Text

Technical Requirements Manual, Affected Pages, Revision 15
	ML15335A497
Person / Time
Site:	Monticello
Issue date:	11/20/2015
From:	; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Reactor Regulation
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ML15335A486	List: L-MT-15-088, Technical Requirements Manual, Technical Specification Bases, Affected Pages, Revision 39; ML15335A482; ML15335A493; ML15335A494; ML15335A496; ML15335A497; ML15335A498; ML15335A499; ML15335A504; ML15335A505; ML15335A506; ML15335A507; ML15335A508; ML15335A509; ML15335A510; ML15335A511; ML15335A512;
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TRM Control Rod Block Instrumentation 3.3.2.1 Monticello 3.3.2.1-1 Revision 15 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation TLCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.2.1-1 ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE-------------

Only applicable to Functions 1, 2, 3 and 5.

One or more Functions with one required channel inoperable.

A.1 Restore channel to OPERABLE status.

7 days B. ------------NOTE-------------

Only applicable to Functions 1, 2, 3, and 5.

One or more Functions with two required channels inoperable.

B.1 Place channel in the tripped condition.

OR B.2 Suspend control rod withdrawal.

Immediately Immediately C. One or more required Function 4 channels inoperable.

C.1 Place channel in the tripped condition.

OR C.2 Suspend control rod withdrawal.

Immediately Immediately

TRM Control Rod Block Instrumentation 3.3.2.1 Monticello 3.3.2.1-2 Revision 15 SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

1.

Refer to Table 3.3.2.1-1 to determine which TSRs apply for each Control Rod Block Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillance, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability.

SURVEILLANCE FREQUENCY TSR 3.3.2.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TSR 3.3.2.1.2

NOTE-------------------------------

1. For Function 1.b, not required to be performed if SRM detectors are secured in the full-in position.

2. For Function 2.a and 2.b, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

7 days TSR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST.

92 days TSR 3.3.2.1.4 Perform CHANNEL CALIBRATION.

92 days TSR 3.3.2.1.5

NOTE-------------------------------

1. Neutron detectors are excluded.

2. For Function 2.a and 2.b, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION.

24 months TSR 3.3.2.1.6 Perform CHANNEL CALIBRATION.

24 months TSR 3.3.2.1.7 Perform CHANNEL FUNCTIONAL TEST.

184 days

TRM Control Rod Block Instrumentation 3.3.2.1 Monticello 3.3.2.1-3 Revision 15 Table 3.3.2.1-1 (Page 1 of 2)

Control Rod Block Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Source Range Monitors

a.

Upscale 2(a), 5 1

TSR 3.3.2.1.1 TSR 3.3.2.1.2 TSR 3.3.2.1.5 1.16 x 105cps

b.

Detector Not Fully Inserted 2(b), 5(b) 1 TSR 3.3.2.1.2 TSR 3.3.2.1.5 NA

2.

Intermediate Range Monitors

a.

Downscale 2(c), 5(c) 2(d)

TSR 3.3.2.1.1 TSR 3.3.2.1.2 TSR 3.3.2.1.5 3/125 divisions of full scale

b.

Upscale 2, 5 2(d)

TSR 3.3.2.1.1 TSR 3.3.2.1.2 TSR 3.3.2.1.5 109.5/125 divisions of full scale

3.

Average Power Range Monitors

a.

Simulated Thermal Power - High 1

3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 0.61W + 61.2% RTP(e) and < 110% RTP

b.

Downscale 1

3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 2/125 divisions of full scale

c.

Neutron Flux - High (Setdown) 2 3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 15%

4.

Scram Discharge Volume

a.

East Water Level High 1, 2 1

TSR 3.3.2.1.3 TSR 3.3.2.1.4 40 gal

b.

West Water Level High 1, 2 1

TSR 3.3.2.1.3 TSR 3.3.2.1.4 40 gal (a)

With IRMs on Range 6 or below.

(b)

With SRM channel count rate < 100 cps and IRMs on Range 2 or below.

(c)

With IRMs on Range 2 or above.

(d)

There must be at least one OPERABLE IRM channel monitoring each core quadrant.

(e) 0.55(W - Delta W) + 55.5% when Technical Specification 3.3.1.1 Function 2.b, is reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating. The value of Delta W is defined in the COLR.

Single loop operation is not permitted while operating in the MELLLA+ operating domain.

(f)

Each APRM channel provides input to both trip systems.

TRM Control Rod Block Instrumentation 3.3.2.1 Monticello 3.3.2.1 Last Revision 15 Table 3.3.2.1-1 (Page 2 of 2)

Control Rod Block Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5.

Average Power Range Monitors (Automated Backup Stability Protection (BSP))

a.

Slope 1(g) 3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 1.3

b.

Constant Power Line 1(g) 3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 30% RTP

c.

Constant Flow Line 1(g) 3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 58.8% Rated Drive Flow (RDF)

d.

Flow Breakpoint 1(g) 3(f)

TSR 3.3.2.1.6 TSR 3.3.2.1.7 34.5% RDF (f)

Each APRM channel provides input to both trip systems.

(g)

Required only when the Automated BSP Scram Region is implemented in accordance with Technical Specification 3.3.1.1.

TRM Safety/Relief Valves Out-of-Service 3.4.4 Monticello 3.4.4-1 Revision 15 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 Safety/Relief Valves (S/RVs) Out-of-Service TLCO 3.4.4 The safety function of eight S/RVs shall be OPERABLE.

APPLICABILITY:

MODE 1, when operating in the MELLLA+ operating domain.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more S/RVs inoperable.

A.1 Restore eight S/RVs to OPERABLE status.

14 days B. Required Action and associated Completion Time of Condition A not met.

B.1 Exit the Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

Operating Domain.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Required Action and associated Completion Time of Condition B not met.

C.1 Enter TLCO 3.0.3.

Immediately SURVEILLANCE REQUIREMENTS There are no additional surveillance requirements beyond those specified in Technical Specification 3.4.3.

Control Rod Block Instrumentation B 3.3.2.1 Monticello B 3.3.2.1-1 Revision 15 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR remains above the Safety Limit (Technical Specification 2.1.1). The trip logic for this function is 1 out of n; e.g., any trip on one of the four APRM's, eight IRM's, four SRM's, or four scram discharge volume water level channels will result in a rod block. For each Control Rod Block Function, there are two trip systems. The scram discharge volume water level instrumentation includes two sensors on each of the two scram discharge volumes. This assures that no control rod is withdrawn unless enough capacity is available in either scram discharge volume to accommodate a scram. The setting is selected to initiate a rod block no later than the scram that is initiated on scram discharge volume high water level.

The minimum instrument channel requirements for the IRM may be reduced by one for a short period of time to allow for maintenance, testing, or calibration. See Section 7.3 FSAR.

The APRM Simulated Thermal Power - High rod block (Refs. 3 and 4) is referenced to flow and prevents operation significantly above the licensing basis power level especially during operation at reduced flow. For operation at low power (i.e., MODE 2), the APRM Neutron Flux -

High (Setdown) Function (Ref. 3) is capable of generating a rod block to prevent fuel damage resulting from abnormal operating transients in this power range. The APRMs provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The operator will set the APRM rod block trip settings no greater than that stated in Table 3.3.2.1-1. However, the actual setpoint can be as much as 3% greater than that stated in Table 3.3.2.1-1 for recirculation driving flows less than 50% of design and 2%

greater than that shown for recirculation driving flows greater than 50% of design due to the deviations that could be caused by inherent instrument error, operator setting error, drift of the setpoint, etc.

The APRM Backup Stability Protection (BSP) Flow-Bias rod blocks are active when the Automated Backup Stability Protection (ABSP) function is enabled. The BSP Flow-Bias rod blocks provide a rod block for reactivity transients when operating at low recirculation flows with the OPRMs out of service. These rod blocks provide a warning of potential ABSP scrams. The constant flow line and flow breakpoint are in terms of rated (recirculation) drive flow or RDF (see Ref. 5). Addition of these rod block functions was approved by Amendment No. 180 (Ref. 6).

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.

Analysis of the worst case accident results in rod block action before MCPR approaches the Safety Limit (Technical Specification 2.1.1).

A downscale indication of an IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus control rod motion is prevented. The downscale IRM rod block assures that there will be proper overlap between the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation. The downscale IRM rod block is set at 3/125 of full scale.

Control Rod Block Instrumentation B 3.3.2.1 Monticello B 3.3.2.1 Last Revision 15 BASES Although the operator will set the setpoints within the trip settings specified in Table 3.3.2.1-1, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. The deviations could be caused by inherent instrument error, operator setting error, drift of the set point, etc. Therefore, these deviations have been accounted for in the various transient analyses.

Trip Function Deviation IRM Downscale

- 2/125 of Scale IRM Upscale

+ 2/125 of Scale APRM Downscale

- 2/125 of Scale APRM Upscale

+ 3% for recirculation driving flows < 50% of design

+ 2% for recirculation driving flows > 50% of design Scram Discharge Volume-High Level

+ 1 gallon The instrumentation in this section will be functionally tested and calibrated at regularly scheduled intervals. The 184 day CHANNEL FUNCTIONAL TEST and 24 month CHANNEL CALIBRATION surveillance frequencies for the APRM Simulated Thermal Power - High, APRM Downscale, and APRM Neutron Flux - High (Setdown) rod block functions are consistent with the NUMAC PRNMS design assumptions (Refs. 1 and 2). Although this instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the same design reliability goals are applied. Where applicable, sensor checks are specified on a once/12 hours basis.

REFERENCES

1.

NEDC-32410P-A, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, October 1995.

2.

NEDC-32410P-A, Supplement 1, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, November 1997.

3.

Amendment No. 159, Issuance of Amendment Re: Request to Install Power Range Neutron Monitoring System, dated February 3, 2009. (ADAMS Accession No. ML083440681)

4.

Calculation 08-052, Instrument Setpoint Calculation - Average Power Range Monitor (APRM) Flow Biased PRNM Setpoints for EPU, Revision 2.

5.

Calculation 12-043 Average Power Range Monitor NUMAC PRNM Setpoints - MELLLA+ Automatic Backup Stability Protection (ABSP),

Revision 0.

6.

Amendment No. 180, Monticello Nuclear Generating Plant - Issuance of Amendment No. 180 to Renewed Facility Operating License Regarding MELLLA+, dated March 28, 2014. (ADAMS Accession No. ML14035A248)

TRM Safety/Relief Valves Out-of-Service B 3.4.4 Monticello B 3.4.4-1 Revision 15 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 Safety/Relief Valves (S/RVs) Out-of-Service BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

Technical Specification LCO 3.4.3, "Safety/Relief Valves (S/RVs),

provides safety mode requirements for the limiting design basis event of closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux. This TLCO imposes an additional restriction that all eight S/RVs be OPERABLE to meet reactor vessel overpressure protection limits for operation within the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain for an Anticipated Transient Without Scram (ATWS) event. Reactor vessel pressure is required to remain within ASME Code Service Level C limits of 1500 psig for the postulated ATWS event in the MELLLA+ operating domain.

APPLICABLE The overpressure protection system must accommodate the most severe SAFETY pressurization transient. Evaluations have determined that the most ANALYSES severe design basis event is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, five S/RVs are assumed to operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure as described in Technical Specification 3.4.3.

However, two additional S/RVs are required to be OPERABLE to provide additional relief capability per Technical Specification 3.4.3.

For the purpose of the ATWS analyses occurring within the MELLLA+

operating domain, eight S/RVs are assumed to operate in the safety mode (Ref. 2). Consequently, the S/RV Out of Service (SRVOOS) flexibility option is not permitted during operation in the MELLLA+

operating domain. Analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Service Level C Code limit of 1500 psig. This TLCO helps to ensure that this acceptance limit of 1500 psig is met if an ATWS were to occur while operating in the MELLLA+ operating domain.

TRM Safety/Relief Valves Out-of-Service B 3.4.4 Monticello B 3.4.4-2 Revision 15 BASES TLCO The safety mode of eight S/RVs are required to be OPERABLE to satisfy the assumptions of the MELLLA+ safety analysis (Ref. 2). The requirements of this TLCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (valve safety function).

The S/RV setpoints are established to ensure that the ASME Code Service Level C limit on peak reactor pressure is satisfied. Operation with fewer than eight valves OPERABLE, or with setpoints outside the ASME limits, could result in a more severe reactor response to an ATWS than predicted, possibly resulting in the ASME Code Service Level C limit on reactor pressure being exceeded for an ATWS event that originates within the MELLLA+ operating domain.

APPLICABILITY In MODE 1 all eight S/RVs must be OPERABLE in the MELLLA+

operating domain, since considerable energy may be in the reactor core and the limiting ATWS event is assumed to occur in this MODE. The lower end of the MELLLA+ operating domain is approximately 70.2% of Rated Thermal Power (RTP), which is not achievable in the other operating modes. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

ACTIONS A.1 The TLCO requires eight S/RVs to be OPERABLE to provide overpressure protection for a postulated ATWS event in the MELLLA+

operating domain. With less than the number of S/RVs specified OPERABLE, an overpressure event could result in violation of the ASME Code Service Level C limit on reactor pressure based on the licensing basis overpressure analysis ATWS in the MELLLA+ operating domain.

The Required Action and associated Completion Time are consistent with Section 9.3.1.1 and Appendix B, Condition 12.18.d, of Reference 2. For this reason, continued operation with an S/RV inoperable is permitted for a limited time.

The 14 day Completion Time to restore inoperable S/RVs to OPERABLE status is based on the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. This Required Action aligns with the the Required Action and associated Completion Time for Technical Specification 3.4.3 when an S/RV is inoperable.

TRM Safety/Relief Valves Out-of-Service B 3.4.4 Monticello B 3.4.4 Last Revision 15 BASES ACTIONS (continued)

B.1 If the safety function of the inoperable S/RVs cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a condition in which the TLCO does not apply. To achieve this status, the MELLLA+ operating domain must be exited within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time is reasonable, based on similar plant operating experience, to exit the MELLLA+ operating domain from full power conditions in an orderly manner and without challenging plant systems.

C.1 If the MELLLA+ operating domain cannot be exited within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the assumption on the number of S/RVs credited in the safety analyses to provide to an overpressure protection for an ATWS event in the MELLLA+

operating domain is not met and the unit is in a condition outside the accident analyses. Therefore, TLCO 3.0.3 must be entered immediately.

SURVEILLANCE REQUIREMENTS There are no surveillance requirements associated with this TLCO.

REFERENCES

1.

USAR, Section 14.5.1.

2.

NEDC-33453P, Revision 1, Maximum Extended Load Limit Analysis Plus (MELLLA+) Safety Analysis Report.

3.

Amendment No. 180, Monticello Nuclear Generating Plant - Issuance of Amendment No. 180 to Renewed Facility Operating License Regarding MELLLA+, March 28, 2014. (ADAMS Accession No. ML14035A248).