ML020630032
| ML020630032 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/14/2002 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| B18556 | |
| Download: ML020630032 (139) | |
Text
{{#Wiki_filter:Dominion Nuclear Connecticut, Inc. Dom inion -.... Millstone Power Station Rope Ferry Road Waterford, CT 06385 FEB I 4 2002 Docket Nos. 50-336 50-423 B18556 RE: 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit Nos. 2 and 3 Technical Specifications Change Requests 2-21-01 and 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby proposes to amend Operating Licenses DPR-65 and NPF-49 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit Nos. 2 and 3, respectively. For Millstone Unit No. 2, DNC is proposing to change Technical Specifications 3/4.4.9.1, "Pressure/Temperature Limits," 3/4.7.2.1, "Steam Generator Pressure/Temperature Limitation," 3/4.7.5.1, "Flood Level," 3/4.7.7.1, "Sealed Source Contamination," and 3/4.7.8, "Snubbers." Index pages VIII and XIII will also be changed consistent with the relocation of the identified technical specifications. For Millstone Unit No. 3, DNC is proposing to change Technical Specifications 3/4.4.9.1, "Pressure/Temperature Limits," 3/4.7.2, "Steam Generator Pressure/Temperature Limitation," 3/4.7.6, "Flood Protection," 3/4.7.10, "Snubbers," 3/4.7.11, "Sealed Source Contamination," and 3/4.7.14, "Area Temperature Monitoring." Index pages viii, x, xiv and xv for Millstone Unit No. 3 will also be changed consistent with the relocation of the identified technical specifications. The Bases of the affected technical specifications will be modified to address the proposed changes. The proposed changes will relocate selected Millstone Unit Nos. 2 and 3 technical specifications related to the Reactor Coolant System and Plant Systems to the respective Technical Requirements Manual (TRM). Information which is relocated to the TRM will be maintained in accordance with the provisions of 10 CFR 50.59.
U.S. Nuclear Regulatory Commission B 18556/Page 2 provides a discussion of the proposed changes and the Safety Summary. provides the Significant Hazards Consideration. Attachment 3 provides the marked-up version of the appropriate pages of the current Technical Specifications for Millstone Unit No. 2. Attachment 4 provides the retyped pages of the Technical Specifications and associated Bases for Millstone Unit No. 2. Attachment 5 provides the marked-up version of the appropriate pages of the current Technical Specifications for Millstone Unit No. 3. Attachment 6 provides the retyped pages of the Technical Specifications and associated Bases for Millstone Unit No. 3. Environmental Considerations DNC has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.22. DNC has determined that the proposed changes meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that the changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or that changes a surveillance requirement, and that the amendment request meets the following specific criteria. (i) The proposed changes involve no Significant Hazards Consideration. As demonstrated in Attachment 2, the proposed changes do not involve a Significant Hazards Consideration. (ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released off site. The proposed changes will relocate selected Millstone Unit Nos. 2 and 3 technical specifications related to the Reactor Coolant System and Plant Systems to the respective TRM. However, the operability requirements for equipment associated with these technical specifications will remain the same. The proposed changes are consistent with the design basis of the plant. The proposed changes will not result in an increase in power level, will not increase the production of radioactive waste and byproducts, and will not alter the flowpath or method of disposal of radioactive waste or byproducts. Therefore, the proposed changes will not increase the type and amounts of effluents that may be released off site. (iii) There is no significant increase in individual or cumulative occupational radiation exposure.
U.S. Nuclear Regulatory Commission B18556/Page 3 The proposed changes will relocate selected Millstone Unit Nos. 2 and 3 technical specifications related to the Reactor Coolant System and Plant Systems to the respective TRM. However, the operability requirements for equipment associated with these technical specifications will remain the same. The proposed changes will not result in changes in the configuration of the facility. There will be no change in the level of controls or methodology used for processing radioactive effluents or the handling of solid radioactive waste. There will be no change to the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from the proposed changes. Conclusions The proposed changes were evaluated and we have concluded that they are safe. The proposed changes do not involve an impact on public health and safety (see the Safety Summary provided in Attachment 1) and do not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92 (see the Significant Hazards Consideration provided in Attachment 2). Site Operations Review Committee and Nuclear Safety Assessment Board The Site Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with the determinations. Schedule We request issuance of these amendments for Millstone Unit Nos. 2 and 3 prior to January 31, 2003, with each amendment to be implemented within 90 days of issuance. State Notification In accordance with 10 CFR 50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut. There are no regulatory commitments contained within this letter.
U.S. Nuclear Regulatory Commission B18556/Page 4 If you should have any questions on the above, please contact Mr. Ravi Joshi at (860) 440-2080. Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC. J. Alan[Pic~e Site ViWe.President - Millstone this/ d f20 Sworn to and subscribed before me Notary Pir ic My Commission expires SANDRAJ. ANTON NOTARY PUBLIC COMMISSION EXPIRES MAY31,2005 Attachments (6) cc: H. J. Miller, Region I Administrator J. T. Harrison, NRC Project Manager, Unit No. 2 NRC Senior Resident Inspector, Unit No. 2 V. Nerses, NRC Senior Project Manager, Millstone Unit No. 3 NRC Senior Resident Inspector, Millstone Unit No. 3 Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Docket Nos. 50-336 50-423 B18556 Attachment 1 Millstone Nuclear Power Station, Unit Nos. 2 and 3 Technical Specifications Change Requests 2-21-01 and 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Discussion of Proposed Changes
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 1 Technical Specifications Change Requests 2-21-01 and 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Discussion of Proposed Changes
Background
Dominion Nuclear Connecticut, Inc. (DNC) hereby proposes to amend Operating Licenses DPR-65 and NPF-49 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit Nos. 2 and 3. DNC is proposing to relocate selected technical specifications for Millstone Unit Nos. 2 and 3. For Millstone Unit No. 2, DNC is proposing to relocate part of Technical Specification 3/4.4.9.1, "Pressure/Temperature Limits," and all of Technical specifications 3/4.7.2.1, "Steam Generator Pressure/Temperature Limitation," 3/4.7.5.1, "Flood Level," 3/4.7.7.1, "Sealed Source Contamination," and 3/4.7.8, "Snubbers," to the Technical Requirements Manual (TRM) for Millstone Unit No. 2. For Millstone Unit No. 3, DNC is proposing to relocate part of Specification 3/4.4.9.1, "Pressure/Temperature Limits," and all of Technical Specifications 3/4.7.2, "Steam Generator Pressure/Temperature Limitation," 3/4.7.6, "Flood Protection," 3/4.7.10, "Snubbers," 3/4.7.11, "Sealed Source Contamination," and 3/4.7.14, "Area Temperature Monitoring," to the TRM for Millstone Unit No. 3. The Bases of the associated technical specifications will also be moved to the Millstone Unit No. 2 or 3 TRM, as applicable. Additional background information will be included, as necessary, to explain these changes. The Millstone Unit Nos. 2 and 3 TRMs include information which has been relocated from Technical Specifications or material which has been judged to warrant administrative control. Modifications to the TRM, which is maintained as a controlled document, are performed pursuant to the provisions of 10 CFR 50.59. The TRM is referenced by both the Millstone Unit No. 2 and 3 Final Safety Analysis Reports (FSAR). The proposed changes are described below: Technical Specification 3/4.4.9 The requirements of Surveillance Requirement 4.4.9.1b and Table 4.4-5 of Millstone Unit No. 2 Technical Specifications and Surveillance Requirement 4.4.9.1.2 and Table 4.4-5 of Millstone Unit No. 3 Technical Specifications will be relocated to the respective facility's TRM where future changes will be controlled in accordance with 10 CFR 50.59. The text for the surveillances will be replaced with the word "DELETED," and the tables will be deleted and replaced with, "This page intentionally left blank." The relocation of these surveillance requirements to the respective facility's TRM is in accordance with
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 2 the Standard Technical Specifications for Westinghouse and Combustion Engineering Plants, (1), (2) and the guidance provided in Generic Letter (GL) 91-01.(') Millstone Unit No. 2 Technical Specifications 3/4.7.2, 3/4.7.5.1, 3/4.7.7.1, 3/4.7.8, and Millstone Unit No. 3 Technical Specifications 3/4.7.2, 3/4.7.6, 3/4.7.10, 3/4.7.11, 3/4.7.14 Millstone Unit Nos. 2 and 3 Technical Specifications will be relocated to the respective facility's TRM where future changes will be controlled in accordance with 10 CFR 50.59. The text on the corresponding pages will be deleted and replaced with, "This page intentionally left blank." Index Pages Index pages VIII and XIII for Millstone Unit No. 2 and index pages viii, x, xiv and xv for Millstone Unit No. 3 will be revised to reflect the relocation of the identified technical specifications to the respective facility's TRM. Technical Specification Bases The proposed changes to the Bases for Millstone Unit Nos. 2 and 3 Surveillance Requirements 4.4.9.1b and 4.4.9.1.1 will delete the corresponding text. The proposed changes to the Bases for Millstone Unit No. 2 Technical Specifications 3/4.7.2, 3/4.7.5.1, 3/4.7.7.1, 3/4.7.8, and Millstone Unit No. 3 Technical Specifications 3/4.7.2, 3/4.7.6, 3/4.7.10, 3/4.7.11, 3/4.7.14 will delete the text associated with each section and replace the section titles with the word, "DELETED." The deleted BASES will be relocated to the respective TRM for each unit. Safety Summary 10 CFR 50.36c(2)(ii) contains the requirements for items that must be in Technical Specifications. This regulation provides four (4) criteria that can be used to determine the requirements that must be included in the Technical Specifications. Items not meeting any of the four criteria can be relocated from Technical Specifications to a Licensee controlled document. The Licensee can then change the relocated requirements, if necessary, in accordance with 10 CFR 50.59. This should result in significant reductions in time and expense to modify requirements that have been "(1)"Standard Technical Specification, Westinghouse Plants," NUREG-1431, Rev. 2, April, 2001. "(2)"Standard Technical Specification, Combustion Engineering Plants," NUREG-1432, Rev. 2, April, 2001. (3) Nuclear Regulatory Commission, Generic Letter 91-01, "Removal of The Schedule For The Withdrawal of Reactor Vessel Material Specimens From Technical Specifications," January 4, 1991.
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 3 relocated while not adversely affecting plant safety. The criteria, and an evaluation of each technical specification proposed for relocation, are provided below. Technical Specification 3/4.4.9 The relocation of the requirements of Surveillance Requirement 4.4.9.1b and Table 4.4-5 of Millstone Unit No. 2 and Surveillance Requirement 4.4.9.1.2 and Table 4.4-5 of Millstone Unit No. 3 to the respective facility's TRM, where future changes will be controlled in accordance with 10 CFR 50.59, is in accordance with the Standard Technical Specifications for Westinghouse and Combustion Engineering Plants, and the guidance provided in GL 91-01. Technical Specifications (TSs) include limiting conditions for operation that establish pressure and temperature limits for the reactor coolant system. The limits are defined by TS figures that provide an acceptable range of operating temperatures and pressures for heatup, cooldown, criticality, and inservice leak and hydrostatic testing. These limits are generally valid for a specified number of effective full-power years. Reactor vessel material surveillance, through periodic withdrawal and testing of the reactor vessel material specimens, ensures the availability of data to update the inservice operating temperature and pressure limits. This surveillance duplicates the requirements of 10 CFR 50, Appendix H to monitor changes in the fracture toughness of the ferritic materials in the reactor vessel beltline. Surveillance Requirement 4.4.9.1b of Millstone Unit No. 2 and Surveillance Requirement 4.4.9.1.2 of Millstone Unit No. 3 require the reactor vessel material irradiation surveillance specimen to be removed and examined in accordance with 10 CFR 50, Appendix H, and the results be considered in the development of the TS Figures. Improved TS 3.4.3 deletes this requirement on the basis that 10 CFR 50, Appendix H requires that the reactor vessel material irradiation surveillance specimen to be removed and examined to monitor changes in fracture toughness properties and that the information be used to update pressure and temperature limits in the TS. Inclusion of these requirements in the TS constitutes a duplication of the requirements outlined in 10 CFR 50, Appendix H. Millstone Unit Nos. 2 and 3 are required to comply with the provisions of 10 CFR 50, Appendix H, therefore, these Surveillance Requirements are not required to be in the TS. The relocation of these surveillance requirements to the respective facility's TRM is consistent with the Standard TSs for Westinghouse and Combustion Engineering Plants (NUREG-1431 and NUREG-1432), and the guidance provided in GL 91-01. Technical Specification 3/4.7.2.1 for Millstone Unit No. 2 and 3/4.7.2 for Millstone Unit No. 3 TS 3/4.7.2.1 for Millstone Unit No. 2 and 3/4.7.2 for Millstone Unit No. 3 are proposed to be relocated to the respective facility's TRM. These specifications provide limitation on steam generator pressure and temperature which ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 4 toughness stress limits. For Millstone Unit No. 2, the limitations of 70 OF and 200 psig are based on a steam generator nil ductility transition temperature (RTNDT) of 50 °F and are sufficient to prevent brittle fracture. For Millstone Unit No. 3, the limitations of 70 OF and 200 psig are based on a steam generator RTNDT of 60 OF and are sufficient to prevent brittle fracture. Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The limitations imposed by these two specifications on steam generators temperature and pressure are provided to prevent brittle fracture in the steam generators and do not describe a limitation on instrumentation that is used to detect, and indicate in the control room, an abnormal degradation of the reactor coolant pressure boundary. These specifications do not meet Criterion 1. Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The limitations imposed by these two specifications on steam generators temperature and pressure are provided to prevent brittle fracture in the steam generators. The limiting temperature and pressure values described in these specifications do not provide direct input to Reactor Protection System or Engineered Safety Features Actuation System functions, nor do they represent process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Criterion 3 A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These specifications do cover a component (steam generator) that is part of the primary success path (heat removal) which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the specification limits apply to conditions when the Reactor Coolant System temperature is
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 5 unusually low (< 70 0 F). Under these conditions, the Steam Generators are not required to function to mitigate any DBA or transient. Therefore, this specification does not satisfy criterion 3. The Steam Generator heat removal function is covered in the Reactor Coolant Loop specifications. Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The limits covered by these specifications have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. This TS does not cover a SSC requiring risk review/unavailability monitoring. This specification does not meet Criterion 4. These TSs do not fulfill any one or more of the 10 CFR 50.36c(2)(ii) criteria on items for which TSs must be established. Therefore, these TSs can be relocated to the TRM. Technical Specification 3/4.7.5.1 for Millstone Unit No. 2 and 3/4.7.6 for Millstone Unit No. 3 TS 3/4.7.5.1 for Millstone Unit No. 2 and 3/4.7.6 for Millstone Unit No. 3 are proposed to be relocated to the respective facility's TRM. TS 3/4.7.5.1 for Millstone Unit No. 2 ensures that one service water pump motor will be protected against flooding to a minimum elevation of 28 feet. The service water pump motors are normally protected against water damage to an elevation of 22 feet. If the water level is exceeding plant grade level or if a severe storm is approaching the plant site, one service water pump motor will be protected against flooding to a minimum elevation of 28 feet to ensure that this pump will continue to be capable of removing decay heat from the reactor. Action to provide pump motor protection will be initiated when the water level reaches plant grade level in order to ensure operator accessibility to the intake structure. TS 3/4.7.6 for Millstone Unit No. 3 ensures that the service water pump cubicle watertight doors will be closed and the pump cubicle sump drain valves will be closed before the water level reaches the critical elevation of 14.5 feet Mean Sea Level (MSL). Elevation 14.5 feet MSL is the floor elevation of the service water pump cubicle. Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. These TSs cover protective actions against flooding of the service water pump motors. They do not cover installed instrumentation which is used to detect, and
U.S. Nuclear Regulatory Commission B1 8556/Attachment 1/Page 6 indicate in the control room, an abnormal degradation of the reactor coolant pressure boundary. These specifications do not meet Criterion 1. Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs cover protective actions against flooding of the service water pump motors. They do not cover a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These specifications do not satisfy criterion 2. Criterion 3 A SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs ensure that service water pump motors are protected against flooding conditions. These TSs do not affect the ability of a SSC, that is part of the primary success path and which functions or actuates to mitigate a design basis accident or a transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, from performing its safety function. These specifications do not satisfy criterion 3. Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The limitations covered by these TSs have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. These specifications do not meet Criterion 4. These TSs do not fulfill any one or more of the 10 CFR 50.36c(2)(ii) criteria on items for which TSs must be established. Therefore, these TSs can be relocated to the TRM. Technical Specification 3/4.7.8 for Millstone Unit No. 2 and 3/4.7.10 for Millstone Unit No. 3 TS 3/4.7.8 for Millstone Unit No. 2 and 3/4.7.10 for Millstone Unit No. 3 are proposed to be relocated to the respective facility's TRM. The snubbers are required to be operable to ensure the structural integrity of the Reactor Coolant System and all other safety
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 7 related systems is maintained during and following a seismic or other event initiating dynamic loads. The restraining action of the snubbers ensures that the initiating event failure does not propagate to other parts of the failed system or to other safety systems. Snubbers also allow thermal expansion of piping and nozzles to eliminate excessive thermal stresses during heatup or cooldown. Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. These TSs do not cover installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. These specifications do not satisfy criterion 1. Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs only require snubbers to be operable. Snubbers are not a design feature that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These specifications do not satisfy criterion 2. Criterion 3 A SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs only require snubbers to be operable. Snubbers are not required to mitigate any DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These specifications do not satisfy criterion 3. Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. Snubbers have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. These specifications do not meet Criterion 4.
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 8 These TSs do not fulfill any one or more of the 10 CFR 50.36c(2)(ii) criteria on items for which TSs must be established. Therefore, these TSs can be relocated to the TRM. Technical Specification 3/4.7.7 for Millstone Unit No. 2 and 3/4.7.11 for Millstone Unit No. 3 TS 3/4.7.7 for Millstone Unit No. 2 and 3/4.7.11 for Millstone Unit No. 3 are proposed to be relocated to the respective facility's TRM. The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. Leakage of sources excluded from the requirements of this specification represent less than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested. Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are not frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shield mechanism. Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. These TSs provide limitations on sealed source removable contamination to ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. Therefore, these TSs do not cover installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. These specifications do not satisfy criterion 1. Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. They are not applicable to a process variable, design feature, or operating restriction that is an initial condition of a DBA or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, these specifications do not meet Criterion 2.
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 9 Criterion 3 A SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These TSs cover limitations on sealed source removable contamination to ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. They do not cover a structure, system, or component that is part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, these specifications do not satisfy criterion 3. Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The limitations covered by these TSs have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. These specifications do not meet Criterion 4. These TSs do not fulfill any one or more of the 10 CFR 50.36c(2)(ii) criteria on items for which technical specifications must be established. Therefore, these TSs can be relocated to the TRM. Technical Specification 3/4.7.14 for Millstone Unit No. 3 TS 3/4.7.14 for Millstone Unit No. 3 is proposed to be relocated to the TRM. The area temperature limitations imposed by this TS ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of +/-2.2 0 F. Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. This TS does not cover installed instrumentation that is used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. This specifications do not satisfy criterion 1.
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 10 Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Area Temperatures are not process variables that are an initial condition of a DBA or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These temperature limits are requirements of the Equipment Environmental Qualification (EEQ) program. Area Temperatures do not satisfy criterion 2. Criterion 3 A SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Area Temperature Monitoring is not a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Area Temperature monitoring does not satisfy criterion 3. Criterion 4 A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The limitations covered by this TS have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. This specification does not meet Criterion 4. This TS does not fulfill any one or more of the 10 CFR 50.36c(2)(ii) criteria on items for which TSs must be established. Therefore, this TS can be relocated to the TRM. Index Pages Revision of Index Pages VIII and XIII for Millstone Unit No. 2 and Index Pages viii, x, xiv and xv for Millstone Unit No. 3 are administrative changes. These changes are consistent with the changes previously discussed. Therefore, the proposed changes will have no adverse effect on plant safety.
U.S. Nuclear Regulatory Commission B18556/Attachment 1/Page 11 Technical Specification Changes - Bases The information contained in the Bases of the affected TSs will not be modified as a result of the proposed TS changes. The proposed changes will not result in any new approaches to plant operation. Therefore, the proposed Bases changes will not adversely affect public safety. The relocation of the requirements for the applicable TSs to the TRM will not result in any new approaches to plant operation and will not adversely affect any accident mitigation equipment. The plant response to the DBAs will not change. Therefore, the proposed changes will not adversely affect public health and safety. Thus, the proposed changes are safe.
Docket Nos. 50-336 50-423 B18556 Attachment 2 Millstone Nuclear Power Station, Unit Nos. 2 and 3 Technical Specifications Change Requests 2-21-01 and 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Significant Hazards Consideration
U.S. Nuclear Regulatory Commission B1 8556/Attachment 2/Page 1 Technical Specifications Change Requests 2-21-01 and 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Significant Hazards Consideration Description of License Amendment Request Dominion Nuclear Connecticut, Inc. (DNC) is proposing to relocate selected technical specifications for Millstone Unit Nos. 2 and 3. For Millstone Unit No. 2, DNC is proposing to relocate part of Technical Specification 3/4.4.9.1, "Pressure/Temperature Limits," and all of Technical Specifications 3/4.7.2.1, "Steam Generator Pressure/Temperature Limitation," 3/4.7.5.1, "Flood Level," 3/4.7.7.1, "Sealed Source Contamination," and 3/4.7.8, "Snubbers," to the Technical Requirements Manual (TRM) for Millstone Unit No. 2. For Millstone Unit No. 3, DNC is proposing to relocate part of Specification 3/4.4.9.1, "Pressure/Temperature Limits," and all of Technical Specifications 3/4.7.2, "Steam Generator Pressure/Temperature Limitation," 3/4.7.6, "Flood Protection," 3/4.7.10, "Snubbers," 3/4.7.11, "Sealed Source Contamination," and 3/4.7.14, "Area Temperature Monitoring," to the TRM for Millstone Unit No. 3. Basis for No Significant Hazards Consideration In accordance with 10 CFR 50.92, DNC has reviewed the proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed technical specification changes will relocate to the TRM the following items: surveillance requirements for the withdrawal of reactor vessel material irradiation specimens of Millstone Unit Nos. 2 and 3 which are part of the Pressure/Temperature Limits technical specifications, Millstone Unit Nos. 2 and 3 technical specifications covering Steam Generator Pressure/Temperature Limitation, Flood Level, Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 3 technical specification covering Area Temperature Monitoring will be relocated to the TRM. Since the relocated requirements remain the same, the proposed changes will have no effect on plant operation, or the availability or operation of any accident mitigation equipment. Therefore, the relocation of the requirements associated with these technical specifications will not impact an accident initiator and cannot cause an accident. These changes will not increase the probability or consequences of an accident previously evaluated.
U.S. Nuclear Regulatory Commission B 18556/Attachment 2/Page 2
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed technical specification changes will relocate the requirements of selected Millstone Unit Nos. 2 and 3 technical specifications as described above to the TRM. The proposed changes do not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. Since the requirements remain the same, the proposed changes do not alter the way any system, structure, or component functions and do not alter the manner in which the plant is operated. The proposed changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Involve a significant reduction in a margin of safety.
The proposed technical specification changes will relocate to the TRM the following items: surveillance requirements for the withdrawal of reactor vessel material irradiation specimens of Millstone Unit Nos. 2 and 3 which are part of the Pressure/Temperature Limits technical specifications, Millstone Unit Nos. 2 and 3 technical specifications covering Steam Generator Pressure/Temperature Limitation, Flood Level, Sealed Source Contamination, and Snubbers. Also the Millstone Unit No. 3 technical specification covering Area Temperature Monitoring will be relocated to the TRM. Since the proposed changes are solely to relocate the existing requirements, the proposed changes will have no effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the Design Basis Accidents will not change. Therefore, there will be no reduction in a margin of safety.
Docket Nos. 50-336 50-423 B18556 Attachment 3 Millstone Nuclear Power Station, Unit No. 2 Technical Specifications Change Request 2-21-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Marked Up Pages
Nuclear Regulatory Commission U.S. U.S. Nuclear Regulatory Commission B18556/Attachment 3/Page 1 List of Affected Pages Technical Specification Affected Page with Section Number Title of Section Amendment Number Index Page VIII Amend. No. 238 Index Page XIII Amend. No. 240 3/4.4.9.1 Pressure/Temperature Limits 3/4 4-18, Amend. No. 218 3/4 4-20, Amend. No. 94 3/4.7.2.1 Steam Generator 3/4 7-10, Original issue Pressure/Temperature Limitation 3/4.7.5.1 Flood Level 3/4 7-13, Amend. No. 101 3/4 7-14, Original issue 3/4 7-15, Original issue 3/4.7.7.1 Sealed Source Contamination 3/4 7-19, Amend. No. 202 3/4 7-20, Original issue 3/4.7.8 Snubbers 3/4 7-21, Amend. No. 160 3/4 7-22, Amend. No. 244 3/4 7-22a, Amend. No. 239 3/4 7-22b, Amend. No. 160 3/4 7-32, Amend. No. 160 3/4.4.9 Bases Pressure/Temperature Limits B 3/4 4-7, Amend. No. 218 3/4.7.2 Bases Steam Generator B 3/4 7-3a, Amend. No. 238 Pressure/Temperature Limitation 3/4.7.5 Bases Flood Level B 3/4 7-4a, Amend. No. 236 3/4.7.7 and 3/4.7.8 Bases Sealed Source Contamination B 3/4 7-5, Amend. No. 202, B 3/4 and Snubbers 7-6, Amend. No. 244
2 INDEX CAEust 12, 199g* LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE .......... ..................... 3/4 7-1 Safety Valves .......... ..................... 3/4 7-1 Auxiliary Feedwater Pumps ..... ............... 3/4 7-4 Condensate Storage Tank ..... ................ 3/4 7-6 Activity . . . . . . . . . . . . . . . . . . . . . . 3/4 7-7 Main Steam Line Isolation Valves .......... 3/4 7-9 Main Feedwater Isolation Components (MFICs)...... 3/4 7-9a Atmospheric Dump Valves ...... ................ 3/4 7-9c Steam Generator Blowdown Isolation Valves ......... 3/4 7-9d 3/4.7.2 STE[A CENlrATOnR PREnlUREFTnEPERA:TRE LIMITATION 3/4 7-10 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYST i 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM ...... .. 9 3/4 7-12 3/4.7.5 FLOO LEVEL .4.................... 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......... 3/4 7-16
-RGI GONhIDE ATAMT10&1ATN . .l 3/4.7.7 3/4 7-19 3/4.7.8 3/47-21 DELETED .S.O.R.C . .C.O.TAITO.............. ...
3/4.7.9 3/4 7-33 3/4.7.10 DELETED ........ ... ........................ 3/4 7-33 3/4.7.11 ULTIMATE HEAT SINK ................. 3/4 7-34 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES .... ....... 3/4 8-1 Operating ...... ............. 3/4 8-1 Shutdown . . . . . . . . . . . . . 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS 3/4 8-6 A.C. Distribution - Operating . . 3/4 8-6 A.C. Distribution - Shutdown . . . 3/4 8-7 D.C. Distribution - Operating . . . 3/4 8-8 D.C. Distribution - Shutdown . . . 3/4 8-10 D.C. Distribution (Turbine Battery) Operating 3/4 8-11 MILLSTONE - UNIT 2 VIII Amendment No. 77, 7A, y, log,
afY-1O-20OO INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...... ... ................ /4 7-1 3/4.7.2 STEAM GENEDATOR PRESSUREI LMTATIN' TMPERATU 3/4 7-3 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM ..... ... B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM ......... ................. ... B 3/4 7-4 3/4.7.5 FLOOD +VEEt .. ........ . ....... .. B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM B 3/4 7-4 3/4.7.7 SEALED SOURCE CONTAMIIATIO, . D E 'C..B 3/4 7-5 3/4.7.8 3/4.7.8 SeUBBER&............... SNU..... BER _... .... .... .......
. . . . . . . . . . . . .. . B 3/4 3 4 7 3/4.7.9 DELETED ............. S. . B 3/4 7-6 3/4.7.10 DELETED ............. B 3/4 7-7 3/4.7.11 ULTIMATE HEAT SINK . . . B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS ..... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...... B 3/4 9-1 3/4.9.2 INSTRUMENTATION ........... B 3/4 9-1 3/4.9.3 DECAY TIME ......... B 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS . . B 3/4 9-1 3/4.9.5 DELETED .... ............ B 3/4 9-1 3/4.9.6 DELETED .... ............ B 3/4 9-2 3/4.9.7 DELETED ..................... B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLING RECIRCULATION . . B 3/4 9-2 MILLSTONE - UNIT 2 XIII Amendment No. X, 77, M , 7,9, Im, w. 7,4
0 n-7) rfo cA 4 ,e-July 1, 1998 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 Reactor Coolant System (except the pressurizer) temperature, pressure, and heatup and cooldown rates shall be limited in accordance with the limits specified in Table 3.4-2 and shown on Figures 3.4-2a and 3.4-2b. APPLICABILITY: At all times.* I ACTION:
- a. With any of the above limits exceeded in MODES 1, 2, 3, or 4, perform the following:
- 1. Restore the temperature and/or pressure to within limit within 30 minutes.
AND
- 2. Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System remains acceptable for continued operation within 72 hours. Otherwise, be in at least MODE 3 within the next 6 hours and in MODE 5 with RCS pressure less than 300 psia within the following 30 hours.
- b. With any of the above limits exceeded in other than MODES 1, 2, 3, or 4, perform the following:
- 1. Immediately initiate action to restore the temperature and/or pressure to within limit.
AND
- 2. Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System is acceptable for continued operation prior to entering MODE 4.
I
- See Special Test Exception 3.10.3.
MILLSTONE - UNIT 2 3/4 4-17 Amendment No. #A, Y, 777, J?, 218
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1 shall be
- a. The Reactor Coolant System temperature and pressure per 30 minutes determined to be within the limits at least once leak and during system heatup, cooldown, and inservice hydrostatic testing operations.
- b. The reactor vessel terial iraat surveillance ecimens shal be removed and xamined, to dete ine changes in aterial 4.4-3. The prope ties, at the in rvals shown in Ta le to update Tab result of these exami tions shall be us d 3.4-2 aadFigures 3.4-a and 3.4-2b.
3/4 4-18 Amendment Ho.i# MILLSTONE - UNIT 2
To (i~cy~~ MA fin-Ak (A.e... July 1, 1998 TABLE 3.4-2 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITS Cooldown Heatup Indicated Cold Leg Limit Indicated Cold Limit Temperature Leg Temperature
< 100F < 5*F/hour if RCS not < 220°F < 30°F/hour vented.
100°F < T < 230°F < 30°F/hour 220°F < T < 275"F if RCS not < 50°F/hour vented.
< 190°F < 50°F/hour if RCS vent > > 275°F < lO0°F/hour 2.2 square inches.
< 230"F < 50°F/hour during unanticipated temperature excursions.
> 230°F < 80°F/hour Inservice Hydrostatic and Leak Testing Indicated Cold < 5OF/hour for Leg Temperature 1 hour prior to and during inservice hydrostatic and leak testing operations above the heatup limit curve.
MILLSTONE - UNIT 2 3/4 4-19 Amendment No. 218
July 1, 1998 2500 ydrostc: . . . .
"LeakTest Limit..... ...... . . ..........
2000-1 en a t Li t ..... ............ 21000
...n.H S.....- ...
a ccepL mtae ,ý............ .. .... ... ........
..... .. . .".... . . ....... . . 3p doe Indicated Maximum S.. mperature opcratg T>L5 00SFT4
.. °
-.[ ...
J! ITemp =70 F MinBot-u . .'. .. . . 200-F < T S 500°F 3 .
.. .. ... J......... J.... ."' ..
' "i ' '
..... *l °° 2 S 0 50 100 150 200 250 300 350 400 450 500 550 Indicated Cold Ltg Temperature (°I9 Millstone Unit 2 Reactor Cool ant System Heatup Limitations for Up to 20 EFPY Figure 3.4-2a MILLSTONE - UNIT 2 3/4 4-19a AMmnAMnnIf U~ !)1 0 T# ..... r~I*' II 5," I .I b I . L.O
F() Ck4 July 1, 1998 2500
- ..... ---- i..... ..... ................ .....
2000
*t Uacpal .......... -0** i.......... *. . i i
. ..... ............. ---- .... . g
......... Unacceptable ......
--- O e a t on- ....... ..... ----- ....-- ..... ..... -..... -----...i ...... .. ....
0150 ~~~~~.. ..... . ..... ..... ..... ........... .....!...... .-- L... . ..!.. ..... ! .. . . ..
.i ...
I:* . ... ... ..........
...... ....... i.........
S..
. .... , .... .... *. . ........ i .. ......
E .. i .......... ;....... .. *.. . . .. S......-*. .... ..... . . ....* .. ....... ..... .. ....- --- TOrperationcrti
.ii --- ...
"105100 ----
S...... .... ; ........ ..... 350 S................... Indicated e 400 Mru 45 5 oo. LCmta ios L--.fp 0 No. of RCPld.. 50 F/hr ,i Temperature o3r4t-2 500 U 2 5An t . S- 200OF < T <5000F .. 3:
.... :- ......... Mvin- BOlt-UP : 150°F < T:5 200*F 2 Tp 70°F .. :... T< 150OF0 0 [ ... "...-.....
0 50 100 150 200 250 300 350 400 450 500 550 Indicate Cold L-eg Temperature (7F) Millstone Unit 2 Reactor Coolant System Cooldown Limitations for Up to 20 EFPY Figure 3.4-2b MILLSTONE - UNIT 2 3/4 4-19b Amendment No. 218
April 10, !9C RIIStckTrl)' PC Ay Lr6LA MILLSTONE - UNIT 2 314 4-20 Amendment No.,4K
PLANT SYS TEMS 3/4.7.2 STEAM GE~k,4TOR PRESSURE/TEMPERA11JR 'IMITATION LIMITING CONDITION FOR 0ORATION 3.7. .1 The temperatures of bo h the primary and second y coolants in the s am generators shall be > °F when the pressure of ither coolant in the 'eam generator is > 200 ps APPLICABIL :ALL MODES. With the requirem ts of the above specifica on not satisfied:
- a. Immediately educe the steam generator ressure to < 200 psig, s,
- b. Perform an analy is to determine the effect f the overpres surization on the tructural integrity of the team generator.
Determine that the ear generator remains acce able for continued operation p *or to increasing its tempe atures above 2000F. K SURVEILLANCE R UIREMENTS x 4.7.2.1 The tempera ures of both the primary a secondary coolants in the steam generators aall be determined to be > °F at least once per hour when pressures in hhe steam generators are % 0 psig and Tavg is m M 200*F. avg N a - - ---- --- - - MILLS1TONE - UNIT 2 3/4 7-10 K "
PLANT SS '*3/4.7.5 FLOOD EVEL LLIMITIN C II O O NI CODITI IMITIG FOR OPERATI0\
- 3. .5.1 At least one PERABLE service wat pump motor shall b pro tec d against floodin to a minimum elevat n of 28 feet Mean Se Level USGS atumif either:
- a. The water level, including wave crest eight, is exceeding plant grade level 14.0 feet Mean Sea vel USGS datum), or
- b. ree or more of the following conditions are occurring s1 ultaneously:
- 1. The center of a st as determined by adar, recon aissance or forecas ed track projection, is presently 1 ated within the cr"tical area as define on Figure
- 3. 1.
- 2. The p ojected track of a torm approaching the facility as det rmined by radar, re onnaissance or forec ted track p ojection, lies betw n 130' and 350°%
- 3. The centr pressure of the s,- rm' is or is forecas d to be < 28.0 i. Hg; or the measur d 15 minute average md spied at nom'nal elevation 389 o the meteorological tower exceeds 60 mph.
- 4. The 15 minute a rage wind direction at nominal elevation 389 on the meteo logical tower is wi hin the sector from 1500 clockwise to 000.
APPLICABILITY: ALMODES ACTION: With the water level ceeding either pl nt grade or with thr e or more exceede simul of the above specified eteorological con itions being tect at least one se ice taneously, immediate ini iate action to pr n ooding to a mini m elevation of 28 f t; water pump motor against complete this protective a ion within 2 hou. MILLSTONE - UNIT 2 3/4 7-13 Amendment No.
c_1q_ IL PLANT SYStMS SURVEILLANCE ýCUREMENTS 44.7.5.1.1 The wat level shall be dete mined to be below pla t gtade Sleast once per h r when the eye of a urricane is within 15 miles o facility., ethe e 4.7. .1.2 The above sp ified meteorological onditions shall be d er mine at least once per 2 hours when a hurrica eye is within 150 mi s of th facility. The mete ological conditions hall be determined fr forecas s obtained from the onnecticut Valley. El trical Exchange (CONVEX and from the site me erological instrument tion. I4I T~t' PP~ j~~t~d,,,LL( LFT 1Lli MILLSTONE - UNIT 2 314 7-14 , . MILSTOE'UI,2-14Voel,,,
AuguL 1, 1975 5,&.. Critical Area L¶ , kFigure 3.7-1 MILLSTONE - UNIT 2 3/4 7-15
3 \X..7SEAEDSORCE\\NANINATIO LIMKING CONDITION FOR OPERP ON ION 3.7-7.1 ch sealed source cont ining radioactive mater 1 either in excess of 100 micro uries of beta and/or a emitting material 5 microcuries of alpha emittin material shall be fre of > 0.005 microcuries f removable APPLICABILITY: A LL TIMES.
- a. Each sealed so rce with removable con ination in excess of e above limit sha be immediately withdr n from use and:
- 1. Either deconta mated and repaired, o
- 2. Disposed of in a ordance with Coimissic Regulations.
- b. The provisions of Specifi ation 3.0.3 are not ap icable.
SURVEI LLANCE QUIREMENTS 4.7.7.1.1 Test R~ea*rements - Each sealed so ce shall be tested for eakage and/or contamination by: aa. The licensee, or ab. Other persons sp cifically authorized by te Comission or an _AAgreement State. The test ethod shall have a d ection sensitivity of at east 0.005 microcurie per test sample. 4.7.7.1.2 r - Each tegory of sealed sources hall be tested at the requen es described below. sources f r a lsealed\ per\ ixmonths
- a. S e n l] -
A use least once sta At (exludin tuD sources l previously Vo to su *ected core__ m a t e r i a l i:. c o n t a i n i ng Nr~d i oa t v oa.e-- -- rOX5 Pt'o6~ý f1Trj-0A- LE~F- BLr9ANk: MILLSTONE - UNIT 2 3/4 7-19 I3Amendment No. ;A;, jtj 0286
-Auguzt 1.,1I,975- p S XVEILLANCE REX ET IR Continue\\
LPLANT SYSTEMS I. With a h If-life greater an 30 days (exc ding Hydrogen 33), and
- 2. In any form ther than gas.
- b. tored sources not -n use - Each seal d source shall tested pior to use or tran Ifer to another Ii ensee unless test d wihin the previous s months. Sealed ources transfer wit out a certificate dicating the las test date shall test ddprior to being p ced into use.
- c. Startu sources - Each sea ed startup source hall be tested prior tkbeing subjected to core flux and fol wing repair or maintena ce to the source.
CcC Ir 1r4 -F(-~ T& NTY AL Y LfE-F-r 6 Lrt-r1 "A, IMILLSTONE
- UNIT 2 3/4 7-20
P'AN___I SYSTEMS S314.7.8 SNUBBERS \ 3.7.x All snubbers shall be *ERABLE. The only snubberss eecluded from the requir ments are those installe on nonsafety-related system and then only if their f Hlure or failure of the s stem on which they are insta led would have no advers effect on any safety-re ted system. APPLICABILI MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers cated on systems requi d OPERABLE in those MOD With one or more sn bers inoperable within hours replace or restore th inoperable snubber(s to OPERABLE status and p rform an engineering evaluati n per Specification 4.7. .d on the attached compo nt or declare the attached system inoperable and f low the appropriate ACT statement for the system. S,,RVE7I,, AN(:. REQUITREMENTS
- 4.7.8 Each snubber shall be emonstrated OPERABLE by erformance of the liowing augmented inservice i pection program and the equirements of Sp cification 4.0.5.
- a. Inspection Types used in this specification, "type of snubber" shall m an snubbers of t same design and manufacturer, irrespective of capacit
- b. Visua Inspections Snubbers are categorized as inaccessib e or accessible during r ctor operation. Each of these categories (i ccessible and accessible may be inspecte independently according to te schedule determined by Table 4.7-3., The visual inspection interv for each type of snubb shall be deter 'ned based upon the critervia rovided in Table 4.7-3 a the first inspec ion interval determined usin this criteria shall be based upon the pr ious inspection interval as stablished by the requirements in ef t before Amendment 160.
c Visual Inspection Acce ance Criteria Visual inspections shall rify that (1) the snubber h no visible dications of damage or im aired OPERABILITY, (2) attac ments to the fo dation or supporting str ture are functional, and (3 fasteners for he attachment of the snub r to the component and to e snubber ancho ge are functional. Snub rs which appear inoperable a result of visu 1 inspections shall be cl sified as unacceptable and ay be reclassi 'ed acceptable for the pu ose of establishing the nex visual inspection interval, provided that ( the cause of the rejectio is clearly esta lished and remedied for at particular snubber and r other snubber irrespective of type MILLSTONE - UNIT 2 3/4 7-21 Amendment No...... ,-- , d be8, 96, e~'~h/l~ __________________Correc~ted by letter ýdated 8126192
-rj ~ T"N 7-C"e te.- 14rtoNAZy £CFT -S1A1A'-v
iPLANT SYSTEMS SU*EILLANCE REQUI
*that may be genericall *susceptible; and (2)\the affected snubber is
*unctionally tested in tle as-found condition nd determined 0 RABLE per Specificatio 4.7.8.e or 4.7.8.f, applicable. All snu ers found connected to an inoperable common draulic fluid rese oir shall be counted a unacceptable for det ining the next inspecion interval. A review d evaluation shall b performed and documentd to justify continued peration with an una eptable snubber. continued operation c not be justified, th snubber m°et.\\
shall be de ared inoperable and th ACTION requirements all be
- d. Snubber Tests At least once per ei hteen (18) months dunl shutdown, a represe tative sample (10% of he total of each type f snubber, mechanica and hydraulic, in use i the plant) shall be t ted either in place or in a bench test. For ach snubber that does ot meet the test cceptance criteria of Spe *fication 4.7.8.e or 4. .8.f, as a licable, an additional 5 f that type of snubbe shall be ers are Testin shall continue until no a ditional inoperable sn bbers are found wi in a sample or until all nubbers have been tes d. The representa ive sample selected for t ting shall include th various configuratio s, and the range of size d capacity of snubbe Snubbers identi ied as "Especially Diffic t to Remove" or in h Radiation Zones ring Shutdown" shall also be included in the representative sam e.*
In addition to the re lar sample, in locations here snubbers had failed the previous tes due to operational or en *ronmental condi tions (excessive vibratio water hammer, high radi ion, extreme eat or humidity, etc.), th snubbers currently inst led in these I cations shall be tested du ng the next test period. est results of hese snubbers may not be i luded for the resampling. All repl ement snubbers shall have en tested prior to inst lation.
- Permanent or other exe tions from functional tes ng for individual snubbe s in these categories may granted by the Commission nly if a justifiable basis for exemption is pre nted.
qILTN P-U1NT34 LN-F-r J3L*'Z, 7-2T2ANmedmn MILLSTONE - UNIT 2 3/4 7-22 Amendment No.JJ , . 0482 7F 7f 7,Jf
PLAN[ YSTEMS SURVEILLA CE REQUJIREMENTS (\Ctinued) If ny snubber selecte for testing either fails to lock-up or fails to m e (i.e., frozen i place), the cause will b evaluated and if cause by manufacturer de ign deficiency, all snubb rs of the same design bject to the same efect shall be tested re rdless of location r difficulty or re oval. This testing requirrment shall be indepen nt of the require nts stated above for snu ers not meeting the est acceptance cri er-ia. For the snubbe s) found inoperabl an engineering evaluati n shall be performed on e component whic are supported by the snu ber(s). The purp e of this engineering evaluation shall be to determine if the co onents supported the snubber (s) were adversely affected b the inoperability f the snubber (s) in orde to ensure that the sup rted component re ins capable of meeting the designed service.
- e. H raulic Snubbers Functio I Test Acceptance riteria The hdraulic snubber functio 1 test shall veri that:
- 1. A ivation (restraining ac on) is achieved wi hin the spe ified on.
comprssi range of velocity r acceleration in th tension and
- 2. Snubber leed, or release rate, specifiedrange in compression or ere required, esion. is 'thin the
- f. Mechanical Snubbe s Functional Test Accept nce Criteria*
SThe mechanical snubbr functional test shall erify that: 1I. The_ forc hat lini ijates free movement of t*e snubber rod in Seither tension or c ~pression is less than t] specified Smaximum drag force.* ..
- 2. Activation (restraining ction) is achieved withi the
_ .co ecified range of veloci or acceleration in bot tension and ression.. . . . \
- g. Snubber Se vice Life Monitoring A record of te service life of each s ubber, the date at whic the designated ser ice life commences and t installation and maintenance reco ds on which the designat d service life is based shall be maintain as required by Quality ssurance Program Topica eport.
- Mechanical snub\befunctional test a ceptance criteria shall ecome effective upon installation o snubber testing e ipment but not later th June 30, 1985.
MILLSTONE - UNIT 2 3/4 Amndmet N.
-22 ~ IF, IgoI, O
i PLANT SYSTEMS SURVEýILANCIE REQUIREMEN\TContinued) n Concurrent ent with e first inservice isual inspection d at\aea least n0nceec per e 18 montn thereafter, a the installation and intenance t ' st intenance r cords c 0 s fo for eachth ubber th a bberr shall be r viewed to verify t hat the inn icated rpae service li has not been exce ed or will not be e ceeded prii to0t ther next sc duled snubber se 'ce life review. i f IF the the f uled indi c tedd s 1Ife service life u ill 11 be exceeded pr r to the next lext _sche Ul ed i ew servi xte life review, the snubber serviceF li e shall. be reevaluu ed
ýe reeval or the cruj nubbber theI ri ew ' be shall r snubb thests re Ins e or reconditi ed so as to extend laced xt 'ts ts serviceI tr ife e beyond 1-sfc the d te of the next sc eduled service y j Ii li s
review. his reevaluation, 1 ed eplacement or rec ditioning shall be
\indicated the records.
Re PtA, Wl't-j, L t /ý'-(A TT& "-ý9LL y ( 04T-5 MILLSTONE UNIT 2 3/4 7-22b Amendment Nos. go, 1J0,X4 0017
TABLE 4.7-3 SNUBBER VISUAL INSPEC ON INTERVAL S* * * ~ NUMBER OF UNACCEP *BLE SNUBBERS / ozrp or- Populatio Column A Column B Column C or Cate ;*ry Extend Interval/ Repeat Interval Reduce Interva M*3 Notes- Iind 1 2) (Notes 3 and *"(Notes-4-andS (Notes 5 and4)/ 200 150 30 8 0 2 5 13 3 00 300 5 12 25 1: The next visual in ection interval for a s bber population
- / te I h .n , "1='sz{)e o or ca gory size shall
;~oysz hl be de deter t e rm eded S2ote based upon the rvious inspection interv and the number of u cceptable snubbers found that interval, ubbers may be categoriz , based upon their ac ssibility during ring as accessible r inaccessible. These power op aton, tegories may be exami that decisio must be made and docume ed before any inspect' ned separately or jointly However, and that decision the basis on which to determine t next inspection inte a for that category. shal be used as
- Note 2: Interpo tion between populatio or category sizes a the number of unaccep** le snubbers is permis ible. Use next lower i eger for the value o the limit oe 3 inte r includes a fractional alue of unacceptable nubbers as determined for Columns , B, or C if that by interpolation.
SNote 3: the number of unaccept le snubbers is equal o or less than the num inspection interval may e twice the previous r in Column A, the next terval but not greater an 48 months. 4 If the number of una eptable snubbers is e al to or. less than I A / the number in Colum A, the next inspecti th umber in Column B but greate than interval shall be the ame as the previous interval. N e 5: If the number ins ection mt ovalunacceptable snubber shall be two-thi s ofisthe equal to or great previous inte al. than the number However, if the in Column the next number o nacceptable snubber
%bal I s ss than the be number i Column ut greaterl athnubriCoun8 netneval shall b re uced proportionally interpolation, that the revious interval factor th is one-third of the atmo of the difference etween shal e reduced by a during e previous interval e number of unacce t le snubbers found the number in Colum to the difference in the num rs in Columns B and C.
Note 6: The rovisions of Specif' ation 4.0.2 are appli able for all inspection interv s up to and including 4 months.
July 1, 1998 REACTOR COOLANT SYSTEM A/h/ BASES Reducing T'vg to < 515°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with iodine spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.0 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. In addition, during heatup and cooldown evolutions, the RCS ferritic materials transition between ductile and brittle (non-ductile) behavior. To provide adequate protection, the pressure/temperature limits were developed in accordance with the IOCFR50 Appendix G requirements to ensure the margins of safety against non-ductile failure are maintained during all normal and anticipated operational occurrences. These pressure/temperature limits are provided in Figures 3.4-2a and 3.4-2b and the heatup and cooldown rates are contained in Table 3.4-2. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermally induced compressive stresses at I the inside wall tend to alleviate the tensile stresses induced by the internal I pressure. Therefore, a pressure- temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location. The heatup analysis also covers the determination of pressure temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients estab lished during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermally induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis. MILLSTONE - UNIT 2 B 3/4 4-5 Amendment No. 218
1..1 ..
- 1 f*rlrl Julye1, IV1y REACTOR COOLANT SYSTEM BASES The heatup and cooldown limit curves (Figures 3.4-2a and 3.4-2b) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to the maximums described in Technical Specification 3.4.9.1, Table 3.4-2. The heatup and cooldown curves were prepared based uponI the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figures 3.4-2a and 3.4-2b.
Verification that RCS pressure and temperature conditions are within the limits of Figures 3.4-2a and 3.4-2b and Table 3.4-2, at least once per 30 minutes, is required when undergoing planned changes of > 10"F or > 100 psi. This frequency is considered reasonable since the location of interest during cooldown is over two inches (i.e. 1/4 t location) from the interface with the reactor coolant. During heatup the location of interest is over six inches from the interface with the reactor coolant. Thi.s combined with the relatively large heat retention capability of the reactor vessel ensures that small temperature fluctuations such as those expected during normal heatup and cooldown evolutions do not challenge the structural integrity of the reactor vessel when monitored on a 30 minute frequency. The 30 minute time interval permits assessment and correction for minor deviations within a reasonable time. During RCS heatup and cooldown the magnitude of the stresses across the reactor vessel wall are controlled by restricting the rate of temperature change and the system pressure. The RCS pressure/temperature limits are provided in Figures 3.4-2a and 3.4-2b, and the heatup and cooldown rates are contained in Table 3.4-2. The following guidelines should be used to ensure compliance with the Technical Specification limits.
- 1. When changing RCS temperature, with any reactor coolant pumps in operation, the rate of temperature change is calculated by using RCS loop cold leg temperature indications.
This also applies during parallel reactor coolant pump and shutdown cooling (SDC) pump operation because the RCS loop cold leg temperature is the best indication of the temperature of the fluid in contact with the reactor vessel wall. Even though SDC return temperature may be below RCS cold leg temperature, the mixing of a large quantity of RCS cold leg water and a small quantity of SDC return water will result in the temperature of the water reaching the reactor vessel wallbeing very close to RCS cold leg temperature.
- 2. When changing RCS temperature via natural circulation, the rate of temperature change is calculated by using RCS loop cold leg temperature indications.
- 3. When changing RCS temperature with only SDC in service, the rate of temperature change is calculated by using SDC return temperature indication.
MILLSTONE - UNIT 2 B 3/4 4-6 LAmendment No. Y;, 1P*, J77, 218
Fe t bVW aI July 1, 1998 REACTOR COOLANT SYSTEM BASES
- 4. During the transition from natural circulation flow, to forced flow with SDC pumps, the rate of temperature change is calculated by using RCS loop cold leg temperature indications. SDC return temperature should be used to calculate the rate of temperature change after SDC is in service, RBCCW flow has been established to the SDC heat exchanger(s), and SDC return temperature has decreased below RCS cold leg temperature.
"15. During the transition from parallel reactor coolant pump and SDC pump operation, the rate of temperature change is calculated by using RCS loop cold leg temperature indications until all reactor coolant pumps are secured. SDC return temperature should be used to calculate the rate of temperature change after all reactor coolant pumps have been secured.
- 6. The temperature change limits are for a continuous one hour period.
Verification of operation within the limit must compare the current RCS water temperature to the value that existed one hour before the current time. If the maximum temperature increase or decrease, during this one hour period, exceeds the Technical Specification limit, appropriate action should be taken.
- 7. When a new, more restrictive temperature change limit is approached, it will be necessary to adjust the current temperature change rate such that as soon as the new rate applies, the total temperature change for the previous one hour does not exceed the new more restrictive rate.
The same principle applies when moving from one temperature change limit curve to another. If the new curve is above the current curve (higher RCS pressure for a given RCS temperature), the new curve will reduce the temperature change limit. It will be necessary to first ensure the new more restrictive temperature change limit will not be exceeded by looking at the total RCS temperature change for the previous one hour time period. If the magnitude of the previous one hour temperature change will exceed the new limit, RCS temperature should be stabilized to allow the thermal stresses to dissipate. This may require up to a one hour soak before RCS pressure may be raised within the limits of the new curve. If the new curve is below the current curve (lower RCS pressure for a given RCS temperature), the new curve will allow a higher temperature change limit. All that is necessary is to lower RCS pressure, and then apply the new higher temperature change limit.
- 8. When performing evolutions that may result in rapid and significant temperature swings (e.g. placing SDC in service or shifting SDC heat exchangers), the total temperature change limit for the previous one hour period must not be exceeded. If a significant temperature change is anticipated, and an RCS heatup or cooldown is in progress, the plant should be stabilized for up to one hour, before performing this type of evolution. Stabilizing the plant for up to one hour will allow the thermal stresses, from any previous RCS temperature change, to dissipate.
This will allow rapid RCS temperature changes up to the applicable Technical Specification temperature change limit. MILLSTONE - UNIT 2 B 3/4 4-6a Amendment No. 218
July 1, 1998 REACTOR COOLANT SYSTEM - BASES
- 9. Additional margin, to prevent exceeding the Appendix G limits when RCS temperature is at or below 230°F, can be obtained by maintaining RCS pressure below the pressure allowed by the 50"F/hr cooldown curve provided on Figure 3.4-2b. This will ensure that if a greater than anticipated temperature excursion occurs during short duration evolutions, the margins of safety required by Appendix G will not be exceeded. Examples of plant evolutions that may result in unanticipated temperature excursions include placing SDC in service without parallel RCP operation, securing RCPs when SDC is already in service, shifting SDC heat exchangers, and switching SDC pumps. Establishing a lower RCS pressure, will minimize the probability of exceeding Appendix G limits.
If the 50°F/hr cooldown curve is used to evaluate unanticipated temperature excursions while limited to a cooldown rate of 30°F/hr, the RCS cooldown rate must be restored to within the 30°F/hr limit as soon as practical. This may require a soak period to allow the thermal stresses, from the previous RCS temperature change, to dissipate. The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table 4.6-1 of the Final Safety Analysis Report. Reactor operation and resultant fast neutron irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, can be predicted using the methods described in Revision 2 to Regulatory Guide 1.99. The heatup and cooldown limit curves shown on Figures 3.4-2a and 3.4-2b include predicted adjustments for this shift in RTNDT at the end of the applicable service period, as well as adjustments for possible uncertainties in the pressure and temperature sensing instruments. The adjustments include the pressure and temperature instrument and loop uncertainties associated with the main control board displays, the pressure drop across the core (RCP operation), and the elevation differences between the location of the pressure transmitters and the vessel beltline region. In addition to these curve adjustments, the LTOP evaluation includes adjustments due to valve stroke times, PORV circuitry reaction times, and valve discharge backpressure." The actual shift in RINDT of the vessel material is established periodically during operation by removing and evaluating, in accordance with IOCFR50 Appendix H, reactor vessel material irradiation surveillance specimensi installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are similar, the measured transition shift for a sample can be correlated to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figures 3.4-2a and 3.4-2b for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing. MILLSTONE - UNIT 2 B 3/4 4-6b Amendment No. 218
REACTOR COOLANT SYSTEM BASES The maximum RTNDT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 50°F. The Lowest Service Temperature limit is based upon this RTNDT since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + ]00°F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement. Included in this evaluation is consideration of flange protection in accordance with 10 CFR 50, Appendix G. The requirement makes the minimum temperature RTNDT plus 90*F for hydrostatic test and RTNDT plus 120°F for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure. Since the flange region RTNDT has been calculated to be 30°F, the minimum flange pressurization temperature during normal operation is 150°F (161°F with instrument uncertainty) when the pressure exceeds 20% of the preservice hydrostatic pressure. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement. To establish the minimum boltup temperature, ASME Code Section XI, Appendix G, requires the temperature of the flange and adjacent shell and head regions shall be above the limiting RTNDT temperature for the most limiting material of these regions. The RTNDT temperature for that material is 30°F. Adding 10.5°F, for temperature measurement uncertainty, results in a minimum boltup temperature of 40.5°F. For additional conservatism, a minimum boltup temperature of 70*F is specified on the heatup and cooldown curves. The head and vessel flange region temperature must be greater than 70°F, whenever any reactor vessel stud is tensioned. The number f reactor vess 1 irradiation sur eillance specim ns and the freq encies for *emoving and tes ing these specime s are provided in Table 4.4-3 to assure cmpliance with t e requirements o ppendix H to 10 CFR Part 50. R1eoval of reactor vessel irradiatio surveillance specime s does not'onstitute a COR ALTERATION per S cification 1.1 The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. Verification that pressurizer temperature conditions are within the limits of LCO 3.4.9.2, at least once per 30 minutes, is required when undergoing planned changes of > IOF. The 30 minute time interval permits assessment and correction for temperature deviations within a reasonable time. MILLSTONE - UNIT 2 B 3/4 4-7 Amendment No. 9, 79, 9, #9'
por 0-ýt2 July 1, 1998 REACTOR COOLANT SYSTEM ,lu c/* BASES The Low Temperature Overpressure Protection (LTOP) System provides a physical barrier against exceeding the IOCFR50 Appendix G pressure/temperature limits during low temperature RCS operation either with a steam bubble in the pressurizer or during water solid conditions. This system consists of either two PORVs (each PORV is equivalent to a vent of approximately 1.4 square inches) with a pressure setpoint < 415 psia, or an RCS vent of sufficient size. Analysis has confirmed that the design basis mass addition transient discussed below will be mitigated by operation of the PORVs or by establishing an RCS vent "of sufficient size. The LTOP System is required to be OPERABLE when RCS cold leg temperature is at or below 275°F (Technical Specification 3.4.9.3). However, if the RCS is in MODE 6 and the reactor vessel head has been removed, a vent of sufficient size has been established such that RCS pressurization is not possible. Therefore, an LTOP System is not required (Technical Specification 3.4.9.3 is not applicable). The LTOP System is armed at a temperature which exceeds the limiting 1/4t RTNDT plus 90°F as required by NUREG-0800 (i.e., SRP), Branch Technical Position RSB 5-2. For the operating period up to 20 EFPY, the limiting 1/4t RTNDT is 145"F which results in a minimum LTOP System enable temperature of at least 263°F when corrected for instrument uncertainty. The current value of 275°F will be retained. The mass input analysis performed to ensure the LTOP System is capable of protecting the reactor vessel assumes that all pumps capable of injecting into the RCS start, and then one PORV fails to actuate (single active failure). Since the PORVs have limited relief capability, certain administrative restrictions have been implemented to ensure that the mass input transient will not exceed the relief capacity of a PORV. The analysis has determined two PORVs (assuming one PORV fails) are sufficient if the mass addition transient is limited to the inadvertent start of one high pressure safety injection (HPSI) pump and two charging pumps when RCS temperature is at or below 275°F and above 190°F, and the inadvertent start of one charging pump when RCS temperature is at or below 190°F. The assumed active failure of one PORV results in an equivalent RCS vent size of approximately 1.4 square inches when the one remaining PORV opens. Therefore, a passive vent of at least 1.4 square inches can be substituted for the PORVs. However, a vent size of at least 2.2 square inches will be required when venting the RCS. If the RCS is depressurized and vented through at least a 2.2 square inch vent, the peak RCS pressure, resulting from the maximum mass input transient allowed by Technical Specification 3.4.9.3, will not exceed 300 psig (SDC System suction side design pressure). When the RCS is at or below 190°F, additional pumping capacity can be made capable of injecting into the RCS by establishing an RCS vent of at least 2.2 square inches. Removing a pressurizer PORV or the pressurizer manway will result in a passive vent of at least 2.2 square inches. Additional methods to establish the required RCS vent are acceptable, provided the proposed vent has been evaluated to ensure the flow characteristics are equivalent to one of these. Establishing a pressurizer steam bubble of sufficient size will be sufficient to protect the reactor vessel from the energy addition transient associated with the start of an RCP, provided the restrictions contained in Technical Specification 3.4.1.3 are met. These restrictions limit the heat MILLSTONE - UNIT 2 B 3/4 4-7a Amendment No. 218
KtALIUK WUULAfKl SY51LM March 30, 2000 BASES input from the secondary system. They also ensure sufficient steam volume exists in the pressurizer to accommodate the insurge. No credit for PORV actuation was assumed in the LTOP analysis of the energy addition transient. The restrictions apply only to the start of the first RCP. Once at least one RCP is running, equilibrium is achieved between the primary and secondary temperatures, eliminating any significant energy addition associated with the start of the second RCP. The LTOP restrictions are based on RCS cold leg temperature. This temperature will be determined by using RCS cold leg temperature indication when RCPs are running, or natural circulation if it is occurring. Otherwise, SDC return temperature indication will be used. Restrictions on RCS makeup pumping capacity are included in Technical Specification 3.4.9.3. These restrictions are based on balancing the requirements for LTOP and shutdown risk. For shutdown risk reduction, it is desirable to have maximum makeup capacity and to maintain the RCS full (not vented). However, for LTOP it is desirable to minimize makeup capacity and vent the RCS. To satisfy these competing requirements, makeup pumps can be made not capable of injecting, but available at short notice. A charging pump can be considered to be not capable of injecting into the RCS by use of any of the following methods and the appropriate administrative controls.
- 1. Placing the motor circuit breaker in the open position.
- 2. Removing the charging pump motor overload heaters from the charging pump circuit.
- 3. Removing the charging pump motor controller from the motor control center.
A HPSI pump can be considered to be not capable of injecting into the RCS by use of any of the following methods and the appropriate administrative controls.
- 1. Racking down the motor circuit breaker from the power supply circuit.
- 2. Shutting and tagging the discharge valve with the key lock on the control panel (2-SI-654 or 2-SI-656).
- 3. Placing the pump control switch in the pull-to-lock position and removing the breaker control power fuses.
- 4. Placing the pump control switch in the pull-to-lock position and shutting the discharge valve with the key lock on the control panel (2-SI-654 or 2-SI-656).
These methods to prevent charging pumps and HPSI pumps from injecting into the RCS, when combined with the appropriate administrative controls, meet the requirement for two independent means to prevent pump injection as a result of a single failure or inadvertent single action. MILLSTONE 0648
- UNIT 2 B 3/4 4-7b Amendment No. 779, X77, 243
REACTOR COOLANT SYSTEM March 30, 2000 BASES These methods prevent inadvertent pump injections while allowing manual actions to rapidly restore the makeup capability if conditions require the use of additional charging or HPSI pumps for makeup in the event of a loss of RCS inventory or reduction in shutdown margin. If a loss of RCS inventory or reduction in shutdown margin event occurs, the appropriate response will be to correct the situation by starting RCS makeup pumps. If the loss of inventory or shutdown margin is significant, this may necessitate the use of additional RCS makeup pumps that are being maintained not capable of injecting into the RCS in accordance with Technical Specification 3.4.9.3. The use of these additional pumps to restore RCS inventory or shutdown margin will require entry into the associated action statement. The action statement requires immediate action to comply with the specification. The restoration of RCS inventory or shutdown margin can be considered to be part of the immediate action to restore the additional RCS makeup pumps to a not capable of injecting status. While recovering RCS inventory or shutdown margin, RCS pressure will be maintained below the Appendix G limits. After RCS inventory or shutdown margin has been restored, the additional pumps should be immediately made not capable of injecting and the action statement exited. An exception to Technical Specification 3.0.4 is specified for Technical Specification 3.4.9.3 to allow a plant cooldown to MODE 5 if one or both PORVs are inoperable. MODE 5 conditions may be necessary to repair the PORV(s). 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a. MILLSTONE 0648
- UNIT 2 B 3/4 4-7c Amendment No. 71?, 7X9, 243,
PLANT SYSTEMS January 10, 2002 BASES PTSCR 2-18-01 a feedwater isolation signal since the steam line break accident analysis credits them in prevention of feed line volume flashing in some cases. Feedwater pumps are assumed to trip immediately with an MSI signal. 3/4.7.1.7 ATMOSPHERIC DUMP VALVES The atmospheric dump valve (ADV) lines provide a method to maintain the unit in HOT STANDBY, and to replace or supplement the condenser steam dump valves to cool the unit to Shutdown Cooling (SDC) entry conditions. Each ADV line contains an air operated ADV, and an upstream manual isolation valve. The manual isolation valves are normally open, and the ADVs closed. The ADVs, which are normally operated from the main control room, can be operated locally using a manual handwheel. An ADV line is OPERABLE if local manual operation of the associated valves can be used to perform a controlled release of steam to the atmosphere. This is consistent with the LOCA analysis which credits local manual operation of the ADV lines for accident mitigation. 3/4.7.1.8 STEAM GENERATOR BLOWDOWN ISOLATION VALVES The steam generator blowdown isolation valves will isolate steam generator blowdown on low steam generator water level. An auxiliary feedwater actuation signal will also be generated at this steam generator water level. Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main feedwater. The steam generator blowdown isolation valves will also close automatically upon receipt of a containment isolation signal or a high radiation signal (steam generator blowdown or condenser air ejector discharge). 3/4.7.2 REA CTOR.BUILDN G CLOS COOLING WATE ED SYSTEM The limitation on eam generator pr sure and temperatux ensures that them pres ure andenginer the steam Fgnerators do not edreedthe maximum allow le fracture touhn ss stress limits. The limitations ofs70y F and 200 psig a based on a s ream dnnerator RTNDT ofthe t andt are ufficient t o freven brittle s fracture. op 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM The OPERABILITY of the Reactor Building Closed Cooling Water (RBCCW) System ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. The RBCCW loops are redundant of each other to the degree that each has separate the other.controls In the and power of supplies event a design and theaccident, basis operationone of RBCCW one does loopnot is depend requiredon to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. To ensure this requirement is met, two RBCCW loops must be OPERABLE, and independent to the extent necessary to ensure that a single failure will not result in the unavailability MILLSTONE - UNIT 2 B 3/4 7-3a Amendment No. XX *11p
PLANT SYSTEMS January 10, 2002 BASES PTSCR 2-18-01 3/4.7.4 SERVICE WATER SYSTEM (Continued) The Technical Specification Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the accident analysis are met and that subsystem OPERABILITY is maintained. The purpose of the service water pumps differential pressure test, Surveillance Requirement 4.7.4.1.a.2, a substantial flow test, is to ensure that the pumps have not degraded to a point where the accident analysis would be adversely impacted. The surveillance requirement acceptance criteria for the service water pumps was developed assuming a 7% degraded pump from the actual pump curves. Flow and pressure measurement instrument inaccuracies for the service water pumps have been accounted for in the design basis hydraulic analysis. It is not necessary to account for flow and pressure measurement instrument inaccuracies in the acceptance criteria contained in the surveillance procedure. 3/4.7.5 F.E. EI V.. _*xe~i dLmage
* [Lte The servic water pump mo ors are normally pltected against ater damage to n elevation f 22 feet. If the water level is xceeding plant rade level or f a severe st rm is approach ng the plant site, one service wate pump mnoto* will be prot cted against f ooding to a minimu elevation of 28 feet to ensur* that this p ~p will continu* to be capable of *emoving decay he t from the re ctor. In or r to ensure op rator accessibilit to the intake s ructure action *o provide pun~ motor protection will be initiatd when the water level reaches lant grade 1 fel.
3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation. cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem the or less whole body, or its equivalent. This limitation is consistent with requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. The [CO is modified by a footnote allowing the control room boundary to be opened intermittently under administrative controls. For entry and exit through doors the administrative control of the opening is performed by the coole bthsssean2)tecnrlrowilrmihaiaefr person(s) entering or exiting the area. consist of stationing a dedicated For other openings, these controls individual at the opening who is in constant communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated. The control room radiological dose calculations use the conservative minimum acceptable flowIof 2250 cfm based on the flowrate surveillance requirement of 2500 cfm + 10%. MILLSTONE - UNIT 2 B 3/4 7-4a i No. m ent Amement , ,
PLANT SYSTEMS 8 BASES 3/4.7.7 SEAL.LED SO 'R-E GO .TAMA ......... hbe lmtations on s aled source removabX contamination nsure that the total ody or individual or n irradiation does exceed allowabi lmts in the event ingstlion or inhoal tioon ofthe source aterial The 1 itaýtions on removab cotmination fo orces requiring INe k eting, inc in alpha emitters, *Is based on 10 CFR 70. 9(a)(3) limits for p tonium. Leakage f9sources 1p-excluded om the requireen ts of this specifica 'on represent 1s than one maximum ermissible body bur n for total body i adiation if te sore aterial is in aled or ingested. Sealed sour es are classified mnt three groups accordi to their use,\wh Sur illance Requi ements commensureate ith the probability o damage to asoc in t t group. Th e sources which are n t frequently handled re required to be teste more often th n those which are not. Sealed sources which are continuously Xenclose within a s ielded mechanism (i. ., sealed \ sources w thin radiation mornitor~ g or boron me uring devices) are co sidered to be stored a d need not be tested unl ss they are r oved from the shield chanism.
*IIUUUL*J 3/4.7.8 3/4.7.8 *JB*ERSDEr-D All sn bers are required OPERABLE to ensur that the struc ral integrity of the reactor coolaht system and all other safety reiated systems is maintained durng and following a seismic or other vent initiating ynamic loads. Snubbe s excluded from is inspection p gram are those installed on nonsafety-relate systems and then nly if their fai re or failure o the system n which they ar installed would ve no adverse e ect on any sa ty-related s'tem.
A list of inv ual snubbers with etailed informat n of snubber 1 cation and size and of system ffected shall be a ilable at the pla t in accordance with S ction 50.71(c) o 10 CFR Part 50. e accessibility each snubber s all be det mined and appr ed by the Plant Operations Revie Committee.. e determina ion shall be ba d upon the existin radiation levels and the expecte time to prform a visual spection in each ubber location well as other factors ass ciated with acce ibility during pa t operation e.g temperature, atmosphere, cation, etc.), a dthe reconinendatio of Regul a~t otryud . gnd 8.10. Th addition or dele ion of any hydraulic or mechanic snubb shall bmade in acco dance with Sectio 50.59 of 10 CFR Pa t50. In\evel The visual spection frequen is based upon ma tading a consta tilevel of s ubber protection to systems. Therefore, the req 'red inspection i terval vanie inversely w th the observed nubber failures an is determined b the number f inoperable snubbers found d ing an inspection. Inspections perfo ed instterval. before hat interval has elapsed may be used as a new reference point no determine the next ins ection. However, the results of sc early Inspection performed efore the or' inal required t e interval has elIa sed (nominal time less 25%.) m y not be us to lengthen th required inspectio interval. Any inspection w ose results r quire a shorter spection interval w 1 override the previous sche le. MILLSTONE - UNIT 2 B 3/4 7-5 Amendment Nos. , ,, hJ*, 71, 7*729rr evised NRC by letter dated November 24, 1998
PLANT SYSTEMS PLAN-h-3I , 2OOO-SYSEMSMe
- BASES When the cause ýf the rejection of a snubber is clea ly established and remedied for that snu ber and for any ot r snubbers that ay be generically usceptible, that snub er may be exempted from being countl as inoperable.
G erically susceptible nubbers are those hich are of a s ecific make or mo 1 and have the same sign features dire tly related to rejection of the f snub er by visual inspecti , or are similarly located or expose -to the same envir mental conditions suc as temperature, ra jation, and vibra ion. Whe a snubber is foun inoperable, an ngineering evalu tion is performed, in addition to the termination of the snubber mode of ailure, in order t determine if any s ety-related compon nt or system ha been adversely a ected by the inoperabi ity of the snubber. The engin ering evaluation shal determine whether or not the snub er mode of failur has imparted a significant effect or egradation on t supported compone t or system. To provide ass rance of snubber relia ility, a represent tive sample of the installed snubbe will be tested duin plant shutdowns at eighteen (18) onth intervals. Obs rved failures of thes sample snubbers s all require t ting of additional u its. Hydraulic snubbers ad mechanical snubbers may each be trea d as a diffe ent entity for the ab e surveillance program The service life of a s ubber is evaluated vi manufacturer inpu and informati n through considera ion of the snubber rvice conditions nd associated installation and main enance records (newly i stalled snubber, se 1 replaced, s ing replaced, in radiation area, in hi h temperature area, etc.... ) T requirement to ni r the snubber service ife is included to ensure that t snubbers periodicall undergo a performance valuation in view of their age an operating conditions. These records will pr ide statistical .bases for future onsideration of snub en service life. The r uirements for "the maintenance o records and the snubb r service life review a e not intend ed to affect plant eration.
\
3/4.7.9 DELETED MILLSTONE - UNIT 2. B 3/4 7-6 Amendment Nos. JJ, 7A, p, JJ, 0483 79J, 70
Docket Nos. 50-336 50-423 B18556 Attachment 4 Millstone Nuclear Power Station, Unit No. 2 Technical Specifications Change Request 2-21-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Retyped Pages
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ..... ................ 3/4 7-1 Safety Valves ....... ................ 3/4 7-1 Auxiliary Feedwater Pumps ............. 3/4 7-4 Condensate Storage Tank .... ........... 3/4 7-6 Activity . . . . . . . . . . . . . . . . . 3/4 7-7 Main Steam Line Isolation Valves ..... 3/4 7-9 Main Feedwater Isolation Components (MFICs) 3/4 7-9a Atmospheric Dump Valves .... ........... 3/4 7-9c Steam Generator Blowdown Isolation Valves . 3/4 7-9d 3/4.7.2 DELETED . . . . . . . . . . . . . . . . . 3/4 7-10 I 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM ........... 3/4 7-12 3/4.7.5 DELETED ........ .................. 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM . 3/4 7-16 3/4.7.7 DELETED ........ .................. 3/4 7-19 3/4.7.8 DELETED ........ .................. 3/4 7-21 3/4.7.9 DELETED ........ .................. 3/4 7-33 3/4.7.10 DELETED ........ .................. 3/4 7-33 3/4.7.11 ULTIMATE HEAT SINK . ........................ ..... .3/4 7-34 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES . . . . . . . . . . . 3/4 8-1 Operating ..... .............. 3/4 8-1 Shutdown . . . . . . . . . . . . . 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS . 3/4 8-6 A.C. Distribution - Operating . . . 3/4 8-6 A.C. Distribution - Shutdown . . . 3/4 8-7 D.C. Distribution - Operating . . . 3/4 8-8 D.C. Distribution - Shutdown . . . 3/4 8-10 D.C. Distribution (Turbine Battery) Operating 3/4 8-11 MILLSTONE - UNIT 2 VIII 77, M 9, 1,9, 0772 10* Amendment X, 790, No. 101, 11, ?Xl, !779 7,
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...... ................. B 3/4 7-1 3/4.7.2 DELETED . . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM ............ B 3/4 7-4 3/4.7.5 DELETED .......... ................... B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM B 3/4 7-4 3/4.7.7 DELETED . . . . . . . . . . . . . . . . . . B 3/4 7-5 II 3/4.7.8 DELETED .......... ................... B 3/4 7-5 3/4.7.9 DELETED .......... ................... B 3/4 7-6 3/4.7.10 DELETED .......... ................... B 3/4 7-7 3/4.7.11 ULTIMATE HEAT SINK . .. . .. .. . .. . . B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS . . . S. . . . . . . .. . B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .... ............. B 3/4 9-1 3/4.9.2 INSTRUMENTATION ..... ............... B 3/4 9-1 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS ......... B 3/4 9-1 3/4.9.5 DELETED ........ .................. B 3/4 9-1 3/4.9.6 DELETED ........ .................. B 3/4 9-2 3/4.9.7 DELETED ........ .................. B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLING RECIRCULATION B 3/4 9-2 MILLSTONE - UNIT 2 XIII Amendment No. 0, 77, M 9, M, 0773 XInf, 19, 700,
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1
- a. The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
- b. DELETED MILLSTONE - UNIT 2 3/4 4-18 Amendment No. jj$,
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- UNIT 2 3/4 7-22a Amendment No. g, 11g, Ifo, 77g,
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- UNIT 2 3/4 7-22b Amendment No. g, 110, 7AR,
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- UNIT 2 3/4 7-32 Amendment No. 77, 99, J1,
REACTOR COOLANT SYSTEM BASES The maximum RTNDT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 50*F. The Lowest Service Temperature limit is based upon this RTNDT since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100'F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement. Included in this evaluation is consideration of flange protection in accordance with 10 CFR 50, Appendix G. The requirement makes the minimum temperature RTNDT plus 90'F for hydrostatic test and RTNDT plus 120°F for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure. Since the flange region RTNDT has been calculated to be 30°F, the minimum flange pressurization temperature during normal operation is 150'F (161°F with instrument uncertainty) when the pressure exceeds 20% of the preservice hydrostatic pressure. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement. To establish the minimum boltup temperature, ASME Code Section XI, Appendix G, requires the temperature of the flange and adjacent shell and head regions shall be above the limiting RTNDT temperature for the most limiting material of these regions. The RTNDT temperature for that material is 30*F. Adding 10.5°F, for temperature measurement uncertainty, results in a minimum boltup temperature of 40.5°F. For additional conservatism, a minimum boltup temperature of 70°F is specified on the heatup and cooldown curves. The head and vessel flange region temperature must be greater than 700 F, whenever any reactor vessel stud is tensioned. The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. Verification that pressurizer temperature conditions are within the limits of LCO 3.4.9.2, at least once per 30 minutes, is required when undergoing planned changes of > 10°F. The 30 minute time interval permits assessment and correction for temperature deviations within a reasonable time. MILLSTONE 0780
- UNIT 2 B 3/4 4-7 Amendment No. X, 70, 99, 710,
PLANT SYSTEMS BASES a feedwater isolation signal since the steam line break accident analysis credits them in prevention of feed line volume flashing in some cases. Feedwater pumps are assumed to trip immediately with an MSI signal. 3/4.7.1.7 ATMOSPHERIC DUMP VALVES The atmospheric dump valve (ADV) lines provide a method to maintain the unit in HOT STANDBY, and to replace or supplement the condenser steam dump valves to cool the unit to Shutdown Cooling (SDC) entry conditions. Each ADV line contains an air operated ADV, and an upstream manual isolation valve. The manual isolation valves are normally open, and the ADVs closed. The ADVs, which are normally operated from the main control room, can be operated locally using a manual handwheel. An ADV line is OPERABLE if local manual operation of the associated valves can be used to perform a controlled release of steam to the atmosphere. This is consistent with the LOCA analysis which credits local manual operation of the ADV lines for accident mitigation. 3/4.7.1.8 STEAM GENERATOR BLOWDOWN ISOLATION VALVES The steam generator blowdown isolation valves will isolate steam generator blowdown on low steam generator water level. An auxiliary feedwater actuation signal will also be generated at this steam generator water level. Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main feedwater. The steam generator blowdown isolation valves will also close automatically upon receipt of a containment isolation signal or a high radiation signal (steam generator blowdown or condenser air ejector discharge). 3/4.7.2 DELETED 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM The OPERABILITY of the Reactor Building Closed Cooling Water (RBCCW) System ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. The RBCCW loops are redundant of each other to the degree that each has separate controls and power supplies and the operation of one does not depend on the other. In the event of a design basis accident, one RBCCW loop is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. To ensure this requirement is met, two RBCCW loops must be OPERABLE, and independent to the extent necessary to ensure that a single failure will not result in the unavailability MILLSTONE - UNIT 2 B 3/4 7-3a Amendment No. 71g, 77ý, 7, 0796
PLANT SYSTEMS BASES 3/4.7.4 SERVICE WATER SYSTEM (Continued) The Technical Specification Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the accident analysis are met and that subsystem OPERABILITY is maintained. The purpose of the service water pumps differential pressure test, Surveillance Requirement 4.7.4.1.a.2, a substantial flow test, is to ensure that the pumps have not degraded to a point where the accident analysis would be adversely impacted. The surveillance requirement acceptance criteria for the service water pumps was developed assuming a 7% degraded pump from the actual pump curves. Flow and pressure measurement instrument inaccuracies for the service water pumps have been accounted for in the design basis hydraulic analysis. It is not necessary to account for flow and pressure measurement instrument inaccuracies in the acceptance criteria contained in the surveillance procedure. 3/4.7.5 DELETED 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. The LCO is modified by a footnote allowing the control room boundary to be opened intermittently under administrative controls. For entry and exit through doors the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in constant communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated. The control room radiological dose calculations use the conservative minimum acceptable flow of 2250 cfm based on the flowrate surveillance requirement of 2500 cfm + 10%. MILLSTONE 0797
- UNIT 2 B 3/4 7-4a Amendment No. 11p, 1f7,
PLANT SYSTEMS BASES 3/4.7.7 DELETED 3/4.7.8 DELETED MILLSTONE - UNIT 2 B 3/4 7-5 Amendment Nos. JJ, ;, P, 11f, 0782
PLANT SYSTEMS BASES 3/4.7.9 DELETED MILLSTONE - UNIT 2 B 3/4 7-6 Amendment Nos. JJ, X, y, j7?, 07R2 71J, g%;9,
Docket Nos. 50-336 50-423 B18556 Attachment 5 Millstone Nuclear Power Station, Unit No. 3 Technical Specifications Change Request 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Marked Up Pages
U.S. Nuclear Regulatory Commission B1 8556/Attachment 5/Page 1 List of Affected Panes Technical Specification Affected Page with Section Number Title of Section Amendment Number Index Page viii Amend. No. 115 Index Page x Amend. No. 160 Index Page xiv Amend. No. 136 Index Page xv Amend. No. 192 3/4.4.9.1 Pressure/Temperature Limits 3/4 4-33, Amend. No. 197 3/4 4-36, Amend. No. 197 3/4.7.2 Steam Generator 3/4 7-10, Original issue Pressure/Temperature Limitation 3/4.7.6 Flood Protection 3/4 7-14, Amend. No. 144 3/4.7.10 Snubbers 3/4 7-22, Amend. No. 71 3/4 7-23, Amend. No. 167 3/4 7-24, Amend. No. 16 3/4 7-25, Original issue 3/4 7-26, Amend. No. 173 3/4 7-27, Amend. No. 100 3/4 7-28, Amend. No. 100 3/4 7-29, Amend. No. 100 3/4.7.11 Sealed Source Contamination 3/4 7-30, Amend. No. 100 3/4 7-31, Amend. No. 100 3/4.7.14 Area Temperature Monitoring 3/4 7-32, Amend. No. 141 3/4 7-33, Amend. No. 100 3/4 7-34, Amend. No. 100 3/4 7-35, Amend. No. 182 3/4.4.9 Bases Pressure/Temperature Limits B 3/4 4-12, Amend. No. 197 3/4.7.2 Bases Steam Generator B 3/4 7-7, Amend. No. 151 Pressure/Temperature Limitation 3/4.7.6 Bases Flood Protection B 3/4 7-10, Amend. No. 144 3/4.7.10 Bases, 3/4.7.11 Bases, Snubbers, Sealed Source B 3/4 7-23, Amend. No. 184, B 3/4.7.14 Bases Contamination, Area 3/4 7-24, Amend. No. 136, B 3/4 Temperature Monitoring 7-25, Amend. No. 136
INDEX May-26--4-99& LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >IyCi/gram DOSE EQUIVALENT 1-131 ........ ............. ...... .3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ................. .......................... 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ...... ................ .. 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY ............................. 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 10 EFPY ........... ................. 3/4 4-35 TABLE 4.4-5 R-ACT-OfR VESSEL MATERIAL ,UR 1 [tLANE PROGRA?' -"C-LE j, WITHDAWA ....... C.H. .............. .... 3/4 4-36 Pressurizer ............. ........................ .. 3/4 4-37 Overpressure Protection Systems ..... .............. .. 3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (FOUR LOOP OPERATION) ....... 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION) .......... 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY ................. 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ..... ............... 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............. ...................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,,g GREATER THAN OR EQUAL TO 350°F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ..... ............... P 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ..... .......... 3/4 5-10 or 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity ................ 3/4 6-1 Containment Leakage ................. 3/4 6-2 Containment Air Locks ................ 3/4 6-5 Containment Pressure ................. 3/4 6-7 MILLSTONE - UNIT 3 viii Amendment No. M P7,, A15
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEIUIREMENTS SECTION
-PAGE TABLI"E3.7-3 STEAM LINE :SAFETY VALVES PER LOOP ... ......... ... 3/4 7-3 Auxiliary Feedwater System ............. ....... 3/4 7-4 Demineralized Water Storage Tank .... ......... ... 3/4 7-6 S. Specific Activity- - .- ... .... - -.. . .-. 3/4 7-7 TABL E 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ...... ................ ... 3/4 7-8 Main Steam Line Isolation Valves .... ......... ... 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines 3/A 7-9a 3/4. 7.2 STEAMI GENERATOR PRESUR[,_'T[iPERATURE 1i 3/4. 7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM .... 3/4 7-11 3/4. 7,4 SERVICE WATER SYSTEM ...... ................ ... 3/4 7-12 3/4. 7.5 ULTIMATE HEAT SINK 3/4 7-13 3/4. 7.6 ....... O FLOOD , . . . . 3/4 7-14 3/4. 7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ..... ... 3/4 7-15 3/4. 7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM . . 3/4 7-18 3/4. 7.9 AUXILIARY BUILDING FILTER SYSTEM .... ..........
3/4 7-20 3/4. 7.10 SNUBEj5. ... . . . . . . .......... ....... ... 3/47-2 TABI .E 4.7-2 SNUBBER VISURL NSPECTIONINTERV-, --
,-.3/4 7-27 FIGURE 4.7-1 SAMPLE PAN 2) FO SNU EL FUNCTIONAL TEST .... 3/4 7-29 3/4..7.11 SE.A.LED SOURCECO..AINATION.. ..... ..........
3/4 7-30 3/4. .7.12 DELETED Tab1 e 3.7-4 DELETED Tabl e 3.7--5 DELETED
.7.13 .DELETED ." "- - .- "
3/4..7.14 AREA. TEMPERATUREMON*-, ITORIN1 3/4 7-32 TABILE 3.7-6 AREAT.E.PERATURE MO;'IT " r' E V 3/4 7-33 MILLSTONE - UNIT 3 X Amendment No. lopB*,X* 0583
INDEX April 10, 19g( BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>lMeV) AS A FUNCTION OF FULL POWER SERVICE LIFE .......... ................. B 3/4 4-10 3/4.4.10 STRUCTURAL INTEGRITY ......... .................. .. B 3/4 4-15 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ....... .............. .. B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ....... ..... ...................... .. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ..... ............... .. B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK ....... .............. .. B 3/4 5-2 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ... ......... .. B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ........ ................... .. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .... .......... .. B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES ....... .............. .. B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL..... .. ............... .. B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM ... ......... .. B 3/4 6-3b 3/4.6.6 SECONDARY CONTAINMENT ........... .................. B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ........ ... ....................... B 3/4 7-1 3/4.7.2 STEAN GENERATOR PRESSURE/TEMPERATURE LIMITTIffb B 3/4 7-7 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ...... B 3/4 7-7 3/4.7.4 SERVICE WATER SYSTEM .................. B 3/4 7-7 3/4.7.5 ULTIMATE HEAT SINK ..... . ............. B 3/4 7-8 3/4.7.6 FLOOD PROTETION, .. ........ ... B 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............ B 3/4 7-10 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM .......... B 3/4 7-17 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ............ B 3/4 7-23 3/4.7.10 SNUftS L-................... B 3/4 7-23 MILLSTONE - UNIT 3 xiv Amendment No. #p, py, jjý, 17y, d,
INDEX Jefuaq6'2N P9 BASES SECTION PAGE 3/4.7.11OVUVRL ENI1 . . . . . . . . . . . B 3/4 7-25 3/4.7.12 DELETED 3/4.7.13 DELETED tC/e--* rT 3/4.7.14 AREA TEMPERATURE-MONITORING. ............ ..... B 3/4 7-25 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION ............ B 3/4 8-1 3/4.8.4 DELETED ............ ....................... . . . B 3/4 8-3 I 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ........ ................... ... B 3/4 9-1 3/4.9.2 INSTRUMENTATION .......... ..................... ... B 3/4 9-1 3/4.9.3 DECAY TIME ..... ..................... ......... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS .... ............ ... B 3/4 9-1 3/4.9.5 COMMUNICATIONS ......... ..... ..................... B 3/4 9-1 3/4.9.6 REFUELING MACHINE ........ .................... ... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS ..... .......... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ........ .. B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ..... ... B 3/4 9-7 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL ......... ....... ..................... B 3/4 9-8 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM .... ........... ... B 3/4 9-8 3/4.9.13 SPENT FUEL POOL - REACTIVITY ....... .............. ... B 3/4 9-8 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN .... ............ ... B 3/4 9-9 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ........ ... ..................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS . B 3/4 10-1 3/4.10.3 PHYSICS TESTS ........ ... ...................... .. B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS .......... .................. B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN .............. .. B 3/4 10-1 MILLSTONE - UNIT 3 xv -Amendment No. f, J , 77, 779, 79,
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSUREITEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.1 Reactor Coolant System (except the pressurizer) temperature, pressure, and heatup and cooldown rates of ferritic materials shall be limited in accordance with the limits shown on Figures 3.4-2 and 3.4-3. In addition, a maximum of one reactor coolant pump can be in operation when the lowest unisolated Reactor Coolant System loop wide range cold leg temperature is
< 160'F.
APPLICABILITY: At all times. ACTION:
- a. With any of the above limits exceeded in MODES 1, 2, 3, or 4, perform the following:
- 1. Restore the temperature and/or pressure to within limit within 30 minutes.
AND
- 2. Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System remains acceptable for continued operation within 72 hours. Otherwise, be in
..at least MODE 3 within the next 6 hours and in MODE 5 with RCS pressure less than 500 psia within the following 30 hours.
- b. With any of the above limits exceeded in other than MODES 1, 2, 3, or 4, perform the following:
- 1. Immediately initiate action to restore the temperature and/or pressure to within limit.
AND
- 2. Perform an engineering'evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System is acceptable for continued operation.:prior to entering MODE 4.
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup and cooldown operations, and during the one-hour period prior to and during inservice leak and hydrostatic testing operations. 4.4.9.1.2 Ie reactor N*ssel materi lirradiation s cveillance specimen s,11 be remov d and examii4d, to determX~ne changes in terial prptes, as equired by S CFR Part Appendix H, in accordance ith the schelule in TL le 4.4-5. e results these exami tions shall menused to upJdate lFigure, 3.4-2 and *.4-_3 as reqNured. MILLSTONE - UNIT 3 3/4 4-33 Amendment No. JX7,
August 27, 2001 Millstone 3 Reactor Coolant System "teatup Limitations for Fluence up to 1.97E+19 n/cm (32 EFPY) Foj- r Cbj, 1 5 2500-1 2000 1500 11 U, 0 0 i i- .. - -
.4 - -I -4 .-- 4 I4 . 4I 4 I I J. ~44 -
1JMMm I I-1117i I tA)U I I __ - - - 7 { II FL . I
>*'1 b-)<1 I [[ II .
I I-I-- I4l I I SUnacceptable 1 i1 . Operation I:I I li-Hi{
- I i-,._ -' I
`IHeatup At A Maximum of 40'F in Any 1-Hour Period Up To 160°F,
-I t!t It and At A Maximum of 80'F in Any Hour Period Above 1600F.
I-,.- VV- 1 1, 1 F II 500 Aff" Hydrostatic and Leak Test p eration: Temperature Changes teiA Limited To A Maximum of 5'F in N AnyI -Hour Period Ft o fl4" F[-.i *[F.ill;** *:.. **: LLII LLI"F.IT.II 11I f$;[I liii Fl I I I *l.*4J* B*t* *., *- . *.* t I : * *E I I I - i L I -
- I -
V - 0 50 100 150 200 250 300 350 400 SIndicated Cold Leg Temeprature (TF) FIGURE 3.4-2 MILLSTONE - UNIT 3 3/4 4-34 Amendment No. Yj, Tji, 197
Millstone 3 Reactor Coolant System August27,2001 Cooldown Limitations for Fluence up to 1.97E+19 n/cm (32 EFPY) 2500 2000 1500 0 a, a 0 02 0) 0 -D 0 0) a -a 1000 Cooldown At A Maximum of 80 0 F In Any 1-Hour Period To 160 0 F, Then At A Maximum of 20°F In Any 1-Hour Period Below 160OF 500 0 0 50 100 150 200 250 300 350 400 Indicated Cold Leg Temeprature (°F) FIGURE 3.4-3 MILLSTONE - UNIT 3 3/4 4-35 Amendment No. 99, 0*7, 197
ymugus! 2001---o-- 4TLY L EFT MILLSTONE -'UNIT 3 3/4 4-36 Amendment No,. Of, J*7, iY7
PLANT ScCEMS 3/4.7.2 S AM GENERATOR PRESSUR TEMPERATURE LIMITATI* LIMITING CONDIT qN FOR OPERATION 3.7.2 The temperatur s of both the reactor d secondary.oolants in the tstm generators shall e greater than 70*F wh n the pressure cool nt in the steam gen ator is greater than of ei er 0 psig. APPLICA ILITY: At 'all time ACTION: With the req rements of the above specification not sati ied:
- a. R~ed.uct the steam generator ressure of the applicab less thn or equal to 200 psi side to within 30 minutes, an
- b. Perform an engineering evaluatio to determine the effec of the overpre -urization on the str tural integrity of the steam generat r. Determine that th steam generator acceptable for ontinued operation p ior to increasingremains its temperatures ab eo 200-F.
4.7.2 The pres re in each side o ~the steam generator sha 1l be determined to be less than 200 sig at least once r hour when the temper re of either the reactor or sec ndary coolant is le s than 70*F. MILLSTONE - UNIT 3 3/4 7-10 j- AsNO
7t2t7`ý POWTSYSTMS 3/4.7\. FLOROTEC 0 LIMITIN\G ODITION FOR OPE ION 3.7.6 Flood p tection shall be ovided for the ser ice water pump cu nd components en the water level xceeds 13 feet Mea Sea Level, iJSGS athe Unit 3 tnt e structure. PP IT: At a times. ACTION: With the ter level at gr ter than 13 feet a ve Mean Sea Level, SGSc4 at the UniL t intake structu e, shut the waterti t doors of both se vic pump cubicl e and close the p p cubicle sump drai valves within 15 *nt 'I SURVEILLANCE RE IREMENTS 4.7. The water leve at the Unit 3 i ake structure shall e determined to bi withi the limits by:
- a. Measurement at east once per 24 ours when the water vel is below levation 8 feet bove Mean Sea Le 1, USGS datum, and
- b. Me urement at leas once per 2 hours hen the water level I equal to o above elevation 8 feet above Mean Sea Level, USGS datu 4'Urs I4 e ,- t.-----Ny
*)c .............. --
MILLSTONE - UNIT 3 3/4 7-14 Amendment No. A$
PLANT SYS MS 1i29 PLT3Z4.7.10 SN BERS I TMTTNA r':-nNnTT N') FnR fnPFRATTnl \ 3.7.10 All snubbe shall be OPERABLE. The only snubbers xcluded from the ,ýequirements are th e installed on non fety-related syste and then only Stheir failure or f ilure of the system n which they are stalled would ha no adverse effect n any safety-relat system. APPLI BILITY: MODES 1, 3, and 4. MODES and 6 for snubbers ocated on system required OPERABLE 1 those MODES. ACTION: With one or ore snubbers inoper le on any system, ithin 72 hours rep ce or restore th inoperable ubber s) to OPERABLE atus and perform engineering eva uation per Specific tion 4.7.10g. on t attached componen or declare the tached system mnope able And follow th appropriate ACTION statement for that system.
- 4. .10 Each snubber s 11 be demonstrated0PERABLE by perfor ance of the following augmented inse ice inspection pr qram and the requ cements of As used in this sp cification, "type f snubber" shall ean Xnubbers of the same design and manufac rer, irrespective ~f Snubbe icateorized as inaccessible or acPessible during racto en to thefoundation Eachogor iesuo (ing srcessible and fucional, man (3 inapeter thdepenatt yachn of ig to the l isua1 inspectionsshlveiyta(tesnbrhso
~to th component and to the snubber anchorag, are functional.
Snubber ,whi a ineabop able as a result of vis ual inspections shtevall be r classif eda fsnuacceptallbleaderecl assified acceptubl for thepurposnceo salext visual inspection ntehal s providd l tat (1) thecause of rejection is clearly e tablished and rem ied for that particu r snubber and for other nnubbers irrespect e of MILLSTONE - UNIT 3- /4-2Aenment No. 32*"2~ IA"
type that ma be generically s sceptible; snubber is fu ctionally tested 'n the and (2 the affected determined OPE BLE per Specificati n as-found ndition and onnected to an i operable common hy aulic 4.7.10.f. All s bbers found unted as unac eptable for -dete iningfluid reserv ir shall be i erval. A revie and evaluation sha the next inspection to *ustify continu be performed and cumented operation contnued operation cannot be with a unacceptable snu er. if justifi *, the snubber s decl ed inoperable a the ACTION requir ments shall be me all be
- d. Transie t Event Inspect n An inspe tion shall be erformed of all ubbers attached sections o systems that ve experienced une ected, potential o damaging tr nsients as dete ined from and a visua inspection of a review f operational data such an even e systems within months following In addition o acceptance cri ria, freedom-of satisfying the *sual inspection tion of mechanica snubbers shall be verified us g at least one of the following:
induced snubber ovement; or (2 evaluation (1) manually piston setting; o (3) stroking t mechanical of lace snubber In-snubbe through its full range of travy.
- e. nctional Tests Du ing the first refueling shutdown and INT RVAL thereafter,* a representative least once each RE UELING type shall be tested usi s mple of snubbers o each one of the fol owing sample plans. The sampl plan for each type all be selected r and c not be changed nor to the test pe od Admini rator shall be duro. g the test peni d. The NRC Regio I fied in writing of the sample pla selecte for each snubber t e prior to sample p n used in the prior est period the test period or the shall be implemented:
- 1) At le st 10% of the total f each type of nubber shall be functi nally tested either each sn bber of a type that -place es not or in a ench test. For meet the unctional test acc-tan ecriteria of Specif ic ti on 4.7.Of of t at pe of-snubber shall bfunctionally ., a additional 5%
more fail es are found or until 11 snubbers te ted until no of tht type have been funct nally tested; or
- Excep the survei ance related to sn bber functional testing ue no later than March 0, 1999 ma be deferred until he end of the next refu ling outage or no lat than Septe ber 10, 1999, whi ever is earlier.
ILLSTONi - UNIT 3 J3/4 7-23 Amendment $7, 7$, lo177, #7 I I3 r1+U Pi96C l:I1Thrtc, y LCFr ,cLA
Ann 1 7 1988 0 PLANT SYSTE (s, URVEILLANCE RE IREHENTS (Continu 8)
- e. Functiona Tests (Continued)
- 2) A repres ntative sample or acb type of snubber hall be func tionally sted In accordane with Figure 1i.7-1. "'C Is the total numbe of snubbers of a ype found not meetfi the accept ance require nts of Specificat n 4.7.10f. The cum ative number of snub rs of a type test d is denoted by ON". Test sults shall beplotted 3equentia y.>in the order of a ple a signment (i.e. ach snubber shal be plotted by Its a igned or r in the random sample, not by order of testing). if atV any lime the point p tted falls in the "Accept* region. te *in&
of: n bers of that ty e may be terminat, - When the point plotte Iles In the "Co inue Testing* reg on, additional snubbers f that type sha 1 be tested until he point falls in the 'Accep " region or the "Reject" region, o all the snubbers of that typ have been teste ; or
- 3) An initial rep sentative sampi of 55 snubbers s 1I be func tionally tested. For each snubb type which does ot meet. the functional test a ceptance criteri another sample at least one-half the size othe initial sa It shall be teste until (
the total number te~s ed Is equal to einitial sample - ze itipie by the fac r. 1 - C/2, vhe "' Is the numbe of anbrs found which do not nee the fun ional test accep nce crit ra. The result f znthis sample p n shall be plotte using n "Accept" line hi h follows the eq tion N i55
- C/2 ). Each snubber point hould be plotte as soon as the snubber tested. If the po nt plotted falls n or below the "Accept" 'ne, testing of that ype of snubber be terminated. If the point plotte falls above the "Accept line, testing must ontinue until the pnt falls In the Accept" region or all e snubbers of that ype have been te ed.
s=ting equipment fail re during functlona testing may inva *date t day's testing and a low that day's test g to resume anew at a Ilat. time provided all a ubber3 tested with te failed equipine durin the day of eQuipment failure are reteste . The representa ive sample elected for the func ional test sample p ns shall be randomly elected fro= the nn bers of each type a d reviewed befor bebhetesting. The rev c shall ensure, as a as practicable, that they are repre ntative of the Yar us configuratlo . operating environi nts, range of size, nd capacity of mnubbers of each type. Snubbers laced in the aaue cation as snubbers which lted the previous fu ctional test shall e retested at the time of th next functional tes but shall not be i luded in the sample plan. during the function I testing, addition I sampling is required due to failure of On one type of anubb , the functional test. result shall be rev iewed a that time to deter inc if additional samples a ould be limited to t type of snubber w "ch isS failed the functional testing. MILLSTONE - UNIT 3 3/4 7-2J Amendment No- /-i t T1_iTLt*. &F-
SYSIMS MPLANI SURVEILLANCE E UIREMENTS (Contir ed f- Function I Test Acceptance iteria The snubber unctional test sha verify that: Activatio (restraining action is achieved with the specified r ge. in both tension nd compression;
- 2) nubber bleed, r release rate whe required, is pre nt in b h tension and ompression, within he specified rang
- 3) For chanical snubb rs, the force requi ed to initiate or mainta motion of th snubber is within e specified range in both irections of tvel; and
- 4) For snubber specifically quired not to disp ce under continuous lo d, the ability of the snubber to w hstand load without displa ment.
Testi methods may be sed to measure arameters indirect or paramet rs other than th e specified i -those results can b correlate to the specifie parameters th ugh established met ods.
- g. Functional st Failure Anal .s An engineering valuation shall made of each ilure to meet the functional test a ceptance criteri to determine e cause of the ailure. The resu s of this evalua ion shall be'u d, if applicable, i selecting snubber to be tested in n effort to de rmine the OP subj BILITY teto__the of other-s same Tai.re ubbersmode."
irrespect-ve_ of type
, whic may be e...
For the ubbers found nop rable, an engine ring evaluation shall be perform d on the componen to which the i perable snubbe are attached. e purpose of this engineering- eval ation shall be determine if e components to ich the inoperab e snubbers are attached were a versely affected the inoperabilty of the snubb s 9nd7gned order tose.rvice. ensu
- that the
.. compone _* t remains
_* capabl of meeting etn h the If a snubber selecte for functional tsting either fal s,to
-lock u or fails to mov i.e.., frozen-in lacel, the cause ill be evaluate and, if caused manufacturer or design deficienc all snubbers the same type sl bject to the sam de-fect shall bhe unc tionallyte ted This testi-" requirement sha 1be independent of the requirm ts stated in Spejif ication 4.7.1o0 for snubbers no meeting the fu tional test acceptance criteria.
MILLSTONE - UNIT 3 3/4 7-25 Mo )
PLANT S AugYs SURVEILL CE REUREEkS (Continued) ' h.! toal Te~stu of Re aired nd Re laced %Sribers Snubb rs which fai the visual ins ection or the nctional test accept ce criteria hall be repair or replaced. eplacement snubbers and snubbers hich have rep irs which might ffect the functiona test result shall be teste to meet the fu tional test criteria b ore installa ion in the unit Mechanical sn bers shall have met the acceptance c*teria subseque to their most ecent service, and e freedom-o otion test mus have been per ed ithin 12 mont before bein installed in t unit.
- i. Sn bbr Service Li e Pro ram The s vice life of draulic and echanical snubb s shall be monito d to ensure th t the servic life is not exc eded between surveill nce inspection The maximu expected servi life for various s ls, springs, a d other crit al parts shall e deter ined and e tablished base on engineeri g information a shall be tended or ortened based n monitored st results and ailure hi tory. Cri 'caI parts sha1 be replaced o that the maxi m ser ice life wi 1 not be excee d during a p iod when the snrbber is r uired to b OPERABLE. Th parts replac ents shall be d u mente and the do mentation shal be retained accordance wit Quality ssurance ogram Topical port. e&
I.. MILLSTONE - UNIT 3 3/4 7-26 Amendment No. P
]ABLL 4./-Z B E VSUAL INSPECTION I FERVAL SN NUMBER OF UNACCEPTAB SNUBBERS Populat n Column A Column B Column C or Catego Extend Interva Repeat Interval Reduce Interval (Notes aotes 4 and 6 Notes 5 and 6) 1 0 0 1 vv 800 0 2 100 0 1 4 150 0 \38 200 2 5 13 300 5 1 25 7* *S0122 2O20 40 24
- 78 48 I
1000 or *eater 2\9 X*56 109 Note 1: Th next visual inspecti interval for a snub er population or cate ory size shall be dete mned based upon the pre ious inspection inter I and the number of un cceptable snubbers fo d during that Z interval. Snubbers may b -categorized, based upon their
-4 accessib ity during power operat on, as accessible or i ccessible.
These cateories may be examined eparately or jointly. However, the license must make and doc nt that decision be re any L *inspection an shall use that decisi as the basis upon wich to determine the n t inspection interval or that category. Note 2: Interpolation betw n population or catego sizes and the number unacceptable snubbe is permissible. Use ext lower integer fo e value of the i it for Columns A, B, C if that integer i luded a fractional v ue of unacceptable sn bers as determined by terpolation. Z Note 3: If the umber of unacceptabi snubbers is equal to less than the number i Column A, the next spection interval ma be twice the "previous i terval but no greate than 48 months. T Note : If the number f unacceptable snub rs is equal to or less than the number in Colu B but greater an number in Column.A, e next inspection inter 1 shall be the same s the previous interv Note 5: If the number of un ceptable snubbers i equal to or greater t n e number in Col C, the next insp tion interval shall t -thirds of the pre ous interval. Howe r, if the number. of una eptable snubbers i less than the n r in Column C but great than the number i Column B, the nex interval shall be reduce proportionally by Iterpolation, that s, the previous interval hall be reduced by a actor that is one-tb rd of the'ratio of the di erence between the n er of unacceptable nubbers found during the revious interval an the number in Col B to the difference in the numbers in Colum B and C. MILLSTONE - UNIT 3 314 7-27 Amendment No. 01,194 0261
1 anuarv 3 lfflS ', TABLE .7-2 SNUBBER VISUAL INS TION INTERVAL Not The provis* ns of Specification 4.0. are applicable or all inspection i tervals up to and includi 48 months. t'OG& ZLNT lTJZcNALLL( L-Cr LLYIHIK 0ILMONE 0261
- UNIT 3 3/4 7-28 Amendment Ho. J, 4
no~aTZT99B7ý MILSOEN!13 3/4 7-2~9-'~ mnmn No. Amendment o If. #0 0281
0Q_-Q_ PLANT SYSTEMS 3L4.7.11 SEAh SOURCE CONTANIN N LIMITING CONDITIO FOR OPERATION
- 3. 11 Each sealed sour containing radioac *ve material either in e ess of 100 'croCuries of beta a d/or gamma emitting material or 5 microCuri of alpha itting material s 11 be free of eater than or equal to 0.005 m* roCurie of removable ntamination.
k APPLICABILI :At all times. S ACTION:
- a. With a aled source having re vable contamination excess of the
*- ~ ~~either: ousan LL above li its, immediately with w the sealed source from use and
- 1. Decontam ate and repair the se ed source, or
- 2. Dispose of e sealed source in a ordance with Comnmis 'on Regulations.
- b. The provisions of Speci cation 3.0.3 are not licable.
SURVEILLANCE RE IREMENTS 4.7.11.1 Test Requir nts - Each sealed sourc shall be tested for akage and/or contamination by:
- a. The licensee, or
- b. Other persons speci ically authorized by t Commission or an Agreement State.
-* The test thod shall have a det tion sensitivity of least 0.005 SmicroCurie prtest sample. S4.7.11.2 Test requencies - Each cate ry of sealed sources xcluding startup sources a fission detectors previou subjected to core flu shall be tested at the fr uency described below.
- a. Sources in e - At least once per 6 mo hs for all sealed sour s containing ra ioactive materials:
- 1) With a half- ife greater than 30 days (e luding Hydrogen 3),
and
- 2) In any form other han gas.
C MILLSTONE 0261
- UNIT 3 3/4 7-30 Amendment No. ,7,/¶(6 4kJ.
PLANT SYSTEMSJau.,,!9S
ýWRYILANE REUI*3OM*S (Continued)
- b. Stored sou es not in use- Ea sealed source an fission detector shall be tes ed prior to use or ransfer to anothe licensee unless tested withi the previous 6 mon s. Sealed sources and fission detectors tran ferred without a ce ificate indicatin the last test date shall be tsted prior to being laced into use; a C. Startup sources a 9--
fission detectors Each sealed start source and fission detect shall be tested w hiln 31 days prior o being ubjected to core or installed in t e core and followi repair maintenance to th source. 4.7.11.3 Repo s - A report shall prepared and subm ted to the Commiss n on an annual ba s if sealed source r fission detector akage tests reveal the presence of ater than or equal o 0.005 microCurie f removable contamination.
-ILLSTONE N- i, AtA L-CTUT 9'1473 MILST ta-rET 3 3/4 7-31 Awmenent No4006
3/4.7o14\ ARE RMONTOIN \EPR
.7.14 The te *eeded.
erature limi of each area hown in Table 7-6 shall not be ABLT: Wh never the equ ment in an a ected area is quired to be OPER LE. With one o more areas ex eedlng the tem erature llmit(s shown in Table .7-6:
- a. less than 20 and for less than 8 hours, cord the cumula iye t e and the amou t by which the emperature in te affected area s ex eded the limit s).
- b. By le s than 20'F an for greater t n or equal to 8 ours, prepare and s bmit to the Commission w hin 30 days, ursuant to Specifi ation 6.9.2, a ecial Report hat provides a cord of the cumulati time and the amount by whi h the temperat re in the affected rea(s) exceed' d the limit( and an an sis to demonstrate the continued PERABILITY of he;affected eq ipment.
The provisio s of Specifica on 3.0.3 are n applicable. C. With one or mo areas exceedi the temperatu e limit(s) show in Table 3.7-6 by eater than or e ual to 20"F, p pare and submi a ecial Report a required by A ION b. above a d within 4 hour e ther restore th area(s) to wit n the tempera re limit(s) or de are the equipmet in the affect area(s) inope ble. 4.7.1 Th tempea re in each of areas shown Tbe376sale determidto be withth its limits:
- a. At least onc per seven days en the alarm is PERABLE, and;
- b. least once r 12 hours when he alarm is mnop able.
( Pn ~ Thr T6N~t~?.I9U.YL&ST 13LJ'9k-// MILLSTONE 1XII
- UNIT 3 3/4 7-32 Amendment No. J7, Y, ¢%
,a3,-,v ,3.9495 --
AREA TEMPER ft RE MONITORING AREA TEMPERATURE LIMIT ('FI
- 1. AUXILIARY BUILDING AB-02, VCT and Boric id Transfer Pump Area, El 43'6" *< 120 AB- ,Charging Pump Area, 1 24'6* < 110 AB-04, eneral Area, El 66'6" 120 AB-06, Ge ral Area, El 43'6" < 0 AB-07, Genera Area, El 4'6" < 120 AB-08, General A a (East), El 4'6" < 120 AB-09, General Area outh), El 4'6" < 120
-10, General Area, El '6" < 120 AB- , General Area, El 43'<12 AB-13, eneral Area (North), E 4'6" 2 AB-16, Su lemental Leak Collectio Filter Area, El 666'612 AB-19, MCC/Rod lye Area, El 24'5'<
AB-Ze, MCC Air Con tioning Room, El 66'6 < 120
-22, Rod Drive Area, 1 43'6U < 120 AB-2 Charging Pump Area, 1 24'6* < 110 AB-26, CCW Pump Area, El 24 < 110 AB-29, Gen al Area (Southeast), 246'6<12 AB-33, Boric A d Tank Area, El 43'6*
AB-35, Boric Acid ank Area, El 43'6* < 120 AB-39, Fuel Building d Auxiliary Building Filter Area, El '6N < 120 1k11zS f lV 3' C 1t N cTr &T +/- Wt9 /L L( L C
-rT ) 3L n l vl/
ý MILLSTONE - UNIT 3 3/4 7-33 8zerdment Ne./o s "rR11
AREA ILEPEARIUKl LIM]I t-ý)
- 2. CONTROL UILDING CB-01, Switchgea and Battery Rooms, E 4'60 04 C -02, Cable Spread g Room, El 24'6" < 11 CB-0 Control and Con ter Rooms, El 47'6" < 95 CB-04, hiller Room, El 6 '6" -< 104 CB-05, Me anical Equipment om, El 64-6" < 104
- 3. CONTAI ENT CS-01, Inside Cane Wall, El all e ept CS-03 and CS-04 < 120 CS-02, Outside Cra eWall, El all 120 CS-03, Pressurizer Cu ile, El all < 0 S-04, Inside Crane wall El 51'46 except -03 and steam <__120 generator en osur i
- 4. INTAKE STRUCTURE CW-01, ntire Building < 110
- 5. DIES GENERATOR BUIDING DG-A1, Entir Building < 120 S-01, HVAC and MC Area, El 36'6' <
ES 2, SIH Pump Area, 1 216'6 < 110 ES-03, Pipe Tunnel Area, 1 4'6' < 110 ES-04, S Cubicles, El all < 110 ES-05, RSS ubicles, El all <110 ES-06, Motor iven Auxiliary Fee ater Pump Area, El 24'-6n 110 ES-07, Turbine Dri en Auxiliary Feedwa r Pump < 1 Area, El 24' K MILLSTONE - UNIT 3 0261 3/4 7-34 f Amendment No.
- I
TABI\3.7-6 (Contin ýd AREA TEMP
- URE MONITORING AREA T PERATURE LIMIT (OF)
- 7. FUEL BUIL NG
-02, Fuel Poo Pump Cubici El 24'6" < 119 FB- 3, General Are ,El 52'4" < 08
- 8. OIL VAULT FV-O1, D sel Fuel Oil ult < 95
- 9. HYDROGE RECOMBINER BU DING HR-O1, Recombi er Skid Area, 24161 < 125 HR-02, Controls Aea, El 24'6" _ 110 HR-03, Sampling Are El 24'6" <1 R-04, HVAC Area, El 3 '6" < 110
- 10. MAIN STEAM VALVE BUI ING MS-0*, Areas above El. 58'0" < 140 MS-02, A as below El. 58'O" < 140
- 11. TURBI BUILDING TB-O1, Entire uilding <1 5 I TUNNEL T-0 Pipe Tunnel- uxiliary, Fuel and ESF Building < 112
- 13. YAR YD-O1, Yar < 5
. , ii h MILLSTONE - UNIT 3 3/4 7-35 Amendment No. Y7, 199, /1
REACTOR COOLANT SYSTEM August 27, 2001 BASES L SPECIFIC ACTIVITY (Continued) Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical time necessary perform the sampling, transport the sample, and perform the analysis of aboutto 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than I hour, about 2 hours, about I day, about I week, and about I month. Reducing Tav, to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time take corrective action. A reduction in frequency of isotopic analyses following to power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEPT THE PRESSURIZER) BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due-to system pressure and temperature changes. These loads are introduced startup ý-(heatup) and shutdown (cooldown) operations, power transients, by and reactor trips. This LCO limits the pressure-and temperature changes during RCS:* heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup, cooldown, I inservice leak and hydrostatic.(ISLH) testing, and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational requirements during heatup or. cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. A heatup or cooldown is defined as a temperature increase or decrease of greater than or equal to 10F in any one hour period. This definition heatup and cooldown is based upon the ASME definition of isothermal conditions of described in ASME, Section XI, Appendix E. MILLSTONE - UNIT 3 B 3/4 4-7 Amendment No. 0*7, 197
REACTOR COOLANT SYSTEM BASES t..L 644,j , -August27, 200f PRESSURE/TEMPERATURE LIMITS (continued) Steady state thermal conditions exist when temperature increases or decreases are <10F in any one hour period and when the plant is not performing a planned heatup or cooldown in accordance with a procedure. The LCO establishes operating limits that provide a margin to brittle failure, applicable to the ferritic material of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the Pressurizer, which has different design characteristics and operating functions which.are addressed by LCO 3.4.9.2, "Pressurizer". The P/T limits have been established for the ferritic materials of the RCS considering ASME Boiler and Pressure Vessel Code. Section XI, -Appendix G (Reference 1) as modified by ASME Code Case N-640 (Reference 2), and the additional requirements of 10 CFR 50 Appendix G (Reference 3). Implementation of the specific requirements provide adequate margin to brittle fracture of ferritic materials during normal operation, anticipated operational occurrences, and system leak and hydrostatic tests. The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases. The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 1 (Ref. 5). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6). The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions. of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations may be more restrictive, and thus; the curves are composites of the most restrictive regions. The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The P/T limits include uncertainty margins to ensure that the calculated limits are not inadvertently exceeded. These margins include gauge and system loop uncertainties, elevation differences, containment pressure conditions and system pressure drops between the beltline region of the vessel and the pressure gauge or relief valve location. MILLSTONE - UNIT 3 B 3/4 4-8 Amendment No. fg, X71, 197
RLACIUR COOLANI SYSTEM August 27, 2001 BASES N6 &Y-L PRESSURE/TEMPERATURE LIMITS (continued) The criticality limit curve includes the Reference I requirement that it be
> 40*F above the heatup curve or the cooldown curve, and not less than 160'F above the minimum permissible temperature for ISLH testing. This limit provides the required margin relative to brittle fracture. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4, "Minimum Temperature for Criticality."
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the ferritic RCPB materials, possibly leading to a nonisolable leak or loss of coolant accident. I In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code, Section XI, Appendix E (Ref. 7) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICABLE SAFETY ANALYSIS The P/T limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Reference 1, as modified by Reference 2, combined with the additional requirements of Reference 3 provide the methodology for determining the P/T limits. Although the P/T limits are not derived from any DBA, the P/T I limits are acceptance limits since they preclude operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of IOCFR50.36(c)(2)(ii). LCO The LCO limits apply to the ferritic components of the RCS, except the Pressurizer. These limits define allowable operating regions while providing I margin against nonductile failure for the controlling ferritic component. The limitations imposed on the rate of change of temperature have been established to ensure consistency with the resultant heatup, cooldown, and ISLH testing P/T limit curves. These limits control the thermal gradients (stresses) within the reactor vessel beltline (the limiting component). Note that:while these limits are to provide protection to ferritic components within the reactor coolant pressure boundary, a limit of 100°F/hr applies to the reactor coolant pressure boundary (except the pressurizer) to ensure that operation is maintained within the ASME Section III design loadings, stresses, and fatigue analyses for heatup and cooldown. MILLSTONE - UNIT 3 B 3/4 4-9 Amendment No. WF7, 197
REACTOR COOLANT SYSTEM August 27, 2001 BASES /.JIo k - y -v
, 1, 4_.
PRESSURE/TEMPERATURE LIMITS (continued) Violating the LCO limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure of ferritic RCS components using ASME Section XI, Appendix G, as modified by Code Case N-640 and the additional requirements of I IOCFR50, Appendix G (Ref. 1). The P/T limits were developed to provide requirements for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, in keeping with the concern for nonductile failure. The limits do not apply to the Pressurizer. During MODES I and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LCO 3.2.5, "DNB Parameters"; LCO 3.2.3.1 and 3.2.3.2, "RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor - Four Loops Operating/Three Loops Operating"; LCO 3.1.1.4, "Minimum Temperature:for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODES I and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent. ACTIONS Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Allowed Outage Times (AOTs) reflects the urgency of restoring the parameters to within,the I analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner. Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components. MILLSTONE - UNIT 3 B 3/4 4-I0- Amendment No. JX7, 197L
REACTOR COOLANT SYSTEM August 27, 2001 BASES H,.) b r-[vF ,o* >L PRESSURE/TEMPERATURE LIMITS (continued) ASME Code, Section XI, Appendix E (Ref. 7), may be used to support the I evaluation. However, its use is restricted to evaluation of the vessel beltline. The 72 hour AOT when operating in MODES 1 through 4 is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing to operate. This evaluation must be completed whenever a limit is exceeded. Restoration within the AOT alone is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. If the required remedial actions are not completed within the allowed times, the plant must be placed in a lower MODE or not allowed to enter MODE 4 because I either the RCS remained in an unacceptable P/T region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. In. reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased. If the required evaluation for continued operation in MODES I through 4 cannot I be accomplished within 72 hours or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in the Action statement. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions. Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours and to MODE 5 with RCS pressure < 500 psia within the next 30 hours. Completion of the required evaluation following limit violation in other than MODES I through 4 is required before plant startup to MODE 4 can proceed. The AOTs 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 REQUIREMENTS Verification that operation is within the LCO limits as well as the limits of Figures 3.4-2 and 3.4-3 is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This frequency- is'. considered reasonable in view of the control room indication available to monitor RCS status. MILLSTONE - UNIT 3 B 3/4 4-11 Amendment No. OY, $*, J$7, 197
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (continued) Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. This Surveillance Requirement is only required to be performed during system heatup, cooldown, and ISLH testing. No Surveillance Requirement is given for criticality operations because LCO 3.1.1.4 contains a more restrictive requirement. It is not necessary to perform Surveillance Requirement 4.4.9.1.1 to verify compliance with Figures 3.4-2 and 3.4-3 when the reactor vessel is fully detensioned. During refueling, with the head fully detensioned or off the reactor vessel, the RCS is not capable of being pressurized to any significant value. The limiting thermal stresses which could be encountered during this time would be limited to flood-up using RWST water as low as 40°F. It is not possible to cause crack growth of postulated flaws in the reactor vessel at normal refueling temperatures even injecting 40°F Water. examine tý;reactor and accordance vessel\Raterial ~2 i radiat~on surveillanc* SuY~eillance h1e to remo~ specimens Requ' ement is in ,th the requirem its of 10 FR50, Appendix H. REFERENCES
- 1. ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, "Fracture Toughness for Protection Against Failure," 1995 Edition.
- 2. ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curmes," dated February 26, 1999.
- 3. 10 CFR 50 Appendix G, "Fracture Toughness Requirements."
- 4. ASTM E 185-82, "Standard Practice for Conducting.Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706."
- 5. 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance!Program Requirements."
- 6. Regulatory Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor-Vessel Materials," dated May 1988.
- 7. ASME Boiler and Pressure Vessel Code, Section XI, Appendix E, "Evaluation of Unanticipated Operating Events," 1995 Edition.
PRESSURIZER BACKGROUND The Pressurizer is part of the RCPB, but is not subject to the same restrictions as the rest of the RCS. This LCO limits the temperature changes of the Pressurizer and allowable temperature differentials, within the design assumptions and the stress limits for cyclic operation. MILLSTONE - UNIT 3 B 3/4 4-12 Amendment No. 99, 07, Wi
PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES The OPERABILITY of the steam generator atmospheric relief bypass valve (SGARBV) lines provides a method to recover from a steam generator tube rupture (SGTR) event during which the operator is required to perform a limited cooldown to establish adequate subcooling as a necessary step to limit the primary to secondary break flow into the ruptured steam generator. The time required to limit the primary to secondary break flow for an SGTR event is more critical than the time required to cooldown to RHR entry conditions. Because of these time constraints, these valves and associated flow paths must be OPERABLE from the control room. The number of SGARBVs required to be OPERABLE from the control room to satisfy the SGTR accident analysis requires consideration of single failure criteria. Four SGARBV are required to be OPERABLE to ensure the credited steam release pathways available to conduct a unit cooldown following a SGTR. For other design events, the SGARBVs provide a safety grade method for cooling the unit to residual heat removal (RHR) entry conditions should the preferred heat sink via the steam bypass system or the steam generator atmospheric relief valves be unavailable. Prior to operator action to cooldown, the main steam safety valves (MSSVs) are assumed to operate automatically to relieve steam and maintain the steam generator pressure below design limits. Each SGARBV line consists of one SGARBV and an associated block valve (main steam atmospheric relief isolation valve, 3MSS*MOV]8A/B/C/D). These block valves are used in the event a steam generator atmospheric relief valve (SGARV) or SGARBV fails to close. Because of the electrical power relationship between the SGARBV and the block valves, if a block valve is maintained closed, the SGARBV flow path is inoperable because of single failure consideration. The bases for the required actions can be found in NUREG 1431, Rev. 1. The LCO APPLICABILITY and ACTION statements uses the terms "MODE 4 when steam generator is relied upon for heat removal" and "in MODE 4 without reliance upon steam generator for heat removal." This means that those steam generators which are credited for decay heat removal to comply with LCO 3.4.1.3 (Reactor Coolant System, Hot Shutdown) shall have an OPERABLE SGARBV line. See Bases Section 3/4.4.1 for more detail. 3/4.7.2 STEAI GENERA*TOR PR[..UR[i... P[.RTURE LiMl,,AON bEL& t The limitation on stea generator pressure nd temperature ens res that the press re-induced stresse in the steam genera ors do not exceed he maximum allowable racture toughness ress limits. The I {itations of 70'F nd 200 psig ar based on a steam nerator RTNDT of 60° and are sufficien to prevent brit le fracture. MILLSTONE - UNIT 3 8 3/4 7-7 Amendment No. ,Jp, ;jf, JO
7,L2 8 /- BES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section. Sarveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature. The 24-hour frequency is based on operating experience related to trending of the parameter variations during the applicable modes. This surveillance requirement verifies that the average water temperature of the UHS is less than or equal to 75"F. Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70OF during the applicable modes. The intent of this Surveillance Requirement is to Increase the awareness of plant personnel regarding UHS temperature trends above 70"F. The frequency is based on operating experience related to trending of the parameter variations during the applicable modes. 3/4.7.6 FL,99 PRtOTfCTIO? LbEG-r* The limitation on floo protection ensure that the service ater pump cubicle wa ertight doors will\be closed and the lmp cubicle sump d~ain valves(I* will be clo ed before the wate level reaches the ritical elevation of 14.5 I ) feet Mean Se Level. Elevation\4.5 feet MSL is th' floor elevation f the Ikj/ service water pump cubicle. 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity. Additionally, the system provides temperature control for the control room during normal and post-accident operations. The control room emergency ventilation system is comprised of the control room emergency air filtration system and a temperature control system. The control room emergency air filtration system consists of two redundant systems that recirculate and filter the control room air. Each control room emergency air filtration system consists of a moisture separator, electric heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system. Normal Operation A portion of the control room emergency ventilation system is required to operate during normal operations to ensure the temperature of the control room is maintained at or below 950F. MILLSTONE - UNIT 3 B 3/4 7-10 LAmendment No. 17J, 171, i%'
PLANT SYSTEMS BASES 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM The OPERABILITY of the Auxiliary Building Filter System ensures that radioactive materials leaking from the equipment within the charging pump, component cooling water pump and heat exchanger areas following a LOCA are filtered prior to reaching the environment. The charging pump/reactor plant component cooling water pump ventilation system must be operational to ensure operability of the auxiliary building filter system and the supplementary leak collection and release system. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. Laboratory testing of methyl iodide penetrationU shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage. 3/4.7.10 -S*$I)U 5 ELEt6 D All snubb s are require OPERABLE to en re that the stuctural integrity of he Reactor Cpolant System a d all other sa ty-related sys ems is main tain during and ollowing a se mic or other e nt initiating\dynamic loads. For t purpose of eclaring the fected system ERABLE with t inoperable snubbe s), an engin ring evaluatin may be perfor ed, in accorda ce with Section 0.59 of 10 C Part 50. Snubb s are classi fed and groupe by design and nufacturer bu not by ize. Snubb rs of the sam manufacturer ut having diffe ent internal chanisms ar classified a different typ s. For example, mechanical sn bers u iizing the me design fea ures of the 2 ip, lO-kip and 00-kip capaci man factured by ompany "A ar of the same pe. The same d ign mechanica snub ers manufac red by Compa 1B" for the p rposes of this chnical Speci ication woul be of a diffe ent type, as uld hydraulic s ubbers from either anufacturer. A li t of indivi al snubbers wi detailed in rmation of snub r location and size a d of system ffected shall available a the plant in acc rdance with Sectio 50.7FI(c) a 10 CFR Part 50. The accessibity of each snu er shall be det mined and a roved by the P nt Operation Review Committee The determinat ion hall be bas !ypon the existing radiation evels and the expected time tperform a visual inspection in each snubb location as we as other factors associated w~th accessibilit during plant perations (e.g., MILLSTONE - UNIT 3 B 3/4 7-23 Amendment No. 97, J71, 70,04
PLANT SYSTEMS A BASES 3/4.7.10 SNUBBERS (Continued) temperature, atmosphere, location(etc.), and the recommenda ions of Regulatory Guides 8.8 an 8.10. The addition\or deletion of any hydrauli or mechanical snubber shall e made in accordance ith Section 50.59 of 10 C Part 50. The visual spection frequency i based upon maintaining a nstant level of snubber otection to each sa ty-related system during a earthquake or severe transient Therefore, the req Jred inspection interval va ies inversely with the o served snubber failu s on a given system and is determined by the number of inop rable snubbers found uring an inspection of eac system. In order to establish he inspection frequen for each type of snubber n safety-related system, it was assumed that t frequency of snubber fai res a d initiating events is onstant with tim an that the failure of any sn bber on that system could cause the system to be unp tected and to result in fa lure durng an assumed initiatin event. Inspections erformed before that inter has lapsed may be used as a new reference point t determine thenext inspe tion. However, the resu ts of such early insp tions performed before the or inal required time int val has elapsed (nomi al time less 25%) may not be used o lengthen the require inspection interval. Any inspection whose results quire a shorter inspec *on interval will over ide the previous schedule. `J *The ac ptance criteria are to e used in the visual i spection to determine OP BILITY of the snubbers. For example, if a fl id port of a hydraulic snu er is found to be uncov ed, the snubber shall e declared L inoperable and hall not be determined ERABLE via functional esting. To provide a surance of snubber functional reliability, one f three functional testing methods is used with the stated acceptance crit ia:
- 1. Function lly test 10% of a type o snubber with an additi nal 5%
tested fo each functional testing ailure, or
- 2. Functionall test a sample size and d termine sample accepta e or rejection us g Figure 4.7-1, or
\ 3. Functionally t t a representative sampl size and determine sam e acceptance or r *ection using the stated euation.
Q ligure 4.7-1 was develo d using "Wald's Sequentia Probability Ratio V Plan" a described in "Quality ontrol and Industrial Sta Jstics" by Acheson Duncan. Per nent or other exemption from the surveillance pro ram for individual snubbers m be granted by the Comnssion if a justifiable bass for exemption is presente and, if applicable, snu ber life destructive testi was performed to qualify t snubbers for the appli able design conditions at ither the com pletion of th ir fabrication or at a s bsequent date. Snubbers s exempted MILLSTONE - UNIT 3 B 3/4 7-24 Amendment Nos. Xf*, 9*7, MX, 00 0509
PLANT SYSTEMS BASES 3/4.7.10 SNUBBERS (Continued) shal .be listed in the \ist of individual s ubbers indicating th extent of the r pce , pr n re l c d i h g r diton ar axnh g e p r t re ar , en~~s t e lifbe rofs nuberio ervica~ i l sa une rgolie p vafomanufc eva tue r uatio n et.. Th*r uieett moio tesnbe dve baesform sevc lfe s nlddto
'onthru ghconsideration o f snubber h service conitone.
3 . / . 4 . ..7. ..1 .1~~~ . . . . .. . . . .. e J[*LLU~ [ *I ~ 111*mLl nUE dL a al t h r bl i i y o a ito installtedioreotn re t eaithoe mn u a e wt R qi do hand which *r~eo not. . le ni aseh wit vet l ne y a e em v d fr m'yistalled t e s els e nubcess
,ec repacdiaio u l s th t d boion meadiationges are an nee spitring otbe te relcd ionhighdernedatobe atrea, ber se.n^ vIc t
. h/ re ln Tc*.Te
. ... ... . . .. t ... 0 . ... E r ic 1 The area teihperature limitation ensure that safety related equipm nt will not e subjected o temperatures in e ccess of their envir nmental qualii cation tempe atures. Expksure to excessive t peratures may degr de equipment a d can cause loss of its\OPERABILITY. The t nperature limits in lude an allowa ce for ins rument error\of +2.2"F.
MILLSTONE - UNIT 3 B 3/4 7-25 Amendment Nos. pp, Ff, lfl jJJ, ,q
Docket Nos. 50-336 50-423 B18556 Attachment 6 Millstone Nuclear Power Station, Unit No. 3 Technical Specifications Change Request 3-18-01 Relocation of Selected Technical Specifications Related to the Reactor Coolant System and Plant Systems Retyped Pages
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1aCi/gram DOSE EQUIVALENT 1-131 ........ .................. 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . ......... ................... 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ..... ............... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY ............... 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 10 EFPY . . . . . . . . . . . . . . . 3/4 4-35 TABLE 4.4-5 DELETED . . . . . . . . . . . . . . . . . . . . . . 3/4 4-36 Pressurizer ........ ... ...................... 3/4 4-37 Overpressure Protection Systems .... ............. 3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COL D OVERPRESSURE SYSTEM (FOUR LOOP OPERATION) ...... 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COL D OVERPRESSURE SYSTEM (THREE LOOP OPERATION) ......... 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY ................ 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ..... .............. 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............ ...................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350'F ........ 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK ..... .............. 3/4 5-9 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS .... ........ 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity . . . . . . . . . . . . . . . . 3/4 6-1 Containment Leakage . . . . . . . . . . . . . . . . . 3/4 6-2 Containment Air Locks . . . . . . . . . . . . . . . . 3/4 6-5 Containment Pressure ...... .................. 3/4 6-7 MILLSTONE - UNIT 3 Viii Amendment No. X, 97, 9, X19, 0870
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP ....... 3/4 7-3 Auxiliary Feedwater System ........... 3/4 7-4 Demineralized Water Storage Tank ........ 3/4 7-6 Specific Activity ............... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM .............. 3/4 7-8 Main Steam Line Isolation Valves ........ 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines 3/4 7-9a 3/4.7.2 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM . . 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM .............. 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ............... 3/4 7-13 3/4.7.6 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 3/4 7-15 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ........ 3/4 7-20 3/4.7.10 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-22 TABLE 4.7-2 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-27 FIGURE 4.7-1 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-29 3/4.7.11 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-30 3/4.7.12 DELETED Table 3.7-4 DELETED Table 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-32 TABLE 3.7-6 DELETED . . . . . . . . . . . . . . . . . . . . 3/4 7-33 MILLSTONE 0871
- UNIT 3 X Amendment No. fg, F4, lo, jf,
INDEX BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>lMeV) AS A FUNCTION OF FULL POWER SERVICE LIFE ...... ................. .. B 3/4 4-10 3/4.4.10 STRUCTURAL INTEGRITY ......... .................. .. B 3/4 4-15 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ..... .............. .. B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............ ...................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ........ ............... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK ......... .............. B 3/4 5-2 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ..... ......... B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ........ ................... .. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .... .......... .. B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES ..... .............. .. B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL ...... ................. .. B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM ..... ......... B 3/4 6-3b 3/4.6.6 SECONDARY CONTAINMENT ........ .................. .. B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...... .......... ................ .. B 3/4 7-1 3/4.7.2 DELETED ........... ......................... .. B 3/4 7-7 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM .... ...... B 3/4 7-7 3/4.7.4 SERVICE WATER SYSTEM ......... .................. .. B 3/4 7-7 3/4.7.5 ULTIMATE HEAT SINK ............ ................... B 3/4 7-8 3/4.7.6 DELETED ........... ......................... .. B 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ........... .. B 3/4 7-10 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM ........... B 3/4 7-17 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ..... ............ .. B 3/4 7-23 3/4.7.10 DELETED ........... ......................... .. B 3/4 7-23 MILLSTONE 0872
- UNIT 3 xiv Amendment No. #$, g, jjý, Ily, Jý9,
INDEX BASES SECTION PAGE 3/4.7.11 DELETED . . . . . . . . . . . . . . ... . . . . . . . . B 3/4 7-25 3/4.7.12 DELETED 3/4.7.13 DELETED 3/4.7.14 DELETED ....... ................ ... . . . . . . . . B 3/4 7-25 I 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION ..... B 3/4 8-1 3/4.8.4 DELETED ....... ................ B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ........ .................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION .......... ...................... B 3/4 9-1 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ..... ............ B 3/4 9-1 3/4.9.5 COMMUNICATIONS . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.6 REFUELING MACHINE .......... .................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS .... ......... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ......... B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ..... B 3/4 9-7 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-8 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM .... ............ B 3/4 9-8 3/4.9.13 SPENT FUEL POOL - REACTIVITY .............. B 3/4 9-8 3/4.9.14 SPENT FUEL POOL - STORAGE PATTERN ..... ............ B 3/4 9-9 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .......... ...................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS B 3/4 10-1 3/4.10.3 PHYSICS TESTS .............. ...................... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS ........ ................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ... ........... B 3/4 10-1 MILLSTONE - UNIT 3 xv Amendment No. F p, 7,9, 107, 179, 0872
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.1 Reactor Coolant System (except the pressurizer) temperature, pressure, and heatup and cooldown rates of ferritic materials shall be limited in accordance with the limits shown on Figures 3.4-2 and 3.4-3. In addition, a maximum of one reactor coolant pump can be in operation when the lowest unisolated Reactor Coolant System loop wide range cold leg temperature is < 160 0 F. APPLICABILITY: At all times. ACTION:
- a. With any of the above limits exceeded in MODES 1, 2, 3, or 4, perform the following:
- 1. Restore the temperature and/or pressure to within limit within 30 minutes.
AND
- 2. Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System remains acceptable for continued operation within 72 hours. Otherwise, be in at least MODE 3 within the next 6 hours and in MODE 5 with RCS pressure less than 500 psia within the following 30 hours.
- b. With any of the above limits exceeded in other than MODES 1, 2, 3, or 4, perform the following:
- 1. Immediately initiate action to restore the temperature and/or pressure to within limit.
AND
- 2. Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System is acceptable for continued operation prior to entering MODE 4.
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup and cooldown operations, and during the one-hour period prior to and during inservice leak and hydrostatic testing operations. 4.4.9.1.2 DELETED MILLSTONE 0873 UNIT 3 3/4 4-33 Amendment No. W*7, 197,
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- UNIT 3 3/4 4-36 Amendment No. 90, 107, 797,
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- UNIT 3 3/4 7-23 Amendment 9, 79, J9, X77, 797,
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- UNIT 3 3/4 7-32 Amendment No. R7, 9A, J0, 141,
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- UNIT 3 3/4 7-35 Amendment No. ?7, lo, 10,
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (continued) Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. This Surveillance Requirement is only required to be performed during system heatup, cooldown, and ISLH testing. No Surveillance Requirement is given for criticality operations because LCO 3.1.1.4 contains a more restrictive requirement. It is not necessary to perform Surveillance Requirement 4.4.9.1.1 to verify compliance with Figures 3.4-2 and 3.4-3 when the reactor vessel is fully detensioned. During refueling, with the head fully detensioned or off the reactor vessel, the RCS is not capable of being pressurized to any significant value. The limiting thermal stresses which could be encountered during this time would be limited to flood-up using RWST water as low as 40°F. It is not possible to cause crack growth of postulated flaws in the reactor vessel at normal refueling temperatures even injecting 40°F Water. REFERENCES
- 1. ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, "Fracture Toughness for Protection Against Failure," 1995 Edition.
- 2. ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," dated February 26, 1999.
- 3. 10 CFR 50 Appendix G, "Fracture Toughness Requirements."
- 4. ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706."
- 5. 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
- 6. Regulatory Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," dated May 1988.
- 7. ASME Boiler and Pressure Vessel Code, Section XI, Appendix E, "Evaluation of Unanticipated Operating Events," 1995 Edition.
PRESSURIZER BACKGROUND The Pressurizer is part of the RCPB, but is not subject to the same restrictions as the rest of the RCS. This LCO limits the temperature changes of the Pressurizer and allowable temperature differentials, within the design assumptions and the stress limits for cyclic operation. MILLSTONE 0878
- UNIT 3 B 3/4 4-12 Amendment No. 4, W*7, 197,
PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES The OPERABILITY of the steam generator atmospheric relief bypass valve (SGARBV) lines provides a method to recover from a steam generator tube rupture (SGTR) event during which the operator is required to perform a limited cooldown to establish adequate subcooling as a necessary step to limit the primary to secondary break flow into the ruptured steam generator. The time required to limit the primary to secondary break flow for an SGTR event is more critical than the time required to cooldown to RHR entry conditions. Because of these time constraints, these valves and associated flow paths must be OPERABLE from the control room. The number of SGARBVs required to be OPERABLE from the control room to satisfy the SGTR accident analysis requires consideration of single failure criteria. Four SGARBV are required to be OPERABLE to ensure the credited steam release pathways available to conduct a unit cooldown following a SGTR. For other design events, the SGARBVs provide a safety grade method for cooling the unit to residual heat removal (RHR) entry conditions should the preferred heat sink via the steam bypass system or the steam generator atmospheric relief valves be unavailable. Prior to operator action to cooldown, the main steam safety valves (MSSVs) are assumed to operate automatically to relieve steam and maintain the steam generator pressure below design limits. Each SGARBV line consists of one SGARBV and an associated block valve (main steam atmospheric relief isolation valve, 3MSS*MOV18A/B/C/D). These block valves are used in the event a steam generator atmospheric relief valve (SGARV) or SGARBV fails to close. Because of the electrical power relationship between the SGARBV and the block valves, if a block valve is maintained closed, the SGARBV flow path is inoperable because of single failure consideration. The bases for the required actions can be found in NUREG 1431, Rev. 1. The LCO APPLICABILITY and ACTION statements uses the terms "MODE 4 when steam generator is relied upon for heat removal" and "in MODE 4 without reliance upon steam generator for heat removal." This means that those steam generators which are credited for decay heat removal to comply with LCO 3.4.1.3 (Reactor Coolant System, Hot Shutdown) shall have an OPERABLE SGARBV line. See Bases Section 3/4.4.1 for more detail. 3/4.7.2 DELETED MILLSTONE - UNIT 3 B 3/4 7-7 Amendment No. 17f, If?, 101, 0879
PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section. Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature. The 24-hour frequency is based on operating experience related to trending of the parameter variations during the applicable modes. This surveillance requirement verifies that the average water temperature of the UHS is less than or equal to 75'F. Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70°F during the applicable modes. The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70'F. The frequency is based on operating experience related to trending of the parameter variations during the applicable modes. 3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity. Additionally, the system provides temperature control for the control room during normal and post-accident operations. The control room emergency ventilation system is comprised of the control room emergency air filtration system and a temperature control system. The control room emergency air filtration system consists of two redundant systems that recirculate and filter the control room air. Each control room emergency air filtration system consists of a moisture separator, electric heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system. Normal Operation A portion of the control room emergency ventilation system is required to operate during normal operations to ensure the temperature of the control room is maintained at or below 95°F. MILLSTONE 0880
- UNIT 3 B 3/4 7-10 Amendment No. 11g, 17f, jgf,
PLANT SYSTEMS BASES 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM The OPERABILITY of the Auxiliary Building Filter System ensures that radioactive materials leaking from the equipment within the charging pump, component cooling water pump and heat exchanger areas following a LOCA are filtered prior to reaching the environment. The charging pump/reactor plant component cooling water pump ventilation system must be operational to ensure operability of the auxiliary building filter system and the supplementary leak collection and release system. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage. 3/4.7.10 DELETED I MILLSTONE - UNIT 3 B 3/4 7-23 Amendment No. 97, 719, 119, J9, 0881
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- UNIT 3 B 3/4 7-24 Amendment Nos. If, 97, 119, 179,
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