ML101110179
| ML101110179 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/14/2010 |
| From: | Belcher S Constellation Energy Nuclear Group, Nine Mile Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME2253 | |
| Download: ML101110179 (13) | |
Text
Sam Belcher P.O. Box 63 Vice President-Nine Mile Point Lycoming, New York 13093 315.349.5200 315.349.1321 Fax CENG a joint venture of 0 Constellation nergy" OEnemgy- ,eDF NINE MILE POINT NUCLEAR STATION April 14, 2010 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 Pursuant to 10 CFR 50.90: Revisions to Primary Coolant System Pressure Isolation Valve Requirements Consistent with Standard Technical Specifications - Response to NRC Request for Additional Information (TAC No. ME2253)
REFERENCES:
(a) Letter from T. A. Lynch (NMPNS) to Document Control Desk (NRC), dated September 18, 2009, License Amendment Request Pursuant to 10 CFR 50.90:
Revisions to Primary Coolant System Pressure Isolation Valve Requirements Consistent with Standard Technical Specifications - Technical Specification Sections 3.2.7.1 and 4.2.7.1 (b) Letter from R. V. Guzman (NRC) to S. L. Belcher (NMPNS), dated March 18, 2010, Request for Additional Information Regarding Nine Mile Point Nuclear Station, Unit No. 1 - License Amendment Request Re: Revisions to Primary Coolant System Pressure Isolation Valve Requirements Consistent with Standard Technical Specifications (TAC No. ME2253)
Nine Mile Point Nuclear Station, LLC (NMPNS) hereby transmits supplemental information requested by the NRC in support of a previously submitted request for amendment to Nine Mile Point Unit 1 (NMP1)
Renewed Operating License DPR-63. The initial request, dated September 18, 2009 (Reference a) proposed to revise Technical Specification Sections 3.2.7.1 and 4.2.7.1, "Primary Coolant System
- Aoo(
Document Control Desk April 14, 2010 Page 2 Pressure Isolation Valves," to incorporate requirements that are consistent with Section 3.4.5 of the Improved Standard Technical Specifications, NUREG-1433, Revision 3.1. The supplemental information, provided in Attachment I to this letter and in Attachments 2 and 3 referenced therein, responds to the request for additional information (RAI) documented in the NRC's letter dated March 18, 2010 (Reference b).
In the Significant Hazards Consideration section of Reference (a), the description of the proposed amendment stated that the existing pressure isolation valve (PIV) leakage test frequencies were being replaced with a reference to the Inservice Testing (IST) Program. Based on the supplemental information provided in this letter, the amendment description is modified to state that the existing PIV leakage test frequencies are being replaced with a test frequency of "24 months." This is considered an administrative change since the 24-month frequency is based on the requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), which is also the basis for preparation of the previously-referenced IST Program. The NMPNS responses to the three standards set forth in 10 CFR 50.92 that were provided in the Significant Hazards Consideration section of Reference (a) are not otherwise affected.
Pursuant to 10 CFR 50.91(b)(1), NMPNS has provided a copy of this supplemental information to the appropriate state representative. This letter contains no new regulatory commitments.
Should you have any questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.
Very truly yours,
Document Control Desk April 14, 2010 Page 3 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 supplemental information 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.
$1 Subscribed and sworn before me, A Notary Public in and for the State of New York and County of 2,* ,this __ day of ,2010.
WITNESS my Hand and Notarial Seal:
Notary Public My Commission Expires:
91/(D Igo 1- LaM. Dtao Date NO" Aft h) Sat o fNewYork Oswego County Rg. No.D06029220 SB/DEV MY Commisson Expkue 9N11212013 Attachments: 1. Nine Mile Point Unit 1 - Response to NRC Request for Additional Information Regarding the Proposed Revisions to the Technical Specification Requirements for Primary Coolant System Pressure Isolation Valves
- 2. Revision to Proposed Technical Specification Changes (Mark-up)
- 3. Revised Changes to Technical Specification Bases cc: S. J. Collins, NRC R. V. Guzman, NRC Resident Inspector, NRC A. L. Peterson, NYSERDA
ATTACHMENT 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED REVISIONS TO THE TECHNICAL SPECIFICATION REQUIREMENTS FOR PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Nine Mile Point Nuclear Station, LLC April 14, 2010
ATTACHMENT 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED REVISIONS TO THE TECHNICAL SPECIFICATION REQUIREMENTS FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES By letter September 18, 2009, Nine Mile Point Nuclear Station, LLC (NMPNS) requested an amendment to the Nine Mile Point Unit 1 (NMPI) Renewed Facility Operating License DPR-63. The proposed amendment would revise Technical Specification (TS) Sections 3.2.7.1 and 4.2.7.1, "Primary Coolant System Pressure Isolation Valves," to incorporate requirements that are consistent with Section 3.4.5 of the Improved Standard Technical Specifications, NUREG-1433, Revision 3.1. This attachment provides supplemental information in response to the request for additional information documented in the NRC's letter dated March 18, 2010. Each individual NRC question is repeated (in italics), followed by the NMPNS response.
RAI-1
Provide the specific Surveillance Requirement (SR) frequency intervalsfor Technical Specification (TS) 3.2.7.1/4.2.7.1 as related to the generic terminology in the applicable American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and explain how the SR frequency intervals are incorporatedinto the TSsfor NMP1.
Response
NMPNS is revising the proposed change to TS Surveillance Requirement (SR) 4.2.7.1.a. Rather than referencing the Inservice Testing (IST) Program for the reactor coolant system pressure isolation valve (PIV) leakage test frequency, NMPNS proposes to specify the PIV leakage test frequency as "24 months."
The 24-month leakage test frequency is established in accordance with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 2004 Edition, which is the applicable code for the NMP1 fourth ten-year IST interval. In accordance with the OM Code, ISTC-1300, the PIVs are Category A valves since their seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function. For Category A valves other than containment isolation valves, the OM Code, ISTC-3630, specifies a test frequency of at least once every 2 years. Since the OM Code is also the basis for preparation of the previously referenced IST program, the proposed change to TS SR 4.2.7.1 .a is considered administrative in nature.
Specifying the leakage test frequency as 24 months rather than 2 years is consistent with the terminology used in the Improved Standard Technical Specifications, NUREG-1433. The P1V test frequency will actually appear in the NMP1 surveillance scheduling and tracking database as 730 days. The provisions of NMP1 TS SR 4.0.2, which allow a maximum surveillance interval extension of 25% of the specified interval, would be applicable to the PIV leakage test frequency. provides revised markups of existing TS pages to show the proposed changes. Only Insert 5 for TS Page 116 is affected. Revisions to the proposed TS Bases that reflect the proposed TS changes are provided in Attachment 3 for information only. Only new TS Bases Page 117b is affected. The revisions are indicated by a vertical bar drawn in the right hand margins.
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ATTACHMENT 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED REVISIONS TO THE TECHNICAL SPECIFICATION REQUIREMENTS FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES
RAI-2
TS 4.2.7.1. a is being changed to reference the Inservice Testing (IST) Programfor pressure isolation valve (PIV) test frequency. The fourth interval IST Program Plan (Agencywide Documents Access Management System Accession No. ML081900568) states the frequency is per the TSs. Explain how the frequency is determined using the IST Program.
Response
As described in the response to RAI-1 above, NMPNS is revising the proposed change to TS SR 4.2.7.1 .a to specify the PIV leakage test frequency as "24 months" rather than referencing the IST Program. The 24-month test frequency is based on the ASME OM Code frequency for Category A valves other than containment isolation valves. Following NRC approval of the license amendment request, appropriate changes to the IST Program documentation will be incorporated as part of normal license amendment implementation activities.
2 of 2
ATTACHMENT 2 REVISION TO PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
The current versions of the following NMP1 Technical Specification (TS) pages have been marked-up by hand to reflect the revised proposed changes:
116 117 Nine Mile Point Nuclear Station, LLC April 14, 2010
(LIMITING 4
CONDITION FOR OPERATION i SURVEILLANCE REQUIREMENT 3.2.7.1 N COOLANT SYSTEM PRESSURE ISOLATION 4.2.7.1 COOLANT SYSTEM PRESSURE ISOLATION Applicability:
Applies to the operating status of isolation valves for Applies to the periodic testing of coolant systems connected to the coolant system. system pressure isolation valves.
Obiective: Obiective:
To increase the reliability of coolant system To increase the reliability of: coolant system pressure isolation valves thereby reducing the pressure isolation valves thereby reducing the potential of an intersystem loss of coolant accident. potential of an intersystem loss of coolant accident.
Specification:, Soecificationr: +
- a. The integrity of pressure isolation valveoto Y shall be demonstrated. Valve leakage shall,605. tme R,-,d59 .
. If SpecificationRcannot be met, an orderly shutdown shall bkinitiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the reactor shall be in e cold shutdown condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. , ']
rSs;L
,a, To s sfy ALAR equireme s, leakage ay be meard .r i rectly (as fr m the per rmance of essure ind'ators) if accomplis d in acc ance with pproved pr edures and sup ted by c putations owing tha e method capab of demo rating val complianc with the le, ge criteri .
AMENDMENT NO. 4,42" 116 1
TABLE 3.2.7.1
- 1. Core Si
- 3. Shutdown _50.375 gpm
_<0.375 gpm
_50.375 gpm (a) 1. Leakage rates shall AMENDMENT NO. W17 , 4F!14 117
INSERT 1 (for TS Page 116; Specification 3.2.7.1)
------------ NOTES ----------------------
- 1. Separate specification entry is allowed for each flow path.
- 2. Enter applicable specifications for systems made inoperable by PIVs.
INSERT 2 (for TS Page 116; Specification 3.2.7.1 .a) be within limit during the power operating and hot shutdown reactor operating conditions.
INSERT 3 (for TS Page 116; New Specification 3.2.7.1 .b)
- b. If one or more flow paths with leakage from one or more PIVs is not within limit:
- 1. Isolate the high pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and
- 2. Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Each valve used to satisfy Specifications b.1 and b.2 above must have been verified to meet Specification 4.2.7.1 .a and be in the reactor coolant system boundary or the high pressure portion of the system.
INSERT 4 (for TS Page 116; Specification 4.2.7.1)
------------- NOTE ---------------------
Not required to be performed in the hot shutdown reactor operating condition.
INSERT 5 (for TS Page 116; Specification 4.2.7.1 .a)
The equivalent leakage of each reactor coolant system PIV shall be verified to be <- 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a reactor coolant system pressure > 1010 psig and
_<1050 psig, at a frequency in a.. dan.e
.. with the ffisr.... Testing Pr-e in of 24 months.
I
ATTACHMENT 3 REVISED CHANGES TO TECHNICAL SPECIFICATION BASES New proposed NMP1 Technical Specifications (TS) Bases (Pages 117a and 117b) that reflect the proposed TS changes are provided for information only.
Nine Mile Point Nuclear Station, LLC April 14, 2010
NEW TS BASES PAGE BASES FOR 3.2.7.1 AND 4.2.7.1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE (PIV) LEAKAGE The function of reactor coolant system (RCS) PIVs is to separate the high pressure RCS from an attached low pressure system. This protects RCS pressure boundary described in 10 CFR 50.2 and 10 CFR 50.55a(c) (Refs. 1 and 2). The PIVs, which are listed in the NMP1 UFSAR (Reference 3), are designed to meet the requirements of Reference 4. During their service lives, these valves can exhibit varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. Leakage through these valves is not included in any allowable leakage specified in Specification 3.2.5, "Reactor Coolant System Leakage."
The RCS PIV Specification allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety. Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed event that could degrade the ability for low pressure injection. In Reference 5, it-was concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.
This Specification applies in the power operating and hot shutdown reactor operating conditions because the PIV leakage potential is greatest when the RCS is pressurized. In the cold shutdown, refueling, and major maintenance reactor operating conditions, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.
Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these conditions.
Note 1 to Specification 3.2.7.1 has been provided that allows separate Condition entry for each affected RCS PIV flow path because the actions in Specification 3.2.7.1 .b provide appropriate compensatory measures for separate, affected RCS PIV flow paths. Note 2 to Specification 3.2.7.1 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems.
If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows time for these actions and restricts the time of operation with leaking valves.
Specification 3.2.7.1 .b.2 specifies that the double isolation barrier of two valves be restored by closing another valve qualified for isolation or restoring one leaking PIV. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit considers the time required to complete the action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the connected low pressure piping when overpressurized to reactor pressure (Ref. 6).
Revision 1 17a
NEW TS BASES PAGE BASES FOR 3.2.7.1 AND 4.2.7.1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE (PIV) LEAKAGE Valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB or the high pressure portion of the system.
If leakage cannot be reduced or the system isolated, the plant must be brought to an operating condition in which the Specification does not apply. To achieve this status, an orderly shutdown must be initiated within one hour and the plant be brought to the cold shutdown condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment.
Performance of leakage testing on each reactor coolant system PIV is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve size up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.
Reference 4 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential). The observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one-half power.
The 24-month Frequency required by the Inservice Testing Program is withiR based on the ASME OM Code Frequency. Specification 4.2.7.1 is modified by a Note that states the leakage Surveillance is not required to be performed in the hot shutdown condition. Entry into this condition is permitted for leakage testing at high differential pressures with stable conditions that are not possible in the cold shutdown or refueling conditions.
References:
- 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 4. ASME Code for Operation and Maintenance of Nuclear Power Plants.
- 5. Letter from T. A. Ippolito (NRC) to D. P. Dise (NMPC) dated April 20, 1981, "Order for Modification of License Concerning Primary Coolant System Pressure isolation Valves," included attached Technical Evaluation Report TER-C5257-237, Rev. 1, dated March 20, 1981.
- 6. NEDC-31339, "BWR Owners Group assessment of Emergency Core Cooling System Pressurization in Boiling Water Reactors,"
November 1986.
Revision 117b