ML16252A220

Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing

ML16252A220
Person / Time
Site:	Brunswick
Issue date:	08/29/2016
From:	William Gideon Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TSC-2016-05, TSTF-545, Rev 3
Download: ML16252A220 (76)

Text

e/_~DUKE William R. Gideon Vice President

~ ENERGY Brunswick Nuclear Plant P .0. Box 10429 Southport, NC 28461 o: 910.457.3698 10 CFR 50.90 AUG 2 9 2016 Serial: BSEP 16-0035 TSC-2016-05 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 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

Pursuant to 1O CFR 50.90, Duke Energy Progress, Inc., (Duke Energy), is submitting a request for an amendment to the Brunswick Steam Electric Plant (BSEP) Technical Specifications (TS) for Units 1 and 2. The proposed change revises the TS to eliminate the Section 5.5, "lnservice Testing Program." A new defined term, "lnservice Testing Program," is added to the TS Definitions section. This request is consistent with TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing." (ADAMS accession number ML15314A305)

Pursuant to 1O CFR 50.55a(z), the application also proposes an alternative to the testing frequencies in the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, by adoption of Code Case OMN-20, "lnservice Test Frquency," for the current 10-year lnservice Testing (IST) interval.

Enclosure 1 provides a description and assessment of the proposed TS changes. Enclosure 2 provides the existing TS pages marked up to show the proposed changes. Enclosure 3 provides revised (clean) TS pages. Enclosure 4 provides BSEP Unit 1 TS Bases pages marked up to show the associated TS Bases changes and is provided for information only. Enclosure 5 provides a description and assessment of the proposed alternative to the ASME code.

Approval of the proposed amendment and relief request is requested by March 1, 2017. Once approved, the amendment shall be implemented within 90 days.

• In accordance with 1O CFR 50.91, a copy of this application, with enclosures, is being provided to the designated North Carolina Official.

U.S. Nuclear Regulatory Commission Page 2 of 3 If you should have any questions regarding this submittal, please contact Mr. Lee Grzeck, Manager- Regulatory Affairs, at (910) 457-2487.

I decl~under penalty of perjury, that the foregoing is true and correct. Executed on

• . '2."" , 2016.

Si~

William R. Gideon SWR/swr

Enclosures:

1. Description and Assessment of Technical Specifications Changes

2. Proposed Technical Specification Changes (Mark-Up), BSEP Units 1 and 2

3. Revised Technical Specification Pages, BSEP Units 1 and 2

4. Proposed Technical Specification Bases Changes (Mark-Up), BSEP Unit 1 - Information Only

5. Description and Assessment of the Proposed Alternative to the ASME Code I

U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with enclosures):

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: Ms. Michelle P. Gatts, NRC Senior Resident Inspector 84 70 River Road Southport, NC 28461-8869 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission 4325 Mail Service Center Raleigh, NC 27699-4300 Mr. W. Lee Cox, Ill, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, N'c 27699-1645 lee.cox@dhhs.nc.gov Chair - North Carolina Utilities Commission (Electronic Copy Only) 432 S Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net

BSEP 16-0035 Enclosure 1 Description and Assessment of Technical Specifications Changes

Enclosure 1 DESCRIPTION AND ASSESSMENT OF TECHNICAL SPECIFICATIONS CHANGES

1.0 DESCRIPTION

The proposed change eliminates the Technical Specifications {TS), Section 5.5, "lnservice Test (IST) Program," to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code),

Case OMN-20, "lnservice Test Frequency." A new defined term, "lnservice Testing Program,"

is added to TS Section 1.1, "Definitions." The proposed change to the TS is consistent with TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to_Section 5.5 Testing."

2.1 ASSESSMENT 2.2 Applicability of Published Safety Evaluation Duke Energy Progress, Inc., (Duke Energy) has reviewed the model safety evaluation provided in NRG letter to the Technical Specifications Task Force, "Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-545, Revision 3, 'TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing,' (TAC No.

MF3349)," (ADAMS Accession Nos. ML15314A365 and ML15314A305). This review included a review of the NRG staff's evaluation, as well as the information provided in TSTF-545. Duke Energy concluded that the justifications presented in TSlF-545 and the model safety evaluation prepared by the NRG staff are applicable to Brunswick Steam Electric Plant (BSEP) Units 1 and 2, and justify this amendment for the incorporation of the changes to the BSEP Unit 1 and Unit 2 TSs.

BSEP Unit 1 was issued a construction permit on February 7, 1970, and the provisions of 1O CFR 50.55a(f)(1) are applicable.

BSEP Unit 2 was issued a construction permit on February 7, 1970, and the provisions of 1O CFR 50.55a(f)(1) are applicable.

2.3 Variations BSEP is proposing the following variations from the TS changes described in the TSTF-545 or the applicable parts of the NRG staff's model safety evaluation dated December 11, 2015. These variations do not affect the applicability of TSTF-545 or the NRG staff's model safety evaluation to the proposed license amendment.

'

1. The generic BWR-4 specification, on page 3.0-2, renumbers TS number 5.5.12 to 5.'5.11.

Duke Energy will not incorporate this change because the IST specification will be deleted without renumbering the subsequent specifications.

2. The generic BWR-4 specification, on page 3.1.7-4, numbers the pump flow surveillance as SR 3.1.7.7. This surveillance is numbered 3.1.7.6 in the BSEP TS.

3. The generic BWR-4 specification, on page 3.4.5-2, proposes a change to the specification for RCS PIV Leakage. Duke Energy will not incorporate this change because the BSEP TSs do not have this Limiting Condition for Operation (LCO).

4. The generic BWR-4 specification, on page 3.5.1-5, proposes a change in Emergency Core Cooling System (ECCS) pump, surveillance frequency to agree with the IST program. Duke Energy will not incorporate this change because BSEP has already committed to a 92-day surveillance frequency.

5. The generic BWR-4 specification, on page 3.5.2-4, proposes a change in ECCS pump surveillance frequency to agree with the IST program. Duke Energy will not incorporate this change because BSEP has already committed to a 92-day surveillance frequency.

6. The generic BWR-4 specification, on page 3.6.1.3-10, numbers the Primary Containment Isolation Valve (PCIV) isolation time surveillance as SR 3.6.1.3.6. This* surveillance is numbered 3.6.1.3.4 in the BSEP TS.

7. The generic BWR-4 specification, on page 3.6.1.3-11, numbers the Main Steam Line Isolation Valve (MSIV) isolation time surveillance as SR 3.6.1.3.8. This surveillance is numbered 3.6.1.3.5 in the BSEP TS.

8. The generic BWR-4 specification has no entry for removing and testing the explosive squib valves for the Traversing lncore Probe {TIP) system. The BSEP specifications address this requirement in TS surveillance requirement 3.6.1.3.8.

9. The generic BWR-4 specification, on page 3.6.2.3-2, proposes a change in Residual Heat Removal (RHR) pump surveillance frequency to agree with the IST program. Duke Energy will not incorporate this change because BSEP has already committed to a 92-day surveillance frequency.

10. The generic BWR-4 specification, on page 3.6.2.4-2, proposes a change in RHR pump surveillance frequency to agree with the IST program. Duke Energy will not incorporate this change because the BSEP TS does not have an LCO for the RHR suppression pool spray mode.

11. The generic BWR-4 specification, on page 3.6.4.2-4, proposes a change in Secondary Containment Isolation Valve (SCIV) surveillance frequency for SCIV isolation to agree with the IST program. Duke Energy will not incorporate this change because the BSEP TS identifies secondary containment isolation dampers, not valves, and BSEP is already committed to a 24-month surveillance frequency.

12. The generic BWR-4 specification, on page 5.5-5, deletes the section for the lnservice Testing Program. In tpe BSEP TS, the "lnservice Testing Program" is; numbered 5.5.6. The section number will rernain unchanged, and the text will be replaced with "Deleted." This is an editorial decision to avoid changing many Duke Energy internal documents to update the TS reference.

13. The generic BWR-4 specification, on pages 5.5-5 through 5.5-19, renumbers several administrative specifications. Duke Energy will not incorporate the renumbering changes for the reasons stated in the following table:

Page 2

Explanation of Numbering Changes Not Incorporated Generic Specification Specification Proposed TSTF Change Reason for Not Page No. Incorporating 5.5-5 Ventilation Filter Renumber step Editorial; previous Testing Program 5.5.8 to 5.5.7 step will not be renumbered.

5.5-7 Explosive Gas and Renumber step Editorial; previous Storage Tank 5.5.9 to 5.5.8 step will not be Radioactivity renumbered.

Monitoring Program 5.5-8 Diesel Fuel Oil Renumber step Editorial; previous Testing Program 5.5.10 to 5.5.9 step will not be renumbered.

5.5-9 Technical Renumber step Editorial; previous Specifications (TS) 5.5.11to5.5.10 step will not be Bases Control renumbered.

Program 5.5-9 Safety Function Renumber step Editorial; previous Determination 5.5.12 to 5.5.11 step will not be Program renumbered.

5.5-10 Primary Renumber step Editorial; previous Containment 5.5.13 to 5.5.12 step will not be Leakage Rate renumbered.

Testino Prooram 5.5-13 Battery Monitoring Renumber step This program does and Maintenance 5.5.14 to 5.5.13 not exist in current Program BSEPTS.

5.5-15 Control Room Renumber step Editorial; previous Envelope (CRE) 5.5.15 to 5.5.14 step will not be Habitability renumbered.

Proo ram 5.5-16 Setpoint Control Renumber step This program does Program 5.5.16 to 5.5.15 not exist in current BSEP TS.

5.5-18 Surveillance Renumber step This program does Frequency Control 5.5.17 to 5.5.16 not exist in current Program BSEPTS.

Page 3

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Duke Energy requests adoption of the Technical Specification (TS) changes described in TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the BSEP Unit 1 and Unit 2 TS. The proposed change revises the TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals," to delete the "lnservice Testing (IST) Program" specification. Requirements in the IST Program are removed, as they are duplicative of requirements in the American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code, as clarified by Code Case OMN-20, "lnservice Test Frequency." Other requirements in Section 5.5 are eliminated because the Nuclear Regulatory-Commission (NRG) has determined their appearance in the TS is*

contrary to regulations. A new defined term, "lnservice Testing Program," is added, which references the requirements of Title 1O of the Code of Federal Regulations (1 O CFR), Part 50, paragraph 50.55a(f). Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 1O 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 revises TS Chapter 5, "Administrative Controls," Section 5.5, .

"Programs and Manuals," by eliminating the "lnservice Testing Program" specification.

Most requirements in the lnservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, "lnservice Test Frequency." The remaining requirements in the Section 5.5 IST Program are eliminated because the NRG has determined their inclusion in the TS is contrary to regulations. A new defined term, "lnservice Testing Program," is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. lnservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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

Page 4

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

Response: No The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged.

However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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

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

Response: No The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 1O CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS SR 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consid~ration under the standards set forth in 10 CFR.50.92(c), and, accordingly, a finding of "rl'o significant hazards consideration" is justified ..-

, '

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1O CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in Page 5

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 change meets the eligibility criterion for categorical exclusion set forth in 1O CFR 51.22(c)(9).

Therefore, pursuant to 1O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Page 6

BSEP 16-0035 Enclosure 2 Proposed Technical Specification Changes (Mark-Up)

BSEP Units 1 and 2

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Submersion , and Ingestion ," 1989 and FGR 12, "External (continued) Exposure to Radionuclides in Air, Water, and Soil ," 1993.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval from SYSTEM (ECCS) RESPONSE when the monitored parameter exceeds its ECCS initiation TIME setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e ., the valves travel to their required positions, pump discharge pressures reach their requ ired values , etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential , overlapping , or total steps so that the entire response time is measured .

ISOLATION INSTRUMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e.g .,

de-energization of the MSIV solenoids) . The response time may be measured by means of any series of sequential ,

overlapping , or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell , such as that from pump seals or valve packing , that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE ;

(continued)

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee program that PROGRAM fulfills the requirements of 10 CFR 50.55a(f).

Brunswick Unit 1 1.1-3 Amendme nt No. 221 I

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate ~ 41 .2 gpm at a In accordance with discharge pressure ~ 1190 psig . the IRsel"t'iee TestiR ~ Pre ~ raffi SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24 months on a into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.8 Verify sodium pentaborate enrichment is ~ 47 atom Prior to addition to percent B-1 O. SLC tank INSERVICE TESTING PROGRAM Brunswick Unit 1 3.1-22 Amendment No. 227

SRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4 .3 Safety/Relief Valves (SRVs)

LCO 3.4 .3 The safety function of 10 SRVs shall be OPERABLE.

APPLICABILITY : MODES 1, 2, and 3.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required SRVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable .

AND A .2 Be in MODE 4. 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 3.4 .3.1 Verify the safety function lift setpoints of the required In accordance with 1O SRVs are as follows: the IAseFViee Testi A ~ PFe ~ FBffi Number of Setpoint SRVs .(Q_§_jgl INSERVICE TESTING PROGRAM 4 1130 +/- 33 .9 4 1140 +/- 34.2 3 1150 +/- 34 .5 (continued)

Brunswick Unit 1 3.4-5 Amendment No. 203

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1 .3.2 ------------------------------N()TES-------------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary M()DE 2 or 3 from containment and not locked , sealed , or otherwise M()DE 4 if primary secured and is required to be closed during accident containment was conditions, is closed . de-inerted while in M()DE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) 31 days shear isolation valve explosive charge.

SR 3.6 .1.3.4 Verify the isolation time of each power operated and In accordance each automatic PCIV, except for MS IVs, is within with the IFtseFYiee limits. Testi Ft ~ Pre ~ reFfl

~

SR 3.6.1 .3.5 Verify the isolation time of each MSIV is ~ 3 seconds In accordance with and :=; 5 seconds. the IFtserviee Testi Ft ~ Pre ~ reFfl

)

( (continued) j~ INSERVICE TESTING PR()GRAM INSERVICE TESTING PR()GRAM Brunswick Unit 1 3.6-12 Amendment No. 203

PC IVs 3 .6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 24 months position on an actual or simulated isolation signal.

SR 3.6.1.3.7 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3 .8 Remove and test the explosive squib from each shear In accordance with isolation valve of the TIP System . the IAsef'ltliee Testi A!jl Pre!j! FSA'I Ill"'~

SR 3.6.1.3.9 Verify leakage rate through each main steam line is In accordance with

~ 100 scfh and the combined leakage rate of all four the Primary main steam lines is < 150 scfh when tested at / Containment

~ 25 ps1g . Leakage Rate Testing Program I

L INSERVICE TESTING PROGRAM Brunswick Unit 1 3.6-13 Amendment No. 239

Programs and Man uals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

h. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following :

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin , and

2. For iodine-131 , iodine-133 , tritium , and for all radionuclides in particulate form with half lives > 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ ;

i. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual and quarterly doses to a member of the public from iodine-131 , iodine-133 , tritium , and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 1O CFR 50, Appendix I; and

k. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uran ium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR Table 5.3.3-2 , cyclic and transient occurrences to ensure that components are maintained with in the design limits.

5.5.6 IAsef\*iee TestiAa Proijrar:i::i ( Deleted J Tl'lis 19re§f8ffi 19reviaes eeAtrels fer iAseFViee testiA§ ef /\SM~ Gase Glass 1, 2, afla 3 J9tlffiJ9S aAs va lves.

a. Testifl§ Freeit1eAeies 01919lieasle te tl'le /\SM~ Gose for 013erati0As aAs MaiRteAaAee ef ~~t1elear Pewer PlaAts (ASM~ OM Gose) aAel a1313lisal3lo

/\88eA8a are as follows:

(contin ued)

Brunswick Un it 1 5.0-9 Amendment No. 247 I

Programs and Manuals 5.5 5.5 Programs and Manuals 6.6.6 IFtsef'tiee TestiAa PrearaFR ~seAtiAl.leel)

ASME OM Geele aFtel a1:31:3lieal3le R:e~1:1ireel Fre~1:1eF1eies 14<eleleF1ela terFRiFlele§y fer fer 1:3el'ferFRiF1§ iFtsef'tiee iFlsef'i'iee testiFI§ aetivities testiFI§ aetivities

'A'eelcly At least eFtee 1:3er 7 elays At least eFtee 1:3er 61 elays Ottefterly er every At least eFtee 1:3er 92 elays 6 FReAtAS eefl'liaF1F11:1ally er At least eFtee 1:3er 184 elays every e RXleAtRe

~\'0F)' Q RX18AtAs At least eAee 1:3er 27e elays Yearly er 8F1F11:1ally At least eFtee 1:3er 6ee elays 8ieF1F1ially er every l\t least eFtee 1:3er 761 elays 2 years

13. Tlote 1:3revisieF1s ef SR 6.0 .2 are a1:31:3lieaele te tlote aeeve re~1:1ireel Fre~1:1eAeies aAel te etloter AerFRal aAel aeselerateel Fre~l.leAsies Sf)esifieel as 2 years er less iFI tlote IFtsef'tiee TestiFI§ Pre§FBFR fer f)erferfl'liA§ iAsef'i'ise testiA~ astivities;

s. Tlote fJFevisieAs sf eR 6.0 .6 are Bflfllisal31e ta iAsef'tise testiA~ astivities;

-aR&-

el . ~~et'"1iF1§ iFI tlote 10.SME OM Geele slotall ee eeF1str1:1eel te s1:11:3erseele tlote re~l.lireFReAts ef aAy Te .

5.5.7 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months ; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber filter bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing ; and , following significant painting , fire , or chemical release in any ventilation zone communicating with the system .

(continued)

Brunswick Unit 1 5.0-10 Amendment No. 247

Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air inleakage into the CRE . These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis .

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

Brunswick Unit 1 ( 6.9 178 ) Amendment No. 248 I

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Submersion , and Ingestion," 1989 and FGR 12, "External (continued) Exposure to Radionuclides in Air, Water, and Soil ," 1993.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM(ECCS)RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential , overlapping ,

or total steps so that the entire response time is measured .

ISOLATION INSTRUMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e .g.,

de-energization of the MSIV solenoids) . The response time may be measured by means of any series of sequential ,

overlapping , or total steps so that the entire response time is measured .

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell , such as that from pump seals or valve packing , that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE ;

(continued)

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee program that PROGRAM fulfills the requirements of 10 CFR 50.55a(f) .

Brunswick Unit 2 1.1-3 Amendment No . 246

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate ~ 41 .2 gpm at a In accordance with discharge pressure ~ 1190 psig the IAseF¥iee

+es~i A ~ 1=2 Fe ~ Fe~

SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24 months on a into reactor pressure vessel STAGGERED TEST BASIS SR 3.1.7.8 Verify sodium pentaborate enrichment is~ 47 atom Prior to addition to percent B-1 O. SLC tank INSERVICE TESTING PROGRAM Brunswick Unit 2 3.1-22 Amendment No . 255 I

SR Vs 3.4 .3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs)

LCO 3.4 .3 The safety function of 10 SRVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required SRVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND A.2 Be in MODE 4 . 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 3.4.3.1 Verify the safety function lift setpoints of the required In accordance with 1O SRVs are as follows: the IAseFYiee TestiA!jl Pre !jj rs~

Number of Setpoint SRVs .{Q_filg)_ INSERVICE TESTING PROGRAM 4 1130 +/- 33.9 4 1140 +/- 34.2 3 1150 +/- 34.5 (continued)

Brunswick Unit 2 3.4-5 Amendment No. 233

PC IVs 3.6.1 .3 s URVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.2 -------------------------------N0 TES------------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PC IVs that are open under administrative controls.

---

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed , or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6 .1.3.3 Verify continuity of the traversing incore probe (TIP} 31 days shear isolation valve explosive charge.

SR 3.6.1.3.4 Verify the isolation time of each power operated and In accordance each automatic PCIV, except for MSIVs, is within with the h~seFYiee limits. TestiR~ Pre~reffl

~

-

SR 3.6.1 .3.5 Verify the isolation time of each MSIV is ~ 3 seconds In accordance with and ::; 5 seconds. the IRSeFYiee

,J TestiR~ Pre~reffl _ _

-

(continued)

"II I

'-..... INSERVICE TESTING PROGRAM INSERVICE TESTING PROGRAM Brunswick Unit 2 3.6-12 Amendment No. 233

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1 .3.6 Verify each automatic PCIV actuates to the isolation 24 months position on an actual or simulated isolation signal.

SR 3.6.1.3.7 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3.8 Remove and test the explosive squib from each shear In accordance with isolation valve of the TIP System. the IAseF¥iee Tes~iA~ Pre~rsFR SR 3.6.1.3.9 Verify leakage rate through each main steam line is In accordance with s 100 scfh and the combined leakage rate of all four the Primary main steam lines is s 150 scfh when tested at Containment

25 psig . Leakage Rate Testing Program INSERVICE TESTING PROGRAM Brunswick Unit 2 3.6-13 Amendment No. 267

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

h. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following :

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin , and

2. For iodine-131 , iodine-133, tritium , and for all radionuclides in particulate form with half lives > 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ ;

i. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary , conforming to 10 CFR 50 , Appendix I;

j. Limitations on the annual and quarterly doses to a member of the public from iodine-131 , iodine-133, tritium , and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

k. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources , conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR Table 5.3.3-2, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 IAseFViee TestiRa PFoaFaFR (Deleted J H1is pro~raffi provieles eoRtrols for iRservioe testiR§ of /\~Me Ceele Class 1, 2, 8R6 3 J31:JFAJ3S aAet va lves.

a. TestiR~ FFCEjl:leReies applioal31e to tl=ie A~M~ Ceele fer 013eratieRs aREI MaiRteRBRee of ~Jl:lelear Power PlaRts (A~M~ OM Ceele) anEI a1313lisaele J\eleleRela are as follows:

(continued)

Brunswick Unit 2 5.0-9 Amendment No. 275 I

Programs and Manuals 5.5 5.5 Programs and Manuals e.e.e IRSBFVise TestiRE! PF0§!F9~ (seritiriwerol)

AelVI~ OIVI Geele BREI e1313liee8le Reeii.1iFeel rF0Efi.JeReies AeleleRela teFR'liRele~y feF feF 13eFf8FR'liR~ iRSSFViee iRseFViee tostiR~ eetivities testiR~ aetivities

\A,teekly At least eRse 130F 7 elays IVleRtRly At least eRse 130F d1 elays QwaFterly er ovoFy At least erise 13eF Q2 Elays eeR'liBRFli.JBlly SF At least ORSO 13er 184 Elays e¥OF)' e ~0RtRS

~Vel)' Q R'l0RtRS At least eRee 13eF 27e Eleys YeaFly BF BRRi.Jally At least eRee 13eF dee Eleys BieAAially er evePf At least eRee 13eF 7d1 eleys 2 yoaFs B. Tl"le f3F0V iSi8RS ef eR d.Q.2 BFe e1313li0e8le te tl"le BB8'18 F8Ef i.JiFeel rFeeii.10Rsios aREI te etl"leF RBFR'lal BREI asseleFeteEI rFeeii.10Reios s13esifieEI as 2 yeaFs er less iR tl"le IRseFVise TestiR~ PFe~FaR'l feF 13erieFR'liR~

iRseFViee testiR~ eetivities;

e. Tl"le 13Fevisi0Rs ef eR d.Q.d are e1313liee8le te iRseFViee testi R~ aeti'1ities;

-aR&-

el . ~'etl"liR~ iR tl"le AelVI~ OIVI GeEle sl"lall 80 e0Rstri.10EI te si.1130Fs0Ele tl"le reE1i.1iro~erits ef ariy Te .

5.5.7 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems .

Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber filter bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing ; and , following significant painting , fire , or chemical release in any ventilation zone communicating with the system .

(continued)

Brunswick Unit 2 5.0-10 Amendment No . 275 I

Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis .

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

Brunswi ck Unit 2 ( e.Q He J Amendment No . 276 I

BSEP 16-0035 Enclosure 3 Revised Technical Specification Pages BSEP Units 1 and 2

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Submersion, and Ingestion," 1989 and FGR 12, "External (continued) Exposure to Radionuclides in Air, Water, and Soil," 1993.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval from SYSTEM (ECCS) RESPONSE when the monitored parameter exceeds its ECCS initiation TIME setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, OV!3rlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 1O CFR 50.55a(f).

ISOLATION INSTRUMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e.g.,

de-energization of the MSIV solenoids). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywall, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

,

(continued)

Brunswick Unit 1 1.1-3 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate ~ 41.2 gpm at a In accordance with discharge pressure ~ 1190 psig. the INSERVICE TESTING PROGRAM SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24 months on a into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.8 Verify sodium pentaborate enrichment is~ 47 atom Prior to addition to percent B-10. SLC tank Brunswick Unit 1 3.1-22 Amendment No.

SRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs)

LCO 3.4.3 The safety function of 1O SRVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required SRVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND A.2 Be in MODE4. *3e hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the required In accordance with 10 SRVs are as follows: the INSERVICE TESTING Number of Setpoint PROGRAM SRVs !J2§.igl 4 1130 +/- 33.9 4 1140 +/- 34.2 3 1150 +/- 34.5 (continued)

Brunswick Unit 1 3.4-5 Amendment No.

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.2 ------------------------------NOTES-------------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing in core* probe (Tl P) 31 days shear isolation valve explosive charge.

SR 3.6.1.3.4 Verify the isolation time of each power operated and In accordance with each automatic PCIV, except for MS IVs, is within the INSERVICE limits. TESTING PROGRAM ,_

SR 3.6.1.3.5 Verify the isolation time of each MSIV is ::::: 3 seconds In accordance with and ~ 5 seconds. the INSERVICE TESTING PROGRAM (continued)

Brunswick Unit 1 3.6-12 Amendment No.

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 24 months position on an actual or simulated isolation signal.

SR 3.6.1.3. 7 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3.8 Remove and test the explosive squib from each shear In accordance with isolation valve of the TIP System. the INSERVICE TESTING PROGRAM SR 3.6.1.3.9 Verify leakage rate through each main steam line is In accordance with

100 scfh and the combined leakage rate of all four the Primary main steam lines is

;; 150 scfh when tested at Containment

o: 25 psig. Leakage Rate Testing Program Brunswick Unit 1 3.6-13 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

h. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following:

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ;

i. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents reieased from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

k. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR Table 5.3.3-2, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Deleted (continued)

Brunswick Unit 1 5.0-9 Amendment No

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber filter bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and, followin_g significant painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.7.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.7.d and 5.5.7.e shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass< 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 1, Positions C.5.a and C.5.c, and ANSI N510-1975 at the system flowrate specified below:

ESF Ventilation System Flowrate (cfm)

Standby Gas Treatment (SGT) System 2700 to 3300 Control Room Emergency Ventilation 1800 to 2200 (GREV) System (continued)

Brunswick Unit 1 5.0-10 Amendment No

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program CVFTP) (continued)

b. Demonstrate for each of the ESF systems that an in place test of the charcoal adsorber shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 1, Positions G.5.a and G.5.d, and ANSI N510-1975 at the system flowrate specified below:

ESF Ventilation *System Flowrate (cfm)

SGT System 2700 to 3300 GREV System 1800 to 2200

c. 1) Demonstrate for the SGT System that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 1, Position G.6.b, and tested in accordance with ASTM D3803-1989, at a temperature of 30°G, a face velocity of 61 fpm, and a relative humidity of 70% within the tolerances provided in Table 1 of ASTM D3803-1989, shows the methyl iodide penetration < 0.5%.

2) Demonstrate for the GREV System that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 1, Position G.6.b, and tested in accordance with ASTM D3803-1989, at a temperature of 30°G and a relative humidity of 95% within the temperature and humidity tolerances provided in Table 1 of ASTM D3803-1989, meets the acceptance criteria of < 5.0% penetration of methyl -

iodide.

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilter (SGT only), and the charcoal adsorbers is less than or equal to the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P (inches wg) Flowrate (cfm)

SGT System 8.5 2700 to 3300 GREV System 5.25 1800 to 2200 (continued)

Brunswick Unit 1 5.0-11 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program NFTP) (continued)

e. Demonstrate that the heaters for each of the SGT subsystems dissipate :::::

16.67 kW under a degraded voltage condition when tested in accordance with ANSI N510-1975.

5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the Main Condenser Offgas Treatment System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each outdoor liquid radwaste tank that is not surrounded by liners, dikes, or walls, capable of holding the tank's contents and that does not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is ::;; 1O Curies, excluding tritium and dissolved or entrained gases.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

5.5.9 Diesel Fuel Oil Testing Program A diesel fuel oil testing program shall establish required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM

.Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has not become contaminated with other products during transit, thus altering the quality of the fuel oil; (continued)

Brunswick Unit 1 5.0-12 Amendment No.

Programs and Manuals 5.5

. 5.5 Programs and Manuals 5.5.9 Diesel Fuel Oil Testing Program (continued)

b. Kinematic viscosity is within limits for ASTM 2-D fuel oil when tested every 92 days; and

c. Total particulate concentration of the fuel oil is ::; 10 mg/I when tested every 31 days in accordance with the applicable ASTM Standard. .

.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.

5.5.10 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 1O CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.1 O.b.1 or 5.5.1 O.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

(continued)

Brunswick Unit 1 5.0-13 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

a. The SFDP shall contain the following:

1. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; 2.. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4.

• Other appropriate limitations and remedial or compensatory actions.

b. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed .in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may

• exist when a support system is inoperable, and: *

1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

2. A required system redundant to system(s) in tum supported by the inoperable supported system is also inoperable; or

3. A required system redundant to support system(s) for the supported systems described in b.1 and b.2 above is also inoperable.

(continued)

Brunswick Unit 1 5.0-14 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDPl (continued)

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.12 Primary Containment Leakage Rate Testing Program A primary containment leakage rate testing program shall establish requirements to implem-ent the leakage rate testing of the primary containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option Bas modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995, as modified by the following exceptions:

a. The visual examination of concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

b. The visual examination of the metallic shell, penetrations, and appurtenances intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where ~elief has been authorized by the NRC.

c. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at Pa as specified in Nuclear Energy Institute Guideline 94-01, Revision O;

d. Reduced duration Type A tests may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

e. Performance of Type C leak rate testing of the hydrogen and oxygen monitor isolation valves is not required; and (continLed)

Brunswick Unit 1 5.0-15 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

f. Performance of Type C leak rate testing of the main steam isolation valves at a pressure less than Pa instead of leak rate testing at Pa as specified in ANSI/ANS 56.8-1994.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 49 psig.

The maximum allowable primary containment leakage rate, La, shall be 0.5% of primary containment air weight per day at Pa*

Leakage rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is :s; 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for Type B and C tests and

s; b.75 La for Type A tests.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is :s; 0.05 La when tested at~ Pa.

2) For each air lock door, leakage rate is :s; 5 scfh when the gap between the door seals is pressurized to ~ 10 psig.

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

5.5.13 Control Room Envelope Habit_ability 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 Emergency Ventilation (GREV) System, 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 occupancy of the CRE under design basis accident (OBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the :duration of the accident. The program shall include the following elements:

(continued)

Brunswick Unit 1 5.0-16 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued)

a. The definition of the CRE and the CRE boundary.

b. Requirements for marntaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods ~nd at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the GREV System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and

,measuring CRE pressure and assessing the CRE boundary as required

• by paragraphs c and d, respectively.

Brunswick Unit 1 5.0-17 Amendment No.

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 Submersion, and Ingestion," 1989 and FGR 12, "External (continued) Exposure to Radionuclides in Air, Water, and Soil," 1993.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM(ECCS)RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pres~ures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

ISOLATION INSTRWMENTATION The ISOLATION INSTRUMENTATION RESPONSE TIME RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves receive the isolation signal (e.g.,

de-energization of the MSIV solenoids). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and*

conducted to a sump or collecting tank; or

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; (continued)

Brunswick Unit 2 1.1-3 Amendment No.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate;;::: 41.2 gpm at a In accordance with discharge pressure ;;::: 1190 psig the INSERVICE TESTING*

PROGRAM SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24 months on a into reactor pressure vessel STAGGERED TEST BASIS SR 3.1.7.8 Verify sodium pentaborate enrichment is~ 47 atom Prior to addition to percent 8-10. SLC tank Brunswick Unit 2 3.1-22 Amendment No.

SRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (SRVs)

LCO 3.4.3 The safety function of 10 SRVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required SRVs A.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND A.2 Be in MODE 4. 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 3.4.3.1 Verify the safety function lift setpoints of the required In accordance with 10 SRVs are as follows: the INSERVICE TESTING Number of Setpoint PROGRAM SRVs .(Q§!g).

4 1130 +/- 33.9 4 1140 +/- 34.2 3 1150 +/- 34.5 (continued)

Brunswick Unit 2 3.4-5 Amendment No.

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.2 -------------------------------NOTES------------------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) 31 days shear isolation valve explosive charge.

SR 3.6.1.3.4 Verify the isolation time of each power operated and In accordance each automatic PCIV, except for MS IVs, is within with the limits. INSERVICE TESTING PROGRAM SR 3.6.1.3.5 Verify the isolation time of each MSIV is;;::: 3 seconds In accordance with and ~ 5 seconds. the INSERVICE TESTING PROGRAM (continued)

Brunswick Unit 2 3.6-12 Amendment No.

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 24 months position on an actual or simulated isolation signal.

SR 3.6.1.3.7 Verify a representative sample of reactor 24 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.

SR 3.6.1.3.8 Remove and test the explosive squib from each shear In accordance with

. isolation valve of the TIP System. the INSERVICE TESTING PROGRAM SR 3.6.1.3.9 Verify leakage rate through each main steam line is In accordance with

~ 100 scfh and the combined leakage rate of all four the Primary main steam lines is ~ 150 scfh when tested at Containment

~ 25 psig. Leakage Rate Testing Program Brunswick Unit 2 3.6-13 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

h. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be limited to the following:

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ;

i. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

k. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the UFSAR Table 5.3.3-2, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 Deleted (continued)

Brunswick Unit 2 5.0-9 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.

Tests described in Specification 5.5.7.a and 5.5.7.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber filter bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the system. -

Tests described in Specification 5.5.7.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and, following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.7.d and 5.5.7.e shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

a. Demonstrate for each of the ESF systems that an in place test of the HEPA filters shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 1, Positions C.5.a and C.5.c, and ANSI N510-1975 at the system flowrate specified below:

ESF Ventilation System Flowrate (cfm)

Standby Gas Treatment (SGT) System 2700 to 3300 Control Room Emergency Ventilation 1800 to 2200 (GREV) System (continued)

Brunswick Unit 2 5.0-10 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 1, Positions C.5.a and C.5.d, and ANSI N510-1975 at the system flowrate specified below:

ESF Ventilation System Flowrate (cfm)

SGT System 2700 to 3300 CREV System 1800 to 2200

c. 1) Demonstrate for the SGT System that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 1, Position C.6.b, and tested in accordance with ASTM D3803-1989, at a temperature of 30°C, a face velocity of 61 fpm, and a relative humidity of 70% within the tolerances provided in Table 1 of ASTM D3803-1989, shows the methyl iodide penetration < 0.5%.

c. 2) Demonstrate for the CREV System that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 1, Position C.6.b, and tested in accordance with ASTM D3803-1989, at a temperature of 30°C and a relative humidity of 95% within the temperature and humidity tolerances provided in Table 1 of ASTM D3803-1989, meets the acceptance criteria of < 5.0% penetration of methyl iodide.

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilter (SGT only), and the charcoal adsorbers is less than or equal to the value specified below when tested at the system flowrate specified as follows:

ESF Ventilation System Delta P (inches wg) Flowrate (cfm)

SGT System 8.5 2700 to 3300 CREV System 5.25 1800 to 2200 (continued)

Brunswick Unit 2 5.0-11 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate that the heaters for each of the SGT subsystems dissipate

~ 16.67 kW under a degraded voltage condition when tested in accordance with ANSI N510-1975.

5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the Main Condenser Offgas Treatment System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each outdoor liquid radwaste tank that is not surrounded by liners, dikes, or walls, capable of holding the tank's contents and that does not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is ::;; 1O Curies, excluding tritium and dissolved or entrained gases.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

5.5.9 Diesel Fuel Oil Testing Program A diesel fuel oil testing program shall establish required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has not become contaminated with other products during transit, thus altering the quality of the fuel oil; (continued)

Brunswick Unit 2 5.0-12 Amendment No.

Programs and Manuals .

5.5 5.5 Programs and Manuals 5.5.9 Diesel Fuel Oil Testing Program (continued)

b. Kinematic viscosity is within limits for ASTM 2-D fuel oil when tested every 92 days; and

c. Total particulate concentration of the fuel oil is~ 10 mg/I when tested every 31 days in accordance with the applicable ASTM Standard.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.

5.5.10 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.1 O.b.1 or 5.5.1 O.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

(continued)

Brunswick Unit 2 5.0-13 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

a. The SFDP shall contain the following:

1. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

2. Provisions for ensuring the' plant is maintained in a safe condition if a loss of function condition exists;

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4. Other appropriate limitations and remedial or compensatory actions.

b. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

3. A required system redundant to support system(s) for the supported systems described in b.1 and b.2 above is also inoperable.

(continued)

Brunswick Unit 2 5.0-14 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.12 Primary Containment Leakage Rate Testing Program A primary containment leakage rate testing program shall establish requirements to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option Bas modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995, as modified by the following exceptions:

a. The visual e_xamination of concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

b. The visual examination of the metallic shell, penetrations, and appurtenances intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

c. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at Pa as specified in Nuclear Energy Institute Guideline 94-01, Revision O;

d. Reduced duration Type A tests may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

e. Performance of Type C leak rate testing of the hydrogen and oxygen monitor isolation valves is not required; and (continued)

Brunswick Unit 2 5.0-15 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

f. Performance of Type C leak rate testing of the main steam isolation valves at a pressure less than Pa instead of leak rate testing at Pa as specified in ANSI/ANS 56.8-1994.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 49 psig.

The maximum allowable primary containment leakage rate, La, shall be 0.5% of

• primary containment air weight per day at Pa*

Leakage rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is :::;; 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for Type B and C tests and

0.75 La for Type A tests.

5.5.12 Primary Containment Leakage Rate Testing Program (continued)

b. Air lock testing acceptance criteria are:

1) Overall* air lock leakage rate is:::;; 0.05 La when tested at~ Pa.

2) For each air lock door, leakage rate is :::;; 5 scfh when the gap between the door seals is pressurized to ~ 10 psig.

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

5.5.13 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 Emergency Ventilation (GREV) System, 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 occupancy of the CRE under design basis accident (OBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident. The program shall include the following elements:

(continued)

Brunswick Unit 2 5.0-16 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued)

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement at designated locations, of the CRE pressure relative to external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the CREV System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

Brunswick Unit 2 5.0-17 Amendment No.

BSEP 16-0035 Enclosure 4 Proposed Technical Specification Bases Changes (Mark-Up), BSEP Unit 1 Information Only

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 throu h 3.10 and a I at all times ,

h rwi SR 3.0.2 and SR 3.0.3 apply in Chapter 5 on/ when invoked b a Cha ter 5 S ecification.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components , and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when :

a. The systems or components are known to be inoperable, although still meeting the SRs; or

b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified . The SRs associated with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

(continued)

Brunswick Unit 1 B 3.0-12 Revision No. 50

SR Applicability B 3.0 BASES SR 3.0.2 SR 3.0.2 permits a 25% extension of the interval specified in the (continued) Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities) .

____.-.--..~The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs.

The exceptions to SR 3.0 .2 are those Surveillances for which the 25%

extension of the interval specified in the Frequency does no,..

t _.

a....._1.._._ _......,

These exce tions are stated in the individual S ecifications -AA--&

Examples of where SR 3.0.2 does not apply~ are the inservice testing of pumps and valves in accordance with applicable American Society of Mechanical Engineers 0 eration and Maintenance Code, as re uired b 10 CFR 50.55a and fo a Surveillance with a Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the TS . The TS cannot in and of themselves extend a test interval specified in the regulations (directly or by reference. )

As stated in SR 3.0.2 , the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ... " basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action ,

whether it is a particular Surveillance or some other remedial action , is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified .

(continued)

When a Sectibn 5.5, " Programs and Manuals," specificatiorl states that the provisions of SR 3.0.2 are applicable, a 25% extension of the testing interval, whether stated in the specification or incorporated by reference, is permitted.

Brunswick Unit 1 B 3.0-14 Revision No. 50

SR Applicability B 3.0 BASES (continued)

SR 3.0 .3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0 .2, and not at the time that the specified Frequency was not met.

. - - - - - -" lThis delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions ,

adequate planning , availability of personnel , the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements .

When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions , operating situations , or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading , or in accordance with 10 CFR 50 , Appendix J, as modified by approved exemptions, etc.) is discovered not to have been performed when specified , SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified , the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 also provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence . Use of the delay period established by SR 3.0 .3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the (continued)

Wheh a Section 5.5, "Programs and Manuals," specification states that the provisions of SR 3.0.3 are applicable, it permits the flexibility to defer declaring the testing requirement not met in accordance with SR 3.0.3 when the testing has not been completed within the testing interval (including the allowance of SR 3.0.2 if invoked by the Section 5.5 specification).

Brunswick Unit 1 B3.0-15 Revision No. 50

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1 .7.5 REQUIREMENTS (continued) This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron (measured in weight% sodium pentaborate) exists in the storage tank. SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. SR 3.1 .7.5 must also be performed anytime the temperature is restored to within the limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred during the time period temperature was outside the limits of the Figure.

The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between Surveillances.

SR 3.1.7.6 Demonstrating that each SLC System pump develops a flow rate

~ 41 .2 gpm at a discharge pressure ~ 1190 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction ,

cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY and detect inci ient failures b indicatin abnormal erformance. The Fre uenc of this Surveillance is in accordance with the IRseFViee TestiR ~ PFe ~ Fa~

/NSERVICE TESTING PROGRAM.

SR 3.1.7.7 This Surveillance ensures that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch (continued)

Brunswick Unit 1 B 3.1.7-5 Revision No. 31 I

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 1O 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 durin 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 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 (continued)

Brunswick Unit 1 B 3.4 .3-3 Revision No. 31

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5 .1.2 (continued)

REQUIREMENTS The 31 day Frequency of this SR was derived from the IAseFVise Testi R ~

t+"iFeaffifR- INSERVICE TESTING PROGRAM requirements for performing va ve es mg a eas once every ays. e requency o ays 1s further justified because the valves are operated under procedural control and because improper valve position typically only affects a single subsystem. This Frequency has been shown to be acceptable through operating experience.

In MODE 3 with reactor steam dome pressure less than the RHR shutdown cooling isolation pressure, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes the period when the required RHR pump is not operating and the period when the system is being realigned to or from the RHR shutdown cooling mode. At low reactor pressure and with a low decay heat load associated with operation in MODE 3 with reactor steam dome pressure less than the RHR shutdown cooling isolation pressure , a reduced complement of low pressure ECCS subsystems should provide the required core cooling in the unlikely event of a LOCA, thereby, allowing operation of the shutdown cooling mode of the RHR System, when necessary.

SR 3.5.1.3 Verification every 31 days that ADS pneumatic supply header pressure is

?: 95 psig ensures adequate pneumatic pressure for reliable ADS operation . The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least three valve actuations can occur with the drywell at 70% of design pressure. The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of :?: 95 psig is provided by the non-interruptible Reactor (continued)

Brunswick Unit 1 B 3.5.1-11 Revision No. 31 I

ECCS-Operating B 3.5. 1 BASES SURVEILLANCE SR 3.5.1.3 (continued)

REQUIREMENTS Instrument Air System, the Pneumatic Nitrogen System , or the Nitrogen Backup System . This SR may be satisfied by verifying the absence of all associated pneumatic low pressure alarms. The 31 day Frequency takes into consideration administrative controls over operation of the pneumatic systems and alarms for low air and nitrogen pressure.

SR 3.5.1.4 Verification every 31 days that the RHR System cross tie valve is locked closed ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem . If the RHR System cross tie valve is open , both LPCI subsystems must be considered inoperable. The 31 day Frequency has been found acceptable, considering that this manual valve is under strict administrative controls that will ensure the valve continues to remain locked closed .

SR 3.5.1 .5 Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will close when required. Upon initiation of an automatic LPCI subsystem injection signal , these valves are required to be closed to ensure full LPCI subsystem flow injection in the RPV.

De-energizing the valves in the closed position will also ensure the proper flow path for the LPCI subsystem . Acceptable methods of de-energizing a valve include de-energizing breaker control power, racking out the breaker or removing the breaker.

The specified Frequency is once each reactor startup before THERMAL POWER is > 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed if the last performance was more than 31 days ago. Verification prior to or during

.

• 0 *

• normal l~sef'l1iee TestiFl~ Pre~F8FA INSERVICE TESTING PROGRAM generic valve cycling Frequency of (continued)

Brunswick Unit 1 B 3.5.1-12 Revision No. 31 I

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6. SR 3.5.1.7. and SR 3.5.1.8 (continued)

REQUIREMENTS Therefore, SR 3.5.1.7 and SR 3.5 .1.8 are modified by Notes that state the Surveillances are not required to be performed until 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the reactor steam pressure is adequate to perform the test.

ti'le IRseFViee Testil'l!:J PF8!:JF8ffl INSERVICE TESTING PROGRAM requirements. The 24 month Frequency for SR 3.5.1.8 is based on the nee to pe orm t e urve1 ance un er t e con 1t1ons t at app y Just prior to or during a startup from a plant outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency is considered to be acceptable from a reliability standpoint.

SR 3.5.1.9 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI , CS , and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions . This SR also ensures that the HPCI System will automatically restart on an RPV low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low level signal or a suppression pool high water level signal.

The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1, "ECCS Instrumentation," overlaps this Surveillance to provide complete testing of the assumed safety function .

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle) . Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.5.1-14 Revision No. 44 I

RCIC System B 3.5.3 BASES (continued)

SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the reactor vessel upon demand. This SR will also prevent water hammer in the piping following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the RCIC System piping, the procedural controls governing system operation , and operating experience.

SR 3.5.3.2 Verifying the correct alignment for manual, power operated , and automatic valves in the RCIC flow path provides assurance that the proper flow path exists for RCIC System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves are verified to be in the correct position prior to locking, sealing, or securing . A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition to the accident position in the proper stroke time.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position . This SR does not apply to valves that cannot be inadvertently misaligned , such as check valves .

This SR also includes the steam flow path for the turbine and the flow controller position.

The 31 day Frequency of this SR was derived from the IRs0F¥ie0 Testi R ~

Pre!f:) F8 FR INSERVICE TESTING PROGRAM requirements for performing further justified because the valves are operated under procedural control and because improper valve position typically affects only the RCIC System. This Frequency has been shown to be acceptable through operating experience.

(continued)

Brunswick Unit 1 B 3.5.3-4 Revision No. 31 I

RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3 .3 and SR 3.5.3.4 (continued)

REQUIREMENTS inlet is OPERABLE, Note 1 to SR 3.5.3 .3 requires the high pressure test to be performed with the turbine steam being supplied with reactor steam from the Main Steam System.

A 92 day Frequency for SR 3.5.3.3 is consistent with the IAseFVise TesUA§ Pre§F8FA INSERVICE TESTING PROGRAM requirements. The mon requency or . . . 1s ase on e nee o pe orm e Surveillance under conditions that apply just prior to or during a startup from a plant outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore , the Frequency is considered to be acceptable from a reliability standpoint.

SR 3.5 .3.5 The RCIC System is required to actuate automatically in order to verify its design function satisfactorily . This Surveillance verifies that, with a required system initiation signal (actual or simulated) , the automatic initiation logic will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence ;

that is, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures the RCIC System will automatically restart on an RPV low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low level signal. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2, "Reactor Core Isolation Cooling (RCIC) System Instrumentation," overlaps this Surveillance to provide complete testing of the assumed design function.

While this Surveillance can be performed with the reactor at power, operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.5.3-6 Revision No. 31 I

PC IVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.3 (continued)

REQUIREMENTS administrative controls , such as those that limit the shelf life of the explosive charges , must be followed . The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity.

SR 3.6.1.3.4 Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.5. The isolation time test ensures that each valve will isolate in a time period less than or equal to that assumed in the safet anal ses. The isolation time and Fre uenc of this SR are in accordance with the requirements of the ll"l!ef'Viee Testif'I§ Pre§reffl INSERVICE TESTING PROGRAM.

SR 3.6.1.3.5 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the OBA and transient analyses . This ensures that the calculated radiological consequences of these events remain within 1O CFR 50.67 limits. The Fre uenc of this SR is in accordance with the requirements of the IRSeRt*iee Testifl§ Pre§reffl INSERVICE TESTING PROGRAM.

SR 3.6.1 .3.6 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a OBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. This SR includes verifying that each automatic PCIV in the Containment Atmosphere Dilution System flow path will actuate to its isolation position on the associated Group 2 and 6 primary containment isolation signals .

The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1 , "Primary Containment Isolation Instrumentation ," overlaps th is SR to provide (continued)

Brunswick Unit 1 B 3.6.1.3-11 Revision No. 31 I

PC IVs B 3.6 .1.3 BASES SURVEILLANCE SR 3.6.1.3.8 REQUIREMENTS (continued) The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be

(

from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The

==~~t~~: :;:;!~~cg~~i/~E~~i~7N~Q~%~mi~~~yhe SR 3.6.1.3 .9 The analyses in References 2, 6, 7, and 8 are based on leakage that is less than the specified leakage rate. Leakage through each main steam line must be ~ 100 scfh when tested at ~ P1 (25 psig) . The combined leakage rate for all four mains steam lines must be ~ 150 scfh when tested at ~ 25 psig in accordance with the Primary Containment Leakage Rate Testing Program . The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(0) to implement the requirements of 10 CFR Part 50 , Appendix J, Option B (Ref. 9) , and conforms with Regulatory Guide 1.163 (Ref. 10) and Nuclear Energy Institute (NEI) 94-01 (Ref. 11) except for the following :

a. Local leak rate testing of the MS IVs may be performed at a pressure less than Pa. This is an exemption from the requirements of 1O CFR 50 Appendix J (Ref. 9). The basis for this exemption is described in Reference 12.

The Frequency is required by the Primary Containment Leakage Rate Testing Program .

(continued)

Brunswick Unit 1 B 3.6.1.3-13 Revision No. 46

Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1 .5 BASES SURVEILLANCE SR 3.6.1 .5.1 (continued)

REQUIREMENTS verifying the absence of the Nitrogen Backup System low pressure alarms . The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on engineering judgment in view of the fact that adequate indication of pressure is available to the operator and the Frequency has also been shown to be acceptable through operating experience.

SR 3.6.1 .5.2 Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position . The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel , and has been shown to be acceptable through operating experience.

Two Notes are added to this SR. The first Note allows reactor building-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR.

These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note is included to clarify that vacuum breakers open due to an actual differential pressure are not considered as failing this SR.

SR 3.6.1 .5.3 Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position . This SR ensures that the safety analysis assumptions are valid . This is accomplished by manually verifying that each mechanical vacuum breaker is free to open and verifying each pneumatic butterfly valve o erates throu h at least one com lete c cle of full travel. The 92 da Frequency of this SR was developed based upon IRseFViee Testir:i 9 Pre !!!j FBffl JNSERVICE TESTING PROGRAM requirements to perform valve testing at least once every 92 days.

(continued)

Brunswick Unit 1 B 3.6 .1.5-7 Revision No. 31 I

Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 (continued)

REQUIREMENTS and drywell is maintained > 0.5 times the initial differential pressure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without nitrogen makeup. The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel and procedural controls to ensure the drywell is normally maintained at a higher pressure than the suppression chamber, and has been shown to be acceptable through operating experience. This verification is also required within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after any discharge of steam to the suppression chamber from any source , and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after an operation that causes any of the vacuum breakers to open.

A Note is added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.

SR 3.6.1 .6.2 Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This is accomplished by verifying each required vacuum breaker operates through at least one complete cycle of full travel. This SR ensures that the safet anal sis assum tions are valid . The 92 day Frequency of this SR was developed, based on IAseFViee TestiR!iJ

/NSERVICE TESTING PROGRAM re uirements to erform valve testing at least once every 92 days. In addition , this functional test is required within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a discharge of steam to the suppression chamber from the SRVs.

SR 3.6.1.6.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid . The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency has been demonstrated to be acceptable, based on operating experience, and is further justified because of other surveillances performed more frequently that convey the proper functioning status of each vacuum breaker.

REFERENCES 1. UFSAR, Section 6.2 .

2. 10 CFR 50 .36(c)(2)(ii).

Brunswick Unit 1 B 3.6.1 .6-5 Revision No. 65

SW System and UHS B 3.7 .2 BASES SURVEILLANCE SR 3.7.2.4 (continued)

REQUIREMENTS The 92 day Frequency was chosen to provide additional assurance that the capability to provide cool ing water to each DG under accident

.. . . . . . .

l "~el"V' i ee Te~ti " ~ Pre~ re ~ /NSERVICE TESTING PROGRAM Frequency for testing of valves .

To minimize testing of the cooling water supply valves to each DG , Note 1 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. Note 2 indicates that isolation of the SW System to a DG renders the DG inoperable but does not affect the OPERABILITY of the SW System . As such, if the automatic transfer of the cooling water supply valves associated with a DG fails this Surveillance, the DG should be considered inoperable. However, the SW System is still OPERABLE.

It is not necessary to declare the DG inoperable if the service water supply valves to the affected DG are administratively controlled to ensure cooling water is supplied to the DG and two NSW pumps are operable on the corresponding NSW header that the DG is aligned to. This ensures that a single active failure will not result in more than one DG not receiving cooling water (Ref. 5) .

SR 3.7.2.5 This SR verifies that the automatic isolation valves of the SW System will automatically align to the safety or emergency position to provide cooling water exclusively to the safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal. This SR also verifies the automatic start capability of the required NSWpumps.

Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency.

Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.7 .2-13 Revision No. 31 I

BSEP 16-0035 Enclosure 5 Description and Assessment of the Proposed Alternative to the ASME Code

ATTACHMENT 5 DESCRIPTION AND ASSESSMENT OF THE PROPOSED ALTERNATIVE TO THE ASME CODE Request in Accordance with 1O CFR 50.55a(z)(2)

Alternative Due to Hardship Without a Compensating Increase in Quality and Safety

1.0 DESCRIPTION

The request is to adopt a proposed alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code by adoption of approved Code Case OMN-20, "lnservice Test Frequency."

2.0 ASSESSMENT Technical Evaluation of the Proposed Alternative to the OM Code Section IST of Division 1 of the OM Code, which is incorporated by reference in 10 CFR 50.55a(a), specifies component test frequencies based either on elapsed time periods (e.g., quarterly, 2 years) or on the occurrence of a plant condition or event (e.g., cold shutdown, refueling outage).

ASME Code Case OMN-20, "lnservice Test Frequency," has been approved for use by the ASME OM committee as an alternative to the test frequencies for pumps and valves specified in ASME OM Division 1: Section IST, 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

Code Case OMN-20 is not referenced in the latest revision of Regulatory Guide 1.192 (August 2014) as an acceptable OM Code Case to comply with 1O CFR 50.55a(f) requirements as allowed by 1O CFR 50.55a(b)(6). The proposed alternative is to use Code Case OMN-20 to extend or reduce the IST frequency requirements for the fourth 10-year IST interval or until OMN-20 is incorporated into the next revision of Regulatory Guide 1.192.

ASME Code Components Affected The Code Case applies to pumps and valves specified in ASME OM Division 1: Section IST, 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code. Frequency extensions may also be applied to accelerated test frequencies (e.g., pumps in Alert Range) as specified in OMN-20.

For pumps and valves with test periods of 2 years or less, the test frequency allowed by OMN-20 and the current TS lnservice Testing Program (as modified by SR 3.0.2 and EGM 2012:-001) are the same. For pumps and valves wit~ test frequencies greater than 2 years, OMN-20 allows the test frequency to be extended by 6 months. The current TS lnservice Testing Program does not allow extension of test frequencies that are greater than 2 years.

Applicable Code Edition and Addenda

ASME Code Case OMN-20 applies to ASME OM Division 1: Section IST, 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code.

The BSEP Code Edition and Addenda that are applicable to the current program interval are the 2001 Edition through the 2003 Addenda of the ASME Code,Section XI. The current interval ends on May 10, 2018.

Applicable Code Requirement

This request is made in accordance with 1O CFR 50.55a(z)(2), and proposes an alternative to the requirements of 1O CFR 50.55a(f), which requires pumps and valves to meet the test requirements set forth in specific documents incorporated by reference in 1O CFR 50.55a(a).

ASME Code Case OMN-20 applies to Division 1: Section IST of the ASME OM Code and associated addenda incorporated by reference in 1O CFR 50.55a(a).

Reason for Request

The IST Program controls specified in Section 5.5 of TS provide: a) a table specifying certain IST frequencies; b) an allowance to apply SR 3.0.2 to inservice tests required by the OM Code and with frequencies of two years or less; c) an allowance to apply SR 3.0.3 to inservice tests required by the OM Code; and d) a statement that, "Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS." In Regulatory Issue Summary (RIS) 2012-10, "NRC Staff Position on Applying Surveillance Requirement 3.0.2 and 3.0.3 to Administrative Controls Program Tests," and Enforcement Guidance Memorandum (EGM) 2012-001, "Dispositioning Noncompliance with Administrative Controls Technical Specifications Programmatic Requirements that Extend Test Frequencies and Allow Performance of Missed Tests," the NRC stated that items b, c, and d of the TS IST Program were inappropriately added to the TS and may not be applied (although the EGM allows licensees to continue to apply those paragraphs pending a generic resolution of the issue).

In RIS 2012-1 O and EGM 2012-001, the NRC stated that the current TS allowance to apply SR 3.0.2 and SR 3.0.3 to the lnservice Testing Program would no longer be permitted. In response, OMN-20, which provides allowances similar to SR 3.0.2, was approved and is proposed to be used as an alternative to the test periods specified in the OM code. The proposed alternative substitutes an approved Code Case for the existing TS requirements that the NRC has determined are not legally acceptable as a TS allowance. This proposed alternative provides an equivalent level of safety as the existing TS allowance, while maintaining consistency with 10 CFR 50.55a and the ASME OM Code.

Proposed Alternative and Basis for Use The proposed alternative is OMN-20, "lnservice Test Frequency," which addresses testing periods for pumps and valves specified in ASME OM Division 1: Section IST, 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

This request is being made in accordance with 1O CFR 50.55a(z)(2), in that the existing requirements are considered a hardship without a compensating increase in quality and safety for the following reasons:

1) For IST testin.g periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the IST testing periods by up to 25%. The period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would Page 2

cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. The test period extension and the statements regarding the appropriate use of the period extension are equivalent to the existing TS SR 3.0.2 allowance and the statements regarding its use in the SR 3.0.2 Bases. Use of the SR 3.0.2 period extension has been a practice in the nuclear industry for many decades, and elimination of this allowance would place a hardship on BSEP when there is no evidence that the period extensions affect component reliability.

2) For IST testing periods of greater than 2 years, OMN-20 allows an extension of up to six months. The ASME OM Committee determined that such an extension is appropriate.

The six-month extension will have a minimal impact on component reliability considering that the most probable result of performing any inservice test is satisfactory verification of the test acceptance criteria. As such, pumps and valves will continue to be adequately assessed for operational readiness when tested in accordance with the requirements specified in 1O CFR 50.55a(f) with the frequency extensions allowed by Code Case OMN-20.

3) As stated in EGM 2012-001, if an lnservice Test is not performed within its frequency, SR 3.0.3 will not be applied. The effect of a missed lnservice Test on the Operability of TS equipment will be assessed under the licensee's Operability Determination Program.

Duration of Proposed Alternative The proposed alternative is requested for the current 10-year IST interval or until Code Case OMN-20 is incorporated into a future revision of Regulatory Guide 1.192, referenced by a future revision of 1O CFR 50.55a, whichever occurs first.

Precedents The NRC approved the use of OMN-20 for North Anna on March 27, 2014 (NRC ADAMS Accession Number ML14084A407).

Page 3

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