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
11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation TECHNICAL SPECIFICATIONS TASK FORCE A JOINT OWNERS GROUP ACTIVITY TSTF 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. 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)
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
Page 3 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 2
27 45 In some locations, the SE refers to the SR 3.6.4.1.3 note as 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."
5 6
8 9
10 11-46 23-25 3-4, 46 46-47 25-26 The SE discusses only radiological consequence analysis using alternative source term (AST), 10 CFR 50.67, and Regulatory Guide 1.183.
TSTF-551 is applicable to all BWR plants regardless of whether the Loss of Coolant Accident and Fuel Handling Accident analyses are based on 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 2
27 45 In some locations, the model SE refers to the SR 3.6.4.1.3 note as 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 4 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 2
29 41 In the model SE, the TSTF recommends not referring to the Standard Technical Specifications NUREGs, but to instead refer to the applicable BWR types.
3 4
5 6
9 9
39-49 1-20 12 39-40 4-5 25-26 The model SE discusses only radiological consequence analysis using alternative source term (AST), 10 CFR 50.67, and Regulatory Guide 1.183. TSTF-551 is applicable to all BWR plants regardless of whether the Loss of Coolant Accident and Fuel Handling Accident analyses are based on 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."
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.
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 KHsueh RDennig JDanna DATE 3/21/2017 6/5/17 12/6/2016 5/19/2017 OFFICE OGC*
DSS/STSB DSS/STSB*
NAME BHarris MHoncharik JMWhitman DATE 6/29/17 6/29/17 7/2017
Technical Specifications Task Force Project No. 753 cc:
Technical Specifications Task Force c/o EXCEL Services Corporation 11921 Rockville Pike, Suite 100 Rockville, MD 20852 Attention: Brian D. Mann E-mail: brian.mann@excelservices.com James R. Morris Diablo Canyon Power Plant Building 104/5/21A P.O. Box 56 Avila Beach, CA 93424 E-mail: james.morris@pge.com Lisa L. Williams Energy Northwest Columbia Generating Station PO Box 968 Mail Drop PE20 Richland, WA 99352-0968 E-mail: llwilliams@energy-northwest.com Otto W. Gustafson Entergy Nuclear Operations, Inc.
Palisades Nuclear Power Plant 27780 Blue Star Memorial Highway Covert, MI 49043 E-mail: ogustaf@entergy.com Jordan L. Vaughan Duke Energy EC2ZF / P.O. Box 1006 Charlotte, NC 28202 Email: jordan.vaughan@duke-energy.com Jason P. Redd Southern Nuclear Operating Company 42 Inverness Center Parkway Bin B234 Birmingham, AL 35242-4809 E-mail: jpredd@southernco.com
ENCLOSURE 1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 2
TSTF-551, REVISION 3, 3
REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS 4
5
1.0 INTRODUCTION
6 7
By letter dated September 3, 2015 (Agencywide Document Access and Management System 8
(ADAMS) Accession No. ML14304A034 ML15246A131), the Technical Specifications (TS) Task 9
Force (TSTF) submitted traveler TSTF-551, Reactor Pressure Vessel Water Inventory Control, 10 Revision 0, for U.S. Nuclear Regulatory Commission (NRC) review and approval. By letter 11 dated January 26, 2016, the TSTF submitted Revision 1 to traveler TSTF-551 (ADAMS 12 Accession No. ML15246A131ML16026A026), and by letter dated May 12, 2016, the TSTF 13 submitted Revision 2 to the traveler (ADAMS Accession No. ML16133A536). By letter dated 14 October 3, 2016 (ADAMS Accession No. ML16277A226), the TSTF submitted Revision 3 of the 15 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]2 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 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.
access opening to be open to permit entry and exit ingress and egress similar to the 1
corresponding statements in NUREG-1434.
2 3
Throughout this safety evaluation (SE), items that are enclosed in square brackets signify 4
plant-specific nomenclature or values. Individual licensees would furnish plant-specific 5
nomenclature or values for bracketed items when submitting a license amendment request 6
(LAR) to adopt the changes described in this SE.
7 8
2.0 REGULATORY EVALUATION
9 10 2.1 SYSTEM DESCRIPTION 11 12 The [secondary] containment is a structure that encloses the primary containment, including 13 components that may contain primary system fluid. The safety function of the [secondary]
14 containment is to contain, dilute, and hold up fission products that may leak from primary 15 containment following a design basis accident (DBA) to ensure the control room operator and 16 offsite doses are within the regulatory limits. There is no redundant train or system that can 17 perform the [secondary] containment function should the [secondary] containment be 18 inoperable.
19 20 The [secondary] containment boundary is the combination of walls, floor, roof, ducting, doors, 21 hatches, penetrations and equipment that physically form the [secondary] containment. A 22 routinely used [secondary] containment access opening contains at least one inner and one 23 outer door in an airlock configuration. In some cases, [secondary] containment access 24 openings are shared such that there are multiple inner or outer doors. All [secondary]
25 containment access doors are normally kept closed, except when the access opening is being 26 used for entry and exit of personnel, equipment, or material.
27 28
[Secondary] containment operability is based on its ability to contain, dilute, and hold up fission 29 products that may leak from primary containment following a DBA. To prevent ground level 30 exfiltration of radioactive material while allowing the [secondary] containment to be designed as 31 a mostly conventional structure, the [secondary] containment requires support systems to 32 maintain the pressure at less than atmospheric pressure. During normal operation, non-safety 33 related systems are used to maintain the [secondary] containment at a slight negative pressure 34 to ensure any leakage is into the building and that any [secondary] containment atmosphere 35 exiting is via a pathway monitored for radioactive material. However, during normal operation it 36 is possible for the [secondary] containment vacuum to be momentarily less than the required 37 vacuum for a number of reasons, such as during wind gusts or swapping of the normal 38 ventilation subsystems.
39 40 During emergency conditions, the SGT system is designed to be capable of drawing down the 41
[secondary] containment to a required vacuum within a prescribed time and continue to maintain 42 the negative pressure as assumed in the accident analysis. The leak tightness of the 43
[secondary] containment together with the SGT system ensure that radioactive material is either 44 contained in the [secondary] containment or filtered through the SGT system filter trains before 45 being discharged to the outside environment via the elevated release point.
46 47 2.2 CHANGES TO THE STS 48 49 The proposed changes would allow the [secondary] containment vacuum limit to not be met 1
provided the SGT system remains capable of establishing the required [secondary] containment 2
vacuum and revises NUREG-1433 to permit [secondary] containment access opening to be 3
open to permit entry and exit ingress and egress similar to the corresponding statements in 4
5 6
Corresponding changes are proposed to the STS Bases. A summary of the revised STS Bases 7
and the NRC staffs evaluation of the revised Bases are provided in an attachment to this SE.
8 9
2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 10 11 Surveillance requirement (SR) 3.6.4.1.1 requires verification that [secondary] containment 12 vacuum is [0.25] inch of vacuum water gauge. This SR would be modified by a note that 13 states:
14 15 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 16 standby gas treatment (SGT) subsystem is capable of establishing 17 the required [secondary] containment vacuum.
18 19 This change is applicable to NUREG-1433 and -1434.
20 21 2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 22 23 SR 3.6.4.1.3 requires verification that one [secondary] containment access door in each access 24 opening is closed. This SR would be modified by adding the following phrase to the end of the 25 SR statement, except when the access opening is being used for entry and exit.
26 27 This change is applicable to NUREG-1433 only. This provision already exists in NUREG-1434, 28 Revision 4.
29 30 2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 31 32 An editorial change is made to SR 3.6.4.1.4 in which the words standby gas treatment are 33 replaced with the initialism SGT.
34 35 2.3 APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 36 37 Section IV, The Commission Policy, of the Final Policy Statement on Technical Specifications 38 Improvements for Nuclear Power Reactors (58 Federal Register 39132), dated July 22, 1993, 39 states in part:
40 41 The purpose of Technical Specifications is to impose those 42 conditions or limitations upon reactor operation necessary to 43 obviate the possibility of an abnormal situation or event giving rise 44 to an immediate threat to the public health and safety by 45 identifying those features that are of controlling importance to 46 safety and establishing on them certain conditions of operation 47 which cannot be changed without prior Commission approval.
48 49
[T]he Commission will also entertain requests to adopt portions of 1
the improved STS [(e.g., TSTF-551)], even if the licensee does 2
not adopt all STS improvements 3
4 The Commission encourages all licensees who submit Technical 5
Specification related submittals based on this Policy Statement to 6
emphasize human factors principles 7
8 In accordance with this Policy Statement, improved STS have 9
been developed and will be maintained for [BWR designs]. The 10 Commission encourages licensees to use the STS as the basis for 11 plant-specific Technical Specifications 12 13
[I]t is the Commission intent that the wording and Bases of the 14 improved STS be used [] to the extent practicable.
15 16 As described in the Commissions Final Policy Statement on Technical Specifications 17 Improvements for Nuclear Power Reactors, recommendations were made by NRC and industry 18 task groups for new STS that include greater emphasis on human factors principles in order to 19 add clarity and understanding to the text of the STS, and provide improvements to the Bases of 20 STS, which provides the purpose for each requirement in the specification. Subsequently, 21 improved vendor-specific STS were developed and issued by the NRC in September 1992.
22 23 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) 24 requires an applicant for an operating license to include in the application proposed TS in 25 accordance with the requirements of 10 CFR 50.36. The applicant must include in the 26 application, a summary statement of the bases or reasons for such specifications, other than 27 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these technical 28 specification bases shall not become part of the technical specifications.
29 30 Additionally, 10 CFR 50.36(b) requires:
31 32 Each license authorizing operation of a utilization facility will 33 include technical specifications. The technical specifications will 34 be derived from the analyses and evaluation included in the safety 35 analysis report, and amendments thereto, submitted pursuant to 36 10 CFR 50.34 [Contents of applications; technical information].
37 The Commission may include such additional technical 38 specifications as the Commission finds appropriate.
39 40 The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required 41 by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are 42 the lowest functional capability or performance levels of equipment required for safe operation 43 of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the 44 licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the 45 condition can be met.
46 47 The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 48 which are requirements relating to test, calibration, or inspection to assure that the necessary 49 quality of systems and components is maintained, that facility operation will be within safety 1
limits, and that the LCOs will be met.
2 3
Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for 4
an amendment must be filed with the Commission, fully describing the changes desired, and 5
following as far as applicable, the form prescribed for original applications.
6 7
Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the 8
applicant, the Commission will be guided by the considerations which govern the issuance of 9
initial licenses to the extent applicable and appropriate.
10 11 The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of 12 NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 13 Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425). As 14 described therein, as part of the regulatory standardization effort, the NRC staff has prepared 15 STS for each of the light-water reactor nuclear designs.
16 17 NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative 18 Source Terms, Revision 0, dated July 2000, provides guidance to the NRC staff for the review 19 of AST amendment requests. SRP 15.0.1 states that the NRC reviewer should evaluate the 20 proposed change against the guidance in RG 1.183.
21 22 Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design 23 Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000, provides acceptable 24 methodology for analyzing the radiological consequences of several design basis accidents to 25 show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable 26 application of alternate source term (AST) (also known as the accident source term) submittals, 27 including acceptable radiological analysis assumptions for use in conjunction with the accepted 28 AST.
29 30 10 CFR 50.67, Accident source term, states that:
31 32 (i)
An individual located at any point on the boundary of the 33 exclusion area for any 2-hour period following the onset of 34 the postulated fission product release, would not receive a 35 radiation dose in excess of 0.25 Sv (25 rem) total effective 36 dose equivalent (TEDE),
37 (ii)
An individual located at any point on the outer boundary of 38 the low population zone, who is exposed to the radioactive 39 cloud resulting from the postulated fission product release 40 (during the entire period of its passage), would not receive 41 a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and 42 (iii)
Adequate radiation protection is provided to permit access 43 to and occupancy of the control room under accident 44 conditions without personnel receiving radiation exposures 45 in excess of 0.05 Sv (5 rem) TEDE for the duration of the 46 accident.
47 48 In the evaluation of plant-specific LARs adopt TSTF-551 changes, the NRC staff will confirm the 1
current licensing basis, which reflects the AST methodology for analyzing the radiological 2
consequences of the design basis accidents using RG 1.183. The NRC staff will also consider 3
relevant information in the updated Final Safety Analysis Report (FSAR), which describes the 4
DBAs and evaluation of their radiological consequences for a specific licensee.
5 6
3.0 TECHNICAL EVALUATION
7 8
3.1 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.1 9
10 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 11 an analysis demonstrates that one SGT subsystem is capable of establishing the required 12
[secondary] containment vacuum. During normal operation, conditions may occur that result in 13 SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external 14 pressure or loss of the normal ventilation system that maintains [secondary] containment 15 vacuum may affect [secondary] containment vacuum. These conditions may not be indicative of 16 degradations of the [secondary] containment boundary or of the ability of the SGT system to 17 perform its specified safety function.
18 19 The note provides an allowance for the licensee to confirm [secondary] containment operability 20 by confirming that one SGT subsystem is capable of performing its specified safety function.
21 This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.
22 While the duration of these occurrences is anticipated to be very brief, the allowance is 23 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]
24 containment to be inoperable per Condition A of LCO 3.6.4.1.
25 26 The NRC staff intends to evaluate the impact of this note on the licensees design basis 27 radiological consequence dose analyses to ensure that the proposed change will not result in an 28 increase in the dose consequences and that the resulting calculated doses remain within the 29 current radiological consequence analysesthe design criteria specified in 10 CFR 50.67 and the 30 accident specific design criteria outlined in RG 1.183.
31 32 The proposed addition of the note to SR 3.6.4.1.1 does not change the STS requirement to 33 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]
34 containment can be drawn down to [0.25] inch of vacuum water gauge in [120] seconds 35 using one SGT subsystem. SR 3.6.4.1.5 requires verification that the [secondary] containment 36 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 37 a flow rate [4000] cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment 38 (SGT) System, must be met; otherwise a licensee shall shut down the reactor or follow any 39 remedial action permitted by STS until the condition can be met.
40 41 As discussed above, [secondary] containment operability is based on its ability to contain, dilute, 42 and hold up fission products that may leak from primary containment following a DBA. To 43 prevent ground level exfiltration of radioactive material the [secondary] containment pressure 44 must be maintained at a pressure that is less than atmospheric pressure. The [secondary]
45 containment requires support systems to maintain the control volume pressure less than 46 atmospheric pressure. Following an accident, the SGT system ensures the [secondary]
47 containment pressure is less than the external atmospheric pressure. During normal operation, 48 non-safety related systems are used to maintain the [secondary] containment at a negative 49 pressure. However, during normal operation it is possible for the [secondary] containment 1
vacuum to be momentarily less than the required vacuum for a number of reasons. These 2
conditions are not indicative of degradations of the [secondary] containment boundary or of the 3
ability of the SGT system to perform its specified safety function. Since the licensee meets the 4
requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the LCO or is following the Actions of TS 5
LCO 3.6.4.3, and the licensees analysis confirms [secondary] containment operability by 6
confirming that one SGT subsystem is capable of performing its specified safety function, then 7
there is reasonable assurance that the [secondary] containment and SGT subsystem will 8
maintain the vacuum requirements during a DBA.
9 10 Therefore, the NRC staff has determined that: if the conditions do not affect (1) the ability to 11 maintain the [secondary] containment pressure during an accident, at a pressure that is less 12 than atmospheric, and (2) the time assumed in the accident analyses to draw down the 13
[secondary] containment pressure, then the [secondary] containment can perform its safety 14 function and may be considered TS operable. This is evident by being able to successfully 15 perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to 16 establish and maintain the required vacuum in the [secondary] containment as assumed in the 17 accident analyses.
18 19 If the specified safety functions of the [secondary] containment and SGT subsystem can be 20 performed in the time assumed in the accident analysis, then the fission products that bypass or 21 leak from primary containment, or are released from the reactor coolant pressure boundary 22 components located in [secondary] containment prior to release to the environment, will be 23 contained and processed as assumed in the design basis radiological consequence dose 24 analyses. If the above statement is true for a plant-specific amendment, then the NRC staff 25 finds that the proposed change does not affect the current radiological consequence analyses.
26 Therefore, the NRC staff concludes this change is acceptable with respect to the radiological 27 consequences of DBAs.
28 29
3.2 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.3 30 31
[NOTE: The proposed change is not applicable if the radiological dose consequence analysis 32 assumes the [secondary] containment pressure is below atmospheric pressure prior to or 33 coincident with the time at which the accident or event occurs. Such an analysis assumption 34 would require a revised radiological dose consequence analysis considering the new release 35 point (the open [secondary] containment doors), with appropriate atmospheric dispersion 36 factors, and any other necessary revisions to the accident or event analysis.]
37 38 The NRC staff review of SR 3.6.4.1.3 was limited to the request to provide an allowance for the 39 brief, inadvertent, simultaneous opening of redundant [secondary] containment access doors 40 during normal entry and exit conditions. Planned activities that could result in the simultaneous 41 opening of redundant [secondary] containment access openings, such as maintenance of a 42
[secondary] containment personnel access door or movement of large equipment through the 43 openings that would take longer than the normal transit time, will be considered outside the 44 scope of the NRC staff's review.
45 46 The NRC staff reviewed the changes to SR 3.6.4.1.3. The NRC staff determined that the SR 47 continues to provide appropriate confirmation that [secondary] containment boundary doors are 48 properly positioned and capable of performing their function in preserving the [secondary]
49 containment boundary. The NRC staff determined that the SRs continue to appropriately verify 1
the operability of the [secondary] containment and provide assurance that the necessary quality 2
of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
3 4
Additionally, the NRC staff evaluated the impact of modifying STS to allow [secondary]
5 containment access openings to be open for entry and exit on the design basis radiological 6
consequence dose analyses to ensure that the modification will not result in an increase in the 7
radiation dose consequences and that the resulting calculated radiation doses will remain within 8
the current radiological consequence analysesthe design criteria specified in 10 CFR 50.67 and 9
the accident specific design criteria outlined in RG 1.183. The NRC staff review of these DBAs 10 determined that there are two DBAs that take credit for the [secondary] containment, and are 11 possibly impacted by the brief, inadvertent, simultaneous opening of both an inner and outer 12 access door during normal entry and exit conditions, the loss-of-coolant accident (LOCA) and 13 the fuel handling accident (FHA) in [secondary] containment.
14 15 3.2.1 LOCA 16 17 Following a LOCA, the [secondary] containment structure is maintained at a negative pressure 18 ensuring that leakage from primary containment to [secondary] containment can be collected 19 and filtered prior to release to the environment. The SGT system performs the function of 20 maintaining a negative pressure within the [secondary] containment, as well as collecting and 21 filtering the leakage from primary containment. The SGT system is credited for mitigation of the 22 radiological releases from the [secondary] containment. In the LOCA analysis, the [secondary]
23 containment draw down analysis assumes that SGT system can draw down the [secondary]
24 containment within [5 minutes]. STS SR 3.6.4.1.4 requires one SGT subsystem to draw down 25 the [secondary] containment, to greater than or equal to [0.25] inches of vacuum water gauge in 26 a maximum allowable time of [120] seconds.
27 28 Conservatively, the DBA LOCA radiological consequence analysis in [UFSAR Chapter 15]
29 assumes that following the start of a DBA LOCA the [secondary] containment pressure of [0.25]
30 inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed that 31 releases into the [secondary] containment prior to the [10]-minute draw down time leak directly 32 to the environment as a ground level release with no filtration. After the assumed [10]-minute 33 draw down these releases are filtered by the SGT system and released via the SGT system 34 exhaust vent.
35 36 Based on this information, the NRC staff concludes that the DBA LOCA analysis has sufficient 37 conservatism by assuming a draw down time of [10] minutes from the start of the DBA LOCA.
38 Margin exists to ensure that the [secondary] containment can be reestablished during a brief, 39 inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 40 assurance that a failure of a safety system needed to control the release of radioactive material 41 to the environment will not result. The brief, inadvertent, simultaneous opening of the 42 secondary containment access doors does not impact the design bases and will not result in an 43 increase in any on-site or off-site dose.
44 45 Based on the above discussion, the NRC staff finds that the proposed change to the STSs does 46 not impact the design basis LOCA radiological consequence analysis, will not result in an 47 increase in any onsite or offsite dose, and is consistent with regulatory requirements and 48 guidance identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the 49 proposed change to the STSs will continue to comply with these criteria and that that the 1
estimates of the dose consequences of the postulated DBAs will comply with the current 2
radiological consequence analysesthe requirements of 10 CFR 50.67 and the accident specific 3
dose guidelines specified in RG 1.183. Therefore, the proposed changes are acceptable with 4
regard to the radiological consequences of the postulated DBAs.
5 6
3.2.2 FHA in [Secondary] Containment 7
8 During normal operation, non-safety related systems are used to maintain the [secondary]
9 containment at [0.25] inches of vacuum water gauge to ensure that any leakage is into the 10 building and that any [secondary] containment atmosphere exiting the building is via a 11 monitored pathway. The refueling floor, which is inside the [secondary] containment, is 12 maintained at a negative [0.25] inches of vacuum water gauge by normal operating ventilation 13 systems. The refueling floor exhaust ductwork in the [secondary] containment is equipped with 14 radiation monitors to detect a fuel handling accident. When a radiological release is sensed by 15 the radiation monitors, a [secondary] containment isolation signal is generated. This initiates 16 the SGT system and the normal ventilation system isolates. The radiation monitor is positioned 17 such that it will detect the release and send a closure signal to the [secondary] containment 18 isolation dampers.
19 20 Following a FHA, the [secondary] containment structure is maintained at a negative pressure by 21 the SGT system ensuring that fission products released from the spent fuel pool to [secondary]
22 containment can be collected and filtered prior to release to the environment. In the FHA 23 analysis, the [secondary] containment draw down analysis demonstrates that SGT system can 24 draw down the [secondary] containment within [5 minutes]. The SGT system is credited for 25 mitigation of the radiological releases from the [secondary] containment. STS SR 3.6.4.1.4 26 requires one SGT subsystem to draw down the [secondary] containment, to greater than or 27 equal to [0.25] inches of vacuum water gauge in a maximum allowable time of [120] seconds.
28 29 Conservatively, the DBA FHA radiological consequence analysis in [UFSAR Chapter 15]
30 assumes that following the start of a DBA FHA the [secondary] containment pressure of 31
[0.25] inches of vacuum water gauge is achieved at approximately [10] minutes. It is assumed 32 that releases into the [secondary] containment prior to the [10]-minute draw down time leak 33 directly to the environment as a ground level release with no filtration. After the assumed 34
[10]-minute draw down these releases are filtered by the SGT system and released via the SGT 35 system exhaust vent.
36 37 Based on this information, the NRC staff concludes that the DBA FHA analysis has sufficient 38 conservatism by assuming a draw down time of [10] minutes from the start of the DBA FHA.
39 Margin exists to ensure that the [secondary] containment can be reestablished during 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 44 an 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 FHA radiological consequence analysis, will not result in an increase 48 in any onsite or offsite dose, and is consistent with regulatory requirements and guidance 49 identified in Section 2.3 of this safety evaluation. The NRC staff finds, that the proposed change 1
to the STSs will continue to comply with these criteria and that that the estimates of the dose 2
consequences of the postulated DBAs will comply with the current radiological consequence 3
analysesthe requirements of 10 CFR 50.67 and the accident specific dose guidelines specified 4
in RG 1.183. Therefore, the proposed changes are acceptable with regard to the radiological 5
consequences of the postulated DBAs.
6 7
3.3 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.4 8
9 The changes to SR 3.6.4.1.4 are editorial only and do not change any technical aspects of SR 10 3.6.4.1.4. The NRC staff determined that the change is acceptable.
11 12
4.0 CONCLUSION
13 14 The NRC staff reviewed traveler TSTF-551, Revision 3, which proposed changes to 15 NUREG-1433, Volumes 1 (STS) and 2 (Bases) and NUREG-1434 Volumes 1 (STS) and 2 16 (Bases). The NRC staff determined that the proposed changes to the STS met the standards 17 for TS in 10 CFR 50.36(b). The proposed SRs assure that the necessary quality of systems 18 and components is maintained, that facility operation will be within safety limits, and that the 19 LCOs will be met, and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the STS were 20 reviewed for technical clarity and consistency with customary terminology and format in 21 accordance with SRP Chapter 16.
22 23 The proposed bases, which will be added to future revisions to NUREG-1433, Volume 2, and 24 NUREG-1434, Volume 2, satisfy the Commissions Policy Statement by addressing the 25 questions specified in the policy statement, and cite references to appropriate licensing 26 documentation to support the Bases.
27 28 Additionally, the NRC staff has evaluated the impact of the proposed changes on the design 29 basis radiological consequence analyses against the regulatory requirements and guidance 30 identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the 31 changes to the STSs will continue to comply with the current radiological consequence 32 analysesrequirements of 10 CFR 50.67 and the guidelines specified in RG 1.183. Therefore, 33 the proposed changes are acceptable with regard to the radiological consequences of the 34 postulated DBAs.
35 36 37 Technical contacts:
Kristy Bucholtz, NRR/DRA/ARCB 38 Nageswara Karipineni, NRR/DSS/SBPB 39 40
Attachment:
Basis for Accepting the Proposed Changes to the Standard Technical 41 Specification Bases, Volume 2 of NUREGs-1433 and -1434 42 43 Date:
44 45 46 47
ATTACHMENT ATTACHMENT 1
2 BASIS FOR ACCEPTING THE PROPOSED CHANGES TO THE STANDARD TECHNICAL 3
SPECIFICATION BASES, VOLUME 2 OF NUREGS-1433 AND -1434 4
5
1.0 INTRODUCTION
6 7
Traveler TSTF-551 proposes changes to Standard Technical Specifications, General Electric 8
BWR/4 Plants, BWR/4 NUREG-1433, Volume 2, Bases, Revision 4.0, April 2012, ADAMS 9
Accession No. ML12104A193 and Standard Technical Specifications, General Electric BWR/6 10 Plants, BWR/6 NUREG-1434, Volume 2, Bases, Revision 4.0, April 2012, ADAMS Accession 11 No. ML12104A196. The changes would be incorporated into future revisions of NUREG-1433, 12 Volume 2, and NUREG-1434, Volume 2. A summary of the changes and the NRC staffs 13 evaluation of those changes are presented in this attachment.
14 15
2.0 REGULATORY EVALUATION
16 17 2.1 APPLICABLE REGULATIONS AND GUIDANCE 18 19 The regulation at 10 CFR 50.36(a)(1) states that each applicant for a license authorizing 20 operation of a production or utilization facility shall include in his application proposed technical 21 specifications in accordance with the requirements of this section. A summary statement of the 22 bases or reasons for such specifications, other than those covering administrative controls, shall 23 also be included in the application, but shall not become part of the technical specifications.
24 25 In its Final Policy Statement on Technical Specifications Improvements for Nuclear Power 26 Reactors, the Commission presented its policy on the scope and purpose of the Technical 27 Specifications. The Commission explained how implementation of the policy statement through 28 implementation of the improved STS is expected to produce an improvement in the safety of 29 nuclear power plants through the use of more operator-oriented TS, improved TS Bases, 30 reduced action-statement-induced plant transients, and more efficient use of NRC and industry 31 resources.
32 33 The Final Policy Statement provides the following description of the scope and the purpose of 34 the Technical Specification Bases:
35 36 Appropriate Surveillance Requirements and Actions should be 37 retained for each LCO which remains or is included in the 38 Technical Specifications. Each LCO, Action, and Surveillance 39 Requirement should have supporting Bases. The Bases should at 40 a minimum address the following questions and cite references to 41 appropriate licensing documentation (e.g., FSAR, Topical Report) 42 to support the Bases.
43 44
- 1. What is the justification for the Technical Specification, i.e., which 45 Policy Statement criterion requires it to be in the Technical 46 Specifications?
47 48
- 2. What are the Bases for each LCO, i.e., why was it determined to 1
be the lowest functional capability or performance level for the 2
system or component in question necessary for safe operation of 3
the facility and, what are the reasons for the Applicability of the 4
LCO?
5 6
- 3. What are the Bases for each Action, i.e., why should this remedial 7
action be taken if the associated LCO cannot be met; how does 8
this Action relate to other Actions associated with the LCO; and 9
what justifies continued operation of the system or component at 10 the reduced state from the state specified in the LCO for the 11 allowed time period?
12 13
- 4. What are the Bases for each Safety Limit?
14 15
- 5. What are the Bases for each Surveillance Requirement and 16 Surveillance Frequency; i.e., what specific functional requirement 17 is the surveillance designed to verify? Why is this surveillance 18 necessary at the specified frequency to assure that the system or 19 component function is maintained, that facility operation will be 20 within the Safety Limits, and that the LCO will be met?
21 22 Note: In answering these questions the Bases for each number 23 (e.g., Allowable Value, Response Time, Completion Time, 24 Surveillance Frequency, etc.), state, condition, and definition (e.g.,
25 operability) should be clearly specified. As an example, a number 26 might be based on engineering judgment, past experience, or 27 PSA insights; but this should be clearly stated.
28 29 The NRC staff used the guidance contained in the Final Policy Statement during its review of 30 the proposed changes to the Bases.
31 32
2.2 DESCRIPTION
OF CHANGES 33 34 Volume 2 of NUREGs-1433 and -1434 contain the Bases for each Safety Limit and each LCO 35 contained in Volume 1. The Bases for each LCO is organized into sections:
36 37
=
Background===
38 Applicable Safety Analyses, LCO, and Applicability 39 Actions 40 Surveillance Requirements 41 References 42 43 The Bases for SR 3.6.4.1.1 in NUREGs-1433 and -1434 is being revised, and the Bases for 44 SR 3.6.4.1.3 in NUREG-1433 is being revised. The following discussion provides a summary of 45 the revised Bases, followed by the NRC staffs evaluation of the revised Bases.
46 47
3.0 TECHNICAL EVALUATION
1 2
3.1 REVISION TO SR 3.6.4.1.1 BASES 3
4 The Bases for SR 3.6.4.1.1 is revised by the addition of a description of the modification to the 5
applicability of the SR acceptance criterion. The revised Bases describe conditions that could 6
lead to the required vacuum not being met and provides a discussion of why these conditions 7
do not indicate a change in the leaktightness of the [secondary] containment boundary. It also 8
provides a description of the analysis needed to determine whether one train of SGT could 9
establish the assumed [secondary] containment vacuum in the unlikely event of an accident 10 occurring.
11 12 The NRC staff reviewed the revised Bases and determined that it adequately provides the basis 13 for the SR, and provides an appropriate description of the note which modifies the SR.
14 15 3.2 REVISION TO SR 3.6.4.1.3 BASES 16 17 The Bases for SR 3.6.4.1.3 are revised in their entirety to describe that the verification of one 18 door being closed is necessary to provide assurance that exfiltration from the [secondary]
19 containment does not occur. The revised bases also provide an explanation that the intent is 20 not to breach the [secondary] containment boundary, but the access openings may be used for 21 entry and exit.
22 23 The NRC staff reviewed the revised Bases and determined that it adequately provides the 24 purpose and the basis for the SR.
25 26
4.0 CONCLUSION
27 28 The NRC staff determined that TS Bases changes are consistent with the proposed TS changes 29 and provide an explanation and supporting information for each of the SRs. Therefore, the NRC 30 staff determined that the revised Bases are consistent with the Commission's Final Policy 31 Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 32 1993 (58 Federal Register 39132).
33 34
ENCLOSURE 2 DRAFT MODEL SAFETY EVALUATION 1
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
[NOTE: Throughout this safety evaluation (SE), items that are enclosed in square brackets 6
signify plant-specific nomenclature or values to be taken from the licensee's submittal.]
7 8
1.0 INTRODUCTION
9 10 By application dated [enter date], (Agencywide Documents Access and Management System 11 (ADAMS) Accession No. [MLXXXXXXXXX], [name of licensee] (the licensee) requested 12 changes to the technical specifications (TS) for [name of facility]. Specifically, the licensee 13 requested changes to the TSs to adopt Technical Specifications Task Force (TSTF) traveler, 14 TSTF-551, Revision 3, Revise Secondary Containment Surveillance Requirements, dated 15 October 3, 2016 (ADAMS Accession No. ML16277A226). The NRC approved the traveler on 16
[Month, Day, 2017 (ADAMS Accession No. MLXXXX]).
17 18 The proposed changes would allow the [secondary] containment vacuum limit to not be met 19 provided that the standby gas treatment (SGT) system remains capable of establishing the 20 required [secondary] containment vacuum within the [specified time] and revises the TS to 21 permit [secondary] containment access opening to be open to permit ingress and egress.
22 23
2.0 REGULATORY EVALUATION
24 25 2.1 SYSTEM DESCRIPTION 26 27 The [secondary] containment is a structure that encloses the primary containment, including 28 components that may contain primary system fluid. The safety function of the [secondary]
29 containment is to contain, dilute, and hold up fission products that may leak from primary 30 containment following a design basis accident (DBA) to ensure the control room operator and 31 offsite doses are within the regulatory limits. There is no redundant train or system that can 32 perform the [secondary] containment function should the [secondary] containment be 33 inoperable.
34 35 The [secondary] containment boundary is the combination of walls, floor, roof, ducting, doors, 36 hatches, penetrations and equipment that physically form the [secondary] containment.
37 Routinely used [secondary] containment access openings contain at least one inner and one 38 outer door in an airlock configuration. In some cases, [secondary] containment access 39 openings are shared such that there are multiple inner or outer doors. All [secondary]
40 containment access doors are normally kept closed, except when the access opening is being 41 used for entry and exit of personnel, equipment, or material.
42 43
[Secondary] containment operability is based on its ability to contain, dilute, and hold up fission 1
products that may leak from primary containment following a DBA. To prevent ground level 2
exfiltration of radioactive material while allowing the [secondary] containment to be designed 3
as a mostly conventional structure, the [secondary] containment requires support systems to 4
maintain the pressure at less than atmospheric pressure. During normal operation, non-safety 5
related systems are used to maintain the [secondary] containment at a slight negative pressure 6
to ensure any leakage is into the building and that any [secondary] containment atmosphere 7
exiting is via a pathway monitored for radioactive material. However, during normal operation it 8
is possible for the [secondary] containment vacuum to be momentarily less than the required 9
vacuum for a number of reasons, such as during wind gusts or swapping of the normal 10 ventilation subsystems.
11 12 During emergency conditions, the SGT system is designed to be capable of drawing down the 13
[secondary] containment to a required vacuum within a prescribed time and continue to 14 maintain the negative pressure as assumed in the accident analysis. For [name of facility], the 15 SGT must be able to establish the required vacuum within [insert time requirement]. The leak 16 tightness of the [secondary] containment together with the SGT system ensure that radioactive 17 material is either contained in the [secondary] containment or filtered through the SGT system 18 filter trains before being discharged to the outside environment via the elevated release point.
19 20 2.2 PROPOSED TECHNICAL SPECIFICATION CHANGES 21 22 The proposed changes would allow the [secondary] containment vacuum limit to not be met 23 provided the SGT system remains capable of establishing the required [secondary]
24 containment vacuum. The proposed changes would also allow for the temporary opening of the 25 inner and outer doors of [secondary] containment for the purpose of ingress and egress (i.e.,
26 normal opening and prompt closure of a door for transit).
27 28 2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 29 30
[NOTE: This change is applicable to all BWR types.NUREGs-1433 and -1434.]
31 32 Surveillance requirement (SR) 3.6.4.1.1 requires verification that [secondary] containment 33 vacuum is [0.25] inch of vacuum water gauge. This SR would be modified by a note that 34 states:
35 36 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 37 standby gas treatment (SGT) subsystem is capable of establishing 38 the required [secondary] containment vacuum.
39 40 2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 41 42
[NOTE: This change is applicable to BWR/2, BWR/3, BWR/4, and BWR/5 plants NUREG-1433 43 only.]
44 45 SR 3.6.4.1.3 requires verification that one [secondary] containment access door in each 46 access opening is closed. This SR would be modified by adding the following phrase to the end 47 of the SR statement, except when the access opening is being used for entry and exit.
48 49 2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 1
2 An editorial change is made to SR 3.6.4.1.4 in which the words standby gas treatment are 3
replaced with the initialism SGT.
4 5
2.3 REGULATORY REQUIREMENTS AND GUIDANCE 6
7 The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) 8 requires an applicant for an operating license to include in the application proposed TS in 9
accordance with the requirements of 10 CFR 50.36. The applicant must include in the 10 application, a summary statement of the bases or reasons for such specifications, other than 11 those covering administrative controls. However, per 10 CFR 50.36(a)(1), these technical 12 specification bases shall not become part of the technical specifications.
13 14 Additionally, 10 CFR 50.36(b) requires:
15 16 Each license authorizing operation of a utilization facility will 17 include technical specifications. The technical specifications will 18 be derived from the analyses and evaluation included in the safety 19 analysis report, and amendments thereto, submitted pursuant to 20 10 CFR 50.34 [Contents of applications; technical information].
21 The Commission may include such additional technical 22 specifications as the Commission finds appropriate.
23 24 The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required 25 by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (LCOs), which are 26 the lowest functional capability or performance levels of equipment required for safe operation 27 of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the 28 licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the 29 condition can be met.
30 31 The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 32 which are requirements relating to test, calibration, or inspection to assure that the necessary 33 quality of systems and components is maintained, that facility operation will be within safety 34 limits, and that the LCOs will be met.
35 36 The NRC staffs guidance for review of TSs is in Chapter 16, Technical Specifications, of 37 NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 38 Nuclear Power Plants (SRP), dated March 2010 (ADAMS Accession No. ML100351425).
39 40 NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative 41 Source Terms, Revision 0, dated July 2000 (ADAMS Accession No. ML003734190), provides 42 guidance to the NRC staff for the review of alternate source term (AST) amendment requests.
43 SRP 15.0.1 states that the NRC reviewer should evaluate the proposed change against the 44 guidance in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for 45 Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000 46 (ADAMS Accession No. ML003716792).
47 48 RG 1.183 provides acceptable methodology for analyzing the radiological consequences of 1
several design basis accidents to show compliance with 10 CFR 50.67. RG 1.183 provides 2
guidance to licensees on acceptable application of AST (also known as the accident source 3
term) submittals, including acceptable radiological analysis assumptions for use in conjunction 4
with the accepted AST.
5 6
10 CFR 50.67, Accident source term, states that:
7 8
(i)
An individual located at any point on the boundary of the 9
exclusion area for any 2-hour period following the onset of 10 the postulated fission product release, would not receive a 11 radiation dose in excess of 0.25 Sv (25 rem) total effective 12 dose equivalent (TEDE),
13 (ii)
An individual located at any point on the outer boundary of 14 the low population zone, who is exposed to the radioactive 15 cloud resulting from the postulated fission product release 16 (during the entire period of its passage), would not receive 17 a radiation dose in excess of 0.25 Sv (25 rem) TEDE, and 18 (iii)
Adequate radiation protection is provided to permit access 19 to and occupancy of the control room under accident 20 conditions without personnel receiving radiation exposures 21 in excess of 0.05 Sv (5 rem) TEDE for the duration of the 22 accident.
23 24
3.0 TECHNICAL EVALUATION
25 26 The NRC staff evaluated the licensees application to determine if the proposed changes are 27 consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of 28 this safety evaluation (SE). In determining whether an amendment to a license will be issued, 29 the Commission is guided by the considerations that govern the issuance of initial licenses to 30 the extent applicable and appropriate. In making its determination as to whether to amend the 31 license, the NRC staff considered those regulatory requirements that are automatically 32 conditions of the license through 10 CFR 50.54.
33 34 The regulation at 10 CFR 50.36(a)(1) states, in part: A summary statement of the bases or 35 reasons for such specifications shall also be included in the application, but shall not become 36 part of the technical specifications. Accordingly, along with the proposed TS changes, the 37 licensee also submitted TS Bases changes that correspond to the proposed STS changes for 38 information only.
39 40
3.1 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.1 41 42 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 43 an analysis demonstrates that one SGT subsystem is capable of establishing the required 44
[secondary] containment vacuum. During normal operation, conditions may occur that result in 45 SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external 46 pressure or loss of the normal ventilation system that maintains [secondary] containment 47 vacuum may affect [secondary] containment vacuum. These conditions may not be indicative 48 of degradations of the [secondary] containment boundary or of the ability of the SGT system to 1
perform its specified safety function.
2 3
The note provides an allowance for the licensee to confirm [secondary] containment operability 4
by confirming that one SGT subsystem is capable of performing its specified safety function.
5 This confirmation is necessary to apply the exception to meeting the SR acceptance criterion.
6 While the duration of these occurrences is anticipated to be very brief, the allowance is 7
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 8
[secondary] containment to be inoperable per [Condition A of LCO 3.6.4.1 or the corresponding 9
Condition for the plant-specific TS].
10 11 The NRC staff has evaluated the impact of this note on the licensees design basis radiological 12 consequence analyses to ensure that the proposed change will not result in an increase in the 13 dose consequences and that the resulting calculated doses remain within the current 14 radiological consequence analysesdesign criteria specified in 10 CFR 50.67 and the accident 15 specific design criteria outlined in RG 1.183.
16 17 The proposed addition of the note to SR 3.6.4.1.1 does not change the TS requirement to meet 18 SR 3.6.4.1.4 and SR 3.6.4.1.5. SR 3.6.4.1.4 requires verification that the [secondary]
19 containment can be drawn down to [0.25] inch of vacuum water gauge in [120] seconds 20 using one SGT subsystem. SR 3.6.4.1.5 requires verification that the [secondary] containment 21 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 22 a flow rate [4000] cubic feet per minute. In addition, TS LCO 3.6.4.3, Standby Gas Treatment 23 (SGT) System, must be met; otherwise the licensee shall shut down the reactor or follow any 24 remedial action permitted by TSs until the condition can be met.
25 26 As discussed above, [secondary] containment operability is based on its ability to contain, 27 dilute, and hold up fission products that may leak from primary containment following a DBA.
28 To prevent ground level exfiltration of radioactive material the [secondary] containment 29 pressure must be maintained at a pressure that is less than atmospheric pressure. The 30
[secondary] containment requires support systems to maintain the control volume pressure 31 less than atmospheric pressure. Following an accident, the SGT system ensures the 32
[secondary] containment pressure is less than the external atmospheric pressure. During 33 normal operation, non-safety related systems are used to maintain the [secondary]
34 containment at a negative pressure. However, during normal operation it is possible for the 35
[secondary] containment vacuum to be momentarily less than the required vacuum for a 36 number of reasons. These conditions may not be indicative of degradations of the [secondary]
37 containment boundary or of the ability of the SGT system to perform its specified safety 38 function. Since the licensee meets the requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the 39 LCO or is following the Actions of TS LCO 3.6.4.3, and the licensees analysis confirms 40
[secondary] containment operability by confirming that one SGT subsystem is capable of 41 performing its specified safety function, then there is reasonable assurance that the 42
[secondary] containment and SGT subsystem will maintain the vacuum requirements during a 43 DBA.
44 45 Therefore, the NRC staff has determined that: if the conditions do not affect (1) the ability to 46 maintain the [secondary] containment pressure during an accident, at a pressure that is less 47 than atmospheric, and (2) the time assumed in the accident analyses to draw down the 48
[secondary] containment pressure, then the [secondary] containment can perform its safety 49 function and may be considered TS operable. This is evident by being able to successfully 1
perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to 2
establish and maintain the required vacuum in the [secondary] containment as assumed in the 3
accident analyses.
4 5
Furthermore, because the specified safety functions of the [secondary] containment and SGT 6
subsystem can be performed in the time assumed in the licensees accident analysis, then the 7
fission products that bypass or leak from primary containment, or are released from the reactor 8
coolant pressure boundary components located in [secondary] containment prior to release to 9
the environment, will be contained and processed as assumed in the licensees design basis 10 radiological consequence dose analyses. The NRC staff finds that the proposed change does 11 not affect the current radiological consequence analyses and concludes that the proposed 12 change is acceptable with respect to the radiological consequences of DBAs.
13 14
3.2 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.3 15 16
[NOTE: The proposed change is not applicable if the radiological dose consequence analysis 17 assumes the [secondary] containment pressure is below atmospheric pressure prior to or 18 coincident with the time at which the accident or event occurs. Such an analysis assumption 19 would require a revised radiological dose consequence analysis considering the new release 20 point (the open [secondary] containment doors), with appropriate atmospheric dispersion 21 factors, and any other necessary revisions to the accident or event analysis.]
22 23 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, 24 inadvertent, simultaneous opening of redundant [secondary] containment access doors during 25 normal entry and exit conditions. Planned activities that could result in the simultaneous 26 opening of redundant [secondary] containment access openings, such as maintenance of a 27
[secondary] containment personnel access door or movement of large equipment through the 28 openings that would take longer than the normal transit time, will be considered outside the 29 scope of the NRC staff's review.
30 31 The NRC staff reviewed the changes to SR 3.6.4.1.3. The NRC staff determined that the SR 32 continues to provide appropriate confirmation that [secondary] containment boundary doors 33 are properly positioned and capable of performing their function in preserving the [secondary]
34 containment boundary. The NRC staff determined that the SRs continue to appropriately verify 35 the operability of the [secondary] containment and provide assurance that the necessary 36 quality of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
37 38 Additionally, the NRC staff evaluated the impact of modifying the TS to allow [secondary]
39 containment access openings to be open for entry and exit on the licensees design basis 40 radiological consequence dose analyses to ensure that the modification will not result in an 41 increase in the radiation dose consequences and that the resulting calculated radiation doses 42 will remain within the design criteria specified in the current radiological consequence 43 analyses10 CFR 50.67 and the accident specific design criteria outlined in RG 1.183. The NRC 44 staff review of these DBAs determined that there are two DBAs that take credit for the 45
[secondary] containment, and are possibly impacted by the brief, inadvertent, simultaneous 46 opening of both an inner and outer access door during normal entry and exit conditions, the loss 47 of coolant accident (LOCA) and the fuel handling accident (FHA) in [secondary] containment.
48 49 3.2.1 LOCA 1
2 Following a LOCA, the [secondary] containment structure is maintained at a negative pressure 3
ensuring that leakage from primary containment to [secondary] containment can be collected 4
and filtered prior to release to the environment. The SGT system performs the function of 5
maintaining a negative pressure within the [secondary] containment, as well as collecting and 6
filtering the leakage from primary containment. The licensee credits the SGT system for 7
mitigation of the radiological releases from the [secondary] containment. In the LOCA 8
analysis, the [secondary] containment draw down analysis assumes that SGT system can 9
draw down the [secondary] containment within [5 minutes]. TS SR 3.6.4.1.4 requires one 10 SGT subsystem to draw down the [secondary] containment, to greater than or equal to [0.25]
11 inches of vacuum water gauge in a maximum allowable time of [120] seconds.
12 13 Conservatively, the DBA LOCA radiological consequence analysis in [Updated Final Safety 14 Analysis Report (UFSAR) Chapter 15] assumes that following the start of a DBA LOCA the 15
[secondary] containment pressure of [0.25] inches of vacuum water gauge is achieved at 16 approximately [10] minutes. The license assumes that releases into the [secondary]
17 containment prior to the [10]-minute draw down time leak directly to the environment as a 18 ground level release with no filtration. After the assumed [10]-minute draw down these releases 19 are filtered by the SGT system and released via the SGT system exhaust vent.
20 21 Based on this information, the NRC staff concludes that the licensees DBA LOCA analysis has 22 sufficient conservatism by assuming a draw down time of [10] minutes from the start of the DBA 23 LOCA. Margin exists to ensure that the [secondary] containment can be reestablished during 24 a brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 25 assurance that a failure of a safety system needed to control the release of radioactive material 26 to the environment will not result. The brief, inadvertent, simultaneous opening of the 27
[secondary] containment access doors does not impact the design bases and will not result in 28 an increase in any on-site or off-site dose.
29 30 Based on the above discussion, the NRC staff finds that the licensees proposed change to the 31 TSs does not impact the licensees design basis LOCA radiological consequence analysis and 32 will not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that 33 this change is acceptable with respect to the radiological consequences of the DBAs.
34 35
[The licensee was approved for AST methodology and the radiological dose consequences 36 analyses for DBAs via license amendment[s] [insert license amendment number] for [name 37 of facility].] The NRC staff reviewed the impact of the proposed changes to [name of facility]
38 TS, on all DBAs currently analyzed in the [name of facility] Updated Final Safety Analysis 39 Report (UFSAR) that could have the potential for significant dose consequences. [Chapter 15]
40 of the [name of facility] UFSAR describes the DBAs and their radiological consequence 41 analysis results.]
42 43 3.2.2 FHA in [Secondary] Containment 44 45 During normal operation, non-safety related systems are used to maintain the [secondary]
46 containment at [0.25] inches of vacuum water gauge to ensure that any leakage is into the 47 building and that any [secondary] containment atmosphere exiting the building is via a 48 monitored pathway. The refuel floor, which is inside the [secondary] containment, is 49 maintained at a negative [0.25] inches of vacuum water gauge by normal operating ventilation 1
systems. The refueling floor exhaust ductwork in the [secondary] containment is equipped with 2
radiation monitors to detect a fuel handling accident. When a radiological release is sensed by 3
the radiation monitors, a [secondary] containment isolation signal is generated. This initiates 4
the SGT system and the normal ventilation system isolates. The radiation monitor is positioned 5
such that it will detect the release and send a closure signal to the [secondary] containment 6
isolation dampers.
7 8
Following a FHA, the [secondary] containment structure is maintained at a negative pressure 9
by the SGT system ensuring that fission products released from the spent fuel pool to 10
[secondary] containment can be collected and filtered prior to release to the environment. In 11 the FHA analysis, the [secondary] containment draw down analysis demonstrates that SGT 12 system can draw down the [secondary] containment within [5 minutes]. The licensee credits 13 the SGT system for mitigation of the radiological releases from the [secondary] containment.
14 TS SR 3.6.4.1.4 requires one SGT subsystem to draw down the [secondary] containment, to 15 greater than or equal to [0.25] inches of vacuum water gauge in a maximum allowable time of 16
[120] seconds.
17 18 Conservatively, the DBA FHA radiological consequence analysis in [UFSAR Chapter 15]
19 assumes that following the start of a DBA FHA the [secondary] containment pressure of 20
[0.25] inches of vacuum water gauge is achieved at approximately [10] minutes. The license 21 assumes that releases into the [secondary] containment prior to the [10]-minute draw down 22 time leak directly to the environment as a ground level release with no filtration. After the 23 assumed [10]-minute draw down these releases are filtered by the SGT system and released 24 via the SGT system exhaust vent.
25 26 Based on this information, the NRC staff concludes that the licensees DBA FHA analysis has 27 sufficient conservatism by assuming a draw down time of [10] minutes from the start of the DBA 28 FHA. Margin exists to ensure that the [secondary] containment can be reestablished during 29 brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable 30 assurance that a failure of a safety system needed to control the release of radioactive material 31 to the environment will not result. The brief, inadvertent, simultaneous opening of the 32
[secondary] containment access doors does not impact the design bases and will not result in 33 an increase in any on-site or off-site dose.
34 35 Based on the above discussion, the NRC staff finds that the licensees proposed change to the 36 TSs does not impact the licensees design basis FHA radiological consequence analysis and will 37 not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that this 38 change is acceptable with respect to the radiological consequences of the DBAs.
39 40 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, 41 inadvertent, simultaneous opening of redundant [secondary] containment access doors during 42 normal entry and exit conditions. Planned activities that could result in the simultaneous 43 opening of redundant [secondary] containment access openings, such as maintenance of a 44
[secondary] containment personnel access door or movement of large equipment through the 45 openings that would take longer than the normal transit time, will be considered outside the 46 scope of the NRC staff's review.
47 48 3.2.3 Conclusion 49 1
As described above, the NRC staff reviewed the technical basis provided by the licensee to 2
assess the radiological impacts of the changes to the [secondary] containment in the licensees 3
TSs. The NRC staff finds that the licensee proposed change to SR 3.6.4.1.3 is consistent with 4
regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds, 5
with reasonable assurance that the licensees change to the TSs will continue to comply with 6
these criteria and that that the licensee's estimates of the dose consequences of a design basis 7
LOCA and FHA will comply with the requirements of the current radiological consequence 8
analyses10 CFR 50.67 and the accident specific dose guidelines specified in RG 1.183.
9 Therefore, the proposed changes are acceptable with regard to the radiological consequences 10 of the postulated DBAs.
11 12
3.3 PROPOSED CHANGE
TO SURVEILLANCE REQUIREMENT 3.6.4.1.4 13 14 The changes to SR 3.6.4.1.4 are editorial only and do not change any technical aspects of 15 SR 3.6.4.1.4. The NRC staff determined that the change is acceptable.
16 17 3.4 VARIATIONS FROM THE APPROVED TRAVELER 18 19
[NOTE: Technical reviewers and/or project manager to discuss variations from the approved 20 traveler and whether they are acceptable. Choose the applicable paragraphs based on 21 information provided in the LAR.]
22 23
[The licensee is not proposing any variations from the TS changes described in TSTF-551 or 24 the applicable parts of the NRC staffs safety evaluation of TSTF-551.]
25 26
[The licensee is proposing the following variations from the TS changes described in TSTF-551 27 or the applicable parts of TSTF-551 or the NRC staffs safety evaluation. These variations do 28 not affect the applicability of TSTF-551 or the NRC staff's safety evaluation to the proposed 29 license amendment.]
30 31
[The [PLANT] TS do not contain an SR equivalent to SR 3.6.4.1.1 modified by TSTF-551.
32 Therefore, the addition of the SR 3.6.4.1.1 Note is not applicable.]
33 34
[The [PLANT] TS already contains an allowance similar to that made to SR 3.6.4.1.3.
35 Therefore, the proposed change does not contain this portion of TSTF-551.]
36 37
[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical 38 Specifications on which TSTF-551 was based. Specifically, [describe differences between the 39 plant-specific TS numbering and/or titles and the TSTF-551 numbering and titles.] These 40 differences are administrative and do not affect the applicability of TSTF-546 to the [PLANT]
41 TS.]
42 43
[The Traveler discusses the applicable regulatory requirements and guidance, including the 44 10 CFR 50, Appendix A, General Design Criteria (GDC). [PLANT] was not licensed to the 45 10 CFR 50, Appendix A, GDC. The [PLANT] equivalents of the referenced GDC are [discussion 46 from licensee's application.] These differences do not alter the conclusion that the proposed 47 change is applicable to [PLANT].]
48 49 3.54
SUMMARY
1 2
The NRC staff reviewed the proposed changes and determined that changes to the TS meet the 3
standards for TS in 10 CFR 50.36(b). The proposed SRs assure that the necessary quality of 4
systems and components is maintained, that facility operation will be within safety limits, and 5
that the LCOs will be met, and satisfy 10 CFR 50.36(c)(3). Additionally, the changes to the TS 6
were reviewed for technical clarity and consistency with customary terminology and format in 7
accordance with SRP Chapter 16.
8 9
Additionally, the NRC staff has evaluated the impact of the proposed changes on the design 10 basis radiological consequence analyses against the regulatory requirements and guidance 11 identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the 12 licensees change to the TSs will continue to comply with the requirements of the current 13 radiological consequence analyses10 CFR 50.67 and the guidelines specified in RG 1.183.
14 Therefore, the proposed changes are acceptable with regard to the radiological consequences 15 of the postulated DBAs.
16 17
4.0 STATE CONSULTATION
18 19 In accordance with the Commission's regulations, the [Name of State] State official was notified 20 of the proposed issuance of the amendment on [enter date]. The State official had [no]
21 comments. [If comments were provided, they should be addressed here].
22 23
5.0 ENVIRONMENTAL CONSIDERATION
24 25
[Note: This section is to be prepared by the PM. As needed, the PM should coordinate with 26 NRRs Environmental Review and Projects Branch (RERP) to determine the need for an EA.
27 Specific guidance on preparing EAs and considering environmental issues is contained in NRR 28 Office Instruction LIC-203, Procedural Guidance for Preparing Categorical Exclusions, 29 Environmental Assessments, and Considering Environmental Issues.]
30 31 The amendment changes requirements with respect to the installation or use of facility 32 components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
33 The NRC staff has determined that the amendment involves no significant increase in the 34 amounts and no significant change in the types of any effluents that may be released offsite, 35 and that there is no significant increase in individual or cumulative occupational radiation 36 exposure. The Commission has previously issued a proposed finding that the amendment 37 involves no significant hazards consideration, and there has been no public comment on such 38 finding published in the Federal Register on [DATE (XX FR XXX)]. Accordingly, the 39 amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
40 Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment 41 need be prepared in connection with the issuance of the amendment.
42 43
6.0 CONCLUSION
44 45 The Commission has concluded, based on the considerations discussed above, that: (1) there 46 is reasonable assurance that the health and safety of the public will not be endangered by 47 operation in the proposed manner, (2) there is reasonable assurance that such activities will be 48 conducted in compliance with the Commission's regulations, and (3) the issuance of the 49 amendment will not be inimical to the common defense and security or to the health and safety 1
of the public 2
3 Principal Contributors:
Kristy Bucholtz, NRR/DRA/ARCB 4
Nageswara Karipineni, NRR/DSS/SBPB 5
Margaret Chernoff, NRR/DSS/STSB 6
7 Date:
8 9