ML110420010
| ML110420010 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/27/2011 |
| From: | Beaver B Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DUK110270027 | |
| Download: ML110420010 (11) | |
Text
PRIORITY Normal DISPOSITION OF THE ORIGINAL DOCUMENT WILL BE TO THE TRANSMITTAL SIGNATURE UNLESS RECIPIENT IS OTHERWISE IDENTIFIED BELOW
- 1) 01749 L C GIBBY - MG01VP
- 2) 01820 J R ELKINS. EC081
- 3) 02388 BOB SCHOMAKER LYNCHBG, VA
- 4) 02532 RESIDENT NRC INSPECTOR MG01VP
- 5) 02546 WC LIBRARY - MG01WC
- 6) 03044 MCG DOC CNTRL MISC MAN MG05DM
- 9) 03744 OPS TRNG MGR. MG03OT
- 10) 03759 U S NUC REG WASHINGTON, DC
- 11) 03796 SCIENTECH DUNEDIN, FL
- 12) 04698 D E BORTZ EC08G
- 13) 04809 MCG PLANT ENG. LIBR. MG05SE
- 14) 05262 J L FREEZE MG011E
- 15) 05606 J C MORTON MG01EP Duke Energy DOCUMENT TRANSMITTAL FORM REFERENCE MCGUIRE NUCLEAR STATION RECORD RETENTION # 581188 TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS BASES Page 2 of 3 Date:
01/27/11 Document Transmittal #:
DUK110270027 QA CONDITION D Yes
- No OTHER ACKNOWLEDGEMENT REQUIRED
Duke Energy McGuire DCRM MGO2DM 13225 Hagers Ferry Road Huntersville, N.C.
28078 Rec'd By Date DOCUMENT NO QA CONDI REV #/DATE DISTR CODE 11 2
3 4
5 6
7 8
9 10 11 12 13 14 15 TOTAL MEMO 1 PAGE TSB LIST OF EFFECTIVE SECTIONS 3 PAGES TSB 3.1.1 6 PAGES NA NA NA 01/18/11 104 01/18/11 112 12/20/11 MADM-04B v1 V1 x
V1 V1 V3 V1 V1 35 REMARKS: PLEASE UPDATE ACCORDINGLY R T REPKO VICE PRESIDENT MCGUIRE NUCLEAR STATION BY:
BC BEAVER MG01RC BCB/TLC 0
January 18, 2011 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 TSB LOES (entire doc) Rev 103 TSB 3. 1.1 (entire doc) Rev 73 TSB LOES (entire doc) Rev 104 TSB 3.1.1 (entire doc) Rev 112 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 112 12/20/10 B 3.1.2 Revision 10 9/22/00 B 3.1.3 Revision 10 9/22/00 B 3.1.4 Revision 0 9/30/98 B 3.1.5 Revision 19 1/10/02 B 3.1.6 Revision 0 9/30/98 B 3.1.7 Revision 58 06/23/04 B 3.1.8 Revision 0 9/30/98 B 3.2.1 Revision 74 5/3/06 B 3.2.2 Revision 10 9/22/00 B 3.2.3 Revision 34 10/1/02 B 3.2.4 Revision 10 9/22/00 B 3.3.1 Revision 108 8/2/10 B 3.3.2 Revision 99 3/9/09 B 3.3.3 Revision 100 4/13/09 B 3.3.4 Revision 57 4/29/04 B 3.3.5 Revision 11 9/18/00 B 3.3.6 Not Used - Revision 87 6/29/06 B 3.4.1 Revision 51 1/14/04 B 3.4.2 Revision 0 9/30/98 B 3.4.3 Revision 44 7/3/03 B 3.4.4 Revision 86 6/25/07 B 3.4.5 Revision 86 6/25/07 McGuire Units 1 and 2 Page I Revision 104
Page Number B 3.4.6 B 3.4.7 B 3.4.8 B 3.4.9 B 3.4.10 B 3.4.11 B 3.4.12 B 3.4.13 B 3.4.14 B 3.4.15 B 3.4.16 B 3.4.17 B 3.4.18 B 3.5.1 B 3.5.2 B 3.5.3 B 3.5.4 B 3.5.5 B 3.6.1 B 3.6.2 B 3.6.3 B 3.6.4 B 3.6.5 B 3.6.5-2 B 3.6.6 B 3.6.7 B 3.6.8 B 3.6.9 B 3.6.10 B 3.6.11 B 3.6.12 B 3.6.13 B 3.6.14 B 3.6.15 Amendment Revision Date Revision 86 Revision 86 Revision 41 Revision 0 Revision 102 Revision 102 Revision 102 Revision 86 Revision 102 Revision 82 Revision 57 Revision 0 Revision 86 Revision 70 Revision 102 Revision 57 Revision 70 Revision 0 Revision 53 Revision 98 Revision 87 Revision 0 Revision 0 Revision 6 Revision 110 Not Used - Revision 63 Revision 63 Revision 63 Revision 109 Revision 78 Revision 53 Revision 104 Revision 64 Revision 0 6/25/07 6/25/07 7/29/03 9/30/98 8/17/09 8/17/09 8/17/09 6/25/07 8/17/09 9/30/06 4/29/04 9/30/98 6/25/07 10/5/05 8/17/09 4/29/04 10/5/04 9/30/98
.2/17/04 3/24/09 6/29/06 9/30/98 9/30/98 10/6/99 9/22/10 4/4/05 4/4/05 4/4/05 9/20/10 9/25/06 2/17/04 6/28/10 4/23/05 9/30/98 McGuire Units 1 and 2 Page 2 Revision 104
Page Number B 3.6.16 B 3.7.1 B 3.7.2 B 3.7.3 B 3.7.4 B 3.7.5 B 3.7.6 B 3.7.7 B 3.7.8 B 3.7.9 B 3.7.10 B 3.7.11 B 3.7.12 B 3.7.13 B 3.7.14 B 3.7.15 B 3.7.16 B 3.8.1 B 3.8.2 B 3.8.3 B 3.8.4 B 3.8.5 B 3.8.6 B 3.8.7 B 3.8.8 B 3.8.9 B 3.8.10 B 3.9.1 B 3.9.2 B 3.9.3 B 3.9.4 B 3.9.5 B 3.9.6 B 3.9.7 Amendment Revision 40 Revision 102 Revision 105 Revision 102 Revision 57 Revision102 Revision 0 Revision 101 Revision 107 Revision 109 Revision 75 Revision 109 Revision 28 Revision 85 Revision 66 Revision 66 Revision 0 Revision 111 Revision 92 Revision 103 Revision 100 Revision 41 Revision 0 Revision 20 Revision 41 Revision 24 Revision 41 Revision 68 Revision 41 Revision 108 Revision 84 Revision 59 Revision 41 Revision 88 5/8/03 8/17/09 2/22/10 8/17/09 4/29/04 8/17/09 9/30/98 9/17/09 6/23/10 9/20/10 6/12/06 9/20/10 5/17/02 2/26/07 6/30/05 6/30/05 9/30/98 10/23/10 1/28/08 12/15/08 4/13/09 7/29/03 9/30/98 1/10/02 7/29/03 2/4/02 7/29/03 9/1/05 7/29/03 8/2/10 2/20/07 7/29/04 7/29/03 9/5/07 Revision Date McGuire Units I and 2 Page 3 Revision 104
SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.
SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or trip of all shutdown and control rods, assuming that the single rod cluster control assembly of highest reactivity worth is fully withdrawn. However, with all rod cluster control assemblies verified fully inserted by two independent means, it is not necessary to account for a stuck rod cluster control assembly in the shutdown margin calculation.
The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Rod Control System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Rod Control System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.
During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.
McGuire Units 1 and 2 B 3.1.1-1 Revision No. 112
SDM B 3.1.1 BASES APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY ANALYSES analyses. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on a reactor trip.
The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:
- a.
The reactor can eventually be made subcritical from all operating conditions, transients, and Design Basis Events;
- b.
The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AQOs, and < 280 cal/gm energy deposition for the rod ejection accident); and
- c.
The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
One limiting accident for the SDM requirements is based on a main steam line break (MSLB) in Mode 2, as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a break of a main steam line upstream of the Main Steam Isolation Valve initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post-trip return-to-power may occur; however, no fuel damage occurs as a result of the post-trip return-to-power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.
A potentially more limiting MSLB accident could occur for a steam line break outside containment when in Mode 3 with the low pressurizer pressure signal for safety injection actuation blocked. In this scenario, feedwater would not automatically isolate and the peak heat fluxes associated with the return-to-power may increase to values significantly McGuire Units 1 and 2 B 3.1.1-2 Revision No. 112
SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) greater than those in the accident analysis (Ref. 2). Therefore, when safety injection is blocked, administrative controls on boron concentration are required to prevent a return-to-power following a steam line break.
In addition to the limiting MSLB transient, the SDM requirement must also protect against:
- a.
Inadvertent boron dilution;
- b.
An uncontrolled rod withdrawal from subcritical or low power condition; and
- c.
Rod ejection.
Each of these events is discussed below.
In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.
Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.
The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure.
The ejection of a rod also produces a time dependent redistribution of core power. SDM satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.
Transients which are made less severe by the rapid insertion of control rod negative reactivity are also affected by the magnitude of the SDM limit. This is because the safety analyses assume a change in the rate of insertion of this negative reactivity when the SDM limit is reached. While the SDM is less than the limit value, the negative reactivity from the McGuire Units 1 and 2 B 3.1.1-3 Revision No. 112
SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) control rods is assumed to be inserted as quickly as the rod worth vs.
time curves shown in Reference 5. When the SDM limit value is reached, the rate of negative reactivity insertion is decreased so that it is only fast enough to compensate for any positive reactivity insertion, e.g., from the cooling of the fuel and moderator (which normally have negative temperature coefficients). This methodology is conservative in that it does not take credit in the safety analyses, even temporarily, for a SDM greater than the limit value.
LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.
The MSLB (Ref. 2) and the boron dilution (Ref. 4) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 5).
For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2 with keff > 1.0, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits,"
and LCO 3.1.6.
ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly.
A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.
In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid McGuire Units 1 and 2 B 3.1.1-4 Revision No. 112
SDM B 3.1.1 BASES ACTIONS (continued) storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.
In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/min when the RCS boron concentration is 1000 ppm and approximately 35 pcm/min when the RCS boron concentration is 100 ppm. If the emergency boration path is used, the CVCS is capable of inserting negative reactivity at the rate of 65 pcm/min when the RCS' boron concentration is 1000 ppm and 75 pcm/min when the RCS boron concentration is 100 ppm. Therefore, if SDM had to be increased by 1%
Ak/k or 1000 pcm, normal makeup path at 1000 ppm could restore SDM in approximately 33 minutes. At 100 ppm, SDM could be restored in approximately 29 minutes. In the emergency boration mode at 1000 ppm, the 1% Ak/k could be restored in approximately 15 minutes. With RCS boron concentration at 100 ppm, SDM could be increased by 1000 pcm in approximately 13 minutes using emergency boration. These boration parameters represent typical values and are provided for the purpose of offering a specific example.
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.
In MODE 2 with keff < 1.0 and MODES 3, 4, and 5, SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:
- a.
- b.
Control bank position;
- c.
RCS average temperature;
- d.
Fuel burnup based on gross thermal energy generation;
- e.
Xenon concentration; McGuire Units 1 and 2 B 3.1.1-5 Revision No. 112
SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
- f.
Samarium concentration; and
- g.
Isothermal temperature coefficient (ITC).
Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.
The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation.
REFERENCES
- 1.
10 CFR 50, Appendix A, GDC 26.
- 2.
UFSAR, Section 15.1.5.
- 3.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4.
UFSAR, Section 15.4.6.
- 5.
McGuire Units 1 and 2 B 3.1.1-6 Revision No. 112