L-81-347, Forwards Description of Design & Results of Analyses of Reactor Coolant Gas Vent Sys,In Response to NUREG-0737,Item II.B.1.Development of Operating Procedures Postponed Until Design Approved
| ML17212A518 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/10/1981 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-TM L-81-347, NUDOCS 8108170268 | |
| Download: ML17212A518 (29) | |
Text
nEGULAT INFORMATION'ISTRIBUTIO>>Q>>STEMl (RIOS)
ACCESSION< NBR!8]08170268 DOC ~ DAlTEt.']/08/10 NOTARIZED:" NO FACIL>>:50; 335 St','ocke Plant'E Unit 1<, Flor ida Power L Light'os,'.
AUTH!.NAMEI AUTHOR AFF IL'IATION" UHRI6 E R ~ E!s Florida Power>>
L Light Co, REC IP ~ NAMEl REC IPJ ENT'F F IL>>I ATION" E>>ISENHUT'<D.G.
Division'f Licensing OOCKEIT-'"
05000335>>
SUBJECT:
- Forwards description of design L results of analyse's of-reactor coolant gas vent>> syscin response to NUREG~0737EItemI II>>,B',] ~ Devel opment>> of operating, procedur es postponed>> unti l design>> appr oved ~,
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FLORIDAPOWER & LIGHTCOMPANY August 10, 1981 L-81-347 Office of Nuclear Reactor Regulation Attention:
Mr. Darrell G. Eisenhut, Director Division of Licensing U.
S.
Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Eisenhut:
Re:
St. Lucie Unit 1
Docket No. 50-335 Post-TMI Requirements Reactor Coolant S stem Vents H~~ ]
AUGl4lS8t~
8 Qsh Please find attached our St. Lucie Unit 1 submittal in response to NUREG-0737 item 11.8.1.
The submittal contains a description of the design and results of analyses of the Reactor Coolant Gas Vent System.
As we previously stated'n our letter of December 23, 1980 (L-80-418),
we have postponed development of operating procedures until such a time that the design is approved.
Procedures for operation and technical specifications will then be developed and submitted.
Very trul
- yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/PKG/ras cc:
Mr. J.
P. O'Reilly, Region II Harold F.
Rei s, Esquire r(
8108i70268 8i08i0 PDR ADOCK 05000335 P
PDR PEOPLE... SERVING PEOPLE
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REACTOR COOLANT GAS VENT SYSTEM DESCRIPTION OF DESIGN AND RESULTS OF ANALYSES ST.
LUCIE UNIT 1 DOCKET NO. 50-335 I.
INTRODUCTION Pursuant to the requirements of NUREG-0737, Item II.B.l, a des-cription of the design,
- location, size and power supply for the vent system along with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe has been prepared for St. Lucie Unit 1 and appears below.
The purpose of the Reactor Coolant Gas Vent System (RCGVS) is to vent non-condensible gases from the Reactor Coolant System (RCS) which may inhibit core cooling during natural circulation.
The vents in this system are part of the reactor coolant pressure boundary
- and, as
- such, they have been designed to conform to the requirements of Appendix A to 10CFR50, "General Design Criteria,"
and Item II.B.1 of NUREG-0737.
The RCGVS does not lead to an unacceptable increase in the probability of a loss-of-coolant accident nor challenge containment integrity.
II'ESIGN BASES The RCGVS is designed to:
1.
provide vent paths during normal and accident conditions for the reactor vessel head and pressurizer steam space which are remotely operated from the control room; 2.
provide vent paths from the prim'ary system to the quench tank or to the containment atmosphere in an area allowing good mixing; 3.
provide positive indication of valve position in the control room; 4.
provide protection against single failure in the power and control of the vent valving; 5.
Seismic Category I and safety related requirements; 6.
limit reactor coolant mass loss to below the definition of a LOCA in 10CFR50, Appendix A; 7.
be operable following design basis events and loss of off-site or onsite AC power; 8.
permit venting of superheated
permit refueling and maintenance operations without interference;
IX ~
0 DESIGN BASES (Continued) 10.
IXI. SYSTEN DESIGN DESCRIPTION The Figure 1.
The system can be described as follows:
provide the capability to monitor RCGVS leakage during power operation.
1 ~
~
t W
1 RCGVS is schematically illustrated on the flow diagram, a ~
Vent paths are provided for the reactor vessel head or pressurizer steam space to either the containment -atmosphere or the pressurizer quench tank.
Venting.is accomplished through the appropriate line-up of any of six-(6) solenoid valves remotely operated from the control room.
The valves are located at a central station on the pressurizer cubicle north wall.
Positive indication of valve position is also provided on a
control panel (see Section XV.2.C) in the control room activated by reed switch assemblies on the solenoids.
The solenoid valves represent.
the active components in the RCGVS.
As such the power supply for these valves is from emergency sources (i.e, station batteries).
Additionally, the system is designed for a single active failure and thus parallel vent paths with valves powered from alternate power sources are provided.
c ~
The'olenoid operated valves are normally closed and de-signed to fail closed.
They are powered from redundant safety grade 125V DC power supplies.
Power is required to open the valves and they are spring loaded such that loss of power to the solenoid will result in a spring closing the
- valve, thus minimizing the probability of a
vent path failing to close once opened.
Sufficient redundancy in the quantity of valves, vent'aths, power supplies and valve controls are provided such that a single active failure of any component will not cause inoperability of the entire RCGVS.
Power is removed from the solenoid valves to minimize the probability of inadvertent operation of the RCGVS.
FPL will develop administrative procedures for reconnection of power in the event that operation of the RCGVS is required.
Vent points utilize'd are existing vents previously used for manual venting during normal start-up and shutdown opera-tions.
The reactor vessel vent is located directly on the reactor vessel head and the pressurizer vent is located on the high point of piping upstream of the Power Operated Relief Valves (PORV).
IXX. SYSTEN DESIGN DESCRIPTION (continued) '0 The vent flow rate capability is based on the following criteria:
The system is designed primarily for venting hydro-gen.
The vent rate is sufficient to vent one-half of the RCS volume of hydrogen in standard cubic feet'n one hour at RCS pressures greater than 50 psia.
ii.
Coolant liquid loss through the vent will not exceed makeup capacity of one (1) charging pump in the event of a
Safety Class 2 pipe break or inadvertent valve operation, thus limiting leakage to less than the LOCA definition of 10CFR50, Appendix A.
iii. The vent mass rate will not result in heat loss from the RCS in excess of the normal pressurizer ho~ter capacity.
-Venting will n'o't result kn uncontrollable pressurizer
'pressure'or level changes.
To meet the vent rate requirements above, of which III.f.ii is the most limiting, flow restricting orifices have been sized and installed at the existing vent points described in III.e.
These orifices also represent the division between Safety Class 1, Reactor Coolant Pressure Boundary and Safety Class 2.-,.portions of the system.
Vent paths are provided to both the pressurizer quench tank and containment atmosphere.
The quench tank path allows for controlled venting of non-condensible gases in that the discharge into the quench tank is below the tank water level, thus promoting cooling of the gas or water vapor, and from the tank non-condensible gases can be discharged to the gaseous waste management system.
The vent to containment atmosphere terminates in an area where good air mixing and cooling exist.
In the unlikely event of generation of large quantities of gas combined with a failure in the vent path to containment atmosphere, gases can be discharged to the c(uench tank where with sufficient pressure a rupture disc on the tank will fail and release the gases to the containment atmosphere at approximately the same location.
The addition of the RCGVS does not affect the rate of generation of combustible gases following a postulated Loss of Coolant Accident (LOCA) as previously analyzed in the safety analysis report in accordance with the requirements of 10CFR50.44.
Additionally,'ent paths to containment atmosphere discharge to open areas of the containment of good air mixing.
XXX. SYSTEM DESXGN DESCRIPTION (continued)
Pursuant to Technical Specification 3.4.6.2, Paragraph d,
a method of leakage detection is provided to identify and ensure that any leakage in the RCGV system is identifiable.
This allows continued power operation at leak rates greater than 1
gpm but less than 10 gpm.
Remote leakage detection is accomplished through the use of a
pressure indicator mounted in the control room which will alarm on high pressure indicating a leak through one of the primary RCGVS valves (V1441 thru V1444).
A solenoid valve (V1449) is then opened and the leakage is discharge to an accumulator and eventually drained to the containment's graduated sump where it can be measured.
k.
The leakage detection piping may also be utilized for venting the RCS during normal start-up and shutdown operations.
A bypass line upstream of the orifice on the pressurizer vent is provided to faciliate venting during normal shutdown operations.
IVo MAJOR COMPONENTS
.1 Mechanical Com onents Major system components are listed, along with a brief des-cription, in Table l.
a.
Piping:
All piping is manufactured from austenitic stainless steels and is Nuclear Safety qualified. Safety Class 1
piping forming part of the reactor coolant pressure boundary is qualified in accordance with ASME Section III, 1977 Edition, Summer 1978 Addenda.
System piping is flanged where required to facilitate removal of com-ponents that might interfere with refueling operations.
b.
Manual Valves:
Manual valves are manufactured from austenitic stain-less steel and qualified in accordance with ASME Sec-tion XIX, 1977 Edition, Winter 1978 Addenda.
Addition-ally, all valves are also seismically qualif.ied.
IV.
MAJOR COMPONENTS (continued) co Solenoid Valves:
The solenoid valve bodies are austenitic stainless steel forgings qualified in accordance with ASME Sec-tion III, 1977
- Edition, Winter 1978 Addenda.
The solenoids and electrical appurtenances are also quali-fied to IEEE 382-1972, 323-1974 and 344-1975.
Addi-tionally, all valves are seismically qualified and designed to fail closed on loss of electric power to the solenoid.
d)
Orifices:
The two (2) orifices are austenitic stainless steel bar stock with a 1/4 inch hole drilled through, sized to provide the appropriate flow rates required by the RCGVS.
Fabrication was in accordance with ASME Section III, 1977 Edition, Summer 1978 Addenda.
e.
Accumulator:
The accumulator is a length of 8 inch nominal diameter Schedule 40 pipe capped at both ends and is non-nuclear safety related.
Its function during leakage detection is to allow for the expansion and condensation of steam leaking through any of the primary RCGVS valves.
A drain from the accumulator discharges to the contain-ment sump where leakage can be measured.
Charcoal Filter:
The charcoal filter is a length of 4 inch Schedule 160 pipe filled with charcoal and connected to the vent line on the accumulator.
The purpose of the filter is to prevent carry-over of radioactive particles from the RCGVS to the plant vent system during leakage detection or normal start-up/shutdown operations.
The heavy wall of the filter pipe is for radiation shielding purposes and not service conditions.
The filter is also non-nuclear safety related.
.2 Instruments and Controls a
~
Pressure Instrumentation:
The pressure instrumentation is provided for leakage detection only and as such is not required for post-accident conditions.
For this reason the pressure in-strumentation is not Class IE qualified.
The locally mounted pressure transmitter and associated tubing are
4 Zt C
-IV.
MAJOR COMPONENTS (continued)
'a;"-'{cont,inued) seismically supported such that in the unlikely event of a design seismic occurence the function of the pres-sure transmitter may fail but the integrity of the
'=,"--'=:"-pressure retaining components will be maintained.
=.:-.-;-'~">:-.Power for the pressure instrumentation is also from
.:, "emergency sources.',
b.
- Control Switches:
The" valve control switches provided are of a
modular plug-in design with key-locks provided.
Each module is
',. provided with two (2) indicating lights, red and green, which are tested by pressing the light bulb.
Control switches are qualified in accordance with IEEE 323-1974 and 344-1975.
c..
Control Room Auxiliary Console:
,;-.-. A new control panel, the Control Room Auxiliary Con-
,- sole, is provided in the control room and is utilized
.for the RCGVS instruments and controls.
The panel is
-"-" .=-'-'--.:-'safety related and of seismic design and
'--'.construction.
A graphic display is provided allowing
-the operator to observe the various flow paths
- =... ';:,, =,
,.:-, ~;,...'-.:><-available prior to activation of any solenoid valve.
,V.'RINCIPAL MODES OF OPERATION
-'illing of the RCS prior to plant start-up can be accom-plished either manually or remotely using the RCGVS.
For manual operation, manual valves V1454 and V1455 are opened to the accumulator, thus bypassing the pressurizer seni8te/inariual, vent path.
Reactor vessel venting can be accomplished by opening either V1441 or V1442 through the accumulator via V1449 or through the pressurizer vent
- path, via V1443 or
- V1444, to the manual bypass.
If it is decided not to use the manual bypass, filling is accomplished by lining up the
'ressurizer vent valve (V1443 or V1444) and the reactor head vent valve (V1441 or V1443) directly to the accumulator via V1449.
'During start-up operations, fluid or gases released from the
.=. -RCS are directed to the accumulator.
Potentially contamin-ated fluids are drained from the accumulator directly to a
';. floor drain which discharges to the containment sump.
- Any
..- contaminated gases released are vented from the accumulator,
.'.-through the charcoal filter to the containment purge header.
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V.
PRINCIPAL MODES OF OPERATION (continued)
~ 2 Normal Operation This system is not intended for use during normal power operation and administrative controls are provided to mini-mize the possibility of inadvertant operation.
Addition-
- ally, power is removed from all valves during normal plant conditions.
During normal operation, leakage detection is maintained by use of the pressure instrumentation.
A rise in pressure will indicate leakage past any of valves V1441 thru V1444.
Technical Specifications require the ability to identify leakage above 1
gpm from the RCS to allow continued reactor operation.
This may be accomplished in the RCGVS using either of two methods:
a.
If the pressure increase is slow enough, the leakage rate can be determined by observing the rate of pres-sure increase per unit time.
b.
The leakage can be diverted through V1449 to the accum-ulator.
When initial transients have subsided, the ac-cumulator discharge to the containment sump will allow for leakage measurement by the existing leakage detection system.
Verification of the leakage source may be accomplished by periodically closing V1449 and observing change in leakage indicated by sump instrumentation, containment radiation or charging versus letdown flows.
4 3 Refueling 'Shutdown Procedure for the RCGVS during plant refueling shutdown (Mode 5 to Mode 6) are basically the same as for start-up.
- However, reactor head vent valves, V1441 or V1442, need not be open for refueling shutdown operations.
~ 4 Accident Conditions Use of the RCGVS during accident conditions in which large quantities of non-condensible gases are generated vary de-pending on the rate of gas generation and on vent paths available assuming design basis accidents.
The primary method of controlled venting for low gas genera-tion. rates is through the quench tank.
Reactor and/or pres-surizer vent valves could be lined up with V1445 and the gas released to the quench tank.
Monitoring of quench tank pressure is highly recommended during this mode of opera-tion.
From this point the gas could be discharged to the gaseous waste management system.
through V1445 and the quench tank rupture disc.
For extremely low gas generation rates
-in the reactor ves-
- sel, the capability exists to vent the vessel by lining up a reactor vent valve, V1441 or V1442, with a pressurizer vent
- valve, V1443 or V1444, and allowing gases to "bubble" to the pressurizer at rates dependent on pressure differential due to pressurizer level, system pressure/temperature and reactor coolant pumps status.
Although this ca abilit exists this vent ath is o
d.
p n t ptfeferre VX SAFETY EVALUATION t
PRXNCIPAL NODES OF OPERATXON (continued)
.4 (Continued)
For high gas generation
- rates, gases may be vented to the containment atmosphere by opening V1446.
Should V1446 fail, vent to containment atmosphere can, still be accomplished Transients:
During normal plant operation the reactor vessel vent and pressurizer vent up to the first isolation valves will be subject to existing RCS transients previously analyzed in the Safety Analysis Report except that flow rates.will be zero
(.0) and temperatures will decrease to containment am-bient due to heat transfe from the piping as it extends from the vessel head and pressurizer.
Downstream of the first;isolation valves no transients will occur during nor-mal operation except in the event of valve leakage.
Plant start-up and shutdown is a normal event and the RCGVS is designed for 500 cycles of this transient.
Pressures and temperatures encountered during start-up and shutdown are far below those used in the design of this system.
The post accident venting of hydrogen transient is the de-sign basis for the RCGVS and is classified as an emergency with 20 cycles occuring over the lxfetime of the plant.
0 2 Sin le Active Failure Anal sis Single active failure for the RCGVS would incorporate either failure of a
power operated'alve or loss of A or B power supply (LOOP concurrent. with a diesel generator failure).
Redundancy in valves for the reactor vessel vent and pres-surizer vent (V1441 thru V1444) will allow continued system operation in the event of such.
an occurence.
Failure of V1445 or V1446 however, though not prohibiting venting, will dictate what vent path is available.
Table 2 is a listing of possib'le component failures and the impact, on operation of the RCGVS.
VI.
SAFETY EVALUATION (continued)
'\\
- 3 Seismic Anal sis All.. components, piping and supports in the
- RCGVS, in the Nuclear Safety Class 1
and 2 piping, are specified and de-
'. signed as Seismic Category I.
Piping has been analyzed and
's'upported in accordance with St.
Lucie Unit 51 seismic cri-teria.
All valves have been analyzed and tested for oper-ability during a seismic event by manufacturers.
The seis-
'. mic-'nalysis is consistent with previous plant design and
. construction.
.4 Pi e Break Anal sis The separation between Safety Class 1 (Reactor Coolant Pres-sure Boundary) and Safety Class 2 portions of the RCGVS is accomplished by the use of the restricting orifices describ-ed in III.g.
A pipe break downstream of these orifices, in the Class 2 piping, would result in a
mass loss less than a
. LOCA as defined in
- 10CFR50, Appendix A and thus a separate analysis of inadvertent system operation or pipe breakage is not required to meet 10CFR50.46.
The. unlikely event of a pipe break in the Safety Class 1
portions of the RCGVS would result in a
mass loss greater than the minimum defined as a
LOCA.
The Emergency Core Cooling System (ECCS) as designed in accordance with 10CFR 50.46..is sized for pipe line breaks in the.reactor coolant pressure boundary with blowdown areas as large as 9.82 square feet (reference Chapter 6 of the Final Safety Analy-sis Report).
All Class 1 piping in the RCGVS is 3/4 inch
..--nominal pipe size and a break in any of these lines will not burden the existing ECCS system.
Additionally, routing of the RCGVS piping is such that it is protected from pipe whip and jet impingement effects from postulated pipe breaks in RCS cold leg piping, branch lines to the -cold leg and non-RCS piping.
With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed
...,.safety
. question; (i) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or mal-
.,function of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is
'-. reduced.
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VI.
0-..::.
-:=-
~
SAFETY EVALUATION (continued
)
The Reactor Coolant Gas Vent System (RCGVS) is an addition to the Reactor Coolant System (RCS) mandated by the NRC under the pro-visions of 10CFR, Part 50.109, as a result of investigations into the accident at Three Mile Island.
The system and component design criteria for the RCS are established in Chapter 5 of the St.
Lucie Unit 41 Final Safety Analysis Report and have been utilized as the basis for the design and implementation of the RCGVS.
Therefore the probability of occurence or the conse-quences of an accident previously evaluated in the safety analy-sis report,
- namely, the small break LOCA, has been addressed and.
is not increased.
The RCGVS is an addition to the RCS with the dedicated function of mitigating the consequences of an accident not previously evaluated in the safety analysis report and does not in itself represent the source of such an accident or any other type not previously evaluated.
Additionally, because the primary function of the RCGVS is to mitigate the consequences of an accident not previously addressed in the safety analysis re-port the margin of safety as defined in the basis for any tech-nical specification is not
- reduced, but increased.
The design philosophy of the RCGVS is consistent with that of all other Nuclear Safety Class systems of St.
- Lucie, and with all applic-able
- codes, regulations and regulatory guides.
It is also con-sistent with the requirements of NUREG-0737 "Clarification of TNl Action Plan Requirements",
Part II.B.l issued by the NRC October 31, 1980.
VII. TESTING AND INSPECTION All components, except pressure instrumentation, have been speci-fied and purchased as Seismic Category 1 and Nuclear Safety Class (where required).
Vendors have substantiated either through
- test, calculational and/or operational data that system compon-ents will remain operable under the design seismic loads.
Ven-dors have tested and inspected all safety class equipment in ac-cordance with applicable ASME and IEEE codes.
A periodic operational testing and inspection program shall also be developed by FPL in accordance with ASME Section XI, Subsec-tion INV, to ensure system operability after installation.
4
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'.TABLE 1
- MAJOR SYSTEM COMPONENTS 1.
Valves (All materials austenitic stainless
- steel, for Code Qualification, See Section IV).
lA)
Power Operated Size lff lff lff lff 1 II 1 II lff V14 41 V1442 V1443 V1444 V1445 V1446 V1449 lB)
Manual
~Ta No.
0 erator Solenoid Solenoid Solenoid Solenoid Solenoid Solenoid Solenoid Design
-Pressure Temp
-2500psig 700'F 2500psig 700'F 2500psig 700'F 2500psig 700 F
2500psig 700'F 2500psig 700'F 2500psig 700'F Safety Class 2
2 2
2 2
2 2
Power Su pl (A/B)
A B
A B
B A
B 2 ~
Taca No.
Vl239 V1447 V1450 V1452 V1453 V1454 V1455 Orifices Size 3/4 ff 1 ff 1 ff
] ff 1 II 3/4 ff 3/4 ff Ratinca 15005 15005 15000 15005 15005 15005 15004
.-Safety Class 1
= 2-2.2-..
. 2....
1 1
e'osition Locked Open Normally Open Locked Open Locked Closed Locked Closed Locked Closed Locked Closed 3.
Ta No.
I-SO-01-58 I-SO-01-59 Accumulator Material Diameter Length Thickness Tag No.
Code Size
Material 3 4"xlffxl 4" ID 316 SS 3/4ffxlffxl/4ffID 316 SS (Non-Nuclear Safety Related) 304 Stainless Steel 8 inch 29 inches Schedule 40 Pipe 8-RC-865 B31.1 Des Pressure 2485psig 2485psig ign Safety
~Tem Class 650 D
1 675oF 1
4
~
Charcoal Filter (Non-Nuclear Safety Related)
Material Diameter Leng th Thickness Tag No.
Code 304 Stainless Steel 4 inch 4 I 0ff Schedule 160 Pipe 4-RC-864 B31.1
(
TABLE 2 FAILORE NSES EFFECTS ANALYSIS FOR THE REACTOR COOLANT GAS VENT SYSTEM I~
Pressure Indicator*
P IA-1117'a I lure Mode
- a. Spurious high press Ind1cat ion/alarm
- b. Spurious low press Indication Cause E Iectro~chan Ica I failure,setpolnt dr Ift E Iectro-meehan Ica I fa I lure, setpo int drift.
Symptoms and Local Effects Includin De endent Fa1 lures IIo Impact on normal operation.
Loss ot ability to detect leakage Into the vent system piping.
No Impact on normal operation.
Loss of ability to detect leakage Into the vent system piping.
Method of Detect Ion Valve position Indicat-ionn ln the C.R Valve position Indica-tion In the C.R.
Inherent Compensating Prov Is Ion Page I of 3 Remarks and Other Effects 2..
Quench Tank IsolatIon
- a. Falls Open Valve VI445
(
bi Fails Closed Mechanical blnd-lng, seat leakage Mechanical fa II-
- ure, loss of
- power, Inability to Isolate quench tank from the reactor coolant gas vent No Impact on normal operation.
'nability to vent pressurizer or reactor to quench tank.
Valve position Indica-tion In the C.Ri system.
Valve position Indlca-tlon ln the C.R.
None Redundant Isolation valves to the re-actor vessel and pressurizer pre-clude uncontrolled venting to the quench tank.
Venting to the containment Is possible, If n aces saryr 3
Pressure Instrument
- a. Falls Open Isolation Valve V1447
- b. Falls Closed Hechan Ica I bind<<r lng, seat leakage Hechan ical fa I I-ure Loss of ability to detect seat leakage from the pressurizer and reactor Isolation valves Into the reactor coolant gas vent system piping.
Operator Operator Redundant valves
. None Un I lke ly event since valve Is normally open and has only a manual operators Contalnaont Isolation
- a. Falls Open Valve YI446
- b. Falls Closed Hechan Ica I Bindi-ngg, seat leakage Meehan Ice I Fa II-ure, loss of power.to valval'nability to Isolate reactor coolant vent system from containment No Impact on normal operation.
Inability to vent pressurizer or reactor to conte lnaint, Nigh containment press f humidity lf venting Is In progress.
Valve position Indication In the C.R, Ya Ive position indicati-onon In the C,R, Operator Red(w(dent Isolation valves provided to preclude uncontro II-ed venting to RCS Yentlng to the quench tank Is possIble, If necessarye
'i 4
I TABLE 2 FAILURE MODES EFFECTS ANALYSIS FOR THE REACTOR COOLANT GAS VENT SYSTEM No.
Naae 5,
Pressurizer Vent Isolation Valve
~
VI443 or VI444
'I I
Fa I lure Mode
- a. Fal I, Open
- b. Falls Closed Cause Meehan lcd I 8 lnd-
. Ing, seat leak-dge Mechanical fa i I-ure, loss of power Sydptors and Local. Effects Inc Iud ln De en dent Fa I lures No lepdct on normal operation, Indbl I ity to vent the reactor vessel without also venting pressurizer.
Inability to vent the pressurizer Method of Detect lon Valve position Indica-tion In the C.R, PIA-1117 high pressure Indication.
Valve position In the C.R.
Isolation valve Operator.
None Parallel redundant Isolation va Ive, Redundant Isolation valves to contaln-aent VI446 6 quench tank VI445 precludes uncontrolled venting of the pressurizer Parallel Isolation valve dl lows venting of tho pressurizer, Pa90 2 of 3 Inherent Compensating Rawrks and Prov lslon Other Effocts
~
~
6 Reactor Vessel i Vent Isolation Valve VI441 or VI442 a, Falls Open
- b. Falls Closed Mechanical Bind-ing, seat leakage.
No impact on nornnl operation.
Unable to vent pressurizer without also venting the reactor vessel
~
Mechanical fa1 I-Inabl I lty to vent the reactor
- ure, loss of poweri vessel.
Valve position Indica-tion ln the C.R.
P IA-1117 high pressure In d I cat Ion, Valve position In the C.Ri Operator Parallel redundant lsoldtlon vdlve Redundant Isolation valves to containm-entt VI446 d YI445 precludes uncon-trol led venting of the I edciol vessel
~
Parallel isolation valve allows venting of the reactor vessels 7,
Pos itIon Indicator for YI441 6 YI442 False Indication of va Ive pos ltIon E Ioctro-sec hen lca I fa Ilure.
Loss of ability to detect valve position In reactor vessel vent I lne, Pressure Gauge PIA-117 Indication shows valve Is opened.
8.
Position Indicator for YI443 6 VI444 Fa Ise Ind1cat lon of, El ectro-meehan lca I va Ive position fa I lure.
Loss of ability to detect valve position In pressurizer vent I lnei Pressure Gauge P IA-II 17, Indication shows valve Is opened.
9, Position Indicator for YI445.
False Indication of valve position E Iectro~chan ica I fal lure Loss of ability to detect valve position ln quench tank vent I lne.
quench tank temp 6 pressure verify valve positions Press gauge PIA-1117 C-I t
g ~
4
TASLE 2 FAIL'NIE MODES EFFECTS ANALYSIS FOR THE REACTOR COOLANT GAS VENT SYSTEM 10 Position Indicator for VI446 Fs I lure Mode False Indication of vol ve posit ion Cause E Iectro-meehan Ica I failure Symptoms nnd Local Effocts Inc iud ln De endent Fal lures Loss of ability to detect valve position In containment vont I lne.
Method of Detect Ion Containment pressure/
humidity/radiation levels verify conte ln-ment valve position.
Press gnuse PIA-III7.
Inherent Compensating Prov Is Ion tkPne Page 3 of 3 Remarks nnd Other Effects 11.
Vent d Drain Valves
- s. Seat lenknge YI452 d VI453
- b. Falls Closed Contamination, mechanical damage Mechnn Ics I binding No impact on system operation, No Impact on normal operations, Inability to drain affected I lne section or tost Isolation valves per Iwf ASME Xlr Operator Drain lines sre blind flsnged None 12w Leakage Detect lon Valve VI449
- n. Falls Open bw Feil Closed Meehan lca I binding, Insbl I ity to Isolate leakage seat leakage detection system from RCGVS.
Mechanical failure, No Impact on system operation.
loss of power Loss of ability to measure lonknge remotelyw Valve position Indlcaww tlcn In C.R.
Valve position Indica-tion In C.R.
None Leakage detection system represents anothor path to conte lnmentP though not recoPPPPNnded to be used ss suchw 13.
Pos ltion Indicator for VI449 4
4 4
False Indication of vn Ivs posit ion P
P Electro-PPechnnlcal Loss of abl I lty to detect vs Ive fs I lure position in leakage detectIon I lne, 4
Ih 4
Drain from leakage detection system to grsdunted sump, in-crease In sump level shows valve Is open.
PP" 4
4
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4.
P P
P
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r 44 4IP
L-RC-SS) 8 0
I 1-ac-854 2-xc-856 1-ac-a63 pc c
0 49 ACCUHULATOR (8-xc-865) 3/4-ac-461 gee3
+P0 L-RC 458 VL454 V145 3/4-aC-131 4-ac-Lo1~
OIARrllAL 1 RC.862 P II:IRK V1452~
[~3IIHT 3lf333 ~Ilk 3C 4-xc-864 1 Rcv859 L-xc-dlo c
n 5
HI (A)
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P~c VL443 C
0 S
Hl (8)
II 5QG3LR 2 L/2 LS VL407 r
a 4-ac-82$
Lpc L-xc-<<9 L-xc-848 3/4-XC 860
~ VL239 LO v1402 1
RC 864
~ tao( toaV "N' Rc 824
~xcrKO KRM RxlSTLKC Lo RC 822 RRLlat VALVC DLSCKARCR QRADRR TD toav "a" A
C 1
1 RC 868
~2-xc-283
~Z-V-OLOOO(1342)
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1 V~ASSR KfrrLad 1-ac-454 2 RC-453 V1445~
05 H pc 1 xc-85L~
Allr 1-ac-a52 1 RC 446 1-RC-846 1-RC 455 33 ~ 33 L-ac-84&
LC 3/4 RC-136 A
1-80-01-54~
VI45U C
0 A)
PCCVL44L S
HIvvQ (8) g ~VL442 1-xc-867 1 ac-460~
1 ac-450 vlaal+
RQASCO DEAVICCIINCORPOIIATXD HCW Yr3313r
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Lr PC RRACTOR COOIAMT CA6 VRIIT STSTITI PLAW DIAGRAM PLCORC REv. N.
1 DATE 4-20-44
II.B.1 REACTOR COOLANT SYSTEH VENTS Position CONPARISON.OF 0737 RE UIRBiENTS/RCVGS SYSTEM DESCRIPTION REFERENCE Each applicant and licensee shall install reactor coolant system (RCS) and reactor vessel heed high point vents remotely operated from the control room.
Although the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity.
Since these vents form a part of the reactor coolant pressure
- boundary, the design of the events shall conform to the requirements of Appendix A to 30 CFR Part 50, "General Design Criteria."
The vent system shall be designed with sufficient redundancy.
that assures a low probability of inadvertent or irreversible actuation.
Each licensee shall provide the following information concerning the design and operation of the high point vent system:"
(1)
Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for loss-of"coolant accidents 1nitiated by a break in the vent p1pe.
The results of the analyses should demonstrate compliance with the acceptance criteria of 10 CFR 50.46.
(2)
Submit procedures and supporting analysis for operator use of the vents that also include the information available to the operator for initiafing or terminating vent usage.
Chanoes to Previous Reouirements and Guidance
- Generic requirement incorporated in entire package.
For 10CFR50.46 compliance, see III.f.iiand VI.4 Procedures will be prepared upon Commission approval of system design (reference FPL letter to NRC L-80-418 dated 12/23/80).
(1)
The probability of a valve failing to close, once opened, should be minimized.
(2)
Es ablishes environmental qualification (Commission Order, Hay 23, 1980).
(3)
Establishes provisions for testing.
(4)
Delete requirements of September 27, 1979 letter from Vassallo to appli-cants stating that vents shall satisfy single-failure criteria of IEEE-279.
Vent systems are not required to have redundant paths.
A degree of redundancy should be provided by powering different vents from different emergency buses.
(5)
Documentation date changed to July 1, 1981 and implementation date to July 1, 1982.
Clarification does not change NRC concept of requirement, but provides more detail on scope.
The dates have been revised to provide time for procurement and installation.
It was tne intent of the October 30, 1979 letter to delete the requirement to meet the criteria of 10 CFR 50.44 and SRP 6.2.5 for beyond-design-basis events.
The analysis requirements of Position 2 in the September 13, 1979 letter are therefore unnecessary.
3-55 Not requirements, see Clarification Section on next sheet.
A.
(2)
, (3)
(4)
(5)
(6)
General Procedures addressing the use of the reactor coolant system vents should define the conditions under which the vents should be used as well as the conditions under which the vents should not be used.
The procedures should be directed toward achieving a subs antial increase in the plant being able to maintain core cooling without loss of containment integrity for events beyond the design basis.
The use of vents for accidents within the normal design basis must not result in a violation of the requirements of 10 CFR 50.44 or 10 CFR 50.46.
The size of the reactor coolant vents ismot a critical issue.
The desired venting capability can be achieved with vents in a fairly broad spectrum of sizes.
The criteria for sizing a vent can be developed fn several ways.
One approach, which may be considered, is to specify a volume of noncondensible gas to be vented and in a specific venting time.,
For containments particularly vulnerable to failure from large hydrogen releases over a shor t period of time, the necessity and desirability for contained venting outside the containment must be considered (e.g., into a decay gas collection and storage system).
Where practical, the reactor coolant system vents should be kept smaller than the size corresponding to the definition of LOCA (10 CFR 50, Appendix A).
This will minimize the challenges to the emergency core cooling system (ECCS) since the inadvertent opening of a vent smaller than the LOCA definition would not require ECCS actuation, although it may result in learage beyond technical specification limits.
On PMRs, the use of new or exis ing lines whose smallest orifice is larger than the LOCA definition will require a valve in series with a vent valve that can be closed from the control room to terminate the LOCA that would result if an open vent valve could not be reclosed.
A positive indication of valve position should be provided in the control room The reactor coolant vent system shall be operable from the control room.
Since the reactor coolant system vent will be part of the reactor coolant system pressure boundary, all requirements for the reacto~ pressure boundary must be ret, and, fn addition, sufficient redundancy should be incorporated into the design to minimize the probability of.an inadvertent actuation of the system.
Administrative procedures, may be a viable option to meet the single-failure criterion, For vents larger than the
~car HHz I
The important safety function enhanced by this venting capability fs core cooling.
For events beyond the present design basis; this venting capability vill suustantially increase the plant's ability'to deal with large quantities of nonccndensfble gas which could interfere with core cooling..
- Inherent in system design, see III.
- Procedures will be prepared upon Commission approval of system design.
For 10CFR50.44 compliance see III.i. For compliance to 10CFR50.46 see III.f.iiand VI.4.
- Paragraphs Ill.fand III.h.
- Paragraphs, III.f.ii,III.g and III.c.
-'Paragraph III.a.
Paragraph III.a.
Paragraph III.g, IV.l.a, IV.ld, VI.4 and VII address
.RCS pressure boundry design.
System redundancy is discussed in Paragraphs III.b, III.c and III.d.. For 10CFR50.46 compliance see Paragraphs Ill.f.iiand VI.4.
3-56
(8)
LOCA definition, an analysis is required to demonstrate compliance with 10 CFR 50.46.
The probability of a vent path failing to close, once opened, should be minimized; this fs a new requirement.
Each vent must have its power supplied from an emergency bus.
A single failure within the power and control aspects of the reactor coolant vent system should not prevent isolation of the entire vent system when required.
On BWRs, block valves are not required in lines with safety valves that are used for venting.
~ r'I qV ~,
Paragraph III.c (9)
Vent paths from the primary system to within contafnment should go to those areas that provide good mixing with containment air.
- Paragraph III.h (10)
(12)
The reactor coolant vent system (f.e., vent valves, block valves, position indication devices, cable terminations, and piping) shall be seismically and environmentally qualified fn accordance with IEEE 344-1975 as supple.
mented by Regulatory Guide 1.100, 1.92 and SEP 3.92, 3.43, and 3.10
'nvfronmental qualifications are in accordance with the Hay 23, 1980 Commission Order and Hemorandum (CLI-80-21).
Provisions to test for operability of the reactor coolant vent system should be a part of the design.
Testing should be performed in accor'dance.
with subsection IWV of Sectfon XI of the ASHE Code for Category 8 valves.
It is important that the dfsplays and controls added.to the control room as a result of t)Tfi'requiremen7 nof increase the potential for operator-error.
A human-factor analysis should be performed taking into considera-tion:
I I
'- Por eahipment qualification see Section IV
'ectiop VII Paragraph IV.2.c.
(a) the use of this information by an operator during both normal and abnormal plant conditions, (b) integration into emergency procedures, (c) integration into operator training, and (d) other alarms during emergency and need for prfor1tizatfon of alarms.
B.
BWR Design Considerations Since the BWR owners'roup has suggested that the present BWR designs have an inherent capability to vent, a question relating to the capability of existing systems arises.
The ability of these systems to vent the RCS of noncondensible gas generated during an accident must be demonstrated.
Because of differences among the head vent systems for BWRs, each licensee or applicant should address the specific design features of this plant and compare them <<1th the generic venting capability proposed by the BWR owners'roup.
In addition, the abil1ty of these systems to meet the same requirements as the PWR vent system must be documented.
In addition to RCS venting, each BWR licensee should address the ability to vent other systems, such as the fsolat1on condenser which may be
- Not applicable to St. Lucle Unit 81 II.8.1-3 3 57 P
~(
required to maintain adequate core cooling. If the production of a large amount of noncondensible gas would cause the loss of function of such a
- system, remote venting of that system is required.
The qual1fications of such a venting system should be the same as that required for PMR venting systems.
4, C.
PMR Vent Oesign Considerations (1)
Each PWR licensee should provide the capabil1ty to vent the reactor vessel head.
The reactor vessel head vent should be capable of venting noncondensible gas frcm the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and cold legs (through head Jets and other leakage paths).
(2)
Additional venting capability is required for those portions of each hot leg that cannot be vented through the reactor vessel head vent or pres"'urizer.
It is impractical to vent each of the many thousands of tubes.
in a U-tube steam generator; however, the staff believes that a procedure can be developed that assures sufficient liquid or steam can enter the U-tube region so that decay heat can be effectively removed from the RCS.
Such operating procedures should incorporate this consideration.
't (3)
Venting of the pressurizer is required to assure its availability for system pressure and volume control.
These are 1mportant considerations, especially during natural circulation.
'ADolkcabilit This requirement applies to all operating reactors and applicants for operating license.
lmolementation Installation should take place by July 1, 1982.
Until staff approval is obtained, installation may proceed; but operating procedures should not be implemented and valves should be placed in a condition so as to minimize the potential for inadvertent actuation (e.g.,
remove power).
Tvoe of Review A preimplementation review will be performed prior to authorizing use of the vent.
Oocumentation Reouired Sy July 1, 1981, the licensee shall provide the following information on the reactor coolant vent system for staff review:
(1)
The information requested in items 1 and 2 under "Position";
~
~
- Paragraphs III.a and III.c I p I
Procedures for venting system generator tube bundles have been prepared by CE.
(Reference CE Report CEN-128).
- Paragraphs III.a and III.e.
See above.
~
~
3-58
(2)
A discussion of the design with respect to coriformance to tho design criteria discussed under "Clarification," inclpding deviations, if any, with adequate justification for such deviations;
- and, (3)
Supporting information including logic diagrams, electrical schematics,:
piping and instrumentation
- diagrams, test procedures, and technical specifications.
Technical Specification Changes Reouir ed Changes to. technical specifications will be required.
References"'UREG-0660 Commission Orders, May 23, 1980 (CLI-80-21)
Letter from D.
G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated'eptember 13, 1979.
Letter from D. 8. Vassallo, NRC, to All Pending Operating License Applicants, dated September 27, 1979.
Letter from H.
R. Denton, NRC, to All Operating Nuclear Power Plants, dated Dctober 30; 1979:
- See hbove See Figure 1.
Precedures and technical specification will be prepared upon Conmission approval of system design (Reference FPL letter to NRC L-80-418 dated 12/23/80)
- See above.
II.8.1 5 3-59
I"