ML22240A011
Text
GTST AP1000- B01-2.0, Rev. 1
Advanced Passive 1000 (AP1000)
Generic Technical Specification Traveler (GTST)
Title:
Changes related to Section 2.0, SAFETY LIMITS (SLs)
I. Technical Specifications Task Force (TSTF) Travelers, Approved Since Revision 2 of STS NUREG-1431, and Used to Develop this GTST
TSTF Number and
Title:
None
NA
NRC Approval Date:
NA
TSTF Classification:
NA
Date report generated:
Tuesday, June 02, 2015 Page 1 GTST AP1000- B01-2.0, Rev. 1
II. Reference Combined License (RCOL) Standard Departures (Std. Dep.), RCOL COL Items, and RCOL Plant-Specific Technical Specifications (PTS) Changes Used to Develop this GTST
RCOL Std. Dep. Number and
Title:
None
RCOL COL Item Number and
Title:
None
RCOL PTS Change Number and
Title:
None
Date report generated:
Tuesday, June 02, 2015 Page 2 GTST AP1000- B01-2.0, Rev. 1
III. Comments on Relations Among TSTFs, RCOL Std. Dep., RCOL COL Items, and RCOL PTS Changes
This section discusses the considered changes that are: (1) applicable to operating reactor designs, but not to the AP1000 design; (2) already incorporated in the GTS; or (3) superseded by another change.
None
Date report generated:
Tuesday, June 02, 2015 Page 3 GTST AP1000- B01-2.0, Rev. 1
IV. Additional Changes Proposed as Part of this GTST (modifications proposed by NRC staff and/or clear editorial changes or deviations identified by preparer of GTST)
In the Safety Limit Violations section of the Bases, (Ref. 4) is added at the end of the second paragraph.
APOG Recommended Changes to Improve the Bases
Throughout the Bases, references to Sections and Chapters of the FSAR do not include the FSAR clarifier. Since these Section and Chapter references are to an external document, it is appropriate to include the FSAR modifier. (DOC A003)
In the Applicable safety Analyses section of the Bases for Section 2.1.2, clarify the equivalency of RCS depressurization valves and ADS valves; and of Turbine Bypass System and Steam Dump System.
Date report generated:
Tuesday, June 02, 2015 Page 4 GTST AP1000- B01-2.0, Rev. 1
V. Applicability
Affected Generic Technical Specifications and Bases:
Section 2.0, SAFETY LIMITS (SLs)
Changes to the Generic Technical Specifications and Bases:
In the Applicable Safety Analyses section of the Bases for Section 2.1.2, Automatic Depressurization System [ADS] valves was added in parenthesis next to RCS depressurization valves to relate their equivalency. Similarly, Steam Dump System was added in parenthesis next to Turbine Bypass System to relate their equivalency. (APOG comment)
In the Safety Limit Violations section of the Bases, (Ref. 4) is added at the end of the second paragraph.
Date report generated:
Tuesday, June 02, 2015 Page 5 GTST AP1000- B01-2.0, Rev. 1
VI. Traveler Information
Description of TSTF changes:
NA
Rationale for TSTF changes:
NA
Description of changes in RCOL Std. Dep., RCOL COL Item(s), and RCOL PTS Changes:
NA
Rationale for changes in RCOL Std. Dep., RCOL COL Item(s), and RCOL PTS Changes:
NA
Description of additional changes proposed by NRC staff/preparer of GTST:
The acronym FSAR is added to modify Section and Chapter in references to the FSAR throughout the Bases. (DOC A003)
The last paragraph of the Applicable Safety Analyses section of the Bases for Section 2.1.2 was revised as follows:
More specifically, no credit is taken for the operation of the following:
- a. RCS depressurization valves (Automatic Depressurization System [ AD S ]
valves):
- c. Turbine Bypass System (Steam Dump System);
- d. Reactor Control System;
- e. Pressurizer Level Control System; or
- f. Pressurizer Spray.
In the second paragraph of the Safety Limit Violations section in the Bases, reference to 10 CFR 50.34, which is 4th reference in the References section, was added as follows:
Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for abnormal radioactive releases (Ref. 4).
Date report generated:
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Rationale for additional changes proposed by NRC staff/preparer of GTST:
Since Bases references to FSAR Sections and Chapters are to an external document, it is appropriate to include the FSAR modifier.
The changes in the Applicable Safety Analyses section of the Bases relate the equivalency of the terminology used and correct the description for RCS depressurization.
Reference to 10 CFR 50.34 was missing. This omission is corrected by adding Ref. 4 in the Safety Limit Violations section. 10 CFR 50.34(a)(1)(ii)(D) refers to 10 CFR 100, Reactor Site Criteria, limits.
Date report generated:
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VII. GTST Safety Evaluation
Technical Analysis:
Changes to the Applicable Safety Analyses section of the Bases for Section 2.1.2
The changes addressed in this section clarify RCS depressurization and equivalency of the terms used in the FSAR. These changes provide clarity and improve understanding, and are therefore acceptable.
Remaining changes
The remaining changes are editorial, clarifying, grammatical, or otherwise considered administrative. These changes do not affect the technical content, but improve the readability, implementation, and understanding of the requirements, and are therefore acceptable.
Having found that this GTSTs proposed changes to the GTS and Bases are acceptable, the NRC staff concludes that AP1000 STS Section 2.0 is an acceptable model Specification for the AP1000 standard reactor design.
References to Previous NRC Safety Evaluation Reports (SERs):
NA
Date report generated:
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VIII. Review Information
Evaluator Comments:
Reference to 10 CFR 50.34 and 10 CFR 100 should preferably be clarified. In NUREG-1431, Rev.4, 10 CFR 100 is referred in the Background section in the Bases; no reference is made to 10 CFR 50.34. Here, 10 CFR 50.34 is referred to in the Safety Limit Violations section in the Bases; no reference is made separately to 10 CFR 100. Section 2.0 of the STS for Westinghouse reactors in NUREG-1431 and here, for AP1000, is nearly identical; it will be appropriate to resolve this difference.
Pranab K. Samanta Brookhaven National Laboratory 631-344-4948 samanta@bnl.gov
Review Information:
Availability for public review and comment on Revision 0 of this traveler approved by NRC staff on 4/28/2014.
APOG Comments (Ref. 7) and Resolutions
(Internal #3) Throughout the Bases, references to Sections and Chapters of the FSAR do not include the FSAR modifier. Since these Section and Chapter references are to an external document, it is appropriate to include the FSAR modifier. This is resolved by adding the FSAR modifier as appropriate.
(Internal #40) 2.0, Pg. 25, In the Applicable Safety Analyses section of the Bases for Section 2.1.2, the APOG appears to suggest making the first list item a. RCS depressurization valves a part of the lead-in for the list. This contradicts comment #41 and was considered not applicable.
(Internal #41) 2.0, Pg. 25, In the Applicable Safety Analyses section of the Bases, equivalency of RCS depressurization valves and Automatic Depressurization System [ADS] valves; and of Turbine Bypass System and Steam Dump System was clarified.
NRC Final Approval Date: 6/02/2015
NRC
Contact:
T. R. Tjader United States Nuclear Regulatory Commission 301-415-1187 Theodore.Tjader@nrc.gov
Date report generated:
Tuesday, June 02, 2015 Page 9 GTST AP1000- B01-2.0, Rev. 1
IX. Evaluator Comments for Consideration in Finalizing Technical Specifications and Bases
None
Date report generated:
Tuesday, June 02, 2015 Page 10 GTST AP1000- B01-2.0, Rev. 1
X. References Used in GTST
- 1. AP1000 DCD, Revision 19, Section 16, Technical Specifications, June 2011 (ML11171A500).
- 2. Southern Nuclear Operating Company, Vogtle Electric Generating Plant, Unit 3 and 4, Technical Specifications Upgrade License Amendment Request, February 24, 2011 (ML12065A057).
- 3. TSTF-GG-05-01, Technical Specification Task Force (TSTF) W riter's Guide for Plant-Specific Improved Technical Specifications, Revision 1.
- 4. RAI Letter No. 01 Related to License Amendment Request (LAR) 12- 002 for the Vogtle Electric Generating Plant, Units 3 and 4 Combined Licenses, September 7, 2012 (ML12251A355).
- 5. Southern Nuclear Operating Company, Vogtle Electric Generating Plant, Units 3 and 4, Response to Request for Additional Information Letter No. 01 Related to License Amendment Request LAR 002, ND 2015, October 04, 2012 (ML12286A363 and ML12286A360).
- 6. NRC Safety Evaluation (SE) for Amendment No. 13 to Combined License (COL) No. NPF-91 for Vogtle Electric Generating Plant (VEGP) Unit 3, and Amendment No. 13 to COL No.
NPF-92 for VEGP Unit 4, September 9, 2013 (ADAMS Package Accession No. ML13238A337), which contains:
ML13238A355, Cover Letter - Issuance of License Amendment No. 13 for Vogtle Units 3 and 4 (LAR 12-002).
ML13238A359, Enclosure 1 - Amendment No. 13 to COL No. NPF-91 ML13239A256, Enclosure 2 - Amendment No. 13 to COL No. NPF-92 ML13239A284, Enclosure 3 - Revised plant-specific TS pages (Attachment to Amendment No. 13)
ML13239A287, Enclosure 4 - Safety Evaluation (SE), and Attachment 1 - Acronyms ML13239A288, SE Attachment 2 - Table A - Administrative Changes ML13239A319, SE Attachment 3 - Table M - More Restrictive Changes ML13239A333, SE Attachment 4 - Table R - Relocated Specifications ML13239A331, SE Attachment 5 - Table D - Detail Removed Changes ML13239A316, SE Attachment 6 - Table L - Less Restrictive Changes
The following documents were subsequently issued to correct an administrative error in Enclosure 3:
ML13277A616, Letter - Correction To The Attachment (Replacement Pages) - Vogtle Electric Generating Plant Units 3 and 4-Issuance of Amendment Re:
Technical Specifications Upgrade (LAR 12- 002) (TAC No. RP9402)
ML13277A637, Enclosure 3 - Revised plant-specific TS pages (Attachment to Amendment No. 13) (corrected)
Date report generated:
Tuesday, June 02, 2015 Page 11 GTST AP1000- B01-2.0, Rev. 1
- 7. APOG-2014- 008, APOG (AP1000 Utilities) Comments on AP1000 Standardized Technical Specifications (STS) Generic Technical Specification Travelers (GTSTs), Docket ID NRC -
2014- 0147, September 22, 2014 (ML14265A493).
Date report generated:
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XI. MARKUP of the Applicable GTS Subsection for Preparation of the STS NUREG
The entire section of the Specifications and the Bases associated with this GTST is presented next.
Changes to the Specifications and Bases are denoted as follows: Deleted portions are marked in strikethrough red font, and inserted portions in bold blue f ont.
Date report generated:
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SLs 2.0
2.0 SAFETY LIMITS (SLs)
2.1 SLs
2.1.1 Reactor Core SLs
In MODES 1 and 2, the combination of THERMAL POW ER, Reactor Coolant System (RCS) highest loop cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.14 for the W RB-2M DNB correlation.
2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MW D/MTU of burnup.
In MODES 1, 2, 3, 4, and 5 the RCS pressure shall be maintained 2733.5 psig.
2.2 SL Violations
2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
AP1000 STS 2.0- 1 Amendment 0Rev. 0 Revision 19 Date report generated:
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Reactor Core SLs B 2.1.1
B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core Safety Limits (SLs)
BASES
BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not to be exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur, and by requiring that the fuel centerline temperature stays below the melting temperature.
The restriction of this SL prevents overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Fuel centerline melting occurs when the local LHR or power peaking in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (Zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.
This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
The proper functioning of the Protection and Safety Monitoring System (PMS) and steam generator safety valves prevents violation of the reactor core SLs.
AP1000 STS B 2.1.1-1 Amendment 0Rev. 0 Revision 19 Date report generated:
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Reactor Core SLs B 2.1.1
BASES
APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the following fuel design criteria:
- a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
- b. The hot fuel pellet in the core must not experience centerline fuel melting.
The Reactor Trip System (RTS) setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, I, and THERMAL POW ER level that would result in a departe frle eogataNB of lesshanhe limitrlthe existence of fltabilities.
Aatic enforcementf thesreactes ihe oprie operatif tMptm generorafyv.
T reprtigreqrt for estlisTp spois. iCl 4. oCprseI Temrur and clte from Nucateog EaNBFi t assumtitionsf tafy analysndicn F S AR Section 7.2, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.
SAFETY LIMITS The figure provided in the COLR shows the loci of points of THERMAL POW ER, RCS pressure, and cold leg temperature for which the minimum DNBR is not less than the safety analysis limit, that fuel centerline temperature remains below melting, or that the exit quality is within the limits defined by the DNBR correlation.
The reactor core SLs are established to preclude violation of the following fuel design criteria:
- a. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and
- b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.
AP1000 STS B 2.1.1-2 Amendment 0Rev. 0 Revision 19 Date report generated:
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Reactor Core SLs B 2.1.1
BASES
SAFETY LIMITS (continued)
The reactor core SLs are used to define the various PMS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
To ensure that the PMS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower T reactor trifunctions.Th isItt be demonstrt torit qualitysitn timits defi the aNBoorrit lvtemperate and lvpow T reactor trip protection functions continue to provide protection if local hot leg streams approach saturation temperature. Appropriate functioning of the PMS ensures that for variations in the THERMAL POW ER, RCS Pressure, RCS cold leg temperature, RCS flow rate, and I that the reactor core SLs will be sisfied duritte operionIm opiransientsI and Alls.
fCABfifTv p1 only appliMlabpausthese arey Mlabpwch tctsritic. omic prottion functions reqred t lmboABilabp 1 to ensur iitn teacte.Tteam gat saf etls autaticrtictionserve to prentCp heatup tthe reactorore pitis orotie actor trifunctic fcto Mlab P. points fetrifctions are specifil P.P.1I tripystemoF fnstrti. f Mlabp PI 4I RIIpplicabilitys reqred sithe reactors generingigfict TeboMAi mlt bo.
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T alliimefr rogzmrtancef iing tt to alabfrihere thisi is t applicabl and rrobability of fuelage.
cbNCbp xI daC
FS AR Section 7.2, Reactor Trip.
AP1000 STS B 2.1.1-3 Amendment 0Rev. 0 Revision 19 Date report generated:
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B 2.0 SAFETY LIMITS (SLs)
B 2.1.2 Reactor Coolant System (RCS) Pressure SL
BASES
BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, Reactor Coolant Pressure Boundary, and GDC 15, Reactor Coolant System Design (Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, Reactivity Limits (Ref. 1),
reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.
The design pressure of the RCS is 2500 psia (2485 psig). During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).
Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases.
APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the reactor high pressurizer pressure trip have settings ANALYSES established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and
AP1000 STS B 2.1.2-1 Amendment 0Rev. 0 Revision 19 Date report generated:
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BASES
APPLICABLE SAFETY ANALYSES (continued)
hence valve size requirements and lift settings, is a complete loss of external load with loss of feedwater flow, without a direct reactor trip.
During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings.
The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs.
The reactor high pressurizer pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressurizer pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.
More specifically, no credit is taken for operation of the following:
- a. RCS depressurization valves (Automatic Depressurization System [ADS] valves);
- c. Turbine Bypass System (Steam Dump System);
- d. Reactor Control System;
- e. Pressurizer Level Control System; or
- f. Pressurizer spray.
SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel, piping, valves, and fittings under the ASME Code,Section III, is 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2733.5 psig.
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 since the reactor vessel closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.
AP1000 STS B 2.1.2-2 Amendment 0Rev. 0 Revision 19 Date report generated:
Tuesday, June 02, 2015 Page 19 GTST AP1000- B01-2.0, Rev. 1
BASES
SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, VIOLATIONS the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for abnormal radioactive releases (Ref. 4).
The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.
If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
- 2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
- 3. ASME Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.
- 4. 10 CFR 50.34.
- 5. FS AR Section 7.2, Reactor Trip.
AP1000 STS B 2.1.2-3 Amendment 0Rev. 0 Revision 19 Date report generated:
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XII. Applicable STS Subsection After Incorporation of this GTSTs Modifications
The entire subsection of the Specifications and the Bases associated with this GTST, following incorporation of the modifications, is presented next.
Date report generated:
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SLs 2.0
2.0 SAFETY LIMITS (SLs)
2.1 SLs
2.1.1 Reactor Core SLs
In MODES 1 and 2, the combination of THERMAL POW ER, Reactor Coolant System (RCS) highest loop cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.14 for the W RB-2M DNB correlation.
2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MW D/MTU of burnup.
In MODES 1, 2, 3, 4, and 5 the RCS pressure shall be maintained 2733.5 psig.
2.2 SL Violations
2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
AP1000 STS 2.0- 1 Rev. 0
Date report generated:
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Reactor Core SLs B 2.1.1
B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core Safety Limits (SLs)
BASES
BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not to be exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur, and by requiring that the fuel centerline temperature stays below the melting temperature.
The restriction of this SL prevents overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Fuel centerline melting occurs when the local LHR or power peaking in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (Zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.
This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
The proper functioning of the Protection and Safety Monitoring System (PMS) and steam generator safety valves prevents violation of the reactor core SLs.
AP1000 STS B 2.1.1-1 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 23 GTST AP1000- B01-2.0, Rev. 1
Reactor Core SLs B 2.1.1
BASES
APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the following fuel design criteria:
- a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
- b. The hot fuel pellet in the core must not experience centerline fuel melting.
The Reactor Trip System (RTS) setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, RCS Flow, I, and THERMAL POW ER level that would result in a departe frle eogataNB of lesshanhe limn d prlthe existence of fltabilities.
Aatic enforcementf thesreactes prihe oprie operatif tMptm generorafyv.
T reprtigreqrt for estlisTp s pois. iCl P.4. oCprseI Temrur and clte from Nucateog EaNBFsI assumtitionsf tafy analysndicn cpAo ption T.OIef. OF pride metrictive limitso ensthat are not exc
pAcv ifMfTp T fige prided in the Cl showsifsf TeboMA mlt boICp prseId leg temperatforch the mimNBosesshanafy analysisimitI tt fuel centline trurrtingI ort the exit qualitys witn timits defined NBo crion.
Teacte are establisto prlviolif the fli fuelignriti
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AP1000 STS B 2.1.1-2 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 24 GTST AP1000- B01-2.0, Rev. 1
Reactor Core SLs B 2.1.1
BASES
SAFETY LIMITS (continued)
The reactor core SLs are used to define the various PMS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
To ensure that the PMS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower T reactor trifunctions.Th isItt be demonstrt torit quality ishhes fi tNBoorrit lvtemperate and lvpow T reactor trip protection functions continue to provide protection if local hot leg streams approach saturation temperature. Appropriate functioning of the PMS ensures that for variations in the THERMAL POW ER, RCS Pressure, RCS cold leg temperature, RCS flow rate, and I that the reactor core SLs will be sisfied duritte operionIm opiransientsI and Alls.
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AP1000 STS B 2.1.1-3 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 25 GTST AP1000- B01-2.0, Rev. 1
B 2.0 SAFETY LIMITS (SLs)
B 2.1.2 Reactor Coolant System (RCS) Pressure SL
BASES
BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, Reactor Coolant Pressure Boundary, and GDC 15, Reactor Coolant System Design (Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, Reactivity Limits (Ref. 1),
reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.
The design pressure of the RCS is 2500 psia (2485 psig). During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).
Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases.
APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the reactor high pressurizer pressure trip have settings ANALYSES established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and
AP1000 STS B 2.1.2-1 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 26 GTST AP1000- B01-2.0, Rev. 1
BASES
APPLICABLE SAFETY ANALYSES (continued)
hence valve size requirements and lift settings, is a complete loss of external load with loss of feedwater flow, without a direct reactor trip.
During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings.
The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs.
The reactor high pressurizer pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressurizer pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.
More specifically, no credit is taken for operation of the following:
- a. RCS depressurization valves (Automatic Depressurization System
[ADS] valves);
- c. Turbine Bypass System (Steam Dump System);
- d. Reactor Control System;
- e. Pressurizer Level Control System; or
- f. Pressurizer spray.
SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel, piping, valves, and fittings under the ASME Code,Section III, is 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2733.5 psig.
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 since the reactor vessel closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.
AP1000 STS B 2.1.2-2 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 27 GTST AP1000- B01-2.0, Rev. 1
BASES
SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, VIOLATIONS the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for abnormal radioactive releases (Ref. 4).
The allowable Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.
If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
- 2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.
- 3. ASME Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.
- 4. 10 CFR 50.34.
- 5. FSAR Section 7.2, Reactor Trip.
AP1000 STS B 2.1.2-3 Rev. 0
Date report generated:
Tuesday, June 02, 2015 Page 28