ML20032A205
| ML20032A205 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/15/1981 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8110280479 | |
| Download: ML20032A205 (24) | |
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Vice President - Steam Production OELD W
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Duke Power Company 0IE (3)_
P.O. Box 33189 bec: L/PDR Charlotte, horth Carolina 28242 NRC/PDR NSIC/ TIC / TERA
Dear t;r. Parker:
ACRS (.16)
Subject:
Request for Additional Information In the performance of the Catawba station licensing review, the staff has identified concerns with regard to the following areas:
1.
Auxiliary systems (Enclosure 1) 2.
Siting analysis - Safety (Enclosure 2) 3.
Radiological Assessment - Environmental -(Enclosure 3) 4.
Raciological Assessn.ent - Tiil issues (Enclosure 4) 5.
Radiological Assessaent - Non-Tril issues (Enclosure 5) 6.
Environmental Engineering (Enclosure 6)
Our review in other areas will be completed in the near future
- and we will send you separate requests for additional infomation related to those areas.
We request that you provide tr.a information herein requested no later than December 1,1981.
If you require any clarification of this request. please contact the project manager, Kahtan Jabbour, at (301) 492-7821.
Sincerely, Elinor G. Adensam, Branch Chief Licensing Branch #4 Division of Licensing
Enclosures:
As stated cc: See next page g 05h0 79 O 3-0
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NRC FORM 318 9tO<80i NRCM O240 OFFICIAL RECORD COPY.
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OCT 151981 Docket Nos.:
50-413/414 Mr. William 0. Parker, Jr.
Vice President - Steam Production Duke Power Company P.O. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
Subject:
Request for Additional Information In the performance of the Catawba station licensing review, the staff has identified concerns with regard to the following areas:
1.
Auxiliary systems (Enclosure 1) 2.
Siting analysis - Safety (Enclosure 2) 3.
Radiological Assessment - Environmental (Enclosure 3) 4.
Radiological Assessment - TMI issues (Enclosure 4) 5.
Radiological Assessment - Non-TMI issues (Enclosure 5) 6.
Environmental Engineering (Enclosure 6)
Our review in other areas will be completed in the near future; and we will send you separate requests for additional information related to those areas.
We request that you provide the information herein requested no later than December 1, 1981.
If you require any clarification of this request, please contact the project manager, Kahtan Jabbour, at (301) 492-7821.
Sincerely, f-n
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m Elinor G. Adensam, Branch Chief t.icensing Branch #4 Division of Licensing
Enclosures:
As stated cc: See next page
CATAWBA Mr. William 0. Parker Vice Presideit - Steam Production Duke Power Company P.O. Box 33189 Charlotte, North Carclina 28242 cc: William L. Porter, Esq.
North Carolina Electric Membership Duke Power Company Corp.
P.O. Box 33189 3333 North Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Ralcigh, North Carolina 27611 J. Michael McGarry, III, Esq.
Debevoise & Liberman Saluda River Electric Cooperative, 1200 Seventeenth Street, N.W.
Inc.
Washington, D. C.
20036 207 Sherwood Drive Laurens, South Carolina 29360 North Carolina MPA-1 P.O. Box 95162 Janes W. Burch, Director Raleigh, North Carolina 27625 Nuclear Advisory Counsel 2600 Bull Street Mr. R. S. Howard Columbia, South Carolina 29201 Power Systems Division Westinghouse Electric Corp.
Mr. Peter K. VanDoorn P.O. Box 355 Route 2, Box 179N l
Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, J r.
NUS Corporation
?536 Countryside Boulevard Clearwater, Florida 3?515 Mr. Jesse L. Riley, Pitsident Carolina Environmental Study Group 854 Henley Place l
Charlotte, North Carolina 28208 Richard P. Wilson, Esq.
Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 l
Columbia, South Carolina 29211 Walton J. McLeod, J r., Esq.
General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, South Carolina 29201
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ENCLOSURE 1 AUXILIARY SYSTEKS BRANCH REQUESTS FOR ADDITIONAL INFORMATION CATAWBA NUCLEAR STATION DOCKET NOS. 50-413/414 41 0.3 You have identified a static probable maximum flood elevation of 592.4 (3.4.1) ft. msl in Section 2.4.1.1 and Table 2.4,3.9 of the FSAR.
Clarify the apparent discrepancy which appears in FSAR Section 9.2.1.2.2 where a probable maximum flood elevation of 598.6 ft. ms1 is identified.
410.4 You have indicated that you provide missile protection by (1) separating (3.5.1.1) redundant safety-related systems, (2) restricting failure modes to safety-related ASME Section III components, (3) assuming non-safety-related components are of insufficient size and energy to do significant damage, and (4) not postulating the pressure boundary failure of ASME Section III components. We cannot verify your position that non-safety-related components are of insufficient size and energy to do significant damage to safety related components.
Provide an analysis which identifies what protection is provided for all safety-related equipment from missiles generated by non-safety-related sources.
Consider potential missiles generated by non-safety related pressurized and rotating sources inside protective structures which house safety-related equipment.
List and describe the size and energy of these non-safety-related missile sources which can impact safety-related equipment and identify the method of protection for the safety-related equipment 41 0. 5 Provide a list of all safety-related components which are located (3.5.2) outdoors or are exposed to torrado generated missiles.
Describe the protection afforded each of these components to prevent their being
. damaged by tornado. generated missiles in accordance with the guidelines of Regulatory Guide 1.117. " Tornado Design Classification." Include in this list a description of all HVAC system air intekes and exhausts including the protection afforded safety related equipment near these openings, as well as emergency diesel exhausts, freight doors, any other openings in structures housing safety related equipment, and any exposed piping. Identify the locations of these components on plant arrangement drawings.
41 0.6 Provide justification and documentation for your position that non-liquid
( 3. 6.1 )
and non-water carrying piping systems over 200*F not be considered high energy systems. The information provided in FSAR Table 3.6.1-3 is insufficient justification.
It is our position that all plant fluid systems which operate at temperatures exceeding 200 F or pressuresexceeding 275 psig during normal plant conditions be considered as high energy systems and protection afforded for safety related systems from their failure. Provide the results of the necessary analysis to assure com-pliance with the guidance of Standard Review Plan (SRP) Section 3.6.1.
41 0.7 Provide an analysis of the effects on safety related systems of failures (3. 6.1 )
in all high-or moderate-energy piping systems in accordance with the applicable criteria of Standard Review Plan (SRP) Section 3.6.1 and identify the criteria you are applying to your plant design. Expand Tables 3.6.1-1 and 3.6.1-2 to correlate interactions between systems.
For each safety-related system, indicate which high-and/or moderate-energy system can affect its safety function and identify the specific protective method provided for the safety-related system from the postulated failures.
Include area layout drawings depicting the locations of
i failures affecting safety-related systems giving dimensions, locations, and pre:ective method for each postulated break or crack in a high-or w derate-energy system.
Include the assumptions us.J in your analysis such as flowrate through postulated cracks, room volumes, sump capacities, and floor drainage system capacities.
Your analysis should verify the capability to sustain any postulated high energy piping system failure concurrent with any single active failure.
41 0. 8 We require that the structures housing main steam and feedwater lines (3.6.1) and isolation valves for those lines be designed to consider the environmental effects (pressure, temperature and h6midity) and potential flooding consequences from an assumed crack of one square foot in area.
We are unable to determire the physical arrangement of your doghouse and steam t0nnels relative to the main steam and feedwater lines, isolation valves and their communication with auxiliary feedwater system' components. We reouire that essential equipment located within structures housing the main steam and feedwater lines, or within structures communi-cating with such lines, be capable of operating in the environment resulting from the assumed crack ' described above, and that structural failures resulting from such cracks should not jeopardize the safe shut-down of the plant.
This capability should be assured assuming any con-current single active failure.
l Provide a subcompartment pressure and temperature analysis including I
l layout drawings of the structures housing steam lines between contain-ment and the turbine building to confirm that the design of the structures conforms to our position as outlined above.
L
... We reouest that you evaluate the design against this staff position and advise us as to the outcome of your review, including any design changes which may be required. The evaluation should include a veri-fication that the methods used to calculate the pressure buildup in the subcompartments outside of the containment for postulated breaks are the same as those used for subcompartments inside the containment.
Also, the allowance for structurcl design margins (pressure) should be the same.
If different methods are used, justify that your method provides adequate design margins an'd identify the margins that are available. When you submit the results of your evaluation, identify the computer codes used, the assumptions used for mass and energy releases including sufficient design data so that we may perform indepen-dent calculations and the results of the analysis including pressure /
temperature vs. time profiles.
The peak pressure and temperatures resulting from the postulated break of a high energy pipe located in compartments or buildings is dependent on the mass and energy ficws during the time of the break. You have l
not provide the information necessary to determine what tenninates l
l the blowdown or to detennine the length of time blowdown exists.
For each pipe break or leakage crack analyzed, provide the total blowdown time and the mechanism used to determine or limit the blowdowm time of l
flow so that the environmental effects will not affect safe shutdown of the facility.
1 i
. In addition to the above, provide a similar subcompartment pressurization and environmental effects analysis for other high energy lines outside containment (such as CVCS letdown, charging, auxiliary steam, steam generator blowdown, etc.) which verifies that safety related equipment in the vicinity of these lines is protected against any structural failures and is qualified for the resulting environment. A full double ended rupture at the postulated break locations should be assumed for these high energy lines.
410.9 Expand and clarify your discussion of the reactor coolant pressure (5.2.5) boundary (RCPB) leakage detection systems. Describe how your systems meet each of the positions of Regulatory Guide 1.45.
Provide information to show how total identified leakaga flowrate is measured.
Indicate the method to be used to obtain an accuracy of 1 gpm or better in one hour for unidentified leakage by the cyclic operation of sump pumps.
Describe more fully the use of the volume control tank level for monitoring RCPB leakage flowrate.
For each source of intersystem leakage describe the three separate detection methods as recommended in Regulatory Guide 1.45, position C3.
410.10 Provide the specific K values determined in your criticality analysis eff (9.1.1) for the new fuel storage arrangement with the associated assumptions and input parameters.
Clarify your assumption regarding water moderation when maximizing K Also, verify that the new fuel storage racks are eff.
capable of maintaining a K f 0.98 or less under optimum moderation eff (foam, small droplets, spray or fogging) or identify the means provided for preventing such a condition in the new fuel storage vault, n_-.-..._...-....-
. Describe the seismic and tornado qualification of the railroad freight door into the new fuel storage building and the features inside the building to prevent tornado missiles, entering through a damaged or missing freight door from reaching safety-related equipment.
j 410.11 Provide a drawing of the Catawba spent fuel pool storage racks.
Describe'
((9.1.2) the measures taken to prevent the placement of Catawba fuel assemblies into positions fitted with spacers to accommodate Oconee fuel assmLlies.
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Describe the consequences of an error in placing Oconee fuel into posi-tions not fitted with Oconee fuel spacers.
It is our position that such errors not result in a K f greater than 0.95 for the spent eff fuel storage arrangement.
410.12 Provide infonnation on the spent fuel pool water temperature following (9.1.3) the loss of one cooling train assuming the maximum heat load condition with Catawba fuel and with non-Catawba fuel.
Present an analysis of temperature vs. time for these two conditions.
Describe the dispositian of the heat load to all heat sinks including the spent fuel pool heat exchanger, the spent: fuel pool water purification system, any waste or drains, and the environment of the fuel building.
Discuss the effects of the heat and humidity load on the fuel building ventilation system.
s 410.13 Verify that the spent fuel pool liner is seismic Category I, or that it (9.1.3) is designed to remain in place and retain its leak tight integrity in a SSE.
It is our position that the liner not' fail in a manner which could result in mechanical damage to the spent fuel or result in an inability to maintain proper spent fuel pool cooling.
. 410.14 Commit to implement the interim actions of NUREG-0612 " Control of Heavy (9.14)
Loads at Nuclear Power Plants" prior to receipt of your operating license.
Provide an analysis of the effects of dropping heavy loads other than the speht fuel cask. The analysis should satisfy the evaluation criteria of NUREG-0612, Section 5.1, and consider the consequences of dropping the reactor vessel head and vessel internals during preparation for or completion of fuel handling.
In addition, the lower load block of both the containment building polar crane and the fuel building crane should be considered as a heavy load and an analysis of the consequences of their falling included in this analysis.
410.15 Provide a listing of all light loads (those of weight less than one fuel (9.1.4) assembly) carried over the open reactor vessel or the spent fuel pool including their kinetic energy on impact with spent fuel and discuss the consequences of dropping of these loads on stored fuel.
It is our position that dropping of these loads not result in release of radio-activity in excess of that assumed in the design basis fuel handling accident.
410.16 Figure 9.5.4-10 indicates that the nuclear service water yard piping (9.2.1) crosses the non-safety-related condenser circulating water system piping.
Describe the protective features provided for the nuclear service water piping at these crossover points against the effects resulting from failed condenser circulating water pipe in a SSE.
It is our position that the integrity of the nuclear service water system be maintained in a SSE.
410.17 Your component cooling water system has a single common supply line
-(9.2.2).
and a single common return line to the reactor building components cooled by this system. These lines contain electrically-operated containment isolation valves. Cooling water to all four reactor
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coolant pump -(RCP) seals and motor bearings which require continuous cooling during all modes of operation would be lost on failure of one of these valves. Loss of cooling could result in. motor bearing failure and locking of all pump rotors.
Indicate the consequences of the loss of motor bearing and seal thermal barrier cooling to all pumps and show that, without operator intervention, this event will not cause fuel damage or damage to the reactor coolant system pressure boundary with i
consequences greater than those assumed in the design basis RCP locked rntor accident.
If this can not be demonstrated, then it is our position that you verify by J
test that the RCPs can function satisfactorily for 30 minutes without com-ponent cooling water flow and provide redundant safety grade indication i
of the loss of component cooling water to the pumps in order to assure that the operator will have sufficient time to trip them. Al terna tively, you may provide redundant :omponent cooling water supply and return lines to/from the RCPs, or an automatic trip of the RCPs on loss of component cooling water flow.
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Indicate the effect of including the maximum worst case spent fuel
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(9.2.5) pool cooling load on your analysis of the heat rejection capabilities of the standby nuclear service water pond (SNSWP).
It is unclea.- if this load has been considered in ycur analysis of the 30-day heat 1
inputs to the SNSWP from the station auxiliary systems. Also provide a drawing showing the physical locations of the SNSWP intake and dis-charge structures, the nuclear service water pump house, and the inter-connecting pipe routing.
_9-410.19 Provide drawings that indicate the relationship of the auxiliary feedwater (9.2.6) condensate storage tank to the condensate storage system and to the auxiliary feedwat7r system.
FSAR Figures 9.2.6-3 and 10.4.9-1 concerning this tank and system d6 not agree.
410.20 Provide additional information conCErning the discussion in FSAR Section
( 9. 3.1 )
9.3.1.3 which describes the need forair operated valve operation in the event of a control room evacuation coincident with a station blackout in order to bring the station to a hot standby ccadition from the auxiliary shutdown panel.
Describe any special provisions made to assure the reliable delivery of instrument air to these valves under this coincident event.
Identify the valves requiring this capability.
41 0.21 Describe the means provided for assuring that instrument air quality is (9. 3.1 )
within the necessary limits to assure proper functioning of all air operated valves and instrumentation in safety related systems.
41 0.22 Provide an analysis to demonstrate that drainage of leakage water away (9.3.3) from safety-rela'r ' components or systems is adequate for worst case i
flooding resulting from pipe breaks or cracks in high-or moderate-energy piping or postulated failure in all non seismic Category I piping near these safety-related components or systems. The analysis must show that drainage by natural routes such as stairwelis of equipment hatches or by the non-seismic Category I drainage system under failed conditions I
is adequate to prevent the loss of function of safety-related components and systems.
Indicate how interconnected drains serving redundant safety-l related eouipment or cuoicles can be prevented from allowing leakage from one failed redundant train from backflowing and fl oding out the other l
train.
In those cases where separate drains are provided for redundant
. l safety-related components or systers, provide an analysis that demon-strates that the compartment and/or area drains serving these components l
or systems have been sized for maximum leakage flow conditions.
It is our position that unless drainage capability by natural or by failed non-seismic Category I drainage system! can be demonstrated that you provide the following for all areas housina redundant safety-related equipment:
1.
Leak detection sumps shall be eouipped with redundant safety grade alarms which unnunciate in the control room. Verify that if operator action is required on receipt of the alarm that flooding of redundant safety related equipment will not occur within 30 minutes; 0]!
2.
Provide separate watertight ruoms and independent drainage paths with leak detection sumps for each redundant safety-related system.
410.23 Describe your provisions for assuring a proper control room environment (9.4.1) during long-term pressurization of the control room area following i
isolation from the outside intakes due to high radiation, chlorine or smoke.
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410.24 Describe the safety classification code and quality assurance criterio
( 9. 4.1 )
for the chlorine, smoke and radiation detectors on the intake duct of the control room area ventilation system.
It is our position that since these redundant detectors are vital for quick isolation of the control room ventilation system, these detect 6rs should be seismic Category I as recommended by Regulatory Guide 1.29, Position C1k.
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410.25 Provide a description of the ventilation systems and/or room coolers (9.4.3) provided to maintain the environment for safety-related equipment in the auxiliary building and fuel building within allowable equipent qualification limits during the 16ss of all non-safety-related ventilation equipment. Assume a ?oss of offsite power and concurrent single failure in these HVAC systems and verify that this condition will not affect plant safety. Discuss each system individually. Describe the source of air, its expected maximum temperature, routing an'd exhaust system. Make clear how non-satety-related system failures affect the safety-relitted systems functions. Relate your description to specific drawiags you have made or will make available.
Also indicate the location of air intakes and exhaustson drawings and i
describe missile and tornado protection ror safety related equipment near these openings.
Provide drawings of the auxiliary shatdown panel room ventilation system.
410.26 Describe the ventilation system for the nuclear service water pump house
( 9. 4. 5 )
l and verify that a proper envi onment is maintained for safety related components within the pump house tssuming a loss of offsite power and any concurrent single active failure.
41037 Verify the operability of the power operated atmospheric relief valves (10. 3.1 )
It is our position that either a test of this function be performed during low-power testing of the plant and that plant safety not be affected assuming the valves are not opened l
for 30 minutes or verification that the relief valves can be opened l
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12-from the control room be provided including a description of the design providing this capability.
Describe the purpose of the block valve located between the main steam line and the power-operated relief valve.
Indicate the normal and failed position of this valve, how the opening of this valve can be achieved from the control room, and the reason why it is not locked open.
410.28 Provide drawings of the condenser circulating water system which will (10.4.5) enable the evaluation of the flooding consequences to safety relate) systems assuming a break in the system (such as would occur in a SSE).
Indicate the elevations of major components of the system including the cooling tower basin, isolation valves, pumps and main condenser.
Describe the consequences of breaks in the condenser circulating water lines at the connection to tne condenser and upstream and downstream of the isolation valves. We are concerned with the gravity draining of the i
cooling tower basin due to a failure of the condenser circulating water lines in the turbine building coupled with the inability to isolate the break from the cooling tower basin and the potential for this water 5
to reach safety-related equipment through turbine building penentrations i
or other openings in structures housing safety related equipment.
J Indicate the volumes of the turbine building basement, water volume available in the cooling tower bacin, and comon water elevation attained in the event of a non-isolable circulating water system failure, and the protection afforded safety related systems from flooding as a result c f this failure.
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i 4.J.29 Comit to perfonn a water hamer test in accordance with the recommendation (10.4.7) 4 of NUREG/CR-1606 "An Evaluation of Condensation-Induced Water Hammer in Preheat Steam Generators." This test shot.H be performed as follows:
"Run the plant at approximately 25% of the full power by using feedwater through the top-feed nozzle at the lowest feedwater temperature that the Standard Operating Procedure (50P) allows. Switch the feedwater at that temperature from the top-feed nozzle to the main feed nozzle by following the SOP. Observe and record the trar31ent that follows."
l 410.30 Describe your method of providing adequate auxiliary feedwater flow to (10.4. 9) the steam generators in the event of the loss of both normal and emer-gency AC power for a two hour period in accordance with the criteria of Branch Technical Position (ETP) ASB 10-1.
It is our position that this method should rely only upon safety-related systems and equipment.
Describe the failed position of electrically-operated valves in the suction lines to the auxiliary feedwater pumps, pump discharge lines to the steam generators, atmospheric dump valves and steam supply valves to the turbine driven auxiliary feedwater pump.
Describe the function of the emergency DC electrical supply in providing this assured supply of auxiliary feedwater.
41 0.31 Provide drawings that describe the standby shutdown system and its (10.4.9) function in providing an independent source of water for the auxiliary feedwater system. Indicate and describe the safety classification l
l of the water lines from the condenser circulating water system to the auxiliary feedwater system and assure that failure of these lines will not affect the safety function of the auxiliary feedwater system, l
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410.32 Describe the function of the main feedwater bypass feed and the main (10. 4. 9) feedwater tempering flow in re' i the auxiliary feedwater system.
Discuss the effects that failt.,
hese lines would'have on the safety function of the auxiliar, rater system and the procedures followed in operation of these lines during normal plant, start-up, shutdown, and emergency conditions.
410.33 Provide a response to our March 10, 1980 letter concerning your auxiliary (10.4. 9) feedwater system (AFS) design (TMI-2 Task Action Plan, NUREG-0737, Item II.E.1.1).
This response should include the following:
l.
A detailed point-by-point eview of your AFS design against Standard Review Plan Section 10.4.9 and Branch Technical Position ASB 10-1.
2.
A reliability evaluation similar to that performed for operating plants (refer to Enclosure 1 of the March 10, 1980, letter) and discussed in NUREG-0611.
3.
A point-by-point review of your AFS design, technical specifications and operatir.; procedures against the generic short-term and long-term re'uirements discussed in the March 10, 1980, letter.
4.
An evaluation of the design basis for the AFS flow requirements and verification that your AFS will' meet these requirements (refer to of the March 10, 1980, letter).
ENCLOSURE 2 4
I 311.0 SITING ANALYSIS BRANCH 311.1 Please provide us with a copy of the August 28, 1973 agreement between the Duke Power Company and the Concord Cemetery Association regarding tha ownership l
and control of the one acre of land in the exclusion area designated as the Concord Church cemetery.
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ENCLOSURE 3 RAP,'s Ql's CONCERNING CATAWBA ENVIRONMENTAL REPORT 470.5 ER rable 2.1.3-1 lists the location of the rearest milk cow, milk goat, garden, residence and site boundary.
Since these locations are based on a survey several years ago, confirm and/or update ER Table 2.1.3-1.
In a similar fashion, provide the location of the nearest meat animals eithin 0-8 Km for the 16 sectors.
ENCLOSURE 4 471.0 RADIOLOGICAL ASSESSMENT BRANCH 471.13 The information contained in FSAR Table 1.9-1, Item II.B.2, and FSAR Chapter 12.3.2 does not describe vital arease degrees of ' occupancy of these arease and anticipated doses to personnel occupying these areas, as outlined in NUREG-0737's Documentation Required Items (3), (4) and Clarification Items (3)(a),(b) sect.ons. Provide additional information as outlined in the referenced sections which (a) clearly identifies post-accident vital areas; (b) describes why certain areas were not selected as vital areas; (c) describe whether occupancy of the vital areas will be continuous or infrequenti and (d) provides the doses anticipated to be received for continuous occupancy or infrequent access in the vital areas identified (parti;utarty sampling areas).
(e) Access routes and alternate routes to vithi areas shculd also be provided.
(f) Also verify that the dose rate maps represent the maximum source term at T=0.
(g) Describe whether or nct additional shieldings remote operations, or other plant modifications are needed to meet GDC 19 criteria.
471.14 Specify the exact location of the containment high range radiation monitors in accordance with the NUREG-0737, Clarificationi Item (3) of Item II.B.2 (as committed in FSAR Table 1.9-1, p.10).
471.15 Provide a description of the in plant post accident radioiodine sampling and analysis system specifying the number and location of samplersi sample flushing metheds, and sample analysis equipment type and location.
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ETCLOSURE 5'
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RADIOLOGICAL ASSESSMENT BRANCH NON-TMI ISSUES 471.16 This information was previously requested by Q331.3 (submitted to DL on 5/14/79). Recently, potential access and shielding problems have been noted at other ice condensor type plants. Provide additional detailed information regarding spent fNel transfer tube shielding and access controls as outlined in Regulatory Guide 1.70 (Sections 12.3.1, and 12.3.2).
Provide diagrams and information which details how all portions of the Catawba 182 fuel transfer tube shielding and administrative and operational controls meet:the following branch positions:
Control of Access to Spent Fuel Transfer Areas Att accessible portions of the spent fuel transfer tube and/or canal must be shielded during fuel transfer. Use of removable shielding for this purpose is acceptable. This shielding shall be such that the resultant contact radiation levels shall be no greater than 100 rads per hour. ALL accessible portions of the spent fuel transfer tube shall be clearly marked with a sign stating that potentially lethat radiation fields are possible during fuel transfer. If remov.able shielding is used for the fuel transfer tubesi it must also be explicitly marked as above.
If other than permanent shielding is used, local audible and visible alarming radiation monitors must be installed to alert personnel if temporary fuel transfer tube shielding is removed during fuel transfer operations.
471.17 The individual selected for the position of Station Health Physicist does not appear to meet the educationi experience, or training criteria outlined in Regulatory Guide 1.8 or the alternative Duke positions outlined in Section l
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13.1.3 of the FSAR. In order for the staff to evaluate Duke's alterrate selection of a " qualified" individual in accordance with Section 12.5.1 of Regulatory Guide 1.70s additional information regarding the selection and approval criteria used by the System Health Physicist and Manager, Nuclear Production to appoint the Station Health Physicist at Catawba is needed.
Detailed information regarding the appointee's operatino reactor professional experience, to include outage experience, number and types of personnel supervised, academic cridits in health physics; abilities in engineering, sciencer and mathematics applicable to radiation protection; and formal professional / technical training and experience in applied radiation protection (e.g., details of Table 13.1.3-1, page 4 of FSAR), should be provided.
471.18 You should describe your plan to provide backup coverage in the event of the absence of the RPM and outline the qualifications of the individual who will act as the backup. The December 1979 revision of ANSI 3.1 specifies that the temporary replacement for an RPM should have a BS degree in science or engineering, 2 years experience in radiation protection, I year of which should be nuclear power plant experiencer 6 months of which should be on-site.
The statement in Section 13.1.3 of the FSAR does not provide adequate infor-mation.
471.19 Provide additional information regarding the sensitivity and function of portal mo6itors in the Catawba contamination control program as outlined in Regulatory Guide 1.70, Section 12.5.3.
The Health Physics Appraisal Program results
indicated that most portal monitors (e.g., GM T).7e) are not sensitive enough under normal operating conditions to detect significant levels of-contamination on personnel. Descriptions should provide sufficient deta'i to verify that portal monitors and related i scedures to be used at Catawba for contamination control can detect significant levels of personnel contamination (e.g.i equivalent to HP-210 type probes used in direct f risks in low background areas).
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ENCLOSURE 6 EEB Requests for Additional Information Catawba ijuclear Station Environmental Review 290.18 Provide the bases for the statement in Section 5.1.4 of the ER that
"...offsite noise will not be a problem."
Indicate the specific receptor areas considered, and the consideration given in the analysis to ambient and plant noise levels (indicate the levels for these noises that were used in the analysis).
290.19 Indicate the current status of the NPDES permit for the Catawba Nuclear Station.
Indicate those limitations which are being adjudicated and the applicable limits presently in effect and the alternate limits which are being requested.
290.20 Provide a copy of the chlorine minimization analysis which has been performed for the Catawba Nuclear Station by Duke Power Company.
290.21 Lake Wylie water quality data is provided in Tables 2.4.1-2 and 2.4.1-4 Indicate whether these data represent conditions over time at a particular lake location or conditions at several lake locations. Also indicate whether the data are surface measurements only, or are depth composites.
290.22 Indicate the thicknesses and permeabilities of the linings of the Conventional Waste Water Treatment System (CWWS) ponds.
290.23 Discuss the classification of the various wastes from the station (e.g.,
sanitary wastes, waste water discharged to the CWWS) under the Resource Conservation and Recovery Act (RCRA).
Indicate the testing, disposal and monitoring provisions currently planned or to be required to assure compliance wi th RCRA.
290.24 Indicate whether the operational phase sewage treatment system will be the same as the presently installed system or will be of another design.
If the system will be of another design, describe its design and operation.
291.4 ER-5.3-2 gives drift in mg/l.
Convert to kg/ha/yr and indicate location of highest drift deposition.
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