptions for use in conjunction 5 with the accepted AST.
6 7 10 CFR 50.67, Accident source term, states that:
8 9 (i) An individual located at any point on the boundary of the 10 exclusion area for any 2-hour period following the onset of 11 the postulated fission product release, would not receive a 12 radiation dose in excess of 0.25 Sv (25 rem) total effective 13 dose equivalent (TEDE),
14 (ii) An individual located at any point on the outer boundary of 15 the low population zone, who is exposed to the radioactive 16 cloud resulting from the postulated fission product release 17 (during the entire period of its passage), would not receive 18 a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and 19 (iii) Adequate radiation protection is provided to permit access 20 to and occupancy of the control room under accident 21 conditions without personnel receiving radiation exposures 22 in excess of 0.05 Sv (5 rem) TEDE for the duration of the 23 accident.
24 25
3.0 TECHNICAL EVALUATION
26 27 The NRC staff evaluated the licensees application to determine if the proposed changes are 28 consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of 29 this safety evaluation (SE). In determining whether an amendment to a license will be issued, 30 the Commission is guided by the considerations that govern the issuance of initial licenses to 31 the extent applicable and appropriate. In making its determination as to whether to amend the 32 license, the NRC staff considered those regulatory requirements that are automatically 33 conditions of the license through 10 CFR 50.54.
34 35 The regulation at 10 CFR 50.36(a)(1) states, in part: A summary statement of the bases or 36 reasons for such specifications shall also be included in the application, but shall not become 37 part of the technical specifications. Accordingly, along with the proposed TS changes, the 38 licensee also submitted TS Bases changes that correspond to the proposed STS changes for 39 information only.
40 41
3.1 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.1 42 43 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 44 an analysis demonstrates that one SGT subsystem is capable of establishing the required 45 [secondary] containment vacuum. During normal operation, conditions may occur that result in 46 SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external 47 pressure or loss of the normal ventilation system that maintains [secondary] containment 48 vacuum may affect [secondary] containment vacuum. These conditions may not be indicative
1 of degradations of the [secondary] containment boundary or of the ability of the SGT system to 2 perform its specified safety function.
3 4 The note provides an allowance for the licensee to confirm [secondary] containment operability 5 by confirming that one SGT subsystem is capable of performing its specified safety function.
6 This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.
7 While the duration of these occurrences is anticipated to be very brief, the allowance is 8 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 9 [secondary] containment to be inoperable per [Condition A of LCO 3.6.4.1 or the corresponding 10 Condition for the plant-specific TS].
11 12 The NRC staff has evaluated the impact of this note on the licensees design basis radiological 13 consequence analyses to ensure that the proposed change will not result in an increase in the 14 dose consequences and that the resulting calculated doses remain within the current 15 radiological consequence analysesdesign criteria specified in 10 CFR 50.67 and the accident 16 specific design criteria outlined in RG 1.183.
17 18 The proposed addition of the note to SR 3.6.4.1.1 does not change the TS requirement to meet 19 SR 3.6.4.1.4 and SR 3.6.4.1.5. SR 3.6.4.1.4 requires verification that the [secondary]
20 containment can be drawn down to [0.25] inch of vacuum water gauge in [120] seconds 21 using one SGT subsystem. SR 3.6.4.1.5 requires verification that the [secondary] containment 22 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 23 a flow rate [4000] cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment 24 (SGT) System, must be met; otherwise the licensee shall shut down the reactor or follow any 25 remedial action permitted by TSs until the condition can be met.
26 27 As discussed above, [secondary] containment operability is based on its ability to contain, 28 dilute, and hold up fission products that may leak from primary containment following a DBA.
29 To prevent ground level exfiltration of radioactive material the [secondary] containment 30 pressure must be maintained at a pressure that is less than atmospheric pressure. The 31 [secondary] containment requires support systems to maintain the control volume pressure 32 less than atmospheric pressure. Following an accident, the SGT system ensures the 33 [secondary] containment pressure is less than the external atmospheric pressure. During 34 normal operation, non-safety related systems are used to maintain the [secondary]
35 containment at a negative pressure. However, during normal operation it is possible for the 36 [secondary] containment vacuum to be momentarily less than the required vacuum for a 37 number of reasons. These conditions may not be indicative of degradations of the [secondary]
38 containment boundary or of the ability of the SGT system to perform its specified safety 39 function. Since the licensee meets the requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the 40 LCO or is following the Actions of TS LCO 3.6.4.3, and the licensees analysis confirms 41 [secondary] containment operability by confirming that one SGT subsystem is capable of 42 performing its specified safety function, then there is reasonable assurance that the 43 [secondary] containment and SGT subsystem will maintain the vacuum requirements during a 44 DBA.
45 46 Therefore, the NRC staff has determined that: if the conditions do not affect (1) the ability to 47 maintain the [secondary] containment pressure during an accident, at a pressure that is less 48 than atmospheric, and (2) the time assumed in the accident analyses to draw down the 49 [secondary] containment pressure, then the [secondary] containment can perform its safety
1 function and may be considered TS operable. This is evident by being able to successfully 2 perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to 3 establish and maintain the required vacuum in the [secondary] containment as assumed in the 4 accident analyses.
5 6 Furthermore, because the specified safety functions of the [secondary] containment and SGT 7 subsystem can be performed in the time assumed in the licensees accident analysis, then the 8 fission products that bypass or leak from primary containment, or are released from the reactor 9 coolant pressure boundary components located in [secondary] containment prior to release to 10 the environment, will be contained and processed as assumed in the licensees design basis 11 radiological consequence dose analyses. The NRC staff finds that the proposed change does 12 not affect the current radiological consequence analyses and concludes that the proposed 13 change is acceptable with respect to the radiological consequences of DBAs.
14 15
3.2 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.3 16 17 [NOTE: The proposed change is not applicable if the radiological dose consequence analysis 18 assumes the [secondary] containment pressure is below atmospheric pressure prior to or 19 coincident with the time at which the accident or event occurs. Such an analysis assumption 20 would require a revised radiological dose consequence analysis considering the new release 21 point (the open [secondary] containment doors), with appropriate atmospheric dispersion 22 factors, and any other necessary revisions to the accident or event analysis.]
23 24 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, 25 inadvertent, simultaneous opening of redundant [secondary] containment access doors during 26 normal entry and exit conditions. Planned activities that could result in the simultaneous 27 opening of redundant [secondary] containment access openings, such as maintenance of a 28 [secondary] containment personnel access door or movement of large equipment through the 29 openings that would take longer than the normal transit time, will be considered outside the 30 scope of the NRC staff's review.
31 32 The NRC staff reviewed the changes to SR 3.6.4.1.3. The NRC staff determined that the SR 33 continues to provide appropriate confirmation that [secondary] containment boundary doors 34 are properly positioned and capable of performing their function in preserving the [secondary]
35 containment boundary. The NRC staff determined that the SRs continue to appropriately verify 36 the operability of the [secondary] containment and provide assurance that the necessary 37 quality of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
38 39 Additionally, the NRC staff evaluated the impact of modifying the TS to allow [secondary]
40 containment access openings to be open for entry and exit on the licensees design basis 41 radiological consequence dose analyses to ensure that the modification will not result in an 42 increase in the radiation dose consequences and that the resulting calculated radiation doses 43 will remain within the design criteria specified in the current radiological consequence 44 analyses10 CFR 50.67 and the accident specific design criteria outlined in RG 1.183. The NRC 45 staff review of these DBAs determined that there are two DBAs that take credit for the 46 [secondary] containment, and are possibly impacted by the brief, inadvertent, simultaneous 47 opening of both an inner and outer access door during normal entry and exit conditions, the loss 48 of coolant accident (LOCA) and the fuel handling accident (FHA) in [secondary] containment.
49
1 3.2.1 LOCA 2
3 Following a LOCA, the [secondary] containment structure is maintained at a negative pressure 4 ensuring that leakage from primary containment to [secondary] containment can be collected 5 and filtered prior to release to the environment. The SGT system performs the function of 6 maintaining a negative pressure within the [secondary] containment, as well as collecting and 7 filtering the leakage from primary containment. The licensee credits the SGT system for 8 mitigation of the radiological releases from the [secondary] containment. In the LOCA 9 analysis, the [secondary] containment draw down analysis assumes that SGT system can 10 draw down the [secondary] containment within [5 minutes]. TS SR 3.6.4.1.4 requires one 11 SGT subsystem to draw down the [secondary] containment, to greater than or equal to [0.25]
12 inches of vacuum water gauge in a maximum allowable time of [120] seconds.
13 14 Conservatively, the DBA LOCA radiological consequence analysis in [Updated Final Safety 15 Analysis Report (UFSAR) Chapter 15] assumes that following the start of a DBA LOCA the 16 [secondary] containment pressure of [0.25] inches of vacuum water gauge is achieved at 17 approximately [10] minutes. The license assumes that releases into the [secondary]
18 containment prior to the [10]-minute draw down time leak directly to the environment as a 19 ground level release with no filtration. After the assumed [10]-minute draw down these releases 20 are filtered by the SGT system and released via the SGT system exhaust vent.
21 22 Based on this information, the NRC staff concludes that the licensees DBA LOCA analysis has 23 sufficient conservatism by assuming a draw down time of [10] minutes from the start of the DBA 24 LOCA. Margin exists to ensure that the [secondary] containment can be reestablished during 25 a brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 26 assurance that a failure of a safety system needed to control the release of radioactive material 27 to the environment will not result. The brief, inadvertent, simultaneous opening of the 28 [secondary] containment access doors does not impact the design bases and will not result in 29 an increase in any on-site or off-site dose.
30 31 Based on the above discussion, the NRC staff finds that the licensees proposed change to the 32 TSs does not impact the licensees design basis LOCA radiological consequence analysis and 33 will not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that 34 this change is acceptable with respect to the radiological consequences of the DBAs.
35 36 [The licensee was approved for AST methodology and the radiological dose consequences 37 analyses for DBAs via license amendment[s] [insert license amendment number] for [name 38 of facility].] The NRC staff reviewed the impact of the proposed changes to [name of facility]
39 TS, on all DBAs currently analyzed in the [name of facility] Updated Final Safety Analysis 40 Report (UFSAR) that could have the potential for significant dose consequences. [Chapter 15]
41 of the [name of facility] UFSAR describes the DBAs and their radiological consequence 42 analysis results.]
43 44 3.2.2 FHA in [Secondary] Containment 45 46 During normal operation, non-safety related systems are used to maintain the [secondary]
47 containment at [0.25] inches of vacuum water gauge to ensure that any leakage is into the 48 building and that any [secondary] containment atmosphere exiting the building is via a 49 monitored pathway. The refuel floor, which is inside the [secondary] containment, is
1 maintained at a negative [0.25] inches of vacuum water gauge by normal operating ventilation 2 systems. The refueling floor exhaust ductwork in the [secondary] containment is equipped with 3 radiation monitors to detect a fuel handling accident. When a radiological release is sensed by 4 the radiation monitors, a [secondary] containment isolation signal is generated. This initiates 5 the SGT system and the normal ventilation system isolates. The radiation monitor is positioned 6 such that it will detect the release and send a closure signal to the [secondary] containment 7 isolation dampers.
8 9 Following a FHA, the [secondary] containment structure is maintained at a negative pressure 10 by the SGT system ensuring that fission products released from the spent fuel pool to 11 [secondary] containment can be collected and filtered prior to release to the environment. In 12 the FHA analysis, the [secondary] containment draw down analysis demonstrates that SGT 13 system can draw down the [secondary] containment within [5 minutes]. The licensee credits 14 the SGT system for mitigation of the radiological releases from the [secondary] containment.
15 TS SR 3.6.4.1.4 requires one SGT subsystem to draw down the [secondary] containment, to 16 greater than or equal to [0.25] inches of vacuum water gauge in a maximum allowable time of 17 [120] seconds.
18 19 Conservatively, the DBA FHA radiological consequence analysis in [UFSAR Chapter 15]
20 assumes that following the start of a DBA FHA the [secondary] containment pressure of 21 [0.25] inches of vacuum water gauge is achieved at approximately [10] minutes. The license 22 assumes that releases into the [secondary] containment prior to the [10]-minute draw down 23 time leak directly to the environment as a ground level release with no filtration. After the 24 assumed [10]-minute draw down these releases are filtered by the SGT system and released 25 via the SGT system exhaust vent.
26 27 Based on this information, the NRC staff concludes that the licensees DBA FHA analysis has 28 sufficient conservatism by assuming a draw down time of [10] minutes from the start of the DBA 29 FHA. Margin exists to ensure that the [secondary] containment can be reestablished during 30 brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 31 assurance that a failure of a safety system needed to control the release of radioactive material 32 to the environment will not result. The brief, inadvertent, simultaneous opening of the 33 [secondary] containment access doors does not impact the design bases and will not result in 34 an increase in any on-site or off-site dose.
35 36 Based on the above discussion, the NRC staff finds that the licensees proposed change to the 37 TSs does not impact the licensees design basis FHA radiological consequence analysis and will 38 not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that this 39 change is acceptable with respect to the radiological consequences of the DBAs.
40 41 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, 42 inadvertent, simultaneous opening of redundant [secondary] containment access doors during 43 normal entry and exit conditions. Planned activities that could result in the simultaneous 44 opening of redundant [secondary] containment access openings, such as maintenance of a 45 [secondary] containment personnel access door or movement of large equipment through the 46 openings that would take longer than the normal transit time, will be considered outside the 47 scope of the NRC staff's review.
48 49 3.2.3 Conclusion
1 2 As described above, the NRC staff reviewed the technical basis provided by the licensee to 3 assess the radiological impacts of the changes to the [secondary] containment in the licensees 4 TSs. The NRC staff finds that the licensee proposed change to SR 3.6.4.1.3 is consistent with 5 regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds, 6 with reasonable assurance that the licensees change to the TSs will continue to comply with 7 these criteria and that that the licensee's estimates of the dose consequences of a design basis 8 LOCA and FHA will comply with the requirements of the current radiological consequence 9 analyses10 CFR 50.67 and the accident specific dose guidelines specified in RG 1.183.
10 Therefore, the proposed changes are acceptable with regard to the radiological consequences 11 of the postulated DBAs.
12 13
3.3 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.4 14 15 The changes to SR 3.6.4.1.4 are editorial only and do not change any technical aspects of 16 SR 3.6.4.1.4. The NRC staff determined that the change is acceptable.
17 18 3.4 VARIATIONS FROM THE APPROVED TRAVELER 19 20 [NOTE: Technical reviewers and/or project manager to discuss variations from the approved 21 traveler and whether they are acceptable. Choose the applicable paragraphs based on 22 information provided in the LAR.]
23 24 [The licensee is not proposing any variations from the TS changes described in TSTF-551 or 25 the applicable parts of the NRC staffs safety evaluation of TSTF-551.]
26 27 [The licensee is proposing the following variations from the TS changes described in TSTF-551 28 or the applicable parts of TSTF-551 or the NRC staffs safety evaluation. These variations do 29 not affect the applicability of TSTF-551 or the NRC staff's safety evaluation to the proposed 30 license amendment.]
31 32 [The [PLANT] TS do not contain an SR equivalent to SR 3.6.4.1.1 modified by TSTF-551.
33 Therefore, the addition of the SR 3.6.4.1.1 Note is not applicable.]
34 35 [The [PLANT] TS already contains an allowance similar to that made to SR 3.6.4.1.3.
36 Therefore, the proposed change does not contain this portion of TSTF-551.]
37 38 [The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical 39 Specifications on which TSTF-551 was based. Specifically, [describe differences between the 40 plant-specific TS numbering and/or titles and the TSTF-551 numbering and titles.] These 41 differences are administrative and do not affect the applicability of TSTF-546 to the [PLANT]
42 TS.]
43 44 [The Traveler discusses the applicable regulatory requirements and guidance, including the 45 10 CFR 50, Appendix A, General Design Criteria (GDC). [PLANT] was not licensed to the 46 10 CFR 50, Appendix A, GDC. The [PLANT] equivalents of the referenced GDC are [discussion 47 from licensee's application.] These differences do not alter the conclusion that the proposed 48 change is applicable to [PLANT].]
49
1 3.54
SUMMARY
2 3 The NRC staff reviewed the proposed changes and determined that changes to the TS meet the 4 standards for TS in 10 CFR 50.36(b). The proposed SRs assure that the necessary quality of 5 systems and components is maintained, that facility operation will be within safety limits, and 6 that the LCOs will be met, and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the TS 7 were reviewed for technical clarity and consistency with customary terminology and format in 8 accordance with SRP Chapter 16.
9 10 Additionally, the NRC staff has evaluated the impact of the proposed changes on the design 11 basis radiological consequence analyses against the regulatory requirements and guidance 12 identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the 13 licensees change to the TSs will continue to comply with the requirements of the current 14 radiological consequence analyses10 CFR 50.67 and the guidelines specified in RG 1.183.
15 Therefore, the proposed changes are acceptable with regard to the radiological consequences 16 of the postulated DBAs.
17 18
4.0 STATE CONSULTATION
19 20 In accordance with the Commission's regulations, the [Name of State] State official was notified 21 of the proposed issuance of the amendment on [enter date]. The State official had [no]
22 comments. [If comments were provided, they should be addressed here].
23 24
5.0 ENVIRONMENTAL CONSIDERATION
25 26 [Note: This section is to be prepared by the PM. As needed, the PM should coordinate with 27 NRRs Environmental Review and Projects Branch (RERP) to determine the need for an EA.
28 Specific guidance on preparing EAs and considering environmental issues is contained in NRR 29 Office Instruction LIC-203, Procedural Guidance for Preparing Categorical Exclusions, 30 Environmental Assessments, and Considering Environmental Issues.]
31 32 The amendment changes requirements with respect to the installation or use of facility 33 components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
34 The NRC staff has determined that the amendment involves no significant increase in the 35 amounts and no significant change in the types of any effluents that may be released offsite, 36 and that there is no significant increase in individual or cumulative occupational radiation 37 exposure. The Commission has previously issued a proposed finding that the amendment 38 involves no significant hazards consideration, and there has been no public comment on such 39 finding published in the Federal Register on [DATE (XX FR XXX)]. Accordingly, the 40 amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
41 Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment 42 need be prepared in connection with the issuance of the amendment.
43 44
6.0 CONCLUSION
45 46 The Commission has concluded, based on the considerations discussed above, that: (1) there 47 is reasonable assurance that the health and safety of the public will not be endangered by 48 operation in the proposed manner, (2) there is reasonable assurance that such activities will be 49 conducted in compliance with the Commission's regulations, and (3) the issuance of the
1 amendment will not be inimical to the common defense and security or to the health and safety 2 of the public 3
4 Principal Contributors: Kristy Bucholtz, NRR/DRA/ARCB 5 Nageswara Karipineni, NRR/DSS/SBPB 6 Margaret Chernoff, NRR/DSS/STSB 7
8 Date:
9