NL-13-003, Proposed Technical Specification Changes Regarding RWST Temperature and Containment Pressure in Containment Integrity Analysis

Proposed Technical Specification Changes Regarding RWST Temperature and Containment Pressure in Containment Integrity Analysis

ML13042A224
Person / Time
Site:	Indian Point  (DPR-064)
Issue date:	01/28/2013
From:	Ventosa J Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-003
Download: ML13042A224 (27)

Text

Enterqy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 John A Ventosa Site Vice President Administration NL-13-003 January 28, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Proposed Technical Specification Changes Regarding RWST Temperature and Containment Pressure in Containment Integrity Analysis Indian Point Unit Number 3 Docket No. 50-286 License No. DPR-64

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc, (Entergy) hereby requests a change to the Technical Specifications for Indian Point Nuclear Generating Unit No. 3 (IP3). The proposed change will revise Technical Specification surveillance requirement (SR) 3.5.4.1, to limit the maximum Refueling Water Storage Tank (RWST) temperature to *1 050F; Technical Specification 3.6.4 Limiting Condition for Operation to limit containment pressure to *++1.5 psig if RWST temperature is > 950F or containment temperature is >125 F; Technical Specification SR 3.6.3.9 is being deleted and Technical Specification 5.5.15 Containment Leakage Rate Testing Program is being changed to specify the re-analysis value of peak containment pressure. A re-analysis of the Large Break Loss of Coolant Accident with an RWST initial temperature of 1050F and containment initial pressure of 1.5 psig was performed to address mass and energy release errors for containment integrity identified in Nuclear Safety Advisory Letter 11-5. Sensitivity studies establish the bases for the proposed changes. The current analysis of record uses an RWST initial temperature of 1 10OF and containment initial pressure of 2.5 psig. Since the current Technical Specifications have been determined to be non-conservative, administrative controls will be in place (when RWST temperature exceeds 950F or containment temperature exceeds 1250F) to operate the plant consistent with the proposed Technical Specifications.

Entergy has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1) using the criteria of 10 CFR 50.92(c) and determined that this proposed change involves no significant hazards, as described in Attachment 1. The marked up Technical Specification pages showing the proposed changes are provided in Attachment 2. The associated Bases changes are provided in for information. A copy of this application and the associated attachments are being submitted to the designated New York State official in accordance with 10 CFR 50.91.

N -

NL-1 3-003 Docket 50-247 Page 2 of 2 Entergy requests approval of the proposed amendment within 12 months and an allowance of 30 days for implementation. There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.

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

2013.

Sincerely, JAV/ai Attachments:

1.

Analysis of Proposed Technical Specification Changes Regarding RWST Temperature and Containment Pressure

2.

Marked Up Technical Specification Pages for Proposed Changes Regarding RWST Temperature and Containment Pressure

3.

Marked Up Technical Specification Bases Pages for Proposed Changes Regarding RWST Temperature and Containment Pressure 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 1 TO NL-13-003 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING RWST TEMPERATURE AND CONTAINMENT PRESSURE ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

NL-1 3-003 Docket No. 50-286 Page 1 of 8

1.0 DESCRIPTION

Entergy Nuclear Operations, Inc (Entergy) is requesting an amendment to Operating License DPR-64, Docket No. 50-286 for Indian Point Nuclear Generating Unit No. 3 (IP3): The proposed amendment will revise Technical Specifications (TS) 3.5.4 surveillance requirement (SR) 3.5.4.1, to limit the maximum Refueling Water Storage Tank (RWST) temperature to _!105 0F; TS 3.6.4 Limiting Condition for Operation (LCO) to limit containment pressure to _<+11.5 psig if RWST temperature is >950F or containment temperature is >1250F; TS 3.6.3 SR 3.6.3.9 is being deleted and TS 5.5.15 Containment Leakage Rate Testing Program is being changed to specify the re-analyzed value of peak containment pressure.

A re-analysis of the large break loss-of-coolant accident (LOCA) was performed to correct methodology errors in the long-term mass and energy (M&E) releases for containment integrity analysis. The RWST temperature and containment pressure changes are necessary as a result of the re-analysis to maintain the peak containment pressure at about the same value as the current analysis of record.

The specific proposed changes are listed in the following section.

2.0 PROPOSED CHANGE

S The proposed TS changes are as follows:

Change SR 3.5.4.1 from SURVEILLANCE FREQUENCY SR 3.5.4.1 -----------

NOTE ----------------------------

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Not required to be performed when ambient air temperature is > 35 0F and < 11 0°F if heating steam supply isolation valves are locked closed.

Verify RWST borated water temperature is > 35 0F and

_ 110°F.

To

NL-1 3-003 Docket No. 50-286 Page 2 of 8 SURVEILLANCE FREQUENCY SR 3.5.4.1

---

NOTE ---------------------------------

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Not required to be performed when ambient air temperature is > 350F and < 950F if heating steam supply isolation valves are locked closed.

Verify RWST borated water temperature is _>

350F and

< 1050F.

Change TS LCO 3.6.4 from Containment pressure shall be >_ -2.0 psig and <_ +2.5 psig.

To Containment pressure shall be maintained as follows:

a. If RWST temperature is > 950F or containment temperature is > 1250F, Containment pressure shall be _> -2.0 psig and _< +1.5 psig.

b. If RWST temperature is < 950F and containment temperature is < 125 0F, Containment pressure shall be _> -2.0 psig and _< +2.5 psig.

Delete SR 3.6.3.9 SR3..Vorify the combined leakage rate for all conRtainmen in accordance with bypass leakage paths is:!* 0.6 L. when pressurizedt the Containmoent 42.42 sig.

eakage-Rate TeetiRg PFog~aFR Change 5.5.15, Containment Leakage Rate Testing Program, from The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42.0 psig. The containment design pressure is 47 psig.

To

NL-1 3-003 Docket No. 50-286 Page 3 of 8 The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42.38 psig. The containment design pressure is 47 psig.

The marked up Technical Specification pages showing these changes are in Attachment 2. The associated changes to the Technical Specification Bases, to be made after approval using the 10 CFR 50.59 process, are in Attachment 3 for information.

3.0 BACKGROUND

Nuclear Safety Advisory Letter 11-5 (NSAL-1 1-5, Reference 1) identified Westinghouse methodology errors in the long-term mass and energy (M&E) releases during a large break loss-of-coolant accident (LOCA). These impacted containment integrity analysis for Indian Point Unit 3 (IP3).

The four issues listed below impact the IP3 long-term LOCA M&E release calculation utilizing the Westinghouse containment analysis methodology; The reactor vessel modeling did not include all the appropriate vessel metal mass available from the component drawings. This discrepancy results in an inaccurate vessel metal mass that affects the amount of reactor vessel stored energy initially available in the M&E model.

The reactor vessel model did not include the appropriate amount of vessel metal mass in the reactor vessel barrel/baffle downcomer region. Differences were identified in the calculated metal mass and surface area input values. Increases in the barrel/baffle metal mass impact the initial energy stored within the reactor vessel.

The long-term LOCA M&E release analysis was initialized at a non-conservative (low) steam generator (SG) secondary pressure condition. This input value determines the initial SG secondary side temperature and pressure used in the long-term LOCA M&E release calculations. The pressure at the exit of the SG outlet nozzle was incorrectly used as the SG secondary side pressure, as opposed to the correct, higher tube bundle pressure.

An error was found in the EPITOME computer code that is used to determine the M&E release rate during the long-term (i.e., post-reflood) SG depressurization phase of the LOCA transient. The error results in an underestimated energy release in the long-term, post-reflood phase of the transient.

The analysis of record (AOR) peak containment pressure is 40.38 psig for the double-ended hot leg (DEHL) break and 42.00 psig for the double-ended pump suction (DEPS) break, respectively (Reference 4).

4.0 TECHNICAL ANALYSIS

Westinghouse re-analyzed the containment integrity analysis with the errors identified in Section 3.0 corrected in the long-term LOCA M&E model. Further, the re-analysis assumed a RWST initial temperature of 105°F and a containment initial pressure and temperature of 1.5 psig and 130TF,

NL-13-003 Docket No. 50-286 Page 4 of 8 respectively.

The analysis of record for containment integrity is based on the limiting single failure of 32 emergency diesel generator (EDG) coincident with loss of offsite power. As noted in FSAR Section 14.3.6.2.2, "The minimum ECCS case was based upon a diesel train failure (which leaves available as active heat removal systems one containment spray pump and four RCFCs)".

With the error corrections of NSAL-1 1-5, and assuming the same initial conditions as the analysis of record (RWST temperature = 100F, containment pressure = 2.5 psig, containment temperature

= 1300F), a peak containment pressure of 44.26 psig was calculated for the DEPS in Reference 2.

While this pressure is well below the containment design pressure of 47 psig, it was desired to maintain peak containment pressure at or about the current analysis of record value for purposes of other programs. Consequently, sensitivity studies were performed in Reference 3 to evaluate the impacts of initial RWST temperature, initial containment pressure, and initial containment temperature on peak containment pressure. These studies showed that with an initial RWST temperature of 1050F, initial containment pressure of 1.5 psig, and initial containment temperature of 1300F, the peak containment pressure would be 39.71 psig for the DEHL break and 42.38 psig for the DEPS break. As shown in Table 1 below, the most limiting peak containment pressure is slightly higher than the analysis of record, and will have an insignificant impact on other programs.

There are no changes to design, and the revised analysis is consistent with the plant configuration for equipment availability and the peak containment pressure remains well below the design pressure of 47 psig.

Table 1 -Comparison of Peak Containment Pressure Error Correction and Peak Containment Pressure Analysis of Record [psig]

RWST=105 0 F, containment pressure 1.5psig

[psig]

Double-Ended Hot Leg (DEHL) Break 40.38 39.71 Double-Ended Pump Suction (DEPS) Break 42.00 42.38 The sensitivity studies also demonstrated that a 50F decrease in initial RWST temperature would result in 0.46 psi reduction in peak containment pressure, a 50F decrease in initial containment/accumulator temperature would result in 0.57 psi reduction in peak containment pressure and a 0.25 psi decrease in initial containment pressure would result in 0.30 psi decrease in peak containment pressure. As mentioned above, the case in Reference 2 which had initial conditions of 1 10°F for RWST temperature, 2.5 psig for initial containment pressure and 130°F for initial containment/accumulator temperature resulted in a peak containment pressure of 44.26 psig.

Reducing the RWST temperature for this case to 950F would result in a peak containment pressure of 42.88 psig [44.26-(3x0.46)] and reducing the containment/accumulator temperature from 130OF to 1250F would result in a peak containment pressure of 42.31 psig (42.88-0.57). Thus, with a RWST temperature _<95 0F and containment /accumulator temperature of *1 250F, an initial containment pressure of 2.5 psig would result in an acceptable peak containment pressure. It should be noted that while the accumulator is located in the containment, due to its lower elevation, it is expected to be at a lower temperature than the containment average temperature.

NL-13-003 Docket No. 50-286 Page 5 of 8 Section 5.5.15 for the Containment Leakage Rate Testing Program has been revised to reflect the change in the peak calculated pressure from 42.0 psig to 42.38 psig. The peak pressure is not an accident initiator. The increase in peak pressure does not result in an increase in doses since the pressures used for the Type A, Type B and Type C tests are above 42.38 psig. The containment Appendix J test also exceeded this value. The increase in calculated pressure does not affect systems, components or tests.

Based on the above, the surveillance requirement for RWST found in TS 3.5.4 has been lowered to an acceptance criteria of 1050F as the maximum allowable temperature for which an analysis has been done. The note concerning the temperature at which this monitoring must be performed has also been decreased to 950F since this is the RWST temperature at which the containment pressure must be reduced. The RWST will not exceed this temperature until the outside temperature reaches it. This change does not affect the probability of an accident and is consistent with accident analyses so no doses would increase. There are no changes to the operation of any systems or components or any tests.

Based on the above, the surveillance requirement for Containment Air Temperature found in TS 3.6.5 LCO and Condition B need not be lowered to an acceptance criteria of 125°F since the pressure limit of 42.38 psig is met with a Containment air temperature of 1300F, a RWST temperature of 1050F and the initial Containment air pressure of 1.5 psig (as required with the RWST above 950F or the Containment air temperature above 125°F). There are surveillances every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for RWST and Containment temperature.

Based on the above, the Containment pressure limits will be maintained using two sets of values.

When the RWST temperature is below 950F and the Containment temperature is below 1250F, the peak internal pressure can be < 2.5 psig. When either temperature exceeds that value, the containment pressure must be < 1.5 psig. This change does not affect the probability of an accident and is consistent with accident analyses so no doses would increase. There are no changes to the operation of any systems or components or any tests.

Indication of containment pressure is available in the control room with an uncertainty of +/- 1.5 psi with all indicators operable. With this indication, containment purging from the control room at _ 1 psi currently satisfies the accident analysis initial condition of 2.5 psig when RWST temperature is 950F and containment/accumulator temperature is *1 250F. When RWST temperature is > 950F or containment/accumulator temperature is >1250F, containment pressure indication with a higher accuracy instrument will assure the accident initial conditions are maintained. This will be controlled by a local mounted high accuracy containment pressure indicator. With this indication, which will have an uncertainty of at least +/- 0.5 psi, containment purging at _ 1.0 psi would satisfy the accident analysis initial conditions. The TS change allows for continued monitoring of the containment pressure from the control room except maybe for a few hot days during the summer, when operators would be required to monitor the containment pressure from the locally mounted instrument.

SR 3.6.3.9 is being deleted as it is redundant to TS 5.5.15. When IP3 converted from custom TS to Standard Technical Specifications (STS) (Reference 9), SR 3.6.3.9 was modified. STS SR 3.6.3.11 was intended to measure bypass leakage from a shield building. Specifically, STS SR 3.6.3.11 which states "Verify the combined leakage rate for all shield building bypass leakage paths is < [ La ] when pressurized to _Ž [ psig]" was modified to state "Verify the combined leakage rate for all containment bypass leakage paths is < 0.6La when pressurized to >_ 42.42 psig." IP3 has no shield building, and the STS SR for bypass leakage from a shield building does not apply.

NL-13-003 Docket No. 50-286 Page 6 of 8 Since leakage rate testing of the containment is specified in TS 5.5.15, Containment Leakage Rate Testing Program, SR 3.6.3.9 is redundant and may be deleted.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Entergy has evaluated the safety significance of the proposed changes to the Indian Point 3 Technical Specifications. The proposed changes have been evaluated according to the criteria of 10 CFR 50.92, "Issuance of Amendment". The changes to SR 3.6.3.9 and TS 5.5.15 are considered editorial changes, in that SR 3.6.3.9 is being deleted due to not being applicable to IP3 and redundancy to TS 5.5.15 and TS 5.5.15 is being changed to specify the re-analyzed value of peak containment pressure. Entergy has determined 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?

No. The proposed change would not change the current limiting EDG failure but limits the RWST temperature to *1 050F and containment pressure to *1.5 psig (when RWST temperature is >950F or containment/accumulator temperature is >1250F). The proposed change also removes a redundant TS for Containment testing and corrects the peak pressure in the containment testing program. The initial conditions assumed in accident analysis are not accident initiators so the probability of an accident does not increase. The change in initial conditions compensates for the error corrections and maintains the post accident containment pressure within 0.38 psig of the current value and within Containment testing limits and therefore does not increase the probability or consequences of a previously evaluated accident. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The change to the initial conditions assumed in the analysis for peak containment pressure, the removal of a redundant Technical Specification and the correction to the peak pressure limit in the Containment testing program do not create the possibility of a new or different accident. There are no changes to design or operating procedures that could create a new or different kind of accident since the changes only affect the initiating conditions. The revised analysis is consistent with the available equipment following the postulated worst case single failure.

Therefore the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. The, change in peak containment pressure is from 42 psig to 42.38 psig as a result of the error corrections of NSAL-1 1-5 and change to the initial conditions for the RWST temperature and containment pressure. There is an insignificant impact on other programs

NL-1 3-003 Docket No. 50-286 Attachment I Page 7 of 8 due to change in peak containment pressure, which remains well below the containment design pressure of 47 psig. Therefore there is no significant reduction in a margin.

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

Based on the above, Entergy concludes that the proposed amendment to the Indian Point 3 Technical Specifications 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 Regulatory Requirements / Criteria The plant will continue to meet Criterion 2 of 10 CFR 50.36 which says "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." The current version of TS SR 3.5.4.1 and TS LCO 3.6.4 are being changed to be consistent with the input assumptions used in the re-analysis (Reference 3) to address errors identified in NSAL-1 1-5. The change to SR 3.6.3.9 and TS 5.5.15 are editorial in that SR 3.6.3.9 is being deleted due to redundancy to TS 5.5.15 and TS 5.5.15 is being changed to specify the re-analyzed value of peak containment pressure. The same codes and methods were used in the re-analysis as was done for Stretch Power Uprate License Amendment Request (LAR) of Reference 4 which contained a proprietary report, Reference 5, and which was approved as Amendment 225 by the NRC in Reference 6.

Section 6.5.3.5.1 of Reference 4 stated: "The associated single failure assumption is the failure of a diesel to start, resulting in one train of ECCS and containment safeguards equipment being available. This combination results in a minimum set of safeguards equipment being available."

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 operation 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.

5.3 Environmental Considerations The proposed changes to the IP3 Licensing Basis 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 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 A Technical Specification revision to change the RWST temperature was found for Indian Point

NL-1 3-003 Docket No. 50-286 Page 8 of 8 Unit 2, which revised the RWST temperature from 1 00°F to 11 0°F at the time of the Stretch Power Uprate submittal (Reference 7) and approved as Amendment 241 by the NRC in Reference 8.

7.0 REFERENCES

1.

Nuclear Safety Advisory Letter, "Westinghouse LOCA Mass and Energy Release Calculation Issues," NSAL-11-5, dated July 25, 2011.

2.

Letter from Edward P. Shields (Westinghouse) to Nasser Nik (Entergy), "LOCA Mass and Energy Analysis," INT-12-3, dated February 24, 2012.

3.

Letter from Edward P. Shields (Westinghouse) to Nasser Nik (Entergy), "LOCA Mass and Energy Analysis," INT-12-8, dated April 23, 2012.

4.

Entergy Letter NL-04-069 to NRC, "Proposed Changes to Technical Specifications:

Stretch Power Uprate (4.85%) and Adoption of TSTF-339," dated June 3, 2004.

5.

Indian Point Nuclear Generating Unit No. 3 Stretch Power Uprate NSSS and BOP Licensing Report, WCAP-16212-P, dated June 2004.

6.

NRC Letter to Entergy, Indian Point Nuclear Generating Unit No 3 - Issuance of Amendment Re: 4.85 Percent Stretch Power Uprate and Relocation of Cycle-Specific Parameters (TAC No. MC3552), March 24, 2005.

7.

Entergy Letter NL-04-005 to NRC, "Proposed Changes to Technical Specifications:

Stretch Power Uprate Increase of Licensed Thermal Power (3.26%)," dated January 29, 2004.

8.

NRC Letter to Entergy, Indian Point Nuclear Generating Unit No 2 - Issuance of Amendment Re: 3.26 Percent Power Uprate (TAC No. MC1865), October 27, 2004.

9.

NRC Letter to Entergy Issuing Amendment For Conversion to Improved Standard Technical Specifications (TAC No. MA4359), dated February 27, 2001.

ATTACHMENT 2 TO NL-13-003 MARKED UP TECHNICAL SPECIFICATION PAGES FOR PROPOSED CHANGES REGARDING RWST TEMPERATURE AND CONTAINMENT PRESSURE Changes indicated by lineout for deletion and Bold/Italics for additions Unit 3 Affected Pages:

3.5.4-2 3.6.4-1 3.6.3 -6 5.0-31 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1

---

NOTE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Not required to be performed when ambient air temperature is

> 350F and < 4--

950F if heating steam supply isolation valves are locked closed.

Verify RWST borated water temperature is 35°F and

• 14 105'F.

SR 3.5.4.2 Verify RWST borated water level is 7 days

Ž 35.4 feet.

SR 3.5.4.3 Verify RWST boron concentration is 31 days

Ž 2400 ppm and

• 2600 ppm.

SR 3.5.4.4 Perform CHANNEL CHECK of RWST level 7 days 3R 3.5.4.5 Perform CHANNEL CALIBRATION of RWST level 184 days switch and ensure the low level alarm setpoint is 10.5 feet and *12.5 feet.

3R 3.5.4.6 Perform CHANNEL CALIBRATION of RWST level 18 months transmitter and ensure the low level alarm setpoint is

Ž10.5 feet and *12.5 feet.

INDIAN POINT 3 3.5.4-2 Amendment 2-4-4

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be maintained as follows:

Pf A ll~

• RiR ;;R i5 4

P

',2F

a. If RWST temperature is

> 950F or containment temperature is

> 1250F, Containment pressure shall be

&> -2.0 psig and ! +1.5 psig.

b.

If RWST temperature is

! 950F and containment temperature is

< 1250F, Containment pressure shall be

> -2.0 psig and * +2.5 psig.

APPLICABILITY:

MODES 1, 2, 3,

and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment A.1 Restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> pressure not containment within limits, pressure to within limits.

B.

Required Action B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and associated Completion Time AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limits.

INDIAN POINT 3 3.6.4-1 Amendment 2-0-5

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.5 Verify the isolation time of each In accordance automatic power operated containment with the isolation valve is within limits.

Inservice Testing Program SR 3.6.3.6 Verify each automatic containment 24 months isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

SR 3.6.3.7 Verify each 10 inch containment 24 months pressure relief line isolation valve is blocked to restrict valve opening to

• 60 degrees.

SR 3.6.3.8 Perform one complete cycle of each 24 months manually operated containment isolation valve on essential lines.

sR 3.6.31.Voify the eomdbin"d leg*o...o rt. for In acoordanco all c.nta.in...nt bypass leakage paths is with the 9 0.6 L-- when prossurizcd to Ž 42.42 Containmont PLoakaej Ratc Testing Program SR 3.6.3.47G9 Verify leakage rate into containment In accordance from isolation valves sealed with with the service water system is within limits.

Containment Leakage Rate Testing Program INDIAN POINT 3 3.6.3-6 Amendment 245

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate Testing Program (continued) cooler unit when pressurized at > 1.1 Pa.

This limit protects the internal recirculation pumps from flooding during the 12-month period of post accident recirculation.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10CFR50, Appendix J.

The calculated peak containment internal pressure for the design basis loss of coolant accident, P,, is 42.438 psig.

The containment design pressure is 47 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of primary containment air weight per day.

5,5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE)

Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation System (CRVS),

CRE occupants can control the reactor safely under normal conditions and maintain it in a

safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5

rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

(continued)

INDIAN POINT 3 5.0-31 Amendment 2-39

ATTACHMENT 3 TO NL-13-003 MARKED UP TECHNICAL SPECIFICATION BASES PAGES FOR PROPOSED CHANGES REGARDING RWST TEMPERATURE AND CONTAINMENT PRESSURE Changes indicated by lineout for deletion and Bold/Italics for additions Unit 3 Affected Pages:

B 3.5.4-3 B 3.5.4-6 B 3.6.2-2 B3.6.3-1 0 B3.6.3-1 1 B3.6.3-16 B3.6.3-17 B 3.6.4-1 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

RWST B 3.5.4 BASES APPLICABLE SAFETY ANALYSIS (continued) to assure subcriticality.

The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

The RWST level required by Technical Specifications includes allowances for instrument accuracy, the unusable volume in the RWST, and the maximum volume expected to remain in the RWST when the plant is switched from the injection to recirculation modes of operation.

The upper limit on boron concentration of 2600 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA.

The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident.

In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of 35°F.

If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature.

The upper temperature limit of 4-O-105°F is used in the LOCA containment integrity analysis.

Exceeding this temperature will result in higher containment pressures due to reduced containment spray cooling capacity.

The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability.

For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.

Following a LOCA, switchover from the injection phase to the recircuIation phase must occur before the RWST empties to prevent damage to the pumps and a loss of cooling capability.

For similar reasons, switchover must not occur before there is sufficient water in the containment to support recirculation pump suction.

Furthermore, early switchover must not occur to ensure that sufficient borated water is injected from the RWST.

(continued)

INDIAN POINT 3 B 3.5.4 -3 Revision -.

RWST B 3.5.4 BASES ACTIONS C.l (continued)

With the RWST inoperable for reasons other than Condition A (e.g., water volume) or B (e.g.,

two level alarms inoperable),

it must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In this Condition, neither the ECCS nor the Containment Spray System can perform its design function.

Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required.

The short time limit of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

D.1 and D.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.5.4.1 The RWST borated water temperature should be verified every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be within the limits assumed in the accident analyses band.

This Frequency is sufficient to identify a temperature change that would approach either limit and has been shown to be acceptable through operating experience.

The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST (a reduced temperature is used to assure restrictions on allowable Containment operating pressures are met) and the heating steam isolation valves are locked closed.

With ambient air temperatures within the band, the RWST temperature should not exceed the limits.

(continued)

INDIAN POINT 3 B 3.5.4--6 Revision 2

Containment Air Locks B 3.6.2 BASES BACKGROUND The containment air locks form part of the containment (continued) pressure boundary.

As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limit in the event of a DBA.

Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analyses.

APPLICABLE SAFETY ANALYSES The DBAs that result in a release of radioactive material within containment are a loss of coolant accident and a rod ejection accident.

In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage.

The containment was designed with an allowable leakage rate of 0.1% of containment air weight per day (Ref. 2).

This leakage rate is defined in 10 CFR 50, Appendix J, Option B (Ref.

1), as La =

0.1% of containment air weight per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure Pa = 42.438 psig following a DBA (LBLOCA or MSLB).

This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.

The containment air locks satisfy Criterion 3 of 10 CFR 50.36.

LCO Each containment air lock forms part of the containment pressure boundary.

As part of the containment pressure

boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA.

Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

Each air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE.

The interlock allows only one air lock door of an air lock to be opened at one time.

This provision ensures that a gross breach of containment does not (continued)

INDIAN POINT 3 B 3.6.2 -- 2 Revision 4

Containment Isolation Valves B 3.6.3 BASES ACTIONS C.l and C.2 (continued) flow path must be verified to be isolated on a periodic basis.

This periodic verification is necessary to assure leak tightness of containment and that containment penetrations requiring isolation following an accident are isolated.

The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.

Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system.

This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system.

The closed system must meet the requirements of Reference 3.

Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted.

Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper

position, is small.

D.1 With the containment bypass leakage rate not within limit of SR 3.6.3.9, TS 5.5.15, the assumptions of the safety analyses are not met.

Therefore, the leakage must be restored to within limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Restoration can be accomplished by isolating the penetration(s) that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or 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, the leakage rate is assumed (continued)

INDIAN POINT 3 B 3.6.3 -

10 Revision 0

Containment Isolation Valves B 3.6.3 BASES ACTIONS D.1 (continued) to be the lesser actual pathway leakage of the two devices.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration(s) and the relative importance of containment bypass leakage to the overall containment function.

With the hydrostatically tested valve leakage not within limit of SR 3.6.3.-i49, the potential exists for flooding the Containment Recirculation Pumps during long term post-accident cooling.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable because of the low probability of an event occurring during this period.

E.1 and E.2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.6.3.1 Each 36 inch containment purge supply and exhaust isolation valve (FCV-1170, FCV-1171, FCV-1172, and FCV-1173) is required to be verified sealed closed at 31 day intervals.

This Surveillance is designed to ensure that a gross breach of containment is not caused by an inadvertent or spurious opening of a containment purge valve.

Detailed analysis of the purge valves failed to conclusively demonstrate their ability to close during a LOCA in time to limit offsite doses.

Therefore, these valves are required to be in the sealed closed position during MODES 1, 2, 3,

and 4.

A containment purge valve that is sealed closed must have motive power to the valve operator removed.

This can be accomplished by de-energizing the source of electric power or by removing the air supply to the valve operator.

(continued)

INDIAN POINT 3 B 3.6.3-1ii Revision -

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.9 This SR ensur-es that: the comfbined lcakage rate of all elntainment leakage paths is less tbhan or equal t the speiafied leakage rate far the ose aths thato airc ot sealedl by the Isolation Valve Seal Waterc Systemf orn sealcd by the RHR systemR or scolodi by the scrvioc wat-er tcm.

This prcavidcts assutranc that the assum.tions in tc safety analysis are met-.

The leakage rate of eacht bypeass leakagýe p9ath is assumfedl to be the mfaxifmum pathway leakage (leakage thruugh the worse of the two isolatio n

valves) unless the penetr.atin is isolated by use of one closedi andl do activat-ed auatomfatic valve, closedl mfanual valve, or blinde flange;" in this ease, the leakage rat-e of the isolated bypass leakage path is assmed to be thei actual p.athway leakage throu.gh the iselation device.

If both isolation valves in the penetrýation are closedl, the&

actual leakage rate is the lesser-leakage rate of the two valvesd.

This testing is p9ierffored i:n accordianec with the&

requiriiements, Fr~equency andl aceepltanee craiter~ia required by Specification 5.5.15, Containmfent-Leakage Rate Testing Programff.

This progrjLamf was est-ablimshedi to fimplemfent the leakeage r~at-e testimng of the containmfent as r-equiraed by 10 CERý 59.54(e) and 10 CER 50, Appenedir J, Option B, as modiified b9y !P3 specific approeved exemptions. This p9rogram conformffs to guaidelines containedi in Reguilatory Cuaide 1.163, "Per~iforance Basedi Containm~ent Lýeak Test Proegrcam, diated Septembler 1995."

In the event containmnent isolation valve leakage r-esulits in exceedeing the over-all containmfent leakage rcate, entrLy into the applicable Conditions and Reqluired Actions of LCO 3.6.1 is r~equired6.

SR 3.6.3.449 The Containment Leakage Rate Testing Program includes verification that inleakage rate from the containment isolation valves sealed with service water is maintained at a level that will prevent flooding the internal recirculation pumps for the full 12-month period of post accident recirculation.

This inleakage test has specific acceptance criteria (< 0.36 gpm per fan cooler unit when pressurized at > 1.1 P,) specified in the (continued)

INDIAN POINT 3 B 3.6.3 -

16 Revision 0

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.3.4--9 (continued)

Containment Leakage Rate Testing Program and the results for this inleakage test are not counted against the acceptance criteria for the Type B and C tests that are also performed as part of the SR.

REFERENCES

1.

FSAR, Section 14.

2.

FSAR, Section 6.

3.

Standard Review Plan Section 6.2.4.

4.

FSAR, Section 5.2.

5.

Generic Issue B-24.

6.

Safety Evaluation Report for IP3 Amendment 195.

INDIAN POINT 3 B 3.6.3 -

17 Revision 0

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB).

The containment can withstand an internal vacuum of 3 psig.

The 2.0 psig vacuum specified as an operating limit avoids any difficulties with motor cooling.

Containment pressure is a process variable that is monitored and controlled.

The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis.

Should operation occur outside these limits coincident with a Design Basis Accident (DBA),

post accident containment pressures could exceed calculated values.

APPLICABLE SAFETY ANALYSES Containment internal pressure is an initial condition used in the DBA analyses to establish the maximum peak containment internal pressure.

The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients. Cycle specific analysis results indicate that the worst case peak containment pressure could result from either a loss of coolant accident or a steam line break inside containment (Ref.

1).

The initial pressure condition used in the containment analysis was +21.5 psig.

Sensitivity studies have demonstrated that if RWST temperature is

! 95°F and containment/accumulator temperatures are 51250F, an initial containment pressure of +2.5 psig would also be acceptable. This analysis concluded that the containment design pressure of 47 psig would not be exceeded for either a major loss-of-coolant accident or for a main steam line break accident.

The containment analysis results are presented in Reference 1 and the current value for peak containment pressure is listed in Specification 5.5.15, "Containment Leakage Rate Testing Program."

(continued)

INDIAN POINT 3 B 3.6.4 -

1 Revision 0

ATTACHMENT 9.4 NRC SUBMiTTAL REVIEw FORM Sheet 1 of 2 Letter#: NL-12-xxx (IP3 TS Change)

Response Due: 10-17-2012

Subject:

. Proposed Technical Specification Date Issued for Review: 10-26-2012::.....

Changes Regarding RWST Temperature and Containment Pressure / Temperature in Containment Integrity Analysis Correspondence Preparer Phone #: A Irani x6618 Section I Letter Concurrence and Agreement to Perform Actions POSITION NAME..

Action Signature (concurrence, certification, etc.)

(sign,.interoffice memo, e-mail, or telecom)

S Prussman - Licensing Concurrence e_

A-_a C- &o1,d,".

i,*

A-cb!.

J Chang - Fuels & Analysis Concurrence/Certification

///4p/*-c/2.

J Hill - I&C Concurrence/Certification

£_- MCoA. t o,,J.4A4 R Walpole - Man Licensing Concurrence M Lewis - Ops Concurrence/Cec,4ifeie*

(,, 0 P Conroy-NSA Concurrence ell T McCaffrey - Design Man Concurrence g

c,,JLo--L D"

+A.. t Gef-l-"

r 1

IIc OSRC Recommendation A

.pove -- i t

^

11 J Ventosa - Site VP Signature M Woodby - Dir Eng Information V APDRE zI RBti

- Systems Man Information M Tesoriero - P& C Man Information COMMENTS EN-LI-1 06 Rev. 9

Section II Section 11 Corre.n.ndenr.

Screenino Does this letter contain commitments? If "yes," identify the commitments with due Yes El dates in the submittal and in Section III. When fleet letters contain commitments, a No PCRS LO (e.g., LO-LAR, LO-WT) should be initiated with a CA assigned to each applicable site to enter the commitments into the site's commitment management system.

Does this letter contain any information or analyses of new safety issues performed at NRC Yes El request or to satisfy a regulatory requirement? If "yes," reflect requirement to update the No z

UFSAR in Section II1.

Does this letter require any document changes (e.g., procedures, DBDs, FSAR, TS Bases, Yes El etc.), if approved? If "yes," indicate in Section III an action for the responsible No 0l department to determine the affected documents. (The Correspondence Preparer may indicate the specific documents requiring revision, if known or may initiate an action for review.)

Does this letter contain information certified accurate? If "yes," identify the information Yes

[]

and document certification in an attachment. (Attachment 9.5 must be used.)

No El ATTACHMENT 9.4 NRC SUBMITTAL REVIEW FORM Sheet 2 of 2 Section III Actions and Commitments Required Actions Due Date Responsible Dept.

Note: Actions needed upon approval should be captured in the appropriate action tracking system LO-LAR-2012-157 CA#4 10/30/2013 Licensing Commitments Due Date Responsible Dept.

Note: When fleet letters contain commitments, a PCRS LO should be initiated with a CA assigned to each applicable site to enter the commitments into the site's commitment management system.

Section IV Final Document Signoff for Submittal Correspondence Preparer A Irani aA-

ýA 6L^ý-

ILl12-ol3 Final Submittal Review (optional)

Responsible Department Head z N EN-LI-106 Rev. 9