ML20213D835
| ML20213D835 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/09/1981 |
| From: | Speis T Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0396, CON-WNP-396 NUDOCS 8110230504 | |
| Download: ML20213D835 (95) | |
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OCT ; sal Docket No. 50-397 t
MEMORANEUM FOR: R. Tedesce, Assistant Director for Licensing. DL FROM: T. Spets, Assistant Director for Reactor Safety, DSI
SUBJECT:
DRAFT SAFETY EVALUATION REPORT FOR WASHINGTON PUBLIC -
POWER SUPPLY SYSTEM (WPPSS) NUCLEAR PROJECT NO. 2 Plant Name:
Licensing Stage:
Docket tb.: 50-397 Washington OL Public Power Supply System Nuclear Projec[P Responsible Branch: LB #1 g[ cp,,f ,,f,6 .,
, .c Project Manager: R. Auluck ' 7 ' '7i >;
Requested Completicn Date: October 21, 1981 Review Status: Additional information is required ti- gO[1 g IE87 l
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.". " 8cE c%t 'dyp, Our draft input for the safety Evaluation Report (SER) for the n ington- s' Public Power Supply System, Nuclear Project No. 2 (WNP-2) is enci 'ed14Wid" final draft for this input is currently scheduled for February 12,1982. ~
This draft report has been prepared by the Containment Systems Branch after having reviewad the Final Safety Analysis Report (FSAR), the Design Assess-ment Report, the applicant's plant unique pool dynamic loads reports and the commitments, made by the applicant during the September 14-17, 1981 meetings, that will be incorporated in Amendment 19 to the FSAR. The Brookhaven Na-tional Laboratory assisted the staff in reviewing the piant-unique pool dy-namic loads reports.
The following summarizes the outstanding items identified in the draft SER.
Additional information from the applicant is needed for us to complete our i review.
I
- 1. Forces and Moments Across Pressure Vessel l
The applicant's analysis of the subcompartment pressure response does not in-clude the forces and moments on the reactor pressure vessel. During the September 14-17, 1981 meeting, the applicant indicated that they will provide this information by October 2, 1981. We have not received the above mentioned information. We will report our finding regarding this item upon receipt of j the applicant's submittal. ,
l 2. Plant-Unique Steam Condensation Load Specification l
In August 1981, the staff issued NUREG-0808, " Mark II containment Program Load j E, valuation and Acceptance Criteria." In its letter of August 13,1981, the l applicant stated that the generic condensation oscillation load specification,
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GCT ew R. Tedesco pl ant. The applicant also stated that they will submit a detailed report by December 15, 1981 to show that the chugging load specification is the governing load for structures, piping and equipment in WNP-2. We informed the applicant, during the September 14-17, 1981 meeting, that this review of plant unique condensation oscillation load specification might require a 6-12 month review. We will report any progress regarding this item upon receipt of the applicant's report.
- 3. X-Quercher Air Clearing Load We have reviewed the applicant's proposed plant unique SRV air clearing load methodology and expressed some concerns. The applicant indicated that they will respond to our concerns by December 15, 1981. We will report our find-ing regarding this item upon receipt of the applicant's response.
- 4. Pool Temperature Transient NUREG-0487 provides that the applicant should perform plant unique ardyses for pool temperature responses to transients involving SRV operations, to demonstrate that the plant will operate within the limit of 210 F. These analyses are currently scheduled for submittal during the week of December 15, 1981. We will report our findings in a revision to this SER.
It should be noted that we have requested the applicant to perform an in-plant SRV test to verify the boundary pressure loads and the difference between bulk and local temperature. The applicant indicated in the September 14-17, 1981 meetings that they will provide us with their position regarding this test by December 15, 1981.
- 5. TMI-Related Issues This information will be submitted to the staff by December 15, 1981. We will report our finding regarding these items in a revision to this SER. ,
- 6. Secondary Containment Bypass Leakage i We have requested the applicant to provide the flashing analysis that was done l to justify eliminating the feedwater lines as bypass leakage paths. The appli-cant indicated during the September 14-17, 1981 meetings that the analysis will be provided by September 25, 1981. We have not yet received this analysis. We will report our finding regarding this item upon receipt of the applicant's /
submittal. ',
- 7. Containment Isolatio3 We have requested the applicant to design the recirculation system control l valves' hydraulic lines to satisfy the requirements of GDC 55. The applicant l . indicated that they will provide their position to us by September 25, 1981. ,
In addition, we have requested the applicant to design the recombiner scrubber omco . . . . . . . . . . .. .. .. .l. .l. ..
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return line to the suppression pcol to satisfy the requirements of !!DC SS. We have not received the applicant's position regarding this item. We will ree port our finding regarding this item upon receipt of the applicant's response to our concern.
- 8. confirmatory Analyses Except for the arci of subcompartment analyses, we have received sufficient informatien to conduct our own confimatory analyses. Ua will report our conclusion regarding our independent analyses in a revision to the SER, {
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- 9. Hydrogen Recombiner System The applicant has proposed to use a catalytic hydrogen recombiner subsystem manufactured by Air Products and Chemicals, Inc. (AP).- Both the Chemical Engineering and Equipment Qualification Branches are conducting their reviews of this system to determine the acceptability of AP reconbiner systems. He will report our finding regarding this issue in a revision to this input for the SER.
It should be noted that the applicant's submittal of additional information, regarding some of the major areas of review, is scheduled for December 15, 1981 which does not give the staff enough tice to meet the SER schedule of Fenruary 12, 1981. He recommand that you arrange to have WPPSS accelerate the submittal by at least two weeks or that the SER date be delayed correspondingly.
C.timi c1gned F7
- 1. a a. B,ht1ar Thenis P. Spels, Assistant Of rector
- for Reactor Safety 1
Division of Systems Integration
Enclosure:
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. b Contairiment Systems Branch Draft Safety Evaluation Report Washington Public Power Supply System NUCLEAR PROJECT NO. 2 Occket No. 50-397 6.2 Containrient Systems the contai6 ment systems for Washington Public Power Supply System (WPPSS)
Nuclear Project No. 2 include a Mark II type containment structure as the primary containment, a secondary containment, surrounding the primary contain-ment and nousing equipment essential to safe shutdown of the reactor and fuel storage facilities.
The secondary containment is designed to confine the leakago frem the primary contai7 ment of airborne radioactive materials.
6.2.1 (catainment FunctionO Desion _
The brimary containment is a free standing steel pressure vessel enclosed in a reiaforced ccacrtte. biological shield wall and separated t,y compressible isolation material. Figure 6.1 shows the principal features of the Mark II over-anc-under contuivient cenfiguration. The drywell is in the form of a truncated feet. cone, closed by a steel dcce with a net free volume of 200,540 cubic This dry wl1 touses the reactor vessel, the reactor coolant recirculation loops arid other branch connecticns of the reactor primary system. The drywell is designed fo'r aa interral pressure af 45 pounds per square inch gauge and a
'emperature of 340 degrats Fanrenheit.
The pressure . suppression chamcer !s in the shape of a cylinder located below the drywell and separateo from the drywell by a reinforced concrete slab. This scopression pcci) with ;;etchfree ybervolaes consis4Sfof an air region and a water region (suppression respectively. 141,184 cubic feet and 112,197 cubic feet, The suppression chamoer is designed for an internal pressure of 45 pounos per square inch gauge ana a ten.perature of 275 degrees Fahrenheit .
The suppression pool serves as a . heat sink for postulated transients and acticente and as a source of cooling water for the emergency core cooling system.
In the case of transients that result in a loss of the main heat sink, energy would be transferred to the pool by the discharge piping from the reactor system's safety / relief valves.
In tne event of a loss-of-coolant accident within the drywell, the drywell atmosphere is vented to the suppres-sion chaxber, through a series of downcomers to provide the energy transfer path.
The vent system consists of 102 straight down oipes (downcomers), 84 downcomers each has a 24-inch ncminal diameter.
nominal diameter. The other 18 downccmers each has a 28-inch pool and are submerged about 12 feet in the pool Thewater.These purpose of the pipes extene f downcemers is to channel the mass and energy released from a postulated break 6-1 W +
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, a in the primary system in the form of uncondensed steam and water. The mixture which also includes air is directed into the suppression pool where the steam is condensed and air released to the air space above the pool. To satisfy the design basis as a low leakage barrier, the primary containment system is designed for a leakage rate of 0.5 percent of the containment volume per day at the design pressure of 45 pounds per square inch gauge.
6.2.1.1 Containment Analysis Our review of the containment included the temperature and pressure responses of the drywell and wetwell to a spectrum of loss-of-coolant accidents; suppres-sion pool dynamic effects during a loss of-coolant accident and following the actuation of one or more reactor coolant system pressure relief valves; the capability to withstand the effects of steam bypass from the drywell directly to the air region of the suppression pool; and the external pressure capability of the containment. In addition, our review considered the applicants' pro-posed design bases and design cr'iteria for the containment and the analyses and test data in support of the criteria and bases.
Our review included consideration of the loads resulting from pool dynamic-related phenomena. Following a loss-of coolant accident, an air steam mixture will be forced from the drywell through the vent system into the sup-presson pool. The air component of the vent flow forms high pressure bubbles in the pool. Motion of the air bubbles results in an upward acceleration of the pool surface which can impact internal containment structures. Additional containment loads result as the steam portion of the vent flow condenses in the pool. Actuation of relief valves also results in containment loads.
Pressure waves are generated within tne suppression pool when the relibf valves discharge air and steam into the pool water.
Figure 6.2 illustrates the drywell and suppression chamber pressure response as a function of time following a recirculation line break, which has been identified as the design basis loss-of-coolant accident and the break yielding the limiting differential pressure across the drywell floor Figure 6.3. The pressure response will be discussed as short-term and long-term transients.
These transients are discussed separately in subsections of this report which follow.
l Figure 6.4 illustrates the drywell and suppression chamber pressure response as a function of time following a steam line break.
The drywell floor serves as the barrier separating the drywell from the suppression chamber. It is a conventional reinforced concrete structure which is supported by steel beams and a series of concrete columns. A flexible, pressure tight seal between the containment wall and the periphery of the diaphragm floor in tne form of 270* cmega shaped stainless steel. is provided to close the drywell floor to the primary containment gap to eliminate suppression pool bypass leakage.
l Our evaluation of the floor design is included in Subsections "Short-Term i Pressure Response," and "Drywell External Pressure and Floor Reverse Pressure,"
which follow. ~
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The drywell is divided into subcompartments by internal structures. Our evaluation of the subcompartment designs is discussed in Subsection "Subcompartment Pressure Analysis," which follows.
6.2.1.2 Review of Boilina Water Reactor Containment Technology Pressure suppression designs of reactor containments in the United States that are different frcm the Mark II containment are the Mark I or "lightbulb-torus,"
and the Mark III. A comparison of design parameters for the Mark I, II, and III containment types is provided in Table 6.1. -
The Mark I containment, Figure 6.5, is the first widely used design for boiling water reactor pressure suppression containments. In the Mark I design, Figure 6.5, the drywell consists of an inverted lightbulb-shaped vessel, and the suppression chamber is a torus-shaped steel vessel located below and encircling the drywell.
The vent systems consist of vent pipes, vent header, and downcomers. The typical design pressure for both drywell and pressure suppression chamber is 56 pounds per square inch gauge, except for Oyster Creek and Nine Mile Point I where the pressure suppression chamber design pressure is 35 pounds per square inch gauge.
The Mark III design, Figure 6.6, is the latest boiling water reactor pressure suppression containment design. In this design, the containment (pressure suppression chamber) cocpletely surrounds the drywell. The suppression pool is a 360 degree annular pool located in the bottom of the containment and retained between the containment wall and the drywell weir wall.
The wetwell and drywell of the Mark I and II designs are connected by a vent system which enters the suppression pool vertically at a constant submergence.
The Mark III design uses a horizontal vent system at variable submergence. In both the Mark I and II containments, the peak drywell pressure occurs in the range of approximately ten to fifty seconds following the onset of the accident, after the vent clearing process and during the vent flow part of the transient.
Wetwell peak pressures occur in about the same time frame for the Mark I and II designs, due primarily to the compression of drywell air which is carried over to the wetwell.
l The peak drywell pressure in the Mark III design occurs at about one second l during the vent clearing process. The peak wetwell pressure for a Mark III design occurs in the long term. The peak pressure is primarily a function of.
the capacity of the containment heat removal system. This peak pressure is not determined by the compression of drywell air carried over to the wetwell as in the case of the Mark I and II designs. This is due to the large volume of the wetwell in the Mark III design. The wetwell free volume for a Mark III design is about five times the drywell volume.
Mark II containments also experience a short-term drywell floor pressure differential during which the vent can flow occur either at the time of vent clearing or later, transient.
Generally, the Mark II plants with a relatively differential pressures. The WPPSShave small break-area-to-vent-area ratio vent clearing controlled peak floor Nuclear Project No. 2 containment falls in this category.
In the long term, both the drywell and wetwell reach a secondary i
peak pressure due to continued decay heat generation; however, this transient l
l 6-7
Reactor Containment Designs Comparison of Boiling Water Mark III 4 6.1 Mark II (Grand Gulf)
Mark I (WNP-2)
(Hatch 2) standing reinforced free concrete steel shell steel 270,128
,f construction 200,540 30 146,266 45 olume (cubicd feet) 56 NA jn pressure (poun s 0.5
. squtre inch gauge) 1.2 rate (percent per day) standing steel-lined free reinforced steel shell steel concrete 1811 (containment) 8 e'of construction 1.4 x 10 141,184 137,073 109,714 112,197 15 r volume (cubict) feet)90,550 45 301 volume (cubic fee56 0.35 0.5 esign pressure r day) (poundsper 1.2 3462 square inch 4025 (gauge)
_eak rate (percent pe 2537 3.54 Thermal power (megawatts 3.106 th2rmal 4.378 552 309 loss-of-coolant accidentbraak 216 area 0.0105 (square 006 feet)
Vent area (square feet) 0.0202
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Table 6.1 Comparison of Boiling Water Reactor Containment Designs Drywell Mark I Mark II Mark III (Hatch 2) (WNP-2) (Grand Gulf)
Type of construction steel shell free standing reinforced steel concrete Air volume (cubic feet) 146,266 200,540 270,128 Design pressure (pounds 56 45 30 per square inch gauge)
Leak rate (percent per day) 1.2 0.5 NA Wetwell Type of construction steel shell free standing steel-lined steel reinforced concrete (containment)
Air volume (cubic feet) 109,714 141,184 1.4 x 106 Pool volume (cubic feet) 90,550 112,197 137,073 Design pressure (pounds 56 45 15 per square inch (gauge)
Leak rate (percent per day) 1. 2 0.5 0.35 Thermal power (megawatts 2537 3462 4025 thermal Loss-of-coolant accident 4.378 3.106 3.54 break area (square feet)
Vent area (square feet) 216 309 552 Break area / vent area 0.0202 0.0105 .006 r
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is less severe than the short term transient and is therefore not controlling for establishing design pressure. The analytical models, assumptions, and methods used by the General Electric ' Company to evaluate the containment response during the reactor blowdown phase of a loss of-coolant accident are described in NE00-10320, "The General Electric Pressure Suppression Containment Analytical Model."
- 6. 2.1. 3 Short-Term Pressure Resoonse The limiting drywell, suppression chamber and drywell floor pressures occur during the blowdown phase of a loss-of-coolant accident transient. The dura-tion of the blowdown period is about 53 seconds following a postulated break in the recirculation line. In the long term, about two nours after a loss-of-coolant accident, both the drywell and wetwell reach a secondary peak pressure due to continued decay heat generation; however, this transient is less severe-than the short-term and is therefore not controlling for establishing design pressures.
The applicants have performed analyses of varied postulated primary system breaks including recirculation line, main steamline and a spectrum of liquid line and steamline breaks. Results of the analyses indicate that the recircu-lation line break yields the limiting drywell and suppression chamber pressure. The applicants, therefore, conclude that the recirculation line break is the design basis accident for the drywell and suppression chamber. Following' the postulated double-ended rupture of a 24-inch recirculation pump suction line break, the mass and energy released from the primary system pres-surizes the drywell. As the drywell pressure increases, water initially in the downcomers is accelerated cownward. This downward motion continues until the entire water volume is expelled. At this point, the air steam-water mixture i i begins to flow into the suppression pool The steam is condensed in the pool and air is released into the suppression chamber air region. The above process is called the vent clearing transient, which occurs less than one second following onset of the postulated accident. As shown in Figure 6.2, the maximum pressure differential between the drywell and pressure suppression l chamber occurs at the time of vent clearing. A detailed discussion of pool dynamics is presented in Subsection "Pcol Dynamics" which follows. The drywell pressure continually increases until it reaches the calculated peak-pressure of 34.7 pounds'per square inch gauge at 19.02 seconds after the onset of the accident. The drywell pressure then decreases slightly as the rate of energy dumped to the suppression pool via downcomers exceeds the rate of energy released into the drywell from the primary system. Following vent clearing the pressure in the pressure suppression chamber increases at about the same rate as the drywell pressure due to the transfer of steam and noncondensibles from the drywell. The suppression chamber reaches a calculated peak pressure of 27.6 pounds per square inch gauge at 55.0 seconds after the accident. 6-11 l _ ., - - - - -
At about 108 seconds the emergency core cooling system injection water floods the reactor vessel to the icvel of the break and the emergency core cooling system flow cascades into the drywell. This results in condensation of the steam in the drywell and a rapid reduction in the drywell pressure. As soon as the drywell pressure drops below the suppression chamber pressure, the dry-well vacuum breakers will open and noncondensible gases from the suppression pool air volume will flow back into the drywell. The limiting break for the drywell floor is the postulated double-ended, rupture of a recirculation pump suction line. The applicants' calculated peak pressures for the design basis accident are 34.7 and 27.6 pounds per square inch gauge for the drywell and suppression chamber respectively. The design pressure of the drywell and suppression chamber is 45 pounds per square inch gauge, which provides a 22.9 percent design margin for the drywell and 38.6 percent for the suppression chamber. The applicants calculated a peak pressure differential across the drywell floor of 19.4 pounds per square inch at 0.776 seconds, the time of vent clearing. The design pressure is 25 pounds per square inch. The resulting margin for the drywell floor is 22 percent. We performed an analysis of the containment pressure response using the CONTEMPT-LT computer code. Our calculations of the peak pressures confirm those calculated by the applicants. Based on the design margin and our own independent verification of the analytical results, we conclude that the design pressures for the drywell, suppression chamber and the drywell floor are acceptable. 6.2.1.4 Lono-Term Pressure Resconse Following the snort-term blowdown phase of the accident, the suppression pool temperature ar.d suppression chamber pressure will continuously increase due to the input of decay heat and sensible heat into the suppression chamber. Referring to Figure 6.2 between about 20 and 100 seconds after the accident, the drywell pressure has stabilized to approximately five pounds per square inch above the suppression chamber pressure. This differential pressure corres-ponds to the submergence of the downcomers. At a later time, the drywell and suppression chamber pressures will equalize due to the return of air from tne suppression chamber. During this time period, the emergency core cooling system pumps, taking suction from the suppression pool, will have reflooded the reactor pressure vessel. Subsequently, emergency core cooling system water will flow out of the break and flow into the drywell. This relatively cold emergency core cooling system water will condense the7 steam in the drywell and bring the drywell pressure down rapidly as shown in Figure 6.2 at about 100 seconds after onset of the accident. At 10 minutes following the accident, the containment cooling mode for the residual heat removal system is activated and suppression pool water is circulated through the residual heat removal system heat exchangers, establishing an energy transfer path to the service water system and ultimate heat sink. In the long-term analysis, the applicant accounted for potential post-accident energy sources. These include decay heat, sensible heat and metal water reaction energy. The applicants' long-term model also assumes that the suppression chamber atmosphere is saturated and equal to the suppression pool temperature at all times. Therefore, tae suppression chamber pressure is equal 6-12
to the sum of the partial pressure of air and the saturation pressure of water corresponding to the pool temperature. Based on the above assumption, the applicants calculated a peak suppression pool temperature of 220 degrees Fahrenheit for the most limiting residual heat removal system cooling mode; i.e., only one residual heat removal system cooling loop. The calculated long-term, secondary peak suppression chamber pressure is about 18 pounds per square inch gauge. The suppression chamber is designed for 45 pounds per square inch gauge and 275 degrees Fahrenheit. On the basis of our review of the applicants' analysis, we conclude that the suppression chamber design pressure and temperature are acceptable. 6.2.1.5 Drywell External pressure and Floor Reverse Pressure WPPSS Nuclear Project No. 2 plant is equipped with three 24-inch reactor building-to-wetwell vacuum relief lines to prevent excessive vacuum from developing in the primary containment. Nine 24-inch wetwell-to-drywell vacuum relief valves attached to particular downcomers in the suppression chamber are also provided to return noncondensibles from the wetwell to the drywell to prevent excessive upward pressure across the diaphragm floor after LOCA. The applicants have investigated events which may result in drywell external pressure and drywell floor reverse pressures and determined that simultaneous actuation of both drywell sprays following a small line break yields the maximum external pressure differential. The drywell initial conditions were conserva-tively assumed to be 150 degrees Fahrenheit, zero pounds per square inch gauge and 30 percent relative humidity. The suppression chamber was assumed to be at 50 degrees Fahrenheit, zero pounds per square inch gauge and 100 percent relative humidity. The spray water was assumed to be at a temperature of 50 degrees Fahrenheit. The postulated small line break serve to pressurize the drywell to force all noncondensibles (which tends to hold up the depressur- ' ization) to the wetwell air space region. The analyses assumed failure of two wetwell-to-drywell vaccum breaker systems and an additional failure of wetwell-to-drywell or reactor building-to-wetwell vacuum breaker system. The applicants calculated a maximum negative overall containment differential pressure across the containment wall of less than 2 pounds per square inch. The containment I design external differential pressure is 4.0 pounds per square inch. However, since the compressed insulation between the concrete biological shield and the containment exerts a uniform 2 pounds per square inch, the reactor building pressure containment maypressure. be no greater than 2 pounds per square inch above the primary
?
For the drywell floor upward pressure, the applicants analyzed the case described above assuming that the drywell initial conditions to be 135 degrees Fahrenheit, 0.75 pound per square inch gauge and 50 percent relative humidity. The applicants assumed that three out of the total of nine sets of wetwell-to-drywell floor vacuum breakers were not functioning. The vacuum breakers are provided to limit the drywell vacuum conditions. Each set of vacuum breakers consists of two valves in series to reduce potential bypass leakage. For the limiting case the applicants calculate a peak reverse pressure differential of less than 2 pounds per square inch. The design reverse pressure differential of the drywell floor is 6.4 pounds per square inch. Our analysis using the \ i 6-13 l .
CONTEMPT LT computer code are currently under way. We will report our finding regarding these analyses in the SER. 6.2.1.6 Subcompartment Pressure Analysis Within the primary containment, internal structures form subcompartments or restricted volumes which are subject to differential pressures following postu-lated pipe ruptures. In the drywell there are two restricted volumes, the annulus formed by the reactor vessel and.the sacrificial shield, and the drywell head region which is a cavity between the drywell head and the refueling bulkhead plate. Since the suppression chamber is virtually an open space, no restricted volume exists. The applicants performed two pressure response analyses for the drywell head region postulating a break in the reactor core isolation cooling system head spray line and a recirculation suction line. The applicants calculated the pressure differential for the head spray line break using the computer code RELAP-4/M00 5. The mass and energy released for this analysis of 398.2 pounds mass per second and an enthalpy of 474,694 BTU /sec. The maximum calculated pressure differential of 10.45 and 7.05 pounds per square inch are obtained for the head spray and the recirculation suction line respectively. We will report our finding regarding this analysis in the SER. The applicants submitted the results of calculations of pressure differentials across the annulus formed by the sacrificial shield wall and the reactor pressure vessel. The analysis of the annulus for a postulated line break considered design pressure for the shield wall. Separate analyses were conducted by the applicant for two different breaks to determine the limiting case. The postu-lated break analyzed are the recirculation outlet line and feedwater line breaks. For these breaks, the proposed General Electric analytical methodology for determining the short-term mass and energy release for postulated recirculation ano feedwater line breaks was utilized. A comparative evaluation was performed by tt.e staff of the recirculation line breaks. These analyses modeled the essential features of the primary system and assumed the Henry-Fauske critical flow model for subcooled break flow and the Moody slip model for saturated break flow. The results of these analyses indicated that the mass and energy release i rates determined by utilizing the proposed General Electric methodology are conservative. Therefore, we find that mass and energy release rates acceptable for the WPPSS Nuclear Project No. 2 application.
?
The applicants calculated a peak differential pressure within the sacrificial shield annulus of 95 pounds per square inch absolute for the recirculation line break. I t The RELAP 4/M00 5 computer code was used to perform the analyses. The actual l pressure transients calculated by the applicant were utilized in dynamic structural analyses to determine the sacrificial wall design. l The applicants have not provided additional information to allow us to complete t our review of the forces on the reactor pressure vessel affecting the design l of the support skirt. Calculation of the forces and moments across a pressure vessel represents an additional complexity to the problem. Recognizing this 6-14 l
refinement in the analyses, we have developed a Category "A" task action plan (A-2) to further investigate the methods and assumptions used for this analysis. The objective of this task is to develop guidelines for both the assumptions and analytical models to be used. In January 1981, we issued a report, NUREG-0609, " Asymmetric Blowdown Loads on PWR Primary Systems," to address guide-lines for calculating the forces and moments for pressurized water reactors. However, the same assumptions could be used to calculate the pressure outside the pressure vessel for boiling water reactors. We will review the applicants' analysis in light of NUREG-0609, to determine the acceptability of the forces and moments calculated by the applicants when it is submitted. We will report our findings in the SER. 6.2.1.7 Steam Byoass of the Sucoression Pool During a postulated primary system line break inside the drywell, possible bypass leakage paths between the drywell and suppression chamber air space could result in excessive containment pressures. The control of such bypass paths is important to ensure that the design pressure of the containment is not exceeded. There are several potential sources of steam bypass to the sup-pression pool air space in the WPPSS Nuclear Project No. 2. They include cracking of the drywell floor, including the reinforced concrete, the wetwell-drywell vacuum breakers or around the penetrations in the drywell floor. The applicants assumed a maximum allowable leakage area of A/JK equal to 0.05 square feet to exist between the drywell and suppression chamber following a small line break accident. The applicants did analyze the containment to determine the capability for pool bypass leakage. The basic model included passive heat sinks, containment spray manual actuation 30 minutes following the 30 pounds per square inch gauge pressure signal and evaporation due to impingement of spray water on the hot downcomer surface. The applicants' analysis indicates that the 30 pounds per square inch gauge pressure signal was reached in about 43 minutes for the postulated event; 30 minutes later, the drywell and wetwell pressure reached 40 and 35 pounds per square inch gauge respectively. The analyses also indicates that, if the spray were not actuated, the suppression chamber pressure will reach the design value of 45 pounds per square inch gauge in 81 minutes at onset of the accident. Based on the results of these studies, the applicant indicated that the WPPSS Nuclear Project No. 2 bypass capability is adequate and therefore, manual wet-well spray actuation is, adequate since the operator has enough time to initiate the pray before the suppression chamber design pressure is reached. In order to determine the adequacy of the steam bypass leakage capability, we performed an independent confirmatory analysis. Our analysis consisted of a transier.t analysis for a postulated small line break that takes credit of passive heat sinks in the wetwell chamber, and an allowable bypass leakage area of A/4K~ equal to 0.05 square feet. This analysis indicates that the wetwell spray need i not be activated until about 40 minutes following onset of the 30 pound per square inch gauge signal. t Based on the results of this study, which shows that at least 30 minutes is available for the operator to actuate the containment spray before the containment design pressure is reached, and since the applicant has committed to: l 6-15 1
(1) manually actuate the wetwell spray after the 30 pound per square inch gauge drywell pressure signal is reached, (2) perform high pressure leakage test prior to operation, (3) establish the acceptance criterion for the leakage tests to be the measured leakage less than 10 percent of capability, (4) provide redundant position indicators on all vacuum breakers with indication and redundant alarms in the control room, (5) test for operability of the vacuum breakers at monthly intervals, and (6) perform a low pressure test at each refueling outage. we conclude that WPPSS Nuclear Project No. 2 steam bypass capability is adequate. 6.2.1.8 Pool Dynamics 6.2.1.8.a Mark II Pool Ovnamics History In the course of the General Electric Company testing program for the Mark III pressure suppression containment program, new containment loads associated with a postulated loss-of-coolant accident were identified which were not explicitly included in the original design of Mark II containment. These loads result from the dynamic effects of drywell air and steam being rapidly forced into the suppression pool during a postulated loss-of-coolant accident event. Other previously unaccounted for pool dynamic loads result from the actuation of safety / relief valves in the Mark II containment. In view of the potential significance of these loads, it was determined that a reassessment of the Mark II containment system design would be required. A letter was sent by us to each domestic Mark II owner on April 11, 1975 notifying them of the need for this reassessment. As a result of our letter, an "ad hoc" Mark II Owners Group was formed which ' is an organization of all domestic utilities owning Mark II facilities. They have engaged General Electric Company as their pro of the Mark II containment pool dynamic concerns. gram manager A program wasfor resolution developed to l establish generic pool dynamic loads, load combinations and design criteria. The program consists of p a numeer of experimental and analytical tasks. The l tasks form the basis for establishing the generic Mark II pool dynamic loads. i ! 6.2.1.8.b Description of Phenomena Loss-of-Coolant Accident Pool Dynamics: Figure 6.6 shows the sequence of events following a postulated loss-of-coolant accident and the potential loading conditions associated with those events. Following a postulated loss-of-coolant accident, the drywell pressure increases due to blowdown of the reactor system. Pressurization of the drywell causes the water initially in the vent system to be accelerated out through the vents. During this water expulsion process the resulting water jets cause impingement loads on local containment structures. 6-16
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Following vent clearing, an air / steam bubble forms at the vent exit which causes a hydrostatic pressure increase in the pool water resulting in a loading condition on the pool boundaries. The steam condenses in the pool. However, the continued addition and expansion of the drywell air causes the pool volume to swell, resulting in a rise of the pool surface. Upward motion of this slug of water creates a drag load on structures submerged in the pool and impact loads on unsubmerged structures located just above the initial pool surface. i After the pool has risen approximately 1.5 times the initial submergence of the main vents, the rising slug of water breaks apart. Subsequent pool swell involves a two phase air water froth which produces further structural-impingement loads. A gravity induced fallback of the pool returns the pool surface to the post-loss-of-coolant accident elevation. At about the time of slug breakup, the drywell floor can be subjected to an upward load due to an imbalance in pressure between the compressed air in the wetwell free air space and the air space and the air purged drywell volume. Following the pool swell transient, there will be a period of high steam flow through the main vent system. At these high steam flow conditions, the water / steam condensation interface oscillates due to bedble growth and collapse. These condensation oscillations result in an oscillatory load on the pool boundary. At low vent flow rates, the water / steam condensation interface can oscillate back and forth in the vents causing " chugging." The chugging action results in loads on both the downcomer vents and the containment boundaries. Relief / Valve Dynamics: Actuation of safety / relief valves produces transient loadings on components and structures in the suppression chamber region. Prior to actuation, the discharge piping of a safety / relief valve line contains atmospheric air and a column of water corresponding to the line submergence. Following safety / relief valve actuation, pressure builds up inside the piping as steam compresses the air in the line. The resulting high pressure air bubble that enters the pool oscillates in the pool as it goes through cycles of over-expansion and recompression. The bubble oscillations resulting frcm safety / relief valve actuation and discharge cause oscillating pressures throughout the pool, resulting in dynamic loads on pool boundaries and submerged structures. Severe steam condensation vibration phenomena can occur when high pressure, high-temperature steam is continously discharged at high mass velocity into the pool, if the pool is at elevated temperatures. These steam quenching vibrations also result f"n loads on pool boundaries and submerged structures. The characteristics of the safety / relief valve load varies depending on the discharge device (ramshead or quencher) located at the exit of the safety / relief valve line. The WPPSS Nuclear Project No. 2 utilizes the General Electric X quencher device in order to help mitigate pool temperature effects and dynamic forces. 6.2.1.8.c Mark II Owners Group - Generic Program
~
The Mark II Owners Group developed a generic program to establish pool dynamic 1.oads, load combinations and design criteria. The program includes a number 6-18
The characteristics of the safety / relief valve load varies depending on the discharge device (ramshead or quencher) located at the exit of the safety / relief valve ifne. The WPPSS Nuclear Project No. 2 utilizes the General Electric X quencher device in order to help mitigate pool temperature effects and dynamic forces. 6.2.1.8.c Mark II Owners Group - Generic Procram The Mark II Owners Group developed 2 generic program to establish pool dynamic loads, load combinations and design criteria. The program includes a number of analytical and experimental programs to provide the data base to support the proposed loads and load prediction models. The Mark II Owners Group's program consists of a "short-term program" or lead plant effort and a "long-term program." The purpose of the short-term program is to demonstrate that sufficient technical understanding of the pool dynamics phenomena and principles of interest exists to allow the utilization of the loads and methods prescribed in the Dynamic Forcing Function Information Report that was submitted in November 1975 for the licensing of Mark II plants. As a result, in some cases, a bounding interpretation of test data was utilized to assure conservatism in tne Dynamic Forcing Function Information Report loads. The primary purpose of the long-term program is to confirm the loads utilized in the short-term program. Data from a number of experimental programs have been used in conjunction with analytical models to support the short-term program loads specified in the Dynamic Forcing Function Information Report. The major Mark II-related experi-mental programs include the temporary tall test tank (4T) full scale, Electric Power Research Institute 1/13 scale, pressure suppression test facility 1/3 scale, and the Monticello in plant tests. In October 1978, we issued a report, NUREG-0487, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," to address the portion of the Mark II Owners Group's program that provides a generic methodology for establishing design basis loss-of-coolant accident and safety / relief valve i ' loads for the lead Mark II plants (La Salle, Zimmer, Shoreham). The load evaluations were conducted by us and our consultant, Brookhaven National Laboratory. Since the issuance of NUREG-0487, the Mark II owners submitted additional reports in which they proposed alternative load methodologies for use in the evaluation of Mark II plants. We and our consultant studied these alternative methodologies proposed 4y the Mark II Owners Group. As a result of these reports, we have issued Supplement I to NUREG-0487, on September 1980. This supplement contains an evaluation of the proposed alternatives to the lead plant acceptance criteria. In addition, the Mark II Owners Group conducted additional tests in a modified 4T facility to answer questions raised by us regarding the influence of vent length effects on the condensation oscillation loads. These loads are the result of condensing of steam exiting the vent lines. The results of these ' tests indicate that the condensation oscillation and chugging load specifica-tions set forth in NUREG-0487 for the lead Mark II plants need to be modified
. based on conservative interpretation of the new 4T data. We issued Supplement 2 to HUREG-0487 in February 1981 that contains our evaluation of an interim 6-18
e. !. of analytical and experimental programs to provide the data base to support the proposed loads and load prediction models. The Mark II Owners Group's j program consists of a "short-term program" or lead plant effort and a "long-l term program." The purpose of the short-term program is to demonstrate that sufficient technical understanding of the pool dynamics phenomena and principles of interest exists to allow the utilization of the loads and methods prescribed in the Dynamic Forcing Function Information Report that was submitted in i ' November 1975 for the licensing of Mark II plants. As a result, in some cases, a bounding interpretation of test data was utilized to assure conservatism in i the Dynamic Forcing Function Information Report loads. The primary purpose of the long-term program is to confirm the loads utilized in the short-term program. Data from a number of experimental programs have been used in conjunction with analytical models to support the short-term program loads specified in the Dynamic Forcing Function Information Report. The major Mark II related experi-i i i mental programs include the temporary tall test tank (4T) full scale, Electric Power Research Institute 1/13 scale, pressure suppression test facility 1/3 scale, and the Monticello in plant tests. 4' In October 1978, we issued a report, NUREG-0497, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," to address the portion of the Mark II Owners Group's program that provides a generic methodology for establishing design basis loss-of-coolant accident and safety / relief valve loads for the lead Mark II plants (La Salle, Zimmer, Shoreham). The load ! evaluations were conducted by us and our consultant, Brookhaven National j Laboratory. Since the issuance of NUREG-0487, the Mark 11 owners submitted additional , reports in which they proposed alternative luad methodologies for use in the , evaluation of Mark II plants. We and our consultant studied these alternative methodologies proposed by the Mark II Owners Group. As a result of these , reports, we have issued Supplement 1 to NUREG-0487, on September 1980. This supplement contains an evaluation of the proposed alternatives to the lead plant acceptance criteria. In addition, the Mark II Owners Group conducted additional tests in a modified 4T facility to answer questions raised by us regarding the influence of vent i length effects on the condensation oscillation loads. These loads are the result of condensing of steam exiting the vent lines. The results of these t i tests indicate that the condensation oscillation and chugging load specifica-tions set forth in NURE@ 0487 for the lead Mark II plants need to be modified based on conservative interpretation of the new 4T data. We issued Supplement 2 to NUREG-0487 in February 1981 that contains our evaluation of an interim condensation and chugging load specification for the lead Mark II plants 1 (La Salle, Shoreham and Zimmer). In addition, the Mark II Owners Group has proposed new condensation oscillation and chugging load specifications based on data obtained from the modified 4T facility. The new load specifications j evaluation has been issued in NUREG-0808 in August 1981. 6.2.1.8.d WPPSS Nuclear Project No. 2 Desian Assessment Report i
\ .
The applicants submitted a Design Assessment Report on the pool dynamic loads
- for the WPPSS Nuclear Project No. 2 Mark II containment. This report provides i
6-19
a description of the specific application of the generic Mark II pool dynamic loads and methods for WPPSS Nuclear Project No. 2 and the plant unique loads used in assessing the capability of the WPPSS Nuclear Project No. 2 containment and components to pool dynamic phenomena. A summary of our review status for each of the pool dynamic loads is presented in Table 6.2. This table provides a description of each load or phenomenon, the Mark II Owners Group's load specification, and references our review status and the applicants' position on each load. As indicated in Table 6.2, tne applicants agreed to adopt all but three of our generic triteria. These items relate to steam condensation oscillation and chugging loads (Load I.C.2 in Table 6.2) and quencher air clearing loads (Load II.B in Table 6.2). Alternative criteria were proposed by the applicants for these items. Our evaluation of these alternative criteria is provided below. . 6.2.1.8.e Steam Condensation Oscillation Load (Load I.C.2 in Table 6-2) In its letter no. G02-81-239 dated August 13, 1981, the applicants indicated that the generic condensation oscillation load specification definition developed for the Mark II Owners Group, and acceptable for those plants with reinforced concrete containments, has been determined to be excessive for WPPSS Nuclear Project No. 2 plant. The applicants contend that, based on the examination and evaluation of available test data, condensation oscillation loading is less critical than the chugging load and does not represent a governing load for structures, i piping and equipment in WPPSS Nuclear Project No. 2. The applicants indicated that a detailed report will be submitted by December 15, 1981 summarizing the results of these studies. We will report on our findings regarding this load specification upon completing our review of the pertinent information. 6.2.1.8.f Steam condensation Chuacing Load (Load I.C.2 in Table 6-2) In July 1981, a report titled " Chugging Loads - Revised Definition and Application Methodology for Mark II Containments (Based on 4TCO Test Results)," was submitted by the applicants in lieu of the generic chugging load method-ologyfoundacceptable)ythestaffinNUREG-0808. The application methodology for WPPSS Nuclear Project No. 2 containment accounts for the plant-specific parameters governing the response such as vent length, three-dimensional j multivent suppression pool geometr with sloped bottom and the flexibility of suppression pool structural boundary Seven key chugs, having significantly l larger pressure peaks and more powe r '.han the remaining chugs, obtained from the 4TCO tests were chtsen 't en. W r: all 4TCO data base at all frequencies. l These key chugs together with thirteen chugging traces from the same time l windows to which the key chug occurs are used to deduce seven single vent l impulsive acceleration sources used to develop the revised chugging load definition. Each single vent impulsive acceleration source is applied in phase at exit elevation of the three vents in each of the thirty-four radial lines where downcomers'are located in the WPPSS Nuclear Project No. 2. 6-20
l t , i
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Table 6.2 Conformance of WNP-2 Design to NRC Acceptance Criteria t Mark II Dwners Group NRC WNP-2 Position on load or Phenomenon Load Specification Evaluation Acceptance Criteria I. LOCA-Related Hydrodynamic Loads A. Submerged Boundary Loads 24 psi over pressure added to During Vent Clearing N local hydrostatic below vent II.A.1 [3] Acceptable exit (walls and basemat) - linear attentuation to pool surface. B. Pool Swell Loads
- 1. Pool Swell Analytical o, Model bi a) Air Bubble Pressure Calculated by the pool swell i
analytical model (PSAM) used in III.B.3.a.1 [1] Acceptable. calculation of submerged boundary loads. b) Pool Swell Elevation Use PSAM with polytropic II.A.2 [2] Acceptable. exponent of 1.2 to a maximum swell height which is the greater of 1.5 vent submergence or the elevation corresponding to the drywell floor uplift AP per NUREG 0487 criteria I.A.4.
; The associated maximum wetwell air compression is used for i design assessment.
i y v w
I i t i ti. - I Table 6.2 (Continued) t I
- Mark II Owners Group NRC WNP-2 Position on Load or Phenomenon Load Specification Evaluation Acceptance Criteria c) Pool Swell Velocity Velocity history vs. pool III.B.3.a.3,[a] Acceptable.
elevation predicted by the PSAM i used to compute impact loading on small structures and drag on gratings between initial pool N surface and maximum pool eleva-tion and steady-state drag between vent exit and maximum pool elevation. Analytical velocity variation is used up o, to maximum velocity. Maximum A, velocity applies thereaf ter up
" to maximum pool swell. PSAM predicted velocites multiplied by a factor of 1.1.
d) Pool Swell Acceleration predicted by the Acceleration PSAM. Pool acceleration is III.B.3.a.4 [1] Acceptable. utilized in the calculation of acceleration loads on sub-merged components during pool swell. e) Wetwell Air Wetwell air compression is Compression calculated by PSAM. II.A.2 [2] Acceptable. f) Drywell Pressure Methods of NEDM-10320 and NED0- III.B.3.a.6 [1] Acceptable. 20533 Appendix B. Utilized in PSAM to calculate pool swell loads.
I j lable 6.2 (Continued) Hark 11 Owners Group t'RC WNP-2 Position on j Load or Phenomenon toad Specification Evaluation Acceptance Criteria l' 2. Loads on Submergcd Haxim m bubble pressure predicted by the PSAM added III.8.3.b [1] Acceptable. uniformly to local hydrostatic below ven't exit (walls and Y liasement) linear attenuation to pool surface. Applied to walls up to maximum pool elevation.
- 3. Iopact toads a) Small Structures 1.35 s Pressure-Velocity III.8,3.t. ) {1] Acceptatile, correlation for pipes and I a beans based on PSTF impulse g data and f lat pool assumption.
Variable pulse duration. b) LarDe struci.ures None - Plar t unique load where III.B.3.c.6 [1] Acceptable. WMP-2 i appilcable. has no large
; Crite.-ia A.5 {3]
structures in the pool swell zotie.
! c) Gratics No impact load specified. III.B.3.c.3 [1] Acceptable.
P drag vs. open area corre- Criteria A.3 [3] 1ation and velocity vs. i elevatica history from the
! PSAM. P drag multiplied by dynamic load factor.
- 4. Wetwell Air Ccapression .
a) Wall leads Direct application of the PSAM III.B.3.d.1 [1] Acceptable. calculated pressure due to wet-wall coa.pression. b) Iliaphrage ti,3 ward loads 5.2 psid for diaphragm loadings cnly. 2.12.7 [3] Acceptable.
i i
~
t Table 6.2 (Continued) l I Mark II Owners Group NRC WNP-2 Position on j toad or Phenomenon Load Specification i Evaluation Acceptance Criteria
! 5. Asymmetric LOCA Pool Use 20 percent of maximum
~
Criteria A-4 [3] Acceptable. pressure statistically applied II.A.3 [2]
}
to 1/2 of the submerged bubble. C. Steam Condensation and Chugging Loads N
- 1. Downcomer Lateral Loads a) Single Vent Loads use single vent dynamic lateral (24 in.) load developed in NEDE-23806.
2.3.3.2 [3] Acceptable. Criteria B.I.a [3]
. ca b) tioltiple Vent loads Use multivent dynamic lateral d2 (24 in.) load developed in NEDE-21106-P 2.3.3.3 {3]
and NEDE-24791-P. c) Single / Multiple Vent Multiply basic vent loads by l Loads (28 in.) factor f=1.34 2.3.2.1 [3] R.I.b [3] i 2. Submerged Boundary Loads
, a) High/ Medium Steam Flux use method described in Condensation Oscillation ?.2.2.1.3 [3] Addressed in this NEDE-24288-P[4] report.
i Load 9
- b) Low Steam flux Chugging Representative pressure
? Loads 2.2.2.3 [3] Plant unique. ? fluctuation taken from 4TCO WNP-2 chugging I (NEDE-24285-P) test added to report submitted local hydrostatic. July, 1981.
Addressed in this , report. I I e
i Table 6.2 (Continued) Mark II Owners Group NRC WNP-2 Position on 8 Load or Phenomenon Load Specification Evaluation Acceptance Criteria c) Low Steam Flux Chugging Loads (continued)
- unifora loading Use method described ir. NEDE-conditions 24302-P[4]
- asymmetricloaling Representative pressure fluctuation taken from 4TCO test [NEDE-24285-P] applied as described in NEDE-24822-P.
II. SRV-Related Hydrodynamic Loads en A. Pool Temperature Limits for 210 degrees Fahrenheit A3 X quencher NRC Criteria II.1 and Acceptable. u, 11.3 [1] B. Quencher Air Cleaning Loads Mark 11 plants utilizing the Criteria II.2 [1] WNP-2 plant unique four ara quencher, use quencher improved X quencher load methodology described in load definition DFFR. based on CAORSO t test data. Addressed in this report. C. Quencher Arm and Tie-Down Loads Includes vertical and lateral III.C.2.e.2 [1] Acceptable. arm load transmitted to the basemat via the tie-down.
- 1) X quencher armloads Vertical and lateral loads 8
III.C.2.e.1 Acceptable. developed on the basis of bounding assumption for air / water discharge f rom the quencher and conservative com- ') binations of maximum / minimum
' bubble pressure acting on the quencher.
i I I
~
i i Table 6.2 (Continued) t i. Mark II Owners Group NRC i Load or Phenomenon Load Specification WNP-2 Position on
! Evaluation Acceptance Criteria
}* 2) X quencher Tie-down loads II.C.) above plus vertical transient wave and thrust III.C.2.e.2 [1] Acceptable loads. Thrust load calculated using a standard momentum N balance. Vertical and lateral moments for air or water clear-ing are calculated based on conservative clearing assumptions.
, III. LOCA/SRV Submerged Structure l , Loads g A. Air Bubble Loads
- 1. Standard Drag in Accelerating Flow fields Draft Coefficients are Acceptable with the Generic inethodology j
presented in Attachment 1.k of following modifica- acceptable. Plant the Zimmer FSAR cation unique flow fields are consistent with
, 1) Use Cg = C -1 t r. C.2.b from Table C-1 the F3 form la. of NURG-0808 for LOCA loads and wit!
- 2) For noncylindrical II. from Table IV-1 l structures use lift
, of NUREG-0487 Sup-t coefficient for plement 1 for SRV appropriate shape loads. (Amplitudes or Ct = 1.6 for SRV loads veri-fied by CA0RSO data
- 3) lhe standard drag on submerged coefficient for structures) pool swell and SRV
{ j oscillating bubbles should he based on 9
- data for structures with sharp edges.
Table 6.2 (Continued) Mark 11 Owners Group NRC WNP-2 Position on ' Load or Phenomenon Load Specification Evaluation Acceptance Criteria
- 2. Equivalent Uniform flow Structures are segmented into Acceptable Velocity and Acceleration small sections such that 1.0 <
L/D < 1.5. The loads are then applied to the geometrc tenter of each segment. N
- 3. Interference Effects A detailed methodology is Acceptable presented in Attachment 1.k of the Fimmer FSAR.
B. Jet Loads Calculated by the Ring Vortex 2.2.4.3 [3] Hodel.
?' C. Steam Condensation Drag Loads No generic load methodology WNP-2 load Described in 03 provided. specification and NRC the DAR.
review is addressed in this report. IV. Secondary Loads ' A. Sonic Wave load Negligible Load - none Acceptable. Acceptable. specified. B. Ccapressive Wave Load Negligible Load none Acceptable. Acceptabic. i specified. . C. Post Swell Wave Load No generic load provided. Plant unique load Addressed in specification the DAR. addressed in this report. D. Seismic Slosh Load No generic load provided. Plant unique load Addressed in specification the DAR. addressed in this report).
I i l Table 6.2 (Continued) l Mark 11 Dwners Group NRC t Load or Phenomenon Load Specification WNP-2 Position on Evaluation Acceptance Criteria E. Fallback load on Submerged
' Negligible Load none Acceptable.
Boundary specified. Acceptable. F. Thrust Loads Momentum balance. Acceptable. Acceptable. G. Friction Drag Lods on Vgpts Standard friction drag Acceptable.
! Acceptable.
~ calculations.
H. Vent Clearing Loads Negligible Load - none Acceptable. Acceptable. specified. NOTES TO TABLE
. o, [1] NRC Acceptance Criteria set forth in NUREG-0487.
$g
[2] NRC Acceptance Criteria set forth in Supplement 1 of NUREG-0487. [3] NRC Acceptance Criteria set forth in NUREG-0808. i l I i l i
The chug start times in each radial direction are assigned arbitrarily based on the smallest vcriance in one thousand Monte Carlo trials drawn from a uniform distribution of start times having a width of 50 milliseconds. The WPPSS Nuclear Project No. 2 pool pressures thus obtained are compared against JAERI data and found to bound the JAERI data. The staff and its consultant, Brookhven National Laboratory, has completed its review of the applicants' improved load methodology and found it to be acceptable. The staff will issue a NUREG report to discuss its fi . dings regarding all WPPSS Nuclear Project No. 2 plant-unique loads. 6.2.1.8.g Quencher Air Clearing Lead (Load II.B in Table 6-2) The applicant has committed to install a X quencher device designed by the General Electric Company. Subsequent to the issuance of NUREG-0487 and in view of the availability of in plant test data for the X quencher, the applicants have proposed an alternative to our acceptance criteria set forth in NUREG-0487. The alternative load specification was submitted to the staff in a report titled, "SRV Loads - Improved Definition and Application Methodology for Mark II Contain-ment." The improved load definition was derived from test results obtained from Caorso (Italy) in-plant SRV actuation experiments. Summary of results from Tokai-2 (Japan) in plant SRV tests were used by the applicants to confirm the adequacy of the improved load definition. Based on our review of the applicants' improved SRV load definition, we concluded that additional information is required to resolve our concerns. In a meeting that was held with the applicants on September 14-17, 1981, the applicants stated that they will respond to our concerns by December 15, 1981. We will report our findings regarding this item upon receipt of the applicants' response. In addition to our generic review of the Mark II pool cynamic loads, we have reviewed a limited number of pool dynamic loads on a plant unique basis. The basis of our review of these areas are discussed below. (1) Orywell Pressure History (Load I.8.1.f first column in Table 6.2) The crywell pressure nistory is utilized as part of the overall pool swell load methodology. The applicants have based its calculation of the drywell pressure history on the methods described in General Electric , Topical Containment Report NEDO-10320, Analytical "The General Model," and Appendix ElectricWe B of NEDO-20533. Pressure orev Suppressio i reviewed this methedology on a generic basis and concluded it was acceptable. l (2) Large Structure Imoact Loads (Load I.B.3.b in Table 6.2) The applicant nas stated that the WPPSS Nuclear Project No. 2 does not contain any large horizontal structures in the pool swell zone that would be subject to impact loads. Since the applicants have reviewed the as-built plant design and concluded that no large structure exists in the pool swell impact zone, we concur that no load specification is necessary for expansive structures. (3) Post-Swell Wave load and Seismic Slosh load (Load IV.C and D in Table 6.2) l ' These loads nave been determined to be secondary loads in that they are not design controlling. We have reviewed the applicants' evaluations of these loads and find them to be acceptable. 6-29
(4) Steam Condensation Submerged Drag Loads (Load III.C. in Table 6.2) Submerged structures in the WPPSS Nuclear Project No. 2 suppression pool were assessed by the applicant for loads due to main vent steam condensa-tion and due to SRV actuation. A procedure was developed to provide a conservative evaluation of these loads. The approach utilizes the same basic approach that was applied to the generic drag load methodology with several modifications. The source strength for these loads was derived from the 4TCO data. Plant unique flow fields will be defined consistently with the chugging and condensation oscillation boundary loads developed under items 6.2.1.8e and f above. For submerged s.tructure drag loads due to SRV actuation, the applicants indicated that data from Caorso SRV tests are examined to define the spatial distribution and to define time history of the dynamic pressure gradientplant. Caorso measured across column, vent pipes and SRV discharge line in the Based on our review of the applicants' preliminary submittal of these loads, we conclude that it is acceptable pending formal documentation of the information that was presented to us during the September 14-17, 1981 meetings. (5) Pool Temoerature Limit (Phenomenon I.A in Table 6.2) and Safety Relief Valve In-Plant Test We require in Criterion II. A. of NUREG-0487 that the suppression pool local temperature shall not exceed 210 degrees Fahrenheit for all plant tran-sients involving safety relieTvalve operations. The applicants.have not provided plant-unique analyses for pool temperature responses to trans-ients involving safety relief valve operation. These analyses are currently scheduled for submittal in late December 1981. We will report our findings in a supplement to this draft Safety Evaluation Report. We have requested the applicants to perform a comprehensive safety relief valve in plant test which is to be completed prior to commercial operation of the facility. These tests will irclude single and multiple valve tests to confirm the adequacy of the piping system design. In Oddition, we have requested the applicants to utilize information from these tests to establish the difference between local and bulk pool temperatures to demonstrate that a maximum local pool temperature specification of 210 1 degrees Fahrenheit will not be exceeded. Ouring the September 14-17 l would respond to our feques,t by December 1981 meetings the applicant indicated that th i regarding this item in the SER. 15, 1981. We will report our findings In conclusion, we conducted an assessment of the WPPSS Nuclear Project No. 2 against our generic acceptance criteria. where alternative criteria have been proposed. We also reviewed those few areas In addition, we completed our review of pool dynamic loads that were relegated to plant-unique reviews. In each of these areas, we concluded that the pool dynamics load utilized by the applicants are conservative and therefore acceptable, except for: 1 - , ,2 - Steam Condensation oscillation load specification; ! Poolintemperature and transients plant SRV test; and involving safety relief valve discharge 3 - Quencher Air Clearing Load. 6-30 l I l - . . . . . . . . . . . . . . . . .. -.
We will report our finding regarding these items in a suppl 3 ment to this Safety Evaluation Report. 6.2.2 Containment Heat Removal System i The containment residual heat removal system includes the piping, valves and mechanical components which will be used to remove energy from the containment to limit temperature and pressure in the drywell and suppression chamber following a postulated loss-of-coolant accident. The residual heat removal system consists of two complete loops including two heat exchangers and two main system pumps. Each loop is designed such that a failure in one loop cannot cause a failure of the other. In addition, each of the loops and associated equipment is located in a separate protected area of the reactor building to minimize the potential for single failures including loss of onsite or offsite power causing the loss of function of the entire system. The system equipment piping and support structures are designed to seismic Category I criteria. Provisions have been made in the residual heat removal system to permit inservice inspection of system components and functional testing of active components. Operating in the containment cooling mode, the residual heat removal pumps take suction from the suppression pool. Flow is then directed througn the residual heat removal heat exchangers to the suppression pool, the reactor vessel, or the containment spray headers. The location of system and return lines in the suppression pool facilitates. mixing of the return water with the total pool inventory before the return water becomes available to the suction lines. Strainers are provided on the suction line inlets. The applicants provided analyses of the long-term post-accident containment pressure and temperature response assuming various combinations of containment cooling availability. Our evaluation of this analysis is discussed in the subsection, "Long-Term Pressure Response," above. The applicants analyzed the net positive suction head that is available at the residual heat pump inlets assuming the containment will be at atmospheric pressure and the pool at saturation temperature. In addition, the applicants designed the suction piping from the suppression pool so that if any one suction strainer is 50 percent plugged, the maximum required net positive suction head to the residual heat removal system pumps during containment cooling will be provided. The above assumptions are in agreement with the provisions of Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Sumps," and, therefore, are acceptable. The pctential for debris to clog the residual heat removal system suction lines was evaluated. Each residual heat removal system pump is provided with its own suction line and strainer assembly. The pipe insulation used in the drywell, metal reflective insulation, is of a type to minimize the potential for its breaking away from piping and being carried through the vent system into the suppression pool. This design minimizes the potential of clogging the suction line. Therefore, we find the design of the pump suction strainers acceptable. 6-31
We conclude that the containment heat removal system can be operated in such a manner as to provide adequate cooling to the containment following a loss-of-coolant accident and will conform to Criteria 38, 39, and 40 of the General Design Criteria, and is acceptable. 6.2.3 Secondary Containment The secondary containment system includes the structures and systems to be used to control and treat radioactive leakage from the primary containment in the event of a loss of-coolant accident. The reactor building ventilation system controls the pressure in the secondary containment during normal operation to -0.25 inches of water gauge. Following a postulated loss-of-coolant accident, the standby gas treatment system will maintain the secondary containment pr. essure at -0.25 inches of water gauge. The standby gas treatment system is designed to seismic Category I criteria and is located within seismic Category I structures. The standby gas treatment system consists of redundant exhaust fans and filtration trains each consisting of a demister, heating coil, prefilter, high-efficiency particulate air filters and carbon acsorbers. Following a postulated loss-of-coolant accident, the pressure in the secondary containment could increase due to inleakage, air expansion due to equipment heat and the starting time required for the standby gas treatment system. The applicant performed an analysis of the secondary containment pressure transient which considers the above phenomenon. The results of these analyses shcw that approximately two minutes after the loss-of-coolant accident, the secondary containment pressure is reduced back to -0.25 inches of water gauge. Operation of only one of the two redundant standby gas treatment system triins was assumed for the analysis. In calculating the offsite radiological consequences of this pressure transient the applicant has assumed that bypass of the secondary containment Chapter 15 ofoccurs for two minutes after a loss of coolant accident (see this report). i The applicants have committed to perform a leak test of the secondary containment volume to verify the inleakage assumption and the 120-seconds drawdown time to reestablish a -0.25 inch of water gauge pressure. This commitment constitutes an acceptable fulfillment for meeting our requirement stated in Section 6.2.3 of the Standard Review Plan. Although the primary cohtainment is enclosed by the secondary containment, there are systems which penetrate both the primary and secondary containment boundaries creating potential paths through which radioactivity in the primary containment could bypass the leakage collection and filtration systems associ-ated with the secondary containment. A number of these lines contain physical barriers or design provisons which can effectively eliminate leakage such as water seals, containment isolation provision, or vent return lines to controlled regions. The criteria by which potential bypass leakage paths are determined l has been set forth in Bran .1 Technical Position, CSB 6-3, " Determination of Bypass Leakage Paths in Dual Containment Plants." Table 6.2-16 of the Final Safety Analysis Report provides a list of all potential bypass leakage paths
.and the corresponding leakage.
6-32
. . . . . . . . . . . . . . . . -- ~ -.-- - - - . .
We reviewed the design of the secondary containment systems for WPPSS Nuclear Project No. 2. Our review included the applicants' design bases, analysis of the functional capability of the secondary containment system and an evaluation of the systems against the criteria specified in the Branch Technical Position 6-3. Based on this review, we find the design of the secondary containment system acceptable. Summary and Conclusions of Containment Functional Design The applicants calculated the short- and long-term drywell and suppression chamber pressures and temperatures and the drywell floor differential pressure as described above. Based on our review of the applicants' analytical methods we conclude that the drywell and suppression chamber design pressures and temperatures and the drywell floor design pressure are acceptable. We also conclude that the drywell floor design reverse pressure is acceptable. We are currently performing our own confirmatory analysis to verify the acceptability of the design pressures. In addition, we completed our evaluation of the tests and analytical programs comprising the generic Mark II pool dynamic load program. This program forms the basis for the pool dynamic, loss of-coolant accident and safety relief valve loads utilized in the WPPSS Nuclear Project No. 2 Design Assessment Report. We conducted an assessment of the WPPSS Nuclear Project No. 2 load specifications against our generic acceptance criteria. We also reviewed those few areas where alternative criteria have been proposed. In addition, we completed our review of those pool dynamic loads that were relegated to plant unique rcviews. In each of these areas, we concluded that the pool dynamic loads utilized by the applicants are conservative and therefore acceptable except for the steam condensation oscillation load specification, quencher air clearing load and pool temperature transient. We will report our findings in a revision to this Safety Evaluation Report. 6.2.4 Containment Isolation System The containment isolation system includes the containment isolation valves and associated piping and penetrations necessary to isolate the primary containment in the event of a loss-of-coolant accident. Our review of this system included the number of isolation valve locations, the valve actuation signals and valve control features, the positions of the valves under various plant conditions, the protection afforded isolation valves from missiles and pipe whip, and the i environmental design conditions? specified in the design of components. l The design objective of the containment isolation system is to allow the normal t or emergency passage of fluids through the containment boundary while preserv-ing the integrity of the containment boundary to prevent or limit the escape of fission products from a postulated loss-of coolant accident. The applicant specified design bases and design criteria as well as the isolation valve arrangements to be used for isolation of primary containment penetrations. The containment isolation system is designed to automatically isolate the containment atmosphere from the outside environment under accident conditions. l Double barrier protection, in the form of two isolation valves in series or a closed system and isolation valve, are provided to assure that no single active l 6-33
failure will result in the loss of containment integrity. The containment isolation system components, including valves, controls, piping and penetra-tions, are protected from internally or externally generated missiles, water jets and pipe whip (see Section 3.6 of the Safety Evaluation Report). The basis for our acceptance has been the conformance of the containment isolation provisions to the Commission's regulations as set forth in the General Design Criteria, and to applicable regulatory guides, our technical positions, the Standard Review Plan, and industry codes and standards. The containment isolation systems are designed to the American Society of Mechanical Engineers Code, Section III, Class 1 or 2 and are classified as seismic Category I design systems. The containment isolation provisions for the lines penetrating containment conformas Criteria to appropriate. the requirements of Criteria 55, 56 or 57 of the General Oesign As provided by Criteria 55 and 56 of the General Design Criteria, there are containment penetrations whose isolation provisions do not have to satisfy the explicit requirements of the General Design Criteria but can be acceptable on some other defined basis. Most of those penetrations not satisfying the explicit requirements of the General Design Criteria were found acceptable based on their meeting alternative criteria specified in Section 6.2.4, item II of the Standard Review Plan. These alternative acceptance criteria are summarized below: (1) Lines that must remain in service following an accident and lines which should remain in service during normal operation for safety reasons are provided with at least one isolation valve. A second isolation boundary is formed by a closed system outside the containment. (2) Where a closed system outside the containment forms the second isolation boundary, each of the systems and all components which form its boundary are designed to Quality Group B and seismic Category I standards. Valves which isolate the branch lines of these closed systems outside containment are normally closed and under strict administrative control. (3) On some engineered safety features or related system, remote manual valves are used in lieu of automatic valves since these lines must remain in service following an accident. Where remote manual valve are used leakage detection capabilities are provided.
?
(4) On some penetrations the containment isolation provisions consist of two valves in series both of which are outside the containment. The location of a valve inside containment would subject it to more severe environmental conditions (including suppression pool dynamic loads), and it would not be easily accessible for inspection. Those lines which we found acceptable based on the criteria specified in the Standard Review Plan include: Drywell suppression chamber purge supply and return; sample analyzer; equipment, floor drains; suppression pool cleanup; containment spray systems, traversing incore probe system; residual heat 4 6-34
removal suction, pump test line, steam condensing recirculation, minimum recirculation; high pressure core injection pump suction, turbine exhaust, pump minimum recirculation, vacuum breaker; reactor core isolation cooling pump suction, turbine exhaust and turbine exhaust vacuum breaker, pump minimum flow vacuum pump discharge, the reactor building-to-wetwell vacuum breaker and the recirculation system hydraulic lines. Other lines penetrating the containment described below do not meet either the explicit requirements of the General Design Criteria or the alternative Standard Review Plan acceptance bases but meet acceptable isolation criteria on other defined acceptance bases. 6.2.4.1 Feedwater Lines The feedwater line penetrates the drywell to connect with the reactor pressure vessel. It has three isolation valves. The isolation valve inside th'e drywell is a check valve. Outside the primary containment is another check valve. Farther away from the primary containment ~ is a motor-operated check valve. Should a break occur in the feedwater line, the check valves prevent signifi-cant loss of reactor coolant inventory and offer prompt primary containment isolation. During the postulated loss-of-coolant accident, it is desirable to maintain reactor coolant makeup from all sources of supply. For this reason, the outermost valve does not automatically isolate upon a signal from the pro-tection system. Instead of using a leakage detection system in conjunction with use of a remote manual valve, this valve will be' procedurally controlled and remotely closed from the control room to provide long-term leakage protec-tion 20 minutes after a postulated loss of-coolant accident. If informatica indicates a degraded core condition prior to twenty minutes, the operator will take action to close the feedwater block valve at this time. We find this acceptable, since the time required to close this valve is approximately the same as if a leakage detection system were employed. 6.2.4.2 Emeroency Core Cooling System Heat Exchancer Relief Valves These system penetrations meet the two barrier criteria described in the Standard Review Plan with a closed system outside containment and a containment isolation valve. However, the isolation valve consists of a system relief valve, which discharges into the containment. We find use of this relief valve acceptable as an isolation valve since accident pressure seats rather than unseats the valve.
?
6.2.4.3 Containment Purce System The applicants proposed to use the drywell and suppression pool purge system during reactor power operation for the purpose of inerting and deinerting the containment. Nitrogen makeup during plant operation will utilize small feed-lines connected to each of the wetwell and drywell containment purge supply lines between the containment penetrations and the first containment isolation valve. A 2-inch vent bypass line exists in the purge system to bleed-off excess primary containment pressure during operations. We reviewed the design of the containment purge system based on the criteria specified in Branch Technical
. Position CSB 6-4, " Containment Purging During Normal Plant Operation," and find 6-35
that the use of the drywell and suppression pool purge system is acceptable provided that the applicants limit purging to control of containment pressure, inerting and deinerting of the containment and qualifies the valves to the requirements of Branch Technical Position CSB 6-4 as discussed further in Section 22 of the Safety Evaluation Report. It should be noted that the current design does not include a debris screen to protect the valves. The applicants have committed to install the debris screens prior to fuel load. Conclusion Based on the above evaluation, we conclude that the applicants' proposed design of the containment isolation system satisfies the requirements of Criteria 54, 55, 56 and 57 of the General Design Criteria and is acceptable except for the recombiner scrubber return line to the suppression pool where we will require the applicant to provide redundant isolation valve to meet GDC 56. 6.2.5 Combustible Gas Control The combustible gas control systems include piping, valves, and components and instrumentation necessary to detect the presence of combustible gases within the primary containment and to control the concentration of these gases. The scope of our review of the design and functional capability of the combustible gas control system for WPPSS Nuclear Project No. 2 included drawings and descriptive information of the equipment to mix the containment atmosphere, monitor combustible gas concentration, and reduce combustible gas concentrations within the containment following the design basis accident. Our review also included the applicants' proposed design bases for the combustible gas control t ' systems, and the analyses of the functional capability of the system provided to support the adequacy of the design bases. The bases for our acceptance are the conformance of system design and design i ' bases to the Commission's regulations as set forth in the General Design Criteria, and to applicable regulatory guides, branch technical positions, and industry codes and standards. Following a loss-of-coolant accident, hydrogen may accumulate within the containment as a result of corrosion of metals in the containment, metal-water l reaction between the fuel cladding, and as a result of radiolytic decomposition ) of the post-accident emergency cooling water. The applicants analyzed the production and accumulation of hydrogen from the above sources in accordance with the guidelines of 4 ranch Technical Position CSB 6-2, " Control of Combustible Gas Concentrations in Containment." The guideline regarding the metal-water reaction states that hydrogen production is five times the maximum calculated l reaction under 10 CFR 50.46, or that amount that would be evolved from a core-I wide average depth of reaction into the original cladding of 0.23 mils, l whichever is greater, in two minutes. The applicants have committed to inert the containment during operation to maintain a low level of oxygen. This item is discussed further in Section 22 of the Safety Evaluation Report where we address the rulemaking proceeding for consideration of degraded or melted cores. We conclude that the applicant calculated the hydrogen source in accordance with the provisions of Branch Technical Position CSB 6-2 and Regulatory
. Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a loss of-Coolant-Accident."
6-36
. . . . ~ . . - - . - - . . . . - . . - -_ --. ..
bsystem catalytic hydrogen The Chemical t topicalEngineering recombiner su report. and and Chemicals, Inc. has also been The applicants proposed re to reviewing use a the Air Produc manufactured by Air Products backup controlled purge systemd through a t Equipment addition to Qualification the recombiner, Branches a a In The containment atmosphere is purgeadded is instructed tobythe containm assumed proposed. bypass and nitrogen olant gas is accident after the the operatoriod the operator approximately 2.75 hours be Following a postulated ombiner After loss-of-co a half an hour warm-up perA hydrogen procedures to start the recflow. loss-of-coolant can initiate recombiner accident. processhydrogen concentration en concentration started if the containment l tions of the i containment ute initiated hydrog 5 hours maximum drywell te of 150 cubic feet per m n The applicants performed calcu a The applicants calculate a ident. based on a recombiner l flow ra e percent. and oxygen after a loss-of-coolant accconcentration pression chamcer. lessan than 4 Redundant systems are providewithin the drywell and sup inntratio concentration concentration i regarding the hydrogen ge of oxygen and its nitrogen arding function in an these itemsconcein the inerted atmosphere, We recomoiner system the percentaare to perform still await the qualification of theWe will report our findings reg inerted atmosphere. i ce SER. Containment Leakaoe Testingk testing program ndix for compl J to contain-an 6.2.6 applicants' containment quirement lea specified in Appe and We reviewedcontainment the leakage testing re t assurance that theserv l with the Such compliance provides ked during adequa ebeonverified service Maintaining con- a time its. 10 CFR Part rates will beleak-tight
- 50. ment periodically checwithin integrity ancecan that,the specified in the even lim that the leakagebasis limits provides to the reasonable maintain within such leakage containment, assur ss of the limits the such tainment within l leak paths will not be in exce of any radioactfDity release m to assure that atmosphere through potentia specified for the site. containment leak valvetesting progra arrangements aret and system isolation The proposed leak Specifically, we reviewed containment the integrated leakvalves, ra emergency the containment penetrations quirements of Appendix J.system, and designed to satisfy the ntainment main and the local leak testing re incoresteam isolation probeexplicit requirem testing practices suction for the cosystems, traversinger, differ from the core cooling The systemcontrol rod drive system, howevaccep Appendix J. water Main Steam Isolationvalves ires local leak testing of boiling 4),
Valves (see Paragraph ident II.H.(see Appendix J to 10 CFR Part 50 requ l ted to the design basis acc reactors main steam isolationlated containm 6-37
Paragraph III.C.2). Furthermore, Appendix J requires that the measured leak rates be in the summation for the local leak rate tests (see Paragraph III.C.3). The applicants have provided information describing its methods of compliance
~w ith Appendix J to 10 CFR Part 50 relating to leak testing of the main steam isolation valves. The applicants propose to leak test the main steam isolation valves at a reduced pressure and exclude the measured leakage from the combined leak rate for the local leak rate tests. As a result of our review of this information, we have determined that exemption to Appendix J is required and justified.
Our bases for this conclusion are discussed below. Each main steam line is provided with two main steam isolation valves which are position 2d to provide better sealing in the direction of post-accident containment atmosphere leakage. In the event of a loss-of-coolant accident, the main steam leakage control system will maintain a negative pressure between the main steam isolation valves. The effluent will be discharged into a volume where it will be processed by the standby gas treatment system prior to being released to the environs. A radiological' analysis for this potential source of containment atmosphere leakage was performed, based on an assumed leak rate past the inboard main steam isolation valve of 11.5 standard cubic feet per hour, see Section 15.3.2 of the this report. The applicants plan to periodically leak test the main steam isolation valves to assure the validity of the radiological analysis. The design of the main steam isolation valves is such that testing in the reverse direction tends to unseat the valve. Testing of the two valves simul-taneously, between the valves, at design pressure would lift the disc at the inboard valve. This would result in a meaningless test. The proposed test calls for a test pressure of 25.0 pounds per square inch gauge to avoid lifting the disc at the inboard valve. The total observed leakage through both valves (inboard and outboard) is then conservatively assigned to the penetration. We conclude that this procedure is acceptable. Furthermore, excluding the leakage frem the summation for the local leak rate tests is acceptable since the leakage has been accounted for separately in the radiological analysis of the site. We conclude that leak testing the main steam isolation valves in the manner described above is an acceptaole alternative to the requirements of Appendix J. Hydrostatic Testing of Containment Isolation Valves Appendix J to 10 CFR Part 50 requires that valves, unless pressurized with fluid from a seal systedi, shall be pressurized with air or nitrogen for leak testing purposes (see Paragraph III.C.2). There are a number of liquid filled systems, however, that are specifically designed to remain intact following a loss-of-coolant accident and thus provide a water seal for the system isolation valves or assure that only liquid leakage from the containment will occur. The applicants propose to perform hydrostatic testing to determine the leak tightness of the emergency core cooling suction systems isolation valves. For the above systems, the applicant indicated that a liquid inventory will produce a water seal during the post-accident period and only liquid leakage l from the containment will occur. The applicants propose to pressurize valves identified above to determine their leak tightness. 6-38 i l . . - - . . . . - . .. .. l
The combined leakage from all these valves will satisfy the acceptance criteria of 10 CFR Part 100 regarding the site radiological safety analysis and will be included in the plant technical specifications. This leakage will be excluded when determining the combined leakage rate for all penetrations and valves as specified in Paragraph III.C.3 of Appendix J. We reviewed the applicants' proposed hydrostatic testing and concluded that such testing is permissible for the lines identified above since the applicants have shown the: (1) Existence of a water seal, (2) System boundaries are designed to engineered safety feature criteria, and (3) Acceptance criteria of 10 CFR Part 100 are satisfied. Traversing Incore Probe System The traversing incore probe system is equipped with ball valves that provide the guide tubes with shutoff capability. following the cable withdrawal. A shear valve is proviced to cut the cable in the event that the drive cable cannot be withdrawn. The applicants have committed to perform a Type C test on the ball valve. Since the shear valve requires testing to destruction, the applicants are not going to perform periodic Type C tests on these valves. However, statistically chosen samples of the shear valves is tested by the manufacturer. To assure that the shear valve will perform its intendeo function, we have requested, and the acplicants have agreed, to: (1) Verify the continuity of the explosive charge at least once per 31 days. (2) Initiate one of the explosive squibs charge at least once per 18 months. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch wnich has been certified by having one of that batch successfully fired. (3) All charges should be replaced according to the manufacturer recommended l life time. l Based on the above discussion, we conclude that the leak testing of the traversing incore probe jystem is acceptable. Control Rod Drive Appendix J to 10 CFR Part 50 (see Paragraph III.C.3.(a)) allows exclusion (from of leakage from valves that are sealed with fluid from a seal i combined system if the 0.6i.3)luid f leakage rates do not exceed those specified in the t specifications. i
' To assure that system leakage will not exceed 3 gallons per minute (maximum leakage), a piping integrity test is accomplished for leaks of the hydraulic control units (operating pressure 1000 pounds per square inch) during daily 6-39 l
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inspection. In addition, several indicators in the control room will indicate if leakage is excessive. Furthermore, since the reactor pressure vessel and nonseismic portion of the control rod drive system are vented during Type A tests, leakage monitoring of the control rod drive insert and withdraw lines is provided by Type A leakage rate tests. We conclude that leak testing the control rod drive in the manner described above is an acceptable alternative to the requirement of Appendix J. 6.2.6.1 Conclusions We reviewed the applicants' proposed leak testing program and conclude that it meets the requirements of Appendix J to 10 CFR Part 50 with the exemptions noted and is acceptable. r 6-40 _ _ . . . . . _ . . .}}