ML20209C634
| ML20209C634 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 11/23/1983 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1475 NUDOCS 8312070264 | |
| Download: ML20209C634 (30) | |
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MEMORANDUti FOR:
T. flovak, Assistant Director for Licensing, DL FROM:
R. W. Houston, Assistant Diregtor for Reactor Safety, DSI l
SUBJECT:
DRAFT INPUT FOR SAFETY EVALUATION REPORT - WNP-3 Plant Name: UNF-3 Docket No.:
50-503 Licensing Stage: OL Responsible Branch:
LB #3 Projact Manager:
A. Vietti Review Status:
Incomplete The enclosed draft Safety Evaluation Report (SER) has been prepared by the Containment Systems Branch (CSB) after having reviewed the applicable portions of the FSAR. The bases used in the review are contained in SRP Sections 6.2.1 6.2.2, 6.2.3, 6.2.4, 6,2.5, and 6.2.6. provides a summary of the outstanding issues. These issues will be resolved as the applicant responds to our concer.ns.
If there are any questions;, please call C. Li, (X29484) of my staff.
C.'t! T:'3sd gy M.DjQ22qtc.g R. W. Houston, Assistant Director for Reactor Safety Divisicn of Systems Integration
Enclosures:
As stated distribution s
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ENCLOSURE 1 DRAFT SAFETY EVALUATION REPORT UNP, UNIT 3 DOCKET No. 50-508 6.2 Containment Systems The containment systems for the WNP, Unit 3 include dual containment structures, containment heat removal systems, a shield building ven-tilation system, a containment isolation system and a containment combustible gas control system.
The primary and secondary containment structures and their associated systems atL function to prevent or control the release of radioactive fission products which might be roteased folLowing a postulated loss-of-coolant accident (LOCA),
s9condary system pipe rupture, or any other accident releasing radioactive material into the containment atmosphere.
The staff has reviewed the applicant's design, design bases, and safety analyses for the containment and the containment systems provided in the FSAR.
The acceptance criteria used as the basis for our evaluation are contained in Sections 6.2.1, " containment Functional Design," 6.262, " Containment Heat Removat Systems," 6.2.3,
" Secondary Containment Functional Design," 6.2.4, " Containment Isota-t ion Sy st em," 6.2.5, " Combust'Ib le Ga s Control In Containment," and 6.2.6, " Containment Leakage Testing," of the Standard Review Plan I
(SRP), NUREG-0800, dated July 1981.
These acceptance criteria in-j clude the applicable General Design Criteria (Appendix A o f 10 C FR Part 50), Regulatory Guides, Branch Technical Positions, and industry j
l codes and standards as specified in the above cited sections of the ERP.
The results of the staff review are discussed below.
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O 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure The reactor containment is a free-standing steel structure with a net 3
free volume of 3,218,000 ft The containment structure houses the nuclear steam supply system including the reactor, reactor coolant pumps, pressurizer and steam generators, as weLL as certain components of the plant's engineered safety feature systems.
The st rticture is dosigned for an internal pressure of 44 psig and a temperature of 367*F.
The reactor containment is completely enclosed by a shield building with an annulus region between the structures.
Maximum Pressure and Temperature Analysis The applicant has analyzed the containment pressure and temperature responses for postulated reactor coolant system and secondary system pipe ruptures to establish the containment design bases and the condi-tions for environmental qualification of safety related equipment in the containment.
The most limiting single active failure, from the standpoint of predicting the highest containment pressure and temperature, was assumed in the containment analyses.
The applicant's postulated spectrum of breaks in the reactor coolant system (i.e.,
Loss-of-coolant accidents), as described in FSAR Table 6.2.1-1, in-clude double ended slot breaks in the hot leg and in the cold leg at the roactor coolant pump suction and discharge.
For the cold leg breaks both ainimum and mdximum emergency core cooling system (ECCS) flow cases were considered.
The reactor coolant system pipe break spectrum is based on 2-
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that prescribed in CESSAR Section 6.2.1.3, which the staff found accep-
-table in the CESSAR SER cf November, 1981.
The design basis LOCA a,t 2
WNP-3 was determined to be a hot leg slot break of 19.2 ft ; the failure of one train of the containment spray system was the worst single active failure.
The spectrum of braaks postulated for the secondary system includes slot breaks in the main steam Line at five different power levels, from 0%
to 102% of full power.
Slot breaks were determined to be more severe than guitLatine breaks.
The maximum break area that atlowed a pure steam blowdown was chosen as the most limiting break for each power level'.
Main feedwater Line breaks were not included because they re-sult in two phase blowdowns and thus are not as severe as main steam breaks.
Again, the spectrum of main steam Line breaks (MSLB) analyzed for WNP-3 is the same as that prescribed in CESSAR Section 6.2.1.4, which the staff found acceptable in the CESSAR SER of November, 1981.
The MSLB resulting in the highest containment pressure was found to be the four-square-foot slot break at 0% power with a failure of cne spray train.
The peak containment temperature of 367 F was calcu'ated 8
2 to occur fotLowing a 8.78 ft slot break in the main steam Line at 102%
power, with a failure of one spray train.
The applicant has performed the containment pressure and temperature analyses using the CONTEMPT-LT computer code.
Initial conditions and input data, including passive and active heat sink parameters, were conservatively chosen to produce the highest containment pressure.
The highest containment pressure was calculated to be 39.4 psig (versus the containment design pressure of 44 psig), which occurred for the design basis LOCA identified above. t 6
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4 For the Long term containment pressure response, a double ended slot break at the pump suction (with minimum safety injection) was analysed.
The analysis showed the containment pressure would be reduced to appro-ximately 17 psig; i.e.,
less than 50% of the peak calculated value (38.2 psig), in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, in accordance with staff guidelines.
The design basis main steam Line break was analyzed to establish the peak containment tCaperature to be used in developing the temperature profiles for environ-contalLy qualifying safety-related equipment located in containment.
The peak temperature was calculated to be 367 F.
Wo have reviewed the applicant's selection of initial c nditions, input parameter.s, and analytical assumptions and find them tc be acceptable and in conformance with staff guidelines.
Staff confirmatory analyses wore performed for the design basis reactor coolant system break and the design basis main steam Line break using the CONTEMPT-LT/28 computer code.
The results of the confirmatory analysis are in close agreement with the applicant's results, and confirm their acceptability.
Based on our review of the applicant's containment functional analysis, as discussed above we conclude that the applicant has satisf actorily demonst ratec i
the adequacy of the containment functional' design, and has appropriately de-tormined the containment temperatures and pressures to which safety-related Oguipment in containment must be environmentally qualified.
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Protection Against Damage From External Pressure To demonstrate the adequacy of the containment against the maximum ex-l pocted external pressure, the applicant has analyzed the consequences i
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of a postulated inadvertont actuation of the containment heat recovel sy$ tem during norm.at plant operation.
The operation of two spray trains and tuo containment fan coolers was assumed to occur during normal operation; One of the two vacuum breakers was assumed failed.
The applicant calculated a maximum pressure differential of 0.58 pounds per square inch, which is loss than the containment vesset design external (di f f erential) pressure 4
of 0.7 psid.
The initial conditions and assumptions used in the analysis wore chosen to maximize the differential pressure load on the containment.
Besed on our review of the applicant's analysis we find the containment dosign has sufficient margin to accommodate the maximum postulated exter-nal load.
6.2.1.2 Subcompartment Analysis Subcompartment analyses were performed to determine the acceptability of the design differential pressure loadings on containment internal structures from high-energy line rupture accidents.
The applicant's subcompartment analyses included the reactor cavity, and pressurizer and steam generator compartments, where high energy line ruptures were postulated to occur.
A spectrum of pipe areaks was analyzed by the ap-plicant to determine the limiting break that resulted in peak loads on oach of the subcompartment watLs.
The reactor coolant system (RCS) breaks, considered in the WNP-3 subcom-portment analysis (except for one), and the mass and energy release data, wore obtained from CESSAR Section 6.2.1.2.
We find this approach accep-tcble based on the staff CESSAR SER.
The one break that 'dif f ers from those in CESSAR is the guitLotine break of the discharge leg in the reactor 2
coolant system; a 350-in break size is specified in CESSAR, where as _ _
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the b'reak 512e has been reduced to 100-in in the WNP-3 safety s
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The inherent stiffness of the system, together
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this break nres.
Tne 100-in break mass and energy data were cal-eutated based on the methodology described in CESSAR.
The staff hos reviewed the,npplicant's analysis and find the mass and energy release data acceptable, contingent upon the acceptability of the Limiting pipe break size (see CER Chapter 3.0).
The applicant used the RElAP-4 M006 computer program to analyze the pressure transients in the reactor cavity, and the steam generator
.and pressurirGr coonertments.
Separate discussions for each subcom-portment are pfesented below:
En a c t o r_ _Cy;i t y Area ty s i r The reactor esvity is a hfavil'i reinforced concrete structure that oorforms th? dual function of providing reactor vessel support and radiation shielding.
The reactor cavity is essentially a cylindrical annular air space between ths reagtor vessel and the primary shield wolL.
The cavity is boanced at the top and bottom by a neutron shield.
The major vent patns for the reactor cavity are the six piping penetra-tions (two hot Legs and tour cold legs) thecugh the primary shield coll to the ' steam generator compartments.
2 The applicant postulated 100 in, discharge and hot leg guitLotine breaks; the design basis break was found to occur -In the discha ge teg.
The peak differential pressure across the reactor cavity wall was calculated by the applicant to be 29.6 psid, versus o design value of 211.4 pcid. >
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The a'pplicant choge t o Tri.ti the r1 actor equity'nodalization i
f, sonsitivity study of Carolina Powe r '& _ Light on their Shearon f
50-400,-4b1,--402,-403)
Hcrris Plant (Docket Nos.
as a 5 sis for verification of the WNP-3 reactor cavity nodalizafion scheme.
It was concluded,in the Shearon Harn:Is nodalization study that subcompartmen't nodalization models were ietermined principally by physical flow restrictions within each compartment.
These flow restrictions consider the presence of steel and concrete' obstructions, doorways, vent shaf.ts,Tgrating, reactor coolant '
pumps, piping, the, steam generator, the pressurizer, the reactor vossel, and the reactor' cavity missile and neutron shields.
The subcompartment models in[WNP-3 taka intI account alL physical, flow j
cbstructions.
AlL. assumptions utilized by the applicant >Jn'the reactor cavity subcompartkent analysis have been reviewed and found to be appropriate.
In' addition, the staff performed 1a confirmatory analysis using the COMPARE-MOD 1A computer code, and the same nodal model as the applican't.
Although the staff's analysis predicted a higher peak differential ptessure (34.6 psid), the reactor cavity design is adequate for the dif-forential pressure loads from the worst postulated pipe rupture within the reactor cavity.
i The applicant has not provided in the FSAR an analysis of the forces and moments on the reactor vessel due to the differential pressure across the vessel caused by a r_eactor coolant system break within the reactor cavity.
The applicant has indicated that the methodology 7-
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y presented in CESSAR Chapter 3 was used for the force / moment analysis.
However, it is not clear that the generic analysis in the CESSAR is applicable to WNP-3.
This matter wiLL remain en open item until further justification or analysis is provided by the. applicant.
Steam Generator Subcompartment (SGS) Analysis The walts of the steam generator compartment are constructed of reinforced concrete.
The applicant considered a spectrum of primary coolant system pipe breaks for resultant load impact on the SGC watLs.
A steam line break was not postulated because the routing of this line does not pass through the steam generator compartment.
The staff has reviewed the spectrum of breaks postulated by the applicant and finds it acceptable.
The TGC has been nodalized as 20 volumes.
The staff has accepted this nodalization based on the above cited Shearon Harris subcompart-
=ent nodalization sensitivity study.
For the spectrum of breaks analyzed for the steam generator
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compartment, the applicant's analysis shows the design basis i
2 break is the 592-in guitLotine break in the suction Leg.
The analysis results a peak differential pressure of 13.1 psid on the steam generator subcompartment walls versus the design value of 25.9 psid.
Our confirmatory calculation, using the COMPARE-MOD 1A computer code, confirms the acceptabilty of the applicant's l
results.
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sosed on our. review of the applicant's analysis and the results of our confirmatory analysis, we find the design bisis of the steam generator subcompartment walls is adequate.
Pressurizer Subcompartment Analysis The pipe breaks considered for the pressurizer subcompartment include surge line, spray line, and safety relief valve line breaks.
The cost limiting case is the double-ended guillotine break in the surge line.
Based on our review of these breaks, we find the applicant's choice acceptable for the pressurizer compartment analysis.
The pressurizer compartment nodalization consisted of 13 volumes.
The applicant calculated a peak differential pressure of 24.3 psid across the' walls of the subcompartment, compared to the design value of 84 psig.
We have performed a confirmatory analysis using the COMPARE-MOD 1A computer code and our results confirm the acceptability of the applicant's results.
We therefore find the applicant's pres-surizer compartment analysis to be acceptable.
6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-Of-Coolant Accidents For the containment functional analysis, the applicant obtained mass and l
energy release data for postulated loss of coolant accidents from CESSAR Section 6.2.1.3.
We find this approach acceptable based on the staff findings in the CESSAR SER dated November, 1981.
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6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Greaks Inside Containment The applicant used the mass and energy release data provided in CESSAR, Soction 6.2.1.4 for main steam Line breaks.
The CESSAR MSLB mass and onergy release data may be used if certain interface requirements (for example, main steam and feedwater isolation valve closure times and maximum steam Line and feedwater Line volumes) are met by the ap-plicant.
We have confirmed that the WNP-3 MSLB analysis satisfies these interface requirements and, therefore, conclude that use of the CESSAR MSLB mass and energy release data is acceptable.
However, the applicant has not addressed the concerns of IE But Letin 80-04, regarding the impact of runout flow from the feedwater system.
Wo have requested additional information in this regard; this matter witL remain an open item until we can review the applicant's respon'se.
6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on the Emergency Core Cooling System (ECCS)
The applicant has not adequately demonstrated the applicability of the CESSAR minimum containment pressure analysis to the WNP-3 design.
There-fore, a CESSAR interface requirement has not been met.
The applicant has indicated that a WNP-3 plant specific analysis witL be performed.
This matter witL remain an open item until we have had the opportunity to review the applicant's analysis.
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6.2.1.6 Summary and Conclusions We have evaluated the WNP-3 containment functional capability with rospect to the requirements of General Design criteria 16 and 50 of Appendix A to 10 CFR Part 50.
We have found the applicant's analyses of the dynamic pressure loads that act on the containment vessel and subcompartment structures from postulated pipe breaks acceptable, with the folLowing qualifications:
1.
The forces and moments acting on the reactor vessel as a result of the design basis reactor coolant system break within the reactor cavity are needed to complete our review.
Th guidelines of SRP 6.2.1.2 and Section 3.2 of NUREG-0609 should be folLowed.
2.
A minimum containment pressure analysis that is applicable to the WNP-3 plant is needed to complete our review.
3.
The applicant's response to IE ButLetin 80-04 is needed to complete our review.
6.2.2 Containment Heat Removat Systems The function of the containment heat removat systems is to remove heat from the containment atmosphere to limit, reduce and maintain at accep-tcbly low Levels both the containment pressure and temperature fotLowing a LOCA or secondary system pipe rupture.
The containment spray system (CSS) is the only active containment heat removat system at WNP-3 and also serves as a fission product removal and control system (see SER Section 6.5).
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The CSS consists of two redundant and independent 100% - capacity trains, each containing a containment spray pump, a shutdown cooling host exchanger, a spray header, and associated valves, piping and instruments.
Each of the two containment spray pumps has a design flow rate of 5000 gpm of water at a head of 645 feet during the in-joction mode, and a flow rate of 6000 gpm at a head of 600 feet during the recirculation mode.
The CSS is automatically started by the containment spray actuation signal (CSAS) which is initiated by high-high containment pressure.
The CSAS may also be initiated canualLy in the control room.
Upon receipt of a CSAS the containment spray pumps are started, the system isolation valves are opened, and the borated water from the refueling water storage tank (RWST) flows into the containment.
Ful L spray flow from the nozzles is established in about 50, seconds after receipt of the CSAS, assuming 4
' Loss of offsite power; this satisfies the CESSAR interface require-cont of establishing full spray flow in less than 58 seconds.
When the water Level in the RWST reaches a specified Low level, a recir-culation actuation signal (RAS) automatically realigns containment spray pump suction from the RWST to the safety injection system (SIS) rocirculation sump.
The operator must then verify that the appro-priate amount of water has been discharged into the containment, the flow path from the SIS recirculation sump to the containment spray pumps is open, and the minimum flow Lines are isolated.
- Finally, the operator witL manually close the RWST isolation valves from the control room to complete the transition from the injection code to the recirculation mode. '
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The C'SS is designed to Quality Group B and Seismic Category I roquirements.
-The applicant has also provided a failure mode and effects analysis (FMEA) and other information demonstrating the_ ability of the CSS to function folLowing postulated single active failures.
Wo have reviewed the applicant's net positive suction head (NPSH)
I colculations and find that sufficient NPSH wilL be available for the CSS pumps during both the injection and the recirculation modes of operation. The applicant's evaluation of the available NPSH is consistent with the guidelines of Regulatory G u i d e 1.1, R e v. O,
" Net Positive Suction Head for Emergency Core Cooling and Contain-cont Heat Removal System Pump," and is acceptable.
f Rogulatory Guide 1.82 " Sumps for Emergency Core Cooling and Contain-aont Spray System," provides guidelines to be met by reactor building sumps that are designed to be sources of water for the ECCS and the CSS fotLowing a LOCA.
The guidelines address redundancy, location, and arrangement of sumps as welL as provisions to screen out-debris and ensure adequate pump performance.
The applicant's sump design conforms to Regulatory Guide 1.82, revision 0, and we l
find the design acceptable.
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Poriodic testing and inspection of the system active components, i.0., pumps, valves, etc. wiLL be performed in accordance with the in-service inspection requirements of the ASME Code Section XI to ccsure the operability and performance of the system. 4 i
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Besed.on our.a e v i e w, we conclude that the containment spray system satisfies the requirements of General Design criteria 38, 39, 40 cnd the provisions of Regulatory Guides 1.1 and 1.82, Rev. O, and, therefore, is acceptable.
6.2.3 Secondary Containment Functional Design The secondary containment encompasses the annular space betwe,en the concrete shield building a'nd the steel primary containment vessel, the ECCS and mechanical penetration areas, and the fuel handling building.
The shield building is a seismic Category I structure which provider biological shielding, controlled release of airborne rcdioactive materials folLowing an accident, and environmental pro-tection.
The shield building ventilation sysetz (SBVS) is an engineered safety feature designed to maintain a negative pressu're in the annulus fotLowing a LOCA and to filter the airborne fission products which may leak from the primary containment to the annulus fotLowing a LOCA.
During normat operation, the annular subatmospheric pressure of minus 10 inches (water gage) is maintained by the annulus vocuum maintenance system (AVMS).
The regions comprising the secondary containment, other than the annulus, are designed to maintain a sub-otmospheric pressure by the ECCS area / fuel handling building (FHB) filtered exhaust system folLowing a design basis accident.
The SBVS consists of two independent 100% capacity trains; each train is actuated by a separate channel of containment isolatica actuation signal (CIAS) and atL redundant active components are powered from separate ESF bases.
Each train includes one full 14 -
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capac'ity exhaust fan, a filter train (including a demister, electric hosting coil, prefilters, HEPA f'i l t e r, charcoal absorbers, and afte -
HEPA filter), ductwork, valves, and instrumentation and controls.
A failure modes and effects analysis was performed by the applicant to show the system meets the single failure criterion.
AlL components end ductwork are designed to meet seismi c Category I requirements.
The applicant has analyzed the performance of the SBVS using the WATEMPT computer code.
The results of the analysis indicate a negative pressure relative to the outside atmosphere can be maintained in the annulus throughout the transient folLowing a LOCA, thus ensuring no primary containment out-leakage escapes unfiltered directly through the shield building.
However, Appendix 6.2A to the WNP-3 FSAR does not have sufficient information concerning the WATEMPT codes to permit a complete evaluation.
We witL conclude on the applicant's method of analysis of the post-LOCA SBVS performance after we have had an oppor-tunity to review the additional information we have requested from the opplicant.
Preoperational testing of the SBVS will verify the system performance capability to achieve and maintain a negative annutus pressure.
Perio-t dic testing and inspection of the SBVS wilL be included in the plant Tochnical Specifications.
The applicant has also identified systems for which through-line loakage following a LOCA could result in containment bypass leakage.
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The applicant has committed to perform local Leak rate tests on the potential bypass Leak paths in accordance with the requirements of Appendix J to 10'CFR Part 50.
The total poten-tial bypass Leakage rate wilL be Limited to 22 percent ofsthe -
design Leak rate of the containment (0.2 weight percent of the internal net free volume per day at a pressure of 39.4 psig),
or 0.044 w/o per day.
With the exception of the need for additional information con-cerning the post-LOCA shield building annulus pressure transient analysis the staff concludes that the secondary containment systems ceet the requirements of GCC 41, 42, and 43 of Appendix A, 10 CFR Part 50, and therefore, are acceptable.
6.2.4 Contai'nmen't Isolation System The function of the containment isolation system (CIS) is to atlow the normat or emergency passage of fluids through the containment boundary while preserving the ability of the boundary to minimize the release of fission products that may result from a postulated l
accident.
This section, therefore, is concerned with the isolation l
of fluid systems which penetrate the containment boundary, including the design and testing requirements for isolation barriers and actua-l tors.
The isolation barriers include valves, closed pipin.- systems, and blind flanges.
In general, for each penetration at least two bar-riers are required between the containment atmosphere or the reactor coolant system and the outside atmosphere so that failure of a single barrier does not prevent isolation.
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AlL non-essential systems are those systems either automatically iso-Lated by one of the actuation signals (CIAS, SIAS, MSIS) discussed below or normalLy Locked closed.
The containment isolation actuation signal (CIAS) and safety injection actuation signal (SIAS) are initiated by high containment pressure or low pressurizer pressure signals.
The main steam isolation signal (MSIS) is initiated by high containment pressure low steam generator pressure, or high steam generator water level.
These signals can also be initiated manually from the control room.
For the containment purge and vent system, the isolation valves are also isolated by the containment high radiation signal in addition to the CIAS.
We, therefore, conclude that the containment isolation signals provide accep-table diversity.
We have r.e v i e veYkih'e applicant's designation of essential systems.
The essential s/ stems do not require automatic isolation, and if the auto-matic isolation exists, the systems are equipped with override features for remote manual operation from the control room.
Those systems or portions of systems classified by the applicant as essential include the high pressure safety injection system, containment spray system, auxiliary feedwater system, main' steam and feedwater isolation system, chemical volume and control system (CVCS) charging and Letdown lines, CVCS rJactor coolant pump '(RCP) seal injection Lines, RCP component cooling water Lines, instrument and control air system, hydrogen purge system, and RCP seal bleed off Lines.
These systems are considered important to post-accident safe shutdown and valuable in accident mitigation, particularly in the event of a smaLL break LOCA or a secondary system rupture, and, therefore, their \\
class'ification as essential is acceptable.
Provisions are made to allow the operator in the control room to detect leakage from renote manually controlled systems.
These provisions include instrumentation to measure radiation levels, flow rates, pressure and sump water levels in the scfety equipment area and the penetration area.
Wo have reviewed the closure times for the containment isolation valves.
Most valves close in 10 seconds or less.
In particular, the containment purge and vent systems are designed to close in 5 seconds (except for isolation valve 2PV-8019 on penetration 80, which wilL be discussed below).
We conclude that the containment isolation valve closure times, with the exception of valve 2PV-8019, are acceptable.
We have reviewed the containment purge and vent systems against the provisions of. Branch Technical Position (BTP) CSB 6-4, " Containment Purging During Normal Plant Operations."
The 48-inch containment purge valves wiLL be sealed closed during normal operation and be vorified to remain closed at least every 31 days; requirements for this wilL be included in the plant Technical Specifications.
Furthermore, l
os a result of our study of valve Leakage due to seal deterioration, loakage integrity tests of the purge and vent system isolation valves cust be conducted periodically; i.e., over and above the Leak testing requirements of Appendix J.
This requirement, together with the test frequency wiLL be included in the plant Technical Specifications.
We conclude that the 48-inch containmen' purge system design satisfies the provisions of BTP CSB 6-4 and that operation of the system as proposed; i.e., only during shutdown and refueling, is acceptable.
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,.q The containment vent system, consisting of two 8-inch containment penetrations (P-80 and P-81), is designed to close following receipt of a CIAS or high radia~ tion signal.
The containment vent exhaust penetration, P-81, has two isolation valves which are designed to close in Less than 5 seconds and fait closed on loss of operating power; we have found this penetration design acceptable.
- However, the containment vent make-up penetration (P-80) is equipped with a cotor operated isolation valve (2PV-8019) and a check valve (2PV-V021).
The motor-operated valve has a closure time of 10 seconds and is designed to fail "as is" (FAI).
The FAI design is not consistent uith the statement in FSAR Section 6.2.4.2 that the isolation valves on both the containment vent and purge systems are designed to fait close.
Also, the 10-second valve closure time differs from the 5-second assumption used in the radiological dose analysis (FSAR Section 9.4.6.6.6).
The check valve (2PV-V021) in the containment vent make up Line is not an appropriate type of containment isolation valve for use in a line which directly connects the containment atmosphere to the outside environment.
It is the staff's position that the con-l tainment isolation val'ves in Line P-80 should be automatic, power operated valves having less than 5 second closure times, and should l
fait closed on loss of operating power.
This matter wiLL remain an i
open item pending receipt of additional information regarding the applicant's plans for complying with the staff position.
We note that the applicant has not provided an analysis of the reduction in the containment pressure resulting from the partial 19 -
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loss of containment atmosphere through purge and vent system isolation valves, which may be open at the onset of a LOCA, and the consequent impact on the minimum ECCS backpressure determination.
This analysis is catled for in BTP CSB 6-4.
Therefore, this matter wi LL remain an open item pending receipt of an appropriate analysis from the applicant.
Our review has confirmed that the WNP-3 containment isolatinn system meets the explicit requirements of GDC 54, 55, 56, and 57, oxcept as discussed below.
The isolation provisions for the chemical and volume control system (CVCS) charging Line (penetration 41) and the CVCS reactor coolant pump (RCP) seat injection Line (penetration 93) conform to GDC 55 except that the power-operated isolation valves out-side the containment are normally open and are closed by remote-canual control from the control room.
As discus. sed above, the CVCS charging Line and the CVCS reactor coolant pump seat injec-tion Line are essential Lines that have an impact on plant safety.
The isolation valves witL also be subject to Type C leak testing.
l Based on the guidance provided in SRP 6.2.4, we find the use of rer.ote-manual instead of automatic isolation valves acceptable.
We, require, however, that Class 1E emergency power be provided to valves 2CH-VQ040 and 2CH-VQ005 in these lines; because of the cbove this is an open issue.
The safety injection system (SIS) recirculation sump discharge lines (penetrations 23 and 24) have only one isolation valve in oach Line, outside containment.
If isolation valves were provided 1 u
m inside containment, they would be submerged fotLowing an accident.
Since these Lines have an important safety function, system reli-cbility is greater with only one isolation valve in each line.
- Also, the SIS is a closed enginered safety feature system outside contain-Cent whose integrity is appropriately maintained throughout plant life.
Based on the provisions of SRP 6.2.4, we find the single isolation valve design in the SIS recirculation sump discharge lines to be acceptable.
We wilL, however, require the applicant to discuss the adequacy of the criteria used in the design of the piping between the containment and the isolation valve, and the valve itself, and the Leakage control pro-visions on the penetration or in the penetration area.
The reactor building vacuum relief isolation valves (penetrations 65 and 75) are normalLy closed and would only 'open in response to a high vacuum signal in the reactor building.
The inboard check valves ensure that flow would always be into the containment.
Based on the system design and operating requirements we find the containment isola-tion provisions for the reactor building vacuum relief Lines acceptable.
'Wa have reviewed information provided by the applicant to demonstrate compliance with the provisions of NUREG-0737 Item II.E.4.2, "Contain-cont Isolation Dependability."
As previously described, the applicant has complied with the provisions regarding diversity in parameters sensed for initiation of containment isolation, identification of es-santial and non-essential systems, automatic isolation of nonessential systems, and closure of containment purge and vent isolation valves on a high radiation signal.
In addition, the FSAR states that alL power-operated isolation valves wilL remain in their accident position after an accident signal clears until deliberate operator action is taken
- 21
p I
to reopen the valves.
The containment setpoint pressure that initiates containment isolation should be reduced to the minimum value compat'ible with normal operating conditions.
The containment setpoint pressure and the justification for it should be provided by the applicant; this information wilL be reviewed by the staff in conjunction with the development of the plant technical specifications.
Fi na l Ly, the applicant has committed to keep the 48-inch containment purge valves closed during the operational conditions of power operation, startup, hot standby, and hot shutdown and verify that the valves are closed at least every 31 days.
This is acceptable, except that the purge valves should be sealed cosed (either electrically or mechanically) in accordance with SRP 6.2.4.
Except for the two issues identified cbove, namely, justification for the containment isolation setpoint pressure and a commitment to seat close the purge valves, we conclude that the applicant has complied with the provisions of NUREG-0737, Item II.E.4.2.
The containment isolation system meets the provisions of Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water,
Steam, and Radioactive-Waste-Containing Components of Nuclear Power
~
Plants," 1.29, " Seismic Design Classification," and 1.141, " Containment Isolation Provisions for Fluid Systems."
The containment isolation valves are designed in accordance with ASME Section III Class II requirements and Quality Class I.
We conclude that the WNP-3 con-tainment isolation system meets the requirements of GDC 54, 55, and 57, and NUREG-0737 Item II.E.4.2, and conforms to SRP 6.4.2 and CSB BTP 6-4, with the exception of the six issues summarized below:
r,
9
- - + ~
,n 1.
The applicant should upgrade the isolation provisions for the containment vent make up line (P-80); it is the NRC position that redundant power-operated, automatic isolation valves should be provided, which close in less than 5 seconds and fait closed upon loss of power to the valve operator; 2.
The applicant should provide an analysis of the effect purge system operation at the time of a LOCA on the minimum con-tainment pressure analysis for ECCS evaluation.
3.
It is the NRC position that the applicant provide Class IE emergency power to valves 2CH-VQ040 and 2CH-VQ005 in the reactor coolant pump seat injection and chemical and volume control paths; 4.
The applicant should confirm the design adequacy of the piping between the containment and the isolation valve, and the valve, in the SIS recirculation sump discharge Lines (penetrations 23 and 24), and the afitLiated Leakage control provisions.
5.
The applicant should justify that the containment isolation setpoint pressure is the minimum value compatible with normal operating conditions.
6.
The applicant should commit to seat close the 48-inch purge valves during operating modes requiring containment integrity. i
6.265 Combustible Gas Control in Containment FolLowing a LOCA, hydrogen may accumulate within the containment as' a result of:
- 1) hydrogen dissolved in reactor coolant system; 2) metal-water reaction between the zirconium fuel cladding and the reactor coolant; 3) corrosion of metals by emergency core coolant and containment spray solutions; and 4) radiolytic decomposition of the post-accident emergency cooling water.
The applicant has provided a combustible gas control system (CGCS) to monitor and cont ~rol the hydrogen cdncentration in containment fotLowing a LOCA.
The CGCS includes the containment hydrogen analyzers, the containment hydrogen recombiners, and the con-tainment hydrogen purge system.
The hydrogen analyzer system consists of two redundant subsystems, each of which can take samples from six locations within containment and one Location in the shield building annulus.
The hydrogen recombiner system consists of two stationary 100% capacity thermal (electrical) recombiners located within the containment.
Both the hydrogen analyzer system and the hydrogen recombiner system are designed to Safety Class 2 and Seismic Category I standards, and are powered from Class IE power sources.
The recombiner wilL be started manually from the control' room by the operator upon indication of a hydrogen concentration of greater than 3.0 volume percent.
Each of the two Westinghouse electric hydrogen recombiners is capable of processing 100 tefm of containment atmosphere for post-accident hydrogen control.
The staff has reviewed tests that were conducted for a full-scale prototype and a production recombiner.
The tests consisted of l. _ _ _ _ _
proof-of principle testing, testing on a prototype recombiner, environmen-tal qualification testing, and functional tests for a production recombiner.
(These tests are described in WCAP-7820 and its supplements.)
The results of these tests demonstrate that the recombiner is capable of controlling the hydrogen in a post-LOCA containment environment.
A purge system has been provided, in addition to the hydrogen recombiner system, in accordance with Section 50.44 of 10 CFR Part 50.
The purge system consists of two 100 percent capacity exhaust trains and a single cakeup train.
The applicant has analyzed the production and accumulation of hydrogen within the containment from the sources discussed above.
SRP 6.2.5 rocommends that the analysis of hydrogen production should be based
,on the parameters listed in Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design basis for combustible gas control systems.
The applicant has been requested to confirm that their analysis is in conformance with RG 1.7.
The applicant's analysis shows that one electric hydrogen recombiner actuated at a containment hydrogen concentration of 3.0 volume percent is capable of limiting the hydrogen concentration in containment to below the R.G.
1.7 Lower flammability limit of 4.0 volume percent.
The applicant should discuss the emergency procedures that wilL be in effect to guide the operator in actuating the hydrogen analyzer.
?
e The applicant has evaluated the possibility for pocketing of hydrogen in the containment following a LOCA and concluded that pocketing of,
hydrogen is not very likely.
This finding is based on the open, internal design of the containment, the low hydrogen generation rates from the various potential sources, and the effectiveness of hydrogen cass transport by convection.
Based on our review of the applicant's rationale, we agree with the applicant's conclusion that pockets of flammable hydrogen are not likely to form.
We conclude that the CGCS satisfies the design and performance require-cents of Section 50.44' of 10 CFR Part 50, " Standards for Combustible Gas Control Systems in Light Wate Cooled Power Reactors," the guidelines of Regulatory Guide 1.7 and the requirements of GDC 41, 42, 43, and 50, and is acceptable, provided the applicant justifies the hydrogen production analysis, and adequate emergency procedures are in place to guide the operator in actuating the hydrogen analyzers and hydrogen recombiners.
6.2.6 Containment Leakage Testing Program The containment design includes the provisions and features necessary to satisfy the testing requirements of Appendix J to 10CFR Part 50.
The de-c sign of the containment penetration and isolation valves permits periodic Leakage rate testing at the pressure specified in Appendix J to 10CFR Part 50.
Included are those penetrations that have resilient seats and expansion bolLows; i.e., air Locks, emergency hatches, and electrical penetrations.
The containment leakage testing program complies with the requirements of Appendix J to 10CFR Part 50.
Such compliance provides adequate assurance that containment Le'aktight integrity can be verified throughout service i
Lifetime and that the leakage rates wilL be periodically checked during servi.ce on a timely basis to maintain such Leakages within the specified Limits of the Technical Specifications.
The plant's Technical Specifica-tions wilL contain appropriate surveitLan'ce requirements for containment Leak testing, including test frequencies.
Maintaining containment Leakage rates within such limits provides reasonable -
ossurance that, in the event of any radioactivity releases within the con-tainment, the loss of the containment atmosphere through Leak paths wilL not be in excess of acceptable limits specified for the site; i.e.,
the resultant dose wilL be within 10CFR Part 100 guidelines in the event of a design basis LOCA.
We conclude that the applicant's program complies with the requirements of Appendix J and with the requirements of goc 52, 53, and 54, and there-fore, is acceptable.
r a.
.w g.s ENCLOSURE 2 u
DRAFT SAFETY EVALUATION REPORT OUTSTANDING ISSUES WNP-3 CONTAINMENT SYSTEMS BRANCH 1.
The forces and moments acting on the reactor vessel folLowing the design basis reactor coolant system break within the reactor cavity is required.
This information may be either plant (WNP-3) specific or shown to be ap-plicable to WNP-3.
The guidelines of SRP 6.2.1.2 and Section 3.2 of NUREG-0609 should be folLowed.
2.
A minimum containment pressure analysis to suopor-t emergency core'ccoling system capability studies that is applicable to WNP-3 should be provided.
3.
The applicant's response to IE ButLetin 80-04, Main Steam Line Break with Continued Feedwater Addition, is required.
4 Additional information is required concerning the applicant's shield building pr wssure response analysis.
In partciutar, a more detailed i
des:ription of the WATEMPT code is naeded, including the assumptions m.i d e in the analysis, to determiae the conservatism in the recults.
i 5.
The applicant should upgrade the isolatTon provisions for the containment vent ma!.e up Line (P-AO).
It is the staff's position that the containment isolation valves in Lir,s ?-80 should be automatic, power operated valves having less than 5-second closure times, and shoutd fait close upon loss of power to the valve operators.
6 The applicant should commit to seat close the 48-inr5 purge valves either electrically or mechanically.
3 7.
The applicant should justify that the setpoint pressure for contaihment isolation is the minimum value compatible with normal operation condi-
^
tions.
8.
The applicant should provide an analysis of the effect of purge system operation at the time of a LOCA on the minimum containment pressure analysis for the ECCS performance evaluation.
9.
It is the staff position that the applicant provide Class IE emergency power to valves 2CH-VQ040 and 2CH-VQ005 in the reactor coolant pump seat injection and chemical and volume control paths.
- 10. Wi t h regard to the SIS recirculation sump discharge lines (penetration 23 and 24), the applicant should discuss the adequacy of the criteria used in the design of the piping between the containment and the isola-tion valve, and the valve itself, and the Leakage control provisions on the penetration or serving the penetration area.
- 11. The applicant should discuss the basis for the hydrogen production
. analysis and justify any deviations from RG 1.7.