ML19220A436
| ML19220A436 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/18/1968 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Logan J JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| NUDOCS 7904180005 | |
| Download: ML19220A436 (9) | |
Text
e s
Gl>
3:STR:5CT 0N AEC Document 0-r o ' "'a ^
Dockat Jo. 5C-223 00T 18 EE3 5"971-DR Readinz DRL Readinz RP3-2 Readinz 0'4,.
a qr m.* 1 --
Jersey Central Power S Light Company
~i' See;s C.
- tadison Avenue at Punch 3cwl Road
" "- M3^3
- !arristown, New Jersey 07960 R.
S.
Boyd Attention:
Mr. John C. Lo32n
}. { rnblich (2)
~ tea a--- (~)
r" s
Vice President
.occ:
J.
R-3uchanan, 0?:.L Gentlasen:
This refers to your application, dated April 29, 1968, for a con-struction permit and facility license to construct and operate the nuclear power facility designated as Oyster Creek Unit No. 2.
In continuing our review of your proposed facility, we find that addi-tional infor=ation will be necessary.
The attached list ill'is trat es tae type of infor=ation needed on the areas of ener,qency core cooling, contain=ent capability, radioactive waste discosal, and various research and development matters. Our initial request for additional infor=ation, dated Septe=ber 19, 1968, covered the areas of quality control and assurance, contain= eat design, and other structural design
=atters.
It is likely that sc=e of the questions included in t'is request for additional information have either been addressed in other formal submittals or were previously resolved in connection with other cen-struction perrit applicatiens filed with the Cer=1ssion. A s "a
'ndiented in cur ?erterber 19, 1068 letter, you are enceuraced to respond to there queettene F re'2rence to the certinent decunented informarien, including any additional infernation needed for clari-ficatien or ceplification o# these =atter?.
We shall be available to discuss any of the enclosed infornation request if further clarification should be required.
j Sincerely yourn,
/
i i
omc:Nal S?GNED ey
[
Peter s. Marn,
j t
Peter A. Mor:1.9, Director
/
Division of Reactor Licensing l
Enclosure:
/:
Request for Infor=ation
/
cc: James 7. Neely vuv&,2.
mi.ma=,
-a s
3
- mcr> W 2P..
.DRL:RP.. f.
D21 '~
",2E
,' ll?d--.
--/----------
%s 1,
i r
suRNaut >
eve}_1/.dj,.
Ri.T e d e s g g,,,,',,, $', h O in e l2$1ayd ~ g v,
p U!crria-~-
s g/-
/
/
OATE >
.. /--. ?, a..
--l.-
f Form AEC-119 Rev. A C'
- m,y,.,,y ;n.g g-2 g.,
i904180066
JERSEY CENTRAL POWER & LIGHT C0M'AIN OYSTER CREEK UNIT No. 2 DOCKET No. 50-320 REQUEST FOR INFORMATION 1.0 Emergency Core Cooline System (ECCS) 1.1 Describe the results of your analysis on the performance of the ECCS that considers the effect of entrainment of the injected core flooding coolant by the discharged primary coolant using the design basis analysis model for the double-ended break of the primary cold leg piping.
The evaluation should be performed at the ultimate power level of 2772 Mwt.
Describe the margin available with respect to the entrained coolant fraction and resultant fuel clad temperatures.
1.2 Describe your analyses of the ECCS for het and cold leg breaks of less than 0.4 ft2 in break area. Discuss the consequences which would result for bceaks less than 0.4 ft2 if the 1800 psi ECCS signal were inoperative.
What is the range of break sizes for which the containment pressure signal serves as a reliable ECCS actuating signali 1.3 Following the loss-of-coolant accident ;LOCA), the long-term cooling equipment must provide recirculation of cooling water which may contain corrosive elements.
Describe your design considerations that provide protect,1on against detrimental effects to the long-term ecoling systems by corrosion. What corrosive elements and concentration are considered?
1.4 Current R&D is being performed by B&W to determine the effectiveness of the containment spray systems using sodium thiosulf ate to remove radio-active iodine from the containment atmosphere.
If the results of this R&D effort indicate the iodine reduction capability is not sufficient to meet 10 CFR 100 limits, indicate what alternative you would consider, such as a filter system inside the centaitment. Our preliminary calcu-lations indicate that the iodine reduction factor necessary is of the order of 6.5, which is in excess of the minimum capability of your proposed containment spray system.
1.5 The design capacity of the emergency containment cooling coils is 80x106 BTU /hr for each coil at peak reactor building pressure and tem-perature conditions.
Provide a description of any tests to be performed to ensure the cooling unit f ans and cooling coils can meet the design re.quirements, considering the effects of the post-accident environmental conditions and the long-term operating requirements.
1.6 Describe the capability of the f acility design to provide flooding of the cavity around the pressure vessel following a design basis accident including your considerations regarding the time to flood the cavity to the top of the active core.
~
y3 001
1.7 Section 6.1.1 in the PSAR contains the ECCS design bases which includes a criterion of "no clad melt"; however, you state that equipment is sized to limit fuel clad temperatures to 2300*F or less.
This difference between the design bases and design performance requirements should be clarified and made consistent.
1.8 Figure 14-48 in the PShR shows graphically the capability of the subsystems
.of the ECCS to provide sufficient emergency core cooling over the primary system break spectrum.
Because of the different design capabilities of each subsystem to cope with certain size breaks, please describe the design bases and criteria that establish the maximum capability of each subsystem or conbination of subsystems.
2.0 Radiolytic Hydrogen Generation 2.1 Discuss the capability of the ECCS to subcool the core and prevent signi-ficant steam generation af ter blowdown.
Consider the cold leg break sizes which might result in a pressure buildup and determine the time interval required to subcool the core as a function of ninimum ECCS delivery.
This is significant in terms of (1) core cooling, (2) long-term requirements on containment spray system, (3) radiolytic decomposition, and (4) recovery from the accident.
2.2 Describe the results of your analysis of the potential hazard that might result from radiolytic and chemical hydrogen formation under post-accident conditione.
Discuss the effect that inerting might have on this prcblem.
Include an estimate of the total alpha, beta and gamma dose in both the.
core region and in the spray water.
Indicate the extent of hydrogen formation by chemical reaction with exposed reactor and containment materials.
Discuss the sensitivity of the calculations to the length of time that boiling is assumed to continue in the core.
Provide possible alterna-tives you would consider to prevent hydrogen from reaching an explosive concentration in the containment.
3.0 Emer ency Diesel 3.1 Provide a tabulation similar to Table 8-1 in the PSAR showing the required loads on the emergency diesci to bring the reactor from nor-mal operation at power to a cold shutdown condition with loss of off-site power.
Include a discussion of the sequence of events and time required to reach the cold shutdown condition.
4.0 Centainment 4.1 Provide an eTalys!s and discussion of the capability of the containment to withstand the LOCA for hot leg or cold leg breaks with simultaneous gross failure of one steam generator loop.
For this case, indicate the margin between the peak containment pressure, in terms of water-metal energy release permissible, and the 60 psig design pressure.
In addition, discuss the margin that may be available with respect to the 69 psig test pressure.
4 4.2 Location of Unit 2 adjacent to Unit 1 makes pocsible for missiles generated from Unit 1 components to strike the Unit 2 buildings.
Provide an analysis and discussion of these missiles and the protec-tion provided by Unit 2 building designs to prevent any missile from causing an accident which will result in an uncontrolled radioactivity release.
5.0 Fuel Handling 5.1 In the PSAR description of the spent fuel transfer operations, the spent fuel shipping cask will be lifted from the-receiving platform and lowered into the shipping cask pool at the end of the spent fuel storage canal.
During this cask transfer operation the cask could become a potential missile if the crane or cables should fail.
Describe the results of your analysis cons Jering the consequences which would result if the shipping cask were to fall into the fuel storage canal during the cask transferring operations including the radiological effects resulting from damaged fuel rods.
Discuss the facility design features that are provided to prevent this accident.
5.2 Your analysis in the PSAR describing the consequence of a fuel element being dropped or damaged assumes the accident occurs in the spent fuel storage pool in the auxiliary building. Since this accident could occur inside the containment building, provide an analysis and indicate the conditions of the containment assumed during the refueling opera-tions. You have indicated that the pool cover described in the PSAR may not be included in your final pool design.
Give justification for its removal and irlicate if its removal would affect the fuel handling accident analy:cd in the auxiliary building.
6.0 Isolation Valves 6.1 Table 5-1 in the PSAR presents a listing of all the reactor building isolation valves but does not include any specification as to actuation tiae to limit the radioactivity released from the containment.
Provide your design criteria and bases for the listed isolation valves regarding actuation time requirements for opening and closing operation and requirements for operation under assumed design bases accident condi-tions.
Indicate testing program to verify these operation times during installation and during the 40-year plant life.
7.0 Xenon Oscillations 7.1 on page 3-31 of the PSAR it is indicated that axial xenon oscillations are assumed to occur and the resulting oscillations will be compansated e
2
. by repositioning rods.
Figure 3-9 indicates that you intend to control axial xenon oscillations by using partial length control rods; however, no design criteria are provided. Consequently, we will need the following information:
(1) describe the partial length rods, their operation and design criteria with respect to the other control rods; (2) give expected locations in the core and number required; (3) discuss how shutdown margins are affected because full length control rods may be replaced with partial length control rods; and (4) discuss any changes in accident analysis which may result if partial length rods are used.
8.0 Pressure Vessel (nyt) 8.1 Recently reviewed P'4R designs have indicated that the uncertainty could be as much as a factor of two in the calculated maximum nyt at pressure vessel wall. What further calculations or experirents are being done or are planned to confirm the maximum nvt on the pressure vessel wall? Are the flux values given in Table 3-7 for the ultimate power level? Since the surveillance specimens may be used to confirm the maximum nyt at the pressure vessel wall, describe the type of fast neutron dosimetces which'are being considered for this application.
9.0 Technical Scecifications 9.1 Identify and discuss those items that will eventually be classified as technical specifications that may affect the plant design; for example,.
minimum conditions of operation for engineered safety features, emer-gency generators, and in-core flux monitors, etc.
10.0 Pressure Vessels (Desien) 10.1 Provide a tabulation of all the nuclear pressure vessels in the facility that are designed as Class A vessels in accordance with ASME Section III.
The tabulation should include a notation of whether the vessel design is complete, the stage of fabricatica of the vessel, and the extent to whi:h each of the vessels will ecmply with each of the 34 " Tentative Regtlatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressuie Vessels" issued by AEC Press Release No. IN-817, Jaced Auguit 25, 1967.
10.2 For e ach nuclear pressure vessel, provide a discussion that presents the reasc as why compliance is not feasible for each of the 34 criteria that is not met in its entirety.
h b
h
. 11.0 Research and Develocment The PSAR indicates the following areas are included as R&D items:
a.
xenon osc'_ilations b.
thermal and hydraulic programs c.
fuel rod clad failure d.
high burnup fuel tests e.
internal vent valves f.
control rod drive line test g.
once-through steam generator test h.
self-powered detector tests 1.
blowdown forces on internals J.
chemical spray syctem for iodine removal.
In order that the above R&D programs planned, completed or in progress can be evaluated, the following information should be supplied for each of the above areas.
11.1 Identification of the specific area where further technical or design information was needed and why these areas are important to complete the safety evaluation in each area, 11.2 Define the information which will be obtained to answer the above safety questions; define the criteria which will be used to establish if the information developed confirms the safety expectations established at this stage of the design.
If these acceptance criteria are not known at the present time, indicate a schedule for development of such criteria.
11.3 Provide a program schedule of suf ficient detail to show relevancy to the schedule of decision making and construction for the plant.
Include where work is to be done, who is responsible for carrying out the work, and identify the level of activity, 11.4 Identify and describe the bases upon which you conclude that the R&D program is adequate and that reasonable alternates are available if required.
12.0 Waste Handling and Discharze 12.1 Jiscuss the discharge of liquid waste from the site with respect to the sources of radioactive wastes, including a presentation of the following data:
(a) Tabulation of the maximum amounts of each radionuclide which will be contained in the liquid waste process tanks at any one time.
(b) Maximum quantities of each radionuclide which will be discharged annually, as well as the maximum short term and yearly average concentrations of each at the point of discharge.
~
- w.- ~
ta t i)
. 12.2 Provide an analysis which relates primary coolant activity, assumed leakage rate from primary to secondary system, removal and cleanup mechanisms for the secondary system coolant, and the derived activity contained in the secondary system.
Present an analysis of the offsite consequences of a steam line ruptu_e based on the derived secondary-activity.
12.3 Our calculation of the site boundary doses reeulting frem waste gas decay rupture dif s with those certained in Section 11.1.2.5.3.
Please describe tF* model used for caleslating whole body doses off-site as a result c_ the accids al rei ese of gaseous fission products.
Include values for all raram
<rs used la the dose calculations.
12.4 Section 2.4.6 of the PSAR discussed the effect of liquid waste discharges on fish and wildlife in Barnegat Bay.
I r. is not clear how that informa-tion applies to the disc'arge of liquid wastes to the Atlantic Ocean as proposed for the comb _..ed operation of Units 1 and 2.
What studies are planned to determine the effect on fish and wildlife of the discharge into the Ocean? What studies are planned of the types and population of significant ocean marine life in the vicinity of discharges How is the environmental monitoring program described in Section 2.7, which was initiated in March 1966, to be modified to include anticipated operation of Unit 2.
12.5 The equation at the top of page 2-5 of the PSAR was used to calculate the long-term average dispersion of effluent gases from the plant.
We do not believe it appropriate to include in this calculation the com-bined effects of sector averaging and dilutien in the turbulent wake of the building.
Please reevaluate the calculated diffusion values incorporating sector averaging only.
12.6 With rcspect to tritium generaticn and control, please provide the following:
a.
Identify the sources of tritium production in the reactor including the bases for evaluating the production rates.
What uncertainties are associated with t'.c production t
rates?
b.
What factors affect the replacement rate of the primary coolant? Show how the holdup period is estimated.
What is the experimental confirmation af :he holdup periods?
c.
What fraction c_' the tritium produced from each of the sources is released to the primary coolant? How have these values been determined? What are the uncertainties in the values?
48 3C6
- d.
What fraction of the tritium released to the primary coolant is released in the liq.uid wastes?
e.
Describe how the tricium concentrations in the discharge canal will be evaluated for average yearly discharges and for conditions which on a short term might lead to higher concentrations.
f.
How will the tritium concentrations in the discharge canal and in the mixed water source (1;e., flowing river) compare with the tritium already present in the water source?
g.
What measures are necessary to ascertain that the tritium wastes pose no problem censidering the range of the uncertainties in tha valuation?
h.
What quantities of activity could accumulate in the condensate and borated water storage tanks and what would be the conser,uences of failure of these tenks?
13.0 Iodine Reduction Capab4 ;.ity 13.1 Justify the assumption that only 5% of the iodine available for leakage after the loss-of-coolant accident is in a non-removable form.
13.2 The containment atmosphere recirculation system is not equipped with HEPA* fi}ters. What fraction of the non-removable iodines in the contain=ent atmosphere following the LOCA are assumed to be in par-ticulate form? Justify this assumption.
Is the system design amenable to the installation of HEPA and charcoal filters if needed?
- High Efficiency Particulate Absorber 48~3C7
.