NLS2016046, Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify Surveillance Requirement Usage Rule Application to Section 5.5 Testing.

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Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify Surveillance Requirement Usage Rule Application to Section 5.5 Testing.
ML16245A288
Person / Time
Site: Cooper Entergy icon.png
Issue date: 08/26/2016
From: Limpias O A
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2016046
Download: ML16245A288 (51)
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Text

H Nebraska Public Power District NLS2016046 August 26, 2016 Always there when you need us Attention:

Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 50.90

Subject:

Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, "TS Inservice Testing Program Removal and Clarify Surveillance Requirement Usage Rule Application to Section 5.5 Testing" Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Nebraska Public Power District (NPPD) is submitting a request for an amendment to the Technical Specifications (TS) for Cooper Nuclear Station (CNS). The proposed change revises the TS to eliminate Section 5.5.6, "lnservice Testing Program." A new defined term, "Inservice Testing Program," is added to the TS Definitions section. This request is consistent with TSTF-545, Revision 3, "TS Jnservice Testing Program Removal and Clarify Surveillance Requirement Usage Rule Application to Section 5.5 Testing." The use of Code Case OMN-20, "lnservice Test Frequency," was previously approved for use at CNS in a Nuclear Regulatory Commission (NRC) safety evaluation dated February 12, 2016 (ML16014Al 74). NPPD requests NRC approval of the proposed TS change and issuance of the requested license amendment by September 10, 2017. The amendment will be implemented within 60 days of issuance of the amendment.

Attachment 1 provides a description and assessment of the proposed TS changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides TS Bases pages marked up to show the associated TS Bases changes and is provided for information only. This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Renewed Facility Operating License through Amendment 256 dated July 25, 2016, have been incorporated into this request. This request is submitted under oath pursuant to 10 CFR 50.30(b).

COOPER NUCLEAR STATION P.O. Box 98 /Brownville, NE 68321-0098 Telephone:

(402) 825-3811 / Fax:*(402}

825-5211 . www.nppd.com NLS2016046 Page 2 of2 By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(l).

Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b )(1 ). This letter contains no regulatory commitments.

Should you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

I declare under penalty of perjury that the foregoing is true and correct. Executed On: 9 { '2-C, \, l Y, Date /dv Attachments:

1. Description and Assessment of Technical Specifications Changes 2. Proposed Technical Specifications Changes (Mark-up)
3. Revised Technical Specifications Pages 4. Proposed Technical Specifications Bases Changes (Mark-up):;:.

Information Only cc: Regional Administrator w/ attachments USNRC -Region IV Cooper Project Manager w/ attachments USNRC -NRR Plant Licensing Branch IV-2 Senior Resident Inspector w/ attachments USNRC-CNS Nebraska Health and Human Services w/ attachments Department of Regulation and Li censure NPG Distribution w/o attachments CNS Records w/ attachments NLS2016046 Attachment 1 Page 1of5 Attachment 1 Description and Assessment of Technical Specifications Changes Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Technical Specification Pages 1.0 Description

2.0 Assessment

1.1-3 1.1-4 1.1-5 3.1-22 3.4-7 3.5-5 3.5-10 3.6-13 3.6-14 3.6-26 3.6-32 3.6-39 5.0-10 2.1 Applicability of Published Safety Evaluation

2.2 Variations

3. 0 Regulatory Analysis 3.1 No Significant Hazards Consideration Analysis 4.0 Environmental Evaluation NLS2016046 Attachment 1 Page 2of5

1.0 DESCRIPTION

The proposed change eliminates Technical Specifications (TS), Section 5.5.6, "Inservice Testing (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, "Inservice Test Frequency." A new defined term, "Inservice Testing Program," is added to TS Section 1.1, "Definitions." The proposed change to the TS is consistent with TSTF-545, Revision 3, "TS Inservice Testing Program Removal and Clarify Surveillance Requirement Usage Rule Application to Section 5.5 Testing." 2.0 ASSESSMENT

2.1 Applicability

of Published Safety Evaluation Nebraska Public Power District (NPPD) has reviewed the model safety evaluation provided in the Federal Register Notice of Availability dated March 28, 2016, to the Technical Specifications Task Force in a letter dated December 11, 2015 (ML15314A365 and ML15314A305).

This review included a review of the Nuclear Regulatory Commission (NRC) staffs evaluation, as well as the information provided in TSTF-545.

NPPD concluded that the justifications presented in TSTF-545, and the model safety evaluation prepared by the NRC staff are applicable to Cooper Nuclear Station (CNS) and justify this amendment for the incorporation of the changes to the CNS TS. CNS was issued a construction permit on June 4, 1968, and the provisions of 10 CFR 50.55a(f)(l) are applicable.

In a letter to the NRC, dated March 19, 2015 (ML15084A221), NPPD submitted requests for relief to certain ASME OM code requirements for the CNS fifth 10-year IST program interval.

Relief Request RG-01 requested use of Code Case OMN-20 as an alternative to the frequencies of the ASME OM Code. Relief Request RG-01 was approved by the NRC in a safety evaluation dated February 12, 2016 (ML16014Al 74). 2.2 Variations No technical variations are proposed in this amendment request. The following proposed variations are administrative and do not affect the applicability ofTSTF-545 or the NRC staffs model safety evaluation dated December 11, 2015. The CNS TS 1) in some cases utilize a different section title or Surveillance Requirement numbering system and, 2) do not include all the specifications shown on the applicable Standard Technical Specifications (STS), NUREG 1433, pages. 11 Since the CNS TS do not include SR 3.4.5.1, Reactor Coolant System Pressure Isolation Valve Leakage, as shown on the General Electric BWR/4 STS pages in TSTF-545, there will be no corresponding change to the CNS TS.

NLS2016046 Attachment 1 Page 3of5 o General Electric BWR/4 STS SR 3.5.1.7 is numbered SR 3.5.1.6 in the CNS TS.

  • General Electric BWR/4 STS IST Program is section 5.5.7. The equivalent CNS TS section is 5.5.6.

NPPD proposes to retain the TS 5.5.6 reference, now shown as "DELETED," and hot. change the subsequent TS program numbers. "' 3.0 REGULATORY ANALYSIS 3.1 No Significant Hazards Consideration Analysis Nebraska Public Power District (NPPD) requests adoption ofthe Technical Specffications (TS) changes described in TSTF-545, "TS Inservice Testing Program Removal and Clarify * ,* Surveillance Requirement Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications, into the Cooper Nuclear Station TS. The proposed change revises the TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals," to delete the "Inservice 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, "Inservice Test Frequency." Other requirements in Section 5.5 are eliminated because the Nuclear Regulatory Commission (NRC) has determined their appearance in the TS is contrary to regulations.

A new defined term, "Inservice Testing Program," is added, which references the requirements of Title lO of the Code of Federal Regulations (10 CPR), Part 50, paragraph 50.55a(t).

NPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.

NLS2016046 Attachment 1 Page 4of5 2. The proposed change revises TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals," by eliminating the "Inservice Testing Program" specification.

Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, "Inservice Test Frequency." The remaining requirements in the Section 5.5 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to " regulations.

A new defined term, "Inservice Testing Program," is added to the TS, which references the requirements of 10 CFR 50.55a(t).

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. Inservice 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.

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.

NLS2016046 Attachment 1 Page 5of5 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 10 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 a 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, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92( c), and, accordingly, a finding of "no significant hazards consideration" is justified.

  • 4.0 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 10 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 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 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b ), no environmental impact -statement or environmental assessment need be prepared in Connection with the proposed change. ,-

NLS2016046 Attachment 2 Page 1 of 14 Attachment 2 Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages 1.1-3 1.1-4 1.1-5 3.1-22 3.4-7 3.5-5 3.5-10 3.6-13 3.6-14 3.6-26 3.6-32 3.6-39 5.0-10 1 . 1 Definitions DOSE EQUIVALENT 1-131 (continued)

INSERVICE TESTING PROGRAM LEAKAGE The INSERVICE TESTIG PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f). LINEAR HEAT GENERATION RATE (LHGR) Defin iti ons 1.1 1-133, 1-134, and 1-135 actually present. The DOSE EQUIVALENT 1-131 concentration is calculated as follows: D O SE EQUIVALENT 1-131 = (1-131) + 0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). T he dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Value of Ra dionuclide Intake and Ai r Concentration and Do se Conversion Factors for Inh alation , Submersion, an d I ng es tion," 1989. EAKAGE shall be: b. I dentified LEAKAGE 1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump o r collecting ta nk; or 2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere w it h the operation of leakage detection systems or not to be pressure boundary LEAKAGE; Unident ified LEAKAGE All LEAKAGE into the drywell that i s not identified LEAKAG E; c. Total LEAKA GE Sum of the identified and unidentified LEAKAGE; d. Pressure Boundarv LEAKAGE LEAKAGE throug h a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wa ll. or vessel wall. The LHGR shall be the heat gene r ation rate per unit length of fuel ro d. It is the integ r al of the heat flux over the heat transfer area associated with the unit length. LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILIT Y of a logic circuit , (cont in ued) Cooper 1.1-3 Amendment N o. 261 , 255

1. 1 Defi nitio n s LOGI C SYTEM FUN CTIONA L T E S T (continued)

MIN I MUM CRITICAL POWER RA TIO (M CPR) M OD E OPERA BLE -O P E RA B ILITY PRE S SU R E AND TEM PE RATUR E LIM I T S REPORT (PTL.R) RATED TH E RMAL POWER (RTP) REACTOR PROTECTION SYS T EM(R P S)RES P ONSE TIME SHUTDOWN MARGIN (SDM) Cooper Definitions f rom as clos e t o the s enso r as pr acticab le up t o , but not including, the act u a t ed device, to ver"i"iy 0::1E RABILI TY. The LOGIC SYSTEM FUNC T IONAL T EST may b e perfo r med by means o f any series of se qu enti al, overlapping, or t otal system steps so tha the entire logic system is t ested. The MCPR sha ll be th e smalle s t c ritical power rati o (CPR) that exists in the core for e ach class of f ue l. Th e C P R is th at power i n the a s sem b ly that is calculated by a p pli ca t io n of the a pp rop r ia te co rrelation (s) to cause some point in the assembly to experience boiling tr ansition, d i vide d by the actual assemb ly operating power. A MODE shall c orre spond to any o n e inclusive combination of mode switch position, average react or coolant tempe rat ure, a nd reactor vessel head closur e bolt t ensio ning spe cified in Table 1.1-1 with fue l in the re acto r vess el. A system, subsystem, division, co m ponent, or de v ice shall be OPERABLE or have OPERABILITY when it is c a pable of performing it s s pe cified safety function(s) and when all necessary attendant in stru ment ation , controls, norma l or emergency electrical power , coo ling and s eal wate r , l ub ri c ation, and other auxiliary equipment that ar e required for t he sys t em, subsys t em , division , com pon ent, or de vi ce t o perform its specified s a fe t y f u n ct ion(s) are al so c ap abl e of perfo r ming their rel at e d suppo rt funciion(s).

The PTLR is t he unit sp ecifi c do c ument t hat provides the r e actor vessel p re ssu r e and temperature l imits, in clu din g h eatup and co old ow n rates , for the curre nt reac t or vessel fluence period. T h e se pressure and t emperature limit s shall be determined for e ach fluence period in accordance with Specification 5.6.7. RTP shall be a total r ea ct o r core heat tran sfe r rate to the re a ctor coolan t of 2419 MWt. The RPS RESP ONS E TIME s h all be tha t t i me segment fro m the time the sensor co nt ac ts a ctuate to t he t i me t he scram soleno i d valves deenergize. SDM shall be t he amount o f reactivity by w hich the re actor i s sub crit ical or would be subcritical throug h out the cycle assum i n g that: a. The reactor is xe n on free; (continued) Amendment No.-26&-

1. 1 Definitions SHUTDOWN MARGIN (SOM) (c on t inued) S TA GGERED TES T BASIS THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TI M E Cooper b. Definitions
1. 1 T h e moderator tempP.ratu r e 6 8°F , corresponding to th e most reactive state; a nd c. A ll co n trol rod s are f ully in s erted e x cept for the single c o n trol r od of highes t r eact i vity worth, which is assume d t o b e f u lly withd r awn. With co nt rol r ods not c a pable o f being fully inse 1 ed , the r ea c t iv i ty w o rth of t hes e co nt ro l r od s mu s t be ac count e d for in t he de t e r m i n a t io n of SOM. A ST A GGER E D T EST BA S IS sha ll consist of t he testing of one of th e systems, subsystems , channels , or other designated components d uring t he interval specified by th e Surv e illance Frequency , so t hat all systems , subsystems , channels , or other d e signated components are tested durin g n Surveillance Frequency intervals, where n is the tota l number of system s , subsystems.

channels , or other designated components in the associated function. TH E RMAL POWER shall be the total reactor core hecit tra n sfer r ate to t h e c oolant. The TURBINE BYPASS SYSTEM RESPONSE T IME consists of two com p onen t s: a. The time from initi a l movement o f the main turbin e s t op valve or co ntrol va l ve until 80% of the t urbine bypass capac it y i s es tabl i she d; and b. The time from initial movement of the main turbine stop valve or cont r ol valve until initial movement of t he turbine bypass valve. The response time may be measured by m e ans of any s eries of sequential , overl a pping , or total steps s o that the en t i r e respons e time is measured. 1.1-5 Amend m ent No. -254--

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.1.7.6 Verify each SLC subsystem manual valve in the flow path that is not locked, sea l ed , or otherwise secured in position , is in the correct position or can be aligned to the correct position. SR 3.1.7.7 Verify each pump develops a flow 38.2 gpm at a discharge 1300 psig. SR 3.1.7.8 Verify flow thrOUgh one SLC subsystem

.from pump into reactor pressure vessel. SR 3.1.7.9 Verify all heat traced piping between storage tank and pump suction is unblocked. Cooper 3.1-22 SLC System 3.1.7 FREQUENCY 31 days In accordance with the =FestiAg F2F9!JFQFFI 24 months on a STAGGERED TEST BASIS 24 months AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is IN SERVICE STING ROG RAM TE p restored within the limits of Figure 3.1. 7-2 Amendment No. -242-SURVEILLANCE REQUIREMENTS SR 3.4.3.1 SR 3.4.3.2 Cooper SURVEILLANCE Verify the safety function lift setpoints of the SRVs and SVs are as follows: Number of Setpoint SRVs jQfilgL 2 1080 +/- 32.4 3 1090 +/- 32.7 3 1100 +/- 33.0 Number of Setpoint SVs jQfilgL 3 1240 +/- 37.2 Following testing, lift settings shall be within+/- 1%. -----------NOTE----

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after -reactor steam pressure and flow are adequate to perform the test. Verify each SRV opens when manually actuated.

3.4-7 SRVs and SVs 3.4.3 FREQUENCY In accordance with the lnse1"111ee TestiAg PFegFeffl 24 months Amendment No. INSERVICE TESTING PROGRAM ECCS -Operating

3.5.1 SURVEILLANCE

REQUIREMENTS (co n tinued) SR 3.5.1.6 SR 3.5.1.7 SR 3.5.1.8 Cooper SURVEILLANCE Verify the following ECCS pumps develop the specified flow rate aga i nst a system head corresponding to the specified reactor pressure. SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RA TE PUMPS PRESSURE OF Core Spray LPCI 4720 gprn 1 ?:. 15,000 gpm 2 113 psig ?:. 20 psig Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. FREQUENCY In accordance INSERVICE JC-TESTING Pr=egram PROGRAM Verify, with reactor 1020 and.::, 920 psig, the 92 days HPCI pump can develop a flow 4250 gpm against a system head corresponding to reactor pressure. -----*------NOTE--------

--Not req1Jired to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. Verify, with reactor 165 psig, the HPCI pump 24 months can develop a flow 4250 gpm aga i nst a system head corresponding to reactor pressure. (continued) 3.5-5 Amendment No. -242--

SURVEILLANCE REQUIREMENTS (c o nt i n u ed) SR 3.5.2.4 SR 3.5.2.5 SURVE I LLANCE Ver i fy each requ i red ECCS pump de v elops the spe c ified flow rate against a system head corresponding to the specified reactor pressure. SYS T EM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RA TE PUMPS PRESSURE OF cs LPCI ::_ 4720 gpm 7700 gpm 1 ::_ 113 ps i g ::_ 20 ps i g ------------N OTE--------Vessel i nject i on/spray may be excluded. Verify each required ECCS i njection/spray subsystem actuates on an actual or simulated automatic i n iti a ti on signal. ECCS -Shutdown 3.5.2 FREQUENCY In accordance with the INSERVICE TestiRg TESTING PROGRAM 24 months Cooper 3.5-10 Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 Amendment 180 SURVEILLANCE

NO TES--------------------

1. Valves and blind flanges in high radiation areas may be ver i fied by use of administrative means. 2. Not required to be met for PCIVs that are open under administrati v e controls. Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked , sealed, or otherwise secured and is required to be closed during accident conditions is closed. Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge. Verify the isolation time of each power operated, automatic PCIV , except for MSIVs , is within limits. 3.6-13 PC IVs 3.6.1.3 FREQUENCY Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed w i thin the previous 92 days 31 days In accordance

'ith th_e INSERVICE

-

..... ., Pregf8m PROGRAM (continued)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.6.1.3.6 Verify the isolation time of each MSIV is 3 seconds and 5 seconds. SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal. SR 3.6.1.3.8 Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation posit ion on an actua l or simulated instrument line -break. SR 3.6.1.3.9 Remove and test the explos ive squib from each shear isolation valve of the TIP System. SR 3.6.1.3.10 Verify leakage rate through each Main Steam line is :::; 106 scfh when tested at 29 psig. Cooper 3.6-14 PC I Vs 3.6.1.3 FREQUENCY In accordance with the IN I -*'--:i: . JC-TE est1Rg PR SERVICE STING OGRAM Pregfaffi 24 months 24 months 24 months on a STAGGERED TEST BASIS In accordance with the Primary Containment Leakage Rate Testing Program (cont i nued} Amendment No.

RHR Containment Spray 3.6.1.9 SURVEILLANCE REQUIREMENTS SR 3.6.1.9.1 SR 3.6.1.9.2 SR 3.6.1.9.3 Cooper SURVEILLANCE Verify each RHR containment spray subsystem manual, power opera t ed, snd automatic valve in the flow path that is not locked, sealed , or otherwise secured in position , is in the correct position or can be aligned to the correct position. Verify each required RHR pump develops a flow rat2 of > 7700 gpm through th e associated heat exchanger while operating in the suppression pool cooling mode. Verify each spray nozzle is unobstructed. 3.6-26 FREQUENCY 31 days Following maintenance which could result in nozzle blockage INSERVICE TESTING PROGRAM AmendmentNo.-rsa-RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE R EQUIREMENTS SURVEILLANCE FREQUE N CY SR 3.6.2.3.1 Verify each RHR s uppression pool cooling subsystem 31 days manual, power operated, and automatic valve in the flow path that is not locked , sealed, or otherwise secured in position, is i n the correct pos i t i on or can be aligned to the correct posit i on. SR 3.6.2.3.2 Verify each RHR pump de v elops a flow rate > 7700 In accordanoe gpm through the assoc ia ted heat exchange r while wit h the IAeef'liee operating in the suppr essi on pool cooling mode. r.stiAg Rrogr41m t\--\__ INSERVICE TESTING PROGRAM Cooper 3.6-32 Amendment S U RV EIL LA NCE REQUIREMENTS SURVEILLANCE SR 3.6.4.2.1

---NOTES----------------

1. Valves and blind f langes in high radiation areas may be verified b y u se of adm i nistrative means. 2. Not required to be met for SC I V s th at are open under admini s trative control s. ------------Verify e ach secondary cont ai nment i solation manua l valve and blind flange that is not locked , sealed, or otherwise secured and is requir e d to be closed during accident conditions is closed. SR 3.6.4.2.2 Verify the isolation time of each power operated automatic SCIV is within limits. SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation position on an actual or simula t ed actuation signal. Cooper 3.6-39 SCI Vs 3.6.4.2 FREQUENCY 31 days In accordance with th_e 1 NS TE ERV ICE STING OGRAM fest1flg Pregram PR 24 months Amendment Progra m s and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 lnseF\<iso Testing Pregram Cooper Thie 13rograFR 13rovieles 60Rtre l s for iRSOF\'iGe testing of ASME Coao Clase 1 , 2 , ana a pumps ar:id
a. Tostin*J fi:eEJuoncies to tl=te AS Me Coelo fQr Operati'm and Maintonaneo of 12ower Plants (ASMI!: OM Coele) and applicable Addenda are as felle-. .. s: ASME OM GeeJe anel appliGable Addor:ida torFRinolagy for insef\'iee testiflg eft*jYities Weeldy Monthly Quarterly or every 3 n1ontl"ls Semiannually er oYory 6 months Every 9 months Yearly or annually Biennially er e't<ef'Y 2 yeeFS Required for f:lerlor:miA!'J insoF¥ice t e stiea actiiAtios At least once per 7 days At least enee per a 1 elays At least onee 13er 92 elays At least eAee per 184 Elays At loaot eneo per 276 Elays At least onee per 366 days At least enee 13er 7a 1 elays 9. Tl=to provisigns of SR 3.0.2 are applieable to the above r:8quired Frequer:icies 111:1d to otner normel and eeeelereted Freqt1eneies specified 89 2 er less ifl IAserviee TestiAg PFegreffi fer 13erfel'ffliflg iAseFViee lestiAg aefrlitfes;
s. The 13rovisions of aR a.Q.a are a13plisabl0 to i nsoPAce tsstiRg act i vities; d. NotAiAg iA ASME OM Gede st=toll BC eoAstFUed to s1:113ersede the requiremeRt3 of aRy TS. (con ti nued) 5.0-10 Amend m en t No. -24.:l-NLS2016046 Attachment 3 Page 1 of 14 Attachment 3 Revised Technical Specifications Pages Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages 1.1-3 1.1-4 1.1-5 3.1-22 3.4-7 3.5-5 3.5-10 3.6-13 3.6-14 3.6-26 3.6-32 3.6-39 5.0-10

1.1 Definitions

DOSE EQUIVALENT 1-131 (continued)

INSERVICE TESTING PROGRAM LEAKAGE LINEAR HEAT GENERATION RATE (LHGR) Cooper Definitions 1.1 1-133, 1-134, and 1-135 actually present. The DOSE EQUIVALENT 1-131 concentration is calculated as follows: DOSE EQUIVALENT 1-131=(1-131)+0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and A i r Concentration and Dose Conversion Factors for Inhalation , S ubmersion , and Ingestion ," 1989. The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f). LEAKAGE s hall 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; b. Unidentified LEAKAGE All LEAKAGE i nto the drywell that is not identified LEAKAGE; c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. The LHGR shall be the heat generation rate per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. (continued) 1.1-3 Amendment No.

1.1 Definitions

LOGIC SYSTEM FUNCTIONAL TEST MINIMUM CRITICAL POWER RA TIO (MCPR) MODE OPERABLE -OPERABILITY PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) RA TED THERMAL POWER (RTP) REACTOR PROTECTION SYSTEM(RPS)RESPONSE TIME SHUTDOWN MARGIN (SOM) Cooper Definitions 1.1 A LOGIC SYSTEM FUNCTIONAL TEST shall be a tes t of all logic components required for OPERABILITY of a logic circuit, from as close to the senso r as practicable up to , but not i ncluding, the actua t ed device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential , overlapping, or total system steps so that the entire logic system is tested. The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation( s) to cause some point in the assembly to experience bo i ling transition, divided by the actual assembly operating power. A MODE shall correspond to any one inclusive combination of mode switch position , average reactor coolant temperature , and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its spec ified safety function(s) and when all necessary attendant in strumentation, controls , normal or emergency electr i cal power, cooling and seal water , lubrication, and other auxiliary equipment that are required for the system, subsystem , division, componen t, or device to perform its spec ified safety function(s) a r e also capable of performing the i r related support function(s). The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits , including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6. 7. RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2419 MWt. The RPS RESPONSE TIME shall be that time segment from the time the sensor contacts act u ate to the time the scram solenoid valves deenergize. SOM shall be the amount of reactivity by which th e reacto r is subcritical or would be subcritical throughout the operating cycle assum ing t ha t: (continued) 1.1-4 Amendment No.

1.1 Definitions

SHUTDOWN MARGIN (SOM) (continued)

STAGGERED TEST BASIS THERMAL POWER TURBINE BYPASS SYSTEM RESPONSE TIME Cooper a. The reactor is xenon free; Definitions 1.1 b. The moderator temperature is ::::: 68°F , corresponding to the most reactive state; and c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With cont r ol rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM. A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve. 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.

1.1-5 Amendment No. I SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.7.6 SR 3.1.7.7 SR 3.1.7.8 SR 3.1.7.9 Cooper SURVEILLANCE Verify each SLC subsystem manual valve in the flow path that is not locked , sealed , or otherwise secured in position, is in the correct position or can be aligned to the correct position. Verify each pump develops a fl ow rate 38.2 gpm at a discharge pressure 1300 psig. Ver i fy flow through one SLC subsystem from pump into reactor pressure vessel. Verify all heat traced piping between storage tank and pump suction is unblocked.

3.1-22 SLC System 3.1.7 FREQUENCY 31 days In accordance with the INSERVICE TESTING PROGRAM 24 months on a STAGGERED TEST BASIS 24 months Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 Amendment No.

SURVEILLANCE REQUIREMENTS SR 3.4.3.1 SR 3.4.3.2 Cooper SURVEILLANCE Verify the safety function l i ft setpoints of the SRVs and SVs are as follows: Number of Setpoint SRVs (ps i g) 2 1080 +/- 32.4 3 1090 +/- 32.7 3 1100 +/- 33.0 Number of Setpoint SVs (psig) 3 1240 +/- 37.2 Following testing, lift settings shall be within +/- 1 %. -------------------------------NOTE----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. Verify each SRV opens when manually actuated.

3.4-7 SRVs and SVs 3.4.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM 24 months Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.6 SR 3.5.1.7 SR 3.5.1.8 Cooper SURVEILLANCE Verify the following ECCS pumps develop the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray LPCI ?! 4720 gpm 1 ?! 15 , 000 gpm 2 ?! 113 psig ?! 20 psig -------------------------------NOTE--------------------------------

Not required to be performed unt i l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. Verify , with reactor pressure s 1020 and ?! 920 psig , the HPCI pump can develop a flow rate ?! 4250 gpm against a system head corresponding to reactor pressure.

NOTE------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. Verify , with reactor pressure s 165 psig, the HPCI pump can develop a flow rate ?! 4250 gpm against a system head corresponding to reactor pressure. 3.5-5 ECCS -Operating

3.5.1 FREQUENCY

In accordance with the INSERVICE TESTING PROGRAM 92 days 24 months (continued)

Amendment No.

ECCS -Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS continued SR 3.5.2.4 SR 3.5.2.5 Cooper SURVEILLANCE Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs LPCI 4720 gpm 7700 gpm 1 1 113 psig 20 psig --------------------------------NOTE-----

Vessel injection/spray may be excluded.

FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Verify each required ECCS inject i on/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal. 3.5-10 Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 Cooper SURVEILLANCE

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 valve and blind flange that is located inside primary containment and not locked, sealed , or otherwise secured and is required to be closed during accident conditions is closed. Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge. Verify the isolation time of each power operated, automatic PCIV, except for MS IVs, is within limits. 3.6-13 PC I Vs 3.6.1.3 FREQUENCY Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4 , if not performed within the previous 92 days 31 days In accordance with the INSERVICE TESTING PROGRAM (continued)

Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.6 SR 3.6.1.3. 7 SR 3.6.1.3.8 SR 3.6.1.3.9 SR 3.6.1.3.10 Cooper SURVEILLANCE Verify the isolation time of each MSIV is 3 seconds and 5 seconds. Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal. Verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break. Remove and test the explosive squib from each shear isolation valve of the TIP System. Verify leakage rate through each Main Steam line is 106 scfh when tested at 29 psig. 3.6-14 PC I Vs 3.6.1.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM 24 months 24 months 24 months on a STAGGERED TEST BASIS In accordance with the Primary Containment Leakage Rate Testing Program (continued)

Amendment No.

RHR Containment Spray 3.6.1.9 SURVEILLANCE REQUIREMENTS SR 3.6.1.9.1 SR 3.6.1.9.2 SR 3.6.1.9.3 Cooper SURVEILLANCE Verify each RHR containment spray subsystem manual , power operated , and automat i c valve in the flow path that is not locked , sealed , or otherwise secured in position, is in the c orrect position or can be aligned to the correct position. Verify each required RHR pump develops a flow r ate of > 7700 gpm through t he a s sociated heat exchanger while oper a ting in the su p pression pool cooling mode. Verify each spray nozzle is unobstructed.

3.6-26 FREQUENCY 31 days In accordance with the lNSERVICE TEST I NG PROGRAM Following maintenance wh i ch could result in nozzle blockage Amendment No.

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SR 3.6.2.3.1 SR 3.6.2.3.2 Cooper SURVEILLANCE Verify each RHR suppression pool cooling subsystem manual , power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

Verify each RHR pump develops a flow rate > 7700 gpm through the associated heat exchanger while operating in the suppression pool cooling mode. 3.6-32 FREQUENCY 31 days In accordance with the INSERVICE TESTING PROGRAM Amendment No.

SURVEILLANCE REQUIREMENTS SR 3.6.4.2.1 SURVEILLANCE

N()TES----------

1. Valves and blind flanges in high radiation areas may be ver i fied by use of administrative means. 2. Not required to be met for SCIVs t h a t are open under adm i nistrative controls.

SCIVs 3.6.4.2 FREQUENCY Verify each secondary cont a inment isola t ion ma n ual 31 days SR 3.6.4.2.2 SR 3.6.4.2.3 Cooper valve and blind flange that is not locked , sealed, or otherwise secured and is required to be closed during accident conditions is closed. Verify the isolation time of each power ope r ated automatic SCIV is within limits. Verify each automatic SCIV actuates to the isolation position on an actual or simulated actuation signal. 3.6-39 In accordance with the INSERVICE TESTING PR()GRAM 24 months Amendment No.

5.5 Programs

and Manuals (continued) 5.5.6 (Deleted)

Cooper 5.0-1 0 Programs and Manuals 5.5 (cont i nued) Amendment No.

NLS2016046 Attachment 4 Page 1 of 16 Attachment 4 Proposed Technical Specifications Bases Changes (Mark-up)

-Information Only Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages 83.0-13 83.1-44 83.4-16 83.5-10 83.5-12 83.5-13 83.5-28 83.5-29 83.6-26 83.6-27 83.6-44 83.6-49 83.6-54 83.6-66 83.6-80 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 t o all Specifications i n Sec ti ons 3.1through3.10 and apply at all times, unless otherwise stated. 1E----SR 3.0.2 and SR 3.0.3 apply in Chapter 5 when invoked by a Chapter 5 ------------------

--i Specification. L......:.------------------'

SR 3.0.1 Cooper SR 3.0.1 establishes the requi r ement tha t S R s must be met during the MODES or other specified conditi o ns in the Applicability for which the requirements of the LCO apply , unless o t herwise specified i n the individual SRs. This Spec i fication is to ensure that Surveillances are performed to verify t he OPERABILITY of systems and components, and that variables are w i thin spec i fied limits. Failure to meet a Surveillance within the specified Frequancy , 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. Noth i ng in this Specification, however , is to be construed as implying that systems or components are OPERABLE when: a. The systems or compone n ts are known to be inoperable, although still meeting the SRs; or b. The requirements of the Surveillance(s) are k nown to be not met be t ween 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 Spec i fication. Surveillances, inc l uding Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2 , prior to returning equipment to OPERABLE status. 8 3.0-13 09.'1 !l'OQ BASES SLC System B 3.1.7 SURVEILLANCE REQUIREMEN TS (continued)

Cooper positive reactivity effects encountered during power reduction , cooldown of the moderator , and xenon decay. This test con firm s one point on the pump design curve and is indicative of overall performance.

Such ins ervice tests co nfirm component OPERABILITY , and detect incipient failures by indicating abnormal performance.

The Frequency of this Surveillance is in acco dance with the lflsefYiee Testifig Pregrefl'I.

TESTING PROGRAM I SR 3.1.7.8 and SR 3.1. 7.9 These Surveillances ensure that there is a functioning flow pa th from the boron solut ion sto rage tank to the RPV, including the firing of an explosive valve. The replaceme nt charge for the explosive valve shall be from the same manufactured batch as th e one fired or from another batch that has been certified by having one of tha t batch successfu ll y fired. The pump and explosive valve te sted should be alternated such that both complete flow paths are tested every 48 months at al t ernati n g 24 month intervals. The Surveillance may be performed in separa te steps to prevent inject in g boron into the RPV. An accepta b le method for verifying flow from the pump to the RPV is to pump deminera li zed water from a test tank through one SLC subsystem and into the RPV. The 24 month Frequency is based on the ne ed to perform this Surveillance under th e conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency; therefore , the Frequency was conc l uded to be acceptable from a reliability standpoint.

Demonstrat ing that all heat traced piping between the boron solution storage tank and the suct i on inlet to the injection pumps i s unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for ve rifyin g that the suction piping is unblocked i s to manually initiate the system , except the explosive valves , and pump from the storage tank to the t est tank. Upon completion of this verification, the pump suction piping must be flushed with demineralized water to ensure piping between the storage tank and pump suction is unblocked , The 24 month Frequency is acceptable since there is a low probab ilit y that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the temperature verification of this piping B 3.1-44 11/25112 BASES APPLICABILITY ACTIONS SRVs and SVs 8 3.4.3 In MODES 1, 2, and 3, 7 of 8 SRVs and 3 SVs must be OPERABLE , since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The SRVs and SVs may be required to provide pressure relief to limit peak reactor pressure. In MODE 4 , decay heat is low e nough for the RHR System to provide adequate cooling, and reactor pre ssure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or acc idents. In MODE 5, the rea ctor vessel head is unbolted or removed and the reactor is at atmospheric pressure.

The SRV and SV function is not needed during these conditions. A.1 and A.2 With the safety function of one or more of the required SRVs or SVs inopersble , a transient may result in the violation of the ASME Code limit on reactor pressure. If the safety function of one or more of the required SRVs or SVs i s inoperable, the plant must be brought to a MODE i n 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 Comple t ion T im es are reasonable , based on operating experience, to reach required plant conditions from full power condit ions in an orderly manner and w i thout challenging pla nt systems. SURVEILLANCE REQUIREMENTS SR 3.4.3.1 This Surveillance requires that the SRVs and SVs will open at the pressures assumed in the safety analysis of Refe re nce 3. The INSERVICE demonstration of the SRV and SV safety function lift settings must be TESTING PROGRAM r--,,eFtot:me d during shu t down , since this is a bench test , to be done in accordance wr * * . The lift setting pressure Cooper sha ll correspond to ambient condit i ons of the valves at nominal operating temperatures and pressures. The SRV setpoint is +/- 3% for OPERABILITY

however, the valves are reset to +/- 1 % during the Surve ill ance to allow for drift. B 3.4-16 03/06112 BASES SURVEILLANCE REQUIREMENTS INSERVICE TESTING PROGRAM SR 3.5.1.2 (continued}

ECCS -Operating B 3.5.l OPERABILI TY. Also, this SR does not apply to valves that are locked , sealed, or other wise secured in position since these were verified to be in the corr ec t positio n prior to locking, sealing, or securing.

A valve that receives an initia tio n signal is allowed to be in a nonaccident position pro v ided th e valve will automatically repositi on in the proper stroke time. This SR does not r equire a ny testing or valve manipul ation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.

This SR does not apply to va lv es that c annot be inadvertently mis align ed, such as check valves. For the HPCI System, this SR also inc ludes the steam flow pa th for the turbine and the flow controller position.

Frequency of this SR was derived from t he for performing valve tes ting at least on ce every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would only affect 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 actual shutdown cooling permissive p re ssure, the RHR System may be required to operate in the s hu tdown cooling mode to re m ove decay hea t and sensible heat from the reactor. Therefor e , this SR is modified by a Not e that allows LPC I subsystems to be considered OPERABLE during alignment and operation for decay heat rem ov al, if capable of being manually realigned (remote or loc al} to the LPCI mode and not otherwise inoperable.

Al ig nment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. At the low pressures and decay heat loads associated with operation in MODE 3 with reactor steam dome pressure less than the shutdown cooling permissive pressure , a reduced complement of low pressure ECCS subsystems should pro vid e the required cooling, thereby allowing operation of RHR shutdown cooling, when necessary.

*------**--**-----*-. Cooper B 3.5-10 e

BASES ECCS -Operating B 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

INSERVICE TESTING PROGRAM Cooper the LPCI subsystem. Acceptable method s of de-energizing the valve include de-energ izing breaker control power , rack i ng ou t the breaker or r emoving the breake r. The specified Frequency i s once during reactor startup before THERMAL POWER is> 25% RTP. However, this SR i s modified by a Note that states the Surveillance is only requi r ed to be performed if the last performance was more than 31 days ago. Therefore, implementation of

  • e uires this test to be performed during reactor startup before exceeding 25% tion during reactor startup prior to reaching > 25% RTP is an exception to the generic valve cycling Frequency of 92 days , but is considered acceptable due to the demonst rat ed reliability of these valves. If the valve is in operable and in the open position, the associated LPCI subsystem must be declared inoperable.

SR 3.5.1.6. SR 3.5.1.7. and SR 3.5.1.8 The performance requirements of the low pressure ECCS pumps are determined through application of the 1 O C F R 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASME Code for Operation and Maintenance o f Nuclear Power Pla n ts requ i rements for the ECCS pumps) to verify tha t the ECCS pumps will develop the flow rates required by the respective analyses. T he low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference

8. The pump f low rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pum p outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressu re present during a LOCA. The flow tests for the HPCI System are performed at two diff erent pressure ranges such that system capability to provide rated flow against a system head corresponding to reactor pressure is tested at both the higher and lower operating ranges of the system. The required system head B 3.5-12 0 4/28/10 BASES ECCS -Operating B 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

INSERVICE TESTING PROGRAM Cooper should overcome the RPV pressure and associated discharge line losses. Adequate reactor pressure must be available to perform these tests. Additionally , adequa te steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Therefore, sufficient time i s allowed after adequate pressure and flow are achieved to perform these tests. Adequate reactor steam pressure must 920 psig to perform SR 3.5.1.7 145 psig to perform SR 3.5.1.8. Adequate steam flow is represented by turbine bypass valves at least 30% open , or total steam flow 10 6 lb/hr. Reactor startup is allowed p ri or to performing the low pressure Surveillance te st because the rea ctor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily c ompleted and there is no indication or reason to believe that HPCI is inoperable. Therefore , SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes t hat state the Surveillances are not required to be p erformed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for the flow tests after required pressure and flow are reached are sufficient to ach i eve stable conditions fo r testing and provides a reasonable time to complete the SRs. For SR 3.5.1.8 , while adequate pressure can be reached pr i or to the required Applicab ilit y fo r HPCI , the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance of the Note would not apply until entering the Applicability

(>150 psig) with adequate steam flow. The Frequency for SR 3.5.1.6 and SR 3.5.1. 7 is in accordance with the requirements. The 24 month Frequency for SR 3.5.1.8 i s based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, wh i ch is based on the refueling cycle. Therefore , the Frequency was concluded to be acceptable from a reliability standpoint. B 3.5-1 3 11/26112 BASES RCIC System B 3.5.3 SURVEILLANCE REQUIREMENTS (continued)

Coope r valves that cannot be inadvertently misa l igned, such a s check valves. For the RCIC System , t his SR also includes t he steam flow path for the turbine and the flow c ont ro l l er osition. INSERVICE TESTING PROGRAM The 31 day Frequency of this was derived from the h:isen15 0 psig) with adequate steam flow. A 92 day Frequency for SR 3.5.3.3 i s cons i stent w it h the lnservise Testing Pregra!TI requirements. The 24 month Frequency for SR 3.5.3.4 is based on the need to perform the Surveillance u nder condit i ons that apply just prior to or during a startup f rom a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency , wh i ch i s based on the refueling cycle. Therefore , the Frequency was concluded 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 satisfactor il y. This Surveillance verifies that , with a required system initiation signal (actual or simulated), the automatic initiation log i c of the RCIC System will cause the sys tem to operate as designed, including actuation of the sys tem throughout its emergency operat in g sequence; that is , automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low water level (Level 2) signal received subsequent to a n RPV high water level (Level 8) trip and that the suction i s automatically transferred from the ECST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps th i s Surve ill ance to provide complete testing of the assumed design function. The 24 month Frequency is based on the need to perform some of the surveillance procedures which satisfy this SR under the conditions that apply during a plant outage and the potential for an unplanned trans i ent if those part i cular procedures were performed with the reac t o r at power. Operating exper ien ce has shown that t hese componen t s 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.

8 3.5-2 9 11/29112 BA S E S PC I Vs B 3.6.1.3 SURVEILLANCE SR 3.6.1.3.3 (eer=1tiF1uedT REQ U IREMENTS Cooper controls consist of stationing a dedicated opera t or at the controls of the val v e , who is in c ontinuou communicat ion with the control room. In thi s way, the penet ration ca n be ra p idly iso l a t e d when a nee d for primary con t ainment isolation is indi cated. S R 3.6.1.3.4 The trav ers i ng incore probe (TIP) shear is olat io n valves are actuated by explosive charges. Su*vei:iance of explosive charge continuity provides assurance that TIP valves will actua t e whan required.

Other administrative cont rols, such as those th at lim i t the shelf li f e of the explosive charges , must be followed.

The 31 day Frequency is based on operating experience that h as demonstrated the reliability of the e xp lo s ive cha r ge continuity. SR 3.6.1.3.5 Verifying the isolation t ime o f each power operated auto mat i c PCIV i s within limits is required to demonstrate OPERABILITY.

MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the safety analyses.

The isolation time an d Freque ncy of this SR are in acc ordance with the re q uiremen ts of the lneeF¥ioe TeetiAg Pregrafl"F.

SR 3.6.1.3.6 INSERVICE TESTING PROGRAM Veri f ying that t he isolation time of each MSIV is within the speci fie d limits is required to demons t rate OPERABILITY.

The isolation time test ensures that the MSIV will i solate in a t ime period that does not exceed the times as sumed in the OBA and t ransient analyses. This ensures that the (continued) B 3.6-26 MareR 8 , 2000 BASES PC I Vs B 3.6.1.3 SURVEILLANCE R E QUIREMENTS (continued)

INSERVICE TESTING PROGRAM Cooper calcu lat ed radiolog ical consequences of these events remain within 1 O CFR 100 limit s. The Frequency of this SR is in accordance with the requirements of th Testing Program. Automat ic PCIVs close on a prim ary containment is olation s ign a l to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation pos ition on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1 , "Primary Containment Isolation Instrumentation

," overlaps this SR to provide complete testing of the safety function. Th e 24 month Frequency was developed consideri ng it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations wou l d disrupt the normal operation of many critical components. Operating exper i ence has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore , the Frequency was concluded to be acceptable from a rel i ability standpoint.

SR 3.6.1.3.8 This SR requires a demonstration that a rep re sentative sample of reactor instrumentation line excess flow check valves (EFCVs) are OPERABLE by verifying that each valve actuates to the i solation pos ition on an actual or simulated instrument line break. The representative sample consists of an approx im a te ly equal number of EFCVs , such tha t each EFCV is tested at least once every 10 years (nominal). Th i s SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during the postulated instrument line break event. The 24 month Frequency is based on the need to perform the Surveillance under the condit i ons that apply during a plant outage and the potential for an unp l anned transient if the Surve ill ance were performed with the reactor at power. The nom in a l 1 O year interval is based on other performance-based testing programs , such as lnserv ice Testing (snubbers) a nd Option B to 10 CFR 50 , Appendix J. Furthermore , any EFCV failures will be evaluated to determine i f additional testing in that test interval is warranted to ensure overall reliability i s maintained. Operating experience has demonstrated that these components are highly re liabl e and that failures to isolate are very infr equent. Therefore , testing of a representative sample was concluded to be acceptable from a rel i ability standpo int. B 3.6-27 11125/12 BASES Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.7 SURVEILLANCE REQUIREMENTS (cont i nued) REFERENCES Cooper SR 3.6.1.7.2 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 ensures that the safety ana l ysis assumpt i ons are valid. The 92 day Frequency of th i s SR was develop e d based upon IAservise Testing Pregrem requireme n ts t o p erform valve t esting at le once every 92 days. SR 3.6.1.7.3 INS E RVICE T EST I NG PROGRAM Demonstrat i on of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regard i ng vacuum breaker full open differential pressure 0.5 psid is valid. The 24 month Frequency is based on the need to perform some of the surveillance procedures which satisfy this SR under the conditions that apply during a plant outage and the potential for an unplanned transient if those particular procedures were perfo r med with the reactor at power. For this unit , the 24 month Frequency has been shown to be acceptable , based on operating experience, and i s further justified because of other Surveillances performed at shorter Frequencies that convey the proper functioning status of each vacuum breaker. 1. Bodega Bay Preliminary Hazards Summary Report , Appendix I , Docket 50-205 , December 28 , 1962. 2. USAR, Section V-2.3.6. 3. 10 CFR 50.36(c)(2)(ii). B 3.6-44 11/25/12 BASES ACTIONS (co nti nued) SURVEI L LANCE REQU IREME NTS Coop er C.l and C.2 Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 If any Require d Action and a ss o ci ated Co mpl e t ion T i me cannot be met, the plant must be brought to a in wh ich the LCO does not apply. To ach ieve t h is status, t h e p lan t mu st be brought t o at least MODE 3 within 1 2 ho u rs and to MODE 4 w i t h in 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Th e all ow ed Completion Time s are reasonable, b ase d on operating to reach the re quired p lant condi t;ons from full powe r conditions i n an orderly manner and without challenging pl ant systems. SR 3.6.I.8.l Each vacuum breaker is ver ifi ed closed (except when th e vacuum breaker is per fo rmin g it s intended design function) io ens u re that this potenti a l large bypass leakage pat h is not present. This Su r veillance i s performed by observing t he vacuum breaker pos i tion indication or by p erforming a lea k test that confirms that the bypass area between the drywell and suppression chamber is less than or equivalent to a one i n ch diameter hole. If th e bypass test fails, not only must the vacuum breaker(s) be considered op en an d th e appr opriate Conditions and Required Actions of this LCO be entered, but al so the appropriate Cond i tions and Required Ac t ions of LCO 3.6.1.1, Primary Containment, must be en te red. The 14 day Frequency is based on engineer i ng judgment, is con s i d ere d adequate in vi ew of other indications of vacuum br ea k er status available to operatio ns pers onnel, and has been shown to be acceptable through operating experience.

A Note is added to this SR which allows suppre ss io n to-drywell vacuum b r eakers opened in con j unc tion with the performance of a Surveillan c e to not b e consid e red a s failing t his SR. These periods of ope ni ng vacuum breakers are con t rol l ed b y plant procedures and do not represent inope r a b le vacuum b reakers. SR 3.6.1.8.2 Each required vacuum b reak er must be cycled to ens ure that it opens adequately to perform i ts d esign fun ction and returns to the fully closed pos iti on. T h i s ensures t h e s af e ty analys is assumptions are valid. T he 31 day F requen cy of this SR bas ed TestiA! P**!*** TESTING PROGRAM L_ (eonHntted) 8 3.6-49 Revision 0 BASES RHR Containment Spray B 3.6.1.9 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.9.2 INSERVICE TEST I NG PROGRAM REFERENCES Cooper Verifying each require d RHR pump deve l ops a flow rate > 7700 gpm while operating in the suppressio n pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. It is tested in the pool cooling mode to demonst r ate pump OPERABILITY without spraying down equipment in the drywell. Flow is a normal test of centrifugal pump p erformance required by the ASME Code,Section X t (Ref. 4). Th i s test confirms one point on the pump performance curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

The Frequency of this SR is in accordance with the lnserviGe Testing Pregram. SR 3.6.1.9.3 This Surveillance is performed following maintenance which could result in nozzle blockage by introduction of air to verify that the spray nozzles are not obstructed and t hat flow will be provided when req u ired. The Frequency is adequate to detect degradation in performance due to the passive nozzle design and its normal y dry state and has been shown to be acceptable through operating experience. 1. USAR, Chapter XIV, Section 6.3. 2. USAR, Chapter V, Section 2. 3. EE 01-035, EQ Temperature Profile i n Containment based on Small Steam Line Break and DBA-LOCA Analysis. 4. ASME Code for Operation and Maintenance of Nuclear Power Plants. B 3.6-54 02.'22/16 BASES RHR Suppression Pool Cooling B 3.6.2.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.2.3.2 Verifying t hat each RHR pu mp develops a flow rate 7700 gpm while operating in the suppr ess ion pool cooling mode with flow through the associated heat exchanger ensures thc::t pump performance has not degraded dur i ng the cyclo. Flow is a normal test of centrifugal pump p erfo rmance required by ASME Code (Ref. 4). This tes t confirms one point on the pump design curve, and the results are * *

  • era I performance.

Such inservice

  • on irm component OPERABILITY, trend performance, and detect incipient failures by indicating abno rmal performance. The Frequency of this SR is in accordance with the * * .-I N-S--E"""""R-V-IC_E_T-ES_T_l_N

__ G __ P ___ R_O_G __ RA __ M__, REFERENCES

1. USAR, Section XIV-6. 2. 10 CFR 36(c)(2)(ii).
3. NEDC 94-0348, C & D 4. ASME Code for Operat i on and Maintenance of Nuclear Power Plants. tests Cooper B 3.6-66 02122,i1s I

BASES SCIVs B 3.6.4.2 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES Cooper reasons. Therefore, t he p r obability of misalignme n t of these isolation devices, once they h ave been verified to be in the proper position, is l ow. A second Note has bee n included t o clar i fy that SCIVs that are open under administrati v e controls are no t required to meet the SR during th e time the SCIVs a r e o p en. These controls consist of stationing a dedicated operator a t the controls of the valve , who is in continuous communication wi t h the control room. In this way , the penetration can be rap id ly isolated whe n a need for secondary containment isolation is indica t ed. SR 3.6.4.2.2 Verifying that the isolation time of each power operated automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses.

The isolation time and Frequency of this SR are in accordance with the lns&r.<ice Testing Pi:ogram.

PNSERVICE TESTING PROGRAM v SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary containment isolation signal is requi r ed t o minimize l eakage of radioactive material from secondary containment following a OBA or other accidents.

This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provi d e complete testing of the safety function. Th e 24 month Frequency is based on the need to perform t his 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. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

1. USAR, Section V-3.0. 2. USAR, Section XIV-6.0. 3. USAR, Section XIV-6.3. B 3.6-80 e2 1 22 1 1s I
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