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Forwards List of Pressure Isolation Valves for Each Lower Pressure Sys Attached to Rcs,Per Generic Ltr 87-06. Operational Limitations Re Valve Periodic Testing & Notes Concerning Allowable Leakage Limits Also Encl
	ML20215F273
Person / Time
Site:	Ginna
Issue date:	06/11/1987
From:	Kober R; ROCHESTER GAS & ELECTRIC CORP.
To:	Stahle C; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0821, RTR-NUREG-821, TASK-03-05.A, TASK-05-02, TASK-05-10.B, TASK-06-04, TASK-3-5.A, TASK-5-10.B, TASK-5-2, TASK-6-4, TASK-RR GL-87-06, GL-87-6, NUDOCS 8706220310
	Download: ML20215F273 (22)
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' ROGER W KODER Tt t E PsoNr VKE rmesentwr -

ELECSM FHODUCTION~

Awt A RODE Fla 5 4 6-2 700

. June 11, 1987-g U.S.- Nuclear Reg' la tory Commission u

Document Control Desk

' Attn:

Mr. Carl Stahle PWR Project Directorate No.

1-

. Washington, D.C.,

20555 l

Subject:

Periodic Verification of Leak Tight Integrity of

. Pressure Isolation Valves (Generic Letter 87-06)

R.

E.

Ginna Nuclear Power' Plant Docket No. 50-244

Dear Mr. Stahle:

The subject Generic Letter 87-06 dated March 13, 1987 requested within 90 days a list of all pressure isolation valves, and, for each valve, the periodic tests or other measures

-performed to assure the integrity of the valve as an independent barrier at the reactor coolant pressure boundary.

Also requested was. acceptance criteria for leakage, if any, operational limits, if any, and frequency.of test performance.

Accordingly, we are providing a tabulation of the pressure isolation valves for each lower pressure system ' attached to the reactor' coolant sys tem, along with the other inf ormation requested.

The tabulation is presented as Attachment 1.

Operational limitations related to-valve periodic testing are presented in Attachraent 2.

Additional remarks and information dealing with the individual system in'which the valve is located and other assurances demonstrating system integrity are presented in Attachment 3.

Notes regarding allowable leakage limits, when applicable, are found.in Attachment 4, with pertinent abbrevi.

'ations and definitions in Attachment 5.

l The pressure isolation valves tabula ted on Attachment 1 are defined for each interface as the two valves in series within the q

reactor coolant pressure boundary which separate the high pressure reactor coolant system from an attached lower pressure system.

With the exception of the PORY block valves which are normally open and are Category B valves, the pressure isolation valves identified are either closed during normal operation or, if open, are designed for automatic closure upon receipt of a containment isolation signal. identifies each valve's position 0

during normal operation.

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8706220310 870611 PDR ADOCK 05000244 -

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PDR

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Certain valves at the Ginna plant were identified as being i

l Event V configurations.

Pressure isolation valves in the High Pressure and Low Pressure Sa fety Injection System were evaluated f

in a. modification order (Eve.nt V order) dated April 20, 1981.

Subsequently, the Ginna Technical Specifications were. amended to q

implement these test requirements.

i The testing requirements for PIVs that perform a safety related function are covered under the R.E..Ginna Inservice Valve Testing Program.

This test program was developed in accordance j

with the requirements of Article IWV of Section XI of the ASME Code revision specified for the plant in Appendix C of the Quality Assurance Manual.

Identification of the requirements for pressure isolation valves was the subject of correspondence for several topics previously evaluated under the Systematic Evaluation Program (SEP) for R.E.

Ginna.. Requirements for PIVs were examined under SEP Topic V-ll.A, Isolation of High and Low Pressure Systems, Topic V-10.B, RHR Reliability, Topic VI-4, Containment Isolation, and

-Topic III-5.A, Postulated Failure of Piping Inside Containment, j

Assurances demonstrating the integrity of the pressure boundary of high-low pressure systems related to PIVs are summarized in the

-I information in Attachments 2 and 3 and were acceptably resolved under these topics.

References to correspondence for each topic evaluated can be found in the Integrated Plant Safety Assessment Final Report, NUREG-0821 dated December 1982 and Supplement No. 1 to NUREG-0821 dated August 1983.

V truly yours, ag e. b %

Roger W.

Kober Attachments I

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l ATTACHMENT 2 OPERATIONAL LIMITS Note 3 - Valves MV 204A, AOV 371 and AOV 427 cannot be stroked during normal plant operation on a quarterly basis because this test would interrupt the letdown (CVCS) system.

These valves will be stroked at cold shutdowns and refueling outages.

The letdown valves downstream of the pressure breakdown

{

orifices, AOV 200A, AOV 200B, and AOV 202 are not part ot the Inservice Valve Testing Program, however, are stroked during cold shutdowns and refueling outages as part of the low pressurizer level interlock test.

Note 8 - Valves 700 and 721 cannot be stroked during normal plant operation on a quarterly basis because there is an inter-lock system which prevents these valves from opening when 4

the primary system is at operating pressure.

Valves 700, 701, 720 and 721 separate a high pressure system from a low pressure system.

These valves will be stroked at cold-shutdowns and refueling outages.

Note 10 - Valves 853A and 853B cannot be stroked during normal plant I

operation on a quarterly basis because this test requires pressurizing the RHR system to the primary system operation pressure.

These valves will be stroked at cold shutdown and refueling outages.

Leakage testing of check valves 853A and 853B shall be accomplished prior to criticality following (1) refueling, (2) cold shutdown, and (3) maintenance, repair or replace-ment work on the valves.

Leakage may be measured indirectly from the performance of pressure indicators, system volume measurements or by direct measurement.

Minimum test differential shall be greater than 150 psid.

Technical Specifica tion 4. 3. 3.4 defines the allowable leakage rates (Note (a) of Attachment 4).

Note 11 - Valves.. 877A, 877B, 878F, and 878H, are considered passive valves which are not required to change position to accom-plish their specific function.

Stroking these valves would serve no useful function and is therefore not done per IWV-3700-1.

a

1

+

ATTACHMEET 2 (Continu2d) Note 13 - Valves 867A, 867B, 878G, and 878J cannot be stroked during normal operation on a quarterly basis or at cold shutdown condition when the primary system is full.

This test may only be done when the plant is in a refueling shutdown condition with a partially full primary system in order to prevent an overpressurization.

Leakage testing of check valves 867A, 867B, 878G and 878J shall be accomplished prior to criticality, except for low power physics testing, following (1) refueling, (2) cold shutdown, and (3) main-tenance, repairs or replacement work on the valves.

Leakage may be measured indirectly from the performance of pressure indicators, system volume measurements or by direct measurement.

Check valves 878G and 878J shall also be tested for leakage following each satety injection flow test i

(monthly).

Minimum test dif ferential shall be greater than 150 paid.

Technical Specification 4.3.3.4 defines the allowable leakage ra tes.

Note 15 - Category C Relief Valves shall be tested in accordance with the extent and f requency requirements of Paragraph IWV-3510 of Article IWV of Section XI of the Code.

Note 16 - Valves FORV 430 and PORV 431C are the pressurizer power operated relief valves.

B~;ause of the potential to cause a plant transient in the event of a failure of the second in-line valve, these valves are not stroked quarterly.

Operability of these valves shall be verified as follows:

(a) Full stroke exercising during cooldown prior to achieving water solid condition in the pressurizer and during cold shutdown prior to heatup.

(b) Stroke timing to be performed as a minimum once each refueling cycle as a part of the channel calibration specified by Technical Specifications 4.16.lb.

(c) Fail safe actuation testing is permitted by the code to be performed at each cold shutdown it valve cannot be tested during power operation.

(d) Technical Specification 4.16.la and 4.16.10 delineate additional requirements for operability verification of the PORV actua tion channel and valve position.

Note 18 - Valves 852A and 852B cannot be stroked during normal plant opera tion as these valves, when cycled, could subject the Residual heat Removal System to a pcessure in excess of its design pressure if there is a failure of the in-line check valve.

These valves will be stroked at cold shutdown and refueling outages.

Note 23 - Valves MOV 515 and MOV 516 are stroked quarterly except if already closed or during cold and refueling shutdowns.

ATTACIMECT 2 (C ntinu d) 3-Note 28 - Check valves 842A and 842B (accumulator check valves) cannot and should not be exercised during plant ope ra tion.

Exercising of these valves requires the reactor coolant system pressure to be reduced to below accumulator pressure.

Therefore, testing of these valves will be performed after refueling and cold shutdowns, and after maintenance, repair or replacement.

Full stroke testing, which involves discharge of the accumulator through the valve to a partially-filled reactor coolant system will not be utilized since this test mode is considered impractical and unsate.

Valve operability from the normal closed position will be verified by partial stroking prior to leak testing with flow through the safety injection test line.

Leak testing will be performed to assure primary system l

integrity f ollowing each cold and ref ueling shutdown atter achieving normal reactor coolant system pressure and prior to reactor criticality.

Testing will be performed by either (1) closing each accumulator motor operated discharge valve, pressuring the line downstream of the check valves and measuring the upstream leakage, or (2) by measuring accumulator in-leakage by pressurizing the line downstream of the 842 valves.

These valves are not Event V check valves.

Therefore, the test acceptance criterion shall be that of Technical Specification 3.1.5 (Leakage),

which allows total RCS leakage of 10 gpm from known sources or 1 gpm from unidentified sources.

Note 29 - Valves 877A, 877B., 878F and 878H are currently not listed in our valve test program, however, a pair of these valves in each hot leg high head safety injection line (877A and 878F in loop B hot leg and 877B and 878H in loop A hot leg) together form one of the two pressure boundaries required to be tested by Technical Specifica tion 4. 3.3. 3.

Because these valves are normally closed and the piping contains motor operated valves (MOVs) which are also closed and de-energized, the check valves will not move with the possible exception of when the MOVs are required to be opened to i

test the check valves.

Thus, once tested the check valves

)

will remain closed.

An NRC order (Event V) dated April 20, 1981 established an appropriate test frequency for these

)

valves to be once every 40 months or after each opening of

{

the MOVs.

These valves are listed as passive valves in our valve testing program with testing required to meet Technical Specification 4.3.3.3.

Note 33 - Sampling valves 951 and 953 need not be stroked as part of this program as they are not required to perform a safety j

function.

Although considered to be part of the reactor coolant system boundary, 10CFR50.55a(c)(2) specifies that components which are connected to the reactor coolant system and are part of the boundary as defined in 50.2(v)

l ATTACH::ENT 2 (Csstinu:d) of this part need not meet the requirements of paragraph (c)(1) of this section, provided: In the event of postu-lated failure of the component during normal reactor ope ra tion, the reactor can be shut down and cooled down l

in an orderly manner, assuming makeup is provided by the reactor coolant makeup system.

The valves are exercised during routine sampling operations.

I 1 For consistency, these Note numbers correspond to the Note numbers identified in the RG&E Appendix C to the Quality Assurance Manual, Ginna Station Inservice Pump and Valve Testing Program.

o I

ATTACBMENT 3 REMARKS In addition to the remarks below, the air operated and motor operated valves identified in the tabulation are provided with control room position indication.

1. The testing of these valves was previously discussed and acceptably resolved in a modification order concerning Event V valve configurations f rom the NRC, D.M.

Crutchfield, to RG&E, J.E.

Maier, dated April 20, 1981.

They were also part of submittals concerning the required Inservice Valve Testing Program in a letter from NRC, D.M.

Crutchfield, to RG&B, J.E.

Maier dated May 26, 1981 and a response from RG&E, J.E.

Maier, to NRC, D.M.

Crutchfield dated August 26, 1981.

The testing program employed by RG&E was-determined acceptable as required by IWV-3400 of Section XI of the applicable edition of the ASME Code.

Testing of these check valves is also covered by Section 4.3.3 of R.E.

G1nna Technical Specitications.

2. These valves are exercised during leak testing of accumulator discharge check valves 842A and 842B and during leak testing of check valves 867A and 8678, SI discharge to loop B and A cold legs.

Downstream of AOV 839B and AOV 840B, valve 879 is a manual valve in the safety injection test line outside containment and is kept locked shut.

This valve is not required to change position to perform a safety function.

The only requirement is that leak-age through valve 879 be acceptably low.

Therefore, the quarterly stroke test has been deleted.

However, this passive valve is leak tested at ref ueling outages consistent with IWV-3700-1.

A relief valve, RV 887, is provided inside containment in the 3/4-1500#

test line downstream of valves AOV 839B and AOV 840B, and is exercised as required by Section XI Article IWV-3510 of the ASME Code.

3. MOVs 852A and 852B are Category B valves which should not be exercised during power operation.

These valves provide the second of two boundaries between the reactor coolant system (RCS) and the lower pressure residual heat removal (RHR) system.

The valves are normally closed to fulfill their pressure isolation function.

One of the two valves is required to open to mitigate large LOCAs

(> four inch diameter break).

Stroking 852A and 852B during power operation would mean that the boundary between the RCS and the RHR system will be formed by a single check valve.

Thus, stroke test-ing of 852A and 852B is appropriate only during cold and refueling shutdowns consistent with our current valve testing program.

J

ATTACHMINT 3 (Ctntinusd) i Valves 852A and 852B are provided with position indication in the control room.

Acceptably low leakage through each pair of valves is demonstrated during each system pressure test following refueling.

A relief valve, RV 203, capable of passing 200 gpm provides relief from the lower pressure RHR piping to the pres-surizer relief tank.

An increase in leakage through the valves during operation would be detected by discrepancies in makeup and letdown rates, by pressurizer relief tank indications, or by RHR system pressure and temperature indication.

Additional leak testing beyond that already performed should not be required.

4.

Valves MOV 700, 701 (suction to RHR pumps), 720 and 721 (discharge from RHR pumps), are normally closed and in the position to f ul-fill their safety function during power operation.

The valves are motor opera ted with position indication in the control room.

After closing these valves, power is removed at the breaker to prevent inadvertent operation.

Because of the sizing of the valve ope ra to r, it is physically impossible for the valves to be opened with a pressure differential across the valve of 500 psi or more.

Acceptably low leakage through each pair of valves (700, 701 and 720, 721) is demonstrated during each system pressure test fol-lowing refueling.

Both pairs of valves are closed during this system hydrotesting.

System hydrotesting during cold and re-fueling shutdowns consists of maintaining system pressure at 2235 psi for 10 minutes or longer to enable a visual inspection inside containment of specified areas.

A relief valve (RV 203) capable of passing 200 gpm provides relief from the lower pressure RHR piping to the pressurizer relief tank.

An increase in leakage through the valves during operation would be detected by discrep-ancies in makeup and letdown rates, by pressurizer relief tank level indications, or by RHR system pressure and temperature Indication.

Additional discussion on these valves concerning SEP Topic V-ll.A (Isola tion of High-Low Pressure Systems) is included in Mr.

D.M.

Crutchfield's letter to Mr. J.E. Maier dated July 22, 1981 and Mr.

Maier's letter to Mr. Crutchfield dated June 23, 1981.

Testing of these valves on an individual basis would require that test personnel work on these lines at times when the reactor coolant system temperature is greater than 350 F.

The valves are open and in service at lower temperatures.

These conditions would subject personnel to high temperature fluid hazards when connecting or disconnecting test equipment and would give them an unnecessary radiation exposure while working in the containment building near the primary loops.

The dose rate near these valves is approximately 100 mr/hr. during reactor outages.

Therefore, we have concluded that additional leak testing beyond that already performed should not be required.

I, ATTACIMENT 3'(Crntinuad)

" 5. An; orifice' on' each branch line upstream of AOV 200A, AOV 200B, and AOV 202 and downstream of AOV 427 reduces the pressure in the letdown line to a design pressure of 600 psi.

A. relief valve, RV ' 203 inside containment' and set at' 600 psi, can relieve in: excess (140 gpm) of the. combined flow.through the three' valves in the event that closure of downstream AOV.371 caused -a buildup of pressure in' this line.

Radiological consequences of'a failure in this line were examined under SEP Topic XV-16 and it was. determined that only a small fraction offthe 10CFR100 guidelines resulted.

As additional assurance of the integrity of.the letdown line, two downstream' valves, the letdown line containment isolation valve AOV 371 and the non-regenerative heat exchanger inlet manual. valve MV 204A (both inside-containment) are include'd in the Inservice Valve Testing Program and receive leak testing and stroke testing during cold and refueling shutdowns.

AOV 371 allowable leakage limit is 58.07 cc/ min.

The combined allowable leakage limit of manual valves MV 204A and MV 820 (in parallel with MV 204A) is 116.14 cc/ min.

AOV 200A, AOV 200B, and AOV 202 act as pressure boundary valves during reqired leak-testing of AOV 371.

The valves AOV-427, AOV 200A,.AOV 200B, AOV-202 and AOV 371.have control room valve position indication and close upon receipt of a containment isolation signal.

AOV 200B is openLduring' normal plant operation through the-letdown line.

AOV 200A and AOV 202, which are in parallel.with AOV 200B are normally closed.

Letdown line flow indication is also available in the control room.

6. Manual valve'311G provides the second valve interface between the excess letdown heat exchanger and the reactor coolant drain -

ta nk (RCDT)'inside containment.

The RCDT level is documented approximately every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months on the Auxiliary Building Log Sheet.

It also has control room level indication off the Plant Process Computer.

This vent and drain line does not lead outside containment.

Similarly, manual vent and drain valves MV 503 and MV 541 in a 2" line from the crossover leg loop "A"and MV 507-and MV 540 in a 2" line from the crossover leg of loop "B"

provide a path to the Reactor Coolant Drain Tank inside containment.

These valves are not included in the Inservice Valve Testing Program since.these are excluded by IWV-1300 of Section XI of the applicable revision of the ASME Code and are not required for accident mitigation.

These valves are subject to Technical Specification 3.1.5 which limits total RCS leakage to 10 gpm from known sources or 1 gpm from unidentified sources.

I

^

ATTACHMEZT 3 (c ntinu:d) l

7. HCV 123 in the excess letdown line is used to control pressure during system pressure tests prior to returning to power after cold and retueling shutdowns.

Both AOV 310 and HCV 123 are failed closed valves on loss of instrument air.

These valves are subject to Technical Specification 3.1.5 which limits total RCS leakage to 10 gpm from known sources or 1 gpm from uniden-tified sources.

Since HCV 123 is a demand valve, it does not have control room position indication.

However, the excess letdown line, which leads to the excess letdown heat exchanger, has temperature and pressure indication on the main control board.

The line is a 3/4 inch, 2500 lb. rated line which allows a maximum of approx-imately 10 gpm with both AOV 310 and HCV 123 open.

This is within the Technical Specification limit for leakage from known sources.

The excess letdown line connects with the seal return lines from the Reactor Coolant Pumps lA and 1B and leads back to the Volume Control Tank.

The seal return line isolation valve MOV 313 outside containment is stroke and leak tested as part of the Inservice Valve Testing Program during cold and ref ueling shutdowns.

Its allowable leakage limit is 87.11 cc/ min.

.Also, pressure and temperature indication in the seal return line is available which would indicate unacceptable leakage through the excess letdown line.

Because of the pressure breakdown in the lines leading to the Volume Control Tank, it is not subjected to high pressures or stresses and failure is unlikely.

How-ever, a rupture of the Volume Control Tank has been analyzed to define the limit of the exposure that could result from such an occurrence.

As documented in the FSAR, this event would present no undue hazard to the public health and safety and the integrated doses were determined to be well within the limits of 10CFR100.

8. Although included here for completeness, the discharge line of the CVCS is not classified as a low pressure system connected to the RCS since the piping is 2500 lb. rated piping throughout its length back to the positive displacement charging pumps.

RG&E has had no experience with failure of the positive displacement pumps to hold primary system pressure, nor would any failure be anticipated.

To verify that the charging line was not a valid " Event V" concern, RG&E calculated the PWR Check Valve Event Tree (Section 4.4 of WASH-1400) using the charging line config-uration of two check valves in series with a charging pump piston.

Very conservatively, assuming that both check valves were undetected open, and that the probability of the charging pump piston failure was equal to a check valve failure, the 1.4 x 10- ulated f or this configuration was determined to be Q sum calg/ year.

This value is low enough to be of no concern.

ATTACHZENT 3 (C:ntinuad) i In addition to these two check valves in series in each of the three charging lines, a fail closed air operated valve in each line can also be closed to provide system integrity against rupture.

Also, the charging line containment isolation check valve, CV 370B, upstream of the regenerative heat-exchanger located inside containment is leak and stroke tested during cold and refueling shutdowns as part of the Inservice Valve Testing Program.

(Quarterly testing would not be possible since this would interrupt the charging flow.)

The allowable leakage limit is 200 cc/ min.

]

9.

Like the charging line, the alternate charging line is a 2500 lb. rated line back to the charging pumps as described in 8, and therefore, is not classified as a low pressure system-connected to the RCS.

It is included here for completeness.

a The alternate charging line containment isolation (check)

)

valve, CV 383B, upstream of AOV 392B and located inside 1

containment, is a Class A valve and subjected to leak testing during refueling outages and stroke testing during cold and I

refueling shutdowns.

The allowable leakage limit is 170.33 cc/ min.

The alternate charging line is not normally used so

{

isolation valves are normally closed.

Valve CV 383B cannot be stroked during normal plant operation on a quarterly basis because this test would result in substantial radiation exposure to test personnel.

Surveys in the area of the test connection during plant operation indicate neutron fields of approximately 500 MR/hr. and gamma fields of 250 MR/hr.

Total whole body dosage to test personnel is estimated to be 375 mrem.

Also, AOV 3928 functions both as a relief valve as well as an isolation valve, with a setpoint of 250 psid.

10. Under SEP Topic III-5.A, Postulated Failure of Piping Inside Containment, RG&E provided analysis showing that the effects of failure of the charging line check valves would have no unacceptable consequences.

The effects of failure of these valves would result in consequences identical to a letdown line

]

break.

This line was resolved as documented in the Integrated Plant Safety Assessment Report, NUREG-0821 Supplement No.

1, Section 2.5 dated August 1983.

Certain instrumentation cables were modified (rerouted) as a result of the analysis.

In the i

evaluation of the alterna te charging line, it was determined that failure of both LPSI valves due to high energy line break i

and single active failure etfects, together with tailure of the

)

alternate charging line check valves 383A and 383B, would not j

result in any unacceptable consequences as noted in NRC Safety Evaluation Report for Integrated Plant Safety Assessment Report (IPSAR), Section 4.13, in a letter from D.M. Crutchfield to j

RG&E, J.E.

Maier dated June 28, 1983.

{

i

)

ATTACHMENT 3 (COntinusd) 11. Under SEP Topic VI-4, Containment Isola tion System, the charging and alternate charging lines were evaluated related to GDC 55 or 56 requirements.

Each line has a check valve inside containment that is leak tested; CV 370B in the charging line and CV 383B in the alterna te charging linu.

These lines do not have a pos t-accident function.

These lines were found acceptable because:

a) The piping system is designed to operate at 2250 psi, significantly above containment design pressure b) The piping is Seismic Category I c) The charging pumps are positive displacement pumps, and, therefore, leakage back through the pumps is expected to be minimal.

The above was documented in the Integrated Plant Safety Assessment Report for R.E.

Ginna Nuclear Power Plant, Section 4.22 forwarded with a letter to RG&E dated May 27, 1982.

12. The sampling lines are closed during normal operation.

The sampling line from loop "B"

hot leg in which AOV 955 (inside containment) and AOV 966C (outside containment) are located is opened for approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per day for sampling.

The AOVs j

may be closed by operator action from the control room or the remote operation station and close on a containment isolation j

signal.

The sample lines are 1/4 inch or 3/8 inch diameter

]

stainless steel tubing rated at 2500 psi.

The normal makeup system can maintain pressurizer level and pressure at 2250 psia j

up to a maximum break size of 3/8 inch.

Sample valves AOV 951 and AOV 953 need not be stroked as part of the Inservice Valve i

Testing Program as they are not required to perform a safety J

function.

Failure of AOVs 955 (hot leg loop "B"),

AOV 953 (pressurizer 1

liquid) or AOV 951 (pressurizer steam) inside containment is mitigated by automatic closure of diaphragm operated valves AOV 966C, AOV 966B or AOV 966A outside containment with containment isolation signal.

A relief valve (RV 257) in a 2 inch line on the Volume Control Tank is provided for the low pressure portions of the system to maintain system integrity in the event of operator error.

13. The sampling line from the loop "A"

hot leg is connected to the loop "B"

hot leg inside containment between AOV 955 and AOV 966C.

The "A"

loop line, however, contains three normally closed manually operated valves.

This line is 1/4 inch diame te r, therefore, flow from a break in this line can be made up by the makeup system.

l

R a.

1 ATTACBMENT 4 ALLOWABLE LEAKAGE LIMITS l

Mote'a) - Leakage rates less than or equal to 1.0 gpm are considered acceptable.

t

- Leakage rates greater than 1.0 gpm but less than or l

equal to 5.0 gpm are considered acceptable if the latest j

measured rate has not exceeded the rate determined by 1

-the previous test by an. amount that reduces the margin between measured leakage rate and the maximum permis-sible rate of 5.0 gpm. by 50% or greater, j

- Leakage : rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the-latest measured rate exceeded the rate determined by 1

the' previous test by an amount that reduces the margin between measured leakage rate and the maximum permis--

sible rate of 5.0 gpm by 50% or greater.

I

- Leakage rates greater than 5.0 gpm are considered unacceptable.

(From Technical Specifications Section 4.3.3.4)

Note b) -- System pressure of 750 psig is maintained for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months j

demonstrating system integrity Note c).- System pressure of 2235 psi is maintained for 10 minutes "or longer to enable a visual inspection.inside contain-ment of specified areas.

On a 10 year basis, system pressure is maintained for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while a visual inspection inside containment is performed.

l i

l ATTACHMENT 5 ABBREVIATIONS ~AND DEFINITIONS

_j i

l AOV Air Operated Valve l

CV Check Valve C/R Cold Shutdown and Ref ueling

-H Hydro Test HCV Hand Control Valve I

HPSI High. Pressure Safety Injection L

Leak Test j

M

. Monthly MOV Motor Operated Valve MV Manual Valve NA Not Applicable NC Normally Closed NO Normally Open PORV Power Operated-Relief Valve PRT Pressurizer Relief Tank

.PRV-Pressure Relief Valve Q

Qua rte rly R.

Refueling Outage RCDT Reactor Coolant' Drain Tank RHR Residual Heat Removal RV Relief Valve RX.,,

Reactor Vessel i

S' Stroke Test SI Safety Injection 3,

t VCT Volume Control Tank As Defined in IWV-2110 of Section XI of ASME Code:

CAT A VALVE - Valves for which seat leakage is limited to a I

specific maximum amount in the closed position for fulfillment of their function.

CAT B VALVE - Valves for which seat leakage in the closed position is~ inconsequential for fulfillment of their function.

CAT C VALVE - Valves which are self-actuating in response to some characteristic, such as pressure (relief valves) or flow direction (check valves).

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ML20215F273

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Subject:

Dear Mr. Stahle:

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