ML13353A478

Fuel Handling Accident Dose Analysis Final Safety Analysis Report and Technical Specification Revision

ML13353A478
Person / Time
Site:	Watts Bar
Issue date:	12/12/2013
From:	Arent G P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13353A478 (152)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 December 12, 2013 10 CFR 50.34(b)10 CFR 50.67 10 CFR 100 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant Unit 2 -Fuel Handling Accident Dose Analysis Final Safety Analysis Report and Technical Specification Revision

References:

1. NRC letter to TVA dated June 19, 2013, "Watts Bar Nuclear Plant, Unit 1 -Issuance of Amendment to Allow Selective Implementation of Alternate Source Term to Analyze the Dose Consequences Associated with Fuel-Handling Accidents (TAC NO. ME8877)" (ADAMS Accession No. ML13141A564)

2. TVA letter to NRC dated August 5, 2011, "Watts Bar Nuclear Plant (WBN) Unit 2 -Final Safety Analysis Report (FSAR) -Chapter 15.5 Design Basis Dose Accident Analysis" (ADAMS Accession No. ML11222A022)

3. TVA letter to NRC dated September 23, 2011, "Watts Bar Nuclear Plant (WBN) Unit 2 -Final Safety Analysis Report (FSAR) -Chapter 15.5 Fuel Handling Accident (FHA) Dose Analysis" (ADAMS Accession No. ML1 1269A064)This letter provides revised Final Safety Analysis Report (FSAR) discussions and Technical Specification (TS) and Technical Specification Bases (TSB) changes associated with the Design Basis Accident (DBA) discussion for the Fuel Handling Accident (FHA) at Watts Bar Nuclear Plant (WBN) Unit 2. The changes to the WBN Unit 2 documents provide consistency with the recently approved amendment issued for WBN Unit 1 (Reference 1).

U.S. Nuclear Regulatory Commission Page 2 December 12, 2013 WBN Unit 1 submitted a license amendment request to implement the Alternate Source Term (AST) methodology for the FHA. The amendment included TS and TSB changes to remove the requirements for certain safety-related filtration systems to be operable during refueling because no credit was taken for radionuclide removal by those systems in the FHA. By Reference 3, WBN Unit 2 submitted a FHA based on the AST methodology for a dropped fuel assembly in the Auxiliary Building and in the containment when the containment was not isolated.

The Nuclear Regulatory Commission (NRC) determined that the WBN Unit 2 FHA analysis was acceptable in NUREG-0847 Supplemental Safety Evaluation Report (SSER) 25, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2." The WBN Unit 1 FHA Amendment did not include a specific evaluation for the case when the primary containment is closed and the purge system is in operation because the results from the containment closed case are clearly bounded by the containment open case. The WBN 2 FSAR currently includes a dose analysis for the FHA with the containment closed based on Regulatory Guide 1.25 guidance.

This letter provides a revised WBN Unit 2 FHA FSAR Section 15.5.6 discussion consistent with the approved Unit 1 License Amendment Request (LAR). This change removes the discussion of the Regulatory Guide 1.25 analysis for the closed containment case. Changes to the Unit 2 TS and TSB to be consistent with the approved Unit 1 TS and TSB are provided.Changes to WBN Unit 2 FSAR Chapters 6 and 9 that remove the mitigation of an FHA as a design basis for the safety related filtration systems consistent with the WBN Unit 1 Amendment are also provided.Enclosure 1 provides a discussion of changes to the FHA analysis currently described in the FSAR.Enclosures 2 through 5 provide red-line markup and final versions of FSAR Sections in Chapters 6, 9, and 15. Enclosures 6 through 9 provide the red-lined markup and final versions of the WBN Unit 2 TS and TSB consistent with Reference

1. Enclosure 10 shows the deletion of Technical Requirements Manual Section 3.9.1, "Decay Time." This requirement has been moved to a new TS Section 3.9.8, "Decay Time." The FSAR changes will be incorporated in Amendment 111. This is a new regulatory commitment.

If you have any questions, please call me at (423) 365-2004.I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of December, 2013.on Arent Director, Watts Bar Licensing Nuclear Construction U.S. Nuclear Regulatory Commission Page 3 December 12, 2013

Enclosures:

1. WBN Unit 2 Revised FSAR Section 15.5 Fuel Handling Accident Dose Analysis Results 2. WBN Unit 2 -Revised FSAR Section 15.5. -Red-Lined 3. WBN Unit 2 -Revised FSAR Sections 6.2, 6.5, 9.4. and 15.5 -Final 4. WBN Unit 2 -Revised FSAR Sections 6.2, 6.5, and 9.4 -Red-Lined 5. WBN Unit 2 -Revised FSAR Sections 6.2, 6.5, and 9.4 -Final 6. WBN Unit 2 -Revised Technical Specification Red-Line Markup 7. WBN Unit 2 -Revised Technical Specification

-Final 8. WBN Unit 2 -Revised Technical Specification Bases Red-Line Markup 9. WBN Unit 2 -Revised Technical Specification Bases -Final 10. WBN Unit 2 -Revised Technical Requirements Manual Section 3.9.1 U.S. Nuclear Regulatory Commission Page 4 December 12, 2013 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 1 WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis The Watts Bar Nuclear Plant (WBN) Unit 2 Fuel Handling Accident (FHA) was updated to use the Alternate Source Term (AST) described in Regulatory Guide (RG) 1.183 for an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch, or both doors in a personnel air lock, are open. The analysis for a dropped fuel assembly inside containment when the containment air locks and equipment hatch are closed continued to use the methodology of RG-1.25. This change was approved by the NRC as documented in NUREG-0847 Supplement 25, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2." Subsequently, WBN Unit 1 submitted a License Amendment Request (LAR) to selectively implement the AST for the FHA. The NRC approved this request June 19, 2013, as Amendment

92. The WBN Unit 1 LAR presented two cases. Case 1 was the FHA at the Spent Fuel Pool. Case 2 was an FHA in containment with the containment open. The discussion of the FHA with the containment isolated was removed from the Updated Final Safety Analysis Report (FSAR) by Amendment

92. The refueling mode Limiting Condition for Operation and associated Surveillance Requirements for the Purge System and the Auxiliary Building Gas Treatment System were removed from the Technical Specifications (TS) because no credit was taken in the analyses for the filtration units. The approval for these changes was included in Amendment 92.The WBN Unit 2 FSAR, TS and TS Bases are being revised to match those of WBN Unit 1. For WBN Unit 2, the discussion of the RG 1.25 analysis of the containment closed FHA is being removed from the FSAR. A table will be added in WBN Unit 2 FSAR, Section 15.5, providing the results for the containment open case to be consistent with what was done for WBN Unit 1.The WBN Unit 2 FSAR tables will not include the values for fuel with Tritium Producing Burnable Adsorber Rods (TPBARS), because they are not part of the WBN Unit 2 design basis. The WBN Unit 1 analyses were performed using the same meteorological data (X/Q) and wind speeds that form the basis for the WBN Unit 2 FSAR, Section 15.5 dose analyses.

Thus, the results are consistent with the WBN Unit 2 AST approval documented in SSER 25.The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open or closed. The release point for the containment purge system is the WBN Unit 2 shield building stack. The X/Qs are lower for this release point than the normal Auxiliary Building exhaust. In addition, any release from the shield building stack would go through the purge system High Efficiency Particulate Air (HEPA) and charcoal filter assemblies prior to release. Currently, when the purge lines isolate on high radiation, the Auxiliary Building also isolates and the Auxiliary Building Gas Treatment System (ABGTS) is actuated.

The release point for ABGTS is the shield building stacks, and the releases are filtered through HEPA and charcoal assemblies.

Thus, the AST analysis for the FHA in the Auxiliary Building that considers no filtration and no Auxiliary Building isolation is conservative and acceptable as the basis for the containment open evaluation.

When the purge valves close at approximately 12.7 seconds with the containment closed, any further release of radioactivity would be terminated.

If the purge valves did not close and the releases continued from the shield building stack, the results would be bounded by the FHA in the Auxiliary Building.E1-1 Enclosure 1 WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis This change is determined to be acceptable because: 1) If the containment closed case were evaluated using the AST, the results would be bounded by the cases currently presented in the FSAR, and 2) This will bring the WBN Unit 2 FSAR discussion of this event into agreement with the recently approved WBN Unit 1 LAR.As part of this selective implementation of AST, the following changes are assumed in the analysis:* The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.* The gap activity is revised to be consistent with RG-1.183." The decontamination factors were changed to be consistent with RG-1.183.* New onsite (control room) and offsite atmospheric dispersion factors (X/Q) are used." The time to isolate the control room is increased from 20.6 seconds to 40 seconds.* No Auxiliary Building isolation is assumed.* No filtration of the release from the Containment or the spent fuel pool to the environment by the containment purge filters or the ABGTS is assumed.The WBN design includes a secondary containment that is designed to limit any potential radioactive leakage to the outside environment following a Design Basis Accident (DBA). The secondary containment consists of the concrete shield building that encloses the steel primary containment and the portion of the Auxiliary Building called the Auxiliary Building Secondary Containment Enclosure.

The Secondary Containment is described in FSAR Section 6.2.3. The secondary containment structures work in conjunction with safety related ventilation systems and the appropriate isolation of normal ventilation systems to perform its safety functions.

In addition to the descriptions in FSAR section 6.2.3, the air clean-up and filtration systems are described in FSAR Section 6.5. The DBA Loss of Coolant Accident (LOCA) is the accident that generally dictates the basis for the design of the Secondary Containment.

In addition to the LOCA, the FHA analyses performed based on RG-1.25 that were part of the original licensing basis for WBN resulted in safety functions being defined for the Secondary Containment.

If an FHA occurred either in the Auxiliary Building or in primary containment, the ABGTS was required to start and the Auxiliary Building normal ventilation system isolated.

A discussion of the Auxiliary Building Ventilation System is provided in FSAR Sections 9.4.2 and 9.4.3. If the FHA occurred in the primary containment, credit was taken for the Reactor Building Purge Filtration system in the FSAR Chapter 15 dose analysis.

A general description of the Reactor Building Purge System is provided in FSAR Section 9.4.6. The FHA based on the AST does not credit containment or Auxiliary Building isolation.

No credit is taken for the high efficiency particulate and charcoal filter systems associated with the ABGTS and the purge system. The approved WBN Unit 1 amendment removed the TS requirements associated with refuel mode operation for these systems. The WBN Unit 2 FSAR revisions associated with the approved WBN Unit 1 TS changes are provided in this submittal.

FSAR Section 6.2.4.3 on containment isolation discusses administrative controls to manually close the ice blowing penetrations in the event of an FHA in containment.

The updated FHA based on the AST does not require containment isolation to meet dose criteria.

This section has been revised accordingly.

E1-2 Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis The following summarizes the specific changes to the FSAR and TS.1. Update FSAR Section 15.5.6 to remove the RG-1.25 based analysis of a FHA in containment with the containment isolated except for the purge system.2. Revise FSAR Sections 6.2.3.1.1 and 6.2.3.1.3 to remove the FHA as a design basis for the Secondary Containment and ABGTS.3. Revise FSAR Section 6.2.4.3 to remove the administrative requirement to manually isolate the ice blowing penetrations for an FHA.4. Revise FSAR Sections 6.5.1.1.3 and 6.5.1.2.3 to remove the discussion of the Reactor Building Purge System design basis for the FHA.5. Revise FSAR Section 9.4.2.3 to remove the requirement for the Auxiliary Building normal ventilation system to isolate for an FHA.6. Revise FSAR Section 9.4.6 on the Reactor Building Purge System to remove the FHA as a design basis for the system.7. Add new WBN Unit 2 TS 3.9.10 and associated Bases Section to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the reactor core has become sub-critical.

TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.8. Modify WBN Unit 2 TS 3.3.6, "Containment Vent Isolation Instrumentation";

TS 3.3.8,"Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation";

and TS 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)", to eliminate the requirements associated with movement of irradiated fuel assemblies in the containment or the fuel handling area. Modify associated TS Bases.9. Eliminate TS 3.9.4, "Containment Penetrations", and TS 3.9.8, "Reactor Building Purge Air Cleanup Units".10. Modify WBN Unit 2 TS 5.7.2.14 to remove RG-1.52 testing of the Reactor Building Purge HEPA and Charcoal Filter Units.11. Modify WBN Unit 2 TS 5.7.2.20 to incorporate the Control Room dose limit defined in 10 CFR 50.67(b)(2)(iii).

12. Modify TS Bases 3.6.1, "Containment Penetrations";

3.6.2, "Containment Air Locks";and 3.6.3, "Containment Isolation Valves", to eliminate isolation requirements during fuel movement inside containment.

Delete TS Bases 3.9.4.13. Modify TS Bases 3.7.13, "Spent Fuel Pool Level," and 3.9.7, "Reactor Cavity Water Level," to update references associated with AST.14. Remove the decay time restriction on post shutdown irradiated fuel movement from Section 3.9.1 of the Technical Requirements Manual. This restriction has been added to the TS as described in Item 7 above.Enclosures 2 and 3 provide a red-lined mark-up and a final version of FSAR Section 15.5.6 on the FHA. Enclosures 4 and 5 provide a red-lined mark-up and a final version of FSAR Sections 6.2.3, 6.2.4, 6.5, and 9.4. Enclosures 6 and 7 provide the red-lined mark-up and final version of the WBN Unit 2 TS sections as enumerated in the numbered list immediately above.Enclosures 8 and 9 provide the red-lined mark-up and final version of the TS Bases sections associated with TS listed in items 7 through 9 of the list provided above. Enclosure 10 deletes Technical Requirements Manual Section, 3.9.1.E1-3 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 E2-1 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 15.5.6 Environmental Consequences of a Postulated Fuel Handling Accident The analysis of the fuel handling accident considers two thee cases. The f4r6t crae

• for a Fuel Handlina Acciden~t Oncido containim~ent With the containment clocod and the Reactor B3Uilding R d P P d --I L J ----P-urge ýiystem operating. , nis anaiysis is emsecussee in ýiecien 15519 ana is haseeo Regulator, G 1 25 anRd -1 IUREG P_-The first SeGend case is for an accident in the spent fuel pool area located in the Auxiliary Building.

This case is discu..sed in Section 15..6.2 and evaluated using the Alternate Source Term (AST) based on Regulatory Guide 1.183[18],"Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The second thrd case considered is an open containment case for an accident inside containment where there is open communication between the containment and the Auxiliary Building.

This evaluation is also based on the AST digc's-sed in SeACtio 15.5.6.2 and ie-based-eG Regulatory Guide 1.183. The parameters used for this analysis are listed in Table 15.5-20.a.

I I A A P I i l.1~b1 03I-up: naaigACOfI ao n1oua~r uo .I P II I li i II I I A P AA Ine parameters used .. ts analysis are listedinlTabe165-20

-" " The Bases +dr the rtguiatory

'uiuuA 4 .2 O e.i ga-I uatien isru.+/-L a. A ^_ r _ _.. --An ^^ .- --_-1--i-I Ifl mu r~uuu:amur:

'~..uiuu i ..~ ~ina:v~i~.

mu ~cuiuvnm u~uur~ uu nuuru JIIU[ ui.inm r i St., ,~flnShn ~flflS5p5tktpp nasal, a. 9155 ,,pr.,nr.

nraflI 559 *nttpfltflflI 55 nfln .na InYarttaI r.an.,aan

5. *15----U Wfl~WflflUSJU S 1.fl ~5NIfl US *1 I t5 7 .EIAI IS I~ IS 1 IS I 15 V S SVV~Is * ,t.,n artid r.I..na.sar.*

~ *ka # re* s w*#. sat ..,nnam ki.. mn*n *k 5 s rwan* , sal .It;~ to *Ian 0 Rte aGGOU~t P ii I Q'2 In theO iRquiater;

~u:e 4.25b analyscis eaFamae is assume-a gar all mroasn on asseMol1y.

fI gl m II 3 Thno aasomnly eamagga is the nignest PoW8ere assemiqiy in Me core region W 1DOýg_ýnýr ý ý xsý i iýý r m ii i iý ýý -m rO tj nxfflýn ^r ýý n assemnbly are GaIcUlated assuming full poWer operation at the_ enRd Of core life immediately pFeceding shutdeown.

W-uclear core characteristics used in the analysiS are given inTal I1. I 21. A radial peaking factor of 1.65 is used.444 For the Reulat8rV Guide 1.25 analysis all of the aaV ati*it'V in the damaged rodn is released to the seont fue: pool and consists of 109 ncoT the total nobie gases and I raa~oacIGGe 1Roaine O :nenr :nMe FRoS a! Me Uime of the aiccdent with the moiiweVAR gap percentage opti9Aons, wfc- are- boaed_ N,1UK FM, / ,UU0 [24] as 14% tv the Kr 85, 5 VA o If the Xe 133, 2% of the Xe 135, and 12A of the 1 131.(5) Noble gases released in the GOntainFMet are released through the Shield Bu ilding Vent to the enuironment.

(6) In the Regulator; Guide 1.25 analysis the iodine gap ineno; s Rposed cfinrac Species (09.75%) and organic species (0.25%).(7) A fiter efficiency of 90% for inor~ganic iodine and 30% for organic- iodine for the purge air exhaust filters is used sinca8 no relative humidity control is provi~ded.

E2-2 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 (9) No GUAdi ;A taken for- nntu r l docay after the activity has been released to the atmeeSPI4e.

(9) The shedt term (i.e., 0 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) atmoe6Pheri dluio factO~r, at the ecuinarea boundar;and low population zone giyen in Ial r8A 2 are used. The thyroid decoa utilizmel

'GRP 30 [25] iodine dose conyeerion factorc. Doece aro bacod en the dose moedeic prOcented i n Appenidix 1 5A.1D.O.U.Z~M ruuiJ L'uun .'cwun D~5U~ uu Mur; '.iu ,AEuu u. io The analysis of a poctu_':atej fuel h ndllng accGident in the Au*iliary Bguilding rf&ueling Arean Or~~- tlký+ t j,. D ...I,,... .1 4 412 Aiýý + Q~t:TeR~ns (AST)The bases for evaluation are: (1) IF; the Regulatey Guide ,. ,,3 , ,;a " +. The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after plant shutdown.

Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.(2) In the RegulatoryG .183 ., ", Damage was assumed for all rods in one assembly.(3) The assembly damaged is the highest powered assembly in the core region to be discharged.

The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown.

Nuclear core characteristics used in the analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.(4) The Guide 1. 183 analyci' aaume. a All of the gap activity in the damaged rods is released to the spent fuel pool and consists of 8% 1-131, 10% Kr-85, and 5% of other noble gases and other halogens.(5) Noble gases released to the Auxiliary Building spent fuel pool are released through the Auxiliary Building vent to the environment.

(6) In the Regulatory Guide 1.183 analysic" t The iodine gap inventory is composed of inorganic species (99.85%) and organic species (0.15%).(7) in the Regulatery Guide 1.183 analyie,m The overall inorganic and organic iodine spent fuel pool decontamination factor is 200.(8) k, the R.gulatey Guide 1.183 analys ", a All iodine escaping from the Auxiliary Building spent fuel pool is exhausted unfiltered through the Auxiliary Building vent.E2-3 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 (9) The release path for the containment scenario is changed to include 12.7 seconds of unfiltered release through the Shield Building vent, with the remainder of the unfiltered release through the Auxiliary Building vent.(10) No credit is taken for the ABGTS or Containment Purge System Filters in the analysis.(11) No credit is taken for natural decay either due to holdup in the Auxiliary Building or after the activity has been released to the atmosphere.

(12) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [25] iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.1.6..6.3 Fuel Handling Accident Results The r-adatio, do- rcul-te Of the RegulatoI

Guide 1.25 4 .ith the co-nta-in ent nlocd fe lal handling accaident (FH4A) are giV8n in T-abl 15.5 23. For a FHA inside containm~ent, no allo.A.nre has boon made for p bleholdup or Mi*ai in the ;. containment Or i;olation of the ntainmeRt 2n a roeult of a high radiatioFn; ignal froam tha thr ventilationm cem o the cass where containment penetrations are cloced to the Auxiliar-Building.

However, the contafinment purge filterc are Gredited.

Dese equations in TID 11811[23] were ucedt d~ete_:ine t-he dCoI-6. DeOS cenVe~iGn factore in !RP 30 [25] Wores used to detemine thYroid doses in placo of thoce found nTD181 The Wentilation funtionR Of the reactorbuilding purge ventilating system (RBP31VS) is noet a safey reltatd fucion Hot*Rwever the filtration unitg and aseociated exhaucit dch ibosrk do provide a safet related flrto path following a fuel handling acciden~t prior to aultomatic c-1ocuIe-GOf th accoc-aiatod isolation valves. The RAWPV S containc, air cleanup units with prefiltorc, H4EPA filtorc, an~d 2 inhthick charcoal AdcrbeA hccce i -tiiart the auiliar; buildinig gas treatment system SXcept that the la#8r is equipped with 4 inch thick charcoal ndcrebeFS.

An~ytime fuel handling operations are being care on icid the primary containmenRt, either the containment is isolated or the reactor buildin~g purge filtration system or, operational.

The assumptions listed above are, therefore, applic-able to a fuel handling aciet fniepimary Gentaonmef*

The thyroid, gamma, and beta doese for FHAs for the% cleced containment are given in Table 15a.69 23 for the eAcnucMon area boundar; and low population ZonRe. These doese are 10cc than 25% oth10 CFR 1 00.11 "Mitts Of 30 reM. to9 the thyroid, and 25 rem gamma to the whole body. These doese are caIculated using the comAputer code FEFNGOOSE

[16].The whole body, beta, and thyroid decec89 to control room pe-GRconn fro-m the radiation SOurcec diccupcced-above are precented OR T-able 15. fi23. T-he doses are calculated by the COROID rc9mputeFrcode

[17]. Parame-terc for the ontrol room anaiycis a re found in Table 15,.5 14. The deco to whole body is beloew the 10Q CFWR 5-0 Appendix A, GDCQG 190 limit of 5 r.m. forF conrol room nerconRRA and the thymo~d decoA ic elo the !imit of 30 rem._E2-4 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 The radiation dose results of the Regulatory Guide 1.183 fuel handling accident (FHA) are given in Table 15.5-23. Alte ..ate -ourc- term-, (A^T) The AST described in RG 1.183 was selectively used to evaluate the FHA due to an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch or both doors in a personnel air lock are open.As part of this selective implementation of AST, the following assumptions are used in the analysis:* The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11." The gap activity is revised to be consistent with that required by RG 1.183." The decontamination factors were changed to be consistent with those required by RG.1.183.* No Auxiliary Building isolation is assumed.* No filtration of the release from the Containment or the spent fuel pool to the environment by the Containment Purge filters or the ABGTS is assumed.The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open or closed. The release point for the containment purge system is the Unit 2 shield building stack. The X/Qs are lower for this release point than for the normal auxiliary building exhaust. in addition, any .eleae the shield building , tac, Currently, when the purge oR high radiation, the buildiRg alco iclate. and-are filtred through HEPA and, Charcoal a.. emblie. Thu- The AST analysis for the FHA in the Auxiliary Building that considers no filtration is conservative and acceptable as the basis for the containment evaluation.

The thyrFid, gamma, and beta TEDE for FHAs in the Auxiliary and the open containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are leee than 25% of the 10 CFR 100.11 limits of 300 rem to the thyro~d, and 25 rem gamma to the IAho-'le" -body and less than the 10 CFR 50.67 limit of 6.3 2-6 rem TEDE. These doses are calculated using the computer code FENCDOSE [16].The TEDE whole body, beta, and thy,.id doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body i" b"elow the 10DO CF ,-R 5-00 Appendix A, GDCr 10- limit of 5% for control room personnel, and the thyroid doce is below the limit of 30 rFem aRd the 1 OCFR 50.67 limit of 5 rem TEDE.E2-5 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 Tabkl--l&62C Used In Fuel Handl. i Regulator' Guide 1.26* t X Time between plant AchutdWnCA a2nd acciden~t Damage to- fuel agmcembly Fuel aseombly activity A.4IIaYJII~

frel a-e p-nt fuel poo1 All Fedo ,r~r Gap arti~ity in ruptured rods(4)-ROWS-(2)V~i V-aylal peakic' g acOr Form of iodine acti'.ity roloasod me~thyl idn elemental odn 90916 30%Amutof mnixing of activity in Auxiliary Building Non MeteOrology See Table 15.15 14 And] Table 15A 2 (1 ) 10% of the, total radvioactwe iodine S*copt for 129% o-f I131 agnd- 10%9A of total no-ble gases, excopt for 141A far Kr 85, 5% for Xe 133 and 2% for Xe 135 in the damaged rods at the time of the-a6ggidt (2) Reactor Buwilding Purge Ventilation System E2-6 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 Table 15.5-20.a Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.183 Analysis Time between plant shutdown and accident 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gap activity in ruptured rods(l)Radial peaking factor 1.65 Form of iodine activity released to spent fuel pool elemental iodine 99.85%(AST) methyl iodine 0.15%(AST)

Decontamination factor in spent fuel pool AST Overall=200 Filter efficiencies No credit taken Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Tablel5A-2 (1) 8% 1-131, 10% Kr-85, and 5% other gasses and other halogens.E2-7 Enclosure 2 WBN Unit 2 Red-line Markup of FSAR Section 15.5.6 Table 15.5-23 Doses From A Fuel Handling Accident (FHA) (rem)Doses from Fuel Handling Accident Regulatory Guide 1.183 Analyses FHA in Auxiliary Building (rem) or In --nt-ainm ant Contnin mant Onon troml 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 3.OOQI0r-01 4.29F5 01 9.27SE 021-.20E 01 1.9-5E4 A Q016r- -QR6- 64 Seta l.177E*O l.IOE+OO 2.7-3 E 013.33E 01 1.068E*0u.68E+O0, ThYrid ICRPD30 5.514+4 2A3re- 1.4 E l r- I+ I l -4.Q51r+0.32E+O4 TEDE 2.38E+00 6.66E-01 1.02E-00 Doses from Fuel Handling Accident Regulatory Guide 1.183 Analyses FHA in Containment

-Containment Open (rem)2 HR EAB 30 DAY LPZ CONTROL ROOM G m3.94E 01 4.29E 01 Q.278E 021 .20E 01 4 935rE 04r; t5.8 01 Beta 1.1:77E9+00

1. 1 E+Q0 2.7-34E 043.33E 01 4.068E+001.69rE+00 T-hYroid ICRP 30 1.67-7E+90 5.51 E*01 3.663E Q11.51E+01 I1.510Er+001.32E+01Q TEDE 2.38E+00 6.66E-01 1.01 E-00 Deoe frmom Fuel Handling Accidont RogulatoY Guide 1.25 FHA In Ro-anter BRuilding, Containmont Closed (MRm), 2 MR EAB 30 DAY LPZ CONTRO' ROOM Ga.ma 4.102E 014.31E 01 9.629E 024.21E 01 2.6V, 77EV Q"1V2.72E 01 Beta 1.1 8_2E_+00Q1.24E+00 2.746-E-C013.48E 01 2.207-E*002.25E+00 ThYroid ICRP 30 39.42E*004.15GE+01

9. 15 8F+*00QI.

16 E +0-1 5 2090E4006 A IE+00 E2-8 Enclosure 3 WBN Unit 2 -Revised FSAR Section 15.5 Final E3-1 Enclosure 3 WBN Unit 2 -Revised FSAR Section 15.5 Final 15.5.6 Environmental Consequences of a Postulated Fuel Handling Accident The analysis of the fuel handling accident considers two cases. The first case is for an accident in the spent fuel pool area located in the Auxiliary Building.

This case is evaluated using the Alternate Source Term based on Regulatory Guide 1.183118], "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The second case considered is an open containment case for an accident inside containment where there is open communication between the containment and the Auxiliary Building.

This evaluation is also based on the AST and Regulatory Guide 1.183. The parameters used for this analysis are listed in Table 15.5-20.a.

The bases for evaluation are: (1) The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after plant shutdown.

Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.(2) Damage was assumed for all rods in one assembly.(3) The assembly damaged is the highest powered assembly in the core region to be discharged.

The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown.

Nuclear core characteristics used in the analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.(4) All of the gap activity in the damaged rods is released to the spent fuel pool and consists of 8% 1-131, 10% Kr-85, and 5% of other noble gases and other halogens.(5) Noble gases released to the Auxiliary Building spent fuel pool are released through the Auxiliary Building vent to the environment.

(6) The iodine gap inventory is composed of inorganic species (99.85%) and organic species (0.15%).(7) The overall inorganic and organic iodine spent fuel pool decontamination factor is 200.(8) All iodine escaping from the Auxiliary Building spent fuel pool is exhausted unfiltered through the Auxiliary Building vent.(9) The release path for the containment scenario is changed to include 12.7 seconds of unfiltered release through the Shield Building vent, with the remainder of the unfiltered release through the Auxiliary Building vent.(10) No credit is taken for the ABGTS or Containment Purge System Filters in the analysis.(11) No credit is taken for natural decay either due to holdup in the Auxiliary Building or after the activity has been released to the atmosphere.

E3-2 Enclosure 3 WBN Unit 2 -Revised FSAR Section 15.5 Final (12) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [25] iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.15.5.6.3 Fuel Handling Accident Results The radiation dose results of the Regulatory Guide 1.183 fuel handling accident (FHA) are given in Table 15.5-23. The AST described in RG 1.183 was selectively used to evaluate the FHA due to an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch or both doors in a personnel air lock are open. As part of this selective implementation of AST, the following assumptions are used in the analysis:* The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.* The gap activity is revised to be consistent with that required by RG 1.183." The decontamination factors were changed to be consistent with those required by RG.1.183." No Auxiliary Building isolation is assumed." No filtration of the release from the Containment or the spent fuel pool to the environment by the Containment Purge filters or the ABGTS is assumed.The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open or closed. The release point for the containment purge system is the Unit 2 shield building stack. The X/Qs are lower for this release point than for the normal auxiliary building exhaust. The AST analysis for the FHA in the Auxiliary Building is conservative and acceptable as the basis for the containment evaluation.

The thyroid, gamma, and beta doses for FHAs in the Auxiliary and the open containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are less than the 10 CFR 50.67 limit of 6.3 rem TEDE. These doses are calculated using the computer code FENCDOSE [16].The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to control room personnel is below the 10CFR 50.67 limit of 5 rem TEDE.E3-3 Enclosure 3 WBN Unit 2 -Revised FSAR Section 15.5 Final Table 15.5-20.a Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.183 Analysis Time between plant shutdown and accident 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gap activity in ruptured rods(l)Radial peaking factor 1.65 Form of iodine activity released to spent fuel pool elemental iodine 99.85%(AST) methyl iodine 0.1 5%(AST)Decontamination factor in spent fuel pool AST Overall=200 Filter efficiencies No credit taken Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Tablel5A-2 (2) 8% 1-131, 10% Kr-85, and 5% other gasses and other halogens.E3-4 Enclosure 3 WBN Unit 2 -Revised FSAR Section 15.5 Final Table 15.5-23 Doses From A Fuel Handling Accident (FHA) (rem)FHA in Auxiliary Building (rem)2 HR EAB 30 DAY LPZ C TEDE 2.38E+00 6.66E-01 FHA in Containment

-Containment Open (rem)2 HR EAB 30 DAY LPZ TEDE 2.38E+00 6.66E-01 C ONTROL ROOM 1.02E-00 ONTROL ROOM 1.01 E-00 E3-5 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 E4-1 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 6.2.3 Secondary Containment Functional Design Structures included as part of the secondary containment system are the Shield Building of each reactor unit, the Auxiliary Building, the Condensate Demineralizer Waste Evaporator (CDWE) Building and the essential raw cooling water (ERCW) pipe tunnels adjacent to the Auxiliary Building.

Depending on the configuration of the plant, the Primary Containment Building(s) may also be included as a structure which is part of the secondary containment system. This condition exists when the primary containment is open to the Auxiliary Building.The emergency gas treatment system (EGTS) is provided for ventilation control and cleanup of the atmosphere inside the annulus between the Shield Building and the Primary Containment Building.

The Reactor Building purge air system is also available for cleaning up the atmosphere inside the Shield Building Annulus. Refer to Section 9.4.6 for further details relating to the purge air system. The Auxiliary Building Gas Treatment System (ABGTS) provides a similar contamination control capability in the Auxiliary Building Secondary Containment Enclosure (ABSCE),which includes all of the areas listed above.6.2.3.1 Design Bases 6.2.3.1.1 Secondary Containment Enclosures Design bases for the secondary containment structures were devised to assure that an effective barrier exists for airborne fission products that may leak from the primary containment, or the Auxiliary Building fuel handling area, during a loss-of-coolant accident (LOCA),-eF-a-fuel handling a,.id.. t (FH)" .Within the scope of these design bases are requirements that influence the size, structural integrity, and leak tightness of the secondary containment enclosure.

Specifically, these include a capability to: (a) maintain an effective barrier for gases and vapors that may leak from the primary containment during all normal and abnormal events;(b) delay the release of any gases and vapors that may leak from the primary containment during accidents; (c) allow gases and vapors that may leak through the primary containment during accidents to flow into the contained air volume within the secondary containment where they are diluted, held up, and purified prior to being released to the environs; (d) bleed to the secondary containment each air-filled containment penetration enclosure which extends beyond the Shield Building and which is formed by automatically actuated isolation valves; (e) maintain an effective barrier for airborne radioactive contaminants, gases, and vapors originating in the ABSCE during normal and abnormal events. Refer to Sections 3.8.1 and 3.8.4 for further details relating to the design of the Shield Building and the Auxiliary Building.6.2.3.1.3 Auxiliary Building Gas Treatment System (ABGTS)The design bases for the ABGTS are: 1. To establish and keep an air pressure that is below atmospheric within the portion of the buildings serving as a secondary containment enclosure during accidents.

2. To reduce the concentration of radioactive nuclides in air releases from the secondary containment enclosures to the environs during accidents to levels sufficiently low to keep the site boundary and LPZ dose rates below the 10 CFR 100 guideline values.3. To minimize the spreading of airborne radioactivity within the Auxiliary Building following an accidental release in the fuel handling and waste packaging areas.ABGTS is not required to mitigate the consequences of a fuel handling accident.4. To withstand the safe shutdown earthquake.

E4-2 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 5. To provide for initial and periodic testing of the system capability to function as designed (See Chapter 14.0 for information on initial testing of systems).6.2.3.2 System Design 6.2.3.2.1 Secondary Containment Enclosures (1) Shield Building[Text not provided as no changes needed](2) Auxiliary Building[Text not provided as no changes needed](3) Auxiliary Building Secondary Containment Enclosure (ABSCE)The Auxiliary Building secondary containment enclosure (ABSCE) is that portion of the Auxiliary Building and CDWE Building (and for certain configurations, the annulus and primary containment, as discussed below) which serves to maintain an effective barrier for airborne radioactive contaminants released in the auxiliary building during abnormal events. Mechanical and electrical penetrations of this enclosure are provided with seals to minimize infiltration.

[Text for next 3 paragraphs no provided as no changes were needed]During periods when the primary containment and/or annulus of both units are open to the Auxiliary Building, the ABSCE also includes these areas. Qwiig fuel hanidling oporatines OF; this configuration, a high radiation; signal fromA spen~t fuel pool radiation moeniters will ro-sult in -a Contain~mlent Ventilation IseolatiOn (CVI) i 1.1Lu: aA AWM1 aF N6II~j~I11GA IMATIJUR RRA1 AN -+. LWR1. 0Ii1 IaHY, a 1.~ 19A 17i icuiga CVI signal generated by a high radiation Gignaal from. ths containment eua arohaU~t Fadiatoen Menitors vill initiato an Au*iIiarj Building if'tolaio

and etart of ,A BGTS. L' ikeve, a A Containment Isolation Phase A (Sl Signal) from the operating unit or high temperature in the Unit 1 or Unit 2 Auxiliary Building air intake, or manual ABI will cause a CVI signal in the refueling unit. These actions will ensure proper operation of the ABSCE. Both doors of the containment vessel personnel airlocks may be open at the same time during refueling activities while the purge air ventilation system is operating.

During fuel handling operations in this configuration, a high radiation signal from spent fuel pool radiation monitors will result in a Containment Ventilation Isolation (CVI) in addition to an Auxiliary Building isolation and ABGTS start. Similarly, a CVI signal, including a CVI signal generated by a high radiation signal from the containment purge air exhaust radiation monitors, will initiate an Auxiliary Building isolation and start of ABGTS.These are not required functions for the ABGTS and the Reactor Building Purge System filters or for Purge System isolation as no credit for these features is in mitigating a fuel handling accident.

Under .p...a!

cOntr.l., one e.the airloc"k deoFr at each !cGatien v."1I be clocod and the purge air Y-ntilation haRnIdig to .n..rAMe A^BRR-rsC boun.da ;ntegrity.

In the case where E4-3 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 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.6.2.3.2.3 Auxiliary Building Gas Treatment System (ABGTS)The ABGTS is a fully redundant air cleanup network provided to reduce radioactive nuclide releases from the secondary containment enclosure during accidents.

It does this by drawing air from the fuel handling and waste packaging areas through ducting normally used for ventilation purposes to air cleanup equipment, and then directing this air to the reactor unit vent. In doing so, this system draws air from all parts of the ABSCE to establish a negative pressure region in which virtually no unprocessed air passes from this secondary containment enclosure to the atmosphere.

During periods when the primary containment and/or annulus of both units are open to the Auxiliary Building, the ABSCE also includes these areas. The ABGTS has been designed to establish a negative pressure in these additional areas for this configuration.

During fuel handling operations in this configuration, a high radiation signal from the spent fuel pool radiation monitors will result in a Containment Ventilation Isolation (CVI) in addition to an Auxiliary Building isolation and ABGTS start. Similarly, a CVI signal, including a CVI signal generated by a high radiation signal from the containment purge air exhaust radiation monitors, will initiate an Auxiliary Building isolation and start of ABGTS.Likewise, a Containment Isolation Phase A (SI Signal) from the operating unit or high temperature in the Unit 1 or Unit 2 Auxiliary Building air intake, or manual ABI will cause a CVI signal in the refueling unit. These actions will ensure proper operation of the ABSCE. However, as an added precaution to protect the ABGTS operational boundary, operational action is needed to ensure the closure of the containment purge exhaust isolation valves (system valves not containment isolation valves) which are controlled by hand switches.

In the case where 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.E4-4 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 6.2.3.3.3 Auxiliary Building Gas Treatment System (ABGTS)The ABGTS has the capabilities needed to preserve safety in accidents as severe as a LOCA. This was determined by conducting functional analyses of the system to verify that the system has the proper features for accident mitigation which consist of a failure modes and effects analysis, a review of Regulatory Guide 1.52 sections to assure licensing requirement conformance, and a performance analysis to verify that the system has the desired accident mitigation capabilities.

A detailed failure modes and effects analysis is presented in Table 6.2.3-3.The functional analyses conducted on the ABGTS have shown that: 1. The air intakes for the system are properly located to minimize accident effects.The use of the air intakes provided in the fuel handling and waste disposal areas minimizes the spread of airborne contamination that may be accidentally released at these positions in which the probability of an accidental release, e.g., a fuel handling accident, is more likely. This localization effect is provided without reducing the effectiveness of the system to cope with multiple activity released throughout the ABSCE that may occur during a LOCA. Such coverage is accomplished by utilizing the normal ventilation ducting to draw outside air inleakage from any point along the secondary containment enclosure to the fuel handling and waste disposal areas.2. Accident indication signals are utilized to bring the ABGTS into operation to assure that the system functions when needed to mitigate accident effects.Accidents in which this system is needed to preserve safety are automatically detected by at least one of the three instrumentation sets used to generate accident signals that result in system startup.3. System startup reliability is very high. The practice of allowing the automatic startup of both full capacity trains in the system gives greater assurance that one train of equipment functions upon receipt of an accident signal.4. The method adopted to establish and keep the negative pressure level within this secondary containment enclosure minimizes the time needed to reach the desired pressure level. Initially, the full capacity of the ABGTS fans is utilized for this purpose. After reaching the desired operating level, the system control module allows outside air to enter the air flow network just upstream of the fan at a rate to keep the fans operating at full capacity with the enclosed volume at the desired negative pressure level. In this situation, the amount of air withdrawn from the enclosed volume is equal to the amount of outside air inleakage through the ABSCE. In addition, two vacuum breaker dampers in series are provided to admit outside air in case the modulating dampers fail.5. The ABSCE is maintained at a slightly negative pressure to reduce the amount of unprocessed air escaping from this secondary containment enclosure to the atmosphere to insignificant quantities.

In addition, this negative pressure level is less than that which is maintained within the annulus; such that, any air leakage between the Auxiliary Building and the Shield Building is from the Auxiliary Building into the Shield Building.E4-5 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 6. The Train A and Train B air cleanup units are sufficiently separated from each other to eliminate the possibility of a single failure destroying the capability to process Auxiliary Building air prior to its release to the atmosphere.

Two concrete walls and a distance of more than 80 feet separate the two trains. The use of separate trains of the emergency power system to drive the air cleanup trains gives further assurance of proper equipment separation.

The review of the ABGTS conducted to determine its conformance with Regulatory Guide 1.52 has shown that this system, designed prior to issuance of the guide, is in general agreement with its requirements.

Details on compliance with Regulatory Guide 1.52 are given in Table 6.5-2.The performance analysis conducted to verify that the ABGTS has the required accident mitigation capabilities has shown that the system flow rate is sized properly to handle all expected outside air inleakage at a 1/4-inch water gauge negative pressure differential.

This indicates that the nominal flow rate of 9000 cfm is sufficient to assure an adequate margin above the expected ABSCE inleakage (ACU filters are replaced as needed to maintain a minimum flow capability of 9300 cfm under surveillance instructions).

The performance analysis evaluated the capability of the ABGTS to reach and maintain a negative pressure of 1/4-inch water gauge with respect to the outside within the boundaries of the ABSCE. The following was utilized in the analysis: 1. Leakage into the ABSCE is proportional to the square root of the pressure differential.

2. Only one air cleanup unit in the ABGTS operates at the rated capacity.3. The air cleanup unit fan begins to operate 30 seconds after the initiation of an ABI signal, Or a high .adiati.n .ig.Ral (Seo Sotion

4. The initial static pressure inside the ABSCE is conservatively considered to be atmospheric pressure, although the ABSCE is under a negative pressure during normal operation.

5. The effective pressure head due to wind equals 1/8-inch water gauge.6. Initial average air temperature inside the ABSCE equals 140 0 F.7. Atmospheric temperature and pressure are 70°F and 14.4 psia, respectively.

8. ABSCE isolation dampers/valves close within 30 seconds after receiving an ABI or a high radiation signal, S*copt fG" tho fuol handling a..a ".hau.t damp,..hIch& muc- t niaga ;Aithin (W3 soconde.9. The non-safety-related general ventilation and fuel handling area exhaust fans are designed to shut down automatically following a LOCA. Each fan is provided with a safety related Class 1 E primary circuit breaker and a safety related Class E4-6 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 1 E shunt trip isolation switch which is tripped by a signal of the opposite train from that for the primary circuit breaker to ensure that power is isolated from the fan.6.2.4.3 Design Evaluation The containment isolation systems are designed to present a double barrier to any flow path from the inside to the outside of the containment using the double-barrier approach to meet the single-failure criterion.(e) The design configuration for penetrations X-79A (ice blowing), and X-79B (negative return) is temporarily modified in operating Modes 5 and 6 and when the reactor is defueled (Mode 7) to support ice blowing activities.

The normally closed blind flange on each penetration will be opened and temporary piping will be installed in the penetrations.

A 12-inch silicone seal will be installed between the piping segment and the penetration.

Manual isolation valves will be connected to the piping on the inboard and outboard side of the penetrations.

This configuration is being installed to permit ice blowing operations to occur concurrently with fuel handling activities inside containment.

Admo, *,tr.atko

c. ntrol. ;Aill .n.uro tiN,, l.,ur o...f th , t to ;a ful h.nling ...'den. The penetrations will be returned to their normal design configuration prior to entry into Mode 4 operations.

E4-7 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 Engineered Safety Feature (ESF) Filter Systems Four Engineered Safety Feature (ESF) air cleanup systems' units are provided for fission product removal in post-accident environments.

These are: (1) The emergency gas treatment system (EGTS) air cleanup units.(2) The Auxiliary Building gas treatment system (ABGTS) air cleanup units.(3) The Reactor Building purge system air cleanup units.(4) The Main Control Room emergency air cleanup units.6.5.1.1 Design Bases 6.5.1.1.1 Emergency Gas Treatment System Air Cleanup Units The design bases are: (1) To provide fission product removal capabilities sufficient to keep radioactivity levels in the Shield Building annulus air released to the environs during a DBA LOCA sufficiently low to assure compliance with 10 CFR 100 guidelines.

(2) These air cleanup units are a part of the EGTS. See Section 6.2.3.1.2 for the design bases for other portions of this system.6.5.1.1.2 Auxiliary Building Gas Treatment System Air Cleanup Units The design bases are: (1) To provide fission product removal capabilities sufficient to keep radioactivity levels in the Auxiliary Building secondary containment enclosure (ABSCE) air released to the environs during a postulated accident sufficiently low to assure compliance with 10 CFR 100 guidelines.

(2) These air cleanup units are a part of the ABGTS. See Section 6.2.3.1.3 for the design basis for other portions of this system.6.6.4.1.3 Building Pug, Air- Syctom Ai.r C-lonnup Units;The design bases aro: (i ) To pvvvidv fission product remew c vapabilities sufficient to keep nadioavtivity levelc in the primary containment air released to the envire s following 2 fu'Al handling accident wit04hin the coentainmient Gufficiently low to accure compliance with 10 CFR 100 guideline&.

(2) These air cleanup units aro a part of the Ro8actor Building purge air cyctom. Sea Section 9.4.6.1 for the design basis for other portionc of this, system.E4-8 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 A A .2 2 Roac.tar Building Purigo System Air Cleanup Units See Sectio 9.4.6.2 for deecription of the system design of the Reactor. Bu:!ding purge system and_ t-he funcRt*on, operation, and control Of the aiA clan nieWiti that cyctomn.Tmo 50%A capacity &i cleanup units, designed to supply a total of 22,919 cfm (two fn together), are proided_ for ~ac~h Reactor Bu"eiding.

Both unite are located wn the Gamne roomR On Elewation 713 adjacent to the Reactor Building they sewre.Eac.h vair- cleanup unit hA.. a- etotain W te hou..n. e d t a t m......samples, teet flitt -Hg W and aceec flaclilitiec forw miaintenAanceG.

The air treatment cOMPOnente within the huciRng inc ude a prefiteI eGctieo, a HERA filter bank, and a carbon; filter bank. This equi ment 1ieV meVtaalled_

In the order 1lIted. Integral to the hIuIing are test fitinge properly sized and proportioned to permit orderly _anRd_ efficientAIR teeting of the HERA filter and carbon adeorber The HEPA filtor .1 mtaled in the air cleanup u arme 1Q00 ft Units deiged to rem-ve at'o.aet 99Q_.9W% of the particulato.

gremater t m ,.ic.rRn -in di*ameter, and mneet the r e.irt of y .pif ication MIL F 51068. The arbon ad-orber6 meRtalled in the ai, cleanup unite are Type I nit trays, fabricated in accordance wvith AACC St-and-ar-d CS 8T equiremeante.

A A.CC CS-8T has been eupmeredd; and, ANSII/RASE N609 089 epecifee ASE AGll I t be ueod Therefore, all now charo T el shall meet AG 1, Section FV , with the exception thait t-he 199A1 vrinof the; cede be ugod Exitin Typ A!cl~onot have to be repl~aced to meet the AG 1 code if being re~fill~d_.

New replacement ch~arcol adeorbent (forF un newll Vlland refilleVd Type II GVlc) shaIl be tmt the1 ASME AG 1191 eu rent in lieuW f the 1088 Ver,.,n (or later provided proper evaluation justifies adequacy), with the eception that labratoy testing of aderben be in accordance with ASTM_03-8032 10-89. The total numbere. of filtorc and -adeorber uItIR trayc provided in each air cleanup unit are licted in Table 6.5 5.Compliance of the doeign, testing, and- maintenance features Of the Rteac-tor-BuIilding purge svsetom aiF Gleinuo units with Regulatory Guide 1.62 OR- tabuwlated in Table A6.523-j ...........

E4-9 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 I 3D10 .o jiwg6UiNorG LWUIOO 1.*,KO~i, 6 6ocuon ppiscuuiuit; For The eoac-tor MicngPurgo

'VOntilatiOR SotOM (Page f2 G~de AppHiGabi Comment GiM.d Applmrsab"W r_ .om-Aent maedo" ToThe ndex SeOtOG1 O TOe l~do G. !a yes NoJte I G.3e yes Note 14 GAb yes -44 yes GA.G yes -C yes Note 14 G.44 yes -.3h yes C.4~e yes --.4 yes Note144 G-34 yes Note 14 G.2a AGNts&4 ~ e Nete-I!G.2b AGNote-4 C44 AG Note144 G.. yes G.4.rn yes G.2. fie Note4 5 G.4; fe Netes-Q--1 G.2e yes G.3. yes G-.24 yes G-" no hnotel12 A14 G--. A Note 6 C GNote4I .4.a fIo Note-12 G-4 yes GA.b AG Note 17 Re.-. AGe- GA. 44 Note144 G.2. yes GAA~ yes C,2 A Note 9 GA~e yes no.a R Notes 3 & 40 G.6. yes Nete 16.3. AGhnotes43&410 G.5b yes Nete 15 G.. yes Note-l 14 f yes Note 45 G3d yes Note-l 14G.4 yes Note 15 G.6.a yesNotes14,%

If C4b yes Netes-14 & 8 Notes 1. The postulated design basis accident (ID"A for the re~actor buAldin pug Retilation

2. [te ic ad fuel handling accident within the PrimFar; Containment.

23. Each air cleanup Unit cOntains a prefilter bank, HEPIA filter- bank, and carbon adseober bank OR the order licted.4. I ro eor Ph duration or theo a' anpuit operation Mlped~ ToiiGWIng Mei PGStuia;oa UWNF--l++-- ,ILL++ ...... ++ ..... i.iaorni~ioa in rioroi mar~oc tn:c reauiromeni unnocaccar:

DOC~U~U in~ proua~iiivj UT ~UCfl d.ctruct. .events to equipmen.

t area.. in oer.ation duing a SAGA.. penoO .time is-w e~demselysmaeW P 6. No mreue suroe of any

'e to" tr-i air c,,ipanu e-uIImnare " nvlclenep£L. ..... I..4..J M~A ;. : 4 A :.. I,.4.I S* 1------1 ~ ** Iti*E4- 10 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 No-vws Continued 69. The cyctGem d860gn provide for temperature and pF8rawcre differential

!ndicatien to allow'forF peioedic Wyreillanco of the filter traine. Also, iniainof fan operation ic ProvideAd-

~n.the main control roomA.:7. eleted 8_ The am~ount of radfioactive maaterial ceolleted by the fiter and adeo~hrbe banks during the postulated DBA identifid in Noete I is not aufficient to create a radiation hazard when the GI. No 6afet enhancement is for-een by the umeI of low leakage du; Owk in this In the eVent Of a poetulate DL -, al ytmhmuwek Gam,4ng ad*9active materia i6 at a pro66UFe beoIw -at~moe6_pheriG.

Concequently, duct leakage in thie part i6 from the"teidel inte the contaminated air trfeam.10. No of thi6 f i utilized in thi; becaUe , nGotUre i6 considered highly unlikely in the postulated DBA.SI The amo, unt f readioative material -ellectled durig the postulated ORA OR too amall ta raiom the ;ad6rbe rhbank temFperature near the Gcarbon ignition temRperature.

HGewoeF, Water epray6 are pFoYidt;d-

n the event of a charcoal fire.12eCmpliance Wth thi e otion ;6 not a iconcinRg requiremnent.

13. The 6yteFm is-6iZed to maintan ac-eptable air purity in the d uring nRormal fuel g GperatineS.

TWO 6Iytom a;ffect the Geii; g 8f the building purg 6etlto ytom. One of these is the fue! handling accident in the cotanment The other :ethe ventilation required to maintain acceptable air purity in the GGA d" ti th d 4WaS found that the V8At*'atwGF; GaPaGity Re A-d-A-d- tw M.a*Ata*R a2 RG-Af-A WGFkoRg eRV*raRM..ARt_

t-*6 gFSateF than that needed temitigaatm the eft.-Gt6of a fuel haRdI*Rg ag-GidARt.

ThSFSfGFS, the 6ysteFA Wa6sizedfQF the ReFmal vent"atien needs. ginGe haRdling Gpwatffiem Gnly take P'aG8 Wh8A the PWFg8 V8At*'atbQF; SysteM *6 in OpeFatmOn, at le-mg-t- 2-0-09A of the puFg-F;g GapaGity Reed-e-d-te nlaaF; up the GentainmeRt atR;GspheFe iA the pestaGGident peFiGdi6GpeFat*Ag 6hGWId aA aGGdentGGGWF.

Availability Or-, t-hopmfem asswed te peftFm the enly eRgineemd safety faafi wa fi indien assigned to tN66yste 4A'a. ~'w. a smut.. .w 5%5 a., %.WU = S ~ , t C ----Saw it ItS -J SS ~I WV- I~E tSR ItS li

• th i---nof thp ANSI d The ryrttmGGAfQFMed tG thir, v leakage teeting icp~re naccordanco with ANSI N509 19:76. W~heneveF poccible-, part6 or Gcomponent6 uced ac replacements comply with the l-atect issue ofAARSI/ASM ftl~lfl Ar peldmna Fa.eirýaW.An~

famr A. cbmfark, gas hia 4 6;ý 15 939 F ew ;.RtR-_.i 1_. :oDI~ancoe withl ANI9AIh'XAM hNb1U ; net required Pinco tflea ;eFeemwas decigno and Ji*---------aaWtUt a ~ ii a *%fl~ ~tStS~U II at. a WW WU ii ViA tashan niannekia aanenfl tha ne.anaAaa..an aaetlinaA in A~LI~ M~4fl 400fl WWLWi I n IW LWWLWW Wh Wbl i the P t" ARPAE WFI QI 1 ()89.I I

• i v I 1 hense or i I.J, I."l t..* a I -,l""l; Ipem f TVA Q ,-dSG,1d, auct WNn are8 Or welgi and a T~DflIedF or ruepaieru aftor inlu :-2, -i87, monet the weldng roguirements of ANSI/PAM A609 10976. The WorftmanehiP camPle6 are not required to ni'n nnnnrrlnT TFIF~T~~(~

~ a nr m~r~-~'ir

~ .. a.~ '...

• I.I " \ " I"* ~.w a I...., t.aer vs.., a ,5 j.r.naa'ae s we a 4 0 I nkn..nbnr.

• Ganiti no fran a any.., ef tha ..~Ar.l lWlll I il nil VVVlIVI I,_l--__ _l.._nl L.._ :__--.... ... : LI. IL. JI. L. eFy 18 I b t.eFa e6 ti f fh d h r. Rh ll;IR&;gr ith the roquirmenic T -~L.e L., 7(ghU6 nOfr orS8F cyGe poatOn;) rer Mope 6, ana RG i11' awh f E4-11 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 9.4.2.3 Safety Evaluation A fuel handling accident in the Auxiliary Building is detected by the two gamma radiation detectors mounted above the fuel pool, as shown in Figure 9.4-12. The high radiation signals via redundant trains will then shut off the fuel handling and Auxiliary Building general supply and exhaust fans and start the ABGTS, as shown in Figures 9.4-9 and 9.4-10. No credit is taken in the dose or accident analyses for these functions.

To a.ccmplish its :f" funt.ction following 3 fu'-" handling acid.nt,, The fuel handling area ventilation system m44st will accomplish the following functions:

(1) Isolate the normal ventilation pathways between the spent fuel pool and the environment.

(2) Filter the contaminants out of the air by the ABGTS before exhausting it to the environment.

The two redundant radiation monitors (non-safety-related) located above the spent fuel pit assure that the accident is promptly detected and that a high radiation signal is provided to each ventilation train, even if one monitor fails. Also, during refueling operations when containment and/or the annulus is open to the Auxiliary Building ABSCE spaces, a Containment Vent Isolation (CVI) signal from either the operating or refueling unit is procedurally configured to assure that a fuel handling accident in containment is promptly detected and the CVI signal is provided to each ventilation train.In addition, the Auxiliary Building radiation monitor (non-safety related) which monitors the Auxiliary Building exhaust vent is also capable of providing a high radiation signal to the MCR. A high radiation signal from either of the monitors located above the spent fuel pit or a operating or refueling unit CVI signal whenever containment and/or the annulus is open to the Auxiliary Building ABSCE spaces during refueling operations causes the fuel handling area (FHA) and Auxiliary Building general supply and exhaust fans to shut down and their associated dampers to close, as shown in Figures 9.4-9 and 9.4-10. Each of the two FHA exhaust fans has both train A and train B dampers, to ensure building isolation in the event of one damper's failure to close.As an added safety feature, all ABSCE boundary isolation dampers are designed to fail-closed on loss of instrument air or electrical power.E4-12 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 9.4.6 Reactor Building Purge Ventilating System (RBPVS)9.4.6.1 Design Bases The RBPVS is designed to maintain the environment in the primary containment and Shield Building annulus within acceptable limits for equipment operation and for personnel access during inspection, testing, maintenance, and refueling operations, and to provide a filtration path for any through-duct outleakage from the primary containment to limit the release of radioactivity to the environment.

The RBPVS performs three distinct functions, the forced air purge function, the continuous pressure relief function, and the alternate containment pressure relief function.The forced air purge function is performed by a purge supply and purge exhaust system consisting of two trains, each of which is designed to provide 50% of the capacity needed for normal purging. Each train consists of a supply fan, an exhaust fan, a HEPA filter-charcoal adsorber assembly, containment isolation valves and associated dampers and ductwork.

This function provides a means by which containment air may be forcibly exchanged and filtered.The purge function provides a means by which containment air may be forcibly exchanged and filtered.

The purge function of the RBPVS is not a safety-related function.

HowoYor, the filtFation units are roguired to provide a safot reated filtration path following a fuel handling acr.ddon until all conta:nment ico~ation valves are clocod. The safety functions are to assure isolation of primary containment during an accident and to isolate the purge air supply intake upon receipt of an Auxiliary Building Isolation (ABI) signal.In the case of a fuel handling accident the filtration units provide a filtration path following a fuel handling accident until all containment isolation valves are closed. However, neither the filtration nor the isolation functions are credited in the Fuel Handling Dose Analysis.

Thus they are not safety functions for this accident.During Operating Modes 1 thru 5, continuous pressure relief is provided by a passive ducting system which passes through containment penetration X-80, through two 100% redundant containment vent air cleanup units (CVACU) containing HEPA filters and charcoal adsorbers.

Containment air is moved into the annulus by the motive force created by differential pressure between the two spaces. Filtration redundancy allows maintenance on one unit at a time while maintaining an open pathway through the other. This ventilation pathway is isolable using containment isolation valves FCV-30-40 and FCV-30-37 which are closed d-"ng Mede 6 r when containment isolation is required.

This system is not required for handling accident mnitigation aRd is not available for that pWeurpo cic ie t ie ocentially i*rlated by containment icol-atio~n valves duNrng fuel leading or handling artivtioc (Mode 6).The alternate pressure relief function is provided by way of a configuration alignment in the forced air purge system. This function is accomplished by opening lower compartment purge lines (one supply and one exhaust) or one of the two pairs of lines (one supply and one exhaust) in the upper compartment.

During purging mode, the purge air fans may or may not be used. To prevent inadvertent pressurization of containment due to supply and exhaust side ductwork flow imbalances, the supply ductwork airflow may be temporarily throttled as needed.The purge function of the RBPVS is not a safety-related function.

H1owever, the filtration unite ar reqird to pFGYide a safety related filtration path following a fuel handling accidet E4-13 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 The design bases include provisions to: (1) Supply fresh air for breathing and contamination control when the primary containment or annulus is occupied.(2) Exhaust primary containment and annulus air to the outdoors whenever the purge air supply system is operated.(3) Clean up containment exhaust during normal operation by routing the air through HEPA-carbon filter trains before release to the atmosphere to limit potential release of radioactivity to the environment.

(4) Provide a reduced quantity of ventilating air to permit occupancy of the instrument room during reactor operation.

The provisions for 1, 2, and 3 above will apply.(5) Assure closure of primary and secondary containment isolation valves following accidents which result in the initiation of a containment ventilation isolation signal.(6) Assure closure of the system air intake dampers, which form part of the ABSCE (see Section 6.2.3.2.1), upon receipt of a signal for Auxiliary Building isolation.

(7) Provide continuous containment pressure relief path through HEPA-carbon filter trains before release to the atmosphere during normal operations.

Items 5 and 6 above are safety-related functions, except in the case of the fuel handling accident.The primary containment penetrations for the ventilation supply and exhaust subsystems are designed to primary containment structural standards.

These are discussed in detail in Section 6.2.4.The RBPVS is sized to maintain an acceptable working environment within the containment during all normal operations.

The system has the capabilities to provide a filtration path for outleakage from the primary containment, and clean up containment atmosphere following a design basis accident.

It also has provisions to filter air flow exhausted from containment for pressure control, during normal operation.

The controls are designed to have simultaneous starting and stopping of the matching supply and exhaust equipment and to initiate an automatic shutdown and isolation upon receipt of the containment ventilation isolation signal.In addition, RBPVS supply fans will shut down and the ABSCE isolation dampers in purge air supply ducts will close on an ABI signal.The RBPVS ";ir c.oanUp equipmpent acu -Fog that "t"'.ty r.loARod ineido ..ntainm.nt from a rnfueling anccient anRd conWAnment isolation, is proGsed through both HE-Pn A finn tr -and c~arbon adeorbere5 before reloase to the atmoephero.

Fuel handling oporatione inridG the primnary cnetainment are conetrained by the operability requirFement for tho RBPA S air cleanup units-contine inthe plant technica!

specifications.

E4-14 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 The RBPVS components are designed or qualified to meet Seismic Category I requirements, except all purge ductwork within the containment, up to the inboard isolation valves, and the supply air ductwork from the downstream flange of the ABSCE isolation dampers to the upstream flange of the Shield Building isolation valves, which are designed to meet Seismic Category I(L) requirements.

The primary containment exhaust is monitored by redundant radiation detectors which provide automatic RBPVS isolation upon detecting the setpoint radioactivity in the exhaust air stream.The RBPVS isolation valves automatically close upon the actuation of a containment ventilation isolation signal or upon manual actuation from the MCR. In addition, during fuel handling operations in the Auxiliary Building with containment and/or the annulus open to the Auxiliary Building ABSCE spaces, the RBPVS isolation valves will close upon a high radiation signal from the spent fuel pool radiation monitors via a CVI signal from the operating or refueling unit.The system air supply and exhaust ducts are routed through the Shield Building annulus to several primary containment penetrations.

Two air supply locations are provided for each of the upper and lower compartments and one for the instrument room. Air is supplied to areas of low potential radioactivity and is allowed to flow to the air pickup exhaust points in areas of higher potential radioactivity.

The air pickup points, located to exhaust air from the lower compartment and instrument room, also provide an air sweep across the surface of the refueling canal...The purge function of the RBPVS is not a safety-related function-:eweve 7 , and the filtration units are not required to provide a safety-related filtration path following a fuel handling accident.The primary containment isolation valves and intermediate piping of the RBPVS are designed in accordance with ANS safety class 2A; other portions are designated ANS safety class 2B except the purge fans, all purge ductwork within the containment, purge supply air ductwork from the ABSCE boundary, fire protection, and drain piping. The instrument room purge subsystem is not an engineered safety feature, and credit for its operability for a LOCA or a fuel-handling accident is not claimed.A containment ventilation isolation signal automatically shuts down the fans and isolates the RBPVS by closing its respective dampers and butterfly valves. Each RBPVS primary containment isolation valve is designed for fail safe closing within 4 seconds of receipt of a closure signal for containment penetrations (See Tables 6.2.4- 1 through 6.2.4-4 and Figure 6.2.4-21).

The RBPVS primary containment isolation valve locations and descriptions are given in Table 6.2.4-1. Each valve is provided with an air cylinder valve operator, control air solenoid valve, and valve position indicating limit switches.Smoke detectors, located in the Auxiliary Building air intake and the general ventilation supply ducts, shut down the purge air supply and the incore instrument room purge supply fans and their isolation dampers.9.4.6.3 Safety Evaluation Functional analyses and failure modes and effects analysis have shown that the RBPVS meets the containment isolation requirements.

The purFg air filtrtirn units -and- 3cci-ted exh,- i-t ducatwor p...de a sa.t* r.lat.d filtration path following a fuel handling The CVACUs, performing a continuously filtered containment vent function during normal operation, are isolated by the closure of their containment isolation valves; therefore are not operable after E4-15 Enclosure 4 WBN Unit 2 Red-line Markup of FSAR Sections 6.2, 6.5, and 9.4 accidents.

In addition, the .entainmont ventilation is not a...wod to be ...d during Mede-6.A functional analysis of the system shows that: (1) During normal operation, adequate fresh air is provided for breathing and for contamination control when the primary or secondary containment (annulus) is occupied.(2) Primary and secondary containment exhaust air is cleaned up during normal operations and following a fuel handling accident.(3) Purge supply and exhaust fan operations cease and isolation dampers in the intake and exhaust ducting close when the system is in the accident isolation mode of operation.

(4) Three signals cause the system to change from the normal purge mode to the accident isolation mode. These signals, which include manual, SIS auto initiate, and high purge exhaust radiation (automatic), initiate a containment ventilation isolation signal. Additionally, during refueling operations whenever containment and/or the annulus is open to the Auxiliary Building ABSCE spaces, a high radiation signal from the spent fuel pool accident radiation monitors or CVI signal from the operating unit automatically cause the system to change from the purge mode to the accident isolation mode.(5) Discharges from the annulus, during normal operation, which are exhausted through the Auxiliary Building exhaust stack, are monitored at the stack. Although these radiation monitors do not initiate an automatic containment isolation signal, radioactive release limits have been established as a basis for controlling plant discharge during operation.

Radioactive releases from the plant resulting from equipment faults of moderate frequency are within 10 CFR 50 Appendix I and 40 CFR 190 limits as specified in the ODCM (See Section 11.3 for further details).

In addition, analyses have shown that any accident with the potential consequence to exceed the 10 CFR 100 limits, would be detected by other indicators (see item 4 above) and cause an automatic primary and/or secondary containment isolation.

Containmont vent sysctem is not te bhe n Mode 6.E4-16 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final E5-1 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 6.2.3 Secondary Containment Functional Design Structures included as part of the secondary containment system are the Shield Building of each reactor unit, the Auxiliary Building, the Condensate Demineralizer Waste Evaporator (CDWE) Building and the essential raw cooling water (ERCW) pipe tunnels adjacent to the Auxiliary Building.

Depending on the configuration of the plant, the Primary Containment Building(s) may also be included as a structure which is part of the secondary containment system. This condition exists when the primary containment is open to the Auxiliary Building.The emergency gas treatment system (EGTS) is provided for ventilation control and cleanup of the atmosphere inside the annulus between the Shield Building and the Primary Containment Building.

The Reactor Building purge air system is also available for cleaning up the atmosphere inside the Shield Building Annulus. Refer to Section 9.4.6 for further details relating to the purge air system. The Auxiliary Building Gas Treatment System (ABGTS) provides a similar contamination control capability in the Auxiliary Building Secondary Containment Enclosure (ABSCE),which includes all of the areas listed above.6.2.3.1 Design Bases 6.2.3.1.1 Secondary Containment Enclosures Design bases for the secondary containment structures were devised to assure that an effective barrier exists for airborne fission products that may leak from the primary containment, or the Auxiliary Building fuel handling area, during a loss-of-coolant accident (LOCA). Within the scope of these design bases are requirements that influence the size, structural integrity, and leak tightness of the secondary containment enclosure.

Specifically, these include a capability to: (a)maintain an effective barrier for gases and vapors that may leak from the primary containment during all normal and abnormal events; (b) delay the release of any gases and vapors that may leak from the primary containment during accidents; (c) allow gases and vapors that may leak through the primary containment during accidents to flow into the contained air volume within the secondary containment where they are diluted, held up, and purified prior to being released to the environs; (d) bleed to the secondary containment each air-filled containment penetration enclosure which extends beyond the Shield Building and which is formed by automatically actuated isolation valves; (e) maintain an effective barrier for airborne radioactive contaminants, gases, and vapors originating in the ABSCE during normal and abnormal events. Refer to Sections 3.8.1 and 3.8.4 for further details relating to the design of the Shield Building and the Auxiliary Building.6.2.3.1.3 Auxiliary Building Gas Treatment System (ABGTS)The design bases for the ABGTS are: 1. To establish and keep an air pressure that is below atmospheric within the portion of the buildings serving as a secondary containment enclosure during accidents.

2. To reduce the concentration of radioactive nuclides in air releases from the secondary containment enclosures to the environs during accidents to levels sufficiently low to keep the site boundary and LPZ dose rates below the 10 CFR 100 guideline values.3. To withstand the safe shutdown earthquake.

4. To provide for initial and periodic testing of the system capability to function as designed (See Chapter 14.0 for information on initial testing of systems).E5-2 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final E5-3 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 6.2.3.3.3 Auxiliary Building Gas Treatment System (ABGTS)The ABGTS has the capabilities needed to preserve safety in accidents as severe as a LOCA. This was determined by conducting functional analyses of the system to verify that the system has the proper features for accident mitigation which consist of a failure modes and effects analysis, a review of Regulatory Guide 1.52 sections to assure licensing requirement conformance, and a performance analysis to verify that the system has the desired accident mitigation capabilities.

A detailed failure modes and effects analysis is presented in Table 6.2.3-3.The functional analyses conducted on the ABGTS have shown that: 1. The air intakes for the system are properly located to minimize accident effects.The use of the air intakes provided in the fuel handling and waste disposal areas minimizes the spread of airborne contamination that may be accidentally released at these positions in which the probability of an accidental release, e.g., a fuel handling accident, is more likely. This localization effect is provided without reducing the effectiveness of the system to cope with multiple activity released throughout the ABSCE that may occur during a LOCA. Such coverage is accomplished by utilizing the normal ventilation ducting to draw outside air inleakage from any point along the secondary containment enclosure to the fuel handling and waste disposal areas.2. Accident indication signals are utilized to bring the ABGTS into operation to assure that the system functions when needed to mitigate accident effects.Accidents in which this system is needed to preserve safety are automatically detected by at least one of the three instrumentation sets used to generate accident signals that result in system startup.3. System startup reliability is very high. The practice of allowing the automatic startup of both full capacity trains in the system gives greater assurance that one train of equipment functions upon receipt of an accident signal.4. The method adopted to establish and keep the negative pressure level within this secondary containment enclosure minimizes the time needed to reach the desired pressure level. Initially, the full capacity of the ABGTS fans is utilized for this purpose. After reaching the desired operating level, the system control module allows outside air to enter the air flow network just upstream of the fan at a rate to keep the fans operating at full capacity with the enclosed volume at the desired negative pressure level. In this situation, the amount of air withdrawn from the enclosed volume is equal to the amount of outside air inleakage through the ABSCE. In addition, two vacuum breaker dampers in series are provided to admit outside air in case the modulating dampers fail.5. The ABSCE is maintained at a slightly negative pressure to reduce the amount of unprocessed air escaping from this secondary containment enclosure to the atmosphere to insignificant quantities.

In addition, this negative pressure level is less than that which is maintained within the annulus; such that, any air leakage E5-4 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final between the Auxiliary Building and the Shield Building is from the Auxiliary Building into the Shield Building.6. The Train A and Train B air cleanup units are sufficiently separated from each other to eliminate the possibility of a single failure destroying the capability to process Auxiliary Building air prior to its release to the atmosphere.

Two concrete walls and a distance of more than 80 feet separate the two trains. The use of separate trains of the emergency power system to drive the air cleanup trains gives further assurance of proper equipment separation.

The review of the ABGTS conducted to determine its conformance with Regulatory Guide 1.52 has shown that this system, designed prior to issuance of the guide, is in general agreement with its requirements.

Details on compliance with Regulatory Guide 1.52 are given in Table 6.5-2.The performance analysis conducted to verify that the ABGTS has the required accident mitigation capabilities has shown that the system flow rate is sized properly to handle all expected outside air inleakage at a 1/4-inch water gauge negative pressure differential.

This indicates that the nominal flow rate of 9000 cfm is sufficient to assure an adequate margin above the expected ABSCE inleakage (ACU filters are replaced as needed to maintain a minimum flow capability of 9300 cfm under surveillance instructions).

The performance analysis evaluated the capability of the ABGTS to reach and maintain a negative pressure of 1/4-inch water gauge with respect to the outside within the boundaries of the ABSCE. The following was utilized in the analysis: 1, Leakage into the ABSCE is proportional to the square root of the pressure differential.

2. Only one air cleanup unit in the ABGTS operates at the rated capacity.3. The air cleanup unit fan begins to operate 30 seconds after the initiation of an ABI signal.4. The initial static pressure inside the ABSCE is conservatively considered to be atmospheric pressure, although the ABSCE is under a negative pressure during normal operation.

5. The effective pressure head due to wind equals 1/8-inch water gauge.6. Initial average air temperature inside the ABSCE equals 140 0 F.7. Atmospheric temperature and pressure are 70°F and 14.4 psia, respectively.

8. ABSCE isolation dampers/valves close within 30 seconds after receiving an ABI signal.E5-5 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 9. The non-safety-related general ventilation and fuel handling area exhaust fans are designed to shut down automatically following a LOCA. Each fan is provided with a safety related Class 1 E primary circuit breaker and a safety related Class 1 E shunt trip isolation switch which is tripped by a signal of the opposite train from that for the primary circuit breaker to ensure that power is isolated from the fan.6.2.4.3 Design Evaluation The containment isolation systems are designed to present a double barrier to any flow path from the inside to the outside of the containment using the double-barrier approach to meet the single-failure criterion.

(0 The design configuration for penetrations X-79A (ice blowing), and X-79B (negative return) is temporarily modified in operating Modes 5 and 6 and when the reactor is defueled (Mode 7) to support ice blowing activities.

The normally closed blind flange on each penetration will be opened and temporary piping will be installed in the penetrations.

A 12-inch silicone seal will be installed between the piping segment and the penetration.

Manual isolation valves will be connected to the piping on the inboard and outboard side of the penetrations.

This configuration is being installed to permit ice blowing operations to occur concurrently with fuel handling activities inside containment.

The penetrations will be returned to their normal design configuration prior to entry into Mode 4 operations.

E5-6 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 Engineered Safety Feature (ESF) Filter Systems Four Engineered Safety Feature (ESF) air cleanup systems' units are provided for fission product removal in post-accident environments.

These are: (1) The emergency gas treatment system (EGTS) air cleanup units.(2) The Auxiliary Building gas treatment system (ABGTS) air cleanup units.(3) The Reactor Building purge system air cleanup units.(4) The Main Control Room emergency air cleanup units.6.5.1.1 Design Bases 6.5.1.1.1 Emergency Gas Treatment System Air Cleanup Units The design bases are: (1) To provide fission product removal capabilities sufficient to keep radioactivity levels in the Shield Building annulus air released to the environs during a DBA LOCA sufficiently low to assure compliance with 10 CFR 100 guidelines.

(2) These air cleanup units are a part of the EGTS. See Section 6.2.3.1.2 for the design bases for other portions of this system.6.5.1.1.2 Auxiliary Building Gas Treatment System Air Cleanup Units The design bases are: (1) To provide fission product removal capabilities sufficient to keep radioactivity levels in the Auxiliary Building secondary containment enclosure (ABSCE) air released to the environs during a postulated accident sufficiently low to assure compliance with 10 CFR 100 guidelines.

(2) These air cleanup units are a part of the ABGTS. See Section 6.2.3.1.3 for the design basis for other portions of this system.E5-7 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 9.4.2.3 Safety Evaluation A fuel handling accident in the Auxiliary Building is detected by the two gamma radiation detectors mounted above the fuel pool, as shown in Figure 9.4-12. The high radiation signals via redundant trains will then shut off the fuel handling and Auxiliary Building general supply and exhaust fans and start the ABGTS, as shown in Figures 9.4-9 and 9.4-10. No credit is taken in the dose or accident analyses for these functions.

The fuel handling area ventilation system will accomplish the following functions:

(1) Isolate the normal ventilation pathways between the spent fuel pool and the environment.

(2) Filter the contaminants out of the air by the ABGTS before exhausting it to the environment.

The two redundant radiation monitors (non-safety-related) located above the spent fuel pit assure that the accident is promptly detected and that a high radiation signal is provided to each ventilation train, even if one monitor fails. Also, during refueling operations when containment and/or the annulus is open to the Auxiliary Building ABSCE spaces, a Containment Vent Isolation (CVI) signal from either the operating or refueling unit is procedurally configured to assure that a fuel handling accident in containment is promptly detected and the CVI signal is provided to each ventilation train.In addition, the Auxiliary Building radiation monitor (non-safety related) which monitors the Auxiliary Building exhaust vent is also capable of providing a high radiation signal to the MCR. A high radiation signal from either of the monitors located above the spent fuel pit or a operating or refueling unit CVI signal whenever containment and/or the annulus is open to the Auxiliary Building ABSCE spaces during refueling operations causes the fuel handling area (FHA) and Auxiliary Building general supply and exhaust fans to shut down and their associated dampers to close, as shown in Figures 9.4-9 and 9.4-10. Each of the two FHA exhaust fans has both train A and train B dampers, to ensure building isolation in the event of one damper's failure to close.As an added safety feature, all ABSCE boundary isolation dampers are designed to fail-closed on loss of instrument air or electrical power.E5-1 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 9.4.6 Reactor Building Purge Ventilating System (RBPVS)9.4.6.1 Design Bases The RBPVS is designed to maintain the environment in the primary containment and Shield Building annulus within acceptable limits for equipment operation and for personnel access during inspection, testing, maintenance, and refueling operations, and to provide a filtration path for any through-duct outleakage from the primary containment to limit the release of radioactivity to the environment.

The RBPVS performs three distinct functions, the forced air purge function, the continuous pressure relief function, and the alternate containment pressure relief function.The forced air purge function is performed by a purge supply and purge exhaust system consisting of two trains, each of which is designed to provide 50% of the capacity needed for normal purging. Each train consists of a supply fan, an exhaust fan, a HEPA filter-charcoal adsorber assembly, containment isolation valves and associated dampers and ductwork.

This function provides a means by which containment air may be forcibly exchanged and filtered.

The purge function provides a means by which containment air may be forcibly exchanged and filtered.

The purge function of the RBPVS is not a safety-related function.

The safety functions are to assure isolation of primary containment during an accident and to isolate the purge air supply intake upon receipt of an Auxiliary Building Isolation (ABI) signal.In the case of a fuel handling accident the filtration units provide a filtration path following a fuel handling accident until all containment isolation valves are closed.However, neither the filtration nor the isolation functions are credited in the Fuel Handling Dose Analysis.

Thus they are not safety functions for this accident.During Operating Modes 1 thru 5, continuous pressure relief is provided by a passive ducting system which passes through containment penetration X-80, through two 100%redundant containment vent air cleanup units (CVACU) containing HEPA filters and charcoal adsorbers.

Containment air is moved into the annulus by the motive force created by differential pressure between the two spaces. Filtration redundancy allows maintenance on one unit at a time while maintaining an open pathway through the other. This ventilation pathway is isolable using containment isolation valves FCV-30-40 and FCV-30-37 which are closed when containment isolation is required.The alternate pressure relief function is provided by way of a configuration alignment in the forced air purge system. This function is accomplished by opening lower compartment purge lines (one supply and one exhaust) or one of the two pairs of lines (one supply and one exhaust) in the upper compartment.

During purging mode, the purge air fans may or may not be used. To prevent inadvertent pressurization of containment due to supply and exhaust side ductwork flow imbalances, the supply ductwork airflow may be temporarily throttled as needed.The purge function of the RBPVS is not a safety-related function.E5-2 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final The design bases include provisions to: (1) Supply fresh air for breathing and contamination control when the primary containment or annulus is occupied.(2) Exhaust primary containment and annulus air to the outdoors whenever the purge air supply system is operated.(3) Clean up containment exhaust during normal operation by routing the air through HEPA-carbon filter trains before release to the atmosphere to limit potential release of radioactivity to the environment.

(4) Provide a reduced quantity of ventilating air to permit occupancy of the instrument room during reactor operation.

The provisions for 1, 2, and 3 above will apply.(5) Assure closure of primary and secondary containment isolation valves following accidents which result in the initiation of a containment ventilation isolation signal.(6) Assure closure of the system air intake dampers, which form part of the ABSCE (see Section 6.2.3.2.1), upon receipt of a signal for Auxiliary Building isolation.

(7) Provide continuous containment pressure relief path through HEPA-carbon filter trains before release to the atmosphere during normal operations.

Items 5 and 6 above are safety-related functions, except in the case of the fuel handling accident.The primary containment penetrations for the ventilation supply and exhaust subsystems are designed to primary containment structural standards.

These are discussed in detail in Section 6.2.4.The RBPVS is sized to maintain an acceptable working environment within the containment during all normal operations.

The system has the capabilities to provide a filtration path for outleakage from the primary containment, and clean up containment atmosphere following a design basis accident.

It also has provisions to filter air flow exhausted from containment for pressure control, during normal operation.

The controls are designed to have simultaneous starting and stopping of the matching supply and exhaust equipment and to initiate an automatic shutdown and isolation upon receipt of the containment ventilation isolation signal.In addition, RBPVS supply fans will shut down and the ABSCE isolation dampers in purge air supply ducts will close on an ABI signal.The RBPVS components are designed or qualified to meet Seismic Category I requirements, except all purge ductwork within the containment, up to the inboard isolation valves, and the supply air ductwork from the downstream flange of the ABSCE E5-3 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final isolation dampers to the upstream flange of the Shield Building isolation valves, which are designed to meet Seismic Category I(L) requirements.

The primary containment exhaust is monitored by redundant radiation detectors which provide automatic RBPVS isolation upon detecting the setpoint radioactivity in the exhaust air stream. The RBPVS isolation valves automatically close upon the actuation of a containment ventilation isolation signal or upon manual actuation from the MCR. In addition, during fuel handling operations in the Auxiliary Building with containment and/or the annulus open to the Auxiliary Building ABSCE spaces, the RBPVS isolation valves will close upon a high radiation signal from the spent fuel pool radiation monitors via a CVI signal from the operating or refueling unit.The system air supply and exhaust ducts are routed through the Shield Building annulus to several primary containment penetrations.

Two air supply locations are provided for each of the upper and lower compartments and one for the instrument room. Air is supplied to areas of low potential radioactivity and is allowed to flow to the air pickup exhaust points in areas of higher potential radioactivity.

The air pickup points, located to exhaust air from the lower compartment and instrument room, also provide an air sweep across the surface of the refueling canal...The purge function of the RBPVS is not a safety-related function and the filtration units are not required to provide a safety-related filtration path following a fuel handling accident.

The primary containment isolation valves and intermediate piping of the RBPVS are designed in accordance with ANS safety class 2A; other portions are designated ANS safety class 2B except the purge fans, all purge ductwork within the containment, purge supply air ductwork from the ABSCE boundary, fire protection, and drain piping. The instrument room purge subsystem is not an engineered safety feature, and credit for its operability for a LOCA or a fuel-handling accident is not claimed.A containment ventilation isolation signal automatically shuts down the fans and isolates the RBPVS by closing its respective dampers and butterfly valves. Each RBPVS primary containment isolation valve is designed for fail safe closing within 4 seconds of receipt of a closure signal for containment penetrations (See Tables 6.2.4- 1 through 6.2.4-4 and Figure 6.2.4-21).

The RBPVS primary containment isolation valve locations and descriptions are given in Table 6.2.4-1. Each valve is provided with an air cylinder valve operator, control air solenoid valve, and valve position indicating limit switches.Smoke detectors, located in the Auxiliary Building air intake and the general ventilation supply ducts, shut down the purge air supply and the incore instrument room purge supply fans and their isolation dampers.E5-4 Enclosure 5 WBN Unit 2 -FSAR Sections 6.2, 6.5, and 9.4 Final 9.4.6.3 Safety Evaluation Functional analyses and failure modes and effects analysis have shown that the RBPVS meets the containment isolation requirements.

The CVACUs, performing a continuously filtered containment vent function during normal operation, are isolated by the closure of their containment isolation valves; therefore are not operable after accidents.

A functional analysis of the system shows that: (1) During normal operation, adequate fresh air is provided for breathing and for contamination control when the primary or secondary containment (annulus) is occupied.(2) Primary and secondary containment exhaust air is cleaned up during normal operations and following a fuel handling accident.(3) Purge supply and exhaust fan operations cease and isolation dampers in the intake and exhaust ducting close when the system is in the accident isolation mode of operation.

(4) Three signals cause the system to change from the normal purge mode to the accident isolation mode. These signals, which include manual, SIS auto initiate, and high purge exhaust radiation (automatic), initiate a containment ventilation isolation signal. Additionally, during refueling operations whenever containment and/or the annulus is open to the Auxiliary Building ABSCE spaces, a high radiation signal from the spent fuel pool accident radiation monitors or CVI signal from the operating unit automatically cause the system to change from the purge mode to the accident isolation mode.(5) Discharges from the annulus, during normal operation, which are exhausted through the Auxiliary Building exhaust stack, are monitored at the stack. Although these radiation monitors do not initiate an automatic containment isolation signal, radioactive release limits have been established as a basis for controlling plant discharge during operation.

Radioactive releases from the plant resulting from equipment faults of moderate frequency are within 10 CFR 50 Appendix I and 40 CFR 190 limits as specified in the ODCM (See Section 11.3 for further details).

In addition, analyses have shown that any accident with the potential consequence to exceed the 10 CFR 100 limits, would be detected by other indicators (see item 4 above) and cause an automatic primary and/or secondary containment isolation.

E5-5 Enclosure 6 WBN Unit 2 -Revised Technical Specification Red-Line Markup E6-1 Containment Vent Isolation Instrumentation 3.3.6 3.3 INSTRUMENTATION

3.3.6 Containment

Vent Isolation Instrumentation LCO 3.3.6 APPLICABILITY:

The Containment Vent Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.MODES 1, 2, 3, and 4, Durin- of i.,diated f'u-Al apamb_.m; .thin continm Ant.Q v.ACTIONS---------------------

NOTE-Separate Condition entry is allowed for each Function.CONDITION REQUIRED ACTION COMPLETION TIME A. One radiation monitoring A.1 Restore the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> channel inoperable, channel to OPERABLE status.(continued)

Watts Bar -Unit 2 (developmental) 3.3-53 AHI Containment Vent Isolation Instrumentation

3.3.6 ACTIONS

(continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. NOTE --NOTE ----------

Only..,..."';""!b"i MODE 1, One train of automatic actuation 24 -,logic may be bypassed and Required Action B.1 may be delayed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for One or more Functions with Surveillance testing provided the one or more manual or other train is OPERABLE.automatic actuation trains inoperable.

B. 1 Enter applicable Immediately OR Conditions and Required Actions of LCO 3.6.3, Two radiation monitoring "Containment Isolation channels inoperable.

Valves," for containment OR purge and exhaust isolation valves made Required Action and inoperable by isolation associated Completion instrumentation.

Time of Condition A not met.(continued)

Watts Bar -Unit 2 (developmental) 3.3-54 AH I Containment Vent Isolation Instrumentation

3.3.6 ACTIONS

(continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. NOTE G4 Place and mnanFtain4Fe;at~

Only applicablo durn containmont purge and rnevement of iFfadiattedd fle, e~haust valves 9in clocod assemblies Mitin PG6Iitin TG~ffiO .OR One onr rnrmore Functiene With C" 2 mtF pplirable 4F~ed4tl one Or more manuwa! or Cond-itions and Required automatic ctato trains Actfione o-f LCOG 3.9.4, Penotrationcs, fo OR containment purge and TPWo rdainmonitoring mnade inoperable byf ch-1Annel inoperable.ioaio ntumnain OR Required Action and aocatod Completion Time for Cond-ition A noet Watts Bar -Unit 2 (developmental) 3.3-55 AH I Containment Vent Isolation Instrumentation

3.3.6 SURVEILLANCE

REQUIREMENTS

---

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Vent Isolation Function.SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> S R 3.3.6.2 ----------------------

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

This surveillance is only applicable to the actuation logic of the ESFAS instrumentation.

Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.3 ----------------------

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

This surveillance is only applicable to the master relays of the ESFAS instrumentation.

Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.4 Perform COT. 92 days SR 3.3.6.5 Perform SLAVE RELAY TEST. 92 days OR 18 months for Westinghouse type AR and Potter &Brumfield MDR Series relays (continued)

Watts Bar -Unit 2 (developmental) 3.3-56 B Containment Vent Isolation Instrumentation

3.3.6 SURVEILLANCE

REQUIREMENTS (Continued)

SURVEILLANCE FREQUENCY SR 3.3.6.6 ----------------------

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

Verification of setpoint is not required.Perform TADOT. 18 months SR 3.3.6.7 Perform CHANNEL CALIBRATION.

18 months Watts Bar -Unit 2 (developmental) 3.3-57 A Containment Vent Isolation Instrumentation

3.3.6 Table

3.3.6-1 (page 1 of 1)Containment Vent Isolation Instrumentation REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CHANNELS REQUIREMENTS VALUE 1. Manual Initiation 2 SR 3.3.6.6 NA 2. Automatic Actuation Logic 2 trains SR 3.3.6.2 NA and Actuation Relays SR 3.3.6.3 SR 3.3.6.5 3. Containment Purge 2 SR 3.3.6.1 -8 .,E 02 , Exhaust Radiation Monitors SR 3.3.6.4 SR 3.3.6.7< 2.8E-02 pCi/ccN (1.14x10 4 cpm)4. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

J k ?F P I 1.3) W~uFrig rnovomen; wT irraeafi-oa TurnA- -266-8-M-911 W:Rnn conainMon.

A A A A I() NMAW f 1 " 2 ,R, --l Watts Bar -Unit 2 (developmental) 3.3-58 SH I ABGTS Actuation Instrumentation 3.3.8 3.3 INSTRUMENTATION

3.3.8 Auxiliary

Building Gas Treatment System (ABGTS) Actuation Instrumentation LCO 3.3.8 APPLICABILITY:

The ABGTS actuation instrumentation for each Function in Table 3.3.8-1 shall be OPERABLE.According to Table 3.3.8-1.ACTIONS--------------------

NOT Separate Condition entry is allowed for each Ft r --------------------------------------------------------------

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A.1 Place one ABGTS train in 7 days one channel or train operation.

inoperable.

B. One or more Functions with B. 1.1 Place one ABGTS train in Immediately two channels or two trains operation.

inoperable.

AND B. 1.2 Enter applicable Immediately Conditions and Required Actions of LCO 3.7.12,"Auxiliary Building Gas Treatment System (ABGTS)," for one train made inoperable by inoperable actuation instrumentation OR (continued)

Watts Bar -Unit 2 (developmental) 3.3-63 AH ABGTS Actuation Instrumentation

3.3.8 ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)

B.2 Place both trains in Immediately emergency radiation protection mode.C. Required Actien and G4 SuBpeind MeOD em3. t 6eho associated Completion iaccemblie Time for Condition A or B in the fuel handling area.not met durig movement of iradiated fuel acemnbliec i the fu-el han~dling area.0- C. Required Action and Co. 1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A or B AND not met in MOIDE 1, 2, 3, CD.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS

---

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

Refer to Table 3.3.8-1 to determine which SRs apply for each ABGTS Actuation Function.SURVEILLANCE FREQUENCY RR,32 Perform CHANNEL CHECK.11-4& ...... 92-days (continued)

Watts Bar -Unit 2 (developmental) 3.3-64 AH I ABGTS Actuation Instrumentation

3.3.8 SURVEILLANCE

REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.8.3-1

---

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

Verification of setpoint is not required.Perform TADOT. 18 months SRP3.3..4 Pefm, CHA.NNEL CALIB.RATION.

48-.mr,,lth Watts Bar -Unit 2 (developmental) 3.3-65 AH ABGTS Actuation Instrumentation

3.3.8 Table

3.3.8-1 (page 1 of 1)ABGTS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE 1. Manual Initiation 1,2,3,4 2 SR 3.3.8.31 NA (a) 2 SR 3.3.8.3 NA 2. FUe! Pooa Armp (a) 2 RR-4441 4-1191 mRthr Radiation Meniter-s SR 3.3.82 2. 3-Containment Refer to LCO 3.3.2, Function 3.a., for all Phase A initiating functions Isolation and requirements.

---

 :---- :-- AL-- l..--u L =.I!L.I ____ : ......&24 W1ffl 'ARQ on 0M 9[T101P1 Ofo !Wag!io WAG fIflOA 1 A A -6 TUO fal R R14Itfl aroa.Watts Bar -Unit 2 (developmental) 3.3-66 AH ABGTS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)LCO 3.7.12 APPLICABILITY:

Two ABGTS trains shall be OPERABLE MODES 1, 2, 3, and 4, Durinen of irradiated-fuel accomem1 l;ies in the handline area ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ABGTS train A. 1 Restore ABGTS train to 7 days inoperable OPERABLE status.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not met AND MODE rl1, 2, 3, or 1 B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Two ABGTS trains inoperable in MOGDE 1, 2, 3, 4-4.G. Required Action and4 Place OPERABLE medatl accated Completion ABGTS6 train in operationr.

Time of Condition A not mnet.during mo9Vement Of OR iradiated fuel assemblies in the fu-el handling aresa. l Susepod moevement ot immediatl irradiated fuel assembliec in the fue! handl"ng area D-- Tw0BTStan Suseond movyement ot Immediately iAGS~ale U~ig iRadiated fue accomblioc move~ment o-f irradiated fuel in the fuel handling area.-assemblies in the fuel handling al ea,,l;;x Watts Bar -Unit 2 (developmental) 3.7-26 A H ABGTS 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABGTS train for > 10 continuous hours 31 days with the heaters operating.

SR 3.7.12.2 Perform required ABGTS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.3 Verify each ABGTS train actuates on an actual or 18 months simulated actuation signal.SR 3.7.12.4 Verify one ABGTS train can maintain a pressure 18 months on a between -0.25 inches and -0.5 inches water gauge STAGGERED TEST with respect to atmospheric pressure during the post BASIS accident mode of operation at a flow rate > 9300 cfm and < 9900 cfm.Watts Bar -Unit 2 (developmental) 3.7-27 AH I VV I f3n FlII Ii-oI IRra;tRIo 34." 3.0 REFUELING OPERATIONS

3.0.4 Containment

Penetations 1- C Q 3. -9. 4 i II I IO.n ...........

t nn nr.Mnn 701' Z;71tntliwn

~r a. iR no lP~e; uguWmun n lesen ao~nd neld in plaGS By a mAInImumR w TOWr beftl-t, ;'tm bh. Onea dooar. fin eac-rh air. locak caloco-d;i Or capable Of baing cloced PFOY.'ied.A. R-T.S- Ic O-GPESR-ABEI-in acco-ardance-f Awith TS 317.42, and G. Emach penetration provid4ng diroctiaccecc66 fromF the containmon atmoephoro to the outcide atFnocphSro either 1. c-loced by a mau!or automatic-isolation valve, blind flange, or 2. capableof being clod by an OPE-RABLE Containment

'ent IelaR-;System~

NOTE Penetration floW path(c) providing

&dirc accec frmM the A cnainment atm~osphere-to the otieatmocsphser m~ay be unicol-'-tated unFder a2dM6RinitrAtiVe controle provided .A.BGTS i OPERABLE in acodn ewth T R lelvl sw v l ~ l .l ~ V V* i , b ~ b 1 V V V I W W i, uur:na movement OT irraaiarea TUOI accemeiiec wirnin containment.

ACTION'S QN44NREmQUIRED[

ACTION COMh-PLET!ON TIME C. One or mRoFre ntainment A4 Suspend movement et .lmmediately pwentratin Anet in required iradiated fuel aiccemblioG status- Within cOntainmenSFt.

MWRtt BAr Unit 2 (dvlomntl AH t6ontainment w-ene;rat~on 3-"A SURVEI.1'E!lr-rCE RE-QUIREMENTS

________SRURVEILLA.NGCE SR .9.4.4 Verify each Fqured Gntainmont pene-tration is in 7-,days the required status.SR- R2. .4.2 Verify each required cOntRainmnt Vent'icoARtion aIV, I4R-MARth actuates to the icltn oiio na actua! ot simulated actuation signial.Waftv BSa Unit 2 (developmental)

AH

,,#d "P" ,,ahin" S, J i f i.l ..I iRl'r to w -'s 3.9 REFUELING OPERATIONS 33.90.8O R 6-actOr Building Pume Air Cleanuo Unite LGO 3.9 Two Reator Bu~i~na Pure Air Cleanup Units shall be OPERABLEr APPLICABILITY:

Dun.... me.eme.t of irradiated fuel assembies within the containment.

CONDITION REQU-IlRD ACTIONAl COIMPLETION TIME A.. One Re-actr Bu-di Purge A4 lotethe nprbeai seit~Air Cleanup Unit ineperabounit.a AND A-2 Verify the OPERABLE air leIatl GpS~atleR;.

B. Two ,;uld*g Purge 8_4 Su.ep..d mvem...en.

Immediately AiF Clean;up Unite irradiated fuel aIembi2il perable~Within containrent.

URVE!ILLA~CS RRVQUREMENTL, S 1U-IREILLANCE FRQ,\N 11 ------ PerGFoM requiere-d MilEr tetin6tFg in accor-danceQf With te I codnewt Ventilation Filter T-ecting ProgramR eIFTP). the VF.p Watts Bar Unit 2e pm ntal)AH Decay Time 3.9.8 3.9 REFUELING OPERATIONS 3.9.10 Decay Time LCO 3.9.10 APPLICABILITY:

The reactor shall be subcritical for >100 hours.During movement of irradiated fuel assemblies within the containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for A. 1 Suspend all operations Immediately

< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. involving movement of irradiated fuel assemblies within the containment.

TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.10.1 Verify the reactor has been subcritical for > 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Prior to movement of by confirming the date and time of subcriticality.

irradiated fuel in the reactor vessel Watts Bar -Unit 2 (developmental) 3.9-14 H Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP)A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2; ASME N510-1989, and the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR.a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass within acceptance criterion when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION ACCEPTANCE SYSTEM CRITERIA FLOW RATE,,ui,',,, ..... e. 14.00% 14 , ., , , m 1. 4n0o_Emergency Gas < 0.05% 4,000 cfm + 10%Treatment Auxiliary Building Gas < 0.05% 9,000 cfm + 10%Treatment Control Room Emergency

< 1.00% 4,000 cfm + 10%(continued)

Watts Bar-Unit 2 (developmental) 5.0-18 8HI Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass within acceptance criterion when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION ACCEPTANCE SYSTEM CRITERIA FLOW RATE Rea~OF Bildig Puqe 4.90%14,900 dm +10%Emergency Gas Treatment

< 0.05% 4,000 cfm + 10%Auxiliary Building Gas < 0.05% 9,000 cfm + 10%Treatment Control Room Emergency

< 1.00% 4,000 cfm + 10%I (continued)

Watts Bar-Unit 2 (developmental) 5.0-19 Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, and the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of _< 30 0 C and greater than or equal to the relative humidity specified below.METHYL IODIDE RELATIVE ESF VENTILATION SYSTEM PENETRATION HUMIDITY Roictor Building Purge 4 40% 0"0 Emergency Gas Treatment

< 0.175% 70%Auxiliary Building Gas < 0.175% 70%Treatment Control Room Emergency

< 1.0% 70%d. Demonstrate for each of the ESF systems that the pressure drop across the entire filtration unit is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION SYSTEM PRESSURE DROP FLOW RATE,1Buildig

.4 1.7 .tOr 14,000 + 40%Emergency Gas < 7.6 inches water 4,000 cfm + 10%Treatment Auxiliary Building Gas < 7.6 inches water 9,000 cfm + 10%Treatment Control Room Emergency

< 3.5 inches water 4,000 cfm + 10%(continued)

Watts Bar-Unit 2 (developmental) 5.0-20 B8HI Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ASME N510-1989.

ESF VENTILATION SYSTEM AMOUNT OF HEAT Emergency Gas Treatment 20 + 2.0 kW Auxiliary Building Gas Treatment 50 + 5.0 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.7.2.15 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5,"Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures." The program shall include: a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., the system is not designed to withstand a hydrogen explosion);(continued)

Watts Bar-Unit 2 5.0-14 (developmental)

BH Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.19 Containment Leakage Rate Testing Program (continued)

Leakage rate acceptance criteria are: a. Containment overall leakage rate acceptance criterion is _ 1.0 La.During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and _! 0.75 L, for Type A tests.b. Air lock testing acceptance criteria are: 1. Overall air lock leakage rate is 0.05 La when tested at > Pa.2. For each door, leakage rate is _ 0.01 La when pressurized to_> 6 psig.The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.5.7.2.20 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of the applicable regulatory requirement (i.e., 5 rem Total Effective Dose Equivalent (TEDE) for a fuel handling accident or 5 rem whole body or its equivalent to any part of the body) for the duration of the accident.

The program shall include the following elements: a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

Watts Bar-Unit 2 5.0-25 (developmental)

AH Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals Watts Bar-Unit 2 (developmental) 5.0-26 AH Enclosure 7 WBN Unit 2 -Revised Technical Specification Final E7-1 Containment Vent Isolation Instrumentation 3.3.6 3.3 INSTRUMENTATION

3.3.6 Containment

Vent Isolation Instrumentation LCO 3.3.6 APPLICABILITY:

The Containment Vent Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.MODES 1, 2, 3, and 4, ACTIONS-------------------

NOTE --------Separate Condition entry is allowed for each Function.CONDITION REQUIRED ACTION COMPLETION TIME A. One radiation monitoring A.1 Restore the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> channel inoperable, channel to OPERABLE status.(continued)

Watts Bar -Unit 2 (developmental) 3.3-53 H Containment Vent Isolation Instrumentation

3.3.6 ACTIONS

(continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE ---------------------

NOTE ----------

One train of automatic actuation One or more Functions with logic may be bypassed and one or more manual or Required Action B. 1 may be automatic actuation trains delayed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for inoperable.

Surveillance testing provided the OR other train is OPERABLE.Two radiation monitoring B. 1 Enter applicable Immediately channels inoperable.

Conditions and Required OR Actions of LCO 3.6.3,"Containment Isolation Required Action and Valves," for containment associated Completion purge and exhaust Time of Condition A not isolation valves made met. inoperable by isolation instrumentation.(continued)

Watts Bar -Unit 2 (developmental) 3.3-54 H Containment Vent Isolation Instrumentation

3.3.6 SURVEILLANCE

REQUIREMENTS

---

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Vent Isolation Function.SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.6.2 ----------------------

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

This surveillance is only applicable to the actuation logic of the ESFAS instrumentation.

Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS SIR 3.3.6.3 ----------------------

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

This surveillance is only applicable to the master relays of the ESFAS instrumentation.

Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.4 Perform COT. 92 days SR 3.3.6.5 Perform SLAVE RELAY TEST. 92 days OR 18 months for Westinghouse type AR and Potter &Brumfield MDR Series relays (continued)

Watts Bar -Unit 2 (developmental) 3.3-55 B Containment Vent Isolation Instrumentation

3.3.6 SURVEILLANCE

REQUIREMENTS (Continued)

SURVEILLANCE FREQUENCY SR 3.3.6.6 ----------------------

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

Verification of setpoint is not required.Perform TADOT. 18 months SR 3.3.6.7 Perform CHANNEL CALIBRATION.

18 months Watts Bar -Unit 2 (developmental) 3.3-56 A Containment Vent Isolation Instrumentation

3.3.6 Table

3.3.6-1 (page 1 of 1)Containment Vent Isolation Instrumentation REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CHANNELS REQUIREMENTS VALUE 1. Manual Initiation 2 SR 3.3.6.6 NA 2. Automatic Actuation Logic 2 trains SR 3.3.6.2 NA and Actuation Relays SR 3.3.6.3 SR 3.3.6.5< 2.8E-02 pCi/cc 3. Containment Purge 2 SR 3.3.6.1 (1.14x10 4 cpm)Exhaust Radiation Monitors SR 3.3.6.4 SR 3.3.6.7 4. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

Watts Bar -Unit 2 (developmental) 3.3-57 H I ABGTS Actuation Instrumentation 3.3.8 3.3 INSTRUMENTATION

3.3.8 Auxiliary

Building Gas Treatment System (ABGTS) Actuation Instrumentation LCO 3.3.8 APPLICABILITY:

The ABGTS actuation instrumentation for each Function in Table 3.3.8-1 shall be OPERABLE.According to Table 3.3.8-1.ACTIONS--------------------

NOTE-Separate Condition entry is allowed for each Function.CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A. 1 Place one ABGTS train in 7 days one channel or train operation.

inoperable.

B. One or more Functions with B.1.1 Place one ABGTS train in Immediately two channels or two trains operation.

inoperable.

AND B.1.2 Enter applicable Immediately Conditions and Required Actions of LCO 3.7.12,"Auxiliary Building Gas Treatment System (ABGTS)," for one train made inoperable by inoperable actuation instrumentation OR (continued)

Watts Bar -Unit 2 (developmental) 3.3-63 H ABGTS Actuation Instrumentation

3.3.8 ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)

B.2 Place both trains in Immediately emergency radiation protection mode.C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A or B AND not met-C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS

---

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

Refer to Table 3.3.8-1 to determine which SRs apply for each ABGTS Actuation Function.SURVEILLANCE FREQUENCY SR 3.3.8.1 ----------------------

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

Verification of setpoint is not required.Perform TADOT. 18 months Watts Bar -Unit 2 (developmental) 3.3-64 H ABGTS Actuation Instrumentation

3.3.8 Table

3.3.8-1 (page 1 of 1)ABGTS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE 1. Manual Initiation 1,2,3,4 2 SR 3.3.8.3 NA 2. Containment Refer to LCO 3.3.2, Function 3.a., for all Phase A initiating functions Isolation and requirements.

Watts Bar -Unit 2 (developmental) 3.3-65 H ABGTS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)LCO 3.7.12 APPLICABILITY:

Two ABGTS trains shall be OPERABLE MODES 1, 2, 3, and 4, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ABGTS train A.1 Restore ABGTS train to 7 days inoperable OPERABLE status.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not met AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Two ABGTS trains inoperable Watts Bar -Unit 2 (developmental) 3.7-26 H ABGTS 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABGTS train for _> 10 continuous hours 31 days with the heaters operating.

SR 3.7.12.2 Perform required ABGTS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.3 Verify each ABGTS train actuates on an actual or 18 months simulated actuation signal.SR 3.7.12.4 Verify one ABGTS train can maintain a pressure 18 months on a between -0.25 inches and -0.5 inches water gauge STAGGERED TEST with respect to atmospheric pressure during the post BASIS accident mode of operation at a flow rate _> 9300 cfm and < 9900 cfm.Watts Bar -Unit 2 (developmental) 3.7-27 H Decay Time 3.9.8 3.9 REFUELING OPERATIONS 3.9.10 Decay Time LCO 3.9.10 APPLICABILITY:

The reactor shall be subcritical for >_100 hours.During movement of irradiated fuel assemblies within the containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for A.1 Suspend all operations Immediately

< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. involving movement of irradiated fuel assemblies within the containment.

TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.10.1 Verify the reactor has been subcritical for > 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Prior to movement of by confirming the date and time of subcriticality.

irradiated fuel in the reactor vessel Watts Bar -Unit 2 (developmental) 3.9-14 H I Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP)A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2; ASME N510-1989, and the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR.f. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass within acceptance criterion when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION ACCEPTANCE SYSTEM CRITERIA FLOW RATE Emergency Gas < 0.05% 4,000 cfm + 10%Treatment Auxiliary Building Gas < 0.05% 9,000 cfm + 10%Treatment Control Room Emergency

< 1.00% 4,000 cfm + 10%I (continued)

Watts Bar-Unit 2 (developmental) 5.0-18 HI Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

g. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass within acceptance criterion when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION ACCEPTANCE SYSTEM CRITERIA FLOW RATE Emergency Gas Treatment

< 0.05% 4,000 cfm + 10%Auxiliary Building Gas < 0.05% 9,000 cfm + 10%Treatment Control Room Emergency

< 1.00% 4,000 cfm + 10%(continued)

Watts Bar-Unit 2 (developmental) 5.0-19 HI Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

h. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, and the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of _< 30 0 C and greater than or equal to the relative humidity specified below.METHYL IODIDE RELATIVE ESF VENTILATION SYSTEM PENETRATION HUMIDITY Emergency Gas Treatment

< 0.175% 70%Auxiliary Building Gas < 0.175% 70%Treatment Control Room Emergency

< 1.0% 70%i. Demonstrate for each of the ESF systems that the pressure drop across the entire filtration unit is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, the exceptions noted for each ESF system in Tables 6.5-1, 6.5-2, 6.5-3, and 6.5-4 of the FSAR, and ASME N510-1989 at the system flowrate specified below.ESF VENTILATION SYSTEM PRESSURE DROP FLOW RATE Emergency Gas < 7.6 inches water 4,000 cfm + 10%Treatment Auxiliary Building Gas < 7.6 inches water 9,000 cfm + 10%Treatment Control Room Emergency

< 3.5 inches water 4,000 cfm + 10%(continued)

Watts Bar-Unit 2 (developmental) 5.0-20 HI Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.14 Ventilation Filter Testing Program (VFTP) (continued)

Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ASME N510-1989.

ESF VENTILATION SYSTEM AMOUNT OF HEAT Emergency Gas Treatment 20 + 2.0 kW Auxiliary Building Gas Treatment 50 + 5.0 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.7.2.15 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5,"Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures." The program shall include: b. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., the system is not designed to withstand a hydrogen explosion);(continued)

Watts Bar-Unit 2 (developmental) 5.0-14 HI Procedures, Programs, and Manuals 5.7 5.7.2.19 Containment Leakage Rate Testing Program (continued)

Leakage rate acceptance criteria are: c. Containment overall leakage rate acceptance criterion is < 1.0 L,.During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the combined Type B and Type C tests, and < 0.75 La for Type A tests.d. Air lock testing acceptance criteria are: 1. Overall air lock leakage rate is < 0.05 L, when tested at > Pa.2. For each door, leakage rate is < 0.01 La when pressurized to_> 6 psig.The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.5.7.2.20 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of the applicable regulatory requirement (i.e., 5 rem Total Effective Dose Equivalent (TEDE) for a fuel handling accident or 5 rem whole body or its equivalent to any part of the body) for the duration of the accident.

The program shall include the following elements: c. The definition of the CRE and the CRE boundary.d. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

Watts Bar -Unit 2 5.0-25 (developmental)

H Enclosure 8 WBN Unit 2 -Revised Technical Specification Bases Red-Line Markup E8-1 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 reguireS that wheR nemvng irradiated fuel in the A u-liar' Building and!or CnimetWith the Conptainment oene to the A xw~~yBilig BCEsars a si~l#R the sp~ poolDG ral;aon mnitors l"0 RE 90 102 and 103 will initiate a Crntainm8nt

%1 I4;1 4; 1~y Ii I V l' r%%1 VV AAI -V1 4-k I v l ll ll t I, ,llnVU I a csignal from the contafinment purge radi.ation mon;ý.6iiitom 2 REm 90 130, and-131 orI other CVI signal Will *initiat that po~tgnn Of the Auvili~arI Building Isolation (AWl noalyiitiated by the apn'fe po radiation MOnitoVA Cn~t~FIMS I~t~ l~lfef hs A (1sga)frm m s .... UIaII, t3, '...'JI ...............

(........

...... ..... .. ...p. ati.g Unit, high temperatue

+ in the Building air intakes, o manual AB!1 wil! Gauss a CVI signal inthe refueling unit. in the caco Where the containm~ent of both uniRtg- i 6 opent the Au;i~iar; Building cpacoc, a CVI iFn One unit Will uiniiateq a.1 ORi the ether unit in order to maRintain t-hoce spaces open to the ABSCE.r-R T-herefore, the containment Yentilatiwn inetrumentation imueAt remý.ain operable when moving irdae ful--- in the -AwIary Bui~dng if the Gentaiwnmet air locks, penletatienc, equipment hatch, etc. are open to the Auxdliary Building A1BSCE spaces.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-150 GHI Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

APPLICABLE SAFETY ANALYSES The containment isolation valves for the Reactor Building Purge System close within six seconds following the DBA. The containment vent isolation radiation monitors act as backup to the SI signal to ensure closing of the purge air system supply and exhaust valves. They-aFe-alee the prfmary. means for automatically

..lating , ,ntainmont in the event of a fu,-l handling accident du-Rng 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.

WAhenR moving iFradiated fu'el incdocotainm~ent or in the Awxla, Building with con-t-ainmen-t air lockeF orpenetratione open to the Auxiliar, Building ABSCE spaces, Or When mo8Ving fuel in the Auxilia~y Building With the containment equipm~ent hatch open, the proVicieneG to iniOtiAt a the aortion of an 131 oRes !l -* -' b'.. the sieent fuel iaoel radiat~on m i A; r"11i 4 A k 4k 4. 4 r~w" VM7 "'a ........a a7...---.'.,".,..-.

purge mntoe, n the eyent of a fuelA handling accident (FHA) must be in placa and funtoig .AIditienally, a Conitainment lselatien PhaseA (8! 6ignal) from the operating unit, high temperature in the Au~ilrary Building air intakes, or manual1 AB Ril cauce a2 CA VI cigna!in the; reh';'fulg unit. The cnainmen eqipment hatch cannot be open whenoig irrad-iateAd-fiuel incidS containment in accoArd-ance wit-h Tecnhnica Specification 3.9.4.The A.BGTS ic, required to be operable during moevement of iraitdfuel in the Auxiliary Bu~idig during an~y moede and durin~g movyement o establisehd as part of the ABSCE boundar'; (coo TS 3.3.8, 3.7.12, 3.9 01). Whe moin2iraited fuel' incide cntaiRnment, at least one train of the containmn pugsystem must be operating Or the containment FAUct be icela-ted-14.L Whenmving irradiated fuel in the Auxiliary Building during- t.m. wAhen the cnametirc open to the Auxiliar; Building A8G Spacec8, Gtawe purge can be oeadbut GAr-tARO h cyctemA ic not required.

However, whether the containment purge system is operated or- not in thic configuration, all containment ventil2atio icolatian valvecA -and 2accociated intuGtto utrmi perable.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-151 GH Containment Vent Isolation Instrumentation B 3.3.6 BASES APPLICABLE This roqu.ir...ent is ...n..e..a. to en.ure a C .VI can be frm SAFETY ho ....t f pool " ad"atin ...nitor-. in the event of an FHA i the, ANALYSES Auiliar,.

Bui.ding.

Additio;ally, a Containment isolation PhaA (continued) (SI .gna.. from the operating unit, high temperatu.re in the Au-Xiliary B3uildin~g air intakoc, Or manual ABI wil auso a CVI signal in the rphefuein 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 3.3-152 (developmental)

Q H Containment Vent Isolation Instrumentation B 3.3.6 BASES LCO (continued)

3. Containment Radiation The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Vent Isolation remains OPERABLE.For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channel electronics.

OPERABILITY may also require correct valve lineups and sample pump operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.Only the Allowable Value is specified for the Containment Purge Exhaust Radiation Monitors in the LCO. The Allowable Value is based on expected concentrations for a small break LOCA, which is more restrictive than 10 CFR 100 limits. The Allowable Value specified is more conservative than the analytical limit assumed in the safety analysis in order to account for instrument uncertainties appropriate to the trip function.

The actual nominal Trip Setpoint is normally still more conservative than that required by the Allowable Value. If the setpoint does not exceed the Allowable Value, the radiation monitor is considered OPERABLE.4. Safety Iniection (SI)Refer to LCO 3.3.2, Function 1, for all initiating Functions and requirements.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, Safety Injection, and Containment Radiation Functions are required OPERABLE in MODES 1, 2, 3, and 4, and during mov.ement of irradiated fue!

..th.. nmnt; .Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment.

Therefore, the Containment Vent Isolation Instrumentation must be OPERABLE in these MODES. See additional discussion in the Background and Applicable Safety Analysis sections.While in MODES 5 and 6 without fuel handling in the Containment Vent Isolation Instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference 1.Watts Bar -Unit 2 (developmental)

B 3.3-153 (continued)

AH Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.

If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately, and the appropriate Condition entered.A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A applies to the failure of one containment purge isolation radiation monitor channel. Since the two containment radiation monitors are both gaseous detectors, failure of a single channel may result in loss of the redundancy.

Consequently, the failed channel must be restored to OPERABLE status. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.B.1 Condition B applies to all Containment Vent Isolation Functions and addresses the train orientation of the Solid State Protection System (SSPS) and the master and slave relays for these Functions.

It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A. 1.If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation.

A Note has been added above the Required Actions to allow one train of actuation logic to be placed in bypass and to delay entering the Required Actions for up to four hours to perform surveillance testing provided the other train is OPERABLE.

The 4-hour allowance is consistent with the Required Actions for actuation logic trains in LCO 3.3.2, "Engineered Safety Features Actuation System (continued)

Watts Bar -Unit 2 B 3.3-154 (developmental)

A H Containment Vent Isolation Instrumentation B 3.3.6 BASES ACTIONS B.1 (continued)

Instrumentation" and allows periodic testing to be conducted while at power without causing an actual actuation.

The delay for entering the Required Actions relieves the administrative burden of entering the Required Actions for isolation valves inoperable solely due to the performance of surveillance testing on the actuation logic and is acceptable based on the OPERABILITY of the opposite train.A II l -I I LB A J -- i A ~oto in agapa nrannu mar uonaarion

~ :n oni~ aDD:IcaD:e in MUU~ 1.3,-OF 4 G. and G.2 Condition C applies to all Containment vent Inolation Fu, nctions and addrec:ec tho itrin orientation of the SSPS and the macter and slave relay for the Functionem it alo arRAPthefailur of mi raddilation lI mo InAGO Ihanols .Or the toailit to roctore a sinR! faIed c nNO -Wo WIbj t -ttu inWA -4 tJ n alwa o i-our Acin .1 If train ;- rmulIt;le channell iRODSrMrbI.

Or the Re, uirad i i I~tIOfl 3fl6 annociatod Unmointiori I imo OT !SOflflitiOfl A am not met..v m:aint-ain cnn-tainment purge and exhauc-t icolation 4 v21 AR~ in their cleeod poitiGn iG me~t Or the applic-able ConRdit*Aon Of 1C 3.., Cnaimn Ponotratione," are Met for each valve mnade inoeperable by failure 0 icoatin ictrmonatin.The Completfion Ti~me for. these Required.A. Nete statece ha ConditionA G snl appo~abe dOngmvmn of irr~ad~ated fuel accemblioc Within containment.

SURVEILLANCE REQUIREMENTS Al+ Ii-4 kk ý ýA#4,4r +^. +ký Q0 7ýkIý +^. ^!rf 4kýt4 TýkIý 'a 'aR I ,A +r. ;na .Ai h k Or. ,r.i, +a nab. ik, t'rý m rat'Jn fl r-,

SR 3.3.6.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that a gross failure of instrumentation has not occurred.

A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.

It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-155 AH ABGTS Actuation Instrumentation B 3.3.8 BASES 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.WU ir. IM2. rnh ., 4-+n mnu~rAk. Rothnr. Af NIAn man 0Gn. BR49 RjR.AAQG1S-.,,IJ--.1.-.-

---

nItI~It~on t~rnn !'Ht~ I M trmn ir InftI~3tod rr! finn ridI~3tinn rintnrflnc

.--.-... ---..~.--.--.-...-----J...~...--.--.-..------

channel dedicated to that train. There are a total of two channels, one for each train. High radiation detected by any monitor Or a 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 require' that Ahen moving irradiated fuel in the A "-ilhr'y Building andior Containment with the Containment andiel annuluseopen to the Auxiliar; Building ABSCE spc a ;iga! from the spent fuel radiat.iA, R* tamra' RE 90 1 02 4 and103 will i 2iti a Coitainment Ventilation lIelation (CGVP their nRMAlR I ,unction.

in aaaitvon, a signal fom the containment purge raeaiaten monitm 21 RE5 @0 130, and -131 or. ether CVI signal will initiate that pertien of the Auxiliar; Building isolation (A81) normally initiated by the spent fuel pool radiation moenitres.

Additionally, a Containment seolatieR Phase A (SI signal) from! the operating unit, high temp~erature in the Au**i~ar; Buildin~g air intakes, Or manu-al ABI W.11 c~aaUee a2 CVA signal in the Fefuei~ng unit. In the case where the containment of both units is open to9 the Auxiliary Build;ng spacoe, a CVI in one unit wil initiate a CVI in the-ether unit in order to mlaintain thQAe Spares open to the ABSCGE.Therefoem, the cont~ain1ment VSntilat*on inetrumentation must remi operable when meying irradi~ated fu el OR the Au~iliar-y Building if the containment-and/or:

annlulue air locke,, penetratines, equipment hatch, etc.(continued)

Watts Bar- Unit 2 (developmental)

B 3.3-166 G-HI ABGTS Actuation Instrumentation B 3.3.8 BASES are open to the Awdiiar-y Building ABSCE spacoc.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-167 G-H I ABGTS Actuation Instrumentation B 3.3.8 BASES APPLICABLE SAFETY ANALYSES The ABGTS ensures that radioactive materials in the ABSCE atmosphere following a ful h.ndling accident or a LOCA are filtered and adsorbed prior to being exhausted to the environment.

This action reduces the radioactive content in the auxiliary building exhaust following a LOCA-ew 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.

IAIA..,*.; .4; .4 S. I ; ;14 44%Building with-GcontAWFa-inet -air locks or. penetrations open to the Auxiliar;Buwilding ASCGE pa OF, or wen Foving fuel in the Auxiliar,'

Building wit th cntanmnt qupmenct hatch open, the PMV*Aision to ini tiate CVI from the spent fuel pool radiation moniRtorsM and ton inRittiat-e an ABI1 (i.e., the po~tion of An ABI nramally initiated by the spent fuel pool radiation moneitors) from a CVI, icunga CVI geoReated by the containcmenti purge monitors, in the event of a- fu-elI handling accident (FH4A) mu, st be in plac~e and fuIonng Additionally, a Containment Isolation.

Phase A (SI signal) from the operating unit, high temperature i theAuxi-ili-ary Building air intakes, Or manua! AB! will csaurse a .V ina in the refueling unit.The con-Ftainmen~t equipment hatcah cannot be open when mo~ving irradiated fulinside containment inaccordance Mith Tkqchn~Specification 3.9.4.The ABGT-S is required to be operable during FMeOeMAnt Of iraitdful 8 in the Auxiliary Building during any mode and durn' eooto irrad-iateA-d fiuel in the Reactor Building when the Re~actr Buldngi es-t-ablishe-d as part of the ABSRCE bondh .cI Ts 3.3.8, 3.7.112, &3.9.4). Whe moig radiated fuel inside con~tainmqent, at least one train of the containen pug system m~ust be operating or the containIIAmen mus-t be isalated.

hA'hn MeVing irradiated fue! in the Au**iiary Building dur~ing times when the, cneRtainmxent is open to the Au*i~iary Building A1Sr spaces, retiFe purge can beoeaebut ep~teiiG systemR is not required.

HeweyeF, Whether the contaiRnment purge systemp es o perated Or not in this configuration, all containment venti~atieniolto valves an~d associ0ated instwm~entatien must rem~ain operable.This requirem:ent is necessary to ensurze a CVI can be accomplished fromp the spen~t fuel pool radiation monitorsi the AAFnt ofR a FA in the Auxiliary Building.

Additionally, a ConRtainmen-t seainPhasA (SI signal) fromA the opertating unit, high tem~perature inthe Au*iliary, Building air intakes, Or manu a! ABI1 will caus-e 2 CVI signal in the refueling unit. In the c9ase whoe.- the containment of both units is open to the AuI~i~iar.

Building spacos, a CVI Inone untvvl "" it a GXIin the ete unit in order to mnaintain those spaces open to the ABSCrE.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-168 Cs-H I ABGTS Actuation Instrumentation B 3.3.8 BASES 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 Arap Radiation onsure that the radiatio moniRtorig intuetto ecoccar; to initia;te-tha ABGTS rmai4np OPERABLE On QrSadiation moni~tor is daedicatad to each train of ABGTS.FoF camp long systems, channel OPERABILITY inunlvec more than OPERABILITY of channel SelotronceS.

OPERA1BILITY may also reqirecorect~ale ineups, sample pump operation, and filter mo~tot opRato, ac well aG detectnonr OP31ERAB!ITYP, if thece cUppo~ting featuraec are n8eocar; for trip to occur under the GOnditione assumed by the safety analycoc.Only the Allowable Value ic 8epoifie-d for- the FuelA Poe' Are RadiatAion Mon~itora in the CO. The AllMOwabl Value Sepoifiod is mor~e Gencorwative than the analytical limit assumed in the safety an~alyricr in o-rderff to- account't for inebtrUmFent uncaretaintie6 appropriate to the trip function.

The actua' nom~ina! Trip Sotpoint is norm~ally still more GORG-e R a& Pe than that required by the Allowable Value. If thel meacured ~ ..- cepontdoc oto iod theg Allowable Value, the radiation monFitor ic concidered OPERABLE.2. Containment Phase A Isolation Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-169 G-H I ABGTS Actuation Instrumentation B 3.3.8 BASES APPLICABILITY The manual ABGTS initiation must be OPERABLE in MODES 1, 2, 3, and 4 and when " mo"ing irradiated fu--el assemblies in tAh hAndlin area to ensure the ABGTS operates to remove fission products associated with leakage after a LOCA Or a fuel handling accident.

The Phase A ABGTS Actuation is also required in MODES 1, 2, 3, and 4 to remove fission products caused by post LOCA Emergency Core Cooling Systems leakage.HIgh ramaiaen iniiarien me us" t Be i.dER .Kl in any MUiIJ du wrin move-ment of irradiated fuel assemblies in the fue'l handling area to v enuren auolmaniG Iniiaonm-e ki I !A -WnenA mRe pont;Iafrl r TG1a rue1 handlin accident e v cte.While in MODES 5 and 6 without fuel i" PFr.g..., the ABGTS instrumentation need not be OPERABLE cinA" a fuel handling acc,,ident cannot occur. See additional discussion in the Background and Applicable Safety Analysis sections.ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.

If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-170 A-H I ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS A.1 (continued)

Condition A applies to the actuation logic train function from the Phase A Isolation, the radiation monitor funGons, and the manual initiation function.

Condition A applies to the failure of a single actuation logic train, monitor channel-, or manual channel. If one channel or train is inoperable, a period of 7 days is allowed to restore it to OPERABLE status. If the train cannot be restored to OPERABLE status, one ABGTS train must be placed in operation.

This accomplishes the actuation instrumentation function and places the unit in a conservative mode of operation.

The 7-day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable.

The basis for this time is the same as that provided in LCO 3.7.12.B.1.1, B.1.2. B.2 Condition B applies to the failure of two ABGTS actuation logic signals from the Phase A Isolation, t-o radiation monitort, or two manual channels.

The Required Action is to place one ABGTS train in operation immediately.

This accomplishes the actuation instrumentation function that may have been lost and places the unit in a conservative mode of operation.

The applicable Conditions and Required Actions of LCO 3.7.12 must also be entered for the ABGTS train made inoperable by the inoperable actuation instrumentation.

This ensures appropriate limits are placed on train inoperability as discussed in the Bases for LCO 3.7.12.Alternatively, both trains may be placed in the emergency radiation protection mode. This ensures the ABGTS Function is performed even in the presence of a single failure.Cond-ition C applies when the Required Action and associatd Completion Time for Condition A or B have net boon met and irradiated fuel are being moved in the fuel building.

Movement e.irrad-iate-d fuel asscomblies in the fiuel building must be suspend i mmediately to eliminate the potentia!

for evente that could reur A13G~T- actuation.PfGRar of ths aefie ,,.mVoving a cernaonent to a c~afe oecition (continued)

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

A-H ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS (continued) alaii- 0.2C1 and C2 Condition 0 C applies when the Required Action and associated Completion Time for Condition A or B have not been met and the plant is in MODE 1, 2, 3, or 4. The plant must be brought to a MODE in which the LCO requirements are not applicable.

To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS A ,Note ha been added to the SRRTbet l~ thatTal 2RI dotorminecm hic SI apply to which ABGT-S Actuation Functioe SR-3444.Perform~ancA of thA CHANNELr-CHE-CK once. Aevey 12 heure ensmuPReth~a agroce faWlur Of iRG#Wmentatieonha, not occurred.

A HANNElCrl CHEC i1 normally aG cmpaGRIilonf the parameter indicated on one channel to a 6imilar parameter on etner cflaRneiL.

It Ic a'en on tne accumptho ar inctrumsnt channsle monitoring the same parameter sheuld read appro..mately the same value.

t

,, o " Q inetru mant chRnnolc Coul beA Rn iniato of euceessineru ntdfti onRe Gof the channole Or of something even mor~e rOerinouci.

A CHANNEL CHECK wlan detel t gnroni channl faialur; thui, it is key to ,erifying the inctumetat~n cntinuec to operate properly between eacah CH4ANN~EL, but....-,

are "ULined the ... .. staff, based e..a combinti of t Vhe c hanneIl nert uncIrtaintiec, includinl indicatgon and readability.

Ifachannel~icutede the c~ritria, it may beaninication that the concor Or the cigna! procacin eqipet hac drifted outside its 4limt.The Frequency i based on operatingexeiec that demnestrates chann~el failura ic are. Th CHNE CEKcppements

!occ formal, bu-,t- mrea frequent, checkA of channels duriFng nor~mal operational -c -P. o the displays accociated;Awith t-he 1-C0 required channels.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-172 A-H ABGTS Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE SR-3-342 REQUIREMENTS (continued)

Av GT nformed onca ever; 02 days en each required channel tomn oRtnmro An~nnoi fll n % nI Tnr iMonnuou0 i GnR. In n18 6u1m Yerifie the capability of the -intR mentation to provide the ABOG TS ac-tuation.

Th4 Fquny cf 02 days is based On the knoAA1 reliability Gf the monitorig eqipet and has been AhoWn to b accaeptable through operating eprn. Thre a plant peifi whih eFA that- the Instrument channel functionc aS required by Yerifying the as lef and- ac fo-und settng a~ere nsistent with these established by the setpeint mfethedelegy.

S R 3.3.08.31 SR 3.3.8.3 1 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every 18 months. Each manual actuation function is tested up to, and including, the relay coils. In some instances, the test includes actuation of the end device (e.g., pump starts, valve cycles, etc.). The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.SR-4348A A C.ANNIAlI CA' IDRATION is nrftrFmed ever' 18 month#14... .. : ...* I ... .. ..& .- I:-- #%11Allllr~ I kor-A approxlm to:.y at ovor; rTU'-' f,,. 4--1,""-L i.,"L: "r", : :'.r: :1 Gemnpuete chenk of the RlIw,*uding the cen -r. The tet verifies that the ch:anne! respendr, to a meaurwed parameter within the n~ecoccar; range and accuracy.

The FrFequency i bRaed On operatin~g exprieceand-WS concAistent with the typica! industr, refueling cycle..or 3i f S in pecIRc prOgwm nn WR uRiw Ye1+18 HMo we 1RrtUm c~hannel funcAtiGne aG required by verifying the as le-ft- andC- as found setting are concistent Yith these established by the Setpoint mnethodology.

REFERENCES

1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance." (continued)

Watts Bar -Unit 2 (developmental)

B 3.3-173 A-H I Containment B 3.6.1 BASES APPLICABLE Satisfactory leakage rate test results are a requirement for the SAFETY establishment of containment OPERABILITY.

ANALYSES (continued)

The containment satisfies Criterion 3 of the NRC Policy Statement.

LCO Containment OPERABILITY is maintained by limiting leakage to < 1.0 La, except prior to the first start up after performing a required Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met.Compliance with this LCO will ensure a containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.Individual leakage rates specified for the containment air lock (LCO 3.6.2), purge valves with resilient seals, and Shield Building containment bypass leakage (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J, Option B. Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding the acceptance criteria of Appendix J, Option B.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 and 6 to prevent leakage of radioactive material from containment.

The ..quir.ment.

for ...ntainMot durng MODR 6 aro gaddmro-d inGLCO 3.9.4, u"Gntainm.nt o .n.r.atinc.2" (continued)

Watts Bar -Unit 2 (developmental)

B 3.6-3 H I Containment Air Locks B 3.6.2 BASES (continued)

APPLICABLE SAFETY ANALYSES The DBAs that result in a significant release of radioactive material within containment are a loss of coolant accident and a rod ejection accident (Ref. 2). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate (LJ) of 0.25%of containment air weight per day (Ref. 2), at the calculated peak containment pressure of 15.0 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.The containment air locks satisfy Criterion 3 of the NRC Policy Statement.

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

As part of containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.Each air lock is required to be OPERABLE.

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

The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.

Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODE 5 and 6 to prevent leakage of radioactive material from containment.

The requiroM.n-t for th- cfar. t nmt.t air lock, dUrI, MODE 6 are in C 3.9.4, "uAnta-iment Watts Bar -Unit 2 (developmental)

B 3.6-7 (continued)

H Containment Isolation Valves B 3.6.3 BASES (continued)

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5 and 6.The roquiF8montA for. conbinRmonti1;eation valves during MODE 6 aro addressed i~q!LCO 3.9.4, "GContainmont PonoAtr1at-ione;." ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths, to be unisolated intermittently under administrative controls.

These administrative controls consist of stationing a dedicated operator (licensed or unlicensed) at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.

For valve controls located in the control room, an operator (other than the Shift Operations Supervisor (SOS), ASOS, or the Operator at the Controls) may monitor containment isolation signal status rather than be stationed at the valve controls.

Other secondary responsibilities which do not prevent adequate monitoring of containment isolation signal status may be performed by the operator provided his/her primary responsibility is rapid isolation of the penetration when needed for containment isolation.

Use of the Unit Control Room Operator (CRO) to perform this function should be limited to those situations where no other operator is available.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions.The ACTIONS are further modified by third Note, which ensures appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve.In the event the isolation valve leakage results in exceeding the overall containment leakage rate, Note 4 directs entry into the applicable Conditions and Required Actions of LCO 3.6.1.(continued)

Watts Bar -Unit 2 B 3.6-16 (developmental)

H Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.7 REQUIREMENTS Verifying that each 24 inch containment lower compartment purge valve is blocked to restrict opening to < 500 is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses of References 1 and 2. If a LOCA occurs, the purge valves must close to maintain containment leakage within the values assumed in the accident analysis.

At other times when containment when purge al've required to be capable of .. -, ng (e.g., my,. m. t of ir rdit -ed fu, assemblies), pressurization concerns are not present, thus the purge valves can be fully open. The 18-month Frequency is appropriate because the blocking devices are typically removed only during a refueling outage.SR 3.6.3.8 This SR ensures that the combined leakage rate of all Shield Building bypass leakage paths is less than or equal to the specified leakage rate.This provides assurance that the assumptions in the safety analysis are met. The as left bypass leakage rate prior to the first startup after performing a leakage test, requires calculation using maximum pathway leakage (leakage through the worse of the two isolation valves). If the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange, then the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. At all other times, the leakage rate will be calculated using minimum pathway leakage.The frequency is required by the Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria.Although not a part of La, the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.(continued)

Watts Bar -Unit 2 B 3.6-25 (developmental)

BH 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 poo! following a fue! 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 .fro the epent fuel poe'0 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 reguiroc that vhen FRGVing iFradiated fu-el in the Auxiliar; Buildin~g andier Containmenpt With the Containment open to the Auxiliar;

@u ilding ABSCE epacoc, a signal from the spent fuel poo fad-ia~t~io moITere0 0 RE 90 102 and 103 will1 initiate a ContainmFenlt Ve1Rn1.iat*

I!cllatin (CVI) in addition to their fu Itin. In additieI, a cignal from the containment purge rad'AtiAn menitorR 1 RE 90 130 andI 131 or oltherIi CVI signal will iintiate that portion of the ABI noeal1y nintiated by the spent fuel peel radiation manitam. Additionally, a Containment Phaco A (SI signal) from the .perating unit, high 18RMoEraiue in Me Ahwuiiiar:

Suidiain aIr 1InM1486.

Or mianuat ABI (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-63 GH I ABGTS B 3.7.12 BASES BACKGROUND (continued) 9 L.tillll containment of botmh units is ;ep to the AuxiliaFy lBuilding spaceS, a V\I in one uni Iwil intiate al CI inM m th other uni;t in order to maiRtain these spaces open to thA ABSC(E. Therefor, the cOntainment ventilation iTremBGTS sdiscmust remain operable when moVing iradiated fuel in the A.uxiliar Buildin. if the andtainment air ielyk)penetrations, equipmen8t hatch, Mtc. are open to the Au*iliar; Bu:Iding ABSCE spacee. In addition, the ABRG-TS must reanoeable i these containment penetrations are open to the Auxiliar Bulin uring movemenet of ir-radiwate&d-fuel insideS con~t-ain.men.t-.

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 SAFETY ANALYSES The ABGTS design basis is established by the consequences of the limiting Design Basis Accident (DBA), which is a LOCA fuel handling arGide. .The analy is of the fue! handling accident, given.... ... .1 .. ... ....... 1 .. .... :' n i KSMMA e , aL.LU:IIeL

+:1 "J~ 2+ u"I 'Au ' i F49 i A ' ARAH6FA0y W[e damagedJ.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 handlirg ac.!det 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 f4el handling accident ad for a LOCA. The assumptions and- the analysis for a- fuel hand-ling accident.follow the guidanA, provided i* Regulator; Guide 1.25 (Ref. 5) and NUREG-/CP.

5009 (Ref.f 0) The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 6 5).The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

IMA MQf*F9 mditedfi-l isidcontainment inr; the A--iia9 Building YAWi containment air locke or penetrations open to the Auxiia- y Building ABSCE spaces, Or When moving fuel in the Auxiliary Building with the containment equipment hatch open, the p n to initiate a r,%1 fr^m +nh¢ r' 4 f ! i , mr , *, ., ,, ,,ý IA, * "-i --- ADI '-'."-.".-.. .. -... -W cppwn ww pcpcp r= = CPý ^Ezian Ox Rn " n^ý-" inmviýTýý

ý Tný ý ýnT rtjýi ^^i......I-- -*n -- '. Si f.j Ib V fmdi'it*n Mnnitem) frnm , CVI inRG'udR~n aG 1" *nt',tniat by the I .... V ..... Vl IgV coeniainment purge moniltors, in the event 0T a-we nanaing aGciaent (FHMA) must be in p lace andI f ucio ian. Additionally, a Containment

% #I v r i W i icoiarion mnase A tsi sigaij WFro tne Operatlng unit, niqn* w (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-64 GH ABGTS B 3.7.12 BASES APPLICABLE SAFETY ANALYSES (continued) temperSature in the AuxW-iliary Building air intakes, Or manual ABI vill cause a CVI signal in the refueling unit. The containment equipmenAt hatch cannot be open when movingirradiated fuel inside con~t2ainmnt in 2-A accord- ante With Technica!

Specifiation 3.0.4.TheABG-QTS iS required to be operable during moevement of irradiated fue!in the Au-xi'ia Building duýrig an~y mode andd duinmomet irradiated fuel in the Reactor Building when the Reactor Building iS established as part of the ABS-CrE boundary (see TS8 3.3.8, 3.7.12, &3.9.4). Whe moig rad*ated fuelA inside containment, at least one train of he ona~ne pugeSystemA mRust beS operating Or the cOntaimenFlt must be isolated.

WAhen. moGVing irradiated fuel in the Auxiliary Building during timers, ven the containment is open to the Auxiliary Building ABSCE spacas, containment purge can be operated, but operation of th system is not required.

HoeWVer, whe-ther the coentainment purge system is operated Or not in this configuration, all containment ventilation isolation valves and associated instrum~entation must remain operable.

This r8;equiement isneMsarAy to ensure a CVI ca e coplsedfo the spet fel ol radiation moni;tors in; the event of a FHA in the Auxiliaigy Building.

Additionally, A- Con-tainRment Isolation Phase A (S! signal) from the operating unit, high temAperature in the Auxiliary Building air intakes, or1 Manua! ABI will cause a CVI saignal inthe refueling unit. In; the case whe~re the coentainment of both units is open to the Auxiliary Building A-Magces.

aa C R in A one uit will intiat-e A- CVI in the other uit inm order to m~aInaIn tnose spares open to thle ABSGF.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 6) limits in the event of a fuel handling acc-.ident Or LOCA.The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building Auxiliary 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 (developmental)

B 3.7-65 GHI ABGTS B 3.7.12 BASES LCO c. Heater, moisture separator, ductwork, valves, and dampers are (continued)

OPERABLE, and air circulation can be maintained.

APPLICABILITY In MODE 1, 2, 3, or 4, the ABGTS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.In MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

During mo'vement of irradiattd fuel in the fuel handling area, the ABGTS is required to be OPERA.BLE to alloviato the conccquoncacs of a- fu-el handling acc~ident.

See additiena!

diecuccinkin thBckroud anid Applicable Safety Analysis sections.ACTIONS A..1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function.

The 7-day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1 and B.2 in MODE 1, 2, 3, or 4, whn When Required Action A. 1 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-66 AH ABGTS B 3.7.12 BASES ACTIONS (continued)

'J"flnn I~nniiirnr1 Rrtinnhl 1 e--innnt fln rnmnintnri

~'Itflin tfln mAuJirfid Completion Time, duiRng moGYmen8t Of rrFAdiated fuel -as-sembiDec in the fuel handling area, the OPEIRABLE ABRGTS train muc-t be atar+-d remaining train in- OPERABLE, that no undetected failures preyenting cyctem operation All occur, and that any active failure- will be readily d4etteaetedd.

if the system is not plaoad in operation, this acti eq uieccupension c fuel movement, which precluides a fuel accident.

Thic dIGAR not preclude the monvement of fuel assemblies to a safe pocition 1 A.hon twov trans of the AB3GT- are ineporable dug moeent Gf ira ite fual ;arAmbl*ec in theg fuel handling ara ac inmust be taken to place the unit in a cond~ition_

in which the LCO door, not apply. Action must be taken immediately to suseond movyemnent of irradiated fuel assemblies i the fuela han~dling area. TWAicdooc not precludeth fA48movemet of fuel1 to a Raft; nacition.SURVEILLANCE REQUIREMENTS SR 3.7.12.1 Standby systems should be checked periodically to ensure that they function properly.

As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. The system must be operated for _> 10 continuous hours with the heaters energized.

The 31-day Frequency is based on the known reliability of the equipment and the two train redundancy available.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-67 AH ABGTS B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.12.2 This SR verifies that the required ABGTS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 8 7). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ABGTS train starts and operates on an actual or simulated actuation signal. The 18-month Frequency is consistent with Reference 8 7.SR 3.7.12.4 This SR verifies the integrity of the ABSCE. The ability of the ABSCE to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the ABGTS. During the post accident mode of operation, the ABGTS is designed to maintain a slight negative pressure in the ABSCE, to prevent unfiltered LEAKAGE. The ABGTS is designed to maintain a negative pressure between -0.25 inches water gauge and -0.5 inches water gauge (value does not account for instrument error) with respect to atmospheric pressure at a nominal flow rate > 9300 cfm and < 9900 cfm. The Frequency of 18 months is consistent with the guidance provided in NUREG-0800, Section 6.5.1 (Ref. 0 8).An 18-month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 8 7.I REFERENCES

1. Watts Bar FSAR, Section 6.5.1, "Engineered Safety Feature (ESF)Filter Systems." 2. Watts Bar FSAR, Section 9.4.2, "Fuel Handling Area Ventilation System." 3. Watts Bar FSAR, Section 15.0, "Accident Analysis." (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-68 B-HI ABGTS B 3.7.12 BASES REFERENCES (continued)

4. Watts Bar FSAR, Section 6.2.3, "Secondary Containment Functional Design." 5-.W* Ii Jl__ __=L al U A ....... .IL==I I--=Jl=--ý i! ýT^ Lj ý ýrf% ýijm 013r. ýý6 5.-7 6.8 7.0 8.Evaluating the Potontial Radiological COnI~eqUOnce6 Of a Fuel Hand~ing Acciden~t in the Fuel Handling and Storage Facility for Boi~ling anRd Proccuwzed Water Reador~s." Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors." Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance." Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants." NUREG-0800, Section 6.5.1, "Standard Review Plan," Rev. 2, "ESF Atmosphere Cleanup System," July 1981.407 All II, AC 60P 7'Ac~eccmorn ot tno Ic or.Aa imuoneag ISUFRUP Fueli n Lighti WaISF oWQorI- r1acor.. 8- NucioaF -to61auizno I AAA i Gommisee~n, 1-obruary 1988.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-14&-HI Fuel Storage Pool Water Level B 3.7.13 BASES (continued)

REFERENCES

1. Watts Bar FSAR, Section 9.1.2, "Spent Fuel Storage." 2. Watts Bar FSAR, Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System." 3. Watts Bar FSAR, Section 15.5.6 4&.45, "Fuel Handling Accident." 4.Regulatory Guide 1.25, March 107-2, 'Assumnptions Ucod fot Evaluating the Potential Radiological Genceguences of a Fuel Handli.n Acciaent in the -oel Handling and StGaa. Facility for II i R i~oiiin~i and l~rec~unzed

~'?ator I-~oactore.

5.Title 10, Codde of Fed-eral Regulations, Part 100.1 I, "-etermination of Ewwucsoen Aroa, Lewx Population Zonie, and Population CenAtet 6. Regulatory Guide 1.183, "Alternate Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors", July 2000.7. Title 10, Code of Federal Regulations 50.67, "Accident Source Term." (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-14 A Fuel Storage Pool Water Level B 3.7.13 8 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity.

The water also provides shielding during the movement of spent fuel.A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR,Section I 4S 15.5.6 (Ref. 3).APPLICABLE SAFETY ANALYSES The minimum water level in the fuel storage pool meets the assumptions of the fuel handling accident described in Regulatory Guide 425 (Ref.4)1.183 Rev. 6. The Total effective Dose equivalent (TEDE) for control room occupants, individuals at the exclusion area boundary, and individuals within the low population zone will remain with 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref 6)for a fuel handling accident.

The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) limits.According to Reference 3 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident.

With 23 ft of water, the assumptions of Reference 6 4 can be used directly.

In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks; however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small non-conservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.The fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-68 AH Fuel Storage Pool Water Level B 3.7.13 (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-69 AH Fuel Storage Pool Water Level B 3.7.13 BASES (continued)

LCO The fuel storage pool water level is required to be > 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3). As such, it is the minimum required for fuel storage and movement within the fuel storage pool.APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists.ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.

When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended.

This action effectively precludes the occurrence of a fuel handling accident.

This does not preclude movement of a fuel assembly to a safe position.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SIR 3.7.13.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.

The water level in the fuel storage pool must be checked periodically.

The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.(continued)

Watts Bar -Unit 2 B 3.7-70 (developmental)

A Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange.During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool.Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 2 and 8 an.d 2).Sufficient iodine activity would be retained to limit offsite doses from the accident to -,2% of 10 ,FR 100 limits, as providod by the guidanco,, ee the limits defined in 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 8).APPLICABLE SAFETY ANALYSES During movement of irradiated fuel assemblies, the water level in the refueling canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment,-as postulated by Rogulator; Guido 1.25 (Ref. 1). A minimum water level of 23 ft (Regulatory Position 2 of Appendix B to Regulatory Guide 1.183 (Ref. 8)) allows an overall iodine decontamination factor of 200 C.I? 9f 0-f 4~ H1 A ^fsn D i +,, D.,4, 40J n n0I~i , I {of-Ref---)

to be used in the accident analysis fer--ied~in.

This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 8% of the 1-131, 10% of the Kr-85, and 5% of the other noble gases and iodines from the total fission product inventory in accordance with Regulatory Position 3.1 of Regulatory Guide 1.183 (Ref. 8). 40%-of tho, total Afuo81 roed i;din in;enter; (Rof. 1) oxcopt fo r! 131 whirch i The fuel handling accident analysis inside containment is described in Reference

2. With a minimum water level of 23 ft in conjunction with aOl a minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 7 and 8 4 a, 5-).Refueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.(continued)

Watts Bar -Unit 2 (developmental)

B 3.9-20 AH Refueling Cavity Water Level B 3.9.7 (continued)

Watts Bar -Unit 2 (developmental)

B 3.9-21 AH Refueling Cavity Water Level B 3.9.7 BASES (continued)

LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits, as provided by the guidance of Reference 3.APPLICABILITY LCO 3.9.7 is applicable when moving irradiated fuel assemblies within containment.

The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis.

If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.

Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.13, "Fuel Storage Pool Water Level." ACTIONS A. 1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.A.2 In addition to immediately suspending movement of irradiated fuel, actions to restore refueling cavity water level must be initiated immediately.

SURVEILLANCE REQUIREMENTS SR 3.9.7.1 Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.(continued)

Watts Bar -Unit 2 (developmental)

B 3.9-22 A Refueling Cavity Water Level B 3.9.7 BASES (continued)

REFERENCES

1. Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Conoquec.c of a Fuel Handli*n;g AccGidont in the Fuel Handling and Storage Facility for Boiling and PF8oc1unzJEd IA.;tAF Roactem " IUISR Nucle9Ar Roaulator.y Commiccioin, March 23, 1972.2. Watts Bar FSAR, Section 15.5.6 54.5, "Fuel Handling Accident." 3. NUREG-0800, "Standard Review Plan," Section 15.7.4,"Radiological Consequences of Fuel-Handling Accidents," U.S. Nuclear Regulatory Commission.

4. Title 10, Code of Federal Regulations, Part 20.1201(a), (a)(1), and (2)(2), "Occupational Dose Limits for Adults." 5~Malinowski, D. D., Bell, M. j., Duhn, E., and Locante, j., W6:~ Id Kaelaaioggleal conequences e; a i-uel manaiin Accident, December 197-1.6- NUREGICR 5009, "Assessment of the Use of E*tended Bumu Fuel in Light W~ater Power Reactors," U. S. NUcloar Regulatory Commlccion, Februar-y 1988.7. Title 10, Code of Federal Regulations 50.67, "Accident Source Term." 8. Regulatory Guide 1.183, "Alternate Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors", July 2000.(continued)

Watts Bar -Unit 2 (developmental)

B 3.9-14 A Enclosure 9 WBN Unit 2 -Revised Technical Specification Bases Final E9-1 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." APPLICABLE SAFETY ANALYSES The containment isolation valves for the Reactor Building Purge System close within six seconds following the DBA. The containment vent isolation radiation monitors act as backup to the SI signal to ensure closing of the purge air system supply and exhaust valves. 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.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-150 HI Containment Vent Isolation Instrumentation B 3.3.6 BASES LCO The LCO requirements ensure that the instrumentation necessary to initiate Containment Vent Isolation, listed in Table 3.3.6-1, is OPERABLE.5. 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.6. 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 3.3-151 (developmental)

H Containment Vent Isolation Instrumentation B 3.3.6 BASES LCO 7. Containment Radiation (continued)

The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Vent Isolation remains OPERABLE.For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channel electronics.

OPERABILITY may also require correct valve lineups and sample pump operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.Only the Allowable Value is specified for the Containment Purge Exhaust Radiation Monitors in the LCO. The Allowable Value is based on expected concentrations for a small break LOCA, which is more restrictive than 10 CFR 100 limits. The Allowable Value specified is more conservative than the analytical limit assumed in the safety analysis in order to account for instrument uncertainties appropriate to the trip function.

The actual nominal Trip Setpoint is normally still more conservative than that required by the Allowable Value. If the setpoint does not exceed the Allowable Value, the radiation monitor is considered OPERABLE.8. Safety Injection (SI)Refer to LCO 3.3.2, Function 1, for all initiating Functions and requirements.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, Safety Injection, and Containment Radiation Functions are required OPERABLE in MODES 1, 2, 3, and 4. Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment.

Therefore, the Containment Vent Isolation Instrumentation must be OPERABLE in these MODES. See additional discussion in the Background and Applicable Safety Analysis sections.While in MODES 5 and 6, the Containment Vent Isolation Instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference 1.(continued)

Watts Bar -Unit 2 B 3.3-152 (developmental)

H Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.

If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately, and the appropriate Condition entered.A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A applies to the failure of one containment purge isolation radiation monitor channel. Since the two containment radiation monitors are both gaseous detectors, failure of a single channel may result in loss of the redundancy.

Consequently, the failed channel must be restored to OPERABLE status. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.B.1 Condition B applies to all Containment Vent Isolation Functions and addresses the train orientation of the Solid State Protection System (SSPS) and the master and slave relays for these Functions.

It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A. 1.If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation.

A Note has been added above the Required Actions to allow one train of actuation logic to be placed in bypass and to delay entering the Required Actions for up to four hours to perform surveillance testing provided the other train is OPERABLE.

The 4-hour allowance is consistent with the Required Actions for actuation logic trains in LCO 3.3.2, "Engineered Safety Features Actuation System (continued)

Watts Bar -Unit 2 B 3.3-153 (developmental)

H Containment Vent Isolation Instrumentation B 3.3.6 BASES ACTIONS B.1 (continued)

Instrumentation" and allows periodic testing to be conducted while at power without causing an actual actuation.

The delay for entering the Required Actions relieves the administrative burden of entering the Required Actions for isolation valves inoperable solely due to the performance of surveillance testing on the actuation logic and is acceptable based on the OPERABILITY of the opposite train.SURVEILLANCE REQUIREMENTS SR 3.3.6.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that a gross failure of instrumentation has not occurred.

A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.

It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-154 H I ABGTS Actuation Instrumentation B 3.3.8 BASES 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 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.There are a total of two channels, one for each train. 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 ABGTS ensures that radioactive materials in the ABSCE atmosphere following a LOCA are filtered and adsorbed 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.

APPLICABLE SAFETY ANALYSES (continued)

Watts Bar -Unit 2 (developmental)

B 3.3-166 HI ABGTS Actuation Instrumentation B 3.3.8 BASES 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. Containment Phase A Isolation Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-167 H I ABGTS Actuation Instrumentation B 3.3.8 BASES APPLICABILITY The manual ABGTS initiation must be OPERABLE in MODES 1, 2, 3, and 4 to ensure the ABGTS operates to remove fission products associated with leakage after a LOCA. The Phase A ABGTS Actuation is also required in MODES 1, 2, 3, and 4 to remove fission products caused by post LOCA Emergency Core Cooling Systems leakage.While in MODES 5 and 6, the ABGTS instrumentation need not be OPERABLE.

See additional discussion in the Background and Applicable Safety Analysis sections.ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.

If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A applies to the actuation logic train function from the Phase A Isolation and the manual initiation function.

Condition A applies to the failure of a single actuation logic train or manual channel. If one channel or train is inoperable, a period of 7 days is allowed to restore it to OPERABLE status. If the train cannot be restored to OPERABLE status, one ABGTS train must be placed in operation.

This accomplishes the actuation instrumentation function and places the unit in a conservative mode of operation.

The 7-day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable.

The basis for this time is the same as that provided in LCO 3.7.12.(continued)

Watts Bar -Unit 2 B 3.3-168 (developmental)

H ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS (continued)

B.1.1, B.1.2. B.2 Condition B applies to the failure of two ABGTS actuation logic signals from the Phase A Isolation or two manual channels.

The Required Action is to place one ABGTS train in operation immediately.

This accomplishes the actuation instrumentation function that may have been lost and places the unit in a conservative mode of operation.

The applicable Conditions and Required Actions of LCO 3.7.12 must also be entered for the ABGTS train made inoperable by the inoperable actuation instrumentation.

This ensures appropriate limits are placed on train inoperability as discussed in the Bases for LCO 3.7.12.Alternatively, both trains may be placed in the emergency radiation protection mode. This ensures the ABGTS Function is performed even in the presence of a single failure.Cl and C2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met and the plant is in MODE 1, 2, 3, or 4. The plant must be brought to a MODE in which the LCO requirements are not applicable.

To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.(continued)

Watts Bar -Unit 2 (developmental)

B 3.3-169 H I ABGTS Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE REQUIREMENTS SR 3.3.8.1 SR 3.3.8.1 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every 18 months. Each manual actuation function is tested up to, and including, the relay coils. In some instances, the test includes actuation of the end device (e.g., pump starts, valve cycles, etc.). The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.REFERENCES

1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance." (continued)

Watts Bar -Unit 2 (developmental)

B 3.3-170 H I Containment B 3.6.1 BASES APPLICABLE Satisfactory leakage rate test results are a requirement for the SAFETY establishment of containment OPERABILITY.

ANALYSES (continued)

The containment satisfies Criterion 3 of the NRC Policy Statement.

LCO Containment OPERABILITY is maintained by limiting leakage to < 1.0 La, except prior to the first start up after performing a required Containment Leakage Rate Testing Program leakage test. At this time, applicable leakage limits must be met.Compliance with this LCO will ensure a containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.Individual leakage rates specified for the containment air lock (LCO 3.6.2), purge valves with resilient seals, and Shield Building containment bypass leakage (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J, Option B. Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding the acceptance criteria of Appendix J, Option B.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 and 6 to prevent leakage of radioactive material from containment.

Watts Bar- Unit 2 (developmental)

B 3.6-3 (continued)

H Containment Air Locks B 3.6.2 BASES (continued)

APPLICABLE SAFETY ANALYSES The DBAs that result in a significant release of radioactive material within containment are a loss of coolant accident and a rod ejection accident (Ref. 2). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate (La) of 0.25%of containment air weight per day (Ref. 2), at the calculated peak containment pressure of 15.0 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.The containment air locks satisfy Criterion 3 of the NRC Policy Statement.

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

As part of containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.Each air lock is required to be OPERABLE.

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

The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.

Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODE 5 and 6 to prevent leakage of radioactive material from containment.

Watts Bar -Unit 2 (developmental)

B 3.6-7 (continued)

H Containment Isolation Valves B 3.6.3 BASES (continued)

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5 and 6.ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths, to be unisolated intermittently under administrative controls.

These administrative controls consist of stationing a dedicated operator (licensed or unlicensed) at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.

For valve controls located in the control room, an operator (other than the Shift Operations Supervisor (SOS), ASOS, or the Operator at the Controls) may monitor containment isolation signal status rather than be stationed at the valve controls.

Other secondary responsibilities which do not prevent adequate monitoring of containment isolation signal status may be performed by the operator provided his/her primary responsibility is rapid isolation of the penetration when needed for containment isolation.

Use of the Unit Control Room Operator (CRO) to perform this function should be limited to those situations where no other operator is available.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions.The ACTIONS are further modified by third Note, which ensures appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve.In the event the isolation valve leakage results in exceeding the overall containment leakage rate, Note 4 directs entry into the applicable Conditions and Required Actions of LCO 3.6.1.(continued)

Watts Bar -Unit 2 B 3.6-16 (developmental)

H Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.7 REQUIREMENTS Verifying that each 24 inch containment lower compartment purge valve is blocked to restrict opening to < 500 is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses of References 1 and 2. If a LOCA occurs, the purge valves must close to maintain containment leakage within the values assumed in the accident analysis.

At other times when containment pressurization concerns are not present, the purge valves can be fully open. The 18-month Frequency is appropriate because the blocking devices are typically removed only during a refueling outage.SR 3.6.3.8 This SR ensures that the combined leakage rate of all Shield Building bypass leakage paths is less than or equal to the specified leakage rate.This provides assurance that the assumptions in the safety analysis are met. The as left bypass leakage rate prior to the first startup after performing a leakage test, requires calculation using maximum pathway leakage (leakage through the worse of the two isolation valves). If the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange, then the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. At all other times, the leakage rate will be calculated using minimum pathway leakage.The frequency is required by the Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria.Although not a part of La, the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.(continued)

Watts Bar -Unit 2 B 3.6-25 (developmental)

BH 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 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.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 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).(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-63 H ABGTS B 3.7.12 BASES APPLICABLE SAFETY ANALYSES LCO The ABGTS design basis is established by the consequences of the limiting Design Basis Accident (DBA), which is a LOCA. 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 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 LOCA.The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 5).The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

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. 6) limits in the event of a LOCA.The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the Auxiliary 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 c. Heater, moisture separator, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-64 H ABGTS B 3.7.12 BASES LCO (continued)

APPLICABILITY In MODE 1, 2, 3, or 4, the ABGTS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.In MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.ACTIONS A.1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function.

The 7-day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1 and B.2 When Required Action A.1 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.Watts Bar -Unit 2 (developmental)

B 3.7-65 (continued)

H ABGTS B 3.7.12 BASES ACTIONS (continued)

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.

As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. The system must be operated for > 10 continuous hours with the heaters energized.

The 31-day Frequency is based on the known reliability of the equipment and the two train redundancy available.

SR 3.7.12.2 This SR verifies that the required ABGTS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 7).The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ABGTS train starts and operates on an actual or simulated actuation signal. The 18-month Frequency is consistent with Reference 7.(continued)

Watts Bar -Unit 2 B 3.7-66 (developmental)

H ABGTS B 3.7.12 BASES SURVEILLANCE REQUIREMENTS SR 3.7.12.4 (continued)

This SR verifies the integrity of the ABSCE. The ability of the ABSCE to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the ABGTS. During the post accident mode of operation, the ABGTS is designed to maintain a slight negative pressure in the ABSCE, to prevent unfiltered LEAKAGE. The ABGTS is designed to maintain a negative pressure between -0.25 inches water gauge and -0.5 inches water gauge (value does not account for instrument error) with respect to atmospheric pressure at a nominal flow rate > 9300 cfm and < 9900 cfm. The Frequency of 18 months is consistent with the guidance provided in NUREG-0800, Section 6.5.1 (Ref. 8).An 18-month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 7.REFERENCES

1. Watts Bar FSAR, Section 6.5.1, "Engineered Safety Feature (ESF)Filter Systems." 2. Watts Bar FSAR, Section 9.4.2, "Fuel Handling Area Ventilation System." 3. Watts Bar FSAR, Section 15.0, "Accident Analysis." 4. Watts Bar FSAR, Section 6.2.3, "Secondary Containment Functional Design." 5. Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors." 6. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance." 7. Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants." 8. NUREG-0800, Section 6.5.1, "Standard Review Plan," Rev. 2, "ESF Atmosphere Cleanup System," July 1981.(continued)

Watts Bar -Unit 2 B 3.7-67 (developmental)

H Fuel Storage Pool Water Level B 3.7.13 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.

The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity.

The water also provides shielding during the movement of spent fuel.A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Section 15.5.6 (Ref. 3).APPLICABLE SAFETY ANALYSES The minimum water level in the fuel storage pool meets the assumptions of the fuel handling accident described in Regulatory Guide 1.183 Rev. 6.The Total effective Dose equivalent (TEDE) for control room occupants, individuals at the exclusion area boundary, and individuals within the low population zone will remain with 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref 6) for a fuel handling accident.According to Reference 3, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident.

With 23 ft of water, the assumptions of Reference 6 can be used directly.

In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks; however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small non-conservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.The fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement.(continued)

Watts Bar -Unit 2 (developmental)

B 3.7-68 H I Fuel Storage Pool Water Level B 3.7.13 BASES (continued)

LCO The fuel storage pool water level is required to be >_ 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3). As such, it is the minimum required for fuel storage and movement within the fuel storage pool.APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the fuel storage pool, since the potential for a release of fission products exists.ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.

When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended.

This action effectively precludes the occurrence of a fuel handling accident.

This does not preclude movement of a fuel assembly to a safe position.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.

The water level in the fuel storage pool must be checked periodically.

The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.(continued)

Watts Bar -Unit 2 B 3.7-69 (developmental)

A Fuel Storage Pool Water Level B 3.7.13 BASES (continued)

REFERENCES

1. Watts Bar FSAR, Section 9.1.2, "Spent Fuel Storage." 2. Watts Bar FSAR, Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System." 3. Watts Bar FSAR, Section 15.5.6, "Fuel Handling Accident." 4. Deleted 5. Deleted 6. Regulatory Guide 1.183, "Alternate Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors", July 2000.7. Title 10, Code of Federal Regulations 50.67, "Accident Source Term." (continued)

Watts Bar -Unit 2 (developmental)

B 3.7-70 A Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange.During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool.Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 2 and 8). Sufficient iodine activity would be retained to limit offsite doses from the accident to the limits defined in 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 8).APPLICABLE SAFETY ANALYSES During movement of irradiated fuel assemblies, the water level in the refueling canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment.

A minimum water level of 23 ft (Regulatory Position 2 of Appendix B to Regulatory Guide 1.183 (Ref. 8)) allows an overall iodine decontamination factor of 200 to be used in the accident analysis.

This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 8% of the 1-131, 10% of the Kr-85, and 5% of the other noble gases and iodines from the total fission product inventory in accordance with Regulatory Position 3.1 of Regulatory Guide 1.183 (Ref. 8).The fuel handling accident analysis inside containment is described in Reference

2. With a minimum water level of 23 ft in conjunction with a minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 7 and 8 ).Refueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.

Watts Bar -Unit 2 (developmental)

B 3.9-20 (continued)

HI Refueling Cavity Water Level B 3.9.7 BASES (continued)

LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits, as provided by the guidance of Reference 3.APPLICABILITY LCO 3.9.7 is applicable when moving irradiated fuel assemblies within containment.

The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis.

If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.

Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.13, "Fuel Storage Pool Water Level." ACTIONS A._1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.A.2 In addition to immediately suspending movement of irradiated fuel, actions to restore refueling cavity water level must be initiated immediately.

SURVEILLANCE REQUIREMENTS SR 3.9.7.1 Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.(continued)

Watts Bar -Unit 2 (developmental)

B 3.9-21 A Refueling Cavity Water Level B 3.9.7 BASES (continued)

REFERENCES

1. Deleted 2. Watts Bar FSAR, Section 15.5.6, "Fuel Handling Accident." 3. NUREG-0800, "Standard Review Plan," Section 15.7.4,"Radiological Consequences of Fuel-Handling Accidents," U.S. Nuclear Regulatory Commission.

4. Title 10, Code of Federal Regulations, Part 20.1201 (a), (a)(1), and (2)(2), "Occupational Dose Limits for Adults." 5. Deleted 6. Deleted 7. Title 10, Code of Federal Regulations 50.67, "Accident Source Term." 8. Regulatory Guide 1.183, "Alternate Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors", July 2000.Watts Bar -Unit 2 (developmental)

B 3.9-22 (continued)

A Enclosure 10 WBN Unit 2 -Revised Technical Requirements Manual Section 3.9.1 E10-1 Der~ay ke TR 3.9.TR an REFErING1 Ihl/ OPPRATIONS TR 3.0.1 Decay Time ADPPI CARILITYI ThA reactor haIh-al ubp warifinca-far -100 hourc.Durin, moFm'o't of irrdiated fuel n thA M-ctr O ovGo-.CORPT4ONONTIAF PGNQIDlOl N REQUIRED ACTION PLETION TiME A, Reiactor Auharitical fni A-4 Suspend all operatione imm~ediate!y 4 400 hGUwS. inYoIY!g moemeent of irradiated fuel in the TECHNICALI SUR"1 Erll I hlQCE REQUIREMENTS SURVEMLLANCE FREQEN TSRW&-494 Verify the reactor has boon subG~rical for > 100 hoUrc Por to IeRet cl by confirm~ing the date and- thme of cbrtait.Rradated fuel *n the reactor v.essel Wfatte Bar Unit 2 Tec-hnica'a Requ*remnents (developmental)A A I Containment Vent Isolation Instrumentation

3.3.6 Watts

Bar -Unit 2 (developmental) 3.3-53 H Decay Time 3.9.8 Watts Bar -Unit 2 (developmental) 3.9-14 H