ML20198G309
| ML20198G309 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 02/14/1975 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1042 NUDOCS 8605290461 | |
| Download: ML20198G309 (33) | |
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s L: Reading File RSB Reading File FEB 14 W5 Docket Nos.: 50-460/513 i
4 Voss A. Moore, Jr., Assistant Director for LWR's, Group 2, TR REACTOR SAFETY INPUT TO WPPSS SER fi Plant Name: WPPSS Nuclear Plants, Units 1 and 4 Licensing Stage: CP Docket Nos.: 50-460/513 Responsible Branch and Project Leader: LWR 2-3, T. Cox Requested Completion Date: January 23, 1975 Technical Review Branch Involved: Reactor Systems Description of Review: SER Input Review Status: CP Application Incomplete The attached report contains the evaluation performed by the Reactor Systems Branch on WPPSS, Units 1 and 4.
Reactor Systems review included Section 1.5, 4.1, 4.4, 5.1, 5.2.2, 5.3, 5.5, 6.3.1, 6.3.2, 6.3.3, 6.3.4, and Chapter 15.
There remsnins two areas of the WPPSS Units 1 and 4 design which require comitments by the applicant. The first concerns the GDC 34 requirement to achieve nomal shutdown cooling from the control room. The second concerns preoperational testing of the LPSI System from the sump and confiming the flow split of the cavitating venturies in the LPSI System. The staff positions for the first and second areas are delineated in Sections 5.5.7 ar.d 6.3.4, respectively.
Orfsinal Signed bg Victor SteUn Victor Stello, Jr., Assistant Director for Reactor Safety Office of Nuclear Reactor Regulation
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UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
~ FES 14 1915 Docket Nos.: 50-460/513 Voss A. Moore, Jr., Assistant Director for LWR's, Group 2, TR REACTOR SAFETY INPUT TO WPPSS SER Plant Name: WPPSS Nuclear Plants, Units 1 and 4 Licensing Stage: CP Docket Nos.: 50-460/513 Responsible Branch and Project Leader: LWR 2-3, T. Cox Requested Completion Date: January 23, 1975 Technical Review Branch Involved: Recctor Systems Description of Review: SER Input Review Status: CP Application Incomplete The attached report contains the evaluation performed by the Reactor Systems Branch on WPPSS, Units 1 and 4.
Reactor Systems review included Sections l.5, 4.1, 4.4, 5.1, 5.2.2, 5.3, 5.5, 6.3.1, 6.3.2, 6.3.3, 6.3.4, and Chapter 15.
There remains two areas of the WPPSS Units 1 and 4 design which require commitments by the applicant. The first concerns the GDC 34 requirement to achieve normal shutdown cooling from the control room. The second concerns preoperational testing of the LPSI System from the sump and confirming the flow split of the cavitating venturies in the LPSI System. The staff positions for the first and second areas are delineated in Sections 5.5.7 and 6.3.4, respectively.
ff Victor Stello, Jrv,1 Assisfan Dira tor for Reactor Safety Office of Nuclear Reactor Regulation
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1.5 Research and Development The applicant has identified in Section 1.5 of the WPPSS PSAR the research and development (R&D) programs applicable to Units 1 and 4.
These programs, which are to be conducted by B&W, are aimed at verifying the new 17 x 17 (Mark C) fuel assembly design and confirming the design margins of the NSSS. A discussion of the various R&D programs and their objectives are presented in Section 1.5 of the PSAR and is summarized in Table 1.5-1.
The results of the R&D program will be reviewed generically by the staff as progress in these experimental programs is reported. All tests directed toward the veri-fication of the 17 x 17 design are scheduled for completion during 1975, well in advance of the proposed WPPSS fuel loading dates.
l The applicant has referenced Topical Report BAW-10097,
" Mark C (17 x 17) Fuel Assembly - Research and Development" in support of Section 1.5 in the SAR. The scheduled critical l
heat flux tests and inlet flow mixing tests will use bundles which are much shorter in length than the WPPSS fuel assemblies.
These tests are intended to verify the applicability of the B&W-2 CHF correlation to the Mark C fuel assembly design by l
demonstrating that the correlation conservatively predicts the test data for the Mark C geometry and grid design. This l-l
correlation also contains an axial flux shape factor multiplier based on tests on short length rods. The applicability of the B&W-2 correlation with the flux shape factor to actual reactor conditions can be veri-fled by non-uniform axial heat flux CHF tests with full length assemblies. We will require the applicant to demonstrate the applicability of the CHF correlation during the operating license review.
We have concluded that:
(1) the test program as it is outlined in the PSAR will provide the information necessary for the design and safe operation of the WPPSS plants; and (2) in the event any of this research and development work provides unexpected results, appropriate restrictions on operation can be imposed or proven alternate designs such as Mark B fuel assembly design can be utilized to protect the health and safety of the public, and (3) the applicant has met the requirements of 10 CFR Part 50.35(a) 7 l
in regard to needed research and development programs. Our conclusions are based on our review of the safety related mechanical and thermal hydraulic differences between the Mark C and Mark B fuel designs. The Mark B fuel design has been reviewed and approved for previous plants using B&W NSSS such as Oconee Units 1, 2, and 3 and North Anna Units 3 and 4.
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TABLE 1.5-1 s-Babcock & Wilcox'Research and Development Program TESTS PURPOSE
- Assembly Flow Tests Assembly Pressure Drop Hydraulic Loads Dynamics of holddown springs Fuel Rod Vibrations Verify scram times Control Rod, Guide tube, Orifice Rod Fretting and Wear Reactor Vessel Flow Tests Inlet flow mixing / Distribution Vessel pressure drop Assembly Mechanical Tests Vibration and Damping characteristics Load Response Component Mechanical Tests Spacer Grid Spring Characteristics Seismic capability of spacer grids End-fitting characteristics Critical Heat Flux Tests Verify applicability of B&W-2 correlation Rod Burst Tests Rod Swelling and burst characteristics All tests provide input data for Seismic and LOCA analyses.
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4.0 REACTOR 4.1 Sumary Description The fuel design and the WPPSS NSSS will be the same as that reviewed and approved for the Bellefonte Nuclear Plant. Bellefonte was the lead plant for the 205 fuel assembly plants which include Greenwood 2 and 3 and WPPSS 1 and 4.
For comparison, WPPSS is geometrically similar but larger, i.e., 205 fuel assemblies, to North Anna 3 and 4, except WPPSS will use fuel assemblies with a 17 x 17 fuel rod array, while the North Anna core was made up from 15 x 15 fuel rod assemblies. The proposed initial power for the WPPSS core is 3600 megawatts thermal, which is 27% higher than North Anna.
4.4 Thermal and Hydraulic Desian The proposed WPPSS reactors are each designed to operate at core power levels of up to 3600 MWt, which corresponds to l
a gross electrical output of about 1329 MWe. We have reviewed l
l the thermal-hydraulics on the basis of 3600 MWt. A comparison of the thermal and hydraulic design parameters for the WPPSS l
and North Anna 3 and 4 plants is shown in Table 4.4-1.
The principal criterion for the thermal-hydraulic design of a reactor is to prevent fuel rod damage by providing adequate heat transfer for the various core heat generation patterns occurring during normal operation, operational transients, 1
l and accidents.
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Maintenance of nucleate boiling is a basic objective of a themal-hydraulic design. The applicant has demonstrated, through the use of the Babcock & Wilcox BAW-2 correlation, that a departure from nucleate boiling (DNB) heat flux can be avoided if the required departure from nucleate boiling heat flux ratio (DNBR) greater than 1.32, is main-tained for steady state and anticipated transient conditions. The values of minimum DNBR design power and design overpower conditions, shown in table 4.4-1, are greater than the miniiwm DNBR design limit of 1.32; however, the hydraulic analysis was based on vessel model flow tests which are not completely applicable to the WPPSS 205 fuel assembly design. Since the core-inlat flow distribution is dependent on flow conditions in the inlet annulus and themal shield regions, the 205 assembly vessel model flow tests, applicable to the design of WPPSS, will be reviewed at the operating license stage to confirm the acceptability of the thermal-hydraulic cal-culations. The applicant has indicated that the core flow distribution tests for the 205 fuel assembly plants are scheduled for completion in 1975
~ The core power level and the peak linear power density of a PWR are controlling factors in the evaluation of various transients and accidents. For the WPPSS station, the core power level used for the safety evaluation of the plant was 3600 MWt and the linear power density used was 14.74 kw/ft (3760 MWt and 15.36 kw/ft for LOCA analyses). With this assumed core power and linear power density, this facility was shown to comply with existing criteria. The maximum linear power density pemitted during steady state operation and the
maximum linear power density permitted to occur for certain plant operating nianeuvers will be determined during the operating license review and will be consistent with the criteria in effect at that time.
Preservation of nucleate boiling as the mode of heat transfer between the fuel cladding hot spot and coolant not only assures that the cladding temperature is only slightly greater than that of the coolant saturation temperature, but that the fuel centerline temperature will not. reach the melting temperature. The applicant's criteria for overpower protection requires that tne maximum fuel center-line tenperature be less than that of the fuel melting temper-ature,at a peak core power generation rate of 16.51 kw/ft during all modes of operation.
In fulfillment of this objective, the applicant has calculated a fuel centerline temperature of 4470 F at 16.51 kw/ft compared to a fuel melt temperature of 5080 F at beginning of life which reduces linearly with burnup to 4800 F after 43,000 mwd /mtU.
On the basis of our review of the analytical techniques applied to the previously reviewed and approved 15 x 15 core designs, we have concluded that for the 17 x 17 core design, there is reasonable assurance (1) that the proposed thermal-hydraulic design accounts for DNB ansi fuel centerline temperature limitation in a satisfactory manner, and (2) that the conservatism in the thermal-hydraulic desiga system can be verified.
In the
event that sufficient verification cannot be obtained from the test programs or that the analytical methods are not conservative, appropriate restrictions on operations will be established at the operating license stage.
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s TABLE 4.4-1 Comparison of Thermal Hydraulic Desicn Parameters For the WPPSS 1/4 and North Anna 3/4 Plants
.NPPSS l&4 North Anna 3&4 Reactor Core Heat Output (MWt) 3600 2631 System Pressure, Nominal (psia) 2250 2235 Minimum DNBR at Design Power 1.82 1.72 Minimum DNBR at Design Overpower 1.4 1.39 Minimum DNBR for Design Transients
>1.32
>l.32 6
Total Reactor Coolant Flow (10 lb/hr) 150.5 106.86 Effective Flow Rate for Heat Transfer 6
(10 lb/hr) 142.4 103.0 Core Coolan.t Average Velocity 16.2 16.3 6
2 Average Mass Velocity (10 lb/hr-ft )
2.66 2.67 Coolant Temperature ('F)
Design flominal Inlet 572.3 566.3 Average Rise In Core 59.3 61.7 2
Total Heat Transfer Surface in Core (ft )
63,991 40,743 2
Average Heat Flux (BTU /hr-ft )
186,822 214,000 2
Maximum Heat Flux (BTU /hr-ft )
507,044 582,000 Maximum Thermal Output (Kw/ft) 14.74 19.2 Maximum Thermal Output at overpower (Kw/ft)'
16.51 21.5 Maximum Fuel Central Temperature (*F) 100% power 3760 4410 At 112% overpower 4470 4720
- Same as Bellefonte 1/2 and Greenwood 2/3
5.0 REACTOR COOLANT SYSTEM 5.1 Sumary Descr'iption The reactor coolant system for WPPSS consists of two similar heat transfer loops connected in parallel to the reactor pressure vessel.
Each loop contains two reactor coolant pumps and a once-through steam generator.
In addition, the system includes a pressurizer, a reactor coolant drain tank (pressurizer relieftank),andinterconnectingpipingandinstrumentation necessary for operational control. All the above components are located in the containment building.
During operation, the reactor coolant system transfers the heat generated in the core to the steam generators where steam is produced to drive the turbine genenator.
Borated demineralized water is circulated in the reactor coolant system at a flow rate and temperature consistent with achieving the reactor core thermal-hydraulic performance. The water also acts as a neutron moderator and reflector and as a solvent for the neutron absorber used in chemical shim control.
l The reactor coolant system pressure boundary provides a second barrier.against the release of radioactivity generated within the l
reactor (the fuel rod cladding is the primary barrier), and is designed to ensure a high degree of integrity throughout the life j
of the plant.
Reactor coolant system pressure changes caused by load transients are controlled by the use of the pressurizer where water and steam are maintained in equilibrium by electrical heaters or water sprays.
Steam l
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s can be formed (by the heaters) or condensed (by the pressurizer spray) to minimize pressure variations due to contraction and expansion of the reactor coolant. Spring-loaded safety valves and power operated relief valves are mounted on the pressurizer and discharge to the reactor coolant drain tank, where,the steam is condensed and cooled by mixing with water.
The system concept is the same as reviewed and approved for Bellefonte 1/2, Greenwood 2/3, Oconee 1, and North Anna 3/4.
5.2.2 Overpressurization Protection Overpressurization protection in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Article NB-7000
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(1971) is provided by pressure relief of the RCS from two pressurizer safety valves and one electrically actuated relief valve mounted on nozzles on the pressurizer. The pressurizer safety valves discharge through a common header to the reactor coolantdraintank(RCDT). The pressurizer safety valves are sized on the basis of the maximum pressure transient imposed on the RCS resulting from complete loss of main feedwater flow.
The applicant's description of RCS overpressure protection references Topical Report BAW-10043, Supplement 1.
At present, this report is under review by the staff and determination as to the adequacy of the overpressure protection for WPPSS will be made as a part of the overall Anticipated Transients Without Scram (ATHS) review. ATWS is addressed in Section 15.2.1.
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5.3 Thermal Hydraulic System Design The thermal and hydraulic design bases of the RCS are discussed in Sections 4.3 and 4.4.
5.5 Component and Subsystem Design
'5.5.1 Reactor Coolant Pumps and Motors The reactor coolant pump is designed to provide adequate core cooling flow and hence sufficient heat transfer to main-tain a DNBR 1.32, within the parameters of operation.
Sufficient pump rotational inertia is provided by the flywheel to promote continued flow following a loss of forced flow resulting from mechanical or power failures to the pumps such that the reactor neutron power can be reduced before DNB limits are exceeded.
A pump overspeed evaluation has not been submitted to the staff and the applicant has indicated that these studies, as well as the results of an investigation of overspeed protection devices, will be reported for our evaluation for the operating license application.
5.5.2 Steam Generator The steam generator is a vertical straight-tube-and-shell heat exchanger and produces superheated steam at constant turbine throttle pressure over the operating power range. The
primary reactor coolant enters the steam generator upper hemispherical head, flows downward inside the tubes giving up heat to generate steam on the shell side secondary loop.
The tube and tubesheet boundary are designed for reactor coolant side design pressure and temperature. The steam generators must provide a heat sink for the primary reactor coolant system and they are at a higher elevation than the core to improve natural circulation for decay heat removal.
5.5.3 Reactor Coolant Pioing The reactor coolant piping is designed and fabricated to accommodate the system pressures and temperatures attained under all expected modes of plant operation or anticipated system interactions.
i 5.5.4 Main Steamline Flow Restrictors The applicant stated that because of the small inventory of water in the S&W OTSG design, no flow restrictors are re-quired in the main steamline.
i 5.5.7 Decay Heat Removal l
The Decay Heat Removal System (DHRS), is designed to remove decay heat and sensible heat from the RCS and core during the latter stages of cooldown. The system also provides j
auxiliary spray to the pressurizer for complete depressurization, reduces reactor coolant temperature during refueling, and provides the means for filling and draining the refueling i
cavi ty.
In the event of a LOCA, the DHRS serves as part of theECCS(EmergencyCoreCoolingSystem)byprovidinglow
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pressure injection of borated water into the reactor vessel for emergency core cooling.
The decay heat removal system is placed into operation approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after initiation of plant shutdown when the temperature and pressure of the RCS are below 305 F and 600 psig., respectively.
Assuming that two pumps and coolers are in service, and that each cooler is supplied with component cooling water at design flow and temperature, the DHRS is designed to reduce the RCS temperature from 305 F to 140'F within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
The DHRS is provided with two DHR pumps and two DHR coolers arranged in redundant and independent flow paths.
If one of the two pumps or one of the two coolers is not operable, safe cooldown of the plant is not compromised; however, the time required for cooldown is extended. The use of the DHRS as part of the'ECCS is described in Section 6.3. ~
Overpressurization protection of the DHRS is provided in the discharge line by two check valves in series in each of the two DHR discharge trains. A pressure switch with a high pressure alarm is provided between these check valves to alert the operator to a faulty check valve.
Each line has a motor operated valve upstream of the two check valves to provide containment isolatior.
In each of the suction lines from the reactor coolant system, two motor-operated
valves located inside containment and a manual valve positioned outside containment prevent overpressurization of the DHRS. Two relief valves in each DHR train provide pressure relief in the event of DHRS overpressurization during plant cooldown when the DHRS is in operation. The relief valve in the injection line is sized to protect against RC leakage through the check valves. The relief valve in the suction line will prevent the DHRS design pressure (675 psig) from being exceeded by more than 10%
during the wcrst transient. This transient was the in-advertent operation of all three HPI pumps.
The staff requested the applicant to change the manual valves in the suction line of the DHRS to motor operated valves with control and indication in the control room.
Pursuant to the staff's interpretation and implementation of General Design Criterion 34, normal shutdown cooling shall be accomplished from the control room even with the system experiencing a single active failure.
Prior to issuance of construction permits for this facility, we will require that the applicant document its commitment to this requirement.
5.5.10 Pressurizer The pressurizer maintains the RCS pressure during steady state operation and limits pressure changes during transients.
It contains a water volume, sized to provide the ability of the system to experience a reactor trip and not uncover the low
level sensors in the bottom head.
It maintains the RCS pressure high enough so as not to activate the high pressure injection system; and a volume of steam, sized to provide the ability of the system to experience a turbine trip and not cover the level sensor in the upper head.
By means of its connection to the gaseous waste systems, the coolant drain tank provides a means for removing noncondensible gases from the reactor coolant system which could collect in the pressurizer vessel.
The tank design is based on the requirement to absorb the pressurizer effluent should the safety valves lift.
By means of its connection to an external heat exchanger in the liquid waste disposal system, the RCDT will be restored to normal temperature (120*F) after the safety valves relieve by the pump cooler train. The tank is designed for 100 psig pressure and 340*F, and equipped with a rupture disc which is designed to maintain the design pressure by releasing the steam to containment.
5.5.13 Safety and Relief Valves The pressurizer safety valves are bellows sealed, balanced, spring-loaded safety valves which are provided with a supplemental back pressure balancing piston for handling a bellows failure. The pressurizer relief valve is an electrically actuated, electrically controlled, pilot operated, pressure loaded, relief valve.
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o The combined capacity of the pressurizer safety valves 6
is 10 lb/hr., which was based on accepting the maximum insurge resulting from a complete loss of main feedwater flow. This objective is met without reactor trip or any operator action provided that the spray nozzle in the upper section of the pressurizer maintain the steam and water at the saturation temperature which corresponds to the desired reactor coolant system pressure.
During out-surges, as the RCS pressure decreases, some of the water flashes to steam and the electric heaters restore the normal operating pressure. Two ASME code safety valves are connected to the upper pressurized head to prevent system overpressure. A pilot-operated relief valye is also provided to lim'it the lifting frequency of the safety valve discharge to the reactor coolant drain tank, located within containment.
5.5.14 Internals Vent Valves The WPPSS design, like the previous Babcock and Wilcox designs, include vent valves which provide a direct flow path between the upper core region and inlet annulus in the event of a loss-of-coolant accident from an inlet line break. This flow path provides for pressure equalization by the venting of steam to the break and permits the emergency coolant water to reflood the core rapidly. There are 8 vent valves, each with an effective flow diameter of 14 inches. They are located on a common plane in the upper core support weldment above the
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outlet nozzles. The face of the disc is inclined 5 degrees to the vertical to insure a positive seal of the disc due to the differential pressure across the valve. The vent valve design is essentially the same on the North Anna design except the exercising hook is a different shape and the valve is mounted by capture bolts through a flange.
In the thermal-hydraulic analysis of the WPPSS plant for normal operation, the 6pplicant assumed that there was no bypass flow resulting from an open vent valve. At ' present,
there is not adequate instrumentation to detect the system flow change (approximately 5% reduction in core flow), which would result from an open valve. The staff position has not changed from that taken on the Oconee plant. Therefore, the staff requires that one valve less than the minimum detectable number of stuck open vent valves be assumed open and the corresponding core flow penalty be imposed for the thermal-hydraulic redesign of the RCS and core. This analysis should be incorporated into the FSAR and will be evaluated to determine the maximum power level of the system.
Changes to this staff positions will be considered as operating experience from operating plants for which vent valves have been installed c
becomes available.
We conclude, with the conditions as noted above, that the proposed reactor coolant system, subsystems, and component designs are acceptable.
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O-6.3 Emergency Core Cooling System (ECCS) 6.3.1 Design Bases
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The applicant has stated that the WpPSS ECCS will be designed to provide emergency core cooling during those postulated accident conditions where it is assumed that mechanical failures occur in the reactor coolant system piping resulting in, loss of coolant from the reactor vessel greater than the available coolant makeup capacity using normal operating equipment.
The ECCS is also designed to protect against steam line break consequences.
The applicant's design basis is to prevent fuel and cladding damage that would interfere with adequate emergency core cooling and to mitigate the amount of clad-water reaction for any size break up to and including a double-ended rupture of the largest primary coolant line.-
The applicant stated that these requirements will be met even with minimum engineered safeguards available, such as the loss of one emergency power bus together with the unavail-ability of offsite power.
The ECCS to be provided is stated to be of such number, diversity, reliability and redundance that no single failure of ECCS equipment, occurring during a LOCA, will result in inadequate cooling of the reactor core.
Each of the submitted ECCS subsystems is - to be designed to function over a specific range of reactor coolant piping system break sizes, up to and including the
o flow area associated with a postulated double-ended break in the 2
largest reactor coolant pipe (15.75 ft is the double-ended area).
6.3.2
System Design
The ECCS proposed for the WPPSS plants will consist of two core flooding tanks, high pressure injection (HPIS) and.ow pressure injection (LPIS) systems, with provisions for recir-culation of the borated coolant af ter the end of the injection phase. Various combinations of these systems will assure core cooling for the complete range of postulated break sizes.
Following a postulated LOCA, the ECCS will operate initially in the passive accumulator injection mode and the active high pressure injection mode, then in the active low pressure injection mode, and subsequently in the recirculation mode.
The high pressure injection mode of operation, upon actuation of a safety injection signal, will consist of the operation of two y
of three centrifugal charging pumps (rated at 700 gpm each at a design head of 2600 ft) which provide high pressure injection of emergency core coolant into the reactor coolant system cold legs.
For very small breaks, the HPIS will suffice initially to replenish the lost coolant. During normal operation, one of the HPI pumps is operating as part of the Make-Up and Purification System. Low pressure injection provided by the DHRS will consist of two decay heat removal pumps (rated at 5000 gpm each at a design head of 385 ft) which will take their suction through two independent lines from the borated water storage tank for short term cooling. The DHRS dis-charges directly into the reactor vessel through the core flooding
O nozzles located in the vessel wall. Crossover lines containing cavitating venturies between the redundant low pressure lines are provided to ensure that sufficient flow will be available for core cooling if a rupture occurs in the core flooding piping.
All Emergency Safeguard Feature Systems take suction from the borated water storage tank (BWST). The tank has a useable inventory of 680,000 gallons of borated water with a minimum boron concentration of 1800 ppm. This tank is located inside the General Services Building which precludes the necessity of
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tank heaters'. The staff have required the applicant to include a technical specification which requires the temperature of the borated water in the tank to be well above the freezing point before and after the reactor is maue critical.
t When the level in the BUST decreases below a preset level, automatic switchover from the injection to the recirculation mode of emergency core cooling is accomplished by opening the containment sump suction valves. The applicant has indicated that this system is the same as that for Bellefonte. The system will ensure that when the BWST empties, the ECCS pumps will have sufficient suction to preclude damage to the pomps and emergency core cooling is not interrupted. Long term core requirements are provided by the LPIS by recirculating the spilled reactor coolant collected in the containment sump back to the reactor vessel through the core flooding lines. However, should the reactor coolant system pressure be higher than the LPI pump head, the required flow is delivered by the HPIS after the operator has aligned the flow from the discharge of the LP pumps to the suction of the HP pumps.
The passive injection node of operation is provided by the core flooding (CF) system, which protects the core in the event of intermediate and large-sized pipe breaks.
The' coolant is automatically inject.ed when the RCS pressure drops below the core flooding tank prcssure (600 psig).
Each of the two core flooding tanks has a normal water volume of 3
3 1350 ft with 450 ft of nitrogen gas at a normal operating pressure of 600 psig. Each tank is connected by a core flooding line directly to a 9-inch reactor vessel core flooding nozzle.
The driving force for injection of the 1800 ppm borated water is supolied by pressurized nitrogen.
Each core flooding line will contain an electric-motor-operated stop valve for isolation of the CFT during reduced pressure operation and two inline check valves in series to provide isolation during normal operation.
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6.3.3 Performance Evaluation l
The applicant has stated that the emergency core cooling systems have been designed to deliver fluid to the reactor coolant system in order to control the predicted cladding temperature transient following a postulated pipe break and for removing decay heat in the long-term, recirculation mode.
On June 29, 1971, the AEC issued an Interim Policy Statement containing Interim Acceptance Criteria for the performance of the ECCS for light-water cooled nuclear power reactors. The Interim Policy Statement includes a set of conservative assumptions and procedures to be used in conjunction with computer codes to analyze and evaluate the ECCS function for a pressurized water reactor incorporating a dry containment. A public rule making hearing on the Interim Acceptance Criteria for ECCS for light-water cooled nucl, ear power reactors has been conducted.
In accordance with the Interim Policy Statement (IPS), the performance of the ECCS is judged to be acceptable if the course of the LOCA is limited as follows:
1.
The calculated maximum fuel element cladding temperature does not exceed 2,300*F.
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed one percent of the total amount of cladding in the reactor.
3.
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling, and before the cladding is so embrittled as to fail durino o after quenching.
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The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.
The applicant presented a Preliminary Safety Evaluation of the LOCA in accordance with the requirements of the IPS in BAW-10065, Supplement 1 and BAW-10074. This evaluation resulted in a peak clad temperature of 1929 F and showed compliance with the Interim Acceptance Criteria.
As part of the PSAR for the WPPSS 1 and 4 plants, the applicar.t.is required to submit a LOCA analysis performed by an acceptable evaluation model under the ECCS criteria published in the Federal Register on January 4,1974, and show that the plant is in compliance with the same criteria. The applicant has stated that this ECCS analysis will be provided to the staff by April 4,1975. This will be reviewed by the staff and reported in the Supplement to this Safety Evaluation Report.
6.3.4 Tests and Inspections The applicant has proposed preoperational tests that will be performed. He believes that these are in compilance with proposed revision 1 to Regulatory Guide 1.79.
However, the staff does not believe that the tests proposed fully meet the intent of the guide. The staff position is that the
recirculation test of the LPSI system shall demonstrate the capability to realign the valves and injection pumps to recirculate coolant from the sump.
In addition, pre-operational flow tests shall be conducted to verify the sizing of the cavitating venturies to confirm the as-built flow split pe.rformance of the LPSI system. Prior to issuance of construction permits for this facility, we will require that the applicant document its commitment to these require-ments.
6.3.5 Conclusions On the basis of our evaluation, we have concluded that the predicted functional performance of the ECCS for the WPPSS 1 and 4 plants is in accordance with the Commission's Interim Acceptance Criteria. The functional performance of the ECCS pursuant to the ECCS Final Acceptance Criteria will be reviewed by the staff and reported by July 7,1975 in the Supplement to this report.
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15.0 Accident Analyses 15.1 General The submitted safety analysis evaluates the ability of the WPPSS 1 and 4 plant to operate without undue hazards to the health and safety of the public. Two basic groups of events pertinent to safety are investigated by the applicant; abnormal transients and postulated accidents. All transients and accidents have been evaluated for a core power of at least 102", of the rated power. The environmental consequences of. accidents have been evaluated at a core power level of 3763 MWt.
15.2 Abnormal Transients The criterion, adopted to assure that the reactor coolant pressure boundary integrity is maintained, is that the system pressure shall remain below the code pressure limits set forth in ASME code Section III (110% of RCS design pressure). The criterion adopted to ensure that no fuel damage has occurred is that the DNBR must be greater than 1.32 throughout the transient.
The applicant has submitted analyses of abnormal transients and has shown that the integrity of the reactor coolant system pressure boundary has been maintained and that the minimum DNBR l
exceeded 1.32 for all transients. The maximum pressure transient was identified (BAW-10043, Supplement 1), as the complete loss
of main feedwater from full power, resulting in a peak RCS pressure of approximately 2670 psia. This is less than 110%
of the design pressure (2750) and is acceptable.
The evaluation of abnormal transients indicated that the transients presented do not lead to unacceptable consequences and are acceptable for issuance 'of a construction permit.
15.2.1 Accidents The applicant has evaluated a broad spectrum of accidents that might result from postulated failures of equipment, or their maloperation. These highly unlikely accidents (design basis accidents) that are representative of the spectrum of types and physical locations of postulated causes and that involve the various engineered safety feature systems have been analyzed in detail.
The accidents reviewed in Chapter 15 of the SAR include the following:
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(1) Locked Rotor (2) Loss-of-Coolant (LOCA)
(3) Steamline Rupture The locked rotor accident was analyzed by postulating an instantaneous seizure of one reactor coolant pump rotor. The reactor flow would decrease rapidly and a reactor trip would occur as a result of a high power-to-flow signal. The thermal analysis of the hot
..o rod in the core performed using 102% of rated power and nominal flow, pressure, and inlet temperature. The analysis revealed that 3% of the pins experienced a DNBR less than 1.32 and one percent experienced a DNBR less than 1.0.
The applicant concluded that there was no fuel cladding failure since the maximum clad 0
temperature was calculated to be 1060 F.
The loss-of-coolant accident analyses referenced B4W-10065, Supplement 1, "Multinode Analysis of B&W's 205-Fuel Assembly (Mark C Design) Nuclear Plant During a Loss-of-Coolant Accident" and BAW-10074, "Multinode Analysis of Small Breaks for B&W's 205 Fuel-Assembly Nuclear Plants with Internals Vent Valves."
The evaluation model described in the AEC Interim Acceptance Criteria and Amendments for Emergency Core Cooling Systems was used in the break analyses except for the deviations noted in the topical reports. Pursuant to the latest Acceptance Criteria for ECCS published in the Federal Register on January 4,1974, the applicant is required to resubmit the LOCA analyses satisfying the requirements of the new Criteria as delineated in Section 6.3.3.
As stated previously, the applicant has submitted analyses based on the interim acceptance criteria. For a core power of 3760 ffWt, an 8.55-ft.2 double-ended cold leg break at the pump discharge resulted in the highest peak cladding temperature of 1929 F.
The metal-water reaction for the whole core was 0.112%
which is well below the 1% limit specified in the Criteria.
Loss-of-secondary-coolant analyses have bee erformed to determine the effects and consequences due to a double-ended steam line rupture. A 29.53-inch I.D. steam line rupture, between the steam generator and the main steam isolation valves, and a 42-inch steam line rupture downstream of the main steam isolation valves were analyzed. The analyses assumed that the reactor was operating at rated power prior to the accident. The reactor remained subcritical even with the maximum worth control rod assumed withdrawn.
The applicant has referenced the NSSS supplier reports of analyses of ATWS consequences and evaluation of common mode failures.
These reports, BAW-10019 and BAW-10099 are currently being reviewed
.by the Regulatory staff to determine if the consequences of anticipated transients without scram are limited to the criteria specified in WASH-1270. The staff expects to complete this review by e
May 1975. The plant changes required and the schedule for making these changes will be determined and, reported after the generic analysis and review have been completed.
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REFERENCES 1.
BAW-10043, Supplement 1, "Overpressurization Protection for Babcock and Wilcox Pressurized Water Reactors with 205 Fuel Assemblies, December 1973.
2.
BAW-10065, Supplement 1 "Multinode Analysis of B&W's 205-Fuel-Assembly (Mark C Design) Nuclear Plant During a loss-of-Coolant Accident," August 1973.
3.
BAW-10074, "Multinode Analysis of Small Breaks for B&W's 205 Fuel-w Assembly Nuclear Plants with Internals Vent Valves," November 1973.
3.
BAW-10019 " Systematic Failure Study of Reactor Protection Systems,"
September 1970.
5.
BAW-10099, " Babcock and Wilcox Anticipated Transients Without Scram,"
December 1974.
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