Text

Proposed Tech Specs Re Limiting Condition for Operation & Surveillance Requirements for PORV & Block Valves
	ML20065T525
Person / Time
Site:	South Texas
Issue date:	12/21/1990
From:	; HOUSTON LIGHTING & POWER CO.
To:	;
Shared Package
ML20065T515	List: ML20065T513; ST-HL-AE-3642, Proposed Tech Specs Re Limiting Condition for Operation & Surveillance Requirements for PORV & Block Valves;
References
	ST-HL-AE-3642, NUDOCS 9012280171
	Download: ML20065T525 (13)
	v • d • e

_ _ - ______-

=

ENCLOSURE 2 TECIINICAL SPECIFICATION CIIANGES SOUTil TEXAS PROJECT ELECTRIC GENERATING STATION LIST OF AFFECTED PAGES 3/4 4-10 3/4 4-11 3/4 4-31 3/4 4-36 B 3/4 4-15 A1/038.H10 9012280171 901221 PDR ADOCK 05000498 P PDR

_ _ _ _ __ _ . = . _ _ __ _ _ _ _

s~ o c_ l w~

[ } 3 7 A e u n a r n't i

REACTOR COOLANT SYSTEM f%kI. $bV2 9 I PAGE I -OF 3/4. 4. 4 RELIEF, MV !

LIMIT 19G COND1T10N FOR_0PERATION 3.4.4 At1 power-operated relief valves (PORVs) and their associated bicek valves shall be OPERABLE.

~

APPLICABILITY: MODES 1, 2, and 3.

ACTION: , j h ,

a. %1th one er.ame,PORV(s) inoperable, because of excessive seat s leakage or close, within I hour either restore t.he PORV(s) to OPERABLE status the associated block valve (sh otherwise, be in at least

HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months M and in 49ttT SHUTDOWN

withi1 g following g hours. ( },,

E b.

]

With one PORV inoperable due to causes other than excessive seat

~ leakage, within I hour either restore the PORV to OPERABLE ste.us or

- close the associated block valve and remove power from the blo:k valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ~or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cab SHUTDOWNwithinthefollowingghours. . Hert L

c. With both ;r-Mb inoperable due to causes other han excessive seat leakage, within I hour either restore each of t eJMYtv) to 0)ERABLE L - status or close their associated block wM(3) nd remove power from the blockNaluefir) and.be in HOT STANDBY w thin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and gSHUT004gn the following yhours.,,,,94,,.3 With o or more bio valve (s) ino etable, within hour: '"*

l J -

-(1) rest the block lve(s) to 0F BLE status, o close tha I ek valve and remove ower from t block valve (s or c1:se the RV and re e' power f the PORV; d (2) apply t.

or c. ove, as ap opriate, the isolat CT.ON . b.,

,. . i

'*".,a PORV(s)(, ,

k j The provisions'of Specification 3.0.4 are not applicable.

e SOUTH TEXAS - UNITS 1 & 2 3/4 4-10

- INSERT A :

..- rta.d o%

ATfA0liplENT~

~

ST HL AE-%4 D PAGE- N _OF- ? _ _-

d. With-one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve to ';!

operable status or place its associated PORV in closed position; restore the '

block valve to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHlTIDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i

e. With both block valves inoperable, within I hour restore the block valves to j operable status or place the associated PORVs in closed position; restore at least one block valve to OPERABLE status within the next hour; restore the ;

remaining block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; othenvise, bc ;

in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT -

SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e i

m ,maa e a-. ,n. .n,-

, . ,- < < . . . ,--ww.v. a e - e

. - ,, , ,- - w

mBY.S$&;S_W REACTOR-COOLANT SYSTEM NHI*iEivi-ST HL-AE-y.4%

REllEF VAtyg -. PAGE .,3 i

_ 0F P _

SURVElLLggJqu]pfgigis 4.4.4.1 .

In addition to the requirements of.5pecification 4.0.5. each FORY shall be demonstrated OPERABLE at least once per 18 months by:

a. Performing a CHANNEL CALIBRATIOW and on W a. & Mos M \> ~ ~

// p dm, b.

Operating the valve through one complete cycle of full travehb4j'd h,

4.4.4.2 Edh block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with F.:

of ACTION b or c. in Specification 3.4.4.27 Tinvid in order to meet the requirements l

l 1 :

1' l

50llTH TEXAS - UNITS 3 & 2 3/4 4-21

. +

-zadz w

A , .I,tN7 REACTOR COOLANT SYSTEM b g

3/4.4.9 PRE $50RE/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 11M1 TING CONDITl0N FOR OPERATION '

3.4.9.1 ._

The Reactor Coolant System (except the pressurizer) temperature and - -

3.4-2 and 3.4-3 during heatup, cooldown, criticality, and ins hydrostatic testing with:

a.

A maximum heatup of 1' 00'F in arty 1-hour period, ,

b.

A maximum cooldown of 100'F in any 1-hour period, and

c. .

A maximum temperature change of less than or. equal to 10'F it. any -

1-hour period during inservice hydrostatic and leak testing coerations , -

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 evaluaticito of the Reactor Coolant Systemdetermine the effects of the out-of-limit con

.9 acceptable for continued opera; tion or be in at least HOT STANDBY next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SbRVEllLANCEREOUIREMENTS 4.4.9.1.1 . . . .. . ..

determined to be within the limits at least once per 30 minutesThe . . .-

heatup, 'cooldown, and inservice leak and hydrostatic testing ope,during rations. system

.4.4.9.1.2 'The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, .

as required in Table by 10 CFR Part 50, Appendix H, in accordance with the schedule 4.4-5.

Figures 3.4-2)end 3.4-3/The respits of. these examinations shall be us Y o.d L 4-4 SOUTH TEXAS - UNITS I & 2 3/4 4-31

p. s

TAC t'

rwari 19lE%L._

{ST-HL-AE-REAtTOR COOL ANT SYSTEM u PAGE ._si__._. OF 9

{

yERPRES$URE PROTECTION SYSTEMS HM1DNG co@fTION FOR OPERAT10N 3.4.9.3' be OPERABLE: At least one of the following Overpressure Protection Systems shall

a. ..

Two power-operated relief valves (PORVs) with lift settings which do not exceed the Ilmit established in figure 3.4-4, or b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than cr equal to 2.0 squpe inches.

APPLICABILITY: MDDES 4 and 5, ad MODI 6 b.ke4

SSb "

ACTIg '

g q, ,', , , , l, , ",,' ,,',,l, ( ', , ,,

s. With one PORV inoperab1p restore the inoperable PORV to OPERAbtE status.withth 7 days or :,lepressurize and vent the RCS through at m .Jeast : 2.0 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

h W th both PORVs inoperable, depressurize and vent the RCS through at least a 2.0 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .4 c

In the gate event an RCS either pressure the PORVs transient or the RCS vent (s) are used to m 3 and submitted within 30 days. to the Commission, pursuant to Specification 6.9.2a The report shall describe the circumstances initi-ating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e. % pds% c4 g.h% s.o.4 on a 9 abb, t

.. . ...: . . r .

. e . .: . . .;

4verify This PORV ' ACTION operability. may be suspended for up to 7 days to allow f

, During this test period, operation of systems or administratively controlled. components which could result in an RCS mass -

During the ASME stroke testing of two inoperable PORVs, cold overpressurization mitigation will be provided by two RH relief valves associated with two OPERABLE and operating RHR loops whic the auto closure interlock bypassed (or deleted). If one PORV ,

is inoperable overpressure mitigation will be provided by the OPERABLE PORV and one RHR t which has the auto closure interlock bypassed (or de SOUTH TEXAS - UNITS I & 2 3/4 4-36

_ l

~

~ENSE23h%7 ~ __

ST Hl.-AE4441 INSERT B PAGE __ 6- 0F _ 9 __

j

b. With one PORV inoperable in MODES 5 or 6 with the head on the reactor vessel, restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or complete depressurization and venting of the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .*

i

n .

i-

!}

g(os W REACTOR COOLANT SYSTEM 5$UAE-3@F PAGE 7 _0 j l

BASES '

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

- fail ure.- To ensure that mass and heat input transients more severe than those ,,

assumed cannot occur, Technical Specifications require lockout of all high head l

safetyinjectionpu s while in MODE $ and MODE 6 with the reactor vessel head ')

on. All but one hl head safety injection pump are required to be locked out in MODE 4.-- Technic 1 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.

L T 'The Maximum Allowed PORY Setpoint for the COMS will be updated based on-the-I 1.--- resultsi of examinations of reactor vessel material irradiation surveillance

.1#(=dW ( spec with mens performedinas the schedule Table required by 10 CFR Part 50, Appendix-H, and in accordance 4.4-5.

3/4.4.10' STRUCTURAL INTEGRITY '

The, inservice inspection and testing programs for ASME Code' Class 3, 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 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-i Boiler and Pressure Vessel Code,1974 Edition and Addenda through Winter 1975. ;

L 3/4.4.11- REACTOR VESSEL HEAD VENTS

' Retetor. 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 perfore this function.

i. .

)

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-l prevent isolation of the vent path.

The function, capsbilities, 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.

L SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-15 linit 1 - Amentent No. 4 i ._-

ROosw fqq -

INSERT C ST HL AE b .t PAGE J __0F _ S

-w_a Administrative controls and two RHR relief valves will be used to provide cold overpressure protection (COMS) during the ASME stroke testing of two administratively declared inoperable PORVs. During the performance of the PORV functional test, two RHR trains will be OPERABLE and in operation with the auto closure interlock bypassed [or deleted] to provide COMS. Each RHR relief valve provides sufficient capacity to relieve the flow resulting from the maximum charging flow with concunent loss of letdown. With one PORV inoperable, COMS will be provided during the ASME test by the OPERABLE PORV and one RHR relief valve associated with an OPERABLE and operating RHR train which has the auto closure interlock bypassed [or deleted). The RHR pump design developed head, corresponding to the design flowrate of 3400 gpm, is 205 ft and the actual pump developed pressure is 115 psig. This results in actuation of the RHR relief valves at a RCS pressure of approximately 485 psig (600 psig - 115 psig). Therefore two OPERABLE and operating RHR trains at one OPERABLE PORV and one OPERABLE and operating RHR train will provide adequate and redundant overpressure protection. Use of the RHR relief valves will maintain the RCS pressure below the low temperature endpoint of the Technical Specification limit curve (550 psig, ref. Technical Specification fig. 3.4-2). With regard to the MODE 6 applicability of this Technical Specification, the statement "with the head on the reactor vessel" means any time the head is installed with or without tensioning the RPV studs.

marnar l

q Attachment 2 ST HL AE 3642 Page 1 of 4 SIGNIFICANT HAZARDS EVALUATION HL&P has incorporated the recommendations provided in Enclosure A and Enclosure B of Generic Letter 90-06. The description of changes and associated justifications are given in Enclosure 1 of this letter. These changes improve the clarity and accuracy of the Technical Specifications and increase the reliability and availability of the PORVs. HL&P has evaluated the proposed changes to the Technical Specifications and has determined that these changes do not represent a significant hazards consideration based on the criteria established in 10 CFR 50.92(c). Incorporation of the recommendations of Generic Letter 90 06 and performance of the proposed PORV operability test will not:

(1) involve a sienificant increase in the probability or consecuences of an ggcident oreviousiv evaluated in the Safety Analyses Report.

GL 90-06 Amendment:

The incorporation of the changes provided in Enclosure 2 are consistent with the recommendations of Generic Letter 90 06 and as such improve the clarity and accuracy of the Technical Specifications and do not increase the probability or consequences of any accident previously evaluated in the Safety Analysis Report.

PORV Operability Verifica 10D:

Administrative controls and procedures have been structured to aid the

-operator in controlling RCS pressure during low temperature operation.

However to provide a backup to the operator, an automatic system is provided to maintain pressures within allowable limits.

Evaluations presented in the Safety Analyses Report have shown that one pressurizer PORV is sufficient to prevent violation of the limits established by ASME III, Appendix C due to anticipated mass and heat input transients.

Redundant protection against a low temperature overpressure event is provided by using two pressurizer PORVs to mitigate potential pressure transients. The automatic system is required only during low temperature water solid operation when it is manually armed and automatically actuated. The STPECS PORVs are safety-related and Class lE powered. They are designed in accordance with the ASME Code, are qualified via the Westinghouse pump and valve operability program, and are seismically and environmentally qualified.

Low temperature overpressure events have been previously evaluated in section 5.2.2.11 of the STriGS UFSAR. These events result from potential increases in mass or heat input into the Reactor Coolant System (RCS) due to a I charging / letdown flow mismatch or inadvertent Reactor Coolant Pump actuation I with a temperature mismatch between the RCS and the secondary side of the Steam Cenerators of 50 F. The probability of a low temperature overpressure )

event due to these initiators is unchanged since the proposed test does not l l involve any changes to plant systems, equipment or controls. During the ASME I l

A1/038 H19

1 . .

Attachment 2 ST-HL-AE 3642 Page 2 of 4 stroke test of two inoperable PORVs, overpressure protection will be provided by operation of two RHR trains. Each RHR discharge relief valve has sufficient capacity to relieve the flow resulting from the maximum charging flow and concurrent loss of letdown.

These RHR relief valves have a setpoint of 600 psi and will actuate at an RCS pressure of 485 psig due to the 115 psig RHR pump head. Therefore, the two OPERABLE and operating RHR trains, with the RHR auto closure interlock bypassed (or deleted), will provide adequate and redundant cold overpressure protection during the proposed test. If only one PORV is inoperable, redundant cold overpressure protection will be provided by the OPERABLE PORV and one OPERABLE and operating RHR train with the RHR auto closure interlock bypassed [or deleted), Operator action to terminate the overpressure event, actuation of the OPERABLE PORV, actuation of one or both of the RHR discharge relief valves, or actuation of the PORV(s) being tested will assure that the accident consequences remain unchanged. The consequences of a low temperature overpressure event, as previously evaluated in the UFSAR, show that the allowable limits as established by ASME III, Appendix G will not be exceeded and therefore Reactor Pressure Vessel integrity and plant safety will be maintained.

/ During operations with the RCS water solid and the COMS PORV(s) unavailable, administrative controls will be implemented to minimize the potential for and severity of postulated overpressure transients. These controls incorporate the following:

a. When RCS pressure is being maintained %y the low pressure letdown control valve, the normal letdown orifices are bypassed but not isolated,

b. Only one centrifugal charging pump (CCP) will be allowed to be operable; this minimizes the potential for a mass input overpressure transient.

c. Administrative controls will be in place to insure that the High Head Safety Injection (HHSI) pumps will not operate during water solid operations with the PORV(s) inoperabic to minimize the potential for creating a cold overpressure transient.

d. The RPV pressure will be controlled at the minimum value necessary to perform the required testing of the inoperable PORV(s) (325-400 Psig).

e. A Reactor Coolant Pump shall not be started with one or more of the RCS cold leg temperatures less that or equal to 350 F unless the secondary side water temperature of each steam generator is less than 50 F above the RCS cold leg temperature (ref. Technical Specification 3.4.1.4.1.a).

A1/038.H19

g

=

9 Attachment 2 ST-HL-AE 3642 Page 3 of 4

f. The positive displacement pump will be demonstrated inoperable during i the water solid operations to minimize the potential for a mass input overpressure event,

g. The RHR auto closure interlock will be bypassed (or deleted) during water solid operations to prevent the loss of letdown capability which could produce a mass input overpressure transient.

h. The Pressurizer Heaters will be inoperable during water solid operations to minimize the potential for a heat input overpressure transient.

As a result of the above administrative controls, the operability of the OPERABLE PORV, the operability of the RHR discharge relief valve (s), and the expected operation of-the PORV(s) being tested, there is no significant increase in the probability or consequences of a low temperature overpressure event, as previously evaluated in the UFSAR. The allowable limits, as established by ASME III, Appendix G, will not be exceeded and therefore Reactor Pressure ~ Vessel integrity and plant safety will be maintained.

.(2) create the oossibility of a new or different kind of accident from any

-previousiv analvred.

OL 90-06 Amendment:

The incorporation of the changes provided in Enclosure 2 are consistent with the recommendations of Generic Letter 90 06. These changes and the additional changes to allow verification of PORV operability during MODES 5 and 6 increase the clarity and accuracy of the Technical Specifications and do not create the possibility of a new or different kind of accident from any

.previously analyzed.

PORV Operability Verification:

Iow temperature overpressure events resulting from inadvertent mass or heat input into the RCS have been previously evaluated in the STPEGS UFSAR. The use of additional administrative controls during water solid operations with one or both COMS'PORVs inoperable does not result in the creation of a new or different kind of accident. Application of these additional controls while performing the required testing of the inoperable COMS PORV(s) ensures'that

.the potential for a low temperature overpressure event is minimized.

(3) involve a sinnificant reduction in the marnin of safety.

GL 90-06 Amendment:

The incorporation of the changes provided in Enclosure 2 are consistent with the-recommendations of Generic Letter 90-06. These changes and the additional changes to allow verification of PORV operability during MODES 5 and 6 increase the clarity and accuracy of the Technical Specifications and do not involve a significant reduction in the margin of safety.

A1/038.H19

9 Attachment 2 ST-HL AE 3642 Page 4 of 4 PORV Operability Verification:

The margin of safety is provided by the difference between the ASME Appendix G limits and the actual pressure capability of the Nuclear Grade Reactor Pressure Vessel. The margins contained within the ASME Appendix G limits provide assurance that vessel integrity is maintained under all operating conditions. ASME Section III, Appendix G, establishes guidelines and limits for RCS pressure primarily for low temperature conditions (s 350'F). Transient analyses have been performed to determine the maximum pressure for the postulated (credible) worst case mass input and heat input events.

The mass input transient is divided into two parts for plant operation in Mode 4 (> 200'F) and Mode 5 (s 200'F). In Mode 4, the mass input transient assumes the operation of one high-head safety injection (HHSI) pump and one centrifugal charging pump (CCP) delivering normal charging flow through the reactor coolant pump (RCP) seals with letdown isolated. It should be noted that the safety injection (SI) signal which isolates letdown also isolates the normal charging line. In Mode 5, the mass input transient assumes the operation of one CCP delivering flow through the RCP seals with letdown isolated.

The heat input analysis was performed for an inadvertent RCP start assuming that the RCS was water solid at the initiation of the event and that a SG'F mismatch existed between the RCS and the secondary side of the Steam Generators. (At lower temperatures, the mass input case is the limiting transient condition.)

Both heat input and mass input analyses took into account the single failure criteria and therefore, only one PORV was assumed to be available for pressure relief. The evaluation of the transient results concludes that the allowable limits will not be exceeded and therefore cold overpressure transients will not constitute an impairment to vessel integrity and plant safety.

These margins are incorporated into the STPECS Technical Specifications and are unchanged by the proposed PORV test. The administrative limits provided in the Technical Specification figure also contain additional margin due to accounting for possible instrument errors, inaccuracy and sensing delays, and valve opening time. The possibility of a cold overpressure event during the testing of the inoperable PORV(s) is considered remote. Even in the unlikely event that such an event were to occur, prompt operator action, actuation of the OPERABLE PORV, actuation of the RHR relief valve (s), or operation of the PORV(s) being tested will terminate the event befoce reaching the Appendix G limits. Consequently, the margins provided by the ASME III, Appendix G limits will be maintained.

A1/038.N19

- - _ _ - - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ - _ _ _ _ _ - __________________-__ _____--____ _ ______ _-

ST-HL-AE-3642, Proposed Tech Specs Re Limiting Condition for Operation & Surveillance Requirements for PORV & Block Valves

Text

Navigation menu

Search