ML14142A011

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Technical Specification & Technical Specification Bases
ML14142A011
Person / Time
Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 05/05/2014
From: Beaver B
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DUK141250019
Download: ML14142A011 (13)
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Text

DISPOSITION OF THE ORIGINAL DOCUMENT WILL BE TO PRIORITY Normal THE TRANSMITTAL SIGNATURE UNLESS RECIPIENT IS Date: 05/05/14 OTHERWISE IDENTIFIED BELOW Document Transmittal #: DUK141250019

1) 01820 J R ELKINS- ECO81
2) 02361 QATS-MG01MM
3) 02388 BOB SCHOMAKER LYNCHBG, VA Duke Energy QA CONDITION D Yes
  • No
4) 02532 RESIDENT NRC INSPECT MG01NRC DOCUMENT TRANSMITTAL FORM OTHER ACKNOWLEDGEMENT REQUIRED E Yes IF QA OR OTHER ACKNOWLEDGEMENT REQUIRED, PLEASE
5) 03044 MCG DOC CNTRL MISC MAN MG05DM ACKNOWLEDGE RECEIPT BY RETURNING THIS FORM TO:
6) 03614 MCG OPS PROCEDURE GP MG01OP REFERENCE
7) 03744 OPS TRNG MGR. MG03OT MCGUIRE NUCLEAR STATION Duke Energy
8) 03759 U S NUC REG WASHINGTON, DC McGuire
9) 03796 SCIENTECH CLEARW-R, FL DCRM MGO2DM TECHNICAL SPECIFICATIONS (TS)
10) 04809 MCG PLANT ENG. LIBR. MG05SE 13225 Hagers Ferry Road
11) 05262 J L FREEZE MG01IE TECHNICAL SPECIFICATIONS BASES Huntersville, N.C. 28078
12) 05606J C MORTON MG01EP (TSB)
13) 08103 WESTINGHOUSE ELECTRIC CO LLC
14) 09665 JON HTHOMPSON, USNRC RECORD RETENTION #

Rec'd By Page 2 of 2 Date V 7

  • T 7 1 1 7 1/2 7 1 1 ,- - A - -

DOCUMENT NO QA COND REV #/ DATE DISTR CODE 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 TOTAL TSB 3.6.9 NA 131 04/14/14 MADM-04B V1 V1 V1 x Vl V3 V1 V1 V1 V1 V1 V1 V1 29 T.S.B 3.6.16 NA 130 04/07/14

- ________ A ____________________________ A ___ .1___ +/- ___ L ___ L ___ .1___ +/- ___ £ ___ I ___ A ____ L ___ L ___ ___

REMARKS: PLEASE UPDATE ACCORDINGLY.

S D CAPPS LOES REVISION 119 DATED 04/21/2014 VICE PRESIDENT MCGUIRE NUCLEAR STATION BY:

BC BEAVER MGO1RC BCB/BRG

April 21, 2014 MEMORANDUM To: All McGuire Nuclear Station Technical Specification (TS) and Tech Spec Bases (TSB) Manual Holders

Subject:

McGuire TS and TSB Updates REMOVE INSERT TS Bases Manual LOES (rev 118) LOES (rev 119)

TSB 3.6.9 Rev 115 (entire section) TSB 3.6.9 Rev 131 (entire section)

TSB 3.6.16 Rev 115 (entire section) TSB 3.6.16 Rev 130 (entire section)

Revision numbers may skip numbers due to Regulatory Compliance Filing System.

Please call me if you have questions.

Bonnie Beaver Regulatory Compliance 875-4180

McGuire Nuclear Station Technical Specification Bases LOES TS Bases are revised by section Page Number Revision Revision Date BASES (Revised per section) i Revision 87 8/15/07 ii Revision 87 8/15/07 iii Revision 87 8/15/07 B 2.1.1 Revision 51 01/14/04 B 2.1.2 Revision 109 9/20/10 B 3.0 Revision 81 3/29/07 B 3.1.1 Revision 115 3/29/11 B 3.1.2 Revision 115 3/29/11 B 3.1.3 Revision 10 9/22/00 B 3.1.4 Revision 115 3/29/11 B 3.1.5 Revision 115 3/29/11 B 3.1.6 Revision 115 3/29/11 B 3.1.7 Revision 58 06/23/04 B 3.1.8 Revision 115 3/29/11 B 3.2.1 Revision 115 3/29/11 B 3.2.2 Revision 115 3/29/11 B 3.2.3 Revision 115 3/29/11 B 3.2.4 Revision 115 3/29/11 B 3.3.1 Revision 124 10/1/12 B 3.3.2 Revision 122 10/25/12 B 3.3.3 Revision 122 10/25/12 B 3.3.4 Revision 115 3/29/11 B 3.3.5 Revision 115 3/29/11 B 3.3.6 Not Used - Revision 87 6/29/06 B 3.4.1 Revision 115 3/29/11 B 3.4.2 Revision 0 9/30/98 B 3.4.3 Revision 115 3/29/11 B 3.4.4 Revision 115 3/29/11 B 3.4.5 Revision 115 3/29/11 McGuire Units 1 and 2 Page I Revision 119

Page Number Amendment Revision Date B 3.4.6 Revision 115 3/29/11 B 3.4.7 Revision 115 3/29/11 B 3.4.8 Revision 115 3/29/11 B 3.4.9 Revision 115 3/29/11 B 3.4.10 Revision 102 8/17/09 B 3.4.11 Revision 115 3/29/11 B 3.4.12 Revision 115 3/29/11 B 3.4.13 Revision 126 5/1/13 B 3.4.14 Revision 115 3/29/11 B 3.4.15 Revision 115 3/29/11 B 3.4.16 Revision 121 8/5/09 B 3.4.17 Revision 115 3/29/11 B 3.4.18 Revision 86 6/25/07 B 3.5.1 Revision 115 3/29/11 B 3.5.2 Revision 116 8/18/11 B 3.5.3 Revision 57 4/29/04 B 3.5.4 Revision 122 10/25/12 B 3.5.5 Revision 115 3/29/11 B 3.6.1 Revision 53 2/17/04 B 3.6.2 Revision 115 3/29/11 B 3.6.3 Revision 115 3/29/11 B 3.6.4 Revision 115 3/29/11 B 3.6.5 Revision 115 3/29/11 B 3.6.6 Revision 122 10/25/12 B 3.6.7 Not Used - Revision 63 4/4/05 B 3.6.8 Revision 115 3/29/11 B 3.6.9 Revision 131 4/14/14 B 3.6.10 Revision 120 4/26/12 B 3.6.11 Revision 122 10/25/12 B 3.6.12 Revision 115 3/29/11 B 3.6.13 Revision 115 3/29/11 B 3.6.14 Revision 115 3/29/11 B 3.6.15 Revision 125 10/19/12 B 3.6.16 Revision 130 4/7/14 McGuire Units 1 and 2 Page 2 Revision 119

Page Number Amendment Revision Date B 3.7.1 Revision 129 10/24/13 B 3.7.2 Revision 105 2/22/10 B 3.7.3 Revision 102 8/17/09 B 3.7.4 Revision 115 3/29/11 B 3.7.5 Revision 115 3/29/11 B 3.7.6 Revision 127 8/2/13 B 3.7.7 Revision 115 3/29/11 B 3.7.8 Revision 128 10/2/13 B 3.7.9 Revision 120 4/26/12 B 3.7.10 Revision 115 3/29/11 B 3.7.11 Revision 115 3/29/11 B 3.7.12 Revision 115 3/29/11 B 3.7.13 Revision 115 3/29/11 B 3.7.14 Revision 115 3/29/11 B 3.7.15 Revision 66 6/30/05 B 3.7.16 Revision 115 3/29/11 B 3.8.1 Revision 115 3/29/11 B 3.8.2 Revision 92 1/28/08 B 3.8.3 Revision 123 9/29/12 B 3.8.4 Revision 115 3/29/11 B 3.8.5 Revision 41 7/29/03 B 3.8.6 Revision 115 3/29/11 B 3.8.7 Revision 115 3/29/11 B 3.8.8 Revision 115 3/29/11 B 3.8.9 Revision 115 3/29/11 B 3.8.10 Revision 115 3/29/11 B 3.9.1 Revision 115 3/29/11 B 3.9.2 Revision 115 3/29/11 B 3.9.3 Revision 115 3/29/11 B 3.9.4 Revision 115 3/29/11 B 3.9.5 Revision 115 3/29/11 B 3.9.6 Revision 115 3/29/11 B 3.9.7 Revision 115 3/29/11 McGuire Units I and 2 Page 3 Revision 119

Reactor Building B 3.6.16 B 3.6 CONTAINMENT SYSTEMS B 3.6.16 Reactor Building BASES BACKGROUND The reactor building is a concrete structure that surrounds the steel containment vessel. Between the containment vessel and the reactor building inner wall is an annular space that collects containment leakage that may occur following a loss of coolant accident (LOCA). This space also allows for periodic inspection of the outer surface of the steel containment vessel.

The Annulus Ventilation System (AVS) establishes a negative pressure in the annulus between the reactor building and the steel containment vessel under post accident conditions. Filters in the system then control the release of radioactive contaminants to the environment. The reactor building is required to be OPERABLE to ensure retention of containment leakage and proper operation of the AVS. To ensure the retention of containment leakage within the reactor building:

a. The door in each access opening is closed except when the access opening is being used for normal transit entry and exit.
b. The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or O-rings) is OPERABLE.

APPLICABLE The design basis for reactor building OPERABILITY is a LOCA.

SAFETY ANALYSES Maintaining reactor building OPERABILITY ensures that the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analyses.

The reactor building satisfies Criterion 3 of 10 CFR 50.36 (Ref. 1).

LCO Reactor building OPERABILITY must be maintained to ensure proper operation of the AVS and to limit radioactive leakage from the containment to those paths and leakage rates assumed in the accident analyses.

McGuire Units 1 and 2 B 3.6.16-1 Revision No. 130

Reactor Building B 3.6.16 BASES APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of radioactive material from the reactor building. Radioactive material may enter the reactor building from the containment following a LOCA.

Therefore, reactor building OPERABILITY is required in MODES 1, 2, 3, and 4 when a steam line break, LOCA, or rod ejection accident could release radioactive material to the containment atmosphere.

In MODES 5 and 6, the probability and consequences of these events are low due to the Reactor Coolant System temperature and pressure limitations in these MODES. Therefore, reactor building OPERABILITY is not required in MODE 5 or 6.

ACTIONS A.1 In the event reactor building OPERABILITY is not maintained, reactor building OPERABILITY must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period.

B.1 and B.2 If the reactor building cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.16.1 REQUIREMENTS Maintaining reactor building OPERABILITY requires maintaining the door in each access opening closed, except when the access opening is being used for normal transit entry and exit. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

McGuire Units 1 and 2 B 3.6.16-2 Revision No. 130

Reactor Building B 3.6.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.16.2 The ability of a AVS train to produce the required negative pressure within the required times provides assurance that the building is adequately sealed. The negative pressure prevents leakage from the building, since outside air will be drawn in by the low pressure. The negative pressure must be established within the time limit to ensure that no significant quantity of radioactive material leaks from the reactor building prior to developing the negative pressure.

The basis for the times and setpoints are provided in TS Bases 3.6.10, "Annulus Ventilation System (AVS)," Applicable Safety Analysis Section.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.16.3 This SR would give advance indication of gross deterioration of the concrete structural integrity of the reactor building. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.6.16-3 Revision No. 130

HMS B 3.6.9 B 3.6 CONTAINMENT SYSTEMS B 3.6.9 Hydrogen Mitigation System (HMS)

BASES BACKGROUND The HMS reduces the potential for breach of primary containment due to a hydrogen oxygen reaction in post accident environments. The HMS is required by 10 CFR 50.44, "Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1), and Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 2), to reduce the hydrogen concentration in the primary containment following a degraded core accident. The HMS must be capable of handling an amount of hydrogen equivalent to that generated from a metal water reaction involving 75% of the fuel cladding surrounding the active fuel region (excluding the plenum volume).

10 CFR 50.44 (Ref. 1) requires units with ice condenser containments to install suitable hydrogen control systems that would accommodate an amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water. The HMS provides this required capability. This requirement was placed on ice condenser units because of their small containment volume and low design pressure (compared with pressurized water reactor dry containments). Calculations indicate that if hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water were to collect in the primary containment, the resulting hydrogen concentration would be far above the lower flammability limit such that, if ignited from a random ignition source, the resulting hydrogen burn would seriously challenge the containment and safety systems in the containment.

The HMS is based on the concept of controlled ignition using thermal ignitors, designed to be capable of functioning in a post accident environment, seismically supported, and capable of actuation from the control room. A total of 70 ignitors are distributed throughout the various regions of containment in which hydrogen could be released or to which it could flow in significant quantities. The ignitors are arranged in two independent trains such that each containment region has at least two ignitors, one from each train, controlled and powered redundantly so that ignition would occur in each region even if one train failed to energize.

When the HMS is initiated, the ignitor elements are energized and heat up to a surface temperature >_1700 0 F. At this temperature, they ignite the hydrogen gas that is present in the airspace in the vicinity of the ignitor.

The HMS depends on the dispersed location of the ignitors so McGuire Units 1 and 2 B 3.6.9-1 Revision No. 131

HMS B 3.6.9 BASES BACKGROUND (continued) that local pockets of hydrogen at increased concentrations would burn before reaching a hydrogen concentration significantly higher than the lower flammability limit. Hydrogen ignition in the vicinity of the ignitors is assumed to occur when the local hydrogen concentration reaches 8.5 volume percent (v/o) and results in 100% of the hydrogen present being consumed.

APPLICABLE The HMS causes hydrogen in containment to burn in a controlled manner SAFETY ANALYSES as it accumulates following a degraded core accident (Ref. 3). Burning occurs at the lower flammability concentration, where the resulting temperatures and pressures are relatively benign. Without the system, hydrogen could build up to higher concentrations that could result in a violent reaction if ignited by a random ignition source after such a buildup.

The hydrogen ignitors are not included for mitigation of a Design Basis Accident (DBA) because an amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water is far in excess of the hydrogen calculated for the limiting DBA loss of coolant accident (LOCA). The hydrogen ignitors have been shown by probabilistic risk analysis to be a significant contributor to limiting the severity of accident sequences that are commonly found to dominate risk for units with ice condenser containments. As such, the hydrogen ignitors satisfy Criterion 4 of 10 CFR 50.36 (Ref. 4).

LCO Two HMS trains must be OPERABLE with power from two independent, safety related power supplies.

For this unit, an OPERABLE HMS train consists of 34 of 35 ignitors energized on the train.

Operation with at least one HMS train ensures that the hydrogen in containment can be burned in a controlled manner. Unavailability of both HMS trains could lead to hydrogen buildup to higher concentrations, which could result in a violent reaction if ignited. The reaction could take place fast enough to lead to high temperatures and overpressurization of containment and, as a result, breach containment or cause containment leakage rates above those assumed in the safety analyses. Damage to safety related equipment located in containment could also occur.

Revision No. 131 McGuire Units McGuire and 22 Units 11 and B 3.6.9-2 B 3.6.9-2 Revision No. 131

HMS B 3.6.9 BASES APPLICABILITY Requiring OPERABILITY in MODES 1 and 2 for the HMS ensures its immediate availability after safety injection and scram actuated on a LOCA initiation. In the post accident environment, the two HMS subsystems are required to control the hydrogen concentration within containment to near its flammability limit of 4.0 v/o assuming a worst case single failure. This prevents overpressurization of containment and damage to safety related equipment and instruments located within containment.

In MODES 3 and 4, both the hydrogen production rate and the total hydrogen production after a LOCA would be tignificantly less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HMS is low.

Therefore, the HMS is not required in MODES 3 and 4.

In MODES 5 and 6, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations of these MODES. Therefore, the HMS is not required to be OPERABLE in MODES 5 and 6.

ACTIONS A.1 and A.2 With one HMS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days or the OPERABLE train must be verified OPERABLE frequently by performance of SR 3.6.9.1. The 7 day Completion Time is based on the low probability of the occurrence of a degraded core event that would generate hydrogen in amounts equivalent to a metal water reaction of 75% of the core cladding, the length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit, and the low probability of failure of the OPERABLE HMS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE train continues to be OPERABLE.

B. 1 Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ignition capability in the adjacent containment regions that would provide redundant capability by flame propagation to the region with no OPERABLE ignitors.

Required Action B.1 calls for the restoration of one hydrogen ignitor in each region to OPERABLE status within 7 days. The 7 day Completion Time is based on the same reasons given under Required Action A.1.

McGuire Units 1 and 2 B 3.6.9-3 Revision No. 131

HMS B 3.6.9 BASES ACTIONS (continued)

C.1 The unit must be placed in a MODE in which the LCO does not apply if the HMS subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time. This is done by placing the unit in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.9.1 REQUIREMENTS This SR confirms that _>34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.

Therefore, energizing provides assurance of OPERABILITY. The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between regions). The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.9.3 A more detailed functional test is performed to verify system OPERABILITY. Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized. Additionally, the surface temperature of each ignitor is measured to be _>1700°F to demonstrate that a temperature sufficient for ignition is achieved. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

McGuire Units 1 and 2 B 3.6.9-4 Revision No. 131

HMS B 3.6.9 BASES REFERENCES 1. 10 CFR 50.44.

2. 10 CFR 50, Appendix A, GDC 41.
3. UFSAR, Section 6.2.
4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units I and 2 B 3.6.9-5 Revision No. 131

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