Text

Proposed Tech Specs Re RHR Sys Interlock Installation
	ML20080J285
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Site:	Browns Ferry
Issue date:	02/23/1995
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
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ML20080J280	List: ML20080J277; ML20080J285;
References
	NUDOCS 9502270238
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{{#Wiki_filter:- ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, and 3 REVISION TO TECENICAL SPECIFICATION (TS) BASES (TS-348) RESIDUAL HEAT REMOVAL SYSTEN INTERLOCK INSTALLATION I. REASON FOR NOT REVISING BASES TO DISCUSS INSTALLATION OF INTERLOCK 8 In Reference 1, to support a proposed amendment to the Units 1, 2, an.d 3 TS, TVA committed to install electrical interlocks in the Unit 1 and Unit 3 Residual Heat Removal System (RHRS) prior to each unit's restart. The-interlock would be installed between the RHRS shutdown cooling pump suction valves and the corresponding suppression pool return line valves. In Reference 2, NRC requested that TVA revise the Bases for Units 1 and 3 to document the need to install these interlocks. TVA considers that adding the requested information to the Units 1 and 3 Bases is not necessary to ensure that these interlocks are installed. First, TVA is tracking this commitment through its commitment tracking program. This ensures that the commitment will be properly dispositioned. Second, as a test case, TVA evaluated deletion of this commitment as part of its commitment management program.3 TVA's evaluation determined that deleting the commitment would negatively impact the ability of a system, structure, or component to perform its safety function. Therefore, the commitment will remain in effect. Third, for restart of Units 1 and 3, TVA will employ the system plant operability checklist (SPOC) process used successfully during Unit 2 restart. SPOC is a systematic method to ensure that items affecting system operability, l such as commitments, are resolved (completed) prior to declaring the system operable. Since this commitment is tied to the RHRS, TVA cannot declare the RHRS operable until this commitment is completed. Accordingly, TVA considers that existing BFN programs provide adequate assurance that this commitment will be ITVA is participating in an industry pilot program on commitment management. This pmgram allows TVA to revise or delete commitments which have no safety significance. 9502270238 950223 PDR ADOCK 05000259 P PDR

f.: .e. o r properly dispositioned, and the interlocks installed prior to each respective unit's fuel load. Furthermore, adding a commitment to the bases would necessitate two future bases changes to delete these commitments after completion. II. REFERENCES 1. TVA letter.to.NRC dated January 21, 1994, Additional Information Regarding Technical Specification (TS) TS-328 -] 2. NRC letter to TVA dated April 19, 1994, Issuance of.TS Amendments for BFN Units 1, 2, and 3 Revising RHRS Operability Requirements (TS-328) (TAC Nos. M85255, M05256, and M85257) .i 1 l i El-2

'-bn: ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, and 3 L REVISION TO TECNNICAL SPECIFICATION (TS) BASES (T8-348) DESCRIPTION OF AND' REASON FOR CHANGES I. DESCRIPTIOhi OF THE CHANGES A. Bases Section 2.1 - TVA is revising this section of the Units 1, 2, and 3 Bases to make editorial changes that delete unnecessary information. These changes are provided below. 1. Units 1, 2, and 3 page 1.1/2.1-11, delete the last sentence in the first paragraph (the sentence discussing that 3293 MWt is the licensed maximum 1 power level). 2. Unit 1 page 1.1/2.1-16, Bases Section 2.1.L, delete References 3 and 4. B. Bases Section 3.2 - TVA is revising this section of the Units 1, 2, and 3 Bases to make a minor editorial 1-change (Item 1) and to clarify the function of the main steam line space high temperature detection instrumentation (Item 2). These changes are listed below. 1. Unit 1 page 3.2/4.2-66, the fifth paragraph, Unit 2 page 3.2/4.2-66, the fourth full paragraph, and Unit 3 page 3.2/4.2-65, the fifth paragraph, provide a discussion of the high steam flow instrumentation. The second sentence of each paragraph currently reads: The primary function of the instrumentation is to detect a break in the main steam line. This sentence is revised to read as follows: The primary function of the hiqh steam flow-instrumentation is to detect a areak in the main steam line. 2. Unit 1 page 3.2/4.2-66, starting with the last paragraph and carrying over to page 3.2/4.2-67; Unit 2 page 3.2/4.2-66, starting with the fifth full paragraph and carrying over to page 3.2/4.2-67; and Unit 3 page 3.2/4.2-65, starting with the

Lg g,.; +c '*i o' last-paragraph-and carrying over to page,3.2/4.2-66, currently read (may be more than-one paragraph): 9 . Temperature monitoringJinstrumentation is .provided'in the main steam line tunnel.to n detect leaks in-these areas.-: Trips are. provided.on this-instrumentation and.when t exceeded, cause closure of isolation. -valves. The setting of 200*F for the main steam line tunnel:detectorEis low enough to i detect leaks of=the order-of 15 gpm; thus,. it is capable of covering the entire spectrum of breaks. For large breaks,.the high steam flow instrumentation is a backup- .] to the temperature. instrumentation. In the o . event of a loss of. This text is revised to read as follows: Temperature monitoring instrumentation is i provided in the main steam'line tunnel to detect leaks or small breaks in the main steam lines. The trip setting of'200*F.for. the main steam line-tunnel detector is low enough to provide early indication of a-steam line break.. Exceeding the trip. setting causes closure of isolation valves. For large breaks, the high steam tunnel temperatura detection instrumentation is a backup _to'the high steam flow instrumentation. In the event of a loss of. c. ' Bases section 3.5.'F - TVA is revising.this'section of the Units 1, 2, and 3 Bases to clarify the function of the Reactor Core Isolation Cooling System c (RCICS). These changes are listed below. 1. Unit :L page 3.5/4.5-31, Unit 2 page 3.5/4.5-29, and Unit 3 page 3.5/4.5-32, the first sentence 1 currently reads:- The RCICS functions to provide core cooling and makeup water to the. reactor vessel during shutdown and isolation from the main heat sink and for certain pipe break accidents. E2-2

~ f 3 The revised sentence reads as follows: The RCICS functions to provide makeup water to the reactor vessel during shutdown and isolation from the main heat sink to supplement or replace the normal makeup

sources, i

2. Unit 1 page-3.5/4.5-31, Unit 2 page 3.5/4.5-29, and Unit 3 page 3.5/4.5-32, the third sentence currently reads: ll Below-150 psig, RCICS is not required to be OPERABLE since this pressure.is substantially.below that for any events in which RCICS is required to provide core cooling. The revised sentence reads as follows: Below 150 psig, RCICS is not required to be OPERABLE since this pressure is 4 substantially below that for any events in which RCICS is needed to maintain sufficient coolant to the reactor vessel. D. Bases section 3.10.B - TVA is revising this section of the Units 1, 2, and 3 Bases to make a minor editorial change and to incorporate a discussion of " fueled region." These changes are discussed below. 1. Units 1 and 2 page 3.10/4.10-13, and Unit 3 page 3.10/4.10-12, the first paragraph. reads: The SRMs are provided to monitor the core during periods of: station shutdown and to guide the operator during refueling operations and station startup. Requiring two OPERABLE SRMs (FLCs) one in and one adjacent to any core quadrant where fuel or control rods are being moved' assures adequate monitoring of that quadrant during such alterations. This paragraph is revised to read as follows: The SRMs are provided to monitor the core during periods-of unit shutdown and to guide the operator during refueling operations and unit startup. Requiring two OPERABLE SRMs (FLCs) during CORE E2-3 4 w me-- c

7 4 r 7 . g r. - ~ -ALTERATIONS assures' adequate' monitoring of the fueled region (s) and the core quadrant where CORE ALTERATIONS are being performed. The-i fueled region.is'any set of contiguous (adjacent) control cells which contain one or more fuel assemblies. An SRM is considered to be in the fueled region.when one or more of.the four fuel' assembly locations surrounding the SRM dry tube contain a fuel' assembly. An FLC is considered.to be in:the fueled' region if the FLC is positioned such that it is monitoring-the fuel assemblies.in its associated core quadrant, even if the actual position.of the FLC is outside the fueled region. Additionally, the remainder of tha paragraph. beginning with the third sentence (i.e., "Each-SRM._. ") becomes a new paragraph. II. REASON FOR TEE CEANGES A. Bases section 2.1 - TVA is revising'. Units 1, 2,'and-3 Bases.Section 2.1 (Item 1) to delete unnecessary information discussing the licensed maximum power level, since this information is provided in the operating licenseL(License Condition 2.C.'(1)).- TVA is revising Unit 1 Bases Section 2.1.L (Item ~2) to delete two unnecessary references. .j B. Bases section 3.2 - TVA is revising this section of the-Units 1, 2,'and 3' Bases to make a minor editorial ~ change (Item 1), and to clarify the function of the j main steam line space high temperature' detection instrumentation (Item 2). The. revision'also deletes i inaccurate information and makes the bases consistent with the BFN Updated Final Safety Analysis Report' j (UFSAR) and BFN design documents. H UFSAR Sections 7.3.4.7 and 7.3.4.8 provide a' discussion of the basis for the 200*F main steam line space high temperature trip setting and the function of the main steam line space high temperature-detection instrumentation.- Specifically, UFSAR Section 7.3.4.7 states that tiu main steam line space high temperature trip setting is set high enough.to avoid spurious isolation but low enough to provide early indication of a steam line break. UFSAR Section 7.3.4.8 states that the-main steam line-space high temperature detection instrumentation is designed to detect leaks of from one percent to ten percent of rated steam flow. E2-4

6 g i > Furthermore, neither the BFN UFSAR nor-BFN design documents support'the existingfbases_ statement that'the main steam line space high temperature instrumentation can detect leaks as small as 15 gallons per minute (gpm). e For'large breaks, the main steam line space high temperature detection instrumentation senses the break and initiates main steam isolation valve (MSIV)-closure within_ ten. seconds, while the-high flow sensors will initiate MSIV closure within 0.5 seconds. Additionally, for main steam'line break accidents, BFN's accident analysis (BFN UFSAR Chapter 14)' assumes the MSIVs will close on high flow. The accident-analysis does not assume MSIV closure _ on: main. steam line space high temperature. Accordingly, for large breaks, the high temperature instrumentation.is a backup to the high flow instrumentation. C. Bases Section 3.5.F - TVA is_ revising this section of the Units 1, 2, and 3 Bases to clarify the function of-the Reactor Core Isolation Cooling System-(RCICS). This revision also makes the bases consistent with the BFN UFSAR. 1 UFSAR Sections 1.6.1.3.6, 4.7.1, and 7.3.4.1 discuss 'I the function of the RCICS, and the systems needed for post-accident mitigation. Specifically, UFSAR Sections 1.6.1.3.6 and 4.7.1 state that the function of the RCICS is to provide makeup water to the reactor vessel to supplement or replace the normal makeup sources. UFSAR Section 7.3.4.1 states that, while the RCICS is expected.to operate for post-accident mitigation, the j RCICS does not provide'any accident mitigation or safety-related function. D. Bases Section 3.10.B - TVA-is revising this section_of the Units 1,-2,-and 3 Bases in response to an NRC request.. Specifically, in Reference-1, which approved amendments to the BFN Units 1, 2, and 3 TS, NRC requested that TVA revise the bases to incorporate a discussion of " fueled region." NRC also requested that' the discussion'of " fueled region" be consistent with the information TVA provided to NRC in Reference ~2. The revised bases are consistent with the definition of " fueled region" found in Reference 2.- The revised-bases are also consistent with the discussion of- " fueled region" found in NRC's safety evaluation of the TS change (Reference 1). E2-5

.4 l IV. REFERENCES 1. NRC letter to TVA dated April 9, 1993, Issuance-of TS Amendments for BFN Units 1, 2, and 3 Regarding Refueling Interlocks and Core Monitoring (TS-324) (TAC Nos. M84699, M84700, and M84701) l l 2. TVA letter to'NRC dated March 31, 1993,1 Clarification of " Fueled. Region" in TS-324 i 1 1 i l E2-6

' ENCLOSURE 3 TENNESSEE VALLEY. AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFW) UNITS 1, 2, AND 3 REVISION TO TECHNICAL SPECIFICATION (TS) BASES (TS-348) NARKED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2 Unit 3 1.1/2.1-11 1.1/2.1-11 1.1/2.1-11 1.3/2.1-16 3.2/4.2-66 3.2/4.2-65 3.2/4.2-66 3.2/4.2-67 3.2/4.2-66 3.2/4.2-67 3.5/4.5-29 3.5/4.5-32 3.5/4.5-31 3.10/4.10-13 3.10/4.10-12 3.10/4.10-13 II. MARKED PAGES See attached.

2.1 B&831: 4IMITIEG SAFETY SYSTEM SETTINGS RELATED TO FtJEL CLADDING MAY201993 The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the marimsa thermal power of 3293 MWt. The' analyses were based upon plant operation in accordance with Reference 1. I f' itis, 2000 r?; i: ;i: li:r- :d ---i__ ; c;; 1;-;l f:: ::f ::--- in;; inin ris; n.i;, cd ;ti; repre;e.;e _ ;;;;fy ;;n ; ;; n ? ich ;tell r.n t; i;;.;ir. gly s;;;;;;. The transient analyses performed for each reload are described in Reference 2. Models and model conservatisas are also described in this reference. ) i .i a AMENDMENT NO. I g 7 8$,1 1.1/2.1-11

t b c' 2.1 BASES (Ctat'd) ' F. (Deleted) G. & H. Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram The low pressure isolation of,the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825.psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by.the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram j assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.. With the scrans set at 10 percent of valve closure, neutron flux does not increase. I.J.& K. Reactor Low Water Level Setpoint for Initiation of HPCI and RCIC Closing Main Steam Isolation Valves, and Starting LPCI and Core Spray Pumps. These systems maintain adequate coolant inventory and provide core .i cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure. L. References 1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant. Unit 1 (applicable cycle-specific document). 2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version). w 3. lifi tion the -Dimen nal C Tran nt Mod for Bo ng Wa r Rea or," 241 P, Oct er 19 4. tte rom H. Bu hols ) to S. Che (NRC) "Respo NRC quest in: In tio On OD Comput Model,' ) S tembe , 19 BF?f 1.1/2.1-16 AMENDMENT RE I g 7 4 4 . e

V ~q: ;s; . e-3.2 B331 (Cont'd) MW i 91994 J The low reactor water level instrumentation that is-set to trip when. reactor water level is 378 inches above; vessel zero (Table 3.2.5). initiates the LPCI, Core Spray Pumps, contributes to ADS' initiation, and starts.the' diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate. CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to .the complete circumferential break of a 28-inch recirculation line.and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the' ) water level instrianentation and, in addition to. initiating CSCS,. it causes isolation of Groups 2 and 8 isolation valves.. For.the breakt discussed above, tais instrumentation will-initiate CSCS operaties at about the same time as the low water level instrumentation; thus, the results given above are applicable here also. ) ADS provides for aute.satic nuclear steam system depressurization, if .1 I needed, for small breaks 1n the naclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure staan to the suppression pool. ADS initiates when the following conditions exist:. low reactor water-level permissive (level 3), low reactor water level (level 1), high-drywell pressure or the ADS high drywell pressure bypass timer timed out, a and the ADS timer timed out. In addition, at laast one RHR pump or two core spray pumps must be running.. The ADS high drywell pressure bypass timer is addad to meet the requirements of NUtsG 0737, Item II.K.3.18. This timer will bypass the high drywell pressure permissive after a sustained 2ow water level. The worst case condition is a main steam line break outside primary containment with EPCI inoperable. With the ADS high drywell pressure bypass timer analytical limit of 360 seconds, a: Peak Cladding Temperature (PCT) of 1500*F will not be exceeded for the worst case event. This temperature is well below the limiting PCT of 2200'F.. p .s [ Venturia are provided in the main steam lines as a means t.f measuring [A steam flow and also limiting the loss of mass inventory from the vessel j n pt during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the [g$) worst case accid mt, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass'invantory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. ITemperaturemonitoring)instrumentationisprovidedinthemainsteauline tunnel to detect leaks [-- --:x::f, _ _ _: : ::

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a lY y,00 3b 3.2 BASES (Cont'd) F 65 SEP 2 71994 A A Y f n The[setti f 200*F for the main steam line tunnel detector is low I enough to _::::: 1: " f it: ::d:: ;f 1;,,,_., 2;;, it i;._,21. ;; M N /, 44ew] instrumentation is a backap to the :--;:::::::$1on system, radian

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. For large breaks. the hiah stamm ' instrumentation. In i the event of a loss of the reactor building ventila 85 gN## heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an k unnecessary main steam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation. Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig. The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be l OPERABLE. High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery is prevented and fission product release is within limits. The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" H O for high flow and 2 200*F for temperature are based on the same criteria as the HPCI. High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated. The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed. AMENDMENT NO. 212 BFN 3.2/4.2-67 Unit 1

~)Q ' 1 INSERT "A" FOR BFN UNIT'1 BASES PAGE 3.2/4.2-67 l provide;early indication of a steam-line break. Exceeding the t trip setting causes closure of isolation valves. f - l I c. s ) i I I e ? 'h P t i .s t k e I r i i I t i I [ l .1

3.5 DASES (Cont'd) 3.5.F Reactor Core Isolation Cooline System (RCICS) IhN g F## The RCICS functions to provide :::: :::lin;; nd makeup water to the fact reactor vessel during shutdown and isolation from the main heat sinkjend, j psal f:

t in pi;: ir:25 w id =. The RCICS provides its design flow f68 between 150 psig and 1120 psig reactor pressure.

Below 150 psig, RCICS Ma N _is not required to be OPERABLE since this pressure is substantin11y $$WE'I' below that for any events in which RCICS is]aequ M :d tr ;::-~'t _ u l' ; RCICS will continue to operate below 150 psig at reduced flow until it automatically isolates at greater than or equal to 50 psig reactor steam pressure. 150 psig is also below the shutoff head of the CSS and RHRS, thus, considerable overlap exists with the cooling systems ede/ /8 that provide core cooling at low reactor pressure. The minimum required pa/wfe/a NPSH for RCIC is 20 feet. There is adequate elevation head between the I ggefem[ suppression pool and the RCIC pump, such that the required NPSH is ggi available with a suppression pool temperature up to 140*F with no I containment back pressure. fbt Mbf Ve $gd, The ADS, CSS, and RHRS (LPCI) must be OPERABLE when starting up from a COLD CONDITION. Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RERS pumps so they will inject water into the vessel if required. Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is.provided for plant operating flexibility. With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required. 3.5.G Aut - tic Depressurization System (ADS) The ADS consists of six of the thirteen relief valves. It is designed to provide depressurization of the reactor coolant system during a small break loss of coolant accident (LOCA) if HPCI fails or is unable to i maintain the required water level in the reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. Specification 3.5.G applies only to the automatic feature of the pressure relief system. Specification 3.6.D specifies the requirements for the pressure relief j function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function. BFN 3.5/4.5-31 MENDMENT N0. 2 0 5 Unit 1

~ 1 f: c o -3.10 A&SF.E (Cont'd) APR 0 9 3.10.A (Cont'd) ir 1. Refueling interlocks (BFNF FSAR Subsection 7.6) { P' B. Core Manitaring smfrY The SMs are provided to monitor the core during riods of seat 6en shutdown and to guide the operator during refueli perations'and-a '* P a*=*='---*****'** *"= "'" **' J '2"* ) N u - .: n.: 2 :___:. z 1:. m. = r - 3_[_F enBach SRM (FM) is not required to readh 3'cys stil- ~ ~ f alter four fuel assemblies have been loaded adjacent to the SM (FIC) l / if no other fuel assemblies are in the associated core quadrant. These four locations are adjacent to the SRM dry tube. When utilising FIEs, MN the F12s will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint. With four fuel assemblies or fever loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical. Under the special condition of removing the full core with all control rods inserted and electrically disassed, it is permissible to allow SEM count rate to decrease below three counts per second. All i sl moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded. Since there will'be no reactivity additions during this i period, the low number of counts will not present a hasard. When j sufficient fuel has been removed to the spent fuel storage pool to drop 1 the SRM count rate below 3 cps, SRMs will no longer be required to be { OPERABLE. Requiring the SRMs to be functiond.ly tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel I removal. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the i count rate diminishes due to fuel removal. Control rods in cells from which all fuel has been removed and which are outside the periphery of ~ the then existing fuel matrix may be armed electrically and moved for' 1 maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed. REFERENCES 1. Neutron Monitoring System (BFNF FSAR Subsection 7.5) 2. Morgan, W. R., "In-Core Neutron Monitoring System for General f Electric Boiling Water Reactors," General Electric Compemy Atomic Power Equipment Department, November 1968, revised April 1969 (AFED-5706) BFN 3.10/4.10 13 D MNNO.Ig( I Unit 1

l i INSERT "B" FOR BFN UNIT 1 BASE 8 PAGE 3.10/4.10-13 during CORE ALTERATIONS assures adequate monitoring of the fueled region (s) and the core quadrant where CORE ALTERATIONS are being performed. The fueled region is any set of contiguous (adjacent) control cells which contain one or more fuel assemblies. An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the SRM dry tube contain a fuel assembly. An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region. I )

.. ~. 2.1-B&gggs LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INr-Tr MAY 2 01993 The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the maxistan thermal power of 3293 MWt. The analyses were based upon plant operation in accordance with Reference 1." rf" ti r, 2222 T i: the li:-- :f -- ' __ ;;r;; ... ;l !;; ::i 0;_- .";..-; P::I::r "I t - ' t,

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tri rtstr ;:- :r ^'rh

'- rrt 5: M-- ' '7 -- ::f:f. The transient analyses performed for each reload are described in Reference 2. Models and model conservatisms are also described in this reference. N I 1 i i l l BFN 1.1/2.1-11 AMENDMENT NO. 214 j Unit 2

~ o 3.2 BASES (Ctat'd> m09m initiates'the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling'can be accomplished _and the ~ guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instruentation; thus, the results given above are applicable here also. ADS provides for automatic nuclear steam system depressurization; if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool. ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level 1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, and the ADS timer timed out. In addition, at least one RER pump or two core spray pumps must be running. The ADS high drywell pressure bypass timer is added to meet the t requirasents of IRJREC 0737, Item II.K.3.18. This timer will bypass the high drywell pressure permissive after a sustained low water level. The worst case condition is a main steam line break outside primary containment with HPCI inoperable. With the ADS high drywell pressure bypass timer analytical' limit of 360 seconds, a Peak Cladding' Temperature (PCT) of 1500*F will not be exceeded for the worst case event. This temperature is well below the limiting PCT of 2200*F. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel Mb during a steam line break accident. The primary function of the ginstrumentation is to detect a break in the main steam line. For the l worst case accident, main steam line break outside the drywell, a trip f setting of 140 percent of rated steam flow in conjunction eith the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000'F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Referencf frealts la fje spela WNWIJN M##8+ Section 14.6.5 FSAR. y sesel g Temperaturemonitoringkastrumentationisprovidedinthemainsteamline Mg tunnel to detect leaks; i " x ; n:. 2;"; x. ; x '_f;f.:. C ; g y +r 9 1 Ng The[e of 200*F for the main steam line tunnel detector is low OS f = =: 1: ' :: x:= :: n _ a -tm, it i: ---21: ::

x' z: ir: ;::. _ cf i:- '.

For large breaks, the high steem i 2NSfetT %* " BFN pM

d, 3.2/4.2-66 Al g gg g 7 gg, y g Unit 2

$f \\ i

e -.i g =IN8ERT "C" FOR BFN UNIT 2' BASES PAGE 3.2/4.2-66 provide early indication of a steam line break. Exceeding the trip setting causes closure.of isolation valves. l I i s a s \\

p MCJ 3.2 RASES (Cont'd) 'h g Al l gg7 .. f,gse/ Gew instrumentation is a backup to the temposeemse instrumentation. In the event of a loss of the reactor buil ing ventilation systen, radi t g /- heating in the. vicinity of the main steam lines raises the ambient I b ** temperature above 200'F. The temperature. increases can cause an unnecessary main steam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary ~ containment leak rate test or make repairs necessary to regain normal ventilation. Pressure instrumentation is provided to close the main steam isole. tion valves in RUN Mode when the main steam line pressure drops below 825 pais. The HPCI high flow and temperature instrumentation.are provided to detect a break in tne HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE. I Eigh temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. Each trip system consists of two elements.. Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area. The RCIC high flow and high area temperature sensing instrument channels are arranged in the same sanner as the HPCI system. The HPCI high steam flow trip setting of 90 paid and the RCIC high steam flow trip setting of 450" H O have been selected such that the trip 2 setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits. The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature - excursions in the vicinity of the steam supply piping. Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable time period to prevent j core uncovery and maintain fission product releases within 10 CFE 100 Jimits. Bish temperature at the Reactor Water Cleanup (RWCU) System in the main steam valve vault, RWCU pump room 2A, RWCU p o p room 25, RWCU heat awhanger room or in the space near the pipe trench containing EWCU piping could indicate a break in the cleanup system. When high temperature 1 occurs, the cleanup system is isolated. BFN 3.2/4.2-67 Unit 2 l

3.5 BASES (C:nt'd) 3.5.F Reactor Core Isolation Cooline System (RCICS) h/penY The RCICS functions to provide :::: :::lin; :nd makeup water to the reactor // I vesselduringshutdownandisolationfromthemainheatsinkfene-de, j of Ygg f / .....ir. pip; i ::h ;;;id::::. The RCICS provides its design flow between fM 88 W 150 psig and 1120 psig reactor pressure. Below 150 psig, RCICS is not paEfY required to be OPERABLE since this pressure is substantially below that for f,0gr48 any events in which RCICS is]::;;ir:d t: pr: tid: :::: :::lin;. RCICS will continue to operate below 150 psig at reduced flow until it automatically isolates at greater than or equal to 50 psig reactor steam pressure. 150 gg/ fe l psig is also below the shutoff head of the CSS and RHRS, thus, considerable Mg, / overlap exists with the cooling systems that provide core cooling at low reactor pressure. The minimum required NPSH for RCIC is 20 feet. There is TW k',C'PN j l l adequate elevation head between the suppression pool and the RCIC ptmp, l coo / spy /8 such that the required NPSH is available with a suppression pool g, p/pr temperature up to 140*F with no containment back pressure. U The ADS, CSS, and RHRS (LPCI) must be OPERABLE when starting up from a COLD CONDITION. Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required. Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility. With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required. 3.5.G Automatic Deoressurization System (ADS) The ADS consists of six of the thirteen relief valves. It is designed to provide depressurization of the reactor coolant system during a small break loss of coolant accident (LOCA) if HPCI fails or is unable to maintain the required water level in the reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. Specification 3.5.G applies only to the automatic feature of the pressure relief system. Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function. The emergency core cooling system LOCA analyses for small line breaks assumed that four of the six ADS valves were operable. By requiring six BFH 3.5/4.5-29 AMENDMENT N0.19 8 Unit 2

3.10 AMIES (Cont'd) APR 0 9]993 3.10.A (Cont'd) 3 M g RzrERaczS 1. Refueling interlocks (BFNF FSAR Subsection 7.6) M gdI B. core Manienr4a. pp,M The SRMs are provided to monitor the core during pei 'iods of eeutdem shutdownandtoguidetheoperatorduringrefueling{),operationsand Requiring two OPERABLE SRMs (FLCs__.__,__,__'_m.__ gg,deteessostartup. ,,_____________m_4 m___ ..' _.....___L __1__i, K_.11 J JiE_lu- - -..'. _.... '___~~ -,---- ebeese46ees.syEach SRM (FLC) is not required to read 1 3 cps until g ' atter tour ruel assemblies have been loaded adjacent to the SRM (FLC). if no other fuel assemblies are in the associated core quadrant. These Y four locations are adjacent to the SRM dry tube. When utilising FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint. With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical. Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allov SRM count rate to decrease below three counts per second. All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded. Since there will be no reactivity additions during this period, the low number of counts will not present a hasard.' When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE. Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal. The once per-12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due-to fuel removal. Control rods in cells from which all fuel has been removed and which are outside the periphery of the then existing fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed. RzFERENCES 1. Neutron Monitoring System.(BFNF FSAR Subsection 7.5) 2. Morgan, W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706) N E IO.209 BFN 3.10/4.10-13 Unit 2

-o o INSERT "D" FOR BFN UNIT 2 BASES PAGE 3.10/4.10-13 during CORE ALTERATIONS assures adequate monitoring of the fueled region (s) and the core quadrant where CORE ALTERATIONS are being performed. The fueled region is any set of contiguous (adjacent) control cells which contain one or more fuel assemblies. An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the SRM dry tube contain a fuel assembly. An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region.

2.1 A4131

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY 2 0 1993 l The abnormal operational transients applicable.to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned j operating conditions up to the maximum thermal power of 3293 MWt. ) The analyses were based upon plant operation in accordance with l Reference 1. Ir "'_tirn, 2202 5?: i "- li c- :d ----i_ _ ;; _ ;. j in ; ::: ::d 5::= T::_; Nr:12:r Pir-t - ' t, - ' -: ::;:: - :: _ ::::f; ::::: ;:r:: -di d. ' ' 1 :: t 5: ' ri--l r ::f:f. i The transient analyses performed for each reload are described in Reference 2. Models and model conservatisms are also described in this reference. I i 3 I l l l l l i BFN 1.1/2.1-11 AMENDMENT NO.170 Unit 3 1

3 i '3.2 R&AES (Cont'd) El 9 $ The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.5) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSC8 operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circunferential break of a 28-inch recirculation line sad with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is s' diverse signal to the water level instrumentation and, in addition to initiating C8CS, it-causes isolation of Groupe 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate C8CS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also. ADS provides for automatic nuclear steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS.uses six of the 13 NERVs to relieve the high pressure steam to the suppression pool. ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level 1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, i and the ADS timer timed out. In addition, at least one RER pump or two core spray pumps must be running. 1 The ADS high drywell pressure bypass timer is added to meet the requirements of NUREG 0737, Item II.K.3.18. This timer will bypass the-high drywell pressure permissive after a sustained low water level. The I worst case condition is a main steam line break outside primary 'I containment with HPCI inoperable. With the ADS high drywell pressure I bypass timer analytical limit of 360 seconds, a Peak Cladding Temperature (PCT) of 1500*F will not be exceeded for the worst case event. This temperature is well below the-limiting PCT of 2200*F. ,l Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel N'. j during a steam line break accident. The primary function'of the. T instrumentation is to detect a break in the main steam line. For the M8" 4l worst case accident, main steam line break outside the drywell, a trip I f/6he setting of 140 percent of rated steam flow in conjunction with the flow I limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below Reference Sectfon 14.6.5 FSAR. 10 CFR 100 guidelines. ar sameM fredAT IN fAe Juee's sNear 8#M. Temperature monitoring Iinstrumentation is provided in the main steam line-tunnel to detect leakay in t ::: :::::. ?:i;

;;rif:f ::

ir:_ _ rt:ti: " e r n:::::e, :x:: ci::::: 25 12:12:i:

17.

BFN 3.2/4.2-65 ANDIDMDfT NO. I 7 8 Unit 3 1

c. b f 1y SEP 2 7 m 3.2 BASES (Cont'd) rA?p,g frif The[etting f 200*F for the main steam line tunnel detector is low _::::: 1: '- :f t: ::f:: :f 15.;;; t, it i

: !"_: 0*

l enough to,i - tir: :;:: _ _ cf i :^. For large breaks, the h' sh steam J W gI

c:': T In gef'#'hTI*instrumentationisabackuptothetemposeeeeepnstrumentat;,on.

the event of a loss of the reactor building ventilation system, radian M8 %[Y#O 4'ys heating in the vicinity of the main steam lines raises the ambient Q temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Pasmesm6am is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation. Pressure instrusentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 peig. The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic-for the high flow is a 1-out-of-2 logic, and all sensors are required to be l i OPERABLE. High temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery is prevented and fission product release is within limits. The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" water for high flow and 200*F for temperature are based on the same criteria as the EPCI. High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated. The instrumentation which initiates CSCS action is arranged in a-dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed. AM M N0.185 BFN 3.2/4.2-66 Unit 3 e

= -) r - INSERT "E" FOR BFN UNIT 3 BASES PAGE 3.2/4.2-66 provide early' indication of-a steam line break. Exceeding the trip-setting causes closure of isolation valves. a e i + 5 b 4 d k 9

3.5 BASES (Cont'd) 3.5.F Reactor Core Isolation Cooline System (RCICS) MW i 91994 Q The RCICS functions to provide errr cr" '7 --f makeup water to the ~ N UM reactor vessel during shutdown and isolation from the main heat sinkh I fM

fff hr ;xti ;ip; i;::S :::id::tc. The RCICS provides its design flow g perseral between 150 psig and 1120 psig reactor pressure.

Below 150 psig, RCICS Mg is not required to be OPERABLE since this pressure is substantially f $##W" l below that for any events in which RCICS is} ::;z :f i: ; :cid: :::: j _ l i..,,. RCICS will continue to operate below 150 psig at reduced flow until it automatically isolates at greater than or equal to 50 psig reactor steam pressure. 150 psig is also below the shutoff head of the ]{ CSS and RHRS, thus, considerable overlap exists with the cooling systems that provide core cooling at low raactor pressure. The minimum required p gftnIM[Mlj j NPSH for RCIC is 20 feet. There iJ adequate elevation head between the T suppression pool and the RCIC purp, such that the required NPSH is }g jGO available with a suppression poci temperature up to 140*F with no containment back pressure.

[M7 M gg/,

The ADS, CSS, and RHRS (LPCI) must be OPERABLE when starting up from a COLD CONDITION. Steam pressure is sufficient at 150 poig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required. Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility. With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required. 3.5.G Automatic Deoressurizatien System (ADS) The ADS consists of six of the thirteen relief valves. It is designed to provide depressurization of the reactor coolant system during a small break loss of coolant accident (LOCA) if HPCI fails or is unable to maintain the required water level in the reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. Specification 3.5.G applies only to the automatic feature of the pressure relief system. Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function. BrN 3.5/4.5-32 AMENDMEU N0. I 7 8 Unit 3 \\

~ Qp 3.10 A&BTJi (Cont'd) g 3.10.A (Cont'd) .p REFERENCES, g/d gA4/4 1. Refueling interlocks (BFNF FSAR Subsection 7.6) pg 3, care Manitor4ne Oneb The SRMs are provided to monitor the core during periods of e4eMas j shutdown and to guide the operator during refueling(operations and I pMetendom.startup. Requiring two OPERABIA 8RMs (F14sgesh

ff: Z: t: x; zz ;z' z t *- : fr:1 :: W A z f: Z: h' ;

_ _;f ::: r:: :f:;zt: n'ter' Of "rt;ztztfr';;J 2 _-.ti:-:. Each SRM (FLC) is not required to read 1 3 cps until g ter four Iuel assemblies have been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.. These four locations are adjacent to the SRM dry tube. When utilising FLCs, j the FLCs will be located such that the required count rate is achieved 1 without exceeding the SRM upscale setpoint. With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the-configuration will not be critical. i Under the special condition of removing the full core with all control. rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second. All. fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded. Since there will be no reactivity additions during this period, the low nueber of counts will not present a hasard. When i sufficient fuel has been removed to the spent fuel storage pool to drog the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE. Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal. The once per 12 hours verification of the SRM count rate end signal-to-coise ratio ensures their continued OPERASILITY sotil the 4 count rate diminishes due to fuel removal. Control rods in cells from which all fuel has been removed may be armed electrically sad moved for maintenance purposes during full core removal, prowided all.roda.that control fuel are fully inserted and electrically disarmed. EEFERENCES j 1. Neutron Monitoring System (BFTt FSAR Subsection 7.5) 2. Morgen. W. R., In-Con Neutron Monitoring System for General Electric Boiling Wter Reactors," General Electric Company, Atomic j Fower Equipmeet Department, November 1968, revised April 1969 ) (AFED-5706) 4 I AMENDMENTR 168 ~ t BFN 3.10/4.10-12 Unit 3

INSERT "F" FOR BFN UNIT 3 BASES PAGE 3.10/4.10-12 during CORE ALTERATIONS assures adequate monitoring of the fueled region (s) and the core quadrant where CORE ALTERATIONS are being performed. The fueled region is any set of contiguous (adjacent) control cells which contain one or more fuel assemblies. An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the SRM dry tube contain a fuel assembly. An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region. i

l:? e ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1, 2, AND 3 REVISION TO TECHNICAL SPECIFICATION (TS) BASES (TS-348) REVISED PAGES I. AFFECTED PAGE LIST Unit 1 Unit 2-Unit 3 1.1/2.1-11 1.1/2.1-11 1.1/2.1-11 1.1/2.1-16 3.2/4.2-66 3.2/4.2-65 3.2/4.2-66 3.2/4.2-67* 3.2/4.2-66 3.2/4.2-67 3.5/4.5-29 3.5/4.5-32 3.5/4.5-31 3.5/4.5-30* 3.5/4.5-33* 3.5/4.5-32* 3.5/4.5-31* 3.5/4.5-34* 3.5/4.5-33* 3.5/4.5-32* 3.5/4.5-35* 3.5/4.5-34* 3.5/4.5-33* 3.5/4.5-36* 3.5/4.5-35* 3.10/4.10-13 3.10/4.10-12 3.10/4.10-13 3.10/4.10-14* 3.10/4.10-13* 3.10/4.10-14*' 3.10/4.10-14* 3.10/4.10-15* Spillover pages II. REVISED PAGE8 See attached.

2.1" BMEls LIMITIltG SAFETY SYSTDI SETTINGS RELATED TO FUEL ctAnDINC INTEGRITY The' abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the maximum thermal power of 3293 MWt.. The analyses were based upon plant operation in accordance with Reference 1. .d The transient analyses performed for each reload are described in -Reference 2. Models and model conservatisms are also described in this reference. 2 t BFN 1.1/2.1-11 Unit 1 f

. ~. ) ?! I.f-3;..- c ~ i '2.1' B&BA'(Cont d) (F.' (Deleted) LG. & H.: Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram I The low pressure isolation of the main' steam lines at 825'psis was' l provided to protect against' rapid reactor depressurization.and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam-line isolation valves are closed, to provide for reactor shutdown'so that high power operation-at. low' reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. -Operation l of the' reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of' neutron flux scram protection'over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrans set at 10 percent of valve closure, neutron flux does not increase. J I.J.& K. Reactor Lov Water Level Setnoint for Initiation of HPCI and RCIC Closina Main Steam Isolation Valves. and Startina LPCI and Core Sorav Pumns. These systems maintain adequate _ coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14' of the FSAR demonstrate that these conditions result in adequate-safety margins for both the' fuel and the system pressure. L. References i 1. Supplemental Reload Licensing Report of Browns Ferry Nuclear ) Plant, Unit 1 (applicable cycle-specific document). 2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and ) NEDE-24011-P-A-US (latest approved version). I ) BFN 1.1/2.1-16 Unit 1

O 4 3.2 B&EEE (Cont'd) The low reactor water level instrumentation that is set to trip when reactor water level.is 378 inches above vessel zero (Table 3.2.B) -initiates the LPCI, Core Spray Pumps,. contributes to ADS initiation, and starts the diesel generators. These trip setting levels were chosen to-be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus, the results given above are applicable here also. ADS provides for automatic nuclear steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool. ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level 1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, and the ADS timer timed out. In addition, at least one RHR pump or two core spray pumps must be running. The ADS high drywell pressure bypass timer is added to meet the ] requirements of NUREG 0737, Item II.K.3.18. This timer will bypass the high drywell pressure permissive after a sustained low water level. The worst case condition is a main steam line break outside primary containment with HPCI inoperable. With the ADS high drywell pressure bypass timer analytical limit of 360 seconds, a Peak Cladding Temperature (PCT) of 1500*F will not be exceeded for the worst case event. This temperature is well below the limiting PCT of 2200*F. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the high steam flow instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity.to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. BFN 3.2/4.2-66 Unit 1

,',U D s '3.2 B&AE1 (Crnti'd), Temperature monitoring instrumentation is provided in the main steam line tunnel to' detect leaks or small breaks in-the main steam lines. The trip l setting of 200*F for the main steam line tunnel detector is low enough to -provide early indication of a steam line break. Exceeding the trip -setting.causes closure of isolation valves. For large breaks, the high. steam tunne1' temperature detection instrumentation is a backup to the high' steam flow instrumentation. In the event of a loss of the reactor building ventilation system,- radiant heating in.the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Permission is J provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the ' secondary - containment leak rate test or make repairs necessary to regain normal ventilation. Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig. The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this ins *.rumentation results in actuation of HPCI. isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE. High temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system. c 'The HPCI trip settings of 90 psi for high flow and 200*F for high l temperature are such that core uncovery is prevented and fission product 1 release is within limits. The RCIC high-flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" H O for high flow and 2 200*F for temperature are based on the same criteria as the HPCI. High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated. The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed. 1 BPN 3.2/4.2-67 1 Unit 1 1

~ t 3b B&&E1 (Cint'd) 3.5.F Reactor Core Isolation Cooline System (RCICS) s The RCICS functions to provide makeup water to the reactor vessel'during i shutdown and isolation from the main heat sink to supplement or replace the normal makeup sources.. The RCICS provides its design flow between 150'psig and 1120 psig reactor. pressure. Below 150 psig, RCICS is not required to be OPERABLE since this pressure is'substantially below that' ~ for any events in which RCICS is needed to maintain sufficient coolant to the reactor vessel. RCICS will continue to operate below'150 pais at reduced flow until it automatically isolates at greater than or equal to 50 pois reactor steam pressure. ' 150 pais is also below the shutoff head of the CSS and RERS, thus, considerable overlap exists with the cooling systems that provide core cooling at low reactor pressure. The minimum . required NPSH for RCIC is 20 feet. There is adequate elevation head ~ between the suppression pool and the RCIC pump, such that the required j NPSH is available with a suppression pool temperature up to 140*F with no containment back pressure. The ADS, CSS, and RHRS (LPCI) must be OPERABLE when starting up from a COLD CONDITION. Steam pressure is sufficient at 150 pais to run the RCIC-turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required. Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility. With the RCICS inoperable, a seven-day period'to return the system to service is justified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with-failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required. 3.5.G Automatic Deoressurization System (ADS) The ADS consists of six of the thirteen relief valves. It is designed to provide depressurization of the reactor coolant system during a small break loss of coolant accident (LOCA) if HPCI fails or is unable to maintain the required water level in the reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. Specification 3.5.G applies only to the automatic feature of the pressure relief system. Specification 3.6.D specifies the requirements for the pressure relief . function of the valves. It is possible for any number of the valves BFN 3.5/4.5-31 Unit 1

M %9 t1 [3.5 34351 JCint'd) l assigned to the ADS to.be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing. their pressure relief function. e The emergency core cooling system LOCA analyses for small line breaks assumed that four of the six ADS valves were OPERABLE. By requiring six valves.to be OPERABLE,-additional conservatism is'provided to account for the possibility of a= single failure in the ADS system. ~ Reactor operation with one of the six ADS valves inoperable is allowed to continue for fourteen days provided the HPCI, core spray, and LPCI systems are.0PERABLE. Operation with more than one ADS valve inoperable is not-acceptable. With one ADS valve known to be incapable of automatic operation, five valves remain OPERABLE to perform the ADS function. This condition is within the analyses for a small break LOCA and the peak clad temperature is well below the 10 CFR 50.46-11mit. Analysis has shown that four valves'are capable of depressurizing the reactor rapidly enough to maintain peak clad temperature within acceptable limits. H. Maintenance of Filled Discharme Pine If the discharge piping of the core spray, LPCI, HPCIS, and RCICS-are not filled, a water hammer can develop in this piping when the pump and/or pumps-are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification-requires the discharge lines to be filled whenever the system is in an OPERABLE condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable-for Technical Specification purposes. The core spray and RHR system discharge piping high point vent is visually checked for water flow once a month and prior to testing to ensure that the lines are filled. The visual checking will avoid starting the core spray or RHR system with a discharge line not filled. In addition to the visual observation and to ensure a filled discharge line other than prior to testins, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level'in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled. When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly. BFN 3.5/4.5-32l Unit 1

-{ 5< //h n, a ~ L3.5. R& eel ~(Cont'd) 3.5.I. Averane Planar Linmar Heat Generation Rate (APtHcs) jf, This specification assures that_the peak cladding temperature ~ following the postulated design basis loss-of-coolant accident will not exceed'the limit specified in the'10 CFR 50, Appendix K. ~ The' peak cladding temperature following a postulated loss-of-coolant-accident is primarily a function of the average heat generation rate "of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly. Since expected local variations in power distribution a within a fuel assembly affect the calculated peak clad temperature by ] less than i 20*F relative to the peak temperature for a typical fuel i s design, the limit on the average linear. heat generation rate is sufficient'to assure that calculated temperatures _are within the 10 CFR 50 Appendix K limit. 3.5.J. Linear Heat Generation Rate (LHGR) This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet' densification is postulated. The LHGR shall be checked daily during reactor operation at l'25 percent power to determine if fuel burnup, or control rod R . movement has caused changes in-power distribution. For LHCR to be a limiting value below 25 percent of rated thermal power, the largest total peaking would have to be greater than approximately 9.7 which is precluded by a considerable margin when employing'any permissible control rod pattern. 3.5.K. Minimum Critical Power Ratio (MCPR). 4 At core thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum. recirculation. pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at.this point,' operating plant experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit. 3.5.L. APRM Setooints The fuel cladding integrity safety limits of Section 2.1 were based on a total peaking factor within design limits (FRP/CMFLPD 1 1.0). The BFN 3.5/4.5-33 l Unit 1 .i l 1

4 L-.-

3.5 ' BASES (Cont'd) i I

APRM instruments must be adjusted to ensure that the core thermal' limits are not exceeded in a degraded situation when entry. conditions are less conservative than design assumptions. 3.5.M. Core Thermal-Hydraulic Stability The. minimum margin to the onset of thermal-hydraulic instability occurs in Region I_of Figure 3.5.M-1. A manually initiated scram upon entry .into this. region is sufficient to preclude core oscillations which could challenge the MCPR safety-limit. Because the probability of thermal-hydraulic' oscillations is lower and. o. the margin to the MCPR safety limit is greater in Region II than in . i Region I of Figure 3.5.M-1,-an immediate scram upon entry into the-. i region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while' exiting Region II (deltying exit for surveillances is undesirable), an immediate manual scram will be initiated if evidence of thermal-hydraulic instability is observed. Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent psak-to-peak or LPRM oscillations. which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated.- During regional oscillations, the safety limit MCPR is not-approached until APRM oscillations are 30 percent peak-to-peak or larger in j magnitude. In addition, periodic upscale or downscale LPRM alarms will i occur before regional oscillations are large enough to threaten the j MCPR safety limit. Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure i that the MCPR safety limit will not be violated in the event that core 4 oscillations initiate while exiting Region II.. Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur. 3.5.N. References

1. " Fuel Densification Effects on General Eltetric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.

1. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

i

3. Communication:

V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974. BFN 3.5/4.5-34 l Unit 1 6 w -<,-v4 -ir -wi-- -v,.

r.

m- --w-i- --v

4 r , a ' 3.5l BASES (Cont'd) o.. 4. ' Generic Reload Fuel Application, Licensing Topical Report,- EEDE-24011-P-A and Addenda. '5. Letter from R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request For.Information On ODYN Computer Model," September 5, 1980. 4.5 -Core

==d Contai====t Coolinn Syst=== Survei11==ce Frean=aales 4 The testing. interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, Judgment + .and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in'the' case of the EPCI,- automatic initiation during power operation would result-in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant j inventory. To increase the availability of the core.and containment cooling system,-the components which make up the system,:1.e., ' l instrumentation, pumps, valves, etc.,'are tested-frequently. The pumps-l and motor operated injection valves are also tested in accordance'with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing of the pumps'and-injection valves in accordance with Specification 1.0.MM is deemed to be adequate. testing of these' systems. Monthly alignment checks of valves. j that are not locked or sealed in position which affect the ability of the ' i systems to perform.their intended safety function are also verified to be in the proper position. Valves which automatically reposition themselves on an initiation signal.are permitted to be in a position other than j normal to facilitate other operational modes of the system. When components and subsystems are c'zt-of-service, overall core and containment cooling reliability is atintained by OPERABILITY of the l remaining redundant equipment. Whenever a CSCS system or loop is made inoperable, the'cther CSCS systems or loops that are required to be OPERABLE shall be connidered OPERABLE if: they are within the required surveillance testing frequency and there-is no reason to suspect they are inoperable. If.the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for.the system or loop shall apply. t Avermae Planar LHGR. LHCR.

==d MCPR The APLUGR, LEGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate. f f 3.5/4.5-35l BFN Unit 1

w. '.3 I 1~

3.10B&gLi-(Cont'd).

) ) REFERINCES 1. Refueling interlocks (BFNP FSAR Subsection 7.6) B. Core Monitorina The SEMs are provided to monitor the core during periods'of unit shutdown and to guide the operator during refueling operations.and unit startup.. Requiring two OPERABLE SRMs (FLCs) during CORE ALTERATIONS assures adequate monitoring of the fueled region (s) and the core-quadrant where CORE ALTERATIONS are being performed ~. The fueled region is any set of contiguous (adjacent)' control cells which contain one or more fuel assemblies. An SEM is considered to be in the fueled region when one or more of the four fuel assembly locations surrounding the' ~j SRM dry tube contain a fuel assembly. An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring i I the fuel assemblies in its associated core quadrant, even if the actual position'of the FLC'is outside the fueled region. -Each SEM (FLC) is not required to read 2. 3 cps until after four fuel assemblies have been loaded adjacent to the SEM (FLC) if no other fuel assemblies are in the associated core quadrant. These four locations-are adjacent to the SRM dry tube.- When utilizing FLCs, the FLCs will be located such that the required count rate is' achieved without exceeding the SRM upscale setpoint. With four fuel assemblies or fewer loaded around each SEN, even with a control rod withdrawn, the 4-configuration will not be critical. Under the special condition of removing the full core with all control ) rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second. All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded. Since there will be no reactivity. additions during this period, the low number of counts will not present a hazard.. When l sufficient fuel has been removed to the spent fuel storage pool to drop j the SRM count rate below 3 cps, SRMs will no longer be required to be j OPERABLE. Requiring the SRMs to be functionally tested prior to fuel -] removal assures that the SRMs will be OPERABLE at the start of fuel removal. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal. Control rods in cells from which all fuel has been removed and which are outside the periphery.of j the then existing fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that j control fuel are fully inserted and electrically disarmed. .j l REFERENCES 1 1. Neutron Monitoring System (BFNP FSAR Subsection 7.5) BFN 3.10/4.10-13 Unit 1

~_ ^ l E 1 y 3.10 BASES (Cont'd)- .j 2. Morgan, W.. R., "In-Core. Neutron Monitoring System for General -l Electric Boiling Water Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 l (APED-5706) C. Soent Fuel Pool Water j -The design of the spent fuel storage pool'provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures' adequate shielding, cooling, and reactivity control of irradiated fuel. 'An analysis has been performed.- l which shows that a water level at or in excess of eight and one-half-. feet over the top of the stored assemblies will provide shielding such' that the maximum calculated-radiological doses do not exceed the limits of 10 CFR 20..The normal water level provides 14-1/2 feet of' additional water shielding.' The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three . days in the absence of additional water input from the condensate storage tanks. All' penetrations of the fuel pool have been installed at such a height that'their presence does not provide a possible. drainage route that could lower the normal water level more than I one-half foot. The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads. 'If the reactor i core is completely unloaded when the pool contains two previous discharge batches, the temperature may increase to greater than 125'F. ] The RHR system supplemental fuel pool cooling mode will-be used under these conditions to maintain the pool temperature to less than 125"F. ? 3.10.D/4.10.D BASES Reactor Buildina Crane The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. The five-ton has both cab and pendant controls. .i A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed. The testing of the various limits and interlocks assures their proper-l 4 operation when the crane is used. 3.10.E/4.10.E 1pfnt Fuel Cask ? The spent fuel cask design incorporates removable lifting trunnions. The visual inspection of the trunnions and fasteners prior to -l BFN 3.10/4.10-14 l Unit 1 6 ..,,, + n -

i.; l3.10 B& GEE (Cont'd) f y attachment;to the cask assures that no visual damage has occurred l iduring prior handling. The trunnions must be properly attached to the cask for lifting of the cask and the visual inspection assures correct installation. 3.10.F Spent Fuel Caak Handlina - Refuelina Floor 3 Although single failure protection-has.been provided in the design of ~ the 125-ton hoist drum shaft, wire: ropes, hook and lower block assembly i h .on the reactor building crane, the limiting of lift height.of a spent. fuel cask controls the amount of energy available in'a dropped cask accident when the cask is over the refueling floor.. { An analysis has been made which shows that the. floor and support members in the area of cask entry into the decontamination facility can 't satisfactorily sustain a dropped cask from a height of three. feet. The yoke safety links provide single failure protection for the hook and lower block assembly and limit cask rotation. Cask rotation is necessary for decontamination and the safety links are removed during' ] i decontamination. 1 4.10 BASES l l A. Refueline Interlocks j Complete functional testing of all required refueling equipment interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed. By loading each hoist with a weight equal to the fuel assembly, positioning the refueling platform, and withdrawing control rods, the' interlocks can be subjected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently -perform its function. B. Core Monitorina Requiring the SRMs to be functionally tested prior to any CORE ALTERATION assures that the SRMs will be OPERABLE at the start of that alteration. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY. i REFERENCES 1. Fuel Pool Cooling and Cleanup System'(BFNP FSAR Subsection 10.5) l 1 2. Spent Fuel Storage (BFNP FSAR Subsection 10.3) BFN 3.10/4.10-15 l Unit 1 i

. o; 2.1 B&gfdg. LIMITING SAFETY SYSTEM SETTINGS DWLATED TO FUEL cf.AnDING INTEGRITY e The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the maximum thermal power of 3293 MWt. The analyses were based upon plant operation in accordance with Reference 1. The transient analyses performed for each reload are described in Reference 2.. Models and model conservatisms are also described in this reference. 4 L ) BFN 1.1/2.1-11 Unit 2 -a a m r--- y

i 3- "V ^3.2 BASES (Cont'd)l 'O = initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and l starts the diesel generators. These triptsetting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate. .CSCS operation so that postaccident cooling can be accomplished-and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet-the above criteria. The high drywell pressure instrumentation.is a diverse signal to the ~ water level instrumentation and,'in addition to initiating CSCS, it causes isolation of Groups.2 and 8 isolation valves. :For the breaks discussed above, this instrumentation will initiate CSCS operation at. 1 about the same time as the low water level instrumentation; thus, the <I d results given above are applicable here'also, ADS provides for automatic nuclear. steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and the CSS can operate to protect the fuel from overheating. ADS uses six of. the 13 MSRVs to relieve the high pressure steam to the suppression pool.. ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level-(level 1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, and the ADS timer timed out. In addition, at least one-RRR pump or two core spray pumps must be running. 1 i The ADS high drywell pressure bypass timer is added to meet the requirements of NUREG 0737, Item II.K.3.18. This timer will bypass the high drywell pressure permissive after a sustained low water level. The i worst case condition is a main steam line break outside primary containment with HPCI inoperable. With the ADS high drywell pressure-bypass timer analytical limit of 360 seconds, a Peak Cladding Temperature (PCT) of 1500*F will not be exceeded for the worst case event. This temperature is well below the limiting PCT of 2200*F. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel. during a steam line break accident. The primary function of the high steam flow instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain 'l below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. .l 1 Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks or small breaks in the main steam lines. The trip l i setting of 200*F for the main steam line tunnel detector is low enough to provide early indication of a steam line break. Exceeding the trip i setting causes closure of isolation valves. For large breaks, the high steam tunnel temperature detection instrumentation is a backup o the l high steam flow instrumentation. l BFN 3.2/4.2-66 Unit 2 i t

I 'l -T. ~,3 4 3.2 B&BES (Cent'd) .In the' event of a loss of the reactor building ventilation system,' radiant ~ heating in the vicinity of the main steam lines raises the ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation. 1 Pressure instrumentation is provided to close the main steam. isolation i valves-in RUN Mode when the main steam line pressure drops below 825 peig. ) The HPCI high flow and temperature instrumentation are provided.to detect i a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves.. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE. l High temperature in the vicinity of the'HPCI equipment is sensed by. four sets of four binetallic temperature switches. The 16 temperature -l switches are arranged in two trip systems with eight temperature switches i in each trip system. Each trip system consists of two elements. Each j channel contains one temperature switch located in the pump room and'three temperature switches located in the torus area. The RCIC high flow and high area temperature sensing instrument channels are arranged in the same manner as the HPCI system. The HPCI high steam flow trip setting of 90 psid and the RCIC high steam ) flow trip setting of 450" H O have been selected such that the trip ] 2 setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product.. releases within 10 CFR 100 limits. .l 1 The HPCI and RCIC steam line space temperature switch trip settings are j high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the steam supply piping. Additionally, these trip settings ensure that the primary containment isolation steam-supply valves isolate a break within an acceptable time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits. High temperature at the Reactor Water Cleanup (RWCU) System in the main-steam valve vault, RWCU pump room 2A, RWCU pump room 2B, RWCU heat exchanger room or in the space near the pipe trench containing RWCU piping ~ could indicate a break in the cleanup system. When high temperature t occurs, the cleanup system is isolated. 1 I i - BFN 3.2/4.2-67l Unit 2 l l -..m ~+-

3.5i RASES (Crnt'd) ~ 3.5.F Reactor Core' Isolation'Coolina System (RCICS) ' The RCICS functionsito provide makeup water'to the reactor vessel during . d shutdown and isolation'from the main heat sink to supplement or replace the- . normal makeup sources. The RCICS provides its design flow between 150 pais and 1120 psig reactor pressure.. Below 150 psig, RCICS is not required'to - be OPERABLE since this pressure is substantially below.that for any. events ~ in which RCICS.is needed to maintain sufficient coolant to the reactor vessel. RCICS will continue to operate below 150 pois at reduced flow 74 until.it automatically isolates at greater than or equal to 50 psig reactor L steam pressure. 150 psis is also below the shutoff head cHf the CSS and -RHRS, thus, considerable overlap exists with.the cooling systems that.. provide ~ core cooling at low reactor pressure. The minimum required NPSH-for RCIC is 20 feet. There isladequate. elevation head between the a suppression pool,and the RCIC pump, such that the required NPSH is - available with a suppression pool temperature up to 140*F~with no ~ containment back pressure.- The ADS, CSS, and RHRS (LPCI) musc be OPERABLE when starting up from a COLD CONDITION.- Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head ~of the CSS and RHRS pumps so they will inject water into the vessel if required. Considering the low reactor pressure and the availability of.the low-pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative'to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility. With the RCICS inoperable,'a seven-day' period.to return the system to service is justified based on the availability of the HPCIS to cool the ~ core and upon consideration that the average risk associated.with failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required. 3.5.G Automatic Deoressurization System (ADS) The ADS consists of six of the thirteen relief valves. It'is designed to. provide depressurization of the reactor coolant system during a small break - loss of coolant accident (LOCA) if HPCI fails or is unable to maintain the-required water level in the' reactor vessel. ADS operation reduces the reactor vessel pressure to within the operating pressure range of the low-1 3 pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. Specification 3.5.G applies only to the automatic feature of the pressure relief system. ~ l Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function. I BFN 3.5/4.5-29 Unit 2 1 p e t

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3.5-BASES:(C:nt'd) The emergency core coolins' system LOCA analyses for small line breaks assumed that four.of the six ADS valves were operable. ;By requiring ~ six valves to be OPES &RLE, additional conservatism is provided to. account for the possibility of.a single. failure in the ADS system. s i Reactor. operation with one of the six ADS. valves inoperable is allowed to continue for-fourteen days provided the HPCI, core spray, and LPCI systems are OPERABLE. Operation with more than one ADS valve inoperable is not' acceptable. With one ADS valve known to be incapable of automatic operation, five - valves remain OPERABLE to perform the ADS function. This condition is + within the analyses for a small break LOCA and the peak clad. temperature is well below.the 10 CFR 50.46 limit.. Analysis has shown-l that four valves are capable of depressurizing the reactor rapidly enough to maintain peak clad temperature within acceptable limits.. 3.5.H. Maintenance of Filled Discharme Pine If the discharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started..To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an OPERABLE condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be-inoperable for Technical Specification purposes. The core spray and RER system discharge piping high point vent is visually checked for water flow cace a month and prior to testing to ensure that the lines are filled.- The visual checking will avoid starting the core spray or RHR system with a discharge line not filled. In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100. feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not'in service. System discharge pressure indicators are used to determine the water level t above the discharge line high point. The indicators will reflect approximately 30 psig for a water level at the high point and 45 psis for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge' lines are filled. When in their normal standby condition, the suction for the HPCI,and RCIC pumps are aligned to the condensate storage tank, which is-physically at a higher elevation than the HPCIS'and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly. i BFN 3.5/4.5-30l Unit 2

~ 13.5 R&&E1 (Cont'd) -3.5.I.c Averaae Planar Linear Heat Generation Rate (APtnGR) This specification assures that the peak cladding' temperature .following the postulated design basis loss-of-coolant accident will-not exceed the limit specified in the-10 CFR 50, Appendix K. The peak ' cladding temperature following a postulated loss-of-coolant 4 accident is primarily a function of the average heat generation' rate i of all the rods of a fuel assembly at any axial location and is only ' dependent secondarily on the rod-to-rod power distribution within an assembly.. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by 1ess than i 20*F relative to the peak temperature for a typical' fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the-l 10 CFR 50 Appendix K limit.

3. 5.J.- Linear Heat Generation Rate (LHCR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated. 1 The LHGR shall be checked daily during reactor operation at 1 25 percent power to determine if fuel burnup, or control rod { movement has caused channes in power distribution. For LEGR to be a limiting value below 25 percent of rated thermal power, the largest total peaking would have to be greater than approximately.9.7 which is precluded by a considerable margin when employing any permissible control rod pattern. 3.5.K. Minimum Critical Power Ratio (MCPR) At tore thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which asy be employed at this point, operating plant experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit. 3.5.L. APRM Setooints Operation is constrained to the LHGR limit of Specification 3.5.J. This limit is reached when core maximum fraction of limiting power BFN 3.5/4.5-31l Unit 2

L;.' '3.5 BASES (Cont'd) i density (CMFLPD) equals 1.0. For the case where CMFLPD exceeds the fraction of rated thermal power, operation is permitted only at less than 100-percent rated power and only with APRM scram settings as required by Specification 3.5.L.1. The scram trip setting and rod ~ block trip setting are adjusted to ensure that no combination of CMFLPD and FRP will increase the LHGR transient peak beyond that allowed by the.1-percent plastic strain limit.- A 6-hour time period 'to achieve-this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis. 3.5.M. Core Thermal-Hydraulic Stability The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1. A manually initiated scram upon entry into this region is sufficient-to preclude core' oscillations which could challenge the MCPR safety limit. 4 Because the probability of thermal-hydraulic oscillations is lower ) and the margin to the MCPR safety limit is greater in Region II than in Region I-of figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal surveillances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an immediate manual scram will be initiated if evidence of. thermal-hydraulic instability is observed. i Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of thermal hydraulic instability and will be immediately investigated. During regional oscillations, the safety limit MCPR is not approached -j until APRM oscillations are 30 percent peak-to-peak or larger in i magnitude. In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit. Therefore, the criteria for initiating a l manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while exiting Region II. Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur.. i BFN 3.5/4.5-32l Unit 2

1,1 BASES (Cont'd) 3.5 3.5.N. References

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2, NEDO - 24088-1 and Addenda.

2. "BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR$1-01-A. .3. Generic Reload Fuel Application, Licensing Topical Report,. NEDE --24011-P-A and Addenda. 4.5 Core mad conta4===at Cooline Syst=== Surveillance Freaumaales 'The testing interval for the core and' containment cooling systems is based on industry practice, quantitative reliability. analysis, judgment and practicality. -The. core cooling systems have not been designed to be' fully testable during operation. -For example, in the case of the HPCI, automatic initiation during power. operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS' testing during power operation causes an. undesirable loss-of-coolant inventory. To increase the availability of.the core and containment. cooling system, the components which make up the

system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each_ cycle combined with testing of the pumps ara injection valves in accordance with Specification 1.0.MN is deemed to be adequate testing of these systems. Monthly alignment checks of valves that are not locked or sealed in position which affect the ability of the systems to perform their intended safety function are also verified to be in the proper position. Valves which automatically reposition themse1 e. on an initiation signal are permitted to be in a position ot' $2 than normal to facilitate other operational modes of the system.

When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by OPERABILITY of the remaining redundant equipment. Whenever a CSCS system or loop is made inoperable, the other CSCS systems or loops that are required to be OPERABLE shall-be considered OPERABLE if they are.within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply. Averane Planar LHGR. LHGR. and MCPR The APLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate. BFN 3.5/4.5-33 l Unit 2

1 <, 2 j 3.10 AAAIR (Cont'd) 3.10.A (Cont'd)' REFERENCES l' ~1. Refueling interlocks (BFNP FSAR Subsection 7.6) B. Core Monitorina The SRMs are provided to monitor the core during periods of unit shutdown and to guide the operator during refueling operations and unit startup. Requiring two OPEnanut SEMs (FLCs) during CORE ALTERATIONS assures adequate monitoring of the fueled region (s) and the core quadrant where CORE ALTERATIONS are being performed. :The fueled region is any set of contiguous (adjacent) control cells which contain'one or l more fuel assemblies. An SRM is considered to be in the fueled region i when one or.more of the four fuel assembly' locations surrounding the SRM dry tube contain a fuel assembly. An FLC is considered to be in the fueled region if the FLC is positioned such that it is monitoring the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region. Each SRM (FLC) is not required to read 1 3 cps until after four fuel assemblies have been loaded adjacent to the SEM (FLC) if no other fuel assemblies are in the associated core quadrant. These four locations are adjacent to the SRM dry tube. When utilizing FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint. With four fuel assemblies or fewer ~ I loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical. 4 Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SEM count rate to decrease below three counts per second. All fuel moves f during core unloading will reduce reactivity. It is expected that the i SRMs will drop below three counts per second before all of the fuel is J unloaded. Since there will be no reactivity additions during this period, the low number of counts will not present a hazard. When sufficient fuel has been removed to the spent fuel storage pool to drop i the SRM count rate below 3 cps, SRMs will no longer be required to be l OPERABLE. Requiring the SRMs to tHe functionally tested prior. to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the i count rate diminishes due to fuel removal. ' Control rods in cells from which all fuel has been removed and which are.outside the periphery of ) the then existing-fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed. ) l REFERENCES

1. Neutron Monitoring System (BFNP FSAR Subsection 7.5)

{ l BFN 3.10/4.10-13 Unit 2 . -. = - ..-r

t y,. '3.10 RASES.(Cont'd) ~ w 2. Morgan, W.'R.~, "In-Core Neutron Monitoring System for' General. ' Electric Boiling Water Reactors," General Electric Company, Atomic j Power Equipment = Department, November 1968, revised April 1969 (APED-5706) 1 C.- Snent Puel' Pool Water The design of the spent fuel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies. y in the reactor building which ensures adequate shielding,. cooling, and : reactivity control of irradiated fuel..An analysis has been performed which shows that a water level at or in excess of eight and one-half' feet over the top of the stored assemblies will provide shielding such-i that the maximum calculated radiological doses do not' exceed the limits of 10 CFR 20.. The normal water leve1'provides 14-1/2 feet of additional water shielding. The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water. input from the condensate storage tanks. All penetrations of the fuel pool have been installed at such a height that their presence does'not provide a possible drainage route that could lower the normal water level more than one-half foot. The fuel pool cooling system is designed to maintain the pool water-temperature less than 125'F during normal heat loads. If the reactor core is completely unloaded when'the pool contains two previous-discharge batches, the temperature may increase to greater than 125'F. The RER system supplemental fuel pool cooling mode will be used under these conditions to maintain the pool temperature to less than 125*F. D. Reactor Buildina Crane The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. The five-ton has both cab and pendant controls. A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed. The testing of the various limits and interlocks assures their proper operation when the crane is used. E. Soent Puel Cask The spent fuel cask design incorporates removable lifting trunnions. The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling. The trunnions must be properly attached to the i cask for lifting of the cask and the visual inspection assures correct installation. 1 i BFN 3.10/4.10-14l Unit 2 1

L: 2.1-RASESs' LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CIAppilq9 IlffEGRITY The abnormal operational. transients applicable to operation of the-Browns Ferry Nuclear Plant have been analyzed in support-of planned operating conditions up to' the maximum thermal power of 3293 MWt. The analyses were based upon plant operation in accordance with Reference 1. d 1 The transient analyses performed for each reload are described in Reference 2. Models and model conservatisms are also described in this reference. l l i BFN 1.1/2.1-11 i Unit 3

ll:. ~ 7, 2 3.2 B&EES.(Cont'd) The low reactor. water level instrumentation that is set to trip when reactor water'1evel is 378 inches above vessel zero (Table 3.2.B)

l

. initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and. starts the diesel generators.' These trip setting levels were chosen to be high enough to prevent-spurious actuation.but low ~ enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of-10 CFR 100 will not be violated. For large breaks.up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time ~to meet the above criteria. The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves. For the breaks discussed above, this. instrumentation will initiate CSCS operation i at about the same time as the low water level instrumentation; thus, ) the results given above are applicable here also. ADS provides for automatic nuclear steam system depressurization, if needed, for small breaks in the nuclear system so that the LPCI and-the CSS.can operate to protect the fuel from overheating. ADS uses six of the 13 MSRVs to relieve the high pressure steam to the suppression pool. ADS initiates when the following conditions exist: low reactor water level permissive (level 3), low reactor water level (level-1), high drywell pressure or the ADS high drywell pressure bypass timer timed out, and the ADS timer timed out. In addition, at least one RER pump or two core spray pumps must be running. l The ADS high drywell pressure bypass timer is added to meet the requirements of NUREG 0737,- Item II.K.3.18. This timer will bypass the high drywell pressure permissivo after a austained low water level. The worst case condition is a main steam line break outside primary containment with HPCI.insperable. With the ADS high drywell pressure bypass timer analytical limit of 360 seconds, a Peak Cladding Temperature (PCT) of 1500*F will not be axceeded for the worst case event. This temperature is well below the limiting PCT of 2200'F. Venturis are provided in the main steam lines as a means of I measuring steam flow and also limiting the loss of mass inventary from the vessel during a steam line break accident. The primary function of the high steam flow instrumentation is to detect a break l in the main steam line. For the worst case accident, main steam ~line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Reference Section 14.6.5 FSAR. BFN 3.2/4.2-65 Unit 3

4 qq.j 3.2 B&&E1 (C:nt'd) Temperature monitoring instrumentation-is prov'ided in the main steam line tunnel to detect leaks'or small breaks in the main steam lines. The trip l setting of 200*F for the main steam line tunnel detector is low enough to provide early indication of a steam line break. Exceeding the trip setting'causes' closure of isolation valves.' For large breaks, the high steam tunnel temperature detection instrumentation is a backup to the high steam flow instrumentation. In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the. ' ambient temperature above 200*F. The temperature increases can cause an unnecessary main steam line isolation and reactor scram. Paraission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation. Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 pais. J The HPCI high flow and temperature instrumentation are provided to detect-a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE. High temperature in the vicinity of the HPCI equipment is sensed by four sets of four binetallic temperature switches. The 16 temperature switches are arranged in two trip systema with eight temperature switches in each trip system. The HPCI trip settings of 90 psi for high flow and 200*F for high temperature are such that core uncovery'is prevented and fission product release is within limits. The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450" water for high flow.and 200'F for temperature are based on the same criteria as the HPCI. High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated. The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed. i 1 BFN 3.2/4.2-66 Unit 3

s, jg [.L. 1 3.51 B4331 (C:nt'd) 3.5.F Reactor Core Isolation Coolinn System (RCICSI j The RCICS functions to provide' makeup water to the reactor vessel during 4 shutdown and isolation from the main heat sink to supplement or replace. the normal makeup sources. -The-RCICS provides its design flow between .150 pais and 1120 psig reactor pressure. Below 150 pais, RCICS is not-required to.be OPERART.E since this pressure. is substantially below that .for any events in which RCICS is needed to maintain sufficient coolant to. the reactor vessel. RCICS will continue to operate below 150 pais at reduced flow until it autonctically isolates at greater than or equal to- ~ 50-psis reactor steam pressure. 150 peig is also below-the shutoff head of the CSS and RERS, thus, considerable overlap exists with the cooling systems that provide core cooling at low reactor pressure. The minimum required NPSH for RCIC is 20 feet.. There is adequate' elevation head between the suppression pool and the RCIC pump, such that the required NPSH is available with a suppression pool-temperature up to 140*F with no containment back pressure. The ADS, CSS,.and RHRS.(LPCI) must be OPERART.R when starting up from a COLD CONDITION. -Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the j CSS and RHRS pumps so they will inject water'into the vessel if ) required..Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available. The alternative to demonstrate-RCIC OPERABILITY PRIOR TO STARTUP using. auxiliary steam is provided for plant operating flexibility. With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be-OPERABLE when required. 3.5.G Automatic Deoressurization System (ADS) The ADS consists of six of the thirteen relief valves. It is designed to i provide depressurization of the reactor coolant system during a small. 1 break loss of coolant accident (LOCA) if HPCI fails or is unable to .i maintain the required water level'in the reactor vessel. ADS operation j reduces the reactor vessel pressure to within the operating pressure range cf the low pressure emergency core cooling systems (core spray and LPCI) so that they can operate to protect the fuel barrier. i Specification 3.5.G applies only to the automatic feature of the pressure j relief system. Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions BFN 3.5/4.5-32 Unit 3 i

.W \\ 4 L *? i .1 g 3.'5 R&g31'(C:nt'd)~ because of. instrumentation failures, yet be fully capable of performing their pressure relief function. The emergency core cooling system LOCA analyses for small line breaks assumed that four of the six ADS valves were OPERABLE. By requiring six valves to be.0PERABLE, additional conservatism is provided to account for the possibility of a single failure in the ADS system. Reactor operation with one of the six ADS valves inoperable is allowed to continue for fourteen days provided the HPCI, core spray, and LPCI systems ^ are OPERABLE. Operation with more than one ADS valve inoperable is not acceptable. With one ADS valve known to be incapable of automatic operation, five valves remain OPERABLE to perform the ADS function. This condition is within the analyses for a small break LOCA and the peak clad temperature is well below the 10 CFR 50.46 limit. Analysis has shown that four valves are capable of depressurizing the reactor rapidly enough to maintain peak clad temperature within acceptable limits. H. Maintan=nce of Filled Discharae Pine If the discharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever.the system is in an OPERABLE condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification purposes. The core spray and RER system discharge piping high point vent is visually checked for water flow once a month and prior to testing to ensure that-the lines are filled. The visual checking will avoid starting the core-spray or RHR system with a discharge line not filled. In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located- ~ approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service. System discharge preseure indicators are used:to determine the water level above the discharge line high point. The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled. When in their normal standby condition, the auction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly. BFN 3.5/4.5-33l 1 Unit 3 .. - - i

= r W i' '3.5E BAAE1 (Cont'd)' - 3.5.I. Averaae Planar Linear Heat Generation Rate (APLHGR) i This spedification assures that the pe'ak cladding temperature following - the postulated design basis loss-of-coolant accident will not' exceed the i limit'specified in the 10 CFE 50,-Appendix K. The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at.any axial location and is'only. dependent secondarily on the rod-to-rod power distribution within an assembly. Since expected local variations-in power distribution within a fuel assembly affect the calculated peak clad temperature by leso than i 20*F relative to the peak temperature for a. typical-fuel design, the limit on the average-linear heat generation rate is sufficient to assure i that calculated temperatures are within the 10 CFR 50 Appendix K limit. 3.5.J. Linear Heat Generation Rate (LHCR) i This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated. The LHGR shall be checked daily during reactor operation at J,25 percent power to determine if. fuel burnup, or control rod movement has caused changes in power distribution. For LEGR to be a limiting value below 25 percent.of rated thermal power, the largest total peaking would have to be greater than approximately 9.7 which is precluded by a considerable margin when employing any permissible control rod pattern. 3.5.K. Minimum Critical Power Ratio (MCPR) At core thermal power levels less.than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated contrci rod patterns which may be employed at this point, operating plant i experience and thermal hydraulic analysis indicated that the resulting i MCPR value is in excess of requirements by a considerable margin. With ] this low void content, any inadvertent core flow increase would only ] place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is. sufficient since power, distribution shifts are very slow when there have' not been significant power or control-rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached i ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit. 3.5.L. APRM Setooints Operation is constrained to the LHGR limit of Specification 3.5.J. This limit is reached when core maximum fraction of limiting power density (CMFLPD) equals 1.0. For the case where CMFLPD exceeds the fraction of rated thermal power, operation is permitted only at less than-BFN 3.5/4.5-34l Unit 3 1 -, 2 -.-

h, m ..3 3.5 BASRS (Cont'd)' 100-percent rated power and only with APRM scram settings as required-j .by Specification 3.5.L.1. The scram trip setting and rod block trip setting 'are adjusted to ensure that no combination of CMFLPD and'FRP will increase.the LHGR transient peak beyond that allowed by the-one-percent plastic. strain limit. A six-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis. 3.5.M. Core Thermal-Hydraulic Stability The minimum margin to the onset of thermal-hydraulic instability occurs in Region I of Figure 3.5.M-1. A manually initiated scram upon entry into this region is sufficient to preclude core oscillations which could challenge the MCPR safety limit. Because the probability of thermal-hydraulic oscillations is lower and the margin to the MCPR safety limit.is greater in Region II than in Region I of Figure 3.5.M-1, an immediate scram upon entry into the region is not necessary. However, in order to minimize the-probability of core instability following entry into Region II, the operator will take immediate action to exit the region. Although formal survelliances are not performed while exiting Region II (delaying exit for surveillances is undesirable), an ismediate manual scram will be initiated if evidence of thermal-hydraulic instability is observed. Clear indications of thermal-hydraulic instability are APRM oscillations which exceed 10 percent peak-to-peak or LPRM oscillations which exceed 30 percent peak-to-peak (approximately equivalent to APRM oscillations of 10 percent during regional oscillations). Periodic LPRM upscale or downscale alarms may also be indicators of. thermal hydraulic instability and will be immediately investigated. During regional oscillations, the safety limit'MCPR is not approached until APRM oscillations are 30 percent peak-to-peak or larger in magnitude. In addition, periodic upscale or downscale LPRM alarms will occur before regional oscillations are large enough to threaten the MCPR safety limit. Therefore, the criteria for initiating a manual scram described in the preceding paragraph are sufficient to ensure that the MCPR safety limit will not be violated in the event that core oscillations initiate while exiting Region II. Normal operation of the reactor is restricted to thermal power and core flow conditions (i.e., outside Regions I and II) where thermal-hydraulic instabilities are very unlikely to occur. 3.5.N. References

1. Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A and Addenda.

BFN 3.5/4.5-35 l Unit 3

3.5 BAEES (C:nt'd) 2. "BWR Transient Analysis Model Utilizing the RETRAN Program," TVA-TR81-01-A. 3. Generic Reload Fuel Application, Licensing Topical Report, -NEDE-24011-P-A and Addenda. 4.5 -Core and Containment Coolinn Systama Surveillance Freauencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be l fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold j water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps j and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing of the pumps and injection valves in accordance with Specification 1.0.MM is deemed to be adequate testing of these systems. Monthly alignment checks of valves that are not locked or sealed in position which affect the ability of the ) systems to perform their intended safety function are also verified to be in the proper position. Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system. l When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by OPERABILITY of the remaining redundant equipment. Whenever a CSCS system or loop is made iroperable, the other CSCS systems or loops that are required to be OPERABLE shall be considered OPERABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable. If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LC0 and the required surveillance testing for the system or loop shall apply. Average Planar LHGR. LHGR. and MCPR The APLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate. BFN 3.5/4.5-36 l Unit 3

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t j B&&E1 (Cont'd) 3.10 3.10.A (Cont'd)' l REFERENCES 1. Refueling interlocks (BFNP FSAR Subsection 7.6) B. Core Monitorina The SENs are provided to monitor the core during periods of unit shutdown and to guide the operator during refueling operations and unit startup. ' Requiring two OPERABLE SRMs (FLCs) during CORE ALTERATIONS assures adequate monitoring'of the fueled region (s) and the core 1 quadrant where_ CORE ALTERATIONS are being performed. The fueled region l is any. set of contiguous (adjacent) control cells which contain one or more fuel assemblies. An SRM is considered to be in the fueled region when one or more of the four fuel assembly locations currounding the' { SRM dry tube contain a fuel assembly. An-FLC is considered to be in -l the fueled region if the FLC is positioned such that it is monitoring 1 the fuel assemblies in its associated core quadrant, even if the actual position of the FLC is outside the fueled region. Each SRM (FLC) is not required to read 1 3 eps until after four fuel assemblies have been loaded adjacent to the SRM (FLC)-if no other fuel assemblies are in the associated core quadrant. These four locations-are adjacent to the SRM dry tube. When utilizing FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint. With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical. Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM-count rate to decrease below three counts per_second. All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded. Since there will be no reactivity additions during this period, the low number of counts will not present a hazard. When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE. Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal. Control rods in cells from which all fuel has been removed may be armed electrically and moved for. maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed. t REFERENCES 1. Neutron Monitoring System (BFNP FSAR Subsection 7.5) 4 BFN 3.10/4.10-12 Unit 3

go. e ^ 3.10 A&IER (Cont'd) 2. Morgan, W. R., "In-Core Neutron Monitoring System for General Electric' Boiling Water Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706) C. Sonnt Fuel Pool Water The design of the spent fuel'atorage pool provides a storage location-for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity _ control of irradiated fuel. An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide. shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20. The normal water ~1evel provides 14-1/2 feet of additional water shielding. The capacity of the skimmer surge. tanks is available to maintain the water level at'its normal height for three days-in the absence of additional water input from the condensate storage tanks. All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible j drainage route that could lower the normal water level more than one-half foot. The fuel pool cooling system is designed to maintain the pool vater temperature less than 125"F during normal' heat loade.- If the reactor core is completely unloaded when the pool contains two previous discharge batches, the temperatures may increase to greater than 1 125'F. The RHR system supplemental fuel pool cooling mode will be used under these conditions to maintain the pool temperature to less than-125'F. 3.10.D/4.10.D BASES Reactor Buildina Crane i The reactor building crane and 125-ton hoist are required to be OPERABLE for handling of the spent fuel in the reactor building. The controls for the 125-ton hoist are located in the crane cab. -The I five-ton has both cab and pendant controls. A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed. l The testing of the various limits and interlocks assures their proper operation when the crane is used. 3.10.E/4.10.E Soent Fuel Cask The spent fuel cask design incorporates removable lifting trunnions. The visual inspection of the trunnions and fasteners prior to 3.10/4.10-13l BFN Unit 3

g,, E r to o l '3.10' R&&R1 (Cont'd)' attachment to the cask assures that no visual damage has occurred R during prior handling.. The trunnions must be properly attached to the cask for lifting of the cask and the visual inspection assures correct-installation.-- 3.10.F Scent Fuel Cank Handlina - Refuelinn Floor Although single failure protection has been provided in the design of the 125-ton hoist drtas shaft, wire ropes, hook and lower block assembly on the reactor building crane,'the limiting of lift height of a spent, fuel cask controls the amount of enargy available in a dropped cask accident when the cask is over the. refueling floor.- An analysis has been made which shows that the floor and support members in the area of cask entry into the decontamination facility can satisfactorily sustain a dropped cask from a height of three feet. The yoke safety links provide. single failure protection for the hook .and lower block assembly and limit cask rotation. Cask rotation is necessary for decontamination and the safety links are removed during decontamination. 4.10 BAREE A. Refuelina Interlocks Complete functional testing of all required refueling equipment interlocks before any refueling outage will provide positive indication. 'i that the interlocks operate in the situations for which they were designed. By loading each hoist with a weight equal to the fuel-assembly, positioning the refueling platform, and withdrawing control rods, the interlocks can be subjected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be i performed to assure that each redundant logic element can independently perform its function. B. Core Monitorina Requiring the SRMs to be functionally tested prior to any CORE ALTERATION assures that the SRMs will be OPERABLE at the start of that alteration. The once per 12 hours verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY. REFERENCES 1. Fuel Pool Cooling and Cleanup System (BFNP FSAR Subsection 10.5) 2. Spent Fuel Storage (BFNP FSAR Subsection 10.3) I i i I BFN 3.10/4.10-14 l Unit 3

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U.S. Nuclear-Regulatory Commission Page 4 February 23, 1995 PS:JEN:SAW:MTN Enclosures cc (Enclosures): R. R. Baron, BR 4J-C E. S. Christenbury,-ET 11H-K -K. N. Harris, LP 3B-C R. W. Huston,'Rockville Office-C R..D. Machon, PAB 1E-BFN T. J. McGrath, LP 3B-C D. E.-Nunn, LP 3B-C T. W. Overlid,-LP 4E-C E. Preston, POB 2C-BFN T. D. Shriver, PAB 1A-BFN H. L. Williams, EDB 1A-BFN RIMS, WT 3B-K mAtachspec\\ta348.new u 1 p*ad - l y}}

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