ML20217C211
| ML20217C211 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/11/1997 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217C205 | List: |
| References | |
| NUDOCS 9710010265 | |
| Download: ML20217C211 (70) | |
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(*$ NUCLEAR REGULATORY COMMISSION WABHINGTON, D.C. 30086 0001
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 249 License No. OPR-52
- 1. The Nuclear Regulatory Commission (the Comission) has found that:
A. The application for amendment by Tennessee Valley Authority (the licensee) dated June 2. 1995, as revised on March 6, 1997, and supplemented on May 13. 1997 and August 20, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I: B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission: C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Commission's regulations: D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public: and-E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. Tu P 1888!t ?d8?s h o PDR < a
- 2. Accordingly. the license is amended by changes to the Technical Specifications as-indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-52 is hereby amended to read as follows-(2) Technical- Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 249. are hereby incorporated in the license. The licensee shall o Specifications. perate the facility in accordance with the Technical 3.
This license amendment is effective as of its-date of issuance and shall be implemented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION O' . s Frederick J. Hebdon. Director Project Directorate 11 3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical - Specifications Date of Issuance: September 21, 1 97 k
ATTACHMENT TO LICENSE AMENDMENT NO. 2A9 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Apgndix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. *0verleaf and "spillover pages are included to maintain document completeness. REMOVE INSERT 1.0-7 1.0-7 1.0-8 1.0-8 1.1/2.1-1 1.1/2,1-1* 1.1/2.1-2 1.1/2.1-2 1.1/2.1 1.1/2,1-3 1.1/2.1-4 1.1/2.1-4* 1.1/2.1-6 1.1/2.1-6* 1.1/2.1-7 1.1/2,1-7 1.1/2.1-16 1.1/2.1-16 1.1/2.1-16a 1.1/2.1-16a* 3.1/4.1-3 3.1/4.1-3 3.1/4.1-4 3.1/4.1-4* 3.1/4.1-5 3.1/4.1-5 3.1/4.1-6 3.1/4.1-6 3.1/4.1-7 3.1/4.1-7 3.1/4.1-8 3.1/4.1-8 3.1/4.1-9 3.1/4.1-9* 3.1/4.1-10 3.1/4.1-10 3.1/4.1 3.1/4.1-11 3.1/4.1-12 3.1/4.1-12 3.1/4.1-14 3.1/4.1-14 3.1/4.1-15 3.1/4.1-15 3.1/4.1-16 3.1/4.1-16 3.1/4.1-17 3.1/4.1-17 3.1/4.1-18 3.1/4.1-18* 3.1/4.1-19 3.1/4.1-19 3.1/4.1-20 3.1/4.1-20 3.1/4.1-21 3.1/4.1-21 3.2/4.2-1 3.2/4.2-1* 3.2/4.2-2 3.2/4.2-2 3.2/4.2-25 3.2/4.2-25 3.2/4.2-25a 3.2/4.2-25a* 3.2/4.2-26 3.2/4.2-26 3.2/4.2 3.2/4.2-27 3.2/4.2-27a 3.2/4.2-27b* Continued on following page.
l; e ATTACHMENT TO LICENSE AMENDMENT NO.249_
)
FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50 260 List continued from previous page: REMOVE INSERT 3.2/4.2-50 3.2/4.2-50 3.2/4.2-50a 3.2/4.2-50a* 3.2/4.2-59 3.2/4.2-59 3.2/4.2-60 3.2/4.2-60 3.2/4.2-67 3.2/4.2-67* 3.2/4.2-68 3.2/4.2-68 3.2/4.2-73 3.2/4.2-73* 3.2/4.2-73a 3.2/4.2-73a 3.3/4.3-7 3.3/4.3-7* 3.3/4.3-8 3.3/4.3 8 3.3/4.3-17 3.3/4.3-17 3.3/4.3-18 3.3/4.3-18* 3.5/4 5-18 3.5/4.5-18 : 3.5/4.5-19 3.5/4.5 19 3.5/4.5-20 3.5/4.5 20 3.5/4.5-20a 3.5/4.5-20a* 3.5/4.5-22 3.5/4,5-22* 3.5/4.5-22a 3,5/4.5-22a 3.5/4.5-30 3.5/4.5-30* 3,5/4.5-31 3.5/4,5-31 3.5/4.5-32 3.5/4.5-32 3.5/4.5-33 3.5/4.5 33 3.5/4.5-34 3.5/4.5-34** 3.5/4.5-35 3.5/4.5-35** 6.0-26 -6,0 26* 6.0-2Ga 6.0-26a
e 1.0 DEFINITION 2 (Ccnt'd) Q. Deeratino evele - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit. R. Refueline outmoe - Refueling outage is the perior!. of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly
. scheduled outage.
S. CORE ALTERATION - CORE ALTERATION shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Movement of source range monitors, intermediate - range monitors, traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location. T. Remeter Vemmel Preneure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors. U. Thermal Parameters
- 1. Minimum critical Power Ratio (McPk) - Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
- 2. Transition meilina - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur
/ intermittently with neither type being completely stable.
- 3. (Deleted) d
- 4. Ayermee Planar Linear Meat Generation Rate (APLMCR) - The Average Planar Heat Generation Rate is applicable to a specific planar height and is equal to the sum of the linear heat generation rates for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
BFN 1.0-7 Amendment No. 249 Unit 2
4 1.O DEFINITIONS (Ctnt'd) i s. coel.t.d) 4
-V. Instrumentation
- 1. Instmment calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.
- 2. Charnal - A channel is an arrangement of the sensor (s) and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terriaates and loses its identity where individual channel outputs are combined in logic.
- 3. Yn=trumant Punctinn=1 Tant - An inJtrument functional t est mesas the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action.
- 4. Inatrument Check - An instrument check is qualitative deterinination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.
- 5. Leale avstam Punational Tant - A logic system functional test means a-test of all relays and contacts of a logic circuit to.
insure all couponents are operable per design intent. Where practicable, action will go to completion; i.e., pumps will be started and valves operated.
- 6. Trin avatam -
A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip rystems.
- 7. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at.a channel or system level, s- protective Punstian - A system protective action which results from the protective action of the channels monitoring a particular plant condition.
BFN 1.0-8 heendment No.249 Unit 2
=. - . . _ . _ _ _ .- . . . - . . .. . - - - . _ ,
a 1.1/2.1 FUEL CLADDINC INTECRITY I SAFETY L'IMIT LIMITINC EAFETY SYSTEM SETTING l 141 M L CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY
- Acelicability Anelicability Applies to the interrelated Applies to trip settings of variables associated with fuel the instruments and devices thermal behavior. which are provided to prevent the reactor system safety j limits from being exceeded.
ghieetive obioetive To establish 1Laits which To define the level of the I ensure the integrity of the process variables at which fuel cladding. automatic protsetive action i . is initiated to prevent the fuel cladding integrity
- safety limit from being exceeded.
soecifications Soecificationg The limiting safsty system settings shall be as specified below: A. Ih3ymal Power Limits A. Neutron Flux Trie Settinos
- 1. Reactor Pressure >800 1. APRM Flux scram psia and Core Flow Trip setting
> 10% of Rated. (RUN Mode) (Flow Biased)
When the reactor pressure is greater a. When the Mode than 800 psia, the switch is in existence of a minimum the RUN critical power ratio position, the l (MCPR) lese than 1.10 APRM flux shall constitute scram trip violation of the fuel setting cladding integrity shall bei safety limit. BFN 1.1/2.1-1 AMENDMENT NO. 2 4 7 Unit 2
e 1.1/2.1 FUEL C'LABDING INTECRTTY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flur Trie Settir.ca 2.1.A.1.a (Cont'd) Ss(0.66W + 71%) l where: S = Setting in percent of rated thermal power (3293 MWt) W = Loop recirculation flow rate in percent of rated
- b. For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
BFN 1.1/2.1-2 Amendment No. 249 Unit 2
a 1.1/2.1 FUEL c't 1MBING TFPEGRTTY thPETY LIMIT LTMITING EAFETY SYSTEM ERTTING 2.1.A Neutron Flur Trin settinam 2.1.A.1.b (Cont'd) E2IE: These settings assume 4 operation within the basic thermal hydraulic design criteria. These criteria are APLHGR within the limits of specification 3.5.I, LHGR within the limits of specification 3.5.J and MCPR within the limits of Specification 3.5.K. If it
- is determined that any of thesei l design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limits.
]
- c. The APRM Rod Block trip setting shall be less than or equal to the limit specified in the CORE OPERATING LIMITS REPORT.
4 BFN 1.1/2.1-3 Amen & ment No. 249 Unit 2
e i 1 1/2.1 FUEL 6 , INTEGRITY BAFETY LTMTT LTMYTING SAFETY SYSTE!E SETTINC 1.1.A Thermal Power Linita 2.1.A Neutron Flum Trin settinas (cont'd)
- d. Fixed High Neutron Flux scram Trip setting--When the mode switch is in the RUN position, the APM fixed high flux scram trip setting shall bei 55120% power.
- 2. Reactor Pressure 1800 2. APM and IRM Trip settings psia or Core Flow 110% (startup and Hot standby of rated. Modes). ,
when the reactor pressure a. APRM--When the is 5800 psia or core flow reactor mode switch is slot of rated, the core is in the STARTUP thermal power shall not position, the APRM exceed 823 MWt (25% of scram shall be set at rated thermal power). less than or equal to 15% of rated power.
- b. IRM--The IRM scram shall be set at less than or equal to 120/125 of full scale.
/ ( BFN 1.1/2.1-4 RENDMENT NO.14 3 Unit 2
rigur. 2.1-1 extano y[y,, 1.1/2.1-6 AMENDMENT NO. 2 3 2
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o ,. 0 10 20 30 40 50, 60 '70 80 90 100 110 120 Core Coolant Flow Rate (% of Design) APRM Flow Blas Scram vs. Reactor Core Flow BFN Flo 2.1-2 1.1"72.1.7 Amerrhent No. 249 Unit 2
e 2.1 R&gLt (cont'd) F. (Deleted) O. & M. ggin Etnam line feelation en Lew Pressure and Main teamm Line Imelatien Scram The low pressure isolatica ' ' he main steam lines at $25 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. The scram feature that occurs when the main steam line isolation va?.ves close shuts down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity SAFETY LIMIT.
, Operation of the reactor at pressures lower than 325 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity SAFETY LIMIT is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability c2 neutron flux scram protection over the entire range of '
applicability of the fuel cladding integrity SAFETY LIMIT. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase. I.J.& K. Remeter Lew water La>el sateeint for Ynitiation of upcf and Reic clemina Main Etnam Yaelation Valves, and Startina LDet and cere serav _Pmne n . These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system preasure. L. Referaneaa
- 1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 2 (applicable cycle-specific document) .
- 2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NFIE-24011-P-A US (latest approved v$rsion) .
- 3. Maximum Extended Load Line Limit knd ARTS Improvement Program Analyses for Browns Fe ry Nuclear Plant, Units 1, 2, and 3, NEDC-32433P.
BFN 1.1/2.1-16 Amendment No. 249 Unit 2
TH!s PAGE INTENTIONALLY LEFT RIANK BFN M 1.1/2.1-16a DMENT NO. I 81 Unit 2 l
$#k TABLE 3.1.A REACIOR PRCTECTION SYSTEM (SCRAM) INSTRtJMENTATION REQUIRDG3rTS y .
Min. No. of Operable Instr. Modes in which runctim Channels Must Be Ooerable Per Trip Shut- Sta rtup/ System! (11(231 Trin Function Trio Irvel Settino im Refuel f71 Het St amthe Rygg Action (1) . l 1 1 Mode Switch in X X X X 1.A Shutdown 1 Manual Scrae X X X X 1.A IRM (16) 3 High Flux s120/125 Indicated X(22) X (5) 1.A on scale 3 Inoperative Z(22) X (5) 1.A APPM (16) (24) (25) te 3(11) High Flux l p (Flow Biased) See Spec. 2.1.A.1 1 1.A. 1.B. or 1.E I
) 3(11) High Flux !
(Fixed Trip) s 120% X 1.A. 1.B, or 1.E l y 3(11) High Flux s 15% rated power (21) 4(17) (15) 1.A or 1.E [ 4 (13) (21) X (17) X 1.A or 1.E l 3(11) Inoperative 2 2-Out-of-4 (12) (21) X X 1.A or 1.F l Voter [ 2 High Reactor Pressure s 1055 peig X(10) X X 1.A (FIS-3-22AA.BB,C,D) 2 High Drywell Pressure (14) s 2.5 psig Z(8) Z(8) X 1A (PIS-64-56 A-D) 2 Reactor Low Water n Level (14) 2 538' above X X X 1.A (LIS-3-203 A-D) vessel zero e e
~
"'O'4e,&
gg
c to EE et e4 facLE 3.1.A REACTM P90fECTIM SYSTEgl (sCaafs) testatsENTAf ttus aEWieEpEWTS . Nin. No. of Operable Instr. Itsdre in dich Feetten alust te Chamele hereMe For Trip Shut- Stort W system (1)(23) Tris Ftsittlest Trfe Levet settim desg! Refurt (T) feet Stener 3lE1 Actlen (1) 2 sigh unter Levet in weet Screar Slecherpe Yardt R(2) N(2) X X 1.A (LS-85-454-9) S 50 settone 2 Wish unter Lewet in feet Scress Discherpe Yardt
< 50 Geltene N(2) N(2) E X T.A (LS-SS-45E-s) s.s N(4) 1.A er 1.C
- 4 feeln Stesse Line ~<19E Velve Cleeure C*
teetetten volve Cleeure R(4) 1.A or 1.3 p 2 Turbine Centret volve Feet 1550 pelg Cleeure er Turbine Trip Turbine Step N(4) 1.A or 1.9
& i 9E 1 Velve Closure volve Cleeure Turbine First not 1154 pelg N(18) 3(18) I(18) 1.A or 1.3 (19) p 2 2C Stege Pressure j
Perarleelve Q c: (P15-1-81A84, i l 3C PIS-1-91AAS)
= 1.A 4
" 2 Lou Scree ritet 150 pels N(2) X(2) x I 2 Air needer P Pressure EG N
.=4 C_
neorms som TAmta 2. g . A
- 1. There shall be two OPERABLE or tripped trip systems for each function. If the minimum number of OPERABLE instrument channels per trip system cannot be met for one trip system, trip the inoperable channels or entire trip system within one hour, or, alternatively, take the below listed action for that trip function. If the minimum number of OPERABLE instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right hand column of Table) shall be taken. An inoperable channel need not be placed in the tripped condition where this would cause the trip function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within two hours, or take the action listed below for that trip function.
A. Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rode-within four hours. _In refueling mode, suspend all-operations involving core alterations and fully insert all OPERABLE control rods within one hour.
-3. Reduce power-level-to_IRM range and place mode switch in the STARTUP/ MOT Standby position within 8 hours.
C. Reduce turbine load and close main steam line isolation valves within a hours. D. Reduce power to less than 30 percent of rated. E. For the APRM trip functions, if one required APRM channel is inoperable, restore it to OPERABLE status or place the channel in trip within 12 hours. If two or more required APRM channels are inoperable for one or more trip functions, restore trip capability within 1 hour or initiate alternate action listed in the table. F. For the APRM trip functions, if one required voter channel is inoperable, restore the channel to OPERABLE status, place the channel in trip, or place the associated trip system in trip within 12 hours. If one required voter channel la inoperable in both trip systems. restore one channel to OPERABLE status, place one channel in trip, or place one trip system in trip within 4 hours, or initiate alternate action listed in the table. If two required voter channels are inoperable in one trip system, restore trip capability within 1 hour or initt ste alternate action listed in the table.
- 2. The scram discharge volume high water level bypass may be used in SHUTDOWN or REFUEL to bypass both the scram discharge volume high.high water level and scram pilot air header low pressure scram signals in order to reset the reactor protection system trip. A control rod withdraw block is present when these scram signals are bypassed.
- 3. (Deleted)
- 4. typassed when turbine first stage pressure is less than 154 psig.
BFN 3.1/4.1-5 Am m h nt No. 249
'Jnit 2
Y l McTag rom TAmta 2.1.1 (font'd)
- 5. IRMs are bypassed when the reactor mode switch is in the RUN position, j
- 6. The design pomita closure of any two lines without a scram being initiated.
- 7. When the reactor is suberitical and the reactor water temperature is less than 212'F, only the following trip functions need to be OPERABLE:
A. Mcde switch in SHUTDOWN - 4
- 3. Manual scram C. Highikux1RM a
D. scram discharge volume high level E. (Deleted) q F. Scram pilot air header low pressure j 0. Not requited to be OPERABLE when primary containment integrity is not
- required.
- 9. (Deleted) l 4
- 10. Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.
- 11. Each APRM channel provides input to both trip systems. l
- 12. Any combination of APRM upscale or inoperative trips from two or more
- non bypassed APRM trip functions will trip all of the 2 out-of-4 voter
, trip functions.
- 13. Less than the required minimum number of OPERABLE LPRMs will cause an j instrument channel inoperative alarm.
- 14. Channel shared by Reactor Protection System and Primary Containment and l Reactor Vessel Isolation Control System. A channel failure may be a
- channel failure in each system.
l- 15. The APRM 15 percent scram is bypassed in the R'JN Mode.
- 16. Channel shared by Reactor Protection System and Reactor Manual Control
- system (Rod Block Portion) . A channel failure may be a channel failure in each system., .If a channel is allowed to be inoperable per i Table 3.1.A, the corresponding function in that same channel may be j inoperable in the Reactor Manual Control system (Rod Block).
4 i i
! prN 3.1/4.1 6 Amerubent No. 249
! Unit 2 5 y--,,.w-, ,--m,y__y-,,,,,--, -- -y w.. c,y,-.- - -. - . , - - , _ ,-y . - , , . ,_.y , . , . . . . m., ..y, _ - - . - - , , , , , , - - . . , w-.-
NOTER FOR TABLE 1.1.A (Cont' d)
- 17. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
- 10. This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.
- 19. Action 1. A or 1.D shall be taken only if the perudssive f ails in such a manner to prevent the af fected RPS logic from performing its intended function. Otherwise, no action is required.
- 20. (Deleted)
- 21. In the REFUEL Mode unless adequate shutdown margin has been demonstrated per Specification 4.3.A.1 and the one-rod-out control rod block is OPERABLE per Specification 3.10. A.1, whenever any control rod is withdrawn from a core cell containing one or more fuel assemblies either (a) shorting links shall be removed from the RPS circuitry to enable the Source Range Monitor (SRM) noncoincidence high-flux scram function or (b) the indicated APRM trip functions shall be OPERABLE per the requirements applicable in the STARTUP/ HOT STANDBY Mode. If the SRM nonceincidence high flux screr function is enabled, the SRMs shell be QFERABLE per Specification 3.10.B.1. The removal of eight (8) shorting links is required to provide nonceincidence high-flux scram protection from the SRMs.
- 22. Only required with any control rod withdrawn from a core cell containing one or more fuel assemblies. For the IRM High riux Trip runction, the three required IRMs per trip channel is not required if at least four IRMs (one in each core quadrant) are connected to give a nonceincidence, High riux scram. The removal of four (4) shorting links is required to provide noncoincidence high-flux scram protection f rom the IRMs.
- 23. A channel may be placed in an INOPERABLE status for up to 4 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
- 24. The Average Power Range Monitor scram function is varied (Reference rigure 2.1-1) as a function of recirculation loop flow (W) . The trip setting of this function must be maintained in accordance with 2.1.A.
- 25. The APRM flow-biased neutron flux signal is fed through a time constant circuit of approxinately 6 seconds. This time constant may be lowered or equivalently removed (no time delay) without affecting the operability of the flow-biased neutron flux trip channels. The APRM fixed high neutron flux signal does not incorporate the time constant but responds directly to instantaneouc neutron flux.
BrN 3.1/4.1 ~1 Unit 2 Amerximent No. 249
CW TABLE 4.1.A k REACIUR PROTECTION SYSTEM (SCRAM) INSTRtmRNTATION FUNCTIONAL TESTS u MINIMIDI FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUDENT AND CONTROL
- CIRCUITS Gramm f21 Functional Teat Minieman Frasmanacyf31 Mode Switch in Shutdown A Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A Trip channel and Alarm Every 3 Months .
IRM High Flux C Trip Channel and Alam (4) Once/Neek (9) i Innperative C Trip channel and Alarm (4) Once/Neek (S) 't I 4 High Flux (15% Scram) Trip Output Relays (4) (5) Every 6 Months (3) l High F1 tax (Flow Biased) Trip Output Relays (4) (6) Every 6 Monthe l i i i High Flux (Fixed Trip) Trip output Relays (4) (5) Every 6 Months l
. i
."4 Inoperative Trip Output Relays (4) (5) Every 6 Monthe l l N ,
4 \ p" 2-Out-of-4 voter Trip Output Relays (4) (5) Every 6 Monthe ] 4 4 2-Out-of-4 voter Logic (10) Each Refueling Outage l Trip Scram Contactors (11) Once/ Neck l High Reactor Pressure B Trip Channel and Alarm (7) Once/ Month (PIS-3-22AA, BB, C, D) High Drywell Pressure B Trip Channel and Alarm (7) Once/ Month (PIS-64-56 A-D)
-p Reactor Ier Water Level B Trip Channel and Alarm (7) Once/ Month ,
{ , (LIS-3-203 A-D) i i < i l l 1 ,
l1llll ;jI ;l e
)
Gv c n u s e a ) S h e r ( t n , F ) o ) e 1 m 1 ne ht ( ( m e i t se ht ht n a h e h n t r t e n n o n a e l na h n t s
, i e t e R m s 3 e g
/ y e
t 4
/
/e /e /e e r /e e c
n c cs c n e v c c i n i e m e O E e e
) )
7 7 ( ( t e er n r s r m r e r mr m r e t o t e l o l a l o a e f l t A A A A A A A
) l d
de a d f u d d d d n n n r n n na n s a o a a e a a a i r l t t l t t t t i e e n nen t r i w e e e n w v r e r i w cr r w e e e r e a e a r e C F h i i h t i ( D d h t C O D t A. p p p p p p p i i i i i i i 1 r r r r r r r 4 T T T T T T T E L B A T
)
2 ( m A B A A S A A a r G
)
F eC p5 r 4 e-5 e hc 3- e e v , i s leS h c l o e 5 , B(L t V t rd un e rA 2 i) ep sa ma e s Swa ne ei s r u e d , rh ,t F r eA e e1 Scc t t6 e t e e t r1 P8 l e N eA niw v , t ve - C 5 I s oe L ee l n ei p e1 g- ) e i - r3
, t Vt r aS3 v A5 t t cA tI l 3 e e v e SPd i5 e t u t e n4 n et r ( na V oS el L F o- i L t r se t
p i(l P r5 ne rvA e P t8 e e trd e ae c-eS l L m ae Ce er t u eu r il1 F e9 e-el 1 t S e s s r au rs) T E( W S s ne e i er i ia-r rS i n cs2 S eS h g nt t l beI br r i iC a rC u rPP u u wP& e N M T T T L l w* 7. l W'3GcErn :=9NN1* i N lllIli'
MOTER FOR TABLE &.1.A
- 1. Initially the minimum frequency for the indicated tests shall be once
; per month.
I
- 2. A description of the three groups is included in the Bases of this specification.
- 3. Functional tests are not required when the systems are not required to be OPERABLE or are operating (i.e., already tripped). If tests are missed, they shall be performed prior to returning the systems to an OPERABLE status.
- 4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.
- 5. The channel functional test shall include both the APRM channels and l the 2-out-of-4 voter channels. I
- 6. The channel functional test shall include both the APRM channels and
{~ the 2-out-of-4 voter channels plus the flow input function, excluding the flow transmitters.
- 7. Functional test consists of the injection of a simulated signal into
- the electronic trip circuitry in place of the sensor signal to verify
- operability of the trip end alarm functions.
1
- 8. The functional test frequency decreased to once every three months to reduce challenges to relief valves per NUREG 0737, Item II.K.3.16,
- 9. Not required to be performed when entering the STARTUP/ HOT STANDBY Mode from RUN Mode until 12 hours af ter entering the STARTUP/ Hot STANDBY Mode.
l , 10. Functional test consists of simulating APRM trip conditions at the 2-out-of-4 voter channel inputs to check all combinations of two tripped inputs to the 2 out-of-4 voter logic in the voter channels.
- 11. Functional test consists of manually tripping the 2-out-of-4 voter trip outputs, one voter channel at a time, to demonstrate that each scram contactor for each RPS trip system channel (A1, A2, 31, and 32) l operates and produces a half-scram, o
BFN 3.1/4.1-10 haerrbnt No. 249 I Unit 2 l l
g 88 TABLE 4.1.5
*E REACTOR P9tOTECTIOlt SYSTEM (SCRAM) INSTRUpWNT CALIBRATION
^
MINIMUM CALIBRATXOct FREQUENCIES FtNt REACTOR P9tOTECTION INSTRtNIENT maanart e es Instrument Channel Crr- fil f.alibration Minimun Frasmanacyf21 IRM High Flux C Comparison to AP9tM en Controlled Note (4) Startups (6) , APRM High Flux
- Output Signal Heat Balance once/7 Days ]
Flow Bias Signal Calibrate Flow Blas Signal (7) Once/ Operating Cycle h LPRM Signal TIP System Traverse (s) Every 1000 Effective Pull l Power Nours High Reactor Pressure B Standard Pressure Source once/6 Months (9) (PIS-3-22 AA, BB, C, D) Migh Drywell Pressure B Standard Pressure Source Once/18 Monthe (9) (PIS-64-56 A-D) Y Q Reactor Iow Nater Level B Pressure Standard Once/18 Monthe (9) (LIS-3-203 A-D) ' M [., High Noter Invel in Scram
" Discharge voltame Float Switches (LS-85-45-C-F) A Calibrated Nater Column Once/18 Months Electronic Level Switches (LS-85-45 A, B, G, N) B Calibrated Nater Cblumn Once/18 Months (9)
Main Steam Line Isolation valve Closure A Note (5) Note (5) Turbine First Stagu Pressure Permissive (PIS-1-81 A&B, - PIS-1-91 A&B) B Standard Pressure Source Once/18 Months (9) h rbine Stop valve Closure A foote (5) Note (5)
" Turbine Control Valve Fast Closure on h rbine Trip A Standard Pressure Source Once/ Operating Cycle g Inw Scram Pilot, Air Header Pressure (PS 85-35 A1, A Standard Pressure Source once/18 Months e
A2, B1, & B2)
180712 Poa TiefE a.1.a
- 1. A description of three groups is included in the bases of this specification.
- 2. calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERA 3LE status.
- 3. (Deleted) 1 4.
Required frequency is initial startup following each refueling outage, s Er Physical inspection and actuation of these position switches will be performed once per operating cycle.
- 6. On controlled startups, overlap between the IRMs and APRMs vill be verified.
- 7. The flow bias signal calibration will consist of calibrating the analog differential pressure flow sensors once per operating cycle.
Calibration of the flow bias processing system is done once per operating cycle as part of the overall APRM instrumentation calibration.
- 8. A complete TIP system traverse calibrates the LPRM signals to the process computer. The individual LPRM meter readings will be adjusted as a minimum at the beginning of each operating cycle before reaching 100 percent power.
- 9. Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.
ani 3.1/4.1 12 Amentheent No. 249 Unit 2
3.1 R&&Eg The Reactor Protection System automatically initiates a reactor scram to:
- 1. Preserve the integrity of the fuel cladding.
~
- 2. Preserve the integrity of 'the reactor coolant system.
- 3. Minimirs the energy which must be absorbed following a loss of coolant accident, and prevents criticality.
This specification provides the LIMITING CONDITIONS FOR OPERATION necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periode when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations. The reactor protection trip system is supplied, via a separate bus, by its own high inertia, ac motor generator set. Alternate power is available to either Reactor Protection System bus from an electrical bus that can receive standby electrical power. The RPS nonitoring system provides an isolation between nonclass 1E power supply and the class 1E RPS bus. This will ensure that failure of a nonclass 1E reactor protection power supply will not cause adverse interaction to the class 1E Reactor Protection System. The Reactor Protection System is made up of two independent trip systems (refer to Section 7.2, FSAR) . There are usually four channels provided to monitor each critical parameter, with two channels in each trip system. The outputs of the channels in a trip system are combined in a logie such that either channel trip will trip esat trip system. The simultaneous tripping of both trip systems will produce a reactor scram. This system meets the intent of IEEE-279 for Nuclear power Plant Protection Sy tems. The system has a reliability greater than that of a 2 out-of-3 system and somewhat less than that of a 1-out-of-2 system. With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation valve closure and the Turbine Stop valve closure, each trip system logic has one instrument channel. When the minimum condition for operation on the number of OPERABLE instrument channels per untripped protection trip system is met or if it cannot be met and the effected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved; i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor. q The APRM system is divided into four APRh channels and four 2-out-of 4 trip voter channels. Each APRM channel provides input to each of the BFN 3.1/4.1-14 1- ~ h at No. 249 Unit 2
3.1 th1ER (Cont'd) four voter channels. The four voter channels are divided into two groups of two each with each group of two providing input to one RPS trip system. The APRM system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a ' half tripa in all four of the voter units, but no trip inputs to either RPs trip system. A trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip system resulting in a full scram. Each APRM instrument channel receivts input signals from forty three (43) Local Power Range Monitors (LPRMs) . A minimum of twenty (20) LPRM inputs with three (3) per axial level is required for the APRM instrument channel to be OPERARLE. Fewer than the required minimum number of LPRM inputs generates an instrument channel inoperative alarm and a control rod block but does not result in en automatic trip input to the 2 out-of-4 voters. rach protection trip system has one more IRM than is necessary to meet l tte minimum number required per channel. This allows the bypassing of one IRM per protection trip system for maintenance, testing or l calibration. The bases for the scram setting for the IRM, APRM, high i reactor pressure, reactor low water level, MsIV closure, turbine control valve fast closure, and turbine stop valve closure are discussed in specifications 2.1 and 2.2. Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation. A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating 4 status. Reference section 7.2.3.7 FsAR. The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation. The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges. The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. Tha discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping. No credit SFN 3.1/4.1-15 Amendment No. 249 Unit 2
3.1 R&&XA (Cont'd) I was taken for this voluwe in the design of the discharge piping as concerns the amount of water which must be acconnodated during a scram. During normal operatien the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated which would result in slow scram tines or partial control rod insertion. To preclude this occurrence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volune of water reaches 50 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharge water and precludes the situation in which a scram would be required but not be able to perform its function adequately. A source range monitor (SRM) system is alsc provided to supply additional neutron level information during startup but has no scram functions. Reference Section 7.5.4 FSAR. Thus, the IRM is required in the RETUEL (with any control rod withdrawn from a core cell containing one or more fuel assemblies) and STARTUP Modes. In the power range the APRM system provides required protection. Reference Section 7.5.7 FSAR. Thus, the IRM System is not required in the RUN mode. The APRMs and the IRMs provide adequate coverage in *.M STARTUP and intermediate range. The high reactor pressure, high drywell pressure, reactor low water level, low scram pilot air header pressure and scram discharge volume high level scrams are required for STARTUP and RUN modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation. Because of the APRM downscale rod block limit of A 3 percent when in the
- RUN mode and high level flux scram limit of $15 percent when in the STARTUP Mode, the transitfon between the STARTUP and RtIN Modes must be made with the APRM instrumentation indicating between 3 percent and 15 percent of rated power. In addition, the IRM system must be indicating d below the High Flux setting (120/125 of scale) or a scram will occur when in the STARTbP Mode. For norwal operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no " gaps" in the power level indications (i.e. , the power level is continuously monitored from beginning of startup to full power and from full power to SNUTDOWN) . When power is being reduced, if a transfer to the STARTUP mode is made and the IRMs have not been fully inserted (a maloperational but not impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.
The low scram pilot air header pressure trip performs the same function as the high water level in the scram discharge instrument volume for fast fill events in which the high level instrument response time may be inadequate. A fast fill event is postulated for certain degraded control air events in which the scram outlet valves unseat enough to allow 5 ypm per drive leakage into the scram discharge volume but not enough to cause control rod insertion. BFN 3.1/4.1-16 Amendment No. 249 Unit 2
4.1- 46488 The minimum functional testing frequency used in this specification is based on a reliability analysis using the concepts developed in
- reference (1) . This concept was specifically adapted to the one out of-
-two taken twice_ logic of the reactor protection system. The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system. This analysis makes use of
- unsafe failure" rate experience at conventional and nuelaar power plants in a reliability model for the system. An ' unsafe failure" is defined av one which negates channel operability and which, due to its nature, is revealed only when the channel is functionally tested or atteopts to respond to a real signal. Failure such as blown fuses,' ruptured bourdon tubes, faulted amplifiers, faulted cables, etc., which result in
.
- upscale
- or 'downscale" readings on the reactor instrumentation are "safea and will be easily recognised by the operators during operation because they are revealed by an alarm or a scram.
Except for the APRMs which take credit for self test capability, the l channels listed in Tables 4.1.A and 4.1.5 are dividad into three groups for functional-testing. These are: A. on off sensors that provide a scram trip function.
- a. Analog devices coupled with bistable trips that provide a scram function.
C. Devices which only serve a useful function during some restricted mode of operation, such as STARTUP, or for which the only practical test is one that can be performed at SHUTDOWN. The sensors that make up group (A) are specifically selected from among the whole family of industrial on off sensors that have earned an excellent reputation for reliable operation. During design, a goal of 0.99999 probability of success (at the 50 percent confidence level) was adopted to assure that a balanced and adequate design is achieved. The probability of success is primarily a function of the sensor failure rate and the test interval. A three-month test interval was planned for group (A) sensors. This is in keeping with good operating practices, and satisfies the design goal for the logic configuration utilised in the Reactor Protection System. The once.per six-month functional test frequency for the scram pilot air header low pressure trip function is acceptable due to:
- 1. The functional reliability previously 43monstrated by these switches on Unit 2 during cycles 6 and 7 J
- 2. The need for minimising the radiation exposure associated with the functional testing of these switches, and
- 3. The incteased risk to plant availability while the plant is in a half-scram condition during the performance of the functional testing versus the limited increase in reliability that would be obtained by more frequent functional testing.
RFN 3.1/4.1-17 Anandment No. 249 Unit 2 l
4.1 36111 (cont *d) A single failure of one of the scram pilot air header low pressure trip switches would not result in the loss of the trip function. It is highly unithely that two switches in one channel would experience an unde'.ected f ailure during the period between six-month functional tests. To satisfy the long-term objective of maintaining an adequate luvel of safety throughout the plant lifetime, a minimum goal of 0 9999 at the 95 percant confidence level is pro p sed. With the (1-out-of-1) X (2) logic, this requires that each sensor have an availability of 0.993 at the 95 percent confidence level. This level of availabillty may be maintained by adjusting the test interval as a function of the observed failure history.1 To f acilitate the Laplementation of this technique, Figure 4.1-1 is
,ptovided to indicate an appropriate trend in test interval. The procedure is as follows:
- 1. Like sensors aru pooled into one group for the purpose of data acquisition.
- 2. The factor M is the exposure hours and is equal to the number of ,
sensors in a group, u, times the elapsed tLee T (M = nT).
- 3. The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1-1.
- 4. After a trend is established, the appropriate monthly test interval to satisfy the goal will be the test interval to the left of the plotted points.
- 5. A test interval of one month will generally be used initially until a trend is established.
Group (3) devices utilise an analog sensor followed by an amplifier and a bistable trip circuit. The sensor and amplifier are active components and a failure is almost always accompanied by an alarm and an indication of the source of trouble. In the event of failure, repair or substitution can start benediately. An *as-is" failure is one that $ sticks" mid-scale and is not capable of going either up or down in response to an out-of-1Laits input. This type of failure for analog devices is a rare occurrence and is detectable by an operator who observes that one signal does not track the other three. For purp se of analysis, it is assumed that this rare failure will be detected within two hours.
- 1. Reliability of Engineered Safety Features as a Punction of Testing Frequency, I. M. Jacobs, " Nuclear Safety," Vol. 9, No. 4, July-August, 1968, pp. 310-312.
"[{ga 3'*/'**~18 AMEN 0 MENT NO. 2 3 8
[ 4.1 34&31 (cont'd) The bistable trip circuit which is a part of the Group (B) devices can i sustain unsafe failures which are revealed only on test. Therefore, it is necessary to test them periodically. A study was conducted of the instrumentation channels included in the Gron - (R) devices to calculate their
- unsafe" f ailure rates. The analog devices (sanoors and amplifiers) are predicted to have an unsafe failure rate of lets than 20 x 10-6 failure / hour. The bistable trip circuits are predicted to have unskfe failure rate of less than 2 x 10*8 failures /bour. Considerfng the two hour monitoring interval for the ans1% dovk9s as assumed above, and a weekly test interval for the histuST.C. tdp circuits, 'the desigt: reliability goal of 0.99999 is
, att6hteJ with angla margin.
The bistable dulces are monitored during plant operation to record their failure history and establish a test interval using the curve of Figure 4.1-1< -There are numerous identical histable devices used throughout the plant's instrumentation system. Therefore, significant , data on the failure rates for the bistable devices should be accumulated rapidly. Group (c) devices are active only during a given portion of the - operational cycle. For example, the IRM is active during the STARTUP/NOT STAND 3Y and REFVEL (with any control rod withdrawn from a core cell containing one or more fuel assemblies) Modes and inactive during full-power operation. Thus, the only test that is meaningful is the one performed prior to entering-the applicable Mode (i.e., the tests that us performed prior to use of the instrument) . since testing of the IRM functions is not practical in the RUN Mode, testing is not required to be coupleted until 12 hours after entering the STARTUP/ HOT STAND 3Y Mode from the RUN Mode. Twelve hours is based on operating experience and in consideration of providing reasonable tima in which to couplete the test. Calibration frequency of the instrument channel is divided into two groups. These are as follows:
- 1. Passive type indicating devices that can be compared with like units on a continuous basis.
- 2. Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate. For those devices which esploy. amplifiers, etc., drif t specifications call for drift to be loss than 0.4 percent / months i.e., in the period of a month a drift of 0.4 percent would occur thus providing for adequate margin. SFN 3.1/4.1-19 heardment No. 249 Unit 2
4 j 4.1 R&gEA (cont'd) For tM APRM system drift of electronic apparatus is not the only l e% sideration in determining a calibration frequency. Change in power j distribution and loss of chamber sensitivity dictate a calibration every [ , seven days. Calibration on this frequency assures plant operation at or <
- below thermal limits. . .
3 A conparison of Tables 4.1.A and 4.1.3 indicates that two instrument ! }' channels have been included in the latter table. These are mode j switch in sNUTDOWN and manual scram. All of the devices or sensors associated with these scram functions are simple on off switches and, i l hence, calibration during operation is not applicable, i.e., the switch is either on or off. 1 I . . : The APRM and 2 out-of 4 voter channel hardware are provided with a self- i i test capability which automatically checks most of the critical hardware at least once per 15 minute interval whenever the APRM channel is in the operate mode. This provides a_ virtually continuous monitoring of the , essential APRM trip functions. In the event a critical fault is ] detected, an " inoperative
- trip signal results. A fault detected in i
non-critical hardware results in an " inoperative
- alarm. Following receipt of an
- inoperative" trip or alarm signal, the operator can enploy numerous diagnostic testing options to locate the problem.
- - The automatic self-test function is supplemented with a manual APRM trip i functional test, including the 2 out-of-4 voter channels and the 1- interface with the RPS trip systems. In combination with the virtually j continuous self-testing, the manual APRM trip functional test provides
! adequate functional testing of the APRM trip function. Therefore, the
- six-month test frequency for the manual testing provides an acceptable level of availability of the APRM.
l In addition to the above tests, the 2-out-of-4 voter is used to test the j RPS scram contactors. The cutput of each voter channel is tripped to , ! produce a scram signal into esch of the RPS trip system channels (A1, l A2, 31, and 32) to individually operate the respective scram contactors. > i The weekly test interval provides an acceptable level of availability of i the scram contactors. - Each APRM receives the output signals from two analog differential ! pressure flow transducers, one associated with recirculation loop A and ! the other with recirculation loop 3. These differential pressure i signals are converted into representative digital loop flow signals l within the same hardware that performs the APRM functions and are addad L to determine a total recirculation flow. The total recirculation flow v.lue is used by the APRM to determine the flow biased setpoints. Each total recirculation flow signal developed by an APRM is compared in the l hardware that performs the RBM functions to the signals from the l remaining three APRMs. An alarm is given if a preset compare level l setpoint is exceeded. The flow processing is integrated with the APRf
- processing and is covered by the same self-test and alarm functions i
l 1 BFN 3.1/4.1-20 Amendment No. 249 Unit 2 4 w- r ,- ,- e w ~- , - . . ~ . e--,e -.-w,,e- .-w e~re -w-e*w, .-w+--- -reee-e --t- w ea -*-+w+--w-ee-w- - _ e ---n - ev ~'-w--*--------=w--,r--te=erm~w-e * --
_ __ . _ _ . _ _ . _ . _ _ . _ _ _ _ _ . - ~ . _ - _ - - - - - - _ _ . _ . _ _ . . . _ _ . _ . ___ _ _ . _ _ _ 4.1 34453 (cont'd) described earlier. As a 16 nit of the virtually continuous monitoring of the equipment performing the flow processing and the automatic covarison of redundant flow signals, it is acceptable to calibrate this equipment once per operating cycle. 4 The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. The APRM system, which uses the LPRM readings to detect a change in themal power, will be calibrated every seven days using a heat balance to cogensate for this change in sensitivity. The RBM system uses the LPRM reading to detect a localised change in themal power. It applies a correction factor based on the APRM output signal to determine the percent themal power and
- therefore any chage in LPRM sensitivity is co gensated for by the APRM calibration. The technical specification limits of CPR and APLMGR are d determined by the use of the process coquter or other backup methods.
These methods use LPRM readings and TIP data to detemine the power distribution. compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the
- computer calculated LPRM correction factors every 1000 effective full power hours.
As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power. s BFN 3.1/4.1-21 haendment No. 249 Unit 2
l 3.2/4.2 PRQT3CTIVE INSTRUMENTATION I LIMITINC evwnITIONE POR OPERATION SURVIItt.auCE Rgouf n utstTS 3.2 Protmetive Instr"==ntation 4.2 Protective Instr"==ntation Aeolicability &gplicability Applies to the plant instrumentation Applies to the surveillance which initiates and controle requirement of the instrumentation a protective function. that initiates and controls protective function. Obiective Obiective
. To assure the operability of To specify the type and protective instrumentation. frequency of surveillance to be applied to protective instrumentation.
Knacification specification A. ggimary containment and Raaetor A. Primary Containment and Buildina Isolation Functions Reactor Buildina Isolation Functions When primary containment -Instrumentation shall be integrity is required, the functionally tested and ILuiting conditions of calibrated as indicated operation "or the instrumen- in Table 4.2.A. tation that initiates primary containment isolation are System logic shall be given in Table 3.2.A. This functionally tested as includes instrumentation that indicated in Table 4.2.A. initiates isolation of the reactor vessel, reactor building, main steam lines, and initiates the standby gas treatment system.
- 3. core and containment coolina 3. core and contL1DERA1 Svatama - Initiation & Control ggolino Evat=== =
Initiation & Control The limiting conditions for Instrumentation shall be operation for the functionally tested, instrumentation that initiates calibrated and checked as or controls the core and indicated in Table 4.2.5. containment cooling systems are given in Table 3.2.3. BFN 3.2/4.2-1 Unit 2
1.2/&.2 PROTeeffVR f MB TattwouTATTaar fTMtTTMa f Tf aMa 703 DDenhTYoal , atfRVR T T T huet platif newsw?B
),2,3 card and Pantalement Paalina 4,2,3 Sara and Pantalement Pnalina Evat--- = fnitiation & centrol Evat=== = fnitiation & Eantrol (Cont'd) (Cont'd)
This instrumentation must be System logic shall be OPERABLE when the systes(s) it functionally tested as initiates or controls are indicated in Table 4.2.3. required to be OPERABLE as
, specified in section 3.5. Whenever a system or loop is made INOPERABLE because of a required test or calibration, the other 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 that they are INOPERABLE. C. enntrol und alack ietuation C. cnntrol med alack ietuatien The 1Laiting conditions of Instrumentation shall be operation for the functionally tested, instrumentation that calibrated and checked as initiates control rod block indicated in Table 4.2.C. are given in Table 3.2.C. system logic shall be functionally tested as I indicated in Table 4.2.C. r 3FN 3.2/4.2-2 Amendment No. 249 Unit 2
....i..
g 88 TABLE 3.2.c pk INSTRUMENTATION THAT INITIATES ROD BLOCKS Minimum @ rable . Channels Per Trin Panctlan (M banction Trin kiel Earttina 3 (1) APRM Upscale (Flow Bias) , (2) 3 (1) APRM Upscale (Startup Model (8) s12% , l 3(1) APRM Downscale (9) 23% l 3 (1) APRM I6eoperative (1Db) l I 2 (7) RBM Upscale (Power Bias) Im Power Range (13) (14) Interimediate Power Range (13) (14) I High Power Range (13) (14) l 2 (7) RBM Downscale (9) (13) (15) l
" 2 (7) RBM Inoperative (10c) 6 (1) IRM Upscale (8) s108/125 of full scale u
$ 6 (1) IRN Downscale (3)(8) 25/125 of full scale m
6(1) IRM Detector not in Startup Position (8) (11) 6 (1) IRM Inoperative (8) (10a) 3 (1) (6) SRM Upscale (8) 1 1K105 counts /sec. 3 (1) (6) SRM Downscale (4) (8) 23 counts /sec. 3 (1) (6) SRM Detector not in Startup Position (4) (8) (11) 3 (1) (6) SRM Inoperative (8) (loa) 1 Rod Block togic N/A y h 1(12) Migh Water Level in West s25 gal. Scram Discharge Tank
$ (LS-85-45L) 1(12) High Water level in East s25 gal.
Scram Discharge Tank (LS-85-45M) l
- l
l l i THIS PAGE INTENTIONALLY LIFT BLANK l s 4 arx s.2/4.2-2s. AMENDMENT NO. 212 Unit 2
o gIts rom TAatr 3.2.c
;1. The minimum number of OPERABLE channels for each trip function is detailed for the STARTUP and RUN positions of the reactor mode selector switch. The SRM, IRM, and TPRM (STARTUP mode), blocks need not be OPERABLE in "RUN" mode, and che APRM (flow biased) rod blocks need not be OPERABLE in "STARTUP" mode.
With the number of OPERABLE channels less than required by the minimum CPERABLE channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.
- 2. The trip level setting shall be as *specified in the CORE OPERATING LIMITS REPORT.
- 3. IRM downscale is bypassed when it ir on its lowest range.
- 4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range 2. SRMs B and D downscale function is bypassed when IRMs B, D, r, and H are above range 2.
SRM detector not in startup position is bypassed when the count rate is 2100 CPS or the above condition is satisfied.
- 5. During repair or calibration of equipment, not more than one SRM, RBM, or APRM channel nor more than two IRM channels may be bypassed.
Bypassed channels are not counted as OPERABLE channels to meet the minimum CPEhABLE channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.
- 6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.
IRM channels B, r, D, H all in range 8 or above bypasses SRM channels B and D functions.
- 7. The following operational restrcants apply to the RBM only.
- a. Both RBM channels are bypassed when reactor power is 130 percent or when a peripheral (edge) control rod is selected.
- b. The RBM need not be OPERABLE in the 'startup" position of the reactor mode selector switch,
- c. The RBM need not be OPERABLE if either of the following two conditions is mett
, (1) Reactor thermal power is 190 percent of ratad and MCPR is l 21.44, or I (2) Reactor thernal power is <9s percent of rated and MCPR is 11.75. BrN 3.2/4.2-26 h h it No. 249 Unit 2 _ l
wmts pen TAata 1.2.c (cont'd)
- d. Two RBM channels are provided and only one of these may be bypassed l i with the console selector. If the inoperable channel casmot be restored within 24 hours, the inoperable channel shall be placed in the tripped condition within one hour,
- e. With both RSM channels inoperable, place at least one inoperable l rod block monitor channel in thS tripped condition within one hour.
- 8. This function is bypassed when the mode switch is placed in RUN.
- 9. This function is only active when the mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is
. OPERABLE and not high.
- 10. The inoperative trips are produced by the fu11owing functions:
- a. SRM and IRM (1) Local
- operate-calibrate" switch not in operate.
(2) Power supply voltage low. (3) circuit boards not in circuit.
.b. APRM (1) Local APRM chassis mode switch not in operate. l (2) Less than the required minimum number of LPRM inputs, both total and per axial level.
(3) APRM module unplugged. l (4) self-test detected critical fault. l
- c. RBM (1) Local RBM chassis mode switch not. in operate. l (2) RBM module unplugged. l (3) RBM fails to null.
(4) Less than required number of LPRM inputs for rod selected. (5) self-test detected critical fault. !
- 11. Detector traverse is adjusted to 114 A 2 inches, placing the detector lower position 24 inches below the lower core plate.
SFN 3.2/4.2-27 haenhent No. 249 Unit 2
NOTEE FOR TA.RLP M J (Cont'd)
- 12. This function may be bypassed in the SHUTDOWN or REFUEL mode. If this function is inoperable at a time when OPERABILITY is required the channel shall be tripped or administrative controls shall be imediately i.mposed to prevent control rod withdrawal.
- 13. The RBM rod block trip setpoints and applicable power ranges are ,
specified in the CORE OPERATING LIMITS REPORT (COLR). ')
- 14. Less than or equal to the setpoint. allowable value specified in the COLR.
- 15. Greater than or equal to the setpoint allowable value specified in the COLR.
9 BFN 3.2/4.2-27a Amendment No. 249 Unit 2
. THIS PAGE INTDTTIONALLY LEFT BLANK G
h 4 5 ( I BFN 3.2/4.2-27b Unit 2
... . _ . , m _ . ._ _ _ . . _ . . . _. . . . . . ~ - . . . . . - . . . . _ . - . . . . - . . . . . h# TABLE 4.2.c y SURVEILIANCE REQUIREMENTS FOR INSTRIMENU.fION THAT INITIATE ROD BLOCKS Mmetion Mmet l a== 1 Test calibration (171 Instrument C2nack APRM Upscale (Flow Bias) (1) (13) once/ operating cycle once/ day (8) l APRM Upscale (Startup Mode) (1) (13) once/ operating cycle ~ once/ day'(8) l APRM Downscale . (1) (13) once/ operating cycle once/ day (8) , l , APRM Inoperative (1) (13) N/A once/ day (8) , RBM Upecale (Power Blas) (1) (13) once/ operating cycle N/A l RBM Downscale (1) (13) once/ operating cycle N/A l REM Inoperative (1) (13) N/A N/A l IRM Upscale (1) (2) (13) once/3 months once/ day (8) IRM Downscale (1) (2) (13) once/3 months once/ day (8) Y Q IRM Detector Not in Startup Position (2) (once operating cycle) once/ operating cycle (12) N/A a , 9J IRM Inoperative (1) (2) (13) N/A N/A E O SRM Upscale (1) (2) (13) once/3 months once/ day (8) SRM Downscale (1) (2) (13) once/3 months once/ day (8) SRM Detector Not in Startup Position (2) . (once/ operating cycle) once/ operating cycle N/A SRM Inoperative (1) (2) (13) N/A N/A Rod Block Logic (16) N/A N/A y West Scram Discharge once/ quarter' once/18 months N/A Tank Water Level High h (LS-85-45L)
% East Scram Discharge once/ quarter once/18 months N/A
@ Tank Water level High i (LS-85-45M) i
r i THIS PAGE INTENTIONAI.1,Y 1IFT BI.hNK l AMENDMEM NO. 212 3.2/4.2-soa SnSt2
NOTtt FOR TARfPC 4.2.1 TMR&UOM 4.2.L exceet 4.2.D iMD 4.2.K
- 1. For IRMs and SRMs, functional tests shall be performed ence per month.
For APRMs and RBMs, functional tests shall be performed once per 6 months.
- 2. Punctional tests shall be performed before aach startup with a required frequency not to exceed once per week.
- 3. This instrumentation is excepted from the functional test definition.
The functional test will consist of injecting a simulated electrical signal into the measurement channel.
- 4. Tested during logic system functional tests.
- 5. Refer to Table 4.1.B.
- 6. The logic system functional tests shall include a calibration ence per operating cycle of time delay relays and timers nece? -ary for proper functioning of the trip systems. *
- 7. The functional test will consist of verifying continuity across the inhibit with a volt-chmmeter.
- 8. Instrument checks shall be performed in accordance with the definition of instrument check (see section 1.0, Definitions) . .An instrument check is not applicable to a particular setpoint, such as Upscale, but is a qualitative check that the instrument is behaving and/or indicating in an acceptable manner for the particuler plant condition. Instrument check is included in this table for convenience and to indicate that an instrument check will be performed on the instrument. Instrument checks are not required when these instruments are not required to be OPERABLE or are tripped.
- 9. Calibration frequency shall be once/ year.
- 10. Deleted
- 11. Portion of the logic is functionally tested during outage only.
- 12. The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.
- 13. Functional test will consist of applying simulated inputs (see note 3) .
Local alarm lights representing upscale and downscale trips will be verified, but no rod block will be produced at this time. The inoperative trip will be initiated to produce a rod block (SRM and IRM inoperative also bypassed with the mode switch in RUN). The functions that cannot be verified to produce a rod block directly will be verified during the operating cycle. BFN 3.2/4.2-59 Amendment No. 249 Unit 2
ucrrra rom tant.we 4 .2 .A tirn otm x a .2 . L ax e me t 4. 2.n nrn 4. 2. r (Cont ' d)
- 14. (Deleted)
- 15. (Delete'd) q
- 16. Performed during operating cycle. Portions of the logic is checked more frequently during functional tests of the functions that produce a rod block.
17 This calibration consists of removing the function from service and performing an electronic calibration of the channel.
- 18. Functional test is limited to the condition where secondary containment
. integrity is not requ.t ed as specified in sections 3.7.C.2 and 3.7.C 3.
- 19. Functional test is limited to the time where the SGTS is required to meet the requirements of section 4.7.C.1.a.
- 20. (Deleted)
-{
- 21. Logic test is limited to the time where actual operation of the equipment is permissible.
- 22. (Deleted)
- 23. (Deleted)
- 24. This instrument check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).
- 25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer response due to mechanical excitation in the vicinity of the sensor.
BFN 3.2/4.2-60 Amendment No. 249 Unit 2
3.2 3&&11 (Cont'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 repaire 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 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 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. Each trip system consists of two channels. Each channel contains one temperature switch located in the l 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 paid and the RCIC high steam flow trip setting of 450" H 2 O have been selected such that the trip 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 CTR 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 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 occurs, the cleanup system is isolated. BFN 3.2/4.2-67 TS 370 Unit 2 Letter Dated 11/17/95
8 i, 3.2 3 EEE (Cont'd) l 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. The control rod b'Acck functions are provided to generate a trip signal to block rod withdrawal if the monitored power level exceeds a preset value. The trip logic for this function is 1-out-of-n e.g., any trip on one of four APRMs, eight IRMs, or four SRMs will result in a rod block. l When the RBM is required, the minimum instrument channel requirements apply. These requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods. The APRM rod block function is flow biased and provides a trip signal for blocking rod withdrawal whin average reactor thermal power exceeds pre-established limits set to prevent scram actuation. The RBM rod block function provides local protection of the cores i.e., the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern. If the IRM channels are in the worst condition of allowed 'aypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10. A downscale indication is an indication the instrument has failed or the 3 instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented. The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position. For effective emergency core cooling for small pipe breaks, the NPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the NPCI does not operate. The arrangement of the tripping contacts "s such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are BFN 3.2/4.2-68 Amendment No. 249 Unit 2
4.2 A&AEA (cont'd) (7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16, 1968, page 10, Equation (24), Lawrence Radiation Laboratory. Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently. These one-out-of-n trip systems will be tested one at a time in order to take advantage of this inherent improvement in availability. optimizing each channel independently may not truly optLaise the system considering the overall rules of system operation. However, true system
, optistnation is a complex problem. The optimums are broad, not sharp, and optimising the individual channels is generally adequate for the system.
The formula given above minimises the unavailability of a single channel which must be bypassed during testing. The minimination of the unavailability is illustrated by Curve No.1 of Figure 4.2-1 which assumes that a channel has a failure rate of 0.1 x 10~ / hour and 0.5 hours is required to test it. The unavailability is a mintmum at a test interval 1, of 3.16 x 103 hours. If two similar channels are used in a 1-out-of-2 configuration, the test interval for minimum unavailability changes as a function of the rules for testing. The simplest case is to test each one independent of the other. In this case, there is assumed to be a finite probability that both may be bypassed at one time. This case is shown by curve No. 2. Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval. Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability. A more usual case is that the testing is not done independently. If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3. Note that the minimum occurs at about 40,000 hours, much longer than for cases 1 and 2. Also, the minimum is not nearly as low as case 2 which indicates that this method of testing does not take full advantage of the redundant channel. Bypassing both channels for simultaneous testing should be avoided. The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested, and restored. This is shown by curve No. 4. Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel. BFN Unit 2 3.2/4.2-73l AMENDMENT NO. 210
9 4.2 R1112 (Cont'd)
, The best test procedure of all those examined is to perfectly stagger the i tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5. The difference between cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human error.
The conclusions to be drawn are these:
- 1. A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and
. 2. more than one channel should not be bypassed for testing at any one time.
The radiation monitors in the reactor and refueling zones which initiate building isolation and standby gas treatment operation are arranged such that two sensors high (above the high level setpoint) in a single channel ' or ine sensor downscale (below low level setpoint) or inoperable in two channels in the same zone will initiate a trip function. The functional testing frequencies for both the channel functional test and the high voltage power supply functional test are based on a Probabilistic Risk Assessment and system drift characteristics of the Reactor Building ventilation Radiation Monitors. The calibration frequency is based upon the drift characteristics of the radiation monitors. The automatic pressure relief instrumentation can be considered to be a 1-out-of-2 logic system and the discussion above applies also. The RCIC and HPCI system logic tests required by Table 4.2.B contain provisions to demenstrate that these systems will automatically restart on a RPV low water level signal received subsequent to a RPV high water level trip. The electronic instrumentation comprising the APRM rod _ block and Rod Block Monitor functions together-with the recirculation flow instrumentation for flow bias purposes is monitored by the same self-test functions as applied to the APRM function for the RPS. The functional test frequency of every six months is based on this automatic self-test monitoring at 15 minute intervals and on the low expected equipment failure rates. -Calibration frequency of once per operating cycle is based on the drif t characteristics of the limited number of analog components, recognizing that most of the processing is performed digitally without drift of setpoint values. BFN 3.2/4.2-73a Arwedr=nt No. 249
-Unit 2
3.3/4.3 REACTIVITY CONTROL LIMITING NwnfTIONS FOR OPERATION SURVEILLANCE REQUIRDMNTS 3.3.5. Control Rods 4.3.3. Control Rods 3.b (Cont'd) 3.b.2 The Rod Worth Minimiter (RWM)
- 3. Should the RWN become shall be inoperable on a shutdown, demonstrated to be shutdown may continue OPERABLE for a reactor provided that a second shutdown by the licensed operator or other folicwing checks:
technically qualified member of the plant staff is present a. By demonstrating at the console verifying that the control
. compliance with the rod patterns and prescribed control rod Banked Position program. Withdrawal sequence (or equivalent) input to the RwM computer are correctly loaded following any loading of the program into the computer.
- b. Within 8 hours prior to RWM automatic initiation when reducing thermal power, verify proper annunciation of the selection error of at least one out-of-sequence control rod,
- c. Within one hour after RWM automatic initation when reducing thermal power, the rod block function of the RWM shall be verified by moving an out-of-sequence control rod.
"Et 2 AMENDMENT NO. 212 l
0
- 3. 3 /4.1 REA MTVTTY OONTROL LIMITING CONDITIONE FOR ODERATION StfRVETLLANCE REOUTREMENTE 3.3.B. cont'rol Rode 4.3.3. control Rode 3.c. If Specifications 3.3.B.3.b.1 3.b.3 When the RWM is through 3.3.B.3.b.3 cannot not OPERABLE a be met the reactor shall second licensed not be started, er if the operator or other reactor is in the run or technically qualified startup modes at less than member of the plant 10% rated power, control rod staff shall verify movement may be only by that the correct
, actuating the manual scram rod program is or placing the reactor mode followed, position.
- 4. Control rods shall not be 4. Prior to control withdrawn for startup or rod withdrawal for' refueling unless at least startup or during two source range channels refueling, verify that have an observed count rate at least two source equal to or greater than range channels have three counts per second. an observed count rate of at least three ccants per second, f 5. (Deleted) 5. (Deleted) d BFN 3.3/4.3-8 Amendment No. 249 Unit 2
I 3.3/4.3 RLHA (Cont'd) 5.. (Deleted) - C. meram fnaartion Timaa The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fusi damage; i.e., to prevent the MCPR from becoming less than the Safety Limit MCPR. The limiting power transients are given in Reference 1. Analysis of these transients shows that the negative reactivity rates resulting from the scram with the average response of all drives as given in the above specifications provide the required protection and MCPR remains greater than the Safety Limit MCPR. On an early BWR, some degradation of control rod scram performance occurred during plant STARTUP and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDR1443) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked. The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB1443) hae been demonstrated by a series of engineering tests under simulated reactor operating conditions. The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plante using the older model Bnl 3,3/4.3-17 Amendaent No. 249 Unit 2
3.3/4.3 &&&R1 (cont'd) drive with a modified (larger screen site) internal filter which is less prone to plugging. Data has been documented by surveillance reports in various operating plants. These include oyster creek, Monticello, Dresden 2, and Dresden 3. Approximately 5000 drive tests have been recorded to date. l Following identification of the " plugged filter" problem, very frequent scram tests were necessary to ensure proper performance. However, the more frequent scram tests are now considered totally unnecessary and unwise for the following reasons:
- 1. Erratic scram performance has been identified as due to an obstructed drive filter in type "A" drives. The drives in BFNP are of the new
. "B" type design whose scram performance is unaffected by filter condition.
- 2. The dirt load is primarily released during STARTUP of the reactor when the reactor and its systeme are first subjected to flows and pressure and thermal stresses. special attention and measures are now being taken to assure cleaner systems. Reactors with drives identical or similar (shorter stroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram performance. This preoperational and STARTUP testing is sufficient to dotect anomalous drive performance.
- 3. The 72-hour outage limit which initiated the start of the frequent scram testing is arbitrary, having no logical basis other than quantifying a " major outage" which might reasonably be caused by an event so severe as to possibly affect drive performance. This requirement is unwise because it provides an incentive for shortcut actions to hasten returning "on lire" to avoid the additional testing due a 72-hour outage.
BFN 3.3/4.3-1B TS 370 Unit 2 Letter Dated 11/17/95
3 . E /4 . E cost AND cONTgtWMENT COOLTWO BYETEMS LIMITING CONDITTONS FOR OPERATION SURVETLLANCE REOUTRhtENTS 3.5.I Avera'an pinnar Linear Heat 4.5.I Averaen planar tin.ar Heat Ganaration Rate Generation Rate (ADLMCR) During steady-state power The APLHGR shall be checked operation, the Average Planar daily during reactor Linear Heat Generation Rate operation at 1 25% rated (APLHGR) of any fuel assembly thermal power, at any axial location shall not exceed the appropriate rated, flow-dependent or power-
. dependent APLHGR limit provided in the CORE OPERATING LIMITS REPORT. If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is '
being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the COLD SHUTDOWN CONDITION within 36 hours. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. J. Linear Heat Generation Rate fLMCR) J. Linear Heat Generation Rate (LHCR) During steady-state power operation. The LHGR shall be checked the linear heat generation rate daily during reactor fuel (LHGR) of any rod in any fuel operation at 1 25% rated assembly at any axial location shall thermal power. not exceed the appropriate LHGR limit provided in the CORE OPERATING LIMITS REPORT. If at any time during operatien it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is BFN 3.5/4.5-18 Amendnumt No. 249 Unit 2
4 3.E/d.E coWW AMM OOMTATWMENT cooLTMG gYstEMR
- LTMTTTMd counfTinum POR Open1TIOM gtygygtf.f.1Mcy smotf7RhTii i'
J, Linmar Ma'at Manaration Rata (LMMB) J. Linaar Meat Sanaration Rata
.UE311.
3.5.J (Cont'd) not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the COLD SHUTDOWN CONDITION within 36 hours.
. Surveillance and corresponding action shall continue until reactor operation is within the-prescribed limits.
3.5.K Minimum critieni power Ratio 4.5.K Minien= critical pn (MCPR) 88tio fMC881 s The minimum critical power ratio 1. MCPR shall be checked daily (MCPR) shall be equal to or during rcactor power greater than the appropriate operation at 2,25% rated rated, flow-dependent or power- thermal power and following dependent operating limit MCPR any change in power level (OU4CPR) as provided in the CORE or distribution-that would OPERATING LIMITS REPORT. If at cause operation with a any time during steady-state LIMITING CONTROL ROD operation it is determined by PATTERN.
-normal surveillance that the limiting value for MCPR is being 2. The operating limit MCPR l exceeded, action shall be shall be determined as i initiated within 15 minutes to provided in the CORE restore operation to within the OPERATING LIMITS REPORT prescribed limits. If the using:
steady-state MCPR is not returned to within the Frescribed limits within two (2) a. T as defined in the hours, the reactor shall be CORE OPERATING LIMITS brought to the COLD SHUTDOWN REPORT prior to initial CONDITION within 36 hours, scram time measurements surveillance and corresponding for the cycle, performed action shall continue until in accordance with reactor operation is within the Specification 4.3.C.1. prescribed limits,
- b. T as defined in the CORE OPERATING LIMITS REPORT following the conclusion of each-scram-time surveillance BFN 3.5/4.5-19 Unit 2 Amendment No. 249
1.$/4.5 come hun coWTATWWENT coottWS BYSTEun LIMITINS cOMBITIONE FOR OPERATTSW BURVEfLLANCE REOtff REMENTE I 3.5 care and' containment coeline sukant 4.5 core and contain=-nt (Cont'd) coolina svata== (Cont' d)
- 4. 5.K (Cont' d) test required by specifications 4.3.C.1 and 4.3.c.2.
The determination of the limit must be completed within 72 hours of each scram-time surveillance required by specification 4.3.c. L. APRM Setpoints L. 3DJM seteeints h (Deleted) (Deleted) d M. core Thermal-Hvdraulic stability M. core Thermal-Hvdraulie stability
- 1. The reactor shall not be 1. Verify that the reactor is operated at a thwrmal power outside of Region I and II and core flow inside of of Figure 3.5.M-1 Regions I and II of Figure 3.5.M-1, a. Following any increase of more than 5% rated
- 2. If Region I of Figure 3.5.M-1 thermal power while is entered, insnediately initial core flow is initiate a manual scram, less than 50% of rated, l
and
- 3. If Region II of Figure 3.5.H-1 b. Following any decrease is entered of more than 10% rated core flow while initial thermal power is greater than 40% of rated.
BFN 3.5/4.5-20 Amendment No. 249 Unit 2 i 1
\. 3.5/4.5 CDR2 AND CONTAINMENT COOLIMo SYSTEMS , LIMfTIMG CONDITIONS FOR OPERATION SUR':"EILLANCE REOUTREMENTS 3.5 core and containment coolina Syst-- 4.5 core and containment Coolino Systema 3.5.M.3. (Cont'd)
- a. Immediately initiate action and exit the region within 2 hours by inserting control rods or by increasing core flow (starting a recircu-lation pump to exit the region is nQ1 an appropriate
, action), and
- b. While exiting the region, immediately initiate a manual scram if thermal-hydraulic instability is observed, as evidenced by APM oscilla- i tions which exceed 10 percent peak-to-peak of rated or LPM oscillations which exceed 30 percent peak-to-peak of scale. If periodic LPM upscale or downscale alanns occur, immediately check the APM's and individual LPM's for evidence of thermal-hydraulic instability.
's UN 3.5/4.5-20a Unit 2 AMENDMENT NO.17 4
- w. .
l l Figure 3.5.K-1 DELETED ,, BFN 3.5/4.5-22 Unit 2 AMENDMENT NO. 214 l
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o 3.5 3M33 (Cot.t'3) With-the RCICs inoperable, a seven-day period to return the system to service 16- jkJtified based on the availability of the HPCIS to cool the core and upon consideration that the average risk associated with f ailure of the LOICS to cool the core when required is not increased. The surveillance requirements, which are based on industry codes and standards, provide adequate ausurance that the RCICs will be OPERABLE when required. 3.5.0 Automatie Denressurination avatam r?3an The ALs consists of six of the thirteen relief valves. It is designed to provide depressurisation of the reactor coolant system during a amall break loss of coolant accidsnt (LOCA) if HPCI fails or is unable to maintain the required water level in the reacter vocool. 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 3.5.0 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 possib'e for any number of the valves assigned to the ads to be incapable of performing their ADS functions because of instrumentation failurer, 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 OPERABL3. By requiring l six valves to be OPERABLE, additional conservatism is provided to account for the possibility of a single failure in tlas 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 conditien is within the analyses for a small break LOCA and the peak clad temperature is well below the 10 CFR 50.46 limit. Ana'.ysis has shown 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 Discharam 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 minimise 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 EFN 3.5/4.5-30 TS 370 Unit 2 Letter Dated 11/17/95
3.5 34334 (Cont'd) I 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 RNR 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 RNR system with a discharge line not filled. In addition to thu 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 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 NPCI and RCIC pumps are aligned to the condensate storage tank, which is
. physically at a higher elevation than the NPCIS 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.
3.5.I. L/arman pinnar TAnnar name canaration mate ( Apr.unn i 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 CFh 50, Appendix K. The peak cladding temperature following a poteW : ^ t ins-of-coolant accident is primarily a function of the aver &9v F3 - ceratjon u te of
.all the rods of a fuel assembly at any axial Bits .. 'd h tnd y dependent secondarily on the rod-to-rod power %. nr. W.'-in an assembly, since expected local variations tu pm o epirticets a within a fuel assembly affect the calculated peak cid t%rature by less than
.t 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat genantica rata is sufficient to assure that calculated temperatures are within tM 10 CM 30 Appendix K limit.
At less than rated power conditions, the rated 7 4 %R limit is adjusted by a power-dependent correction factor, MAPFAC G). At less than rated flow conditions, the rated APLHGR limit is adjusted by a flow-dependent correction-factor, MAPFAC(F). The most limiting power-adjusted or flow-adjusted value is taken as the APLHGR operating limit for the off-rated condition. BFN 3.5/4.5-31 Anendment No. 249 Unit 2
i, 1 3.5 BASES (Cont'd) The flow-dependent correction factor, MAPFAC(F), applied to the rated APLHGR limit assures that (1) the 10 CFR 50.46 limit would not be exceeded during a LOCA initiated from less than rated core flow conditions and (2) the fuel thermal mechanical design criteria would be met during abnormal operating transients initiated from less than rated core flow conditions. MAPFAC (F) values are provided in the CORE OPERATING LIMITS REPORT. The power-dependent correction factor, MAPFAC(P), . applied to the rated APLHGR limit assures that the fuel thermal mechanical design criteria would be met during abnormal operating transients initiated from less
, than rated power conditions. MAPFAC(P) values are provided in the CORE OPERATING LIMITS REPORT.
3.5.J. Linear want canaration kata (tunn) 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 2 25 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LNGR to be a limiting value below 25 percent of rated thermal power, the largesc 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. Mini == critient power natic (Mcom) 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 des'ignated 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 NCPR when a limiting control rod pattern is approached ensures that-MCPR will be known following a change in power or power shape (regardless of magnitade) that could place operation at a thermal limit. At less than rated power conditions, a power-dependent MCPR operating limit, MCPR(P), is applicable. At less than rated flow conditions, a flow-dependent MCPR operating limit, MCPR(F), is applicable. The most limiting power-dependent or flow-dependent value is taken as the MCPR operating limit for the off-rated condition. BFN 3.5/4.5-32 Amen &nent No. 249 Unit 2 I
3.5- RASE 3 (C%nt'd) l The flow-dependent limit, MCPR(F), provides the thermal margin required to p!otect the fuel from transients resulting from inadvertent core flow incr eases. MCPR(F) values are provided in the CORE OPERATING LIMITS REPJRT. The power-dependent limit, MCPR(P), protects the fuel from the other limiting abnormal operating transients, including localized events such as a rod withdrawal error. MCPR(P) values are provided in the CORE OPERATING LIMITS REPORT. s 3.5.L. APRM natnoints (5)eleted)' ~ 3.5.M. pore Thumm1 -Hvaraulie nenhil f tv 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 in 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 imediate 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. 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. EFN 3.5/4.5-33 haendment No. 2A9 Unit 2
3.5 BAsn3 (Cont'd) 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 proceding 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. Rafarences
- 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-TR81-01-A.
- 3. Generic Reload Fuel Application, Licensing Topical Report, NEDE -
240ll-P-A and Addenda.
- 4. GE Document NEDC-32484P, Rev. 1, S. K. Rhow and C. T. Young, " Browns Ferry Nuclear Plant Units 1, 2, and 3 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," February 1996.
- 5. GE Document GE-NE-513-01755-2, Rev. 1, S. K. Rhow and T. H. Chuang,
" Relaxation of Emergency Core Cooling System Parameters for Browns Ferry Nuclear Plant Units 1, 2, and 3 (Perfoca Program Phase I),"
February 1996. 4.5 cern and containment cooline svat=== surveillance preouancien 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-cnolant 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 and inlection valves in accordance with Specification 1.0.MM it deemed to be admquate 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. BFN 3.5/4.5-34l kmbmt No. 249 Unit 2
4.5 RAsss (Cont'd) i 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 Leo and the required surveillance testing for the system or loop shall apply. Avarman planar fMce. Luca. and McDa The APLHCR, LHCR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due te burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate. ' BFN 3.5/4.5-35l Anendment No. 249 Unit 2 {
total deep dose equivalent exposure received from external l sources shall be assigned to specific major work functions.
- b. Any mainotsam relief valve that opens in response to reaching its setpoint or due to operator action to control reactor pressure shall be reported.
6.9.1.3 MCMTHLY OPERATING REPORT t* Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office, to be submitted no later than the fifteenth of each month , following the calendar month covered by the report. A narrative summary of operating experience shall be submitted in the above schedule. 6.9.1.4 REPORTABLE EVENTS Reportable events, including corrective actions and measures to prevent re-occurrence, shall be reported to the NRC in accordance with Section 50.73 to 10 CFR 50. 6.9.1.5 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. A single submittal may be made for a multi-unit station. The report shall include susunaries, interp etations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the OOCH and (2) Sections IV.B.:2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. BrN 6.0-26 AMENDMENT NO. 2 2 0 Unit 2
I
- l.
- 6.9.1.6 SOURCE TESTS Results of required leak tests performed on sources if the tests
' reveal the presence of 0.005 microcurie or more of removable contamination.
6.9.1.7 CORE OPERATING LIMITS REPORT
- a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following (1) The rated APLHGR limit; the Flow-Dependent APLHOR Factor, MAPFAC(F); and the Power-Dependent APLHOR Factor, MAPFAC(P) for Specification 3.5.I.
(2) The LHGR' limit for Specification 3.5.J. l (3) The rated MCPR Operating Limit; the Flow-Dependent MCIP. Operating Limit, MCPR(F); and the Power-Dependent MCPR Operating Limit, MCPR(P) for Specification 3.5.K/4.5.K. (4) The APRM flow biased rod block trip setting for Specification 2.1.A.1.c and Table 3.2.C. O!) The RBM downscale trip setpoint, high power trip
<stpoint, intermediate power trip setpoint, low power trap satpoint, and applicable reactor thermal power rrmges for each of the setpoints for Table 3.2.C.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
h b BFN 6.0-26a haendment No. 249 Unit 2 .
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