ML17272A267
| ML17272A267 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/18/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 7901290403 | |
| Download: ML17272A267 (12) | |
Text
0 Docket No.:
50-397
Dear f<r. Strand:
Distribution NRC PDR JQIIt 18 197",
Loc DR R.
Boyd D.,Vassallo F. Williams S.
Varga D. Lynch Mr. Neil 0. Strand M. Ser vice Washington Public Power Supply System 300 George Washington Way P.
O.
Box 968 Richland, Washington 99352 D.
Ross J.
Knight R. Tedesco R.
DeYoung V. Moore R. Vollmer M. Ernst R. Denise ELD IE(3) bcc:
J.
SUBJECT:
FIRST ROUND qUESTIONS.ON THE. HNP.-2 OL APPLICATIOil-ASB In our review of your application for an operating license for the WNP-2 facility, ere have identified a need for agditional information which we require to complete,our review
.The specific requests are contained in the enclosure to. this.letter. and are the third set of our round one questions; addit<ona1. requests related to other portions of the WNP-2 facility vli11 be sent over the next few months.
In order to maintain our present tentative schedule, we need,a completely adequate response to all questions..in the enclosure by April 9, 1979.
The attached set of round one questions represents the review effort of the Auxilia~y Systems Branch.
,Where we have been able to formulate statements of staff positions. {RS4), we have included these in our questions.
Please contact us ifyou require any discussion or clarification of the enclosed requests.
Sincer ely, Original signed bye
- s. A. Var@a Steven A. Varga, Chief Light Mater Reactors Branch No. 4 Division of Prospect Management
Enclosure:
Request for Additional Information cc:
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Mashington Public Power Supply System ccs:
Joseph B. Knotts, Jr.,
Esq.
Debevoise 8 Liberman 700 Shoreham Building 806 Fifteenth Street, N.
W-Hashington, D.
C.
20005 Richard g. guigley, Esq.
Washington Public Power Supply System 3000 George Washington Hay P. 0.
Box 968 Richland, Washington 99352 Nepom I Rose Suite 101 Kellogg Building 1935 S.
E. Washington Milwaukie, Oregon 97222 Ns.
Helen Vozenilek 7214 S.
E. 28th Street
- Portland, Oregon 97202 Ns. Susan h1. Garrett 7325 S.
E. Steele Street
- Portland, Oregon 94206 Nicholas Lewis, Chairman'nergy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 JAN 18 1979
0 g r r
">N 18 1979 ENCLOSURE STATEMENT OF REGULATORY STAFF POSITIONS AND RE VEST FOR ADDITIONAL INFORMATION NPPSS NUCLEAR PLANT NO.
2 DOCKET NO.
50-Z97
010. 0 AUXILIARYSYSTEMS BRANCH 010. 10 (3. 4. 1) 010. 11 (3.6)
Demonstrate that all piping and electrical penetrations in safety-related structures that are below the level of the Probable Maximum Flood, are water tight.
Me require you to provide an evaluation of the environmental effects resulting from a postulated failure of the main steam lines and the main feedwater lines.
Your evaluation should demonstrate conformance with our requirements that:
a 0 Those compartments and tunnels which house the main steam lines,'he feedwater lines, including the isolation valves for these lines, are designed to withstand the environ-mental effects (pressure, temperature and humidity) and the potential flooding resulting from a postulated crack equivalent to the flow area of a single-ended pipe rupture in these lines.
b.
The essential equipmept located within these compartments, including the main steam line isolation valves and the feedwater valves and their associated valve operators, are capable of operating in the environment resulting from the crack postulated in Item (a) above.
c.
If the forces resulting from this postulated crack could cause the structural failure of these compartments, the consequent failure of these compartments will not jeopardize the safe shutdown of the plant.
d.
The remaining portion of the pipe in the tunnel between the outboard safety valve and the turbine building meet the guide-lines of Branch Technical Position ASB 3-1, "Protection Against Postulated Piping Failures In Fluid Systems Outside Contain-ment," with respect to the stress levels in this portion of the pipe and with respect to the location of the postulated breakpoints.
Me further require that you submit an analysis of the subcompart-ment pressure buildup following a postulated pipe break, including the structural evaluation of the affected subcomparments, to demonstrate that the design of the pipe tunnel conforms with our positions as stated above.
If you cannot demonstrate conformance with our positions on this matter, indicate any design changes which may be required to comply with our positions.
This evalua-tion should demonstrate that the methods used to calculate the pressure transient in the subcompartments outside of the primary containment, are the same as those used for subcompartments inside the containment for postulated pipe breaks.
Demonstrate that the 010-3
010. 12 010. 13 (3.6) 010. 14 (3.6) 010. 15 RSP margins against a structural failure resulting from the pressure transient, are the same as those in subcompartments inside the primary containment.
If you propose to use methods of analysis for subcompartments outside of containment which are different from those used inside containment, demonstrate that the methods of analysis for subcompartments outside containment provide adequate design margins.
Identify the computer codes and the assumptions regarding the mass and energy release rates which you used in your analysis.
Provide sufficient design data so that we may perform independent calculations.
Provide the results of your evaluation of the jet impingement forces and the environmental effects, including pressure, temperature, humidity and flooding, resulting from a postulated failure of the main steam and main feedwater systems in the turbine building.
This. evaluation should address only those safety-related components, systems and structures, if any, in (or immediately adjacent to) the turbine building (e. g., the walls of the auxiliary building).
For postulated pipe breaks, you have not provided the information required to determine:
(1) the mechanism which terminates the resulting blowdown; or (2) the period of time over which blowdown occurs.
Accordingly, for each postulated pipe break or leakage crack, indicate the time over which blowdown occurs and identify the mechanism which either terminates the blowdown or limits the amount of blowdown flow.
These mass and energy flow rates will be used to evaluate the peak pressures and temperatures in compartments and structures following a postulated break of the high energy pipes inside these structures.
You state in Section 3.6. 1. 1. 1 of the FSAR, that fluid piping systems which the staff would classify as high-energy lines are considered by you to be moderate-energy systems if:
(1) their fluid temperatures are below 200~F and; (2) the fluid pressure is generated by a centrifugal pump instead of a fluid reservoir.
(The staff classification system states that the fluid temperature must be less than 200 F and the fluid pressure must be less than 275 psi for a system to Se designated as moderate-energy.)
Accordingly, demonstrate that these systems do not contain enough energy to cause pipe whip.
Additionally, provide justification for your analysis of flooding based on the moderate-energy crack criteria rather than basing your analysis on the full break required by the high-energy break criteria.
We require that you modify the main steam line isolation valve leakage control system (MSIV-LCS) to satisfy the staff positions contained in Regulatory Guide 1.96, Revision 1, "Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water 010-4
010. 16 (9. 0) 010. 17 (9. 1. 2) 010. 18 (9.i.2) 010. 19 RSP (9.1.2)
Reactor Nuclear Power Plants,"
June 1976:
Specifically, we require that:
a.
the design of the MSIV-LCS permit its actuation within 20 minutes after a postulated loss-of-coolant accident; b.
the leakage control system for the valve stems on the main steam line be designed to the same standards as the MSIV-LCS; and c.
operation of the MSIV-LCS during normal plant operation be prevented by interlocks capable of functioning after a postulated single failure in the interlocking system.
Identify all safety-related equipment that could be exposed to, or affected by, dust storms.
Describe how you propose to assure the proper functioning of this equipment during dust storms.
Provide a description of the methods which will be used to prevent the blockage of vital air supplies to safety-related equipment (e.g.,
clogging of the air filters of the diesel-generators).
In your response to this question, provide a
cross-reference to your response to Item 372.8.
The design of your spent fuel rack includes a neutron absorbing material encapsulated in stainless steel.
However, recent experi-ence at some spent fuel pools has shown that the stainless steel cladding may bow out due to the internal pressure of gases generated by the irradiation of the neutron absorbing material in the spent fuel pool.
This bowing of the steel cladding has caused the spent fuel assemblies to become lodged in the spent fuel racks.
Accord-ingly, describe the method (e. g., venting the stainless steel plates to release any evolved gases) you propose to p'revent this from occur-ring in the WNP-2 spent fuel pool.
In Section
- 9. 1.2 of the FSAR, you list the test results involving radiation, thermal, seismic and borated water testing of the boron carbide plates.
Describe the procedures used for these tests.
Alternatively, provide a cross-reference to any of these test procedures which have previously been accepted by the NRC staff on another application.
In Section
- 9. 1.2.3.3 of the FSAR, you state that the interlocks which prevent the 125 ton crane in the reactor building from traversing the spent fuel pool, are occasionally by-passed.
This by-passing is unacceptable.
. Accordingly, we require you to modify your procedures so that the interlocks on the reactor building crane prevent the crane from traversing over the spent fuel pool whenever there is spent fuel in the pool.
010-5
gC'10.
20 RSP (9.1.2) 010. 21 (9.i.3) 010. 22 (9. 2. 1) 010. 23 RSP (9. 2. 5)
In Section
- 9. 1.2 of the FSAR, you state that a portion of the fuel handling building above the refueling floor is constructed of sheet metal.
Accordingly, we require you to demonstrate that the spent fuel pool is housed in a seismic Category I structure which can withstand the impact of tornado missiles.
'b Provide a cooling system and a source of makeup water for the spent fuel pool which are both designed to seismic Category I criteria in accordance with the staff positions contained in Regulatory Guide
- 1. 13, Revision 1, "Spent Fuel Storage Facility Design Basis,"
December 1975.
Identify which valves are used to isolate that portion of the plant service water system which is not designed to seismic Category I criteria from that portion which is designed to these criteria.
Provide a failure modes and effects analysis for the plant service water system, assuming a seismic event has occurred.
Provide the results of your analysis of the capability of the ultimate heat sink to absorb heat over a thirty-day period following a postulated design basis accident.
Indicate the total heat absorbed in the ultimate heat sink, including the sensible
- heat, the station auxiliary system heat, and the decay heat released by the reactor core.
In particular, provide the following information in both tabular and graphical formats:
a.
The total integrated decay heat.
b.
The heat rejection rate and the integrated heat rejected by the station auxiliary systems, including all operating pumps, ventilation equipment, diesels and other heat sources.
c.
The heat rejection rate and integrated heat rejected due to sensible heat removed from the containment and the primary system.
d.
The total integrated heat rejected; i.e., the sum of Items (a),
(b) and (c).
Additionally, provide the following information:
e.
The maximum allowable temperature of the inlet water taking into account the rate at which heat must be removed, the cooling water flow rate, and the capabilities of the respective heat exchangers.
f.
The required and available net positive suction head (NPSH) at the suction lines of the service water pumps at the minimum water level of the ultimate heat sink.
010-6
This analysis should demonstrate the capability of the ultimate heat sink to provide:
(1) an adequate water inventory; and (2) sufficient heat dissipation which will limit the essential cooling water operating temperatures within the design ranges of system components.
In this regard, we require you to use the methods contained in Branch Technical Position ASB 9-2, "Residual Decay Energy for Light Mater Reactors for Long Term Cooling,"
when evaluating the residual decay energy release rate from the reactor core due to fission product decay and heavy element decay.
Assume an initial cooling water temperature based on the most adverse conditions possible during normal operations.
The meteorological conditions should be established following the guidance contained in Position C. 1 of Regulatory Guide 1.27, Revision 1, "Ultimate Heat Sink for Nuclear Power Plants," tfarch 1974.
010. 24 RSP (9.2.5) 010. 25 (9. 2. 5)
We require that you protect the sprays in the ultimate heat sink from the effects of tornados and tornado missiles.
In the event that a tornado siphons water from the ultimate heat sink (UHS), the make-up water pumps will replenish the UHS.
Demonstrate that the transformers located in the turbine build-ing and the electric cabling which are both required to operate the make-up
- pumps, are protected from tornados and tornado missiles.
010. 26 (9. 2. 5) 010. 27 RSP (9.2.7) 010. 28 (9.3.4)
In Section 9.2.5 of the FSAR, you state that the two ponds which comprise the ultimate heat sink are connected by a siphon that allows water to flow from one pond to the other.
Demonstrate that a failure in this siphon line, or in one of the ponds, will not result in draining of both ponds.
Me require that you protect the standby service water system from tornado missiles.
Describe how flooding of safety-related equipment due to back-flooding through the equipment and floor'drainage
- system, is prevented.
Demonstrate that those portions of the drainage system necessary to prevent backflooding (e.g.,
check valves) are designed to seismic Category I criteria and that their system function will be maintained, assuming a single active failure.
010. 29 (9.4.0)
Demonstrate that the heating, ventilation and air conditioning systems (HVAC) for the engineered safety features,,
are protected from tornado missiles'10-7
010. 30 (9.4.0)
You state in the FSAR that the outdoor design temperature range for the HVAC is O'F to 105 F.
However, you also indicate on Page 9.4-2 of the FSAR, that the extreme outdoor temperature range is -27 F to 115'F.
Provide the results of your analysis which demonstrates that the functional capability of safety-related equipment will not be impaired by the indoor temperatures which would occur during these extreme meteorological conditions.
The effect of extreme low temperatures on safety-related equipment located outdoors should also be discussed.
010. 31 (9.4.4) 010. 32 (9.4. 7) 010. 33 (9.4.7) 010
~ 34 (10.4.5)
The radwaste building chilled water system which is not designed to seismic Category I criteria, is connected to the HVAC system at the control room and to the standby service water system.
Provide your analysis which demonstrates that the potential failure of the radwaste building chilled water system during an earthquake will not cause an unacceptable degradation of the control room HVAC system and the standby service water system.
Demonstrate that the ventilation system of the diesel-generator fuel oil pump room is designed to seismic Category I criteria and receives power from the Class IE busses.
Provide your analysis which demonstrates that the potential failure of the heaters in the diesel-generator HVAC system which are not designed to seismic Category I criteria, will not have an adverse effect on the functional capability of either the diesel-generators or the diesel-generator HVAC system.
Your response to Item 010.09 is unacceptable.
Specifically, your analysis of flooding due to failure of the circulating water system is based on a crack whose area is equal to one-quarter of the pipe diameter times the pipe thickness (0.5 t x 0.5 d).
Provide an analysis of flooding due to a postulated failure of the expansion joint in the circulating water system, assuming a
double-ended guillotine break at this location.
010-8
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