Text

Submittal of Developmental Revision G of the Unit 2 Technical Specification (TS) and Technical Specification Bases
	ML12065A036
Person / Time
Site:	Watts Bar
Issue date:	02/28/2012
From:	Hruby R; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML12065A036 (44)
	v • d • e

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 February 28, 2012 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 10 CFR 50.36 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant (WBN) Unit 2 - Submittal of Developmental Revision G of the Unit 2 Technical Specification (TS) and Technical Specification Bases

Reference:

E-mail from Justin Poole (NRC) to Gordon Arent (TVA) dated January 24, 2012, "Draft Questions on Technical Specifications" This letter provides Developmental Revision G of the WBN Unit 2 TS and TS Bases. provides a listing of the changes made to Revision F that are included in Revision G. Enclosure 2 responds to the questions and comments provided by the reference on the WBN Unit 2 Steam Generator Program TS. Enclosure 3 provides the annotated pages associated with the changes to the WBN Unit 2 TS and TS Bases. provides a clean copy of Revision G on optical media.

There are no new regulatory commitments associated with this letter.

I declare under penalty of perjury that the foregoing is true and correct to the best of my knowledge. Executed on the 2 8 th day of February, 2012.

Respectfully, Raymond A. Hruby, Jr.

General Manager, Technical Services Watts Bar Unit 2 Enclosures (see page 2) , .1

U.S. Nuclear Regulatory Commission Page 2 February 28, 2012

Enclosures:

1. List of Changes to the WBN Unit 2 TS and TS Bases

2. Response to NRC Draft Questions on Technical Specifications

3. WBN Unit 2 TS and TS Bases Developmental Revision G Mark-ups

4. WBN U2 TS and TS Bases Developmental Revision G Optical Media cc (Enclosures):

U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

Enclosure I List of Changes to the WBN U2 TS and TS Bases

Change to SR 3.3.1.3 on NIS power and associated TS Bases SR 3.3.1.3.

, Change to TS Table 3.3.1-1, Source Range Neutron Flux Allowable Value.

Change to TS Table 3.3.1-1, Reactor Coolant Pump (RCP) Undervoltage Allowable Value and Nominal Trip Setpoint, associated TS Bases B 3.3.1, and TS Bases B 3.3.1 References.

Change to TS Table 3.3.1-1, Turbine Trip Low Fluid Oil Pressure Allowable Value and the applicable SR.

Change to TS Table 3.3.2-1, Auxiliary Feedwater Trip of all Turbine Driven Main Feedwater Pumps Allowable Value.

Change to SR 3.4.12.1 and 3.4.12.2, cold overpressure arming temperature.

Change to TS Bases B 3.7.7 providing power and train alignments for the Component Cooling Water System.

Change to TS Bases B 3.3.6 describing Containment Vent Isolation Instrumentation and associated changes to TS Basis B 3.3.8 Auxiliary Gas Treatment System Actuation Instrumentation, TS Bases B 3.7.12 on Auxiliary Building Gas Treatment System performance, and TS Bases B 3.9.8 Reactor Building Purge Air Cleanup Unit performance during refueling operations or fuel movement.

Change to TS Bases B 3.8.4 on DC Sources to update the number of battery chargers needed for two-unit operation and to match current Unit 1 wording.

Change to TS 4.3.1, Criticality describing acceptable spent fuel pool loading patterns.

Changes to Limiting Condition for Operation (LCO) 3.4.17, TS Section 5.7.2.12 and associated Bases B 3.4-13 and B 3.4.17 to incorporate appropriate information from Technical Specification Task Force (TSTF) Traveler 510 R2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," and to match the description to the Chapter 15.5 dose analysis for Main Steam Line breaks.

E1-1

Enclosure 2 Response to NRC Draft Questions on Technical Specifications The draft questions are provided below by number with the associated response below the individual question or related group of questions.

1. Section 5.7.2.12.d.I indicates that 100% of the tubing will be inspected in the first refueling outage. Since Watts Bar Unit 2 may replace steam generators(i.e., have a first refueling outage without replacing steam generatorsand may have a first refueling outage following steam generatorreplacement), please discuss your plans to modify this requirementto ensure both conditions are addressed(e.g., "...duringthe first refueling outage following steam generatorinstallation.").

2. Since your proposedtechnical specificationswere prepared,the NRC staff has approved Technical Specification Task Force Traveler510 (TSTFI -510), Revision 2, "Revision to Steam GeneratorProgramInspection Frequenciesand Tube Sample Selection" (ML110610350). The staff's review is summarized in a Notice of Availability dated October 19, 2011 (MLl12101604). Since this TSTF Traveleraddresses implementation issues associatedwith the inspectionsperiods and addresses other administrative changes and clarificationswhich improve the technicalspecifications, please discuss your plans to incorporatethe TSTF-510, Revision 2, changes into your proposed technical specifications.

Response to questions 1 and 2:

The WBN U2 developmental TS has been revised to incorporate TSTF-510 on steam generator (SG) program inspection frequencies and tube sampling. This revision addresses the information requested in questions 1 and 2.

3. On page B 3.4-73, the applicable safety analyses for operationalleakage are provided.

In this discussion, you indicate that the pre-accidentprimary-to-secondaryleakage for a main steam line break is 1 gallon per minute from one steam generator. The standard technical specification wording indicates that the leakage may increaseto this value during the accident. Please confirm that your accident analysis assumes that the leakage priorto a main steam line break is 1 gallon per minute (in which case the actual rate of leakage during the accident would be greaterthan 1 gallon per minute due to the higher differential pressure during the accident).

Response to question 3:

The discussion on page B 3.4-73 has been revised to state that for the main steam line break (MSLB) the pre-accident primary-to-secondary leakage is 150 gallons per day (gpd) per SG. At the time of the rupture of the steam line, it is assumed that the total primary-to-secondary leak rate in the faulted SG is 1 gallon per minute (gpm). The primary-to-secondary leak rate in the non-faulted SGs is maintained at 150 gpd per SG.

E2-1

Enclosure 2 Response to NRC Draft Questions'on Technical Specifications

4. On page B 3.4-73, the applicable safety analyses for operationalleakage is provided.

There are two paragraphsfrom the standardtechnical specifications that are not included in this section for Watts Bar Unit 2. These paragraphsaddress other accidents that involve secondary steam release to the atmosphereand the assumptions for a steam generatortube rupture [SGTR] accident. Please discuss why these paragraphs were not included in your Bases.

Response to question 4:

One reason that these paragraphs are not included in our Bases is because WBN does not conform to revision 3 of the Standard TS (STS). They match the Unit 1 TS to the extent possible. Unit l's TS generally match a mix of revision 0 and revision 1 of the STS but do not fully conform to those standards.

More importantly, the paragraphs do not reflect the WBN U2 safety analyses. The STS paragraph on SGTR states "The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is only briefly released via safety valves and the majority is steamed to the condenser. The [1 gpm] primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential."

While the write-up in the paragraph above represents the most probable and desirable plant response to a SGTR event, the WBN accident analyses use more conservative assumptions. The WBN dose analysis for the SGTR assumes a relatively large secondary side release from the faulted SG directly to the atmosphere for the period of 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and a smaller atmospheric release for the 2 to 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. In addition, the SGTR analysis is the controlling analysis for control room operator dose due to the assumptions mandated for the Iodine spike cases.

The second paragraph in the STS but not in the WBN U2 specifications states, "The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes the entire [1 gpm] primary to secondary LEAKAGE is through the affected generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits)." While the MSLB is more limiting than the SGTR for offsite dose consequences at WBN, as noted in the preceding paragraph that is not the case for control room doses.

TVA concludes that the current discussion on page B 3.4-73, as amended by response 3 above, correctly reflects the WBN U2 accident dose analyses and that inclusion of the two paragraphs from the STS discussed above would not be appropriate.

5. On page B 3.4-97, the applicablesafety analyses for steam generatortube integrity is provided. Please clarify the following sentence: "In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from E2-2

Enclosure 2 Response to NRC Draft Questions on Technical Specifications 150 gallons per day (gpd)per steam generatorand 1 gallon per minute (gpm) in the faulted steam generator." In addition, the standardtechnical specificationsmake it clear that the leakage may increaseas a result of accident induced conditions to a specific value. This thought is not capturedin your proposal. Pleaseclarify why this was not included in your Bases.

Response to question 5:

The change in primary-to-secondary leakage in a faulted SG at the start of a MSLB is an assumption made in the dose analysis to add conservatism. The assumed higher leak rate is not the result of a specific physical process or analysis. There is no regulatory requirement to assume a total SG leak rate above the TS allowable in the dose analysis.

This accepted regulatory position is shown clearly in Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," which states "For facilities with traditional SG specifications (both per generator and total of all generators), the leakage should be apportioned between affected and unaffected SGs in such a manner that the calculated dose is maximized. For example, for a four loop facility with a limiting condition for operation of 500 gpd for any one generator not to exceed 1 gpm from all generators, it would be appropriate to assign 500 gpd to the faulted generator and 313 gpd to each of the unaffected generators." In this example the tube leak rate does not increase at all after an accident. [Note that the sum of 500 gpd + (3x313 gpd)= 1 gpm.] The approach to post-accident SG tube leakage prescribed in this Regulatory Guide is found in other NRC regulatory guidance as well. Thus, the WBN approach is conservative but not phenomenological. The wording provided in the Unit 2 TSB is the same as is in the Unit 1 TSB for this section. Thus for both technical accuracy and licensing basis consistency, the current wording is appropriate.

E2-3

Enclosure 3 WBN Unit 2 TS and TS Bases Developmental Revision G Mark-ups E3-1

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.3 ---------- -------- NOTES--------------

1. Adjust NIS channel if absolute difference is _>

3%.

2. Required to be performed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months after THERMAL POWER is _> 4.25% RTP.

Compare results of the PDMS measurements to 31 effective full NIS AFD. power days (EFPD)

SR 3.3.1.4 --------------------- NOTE ---------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT. 62 days on a STAGGERED TEST BASIS SR 3.3.1.5 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.1.6 --------------------- NOTE ---------------

Required to be performed within 6 days after THERMAL POWER is > 50% RTP.

Calibrate excore channels to agree with the PDMS 92 EFPD measurements.

(continued)

Watts Bar - Unit 2 3.3-11 (developmental) RG

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

5. Source Range 2 M 2 I, J SR 3.3.1.1 c :1.33 E5 c 1.0 E5 cpt Neutron Flux SR 3.3.1.8 (b)(c)ps SR 3.3.1.11 (bllPS 3 (a), 4 (a),5 (a) 2 J, K SR 3.3.1.1 <- 1.33 6E5 cp 1.0 E5 cpt SR 3.3.1.8 (bXs)

SR 3.3.1.11 (b)(c)

SR 3.3.1.15 3 (9),4 (9),5 () 1 L SR 3.3.1.1 N/A N/A SR 3.3.1.11 (b)11)

6. Overtemperature AT 1,2 4 W SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 1 Note 1 SR 3.3.1.6 (Page 3.3-21) (Page 3.3-21)

SR 3.3.1.7 (bXc)

SR 3.3.1.10 (b)(c)

SR 3.3.1.15

7. Overpower AT 1,2 4 W SR 3.3.1.1 Refer to Refer to SR 3.3.1.7 (bX,) Note 2 Note 2 SR 3.3.1.10 (b)oC) (Page 3.3-22) (Page 3.3-22)

SR 3.3.1.15 8, Pressurizer Pressure

a. Low 1 (h) 4 X SR 3.3.1.1 - 1964.8 psig 1970 psig SR 3.3.1.7 (bXo)

SR 3.3.1.10 (b)(c)

SR 3.3.1.15

b. High 1, 2 4 W SR 3.3.1.1 *2390.2 psig 2385 psig SR 3.3.1.7 (bXc)

SR 3.3.1.10 (b)(c)

SR 3.3.1.15 (continued)

(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b) If the as found channel setpoint is outside its predefined as found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The methodologies used to determine the as found and as left tolerances for the NTSP are specified in FSAR Section 7.1.2.

(f) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(g) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide indication.

(h) Above the P-7 (Low Power Reactor Trips Block) interlock.

Watts Bar - Unit 2 3.3-16 (developmental) G&I

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

9. Pressurizer Water 1 (h) 3 X SR 3.3.1.1 *92.7% span 92% span Level-High SR 3.3.1.7 (bXc)

SR 3.3.1.10 (b)c)

10. Reactor Coolant Flow - 1 (h) 3 per loop N SR 3.3.1.1 _89.7% flow 90% flow Low SR 3.3.1.7 (bXC)

SR 3.3.1.10 (h)b")

SR 3.3.1.15

11. Undervoltage RCPs 1 (h) 1 per bus M SR 3.3.1.9 >_4-7345112 V 48205400kV SR 3.3.1.10 (hoc)

SR 3.3.1.15

12. Underfrequency RCPs I (h) 1 per bus M SR 3.3.1.9 Ž56.9 Hz 57.5 Hz SR 3.3.1.10 (b)(,l SR 3.3.1.15 (continued)

(b) If the as found channel setpoint is outside its predefined as found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The methodologies used to determine the as found and as left tolerances for the NTSP are specified in FSAR Section 7.1.2.

(h) Above the P-7 (Low Power Reactor Trips Block) interlock.

Watts Bar - Unit 2 3.3-17 (developmental) F~GI

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

13. SG Water Level - 1,2 3/SG U SR 3.3.1.1 >16.4% of 17% of narrow Low-Low SR 3.3.1.7 (bxc) narrow range range span SR 3.3.1.10 (b)lc) span SR 3.3.1.15 Coincident with:

a) Vessel AT 1,2 3 V SR 3 .3.1.7 (b)(c) Vessel AT Vessel AT Equivalent to SR 3.3.1.10 (b)(c) variable input variable input power _<52.6% RTP 50% RTP

<50% RTP With a time _1.01 T5 T.

delay (T.) if one (Refer to (Refer to steam generator Note 3, Note 3, is affected Page 3.3-23) Page 3.3-23) or A time delay (T.) *1.01 Tm Tm iftwo or more (Refer to (Refer to steam Note 3, Note 3, generators are Page 3.3-23) Page 3.3-23) affected b) Vessel AT 1,2 3 V SR 3 .3.1.7 (bXc) Vessel AT Vessel AT Equivalent to SR 3.3.1.10 (b)(c) variable input variable input power *52.6% RTP 50% RTP

> 50% RTP with no time delay (T.

and Tm = 0)

14. Turbine Trip

a. Low Fluid Oil 1 0) 3 0 SR 3.3.1.4410 t 4.&38.3 psi 45 psig Pressure (bXc) g SR 3.3.1.14

b. Turbine Stop 1 ) 4 Y SR 3.3.1.10 > 1% open 1% open Valve Closure SR 3.3.1.14 (continued)

(b) If the as found channel setpoint is outside its predefined as found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The methodologies used to determine the as found and as left tolerances for the NTSP are specified in FSAR Section 7.1.2.

(i) Above the P-9 (Power Range Neutron Flux) interlock.

Watts Bar - Unit 2 3.3-18 (developmental) PG I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 8)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

6. Auxiliary Feedwater (continued)

d. Loss of Offsite 1, 2, 3 4 per bus F Refer to Function 4 of Table 3.3.5-1 for SRs and Power Allowable Values. Notes (b) and (c) are applicable to SR 3.3.5.2 for this function.

e. Trip of all 1 0),2 (k) 1 per pump J SR 3.3.2.8 (b)(c) > 4943.3 50 psig I Turbine Driven SR 3.3.2.9 1b)1c) -

Main Feedwater SR 3.3.2.10 Pumps

f. Auxiliary 1, 2, 3 3 F SR 3.3.2.6 A) > 0.5 psig A) 1.2 psig Feedwater SR 3.3.2.9 (b)

Pumps SR 3.3.2.10 B) _ 1.33 psig B) 2.0 psig Train A and B Suction Transfer on Suction Pressure - Low

7. Automatic Switchover to Containment Sump

a. Automatic 1,2, 3, 4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays

b. Refueling Water 1,2,3,4 4 K SR 3.3.2.1 > 155.6 inches 158 inches Storage Tank SR 3.3.2.4 (b)oc) from from (RWST) SR 3.3.2.9 (b)(C) Tank Base Tank Base Level - Low SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection and Coincident with 1,2,3,4 4 K SR 3.3.2.1 > 37.2 inches 38.2 inches Containment SR 3.3.2.4 (b)(c) above above Sump SR 3.3.2.9 (b( (c) el. 702.8 ft el. 702.8 ft Level - High SR 3.3.2.10 (continued)

(b) If the as found channel setpoint is outside its redefined as found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(c) The instrument channel setpoint shall be reset to a value that is within the as left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The methodologies used to determine the as found and as left tolerances for the NTSP are specified in FSAR Section 7.1.2.

(I) Entry into Condition J may be suspended for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when placing the second Turbine Driven Main Feedwater (TDMFW) Pump in service or removing one of two TDMFW pumps from service.

(k) When one or more Turbine Driven Feedwater Pump(s) are supplying feedwater to steam generators.

Watts Bar - Unit 2 3.3-38 (developmental)

COMS 3.4.12 SURVEILLANCE REQUIREMENTS Y

SURVEILLANCE FREQUENCY 4

SR 3.4.12.1 Verify no safety injection pumps are capable of Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after injecting into the RCS. entering MODE 4 from MODE 3 prior to the temperature of one or more RCS cold legs decreasing below &-42250F.

AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter SR 3.4.12.2 Verify a maximum of one charging pump is capable Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after of injecting into the RCS. entering MODE 4 from MODE 3 prior to the temperature of one or more RCS cold legs decreasing below a2,225°F.

AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter SR 3.4.12.3 Verify each accumulator is isolated. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.12.4 -------------------- NOTE---------------

Only required to be performed when complying with LCO 3.4.12.b.

Verify RCS vent open. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for unlocked open vent paths AND 31 days for locked open vent paths (continued)

Watts Bar - Unit 2 3.4-26 (developmental) AG I

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained AND All SG tubes satisfying the tube plugging or repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A. 1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is plugging or repair criteria maintained until the next and not plugged or refueling outage or SG repaired in accordance with tube inspection the Steam Generator Program AND A.2 Plug or repair the Prior to entering affected tube(s) in MODE 4 following accordance with the the next refueling Steam Generator outage or SG tube Program inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not met AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained Watts Bar - Unit 2 3.4-38 (developmental) FG I

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify steam generator tube integrity in accordance In accordance with with the Steam Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube plugging or repair criteria is plugged or MODE 4 following a repaired in accordance with the Steam Generator SG tube inspection Program.

Watts Bar - Unit 2 3.4-39 (developmental) AGI

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. keff < 0.95 if fully flooded with unborated water, which, includes an allowance for uncertainties as described in Sections 4.3.2.7 and 9.1 of the FSAR;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules.

d. Fu., _romb'o.--.th ...-, nrF8hM@R4o . . thO.n... Or.. 3-. 0

e. Fuel assemblies with initial enrichments greater than 3.80 weight percent and less than a maximum of 5 percent enrichment (nominally 4.95 +/- 0.05 percent) may be stored in the spent fuel racks in one of four arrangements with specific limits as identified below:

1. New and sSpent fuel assemblies may be stored in the racks in an all cell arrangement "::ite't -. t.rhr Irctriotiont provided the burnup of each assembly is in the acceptable domain identified in Figure 4.3-3, depending upon the specified initial enrichment.

2. New and spent fuel assemblies may be stored in a checkerboard arrangement of 2 new and 2 spent assemblies, provided that each spent fuel assembly has accumulated a minimum burnup in the acceptable domain identified in Figure 4.3-4.

3. New fuel assemblies may be stored in 4-cell arrays with 1 of the 4 cells remaining empty of fuel (i.e. containing only water or water with up to 75 percent by volume of non-fuel bearing material.

(continued)

Watts Bar - Unit 2 4.0-2 (developmental) PG

Design Features 4.0 ACCEPTABLE DOMAIN 16 FIBURNUP 14 12 10, UNACCEPTABLE

6. BURNUP DOMAIN 6

4-0 -. ~

j5 Ia II 4 Y-I-~

I* I

~

I*I*

4 20 0 3.10.0....3. .4 0.b 50....

Initial Enrichment wt% U-235 FIGURE 4.3-4 MINIMUM REQUIRED BURNUP FOR A 3X2-CHECKERBOARD ARRANGEMENT OF 2 SPENT AND 'AJS&M*- -.- WITH 2 NEW FUEL ASSEMBLIES OF 5% ENRICHMENT (MAXIMUM)

Watts Bar - Unit 2 4.0-9 (developmental) AG I

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following s..eI:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown)T eA4EI-all anticipated transients included in the design specification,) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

(continued)

Watts Bar-Unit 2 5.0-14 (developmental) AG

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Program (continued)

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than an SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG, except for specific types of degradation at specific locations as described in paragraph c. of the Steam Generation Program.9ef-desi bas

.. is t..

1..,-. - faultod

.. cteem onra. or*

minutc8 (@pm4) for the faultad ctoSAm 160 4n

-Aoro gallonc, por day; (gpd) for tho non faltod- stsom gonreatorc.

For design bacis seoieanto that do not -havo a feultod cateem gonr@ator, sacidonti induood loakago is-not to oxod- 16 0

-4d -ar --- FAm laonraFt@r

3. The operational leakage performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube plugging or repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging or repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A - accoccmont of degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

(continued)

Watts Bar-Unit 2 5.0-15 (developmental) i&G

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

5.7.2.12 Steam Generator (SG) Program (continued)

2. lncpot 1004 Oof tho t*08 at coguontial P600616 Of 60 99ofctVO full powor Aonthc. The fictaguontial poriod ha4l the &-G--. "o I~ Ih for m6Or t han -21ofo4 ofl oprIo without being iRncpct-od.After the first refueling outage following SG installation, inspect each SG at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections).

In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever he-leeresults in more frequent (continued)

Watts Bar-Unit 2 5.0-16 (developmental) RG

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued) inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary LEAKAGE.

5.7.2.13 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data; Secondary Water Chemistry Program (continued)

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

(continued)

Watts Bar-Unit 2 5.0-17 (developmental) SGI

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.7 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

5.9.8 PAMS Report When a Report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.9.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.7.2.12, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each I active degradation mechanism,

f. TtO numbor and por...t@g. Of tu... puggod t@ d toThe number and percentage of tubes plugged or repaired to date, and the effective plugging percentage in each SG,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. Th@ offot plugging p.....t... =f. ll plugging in oash &GRepair method utilized and the number of tubes repaired by each repair method.

Watts Bar-Unit 2 5.0-33 (developmental) &GI

RTS Instrumentation B 3.3.1 BASES APPLICABLE 10. Reactor Coolant Flow - Low (continued)

SAFETY ANALYSES, In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop LCO, and could result in DNB conditions in the core because of the higher APPLICABILITY power level. In MODE 1 below the P-8 setpoint and above the P-7 (continued) setpoint, a loss of flow in two or more loops is required to actuate a reactor trip because of the lower power level and the greater margin to the design limit DNBR. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since there is insufficient heat production to generate DNB conditions.

The Reactor Coolant Flow-Low Trip Setpoint and Allowable Value are specified in % indicated loop flow; however, the Eagle-21TM values entered through the MMI are specified in an equivalent %

differential pressure.

11. Undervoltage Reactor Coolant Pumps The Undervoltage RCPs trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two or more RCS loops. The voltage to each RCP is monitored.

Above the P-7 setpoint, a loss of voltage detected on two or more RCP buses will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant Flow - Low Trip Setpoint is reached in two or more RCS loops. The loss of voltage in two loops must be sustained for a length of time equal to or greater than that set in the time delay. Time delays are incorporated into the Undervoltage RCPs channels to prevent reactor trips due to momentary electrical power transients.

The "Allowable Value" of greater than or equal to 5112 V includes the 3.5 cycle relay response time. This is required to ensure a reactor trip occurs before reaching a safety limit.

The undervoltage relay "Nominal Trip Setpoint" value of 5400 V includes the 3.5 cycle relay response time. This is required to meet the reactor trip nominal setpoint of 4830 V (Reference 17).

The LCO requires one Undervoltage RCP channel per bus to be OPERABLE.

(continued)

Watts Bar - Unit 2 B 3.3-23 (developmental) FGI

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS (continued) The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Together these factors demonstrate the change in the absolute difference between NIS and heat balance calculated powers rarely exceeds 2% in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

In addition, control room operators periodically monitor redundant indications and alarms to detect deviations in channel outputs.

S R 3.3.1.3 SR 3.3.1.3 compares the power distribution measurement to the NIS channel AFD output every 31 EFPD. If the absolute difference is > 3%,

the NIS channel is still OPERABLE, but must be readjusted. Ifthe NIS channel cannot be properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f(AI) input to the Overtemperature AT Function. The incore power distribution measurement is obtained using the OPERABLE Power Distribution Monitoring System (PDMS) (Ref. 16).

Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is > 3%. Note 2 clarifies that the Surveillance is required only if reactor power is >. 4.25% RTP and that 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months is allowed for performing the first Surveillance after reaching .4.625% RTP. This surveillance is typically performed at greater than or equal to 50% RTP to ensure the results of the evaluation are more accurate and the adjustments more reliable. Ninety-six (96) hours are allowed to ensure Xenon stability and allow for instrumentation alignments.

The Frequency of every 31 EFPD is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Also, the slow changes in neutron flux during the fuel cycle can be detected during this interval.

(continued)

Watts Bar - Unit 2 B 3.3-54 (developmental) 8G

RTS Instrumentation B 3.3.1 BASES REFERENCES 10. Deleted (continued)

11. WCAP-1 3632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996

12. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

13. Deleted

14. WCAP-14333 P-A, Revision 1, "Probablistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"

October 1998.

15. WCAP-1 5376-P-A, Revision 1, "Risk Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003

16. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994 (Addendum 2, April 2002).

17. TVA Calculation WBPE0689009007, "Demonstrated Accuracy Calculation for Reactor Coolant Pump Undervoltage" Watts Bar - Unit 2 B 3.3-65 (developmental) F

Containment Vent Isolation Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Containment Vent Isolation Instrumentation BASES BACKGROUND Containment Vent Isolation Instrumentation closes the containment isolation valves in the Containment Purge System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. The Reactor Building Purge System may be in use during reactor operation and with the reactor shutdown.

Containment vent isolation is initiated by a safety injection (SI) signal or by manual actuation. The Bases for LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," discuss initiation of SI signals.

Redundant and independent gaseous radioactivity monitors measure the radioactivity levels of the containment purge exhaust, each of which will initiate its associated train of automatic Containment Vent Isolation upon detection of high gaseous radioactivity.

The Reactor Building Purge System has inner and outer containment isolation valves in its supply and exhaust ducts. This system is described in the Bases for LCO 3.6.3, "Containment Isolation Valves."

The plant design basis requires that when moving irradiated fuel in the Auxiliary Building and/or Containment with the Containment open to the Auxiliary Building ABSCE spaces, a signal from the spent fuel pool radiation monitors 0-RE-90-102 and -103 will initiate a Containment Ventilation Isolation (CVI) in addition to their normal function. In addition, a signal from the containment purge radiation monitors 2-RE-90-130, and

-131 or other CVI signal will initiate that portion of the Auxiliary Building Isolation (ABI) normally initiated by the spent fuel pool radiation monitors.

Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE. Therefore, the containment ventilation instrumentation must remain (continued)

Watts Bar - Unit 2 B 3.3-154 (developmental) 8G I

Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

BACKGROUND operable when moving irradiated fuel in the Auxiliary Building ifthe (continued) containment air locks, penetrations, equipment hatch, etc. are open to the Auxiliary Building ABSCE spaces.

APPLICABLE The containment isolation valves for the Reactor Building Purge System SAFETY close within six seconds following the DBA. The containment vent ANALYSES isolation radiation monitors act as backup to the SI signal to ensure closing of the purge air system supply and exhaust valves. They are also the primary means for automatically isolating containment in the event of a fuel handling accident during shutdown. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100 (Ref. 1) limits.

The Containment Vent Isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement.

When moving irradiated fuel inside containment or in the Auxiliary Building with containment air locks or penetrations open to the Auxiliary Building ABSCE spaces, or when moving fuel in the Auxiliary Building with the containment equipment hatch open, the provisions to initiate a CVI from the spent fuel pool radiation monitors and to initiate an ABI (i.e.,

the portion of an ABI normally initiated by the spent fuel pool radiation monitors) from a CVI, including a CVI generated by the containment purge monitors, in the event of a fuel handling accident (FHA) must be in place and functioning. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. The containment equipment hatch cannot be open when moving irradiated fuel inside containment in accordance with Technical Specification 3.9.4.

The ABGTS is required to be operable during movement of irradiated fuel in the Auxiliary Building during any mode and during movement of irradiated fuel in the Reactor Building when the Reactor Building is established as part of the ABSCE boundary (see TS 3.3.8, 3.7.12, &

3.9.4). When moving irradiated fuel inside containment, at least one train of the containment purge system must be operating or the containment must be isolated. When moving irradiated fuel in the Auxiliary Building during times when the containment is open to the Auxiliary Building ABSCE spaces, containment purge can be operated, but operation of the system is not required. However, whether the containment purge system is operated or not in this configuration, all containment ventilation isolation valves and associated instrumentation must remain operable.

(continued)

Watts Bar - Unit 2 B 3.3-155 (developmental) AGI

Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

APPLICABLE This requirement is necessary to ensure a CVI can be accomplished from SAFETY the spent fuel pool radiation monitors in the event of an FHA in the ANALYSES Auxiliary Building. Additionally, a Containment Isolation Phase A (SI (continued) signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit.

LCO The LCO requirements ensure that the instrumentation necessary to initiate Containment Vent Isolation, listed in Table 3.3.6-1, is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate Containment Vent Isolation at any time by using either of two switches in the control room or from local panel(s). Either switch actuates both trains. This action will cause actuation of all components in the same manner as any of the automatic actuation signals. These manual switches also initiate a Phase A isolation signal.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one selector switch and the interconnecting wiring to the actuation logic cabinet.

2. Automatic Actuation Logic and Actuation Relays The LCO requires two trains of Automatic Actuation Logic and Actuation Relays OPERABLE to ensure that no single random failure can prevent automatic actuation.

Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b, SI. The applicable MODES and specified conditions for the containment vent isolation portion of the SI Function is different and less restrictive than those for the SI role. If one or more of the SI Functions becomes inoperable in such a manner that only the Containment Vent Isolation Function is affected, the Conditions applicable to the SI Functions need not be entered. The less restrictive Actions specified for inoperability of the Containment Vent Isolation Functions specify sufficient compensatory measures for this case.

(continued)

Watts Bar - Unit 2 B 3.3-156 (developmental) AG

ABGTS Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Auxiliary Building Gas Treatment (ABGTS) Actuation Instrumentation BASES BACKGROUND The ABGTS ensures that radioactive materials in the fuel building atmosphere following a fuel handling accident or a loss of coolant accident (LOCA) are filtered and adsorbed prior to exhausting to the environment. The system is described in the Bases for LCO 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)." The system initiates filtered exhaust of air from the fuel handling area, ECCS pump rooms, and penetration rooms automatically following receipt of a fuel pool area high radiation signal or a Containment Phase A Isolation signal.

Initiation may also be performed manually as needed from the main control room.

High area radiation, monitored by either of two monitors, provides ABGTS initiation. Each ABGTS train is initiated by high radiation detected by a channel dedicated to that train. There are a total of two channels, one for each train. High radiation detected by any monitor or a Phase A isolation signal from the Engineered Safety Features Actuation System (ESFAS) initiates auxiliary building isolation and starts the ABGTS. These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the Auxiliary Building Secondary Containment Enclosure (ABSCE).

The plant design basis requires that when moving irradiated fuel in the Auxiliary Building and/or Containment with the Containment and/or annulus open to the Auxiliary Building ABSCE spaces, a signal from the spent fuel pool radiation monitors 0-RE-90-102 and -103 will initiate a Containment Ventilation Isolation (CVI) in addition to their normal function. In addition, a signal from the containment purge radiation monitors 2-RE-90-130, and -131 or other CVI signal will initiate that portion of the Auxiliary Building Isolation (ABI) normally initiated by the spent fuel pool radiation monitors. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE. Therefore, the containment ventilation instrumentation must remain operable when (continued)

Watts Bar - Unit 2 B 3.3-170 (developmental) SGi I

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

BACKGROUND moving irradiated fuel in the Auxiliary Building ifthe containment and/or (continued) annulus air locks, penetrations, equipment hatch, etc. are open to the Auxiliary Building ABSCE spaces.

APPLICABLE The ABGTS ensures that radioactive materials in the ABSCE atmosphere SAFETY following a fuel handling accident or a LOCA are filtered and adsorbed ANALYSES prior to being exhausted to the environment. This action reduces the radioactive content in the auxiliary building exhaust following a LOCA or fuel handling accident so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. 1).

The ABGTS Actuation Instrumentation satisfies Criterion 3 of the NRC Policy Statement.

When moving irradiated fuel inside containment or in the Auxiliary Building with containment air locks or penetrations open to the Auxiliary Building ABSCE spaces, or when moving fuel in the Auxiliary Building with the containment equipment hatch open, the provisions to initiate a CVI from the spent fuel pool radiation monitors and to initiate an ABI (i.e.,

the portion of an ABI normally initiated by the spent fuel pool radiation monitors) from a CVI, including a CVI generated by the containment purge monitors, in the event of a fuel handling accident (FHA) must be in place and functioning. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. The containment equipment hatch cannot be open when moving irradiated fuel inside containment in accordance with Technical Specification 3.9.4.

The ABGTS is required to be operable during movement of irradiated fuel in the Auxiliary Building during any mode and during movement of irradiated fuel in the Reactor Building when the Reactor Building is established as part of the ABSCE boundary (see TS 3.3.8, 3.7.12, &

3.9.4). When moving irradiated fuel inside containment, at least one train of the containment purge system must be operating or the containment must be isolated. When moving irradiated fuel in the Auxiliary Building during times when the containment is open to the Auxiliary Building ABSCE spaces, containment purge can be operated, but operation of the system is not required. However, whether the containment purge system is operated or not in this configuration, all containment ventilation isolation valves and associated instrumentation must remain operable.

(continued)

Watts Bar - Unit 2 B 3.3-171 (developmental) AG

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

APPLICABLE This requirement is necessary to ensure a CVI can be accomplished from SAFETY the spent fuel pool radiation monitors in the event of a FHA in the ANALYSES Auxiliary Building. Additionally, a Containment Isolation Phase A (SI (continued) signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE.

The LCO requirements ensure that instrumentation necessary to initiate LCO The LCO requirements ensure that instrumentation necessary to initiate the ABGTS is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate the ABGTS at any time by using either of two switches in the control room. This action will cause actuation of all components in the same manner as any of the automatic actuation signals.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one hand switch and the interconnecting wiring to the actuation logic relays.

2. Fuel Pool Area Radiation The LCO specifies two required Fuel Pool Area Radiation Monitors to ensure that the radiation monitoring instrumentation necessary to initiate the ABGTS remains OPERABLE. One radiation monitor is dedicated to each train of ABGTS.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of channel electronics. OPERABILITY may also require correct valve lineups, sample pump operation, and filter motor operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.

(continued)

Watts Bar - Unit 2 B 3.3-172 (developmental) A

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE is ANALYSES related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for a main steam line break (MSLB) assumes that the Wie-accident primary-to-secondary I LEAKAGE from three steam generators is 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) from one steam generator. ThiS o8k8@8o...... tI..

on romping th SOFA Ifr tho... ido'nt.

-For an SGTR accident, the accident analysis assumes a primary-to-secondary leakage of 150 gpd per steam generator prior to the accident.

Subsequent to the SGTR a leakage of 150 gpd is assumed in each of three intact steam generators and RCS blowdown flow through the ruptured tube in the faulted steam generator. Consequently, the LCO requirement to limit primary-to-secondary LEAKAGE through any one steam generator to less than or equal to 150 gpd is acceptable.

The safety analysis for the SLB accident assumes the entire 1 gpm primary-to-secondary LEAKAGE is through the affected steam generator as an initial condition. The dose consequences resulting from the SLB accident are within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of an off-normal condition. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

(continued)

Watts Bar - Unit 2 B 3.4-73 (developmental) AG I

SG TUBE INTEGRITY B 3.4.17 BASES BACKGROUND The processes used to meet the SG performance criteria are defined by (continued) the Steam Generator Program Guidelines (Ref. 1).

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity,"

limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging or repair criteria be plugged or repaired in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging or repair criteria is repaired or removed from service by plugging. If a tube was determined to satisfy the plugging or repair criteria but was not plugged or repaired, the tube may still have tube integrity.

(continued)

Watts Bar - Unit 2 B 3.4-97 (developmental) AGI

SG TUBE INTEGRITY B 3.4.17 BASES LCO In the context of this Specification, an SG tube is defined as the entire (continued) length of the tube, including the tube wall and any repairs made to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

An SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.7.2.12, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

(continued)

Watts Bar - Unit 2 B 3.4-98 (developmental) AGI

SG TUBE INTEGRITY B 3.4.17 BASES ACTIONS A.1 and A.2 (continued)

Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging or repair criteria but were not plugged or repaired in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging or repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if an SG tube that should have been plugged or repaired, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection.

If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged or repaired prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

(continued)

Watts Bar - Unit 2 B 3.4-100 (developmental) AGI

SG TUBE INTEGRITY B 3.4.17 BASES (continued)

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging or repair criteria. Inspection scope (i.e.,

which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.7.2.12 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.7.2.12 until subsequent inspections support extending the inspection interval.

(continued)

Watts Bar - Unit 2 B 3.4-101 (developmental) AG

SG TUBE INTEGRITY B 3.4.17 BASES SURVEILLANCE SR 3.4.17.2 REQUIREMENTS (continued) During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging or repair criteria is removed from service by plugging. The tube plugging or repair criteria delineated in Specification 5.7.2.12 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging or repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

Steam Generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.

The Frequency of prior to entering MODE 4 following an SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19, Control Room.

3. 10 CFR 100, Reactor Site Criteria.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

Watts Bar - Unit 2 B 3.4-102 (developmental) AG I

CCS B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Component Cooling System (CCS)

BASES BACKGROUND The CCS provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, the CCS also provides this function for various non-essential components, as well as the spent fuel storage pool. The CCS serves as a barrier to the release of radioactive byproducts between potentially radioactive systems and the Essential Raw Cooling Water (ERCW) System, and thus to the environment.

The CCS is arranged as two independent, full-capacity cooling trains, Train A and Train B. Train A in Unit 2 is served by CCS Hx AB and CCS pump 4A.A2A-A. Pump 4.&-42B-B, which is actually Train B equipment, is also normally aligned to the Train A header in Unit 2.

However, pump 4.842B-B can be realigned to Train B on loss of Train A.

Train B is served by CCS Hx C. Normally, only CCS pump C-S is aligned to the Train B header since few non-essential, normally-operating loads are assigned to Train B. However, pump -. 4B-2B-B can be realigned to the Train B header on a loss of the C-S pump.

Each safety related train is powered from a separate bus. An open surge tank in the system provides pump trip protective functions to ensure that sufficient net positive suction head is available. The pump in each train is automatically started on receipt of a safety injection signal, and all non-essential components will be manually isolated.

Additional information on the design and operation of the system, along with a list of the components served, is presented in the FSAR, Section 9.2.2 (Ref. 1). The principal safety related function of the CCS is the removal of decay heat from the reactor via the Residual Heat Removal (RHR) System. This may be during a normal or post accident cooldown and shutdown.

(continued)

Watts Bar - Unit 2 B 3.7-37 (developmental) AG I

ABGTS B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

BASES BACKGROUND The ABGTS filters airborne radioactive particulates from the area of the fuel pool following a fuel handling accident and from the area of active Unit 2 ECCS components and Unit 2 penetration rooms following a loss of coolant accident (LOCA).

The ABGTS consists of two independent and redundant trains. Each train consists of a heater, a prefilter, moisture separator, a high efficiency particulate air (HEPA) filter, two activated charcoal adsorber sections for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system.

A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails.

The downstream HEPA filter is not credited in the analysis. The system initiates filtered ventilation of the Auxiliary Building Secondary Containment Enclosure (ABSCE) exhaust air following receipt of a Phase A containment isolation signal or a high radiation signal from the spent fuel pool area.

The ABGTS is a standby system, not used during normal plant operations. During emergency operations, the ABSCE dampers are realigned and ABGTS fans are started to begin filtration. Air is exhausted from the Unit 2 ECCS pump rooms, Unit 2 penetration rooms, and fuel handling area through the filter trains. The prefilters or moisture separators remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The plant design basis requires that when moving irradiated fuel in the Auxiliary Building and/or Containment with the Containment open to the Auxiliary Building ABSCE spaces, a signal from the spent fuel pool radiation monitors O-RE-90-102 and -103 will initiate a Containment Ventilation Isolation (CVI) in addition to their normal function. In addition, a signal from the containment purge radiation monitors 1-RE-90-130 and

-131 or other CVI signal will initiate that portion of the ABI normally initiated by the spent fuel pool radiation monitors. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI (continued)

Watts Bar - Unit 2 B 3.7-63 (developmental) AG

ABGTS B 3.7.12 BASES BACKGROUND will cause a CVI signal in the refueling unit. In the case where the (continued) containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE. Therefore, the containment ventilation instrumentation must remain operable when moving irradiated fuel in the Auxiliary Building if the containment air locks, penetrations, equipment hatch, etc. are open to the Auxiliary Building ABSCE spaces. In addition, the ABGTS must remain operable if these containment penetrations are open to the Auxiliary Building during movement of irradiated fuel inside containment.

The ABGTS is discussed in the FSAR, Sections 6.5.1, 9.4.2, 15.0, and 6.2.3 (Refs. 1, 2, 3, and 4, respectively).

APPLICABLE The ABGTS design basis is established by the consequences of the SAFETY limiting Design Basis Accident (DBA), which is a fuel handling accident.

ANALYSES The analysis of the fuel handling accident, given in Reference 3, assumes that all fuel rods in an assembly are damaged. The analysis of the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the ABGTS. The DBA analysis of the fuel handling accident assumes that only one train of the ABGTS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the one remaining train of this filtration system. The amount of fission products available for release from the ABSCE is determined for a fuel handling accident and for a LOCA. The assumptions and the analysis for a fuel handling accident follow the guidance provided in Regulatory Guide 1.25 (Ref. 5) and NUREG/CR-5009 (Ref. 10). The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 6).

The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

When moving irradiated fuel inside containment or in the Auxiliary Building with containment air locks or penetrations open to the Auxiliary Building ABSCE spaces, or when moving fuel in the Auxiliary Building with the containment equipment hatch open, the provisions to initiate a CVI from the spent fuel pool radiation monitors and to initiate an ABI (i.e., the portion of an ABI normally initiated by the spent fuel pool radiation monitors) from a CVI, including a CVI initiated by the containment purge monitors, in the event of a fuel handling accident (FHA) must be in place and functioning. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high (continued)

Watts Bar - Unit 2 B 3.7-64 (developmental) 8GI

ABGTS B 3.7.12 BASES APPLICABLE temperature in the Auxiliary Building air intakes, or manual ABI will SAFETY cause a CVI signal in the refueling unit. The containment equipment ANALYSES hatch cannot be open when moving irradiated fuel inside containment in (continued) accordance with Technical Specification 3.9.4.

The ABGTS is required to be operable during movement of irradiated fuel in the Auxiliary Building during any mode and during movement of irradiated fuel in the Reactor Building when the Reactor Building is established as part of the ABSCE boundary (see TS 3.3.8, 3.7.12, &

3.9.4). When moving irradiated fuel inside containment, at least one train of the containment purge system must be operating or the containment must be isolated. When moving irradiated fuel in the Auxiliary Building during times when the containment is open to the Auxiliary Building ABSCE spaces, containment purge can be operated, but operation of the system is not required. However, whether the containment purge system is operated or not in this configuration, all containment ventilation isolation valves and associated instrumentation must remain operable. This requirement is necessary to ensure a CVI can be accomplished from the spent fuel pool radiation monitors in the event of a FHA in the Auxiliary Building. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE.

LCO Two independent and redundant trains of the ABGTS are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure could result in the atmospheric release from the ABSCE exceeding the 10 CFR 100 (Ref. 7) limits in the event of a fuel handling accident or LOCA.

The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE in both trains. An ABGTS train is considered OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function; and (continued)

Watts Bar - Unit 2 B 3.7-65 (developmental) AG I

DC Sources - Operating B 3.8.4 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.4 DC Sources - Operating BASES BACKGROUND The station DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment and preferred AC vital bus power (via inverters). As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system also conforms to the recommendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3).

125 V Vital DC Electrical Power Subsystem The vital 125 VDC electrical power system is a Class 1E system whose safety function is to provide control power for engineered safety features equipment, emergency lighting, vital inverters, and other safety related DC powered equipment for the entire unit. The system capacity is sufficient to supply these loads and any connected non-safety loads during normal operation and to permit safe shutdown and isolation of the reactor for the "loss of all AC power" condition. The system is designed to perform its safety function subject to a single failure.

The 125V DC vital power system is composed of the four redundant channels (Channels I and III are associated with Train A and Channels II and IV are associated with Train B) and consists of four lead-acid-calcium batteries, ei~eight battery chargers (including two pairs of spare chargers), four distribution boards, battery racks, and the required cabling, instrumentation and protective features. Each channel is electrically and physically independent from the equipment of all other channels so that a single failure in one channel will not cause a failure in another channel. Each channel consists of a battery charger which supplies normal DC power, a battery for emergency DC power, and a battery board which facilitates load grouping and provides circuit protection. These four channels are used to provide emergency power to the 120V AC vital power system which furnishes control power to the reactor protection system. No automatic connections are used between the four redundant channels.

(continued)

Watts Bar - Unit 2 B 3.8-50 (developmental) AGI

Reactor Building Purge Air Cleanup Units B 3.9.8 BASES BACKGROUND The Primary containment exhaust is monitored by a radiation detector (continued) which provides automatic containment purge ventilation system isolation upon detecting the setpoint radioactivity in the exhaust air stream. The containment purge ventilation isolation valves will be automatically closed upon the actuation of a Containment Vent Isolation (CVI) signal whenever the primary containment is being purged during normal operation or upon manual actuation from the Main Control Room (Ref. 2). Requirements for Containment Vent Isolation Instrumentation are covered by LCO 3.3.6.

APPLICABLE The Reactor Building Purge Ventilation System air cleanup units ensure SAFETY that the release of radioactivity to the environment is limited by cleaning ANALYSES up containment exhaust during a fuel handling accident before the containment purge exhaust valves are isolated. Reactor Building Purge Ventilation System filter efficiency is one of the inputs for the analysis of the environmental consequences of a fuel handling accident.

Containment isolation can only result in smaller releases of radioactivity to the environment (Ref. 1). The Containment Vent Isolation System ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment (Ref. 2). Containment Vent Isolation Instrumentation is addressed by LCO 3.3.6.

The Reactor Building Purge Air Cleanup Units satisfy Criterion 3 of the NRC Policy Statement.

In addition, during movement of irradiated fuel in the Auxiliary Building when containment is open to the Auxiliary Building spaces, a high radiation signal from the spent fuel pool accident radiation monitors, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will initiate a CVI. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE.

LCO The safety function of the Reactor Building Purge Air Cleanup Unit is related to the initial control of offsite radiation exposures resulting from a fuel handling accident inside containment. During a fuel handling accident inside containment, the Reactor Building Purge Air Cleanup Unit provides a filtered path for cleaning up any air leaving the containment until the containment ventilation is isolated.

(continued)

Watts Bar - Unit 2 B 3.9-28 (developmental) AGI

Reactor Building Purge Air Cleanup Units B 3.9.8 BASES (continued)

LCO The plant design basis requires that when moving irradiated fuel in the (continued) Auxiliary Building and/or Containment with the Containment open to the Auxiliary Building ABSCE spaces, a signal from the spent fuel radiation monitors 0-RE-90-102 and -103 will initiate a CVI in addition to their normal function. In addition, a signal from the containment purge radiation monitors 2-RE-90-130, and -131 or other CVI signal will initiate that portion of the ABI normally initiated by the spent fuel pool radiation monitors. Additionally, a Containment Isolation Phase A (SI signal) from the operating unit, high temperature in the Auxiliary Building air intakes, or manual ABI will cause a CVI signal in the refueling unit. Therefore, the containment ventilation instrumentation must remain operable when moving irradiated fuel in the Auxiliary Building ifthe containment air locks, penetrations, equipment hatch, etc. are open to the Auxiliary Building ABSCE spaces. In addition, the ABGTS must remain operable ifthese containment penetrations are open to the Auxiliary Building during movement of irradiated fuel in side containment. In the case where the containment of both units is open to the Auxiliary Building spaces, a CVI in one unit will initiate a CVI in the other unit in order to maintain those spaces open to the ABSCE.

APPLICABILITY An initial assumption in the analysis of a fuel handling accident inside containment is that the accident occurs while irradiated fuel is being handled. Therefore, LCO 3.9.8 is applicable only at this time. See additional discussion in the Applicable Safety Analysis and LCO sections.

ACTIONS A.1 and A.2 If one Reactor Building Purge Air Cleanup Unit is inoperable, that air cleanup unit must be isolated. This places the system in the required accident configuration, thus allowing refueling to continue after verifying the remaining air cleanup unit is aligned and OPERABLE.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

(continued)

Watts Bar - Unit 2 B 3.9-29 (developmental) B

ML12065A036

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