RA-19-0015, Request for License Amendment: Surveillance Requirements for Safety Relief Valves

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Request for License Amendment: Surveillance Requirements for Safety Relief Valves
ML19063B740
Person / Time
Site: Brunswick  Duke energy icon.png
Issue date: 03/04/2019
From: William Gideon
Duke Energy Progress
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0015
Download: ML19063B740 (31)
v • d • e


Text

William R. Gideon Vice President Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.832.3698 March 4, 2019 Serial: RA-19-0015 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Request for License Amendment - Surveillance Requirements for Safety Relief Valves Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is submitting a request for an amendment to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed change modifies Surveillance Requirement (SR) 3.4.3.2, in TS 3.4.3, "Safety Relief Valves (SRVs)," and SR 3.5.1.11, in TS 3.5.1, "ECCS - Operating," by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened. Additionally, the proposed change revises the Frequency for performing these SRs to be in accordance with the Inservice Testing Program.

The enclosure provides a description and assessment of the proposed change. Attachments 1 and 2 to the enclosure provide the existing TS pages, for Units 1 and 2, respectively, marked to show the proposed change. Attachments 3 and 4 provide revised (i.e., typed) TS pages for Units 1 and 2, respectively. Existing Unit 1 TS Bases pages, marked to show corresponding changes, are provided in Attachment 5 for information only.

Approval of the proposed amendment is requested within one year of completion of the NRCs acceptance review. Once approved, the amendment shall be implemented within 120 days.

In accordance with 10 CFR 50.91, Duke Energy is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.

This document contains no new regulatory commitments. Please refer any questions regarding this submittal to Mr. Art Zaremba, Director - Nuclear Fleet Licensing, at (980) 373-2062.

U.S. Nuclear Regulatory Commission Page 2 of 3 I declare, under penalty of perjury, that the foregoing is true and correct. Executed on March 4, 2019.

William R. Gideon MAT/mat

Enclosure:

Description and Assessment of the Proposed Change Attachment 1: Proposed Technical Specification Changes (Mark-Up) - Unit 1 Attachment 2: Proposed Technical Specification Changes (Mark-Up) - Unit 2 Attachment 3: Revised (Typed) Technical Specification Pages - Unit 1 Attachment 4: Revised (Typed) Technical Specification Pages - Unit 2 Attachment 5: Technical Specification Bases Pages (Mark-Up) - Unit 1 (For Information Only)

U.S. Nuclear Regulatory Commission Page 3 of 3 cc:

U. S. Nuclear Regulatory Commission, Region II ATTN: Ms. Catherine Haney, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Dennis J. Galvin 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Mr. Gale Smith, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov

RA-19-0015 Enclosure Page 1 of 9 Description and Assessment of the Proposed Change

Subject:

Request for License Amendment - Surveillance Requirements for Safety Relief Valves

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES ATTACHMENTS:
1. Proposed Technical Specification Changes (Mark-Up) - Unit 1
2. Proposed Technical Specification Changes (Mark-Up) - Unit 2
3. Revised (Typed) Technical Specification Pages - Unit 1
4. Revised (Typed) Technical Specification Pages - Unit 2
5. Technical Specification Bases Pages (Mark-Up) - Unit 1 (For Information Only)

RA-19-0015 Enclosure Page 2 of 9

1.

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is submitting a request for an amendment to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed change modifies Surveillance Requirement (SR) 3.4.3.2, in TS 3.4.3, "Safety Relief Valves (SRVs)," and SR 3.5.1.11, in TS 3.5.1, "ECCS - Operating," by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened.

2. DETAILED DESCRIPTION 2.1 System Design and Operation The nuclear system pressure relief system includes eleven SRVs each for Units 1 and 2, all of which are located on the main steam lines within the drywell between the reactor vessel and the first isolation valve. The SRVs are Target Rock model 7567F two-stage, pilot-operated safety relief valves. They are mounted on the four main steam lines so that a single accident cannot completely disable a safety, relief, or automatic depressurization function. The SRVs are installed so that each valve discharge is piped through its own discharge line to a point below the minimum water level in the primary containment suppression pool to permit the steam to condense in the pool.

The SRVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve.

Seven of the SRVs serve as Automatic Depressurization System (ADS) valves. Their purpose is to automatically depressurize the reactor vessel thus allowing injection by low pressure emergency cooling sources.

2.2 Current Technical Specification Requirements Currently, SR 3.4.3.2 and 3.5.1.11 read as follows.

SR 3.4.3.2 -------------------------------NOTE---------------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required SRV opens when manually actuated.

SR 3.5.1.11 -------------------------------NOTE---------------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required ADS valve opens when manually actuated.

RA-19-0015 Enclosure Page 3 of 9 The current Frequency for each SR is "In accordance with the Surveillance Frequency Control Program." At present, the Surveillance Frequency Control Program (SFCP) establishes the Frequencies for both SRs as 24 months.

2.3 Reason for the Proposed Change TS surveillance testing of SRVs at BSEP is currently performed at a reactor pressure vessel (RPV) pressure greater than or equal to 945 psig to verify that, mechanically, each valve is functioning properly and no blockage exists in the valve discharge line.

The Boiling Water Reactor Owners' Group (BWROG) Evaluation of NUREG-0737, "Clarification of TMI Action Plan Requirements, Item II.K.3.16, "Reduction of Challenges and Failures of Relief Valves," recommends that the number of SRV openings be reduced as much as possible and that unnecessary challenges to the SRVs be avoided.

Experience in the industry has shown that manual actuation of SRVs during plant operation may create a potential for SRV seat leakage. Potential SRV leakage is routed to the suppression pool. The increased heat and fluid additions to the suppression pool requires more frequent suppression pool cooling and more frequent pump-down operations to control suppression pool temperature and level. Main stage SRV seat leakage also tends to mask the indications of SRV pilot stage seat-leakage. Pilot stage leakage could cause spurious SRV actuation and/or SRV failure to reclose after actuation. Excessive leakage would require plant shutdown to replace the leaking SRV.

Reducing or eliminating the number of manual actuations of the SRVs during plant startup minimizes the potential depressurization and cooldown events due to failure-to-close SRV events as well as minimizing the potential for pilot or main stage leakage of the SRVs.

Implementing this change would still maintain the capability to manually open and close SRVs, as necessary, for the Inservice Testing (IST) Program or as corrective action for SRVs with excessive leakage.

2.4 Description of the Proposed Change The proposed change revises SR 3.4.3.2 to verify each required SRV is capable of being opened versus verifying that each required SRV opens. Similarly, SR 3.5.1.11 is revised to verify each required ADS valve is capable of being opened versus verifying that each required ADS valve opens. Additionally, the proposed change revises the Frequency for performing these SRs to be in accordance with the Inservice Testing Program.

TS Bases associated with these Surveillance Requirements will also be revised to describe the new testing method. Existing Unit 1 TS Bases pages, marked to show these changes, are provided for information only.

3. TECHNICAL EVALUATION The manual actuation tests currently prescribed in TS SRs 3.4.3.2 and 3.5.1.11 provide demonstration of the mechanical operation of the SRVs, and overlaps with other testing to demonstrate that the functions of the SRVs can be performed. These manual actuation tests are currently performed once per operating cycle (24 months) corresponding to start-up from refueling outages. The SRV manual actuation lift test is credited for demonstrating the

RA-19-0015 Enclosure Page 4 of 9 mechanical functioning of the valve for the relief mode and for the automatic depressurization function.

BSEP Units 1 and 2 are in the fifth ten-year inservice test (IST) interval and the code of record is the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (i.e., referred to herein as the OM Code), 2004 Edition through 2006 Addenda. Appendix I, paragraph l-3410(c). of the 2004 OM Code permit the valve disk stroke capability to be verified by mechanical examination or tests. Overlapping tests can be credited to individually test SRV components. The proposed amendment revises the TSs to reflect the testing provisions of the 2004 OM Code, while retaining the current TS capability to manually actuate as an alternative in the surveillance requirements. Additionally, the Frequency of TS SRs 3.4.3.2 and 3.5.1.11 is being revised to be in accordance with the lnservice Testing Program.

In lieu of performing a manual actuation of each SRV once per cycle it is proposed in accordance with the 2004 OM Code to credit overlapping code and TS surveillance requirements (testing) to ensure the capability of the SRV to open. Manual actuation testing will however, be retained as an alternative in the TS surveillance requirements. The proposed revision to the SRs provides an alternative to the current requirement to demonstrate the capability of the relief valves to open when manually actuated during plant startup. This alternative provides another option to satisfy the surveillance requirements allowing a determination to be made that the valve is capable of being opened. Crediting of other testing and verification of electrical and pneumatic connections is in accordance with the 2004 Edition of the ASME OM Code, paragraph l-3410(d). The proposed revisions to the TS Bases describe the testing that will occur to verify the opening capability of the valve. The combination of testing the SRV actuator and solenoid valves and verifications of the capability of the SRV to open provide a complete verification of the functional capability and is in accordance with 2004 Edition of the ASME OM Code. This testing is described in more detail below.

  • The simulated automatic actuation test specified in SR 3.5.1.10 of TS 3.5.1, "ECCS - Operating," and additional surveillances associated with TS 3.3.5.1, "ECCS Instrumentation," demonstrate the ability of various logics and controls to actuate the SRVs up to the point of energizing the solenoids. These tests are currently performed once per operating cycle (24 months).
  • A solenoid valve (SOV) functional test will be performed in-situ for each SRV solenoid valve once per operating cycle (24 months). In the SOV functional test, a test rig with a pressure gauge will be connected downstream of the SOV pneumatic manifold in place of the SRV actuator. Each SOV will be energized, and pneumatic pressure at the downstream connection will be recorded and compared with pneumatic header pressure.
  • An SRV actuator functional test will be performed at an offsite test facility as part of certification testing for each SRV pilot assembly. The functional test will include testing manual mode actuation. The test will apply steam pressure to the SRV at approximately 1000 psig, pressurizing the SRV solenoid to approximately 100 psig, and energizing the SRV solenoid valve with approximately 125 VDC. Parameters such as steam inlet pressure, pilot disc motion, main disc motion, solenoid actuation signal and valve response time are recorded. The current practice of replacing all 11 SRV pilot assemblies each operating cycle (24 months) will be maintained.

RA-19-0015 Enclosure Page 5 of 9

  • SRV setpoint testing is performed using steam at the offsite test facility as part of certification testing for each SRV pilot assembly, at intervals determined in accordance with the IST Program. This test is the existing test required by TS SR 3.4.3.1. In addition to demonstrating that the SRV pilot stage will actuate on high steam pressure in the safety mode, this test overlaps with the pilot assembly actuator functional test to demonstrate that the pilot stage will actuate in the relief mode.
  • Currently, BSEP removes and tests all 11 pilot valves every 24-month refuel cycle and actuates each SRV main stage during plant startup to comply with the ASME OM Code 5-year test requirement. Coincident with this LAR, BSEP will adopt Code Case OMN-17, "Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves."

OMN-17 extends the frequency for 100 percent removal/refurbishment and as-left certification testing interval to three refueling outages or six years plus six months grace.

Extending the test interval to six years would reduce the number of SRVs removed during an individual outage to four, such that the full scope of 11 SRVs would be tested over three refuel cycles while still complying with the other Code requirement to test 20 percent of the valves within any 24-month period. Additionally, the six months grace would allow flexibility in the scheduling of certification testing to account for the variability of refuel outage dates.

10 CFR 50.55a(f) requires that the licensee's IST Program meet the requirements of the ASME OM Code. The current BSEP IST Program has been prepared to meet the requirements of the ASME OM Code, 2004 Edition through 2006 Addenda. The ASME OM Code allows a series of overlapping tests to individually test SRV components. Furthermore, the ASME OM Code, 2004 Edition, no longer requires in-situ SRV testing following maintenance. Rather, Section 1-3410(d) requires that each SRV that has been removed for maintenance or testing and reinstalled shall have the electrical and pneumatic connections verified either through mechanical/electrical inspection or test. The ASME OM Code, 2004 Edition, does not require that an SRV be tested as a unit. For example, the auxiliary actuating device can be tested independently of the main disk assembly. The test methods described above fully meet the requirements of the ASME OM Code, 2004 Edition, for safety and relief valves.

Main stage certification testing demonstrates that the main stage will open and port steam when actuated by the installed pilot stage. Once installed in the plant, improper valve functioning or blockage would arise only through assembly errors or the introduction of foreign material into the piping system. Specific SRV maintenance procedures and plant Foreign Material Exclusion procedures and practices are sufficient to ensure proper functioning and unobstructed steam flow capability without periodic actuation testing. BSEP and fleet procedures establish stringent Foreign Material Exclusion (FME) controls, including installation of protective FME covers over component and system openings. The area above and around the open pipe flanges require full FME tool and materials control. The area is roped off and lanyards are required on all tools.

The Duke administrative procedure for Foreign Material Exclusion has minimum FME requirements for the high risk systems, including pre-job briefings, use of a FME Tools and Materials Log, and the detailed instructions for retrieval of foreign material introduced into a system. BSEP has had no instance of SRV test failure due to loss of FME controls.

The proposed change revises the Frequency for performing SR 3.4.3.2 and SR 3.5.1.11 to be in accordance with the Inservice Testing Program. Specifying the required frequency through the IST Program is not a new technique and occurs throughout the TS (e.g., SR 3.4.3.1 for verifying the safety function lift setpoints of the SRVs). Performing testing in accordance with the IST

RA-19-0015 Enclosure Page 6 of 9 Program retains appropriate legal control over the testing methodology and specified frequency, since performance is required and is governed by a code adopted into the regulation, i.e., 10 CFR 50.55a. Also, future OM Code changes could then be adopted without requiring a corresponding TS change, allowing NRC endorsed code changes to be more rapidly put in place. Additionally, this will allow crediting IST Program tests performed at frequencies other than the current 24-months as established by the BSEP SFCP.

Based on the above technical evaluation, Duke Energy concludes that testing in accordance with the proposed changes to SR 3.4.3.2 and SR 3.5.1.11 will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the ASME OM Code, 2004 Edition, for safety and relief valves which the NRC staff has previously found to be acceptable.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, "Technical specifications," provides the requirements for the content required in the TSs. As stated in 10 CFR 50.36, the TSs include, among other things, Limiting Conditions for Operation (LCO) and Surveillance Requirements (SR) to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The proposed change modifies SR 3.4.3.2, in TS 3.4.3, "Safety Relief Valves (SRVs)," and SR 3.5.1.11, in TS 3.5.1, "ECCS - Operating," by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened and by establishing the frequency of this testing to be in accordance with the Inservice Testing Program. The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the ASME OM Code, 2004 Edition, for safety and relief valves.

10 CFR 50.55a(f), "Inservice testing requirements," requires, in part, that American Society of Mechanical Engineers (ASME) Class 1,2, and 3 components must meet the requirements of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), except where alternatives have been authorized pursuant to paragraphs (a)(3)(i) and (a)(3)(ii) of 10 CFR 50.55a. Currently, BSEP Unit 1 and 2 SR 3.4.3.2 and SR 3.5.1.11 require that each main steam SRV must be manually actuated during a startup once reactor steam pressure and flow are adequate to perform the test. Replacing the manual actuation SRV test method in SR 3.4.3.2 and SR 3.5.1.11 with an alternative requirement to verify that the valves are capable of being opened as determined through a series of overlapping tests will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam and that this testing fully meets the requirements of the ASME OM Code, 2004 Edition, for safety and relief valves.

4.2 Precedent The proposed change is similar to NRC approved license amendments issued to the Detroit Edison Company for Fermi 2 (Reference 1) and to Northern States Power Company -

Minnesota for Monticello (Reference 2).

RA-19-0015 Enclosure Page 7 of 9 4.3 No Significant Hazards Consideration Determination Analysis Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is requesting an amendment to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP),

Unit Nos. 1 and 2. The proposed change modifies Surveillance Requirement (SR) 3.4.3.2, in TS 3.4.3, "Safety Relief Valves (SRVs)," and SR 3.5.1.11, in TS 3.5.1, "ECCS - Operating," by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened. Additionally, the proposed change revises the Frequency for performing these SRs to be in accordance with the Inservice Testing (IST) Program.

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," 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 does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The proposed change revises SR 3.4.3.2 and SR 3.5.1.11 by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and requires the testing to be completed on a frequency in accordance with the IST Program. The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam. This testing fully meets the requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for safety and relief valves. Performing testing in accordance with the IST Program retains appropriate legal control over the testing methodology and specified frequency, since performance is required and is governed by a code adopted into the regulation, i.e., 10 CFR 50.55a. Therefore, the proposed change does not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of event. Further, the proposed change does not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures.

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

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 revises SR 3.4.3.2 and SR 3.5.1.11 by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and updating the frequency to be in accordance with the IST Program.

It does not require any modification to the plant and it does not alter the design

RA-19-0015 Enclosure Page 8 of 9 configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis.

Therefore, the proposed amendment 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?

Response: No The proposed change revises SR 3.4.3.2 and SR 3.5.1.11 by replacing the current requirement to verify the SRVs open when manually actuated with an alternate requirement that verifies the SRVs are capable of being opened through a series of overlapping tests and updating the frequency to be in accordance with the IST Program.

The proposed SR testing will continue to demonstrate proper SRV operation without the need for in-situ testing with reactor steam. It does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(3) will continue to be met.

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

Based on the above, Duke Energy 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.

4.4 Conclusion 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. 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 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.

RA-19-0015 Enclosure Page 9 of 9

6. REFERENCES
1. Letter from the U.S. Nuclear Regulatory Commission to the Detroit Edison Company, "Fermi 2 - Issuance of Amendment to Modify Technical Specification Surveillance Requirements for Safety Relief Valves," dated December 21, 2012, NRC ADAMS Accession No. ML12321A234.
2. Letter from the U.S. Nuclear Regulatory Commission to Northern States Power Company - Minnesota (NSPM), "Monticello Nuclear Generating Plant - Issuance of Amendment Re: Testing of Main Steam Safety/Relief Valves," dated July 27, 2012, NRC ADAMS Accession No. ML12185A216.

RA-19-0015 Enclosure Attachment 1 Proposed Technical Specification Changes (Mark-Up)

Unit 1

SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.3.2 ------------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required SRV opens when manually In accordance with actuated is capable of being opened. the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM Brunswick Unit 1 3.4-6 Amendment No. 276

ECCSOperating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 ---------------------------------NOTE-----------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates In accordance with on an actual or simulated automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.10 ---------------------------------NOTE-----------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated In accordance with automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.11 ---------------------------------NOTE-----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required ADS valve opens when manually In accordance with actuated is capable of being opened. the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM SR 3.5.1.12 ---------------------------------NOTE-----------------------------

Instrumentation response time may be assumed to be the design instrumentation response time.

Verify the ECCS RESPONSE TIME for each ECCS In accordance with injection/spray subsystem is within the limit. the Surveillance Frequency Control Program Brunswick Unit 1 3.5-7 Amendment No. 276

RA-19-0015 Enclosure Attachment 2 Proposed Technical Specification Changes (Mark-Up)

Unit 2

SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.3.2 -------------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required SRV opens when manually In accordance with actuated is capable of being opened. the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM Brunswick Unit 2 3.4-6 Amendment No. 304

ECCSOperating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------------------NOTE--------------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates In accordance with on an actual or simulated automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.10 -------------------------------NOTE--------------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated In accordance with automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.11 -------------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required ADS valve opens when manually In accordance with actuated is capable of being opened. the Surveillance Frequency Control Program INSERVICE TESTING PROGRAM SR 3.5.1.12 -------------------------------NOTE--------------------------------

Instrumentation response time may be assumed to be the design instrumentation response time.

Verify the ECCS RESPONSE TIME for each ECCS In accordance with injection/spray subsystem is within the limit. the Surveillance Frequency Control Program Brunswick Unit 2 3.5-7 Amendment No. 304

RA-19-0015 Enclosure Attachment 3 Revised (Typed) Technical Specification Pages Unit 1

SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.3.2 ------------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required SRV is capable of being opened. In accordance with the INSERVICE TESTING PROGRAM Brunswick Unit 1 3.4-6 Amendment No. 276

ECCSOperating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 ---------------------------------NOTE-----------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates In accordance with on an actual or simulated automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.10 ---------------------------------NOTE-----------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated In accordance with automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.11 ---------------------------------NOTE-----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required ADS valve is capable of being In accordance with opened. the INSERVICE TESTING PROGRAM SR 3.5.1.12 ---------------------------------NOTE-----------------------------

Instrumentation response time may be assumed to be the design instrumentation response time.

Verify the ECCS RESPONSE TIME for each ECCS In accordance with injection/spray subsystem is within the limit. the Surveillance Frequency Control Program Brunswick Unit 1 3.5-7 Amendment No. 276

RA-19-0015 Enclosure Attachment 4 Revised (Typed) Technical Specification Pages Unit 2

SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.3.2 -------------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required SRV is capable of being opened. In accordance with the INSERVICE TESTING PROGRAM Brunswick Unit 2 3.4-6 Amendment No. 304

ECCSOperating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------------------NOTE--------------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates In accordance with on an actual or simulated automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.10 -------------------------------NOTE--------------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated In accordance with automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.11 -------------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Verify each required ADS valve is capable of being In accordance with opened. the INSERVICE TESTING PROGRAM SR 3.5.1.12 -------------------------------NOTE--------------------------------

Instrumentation response time may be assumed to be the design instrumentation response time.

Verify the ECCS RESPONSE TIME for each ECCS In accordance with injection/spray subsystem is within the limit. the Surveillance Frequency Control Program Brunswick Unit 2 3.5-7 Amendment No. 304

RA-19-0015 Enclosure Attachment 5 Technical Specification Bases Pages (Mark-Up) - Unit 1 (For Information Only)

SRVs B 3.4.3 BASES APPLICABILITY be required to provide pressure relief to discharge energy from the core (continued) until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The SRV function is not needed during these conditions.

ACTIONS A.1 and A.2 With less than the minimum number of required SRVs OPERABLE, a transient may result in the violation of the ASME Code limits on reactor pressure. If the safety function of one or more required SRVs is inoperable, 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 MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 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 required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the required 10 SRVs will open at the pressures assumed in the safety analysis of References 1, 2, and 3. The demonstration of the SRV safety function lift settings must be performed during shutdown, since this is a bench test, in accordance with the INSERVICE TESTING PROGRAM. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

SR 3.4.3.2 This Surveillance verifies that each SRV is capable of being opened, which can be determined by either of the following two methods.

(continued)

Brunswick Unit 1 B 3.4.3-3 Revision No. 95

SRVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)

REQUIREMENTS Method 1 Valve OPERABILITY and setpoints for overpressure protection are verified in accordance with the requirements of the ASME OM Code (Ref. 5). Proper SRV function is verified through performance of inspections and overlapping tests on component assemblies, demonstrating the valve is capable of being opened. Testing is performed to demonstrate that:

  • Each SRV main stage opens and passes steam when the associated pilot stage actuates;
  • Each SRV pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized;
  • Each SRV solenoid valves ports pneumatic pressure to the associated SRV actuator when energized; and
  • Each SRV actuator stem moves when dry lift tested in-situ. With exception of the main and pilot stages this test demonstrates mechanical operation without steam.

The solenoid valves and SRV actuators are functionally tested as part of the INSERVICE TESTING PROGRAM. The SRV assembly is bench tested as part of the certification process, at intervals determined in accordance with the INSERVICE TESTING PROGRAM. Maintenance procedures ensure that the SRV is correctly installed in the plant and that the SRV and associated piping remain clear of foreign material that might obstruct valve operation or full steam flow.

Method 2 A manual actuation of each required SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required pressure is achieved to perform this test. Adequate pressure at which this test is to be performed, to avoid damaging the valve, is 945 psig. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.

Therefore, this SR is modified by a Note that states the Surveillance is not Brunswick Unit 1 B 3.4.3-4 Revision No. 94

SRVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)

REQUIREMENTS required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program INSERVICE TEST PROGRAM. Industry operating experience has shown that these components usually pass the SR when performed at the Code required Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 5.2.2.2.

2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Supplement 1, March 1996.
3. UFSAR, Chapter 15.
4. 10 CFR 50.36(c)(2)(ii).
5. ASME Operation and Maintenance (OM) Code.

Brunswick Unit 1 B 3.4.3-5 Revision No. 94

ECCSOperating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 (continued)

REQUIREMENTS This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

SR 3.5.1.10 The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,

solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that excludes valve actuation since the valves are individually tested in accordance with SR 3.5.1.11. This also prevents an RPV pressure blowdown.

SR 3.5.1.11 This Surveillance verifies that each ADS valve is capable of being opened, which can be determined by either of the following two methods.

Method 1 Valve OPERABILITY and setpoints for overpressure protection are verified in accordance with the requirements of the ASME OM Code (Ref. 16). Proper ADS valve function is verified through performance of inspections and overlapping tests on component assemblies, demonstrating the valve is capable of being opened. Testing is performed to demonstrate that:

  • Each ADS valve main stage opens and passes steam when the associated pilot stage actuates;
  • Each ADS valve pilot stage actuates to open the associated main stage when the pneumatic actuator is pressurized;
  • Each ADS valve solenoid valves ports pneumatic pressure to the associated SRV actuator when energized; and (continued)

Brunswick Unit 1 B 3.5.1-15 Revision No. 94

ECCSOperating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 REQUIREMENTS

  • Each ADS valve actuator stem moves when dry lift tested in-situ.

With exception of the main and pilot stages this test demonstrates mechanical operation without steam.

The solenoid valves and SRV actuators are functionally tested as part of the INSERVICE TESTING PROGRAM. The SRV assembly is bench tested as part of the certification process, at intervals determined in accordance with the INSERVICE TESTING PROGRAM. Maintenance procedures ensure that the SRV is correctly installed in the plant and that the SRV and associated piping remain clear of foreign material that might obstruct valve operation or full steam flow.

Method 2 A manual actuation of each required ADS valve is performed to verify that the valve and solenoid are functioning properly and that no blockage exists in the SRV discharge lines. This is demonstrated by the response of the turbine control or bypass valve; by a change in the measured flow; or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required pressure is achieved to perform this SR. Adequate pressure at which this SR is to be performed, to avoid damaging the valve, is 945 psig. Reactor startup is allowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.10 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program INSERVICE TEST PROGRAM. Industry operating experience has shown that these components usually pass the SR when performed at the Code required Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.5.1-16 Revision No. 94

ECCSOperating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 REQUIREMENTS (continued) This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS RESPONSE TIME testing. This exception is allowed since the ECCS instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g.,

sufficient margin exists in the emergency diesel generator start time when compared to the instrumentation response time) (Ref. 14).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. UFSAR, Section 6.3.2.2.3.

2. UFSAR, Section 6.3.2.2.4.
3. UFSAR, Section 6.3.2.2.1.
4. UFSAR, Section 6.3.2.2.2.
5. UFSAR, Section 15.2.
6. UFSAR, Section 15.6.
7. 10 CFR 50, Appendix K.
8. UFSAR, Section 6.3.3.
9. 10 CFR 50.46.
10. (Deleted.)
11. 10 CFR 50.36(c)(2)(ii).
12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),

Recommended Interim Revisions to LCOs for ECCS Components, December 1, 1975.

13. UFSAR, Section 6.3.3.7.
14. NEDO-32291-A, System Analyses for the Elimination of Selected Response Time Testing Requirements, October 1995.
15. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.
16. ASME Operation and Maintenance (OM) Code.

Brunswick Unit 1 B 3.5.1-17 Revision No. 97

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