ML18031A034
| ML18031A034 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/22/1978 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 7812050048 | |
| Download: ML18031A034 (13) | |
Text
'
f Docket Hos. 50-a
-38 II
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Distribution w encl Doc et HRC PDR Local PDR Ll(R 83 File R.
Boyd D. Vassallo, F.. Williams
. Parr
. Niner I
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. Hattson Ross
- Knight, Tedesco DeYoung Moore Vollmer Ernst I
R. Denise R. Hartfield, t<PA OELD IE (3)
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JBuchanan TAbernathy ACRS (16) ter.
Norman ll. Curtis.
'Vice President - Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 NOY g2 ignis
Dear fir. Curtis:
SUBJECT:
SUSQUEHANNA STEAl'l ELECTRIC STATION UllIT thOS.
1 AHD 2-REQUEST FOR ADDITIONAL INFORMATION As a result of our'review of your application for operating licenses for the Susquehanna Steam Electric Plant, we find that we need additional information in.the area of Auxiliary Systems.
,The specific information required is listed. in the Enclosure.'lease inform us of the date when this requested additional information will be available for our review.
Please contact us if you desire any discussion or clarification of the information requested..
Sincerely, Or)jlnal Slgnef bP
- 0. 6. -Pirr
':,."'-'Olan D. Parr, Chief Light Mater Reactors Branch Ho.
3 Division of Project Management
Enclosure:
As Stated cc w/enclosure:
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Mr. Norman W. Curtis NOV 82 87B cc:
Mr. Earle M. Head Project Manager Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Jay Silberg, Esq,.
Shaw, Pittman, Potts 8
Trowbridge 1800 M Street, N.
W.
Washington, D. C.
20036 Mr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.
Pennsylvania Power 5 Light Company 901 Hamilton Street Allentown, Pennsylvania 18101 Robert H. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsylvania 18655
ENCLOSURE RE UEST FOR ADDITIONAL IHFORtlATION SUS UEHAHHA STEAtl ELECTRIC STATION DOCKET HOS.
50-387 AND 50-388
010.0 AUXILIARYSYSTEMS BRANCH 010.6 A single failure of an inboard MSLIV would allow a continuous blowdown (RSP)
(6. 7) of tl e containment atmosphere to the reactor building standby gas treatment system for a specified period of time when the NSIVLCS is initially actuated.
This violates our containment isolation criteria and the corsequences of the blov down are unacceptable.
It is our position that an interlock be provided so that the leakage control system actuation valves can be opened only if the associated inboard 010.7 (RSP)
(6.7)
NSLIV is in a fully closed position.
Revise the FSAR to indicate conformance to our position.
The design criteria for the main steam isolation valve leakage con-trol system indicates that you propose to allow a main steam isolation
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valve (NSIV) leakage rate up to 100 SCFH for each NSIV in each steam-line. It is our position that the design basis leak rate of 100 SCFH is not an acceptable MSIV leakage rate for normal operation.
Therefore, we will still impose a technical specification limit of 11.5 SCFH for the MSIV leak rate and a leak rate verificatioh testing frequency consistent with the plant Technical Specifications used for other operating BWR's.
Revise the FSAR to indicate that the NSIV leak rate for normal operation will be limited to 11.5 SCFH.
010-2 010.8 Confirm that a Keff of less than 0.98 will be maintained with fuel (RSP)
(9. 1. 1) of 'the highest anticipated reactivity in place in the new fuel storage racks and assuming optimum moderation.
010.9 (RSP)
(9.1.2)
The information contained in the Susquehanna FSAR is not of sufficiert detail to support a conclusion that the liner plate for the spent fuel pool is designed to seismic category I. Therefore, we require, that you demonstrate compliance with Regulatory Guides 1.13 and 1.29 by show-ing that a failure of the liner plate as a result of an SSE will not affect any of the following: significant release of radioactive materials due to mechanical damage to the spent fuel; significant loss of water from the pool which could uncover the fuel and lead to release of radioactivity due to heat-up; loss of ability to cool the fuel due to flow blockage caused by a portion or one complete section of the liner plate falling on top of the fuel racks; damage to safety related equipment as a result of pool leakage; or uncon-trolled release of significant quantities of radioactive fluids to the environs.
010.10 (RSP)
(9.1.2)
Confirm that all portions of the structure (reactor bui lding) which serve as a low leakage barrier to provide atmospheric isolation of the spent fuel storage pool and associated fuel handling area are designed to seismic Category I criteria.
010-3 010. 11 (RSP)
(9.1.3)
The spent fuel pool cooling system is a non-seismic system.
This does not meet the guidelines set forth in Regulatory Guide 1.13 and 1.29.
Analyze the design of the spent fuel pool cooling system to show that the pumps and piping are supported so that they are capable of withstanding an SSE, or provide the results of an analysis to show that for the complete loss of fuel pool cooling that would result in pool boiling, a release of significant quantities of radioactivity to the environment will not result.
01 0.12 (9.1.3)
Confirm that a spent fuel pool water temperature of 125'F is main-tained when the fuel pool cooling system is used to cool the emergency heat load.
01 0.13 (9.1.3)
Based on information provided in your FSAR, it appears that either the spent fuel pool is capable of storing over 2 1/2 full cores or has a storage capacity for 4 1/2 full cores.
State the design bases storage capacity provided for the spent fuel pool.
01 0.14 (g.i.3)
The decay heat during normal
( 1/4 full core load, plus previous refueling loads) storage conditions has not been provided.
Assuming that fuel assemblies from 1/4 of a full core load are placed in the pool 7 days after reactor shutdown and the remaining storage spaces are filled with spent fue'I from previous refuelings, reevaluate the spent fuel pool cooling system's and the residual heat removal system's capability using the heat loads determined by the methods set forth in
010-4 Branch Technical Position ASB 9-2, "Residual Decay Energy for Light Mater Reactors for Long Term Cooling."
Also, reevaluate the systems capability for the emergency (1 full core unloaded from the reactor 7 days after shutdown plus the normal refueling load that has been in the pool for 30 days) storage conditior.
For both the normal and emergency storage condition stage the maximum'decay heat load, the maximum spent fuel pool temperature.
and provide the time required to raise the temperature of the pool to boiling assuming the coolinc systems are not available.
Our criteria for safety related cooling systems is that sufficient cooling must be provided for at least 30 days:
(1) to permit simultaneous safe shutdown and cooldown of both nuclear reactor units and maintain them is a safe shutdown condition, or (2) to mitigate the consequences of an accident in one unit and a safe shutdown and cooldown in the other unit and maintain it in a safe shutdown condition.
Expand table 9.2-5 in the FSAR to cover this 30 days time span.
The emergency service water system (ESWS) is designed to take water from the spray pond and provide coolinc to safety related components during safe shutdown and accident conditions.
During safe shutdown or the loss of offsit'e power non-safety related components are cooled by the ESWS.
Demonstrate that the safety function of the system will not be affected assuming a failure in the non-safety related portion of the system coincident with a single failure in the safety
010-5 related portion of the system.
Also, provide an evaluation of the effects of flooding on safety related components.
010.17 (g.2.5)
The reactor building closed cooling water (RBCCM) heat exchanges and turbine building closed cooling water (TBCCW) heat exchanges are not designed to seismic Category I requirements.
However, these components are cooled by the safety related emergency service water system and isolated by a single isolation valve.
This does not meet the single failure criterion.
Revise your design to meet single failure.
010.18 RSP) g.2.7)
In order to permit an assessment of the Ultimate Heat Sink, provide the results of an analysis of the thirty-day period following a design basis accident in one unit and a normal shutdown and cooldown in the remaining unit, that determines the total heat r ejected, the sensible heat rejected, the station auxiliary system heat rejected, and the decay heat release from the reactors.
In submitting the results of the analysis requested, include the following information in both tabular and graphical presentations:
(1) The total integrated decay heat.
(2) The heat rejec:-on rate and integrated heat rejected by the station auxiliary systems, including all operating pumps, ventilation equipment, diesels, spent fuel pool
- makeup, and other heat sources for both units.
(3)
The heat rejection rate and integrated heat r ejected due to the sensible heat removed from containment and the primary system.
010-6
{4)
The total integrated heat rejected due to the above.
(5)
The maximum allowable inlet water temperature taking into account the rate at which the heat energy must be
- removed, cooling water flow rate, and the capabilities of the respective heat exchangers.
(6)
The required and available NPSH to the Emergency and RHR service water pumps at the mi nimum Ultimate Heat Sink water level.
The above analysis, including pertinent backup information,'is required to demonstrate the capability to provide adequate water inventory and provide sufficient heat dissipat.'on to limit essential cooling water operating temperatures within the design ranges of system components.
Use the methods set forth in Branch Technical Position ASB 9-2, "Residual Decay Energy for Light >/ater Reactors for Long Term Cooling," to establish the input due to fission produce decay and heavy element decay.
Assume an initial cooling water temperature based on the most adverse conditions for normal operation.
010-7 01 0.19 (9.2.7)
Sufficient informatior, is not available for us to evaluate t e the plant safe shutdown capabilities from internal flooding of the engineered safeguard service ~ater pumphouse.
For a moderate energy leakage crack in the residual heat removal service water system pip-ing or the emergency service water system piping, determine the effects of flooding on the safety related pumps located within the pump cubicle assuming 30 min. for any operator action.
Also, des-cribe any communication pathways between service water system pumps cubicles for loops A and loop B.
010.20 The Reactor Building chilled water system is designed seismic
'ategory I from the isolation valve outside containment to piping (9.2.12) just inside containment.
Figure 9.2-13B in the FSAR does not show
..'ny safety related valving inside containment for system isolation.
The system and components inside containment and outside the containment penetrations are not seismically designed.
The rupture of these non-sei.smic
- ines, 1'lus a single active failure of the isolation....,
value outside cortainment would cause a breech of containment.
Provide the required isolatior valves inside containment.
010.21 Your FSAR does not evaluate the effects of an expansion Joint failure at the condenser.
Expand the information provided to include an (10.4.4)
010-8 evaluation regarding the effects of possible circulating water system failure inside the turbine building.
Include the following:
(1)
The maximum flow rate through a completely failed expansion joint.
(2)
The potential for and the means provided to detect a
failure in the circulating water transport system barrier such as the rubber expansion joints. Include the design and operating pressures of the various portions of the transport system barrier and their relation to the pressures which could exist during malfunctions and failures in the system (rapid valve closure).
(3)
The time required to stop the circulating water flow (time zero being the instant of failure) including all inherent delays such as operator reaction time, drop out times of the control circuitry and coastdown time.
(4)
For each postulated failure in the circulating water transport system barrier give the rate of rise of water in the associated spaces and total height of the water when the circulating water flow has been stopped or overflows to site grade.
010-9 (5)
For each, flooded space provide a discussion, with the aid of drawings if necessary, of the protective barrier provided for all essential systems that could become affected as a result of flooding.
Include a discussion of the consideration given to passageways, pipe chases and/or the cableways joining the f1ooded space to the spaces containing safety related system components.
Discuss the effect of the flood water on all submerged essential electrical systems and components.