ML20082K130

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Proposed Tech Specs 3.4.9.1 Re Schedule for Withdrawal of Reactor Vessel Matl Surveillance Specimens.Subj TS Table to Be Removed from TS & Placed in Chapter 16 of Updated FSAR
ML20082K130
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 08/23/1991
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20082K126 List:
References
NUDOCS 9108290213
Download: ML20082K130 (8)
v • d • e


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/WCCUM ~1 E T ill- AE-3M PAGE_.1._.0F S _ . . -

INDt'X LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION .P. AGE FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 32 EFPY................................. 3/4 4-33 '

TABLE 4.4-544E+C-TOR VES$[L liATE"d AL 5""VTliiANCC- PROGRA!!

ncRAWAL Scn[ m t .................................... 3/4 4-34 (Tw+

w osd) w % g - @ Pressurizer............................................., 3/4 4-35 l Overpressure Protection Systems.......................... 3/4 4-36 FIGURE 3.4-4 NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM ........................ 3/4 4-38 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-39 3/4.4.11 RF. ACTOR VESSEL HEAD VENTS........ ........... ........... 3/4 4-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T F.... 3/4 5-3 3yg GREATER THAN OR EQUAL TO 350 3/4.5.3 ECCS SUBSYSTEMS - T avg LESS THAN 350 F................... 3/4 5-6 ECSS SUBSYSTEMS - T avg LESS THAN OR EQUAL TO 200 F....... 3/4 5-8 3/4.5.4 (This specification number is not used).................. 3/4 5-9 3/4.5.5 REFUELING WATER STORAGE TANK.................. ... . .... 3/4 5-10 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM . .................... 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/A.6.1 PRIMARY CONTAINMENT Containment Integrity. ............................. . .. 3/4 6-1 Containment Leakage......... ... ....................... 3/4 6-2 Containment Air Locks.. ........ .. ... ............ .... 3/4 6-5 Internal Pressure............ . ... .................... 3/4 6-7 Air Temperature................................... . . .. 3/4 6-8 Containment Structural Integrity......................... 3/4 6-9 Containment Ventilation System........ . ............ ... 3/4 6-12 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System....... ... ........ . . ....... 3/4 6-14 Spray Additive System....... .. ... . .. .. ... .... 3/4 6-15 Containment Cooling System... .. . ...... .. .... 3/4 6-17 yluezvoziB ylv043 PDR ADOCR 05000498 P PDR SOUTH TEXAS - UNITS 1 & 2 viii

! b v$$f553 sT.HL AE 3E 4 9 REACTOR COOLANT SYSTEM PAGE __Il-_ 0F i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REAC10R COOLANT SYSTEM L111LIK_COND11101LE01L0ELRA110N _

3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3 4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100 F in any 1-hour period,
b. A maximum cooldown of 100 F in any 1-hour period, ano
c. A maximum temperature change of less than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the ef fects of the out of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STAN0BY 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 />.

EURVEILLMCLREQUIREMBLS &

4.4.9.1.1 The Reactor Coolant c tem temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, i as required by 10 CFR Part 50, Appendix H, -4n accordeme with the schedah--

in Idule 4.4 5.- The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-31

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ATTACHMENT A l SbHL AE.38 Ljtf (Ws hkh now.kr m,t us,,a) PAGE_ 3___ 0F 5 TABLE 4.4-5

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RE ET5R VESSEL MATERTAI3URVEILTANCE PROTIRAM - WITHDRAWAE'5CHEDULL f' CAPSULE VESSEL LEAD j NUMBER 4CbT10N FACTOR

/'WHtt01fAWAL TIME (Ef PY)

U 58.5* 4# first Refueling V 3 9 X 238'$ 4.00 15 W 121.5" 4.00 t.qdby Z 301.5 4.00 Standby SOUTH TEXAS - UNITS 1 & 2 3/4 4-34

7T1 ACHMENT &

REACTOR COOLANT SYSTEM ST-HL AE - 0FMW '$

PA .- . 4

__.GE.

fB515 PRESSURE / TEMPERATURE LIMITS (Continued) can cause an increase in the RT NDT. Therefore, an adjusted reference tempera-ture, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the value of ART NDT computed by Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heat-up and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT at the end of 32 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. l j Values of ART NDT determined in this manner may be used until the results i from the material surveillance program, evaluated according to ASTM E185, are , ,

available. Capsules will be removed in accordance with the requirements of i ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance speci= n with-

= drewal schedule is shown in Tabic 4.4-5. The lesd facter represents the rela-

--tion; hip between tk fast neutron flux-dea-ity at the ' cation of the capsule

-end-the inner wall of th^ reactor vessel Therefererheresultsobtainedfrom the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cool-down rates are calculated using methods derived from Appendix G in Se: tion III of 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-7924-A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME Sec-tion III as the ieference flaw, amply exceed the current capabilities of inser-vice inspection techniques. Therefore, the reactor operation limit curves de-veloped for this reference crack are conservative and provide suf ficient safety margins for protertion against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, j Kg , for the combined thermal and pressure stresses at any time during heatup l or cooldown cannot be greater than the reference stress intensity factor, K IR' for the metal temperature at that time. K IR is b'ained from the reference SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-8

WiTACHMENT 2 ST-HL-AE-3 WV i REAbTORCOOLANTSYSTEM PAGE 3 - 0F S l

1

- -BASES ,

l LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) overshoot beyond the PORV Setpoint_which can occur as a result of time delays ,

in signal processing and valve opening, instrument uncertainties, and single "

failure. To ensure that mass and heat input transients more severe than those assumed cannot occur Technical Specifications require lockout of all high head safety injection pumps while in. MODE 5 and MODE 6 with the reactor-vessel head

-on. All but one high head safety injection pump are required to be locked out -

in MODE-4. . Technical Specifications also require lockout of the positive displacement pump and all but one charging pump while-in MODES 4, 5, and 6 with the reactor vessel head installed and-disallow start of an RCP if secondary

-temperature is more than 50*F above primary temperature. -

The Maximum Allowed PORV Setpoint for the COMS will be updated based on the results of examinations of reactor vessel- material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance-

-with the schedule in Tcble 1.'-5.

3/4.4.10--STRUCTURAL INTEGRITY The i_nservice 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

l. 10 CFR 50.55a(g) except where specific written relief has_been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i),

Components of the Reactor Coolant System:were designed to-provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Winter 1975.

3/4.4.11 REACTOR VESSEL HEAD VENTS

, Reactor _ vessel head vents are provided to exhaust noncondensible gases -

and/or steam'from the Reactor Coolant System that could_ inhibit natural circulation core cooling... The OPERABILITY of at least two reactor vessel head l- . vent. paths ensures that the capability exists to perform this function.

The.' valve redundancy of the reactor vessel head vent paths serves to mini-mize the probability of inadvertent or irreversible actuation while ensuring

- that a single- failure of _a vent valve, power supply,- or control system does not prevent isolation of.the-vent path.

The function, capabilities, and testing requirements of the reactor vessel

, head' vents:are consistent with the requirements of Item II.B.1 of'NUREG-0737,

" Clarification of THI_ Action Plan Requirements," November 1980, i=

. SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-15 Unit 1 - Arr,andment No. 4

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ATTACllMENT 3 PROPOSED UFSAR CllANGES T3C \ 91- 169. 001

__A TACHMENT 3 STPEGS UFSAR SI-HL AE-37 y4 PAGE I -- O F A

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IbBLE OF CONTENTS CHAPTER 16 TECHNICAL SPECIFICATIONS LIST OF TABLES fatt Iahl.t 11.tl.t Containment Isolation Valves 16.1 2 16.1 1 llo.1 Seuchtr Nes:(( Mderini Sorye,glqnce _

h(ogro,m- WMk& tad Scheiolt TC 16 1 Revision 0

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. ATTACHMENT 3 - '

b I-' g ST HL AE 384 tl PAGE 3 0F L ,

TABLE REACTOR VESSEL MATCRIAL SURVEILLANCE PROGRAM - WITH0RAWAL SCHEDULE CAPSULE . VESSEL LEAD NUMBER LOCATION FACTOR WITH0RAWAL TIME (EFPY) l V 58.5' 4.00 First Refueling i Y 241 3.69 5 i V 61' 3.69 9 i X 238.5 4.00 15 -l W 121.5' 4.00 Standby t Z 301,5' 4.00 5tandby 3 I

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