ML091950415

License Amendment Request to Remove Position Indication for Relief Valves and Safety Valves from Technical Specifications

ML091950415
Person / Time
Site:	Nine Mile Point
Issue date:	07/02/2009
From:	Belcher S Constellation Energy Group, Nine Mile Point
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML091950415 (14)

Text

Sam Belcher P.O. Box 63 Vice President-Nine Mile Point Lycoming, New York 13093 315.349.5200 315.349.1321 Fax 0 Constellation Energy-Nine Mile Point Nuclear Station, LLC July 2, 2009 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 License Amendment Request to Remove Position Indication for Relief Valves and Safety Valves from Technical Specifications In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR 50.90), Nine Mile Point Nuclear Station, LLC (NMPNS) is submitting a request for an amendment to the renewed Facility Operating License DPR-63 for Nine Mile Point Unit 1 (NMP 1).

The proposed amendment would remove position indication for the relief valves and safety valves from NMPI Technical Specification (TS) 3.6.11, Accident Monitoring Instrumentation. The justification for this proposed amendment is that position indication for the relief valves and safety valves do not meet any of the criteria for inclusion in TS provided in the NRC Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. In addition, the proposed amendment would correct an editorial error in the title of Table 4.6.11. An evaluation of the proposed changes is enclosed.

NMPNS requests approval of the proposed license amendment by July 2, 2010, with implementation within 60 days of NRC approval.

In accordance with 10 CFR 50.91, a copy of this application, with enclosure, is being provided to the appropriate state official.

If you should have any questions regarding this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.

Very truly yours,

Document Control Desk July 2, 2009 Page 2 STATE OF NEW YORK TO WIT:

COUNTY OF OSWEGO I, Sam Belcher, being duly sworn, state that I am Vice President-Nine Mile Point, and that I am duly authorized to execute and file this request on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Nine Mile Point employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and syorn before me, a N.otary Public in and for the State of New York and County of Oswego, this :9 day of .ToI, 2009.

WITNESS my Hand and Notarial Seal: 67yX ,/Q, 6ea*.

Notary Public My Commission Expires:

NoMy P inthe State of New York Date 00 =18O Re9. No. 01 CL6029220 MY WExres,q/ log SLB/GNS

Enclosure:

Evaluation of the Proposed Change cc: S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC A. L. Peterson, NYSERDA

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent

.4.3 Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

ATTACHMENT Proposed Technical Specification Change for Nine Mile Point Unit I (Mark-Up)

Nine Mile Point Nuclear Station, LLC July 2, 2009

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the renewed Facility Operating License DPR-63 for Nine Mile Point Unit I (NMP 1).

The proposed amendment would remove position indication for the relief valves and safety valves from NMP ITechnical Specification (TS) 3.6.11, Accident Monitoring Instrumentation.

The proposed change will provide greater flexibility in scheduling maintenance and repairs of this instrumentation commensurate with their importance to safety under the Nine Mile Point Nuclear Station, LLC (NMPNS) 10 CFR 50.65, Maintenance Rule Program.

2.0 DETAILED DESCRIPTION The changes requested by this amendment request are:

I. Delete TS 3.6.11, Table 3.6.11-1, Accident Monitoring Instrumentation, Item 1), Relief Valve Position Indication.

2. Delete TS 3.6.11, Table 3.6.11-1, Accident Monitoring Instrumentation, Item 2), Safety Valve Position Indication.

3. Delete TS 3.6.11, Table 3.6.11-2, Accident Monitoring Instrumentation Action Statements, Action 1.

4. In the title of TS 3.6.11, Table 4.6.11, change the word "Requirement" to "Requirements."

5. Delete TS 3.6.t1, Table 4.6.11, Accident Monitoring Instrumentation Surveillance Requirement, Item (1), Relief valve position indicator (Primary - Acoustic), Relief valve position indicator (Backup - Thermocouple).

6. Delete TS 3.6.11, Table 4.6.11, Accident Monitoring Instrumentation Surveillance Requirement, Item (2), Safety valve position indicator (Primary - Acoustic), Safety valve position indicator (Backup - Thermocouple).

There are no TS Bases section changes associated with this proposed amendment.

The continued functionality of the position indicator instrumentation will be maintained by the existing surveillance and preventative maintenance procedures. Changes to these procedures are subject to the controls of 10 CFR 50.59. In addition, the current classification of the position indicators for the safety valves and relief valves as Regulatory Guide (RG) 1.97 Type D, Category 3 variables will continue to be reflected in the NMP1 Updated Final Safety Analysis Report (UFSAR).

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

3.0 TECHNICAL EVALUATION

The safety valve/relief valve position indicators were added to the NMP1 TS by Amendment 42, on April 13, 1981, in response to an NRC letter dated July 2, 1980, to all boiling water reactor licensees, related to TMI-1 Lessons Learned Category "A" items.

RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Rev. 2, identifies position indication for safety/relief valves (SRV), including automatic depressurization system (ADS) or flow through or pressure in valve lines (SRV position), as a key variable for providing detection of an accident and for boundary integrity indication for the main steam system as a Type D, Category 2 variable.

At NMP1, the equivalent valves are referred to as safety valves (SV), and electromagnetic relief valves (RV). For consistency with the terminology used in other industry documents referenced in this evaluation, these valves will be referred to collectively as SRVs. The SRVs are comprised of 9 SVs connected to the reactor vessel head and 6 RVs, 3 on each of the two main steam lines, upstream of the main steam isolation valves. The SRVs are part of the primary success path in the UFSAR accident analysis because they can actuate to mitigate a Design Bases Accident (DBA) and therefore meet Criterion 3 of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, (Reference I). Accordingly, their operability is required by TS 3.2.8, Pressure Relief Systems-Safety Valves, and TS 3.2.9, Pressure Relief Systems-Solenoid-Actuated Pressure Relief Valves (Overpressurization). Redundant position indication is provided for each valve consisting of RG 1.97 Type D, Category 3 instrumentation (acoustic monitors and tailpipe thermocouples). Currently, SRV position indication operability is required by TS 3.6.11, Accident Monitoring Instrumentation, Table 3.6.11-1. The associated surveillance requirements are located in Table 4.6.11.

At NMPI, SRV position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary (Criterion 1). This is consistent with the Commission's Final Policy Statement which indicates that the first criterion was intended to assure that TS controlled those instruments specifically installed to detect reactor coolant leakage, but not to include instrumentation to identify the source of actual leakage (e.g., valve position indication). SRV position indication is not a process variable, design feature or operating restriction that is an initial condition of a DBA or transient analysis (Criterion 2). While the function of SRVs is part of the primary success path and the SRVs actuate to mitigate a DBA or transient, position indication for the SRVs does not form a part of the primary success path since the UFSAR accident analysis assumes the SRVs function as designed. That is, the accident analysis assumes no operator action based on SRV valve position for the SRVs to perform their primary success path function (Criterion 3). Finally, failure of SRV position indication would not pose a significant challenge to the ability of the operating crew to respond to a DBA transient, since the Emergency Operating Procedures (EOP) provide symptom-based instruction to the crew in mitigating an upset condition of the plant (i.e., level, pressure, and temperature provide EOP direction regardless of SRV status). The loss of this instrumentation has no effect on the probabilistic safety assessment and has not been shown to be significant to public health and safety (Criterion 4).

Consequently, SRV position indication does not meet any of the screening criteria of the Final Policy Statement for inclusion in TS and can be removed from the NMPI TS.

Additional information supporting this conclusion is as follows:

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE Although NMP1 has not adopted Improved TS, it is noted that NUREG-1433, Standard Technical Specifications - General Electric Plants (BWR/4), Rev. 3, does not list SRV Position Indication in Specification 3.3.3.1. "Post-Accident Monitoring Instrumentation."

The reviewers note in TS Table 3.3.3.1-1 of NUREG-1433 states:

Table 3.3.3.1-1 shall be amendedfor each plant as necessary to list.

1. All Regulatory Guide 1.97, Type A instruments and

2. All Regulatory Guide 1.97, Category 1, non-type A instruments specified in the plant's Regulatory Guide 1.97, Safety Evaluation Report NEDO-33160-A (Reference 2) states that for BWRs, reactor pressure vessel (RPV) pressure and suppression pool temperature instrumentation in combination with other instruments (e.g., RPV water level, suppression pool level, containment pressure) satisfy the accident detection and boundary integrity indication purpose as specified in RG 1.97 for the SRV position variable. This alternate instrumentation either meets or exceeds the Category 2 criteria. SRV position indication instrumentation provides backup information and does not need to be classified as a Category 2 variable. Therefore, NEDO-33160-A recommends that, for BWRs, SRV position indication be reclassified as a Type D Category 3 variable.

NEDO-33160-A states that the knowledge of SRV position is not used by the operator to make appropriate decisions in using individual systems important to safety in mitigating the consequences of an accident. SRVs are used as part of the ADS whose indication of successful performance of the safety function is the reduction of RPV pressure to enable use of low pressure Emergency Core Cooling Systems (ECCS) to mitigate the consequences of an accident. Other operator indications and requirements for ADS operation include an increase in suppression pool water temperature, increase in suppression pool water level, and change in RPV water level before the ADS will function. These other indications are included as RG 1.97 Category 1 or 2 variables, but the primary RG 1.97 variables are RPV pressure and suppression pool water temperature. The operator would use SRV position indication as a confirmation of SRV opening; however, this would be a backup to the main safety function for ADS of reduction in RPV pressure to enable use of low pressure ECCS.

BWRs are specifically designed to depressurize the RPV during certain accident scenarios.

Depressurization allows the initiation of low pressure ECCS loops to provide core cooling. Low pressure ECCS initiation and operation is independent of the causes for RPV depressurization (e.g.,

line break or ADS actuation). ECCS initiation and operation are primarily based on RPV water level, RPV dome pressure, containment pressure, and suppression pool temperature. Knowledge of SRV position or its indication does not significantly affect automatic or manual low pressure ECCS operation and thus does not affect short-term or long-term core cooling.

SRV position indication instrumentation provides direct indication of flow through a main steam SRV. This can occur as a result of: (1) planned manual operation of the SRV under normal operating conditions, (2) planned manual opening of the SRV under shutdown conditions (i.e.,

pressure control), (3) primary reactor system pressure exceeding the SRV relief function setpoint during power operation, (4) the effect of a system isolation and reactor shutdown, (5) a spurious actuation of the valve while at power (i.e., stuck open SRV), or (6) manual or automatic operation of the ADS function.

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE Of the above six scenarios, only the usage of SRV position information, as it relates to operation of the ADS function, is associated with an accident within the context of the guidance of RG 1.97, Rev. 2. The function of ADS is to provide the capability to reduce reactor pressure to allow low pressure ECCS to function to provide core cooling.

While SRV position instrumentation provides positive indication of flow through the SRVs, this information is of secondary importance to the operator during ADS operation. Successful indication of ADS actuation is provided by RPV pressure and suppression pool temperature. Flow indication through SRVs does not provide the operator with a unique indication of ADS actuation, a positive indication that an accident has occurred, or essential confirmation that an accident mitigation has occurred.

In summary, based upon the preceding analysis, NMP 1 SRV position indication does not meet any of the screening criteria of the Final Policy Statement for inclusion in TS and may be removed from the NMPI TS.

The proposed change of the word "Requirement" to "Requirements" in the title of Table 4.6.11 is editorial and therefore has no impact on safety.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The primary purpose of Accident Monitoring Instrumentation is to display plant variables that provide information required by the control room operator during accident situations.

This information provides the necessary support for the operator to take manual actions to initiate safety systems and other appropriate systems important to safety.

By letter dated October 31, 1980, the NRC set forth its requirements regarding approved TMI Action Plan items. The October 31, 1980, letter incorporated into one document all TMI-related items approved for implementation by the Commission. This document was published as NUREG-0737, "Clarification of TMI Action Plan Requirements." After NRC staff review of licensees' responses to NUREG-0737, the NRC issued Confirmatory Orders for the implementation of the TMI Action Plan items. NUREG-0737, Item II.D.3 requires, "Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe."

RG 1.97, Rev. 2 describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation for monitoring plant variables and systems during and after an accident. RG 1.97 groups these variables into five types, Types A, B, C, D, and E. Type D variables provide the operator with information on the operation of individual safety systems and other systems important to safety. Type D variables are to help the operator make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident. RG 1.97, Rev. 2, recommends that Type D variables that provide backup information should meet the Category 3 criteria. The 4 of 7

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE Category 3 criterion includes high-quality commercial grade equipment that is designed to withstand the specified service environment.

The NRC provided guidance for the contents of TS in its "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, (Reference 1). In particular, the NRC indicated that certain items could be relocated from TS to owner controlled documents, and identified criteria to be used to determine the functions to be included in the TS. The NRC adopted revisions to 10 CFR 50.36, "Technical Specifications," pursuant to which the rule was revised to codify and incorporate these criteria.

Although NMP1 has not adopted Improved TS, it is noted that NUREG-1433, Standard Technical Specifications - General Electric Plants (BWR/4), Rev. 3, does not list SRV Position Indication in specification 3.3.3.1. "Post-Accident Monitoring Instrumentation."

4.2 Precedent Nine Mile Point Unit 2 - Safety/relief valve position indicators were removed from TS as part of Amendment 69. (Reference 3)

Limerick Generating Station, Units 1 and 2 - Safety/relief valve position indicators were removed from TS as part of Unit 1 Amendment 179, and Unit 2 Amendment 141. (Reference 4) 4.3 Significant Hazards Consideration Nine Mile Point Nuclear Station, LLC (NMPNS) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The failure of the safety/relief valve (SRV) position instrumentation is not assumed to be an initiator or any analyzed event in the Updated Final Safety Analysis Report (UFSAR). The proposed changes do not alter the physical design of the SRVs or any other plant structure, system, or component. The changes would remove the SRV position indicator and surveillance requirements from the NMP1 TS, but would not involve any physical changes to the instrumentation.

The proposed change of the word "Requirement" to "Requirements" in the title of Table 4.6.11 is editorial and therefore has no impact on accident probability or consequence.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

2. Does the proposed amendment create the possibility of a new or different kind accident for any accident previously evaluated?

Response: No. The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions, associated with the operations of the plant. Accordingly the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

The proposed change of the word "Requirement" to "Requirements" in the title of Table 4.6.11 is editorial and therefore does not create the possibility of a new or different kind accident.

Therefore, the proposed amendment does not create the possibility of a new or different kind accident for any accident previously evaluated

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No. This instrumentation is not needed for manual operator actions necessary for safety systems to accomplish their safety function for the design basis accident events.

The instrumentation provides only alarm and SRV position indication, and does not provide an input to any automatic trip function. Several diverse means are available to monitor SRV position.

The proposed change of the word "Requirement" to "Requirements" in the title of Table 4.6.11 is editorial and therefore has no impact margin of safety.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, NMPNS concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operations in the proposed manner, (2) such activities will be conducted in compliance with the Commission's 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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 6 of 7

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. NRC "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," 58 FR 39132, July 22, 1993.

2. Letter from R. C. Bunt (BWROG) to Nuclear Regulatory Commission (NRC), dated October 13, 2006, Final BWR Owners Group Licensing Topical Report NEDO-33160-A, Revision 1, Regulatory Relaxation For The Post Accident SRV Position Indication System Dated October 2006.

3. Letter from G. E. Edison (NRC) to B. R. Sylvia (NMPC), dated September 11, 1995, Issuance of Amendment for Nine Mile Point Nuclear Station, Unit 2 (TAC No. M91295).

4. Letter from T. L. Tate (NRC) to C. M. Crane (Exelon), dated September 27, 2005, Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re: Relocation of Operability and Surveillance Requirements for the Safety/Relief Valve Position Instrumentation (TAC Nos.

MC3454 and MC3455).

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ATTACHMENT PROPOSED TECHNICAL SPECIFICATION CHANGE FOR NINE MILE POINT UNIT 1 (MARK-UP)

Technical Specification pages included in this markup:

269 270 272 Nine Mile Point Nuclear Station, LLC July 2, 2009

TABLE 3.6.11-1 ACCIDENT MONITORING INSTRUMENTATION Minimum Number of Operable Total Number of Channels Sensors or Channels Action (See Table 3.6.11-2) bdJ- dParameters

1) A lleIe V~' lVe Pc tclri

,- ll cii 1,c*(*

A 2/Valve NaiV~ve*

2) PSafety ~Valve rositiom Iridieati.

3) Reactor Vessel Water Level 2 1* 2

4) Drywell Pressure Monitor 2 1 4

5) Suppression Chamber Water 2 I* 4 Level 6) 7)

Deleted Containment High Range 2 1 3 4

Radiation Monitor

8) Suppression Chamber Water 2 I 2 Temperature A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance provided at least one Operable channel is monitoring that Parameter.

AMENDMENT NO.-ýj 269

TABLE 3.6.11-2 ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS ACTION - 1 t ACTION - 2

a. With the number of OPERABLE accident monitoring instrumentation channels less than the total Number of Channels shown in Table 3.6.1 1-1, restore the inoperable channel(s) to OPERABLE status within seven days or be in at least HOT SHUTDOWN within the next 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

With the number of OPERABLE accident monitoring instrumentation channels less than the minimum Channels OPERABLE requirements of Table 3.6.11-1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION - 3

a. With the number of OPERABLE channels less than the total Number of Channels shown in Table 3.6.11-1, prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b. With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable channel(s) to OPERABLE status within seven days of the event, or

2) prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

AMENDMENT NO.-44-2--27 270

TABLE 4.6.11 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT S Parameter Instrument Channel Test Instrument Channel Calibration DdeA ead (1) A Relief vav positiamchceeat (rrirnary Once per Eluarter Once dur~ing each majGr refucling outagc-Aeaustie)

Relief valve posit;ion iniao (Backt - Oncee per quarter Once durin~g each major rofuoling outage (2) A 1-1tva,-'c pesitian kndieator (Prfimary Once per qularter Onc- du1FRin each mnajor refueling outage Aeebistie~

Safety valve pesitian indieater (Beeltup Oncee per quarter Oncc during each majorF refueling outage Thqemfieeebipe)

(3) Reactor vessel water level Once per quarter Once during each major refueling outage (4) Drywell Pressure Monitor Once per month Once during each major refueling outage (5) Suppression Chamber Water Level Monitor Once per quarter Once during each major refueling outage (6) Deleted (7) Containment High Range Radiation Monitor Once per month Once during each major refueling outage (8) Suppression Chamber Water Temperature Once per month Once during each major refueling outage AMENDMENT NO. 442t-,2 272