LD-92-060, Forwards Draft Sys 80+ Tier 1 Design Descriptions & Insps, Tests,Analyses & Acceptance Criteria for Sys 80+ Std Design.Meeting Requested on 920512 to Discuss Encl Design Descriptions
| ML20095L271 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 04/30/1992 |
| From: | Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LD-92-060, LD-92-60, NUDOCS 9205070020 | |
| Download: ML20095L271 (66) | |
Text
_ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ . __
ABB ASrA DROWN PoVE Al April 30,1992 LD 92-060 Docket No. 52 002 U.d. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
System 80+" Tier 1 Design Descriptions and ITAAC's
Reference:
1.etter LD-92-038 from C. B. Brinkman to D. M. Crutchfield, dated March 25, 1992
Dear Sirs:
Enclosed are ten (10) example Tier i design descriptions and asscciated ITAAC's for the System 80+" Standard Design.
ABB Combustion Engineering (ABB-CE) has benefitted from the sample ITAAC's which the NRC and the ITAAC lead design have been developirig. We consider the enclosure to be draft Tier i material because revisions are expected as System 80+ design certificatiori proceeds. We look forward to staff feedback on these initial ITAAC -
submittals and to agreeing with the staff on the format and content of the Tier 1 design descriptions and ITAAC which are most appropriate for the Sydm 80+ design, s We request a meeting with the staff by May 12, 1992 to diutJs Se enchsed design -
descriptions and ITAAC's We have assembled a multii discip;i'.afy team which is proceeding immediately with the preparation of the next set of ITAAC's. In order for us -
to meet the common objective of submitting the bulk of the Tier i deNgn Gsciiptions and the associated ITAAC before the draft Safety Evaluation issues 'm September, it is -
necessary that we receive very prompt feedback from the' staff on the adequacy of the sttached material.-
g g ;) - A.B B Combustion Eggineging Nucipar Power comtm,m rm w getygg; us :
gepu won 8n ::nnec A- ppR 4+ 1-
This submittal fulfills the commitment in the reference letter to supply a set of about 10 representative ITAAC by April 30th. Please contact me or Mr. George Hess of my staff at (203) 285-5218 if you have any questions on the enclosed material.
Very truly yours, COMBUSTION ENGINEERING, INC. l
. 1 4._
C. B. Brinkman Acting Director Nuclear Systems Licensing cbb/ mis dnclosures: As Stated cc: T. Wambach (NRC)
T. Boyce (NRC) i A. Heymer (NUMARC) )
J. Trotter (EPRI) l l
1 l
-. - , = . . . - . _ . .-
S.mliM 80+ 5 l 1.3.2 DESIGN FOR TIIE PROTECTION OF STRUCTURES, COM-PONENTS, EQUIPMENT AND SYSTEMS AGAINST DYNAMIC EFFECTS OF PIPE HREAK AND LEAK HEFORE IIREAK Design Description Vital structures, components, equipment and systems necessary for the successful operation of the safe shutdown systems are protected from the damaging effects of postulated pipe breaks not climinated by leak before break evaluations. Designs which protect these items consider the consequcnces of pipe whip, water spray, jet impingement, Gooding, compartment pressurization, and environmental conditions.
Essential systems protected are those that are needed to safely shut down the reactor or mitigate the consequences of a pipe break for a given postulated piping failure, in addition, systems are protected such that no break violates the following criteria:
1, A postulated pipe break which is not a LOCA will not cause a I.OCA.
I 2.
I The postulated pipe break will not cause unacceptable consequential reactor coolant steam or feedwater line damage.
- 3. The function of safety systems required to perform protective actions to mitigate the consequences of the postulated break will be maintained.
- 4. The ability to place the plant in a safe shutdown condition will be maintained.
Protection of vital equipment is achieved primarily by s+paration of redundant safe j
shutdown systems and by separation of high-energy pipe lines from safe shutdown systems. This redundancy and separation results in a design which requires very few special protective features such as whip restraints and jet denectois to ensure safe
- shutdown capaHlity following a postulated high-energy line break. In general, layout of the facilities follows a multi-step process to ensure adequate separation.
- 1. Safety-related systems are located away from most high-energy piping.
- 2. Redundant safety systenu and subsys: cms are located in separate compartments.
1.3.2 1 4 30-92 l
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,~ - ---, . _ -,. . -- _-- . _.,-.. ---.-_ .- -
,_, DRAFT
- 3. As necessary, specific components are enclosed to maintain the redundancy required foi those systems that must function as a consequence of specific piping failure events.
Protection requirements are met through the protection afforded by the walls, cetumns, floors, abutments, and foundations in many cases. Where adequate protection does not already exist due to separation, additional barriers, deflectors, or shicids are provided u necessary to meet the functional protection requirements.
These additional defenses are designed to withstand the cornbined effects of the postulated failure plus normal operating loads plus carthquake loadirg.
Where protection requirements are not met through existing separation, barriers, or shields, piping restraints are provided as necessary ;a meet those requirements.
llestraints are not provided when it can be shown that the pipe break would not cause unacceptable damage to essential systems or com;mnents.
Analyses of postulated pipe break events are performed to identify those safety. ,
related systems and components that provide protective actions required to mitigate, to acceptable limits, the consequences of the postulated pipe break events. In conducting these facility response analyses, the following are among the criteria used to establish the integrity of systems and components necessary for safe shutdown and rnaintenance of the shutdown condition:
- 1. Each high- or mode' ' :crgy fluid system pipe failure is considered l separately as a singli .ed initiating event occurring during normal plant conditions.
- 2. Offsite power is assumed to be unavailable if an automatic turbine generater trip or automatic reactor trip is a direct result of a postulated piping failute.
l
- 3. Piping systems containing high energy Guids are designed so that the effects of a single postulated pipe break cannot, in turn, cause failures of other pipes or components with unacceptable consequences, i
The design criteria define acceptable types ofisolation for safety-re!ated clc.ments and for high<nergy lines from similar elements of the redundant trains. Separation is 1.3.2 4 30-92
i .
EyHEM 80+
accomplished by. i l
- 1. Routing the two groups through separate compartments, or
- 2. Physically separating the two groups by a specified minimum distance, or
- 3. Sepwigh< dvo groups by structural barriers.
The design criteria assure that a postulated failure of a high energy !ine or a safety-telated element cannot take more than one safe:y-related train ort 9f service.
Posilated pipe ruptures are considered in all plant piping systems not climinated by .
leak before breah evaluations, and the associated potential for damage to required systems and components is evaluated on the basis of the energy in the system. Each postulated rupture is considered separately as a single postulated initiating evu :. For each postulated break, an evaluation is made of the effects of pipe whip, jet impingement, compartment pressurization, environmental conditions, rnd flooding.
These evaluations of the required systems and components demonstrate that the protection requirements above are met.
A leak-before-breal evaluation is performed for Class I piping with a diamete of ten inches or greater (i.e, the reactor coolant system (RCS) main loop piping, surge line shutdown cooling and safety injectien lines) and for the main steam line inside containmmt in order to eliminate the dynamic effects of pipe rupture from the design basis. The .: valuation is intended to meet the requirements of 10 CFR 50, Appendix A. Ocneral Design Criterion (GDC) 4. The evalur. tion is performed using the guidelines of,NUREG 1061, Vol. 3. Piping of this sort is designed to be not particularly susceptible to failure from the effects of corrosion, water hainmer or low-and high-cyck fatigue, or degradation or failure of the r:iping from indirect causes.
In addition. a leak detection system as recommended by Regulatoiy Guide 1.45 capable of detecting a leakage rate of less than 1.0 gpm from the primary system is included in the System 80+ plant design.
Inspections, Tests, Analyses, and Acceptance Criteria Table 13.2-1 specifies the inspe::tions, tests, analyses and associated acceptance criteria for evaluating design for the protection of structures, components, equipment and systems against dynamic effects of pipe break and leak before break.
13.2 4-30-92
~
DRAFT TABLE 13.2-1 DESIGN FOR TIIE PROTECTION OF STRUCTURES. COMPONENTS. EOUIPMENT AND SYSTEMS AGAINST DYNAMIC EFFECTS OF PIPE _ BREAK AND LCMK BEFO_RE _ BREAK Inspections. Tests. Analyses. and Acceptance Criteria
?
l Certified Design Commi2 ment Inspections. Tests. Anahses Acceptance Criteria 4 1 i
- 1. Analytical methods for the dy- 1. Review of the certified design 1. Methods shall be in compliance
}' namic and static analysis of piping specification and the certified with - the requirements of the i systems and the corresponding stress report wi2 be conducted to - ASME Code,Section III.
component -- stress analysis shall be confirm that the piping was de-specified in . a ceitified design signed and analyzed in compliance ll specification for each p! ping sys- with ASME Code,Section III re-
, tem. : The dynamic analysis of pip- quirements.
ing systems shall use a suitable dynamic method, such as time his-2 tory or response spectrum method, or ' an equival:r: ' c.ade load method. For the applied method, the key anal ysis parameters shall be addressed.
4 1.3.2 4-30-92 a
4
_ _ _ _ _ _ _ _ _ - - . - - _ _ _ _ _ _ - - - _ - - - - _ - em " - _
i SYSTEM 80+m TABLE 13.2-1 (Continued) .
DESIGN FOR THE PROTECTION OE STRUCTURES. COMPONENTS. EOUIPMENT AND SYSTEMS AGAINST DYNAMIC EFFECTS OF PIPE BREAK AND LEAK BEFORE BREAK Inspg:tions. Tests. Analyses, and Acceptance Criteria ;
Certified Design Commitment Inspections. Tests. AnalVSc5 AceCDiante Criteria.
- 2. All ASME Code Safety Class 1, 2, 2. - An inspecdon of the certified 2. ASME Code, Section IIIlimits that and 3 piping systems which are es- stress report wili be conducted to protect ti e piping av yipe _ sup-sential for safe shutdown shall be assure that none of the stresses or ports against primary 2ress fail-desig ed to assure that they will deflections of the piping system ures shall be compared with allow-malaain sufficient dimensional exceed values which could lead to able values that preclude impair-5 1 stability to perform their required large reductions in the. cross- ment of functional capability. In function following ' application of. sectional flow area. no case shall stresses exceed values
, all loads to which - they will be allowed for Servia Level D in 2 s bjected dung postulcted events ASME Code,Section III.
requiring their safety function.
E 1
I i
i , i I
1.3.2 4-30-9?
4 I
l -
. SYSTEM 80+ * ,
TABLE 13.2-1 (Continued)
DESIGN FOR TIIE PROTECTION OF STRUCI'URES. COMPONENTS. EOUIPMENT AND SYSTEMS AGAINST DYNAMIC EFFECTS OF PIPE BREAK AND LEAK BEFORE BREAK Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance CriMria '
- 3. Essential piping systems, including 3. Inspections of ASME Code, Sec- 3. The essential functions of struc-required. pipe whip restraints, will tion Ill required documents and tures, systems, and components be designed to protect against the the pipe break analysis report will shall not be precluded by the pos-dynamic effects associated with be conducted to confirm that the ' tulated pipe breaks. For those the postulated rupture of high piping system was designed and components required for safe energy and moderate energy fluid analyzed in compliance with re- shutdown, limits to meet the 4
systems. A pipe break analysis quirements that assure postulated ASME Code requirements for j report will be generated to con- pipe breaks will not unduly impact faulted conditions and limits to j firm that the piping system is the safety of the plant. ensure requbed operability shall acceptable for all postulated be met.
breaks. Piping systems that are qualified for the optional leak-
, before-break . design approach (i.e, RCS main ' loop, surge line shutdown cooling, safety injection i lines, and the main steam line in-
, side containment) may exclude de-sign against the dynamic effects from the postulation of breaks in I
high energy piping.
J.
1.3.2 4-30-92 l
SYSTE.Mgo m DRAFT TABLE 13.2-1 (Continued)
DESIGN FOR TIIE PROTECTION OF STRUCTURES. COMPONENTS. EOUIPMENT AND SYSTEMS AGAINST DYNAMIC EFFECTS OF PIPE BREAK AND LEAK BEFORE BREAK Inspections. Tests. Analyses. and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analvses Acceptance Criteria 4.' The piping, its appurtenances, and 4 Inepections will be conducted of 4 Existence of ASME Code required its supports, will satisfy the ASME ASME Code required documents documer.;s and the Code stamps on Class, Seismic Category, and Qual- and the Code stamp on the com- the components shall be reviewed
- ity Group requirements commen- ponents. to confim that the piping and surate with its classification. compone,ats have been designed, j analyzed, fabricated, and ex-amined in accordance with the applicable requirements.
- 5. Redundant safety systems and 5. Visual inspection of plant re- 5. Visual inspecdon report shall be :
subsystems will be kated in dundant safety systems and sub- prepared to provide confirmation.
separate compartments. systems verifies they are located in separate regions.
1.3.2 4-30-92 4
SYSTEM 80+
1.6.3 ANNULUS VENTILATION SYSTEM Design Description
'Ihe Annulus Ventilation System (AVS) is designed as an enginected safety feature and h credited in~ analyzing design basis accidents. The Annulus Ventilation System reduces the concentration of radi'o^aWviTyin the annulus atmosphere between the primary containment and the secondary containment by tiltration, holdup (decay) and
- recirculation before the air is released to the atmosphere.
The Annuhis Ventilation System (AVS) takes the air from above the primary containment dome, filters it and discharges a portion of the air close to the annulus -
floor and a portion of the air to the atmosphere, . Two redundant filtration trains complete with fans, filters, dampers, ductwork, supports and control systems are provided (see Ig;ure 1.6.3 1). Each train is capable of maintaining the air flow within -
specilled limits at maximum and minimum filter pressure drops. The dampers modulate exhaust air as required to maintain the negative pressure greater than 0.25 inches of water gage within the annulus.
Each filter train consists of a moisture climinator, pre 0lter, electric heater, carbon ,
absorber and a HEPA filter before and after the carbon absorber. Each train is sized to remove the fission pnx!ucts released to the annulus following any of the postulated accidnts. Failure of AVS to perform the intended function will be detected by a unit vent radiation monitor, which monitors the activity level of the system effluent.
Electrical and control component separation is maintained between the redundant-trains to meet single active failure criteria although the ducting inside the annulus is shared. Components of the Annulus Ventilation System (AVS) are designed to withstand the post-accident pressure and temperature transients. Components of.the AVS are designed as Seismic Category I equipment.
Each train is normally not operating and is activated by a containment spray actuation signal. Each train is powered by Class 1E power and backup power from the Emergency Diesel Generator. Indication of damper position and fan operating status -
l is provided in the control room. High temperature conditions for each absorber bed L and high and low differential pressures across filter beds are alarmed in the control i room.-
p Inspections, Tests, Analyses and Acceptance Criteria.
l
- Table 1.63 provides the inspections, tests and/or analyses and associated acceptance - .
criteria.
l 1.63 4-30-92 l
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DR, AFT TABLE 1.63-1 ANNULUS VENTILATION SYSTEM Inspections. Tests. Analy: es, and Acccotance Criteria Certified Desien Commitment Inspections. Tests. Analyses Acceptance Criteria
- 1. The AVS is capable of reducing the 1. Documented records reviews and 1. The filter efficiencies and fan concentration of radioactivity in field test results evaluation will be capacity meet the following re-the annulus to an acceptable level conducted to verify specified quirements.
consistent with safety analysis. parameters.
a) 95%(Elemental and Organic iodine) a) Fiher Efficiencies: a) D O P test will be . conducted to 99% (Particulate) measure filter efficiencies.
95%(Elemental and organic imine) b) <18,000 CFM l .
b) Fan capacity will be tested in the
~
99% (Particulate) straight portion of the duct either upstream or downstream of the fan.
b) Fan Capacity:
< 18000 CFM
- 2. A simplified system configuration 2. Inspections of installation records 2. The system configuration is in is shown in Figure 1.63-1.
together with plant walkdowns will accordance with Figure 1.63-1.
be conducted to confirm that the installed equipment is in com- ;
pliance with the design configur-ation defined in Figure 1.63-1. '
1.6.3 4-30-92 t
e
.- . _ _ . _ _. . .m. . _ ._- .- - . _ - . _ _A_.-. _ __ . . _
SYSTEM 80+5 TABLE L6.3-1 (Continued)
ANNULUS VENTILATION SYSTEM Inspections. Tests._ Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria
' 7 The essential fans, dampers, 3. The sptem tests will be conducted 3. The installed equipment can be interlocks and alarms can be after installation to confirm that the ' powered from Class 1E power or supplied power from class 1E electrical power srpply configur- from the Emergency Diesel power with backup power from ations are in compliance with de- Generator.
Emergency Diesel Generator, sign commitments.
- 4. The AVS is designed to Seismic 4. Evaluation of de. characteristics 4. The essential components of the Category I Requirements. t M construction records will be AVS meet Seismic Category I re-performed toevaluate conformance quireme-ts.
to design requirements.
g
- 5. Essential AVS Components are pro- 5. Evaluation of as-constructed AVS 5. The essential AVS components are tected from malfunctions caused by components will be performed protected from the identified
- j. floods, tornados. internal missiles, against design requirements. harards.
pipe breaks and whip jet impinge-ment and interactions with non-seismic systems. ,
I.
- 6. Each train is activated by.a Con. 6. Field tests will be conducted to 6. Each train is activated by a I tainment Spray Actuation Signal verify this actaation of.cach train Containment Spray Actuatioa upon receiving . . a Containment Signal Spray Actuation Signal.
l 1.63 4-30-92
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SYSTEM 80+m N
Unr%ED g { "[E
- q CONTAINVENT ANMULUS NOCLEAf4 m.
ANNEA ca
= 0
,s UPPER ANNULUS ANNULUS FETER # 1 TO UNIT YENT c,
=
g =
=7 ; , '
ANNULUS .1 TO UNfT VENT FLTER # 2 1I , ,
N g g n /
NGlE1 cs s :
J LOWER
' ANNULUS 1F c, ,
.J NOTES: J
- 1. T}eS DAA4PER IS FOR TORNADO PROTECTION AND FOR ISOLAT!ON WHEN FAN IS OFF FIGURE 1.63 ANNULUS VENTIIATION SYSTEM 1.63 4-30-92
-l
DDBCf SYSTEM 80+ = NMI
.;COWAINufNT ANNULUS NtfCLEAR m ANNE 4
'3 s
UPPER ANNULUS e
ANNULUS ETER # 1
' 0 9 toUNirvtNT Om T
U sm# ~
s l
' )
s NcYE 1 NM FILTER 8 2 TO UNT VENT 1F , g IN E1 9 1I C; , ;
2 LOWER ANNULUS C, ;
a d
NOTES:
t
ANNULUS VENTILATION SYSTEM ;
1 i 1.63 4-30-92
SYSTFM 80+ m also be manually initiated from the Main Control Room.
The SITS, which contain borated water pressurized by a nitrogen cover gas, constitute a passive injection system. No operator action or electrical signal is required for op. ration. Each tank is connected to a DVI nozzle by a separate line containing two check valves which isolate the tank from the RCS during normal operation. When
-- RCS pressure falls below SlT piessure, the check valves open, discharging the _
contents of the tank into the reactor vessel. A remotely operated isolation valve in each SIT discharge line is administratively controlled open to assure SIT injection when needed. To further assure SIT availability, each SIT isolation valve receives an open signal upon an SIAS.
The SITS are located inside the containment to minimize their distance from the RCS, but outside the biological shield to protect them from RCS-generated missiles.
Vent valves are connected to each SIT. Venting may be required to lower SIT pressure to shutdown cooling entry pressure following a LOCA. During normal operation, the vent valves are locked closed and power is removed.
The shutoff head and flow rates of the SI pumps were selected to insure that adequate flow is delivered to the reactor vessel to cool the core during LOCAs and non-LOCA design basis events. El pump capacity is also sufficient to remove decay heat during feed and bleed operations. The SIS is designed to provide net positive suction head (NPSH) greater than the pump vendor's required NPSH for all expected fluid temperature conditions during STS operation. Each SIS pu np has a minimum flow recirculation line to the IRWST to ensure sufficient pump flow during operation against shutoff pressure.
The SIS components and instrumentation can be powered from the plant turbine generator (onsite power), and/or plant startup power source (offsite power), or the <
emergency generators (emergency power). Power connections are through at least two independent buses. One independent electrical bus supplies power to two SI pumps and associated valves. A second independent electrical bus supplies power to l the remaining two SI pumps and associated valves.
Power to the SIS hot leg injection valves is designed such that a single electrical failure cannot cause spurious initiation of hot leg injection flow, nor will a single electrical failure prevent initiation of flow through at least one hot leg injection line.
SIS components required for injection of borated water into the reactor vessel are classified Seismic Category I. Mechanical components meet ASME Boiler and Pressure Vessel Code Section III requirements as follows: IRWST, SI pumps, SITS, piping and valves up to the second check valve from the RCS are Class 2; piping and valves from (and inc;uding) the second check valve from the RCS are Chiss 1.
1.6.5 4-30-92
- c
SYSTEM 80+ 5 SIS components and instrumentation which must operate following a design basis event are designed, built, and qualified to operate in the posteent environment in the compartment where the component or instrument is located.
Adequate physical separation is provided between the redundant piping paths and containment penetrations of the SIS to prevent a failure of one train from affecting other trains. Cabling associated with reduwhvutek of Class 1E circuits for the SIS is physically separated to preserve redundance and prevent a single failure from causing multiple channel malfunctions or interactions between channels.
De SIS permits periodic inspection of important components such as injection nonles, piping, pumps, and valves, and periodic functional testing, including the full operational sequence that brings the system into operation.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.6.51 specifies the inspections, tests, analyses and associated acceptance criteria for the SIS.
1.6.5 43092 E -
SYSTEM 80+* I ASME CODE CLASS ]
lil2l -
LOOP 2 HOT LEG SIAS W.%$ 'M S1T DRAFT OPENS l1l2l SIAS W g i b ,j lp [S y (AMPS)
J REACTOR gH { S!AS ___f 3 -
VESSEL DIRECi SIAS gm T l g i, g ,
STARTS Q__ SIT FLOW TEST INJECTION OPENS l1l2l f 3lAS TO1RWST S1 PUMPS
-*-- l ,
(AMPS) e m ,i 0 i nwC =
IN-CONTA!NMENT 3 IN-CONTAINMENT OUT-CONTAINMENT REFUELING WATER STORAGE TANK l h gni l j OPENS SIAS W g g (AMPS)
' l _
VESSEL SlAS gA V RECIRC./
' DIRECT INJECTION CPENS @ SIT l SIAS FLOW TEST TO IRWST StPUMPS 4 l + , (AMPS) .
t
- Sl AS ___ l STARTS.
LOOP 1 HOT LEG W j FIGURE 1.6.5-1 SAFETY INJECTION SYSTEM
- 9 1.6.5 i
SYSTEM 80+m -
TABLE I.6.5-1 SAFETY INJECTION SYSTEM <
Inspections. Tests. Analyses,_and Acceptance Criteria l
4 g
Certified Desien Commitment . Inspections. Tests. Analyses Acceptance Criteria m 1. The configuration of the SIS includes, 1. Perform walkdown inspections of the 1. The as- built SIS configuration as a minimum, the flow paths and as-built SIS configuration. includes the flow paths and.
4 components shown 'in Figure 1.6.5-1. components shown in Figure 1.6.5-1.
- 2. SIS mechanical equipment is built in 2. Inspec procurement and installation 2. SIS equipment has appropriate ASME, accordance w;th ASME Code, Section records to verify components were ' Section III, Class 1,' 2, or 3 cmie III, requirements. manufactured per the relevant ^SME stamp.
requiiements.
II
- 3. All SIS componens requ; red for in- 3. .See Gener'c Se smic and Environ- 3. See Generic Seismic and Emiron-jecting water into' the reactor vessel mental Quali{ication . inspections, tests, m ental . Qualification acceptance are classified Seismic Category I, and and analyses. criteria.
are qualified for the environment in the locations where they are installed.
i i
4 i
i 1.6.5 - 4-30-92
SYSTEM 80+m '
TABLE 1.6.5-1 (Continued)
SAFETY IN_JFJ_HJON SYSTEM .
Inspections Tests. AmilysAand Acceptance Criteria i
n Certified Design Commitment Inspections. Test Analyses Acceptance Criteria i
4 SIS pumps, valves, and controis nec. 4.a) Perform tests to demonstrate the cap- 4.a) With power supplied from normal, essary for injecting water into the ability of the SIS to automatically standby, and craergency AC power
- reactor vessel can be powered from actuate in response to an SIAS gener- sources- low pressuri7er pressure or-l the normal, standby, or emergency ated by low pressurizer pressure and high containment pressure generate an AC power sources. by high containment pressure. Per- SIAS; an SIAS starts the SI pumps, form these tests with power supplied opens the SI di< charge isolation vahes, from the normal (onsite), standby and sends an open signal to the SIT I
(offsite), and emergency (diesel gen- isolation valves.
erator) AC power sources.
- l b) Perform tests to confirm the SITS and b) SIT isolation va%.s open on SIAS and associated valves will respond to an SIT discharge check vahes open wh-SIAS and will discharge SIT water SIT pressure exceeds RCS pt:ssure.
volume to the RCS.
- 5. The SIS operates in the short term 5. Perform SIS functional tests to deter- 5. The analyses results show that the fol-injection mode to provide core cooling mine as-buik system flow vs. RCS lowing acceptance criteria are met for following a LOCA or non-LOCA pressure and time to rated flow after LOCA and non-LOCA events:
event. an SIAS. l a) For LOCA, the acceptance criteria of 10CFR50.46(b)
A j
4:
lm L6.5 4-30-92 4
e
SYSTEM 80+5 -
TABLE L6.5-1 (Continued) h SAFETY INJECTION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria 1
- Certified Design Commitment Inspections. Test. Analyses Acceptantu Criteria Inspect as-built _ drawings and cal- b) For non-LOCA transients, the accep-culate the volume of unborated water tance criteria of Section 15 of in each SI line prior to an actuation, NUREG-0800, Rev. 3.5.
the IRWST volume, SIT volume, SIT inner diameter, and SIT nozzle ele-vation above the DVI nozzles.
From the results of tue SIT tests in ITA 4.b), calculate the SIT discharge line K-factor.
Compare as-built data from these tests 4) and calculations with the data used in analyses of LOCAs and non-LOCA events. which require SIS operation.
Confirm the analyses results. .
4
- 6. The Sb operates in the ' long term 6. Manually. align the SIS for long term 6. The SIS can be manually aligned for post-LOCA injection mode to provide cooling using simultaneous DVI and long; term cooling using simultaneous
, flow circulation 'through the core. hot leg injection. Perform tests to DVy and hot leg injecticn.
determine the hot leg /DVI flow split with four SI pumps running. Com- The acceptance criteria of pare as-built data from these tests and 10CFR50X,(b) ' are mer.
1 1.6.5 4.-30-92 i
s
SYSTEM 80+* .
TABLE 1.6.5-1 (Continued.J SAITIY INJECTION SYSTEM Inspections. Tests. Analyses.~and Acccotance Criteria i Certified Design Commitment Inspections. Test. Analyses . Acceptance Criteria the tests and calculations 'in ITA . 5 with the data used in the analysis of post-LOCA long-term cooling. Con-firm the analysis results.
- 7. The SI pumps have sufficient NPSH 7. . Perforrn tests to measure suction head 7. Minimum pump NPSH available, as
- during all postulated operating 4 avpliable to each SI pump. . Suction determined based on as-built . con-conditions, wi'l b- taken from the IRWST under ditions, meets or exceeds the pump ma-imam flow conditions in the com- designer's NPSH requirements. .
' tid suction lines (i.e containment spray pumps also running when. test-
{ ing the two SI pumps with shared i
suction lines). Correct the measured
, suction head for the IRWST minimum
-level attainable after an SIAS and the
- maximum IRWST fluid temperature.
1 .
, .8. The' SIS includes a minimum flow 8. The as-built ' system configuration and 8. Min' um flow recirculation, as deter-l ' recirculation path for each SI pump to installation will be iraspected and mim based on as-built conditions, a protect the pump from overheating. minimum flow recirculation verified meejs or exceeds the pump desimer*c by analyses, er a minimum . flow requ!rements.
measurement test will be performed i
)
ii i 1.6.5 4-30-92
SYSTEM 80+5 .
AE TABLE 1.6.5-1 (Continued) dry"""T" g 1
SAFETY INJECTION SYSTEM Inspedions. Tests. Analyses._and Acceptance Criteria m
I i Certified Design Commitment Inspections. Test. Analyses cceptsnee Criteda
' I
- 9. The SIS provides automatic indication 9. Perform tests to verify operation of C. SIAS bypass and SIS status inoperable when SIAS is bypassed and when SIS the ~ STAS bypass and SIS .moperable are automatically indicated.
status is inoperable. indications.
- 10. SIS redundant ' ' piping trrdas and 10. Visual inspections will be performed 10. Four quadrant separation is provided containment penetrations are against construction drawings to for the SIS piping . trains ' and
, physically separated such that failure verify physical separation of' SIS- containment penetrations.
of one train will not ' cause failure of piping trains and containment other trains. penetrations.
I
- 11. The SIS permits periodie inspection of 11. Perform visual inspection of access- . 11. Access is provided for inspection of -
4 important components su,:h as injec- ibility for periodic inspections of the SIS injection nozzles, piping, pumps, tion nozzles, piping, pumps, and . SIS injection nozzles, piping, pumps, and valves.
, valves. and vahes.
?
i 12. The. SIS permits functional testing of 12. Demonstrate that ' the SIS full opera- 12. The initiation signals generate an i the full operational seqecace that tional sequence can be tested by per- SIAS, an SIAS.. actuates the Si pumps brings the system -into operation. forming tests to show that a low pres- and associated valves, and the SIS de-surizer pressure :ondition or high livers flow through either the SI lincs j containment pressure conditio2 gen- to the RCS or the test lines to the IRWST.
d 1
. L6.5 4-30 5
. SYSTEM 80+m .
TABLE 1.6.5-1 (Continued)
. SAFETY INJECfl0N SYSTEM Jnsp Mions. Tests. Analyses. and Acceptance Criteria Certified Desirs Commitment Inspections. Test. Analyses Aca-stance Criteria I
an SIAS, an SIAS actuates the Si pumps and associated valves, and the i SIS delivers flow to tne RCS via the SI lines or to the IR'WST ' via the ' test .
lines. These tests may be combined -
with ITA 4.a).
Il
<n.
1.6.5 30-92
- w
l l
l- . .
SYSTEM 80+m 1.7.1 PLANT PROTECTION SYSTEM =
Design Description
'Ihe System 80+ Plant Protection System (PPS) is a warning and trip system where
- initial warning and trip decisions are implemented with software logic installed in programmable digital devices. The system provides the limit hgic, matrix logic and l
initiation circuits for both the reactor trip functions and the engineered safety features -
functions for the System 80+ Advanced Light Water Reactor design. In the System 80+ design the reactor trip function is implemented by process instrumentation, the PPS and the reactor . rip switchgear. S!milarly, the engineered safety features actuation is implemented by process instrumentation, the PPS, the Engineered Safety l Features. Component Control System, motor starters and actuated devices.
l .
l The primmy functions of the PPS are to: (1) make the logic decisions related to warning and trip conditions of the individualinstrument channels, and (2) make the decisions for reactor trip and initiation of engineered safety features based on coincidence of instrument channel trip conditions.
Reactor Trio Actuation Function i.
The Reactor Trip actuation function of the System 80+ PPS monitors selected plant conditions and automatically effects rapid reactor shutdown (reactor trip)if monitored I conditions exceed safety system setpoints. Initiation of a reactor trip for the following conditions has been demonstrated to meet PPS functional requirements: _
Variable Overpower l High Logarithmic Power Level i High Local Power Density Low Departure from Nucleate Boiling Ratio High Pressurizer Pressure Low Pressurizer Pressure Low Steam Generator Water Level ~
Low Steam Generator Pressure
.High Containment Pressure High Steam Generator Water Level low Reactor Coolant Flow The PPS includes limit logic for simple process-value to set point comparison; the limit logic is implemented in devices referred to as bistable processors. In addition .
the PPS includes limit logic for complex calculations of the departure from nucleate -
i boiling ratio and the local power density, which is implemented in devices referred to as Core Protection Calculators.
1.7.1 4-30 -
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SYSTEM 80+5 1.I1(AF 44 If the setpoint for a trip condition varies with power, the setpoint change is performed automatically within the PPS For each tnp condition, pre-trip alarms are prcvided. These can provide the operator with an opportunity to take control actions to avoid the trip limit condition.
As shown in Figures 1.7.la and 1.7.lb, the PPS includes the following elements:
- p. bistable trip, local coincidence logic, reactor trip initiation logic, ESFJwfinn initiation logic, and automatic testing of PPS logic. The PPS i5 dhided into four channels, and each channel has a pair ofl^able processors, refened to as the Plant Protectior. Calsulator (PPC), and a Core Protection Calculator. The bistable trip processors generate trips based on the measurement channel digitized value exceeding digital setpoint. The trip outputs of the bistable processors and the core protection calculators are sent to the local coincidence logic processors. Each local coincidence logic processor receives four trip signals, one from its associated bistable processor or CPC in the channel and one from each of the equivalent bistable processors or CPCs located in the other three channels. 'Itse coincidence processors evaluate the local coincidence logic based on the state of the four like trip signals and their respective bypasses. A coincidence of two-out-of-four like trip signals is required to generate a r: actor trip or ESF initiation signal. The fourth channel is provided as a spare and allows bypassing of one channel while maintaining a two-out-of-three system.
The coincidence signals are used in the generation of the Reactor Trip Switchgear System (RUS) or Engineered Safety Features-Component Control System (ESF-CCS) initiation. Upon coincidence of two signals indicating one of the trip conditions, the PPS initiates actuation of the Reactor Trip Switchgear (RTSG). The reactor trip switchgear breakers interrupt power to the Control Element Drive Mechanism l (CEDM) coils, allowing all CEA's to drop into the core by gravity. The RTSG can l be tripped manually from the Main Control Panel (MCP) and Remote Shutdown Panel (RSP) independent of the PPS bistable and coincidence processors.
En_gi.ncered Safety Features Actuation Function The Engineered Safety Features actuation function of the System 80+ PPS monitors selected plant cor.Jitions and automatically initiates enginected cafety features actuation signals (ESFAS) to the Engineered Safety Features-Component Control-Systems (ESF-CCS) when monitored conditions exceed safety system setpoints.
The ESF CCS actuates the plant's ESF System components (for example pumps, valves, etc.) The System 80+ PPS initiates the following ESF actuation signals:
! Safety Injection Actuation Signal j Containment Isolation Actuation Signal Containment Spray Actuation Signal Main Steam Isolation Signal Emergency Feedwater Actuation Signal-1 1.7.1 4 30-92
- F r . --- ,- - - -
SYSTEM 80+
Emergency Feedwater Actuation Signal-2 Conditions which can initiate one or more of the above ESF actuations are:
Low Pressurizer Pressure Low Steam Generator Water Level hmotcam Generator Presst re High Containment Pressure High Steam Generator Water Level Low Reactor Coolant Flow High High Containment Pressure Using the same method as desenbed above for the scactor t;ip conditions, these ESF actuation conditions are monitored by the PPS bistable proecssors and trip outputs are sent to the local coincidence processors for de+ection af two or more tripped conditions. The functional logic used in the PPS tra geneste cach of the ESF initiation signals is shown in Figures 1.7.lc,1.7.1d, L7Je end 1.7.lf.
Manual initiation of the ESF actuations. indgendent of the histable and coincidence processors can be performed at either the main Control Panel (MCP) or the Remote Shutdown Panel (RSP). The PPS inter 6ce at .he ML/ provides for manual initiation of all ESF actuation signals. The PPG interface at the RSP provides for manual initiation of the Main Steam Isolation Signal. nc ESF-CCS interfaces provide for initiation of all ESF functions on a trtin or component basis at either the MCP or RSP.
PJS Divisional Senarition and Isolation The PPS is a four diMn system which is designed to provide reliable single failure pr apabilit,, to automatically or manually initiate a reactor scram while maintaining l protection 7ainst unnecessary scrams resulting from single failures in the PPS. All functions et the PPS and the components of the system are saiety-relateri The PPS and the c!:ct,ical equipment of the system are also classified as Safety Class 3, Seismic Category I and as IEEE electrical category Class 1E.
PPS components nnd equipment are separated or segregated from process control system circuits and functions such a to minimize control and protection system interactions. Any necessary interlocks from the PPS to cont-ol systems are through isolation devices.
The PPS remains single-failure proof even when one entire dMsion of channel l sensors is bypassed und/or when one of the four automatic PPS trip bistable processors or CPCs is out-of-service. All equipment within the PPS is designed to fail into a trip initiating state or other safe state on loss of power or input signals or disconnection of portions of the system. The system also includes trip bypasses and 1
1.7.1 4-30-92
sv - DRAFT isolated outputs for display, annunciation or performance monitoring. PPS interfaces to the Power Control System, Discrete Indication and Alarm System, Data Processing
! System, PPC Operator Modules and the Mainte- and Test Panel are electrically I
isolated so that no malfunction of the associated equipment can functionally disable any portion of the PPS. The PPS related equipment is divided into four redundant divisions of sensor (instrument) channels, trip logics and trip actuators, and manual
.- scram controls and scram logic circuitry. %e automatic and manual scram initiation logic systems are independent of each other and use diverse methods and equipment to initiate a reactor scram. Once a reactor trip has been initiated, the breakers in the reactor trip switchgear latch open, assuring that the intended fast insertion of all control rods into the reactor core cannot be compromised bv any action of the normal power control system. After all of the trip conditions have oecn cleared, deliberate operator action is required to manually reclose the trip breakers.
Figure 1.7.lb shows the PPS divisional separation aspects and the signal flow from the process instrumentation to the individual channels for initiation of protection system functions. The PPS includes calculators, logie, and other equipment necessary to rnonitor selected plant parameters and plant conditions and to effect reliable and rapid reactor shutdown (reactor trip) and actuate appropriate ESF System components if monitored conditions approach specified safety system settings. Four measurement channels with electrical and physical separation are provided for each parameter used in the direct generation of trip signals, with the exception of Control Element Assembly (CEA) position which is a two channel measurement. Trip bistable settings associated with initiati on of reactor trip are selected to protect the core fuel design limits and the Reactor Coolant System (RCS) pressure boundary for Anticipated Operational Occurrences, and also to provide assistance in mitigating the consequences of accidents.
Basic System Parameters are:
- a. Number of independent divisions of cquipment 4 l b. Minimum number of sensors per trip variable 4 l (at least one per division) l
- c. Number of automatic trip systems (one per division) 4
- d. Automatic trip logic used for plant sensor inputs 2-outef-4
- e. Number of separate manual trip systems 4
- f. Manual trip logic 2-out-of-4
- g. ESF Actuation Logic Selective 2-out-of-4 1.7.1 4-30-92 g -,
~.- ~ -
SYSTEM 80+ m PPS Interfaces and Testine As shown in Figure 1.7.la, the PPS interfaces with the following:
Class 1E safety process instrumentation, including:
Reed Switch Position Transmission (RSP!Q.(fotEF.A positicia).
Auxiliary Process Cabinet (APC) (for Pressurizer Pressure, RCS Hot Leg Temperature, RCS Cold Irg Temperature, RCP Speed Pulses, and Ex Core Neutron Flux Power).
Manual Actuation Signals for Reactor Trip at the Main Control Panel and the Remote Shutdown Panel.
Manual Actuation Signals for ESF Systems at the Main Control Panel and the Remote Shutdown Panel.
Reactor Trip Switch Gear (RTSG).
Enginected Safet" Feature Component Control System (ESF-CCS).
The Interface and Test Processor portion of the PPS provides optical cable data link interfaces to the following:
Power Control System Discrete Indication and Alarm System Data Processing Systern Plant Protection Calculator Operator's Module in the MCP Plant Protection Calculator Operator's Module in the RSP PPS interfaces for operator interaction, alarm annuncietion and testing (manual and automatic) are shown in Figure 1.7.1g.
l The local and main control room PPS operator's module (one per channel) provides for entering trip channel bypasses, operating bypasses, and variable setpoint resets.
These modules also provide indication of status of bypasses, operating bypasses, bistable trip and pre-trip. 'Ile local operator module provides the man-machine interface during manual testing of bistable trip functions not tested automatically.
The Interface and Testing Processor (ITP), one per channel, communicates with the bistable trip processors, coincidence processors, operator's modules, ESF-CCS, RTSS and ITP's in the other threc channels to monitor, test and control the operational state of the PPS. It also provides selected PPS channel status and test results information to the Data Processing System (DPS), and Discrete Indication. nd Alarm 1.7.1 4 30-92
~
M SYSTEM 80+ m Ph Ul
- System (DIAS),
Inspections, Tests, Analyses, and Acceptan: Criteria Table 1.7.1 1 provides a definition of the visual inspections, tests and/or analyses, together with associated acceptance criteria for the PPS.
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. . PROCESS PROCESS PROCESS PROCESS INST. INST. INST. INST.
^ ^ ^ ^
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PO*ER CONTROL SYSTEM l 190URE 1.7.1g PPS FUNCITON INTERFACE AND 'IPS11NO DIAGRAM 1.7.1 13 - 43092
i -
SY5 I EM 80+*
TABLE I.7.1-1 PLANT PROTECTION SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Inspections. Teds. Ama!vses Acceptance Criteria !
Certified Desirs Cosamitment L '- PPS safety-related software, which L See Generic Software Develep- 1. See Generic Software Develop-is utilized in effeuing individual ment verification actinties (ITA). ment Acceptance Criteda (AC).
sensor channet trip decidocs and trip system coinciaiice trip decisions, has been dewloped and verified, 'he firmware imple-mented and validated and then
. integrated with hardware; all according to a formal documented Pl an.
- 2. Critical . parameter trip setpoints - 2. See Generic Setpoint Methodology 2. See Generic Setpoict Methodokg are based upon values used in an- verification activities (ITA). Acceptance Criteria (AC).
l}. alyses of abnormal operational oc-currences. L Docmnented instru-ment ' setpoint . methodokg has been used to account for uncer-tainties (such as instrument in-accuracies and drift) in order to establish PPS related setpoints.
- 3. PPS equipment is designed to be 3. See Generic EMI/SWC Onafifi- 3. See Generic EMI/SWC Oualifi-protected from effects of noise, cation wrification activities GTA}. cation Acceptance Criteria (AC).
such as electromagnetic inter-ference (EMI), and has adequate surge withstand capability (SWC).
I
\
4-30-92 1.7.1 -
SYSTEM 80+w TABLE 1.7.1-1 (Continued)
PLANT PROTECTION SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Certified Design Comunisment Inspections. Tests. Analyses Acceptance Criteria
- 4. PPS quipment is qualified for 4 See Generie Equipment Qualifi- 4 See Genenc Equipment QuaEfi-seismic loads and appropriate en- cation verification actinies (ITA). cation Acceptance Criteria (AC).
vironment for locations where in-stalled.
- 5. PPS components and equipment 5. Visual field inspections and anal- 5. PPS equ(.nnent installation accep-are kept separam from equipment yses of relationship of installed table i; inspections, analyses associated with process control PPS equipment and of installed and/ or tyts confirm that any fail-systems. equipment of interfacing process are in p ocess control systems can control systems (and/or tests ' of not prevjnt PPS safety functicas.
interfaces) to confirm appropriate isolation methods used to satisfy separation and segregation re-
) quirements.
- 6. Fail-safe failure ~ modes - result 6. Fic!d tests to confirm that trip 6. Acceptabic if trip condiions upon loss of power or discon- conditions and/or bypass hihibits and/or bypass inhibits result epon nection of components. result upon loss of power or dis- loss of power or disconnecion of connection of components. portions of the PPS.
- 7. Provisions exist to liinit access to 7. Visual field inspections of the in- 7. The PPS hardware /firmware will trip setpoints, calibration c.atrols staHed equipment will be used to be considered acceptable if appro-and test points. confirm the existence of approp- priate methods exist to enforce ad-riate administrative coratrols. ministrative control for acxess to sensitive areas.
- n 1.7.1 4-30-92 i
a - ,t . . .
SYSTEM 80+5 TABLE 1.7.1-1 (Continued)
PLANT PROTECTION SYSTEM Inspections. Tests. Analysee and Accepta:sce Criteria Inspect ~ ens. Tests. Analyses Acceptance Criteria Certified Desimo femmitment
- 8. The four redundant divisions of 8. Inspections of fabrication and in- 8. InstaHed PPS equipmer.t will be PPS equipment 'are independent staHation records and construction determined to conform to the doc-from each other except in the area drawings or visual field inspec- - umented description of the design of the required coincidence of trip tions of. the instaI!cd PPS equip- as depicted in Figures L7.lb, c. d.
~
logic decisions ,. and are both elect- ment will be used to confirm the e and f.
rically and physically separated quadruple redundancy of the PPS from each other. and the electrical and physical ;
separation beturen disisions. ,
- 9. It is possible , to conduct veri- 9. , Prt. operational tests will be con- 9. The ins;aHed reactor protection fications of PPS operations, both ducted to confirm that system system ' configuration, controls, on-line and offline, by means of - ' testing such as trip bistable tests, power s'mrces and installations of-
. channel functional tests, channel interfacir'g systems supporu the a) -individual instrument channel
- functional tests, b) trip system calibrations, minhe logic tests, PPSlogic system functional , testing i
functio-ud tests and c) total system reactor trip initiation logic tests, and the operabiEty verification of functional tests.
manual trip test, and c@cred d-sign as foDows:
safety feature ' initiation and ac-tuation logie tests ' can be per- a. InstaBed PPS hardware /Trrmware formed. 'Ihese tests wiH invohr in:tiates trip condkions in aH four simulation of PPS testing modes of PPS automatic trip systems upon operation. Interlocks associated coincidence of trip conditions . in with the restor mode switch posi- two or more instrument c8:annels tions, and with other w aivu.1 associated with the same trip var-and maintenance bypasses or test iable(s).
switches will be tested and annnn-ciation, display and logging fune-tions will be confirmed.
4-30 92 L7.1
SYSTEM 80+*
a e -
TABLE 1.7.1-1 (Continued) e hh PLANT PROTECTION SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Certified Desirs Con mitsment inspections. Tests. Analyses Acceptance Criteria
- b. InstaDed system initiates trip (both under voltage and shunt trip) upon coincidence of trip conditions in two or more of the four PFS auto-matic trip systems.
- c. InstaHed system "mitiates trip om-diuons two (of four) Y manusi trip switches are operated.
- d. Instaned system initiates appro-priate ESFAS actuation signal upon cMW of appropriate trip conditions in two or more of I the four PPS automatic trip sys-tems.
- c. Irat.,11ed system initiates approm riate ESFAS actuation signai if two of the four associated ESF manual initiation switches are operated.
- f. Trip system and ESF icitiation (automatic and manuai) trip om-
% ditions seal-in and protective ad-uation signals are maintamed.
3 1.7.1 - ' I 4-30-92 o
' " ~~' ' " ' ~ ' ' ' ' -
L____ .-..s_-..m_.....-....
t SYSTEM 80+ *
- r i i t i
' 1 TABLE 1.7.1-1 (Continued) i i PLANT PROTECTION SYSTEM t I
Insocctions. Tests. Analyses and Acceptance Critena e
- ' I i
Certified Desis Commitment Insocctions. Tests. Analyses Acceptance Critena i i
- g. InstaIIed system provides isolated [
j status and control sirak to data i logging, display and annunciator systems.
l' l' h. Installed system demonstrates op- '
erational hterlocks (i.e, trip ,
inhibits or permissives) required ,
j for. different condit2ons of reactor l t- operation. ;
l 10. The PPS design provides . prompt 10. ' Preoperational' ' tests wi!I be con The PPS hardware /firmware re-
{} '
protection - against the onset and ducted to measure the PPS and sponse to initiate reactor scram consequences of events or con- . supporting systems response times and Engineered Safety Feature ac-ditions that th eaten the integrity . to: (1) monitor the variation of the tuation will be considered accer-of the fuel barrier. selected processes; (2) detect when table if such response is demon- :
i trip setpoints have been exceeded; strated to be sufficient to assure !
'and, (3) execute the subsequent that the specified acceptable fuel protection actions when coin- design limits are not exceeded. j
]- cidence of trip conditions existf l i !
- t 1
f
!- -l 1
1.7.1 - 3 4-30-92 ;
I i
_ _ _ . _ _ _ _ x __ _ _
j
SYSTIIM 80+ m 1.9.1 SPENT FUEL STORAGE Design Description ne spent fuel storage racks are dealgned to support and protect spent fuel assemblics and assure a geometrically safe <.onfiguratl5n'iiDirespect to criticality. The spent fuel storage rack arrangement is shown in Figure 1.9.1 1. De spent fuel storage rack arrangement is shown in Figure 1.9.12. There are sufficient spent fuel storage racks -
to provide for the licensed storage capacity.
De spent fuel storage racks are designed to maintain a neutron multip!! cation factor less than Keff=.95 for normalloadings including seismic events and impact oue to the drop of a fuel assembly plus its handling tool.
The spent fuel storage racks arc designed to meet the stress acxptance criteria of the ASME B&PV Code,Section III, Subsection NF, Class 1.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.11 specifics the inspections, tests, analyses and associated acceptance criteria for spent fuel storage.
1.9.t .1 -4 30-91
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SYSTEM 80+=
DRAFT TABLE 1.9.1-1 SPENT FUEL S1T) RAGE Inspection. Tests. Analysis and Acceptance Criteria Certified Desis Coasmitment laspectma. Test. Analysis Accustance Criteria L The rack arrangement prm'es L Visual inspection of the rack va- i- L Visual inspection data report pro-storage locations for the licensed fies rack size and capacity. sides verification.
spent fuel storage capacity.
l l
- 2. ' The rack is constructed to ASME 2. Examine certification data reports. 2. Certification data reports show Code NF. compliance with ASME Code, Subsection NF.
j
- 3. Stress limits ' are met for SSE loads 3. Examine the design report 3. Design report prmides actual and drop accidents. stress and shows resuks within allowable Emits.
,and measure the pitch between < 35. Pitch and separation within cells - and separation between drawing limits.
modules.
1.9.1 - 4-30-92
,YSTFM M+
1.9.2.2 COMPONENT COOLING WATER SYSTEM Design Description caoling water system The that is Component designed to remove Coolingheat fromWater the plant's System (CCWS) safety related a is a c!osco looy'nis no components and heat exchangers as required during normal operation, shutdown cooling, refueling, and design basis accident conditions. 'Ile system, in conjunction with the Station Sarvice Water System (SSWS) and the Ultimate Heat Sink (UHS),
is capable of removing heat from the essential components and heat exchangers to en ure a safe reactor shutdown and cooling following postulated eccidents.
The Component Cooling Water System is an intermediate cooling water system between the Reactor Coolant System (RCS) and the Station Service Water System.
The CCWS provides protection against station service water leakage into the Reactor Coolant System. The Component Cooling Water System also provides a barrict to >
the release of radiological contamination into the environment via the Ultimate Heat Sink.
The CCWS has two 100% capacity divisions (see Figure 1.9.2.2). Each divi:; ion is connected to its corresponding SSWS division through the component cooling water heat exchangers. Each divnion has 100ro heat dissipation capacity to obtein safe cold shutdown.
Each division of the CCWS includes two component cooling water heat exchangers, a component cooling water surge tank, two componerit cooling water pumps, piping, valves, controls, and instrumenation. There are no cross connections between the two divisions. A single failure of any component in the CCWS will not impair the '
ability of the CCWS to meet its functional requirements, The temperature of the component cooling water leaving each component cooling water heat exchanger is regulated by a component cooling water bypass control valve.
The component cooling water pumps '.w the capability to supply the plant components and heat exchangers during tn i, unit operation, during unit cooldown, during refueling, and during emergency situations as required.' Inherent system logic is provided to ensuic that flow requirements are met and that a minimum Dow path is provided as necessary.
i The component cooling water surge tanks ensure that required NPSH is provided for l the component cooling water pumps. Each tank allows for expansion and contraction of fluid in the system due to temperature changes and provides a means to moultor fluid leakage into and out of the system. Fluid losses are accommodated by the surge 1.9.2.2 4 30-92
SYSTEM 80+ w IN b b tank volumes. System venting and filling are accomplished by the surge tanks. He tanks are also provided with an adequately sized overflow line to protect against overpressurization. In order to maintain surge tank volume in the event of a break n a system cooling loop composed of non. nuclear safety class piping, levelindications and controls are provided to isolate these portions of the system in the event of abnomRt.Smage tank level.
System water chemistry is controlled for the prevention of long term corrosion. The capability is provided to sample the water, and if acquired, the pil can be adjusted by the addition of chemicals. Org nic fouling and inorganic buildups are controlled by
, proper water treatment. Radiation monitors and system sampling are provided to detect radioactive contamina' ion so that the contaminated water can be processed as liquid waste.
Makeup water to the CCWS is normally ::upplied by the Deminera!! zed Water Makeup System (DWMS). If the DWMS is unavailable, such as during an accident a safety related backup makeup line of Scismic Category I constructica is provided from the Station Service Water System. A removable spool picce is placed on this
. line to prevet the inadvertent addition of station service water.
Instrumentation and controls are provided to auquately monitor and control the CCWS. All non. safety related instrutnentation and controls are designed such that any failure will not caut.e degradation of any essential equipment function. He CCWS instrumentation facilitates automatic operation, remote control, and continuous indication of the system parameters local!y and in the control room.
Control room process indications and alarms are provided to enable the operator to evaluate the CCWS performance and to detect malfunctions.
Each division of the CCWS consists of essential and non-essential cooling loops. The essential cooling loop piping and components are designed in accordance with Safety Class 3 requirements. Containment isolation valves and containroent penetration piping are designed in accordance with Class 2 requirements. All pneumatic valves in!! to pre dctermined safe positions upon the loss of instrument air. All Major CCWS pumps, he at exchangers and surge tanks are designed as a minimum to meet ASME 111 requirements.
The essential portions of the CCWS are designed as Scismic Category I snd are contained in Scismic Category I structures. All essential CCWS components are protected from floods, tornado missile damage, internal missiles, pipe breaks and whip, jet impingement, and interaction with other non-seismic systems in the vicinity. Also, failure of the non-caential portions within the CCWS itself does not cause flow degradation to safety related components.
1.9.2.2 4 30-92
' I
SYSTEM 80+
?OOCy t Each division of the CCWS receives power from its associated Class IE Auxiliary Power System, in the event of a loss of offsite power, the system receives power from the emergency diesel generators.
Inspections, Tests, Analyses, and Acceptance Criteria Tatts 1.9.2.2 provides the inspections, tests, and/or analyses and associated acceptance critcria.
t l
l 1.92.2 -3 4 30-92
.._. . . _ . _ _. . . . _ . _ . , . . _ , . . _ -. 7 . -, . . _ . _ . . . . _ . . . - , . . . _ _ . - - . _ - - - .
SYSTEM 80+m El' ~ ' ~
TABLE L9.2.2 COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desien Commitment Inspections. Tests. Ama'vses Acceptance Criteria
- 1. , The simplified system configuration 1. Inspections of installation records 1. The system configuration is in ac-is shown in Fqure 1.9.2.2. together with plant walkdour.s will cordance with Figure 1.9.2.2.
be conducted.
- 2. Pneumatic control valves fail to - 2. System testing will be conducted to 2. The affected vahrs respond to a their failed safe positions upon the simulate loss of instrument air con- loss of instrument air as designed.
loss of instrument . air. ditions to verify the response of '
valves hav:ng an mstrument air
' supply. l l 1
- 3. Cooling loops composed of non- 3. System tests will be conducted to 3. The affeced valves isolate the nuclear safety ; chss component siraulate . an abnormally low surge eling loops composed of non-cooling water piping are isolated on.. . tank condition ta verify the res- nuclear safety class piping in res-an abnormally low surge tank lewL ponse of valves used to isolate these ponse to an abnormally low surge partions of the system. tank levet ,
- 4. The CCWS is capable ' cf accom- 4. System tests will be conducted after 4.' The CCWS is powered with Class plishing its nuclear safety functions installation to confirm that the IE power from the plant's normal with Class ' IE powa- from the electrical - power supply config- and emergency power sources.
plant's normal and emergency' urations are in compliance with power sources. design.
W I.9.2.2 -
4-30-92
(
SYSTEM 80+m .
v . ex .s f p g Wsii
_ TABLE 1.9.2.2 (Continued)
COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criten]
- 5. Cor.tainment iwlation v.Jves and 5. Review of plant records are made 5. Containment imlation and in-penett:. tion piping meet the spec- to verify compliance. containment piping meets ASME ified ASME Code Class 2 require- Code Class 2 requirement.s ments.
- 6. The essential portions of the CCWS 6. Evaluation of design characteristics 6. The essential components of ti,e are designed to Seismic Category I and co. struction records will be CCWS meet Seisc-ic Category I requirements. performed to evaluate conformance requirements.
to design requiremen:s.
} 7. Tsential CCWS components are ~. Evaluation of the CCWS s apos- 1. The essential components are pro-protected from malfunctio.a caused ents will be performed against tected from the identified hazards.
l by floods, tornados, internal design requirements.
l missiles, pipe breaks and whip, jet l impingement, and interactions with
! non-seismic systems.
i e
1.92 2 4-30-92 I
f I nM
r i3 w> !
SYSTEM 80+
- r ,
DEMINEA AUZED w ATEa SERVICE WATER SYSTEr1 MAKEtJ) CCw MAKEUF SYSTEM l l CCW SURGE l TANK ,,,_"V SSW 5W Cs CCW Pmp A
"hy YID _.,d,b '
I bl &CCW CCW AETURN SUPptY f RC4 s CCW PL'Mp
$* EAT LO ADS O rOatAT q LOADS I i Y CCW F
55w lli SSw DiviS:ON t vi vemvN+t devAw Alius I3 NN5 l D! VISION 2 DEMINEA AU ZED W ATE 4 5 STEM MAKEUP CCW MAKEUP SYSTEM i l
- SURGE ,_ CCW
- X TANK S5W 5w LJ CCW PUMP Z
b/ [] CCW CCW RETUcN FROM CCw pmp SUPPLY HEAT LOADS TO HEAT i
- 42 , , 4 CE# $6T ALL piptNG AND CWCN(Mis "I 5+owM AAE CLASS 3 ASM CCDE.
SSW SSW uuttss oTwpest hcTtt er A FIGURE ISS2 ctAss entAx sm COMPONENT COOLING WATER SYSHBi
'm ID 4-30-92
$_YSTl!M 80+
1.9,6 COMPRESSED AIR SYSTEMS I Design Dese.iption
%e Compressed Air Systerns are non-safety related rystems consisting of the Instrument Air System (IAS), the Station Air System (SAS), and the Breathing Air System (BAS). The Inst'ument Air System supplies compressed air to air operated instrumentation, controls, and valves. The Station Air System supplies compressed air for air operated tools, miscellaneous equipment, and various maintenance purposes. The Breathing Air System supplies compressed air to various locations in the plant, as required, for breathing protection against airborne contamination for-personnel performing certain maintenance and cleaning operations.
The Compressed Air System is not required to achieve a safe reactor shutdown or to mitigate the consequences of an accident. Loss ofinstrument air due to a failure of the instrument air system during an accident, loss of offsite power, or station blackout
. (i.e. complete loss of all AC power) will cause all of the pneumatically operated safety related components to fall to their predetermined safe position. Therefore, failure of the IAS will not prevent any safety-related component or system from performing its intended safety functions.
Each of the four instrument air supply trains is composed of an air intake filter / silencer, an air compressor with intercooler(s), an air receiver, a dryer / filter train, and associated piping and valves as shown in Figure 1.9.61. The IAS is capable of supplying instrument air quality compressed air.
The SAS is composed of two parallel,100% capacity trains of equipment. Each station air supply train consists of an intake filter / silencer, a compressor, an air receiver, a dry::r/ filter, and associated piping and valves as shown in Figure 1.9.6 2.
The BAS !s composed of two parallel,100% capacity trains of equipment as shown in Figure 1.9.6 3. Each breathing air supply train consists of an intake filter / silencer, a breathing air compressor, an air receiver, a breathing air purifier, and associated piping and valves.
The Compressed Air Systems are Class NNS (Non Nuclear Safety) with the exception of the containment isolation valves and associated piping which are Safety Class 2 (ASME Code Class 2) and Scismic Category I.
The normal power source for the Compressed Air Systems is the non. class 1E AC Power Source. The nuclear safety-related electric valve actuators on the Compressed Air System containment isolation valves receive power from the Class IE Alternate AC Source Standby Power Supply to accommodate a loss of offsite power.
4 1.9.6 1- 4 30-92 2_ - - __
q
- - + - DRAFT Inspections, Tests, Analyses, and Accep.'.ance Criteria ,
Table 1.9.6 provides the inspections, tests, and/or ar alyses and their with associated acceptance criteria.
1.9.6 2 43092 F
SYSTEM 80+w TABLE 1.9.6 COMPRESSED AIR SYSTEMS Inspections. Tests. Analyses and Acceptance Criteria Inspections. Tests. Analyses Acceptance Criteria Certified Desirs Commitment
- 1. The simplified system' configuration 1. Inspections of instaIIation records L %c system configurations are in for the IAS is as shown in Figure together with plant walkdowns wiH accordance with Figures 13.6-1, 13.6-1; for the SAS in Figure 13.6- be conducted to confirm that the in- 13.6 2, and 13.6-3.
2; for the BAS in Figure 13.6-3. 'staHed equ;pment is in compliance with the design conf:gurations.
- 2. ' The IAS can ' be powered from the 2. System tests will be conducted after 2. He instaHed equipme-a can be Non-Class - IE Alternate AC Source instaBation to confirm that the elec- powered from the Non-Class IE Al-
}
Standby Power Supply, tric power supply configurations are ternate AC Source Standby Power in compliance with the design. Supply.
I
- 3. Containment isolation valves 'and 3. Review plant records to verify com- 3. The containment isolation values and associated piping meet ASME Code pliance. in-containment piping associated Cia-s 2 requirements.- with the Compressed Gas System conforms to ASME Code Class 2 re-quirements.
1 1.9.6 4-30-92
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1.9.22.9 STATION SERVICE WATER SYSTEM (SSWS) PUMP ,
STRUCTURE Design Description ne S9W9 Unmn Structure interfaces with the station service water system pumps and G design is a site specinc requirement of the applicant. He structure includes the SSWS pump intake structure and foundations and is located in the same vital protection area as the main plant end outside the corridor designated as the potential turbine missile path.
The pump structure design meets Seismic Category I requimnents. The design provides physical barricts that maintain dMsional separation of SSWS components and will withstand the effects of:
- a. natural phenomenon, including a safe shutdown carthquake, tioods, tornadoes, and hurricanes
- b. cxternally and internally generated missiles
- c. fire hazards
%e design includes a safety grade screen system located b: fore the SSWS pump inlets. The design provides the capability for per. odic cleaning and means to limit ingestion of biofouling organic materials and debris, consistent with the fouling limits of the piping and Component Cooling Water System (CCWS) designs. Pump inlet blockage limits are accommodated in the design.
The SSWS pump well is designed to prevent the formation of air vortices for the complete range of operating water lev'Is in the pump well.
Inspections, Tests, Annlyses, and Acceptance Criteria Table 1.9.22.9 provides a definition of the inspections, tests, and/or analyses and associated acceptimte criteria.
1.9.22.9 1 43992
- - _ _ _ - _ _ _ _ _ - - a
s' - + ~ DRAF 1
_ TABLE 1.9.22 9 STATION SERVICE WATER SYSTEM (SSWS) PUMP STRUCTURE Inspections. Tests. Analvsgs and Acceptance Criteria Certified Desima Commitment Inspections. Test. Analysis Accertmace Criteria
- 1. The SSWS pump structure is de- L The design is evaluated for confor- 1. Sci mic Category I requirements are sigacd to Seismic Category I re- mance to Seismic Category I re- met for the design.
quiremerits. - quirements.
- 2. The design prosides . physical bar- 2. Inspection of the physical layout will 2. Divisional separation is preided riers to maintain : divisional sep- be made to evaluate the capability between the SSWS. J aration of SSWS components. for dhisional separation.
- 3. The design provides the capability to 3. An analysis of the. design charac- 3. The design is simu tobe capable to withstand: teristics is made to verify the cap- withstand:
- g" ability to withstand the specirsed
- 1) . Natural phenomena etoditions. 1) Natural ,dienomena
- a. SSE 2) Missiles
- b. Pmis 3) Fire Hazwds
- c. Tornadoes
- d. klurricanes 2)' Externally and internally generated-missiles.
' 3) Fire Hazards
+n 1.9.22.9 4-30-92
[
~
s'"
DRAFT TABLE 1.9.22.9 (Continued)
STATION SERVICE WATER SYSTEM (SSWS) PUMP STRUCTURE Inspections. Tests. Analyses and Acceptance Criteria inspections. Test. Analysis Acceptance Criteria Cenified Desizn Commitment 4 The site layut is inspected to verify 4 & SSWS pump strucrare is within 4 & SSWS pump structure is located the sane vital protected area as the in the same vital protection areas as this requzrement.
plant.
the main plant.
The site layout is inspected to verify. 5. The 35WSis L~= sed cuiside the pro-5 The SSWS pump structure is located 5. ,
outside the turbine missile path. this requirement. ject.:d paths for turbme generated l
- assiles.
- 6. An inspection of the design is made 6. The design includes a safety grade
- 6. . The SSWS pump structure provides screen system in a posidon before a safety grade screen system prior to to wrify the inclusion of a safety g grade screen system located prior to the pump inlets.
the pump inlets.
the pump inlets.
- 7. An analysis of the design if made to 7. The % prevents air wxtices in
- 7. T'je SSWS pump well is designed to verify. that air vertices will not (c- the pump weH for au operational prevent air vortices . over the com-plete range of : operathnal water cur for all operaticnal water levels states, levels in the pump well. in the pump..wcII. -
4-30-92 1.9.22.9 L.
i l
-~ ~
~ -
DRAFT L 1.11.1 LIQUID WASn MANAGEMENT SYSTEM l i
Design Description ,
ne Liquid Waste Management System (LWMS) provides the capability to collect, segregate, store, process, sample, and monitor radioactive liquid waste. The LWMS is a non-nuclear safety (NNS) system containing no sJety class components except for containmer't isolation valves and penetrations which are designed to Refety Class .
- 2 requirements.
The liquid waste is segregated into the following categories:
- a. Equipment drain waste or clean waste -- degassed reactor grade -
radioactive liquid waste -
- b. Floor drain waste or dirty waste - non-reactor grade radioactive liquid h waste
- c. Detergent waste - laundry and het showers
- d. Chemical ,vaste -- non-detergent liquid waste (e.g., decontamination Guids)
Each category of waste is processed by an independent subsystem.
The equipment and Door drain waste subsystems are designed with the provision for .
fihration, decontarninat!on by demineralizers, batch sampling, and recirculation capability for further processing.
The floor drain waste subsystem is designed with the additional ca['n'vility for oil / crud removal, Occculent addition to collection tanks, and pH adjustment.
The chemical waste subsystem is designed with the capabilhy for pH adjustment -
through chemical addition, filtration, batch sampling, and recirculation to floor draial waste subsystem for further processing.
~
The detergent waste subsystem is designed :th the capability for filtration,-
decontamination by demineralizers, batch sampling, and recirculation to Door drain subsystem for further processing.
This system is designed with collection and storage capacity to process the maximum
- expected liquid waste volumes. These liquid waste volumes are calculated, based on anticipated peak daily inpum, utilizing vendor, plant specific, and industry data Each large volume subsystem is provided with one or more parallel collection tanks and 1.11.1 4-30-92
'l__'.. . . . . .
l SYSTEM 80+ m waste monitor or sample tanks. Each small volume subsystem has one collection tank.
Ti.e system is designed to proces:; radioactive liquid waste so that the concentration of the liquid effluents at the potable water source is within 1;mits specified by regulatory directives. Criteria are met by the provisior of design processing capabilities and adequate dilution flow at the plant discharge.
This system is designed so that releases of radioactive me.terials to the environment '
will be controlled and monitored in accordance with 10 CFR 50, Appendix AL (General Design Criteria 60,61, and 64) r : system is designed so that releasc. of processed liquid waste will require an op< o . action. This system is designed with the capability to batch sample and monitor precessed liquid waste prior to release to the environment. The radiation monitor, located upstream of the pir.nt discharge, is designed to terminate Ibc release if a pre-set limit is exceeded.
The LWMS is housed in a Radwaste Facility building.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.11.1 provides the inspections, tests and/or aaalyses and their associated acceptance criteria.
4 1.11.1 43092
=
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SYSTEM 80+w P r"k h ;
_ TABLE 1.11.1 LIOUID WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Design Commitment Inspections. Tests.' Analyses Acceptance Criteria i
L The LWMS is designed with suf- 1. Vendor, plant specific, aad indus - L' The LWMS prevides storage cap-ficient - collection and storage try information ' vi'I be reviewed
. acity to accommodate the maxi-capacity. to verify: mum expected radioactive liquid waste input during normai and an-
. a) . Provision for at least one collec- tidpsted occurrences;-
tion, . waste tronitor, and samp!c tank per subastem.
!j b) Sizing of coIIection, waste monitor and sample tanks.
c) Pro.ision - for emergency . storage capacity by the Steam Generator Drain Tank. ,
i d) Provision of level indication for collection, . waste monitor, and sample tanks e:-
l'11.1' 4-30-92
.__g..m
SYSTEM 80+m ,- g TABLE 1.11.1 (Contirmed)
LIOUID WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria The LWMS is designed with suffi- 2. Vendor records, instalbtion' in- 2. The LWMS designs conforms to 10
. cient . processing - capabilities to- for' nation, and performarece test CFR 20, Appendix B. Table II, results will be reviewed to verify Column 2 and 10 CFR 50, ensure ' the concentration of the liquid effluents at the ' potable that assumptions made during the Appendix I linits.
w" ten source. is within design analysis are conservative with limits, respect to:
a) Process decontamination capability provided.
l b) Dilution - flow available at plant -
discharge.
c) Recirculation capability, prosided.
- 3. ' The LWMS is ' designed to provide 3. Inspection of installation records 3. The LWMS design includes pro.
for 'a controlled . monitored release. together ' with plant . walkdowns vision for controlled monitored -
This will be ensured Lthrough the - will"be conducted to confirm that. release in accordance with 10 CFR capability ~ to sample and monitor batch sampling and radiation mon- 50, Appendix A (General Design each batch prior to release to the' itoring capabilities are provided in ' Criteria 60, 61, and 64)J
. environment. the system ' design upstream of the plant discharge.
1.11.1' 4-30-92
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e 4
- n SYSTEM 80+m d(( l TABLE 1.11.1 (Continued)
LIOUID WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analygs Acceptance Criteria a) Sampling capabiSty is . provided for each collection tank prior to processing and for each waste monitor tank and sample tank prior to batch release.
b) Radiation monitoring is provided upstream a Jant discharge.
- 4. Interfacing systems are designed 4 Inspection of installation records 4. Radioactive liquid waste streams j
' with the capability to segregate together with salk-downs will be are seg:cgated prior to collection radioactive ' liquid waste. prior to conducted to conGrm that the in- and . processmg by independent collection and precessing in the terfac;ng system design config . subsysten.s in the LWMS.
LWMS. urations provide for segregation of the radioactive. liquid waste streams.
~t 1.11.1 4-30-92
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SYSTEM 80+ =
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rk 1.11.2 GASEOUS WASTE MANAGEMENT SYSTEM Design Description The Gaseous Waste Management System (GWMS) provides the capability to collect, store, process, sample, and monitor radicauive gaseous waste. The GWMS is a non-nuclear safety (NNS) system containing no safety class _ components except for containment isolation valves and penetrations which shall be designed Safety Class 2.
The GWMS is a gas delay system. The GWMS is designed to operate continuously, l as well as periodically, at flow rates established by systems feeding the GWMS, such as the Chemical and Volume Control System. 1 This system is designed to include conditioning c.luipment, such as a cooler. condenser for humidity control, charcoal guard bed to protect the charcoal adsorbers from excessive moisture or contamination, and charcoal adsorbers for delay of noble gases.
This system is designed: to _ include dualj hydrogen analyzers, which monitor the concentration of hydrogen and oxygen in the GWMS, and nitrogen purge capability to maintain the concentration of hydrogen and oxygen less than 4% in accordance with to CFR 50, Appendix A (General Design . Criterion 3).
The system is designed to process radioactive giscous waste so that the concentration -
of the gaseous effluents at the exclusion area boundary is within ilmits specified by regulatory directives. Criteria are met by the piovision of sufficient processing capability and adequate dispersion of the effluent released from the unit vent.
This system is designed to ensure releases of radioactive materials to the environment can be controlled and monitored in-accordance with 10 CFR 50, Appendix A (General Design Criteria 60,61, and 64). This system is designed with the capability to continuously monitor processed gaseous waste prior to release to the environment.
The radiation monitor, is designed to automatically isolate the GWMS discharge if a -
pre-set limit is exceeded.' This system is designed so that leakage rates of processing equipment ensure releases of radioactive gases from the GWMS to the environment is controlled.
The GWMS is located in the Radwaste Facility.
-Inspections, Tests, Analyses, and Acceptance Criteria
. Table 1.11.2 provides the inspections, tests and/or analyses and their associated acceptance criteria. '
1.11.2 -. 1 - 430 a l
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4
% .- d mane wm System 80+m r, ,kf f l TABLE 1.11.2 GASEOUS WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria
- 1. Tue GWMS is desigred with pro- 1. Vendor records, site specific in- 1- Conformance to 10 CFR 20, cessing capabilities to ensure the formation, industry data, and pre- Appendix B, Table II, Column I concentration 'of the gaseous efflu- operational tests will be reviewed and 10 CFR 50, Appendix I liteits.
ents at the exclusion area boundary to verify that the bases of the de-is within design limits. sign analysis are conservative with respect to:
a) Delay time for each isotope cal- Delay time for xenon and krypton
) are at least 60 days and 3 days, culated based on:
respectively.
i) Carrier gas flow rate l
Mass of charcoal in absorber. j ii) iii) Absorbtivity of charcoal for each l isotope.
b) Dispersion of effluents at plant unit vent.
l TABLE 1.11.2 (Continued) 1.11.2 . 4-30-92
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_ = _ - .-.-. . .
i System 80+w - E.4mk GASEOUS WWTE MANAGEMENT SYSTEM Inspections, Teig. Analyses and Accepts /nce Criteria Certified Design Commitment Inspections. Tests. Analyses Acce plance Criteria
- 2. The GWMS is designed to lhnit a 2. Vendor records, pre-operational 2. He GWMS conforms to 10 CFR '
buildup of hydrogen - and oxygen to tests, and inspections of instaation 50, Appendix A (General Design explosive limits. records together with plant walk- Criterion 3) requirements for hy-downs will be reviewed, drogen control. Hydrogen and oxygen concentrations will be maintained less than 41
- 3. The GWMS has provisions for con- 3. Inspection of installation records 3. He GWMS conforms to 10 CFR trolled monitored releases. together with plant walkdowns. will 50, Appendi A (General Design be conducted to confirm radiation Criteria 60, 61, and 64).
I monitoring . capabilities are pro-vided. Inspection of vendor Radiation monitoring is provided records and leak tests will be upstream of plant unit' vent.
conducted to confirm installed equipment is leak-tight. Irakage rates from individual components within a tone are-within design limits. l l
l L
1.11.2 4-30-92 l
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