W3P84-0577, Forwards Justification for Exemption from Type C Leak Testing,List of Isolation Valves within Essential Sys & Penetration Isometric Drawings Re Finalization of Tech Specs in Preparation for OL

Forwards Justification for Exemption from Type C Leak Testing,List of Isolation Valves within Essential Sys & Penetration Isometric Drawings Re Finalization of Tech Specs in Preparation for OL

ML20087H447
Person / Time
Site:	Waterford
Issue date:	03/16/1984
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-0787, RTR-NUREG-737, RTR-NUREG-787, TASK-3.D.1.1, TASK-TM W3P84-0577, W3P84-577, NUDOCS 8403210030
Download: ML20087H447 (17)

Text

. s LOUISIANA u2 onAnoNos sineer P O W E R & L I G H T! R O. BOX 6008

• NEW ORLEANS. LOUISIANA 70174

• (504) 366-2345 UTiuSN SYSSE March 16, 1984 W3P84-0577 Q-3-A29.20 Director of Nuclear Reactor Regulation Attention: Mr. G.W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

r Waterford SES Unit 3

? Docket No. 50-382

) Technical Specifications: 10CFR50

. -_ Appendix J Leak Rate Testing and Containment Isolation Valves ATTACHMENTS: (1) .Tustification for exemption from Type C Testing

'(2) List of isolation valves within essential system (3) Penetration Isometric Drawings

Dear Mr. Knighton:

Louisiana Power and Light is currently finalizing and making plant-specific the Technical Specifications in preparation for an operating license. The development and review process has generally proceeded in a very productive manner, however we have reached an impasse on three critical issues under Containment Systems Branch cognizance. These issues potentially impact the fuel loading schedule and subsequent power operation of Waterford 3 if-not satisfactorily resolved.

TYPE C TESTING l

The first issue concerns Type C leakage testing pursuant to 10CFR50, Appendix J.

During the Technical Specification Proof and Review phase, the Containment Systems Branch (CSB) indicated that penetrations subject to Appendix J Type 2 and C testing should be explicitly identified in the Technical Specifications.

Consequently, LP&L proposed an update to the Technical Specifications that added applicable penetraticus for Type B and C testing as previously identified in the FSAR. However, the CM has indicated that additional penetracions should be subject to Type C testing. There are two points which we believe are relevant to this issue.

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W3P84-4577 lQ-3-A29.20 First,~apparentlyitheLCSB is maintainingLa position ~ requiring'nine (9) additional

penetrations to be Type C tested. 'None of these penetrations are required to be AType'C tested in accordance with the definition of " Type C Tests" included in 1 Appendix J of 10CFR Part 50. In addition, all of.these penetrations are normally

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,-water. filled land most are in operation under post _ accident-conditions. In order to postulate'leakagelfrom the containment atmosphere to the outside atmosphe *,

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multipleLfailures or single 1f ailures plus leakage through multiple barrierr

~have;to be' assumed in addition to the postulated LOCA. The penetrations identified in the FSAR for Type C testing adequately ensure that post-accident

^

containment-leakage is less than~that assumad in the1 accident analyses. However, 7 :notwithstanding LP&L's firm-position that these penetrations.do not fit the

[ feriteria of' Appendix J or. applicable NRC guidelines'for requiring leak tests, we

. have; performed calculations which demonstrate that, even in the event of any

' single-active-failure, a water seal will be.present on each penetration thereby preventing montainment leakage. A discussion of the results of these

.: calculations is provided'in Enclosure.l.

^

Second,[ithad~beenLP&L'sunderstandingthatthetestingrequirementsof

Appendix J of-10 CFR Part.50 had been satisfactorily resolved during the FSAR

review. Evidence of this resolution is contained in the Waterford-3 Safety

Evaluation Report.(NUREG-0787) Subsection 6.2.6. LP&L does recognize.that the determination of Containment bypass leakage paths remained open to be resolved as Ep art of the-Technical Specification review process as stated in SER Subsection F >

6 '. 2. 3. .However final determination of bypass leakage-paths should not have 3

. . identified additional penetrations for Type C testing since bypass leakage

, -penetrations are:a subset of Type B and'C penetrations.

!l TThe:nineL(9)-~ penetrations at. issue are not designed with provisions to conduct leak rate testing in accordance with Appendix J. EIt is extremely late in the

- ' licensing-~ process for/the Containment Systems: Branch-to reopen design questions

that)had previously been accepted, and.it is inappropriate that the Technical Specification, process be used as a mechanism for reinterpreting leak testing regulations.- Being that'the CSB's concerns =do not.seem plant-specific to Waterford-3,-LP&L believes'that.these concerns, including their inconsistency with; Appendix J, may be more appropriately addressed in accordance with the

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Commission's Backfit' Policy.

Technical Specification 3/4.6.3 and General Design Criteria (GDC) 57 Systems 1

^ '

-The second issue of' concern to LP&L is the Containment Isolation Valve Technical TSpecification. 'LP&L's concern deals with the applicability of the Limiting y -Condition;forTperatica (LCO) toward1GDC 57 systems (closed systems penetrating

< containment)-and the appropriateness of including. isolation valves within "

essential systens under the jurisdiction of this-Technical' Specification.

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W3P84-0577L Q-3-A29.20l

, - i g .. ;LP&L's positioniis thatjthe LCO/ Action-requirement;of this specification is inappropriate'for penetrations designed.in.accordance with General Design J

- @ Criteria (GDC) 57 of-Appendix.A of 10CFR Part 30. CDC 57 penetrations are

'neitherLpartiof the 3 reactor coolant pressure-boundary nor open to containment fatmosphere / but are connected to a_ closed-seismic Category I system inside

. / containment

• and;are- provided with a containment -isolation . valve outside the -

" containment.--Hence, the two operable isolation barriers are an active isolation

,? . valve _and.'a passive closed piping. system. LP&L's concern with the Action

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,  ; requirements of. Specification 3.6.3 ic that an inoperable-isolation valve in a

_GDC 57 penetration'would require immediate action to initiate plant shutdown

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kbecause:the action statements do not take. credit for the second isolation

. barrier, i.e., closed systems inside containment. However, an inoperable

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isolation barrier (valve)-in a typical GDC 55 and-56 penetration.is allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before initiation of-plant shutdown-is required. This allows expected and funexpected maintenance. activities to be performed on~ containment isolation

-systems.; Prior to revision 3.(Fall /1981) of the Standard Technical

-Specificatione,.the bases for.this specification never explicitly included GDC

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'570 It is LP&L's position that Specification 3.6.3 should not decrease potential 1 U plant availability without a corresponding 1nerease in safety. Consequently, it  !

is requested

that Specification 3.6.3 be revised to allow credit for the

. . containment isolation boundary intrinsic 'to a GDC 57-penetration (i.e., a closed ,

[ e s'ystem), or the~ valves in these penetrations be deleted from the specification.

^

Technical Specification-3/4.6.3 and Essential Systems

" An additional concern with Technical Specification'3/4.6.3 is the inclusion into

.this specification of valves in essential systems. Placing these valves into_.

[ 'this Technical Specification can degrade the overall safety of_the plant as well

.as limit operations. . Inclusion of essential system valves forces the plant to

~

' isolate these' penetrations within four hours if a' valve is inoperable in' order to

. seet'the. action statements of Specification 3.6.3. Isolation of these valves is I

-not'the required post-accident position nor the position of' greater importatce to plant safety. 'These: penetrations are~ designed to be open and operating post-LOCA and usually. get ESFAS Signals (SIAS, CSAS, EFAS) to open in the event of an :

accident.1.The operability of these. valves / systems is ensured by other means

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, -including: separate Technical Specifications for each system (Safety Injection, Containment' Spray, etc.); the provisions and requirements of ASME.Section XI; and j .

jactuation during ESFAS subgroup relay testing (Technical Specification Table

.423-2). ; Inclusion of these valves does not seem to be in the best interest'of-

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plant safety,or operation and is inconsistent with the requirements placed on .

. - previously ~1icensed plants. ' LP&L's position. is that the operability of the essential system containment isolation valves is ensured by other requirements -

, :and thatfthecievel of safety.being verified by the Technical Specifications should be consistent among all the Engineered Safety Feature Systems. Enclosure

-(2).-11sts those isolation valves that LP&L feels are better validated by other  !

TechnicalLSpecifications'and requirements.

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Your assistance.in ensuring that these matters receive the appropriatc high level reviews will be greatly appreciated. 'Should you have any questions or comments, please do not hesitate to contact us. We are prepared to meet with the NRC staff to respond in detail to any. questions.

Very truly-yours,

/

K. W. Cook Nuclear Support & Licensing Manager KWC/RMF/W/ch Attachments

- cc: W.M. Stevenson, E.L. Blake, D.M. Crutchfield, W. Butler, J. Wilson, D.

Hoffman, J. Huang, G.L. Constable i

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~ Enclosure (1)

. Justification for Exemption from' Type C Testing i The_following penetrations have not been resolved:

N '1.- l#23, #24 CCW for Reactor Coolant Pumps and CEDM Coolers

i. ._ ? 2. #27_CVCS Charging Line

3. +#34,:#35 Containment. Spray (4.- .#40,'#41: Shutdown Cooling

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J5. #69,,#70 Hot Leg Injection For penetrations 23 and 24, the Containment Systems Branch at a meecing on l1/17/84 indicated-that an exemption from leak. testing until the first refueling f  ?

- (could be granted due-to lack of testing provisions. This is acceptable at present-to LP&L, however further effort is planned to support a permanent

exemption and to fully address CSB_ concerns.

In regard.to thetother' penetrations, Appendix J to 10CFR Part 50 states:

. " Type ,C Tests'? means - tests intended to neasure containment isolation valve x.

-- .leakageirates. The containment isolation valves included are those that:

1. -Provide a direct connection between the inside and *.he outside atmosphere of the primary reactor containment under normal

-_ operation, such as' purge and ventilation, vacuum relief, and instrument valves;-

2. -Are required to close' automatically upon receipt of a containment

-isolation signal;in response to controls intended to effect containment-isolation;

'3. Are required to operate intermittently under post accident conditions; and

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_4 . .Are in main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling vater power

= reactors.

_.The NRC staff conclusions reached in Section 6.2.6 (Containment Leakage Testing Program) in'the Waterford 3 SER (NUREG-0787_ dated 7/81) state:

~

"The. proposed reactor containment leakage-test probram complies with the

. requirements of~ Appendix ~J to 10 CFR Part 50. Such compliance provides adequate assurance that containment leak-tight integrity can be verified

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.  : periodically throughout . service lifetime _ on a timely basis' to maintain such J fleakage.within the limits of the technical speciffcations.

Maintaining containment leakage rates within such limits provides reasonable

assurat.celthat,'in the event of any radioactivity releases within the containrent, the. loss of the containment atmosphere through the leak paths will not be in excess of acceptable limits specified for the' site.

- Compliance _with the requirements of Appendix J constitutes an acceptable basis for satisfying thel requirements of GDC 52, 53, and 54."

m None of.the disputed valves fit any of the above categories and therefore should

'not require testing. In addition, all of these penetrations are normally water filled and most are in operation under post accident conditions. In order to postulate leakage from the containment atmosphere to the cutside atmosphere, multiple failures or single failures plus leakage through multiple barriers against a water seal would have to be assumed in addition to the LOCA. This enclosure contains-a discussion of each penetratien in question.

Penetrations 34, 35, 40, 41, 69, and 70 form closed, seismic Category I, Safety Class 2 Systems outside containment, have design temperature and pressure greater than the post-LOCA containment environment, and are protected from the effects of pipe rupture and missiles. These systems are maintained full of water during normal plant operation and due to their Post Accident operation are

-subject to a system leak reduction program including periodic leak testing (with fluid) in accordance with Technical Specification 6.8.4.a and NUREG 0737 Item III.D.1.1. These leak tests are-performed at system operating pressures which are much greater than the containment post accident pressure; therefore, leakage due to containment pressure for postulated system failures will be minimal. The

~ discovery of any significant leakage identified during testing requires appropriate maintenance to reduce the leak rate. Leakage from these systems occurs within the Controlled Ventilation Area and the worst-case effects are accounted for in the annlysis of LOCA Dose Consequences (See SER 15.4.7).

~

.In addition to the above arguments, the following provides specific details (penetration isometric drawings are included as Attachment (3)) for each penetration supporting LP&L's position that post LOCA containment leakage paths are not credible for these penetrations and that it is unnecessary to perform Type C tests.

Penetration 27 CVCS Charging Line The charging system receives an automatic SIAS signal to start two charging pumps, align the system to take a suction on the Boric Acid Makeup Tanks and inject borated water into the Reactor Coolant System. Flow through this penetration is guaranteed in this line under post-LOCA conditions, and credit is taken for flow in the small break LOCA analysis. Technical Specification 4.1.2.2.b verifies the flowpath at least every 31 days.

During the subsequent post-LOCA period, the sater in the Boric Acid Makeup Tanks may be depleted. When this happens, charging will be secured and the charging isolation valves will be shut. For most RCS break locations, the charging line will not be exposed to the containment atmosphere, since flow from the Safety Injection system keeps the piping covered with fluid. In the event the break was in the RCS Cold Leg at a position which caused the charging line to be exposed to the containment atmosphere, there are still several barriers available. Inside containment there are 2 valves in series in each charging path and outside containmcnt there are check valves on each charging pump discharge in addition to the containment isolation valve. The charging pumps are positive displacement pumps further minimizing any backflow in the system. All of the piping is of high pressure design, and the valves are subject to ASME section XI testing.

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_ _ JA conservh lve calculation was performed assuming' seat leakage of the Loutboard isolation valve,ineglecting the resistance of the other valves and the'.

f l positive? displacement pumps. . The'outside' isolation valve is a gate valve, thus

~ m f,only_seatileakage has the' potential to;become stem leakage.. The result shows

'q Jthatfa water barrier can be maintained on the isolation valve for greater than 30.

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-days even iffone and one-half times.the valve's design leakage is assumed. .This k calculation and the design leakage specification are based _on_ system design pressure which is_much higher than the relatively low Containment Post Accident

. Pressure.1 <

LP&Lifinds 'it highly tmlikely that a credible leakage path from the -

1. < ' Containment--stmosphere'to'the outside atmosphere exists in this penetration.

Penetrations 34 and 35, Containment Spray

The' Containment Spray System is an ESF system that is required to be in Loperation following a LOCA. During normal' plant operation the system is

-- maintained full of. water from.the RWSP through the pump discharge-header (riser) up to a level of-149.5 ft._MSL (See Technical Specification 3/4 6.2). After the LOCA,tthe system is automatically placed in-operation, supplying water from the RWSP or the Safety. Injection System Suinp inside containment. The outer isolation

.1 valve'is located at'the system lowpoint-(-32 ft. MSL) and thus water is.

maintained around the valve even if..the Containment Spray pump fails to start.

.That is, a head of water willibe applied against_this_ valve from the RWSP (minimum water (level: -2 ft., MSL) and/or the Safety Injection _ Sump (bottom at -16

'ft.:MSL plus post-accident containment pressure) 'due.to their higher elevation,

, , ' Jirrespective of what single-failures are applied. .In-order for leakage to occur La check valve inside the containment and an-isolation valve outside containment

-would have to leak:through the seats under the small differential pressure across cthe valve'. A leakage calculation was performed, which neglected the check valve j . completely. .The_outside containment isolation valve is a gate valve, thus only_

O iseat leakage has.the' potential'to.become valve" stem leakage. The calculation

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. _ conservatively assumed all seat leakage'as stem leakage. The results show that a

<- water barrier can be maintained for greater than 30 days even if 45 times the design seat leakage is' assumed..

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=LP&L fi s'it highly unlikely that leakage can occur through these f

penetrations.

7 Penetrations 40 and 41, Shutdown Cooling The'SDC lines connect the RCS hot legs to the LPSI Pump Suction through-three normally closed isolation valves. Two of these valves are interlocked'such that they_cannot-be opened until the RCS prescure is reduced below the SDC entry fpressure'(377'psig).' Inside the Reactor Auxiliary Building, this closed system ipiping' descends'to the -35 ft. MSL' elevation where it connects to the LPSI Pump

' Suction. This system lowpoint will'be maintained under a head of water due to

the higher elevation of the RWSP and/or the Safety Injection Sump. The SI Sump,

'due'to(the water level-in the sump and containment atmospheric pressure will i-

"E always exert a higher pressure on the RAB side of the low point piping than the l' piping from containment will if. exposed to the containment atmosphere. Should

the SI Sump' isolation valve fail to open, the RWSP at the minimum level achieved

'p . upon RAS :(-2 f t. MSL) will exert approximately 13 psi on this piping. A higher

. pressure (44 psi peak) can. exist-inside the Containment, but this is reduced to m

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,In. order for:leakageito' occur,:these three11n-series valves must leak by the-

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seats.E The-containment: isolation valves are gate valves. thus~only seat leakage-hasithe potential to becoce valve stem: leakage. :The two valves inside (Containment are!RCS boundary' valves ~and tested for leakage in accordance with 1 afechnical: Specification 4.4.5.2.'2 at: 2250 ' psia. - A calculation was performed-

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isesuming.' seat leakage 3of the outboard-isolation valvei neglecting the isolation

. - provided by'theitwo inaseries RCPB valves inside Containment. The:results show

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~ that f a water. barrier can be maintained on' the valve for greater than-30 days even

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fif~9ctimes-the design leakage occurred.; Both the calculation and the design-w leakage;specificationsiare based on system' design pressure which is much greater than the~relatively low Containment Post Accident-Pressure.

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. -LP&L^ finds--it highly unlikely that post accident containment atmosphere

,eakage will occur in this penetration.

4

- Penetrations 69 and'70 Hot Leg Injection.

'. The Hot Leg Injection headers are an integral part of the ECCS system and E ?will be placed ~in operation post-LOCA. . Prior to initiation of hot leg injection.

- the~ isolation valves-outside: containment are closed with HPSI pump discharge

, pressure applied against the outermost-valve. Even if a HPSI pump failed during ,

-post-accident' conditions, pressure isEstill applied to the'RAB side of the

. outermost isolation valve due to the head-of water from the RWSP and/or the' '

= Safety Injection Sump.

Due to the dual considerations of the' water level in the LST' sump in addition to the ambient containment pressure, a higher pressure will always exist on' the RAB side'of the isolation valve'than on the containment side,

~ f thus ensuring- that any: leakage will be into cuatainment, and that in any event a

'waterzseal'is maintained-on the containment side of the valve. Should the sump ,

-isolatien . valve -fail to lopen (a second single-active-failure), the minimum RWSP

. tievel!will still? exert at_least an approximate 13 psi-against the outermost

, ' isolation valve.. Although a higher ambient pressure may exist for a short time

. inside' containment, 2 check valves -(RCS boundry valves)' and the isolation valve

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outside containment:all present-barriers against leakage. The check valves are tested for leakage (intersystem.LOCA bases; Technical. Specification 4.4.5.2.2) at 2250 psia.ifar. greater than the small post-accident differential pressure which may. exist across the isolation valve (less:than approximately 31 psi).

'A calculation was performed assuming seat leakage and stem leakage of the

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. outboard isolation ~ valve and neglecting the isolation provided by the series

[ check valves completely.. .The<results show that a water barrier can be maintained

~for, greater than'30' days even if double-the design leakage occurred. This l.

calculation included stem leakage'that will-be detected and corrected under the .

-leak! reduction program.- Both the~ calculation and the design leakage R

specifications are based on system' design pressure which is much higher than the relativelyjlow containment post-accident pressure.

o, 1LP&L' finds;it' highly unlikely;that a credible leakage path from the l containment' atmosphere to the.outside atmosphere exists in these penetrations.

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Proposed Valves to be Deleted from-Technical Specification Table 3.6-2*

'1. Containment Isolation Acceptable as is

2. Containment PurFe ,

. Acceptable as is

3. -Safety Injection Actuation Signal (SIAS)**

Penetration 26 - CVCS-Letdown (Already listed under (CIAS))

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Penetration 32 and 33 .SI Sump Isolation

4. Main Steam Isolation Signal (MSIS)**

. Penetration 32 and 33 - Emergency Isolation

5. . Manual / Remote Manual Essential Systems:

a.; . Penetration 1 and 2 - 2MS ~ V611A, 2MS - PM629A, 2MS - V612B, 2MS - PM630B

b. Penetration 15 through 22 - CCW to Containment Fan Coolers

c. Penetration 27 C7CS - Charging

d. -Penetration 34 and 35 - Containment Spray

e. Penetrations 36 through 39 - LPSI

f. Penetrations 40 and 41 - Shu:down Cooling

g. . Penetrations 55 through 58 - HPSI

h. Penetrations-69 and'70 - Hot Leg injection 6.- Other a.- Essential Systems: Penetrations 27, 34, 35, 36-39, 55-58, 69 and 70

b. Penetration 1 and 2. These valves are not required for containment f- isolation in accordance with CDC 57 (See revised FSAR Table 6.2-32, and 33).

l- . . .

'* This list is broken down 'into Subsections as shown in the February 16, 1984

Draft version _of Technical Specifications.

• • . LP&L's position is that these Subsections should not be included in this

-Technical Specification.

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I m , . . _ - . _ ,_ . . _ , - . , . _ _ _ . . _ . _ . . - - . , . . _ _ , _ . , _ , . _ _ _ .