ML20149K100

Proposed Tech Specs,Relocating RCS P/T Limits from TS to Proposed P/T Limits Rept IAW Guidance Provided by GL 96-03, Relocation of P/T Limit Curves & LTOP Sys Limits

ML20149K100
Person / Time
Site:	Farley
Issue date:	07/23/1997
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20149K099	List: ML20149K097 ML20149K100 ML20149K103 ML20149K105
References
GL-96-03, GL-96-3, NUDOCS 9707290250
Download: ML20149K100 (71)

Text

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i ENCLOSURE 3 Joseph M. Farley Nuclear Plant Unit 1

, Pressure Temperature Limits Repon

Technical Specification Changes  ;

i Pace Chance Instructions

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IRDEX

DEFINITIONS 1

A s' 'SECTION PAGE l 1.0 DEFINITIONS 1.1 ACTION........................................................... 1-1 1.2 AXIAL FLUX DIFFERENCE............................................ 1-1 1.3 CHANNEL-CALIBRATION.............................................. 1-1 1.4 CHANNEL CHECK.................................................... 1-1 j 1.5 CHANNEL FUNCTION TEST............................................ 1-1 i 1.6 CONTAINMENT INTEGRITY............................................ 1-2

, 1.7 CONTROLLED LEAKAGE............................................... 1-2 1.8 CORE ALTERATION.................................................. 1-2 1.9 DOSE EQUIVALENT I-131............................................ 1-2 1.10 E - AVE RAGE D I S I NTEG RATION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.11' ENGINEERED SAFETY FEATURES RESPONSE TIME......................... 1-3 1.12 FREQUENCY NOTATION............................................... 1-3

! 1.13 V6f5H (Deleted)...................... 1-3

1.14 IDENTIFIED LEAKAGE............................................... 1-3 j 1.15 L!^" 0 ' " art'ncT

Tn:nTMENT CYCTEM (Deleted)....................... 1-4

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1.16 ._.JOn C"'NOSC TC ".AO On0T

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GV6@6M6 (Deleted)...,............................................ 1-4

1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)........................... 1-4 1.18 OPERABLE - OPERABILITY........................................... 1-4 j' 1.19 OPERATIONAL MODE - MODE.......................................... 1-5

! 1.20 PHYSICS TESTS.................................................... 1-5 1

1.21 PRESSURE BOUNDARY LEAKAGE........................................ 1-5 .

l 1.21a-PRESSURE TEMPERATURE LIMITS REPORT (PTLR)....................... 1-5 l 5 1.22 PROCESS CONTROL PROGRAM (PCP).................................... 1-5 l 1.23 PURGE-FURGING.................................................... 1-5 l 1.24 QUADRANT POWER TILT RATIO................................. ...... 1-5 1.25 RATED THERMAL POWER.............................................. 1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME................................ 1-6 4

1.27 REPORTABLE EVENT................................................. 1-6 1

1.28 SHUTDOWN MARGIN.................................................. 1-6 f 1.29 GGb40!FICAT!ON (Deleted)......................................... 1-6 l 1.30 SOURCE CHECK..................................................... 1-6 j 1.31 STAGGEREO TEST BASIS............................................. 1-6 1 1.32 THERMAL POWER.................................................... 1-7

+

j. 1.33 UNIDENTIFIED LEAKAGE............................................. 1-7 1 1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................. 1-7

' 1.35 VENTING.......................................................... 1-7 L.

TABLE 1.1 OPERATIONAL MODES......................................... 1-8

-TABLE 1.2 FREQUENCY NOTATION........................................ 1-9 FARLEY-UNIT 1 I ANENDMENT No.

1 INDEX l ADMINISTRATIVE CONTROLS SECTION pAGE l

Review..................................................... 6-10 Audits..................................................... 6-l'1 Authority.................................................. 6-12 Records.................................................... 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL A c t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 12 Records.................................................... 6-13 6.6 REPORTABLE EVENT ACTION..................................... 6-14 6.7 SAFETY LIMIT VIOLATION...................................... 6-14 6.8 PROCEDURES AND PROGRAMS..................................... 6-14 6.9 REPORTING REOUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report............................................. 6-15a Annual Report.............................................. 6-16 Annual Radiologica] Environmental Operating Report..................................................... 6-17 Annual Radioactive Effluent Release Report..................................................... 6-17 Monthly Operating Report................................... 6-19 Radial Peaking Factor Limit Report......................... 6-19 Annual Diesel Generator Reliability Data Report..................................................... 6-19 Annual Reactor Coolant System Specific Activity Report............................................ 6-20 Annual Sealed Source Leakage Report........................ 6-20 Pressure Tempersture Limits Report (PTLR).................. 6-20a 6.9.2 SPECIAL REPORTS............................................ 6-20a 6.10 RECORD RETENTION........................................... 6-20a 6.11 RAD I ATION PROTECTION PROG RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 a 6.12 HIGH RADIATION AREA........................................ 6-22 FARLEY-UNIT 1 XIX AMENDMENT NO.

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l DEFINITIONS OPERATIONAL MODE - MODE I

1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESCURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System Component body, pipe wall or vessel wall.

PRESSURE TEMPERATURE LIMITS REPORT (PTLR) i 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P/T) limits, including heatup and cooldown rates, for the current reactor vessel  ;

fluence period. These P/T limits sball be determined for each fluence period '

or effective full power years (EFT'ss) in accordance with Specification .

6.9.1.15. Plant operation within these operating limits is addressed in LCO 3.4.10.1, RCS Pressure / Temperature Limits.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that.

processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; burial ground requirements; and other requirements governing the 1 disposal of solid radioactive waste. j l

PURGE - PURGING 1.23 PURGE or PURGING is the controlled proceus of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

FARLEY-UNIT 1 1-5 AMENDMENT NO.

l REACTOR COOLANT SYSTEM "

3/4.4.10 PRESSURE / TEMPERATURE LIMITS '

l REACTOR COOLANT SYSTEM =

LIMITING CONDITION FOR OPERATION I 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited _in accordance with the limits specified in the PRESSURE TEMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

I APPLICABILITY: At all times.

ACTION:

With any of the above limits specified in the PTLR exceeded, restore the l temperature and/or pressure to within the limit within 30 minutes; perform an  !

engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued ,

operation or be in at least iiOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the  !

RCS T avg and pressure to less than 200*F and 500 psig, respectively, within '

the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at least once per l hour during system heatup, cooldown, and inservice leak and hydrostatic  ;

testing operations. I

4. 4.1,0.1. 2 The reactor vessel material irradiation surveillance specimens l shall be removed and examined, to determine changes in material properties, as '

required by 10CFR50, Appendix H. i i

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l Reducing T avg to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary

] coolant is below the lift pressure of the atmospheric steam relief va'ves. l The surveillance requirements provide adequate assurance that excessive i

! specific activity levels in the primary coolant will be detected in sufficient

} time to take corrective action. Information obtained on iodine spiking will '

be used to assess the parameters associated with spiking phenomena. A .

reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.  !

k 3/4.4.10 PRESSURE / TEMPERATURE LIMITS I 1

s The temperature and pressure changes during heatup and cooldown are limited to

be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, asrequiredper10CFRPart50,AppendixG.l a

} 1) The reactor coolant temperature and pressure and system heatup and

cooldown rates (with the exception of'the pressurizer) shall be limited 5

in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR). 'l l

a) Allowable combinations of pressure and temperature for specific

, temperature change rates are below and to the right of the limit i

lines shown in the PTLR. Limit lines for cooldown rates between l j those presented may be obtained by interpolation.

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l b) The PTLR defines limits to assure prevention of nonductile failure l l

! only. For normal operation, other inherent plant characteristics, I i e.g., pump heat addition and pressurizer heater capacity, may i l limit the heatup and cooldown rates that can be achieved over j certain pressure-temperature ranges.

}'

g 2) These limit lines shall be calculated periodically using methods

provided in WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold l Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit i

. Curves. i

, 3) The secondary side of the steam generator must not be pressurized above

{ 200 psig if the temperature of the steam generator is below 70'F.

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REACTOR COOLANT SYSTEM BASES l

4) The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 36 effective .

full power years (EFPY) of service life. The 36 EFPY service life period is  !

chosen such that the limiting RTndt at the 1/4T location in'the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt assuroc that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

FARLEY-UNIT 1 B 3/4 4-7 ANENDMENT NO.

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a Values of ARTndt determined in accordance with WCAP-14040-t:P-A, Revision 2, l may be used until the next results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in the PTLR. The l j heatup and cooldown curves must be recalculated when the ART ndt determined j from the surveillance capsule exceeds the calculated ART ndt for the

! equivalent capsule radiation exposure.

j Allowable pressure-temperature relationships for various heatup and cooldown T

rates are calculated using methods derived from Appendix G in Section XI of l j the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR

, Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2.l l

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are a

provided to assure compatibility of operation with the fatigue analysis

performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater e than or equal to 2.85 square inches ensures that the RCS will be protected from j pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310*F.

Either RHR relief valve has adequate relieving capability to protect the RCS 3 from overpressurization when the transient is limited to either (1) the start j of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures provided measures are

taken to cushion the overpressure effects at RCS temperatures above 250*F, or (2) the start of all operable charging pumps and their injection into a water solid RCS. In the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operabis charging pumps allowed by the Technical Specifications.

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i 3/4.4.11 STRUCTURAL INTEGRITY

. The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 1

components ensure that the structural integrity and operational readiness of

. these components will be maintained at an acceptable leve) throughout the life t of'the plant. These programs are in accordance with Section XI of the ASME

{ Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR

+ Parr 50.55a(g) except where specific written relief has been granted by the-

, Commission pursuant to 10 CFR Part 50.55a(g)(6)(1).

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s FARLEY-UNIT 1 B 3/4 4-14 AMLNDMENT No.

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ADMINISTRATIVE CONTROLS l

l ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT l 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be l included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded (in graphic and tabular format);

(2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the l

limit. Each result should include date and time of sampling and the

! radiciodine concentrations; (3) Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro ci/gm) and one other radiciodine isotope concentration l l (micro ci/gm) as a function of time for the duration of the specific activity

• i above the steady-state level; and (5) The time duration when the specific l j activity of the primary coolant exceeded the radiciodine limit. I ANNUAL SEALED SOURCE LEAKAGE REPORT '

I 6.9.1.14 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fiasion detector leakage tests reveal the  !

l presence of greater than or equal to 0.005 microcuries of removable contamination.

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FARLEY-UNIT 1 6-20 AFFNDMENT NO.

ADMINISTRATIVE CONTROLS PRESSURE TEMPERATURE LIMITS REPORT fPTLR) 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewod and approved by the NRC, specifically those described in WCAP-14040-NP-A, Revision 2, " Methodology Used to Develop Cold overpressure Miticating System Setpoints and RCS Heatup and Cooldown Limit Curves," approved by NAC SER dated October 16, 1995.

The PTLR shall be provided ic the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Commission in accordance with the requirements of 10CFR50.4 within the time period specified for each report. Reports should be submitted to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555.

6.10 RECORD PETENTION i

j In addition to the applicable record retention requirements of Title 10, code j of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

3 6.10.1 The following records shall be retained for at least five years:

a. Records and logs of unit operation covering time interval at each

< power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to j nuclear safety.

c. ALL REPORTABLE EVENTS submitted to the Commission.

i d. Records of surveillance activities, inspections and calibrations j required by these Technical Specifications.

l e. Records of changes made to the procedures required by l Specification 6.8.1.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 1 6-20a AMENDMENT NO.

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1 l ENCLOSURE 4 4

Joseph M. Farley Nuclear Plant Unit 2 Pressure Temperature Limits Report Technical Specification Changes Pane Change Instructions Remove Page Replace Page -

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I I XIX XIX  ;

i 1-5 1-5 3/4 4-27 3/4 4-27 l 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 B 3/4 4-6 B 3/4 4-6 B 3/4 4-7 B 3/4 4-7 B 3/4 4-8 B3/44-8 B 3/4 4-9 B 3/4 4-9

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INDEX DEFINITIONS 4

SECTION PAGE 4

j 1.0 DEFINITIONS E

1.1 ACTION........................................................... 1-1 1 1.2 AXIAL FLUX DIFFERENCE............................................ 1-1 I

1.3 CHANNEL CALIBRATION.............................................. 1-1 1.4 CHANNEL CHECK.................................................... 1-1 l

1.5 CHANNEL FUNCTION TEST............................................ 1-1 i

1.6 CONTAINMENT INTEGRITY............................................ 1-2

1.7 CONTROLLED LEAKAGE............................................... 1-2 i

1.8 CORE ALTERATION.................................................. 1-2 i 1.9 DOSE EQUIVALENT I-131............................................ 1-2 1.10 E - AVE RAGE D I S INTEG RATION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME......................... 1-3 1.12 FREQUENCY NOTATION............................................... 1-3 a<

. 1.13 e.m._e e m_ r_v n_ea.m

_ . m .u m..e m. ee. .n._ c. .a.m.

.- u. e u m..cu.e M (Deleted)...................... 1 3 1.14 IDENTIFIED LEAKAGE............................................... 1-3 1.15 bi^"ID "n0"acT" Tn".'T"5"T CYOTE" (Deleted)....................... 1-4

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, - m 1.16 .u s._ m_ n.- _HANGGG . m. . . . . m,.m

.. . . e. o. n. .e .s... . u.e_ .n. .

SYOT""C (Deleted)................................................ 1-4 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)........................... 1-4 1.18 OPERABLE - OPERABILITY........................................... 1-4 1.19 OPERATIONAL MODE - MODE.......................................... 1-5 1.20 PHYSICS TESTS......................... ............... .......... 1-5 1.21 PRESSURE BOUNDARY LEAKAGE........ ... ........................... 1-5 1.21a PRESSURE TEMPERATURE LIMITS REPORT (PTLR)........................ 1-5 l 1.22 PROCESS CONTROL PROGRAM (PCP).................................... 1-5 1.23 PURGE-PURGING.................................................... 1-5 1.24 QUADRANT POWER TILT RATIO................... .................... 1-5 1.25 RATED THERMAL POWER.............................................. 1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME................................ 1-6 1.27 REPORTABLE EVENT................................................. 1-6 1.28 SHUTDOWN MARGIN.................................................. 1-6 1.29 SOL!DI."ICATIO" (Deleted)......................................... 1-6 1.30 SOURCE CHECK..................................................... 1-6

. 1.31 STAGGERED TEST BASIS............................................. 1-6 1.32 THERMAL POWER.................................................... 1-7 1.33 U N I D E N T I F I E D LE A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 7 1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................. 1-7 1.35 VENTING.......................................................... 1-7 TABLE 1.1 OPERATIONAL MODES........................................... 1-8 TABLE 1.2 FREQUENCY NOTATION.......................................... 1-9 FARLEY-UNIT 2 I AMENDMENT NO.

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1 INDEX ADMINISTRATIVE CONTROLS l

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SECTION PAGE Review.................................................. 6-10 Audits.................................................. 6-11 Authority............................................... 6-12 Records................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities.............................................. 6-12 Records.... ............................................. 6-13 6.6 REPORTABLE EVENT ACTION........................................ 6-14 6.7 SAFETY LIMIT VIOLATION......................................... 6-14 6.8 PROCEDURES AND PROGRAMS........................................ 6-14 6.9 REPORTING REOUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report.......................................... 6-15a Annual Report........................................... 6-16 Annual Radiological Environmental Operating Report...... 6-17 Annual Radioactive Effluent Release Report.............. 6-17 Monthly Operating Report................................. 6-19 Peaking Factor Limit Report............................. 6-19 Annual Diesel Generator Reliability Data Report......... 6-19 Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report..................... 6-20 Pressure Temperature Limits Report (PTLR)............... 6-20a 6.9.2 SPECIAL REPORTS.............................................. 6-20a 6.10 RECORD RETENTION.............................................. 6-20a 6.11 RADIATION PROTECTION PROGRAM.................................. 6-21a 6.12 HIGH RADIATION AREA........................................... 6-22 FARLEY-UNIT 2 XIX AMENDMENT NO.

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DEFINITIONS OPERATIONAL MODE - MODE -

1.19- An OPERATIONAL MODE (i.e.,. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.-

PHYSICS TESTS-PHYSICS TESTS shall be those tests performed to measure the fundamental 1.20 nuclear characteristics of the reactor core and related instrumentation and 1)

described in Chapter 14.0 of the_FSAR, 2) authorized under the provisions .

l of 10 CFR 50.59, or 3) otherwise approved by the Commission.

1 1 PRESSURE BOUNDARY LEAKAGE.

1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube

leakage) through a non-isolable fault in a Reactor Coolant System Component body, pipe wall or vessel wall.

PRESSURE TEMPERATURE LIMITS REPORT (PTLR) l 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific

, document that provides the reactor vessel pressure and temperature (P/T)

?

limits, including heatup and cooldown rates, for the current reactor vessel 9

fluence period. These P/T limits shall be determined for each fluence period l j or effective full-power years (EFPYs) in accordance with Specification ,

6.9.1.15. Plant operation within these operating limits is addressed in LCO I 3.4.10.1, RCS Pressure / Temperature Limits.

t l PROCESS CONTROL PROGRAM (PCP)

I

$ 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,

! sampling, analyses, tests, and determinations to be made to ensure that l

processing and packaging of solid radioactive wastes based on demonstrated

{ processing of actual or simulated wet solid wastes will be accomplished in 1

such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State ,

i regulations; burial ground requirements; and other requirements governing the

disposal of solid radioactive waste.

j PURGE - PURGING I

j 1.23 PURGE and PURGING is the controlled process of discharging air or gas i from a confinement to maintain temperature, pressure, humidity, concentration

or other operating condition, in such a manner that replacement air or gas is  ;

l required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore j detector calibrated output to the average of the upper excore detector cali-j brated outputs, or the ratio of the maximum lower excore detector calibrated j output to the average of the lower excore detector calibrated outputs, i whichever is greater. With one excore detector inoperable, the remaining I three detectors shall be used for computing the average.

FARLEY-UNIT 2 1-5 AMENDMENT NO.

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l RERCTOR COOLANT f.'YSTfj 3/4.4.10 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM i

4 i LIMITING CONDITION FOR OPERATION q-

) 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature

and pressure shall be limited in accordance with the limits specified in the '

PRESSURE TEMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, l criticality, and inservice leak and hydrostatic testing.

l APPLICABILITY: At all times.

ACTION:

With any of the above limits specified in the PTLR exceeded, restore the l l temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. and reduce the RCS T av and pressure to less than 200'F and 500 psig, respectively, within the fol$cwing30 hours.

SURVEILLANCE REQUIREMENTS , 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at least once per l hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

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FARLEY-UNIT 2 3/4 4-27 AMdNDMENT NO.

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REACTOR COOLANT SYSTEM BASES Reducing T avg to less than 50G*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G as required per 10 CFR Part 50 Appendix G. l

1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR). l a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown in the PTLR. Limit lines for cooldown l rates between those presented may be obtained by interpolation.

b) The PTLR defines limits to assure prevention of nonductile failure only. ' tor normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2) These limit lines shall be calculated periodically using methods provided in WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.

FARLEY-UNIT 2 B 3/4 4-6 AMBNDMENT NO.

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REACTOR CCOLANT SXSTEM 4

BASES

4) The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

l Heatup and cooldown limit curves are calculated using the most limiting

.value of the nil-ductility reference temperature, RTndt, at the end of 36 effective full power years (EFPY) of service life. The 36 EFPY service life l period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

FARLEY-UNIT 2 B 3/4 4-7 AMENDMENT NO.

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i REACTOR COOLANT SYSTEM BASES l Values of ARTndt determined in accordance with WCAP-14040-NP-A, Revision 2, may l

be used until the next results from the material surveillance program, evaluated ,

1 according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the PTLR. The heatup and l cooldown curves must be recalculated when the ARTndt determined from the next i surveillance capsule exceeds the calculated ARTndt for the equivalent capsule >

radiation exposure.

{

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of l the ASME Boiler and Pressure Vessel Code as required by Appendix G to i 10 CFR 50 and these methods are discussed in detail in WCAP-14040-NP-A, l Revision 2.

l Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310*F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start {

of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50 F above the RCS cold leg temperatures provided measures are taken to cushion the overpressure ef fects at RCS temperatures above 250"F, or (2) the start of all operable charging pumps and their injection into a water solid RCS. In the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operable charging pumps allowed by the Technical Specifications.

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FARLEY-UNIT 2 B 3/4 4-11 AMENDMENT NO.

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i REACTOR COOLANT SYSTEM '

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REACTOR COOLANT SYSTEM BASES 1

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3/4.4.11 STRUCTURAL INTEGRITY 7

The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR

• Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

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FARLEY-UNIT 2 B 3/4 4-14 AMDNDMENT NO.

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ADMINISTRATIVE CONTROLS ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be included (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radiciodine concentrations; (3)

Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro Ci/gm) and one ,

other radioiodine isotope concentration (micro Ci/gm) as a function of time for )

the duration of the specific activity above the steady-state level; and (5) The '

time duration when the specific activity of the primary coolant exceeded the radiciodine limit. ,

ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.  ;

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i FARLEY-UNIT 2 6-20 AM'NDMENT J NO.

l ADMINISTRATIVE CONTROLS 1

PRESSURE TEMPERATURE LIMITS REPORT fPTLR)

) 6.9.1.15 The reactor coolant system pressure and temperature limits, including i heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1. l l

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, Revision 2, " Methodology Used to Develop Cold  ;

Overpressure Mitigating System Setpoints and F.CS Heatup and Cooldown Limit l Curves," approved by NRC SER dated October 16, 1995. I i

ihe PTLR shall be provided to the NRC upon issuance for each reactor fluence .

period and for any revision or supplement thereto.

l SPECIAL REPORTS i l

6.9.2 Special reports shall be submitted to the Commission in accordance with i the requirements of 10CFR50.4 within the time period specified for each report. l Reports should be submitted to the U. S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555. i I

6.10 RECORD RETENTION l

1 In addition to the applicable record retention requirements of Title 10, Code of j Federal Regulations, the following records shall be retained for at least the minimum period indicated. I 6.10.1 The following records shall be retained for at least five years: I

a. Records and logs of unit operation covering time interval at each l power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. All REPORTABLE EVENTS submitted to the Commission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 2 6-20a AMFNDMENT a NO.

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ENCLOSURE 5 Joseph M. Farley Nuclear Plant Units I and 2 Pressure Temperature Limits Report Technical Specification Changes Marked-Un Paces h

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OEFINITIONS t,

a

! SECTION f%L

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1.0 DEFINITIONS 1.1 I 1.2 ACTION ......................................................... 1-1 AX I A L F LUX D I F F ERENC E . . . . . . . . . . . . . . . . . . . . . . . . . .1-1 .......

! 1.3 CHANNEL l 1.4 CALIBRATION.............................................

CHANNEL CHECK ........................... 1-1 1 1.5 1-1 i 1.6 CHANNEL FUNCTION TEST ................... ...................... 1-1 1.7 CONTAINNENT INTEGRITY ................... ...................... 1-2 j 1.8 CONTROLLED LEAKAGE ............................................. 1-2 1.9 CORE ALTERATION ................................................ 1-2 1 QOSE EQUIVALENT I-131 ................... ......................

........... 1-2 i 1.10 E-AVERAGE DIS!NTEGRATION ENERGY ..................... ..........

.......... 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME ... 1-3

1.12 FREQUENCY NOTATION ......................... ...................

................. 1-3 1.13 GASE0WS-RA9 WASTE-TREATNENT-SYSTEM (Del eted) . . . . . . . . . . . . . . .

1.14 IDENTIFIED LEAKAGE .............................................

1-3

1.15 LIGWIS-RA9 WASTE-TREATMENT-SYSTEM (Deleted) ..................... 1-4 .

! 1.16 NAJGR-GWANGES-TG-RASIGAGTIVE-WASTE-TREATMENT-SYSTEM 1.17 0FFSITE DOSE CALCULATION MANUAL 00CM ........ 1-4 i

1.18 OPERA 8LE - OPERA 81LITY .........(.....)........................... 1-4 1.19 OPERATIONAL MODE - MODE ....................... ................ 1-5 l

1.20 1' PHYSICS TESTS .................................................. 15 M*C 4

j M 2122PRESSURE BOUNDARY LEAKAGE ......................................

........ 15 PROCESS CONTROL PROGRAM l

1.23 PURG E - PURG I ...........................

NG . . . . . . . . ( PC P ) . . ........ . . . . . 1-5 1-5

, 1.24 QUADRANT POWER TILT RATIO ...................................... 15

! 1.25 RATED THERMAL POWER ............................................ ........ 1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME .............................. 1-6 1.27 REPORTABLE EVENT ............................................... 1-6 4

1.28 1.29 SHUTDOWN MARGIN ............................................... .... 1-6 SGkISIFIEATION (Deleted ................... 1-6 i ' 1.30 SOURCE CHECX ..........).........................................

........... 1-6 4

1.31 STAGGERED TEST BASIS ........................................... 1-6 1 1.32 THERMAL POWER .................................................. 1-7 i

1.33 UNIDENTIFIED LEAKAGE ........................................... ......... 1-7 1.34 VENTILATION EXHAUST TREATMENT SYSTEM ........................... 1-7

{, 1.35 VENTING f......................................................- 1-7 l

TABLE 1.1 OPERATIONAL MODES ......................................... 18 i

TABLE 1.2 FREQUENCY NOTATION ........................................ 1-9

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AME E MENT No. 57 J

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, Revies ..................................................... 6-10 l

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, Audits ..................................................... 6 11 l Antherity .................................................. 6 12 i

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Recerts .................................................... 6 12 i 4.5.3 TEOel! CAL REV15 AM CMTNN.

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Activities ................................................. 6-12 j

i Racerts .................................................... 4 13 s.a EP E TAE.1 EV W T ACTIM ...................................... s.14

a. 7 sum unti vietation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . s-14 a . 3 pmIIREES Am N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 14 a.s arpentina aEnutanent  :

i 6.9.1 Rolff!E MPORTS '

4 j Startup Report ............................................. 6-15e b

l Aeneal Report .............................................. 6 16

, Assoal Radielegical Eevireausstal Operaties Report ......g.. 4-17

! Annual Radienstive Effl eent Aaleese tapert . . . . . . . . . . . . . . . . . 6-17 l

nestsly esertting neport ................................... 6-19 4

Radial Peeking Facter Limit Aspert ......................... 6-19 i Aeneal Diesel Generater Anitability Data Aspert ............ 5-19 1

i Aeneal Beester Caelant System Specific Activity Aspert ..... 45 d

i amammt Sealed Seeren Laekage Report ........................

63 I Ide2T2 j 6.9.[SPEIN.MrMTS............................................ 6 a.

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! 8 18 m MTWrls ............................................ 6- a a.11 RE1Affm faM Kffe BODM E ................................ 6-Ils l 4-tt a.11 Mian agapT1ml aaEn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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OEFINITIONS

) ...................................................................,,,,,,,

"~

j OPERATIONAL MODE - MODE '

~

1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor 4

coolant temperature specified in Table 1.1.

PHYSICS TESTS i

i 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions

, of 10 CFR 50.59, or 3) otherwise approved by the Comission.

}

I PRESSURE BOUNDARY LEAKAGE l 1.21 PRESSURE BOUNDARY LEAXAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant Systee Component

! body, pipe wall or vessel wall.

l 11lk 6 3 =- .

gggf PROCESS CONTROL PROGRAM f PCP)

N I

1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that

! )- processing and packaging of solid radioactive wastes based on demonstrated

! processing of actual or simulated wet solid wastes will be accomplished in

! such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State (

l regulations; burial ground requirements; and other requirements governing the j disposal of solid radioactive waste. ,

l >

l PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas ,

from a confinement to maintain temperature, pressure, humidity, concentration  !

or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

, OUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximus upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated

, output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining threatdetectors shall be used for computing the average.

FARLEY-UNIT 1 1-5 AMEEMENT NO 26h h

4 l FARLEY NUCLEAR PLANT - UNIT 1 PTLR SUBMITTAL TECHNICAL SPECIFICAT!ONS MARKUPS -

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! PRESSURE TEMPERATURE LIMITS REPORT Q I

1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document

!- that provides the reactor vessel pressure and temperature (P/T) limits, including heatup and i cooldown rates, for the current reactor vessel fluence period. These P/T limits shall be determined for each fluence period or effective full-power years (EFPYs) in accordance with Specification 6.9.1.15. Plant operation within these operating limits is addressed in LCO 3.4.10.1, RCS Pressure / Temperature Limits.

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REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS

! REACTOR COOLANT SYSTEM LIMIT 5 Mc;Fiso W~TH6 f%%.lR4 i

TeMFb:EATc.jAMr LiMmi,26,AXX(PrLE) )

LIMITING CONDITION FOR OPERATION -

' _/

3.4.10.1 The Reactor Coolant System (except the ressurizer) temperature and i pressure shall be limited in accordance with the .1:i; lin : :h:: :: Fi;;:::

l 2.' 2 : d 3.6 3 during heatup, cooldown, criticality, and inservice leak and

hydrostatic testing,vi e-

iru h:: tup :f 1^^ F in :ny-one-hour ;::ird.

t i _ ___ ,_.._ _ _ _3 2 ... _s enn e  % _,.., ___ 3..; p_ g g_

. 2 may.irue temposu w e 9 enge of less th=a er e':"el to-1A*' 4a say aae h:ur p;ried during in ervice hydreetitie sad le @ t=?'4a- aa-retiaa?

1 at c; th; h;; tup ;;d :scider. li:it ::::::. j l

APPLICABILITY: At all times.

ACTION: JP6f/F'/L-D /A/7ME PT Vith any of the above limits exceeded, restore the temperature and/or prassure

, to within the limit within 30 minutes; perform an engineering evaluatior or ,

inspection to determine the effects of the out-of-limit condition on thw )

fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least 80T STANDBY yithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200 F and 500 psig, respectively, within the foll8ving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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SURVEILLANCE REQUIREMENTS ;ggggg]g;ggggn$

4.4.10.1.1 The Reactor Coolant Sy tem temperature and pressure shall be i determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

FARLEY-UNIT 1 3/4 4-27 AMENDMENT No. 25 p

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M PM Ik6dCAI4LLYL&f-Tf3ULAJK MATERI AL PROPERTY I

' CONTROLLING MATE RI AL LOWER SHELL (PLATE NO. 84819 2)

O.14 WT%

COPPER CONTENT .

j NICKEL CONTENT 0.54 WT%

INITIAL RT NOT 5' F l

T AFTER 18 EFPY  : 1/47,144,4' F

NOT

• 3/47,1213' F 4

1 8 E SERVICE CUR S APPLICASLE FOR HEATUP R ATES UP TO 80 F/HR FOR PERIO UP TO to EFPY i

T I I I I I iil iI e I t 's j J

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\ n J JI I ,

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- \ R f l == ' '

,'  ; , l: ,

j UNACC ABLE s I f ACCEPTABLE M

! OPERATI N r i  ;-

OPERATION @

s i= ,

\ I

/

f l I

I f , 3 J I j

** 'L, I I E y sooO  !. ' '

l l 3 HEATUP RATES 1 1 i $ TO 608 F/HR ', / /

l, ', '

1258  ;

E v

.r I

a w r i r

l g tone ,

/ '

I 5:1 e a

/ s s

l

. O r t

i 2 = '

s - CRITICALITY LIMIT l , ,,,, N BASED ON INSERVICE i'

' " ' HYDROSTATIC TEST

• / EMPERATURE (274* F)

,' F R THE SERVICE PERIOD i '

i- UP :O 16 EFPY t 2so / s t

? >

m j r I 1 1

/

r I I

• ! 0 484 0 0 100 150 200 250 300 M0 500 4

l j INDICATED TEMPERATURE (O F) i

(

I i 4 j FIGURE .4-2 FARLEY. UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS* APPLIC LE

{ FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/4 4-29 AttEN0 MENT NO. ES p 4 ,

l

1IM FW #reuT)odau.hFT 8L6JK u j M ATE RI AL PROPE RT'y B ASIS

• j 1

CONTROLLING MATERI AL e LOWER SHELL IPLATE NO. 58819 23 i COPPE R CONTENT 4 0.14 WT%

l NICKEL CONTENT 034 WT% l INITIAL RT NOT I '

l T

NOT AFTER 16 E FPY 1M. W.4* F 3/47,1213* F CU ES APPLICA8LE FOR COOLDOWN UP TO 100* F/HR FOR THE SER E PERIOD UP TO 16 CFPY 2s00 , , , ,

\~ 6 I e i j , y

', i i 1 f, i

\ I 6) 2250 \

k i

/ ,/

i

\ f < , l t r e i

' l 2#0 \ i i e i , I i T I 3 E J i  ! I

\! 3 Y ,r i i 3 ,3, UNACC ABLE- /r ACCEPTABLE' g 1 OPERATI N r>

'[ OPERATION

, i y i *\ a .

j a: 1500 , \ ,

O \ f r w

i

\ r

< r a

i g 1250 ( '

r, i o

w j v ,

, r, ,

j l I i / .F k I O LDOWN RATES, F

$ _ _CO' \

z O F/HR

~

30

, / \

i i e s

' s y

OQ ,

"'::: 20 500 ____,M- 77 -

<e

\

s

____ .60 G( \

"":~': 100 <

ano , \

/ \ l

\

> w F \

0 O 0 100 150 200 250 300 400 460 s00 INDICATED TEMPERATURE (O F l

l F URE 3.4-3 FARLEY UNIT 1 REACTOR COOLANT SYST'EM COOLDOWN L ITATIONS APPLICABLE FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/4 4-30 AMENDMENT NO.58 yj

• 1 1

~_- - - .

REACTOR COOLANT SYSTEM BASES 33 Reducin'g T av steam generakor tube rupture since the saturation pressure coolant is below the lift pressure of the atmospneric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels inaction.

take corrective the primary coolant will be detected in sufficient time to Information obtained on iodine spiking will oe used to assess the parameters associated with spiking phenomena. A reduction in frecuency justified byofthe isotopic data analyses obtained.following power changes may be permissible if 3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature be consistent withand th pressure changes during heatup and cooldown are limited to y/ >

Vessel Code, Sectio quirements given in the ASE Boiler and pressure Ap dix G as required per 10 CFR Part 50 Appendix G.

XI j

1) The reactor m lant te ure and pressure and system heatup and cooldown rates (with me exception of the pressurizer) accordance with e';um 2.' 2 =d 2.' 2. tW- W 'w

% fD55df4 Td/dWdf6 '

a) Allowable combinations of pressure and t;t_iM  ;.uew G5 h.FbEl-(Pn.

fr:ani temperature showr change rates are below and to the right of the limit lines bta t.imit lines for cooldown rates between those presented may be olation.

b)  !;r_ __. 5 5-1_.. 3.' 2 defin@imits to assure prevention of

- nonductile f ailure characteristics, e.g., pump only. For normal operation, other inherent plant heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2) These lief t H a* < thall he calcula+-d =

ica11v ueinn thnds ornvu.a

&w " TV HCAP NC40WN-AAVi$KNEl@Mt7htDixa%9fbin7Di2 A. Ma7smAb MrgMMmN54D ktSLkd7UP 4D M M ukWW.S.

31 -

ine secunBaVi ai- 4 i.ne iteam generatorWst not tie pressurued abDve 200 psig if the temperature of the steam generator is below 70*F.

FARLEY-UNIT 1 B 3/4 4-6 AMENDMER NO. 25

4 J

i ,

REACTOR COOLANT SYSTEM j BASES

, 3 1

); 4) The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than i 320*F.

S) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler

and Pressure Vessel Code,Section XI.

m f aren . +nopa.,, 7, e -. . 4 . , ef .we f: _iti: .::: , ,, ,_ tu, _ = :=

,... ._. .. .a.. .

. . . . . . , __,__2 4.

4 . s. .n .e v. u e,oc m.3

... .; 4

...... .... 2. . . . . .. . .. . . . . . . . ... 2 ... u. . ..s.

. . . . .. r u. . ... . . . , . . _ , _ .........u...

. . . .. . ......~.,..._.2 me.u ... _24 r ., am

...._2.... .w. , o u e..___ n au.a. .. e. 4.. , ,, , .u.

y m ,. ma,.__

- - - - , v r _ _- -__ u____,-

r_2.- __2 - .

d ,',g.,. i h. l.,.. . .. .d.,. ..dd.V

. .d,b. . . . , . . .'8 -.u 8

, , . , - - - - - -.i...... =.=

_. .w 2,

. = , .

2 ,,._4u 2 4.

,ne, Dr,.w.- -...

m,... , . -e _,_ ,__ __; ,_'_,,

m,,. ., ,, ..., . , , ,.n__.._ .,. .., .... r,_,,_.,.2_._" . . . .. ... . . . . . ,

r.._..__

.r.

. n. . e. . -

I 3G i Heatup and cooldown limit curves are calculated u

• g the most li ing value of the nil-ductility reference temperature, RTn , at the end of ef fective full power years (EFPY) of service life. The EFPY service life period is

! chosen such that the limiting RT t at the 1/4T location in the core region is i greater than the RTndt of the li ting unirradiated material. The selection of such a limiting RTndt assures that all components in ,the Reactor Coolant System will be operated. conservatively in accordance with applicable Code j requirements.

,4

,. u .

l... ....._......,_....a.,,u.....

u...

.... .....a.

.. u.. _4..

.. . w . 4. 4

4. . 4. , o r. g' g.g,. .

...._......u.....-

.u. o..,-....

___..,......c...., ....i.. .u .. , _ T u,.

. . , . . . . . . , _ . . . . , 2.

__ 1 3 .... . .. . . . . . . . . . . n,

_ ...___ re __ ..._ .w.. , . ueu

. . . . , .:_ . ,/ A24. .A ..

4..

4.......

......... . . . . .... .vu___,___ . . . . . . . . .. ,. .o.....; ... .. ... . , _. _. _. _. _, _ . _ ._._. _. .. .. .. .. ._ .. . . u..

1. .u.

or .. . ____. . _ - _.u_

2 _ .. ....

, nus- ~ - , ,. --_, - ~ _ . _ . .. - _ -. .u. - -

--r- ----- -r-- ---

,,.____ __2

...u,,.. . . . . . .

4 - . - - _

4. - .... 4.. ... u.

.. ,. r,.. . . . . .. . . . . .

._ o_..;

,............,e,_..__ __

o ,

.,o 2 .,, e,_..__ . , . .

e,,.

o en , ,

._2 .u. _______2..,___

.1 , . ,

n.._._..,..,__.. n ..n._..,_,_

3 r . 12. , , __._ ., n.,2..., e,____., .. n _ o . .;

, , . . . , , .. . . . n. ,, . . . . . . ., ....... . . . , . . . . . . . . . . . . . . . ... . . , . . . . . -

n_o.

__ m

.__m._.,._ u,,____, m.._ __2 ___,2_. ,,_,.

,,s,. ,__,.a.

.,m, u ._ u..._._.._

.. . . . . . . -,.....p .. . ..... .m .y .mo .onw 3

v,v.ww.u,_o  ;

g, my e u.,.. ..

.. Au_

, a

_,.e,..

eenu

..a . ..... .

..2.

___o..,_.._;

n..._.___

. . ,....~ .. . , . , . . . . . .

_u..,.,..

l

. . n a r. - - -"' ' " - v ' ' ' ' '

, , r'-

' l I

i '

4 4

s 4

4 4

1 4

i t

4

, FARLEY-UNIT 1 B 3/4 4-7 AMENDMENT NO. H -?

I

' 1 8

l d

i l

l REACTOR COOLANT SYSTEM BASES M O M/M -

O-A/8-b MI/OM2, '

l 1

1

.i: un :: may be used until the next l Values results of from ORT,$, determined t e material in surveillance program, evaluated according to l J

] ASTM E185-82, are available. Capsules vill be removed in accordance with i the requirements of ASTM E185-82 and 10 CFR 50, Appendix B. The surveillance specimen withdrawal schedule is shown in "S!". S:::ien 5.' '

l

! The heatup and cooldown curves must be recalculated when the ART i determined from the surveillance capsule exceeds the calculated Mt,,,

4 for the equ valent capsule radiation exposure.

y r/e PILE ,

i Allovable ssure-temperature relationships for various heatup ancf'~~

, cooldownyates are calculated using methods derived from Appendix G in Section in of the ASME Boiler and Pressure Vessel Code as required by i

] Appendix G to 1 'ar* "A =dqb. ods are discussed in detail in

Ot." ? ^ 21 ".

CAP. b-MP A I?6Vhic4J 2.

i TE: ; = =:1 = :hed fx = h ch:in; h = ::; =d : =ld:= 11:i: :=v= i:

i b--^'_upe".

:. pri-^!;lr

,:f_.*he

.. _ 2.....

r------ -- - - - - - ---

ti fr ttur :=

lin = r Ol r.___.,,....__,'rie:..._,___"""")

r- --- ------

def = zuh e:::s :f == ; =:= :f :s: =u :ua==, T, =e : l=c

f L'2? ;; ::::::d :: :ni:: :: th: in:id: :f th: 7::::1 =11 :: r:11 ::

i 1: :5: ::::id: :f the e::::1 rell. Th: dircrri re :f thie ; :: let:d i cr ^ refer =d te in A;;:ndi C Of AS"." S:::10: I!! :: th: refer:n::-

i f1:e, ;ly  :::d th: :=:::: ::;:Sili:le: f in:::vi:: in:p:::ie:

hniq:::. Th:::f:::, th: ::::::: :;::: 10 liti currer devel ped fer j e- -- f r- : = =h ::: nrc::v::ir: =d ;::vid: : ffici :: ::f::7 i =r;in: f= ; :::::1:n 1; ine: n:: d=:il: feile=. T =:ur: the th:

?  ::di::i:n :: brit:1::: : :ff:::: ::: ::::::::d f:: in th: ::1:21::10: Of J th; lizi: :::v::, :h: :::: li:itin; v:12: ef th: til ductility ::f: rent:

=; =: : : = , ".T,, , , i: =d :nd :hi i=lud:: th ::dicti= ind:::d 4

hif:. '"T ,,y , = = =;= din; :: :h: ::d :" :h: ;:: icd-!:: rhi:h h::: p i _e 4 ---le 2.v = = = = = :e .

t I

i

]

e

! Wore we nw sawlala aasetNs oe BG& sunod 3/4.4..to, l As Maomo,14 Malo ib 71M R6& TRM PM8W4 4id 4

roe coundurre 1

J i

, s FARLEY-UNIT 1 B3/4 4 8 AMENDMENT NO. 58 ,

4 4

1 h

5 i

l

p t

4 "1[415 MINK:dTIDM E* *% JA A a M_A

. n n b.

n jS

-4AREM-UNFT--l-REACT ^" YESSEL !^"C::::ESS ^^^^E"J:ES -

E Upper 1 Energy.

, Q Material Cu P Ni Tndt RTndt Component Code No. Type (%) (1) (1) (*F) (*F) [C] NMWD[d]

Closure head done

~

B690 A533.8,C1.1 0.16 0.009 0.50 -30 -20[a] 140 .

j I

Closure head segment -86902-1 A533,B.C1.1 0.17 0.007 0.52 -20 -20[8] 138 -

Closure head flange B6915-1 , C1.2 0.10 0.012 0.64 ] 60[a] 75[a] _

Vessel flange B6913-1 A5 , C1.2 0.17 0.011 0.69 60[a] 60[8] 106[aj _

Inlet nozzle B6917-1 A508, .2 - 0.010 0. 60[8] 60[a] -

110

- 0.008 .80 60[8] 60[a] 80 Inlet nozzle B6917-2 A508, C1. -

O. 0.87 60[a] 60[a3 _ 9g m inlet nozzle 86917-3 'A508, C1.2 -

Outlet nozzle B6916-1 A508, C1.2 - 7 0.77 60[a] 60[a] -

96.5 w 97.5 2 Outlet nozzle B6916-2 A508 C1.2 -

.011 0.78 60[8] 60[a] -

3 L

1

, Outlet nozzle B6916-3 A508 C1.2 -

.009 0.78 60[a] 60[a] -

100 B6914-1 A508 C1.2 0. 0 0.68 30 30[8] 148 -

& Nozzle shell A533,B,C1. 0.13 0.01 0.60 0 0 151.5 97  !

Inter. shell B6903-2 134.5 100 Inter shell 86903-3 A533,B, . 0.12 0.014 .56 10 10 A53 .C1.1 0.14 0.015 0. -20 15 133 90.5 Lower shell B6919-1 97 i Lower shell . 86919-2 3,B C1.1 0.14 0.015 0.56 -10 5 134 Bottom head ring B6912-1 A508, C1.2 - 0.010 0.72 0 10[a] 163.5 -

r Bottom head segment 86 A533,B,C1.1 0.15 0.011 0.52 -3 -30[a] 347 _

Bottom head done 7-1 A533 B.C1.1 0.17 0.014 0.60 -30 -30[a] 143.5 - .

$ inter. shell long. M1.33 Sub Arc Weld 'O.25 0.017 0.21 0[aj [aj _ _

l t

9 weld seam - 0[a]

E Inter. to lowe Gl.18 Sub Arc Weld 0.22 0.011 <0.20[b] 0[a] _ _

l 9 shell weld ans j

1. Lower s I long. Gl.08 Sub Arc Weld 0.17 0.022 <0.20[b] 0[a] 0[a3 _ _

i 5 weld seams i

. [a] Estimate per NUREG'0800 "USNRC Standard Review Plan" Dranch Technical Position MIEll 5-2.

i 4

[b] Estimated (low nickel weld wire used in fabricating vessel weld seams).

T [c] Major working direction. -

[d] Normal to major working direction.

I

1020 hidy,L 9

8 T'

7 /

6 /

5 \ /SURFACE x

4 \ #

\ /

m# f 3 \ > /

y j e pr 1/4T -

2 \ / / /

\/

f\ '

, / /

f f 1019 9 I # \ /

8 # # \ /

/ af \ /

- 6 / / \ / s ' 47' "Eu 5

/

f

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$ 4 I / \ f w I / *//

z 3 I I # \/

E // / A 32 o

//

I f

/

f

/ \

\

5 I / / \

3 [ [

10 18 I

/ , ,

9 / \

8 # / \

7 I / \

6 I ' \

5 I

/ / \

/ \

4 f / \

/ \

3  ! \

,/ \

2 / \

/ \

\

10 17 0 5 10 15 20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

T: CUR ; :/ 4. 4>. ; TAsi :; UTRON TLUCNC: (: 1 N'!) ".: ^ "'Jft:TICf! 0F ~"LL 'D"E" 5:R'!!:: L:I: (:I"')

FARLEY - UNIT 1 B 3/4 4-10 AMENDMENT N0.

1020 M & glOd % V LE T h _ Q -

9 8

'\ /

7

' /

I 6 / -

5 \

/

4 \ /

\ /

\ /

\' /

/

2

SURFACE i N / p -

/

10 19 >

3 9 \ ' / - 1/4T

) W 8 \ # # #

E 7

\ # / _/~

4 i g X L7

; / \ y/ /T/

< u 5 s b 4 / s R /

, D / / \ /

d / / V 3

5 / / /\  !

[ [ [ \ '

p 3/4T I

/ \_- '

i e / / i 1018 / m / s j

9 I I / # \

• I I // \

I f 7 \

Ii A \

3 II "I

// \

/ / \

4 i / / 3 I // \

l / / \

/ /  %

2

// \

// 'N 17 10 0 5 10 15 20 25 30 25 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

"!CUC 2 3/4.4 : IAST N UTCON I' UENCE (D ' "^'!) f 7 '5" f,0

• IUNCTION 07 TULL 70W:R SERVICC (CTTY)

FARLEY - UNIT 1 B 3/4 4-10 A AMENDMENT N

< REACTOR COOLANT SYSTEM p -

BASES V >

j The ASME approach for calculating the allowable limit curves for var us

he up and cooldown rates specifies that the total stress intensity fa or, K,g r the combined thermal and pressure stresses at any time durin heatup or cool wn cannot be greater than the reference stress intensity ctor, K IR' j for the m al temperature at that time. K IR is obtained from t reference j fracture to hness curve, defined in Appendix G to the ASME C e. The K IR

curve is given y the equation

Kgg = 26.78 + .223 exp [0.0145(T-RTNOT + 160)] (1) where K gg is the refere e stress intensity facto as a function of the metal j temperature T and the met nil ductility refer ce temperature RT HDT.

Thus, the governing equation for t a heatup-cooldo analysis is defined in Appendix G of the ASME Code as follows:

l l

CKgg + kit IEIR (2)

Where, K IM is the stress'intensi fact caused by membrane (pressure) stress.

K It is the stress tensity factor caus by the thermal gradients.

K gg is provid by the code as a function o temperature relative to the RT N of the material.

C = 2. for level A and B service limits, and C= .5 for inservice hydrostatic and leak test operat ns.

A any time during the heautp or cooldown transient, K IR is termined by the tal temperature at the tip of the postulat'ed flaw, the approp ate v ue for RTNDT, and the reference fracture toughness curve. The the al '

tresses resulting from temperature grad'ients through the vessel wall are FARLEY-UNIT 1 B 3/4 4-11 AMENDMENT NO a

I I

~ REACTOR COOLANT SYSTEM _

BASES

• a m Wr.Ji d adtars q J

~- ^

i 4

11 l 1culated and then the corresponding thermal stress intensity factor, K ,

l fo the reference flaw is computed. From Equation (2) the pressure str ss l inten ity factors are obtained and from these, the allowable pressur are calcula d.

j l COOLDOWN i For the cal lation of the allowable pressure versus colant temperature  ;

during cooldown, t Code reference flaw is assumed to st at the inside of j the vessel wall. Du ng cooldown, the controlling loc ion of the flaw is 4 always at the inside o the wall because the themal radients produce tensile i stresses at the inside, hich increase with increas g cooldown rates. Allowable l pressure-temperature rela ons are generated for th steady-state and finite j cooldown rate situations. on these relations omposite limit curves are j constructed for each cooldown ate of interes i

1

}

The use of the coa.posite cu in th ooldown analysis is necessary l because control of the cooldown proc u is based on measurement of reactor i coolant temperature, whereas the lim pressure is actually dependent on i

j the material temperature at the ti of th assumed flaw. During cooldown, the l 1/4T vessel location is at a hi er temperat e than the fluid adjacent to the

{ vessel ID. This condition, o course, is not ue for the steady-state situation.

! It follows that at any giv reactor coolant temp ature, the delta T developed during cooldown results a' higher value of K gg at e 1/4T location for j finite cooldown rates an for steady-state operation. Furthermore, if conditions exist su that the increase in K gg exceeds K It the calculated allowable pressu during cooldown will be greater than the teady-state 1

value. '

i The ove procedures are needed because there is no direct cont 1 on j tempera re at the 1/4T location; therefore, allowable pressures may i unkno ngly be violated if the rate of cooling is decreased at various I int als along a cooldown ramp. The use of the composite curve eliminat

th problem and assures conservative operation of the system for the enti l oldowr oeriod. .

l f

FARLEY-UNIT 1 B 3/4 4-12 AMENDMENTNO.h j-

1 l .

{ REACTOR COOLANT SYSTEM BASES p IdIh.Iil dj N 1 M b l

j HEATVP hree separate calculations are required to determine the limit curve for fi e heatup rates. As is done in the cooldown analysis, allowable pressure-tperature relationships are developed for steady-state con tions as well as f nite heatup rate conditions assuming the presence of a /4T defect at the nside of the vessel wall. The thermal gradients ring heatup

produce compress e stresses at the inside of the wall that al viate the j tensile stresses p duced by internal pressure. The metal rature at the l- crack tip legs the c ant temperature; therefore, the K for the 1/4T crack  :

! during heatup is lower n the E gg for the 1/4T crack ring stea@-state j l conditions at the same coo t temperature. During stup, especially at the j end of the transient, condit ns may exist such 't the effects of compressive 3

thermal stresses and different ' s for staa tate and finite heatup rat'es j do,40t offset each other and the essure-t oraturicurvebasedonsteady-l state conditions no longer represent a1 r bound of all similar curves for finite heatup rates when the 1/4T flew considered. Therefore, both cases l

! have to be analyzed in order to assu t at any coolant temperature the l

lower value of the allowable press calcu ted for stea@-state and finita j heatup rates is obtained.

! The second portion of heatup analysis co ens the calculation of

pressure-temperature limi tions for the case in whi a 1/4T deep outside j surface flaw is assumed Unlike the situation at the ssel inside surface,

, the thermal gradients stablished at the outside surfac during heatup produce -

I stresses which are sile in nature and thus tend to rei orce.any pressure

! stresses present. thermal stresses, of course, are ndent on both

! the rate of heat and the time (or coolant temperature) al the heatup l

ramp. Furthe  ; since the thermal stresses, at the outside re tensile and

! increase with ncreasing heatup rate, a lower bound curve canno be defined.

Rather, eac atup rate of interest must be analyzed on an indiv 1 basis.

f. Fo1 ing the generation of pressure-temperature curves for be the i steady- to and finite heatup rate situations, the final limit curves re i prod d as follows. A composite curve is constructed based on a point j poi comparison of the stes @-state and finite heatup rate data. At any

! gi en temperature, the allowable pressure is taken to be the lesser of the .

i ree values taken from the curves under consideration.

i i

FARLEY-UNIT 1 3 3/4 4-13 ,

AMDI0 MENT ND i

l 4

O.

____._____r

) REACTOR COOLANT SYSTEM d

BASES 4

Th: un Of the ^= ::!te r"^ is ="cessary t :=t n nerv:tiv: hutep

! 'tettattaa kat?vse "-it i: pessible 'e- 'a .dttieas te ert:t :=5 th:t :;;r th; n urn e' the heet"a the c^a+"011(a0 caa m iaa switche )

nd th pre ;ter: 'i-it meet it !!' ti n be b r ed n the inid:

f :

! '-a=

l th: ::t:td:

th: ::t criti n! Orit;ri n.

nly:i:  !

l 1

J 4

j Fin:lly, th: 10 tF" Prt 50, ^;; dfr G 8u'- 'hich add-ane the -t:1 t g : nter: i l

itd "- r=' in;e -"st 5 centd:nd.

j Tht: "1 :t:t:0 th:t th:

ir.i-" =:t:-' the !!! ur: 'etd '!!- I l *egie be St '*ist 120er 63;k.y the. the 1 t 14r;:nter:

4t4.; ef or th:sg.:10: r: f1:r.;; -

1 1

, ggege 7.; gen =g;7, 1 l th: ;rn n r: :nnd:

(521 ;:t" 'er Farley Unit-!). 20  ;:rn et c' the pr- --" ice hydr _e:ttti: t::t pr::: r

!." edditica, the --" 10 cro 937g gg no) g gn lt:itin; =t  :;nif t: fracter:  :":!uttte ty 5 ;erfend t: jntify in:

1 t-h:t : ;' 1 i n;;t x :tc. )

Frity Unit 2.

": : rn elt, se:5 fetet"re rt!y i: en ;rf:=d f:r The t Frlty Unit 2 #-'attre !nlytte r" "!t: r: ;;1 tnt'.: t:  !

i N* F:rl:y Unit 1 ei n: the ; rtirr t ?!r :t- tre id"-tica! 'ar beth ;! sts. .

"ind r;n tht: 'rntr : n:!y:i: the !! EF8Y h et"; rd neld = cun n r:

i l

k% t;nt:d5 Q : 10 C'" Prt 50 ". !: n- "h z :: "f; n: 2.i ! qd 2.t-3.

y _

=

Although the pressurizer operates in temperature ranges above those for,whi there is reason for concern of non-ductile failure, operating limits are i provided to assure compatibility of operation with the fatigue analysis l

l performed in accordance with the ASME Code requirements. )

t The OPERA 81LITY of either RHR relief valve or an RCS vent opening of greater 1 l

- than or equal to 2.85 square inches ensures that the RCS will be protected fros l pressure transients which could exceed the limits of Appendix G to 10 CFR Part j 50 when one or more of the RCS cold legs are less than or equal to 310*F.

' l 1

i Either RHR relief valve has adequate relieving capability to protect the RCS l

i from overpressurization when the transient is limited to either (1) the start f(

i of an idle RCP with the secondary water temperature of the steam generator less I

! i than or equal to 50*F above the RCS cold leg temperatures provided measures are (

taken to cushion the overpressure effects at RCS temperatures above 250*F, or I

( (2)y the start of-4 -hr;;1- -n=and their injection into a water solid RCS.

l q -

JM_ _ m I- - )

, 14.ll' STRUCTURAL IiuiEiRITY ' - - ^ ~ _

I IN The inservice inspection and testing programs for ASHE Code Class 1, 2 and 3 j

components ensure that the structural integrity and operational readiness of these of the components planer These will be maintained at an acceptable level throughout the life programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except uhere specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1),

im6cr4 Idht.A56OF THe:.lATT6cTeokJ fM THE! CHd& Ah RJMPs 114 MM_ dis is BAf6D cuTwesre.cr & TM NWIMdM M6eE (F 044A6(6 CNAA& RJMPS .

Au. OWED EN TW6 Teo4JicAL3Ft:GFICATiod, FARLEY-UNIT 1 8 3/4 4-14 AMDEMENT N0. 47,7@

ye a mm Wu.dteo iu TM- SJBBM> JMoolFl6D, APA McVe:D fo th 8 s/4 4-8 Fce CceJTWOW OF h 3/M 1O. 1 k l

]

._. _. ~- ' ~

i 1

AOMINIST3AT:7E CONTROLS eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseesse .....................

,,,,,,,,,,,,,,,,,, l AWUM. REAC"0R COOLANT $? STEM $PEC*?*O AC":7 '? REPORT 6.b l.13 This annual report acityt ty analyses of :ne pr mary toolant nave ex:eededts oniv required when the rI 3pec:!! cat:en ).a.71ur:ng :ne year.

ncluded

ne liitts of '

(1) The folloving :nformation shall se Reactor power history starting =4 hours prior to the first s a.ic . e in .ntch the last the limit vas exceeded (!n grannte and taoular format): (2) Results  !

liitt. results of analysis while limitisotopic analysis for raciotodtne performed p of '

- after the radiciodine activity was reducedwas exceeded and results of one analysis .

to less than the limit. Each resul; {

should include date and time of sampling and the radiciodine concentrations: (1) :ne Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which limit was exceeded i

other radiciodine isotope concentration (micro C1/gs)(4) Graph and one of the I-l the duration of the specific activity above the steady-state levels and (5)asThea function  ;

time durationlimit.

radiciodine when the specific activi:y of the primary coolant excenced :ne j 4

l ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 i annual basis if sealed source or fission detector leakage tests re presence of greater than or equal to 0.005 microcuries of removable contamination.

i bk$ I

$ FOLumahAA4 SPECIAL REPORTS

fl<ds 6.9.2 Special reports shall be submitted to the Commission in accordance with Reports should be submitted to the U. 5. Nuclear Regulatory C Document Control Desk. Vashington. D.C. 20535.

1 4

6.10 RECORD RETENTION t

In addition to the applicable record retention requirements of Title 10. Code of I Federal Regulations, minimum the following records shall be retained for at least the period indicated.  !

6.10.1 I

a. The following records shall be retained for at least five years: l

{ Records level. and logs of unit operation covering time interval at each power '

b.

Records and logs of principal maintenance activities. inspections, repair and nuclear replacement of principal items of equipment related to safety.*

c.

d. ALL REPORTABLE' EVENTS submitted to the Commission.

Records of surveillance activitiss, inspections and calibrations e.

required by these Technical Specifications.

Records 6.8.1. of changes made to the procedures required by Specification f.

Records of cadioactive shipments. 1

g. 1 Records of sealed source and fission detector, leak tests and results. l FARLEY-UNIT 1 6-20 AM NO.

11 m w :r d s.o.u o, c9.2, w o 0.Io w 4 o To % A i

1 i

h

_ . . _ . _ _ _ _ _ _ _ _ . _ _ . . _ . _ _ _ _ _ . _ - ___.____m .- _.___ .. _ _

t

, FARLEY NUCLEAR PLANT - UNIT 1 PTLR SUBMITTAL ,

TECHNICAL SPECIFICATIONS MARKUPS i ,

INSERT 5

{

a >

i t

1 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)  ;

I

! 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup and I cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

]

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in iL WCAP-14040-NP-A, Revision 2, " Methodology Used to Develop Cold Overpressure Mitigating

! System Setpoints and RCS Heatup and Cooldown Limit Curves," approved by NRC SER dated October 16,1995.

i .

1 The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any

revision or supplement thereto.

i h

4 i

(

l

i E

DEFINITIONS t

i SECTION a

f3E

[ 1.0 DEFINITIONS i

i.i

1.2 ACTION ......................................................... 11

1.3 AXIAL CHANNEL FLUX OIFFERENCE ..........................................

CALIBRATION............... 1-1 j 1.4 1-1 1.5 CHANNEL CHECK .................... ... ............................. 1-1 1.6 CHANNEL FUNCTION TEST ..........................................

.......................... 1-1 i 1.7 CONTAINMENT INTEGRITY ............. 1-2 1.8 CONTROLLED LEAKAGE ............................................. 1-2 l

1.9 CORE ALTERATION ................................................

1-2 QOSE EQUIVALENT I-131 ..........................................

i 1.10 E-AVERAGE DISINTEGRATION ENERGY 1.11 ................................

....... 1-21-3 i.12 rREauENCY NOTATION ..................IME ............... .......

ENGINEERED SAFETY FEATURES RESPONSE T

....... 1-3 ep

..................... 1-3 i 1.13 GASE0WS-RA9 WASTE-TREATMENT-SYSTEM (Deleted) ....................

...... 13 l

1.14 1.15 IDENTIFIED LEAKAGE ..................................... 13 A 1 1.16 LIQWIS-RA9 WASTE-TREATMENT-SYSTEM (Deleted) ..................... ........ 1-4 j

1.17 MAJOR-GNANGES-TG-RA910 ACTIVE-WASTE-TREATMENT-SVSTEMS Deleted) . 1-4 i 1.18 0FFSITE DOSE CALCULATION MANUAL OPERA 8LE - OPERA 8ILITY .........(00CM) .........

..............(. 1-4 1 14 l

1.19 OPERATIONAL MODE - M00E ........................................ 1-5 i 1.20 PHYSICS TESTS ............................ ..................... 15 l i

i IM7'I d.21 PRESSURE 22 PROCESS BOUNDARY CONTROL PROGRAMLEAKAGE ......................................

(PCP)

................... .............. 1-5 1-5 l 1.23 PURGE - PURGING ................................ 1-5 l 1.24 QUADRANT POWER TILT RATIO ...................................... 1-5 <

7-l 1.25 RATED THERMAL POWER ............................................ ................ 16

1.26 REACTOR TRIP SYSTEM RESPONSE TIME ...... 16 l 1.27 REPORTABLE EVENT ....................... ....................... 16 1.28 SHUTDOWN MARGIN ................................................

4 1.29 ...................... 1-6 l 59EI91FIGATION (Deleted

. ). . . . . . . . . . . . . . .1-6 i 1.30

SOU RC E CHECK . . . . . . . . .......................... 16......

1 1.31 STAGGERED TEST BASIS ................ 16

1.32 THERMAL POWER ..................................................

........ 1-7 i 1.33 1.34 UNIDENTIFIED LEAKAGE .............................................................. 1-7 VENTILATION EXHAUST TREATMENT SYST ........................... 1-7 i

4

\

1.35 VENTING ..........................EM ........................ ... . 1-7 p

TA8LE 1.1 OPERATIONAL MODES ................................... ..... 1-8 4

TA8LE 1.2 FREQUENCY NOTATION .......................................- 1-9 (USeKri j.21a hPsCf67EMM4Tu& UM M%N M a

J i

i j

)

J FARLEY-UNIT 2 I N No. O h

]

1 4

_m__ _ . _ _ _ .

.- __. . - . _ _ - - . . =._.-- . _ . - . ~ . . - - - - .-.. - - .-. - - - - - .

l l ADMINISTRATIVE CDIITROLE -

l 8ECTIo" . . _ _ . .

ZAGE Review................................................... 6-10 Audits................................................... 6-11 t

- Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 TE5NIch!. REVIEW AND CDIITROL t

)

t Activities............................................... 6-12 i

j Records.................................................. 6-13 4

j 6.6 REPQRIABLE EVENT ACTI05..................................... 6-14 l 6.7 SAFETT LIMIT YTOLATION ..................................... 6-14 f

. 6.8 PROCEDURES AND PROGRAMS...........8......................... 6-14

6.9 REPORTING REOtTTREMENTS ,

6.9.1 ROUTIME REPORTS Startup Report .......................................... 6-15a Annual Report............................................ 6-16 Annual Radiological Envirasamental Operating Report....... 6-17 Annual Radioactive Effluent.amtmaam Report............... 6-17 Monthly operating Report................................. 6-19 Peaking Factor Limit Report.............................. 6-19 l Annual Diesel Generatar Reliability Data Report.......... 6-19 Annual Reactor coolant system specific Activity Report... 6-20 Amamal sealed source Imakage Report...................... 6-20

/ 2 7

6. 9 . 2 SPECIAL EE701t23. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 .10 REcoltD RETENTIM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2Ddil '

6.11 RADIATIM FROTECTIM FBomtAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-214 6 .12 IIGE R&DIATIM REE1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 i rAELEr-nuTT 2 zzz ammimosam? no. 444-MNe&T2 Fffsss/B& TsMR4472/26 UM/71 Eener. 6-20" h

a I

i DEFINITIONS

{' OPERATTONAL MODE . MODE 4

1.19 i

An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor j coolant temperature specified in Table 1.1.

1 i PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental i nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

i t

PRESSURE B0UNDARY LEAXAGE l 1.21 PRESSURE 8OUNDARY LEAXAGE shall be leakage (except steam generator tube  !

! leakage) through a non-isolable fault in a Reactor Coolant System Component body, pipe wall or vessel wall.

i llNd73

• PROCESS CONTROL PROGRAM (PCP)

%sFou.cul%  ;

fA64 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or slaulated wet solid wastes will be accomplished in 1

} such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State ,

i regulations; burial ground requirements; and other requirements governing the i disposal of solid radioactive waste.

j i (

j PURGE - PURGM g '

i 1

i 1.23 PURSE or PURGING is the controlled process of discharging air or gas '

i from a confinement to maintain temperature, pressure, humidity, concentration i or other operating condition, in such a sanner that replacement air or gas is j required to purify the confinement.

i j QUADRANT POWER TILT RATIO

! 1.24 QUADRANT POWER TILT RATIO shall be'the ratio of the maximum upper excore

detector calibrated output to the average of the upper excore detector cali-

! brated outputs, or the ratio of the maximum lower excore detector calibrated j output to the average of the lower escore detector calibrated outputs, whichever

is greater. With one encore detector inoperable, the remaining three detectors j shall be used for computing the average.

3

(~ FARLEY-UNIT 2 1-5 M4T d'

2 l

4

_ . _ . ~ __ _ _ _. . . . _ . . . . . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _

FARLEY NUCLEAR PLANT - UNIT 2 PTLR SUBMITTAL l TECHNICAL SPECIFICATIONS MARKUPS l INSERT 3 PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P/T) limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These P/T limits shall be determined i for each fluence period or effective full-power years (EFPYs) in accordance with Specification 6.9.1.15. Plant operation within these operating limits is addressed in LCO 3.4.10.I, RCS Pressure / Temperature Limits.

i i

l e

i 1

REACTOR C00LAfff SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS -- _

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION UMITs 5R:o A60 DJ T "IeMMITt.lfbUMfB %R:27(PrLT 3.4.10.1 The Reactor Coolant System (except the ressurizer) temperature and pressure shall be limited in accordance with the li :: lin:: :h:r- :n "igu:::

3.'.4 ...d 2.'.-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing. h

^ I

- en' ur h:::up :f 100 F i :ny : : h::: ; :i:d.

i., . o a.xi;. c;;1d;u f 100* " i _ny ::: h::: ;::i:d.

--ximu- : -bes ture ch s ;c -' l--- 4-- --

q" ' *^ 1"** " - y ^ e h;ur p:: icd during in:::ci:: hyd::::::i: :nd 1::h ::::ing :p::::i: :

-ab::: th: 5:::up :nd :::1deu- li-i! tur"e: i 1

APPLICABILITY: At all times.

ACTION:

.Fff///M /A/ $M _ >

Vith any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY yithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200 F and 500 psig, respectively, within the foll8ving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS

..................................... ... ..... ..... .................. 4

.5fk.< tF/t O I d 77 6 P il. 2 4.4.10.1.1 The Reactor Coolant Sys emlemperature a p ure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to ostermine changes in material properties, as required by 10CFR50, Appendix H.

P FARLEY-UNIT 2 3/4 4-27 AMENDMENT NO

1 1I4M Pt/d Idredriodu LY ts:7BuMt 250

/

\

\

LEAK TEST LIMIT l l 1

/ /

I

/

~ / /

2250 : RTg AFTER 14 EFPY [ [ f /

1 i ,

e

\ 1/4T: 152*F ' I /

l i I /

2000 ---. T: 124'F J l r s

, i .<

1 1 ( /

, h i I /

N I r >

^ 1750 i g. 6I i d

/

9 m

UNA EPTABLE f

/

/

1 l

r , -

1 OPER ON e i / I w 1500 x / / / 1 2 s  ; s /

m r

3 r i

/

i r

/

w '

x 1250

U '

r' u

& R. J '

a -HEAT.UP RATES. UP TO 60? i 7 j y i l- \ i /

Q 1000 u \/ A /

/ .

5 / \ r

^ ACCEPTABLE OPERATION :"'

2 >

750 ' ---

r

/ \ s

/ '

., # / \

l' 500

[

/ '

T

, i . .

I I l j , )g CRITICAUTY UMIT BASED ON' ' '-

/ "

250 ' ERVICE HYDROSTATICTEST ::  !

f PERATURE (279'F) FOR THE --

/

s' E PERIOD UP TO 14 EFPY '[  !

f f f I I t I I I I I I t t I t t '

0 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '

0 50 00 150 200 250 300 400 450 500

, INDICATEDTEMPERATURE(DEG F) l-i igure 3.4-2 Farley Unit 2 Rea; tor Coolant System Heatup Limit &t ns l Applicable for the First 14 EFPY. I i

l rarley-Unit 2 3/4 4-29 Amendment No. 55@

.b

l NIS MINEdTIGblLf L6FT St,t.dg 3 l 2500\,. 1 ,

= -

.s ,

I l

xi '/

/' l

\ ,

i

/ 1 I

/ )

2250 '

m AFTER 14 EFPY / ,

i I , \

1 T: 152*F '

/

i i 2000 3/4 .124'F I l

___. 1 ,f r i

\ J j i / ,

~ 1750 'N i

/

.r f

!h, UNAC EPTABLE OPERA N r

/ .-

/

w g

1500 ( / /

h '

y ) ,

m x / / l (f) k / / i w N x 1250 J '

x i i 1 Q. N / /

g L / ./

y V # /

A  % j o 2n ,

i5 m v ACCEPTABLE OPERATION :-

E ONN RAE 7/ ,0  ; ; ;;

< 750 iiii e /

t , . _

___ F/HR. "

\ i i

e N i 0 -s w_._ / / n- i 500 20 - w / / 3 i  !

4o so e / x 60 / ,

f \x l 100 e ,

250 j N , l

< s i

/ x

/ \ e 0 * ' '

O 50 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG. F)

F gure 3.4-3 Farley Unit 2 Reactor Cooling System Cooldown Limitat ns Applicable for the First 14 EFPY, Farley-Unit 2 3/4 4-30 AmndmentNo.55,6

. 4

J REACTOR COOLANT SYSTEM J

BASES

• generator tu Tavb rupture since the saturation pressure of the pMaary( coo Reducing i

below the 1tft pressure of the atmospheric steam relief valves. The i

surveillance' requirements provide. adequate assurance that excessive specific

?

activity levels in the primary coolant will be detected in sufficient time to i

take corrective action. Information obtained on todine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in i frequency of isotopic analyses following' power changes may be permissible if

{

, justified by the data obtained. )

i 3/4.4.10 PRES $URE/ TEMPERATURE LIMIT 5 l The temperature and pressure changes chring heatup and cooldoen are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel j Code. Section Appendix G as required per 10CFR Part 50 Appendix G.

l 1) The reacto coolant tesperature and pressure and ' system heatup and cooldown

! rates (with the exception of the pressuMzer) shall be limited in

! accordance w_ith_f yr;; 3.' * -: 3.1 ? f;- O_ fi. 7. f.;11 ;_n;; ei,idi.,

j P;d@ f%Wkif.e:MF%CATUR6 L[MI'PSgPDET' (PTLR.g l

' a) Allowable codinations of pressure and toeperature for specific ^

{ tasperature change rates are below and to the Mght of the limit. lines show' Limit lines for cooldown rates between those presented may be i

p obt ned hv intergplation.

l_AM PT i b) 1;: ;;; ' :: @_ ..i ! defi inits to assure prevention of -

! nonductile failure only. For normal operation, other inherent plant f characteHstics, e.g., pusy heat addition and pressurizer heater

! capacity, any limit the heatup and cooldown rates that can be achieved

over certain pressure-temperature ranges.

2) These limit 11a== -hall ha enleula+=d meriodically usine mee w = nre "M j lAf HCAP- 1404b.AIP L MS/S EdMeXn&v7YN<.htO Tb' -

P

3) (OVbdAltS.ft}Lh M016471Alb .WSWM@Allt AAQkt3 H6472Jo AkDfmy l psig if the temperature of the steam generator is. belas se p..svurvaes 70T. soove %q);

h 1

c l

i

?

l 4

l

{ FAALEY UNIT 2 8 3/4 4 4

> NENIMENT Nh 4

l 4 i

\.

i I

)

l

., t REACTOR C00t.Altr SYSTEN i BA.SI.S . .. . ..... ... - . .

4)  !

i The pressuriser heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperatute

difference between the pressurizer and the sprey fluid is greater ti'an 320'F. (

I

! 5) i System preservice hydrotests and in-service leak and hydrotests shall be

{ performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code. Sectica II.

j M f- e te e te _'n e ;re--* *-- -! ^c f rr! *!e -tert 4-he re:::= r:r :1 - d

,,4 a .22.4...i e t e --!-M is r : r'_r x - i-t ^ ""._"!?!  !? , -M i:

.,.m.a _...

.....___i.-__--- ,<___

ps;;,e.^i E x add--f ::

A ::; O_z -:..ir-M '$ h;;;d5E;A 315. ^;;E_fiiid3 :te i^,

. 5:::i:: !!! :

,.a.._ . . . . . . ....

, _ ____.__- .<_>_' ___.._t__2 d * *"" 5:iin r,f.7- x_, x; "_,:_ x ,:__

4

_..m.. _ 2____, ,_ o,.. 1

{ "xt:- --f Tr:ld:n ' !-i t Cr_- r ^;rd ' __' * "_ " _ . _ ._ . _, _ _ _ _ _

Bea u r--  %

va e d cooldova limit curves are calculated using the host miting  !

of the nil-ductility reference temperature. RT , a the end of i

I effectivefullpoweryears(EFFT)ofservicelife,.,,T'_j'"'FTservicelifel _.

1 i

periodischosen.suchthatthelimiting17,l,thelimitingunit the core region is greater than the RT at the 1/4T location in i o radiated u terial. Theselectionofsucha11alllagRT assures that all

components in the Reactor Coolant Systen vill Se, operated conservatively i in accordance with applicable Code requirements.

e S. ,: :::nn rx--I r,:::'

- _ , , , , . .__ ___.s._ _ .<__

.t_

'_r.__-..

'_- :'__x_<__ t- td,_t.e d:t---' c t'e,ir init'_11

-- f :--_ tr: - _ a__ -x t:x " .- xt : _____._.u

_ .t,_

opesoti-- ..

_____..____,____'___,_'r

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, __ __, ___ _ _ . _ . , _.__a__

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M "r- _, ,

,1_-:.._. _ " ,

" r n! r'r_'_r." S, _ c 'r !"-

- '_ _ r -' '_ _- ' 4 " ! ' , -". - , r - "

. n.t::_ _ _ _ ...,__,_,___a...._a

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i yAni,gT-UNIT 2 3 3/4 4 7 i AMEWNENT NO. 55 h l

i i

1 4

4 i

l 4

I l

i

,' mC ce-C ::.m sysits ,

%~- _

sAses

' N^#M'h W!W HbdP./do/c _AfP Ag.gicy/2, 1 Values of :R7ngt determined in >': rr:r may results from t. e material surveillance program,be used until evaluated the nextto accorcing

> ASTM E185-32, are available.

l

- t*e recuirements of ASTM E185-82 ,t.and Capsules 10 CFR 50, willAppencia ce removed The in accorcance l witn

, ine bettup and coolcown curves must bc recalculated l wh determined ndt se

• frcmequivalent the next surveillance capsule exceeds the calcu'ated 14 7 8 X

capsule radiatica exposure. *

'/7rd- M k i A110waal ;

coold wn sure-tecoerature relationshi 5 for various hea'ao tse Section ates are calculated using cwthods cerived from Appencia 3 in Appendia G*to of13 tneCF ASPE Soiler and Pressure Vessel C de as. recuired oy d te ele * * * ' d

,' " ~ '

-- M < cuised in ce ta il i n m ;;=,: = r:: .r 2 ,

(AP-J4043-MP-A(_"4fl*>lcRM -

'-e<-- -- , ... --e' -~. "-< .., 2 <.

t_ = : _;= =: r > = e' = : r: " , ':--e < = z , - e1. = "---,

1

= m;3  :- e: = =:n?

enn e e; = e = :,= =;=g= :=,= :"==-  : > e: ",t ; t: = ? =r e x

-f =. . re ; u m

,'27 . .'.;. ..;;..,A

, . . ..=_. ..!;:

4 ^

. , , ,r,

' ,rm. - f C, , _^ f".M -

4 =e. ,em: =>

= m

e sm t= e- "

= : -- =r=:

'J ^ ^ = ? ._ .^ ? ' =. m .! ' =

=,:;'y ==:0 M =r==

r: - = =tt:f = c' frn r-f = ' :; ni =

.e.x-: . ... _=.... . m=e .... =... ... =..:.r n r ut= < >:. . . . . . . . . . . . . .=. =. =. . . .= . m. . ;. u n . .

,.;1.. 4,. ,.o.... .

1 tu ::::;=.=, e: n -; + - - - - .~< < (

e.m.n_,=_=....=.....".....<..,........"_...._.."..._...----

==:4 ; = e: 2 e e :r>:e -- -t --.=~ =2

-' =:?fr' =r:= =: ;; .^ t: d. i 1

dav 7% 7T40 SMAitbd, Au%edRk 6 fS & s S L Ti o r f 3H,4,10, M Moolfl&D, /& ,0 2 N/s R&

R&M /M 8SH 4.14 R>e CodnddirH r a cy-unt a m 4.s mcmaru.33.g x

4

. . - = . - . . - .. . - .. . - - - - - .

l .

1

. I 1 .

8 EEzI s -

4

),f 3e!i m a;- mi, s s a s = 2 = c e x z e:: x . .

i i

t. "  !

i e= .-

~ t r .5 2

---*i i

7773575337 77 i I3TE=

ggag.  %.3,%3*3383%..

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• j

.P I.

2333233**R *2*3t*3 3 2 .

=

=

~~-~~--- g :

- 3*3SS*S T  :

V  :

a. sis a si Ssat.a.gs. a. s s. s. a.  :. -

2 , -- i ee  : 2 l- 2 3:=====3 32:32 2 2

=5 ........ ....... v v 3. 3 al '

2 -

1 .

. .-~ .... ......-

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. X

e  : ;

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.~~~~~~~~ege~e g

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' d, I g{ a EI.EEE.E.E.I.RI.EE.EH.E.ll33

. .. ... g, l r .- j h1j ---~-t-ve--- v I!:

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dessssssa ss.l[ ssg'ssl{

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1s=

i FARLEY-UNIT 2 83/4 4-9 M NDMENT NO i

i 1

i 4

1 1

I 1

REACTOR COOLANT SYSTEM BASES The ASME approach for calculating the allowable limit curves for vario l hea and cooldown rates specifies that the total stress intensity facto ,

Ky , fo the combined thermal and pressure stresses at any time during atup l or cool n cannot be greater than the reference stress intensity fa or, K IR' for the me 1 temperature at that time. K IR is obtained from the eference l fracture tou ess curve, defined in Appendix G to the ASME Cod . The K IR curve is given the equation:

]

KIR = 26.78 + 23 exp [0.0145(T-RTNDT + 160)] (1) where K IR is the referen stress intensity factor a function of the metal ;

temperature T and the meta il ductility refere e temperature RT NDT. Thus,  !

the governing equation for th heatup-cooldown nalysis is defined in Appendix G of the ASME Code as follows CKgg + kit < KIR (2) l Where, K gg is the stress intensity acto caused by membrane (pressure) stress. ,

1 l

K It is the stress in nsity factor cause by the thermal gradients. I K

IR is provided y the code as'a function of erature relative to the RT NDT the material.

C = 2.0 r level A and B service limits, and C=1 for inservice hydrostatic and leak test operati s.

l At ny time during the heautp or cooldown transient, K7g is d ermined by the al temperature at the tip of the postulated flaw, the 'appropr te I val e for RTNDT, and the reference fracture toughness curve. The the 1 resses resulting from temperature gradients through the vessel wall are FARLEY-UNIT 2 8 3/4 4-11 AMhbbT NO.

REACTOR COOLANT SYSTEM BASES w_ - -

= - -

\

ca ulated and then the corresponding thernal stress intensity factor, K ,

for e reference flaw is computed.

From Equation (2) the pressure str ss intenst factors are obtained and from these, the allowable pressur are calculate COOLDOWN For the calcu tion of the allowable pressure versus oolant temperature during cooldown, th Code reference flaw is assumed to. st at the inside of the vessel wall. Our g cooldown, the controlling loc ion of the flaw is always at the inside o he wall because the thermal stresses at the inside, radients produce tensile ich increase with'increas' g cooldown rates. Allowable pressure-temperature relat ns are generated for th steady-state and finite cooldown rate situations. m these relations omposite limit curves are constructed for each cooldown te of interas The use of the composite curv in th cooldown analysis is necessary be-cause control of the cooldown proced e s based on measurement of reactor coolant temperature, whereas the limi i pressure is actually depenefent on the material temperature at the ti of the assumed flaw. During cooldown, the 1/4T vessel location is at a hi er temperat e than the fluid adjacent to the vessel ID. This condition, o course, is not t e for the steady-state situation.

It follows that at any give reactor coolant temp ature, the delta T developed during cooldown results a higher value of KIR at e 1/4T location for finite cooldown rates an for steady-stato operation. Furthermore, if conditions exist su that the increase in KIR exceeds kit the calculated allowable pressur during cooldown will be greater than the teady-state value. ,

The ove procedures are needed because there is no direct con al on tempera re at the 1/4T location; therefore, allowable pressures may unknow gly be violated if the rate of cooling is decreased at various i inter als along a cooldown ramp. The use of the composite curve elimina s .

thi problem and assures conservative operation of the system for the enti c Idown period.

~

FARLEY-UNIT 2 8 3/4 4-12 4A%dCweJr 4.

a i

REACTOR COOLANT SYSTEM BASES b

{ HEATUP bree separate calculations are required to determine the limit cur es 3 for fin te heatup rates. As is done in the cooldown analysis, allowa e pressure- mperature relationships are developed for steady-state c ditions

as well as nite heatup rate conditions assuming the presence of a 1/4T j

defect at the side of the vessel wall. The thermal gradient during heatup produce compress e stresses at the inside of the wall that leviate the f tensile stresses p duced by internal pressure. The meta temperature at the

{ crack tip. lags the co lant temperature; therefore, the IR for the 1/4T crack during heatup is lower an the K yg for the 1/4T cra during steady-state 4

conditions at the same co ant temperature. Durin heatup, especially at the j end of the transient, condit ns may exist such at the effects of compressive

thermal stresses and different R' s for stea -state and finite heatup rates l do not offset each other and the ressure-t erature curve based on steady-

{ state conditions no longer represen a er bound of all similar curves for i finite heatup rates when the 1/4T fla s considered. Therefore, both cases.

have to be analyzed in order to ass e at at any coolant temperature the lower value_of the allowable pres re calc ated for steady-state and finite heatup rates is obtained.

l The second portion of he heatup analysis co erns the calculation of pressure-temperature lim tions for the case in w ch a 1/4T deep'outside surface flaw is assume Unlike the situation at th vessel inside surface, the thermal gradients stablished at the outside surfa during heatup produce stresses which are' nsile in nature and thus tend to r nforce any pressure stresses present. hese thermal stresses, of course, are opendent on both the rate of heat and the time (or coolant temperature) a ng the heatup ramp. Furthe e, since the thermal stresses, at the outsi are tensile and increase with nereasing heatup rate, a lower bound curve cann t be defined.

Rather, eac eatup rate of interest must be analyzed on an ind idual basis.

Fo1 ing the generation of pressure-temperature curves for b h the steady- ate and finite heatup rate situations, the final limit cury are produ d as follows. A composite curve is constructed based on a poi -by-poi comparison of the steady state and finite heatup rate data. At a gi en temperature, the allowable ,nressure is taken to be the lesser of th ree values taken from the curves under consideration.

FARLEY-UNIT 2 8 3/4 4-13 k A O,

j REACTOR COOLANT SYSTEM -

BASES f

1  % .. , .6. . ..... ,,,,o. 4.

i 1i;;;itati::: 5:

..,...... .. .. , ... . 4o. 6. . t . . .

ceer:e ef the haete; 0 tere-=== it is a^t!4ble the centr-ol. for4ondit44ns-to-exist-such Ling-ceaditier switches #-a= +55t-over-the j

th;-cut:id: ad the ; . ester- 'i-it ="!st at a!' ti e k= ba:-* en-analy:S the in:ide4e-th: =:t critit:1 trtt rice- cf f

e i ..i i m .w. in ren o... en a....a4, e o,i. whi,6 .aa...... 6. ..i l

.tThi:

5 5hM ush~e f t5 c! A ir 3 EE '15aae'=a b a?==1 *1==;Usit k5 55Eiid:r:d.

P,21: :t:t :

th:t th: =f ri-"r etal te perature 4f th: clecere-f1::;;

j r;;i::: 5: :t let:t 120*" 'i;5er tht- the 1i?!tia; 8 % 'er the:e P ;i- -h:

th: pre ere Orteed! 20 per --t c' +ke a-asa-wiae hyd-astatic test arettere

! (621 ariq 'ar rirley Unit ?). != idditie=, the =~ 10 CF". " rt 50 a '.e :t:t::

! th:t :

t :;;;fff: fr::ter:  :"-lettie- :- y be pe-#e n d te justify 1:::

1i:ttin; c::;;ir;r;;t:. 5:M " en-suck 'r :turs-a=*!y-t- fer F:r! y U .it 2, j gg th: M EFav 5::te :d :: 14ewn-curves-are i- : td by the-new-10 CT", P:rt "O y h!e 1: :h: = := Figer:: 3.0 ::d 3.?-0.

^ l

_ __~ _

, l i

Although the pressurizer operates in temperature ranges above those for which i there is reason for concern of non-ductile failure, operating limits are i

j provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERA 8ILITY of either RHR relief valve or an RCS vent opening of greater i than or equal to 2.85 square inches ensures that the RCS will be protected from h pressure transients which could exceed the limits of Appendix G to 10 CFR Part

} 50 when one or more of the RCS cold legs are less than or equal to 310'F. ' I i

i Either RHR relief valve has adequate relieving capability to protect the RCS i

' from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less d h than or equal to 50*F above the RCS cold leg temperatures provided measures are j

j j taken to cushion the overpressure effects at RCS temperatures above 250*F, or i j (2) the start ofgharging pumps and their injection into a water solid RCS.

l g /4 3 1 11 STRUCTURAL M r**~! " (At.r_ d ' cibees jta i The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 i

components ensure that the structural integrity and operational readiness of l these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section II of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR i Part 50.55a(g i Commission pur)suant to 10 CFR Part 50.55a(g)(6)(1).except where spe I dk:tT4 1d tnt: CA46 0F THe IdJ6cnod BVMC#Ab4 RJMP5, T4 AdAL456 K W6D j odh 3rae;r oph M44MdM UdMB64 CF OF%d4&6 (Jabd PdMP5 l

c hu.On4:0 BY%TstAlIcbLSnomctncnJs.

4 d

a i

FARLEY-UNIT 2 8 3/4 4-14 AMENOMENT NO. 3 1

i 8 Wars Th filEA6 KAP 45 Idd.dC60 idTie fl)6&& AS M0C)(Ffeo, Af.6 Ma60 To 4

j BN: B BM 4 8 R4CafnudnV 0: 84S65 3/4.4.10.

i

ADMINISTRATIVE CONTROLS ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT 6.9.1.13 'his annual report is only required when the results of specific activity aaalyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be <

included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the fi.rst sample l in which the limit was exceeded (in graphic and tabular format): (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radiciodine concentrations: (3)

Clean-up flov history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro Ci/gm) and one other radioiodine isotope concentration (micro Ci/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.

IdMi'S .

=

w FouoW 45PECIAL REPORTS P4 6 '

6.9.2 Special reports shall be submitted to the Commission in accordance with the requirements of 10CFR50.4 vithin the time period specified for each report.

Reports should be submitted to the U. S. Nuclear' Regulatory Commission, ATTN:

Document Control Desk, Vashington, D.C. 20555.

6.10 RECORD RETENTION In a Fede,ddition to the applicable ral Regulations, record the following retention records requirements shall be retainedoffor Title 10, Code at least the of minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of unit operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. f-

c. ALL REPORTABLE EVENTS submitted to the Commission.  !

(. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes,made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 2 6-20 AMENDMENT NO. O kbre: 56Cped; G9. l.15, GA.2, AUD (,.10 Mo4D To 4 6-20a.

4

J e FARLEY NUCLEAR PLANT - UNIT 2 PTLR SUBMITTAL l TECHNICAL SPECIFICATIONS MARKUPS A

i INSERT 5 i

PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

]

1 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup and 3 j cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

i l The analytical methods used to determine the RCS pressure and temperature limits shall be those

' previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, Revision 2, " Methodology Used to Develop Cold Overpressure Mitigating i System Setpoints and RCS Heatup and Cooldown Limit Curves," approved by NRC SER dated October 16,1995, n

He PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

1 t

1 m

i e

4