NL-13-007, Supplement to Proposed Change to Containment Purge System and Pressure Relief Line Isolation Instrumentation Technical Specifications

Supplement to Proposed Change to Containment Purge System and Pressure Relief Line Isolation Instrumentation Technical Specifications

ML13037A309
Person / Time
Site:	Indian Point
Issue date:	01/24/2013
From:	Ventosa J Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-007
Download: ML13037A309 (9)

Text

Enterqy Nuclear Northeast Indian Point Energy Center

'ý'Ente-rg 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 John A. Ventosa Site Vice President Administration January 24, 2013 NL-13-007 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Supplement to Proposed Change to Containment Purge System and Pressure Relief Line Isolation Instrumentation Technical Specifications Indian Point Nuclear Generating Unit 2 Docket No. 50-247 License No. DPR-26

REFERENCE:

Entergy Letter (NL-12-002) to NRC regarding Proposed Change to Containment Purge System and Pressure Relief Line Isolation Instrumentation Technical Specifications, dated January 11, 2012

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) submitted, Reference 1, a request for an amendment to the Facility Operating License Number DPR-26 for Indian Point Nuclear Generating Unit 2 (IP2). The proposed amendment would revise Technical Specification (TS) Table 3.3.6-1 by replacing the "Allowable Value" "< 3 background" with "Trip Setpoint" "(a) As specified in the IP2 Offsite Dose Calculation Manual."

Following recent discussions with the NRC, Entergy again sought to find and review the multiple submittals for the conversion to Standard Technical Specifications that had not been found due to difficulties in finding specific documentation in the public document room. The intent was to find the specific change from "Trip Setpoint" to "Allowable Value." Entergy was able to locate the documentation and determined that the change was intentional and not an unintentional change as had originally been indicated in the referenced submittal. The purpose of the intentional change is unclear and the proposed revision to the Standard Technical Specification is still considered administrative. A revision to the evaluation of the proposed changes in accordance with 10 CFR 50.91 (a)(1) using criteria of 10 CFR 50.92 (c) was made to reflect this new information and is attached. Changes are identified by bars in the margin. The conclusion does not change that the proposed changes involve no significant hazards considerations. There is no change to the proposed TS changes or the intended changes to the Bases in the referenced submittal. A copy of this application and the associated attachments are being submitted to the designated New York State official in accordance with 1 OCFR 50.91 (a)(1).

NL-13-007 Dockets 50-247 Page 2 of 2 There are no new commitments identified in this submittal. If you have any questions or require additional information, please contact Mr. Robert Walpole, IPEC Licensing Manager at (914) 254-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 2A 2012.

Sincerely, JV/sp Attachments: Analysis of Proposed Technical Specification Change regarding Containment Purge System and Pressure Relief Line Isolation Instrumentation cc: Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORL Mr. William M. Dean, Regional Administrator, NRC Region 1 NRC Resident Inspectors Mr. Francis J. Murray, Jr., President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service

ATTACHMENT TO NL-13-007 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING CONTAINMENT PURGE SYSTEM AND PRESSURE RELIEF LINE ISOLATION INSTRUMENTATION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247

NL-13-007 Dockets 50-247 Attachment 1 Page 1 of 6

1.0 DESCRIPTION

This is a request to amend Operating License DPR-26, Docket 247, for Indian Point Unit 2 (IP2).

The proposed amendment would change the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and revise the current setpoint used for the Containment Purge Systems and Pressure Relief Line isolation. The proposed revision to Table 3.3.6-1 table will change "< 3 background" to allow the allowable value to be as specified in the Offsite Dose Calculation Manual (ODCM) (Reference 1).

2.0 PROPOSED CHANGE

Proposed changes to Section 3.3.6, Table 3.3.6-1 of the Indian Point 2 Technical Specifications are identified below:

Indian Point 2 Technical Specification 3.3.6, Table 3.3.6-1 currently says:

REQUIRED SURVEILLANCE FUNCTION CHANNELS REQUIREMENTS ALLOWABLE VALUE 2.a Gaseous Containment Radiation (R-42) SR 3.3.6.1 <3 x background SR 3.3.6.4 SR 3.3.6.6 2.b Particulate Containment Radiation (R-41) SR 3.3.6.1 < 3 x background SR 3.3.6.4 SR 3.3.6.6 The proposed amendment will revise Table 3.3.6-1 to say:

REQUIRED SURVEILLANCE FUNCTION CHANNELS REQUIREMENTS TRIP SETPOINT 2.a Gaseous Containment Radiation (R-42) 1 SR 3.3.6.1 (a)

SR 3.3.6.4 SR 3.3.6.6 2.b Particulate Containment Radiation (R-41) 1 SR 3.3.6.1 (a)

SR 3.3.6.4 SR 3.3.6.6 (a) As specified in the Offsite Dose Calculation Manual.

No changes to the Bases Section associated with the proposed change are needed.

3.0 BACKGROUND

Both the containment purge supply and exhaust isolation valves (FCV-1 170, FCV-1 171, FCV-1172, and FCV-1 173) and the containment pressure relief line isolation valves (PCV-1 190, PCV-1191 and PCV-1 192) close when high radiation levels are detected by the Containment Air

NL-13-007 Dockets 50-247 Attachment 1 Page 2 of 6 Particulate Monitor (R-41) or Containment Radioactive Gas Monitor (R-42). The Containment Phase A Isolation ESFAS signal and Containment Spray ESFAS signal also cause closure of the containment purge isolation valves and the containment pressure relief isolation valves.

The use of the allowable value / setpoint of "< 3 x background" is not the best specification for the stated purpose. The value is too low creating the potential for unnecessary isolation and requiring adjustment of the setpoint on a more frequent basis as the background changes during the operational cycle.

The Standard Technical Specification (STS) 3.3.6 have the same stated purpose as IP2 and the Setpoint is "< 2 x background." This change is therefore a deviation from the STS but it has been adopted at a number of plants where the problems have been identified. For example:

1. Wolf Creek has the trip setpoint for the containment atmosphere gaseous radioactive monitor set point established such that the actual submersion rate would not exceed 9mR/hr in the Containment.

2. Ginna has the trip setpoint for the containment gaseous and particulate radiation monitors set per the Radiological Effluent Controls Program.

3. Byron / Braidwood has the trip setpoint for the containment radiation - high established such that the actual submersion rate is < 10 mR/hr but allows the trip setpoint to be increased above this value in accordance with the methodology established in the Offsite Dose Calculation Manual.

4. Indian Point 3 has the trip setpoint for the containment gaseous and particulate radiation monitors set per the Offsite Dose Calculation Manual.

5. Diablo Canyon has the trip setpoint for the containment gaseous and particulate radiation monitors set per the Offsite Dose Calculation Manual.

The proposed change from "ALLOWABLE VALUE" to "TRIP SETPOINT" is based on a deviation from the standard technical specification that occurred during the conversion to standard technical specifications (STS). This corrects the TS for consistency with the STS.

4.0 TECHNICAL ANALYSIS

The purpose of this proposed amendment is to restore the term "TRIP SETPOINT" as the heading on the last column of Table 3.3.6-1 and to change the value from "< 3 x background" to reference the Offsite Dose Calculation Manual (ODCM) methodology.

The Standard Technical Specification (STS) requires radiation monitors with a SETPOINT of "< 2 x background" for Containment Purge System and Pressure Relief Line isolation following an accident. When IP2 converted to the STS no change to the term "TRIP SETPOINT" in the heading of Table 3.3.6-1 was proposed in the original submittal (Reference 3). In supplement 3 to the license amendment request (Reference 4), the change was made to revise "TRIP SETPOINT" to "ALLOWABLE VALUE" This change was explained in more restrictive change M.2:

NL-13-007 Dockets 50-247 Attachment 1 Page 3 of 6 "CTS Table 3.5-4, Item 4a, Containment Radioactivity High (R-41/R-42) does not specify an allowable value or trip setpoint for Gaseous Containment Radiation Monitor (R-42) or Particulate Containment Radiation Monitor (R-41). ITS LCO 3.3.6, Table 3.3.6-1, will include an allowable value for these functions which is based on engineering judgment. This is acceptable because ESFAS signals (LCO 3.3.2, Function 3.a) and Containment Spray ESFAS signal (LCO 3.3.2, Function 2) provide the analyzed functions for isolation of the containment purge supply line, containment purge exhaust line and the containment pressure relief line in Modes 1, 2, 3 and 4. Isolation on a high radiation signal is a backup that is not directly credited in the accident analysis... "Section I1.E.4.2.7, Containment Isolation Dependability, of NUREG-0737, "Clarification of TMI Action Plan Requirements," requires that containment purge and vent isolation valves must close on a high radiation signal. This is satisfied by ITS 3.3.6, Functions 2.a (R-42) and 2.b (R-41). However, no specific criteria is established for the allowable value. Indian Point 2 is currently in the process of developing an allowable value of this function using a methodology consistent with Part I of ISA-$67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation" and Regulatory Guide 1.105, "Setpoints for Safety-related Instrumentation." When completed, Indian Point 2 will submit to the NRC a copy of site specific methodology and the results. Upon NRC approval, the allowable values will be incorporated into the Improved Technical Specification submittal."

The Regulatory Guide (RG) guidance identifies an intended purpose that is not applicable to the current situation. The RG notes that "Section 4.3 of ISA-$67.04-1994 states that the limiting safety system setting (LSSS) may be the trip setpoint, an allowable value, or both. For the standard technical specifications, the staff designated the allowable value as the LSSS. In association with the trip setpoint and limiting conditions for operation (LCOs), the LSSS establishes the threshold for protective system action to prevent acceptable limits being exceeded during design basis accidents." In this case, there is no LSSS since, as discussed earlier, the isolation on high radiation is not depended upon to prevent acceptable limits from being exceeded. Also, the allowable value and consequently the trip setpoint is based on engineering judgment and not an analytical limit. Consequently, the change to go from "TRIP SETPOINT" in the STS to "ALLOWABLE VALUE" is not warranted.

For purposes of this amendment the proposed change to restore the STS terminology is regarded as administrative. The use of "trip setpoint" will be consistent with the STS and the change will not affect any acceptable limits from being achieved. Making this administrative change has no bearing on accident precursor conditions or events and therefore cannot lead to an increase in the probability of an accident or create the probability of a new or different type of accident than previously evaluated. Making this administrative change cannot lead to an increase in the consequences of an accident previously evaluated or affect the margin to safety.

The TS 3.3.6 Bases notes "Containment purge system and pressure relief line isolation instrumentation closes the containment isolation valves in the Containment Purge System (containment purge supply line and containment purge exhaust line) and the Containment Pressure Relief Line. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident." The Bases further note that the "safety analyses assume that the containment remains intact with penetrations unnecessary for core cooling isolated early in the event, within approximately 60 seconds." The proposed change to reference a setpoint set by the ODCM methodology will result in a higher setpoint than the "< 3 x background" but will still meet the safety function of the instrumentation. The source term from an accident event (Reference 2) is greater than the setpoints the ODCM setpoint will use.

NL-13-007 Dockets 50-247 Attachment 1 Page 4 of 6 Section 3.2.1 of the ODCM (Reference 1) requires the dose rates for radioactive materials released in gaseous effluents to meet specific limits. When these limits are exceeded the release must be stopped immediately. The release rates are those that would give an annual dose of < 500 mrem whole body or 3 rem skin due to noble gases or that would give an annual dose of 1.5 rem to any organ due to all radionuclide in particulate form with half lives > 8 days. The release rates apply to the site and are conservatively established, as discussed in the ODCM, in order to assure compliance with 10 CFR 20 limits. The current setpoint of "< 3 x background" achieves this purpose but it can also result in unnecessary isolations unless the setpoint is adjusted as the background changes. Changing the setpoint to match the ODCM methodology would eliminate this potential while providing adequate assurance that the results of an accident would not increase because these release rates are below those of the accident analyses.

The use of the ODCM to set the release rate eliminates the need to readjust the trip setpoint as the background levels change inside containment. Further, the risk of spurious trips is eliminated. The margins associated with accident analyses are slightly reduced because a higher level of radioactivity is required to trip the isolation. However, the effect is minimal since the fuel handling accident did not credit isolation, the gas tank ruptures are in the auxiliary building, the SG tube rupture is outside containment, the steam line break is also outside containment and gives the highest dose, and the LOCA source term will go from < 3 x background to the new setpoint in such a short time there is no practical difference.

Changing the setpoint does not affect the manner in which the plant is operated or controlled so there is no possibility of a different type of accident.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) has evaluated the safety significance of the proposed change of Containment Purge and Pressure Relief Line Systems isolation TS. The evaluation was for the change of the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and the proposed revision to the setpoint from "< 3 background" to allow the allowable value to be as specified in the Offsite Dose Calculation Manual (ODCM) according to the criteria of 10 CFR 50.92, "Issuance of Amendment". Entergy has concluded that the subject changes do not involve a Significant Hazards Consideration as discussed below:

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

Response: No. The proposed change will revise the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and change the setpoint requirements from "< 3 background" to allow the setpoint to be as specified in the Offsite Dose Calculation Manual (ODCM). The change to trip setpoint is a correction of an administrative error and will only affect the instrument setting specified. Therefore it does not involve the initiation of anaccident or the consequences. The values for the instrument setting are provided for isolating the Containment Purge and Pressure Relief Systems due to increased source terms and are redundant to containment isolation signals. They

NL-13-007 Dockets 50-247 Attachment 1 Page 5 of 6 have no effect on the probability of an accident previously evaluated. The change in the setting will be negligible for purposes of an accident termination. The ODCM limits are based on 1 0CFR20 limits which are substantially below accident analysis release rates. Therefore the change has a minimum effect on the consequences of such accidents. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of previously evaluated accidents.

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

Response: No. The proposed change will revise the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and change the setpoint requirements. The changes do not affect the system operations, plant operating procedures or affect how the plant is operated. The change does not create the possibility of any equipment failure or effect on other equipment. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No. The proposed change will revise the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and change the setpoint requirements. The change to trip setpoint is correcting an administrative error and has no significant affect on the margin of safety. The proposed change involves changes to existing setpoints for automatic isolation of the Containment Purge and Pressure Relief Systems.

However, the ability of the systems to isolate remains within current evaluations and therefore does not significantly reduce the safety margin. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Entergy Nuclear Operations, Inc. concludes that the proposed amendment presents no 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.

5.2 Applicable Regiulatory Requirements / Criteria The proposed changes have been evaluated to determine whether applicable requirements continue to be met. The Containment Purge and Pressure Relief Line Systems are isolated by radiation monitors as backup to the ESF system to ensure Containment Isolation. This insures meeting the Containment leakage assumption of accident analyses demonstrating compliance with 10 CFR 100. This instrumentation will continue to perform this function and will satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) 5.3 Environmental Considerations The proposed changes to the IP2 Technical Specifications regarding the change of the term "ALLOWABLE VALUE" to "TRIP SETPOINT" and the proposed revision to the setpoint from "< 3 background" to be as specified in the ODCM do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent 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

NL-13-007 Dockets 50-247 Attachment 1 Page 6 of 6 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 PRECEDENCE The proposed change is consistent with the requirements approved at Wolf Creek (50-482), Ginna (50-244), Byron / Braidwood (50-454, 455, 456, and 457), Indian Point 3 (50-286), and Diablo Canyon (50-275).

7.0 REFERENCES

1. Indian Point letter NL-10-045, "2009 Annual Radioactive Effluent Release Report,"

dated April 29, 2010 transmitted Units 1, 2 and 3 Offsite Dose Calculation Manual (ODCM), Revision 2..

2. Indian Point 2 UFSAR, Chapter 14, Tables 14.2-2 (fuel handling accident), 14.2-4 (MSLB), 14.2-5 (VCT), 14.3-43 (LOCA).

3. Entergy Letter to NRC, NL-02-016, "License Amendment Request (LAR 02-005)

Conversion to Improved Technical Specifications," March 22, 2002

4. Entergy Letter to NRC, NL-03-081, "Supplement 3 to the License Amendment Request for Conversion to Improved Technical Specifications," May 19, 2003