ML022070197

License Amendment, Revise Containment Systems Technical Specifications

ML022070197
Person / Time
Site:	Monticello
Issue date:	09/23/2002
From:	Miranda S NRC/NRR/DLPM/LPD3
To:	Forbes J Nuclear Management Co
References
TAC MB3706
Download: ML022070197 (14)

Text

September 23, 2002 Mr. Jeffrey S. Forbes Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT RELATING TO CONTAINMENT SYSTEMS TECHNICAL SPECIFICATION REVISIONS (TAC NO. MB3706)

Dear Mr. Forbes:

The Commission has issued the enclosed Amendment No. 130 to Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 21, 2001, as supplemented April 26, 2002.

The amendment revises TS Sections 3.7/4.7, "Containment Systems," to (1) clarify existing requirements, (2) make editorial changes, (3) revise limiting conditions for operation (LCOs) and surveillance requirements, and (4) add certain LCOs.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Samuel Miranda, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 130 to DPR-22

2. Safety Evaluation cc w/encls: See next page

ML022070197 *Provided SE input by memo OFFICE PDIII-1/PM PDIII-1/LA RORP/SC* SLPB/SC OGC** PDIII-1/SC NAME SMiranda RBouling RDennig SWeerakkody RWeisman LRaghavan DATE 08/16/02 08/16/02 07/18/02 09/11/02 09/20/02 09/23/02 Monticello Nuclear Generating Plant cc:

J. E. Silberg, Esquire Commissioner Shaw, Pittman, Potts and Trowbridge Minnesota Department of Commerce 2300 N Street, N. W. 121 Seventh Place East Washington, DC 20037 Suite 200 St. Paul, MN 55101-2145 U.S. Nuclear Regulatory Commission Resident Inspectors Office Adonis A. Neblett 2807 W. County Road 75 Assistant Attorney General Monticello, MN 55362 Office of the Attorney General 445 Minnesota Street Site Licensing Manager Suite 900 Monticello Nuclear Generating Plant St. Paul, MN 55101-2127 Nuclear Management Company, LLC 2807 West County Road 75 Mr. Roy A. Anderson Monticello, MN 55362-9637 Executive Vice President and Chief Nuclear Officer Robert Nelson, President Nuclear Management Company, LLC Minnesota Environmental Control 700 First Street Citizens Association (MECCA) Hudson, WI 54016 1051 South McKnight Road St. Paul, MN 55119 Nuclear Asset Manager Xcel Energy, Inc.

Commissioner 414 Nicollet Mall Minnesota Pollution Control Agency Minneapolis, MN 55401 520 Lafayette Road St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.

Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 March 2002

NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. DPR-22

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated December 21, 2001, as supplemented April 26, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 130 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 23, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 130 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 157 157 163 163 165 165 166 166 170 170 171 171 6 171a 177 177 180 180 181 181 182 182 6 182a 189 189 190 190

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 130 TO FACILITY OPERATING LICENSE NO. DPR-22 NUCLEAR MANAGEMENT COMPANY LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated December 21, 2001, as supplemented April 26, 2002, the Nuclear Management Company, LLC (the licensee), requested changes to the Technical Specifications (TSs) for the Monticello Nuclear Generating Plant. The proposed changes would modify TS Sections 3.7/4.7, "Containment Systems," to (1) clarify existing requirements, (2) make editorial changes, (3) revise limiting conditions for operation (LCOs) and surveillance requirements (SRs), and (4) add certain LCOs.

The April 26, 2002, supplemental letter provided additional clarifying information that was within the scope of the original notice of the amendment request published in the Federal Register (67 FR 34490, May 14, 2002), and did not change the Nuclear Regulatory Commission (NRC) staffs initial proposed no significant hazards consideration determination.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act requires that applicants for nuclear power plant operating licenses will state:

Such technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization...of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and the mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TSs those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity, as set forth in Statement of Consideration, Technical Specifications for Facility Licenses; Safety Analysis Reports (33 FR 18610, December 17, 1968). Pursuant to 10 CFR 50.36,

TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plants TSs.

On July 22, 1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36 (58 FR 39132). Further, the Final Policy Statement gave guidance for evaluating the required scope of the TSs and defined four guidance criteria to be used in determining which of the LCOs and associated SRs should remain in the TSs. Existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TSs; those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria set out in the Federal Policy Statement in 10 CFR 50.36 (60 FR 36953, July 19, 1995). The four criteria of 10 CFR 50.36(c)(2)(ii) are as follows:

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.

Criterion 2 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.

Criterion 3 A structure, system, or component that is part of the primary success path and 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.

Criterion 4 A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

3.0 EVALUATION In its December 21, 2001, application and April 26, 2002, supplemental letter, the licensee proposed changes to the Monticello current TSs (CTS) for containment systems in three specific areas: (1) suppression pool volume and temperature, (2) containment atmosphere control, and (3) containment isolation.

3.1 Revision to Suppression Pool Volume and Temperature (CTS 3/4.7.A.1)

The licensee proposes to delete the CTS LCO and SR (3/4.7.A.1.f) for the instrumentation that monitors suppression pool water level. This change would eliminate an unnecessary 6-hour shutdown LCO to calibrate this instrumentation. In addition, CTS 3.7.A.1.e would be revised to allow for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to return the suppression pool water level to within the limits specified when the suppression pool water level is outside of its required limits. The licensee would also revise the TS Bases consistent with the changes described above.

The proposed deletion of CTS 3.7.A.1.f is based on the justification that it does not meet the requirements of 10 CFR 50.36(c)(2)(ii) for inclusion in TSs. The instrumentation of CTS 3.7.A.1.f does not meet the criteria established in 10 CFR 50.36(c)(2)(ii) in that this instrumentation is not used for detection or to provide an indication in the control room of a significant abnormal degradation of the reactor coolant pressure boundary. Additionally, this instrumentation is not part of the primary success path and does not provide a safety function or actuation 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; nor has this instrumentation been shown, either through operating experience or probabilistic risk assessment, to be significant to public health and safety. Specifying suppression pool water level meets Criterion 2 of 10 CFR 50.36(c)(2)(ii), is satisfied by CTS 3.7.A.1.e, and is not proposed to be changed. In addition, the proposed deletion of CTS 4.7.A.1.f is based on the reason that without the LCO which it supports, this SR no longer satisfies the requirements of 10 CFR 50.36(c)(3) for inclusion in TSs.

The proposed revision to the wording of CTS 3.7.A.1.e to allow up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore water level is based on the premise that the suppression pool water level does not change rapidly during normal operation. During operations that do change the suppression pool water level, the level of the pool is monitored. Because of the large volume of water in the suppression pool, the pool level changes very slowly. The 2-hour completion time should be sufficient to restore suppression pool water level to within limits. If the water level cannot be returned to within its established band within the 2-hour period, then the reactor will have to be placed in a condition in which this LCO does not apply, which means placing the reactor in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and suspending all activities with the potential for draining the reactor vessel proposed TS 3.7.A.1.f). See Section 3.2 below for evaluation of proposed TS 3.7.A.1.f.

The NRC staff has reviewed the changes associated with CTS 3/4.7.A.1 and finds them acceptable, based upon the above, and consistent with NUREG-1433. Also, the NRC staff has no objection to the licensees proposed TS Bases changes.

3.2 Clarification of Action Statements in CTS Section 3.7.A and Revise Containment Atmosphere Control (CTS 3.7.A.5)

The licensee has proposed the following changes to CTS 3.7.A.1, 3.7.A.3, 3.7.A.4, and 3.7.A.5:

1) Add proposed TS 3.7.A.1.f, which states that if the requirements of CTS 3.7.A.1 cannot be met, place the reactor in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and suspend all activities with the potential for draining the reactor vessel.

2) Add proposed TS 3.7.A.3.c, which states that if the requirements of CTS 3.7.A.3 cannot be met, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3) Add proposed TS 3.7.A.4.f, which states that if the requirements of CTS 3.7.A.4 cannot be met, the reactor shall be placed in a Hot Shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

4) Renumber CTS 3.7.A.6 as proposed TS 3.7.A.5.d, and reword it to state that if the requirements of CTS 3.7.A.5 cannot be met, reduce Thermal Power to # 15-percent rated thermal power (RTP), within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, revise CTS 3.7.A.5.b to reflect when the containment shall be inerted and deinerted to conform to the new action statement, and change the title from Containment Atmosphere Control to Primary Containment Oxygen Concentration.

The licensee has revised the CTS Bases consistent with the changes described above.

Currently, CTS 3.7.A.2.a, 3.7.A.2.b and 3.7.A.2.c contain specific action statements which govern the time allowed before shutdown if LCOs are not met and specifies the time to shutdown. CTS 3.7.A.6 is an action statement which currently applies to all requirements of CTS 3.7.A and requires that the reactor be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the requirements of CTS 3.7.A are not met. Therefore, CTS 3.7.A.2.a, 3.7.A.2.b and 3.7.A.2.c, together with 3.7.A.6, establish two separate and different action statements. In addition, using CTS 3.7.A.6 would allow certain systems required to be operable to be inoperable in Hot Shutdown when containment integrity is required. This overlap in TS applicability results in conflict and confusion on the part of operators as to which action statements are applicable. The proposed changes would resolve the overlap by providing specific action statements for CTS 3.7.A.1, 3.7.A.3, and 3.7.A.4 consistent with the LCO requirements specified for these sections in CTS 3.7.A.

The balance of this change would revise the wording of the CTS 3.7.A.6 action statement and renumbers it as proposed TS 3.7.A.5.d, which will make it specific to CTS 3.7.A.5. In addition, CTS 3.7.A.5.b would be revised to specify when containment shall be inerted and deinerted and change the title of CTS 3.7.A.5 to reflect the substance of this LCO. Inerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in plant startup and deinerted as soon as possible in plant shutdown. As long as reactor power is < 15-percent RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inerted. Furthermore, the probability of an event that generates hydrogen occurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a startup or the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before a shutdown is low enough to justify these windows, during which the primary containment is not inerted. The 24-hour time period is a reasonable amount of time to allow plant personnel to perform the containment atmosphere inerting or deinerting.

If oxygen concentration is $ 4.0 percent by volume at any time while operating, with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must be restored to < 4.0 percent by volume within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24-hour completion time is allowed when oxygen concentration is $ 4.0 percent by volume because of the low probability and long duration of an event that would generate significant amounts of hydrogen occurring during this period. If oxygen concentration cannot be restored to within limits within the required

completion time, the plant must be brought to a mode in which the LCO does not apply. To achieve this status, power must be reduced to # 15-percent RTP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Based on operating experience, the 8-hour completion time is reasonable to reduce reactor power from full power conditions in an orderly manner and without challenging plant systems.

The proposed revisions to these CTS sections will result in each subsection of 3.7.A having its own action statement to govern time allowed before shutdown if LCOs are not met. This will eliminate confusion, the potential for errors, and the current inconsistencies between CTS LCOs. This change is acceptable because each CTS LCO will have its specific action requirement when that particular LCO cannot be met. These changes do not involve equipment modifications or program changes. The changes provide specific action statements which provide for allowed time to place the reactor in a condition in which the LCO is no longer applicable.

The renaming, restructuring and rewording of TS 3/4.7.A.5 is acceptable because it will enhance the CTS by adding an additional LCO which places a specified time limit on oxygen concentration greater than or equal to 4 percent by volume and by making the TS requirements more easily understood, which will help to prevent errors and provide more detailed instructions.

The NRC staff has reviewed the changes associated with CTS 3.7.A.1, 3.7.A.3, 3.7.A.4, and 3.7.A.5 and finds them acceptable, based upon the above, and consistent with the NUREG-1433. Also, the NRC staff does not object to the licensees proposed changes to the TS Bases.

3.3 Revisions to Primary Containment Automatic Isolation Valves (CTS 3/4.7.D)

The licensee proposes to rename CTS 3/4.7.D from Primary Containment Automatic Isolation Valves to Primary Containment Isolation Valves (PCIVs). The licensee also proposes to revise CTS 3.7.D.2 for inoperable PCIVs such that 1) in the event there is one or more penetration flow paths with one inoperable PCIV, reactor operation in the run mode may continue provided that either the inoperable valve(s) is restored to operable status, or at least one valve in each line having an inoperable valve is deactivated in the isolated condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (except allowing 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for MSIVs and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for excess flow check valves); and 2) in the event there is one or more penetration flow paths with two inoperable PCIVs, reactor operation in the run mode may continue provided that either the inoperable valve(s) is restored to operable status, or at least one valve in each line having inoperable valves is deactivated in the isolated condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In both cases, the term deactivated is defined as being electrically or pneumatically disarmed or otherwise secured and in which valves and blind flanges will satisfy the requirements for each type of penetration. Also a footnote applicable to both cases, as well as proposed TS 4.7.D.2, would be added to state that valves closed to satisfy the requirements of this specification may be reopened on an intermittent basis under approved administrative controls. These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.

CTS 4.7.D.2 would be revised to change the interval at which deactivated and isolated valves used to meet the requirements of proposed TS 3.7.D.2 must be recorded from daily to monthly for valves outside of primary containment. For valves inside primary containment, their position shall be recorded prior to entering Hot Shutdown or Startup from the Cold Shutdown condition, if the containment was deinerted during the Cold Shutdown, and if the surveillance has not been performed in the previous 92 days. Also, a footnote would be added to CTS 4.7.D.2 and proposed TS 4.7.D.3 to allow isolation devices inside high radiation areas to be verified closed by use of administrative means.

The proposed change to this section would also move the LCO requirements of CTS 3.7.A.5.c to proposed TS 3.7.D.3 and reword this LCO for clarification. This rewording would state that inerting and deinerting operations permitted by TS 3.7.A.5.b shall be via the 18-inch purge and vent valves (equipped with 40-degree limit stops). All other purging and venting, when primary containment integrity is required, shall be via the 2-inch purge and vent bypass line and the standby gas treatment system. In addition a new action and associated surveillance with regard to purge valve leakage is proposed. The action (proposed TS 3.7.D.3.b) requires that with one or more penetration flow paths with one or more containment purge and vent valves not within purge and vent valve leakage limits, reactor operation in the run mode may continue provided that within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve(s) are restored to operable status, or at least one valve in each line having a purge and vent valve not within leakage limits is deactivated in the isolated position. This requirement may be satisfied by use of one closed and deactivated automatic valve, a closed manual valve, or a blind flange. Proposed TS 4.7.D.3 would require that whenever containment purge and vent valves are isolated to meet the requirements of proposed TS 3.7.D.3.b, the position of the deactivated and isolated valves or isolation devices outside primary containment shall be recorded monthly.

The CTS Bases would be revised, consistent with the changes described above.

Revising the title of CTS Section 3/4.7.D is acceptable because it more closely reflects the substance of the LCOs and SRs within this section of the TSs.

The revision to the action statements for inoperable PCIVs not only expands the scope of the LCO but also provides clarification of what is specifically required for this method of isolation, which is stated to be the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and deactivated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured, use of which is limited by the type of penetration and the number of inoperable valves per penetration. Additionally, the clarification states how to deactivate an inoperable PCIV, which is to electrically or pneumatically disarm or otherwise secure the valve, and this is acceptable based upon engineering judgment, and the importance of primary containment operability during times when primary containment integrity is required.

Changing the interval at which the position of valves deactivated and isolated to meet the requirements of proposed TS 3.7.D.2 must be recorded from daily to monthly, and adding the purge vent valve surveillance (proposed TS 4.7.D.3) to verify valve isolation for meeting the requirements of proposed TS 3.7.D.3.b is acceptable because the devices are operated under administrative controls, and the probability of this misalignment is low. These surveillances (proposed TS 4.7.D.2 and 4.7.D.3) do not require any testing or device manipulation. Rather, they involve verification that those devices located outside primary containment and capable of

being mispositioned are in the correct position. For the devices inside primary containment, the time period specified is prior to entering Startup or Hot Shutdown from Cold Shutdown, if primary containment was de-inerted during Cold Shutdown, and if the surveillance has not been performed in the previous 92 days. This period of time is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is not likely. Additionally, these SRs would be modified by a footnote that applies to devices located in high radiation areas which allows them to be verified by use of administrative means. This note is acceptable because access to these areas is typically restricted. Therefore, the probability of misalignment of these devices, once they have been checked and verified to be in the proper position, is low.

Relocating the requirements of CTS 3.7.A.5.c to proposed TS 3.7.D.3.a places the requirements for limiting the use of the purge and vent valves into a more appropriate TS.

Since the 18-inch purge and vent valves are PCIVs, it is more appropriate for the LCO limiting their use to be located in the PCIV section of TSs, rather than in the Primary Containment Oxygen Concentration section. Rewording this TS for clarification is also acceptable because it reduces confusion on the part of the operators as to when these valves can be used for purging and venting. The existing wording of this TS could be misinterpreted such that purging and venting through the 18-inch purge and vent valves is not allowed when the mode switch is in refuel. This clarification also restricts the use of these valves, in that they may not be used when primary containment is required, except as allowed by TS 3.7.A.5.b.

If the purge valve leakage rate is not within the limit, the assumptions of the relevant safety analyses would not be justified. Therefore, either the leakage must be restored to within limits or the penetration(s) that caused the limit to be exceeded must be isolated by use of one closed and deactivated automatic valve, a closed manual valve, or a blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, then the leakage rate is assumed to be the lesser of the actual pathway leakage of the two devices. The 24-hour completion time for purge valve leakage is acceptable considering the purge valves remain closed so that a gross breach of the containment does not exist.

The proposed changes described above will enhance the LCO and SR wording to provide clearer, more easily understood descriptions of conditions and requirements. This change also fulfills a statement made in LER 2000-010, Revision 1, dated October 13, 2000, promising to submit a TS revision stating that an automatic containment isolation valve that is deactivated in the closed position is to be considered operable. These changes do not involve equipment modifications or program changes, and therefore do not adversely affect the public health and safety.

The NRC staff has reviewed the proposed changes to CTS 3/4.7.D and finds them acceptable, based upon the above, and consistent with NUREG-1433. Also, the NRC staff has no objection to the licensees proposed changes to the TS Bases.

The NRC staff concludes that the proposed changes to the containment-related CTS, based upon the STS and plant-specific considerations, are consistent with the Monticello current licensing basis and the requirements and guidance of the Final Policy Statement and 10 CFR 50.36. Accordingly, the NRC staff finds these changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (67 FR 34490). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: R. Giardina Date: September 23, 2002