ML20198F594
| ML20198F594 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 03/11/1974 |
| From: | Tedesco R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0961, CON-WNP-961 NUDOCS 8605290014 | |
| Download: ML20198F594 (18) | |
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DIF IBUTION:
L Reading L:CS L:APCSB yn 1 1 1974 Docket File (M-4M Docket No. 50-460 Vons A. Ifoore, Assistant Director for Light Water Reactors Group 2, L REQUEST FOR ADDITIONAL INFORMATION, AUXILLUtY AND POWER CONVERSION SYSTEMS BRANCH Plant Name: Washington Public Power Supply System Nuclear Project No.1 (WNP-1)
Licensing Stage: CP Docket Nunber: 50-460 Responsible Branch: LWR 2-3 Project Manneer: Thomas Cox Requested Completion Date: February 22,1974 (Approved Draft to LPM February 22, 1974)
The enclosed first round request for additional information covers those portions of the FSAR for which the Auxiliary and Power Conversion Systems Branch has primary responsibility. All questions on items not presently included in the standard format are identified by a double asterisk.
Robert L. Tedesco, Assistant Director for Containment Safety Directorate of Licensing
Enclosure:
As Stated
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WASHINGTON PUBLIC POWER SUPPLY SYSTEM WASHINGTON NUCLEAR PROJECT-1 DOCKET NO. 50-460 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.62 The PSAR provides a description of design-basis tornado borne missiles considered for this facility.
Expand the spectrum of tornado-missiles considered to include 3" and 12" schedule 40 pipes, 15 ft long with a density of 490 lbs/ft Apply similar criteria as applied to those missiles already considered in the PSAR and provide similar information as for those already included in Table 3.5-3.
3.63 Section 3.5.3.1 gives a velocity equation for stored strained energy (Class I missiles)' that does not appear to be correct. Verify this equation.
3.64 Discuss the effect that secondary missiles may have on safety-related equipment and systems with respect to safe plant shutdown.
l 3.65 Section 3.5.3.2 describes the method used to obtain the data listed i
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in Table 3.5-3 for tornado missiles. This method assumes that the missiles tumble while in flight. What would be the effect on the data assuming that the missiles do not tumble and are at all times oriented such as to have the maximum value of C A while in flight?
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-4 9.0 AJXILPJY SYSTIMS 00 9.17 In regard to potential failures or malfunctions occurring due to freezing, icing, and other adverse environmetnal conditions for those components not housed within temperature controlled areas and which are essential in attaining and maintaining a safe shutdown, identify and discuss the protective measures taken to assure their operation.
cc 9.18 Provide a tabulation of all valves in the reactor pressure boundary and in other seismic Category I systems, as recommended in Regulatory Guide 1.29, e.g., safety valves, relieve valves, stop valves, stop-check valves, con-trol valves whose operation is relied upon either to assure safe plant shutdown or to mitigate the consequences of a transient or accident. The tabulation should identify the system in which it is installed, the type and size of valves, the actuation type (s), and she environmental design criteria to which the valves are qualified, as stated in the design speci-fications.
ca 9.19 Provide the results of an analysis which demonstrate that failure of any non-Category I auxiliary system or component (including associated turbine systems and components) will not have a detrimental effect (such as flood, spray, leaks) on safety related systems or prevent safe shutdown of the plant.
'9.20 Section 9.1.1.1 states that the fuel storage facility and fuel storage racks are capable of withstanding loads imposed by the dead load of the fuel assemblies, impact and handling, and the safe shutdown earthquake (SSE). Describe in more detail the ability of the new fuel racks to with-stand impact loads from dropped masses.
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.. - Mc 9.21 Section 9.1.1.3 states if it is necessary to conduct lif ting operations over the new fuel storage vault during infrequent maintenance of the fuel transfer system using the new fuel unpacking and inspection area overhead bridge crane, the new fuel in the vault will either be protected by a suitable protective cover or the new fuel will be transferred to the spent fuel storage racks prior to conducting such lifting operations.
Provide a description of the protective cover including a discussion of the seismic design, storage facility and its ability to withstand dropped masses.
9.22 Describe in detail the ability of the spent fuel racks to withstand impact loads from dropped heavy masses.
9.23 Provide a description (including discussion of safety features) of each crane and the degree of compliance with OSHA subpart N Materials Hand-ling and Storage of 29 CFR 1910, Section 1910.179. Identify, discuss and provide a basis for any exceptions and/or deviations taken.
9.24 Provide a list of all major tools and servicing equipment used to perform the various reactor vessel servicing and refueling functions, show their storage locations, and indicate the siesmic category of each.
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- 9. 25 What provisions are being made in the design of the spent fuel pool j
and cask pool in the event the cask is dropped in the cask pool? The l
discussion and analysis should include:
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An outline drawing of the cask, cask dimensions and center of
- gravity, b.
The cask weight, assumed drop height, deceleration distance, deceleration. force versus stroke, velocity of impact consideri'ng deceleration caused by pool water.
c.
The maximum possible drop height.
d.
The means provided aside from administrative centrol to limit the drop height to that assumed in the analysis.
t 9,26 Provide the applicable codes and standards which will be used in the design, fabrication, installation and testing of rails, trolleys, hoists, cables, lif ting hooks, special handling fixtures and slings.
9.21 Prov1de a more detailed description of the new and spent fuel racks and discuss the compatibility of the rack assemblics in their cooling medium so as to prevent galvanic or excessive corro'aiou.
9.28 Describe in detail now normal,make-up is provided to the spent fuel coolfug system.
It is not readily apparent in Fig. 9.1-3 how make-up is introduced into the opent fuel pool. Clarify this condition.
9.29 Section 9.1.3.3 of the PSAR states the normal make-qp supply to the spent fuel pool is Category I.
Fig. 9 3-11 indicates the supply from the make-up and purification system to the spent fuel pool is non-category I.
Correct the PSAR to conform wi'th Regulatory Guide 1.13.
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9.30 Provide additional detailed descriptions and drawings of the following equipment:
a.
new fuel transfer cart b.
fuel handling bridges and overhead cranes c.
fuel transfer canal and rotating fuel baskets d.
spent fuel cask and cask loading area
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e, all major tools and equipment used for refueling or fuel transfer operations 9.31 Section 9.2.1.1 provides a list of equipment supplied by the nuclear service water system. Indicate equipment that is essential and non-4 essential including the seismic category of each. Provide similar infor-mation for the equipment supplied by the component cooling water system listed in Section 9.2.2.2.
l 9.32 Describe in more detail how make-up is provided by the Category I l
demineralized water system to the component cooling water surge tanks.
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9.33 Figure 9.2-9 indicates non-Category I make-up lines to the component cooling water and control rod drive surge tanks and Tig.10.1-4 indicates Category I make-up from the demineralized water storage tank. Correct this inconsistency.
9.34' Section 9.2.2.2 lists the components cooled by the CCWS during various modas of operation. The heatup and power modo of operation lists more heat loads than the first six hours of blackout operation. Justify the need for two circulating water pumps during the first six hours of blackout when only one circulating water pump is needed for heatup and power operations.
J C *9. 35 Tha infornation supplied for the ultimate heat sink is not in sufficient j
detail to pernit a meaningful evaluation of the associated safety related problers.
In order to support your claims that the " sink features" I
meets the suggested criteria of Regulatory Guide 1.27, " Ultimate lleat Sink",, the following information would be required:
a.
The results of an analysis supporting your conclusions, in sufficient detail to permit an independent review; b.
A discussion of how the Regulatory positions set forth in Regulatory Guide 1.27 vere'implesented.
Identify each exception taken and provide the bases; c.
A tabulation and plot spsnning a thirty-day period of (1) the total heat rejected, (2) sensible heat rejected, (3) station auxiliary system heat rejected, and (4) decay heat from radio-active caterial. Use the methods set forth in the October 1971 i
draf t Proposed A55 Standard " Decay Energy Releases Rates Following Shutdown of Uranium-Fueled Thermal Reactors", to establish the heat release due to the decay of radioactive material. Assume an equilibrium fuel cycle
- and calculated heat release as follows:
(1)
For the time interval O to 10 seconds, add 20 percent to dhe heat releasad by the fission products to cover the uncertainty in their nuclear properties.
- In this regard tse the ANS formulation for finite operating time.
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(2) -For the time interval 10 to 10 seconds, add 10 percent to the heat released by the fission products to cover the uncertainty in their nuclear properties.
1 (3) For the time interval 0 to 10 seconds, calculate the heat released by the heavy elements (using the best estimate of the production rate) and add 10 percent to cover the uncer-tainties in their nuclear properties.
In submitting the results of the analysis requested, include the following information in both tabular and graphical presentations:
(a) The heat rate and total integrated heat released due to the fission product decay heat.
(b) The heat rate and total integrated heat released due to the heavy elements.
(c) The heat rate and total integrated heat rejected by the Station i
Auxiliary Systems.
(d) The heat rate 2nd total integrated heat rejected due to sensible heat.
(e) The maximum allowable plant inlet water temperature taking into account:
(1) the rate at which the heat must be removed; (11) the water flow rate, and (iii) the capabilities of the respective heat exchangers.
(f) The required NPSH for the water pumps (taking the required water flow rates and temperatures into account).
(g) The maximum available NPSH for the NSW pumps.
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- s 9.36 Provide a detailed description of the Category I demineralized water storage tank and its distribution system including a list of essential and non-essential equipment or systems supplied by the tank. The des-cription should include how the demineralized water storage tank will be protected from tornado missiles, floods, and any extreme environmental conditions.
9.37 Section 9.3.1.4 of the PSAR states that the nuclear air system has sufficient storage capacity to supply the maximan postulated demands of all safety related air operated components from passive accumulators with no compressor running. Provide the results of an analysis which substantiates this capability. The results should include the maximum postulated consumption rate and the decay rate of the accumulators.
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9.38 The single f ailure analysis of the nuclear instrument air system in section 9.3.1.4 states that an air leak in any single component being operated affecting both headers results in insufficient loss of air pressure to prevent operation of the system. Provide the results of an analysis which substantiates this statement.
9.39 Section 9.3.1.4 states that two independent ring headers will be used l
to supply the nuclear instruments and controls. Fig. 9.3-3 indicates l
l a single header from the nuclear air system to the instrument and con-trols. Clarify this inconsistency and upgrade Fig. 9.3-3 to include the details of the distribution system.
9.40 Provide a list of the essential and non-essential equipment supplied by the nuclear air system and include the failure mode of the essential components listed as well as the normal mode. For the essential equip-l 1
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_9 ment also include the mode the equipment assumes during a design basis accident or LOCA in the event of loss of air pressure.
9.41 Correct the inconsistencies in the seismic category and quality classifications between the nuclear sampling system (Fig. 9.3-4 and 3.2-19) and the following components in their respective system's drawings.
demineralizer pre-filter influent, make-up tank liquid space a.
and deborating demineralizer outlet (Fig. 9.3-11) b.
pressurizer steam and liquid spaces (Fig. 3.2-1) generator liquid sample lines missing on Fig. 3.2-1 c.
steam d.
core flooding tanks (Fig. 3.2-39) decay heat removal system (Fig. 3.2-4) e.
f.
containment spray system (Fig. 3.2-5) c*9.42 Describe the alternate paths that will be provided in the design of the sampling systems for obtaining samples from the reactor or containment during accident conditions.
c*9. 4 8 For all componenta needed for safe shutdown cud accident prevention or mitigation, provide a discussion on the floor drainage system serving the area where the equipment is located.
Identify potential sources of water for which a single failure could causa flooding of the areas and, in this event what effect would it have on the safe shutdown of the plant. Discuss the precautions taken in the design to prevent flooding by the above mentioned sources. Identify the means provided by which the operator will be alerted that water is entering the area, or compartment and the methods available for corrective action.
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, S.44 Figure 1.2-7 shows the decay heat removal pump (item No. 29) at a lower elevation than the local sump (item No.169). Discuss the possibility of a local pipe break flooding out the decay heat removal pump.
Ct9.45 Provide additional explanation and assumptions used for the evaluation of the general services building lower elevation drainage systems, sumps and sump pumps as to capability to collect excess liquid due to an emergency flood condition.
cc$.46 Describe the provisions made in the design to limit the radioactive concentration buildup in all chemical, volume centrol and liquid poison system's components and what is the maximum concentration permitted.
c *9.47 Section 9.4 on HVAC states that in the event a pneumatically controlled damper fails to operate, the dampers have a manual over-ride operated from the control room.
Supply a list of the dampers that are pneumatically operated and describe the operation of the manual over-ride in more detail.
C*S.44 Sections 9.4.1.1 and 9.4.9.3 indicate that hydrogen detectors will be located to monitor hydrogen buildup in the battery rooms during a LOCA.
Specify if hydrogen level will be continuously monitored and if the ventilation system will be designed to maintain low hydrogen levels during all modes of operation to prevent explosive mixture accumulation.
- *9.' 49 Recent operating experience indicates diesel engine operation and performance is affected by oxygen content of combustion air.
Reduction of oxygen in intake air caused by sudden meteorological charges or accidental discharges of noxious gases (or diesel exhaust) in the vicinity of air intakes has degraded (and in some instances curtailed) performance of the diesel generating plant.
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In regard to the diesel engine combustica air intake and exhaust systems, discuss the precautionary measures taken to obtain assurance that the oxygen content of the incoming combustion air will not under any meteor-ological and accident conditions be diluted to an extent as to prevent the diesel from developing full rated power or causing engine shutdown.
Include in the discussion the potential of fire extinguishing (gaseous) medium or noxious gases being drawn into the combustion air system of one or all diesel generators, thereby degrading their performance or possibly result in loss of emergency generator and loss of emergency power supply.
9.50 Section 9.5.1.1 states that the original fire protection system will be isolated from the new fire protection system during major catastrophic occurrences, i.e., earthquakes, tornado, flood, by closing the new iso-lation valves.
Indicate the location of these new isolation valves and include them in Fig. 9.5-1.
C09.51 Discuss the potential fire hazards with regard to materials stored in each area including fire protection requirements and discuss the fire risk evalua-tion to be utilized in the design of a fire protection system.
lr 9.52 Provide a discussion of the ef tects on safety related equipment from operation of the fire protection system. Also provide a discussion
'I relating to the reliability of the fire detection equipment in terms I
of sensitivity, mean time between failures, and other operational
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experiences.
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9.53 Discuss the suppression systemc to be provided for smoke and heat i
control, combustible and explosive gas control, toxic and contaminant control. Describe the operating functions of the ventilating and ex-
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haust systems during the period of fire extinguishing and fire control.
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9.54 Provide additional discussion on the precautionary measures taken in the event of a fire in the control roc =.
In this discussion also include:
A discussion of the fiie protection monitor's ability to detect f
a.
smoke and isolate tha ventilation system.of the polluted air.
b.
The anticipated degradation of control room equipment i'f the design temperature levels are exceeded due to a fire inside or outside the control room.
9.55 Discuss the means by which the operator becomes aware of the existance of. fire protection system pipe breaks or Icaks and determines their loca-tion in the system in order that appropriate action may be taken.
9.56 What provisions have been made for the detection of water in the diesel generator fuel oil storage system and how will potential water accumula-tions be dis' posed of?
9.57 Hydrazine, lithium hydroxide and boric acid appear to be stored in the same room, according to Fig. 1.2-4.
Indicate what measures will be taken in the design to prevent spillage, a toxic or explosive atmosphere and describe the fire protection measures for this area.
9.58 Section 9.5.4 states that diesel oil is pumped from the day tank to the diesel. injector supply manifold by two redundant fuel pumps. Section 8.3 states that each diesel engine is provided with a gravity feed day tank. Clarify this inconsistency.
l 9.59 Sec. tion 9.5.4.2 states that the diesel generator fuel oil storage tank vents are located above the design flood level. Discuss how the vents will be designed to:
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Prevent damage due to tornado missiles.
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- b. Prevent water from entering the storage tanks during adverse environmental conditions.
9.60 Section 9.2.2.2 mentions containment piping penetrations are cooled by the CCWS. This is also indicated on Fig. 9.2-8.
Provide:
- a. a list of the containment penetrations which will be cooled by the CCWS.
- b. a typical detail of a water cooled penetration.
- c. a flow dlagram of the CCWS supplying cooling water to the centainment penetrations listed in (a) above 9.61 Paragraph (d) of Section 9.1.2.3 indicates in the event of a cask drop damaging the water tight gate between the pool transfer canal and cask loading area the water will remain at the level of the gate sill. Pro-j vide the depth of water over the fuel rods under these conditions and what are the radioactivity levels under these conditions.
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10.0 STEAM ARID POIIER CONVERSI01i SYSTDI 10.4 Provide a more detailed description of the main steam isoaltion valves including:
a.
type of valve (more detail) b.
how the single failure criteria are ret.
c.
design parameters of the valve (i.e., expected leakage rate, etc.)
d.
method of equalizing pressure across the valve seat.
00 10.5 Discuss the basis for the main steam isolation valve design leakage rates and acceptance criteria for shop and inplant tests.
In addition provide the basis for the adequacy of the valve design to withstand the accident load imposed during a steam line' break.
10.6 Describe the turbine control and overspeed protection system in sufficient detail to permit an evaluation of the degree of independence and redundancy of components and systems for each function.
10.7 2'or the emergency devices and overspeed controls describe and discuss the degree of compliance with each of the items of section 4 of IEEE-279, Nuclear Power Plant Protection Systems.
CO 10.8 Describe the location, physical sepatation, or protective barriers pro-vided the main and auxiliary feedwater pumps to ensure their operation if ficoding or gross failure of adjacent components or structures were CO occur.
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10.9 Describe, with the aid of drnwings, the bulk hydrogen storage facility including its location and distribution system.
Inclu,de the protective measures taken to prevent fires and explosions during eperation, such as purging the generator, as well as during normal operation.
10.10 Provide tb.e locatiot of all safety related equipment located within the turbine building on plan and elevation drawings.
l 10.11 Provide elevation drawings showing the water level in the turbine l
t building at various times af ter a complete rupture of the main conden-ser circulating water rubber expansion joint. For each time increment discuss which, if any, essential systems and components could be i
rendered inoperable. Include in your discussion the consideration given to passageways, pipe chases, cableways, and all other possible flov paths joining the flooded space to other spaces containing essential systems and components. Discuss the effects of the flood waters on all submerged essential electrical systems and components.
10.12 Describe the means provided to detect a failure in the circulating. water system and har and irc what time interval flow will be stopped, consider-ing all factors, e.g., operator reaction time, drop-out time for control circuitry and coastdown.
Ch10.13 Provide the criteria and basis of design that have been used to preclude the consequences of postulated high and moderate energy piping system ruptures ruptures outside the primary containment from having an adverse effect on safety related structures, systems or components, necessary for safe shutdown. Include in the discussion a failure mode and effects analysis
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relating to systems required for safe shutdown to demonstrate that a concurrent single active component failure will not produce an unsafe
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It is suggested that in preparing the above information you should follow the criteria set forth in Mr. J. O' Leary's letter dated July 12, 1973. In addition to the above information, provide the following:
a.
A tabulation listing all high and moderate energy systems as defined in Mr. J. O' Leary's letter dated July 12, 1973.
b.
For each system indicate the means by which protection is afforded, whether:
(1) The piping system is isolated by adequate physical separation and remotely located from safety systems and components, or (2) the piping system is enclosed within structures suitably designed to protect adjoining safety systems and components, or (3) the piping system is provided with suitable restraints and protective measures such that the operability and integrity of structures, safety systems and components that are required to shutdown the reactor safety and maintain the plant in a cold, shutdown condition are not impaired.
c.
If the piping system is described elsewhere in the PSAR, provide reference to such description ha this tabulation, and if not describe'd elsewhere, include a description.
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