L-MT-05-086, Update Monticello Nuclear Generating Plant Technical Specification Bases Change

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Update Monticello Nuclear Generating Plant Technical Specification Bases Change
ML052630318
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 09/16/2005
From: Conway J
Nuclear Management Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-05-086
Download: ML052630318 (13)
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Text

Monticello Nuclear Generating Plant Commitedto NuclearExcelec Operated by Nuclear Management Company, LLC September 16, 2005 L-MT-05-086 Technical Specification 6.8.K U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Monticello Nuclear Generating Plant Docket 50-263 License No. DPR-22 Undate Monticello Nuclear Generating Plant Technical Specification Bases Pages In accordance with the provisions of Monticello Nuclear Generating Plant (MNGP)

Specification 6.8.K, "Technical Specifications (TS) Bases Control Program," the Nuclear Management Company, LLC (NMC) is providing the enclosed changes for inclusion in the MNGP TS Bases.

Enclosure I provides a summary of the TS Bases changes. Enclosure 2 provides the current list of effective pages and records of revision (for information) and a typed copy of the revised TS Bases pages, for entry into the U.S. Nuclear Regulatory Commission Authority copy.

This letter contains no new commitments and makes no revisions to existing commitments.

John T. Conway Site Vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC Enclosures (2) cc: Administrator, Region ll, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC OND I Minnesota Department of Commerce 2807 West County Road 75

  • Monticello, Minnesota 55362-9637 Telephone: 763.295.5151
  • Fax: 763.295.1454

ENCLOSURE 1

SUMMARY

OF TECHNICAL SPECIFICATION BASES CHANGES Following is a summary of the Technical Specification Bases changes (TBSC) forwarded herein. The changes have been processed in accordance with Monticello Nuclear Generating Plant (MNGP) Specification 6.8.K, "Technical Specifications (TS)

Bases Control Program," and have been reviewed by the plant Operations Committee. provides a copy of the revised Bases pages.

1. TSBC-140a Technical Specification Bases Involved - 3.6/4.6 Item D.2 Page affected - 151 Summary of Change: The Bases for the Reactor Coolant System Leakage Detection Instrumentation incorrectly listed the drywell ventilation coolers as routed to the drywell equipment drain sump. This change corrects the Bases to reflect actual equipment configuration.
2. TSBC-141a Technical Specification Bases Involved - 3.3/4.3 Item B.3 Page affected - 88 Summary of Change: The Bases associated with the Rod Worth Minimizer (RWM) imply the sole reason for bypassing it is that it is physically unable to operate. In 2004, the NRC approved a report providing an improvement to the Banked Position Withdrawal Sequence acceptance criteria for control rod movement. This TBSC clarifies that the RWM may be bypassed whenever it is incapable of enforcing the control rod withdrawal / insertion sequence to be used.
3. TSBC-141b Technical Specification Bases Involved - 3.2, 3.7.E and 4.7.E Pages affected - 64, 182a and 190 Summary of Changes: These changes are associated with license amendments 138 and 140. Amendment 138 removed the Combustible Gas Control System (CGCS) from the TS. During the 2005 Refueling Outage (RFO) sealing of the CGCS containment penetrations was completed. Wording pertaining to CGCS in Bases sections 3.7.E and 4.7.E may now be completely removed from the TS.

Amendment 140 removed the reactor head cooling line from the TS. Bases section 3.2 was modified to clarify that the valves in this line provided containment isolation until the penetration for the reactor head cooling line was sealed during the 2005 RFO. Now that the penetration is sealed this statement can be removed.

Page 1 of 2

ENCLOSURE I

4. TSBC-141c Technical Specification Bases Involved - 3.4/4.4 Item A Page affected - 99 Summary of Change: The Bases currently state the design objective of the Standby Liquid Control System is to provide the capability of bringing the reactor from full power to a cold shutdown xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. Sufficient boron to bring the reactor from full power to a 3%delta k subcritical condition is injected (considering the other factors listed) in less than 125 minutes.

Global Nuclear Fuels (General Electric), the fuel vendor, now specifies a different requirement. The NRC approved fuel licensing report, GESTAR II states: "The Standby Liquid Control System (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from a full power and minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon-free state." This TBSC removes the 3% delta k subcritical condition and updates the design objective consistent with the current fuel design / licensing basis.

Page 2 of 2

ENCLOSURE 2 TECHNICAL SPECIFICATION LIST OF EFFECTIVE PAGES, RECORD OF REVISION, AND BASES CHANGES This enclosure provides numerically numbered 'change pages' for the Monticello Nuclear Generating Plant (MNGP) Technical Specification Bases page(s) incorporating the changes described herein. The affected Bases pages are designated with the amendment applicable at the time and the suffix ua."

This enclosure also provides, for information, the current, alphabetically ordered,

'change pages' for two indexes to the MNGP Technical Specifications. The first, "Appendix A, Technical Specifications Record of Revisions" provides a list of effective pages and corresponding amendment numbers. The second, the "Record of Technical Specification Changes and License Amendments," correlates between the amendment numbers and the subject of the amendment or bases changes.

The page(s) included in this enclosure and instructions for insertion into the Technical Specifications are provided below:

Remove the pages listed below Replace the removed pages with the and destroy. pages listed below.

A A B B J J 64 64 88 88 99 99 151 151 182a 182a 190 190 These replacement pages should be entered into the NRC Authority copy of the MNGP Technical Specifications.

9 pages follow

MONTICELLO NUCLEAR GENERATING PLANT APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Amend Amend Amend Page No. Page No. Page No. Paae No.

A 142 35 100a 71 100a 121 0 B 142 36 128 71 a 129b 122 135 C 115 37 128 72 104 123 117 D 115 38 128 76 0 124 121 E 115 39 129b 77 86 125 104 F 115 40 129b 78 0 126 137 G 115 42 138a 79 0 126a 137 H 119 45 0 80 29 127 137 1 130a 46 70 81 3 128 42 J 142 46a 37 82 123 129 122 K 141 47 40 82a 63 130 82 i 128 48 89 83 24 131 122 ii 138 49 140 83a 24 132 39 iii 120 50 128 *84 100a 132a 122 iv 128 50a 117 85 100a 133 106 v 120 51 117 86 100a 134 133 vi 121 51a 140 87 100a 135 133 vii 122 52 128 88 141 a 136 133 1 119 53 128 89 104 137 0 2 70 54 128 90 100a 138 100a 3 21 55 103 91 123 145 118a 4 102 56 102 92 100a 146 135 5 137 57 70 93 122 147 107

.5a 120 58 84 94 106 148 117 6 128 58a 141 95 77 149 100a 7 128 59 140 96 77 150 137a 8 128 59a 140 97 57 151 140a 9 128 60 128 98 56 152 137a 10 128 60a 31 99 141c 152a 137a 11 128 60b 62 100 141c 152b 137a 12 128 60c 30 101 122 153 100a 25a 127 60d 128 102 122 154 129a 25b 127 60e 89 103 122 155 122 25c 127 61 104 104 122 156 141 25d 127 62 117 105 122 157 130 26 5 63 117 106 79 158 141 27 81 63a 117 107 97 159 132 27a 81 64 141b 108 128 160 132 28 128 65 117 109 100a 163 141 29 128 66 119a 110 100a 164 141 30 103 67 117 111 133a 165 130 31 104 68 129b 112 130a, 166 130 32 103 69 129b 113 130a 167 112 33 103 69a 129b 114 133a 168 94 34 83 70 117 115 130a 169 94 A

Amendment No. 142 02/01/05

MONTICELLO NUCLEAR GENERATING PLANT  ;

APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Amend Page No. Page No.

170 141 217 128 171 130 218 120 171a 141 223 119 172 138 224 119 175 107 225 137b 175a 117 226 119 176 100a 229a 63 177 130 229b 138 178 100a 229c 104 179 123a 229d 138 180 130 229e 122 181 130 229u 104 182 130 229v 112 182a 141b 229vv 112 183 117 229w 112 184 100a 229ww 112 185 134 229x 112 188 104 22 9 y 115a 189 130 229z 112 190 141b 230 54 191 0 231 34 192 121 232 119 193 121 233 124 196 126a 234 119 197 121 235 115 198 121 236 115 199 51 243 128 200 129 244 124 201 129 248 59 202 129 249 142 I 203 41 250 128 204 129 251 124 204a 129 252 120 205 129 253 120 206 0 254 136 207 123 255 120 208 63 256 122 209 123 257 122 209a~ 100a 258 134 210 100a 258a 132 211 131 259 120 212 109 260 120 213 99 261 120 216 100a 262 120 B

Amendment No. 142 02/01/05

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS NSP License Revision DPR-22 (REV) No. Amend No. & Date Major Subiect 131 10/02/02 Update the Multiplier Values for Single Loop Operation Average Planar Linear Heat Generation Rate (APLHGR) 132 02/04/03 Conversion to Option B for Containment Leak Rate Testing 133 02/24/03 Revision to Pressure-Temperature CurvesI 133a 03/28/03 Bases Change - Adequate Reactor Steam Flow for HPCI/RCIC Testing 134 03/31/03 One-Time Extension of Containment Integrated Leak-Rate Test Interval 135 04/22/03 Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program 135a 04/24/03 Bases Change - Clarify description of head cooling Group 2 valves 136 06/17/03 Elimination of Requirements for Post Accident Sampling System 136a 09/25/03 Bases Change - Editorial change to define the abbreviation IEFCV."

137 08/21/03 Drywell Leakage and Sump Monitoring Detection System 137a 10/09/03 Bases Change - RCS Leakage Requirements for TS 3.6.4.D 137b 10/14/03 Bases Change - Clarification of system control boundary for ASDS 138 05/21/04 Elimination of Requirements for Hydrogen Recombiners and Hydrogen and Oxygen Monitors 138a 06/10/04 Bases Change - Clarification of Tech Spec Table 4.1.1 Manual Scram 139 06/02/04 Revised Analysis of Long-Term Containment Response and Net Positive Suction Head (Design Bases and USAR change) 140 11/02/04 Revision to Technical Specification Tables 3.2.1 and 3.2.4 140a 01/13/05 Bases Change - Removal of Drywell Vent Coolers from 3.6/4.6 Bases 141 01/28/05 Revision to Technical Specifications Table 3.2.3 and Section 3.7/4.7 141 a 02/24/05 Bases Change - Implement Improved BPWS as Described in NEDO-33091 -A 141 b 03/10/05 Bases Change - Bases Changes for License Amendments 138 and 140 141 c 03/10/05 Bases Change - Removal of 3% Delta-Kfrom Standby Liquid Control Bases 3.4.A/4.4.A 142 02/01/05 Deletion of Requirements for Submittal of Occupational Radiation Reports, Monthly Operating Reports, and Report of Safety/Relief Valve Failures and Challenges J

Amendment No. 142 02/01/05

....... I

. - - -11. - I - _. - - - - . _-__ -__ - - -___ - .. - I.- - . - . -_ .- - -, __ -

In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operators ability to control, or terminate a single operator error before it results in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, and other safety related functions. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required, and (ii) to prescribe the trip settings required to assure adequate performance. This set of Specifications also provides the limiting conditions of operation for the control rod block system.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by protective instrumentation shown in Table 3.2.1 which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines-of 10 CFR 100 are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus, the discussion given in the bases for Specification 3.1 is applicable here.

The low reactor water level instrumentation is set to trip when reactor water level is > 7' on the instrument. This corresponds to a lower water level inside the shroud at 100% power due to the pressure drop across the dryer/separator. This has been accounted for in the affected transient analysis. This trip initiates closure of Group 2 primary containment isolation valves. Reference Section 7.7.2.2 FSAR. The trip setting provides assurance that the valves will be closed before perforation of the clad occurs even for the maximum break in that line and therefore the setting is adequate.

The low low reactor water level instrumentation is set to trip when reactor water level is Ž-48'. This trip initiates closure of the Group 1 and Group 3 Primary containment isolation valves, Reference Section 7.7.2.2 FSAR, and also activates the ECC systems and starts the emergency diesel generators.

3.2 BASES 64 03/10/05 Amendment No.65, 81,100a, 102,117,128,135a, 141b

Bases 3.3/4.3 (Continued):

Should a control rod drop accident result in a peak fuel energy content of 280 cal/gm, less than 660 (7 x 7) fuel rods are conservatively estimated to perforate. This would result in offsite doses twice that previously reported in the FSAR, but still well below the guideline values of 10 CFR 100. For 8 x 8 fuel, less than 850 rods are conservatively estimated to perforate, which has nearly the same consequences as for the 7 x 7 fuel case because of the operating rod power differences.

The RWM provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. Reference Section 7-9 FSAR. It serves as an independent backup of the normal withdrawal procedure followed by the operator. In the 6vent that the RWM is not capable of enforcing a particular control rod withdrawal/insertion sequence when required, it is considered to be inoperable, and a second independent operator or engineer can manually fulfill the operator-follower control rod pattern conformance function of the RWM. In this case, procedural control is exercised by verifying all control rod positions after the withdrawal of each group, prior to proceeding to the next group. Allowing substitution of a second independent operator or engineer in case of RWM inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restricting plant operations. Above 10% power, there is no requirement that the RWM be operable since the control rod drop accident with out-of-sequence rods will result in a peak fuel energy content of less than 280 cal/gm. To assure high RWM availability, the RWM is required to be operating during a startup for the withdrawal of a significant number of control rods for any startup, after May 1, 1974.

4. The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has no scram function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The 3.3/4.3 BASES 88 02/24/05 Amendment No. 0, 400a, 141a

Bases 3.4/4.4:

A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon-free state without taking credit for control rod movement. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a subcritical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin to allow for leakage and imperfect mixing.

The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The ATWS Rule (10 CFR 50.62) requires the addition of a new design requirement to the generic SLC System design basis.

Changes to flow rate, solution concentration or boron enrichment to meet the ATWS Rule do not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution" (natural boron enrichment).

The described minimum system parameters (equivalent to 24 gpm, 10.7% concentration and 55 atom percent Boron-10 enrichment) will ensure an equivalent injection capability that meets the ATWS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being. made. A reliability analysis indicates that the plant can be operated safely in this manner for ten days. For additional margin, the allowable out of service time has been reduced to seven days.

3.4/4.4 BASES 99 03/10/05 Amendment No. 56, 5 7 , 77 , 1 0 4 r 141 c

Bases 3.6/4.6 (Continued!:

that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased leakage. This type of piping is very susceptible to IGSCC. Note also that once leakage is attributed to a specific source, that leakage can be considered to be identified and can be applied against the identified limit, rather than the unidentified limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable to properly reduce the Unidentified Leakage increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety.

The Surveillance Requirement (SR) associated with RCS leakage is acceptable because RCS leakage is monitored by a variety of instruments designed to provide alarms when leakage is Indicated and to quantify the various types of leakage. Sump level and flow rate are typically monitored to determine actual leakage rates; however, other methods may be used to verify leakage.

It is permissible to use preexisting information,.in conjunction with secondary measurements (e.g., drywell pressure and temperature), to verify that leakage'remains within limits by looking for step changes in conditions or to perform calculations to estimate leakage. The complete failure to demonstrate that RCS leakage Is within limits, on the required frequency, constitutes a failure to meet this SR, notwithstanding entrance into conditions and required actions of TS 3.6.D.2.

2. RCS Leakage Detection Instrumentation Two leakage collection sumps are provided inside primary containment. Identified leakage is piped from the recirculation pump seals, valve stem leak-offs, reactor vessel flange leak-off, and bulkhead and bellows drains to the drywell equipment drain sump.

All other leakage is collected In the drywell floor drain sump. Both sumps are equipped with level and flow transmitters connected to recorders in the control room. The Drywell Floor Drain Sump Monitoring System instrumentation consists of one floor drain sump flow integrator, one sump level recorder and one sump fill rate computer point (rate of change). The Drywell Floor Drain Sump Monitoring System is operable when any one of these three channels is operable. An annunciator and computer alarm are provided in the control room to alert operators when allowable leak rates are approached.

Drywell airborne particulate radioactivity is continuously monitored as well as drywell atmospheric temperature and pressure.

The drywell particulate radioactivity monitoring system monitors the drywell for airborne particulate radioactivity. A sudden increase in radioactivity may be attributed to RCPB steam or reactor water leakage. The drywell particulate radioactivity monitoring system is not capable of quantifying leakage rates, but Is sensitive enough to Indicate increased leakage rates. The drywell particulate radioactivity monitoring system provides a backup to the Drywell Floor Drain Sump Monitoring System and Is capable of monitoring leakage at least as low as 1Q`9 PCi/cc radioactivity for air particulate monitoring. Systems connected to the reactor coolant systems boundary are also monitored for leakage by the Process Liquid Radiation Monitoring System.

3.6/4.6 BASES 151 01/13/05 Amendment No. 30, 76, 93, 100a, 11i1, 12, 137a, 140a

With one or. more penetration flow paths with one PCIV inoperable, the affected penetration must be returned to operable status or isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for MSIVs and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for Excess Flow Check Valves (EFCVs)). The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time for MSIVs allows a period of time to restore the MSIVs to operable status given the fact that MSIV closure will result in a potential for plant shutdown. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time for EFCVs is reasonable considering the instrument and the small diameter of the penetration piping combined with the ability of the penetration to act as an isolation boundary. With one or more penetrations with two PCIVs inoperable, either the inoperable PCIVs must be returned to operable status or the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Specification 3.7.D.3 requires the containment to be purged and vented through the standby gas treatment system except during inerting and deinerting operations. This provides for iodine and particulate removal from the containment atmosphere.

Use of the 2-inch flow path prevents damage to the standby gas treatment system in the event of a loss of coolant accident during purging or venting. Use of the reactor building plenum and vent flow path for inerting and deinerting operations permits the control room operators to monitor the activity level of the resulting effluent by use of the Reactor Building Vent Wide Range Gas Monitors.

E. (Deleted) 3.7 BASES 182a 03/10/05 Amendment No. 430,-136a, 141 b

Bases 4.7 (Continued):

will be in the isolation position should an event occur. This required action does not require any testing or device manipulation.

Rather, it involves verification that those devices outside containment and capable of potentially being mispositioned are in the correct position. The completion time of "monthly" for devices outside containment is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low.: For the devices inside primary containment, the time period specified "prior to entering Startup or Hot Shutdown from Cold Shutdown, if primary containment was deinerted while in Cold Shutdown, if not performed in the previous 92 days" is based on engineering judgement and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility.

The surveillance requirements are modified by a footnote allowing both active and passive isolation devices, used to isolate a penetration; that are located in high radiation areas can be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these devices, once they have been verified in the proper position, is low.

The containment Is penetrated by a large number of small diameter Instrument lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27, 1973.

The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.

E. (Deleted) 4.7 BASES 190 03/10/05 Amendment No. 34 , 141 b

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