Text

Amends 198 & 198 to Licenses DPR-32 & DPR-37,respectively, Incorporating Operability & SR for power-operated Relief Valves to Conform W/Gl 90-06
	ML20082V272
Person / Time
Site:	Surry
Issue date:	05/02/1995
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20082V277	List: ML20082V272; ML20082V284;
References
	GL-90-06, GL-90-6, NUDOCS 9505090157
	Download: ML20082V272 (18)
	v • d • e

p-l

./@MCp'g

-[I > M ' $

UNITED STATES

-F NUCLEAR FIEGULATORY COMMISSION

%,, *.. /

WASHINGTON. D.C. 2055M001 l

I i

VIRGINIA ELECTRIC AND POWER' COMPANY DOCKET NO. 50-280 l

E'.QRY_20%ER STATION. UNIT NO.1 6MENDMENT TO FACILITY OPERATING LICENSE i

Amendment No.198 License No. DPR-32 1.

The Nuclear Regulatory Comission (the Comission) has found that:

i A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated June 28, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as i

amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in ::onformity with the application, the I

provisions of the Act, and the rules and regulations of the Comission; 1

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public;-

and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

t e

i i

i 9505090157 950502 PDR ADOCK 05000200 p

,PDR

l

- 2'-

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating Licerse io. DPR-32 is hereby amended to read as follows:

(B)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.198, are hereby incorporated in the '

license. The licensee.shall operate the facility in accordance i

with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COPMISSION j

d A

David B. Matthews, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 2,1995

h,l pf "*%c,,

l n

r

'8-E UNITED STATES

I NUCLEAR REGULATORY COMM SSION

,/ '

WASHING 1rON, D.C. 205554001

...+

I VIRGINIA ELECTRIC AND POWER COMPANY j

DOCKET NO. 50-281 SURRY POWEP. STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE bP 37 o

i l

1.

-The Nuclear Regulatory Comission (the Comission) has found that:

i t

A.

The ~ application for amendment by Virginia Electric and Power Company (the licensee) dated June 28, 1991, complies with the standards and requirements of the Atomic. Energy Act of 1954, as amended -(the Act) and the Commission's rules and regulations set.

l forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; Lj D.

The issuance of.this amendment will not be inimical to the common defense and security or to the health and _ safety of the public; i

and E.

The issuance of this amendment is in accordance with 10 CFR Part l

51 of the Commission's regulations and all applicable requirements have been satisfied.

i l

t i

1 I

i

l l

! 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.198, are hereby incorporated in the license. The Itcensee shall operate the facility in accordance i

with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMISSION o

David B. i.atthews, Director Project Directorate II-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 2,1995 w

_w

^

' ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.198 TO FACILITY OPERATING LICENSE NO. DPR-32 ANENDMENT NO.198 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NGS. 50-280 AND 50-281 Revise. Appendix A as'follows:

Remove Paaes Insert Paaes TS 3.1-4 TS 3.1 TS 3.1-4a TS 3.1-5 TS 3.1-5 TS 3.1-5b TS 3.1-5b TS 3.1-5c TS 3.1-Sc TS 3.1-22 TS 3.1-22 TS 3.1-23 TS 3.1-23 TS 3.1-23a TS 3.1-24 TS 3.1-24 TS 4.1-1 TS 4.1-1 TS 4.1-la TS 4.1-la TS 4.1-9c TS 4.1-9c TS 4.1-9d TS 4.1-9d

r-

}

TS 3.1-4 4.

Reactor Coolant Loops Loop stop valves shall not be clcsed in more than one loop unless the Reactor Coolant System is connected to the Residual Heat Removal System and the Residual Heat Removal System is OPERABLE.

l 5.

Pressurizer a.

The reactor shall be maintained suberitical by at least 1%

until the steam bubble is established and the necessary sprays and at least 125 KW of heaters are operable.

b.

With the pressurizer inoperable due to inoperable pressurizer heaters, restore the inoperable heaters within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Reactor Coolant System temperature and pressure less than 350 F and 450 psig, respectively, within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

?%C c.

With the pressurizer otherwise inoperable, be in at least HOT GHUTDOWN with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Reactor Coolant System temperature and pressure less than 350 F and 450 psig, respectively, within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Amendment Nos.198 and 198

TS 3.1-4a l 6.

Relief Valves Two power operated relief valves (PORVs) and their associated block valves shall be OPERABLE

whenever the Reactor Coolant System average temperature ;s 2350 F.

a.

With one or both PORVs inoperable but capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and maintain power to the associated block valve (s).

Otherwise, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System average temperature in <350 F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one PORV inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either re: tore the PORV to OPERABLE status or capable of being manually cycled or close the associated block valve and remove power from the block valve.

In addition, restore the PORV to OPERABLE status or capable of being manually cycled within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Otherwise, ba in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System average temperature to <350 F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With both PORVs inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore at least 1 PORV to OPERABLE status or capable of being manually cycled.

Otherwise, close the associated block valves and remove power from the block valves.

In addition, be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System average temperature to <350 F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Automatic actuation capability may be blocked when Reactor Coolant System pressure is below 2000 psig.

Amendment Nos.198 and 198

i L

TS 3.1-5 I

I l

d.

With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either l

restore the block valve to OPERABLE status or place the L

associated PORV in manual. In addition, restore the block l

valve to OPERABLE status in the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce -

reactor coolant average temperature to <350 F within the

]

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

e.

With both block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valves to OPERABLE status or place the associated PORVs in manual. Restore at least 1 block valve to OPERABLE status within the next hour or, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce reactor coolant average temperature to <350 F within the I

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

1 7.

Reactor Vessel Head Vents l

a.

At least two Reactor Vessel Head vent paths consisting of two isolation valves in series powered from emergency buses shall be OPERABLE and closed whenever RCS l temperature and pressure ars >350 F and 450 psig.

Amendment Nos.198 and 198

TS 3.1-5b One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor i

shutdown. The requirement for redundant coolant loops ensures the capability to remove core decay heat when the Reactor Coolant System average temperature is I less than or equal to 350 F.

Because of the low-low steam generator water level reactor trip, normal, reactor criticality cannot be achieved without water in the steam generators in reactor coolant loops with open loop stop valves. The requirement for two OPERABLE steam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for Reactor Coolant System temperatures of greater than 350 F.

Each of the pressurizer safety valves is designed to relieve 295,000 lbs per hr. of saturated steam at the valve setpoint. Two safety valves have a capacity greater than the maximum surge rate resulting from complete loss of load.(2)

The limitation specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate.the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.

The requirement for steam bubble formation in the pressurizer when the reactor passes 1% subcriticality will ensure that the Reactor Coolant System will not be solid when criticality is achieved.

The requirement that 125 K,; of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT SHUTDOWN.

l 9

Amendment Nos.198 and 198

TS 3.1-Sc l

l i

The power operated relief valves (PORVs) operate to relieve Reactor Coolant System '

pressure below the setting of the pressurizer code safety valves. The PORVs and their

{

associated block valves may be used by the unit operators to depressurize the Reactor Coolant System to recover from certain transients if normal pressurizer spray is not available. Specifically, cycling of the PORVs is required to mitigate the consequences of a design basis steam generator tube rupture accident. Therefore, whenever a PORV is inoperable, but capable of being manually cycled, the associated block valve -

will be closed with its power maintained. The capability to cycle the PORVs is verified during each refueling outage (and is not required during power operations). These relief valves have remotely cperated block valves to provide a positive shutoff capability should a relief valve leak excessively. The electrical power for both the relief valves and the block valves is supplied from an emergency power source to ensure the ability to seal this possible Reactor Coolant System leakage path.

The accumulation of non-condensable gases in the Reactor Coolant System may result from sudden depressurization, accumulator discharges and/or inadequate core cooling conditions. The function of the Reactor Vessel Head Vent is to remove non-condensable gases from the reactor vessel head. The Reactor Vessel Head Vent is l

designed with redundant safety grade vent paths. Venting of non-condensable gases from the pressurizer steam space is provided primarily through the Pressurizer PORVs.

The pressurizer is, however, equipped with a steam space vent designed with redundant safety grade vent paths.

References (1)

UFGAR Soction 14.2.9 (2)

UFSAR Section 14.2.10 Amendment Nos.198 and 198 i

h TS 3.1-22 Basis By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the limits as specified in technical specification 3.1.F.1 and 3.1.F.4 the integrity of the reactor coolant system is assured under all operating conditions.(1) If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion excnange resin, or adjustment of the hydrogen concentration in the volume control tank.(2) Because of the time dependent nature of any adverse effects arising from oxygen, chlonde, and fluoride concentration in excess of the limits, it is not necessary to shutdown immediately if the condition can be corrected. Thus the period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for corrective action to restore concentrations within the limits has been established. If the corrective action has not been effective at the end of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, then the reactor will be brought to COLD SHUTDOWN and the corrective action will continue.

In restoring the contaminant concentrations to within specification limits in the event such limits were exceeded, mixing of the primary coolant with the reactor coolant pumps may be required. This will result in a small heatup of short duration which will not increase the average coolant temperature above 250 F.

More than one contaminant transient, in any seven consecutive day period, that results in exceeding normal steady state operation limits, could be indicative of unforeseen chemistry control problems.

Such potential problems warrant investigation, correction and measures to insure that the integrity of the Reactor Coolant System is maintained.

References (1)

UFSAR 4.2 (2)

UFSAR 9.2 Amendment Nos.198 and 198

TS 3.1-23 G.

Reactor Coolant System Overoressure Mitiaation Soecification 1.

The Reactor Coolant System (RCS) overpressure mitigating system shall be OPERABLE as described below; a.

Whenever the RCS average temperatura is greater than 350 F, a bubble shall exist in the pressurizer with the necessary sprays and heaters OPERABLE.

b.

Whenever the RCS average temperature is less than or equal to 350 F and the reactor vessel head is bolted:

(1)

A maximum of one charging pump shall be OPERABLE, and (a)

Prior to decreasing RCS average temperature below 350 F verify a maximum of one charging pump is capable of injecting into the RCS, thereafter (b)

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that a maximum of one charging pump is capable of injecting into the RCS, (c)

Two charging pumps may be in operation momentarily during transfer of operation from one charging pump to another.

(2)

The accumulators shall be isolated, (accumulator discharge valves closed and their respective breakers locked open),

and (a) Prior to decreasing RCS average temperature below 350 F verify that each accumulator is isolated, thereafter (b) Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that each accumulator is

isolated, (c) The accumulator discharge isolation valves may be opened for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to perform valve testing.

Amendment Nos.198 and 198

TS 3.1-23a r

(3)

Two Power Operated Relief Valves (PORV) shall be

. OPERABLE with a lift setting of s 385 psig, or (a)

When the PORVs are providing the overpressure protection verify the block valves are open at least every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

(4)

A bubble in the pressurizer with a maximum pressurizer narrow range level of 33% shall be maintained. After the initial 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , two PORVs must also be OPERABLE, or (5)

The RCS shall be vented through one opened PORV or an equivalent size opening.

(a) With the RCS vented through an unlocked vent path verify the path open at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(b) With the RCS vented through a locked vent path verify the path open at least once every 31 days.

2.

The requirements of Specification 3.1.G.1.b (3) may be modified as l follows:

a.

One PORV may be inoperable in INTERMEDIATE SHUTDOWN with the RCS average temperature > 200 F but < 350 F for a period not to exceed 7 days. If the inoperable PORV is not restored to OPERABLE status within 7 days, then completely depressurize the ROS and vent through one open PORV or an equivalent size opening within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

One PORV may be inoperable in COLD SHUTDOWN or REFUELING SHUTDOWN with the reactor vessel head bolted for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the inoperable PORV is not restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months then completely depressurize the RCS and vent through one open PORV or an equivalent size opening within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Amendment Nos.198 and 198

i TS 3.1-24 T

l c.

With both PORV's inoperable, depressurize the RCS within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unless Specification 3.1.G.1.b.(4) is in effect. When the RCS has beer. depressurized, vent the RCS through one open PORV or an equivalent sized opening, or establish the conditions listed below. Maintain the RCS depressurized until both PORV's have been restored to OPERABLE status.

l

.(1)

A maximum pressurizer narrow range level of 33%.

(2)

The series RHR inlet valves open and their respective breakers locked open or an alternate letdown path OPERABLE.

(3)

A maximum of one charging pump is capable of injecting into the RCS.

(4)

Safety injection accumulator discharge valves closed and i

their respective breakers locked open.

d.

When the conditions noted in 3.1.G.2.c.(1) through 3.1.G.2.c.(4) above are required to be established, verify the required conditions are met at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.

In the event that the Reactor Coolant System Overpressure Mitigating

[

System is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.6 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the mitigating system or the administrative controls on the transient and any corrective actions necessary to prevent recurrence.

Basis The operability of two PORV's or the RCS vented through an opened PORV ensures that the Reactor Vessel will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when ti.e Reactor Coolant System average temperature is s 350 F and the Reacter Vessel Head is bolted.

When the Reactor Coolant System average iamperature is >350 F, overpressure protection is provided by a bubble in the pressurizer and/or pressurizer safety valves.

A single PORV has adequate relieving Amendment Nos.198 and 198 i

TS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW Aeolicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Soecification A.

Calibration, testing, and checking of instrumentation channels and interlocks shall be performed as detailed in Tables 4.1-1, 4.1-1 A, and 4.1-2.

B.

Eqqipment tests shall be performed as detailed in Table 4.1-2.A and as deiailed below.

1.

In addition to the requirements of 4.0.5, each Pressurizer PORV and block valve shall be demonstrated CPERABLF. by:

a.

Performing a complete cycle of each PORV with the reactor coolant average temperature >350 F at least once per refueling cycle.

b.

Performing a complete cycle of the solenoid air control valve and check valves on the air accumulators in the PORV control system at least once per refueling cycle.

c.

Operating each block valve through one complete cycle of travel at least once per 92 days. This surveillance is not required if the block valve is closed in accordance with 3.1.6.a. b, or c.

Amendment Nos.198 and 198

TS 4.1-la i

2.

The pressurizer water volume shall be determined to be within its I limit as defined in Specification 2.3.A.3.a at least once per 12-hours whenever the reactor is not subcritical by at least 1% Ak/k.

3.

Each Reactor Vessel Head vent path remote operating isolation l valve not required to be closed by Specification '3.1.A.7a or

'3.1.A.7b shall be demonstrated OPERABLE at each COLD SHUTDOWN but not more often than once per 92 days by operating the valve through one complete cycle of full travel from the control room.

l 4.

Each Reactor Vessel Head vent path shall be demonstrated OPERABLE following each refueling by:

a.

Verifying the manual isolation valves in each vent path are locked in the open por,ition.

b.

Cycling each remote operating isolation valve through at least one complete cycle of full travel from the control room, i

c.

Verifying flow through the reactor vessel head vent system vent paths.

C.

Sampling tests shall be conducted as detailed in Table 4.1-2B.

D.

Whenever containment integrity is not required, only the asterisked items i

in Table 4.1-1 and 4.1-2A and 4.1-2B are applicable.

E.

Flushing of wetted sensitized statinless steel pipe sections as identified in the Basis Section shall be conducted only if the RWST Water Chemistry exceeds 0.15 PPM chlorides and/or florides (CL and or F-).

j Flushing of sensitized stainless steel pipe sections shall be conducted as l detailed in TS Table 4.1-3A and 4.1-38.

Amendment Nos.198 and 198

~

TABLE 4.1-2A (CONTINUED)

MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE 14a. Service Water System Valves in Functional Each Refueling Shutdown 9.9 Uno Supplying Recirculation Spray Heat Exchangers b. Service Water System Valves Functional Each Refueling Shutdown 9.9 Isolating Flow to Non-essential loads on intake Canal Low Level Isobtion 15.

Control Room Ventilation System

Ability to n.aintain positive Each Refueling Shutdown 9.13 pressure for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using a volume of air equivalent to orless than stored in the bottled air supply 16.

Reactor Vessel Overpressure Functional & Setpoint Prior to decreasing RCS 4.3 Mitigating System (except temperature below 350'F backup air supply) and monthly while the RCS is <350*F and the Reactor Vessel Head is bo!!ed CHANNEL CAllBRATION Each Refueling Shutdown l

E 17.

Reactor Vessel Overpressure Setpoint Each Refueling Shutdown 4.3 Mitigating System Backup y

Air Supply

<u 5

18 Power-Operated Relief Valve Functional, excluding valve Monthly 4.3 Control System actuation CHANNEL CAllBRATION Each Refueling Shutdown G.

y v

5 Cp O-8

TABLE 4.1-2A (CONTINUED)

MIN! MUM FREQUENUY FOR EQUlPMENT TESTS UFSAR SECTION DESCRIPTION TEST FREQUENCY REFEREMCE 19.

Prirnary Coolant System Functional 1.

Fedade lestage leseng(aW on eeth waive listed h l

15edication 3.1.C.7e shsE be oceanoplished prior to enterire POWER OPERATION efter overy thee the pierd is pdeced in COLD 94UTDOWN for sufueling, ator each One lhe plant is placedin 00LD 91U1DOMI for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Rtenswg has not been % Insie precedng a monha, and oster to nesumho m valve io sawtes sem e

maintenance, supsir orW work in i "--. - :1 20.

Containment Purge MOV Leakage Functional Semi-Annual (Unit at power or shutdown) l if purge valves are operated during interval (c) 21.

Containment Hydrogen Analyzers a.

CHANNEL FUNCTIONALTEST Once per 92 days l

b. CHANNELCAllBRATION Once per 18 months 1.

Sarnple gas used:

One volume percent (10.25%) hydrogen, balance nitrogen Four voleme percent (10.25%) hydrogen, belance nitrogen 2.

CHANNEL CAllBRATION willinclude startup and operation of the Heat Tracing System 22.

RCS Flow Flow 2 273,000 gpm Once per refuelin;; cycle 14 l

k 23.

RWST Parameters

a. Temperature s 45'F Once per shift l

4

b. Volume 2 387,100 geilons Once per shift g

(a)

To satisfy ALARA requirements, leakage may be measured indrectly (as from the periormance of pressure indicators)if accomplished in T

accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

H ag (b)

Minimum differential test pressure shall not be below 150 poid.

(c)

Refer to Section 1.4 for acceptance criteria.

o, i

See Specification 4.1.D 4

m c6

--

ML20082V272

Text

Attachment:

Attachment:

Navigation menu

Search