ML25233A058

Safety Analysis by the Division of Reactor Licensing U.S. Atomic Energy Commission in the Matter of Carolina Power and Light Company - H. B. Robinson, Unit No. 2

ML25233A058
Person / Time
Site:	Robinson
Issue date:	02/27/1967
From:	US Atomic Energy Commission (AEC)
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Download: ML25233A058 (1)

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DO NOT REMOVE SAFETY ANALYSIS BY THE DIVISION OF REACTOR LI CENSING U. S. ATOMIC ENERGY COMMISSION IN THE MATTER OF CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON UNIT NO. 2 DARLINGTON COUNTY, SOUTH CAROLI NA DOCKET NO. 50-261 February 27 1 1967

1.0 4.0 s.o TABLE OF CONTENTS INTRODUCTION SITE -

COMPARISON WITH RECENT PRESSURIZED WATER REACTORS (PWRs) 3.1 Novel Features 3o2 Plant Systems

3. 3 Plant Layout 3o4 Instrumentation and Control ENGINEERED SAFETY SYSTEMS 4ol Containment Design 4.2 Emergency Core Cooling 4.3 Containment Cooling Systems 4.4 Auxiliary Cooling System 4,5 Iodine Removal System ACCIDENT ANALYSIS 5.1 Mechanical Failures 5,2 Reactivity Insertions

5. 3 Maximum C~ible *Acci.dent 600 CONFORMANCE TO THE GENERAL DESIGN CRITERIA 7.0 RESEARCH AND DEVELOPMENT 8.0 TECHNICAL QUALIFICATIONS 9o0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 10,0 CONCLUSIONS Page 1

2 7

7 8

10 11 12 12 13 15 16 17 18 19 20 21 23 24 24 25 26

Appendix A Appendix B Appendix C Appendix D Appendix E Appendix F Appendix G Appendix H Appendix I ii ACRS Letter to Chairman, AEC, dated February 17, 1967 Report to the AEC Regulatory Staff, U. S. Weather Bureau Report to the AEC Regulatory Staff, U. 8. Geological Survey Report to the AEC Regulatory Staff, U. S. Coast and Geodetic Survey Report to the AEC Regulatory Staff, U. 8. Fish and Wildlife Service Report to the AEC Regulatory Staff, Nathan M. Newmark, Consulting Engineering Service Division of Reactor Licensing Staff, Report on the Structural Design of H. B. Robinson Unit No. 2 Containment Chronology - Review of the Carolina Power and Light Company Application Comparison of H. B. Robinson Characteristics with those of Turkey Point and other Recent PWRs.

1.0 Introduction The Carolina Power and Light Company submitted an application* dated July 12*

1966* to the Atomic Energy Commission for a license to construct and operate a 2094 MW(t) pressurized water reactor designated H.B. Robinson Unit No, 2. to be located in Darlington County, South Carolina.

Unit No. 2 will be situated adjacent to Unit No. 1 which is a conventional fossil fuel fired power plant.

This report presents our analysis of the safety of the proposed nuclear power unit and is based on the Preliminary Facility Description and Safety Analysis Report and seven amendments thereto (the "application").

The issues to be considered* and on which findings will be made by the Director of Regulation and by the Atomic Safety and Licensing Board before the construction permit is issued* are set forth in the Notice of Hearing issued by the Commission and published in the Federal Register on February 24

• 1967

• In our review* we have placed particular emphasis on the adequacy of the con-tainment design and the other engineered safety features which limit the consequenceE of credible accidents.

We have evaluated these features as well as the site*

maximum credible accident* and off-site safety considerations based on the maximum power rating of 2300 MW(t).

Evaluation of the thermal-hydraulic characteristics as they relate to safe operation of the reactor was performed for the initial design power rating of 2094 MW(t).

In the course of our review* we have met with the applicant and its contractors Ebasco Services and Westinghouse Electric Corporation* to discuss the design of the propased plant.

In addition* the Advi ory Committee on Reactor Safeguards (ACRS) has also met with us and with the applicant to discuss the application.

A chron-ology of aubmittale and meetings is presented in Appendix H, The review and evaluation of the proposed design 'and construction plans of the applicant at the construction permit stap.c of the proposed project is the first stage of the regulatory review which will continue throughout the lifetime of the facility.

Prior to issuing operating licenses for the facility, we will review thoroughly the final design to determine that all of the Commission's safety require-ments have been met.

The facility would then be operated only in accordance with the terms of the operating licenses and the Commission's regulations and under the continued scrutiny of the Commission's regulatory staff, Site -

The site of the proposed Ho B. Robinson Unit No. 2 consists of 5000 acres.

including Lake Robinson* a man-made lake about 7 miles in length on Black Creek.

Unit No. 1 is situated between Unit No, 2 and the intake structure on the lake, The site is about 5 miles WNW of Hartsville* South Carolina, and about 25 miles NW of

Florence, The region within 5 miles radius of the site is used primarily for agricultural pursuits.

The distance to the nearest site boundary is approximately 1400 feet, which is the exclusion distance as defined in 10 CFR 100 of the Commis-sion's regulations*

We have requested comments from several consultants concerning the site, and their reports are attached as appendices as listed below:

Consultants Subject Appendix

u. s. Weather Bureau Meteorology B

u. s. Geological Survey Geology & Hydrology C

u. s. Coast & Geodetic Survey Seismology D

u. s. Fish & Wildlife Service Environmental Effects E

and Monitoring 2,1 Population Within one-half mile of the site there are 25.homes 1 clustered generally southwest of the site. The nearest town of any significant size is Hartsville (Pop. 8,762) which is approximately 5 ¢.les to the southeast.

The nearest city is Florence 1 approximately 25 miles southeast 1 with a metropolitan area population estimated to be nearly 30 1000.

The total population within 5 miles is 10 1885 1 and this is expected to increase to approximately 13 1940 by 1986.

On the basis of these figures 1 we consider the low population zone distance to be 5 miles, and the popula-tion center distance 25 miles.

2.2 Meteorology The applicant has presented data on the meteorology of the site from Florence Municipal Airport and from Shaw Air Force Base in Sumter, 35 miles SSW of the site, Based on this information, the only unusual meteorological feature of the area is the frequency and persistence of calms* a eharacteristic of the SE part of the United States.

Although the atmospheric diffusion parameters for routine operation and for accident evaluations will be based on detailed data to be taken at the site prior to oper ation, for the present, the applicant has assumed the standard Pasquill Type F conditions with a 1 meter/second wind speed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for ~valua-tion of off-site consequences af accidents.

The diffusion predicted in this manner i* reasonable* conservative, and consistent with the data obtained at Florence Municipal Airport and Shaw Air Force Base.

The comments received from the u. s.

Weather Bureau confirm that the diffusion parameters used by the applicant are con-aervative, An analysis of extreme winds in the area indicate that there is a probability of a speed of 80 mph occurring once in 50 years, and 85 mph once in 100 years, Accordingly, the facility will be designed for an 85 mph hurricane.

There have been 20 tornados within 30 miles of. the site in the past 50 years, and the containment and auxiliary building (which houses *safety equipment) will be designed to withstand I

a 300 mph tornado and a concurrent atmospheric pressure drop of 3.0 psi, The auxiliary building which houses most of the safety equipment has 8-inch thick con-crete walls which are expected to preclude damage to internal equipment from missiles generated by tornados or hurricanes, We believe these figures are reason-ably characteristic of the conditions associated with a tornado and that the criteria are acceptable; 2,3 Geology The site is located adjacent to the Orangeburg Scarp, which is the boundary between the upper coastal plain depo;it (Middendorf formation) and the lower coastal plain deposits (Black Creek formation) in this area.

Borings at the site indicate there is approximately 30 feet of surface alluvium, with 430 feet of Middendorf formation below it, before reaching the Piedmont crystalline basement rock.

The Middendorf formation is made up of sands, silty and sandy clay, sandstone, and mudstone.

The alluvium portions of the Middendorf formation occurring near the surface have lenses of compressible material, and these have influenced the appli-cant to choose piles for the support of all major structures. The possible earth-quake amplification due to this alluvium is taken into account in assessing the intensity of the largest earthquake which could occur, 2,4 Hydrolog I

Because of the clay lenses and other features of the formations described above, there are a number of restrictions to ground water motion in this area.

This reaults in a condition in which the net ground or surface wuer movement (above the clay) in the vicinity of the site is toward the creek or Lake Robinson, There-fore, if any liquid radioactive effluents were accidentally released in the vicinity of the plant, it would find its way to the creek, or Lake Robinson, rather than to the ground water below the clayo Neither Black Creek nor the Pee Dee River (into which Black Creek flows) is used for public water supply downstream from the site since ample supply of water fr~m deep wells is available.

The applicant has estimated° that the maximum flow of B*lack Creek which has occurred since 1891 was in 1945, and was about 5100 cfs.

The calculated flow obtained by transposing the maximum rainfall observed anywhere within a large area of the site was 23,000 cf, and resulted from a rainfall of 21.4 inches in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

The calculated maximum flow obtained by taking Weather Bureau estimates of maximum probable precipitation and the Black Creek watershed area is 39,000 cf**

The Lake Robinson Dam and spillway are designed to pass a flow of 40,000 cfs at a lake level of 221,67 feet which is 8 feet lower than the height of the dam, and 3 feet lower than the site grade of 225 feeto Therefore, flooding of the ite appears highly improbable.

For the nuclear unit, it is planned to extend the discharge canal some 4 miles upstream along the lakeo The discharge flow in the canal with the two units oper-ating will be approximately 1000 cfso The concentration of fission products in the discharge canal will always conform to 10 CFR Part 20 limits. If the results of the radiological monitoring program of lake water and biota which the applicant intends to perform during plant operation indicate that significant reconcentration of fission products has occurred, the release rate could be lowered.

2

• 5 Seismology There are no identifiable faults in the vicinity of the site and no earthquake epicenter have been located within 15 miles.

There have been two rather intense earthquake* in South Carolina in hi.torical ti-* (Charle1ton in 1886 and Union County in 1913).

The inten1ity at the aite due to th*** u well as other nearby less intense earthquake* hu been 1tudied extensively by the applicant.

Baaed on the1e studies and the 1ite characteriatic** the applicant has selected as the design earthquake a horizontal ground acceleration of O,lg, They have also specified the maximum hypothetical horizontal ground acceleration for which there is to be "no loss of function of systems and equipment vital to safe shutdown" (Class I systems and equipment) as 0,2g, Vertical acceleration is assumed to be 2/3 of the hori-zontale The U, S, Coast & Geodetic Survey has evaluated the site characteristics and agrees that the maximum hypothetical earthquake that could occur at the site is Oo2g.

Therefore* we believe that the design of the systems and equipment vital to safe shutdown to withstand a 0,2g earthquake without loss of function is adequate.

Dynamic confined compression tests and triaxial compression tests have been performed on samples of the soil at the site, It has been found that with dynamic loading there is some reduction in soil strength and increased settling. This will be accounted for in the design and loading of the piles, The applicant has analyzed the seismic response of the earthen dam* its tainter gates 9 and the spillway and concludes that each component can withstand a 0.2g earthquake acceleration with a considerable margin of safety1 We have reviewed this analysis and agree that there is considerable margin to failure of the dam under a 0,2g earthquake.

2e6 Environmental Monitoring A study of environmental radiation levels will begin approximately 12 months prior to plant operation.

While the details have not as yet been formulated* the program will include periodic analysis of samples of crops* lake water from various locations* bottom sediments* fish* and air at various locations including Hartsville.

2,7 Conclusion We believe that all of the iJDPortant characteristics of the site have been adequately considered in the design of H.B. Robinson Unit No, 2.

r oO Copariaon with Recent Preasurized Water Reactor* (PWb)

Ho Bo Robinaon Unit Noo 2 is aimilar to other PWRa recently reviewed by the regulatory ataff: however* fpur feat urea which are novel to thia unit are deacribed belowo In the remaining portion* of thia section 8 feature* of the unit de1ign which we con1ider important to safety are di1cuaaed 0 lol Nov*l Peature1 Th* Bo Bo Robinson design incorporate* four features which are conaidered to be novel compared with other PWR design10 Theae ar*:

(1) a co11111on control room with Unit No. 1 (fossil fuel fired) and location of reactor and turbine-generator control* on a common board* (2) automatic load di~patch* (3) capability to with-atand a net load reduction without reactor scram* and (4) certain feature* of the preatreaaed concrete containment deaign (discussed in Appendix G) 0 We have reviewed the proposed layout for the common control room and have determined that the control boards for each unit are coq,letely independent and are located 30 feet apart in opposite corners of the control roomo The alarms or annunciator* are of a different tone for each board* and a aeparate control operator will be uaigned to each boardo We believe that thi* arrangement is acceptable and that the location of both boards in one room aa propoaed will not reault in con-

• -fuaion during ei,ther normal or abnormal operationo The Carolina Power and Light Company baa a coq,uterized dispatch center which 0 in accord~c* with a pre-aet program* request change in load from all power plants on the grido Theae requests are sent out aa abort impulses on the average of one every 4.8 econda by microwave tranamittera to a receiver at each planto H.B. Robinaon Unit Noo 2 will have a receiver and the neceary control circuitry to automatically signal for turbine throttle-valve operation and consequent changes in reactor powero The *Jste~ incorporates several design features to assure that the operator i,copbant o* lo.ad demands and that limita et by the operator are not exceeded.

The automatic load dispatch system will be designed so that it will in no way interfere with the safety and protective circuitry of the reactor.

We believe this system as presently conceived is acceptable.

Another unique feature of the H.B. Robinson design is the control system which will enable the reactor to remain critical following the loss of net electrical power being delivered to the grid system.

This is accomplished by control circuitry which closes the turbine admission valves relatively slowly without tripping the reactor or turbine.

Excess secondary steam is dumped to the atmosphere and condenser until the control rods have reduced powe~ in about 10 or 15 minutes to the auxil-iary load of about 5% rated core power.

This feature does not present any signifi-cant safety problems since no bypassing or disabling of any core safety instrumenta-tion is required.

The frequency of occurrence of this transient is expected to be very low.

Nevertheless. the release of potentially radioactive steam to the atmos-phere by this mechanism has been analyzed.

Assuming conservatively that 5% of the fuel rods in the core leak and a 0.15 gpm leak from the primary to the secondary exists, 170 steam dumps to atmosphere per year could be tolerated and remain within the 10 CFR Part 20 limits. If the leakage were greater* correspondingly fewer dumps would be allowed.

We believe that the doses from net load rejection are acceptable, and there is no safety problem involved with the plant having the described net load reduction capability.

3.2 Plant systems The thermal-hydraulic parameters* nuclear design characteristics and control and kinetic behavior of the H, B. Robinson plant are essentially identical to those

~f. the Turkey.Point,units and similar to those of Ginna and Indian Point II, A detailed compkriaon.of.the H. B, Robinson plant with the other PWRs is presented in Append!¥ I, The fabrication and quality control procedures for t~e primary system are in accordance with the applicable codes and are the same as for the other PWRs.

Critical areas of the primary system will be available for in-service inspection.

Primary *yatem components have been analyzed to be capable of withstanding, without deformation which would inhibit core cooling, the forces and thert11al transients which could accompany a loss-of-coolant accident and subsequent safety injection of cold water.

The containment isolation valve system meets the criteria applicable to PWRs 0 In essence, these criteria are that a double barrier shall be provided at all pene-trations or weld seams of the liner. The H.B. Robinson plant design also incorpo-rates (1) liner seam channels welded over all liner welds, (2) a penetration pressurization system, and (3) an isolation valve,seal water injection system (all similar to those provided on Indian Point II) to further reduce the leakage rate below the design value of O. 1% of the containment volume per day*

The containment design is such that it can be periodically leak rate teated at the design pressure of 42 psig.

We believe the over-all containment design including the isolation valve system is adequate.

Because the *ite is in a comparatively active seismic region, we have care-fully evaluated the seismic design criteria for all Class I *tructures and compo-nents.

Seismic loadings due to the 0.2g ground accelerations,are con*idered to be primary loads and are considered to be combined with normal and accident loads.

For all structural members of the containment 1 t~e stresses will remain below yield including combined primary and secondary stresses.

In Claes I components, such as piping 1 tanks, etc., the primary plus secondary stre*ses will remain below 120%

of the yield stress of the material.

We believe that th**e criteria will assure "no loss of function" for the Class I systems under maximum earthquake loadings 0 3.3 Plant Layout We have reviewed the over-all plant layout to determine that the equipment vital to safe shutdown (Class I equipment) is properly protected from credible hazards due to earthquake* missiles* environment* and fire.

All Class I systems and safety equipment external to the containment are (1) located inside Class I structures, (2) are otherwise suitably protected from missiles, natural phenomena (including tomados and hurricanes) and fire* or (3) are backed up with equipment that can mitigate the consequences of their failure or provide a similar protective function.

For example* the auxiliary building and the piping tunnel connecting this building to the containment vessel are of Class I design, all pipe trenches leading to water storage tanks outside the auxiliary building are located in Class I 'covered concrete trenches* and the detailed design of the auxiliary building will assure that reinforcing steel is placed to preclude large pieces of concrete from fallinga should significant yielding and cracking occur during an earthquake.

The layout of the primary system inside the containment has been reviewed to ascertain if missiles generated by primary system components might damage the con-

~

tainment liner or safeguards equipment.

The applicant's criterion is that protec-tion of the containment liner and safeguards equipment will be provided for all missiles which could credibly be generated by the primary system.

In general, the concrete shield wall and operating floor provide such protection. It is further stated as a criterion that the failure of a primary pump impeller* rotor, or fly-wheel will not generate a missile which could result in a loss of coolant from the primary or secondary system.

Since the turbine is oriented with its axis of rotation perpendicular to the containment structure, a missile* such as a part of a disc_, could impinge on the containment building or control room.

The applicant has presented a turbine missile analysis which is similar to that provided for the Turkey Point project.

The conclusion of this study is that if the most energetic disc were to leave the turbine casing. it would penetrate less than one foot into the reinforced concrete containment or the control room walls or ceiling.

The concrete in these struc-tures is at least two feet thick, Based on our review of this problem* we believe that a missile generated by turbine failure will not damage the primary system or prevent safe shutdown of the reactor.

The only ventilation system required to operate after an accident is one in the auxiliary building which exhausts from areas in the vicinity of the external recirculation system.

Effluent from these areas is routed through charcoal or particulate filters to the plant vent.

Ventilation is provided in the pipe tunnel adjacent to the containment building and in other potentially contaminated areas and assures a negative pressure in each critical area.

We believe that the auxiliary building ventilation system will prevent a significant increase in potential off-site doses due to leakage from the recirculation system following an accident.

In summary* we believe that the plant layout and protection of Class I equip-ment are acceptable from a safety standpoint.

3.4 Instrumentation and Control The design philosophy and design criteria for instrumentation and control systems are essentially identical to those applied to the Turkey Point units and other Westinghouse PWRs, All nuclear and process system parameters used for deriving a reactor* trip signal are monitored by independent and redundant channels of instrumentation.

These instrumentation channels drive relays. the contacts of which are connected through three independent logic chain* which in tum actuate two redundant trip breakers.

Containment isolation and safety injection signals are also initiated by redundant and independent instrument channels which will be designed to function under all postulated operating conditions, All instrument channels will be testable and fail-safe insofar as is practicable.

Fire protection in the control room and cable vaulting areas is provided* and remote operating stations are provided to enable operation of safety systems and shutdown equipment in the unlikely event that the control room is uninhabitable.

We believe the instrumentation and control systems proposed.for H, B, Robinson Unit No. 2 are acceptable.

4,0 Engineered Safety Systems Engineered safety systems are provided in the design of the H.B. Robinson plant to mitigate the consequences of credible accidents, A comparison and evalua-tion of these features follows, 4,1 Containment Design As stated in Section 3.1 of this analysis, the containment structure is one of the novel features of the proposed design.

Basically* the structure is a vertical right cylindrical concrete structure,

~he concrete is prestressed in the vertical direction and reinforced in the conventional manner circumferentially and in the hemispherical dome and base slab, The design concept is similar to that of Ginna except in details such as its fixed hinge at the base, the grouted steel bars used for prestressing* the hinged reactor sump and the pile foundation.

Various l oading combinations due to internal pressure* temperature* earthquake, hurricane and

~ornado' wind** and snow and ice will be considered in the de1ign using the load factor approach* similar to that used for Ginnaa Indian Point Ila and Turkey Point to establi1h the margin to yield of any structural member under credible loading combinations, The most severe requirement on the structure is that there be no yielding in any structural member under the combined loadings due to the maximum accident pressure of 42 psig and maximum potential earthquake of 0,2g acting simultaneously, We have thoroughly reviewed the proposed design of th~s,tructure and have determined that the design will provide adequate margin to assure the required leak-tight integrity of the vapor container under the simultaneous loadings due to internal pressure and earthquake, Our report concerning all aspects of the struc-tural design is presented in Appendix G, Dr. N, M, Newmark's report on the adequacy of the structure is presented in Appendix F.

To establish the design basis accident pressure of 42 psig* the applicant per-formed an evaluation of consequences of the maximum credibleaccident(loss of coolant) conservatively assuming that no core cooling ia available.

A detailed heat balance as a function of time was performed taking into account heat sources from stored energy* reactor core decay heat* metal-water reaction* and heat absorption in the containment structure* containment spray system* or fan-cooler system, These analyses indicated that the pressure will not exceed 40 psig even if only 50 percent of the containment cooling safeguards operate, We have reviewed the analyses and believe th~t they provide an acceptable method for establishing the containment design pressure.

Thus* we conclude that the design pressure of 42 psig is adequate.

4,2 Emergency Core Cooling The applicant has provided an accumulator system and high pressure and low pressure safety injection systems that are designed to mitigate the course of the thermal transient in the reactor core following a loss-of-coolant accident due to any size break in the primary piping and prevent the occurrence of a subsequent core meltdown.

In the event of small sized pipe ruptures. the primary system may not be rapidly depressurized below 600 psi where the accumulators could inject. In these cases* coolant injection from the high pressure system would be sufficient to protect the core.

The accumulator system proposed consists of three pressurized tanks of borated water 9 each connected through check valves directly to the three inlet pipes of the reactor cooling loops.

This system is identical to that proposed for the Turkey Point units.

Under accident conditions in which pipe failure causes rapid release of the reactor coolant and primary system pressure drop* two out of three of the accumulators will supply the water necessary to re-cover the reactor core to the hot spot and prevent immediate melting of the fuel cladding.

The system acts auto-matically with no outside action or instrument signal requiredo To provide core cooling following injection by the accumulators, a low pressure core deluge system is providedo Both the high and low pressure systems are independent and have separate headers to the primary piping, There are three 300-RJ>m pumps in parallel compris-ing the high pressure system* and two 3000-gpm pumps in parallel comprising the low pressure core deluge system.

Redundant parallel valves and all pumps receive a signal for operation on coincidence of pressurizer low-level and low-pressure signals.

Borated water is supplied from the 400.000 gallon refueling water storage

tank, When the contents of the tank have been expended 45 minutes or more after an accident* recirculation of the injected water from the containment sump through the residual heat exchangers and back to the core region is accomplished by the operator from the control room.

Valving to accomplish this function is redundant such that a single failure would not prevent recirculation.

During the recirculation phase* heat is rejected to the eomponent cooling system and is then rejected to the service water system to the lake.

One low pressure pump and one of the two residual heat exchangers is required for long term heat removal from the core following an accident.

This equipment is operable from either of the two diesel generator emergency power systems.

We have reviewed the ability of the emergency core cooling systems to arrest the core thermal transient and prevent fuel clad melting due to loss-of-coolant for pipe breaks varying in size from less than 3 inches in diameter to the equivalent of a double-ended break in the 29-inch primary loop pipe, A zero or negative moderator reactivity coefficient was assumed, Our review indicates that the emerg-ency core cooling systems can cool the core for this range of breaks and moderator reactivity coefficient with two accumulators and one high pressure and one low pressure pump operating, There is continuing effort to develop detailed information concerning the mechanical stability of the reactor components during a loss-of-coolant transient and to evaluate the influence of the moderator reactivity coeffi-cieDt on the loss-of-coolant accident.

We intend to continue our evaluation of the capability of the accumulators and core cooling equipment as more information be-comes available so that it can be assured that the proposed systems will perform as required, In view of the proposed design and the continuing design and analytical effort, we believe an emergency core cooling system can be developed which will provide adequate core cooling to satisfy the applicant's criteria that for any size break the core will remain essentially intact and coolable, 4,3 Containment Cooling Systems To control the containment pressure and temperature following a loss-of-coolant accident, two independent heat removal systems are provided, Each system acting alone at its rated capacity can prevent overpressurization, The two systems are the containment spray system and the fan-cooler system, and are similar to those provided in other PWRs, The containment spray system consists of two 880 gpm pumps, each connected to a separate spray header, The spray nozzles will deliver a spray of droplets with a mean diameter of 700 microns.

One pump and one spray header is designed to remove 72 x 106 BTU/hour of heat from the containment volume by delivering borated water from the refueling water storage tank to the spray nozzles, After recircula-tion commences, water can be recirculated to this system from the containment vessel sump via a residual heat exchanger.

We have evaluated the heat removal capability and have concluded that with this size spray droplet and the proposed flow rate, the design heat removal capability will be provided.

The pumps are operable from the diesel generators, We believe the design of the spray system is accept ab le o The fan-cooler system consists of four air recirculation units utilizing plate-type finned cooling coils to transfer heat to the service water system.

The units will be in normal operation and are designed to remove heat from the normal con-tainment environment, However, since there is relatively little experience with finned coolers in the steam-air mixture following an accident, a calculational model has been set up to check the performance under accident conditions, We have reviewed the parameters which affect heat transfer to be used in these calculations and believe that the appropriate factors are being accounted for.

Each of the four coolers is designed to remove 36 x 106 BTU/houro We believe that the type of fan-coolers proposed can be designed to remove the required amount of heat from the post accident containment atmosphere, 4,4 Auxiliary Cooling System The component* cooling system provides cooling for the two residual heat exchangers following an accident, Three auxiliary cooling system pumps and two heat exchangers reject heat to the service water system, Only one pump and one heat exchanger are required foll011ing the maximum credible accident to provide necesaary cooling, The service water system provides cooling water from Lake Robinson to the com-ponent cooling heat exchangers* the fan-coolers* and emergency diesels 0 There are four pumps supplying redundant Class I piping systems which are valved such that cooling water can be supplied to the redundant component if a break were to occur anywhere in the system.

One pump is sufficient to supply adequate cooling to accommodate accident loads.

Remote-operated valves and instrumentation for detect-ing leakage of fresh water inside containment are provided for each fan-cooler; thus* dilution of borated water inside the containment can be detected and the defective fan-cooler isolated.

The pumps in each of these systems can be operated from diesel power.

We believe the design of the service water system and component cooling system will assure a supply of cooling water to the engineered safety features which pro-tect the containment after an accident.

4,5 Io.dine Removal System A system which would remove radio-iodine from the containment atmosphere following the maximum credible accident (in which core melting is assumed to occur) is necessary to limit the radiation dose at the site boundary.

In the design of the proposed H.B. Robinson plant* the containment spray system performs this func-tion.

The discharge of each containment spray pump is routed so that sodium thio-sulfate from a single storage tank is forced out of the tank to the spray pump inlet header and is mixed with the borated spray water.

This system is similar to that provided on Indian Point II. The design as now conceived calls for a 5%-by-weight solution of sodium thiosulfate at the spray header which is expected to reduce iodine concentration in the containment by a factor of 8.8 per hour per spray pump o We have reviewed the ability of the sodium thiosulfate solution to absorb elemental* organic* or particulate forms of iodine when aprayed into the contain-ment atmosphere.

It was determined that little experiinental data concerning the effectiveness is availableo Thus* Westinghouse will perform a series of tests this year to demonstrate the removal efficiency.

We intend to review the details of this research and development work as they become available as part of our con-tinuing review of reactor technology.

The parameters which will be studied are as follows:

(1) absorption efficiencies of elemental iodine from various air-steam

mixtures, (2) effect of recirculation, (3) effect of boric acid in the solution 9 (4) nozzle flow rate and droplet size, (5) time and path length of particle

exposure, (6) heat removal capability*

(7) concentration and chemical form of the absorbed iodine* and (8) storage capability of the thiosulfate solution.

Removal efficiencies of organic and particulate forms of iodine will not be studiede If further experimentation on the character of fission products released from the core confirms that (as now believed) only a small fraction of th~ airborne iodine will be in an organic or particulate form following an accident* we believe that the parameters to be studied are adequate, We intend to continue our studies in this area to assure that all appropriate parameters are considered, The appli-cant has stated that if this research and development program does not verify the removal capabilities claimed for the spray system* there is ample room inside the containment for installation of charcoal filters, We have concluded* based on the limited experiments performed to date at ORNL that the proposed system is feasible and that by using the information to be obtained in the proposed Westinghouse experiments an acceptable system can be pro-

vided, 5o0 Accidept Analysis The applicant has described the consequences of various accidents resulting from assumed mechanical failures and reactivity insertions, We believe that all typea of credible accidents have been considered* and we are in general agreement with the consequences described.

We do not expect that accident consequences will be significantly different upon determination of the final thermal* hydraulif* and physics parameters of the core, However* the consequences of these accidents will be evaluated by the applicant when final design details are available* and will be reviewed by us prior to reactor operation, 5,1 Mechanical Failures Mechanical accidents such as loss-of-coolant flow* steam generator tube rupture 9 steam line break* failure of waste hold-up tanks* and fuel handling accidents have been analyzed by the applicant, The applicant's criteria for the systems affected by such accidents set limits on design and operating parameters such that signifi-cant fuel damage or fission product release will not result if the accidents were to occur, We conclude from our review of these accidents that the stated criteria in each case are reasonable and will provide assurance that these accidents will not result in hazard to the public.

The following two incidents are of particular importance to the proposed H, B, Robinson Unit, 5, 1,1 Loss of Off-Site Power To provide core cooling upon loss of off-site power* a steam driven emergency feedwater system and a motor driven feedwater system are provided.

Primary water would be circulated through the core region by natural convection* and the core decay heat would be transferred to the feedwater in the steam generators.

The capacity of each of the feedwater pumping systems is 600 gpm, Feedwater require-ments immediately following shutdown are 600 gpm reducing to a steady state require-ment for 100 gpm 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> later, Thus* ~he capacity of the pumps is adequate, Water is supplied at this rate* or greater* from the condensate storage tank* the service water system* or the two deep well pumps, There is sufficient water int*

condensate storage tank so that the reactor could remain at hot standby for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or at 350°F for 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />so Both diesels are started automatically on loss of off-site power, and one will provide sufficient power to operate the motor driven emergency feedwater pumpso If both diesels fail to operate* the steam driven pump is capable of providing sufficient feedwater.

Battery power will continue to supply the instrumentation and control systems necessary to operate vital systems.

We believe these systems are adequate to accomplish shutdown of the plant following a loss of off-site power.

5.1.2 Loss of the Dam Our analysis of the dam structure indicates that it will not fail under the maximum earthquake loading of 0.2g.

However, if the dam should fail for any reason and service water from the lake is unavailable* the plant could still be shut down safelyo In this case 1 the steam driven emergency feedwater system pumps provide feedwater.

Cooling water for the pump bearin~ coolers is provided by the source of feedwatero Long term cooling of the core requires only that 100 ~pm of water be available after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (when the condensate storage tank is emptied).

This amount of water could be provided by the deep-well pumps, Based on the foregoing 1 we believe the plant can be safely shut down in the event of failure of the damo 5.2 Reactivity Insertions The applicant has presented discussions of various reactivity accidents including the rod withdrawal accident 1 dropped control rod 1 rod ejection accident 1 boron dilution 1 and cold water injection.

The consequences of these accidents do not differ significantly from those analyzed for the Indian Point 2. Ginna, and Turkey Point applications.

The nuclear characteristics of the H, B. Robinson reactor will be similar to those of the Turkey Point units. and other PWRs, We expect the kinetic behavior of the reactor to be confirmed by operation of these units prior to operation of H.B. Robinson.

We are continuing to explore the sensitivity of the results of the reactivity accident studies to variations in the applicant's analytical methods as it pertains to the current class of Westin~house reactors as this information becomes available.

In the unlikely event that undamped spacial xenon o*cillations should occur in the H, B, Robinson core, the applicant believes that such instabilities could be detected and controlled, These oscillations would be detected using lon~ ion chambers located outside the core region.

Control would be accomplished by appro-priate motion of part-length control rods, In our opinion, the proposed scheme appears reasonable; however, we do not have sufficient information at this time in terms of operating experience to justify unqualified acceptance, We will continue to review this area as additional information becomes available, There are uncertainties concerning operation with a positive moderator tem-perature reactivity coefficient, The positive reactivity coefficient affects several areas of reactor operation and safety, However, the applicant has stated that the positive moderator reactivity effect appears to be most significant in the loss-of-coolant incident, Westinghouse is continuing to evaluate the effects of the positive coefficient and has stated that, if necessary, the magnitude of the coefficient will be reduced to a value which will provide for safe operation or safe shutdown (in case of an accident) of the facility, The mechanical design of the fuel elements is such that the addition of fixed neutron absorbers to reduce the coefficient can be easily accomplished, The decision regarding whether or not the fixed absorbers will be used will be made prior to the operating permit stage of review, We will continue to evaluate the progress of Westinghouse in this regard.

5,3 Maximum Credible Accident The maximum credible accident for H, B, Robinson has been shown to be the loss-of-coolant from the primary system due to a double-ended break of a reactor coolant loop pipe, We believe this accident represents the maximum potential for off-site consequences, The ability of the containment and the engineered safety features to limit the consequences of this accident have been discussed earlier in this analysis 0 In analyzing the hypothetical consequences of this accident in terms of off-site doses, it was assumed that full core meltdown resulted.

However, it must be emphasized that if the emergency core cooling systems operate as designed, the release of fission products from the core would be substantially reduced.

In addition, the provisions for containment isolation will also reduce the amount of leakage below 0,1% per day if they operate as designed, Thus, off-site doses might reasonably be expected to be lower than those stated below.

The guideline for site acceptance is the *Commission's site criteria* 10 CFR Part 100, This regulation relates potential radiation doses to the site character-istics, exclusion d~stance and low-population distance.

The criteria state that potential radiatiop doses at the exclusion area boundary during the first two hours following the accident should* not exceed 300 rem to the thyroid or 25 rem to the whole body, Also, these same doses should not be exceeded at the outer edge of the low population zone through the course of the accident, The maximum credible accident has been analyzed considering a com-plete meltdown of the reactor core at end-of-life conditions, The release of fission product inventory from the core is assumed to be 100% of the noble gases, 50% of the halogens, and 1% of the particulateso A 50% plate-out of the halogens was assumed thereby making 25% of the total halogens airborne in the containment.

A containment leakage rate of 0,1% of the contained volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a rate of 0,045% per day was assumed thereafter, Pasquill's Type F meteorological conditions (moderately stable) with a wind speed of one meter per second were used, The effect of a turbulent wake resulting from air flow around the containment building was also considered, The sodium thio-sulfate spray system was assumed to remove elemental iodine by a factor of a.a per houro Five percent of the iodine in the containment was conservatively assumed to be of an organic or other form of iodine which cannot be removed by the spray

system, In our judgment* all of the foregoing assumptions are conservative.

The result of the radiation dose calculations for a two-hour exposure at the 1400-foot exclusion area boundary was a potential 170 rem to the thyro~do The potential dose for the course of the accident (30 days) at the population distance of 5 miles was less than 25 rem to the thyroid, In each case, the potential whole body dose was less than 1 rem.

These doses demonstrate that with only one spray pump operating at less than the design radio-iodine removal rate of a factor of 8e8 per hour* the guidelines of 10 CFR Part 100 would be met, We expect that the experiments to be performed by Westinghouse on the iodine removal system will demon-strate that an adequate removal rate is achievable, On this basis* we believe that the off-site consequences of an MCA would safety of the general public.

6e0 Conformance to the General Design Criteria not endanger the health and In our review of~ B. Robinson Unit No. J we have considered the manner in which the proposed design meets the General Design Criteria for Nuclear Power Plant Construction Permits published for comment by the Commission on November 22, 1965 0 (Press Release No. H-252).

Although certain areas have been pointed out that require additional analysis and evaluation, such as reactivity accident~, xenon atability, moderator reactivity coefficient effects* iodine removal* and emergency core cooling, our review of the design and analysis presented by the applicant for the proposed unit leads us to conclude that adequate consideration has been given to the ~afety aspects of the unit to meet the requirements of the General Design

Criteria,

. eO Research and Development Although the engineering design criteria of the proposed unit have been described* the implementation of some of these criteria are to be achieved from the results of appropriate development programs and analysis.

The development programs are not limited to the H.B. Robinson plant but are being conducted for the current generation of Westinghouse pressurized water reactors.

These development areas are:

(1)

Development of final core design and final thermal-hydraulic and physics parameters.

(2)

Development of analytical methods for analysis of reactivity transients associated with control rod ejection and the positive moderator reactivity coefficient.

(3)

Development of the emergency core cooling system (including accumulators).

(4)

Development of systems for reactor control during xenon instabilities.

(5)

Development of a sodium thiosulfate spray system for iodine removal.

Our evaluation of the information submitted thus far supports our belief that acceptable design details can be evolved from the programs proposed.

At a later state of development* a description of the final design derived on the basis of these programs will be submitted to us for review, 8.0 Technical Qualifications The applicant will operate the reactor when completed, The company employees who will be associated with operation of the unit will participate in a training program to be conducted by Westinghouse, This program will include courses and comprehensive on-site training that will lead to qualifying key personnel for Senior Reactor Operator Licenses and Reactor Operator Licenses, A five-month tour of duty at an operating pressurized water reactor plant for key operating personnel is planned, Westinghouse Electric Corporation is responsible for the over-all design and construction of Ho Bo Robinson Unit No, 2 on a "turnkey basis." Ebasco Services*

Inc ** has contracted with Westinghouse to provide the architectural-engineering design of the plant.

Westinghouse has been directly associated with the design of many nuclear facilities* such as San Onofre* Turkey Point Units 3 and 4* Connecticut Yankee 1 and otherso Each of these facilities is comparable to that proposed for Ho B. Robinson Unit Noo 2o All phases of the design and construction work pro-vided by Ebasco Services will be reviewed and approved by Westinghouse.

Ebasco is one of the largest architect-engineer firms in the country and has had extensive experience in the design and construction of large steam supply systems including nuclear units.

Based on these considerations as well as our evaluation of the qualifications of the responsible personnel* we have concluded that there is reasonable assurance that the applicant and its principal contractors collectively are technically qualifie~ to design and construct the proposed H, Bo Robinson unit, 9.0 Report of the Advisory Committee on Reactor Safeguards During its January and February 1967 meetings* the Advisory Committee on Reactor Safeguards (ACRS) met with representatives of the applicant to review the proposed H.B. Robinson Unit No, 2 0 Previous to this* a Subcommittee of the ACRS had met with the applicant on December 13* 1966* and February 1. 1967.

A copy of the ACRS letter to the Commission concerning the Carolina Power and Light Company application for construction permit for H. B, Robinson Unit No. 2 is attached as Appendix A, The ACRS in this letter of February 17* 1967* made several comments and recommendations concerning the design of the proposed unit which have been discussed in the body of this analysis, We have considered each of these matters and inten -

to implement these recommendations, In addition* they have commented on the importance of steam line isolation valves in preventing the escape of radio-activity and on the quantity of diesel fuel to be stored on site, In regard to the formerD leakage past these valves could be included in the containment leak rate testing program to be developed as a part of the facility technical specifications, In regard to the diesel fuel storage 1 we understand that the applicant will store sufficient fuel for one week's operation of the diesels at the load required for safety equipment, The letter then concluded,

"*,

• the proposed reactor can be built at the Ho B, Robinson site with reasonable assurance that it can be operated without undue risk to the health and safety of the public,"

10

• 0 Conclusions Based on the proposed design of H, B, Robinson Unit No, 2 1 on the criteria 0 principles and design arrangements for systems and components thus far described, which include all of the important safety items 0 on the calculated potential conse-quences of routine and accidental release of radioactive materials to the environs 1 on the scope of the development program which will be conducted* and on the techni-cal competence of the applicant and the principal contractors 1 we have concluded that, in accordance with the provisions of paragraph 50,35(a) 1 10 CFR 50 and 2,104(b)* 10 CFR 2:

1, The applicant has described the proposed design of the facility 1 includinR the principal architectural and engineering criteria for the design 1 and has identified the major features or components for the protection of the health and safety of the public; 2,

Such further technical or design information as may be required to com-plete the safety analysis and which can reasonably be left for later con*

sideration 1 will be supplied in the final safety analysis report; 3,

Safety features or components** which require research and development have been described by the applicant and the applicant has identified, and there will be conducted* a research and development program reasonably designed to resolve any safety questions associated with such features or components; 4,

On the basis of the foregoing* there is reasonable assurance that (i) such safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of construction of the pro-posed facility* and (ii) taking into consideration the site criteria con-tained in 10 CFR Part 100* the proposed facility can be constructed and operated' at the proposed location without undue risk to the health and safety of the public;

s.

The applicant is technically qualified to design and construct the pro-posed facility; and 6 0 The issuance of a permit for the construction of the facility will not be inimical to the common defense and security or to the health and safety of the public,

Appendix A ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, O.C, 20545 Honorable Glenn T. Seaborg Chairman

u. S. Atomic Energy Commiasion Washington, D. c.

FEB 1 7 1967

Subject:

RE'PORT ON H.B. ROBINSON UNlt NO. 2 Dear Dr. Seaborgi At its eighty-first meeting, on January 12*14, and its eighty-second meeting, on February 8-11, 1967, the Advisory Com:nittee on Reactor Safeguards completed its review of the application of the Carolina Power and Light Company to construct H.B. Robinson Unit No. 2 near Hartsville, South Carolina.

An ACRS Subc0Im1ittee met to review this project on December 13, 1966 at Hartsville, S,C., and on February l, 1967 in Washington, D. C.

During its review, the Committee had the benefit of discussions with representatives of the applicant, the Westinghouse Electric Corporation, Ebasco Services, Inc., consultants to these groups, and the AEC Regulatory Staff. The Coimnittee al&o had the benefit of the documents listed.

The unit includes a pressurized water r~ctor to be operated at 2094 MWt. It will be constructed at the H. B. Robinson Station adjacent to Unit No. l, an existing coalfired plant. The H.B. Robinson Station is in Darlington County, approximately five miles from Hartsville and 30 miles from Florence, South Carolin.a. The plant ia located on the shore of Lake Robinson just above the dam that impounds the water of Black Creek.

The containment is a cylindrical eteel*lined concrete structure with a spherical dome and a flat base-slab. The design will permit pressuri*

zation of the containment for test purpoaes aa may be required through-out the life of the plant. The cylindrical wall is prestressed vertically and reinforced circumferentially; the dome and base are reinforced. The tendons will be grouted in place and will not be accessible for surveillance.

The applicant plans to prepare samples of similarly prestreased and grouted tendons, and to expose them to the same general environmental conditions aa those experienced by the containment tendons. Samples will be available for investigation, aa needed, throughout the 11fe of the plant.

I

Honorable Glenn T. Seaborg

• 2 -

FEB 1 7 1967 The site is in a region of moderate seiGmic activity and the plant is being desinned accordingly. In the event that an earthquake produceo significant ground acceleration at this site in the future. an on-site measurement of the shock intensity should provide data valuable in assessing the poasibility of hidden structural damage to vital portions of the facility. A strong*motion accelerograph will be installed.*

The Committee notes that the entire primary system will be inspected to the requirements of the ASME Boiler and Pressure Ve&sel Code.Section III, Class A.

Tho emergency core cooling systems consist of a high head safety injection system. a low head residual heat removal syste.n, and an accumula*

tor injection system. The applicant states that. for all sizes of pipe ruptures of the primary system. a conservative evaluation of the function*

ing of engineered safeguards indicates that there will be no clad melting and leos than one per cent of clad-water reaction. The emergency contain*

ment cooling systems consist of a spray system and a circulating air cool*

ing system.

The water injected into the core or containment is borated; in addition, sodium thiosulfate is injected into the containment spray to reduce the iodine content of tho atmosphere in the unlikaly event of a primary system rupture. Tests are planned to establish the performance and reliability of the sodium thiosulfate injection sub-system. The Com~

mittee recoI:llllends that the AEC Regulatory Staff review details of the test data and the design of the emergency systems as they becane available.

The Committee notos that under certain highly improhable but credible acci*

dent conditions the isolation valves in the steam lines may be an important factor in preventing escope of radioactivity. The CoIImittce is of the opinion 'that a special effort should be made to assure that the6e valves.

and other valves that must be depended on for containment, are effectively tight under accident conditions.

Calculations by the applicant show that th& reactor has a positive modcra*

tor coefficient at some period of the core life. The applicant is con*

tinuing his analysis of all consequences of the positive coefficient and, if necessary, will adjust the core composition to assure aafety. The Comnittee recommends that the Regulatory Staff follow the applicant's studies and conclusions in this respect.

The Committee believes the applicant should store sufficient diesel fuel to permit operation of the emergency diesels aa required by accident condi*

tions for a minimum of ona week.

iwTbe Comnittee believes that the installation of a strong-motion accelero*

graph may ba appropriate for most large power reactors. including those located in aones of relative,eiamic quiesenca.

I

Honorable Glenn T. Seaborg FEB 1 7 1967 The Advisory Committee on Reactor Safeguards believes that the itema mentioned can be resolved during construction and that the proposed reactor can be built at the H. a. Robinson aite with reasonable as*

surance that it can ba operated without undue rbk to the health and aafety of the public.

References:

Sincerely yours.

O?IGii.iAL SIG!~::ill BY N. J. PALLADINO N. J. Palladino Chairman

l. Shaw, Pittman, Potts, Trowbridge & Madden letter dated July 25, 1966 to AEC Division of Reactor Licensing transmitting "Preliminary Facility Desc,ription and Safety Analysis Report", Volumes 1*3.

~,* 2.

Shaw, Pittman, Potts, Trowbridge & Madden letter dated October 10, 1966 to AEC Division of Reactor Licensing transmitting Amendment No. l.

3. "First Supplement to Preliminary Facility Description and Safety Analysia Report" (Amendment No. 2), dated November 28, 1966.

4.

"Second Supplement to Preliminary Facility Description and Safety Analysis Report" (Amendment No, 3), dated December l, 1966.

5, "Third Supplement to Preliminary Facility Description and Safety Analysis Report" (Amendment No, 4), dated December l, 1966.

6.,,.Fourth Supplement to Preliminary Facility Description and Safety.Analysis Report" (Amendment No, 5), dated December 1, 1966,

7. Amendment No. 6, "Fifth Supplement to Preliminary Facility l>escription and Safety.Analyaia Report", dated January 27, 1967.

8. Amendment No. 1. "Sixth Supplement to Preliminary Facility Description and SafQty Analyaia R.epcrt"

• dated February 6, 1967.

APPENDIX o Comnents pn, Carolina Power & Light Company Preliminary Facility Description and Safety Analysis Report Dated July 25, 1966 Prepared by Environmental Meteorology Branch Institute for Atmospheric Sciences October 7, 1966 Relatively speaking, the southeastern portion of the United States and South Carolina i n particular has one of the poorest dilution climates in the country based on low-level inversion frequency (451.) and nighttime winds below 7 miles per hour (701.) as shown by Hosler [l]o However, these poor dilution conditions are largely confined to the nighttime hours and it would be anticipated that the daytime dilution out to distances of the site boundary would be average.

Both tornadoes and hurricanes have been reported within 30 miles of the site.

Although the occurrence of a tornado directly over the site cannot be discounted, the probability is extremely small.

Hurricanes, which have a path width of several hundred miles, as opposed to several hundred feet for tornadoes, have a higher probability of occurring over the site. However, hurricanes usually decrease in intensity after traveling over land and it would be expected that the probability of occurrence of hurricane winds above 75 miles per hour at the site would be extremely small.

The assumption of moderately stable atmospheric diffusion conditions (Pasquill Type F) and a 1 m/sec wind for a 24-hour period with no credit taken for variation of wind direction during this period r!~ults !n a conservative concentration value at the site boundary of 3xl0 Ci/m per Ci/sec released.

If, as is suggested by Slade (2J 9 credit ls taken for the extended release time, concentrations are reduced by a factor of 2.

In addition, a reduction factor of about 5 and 2 at 100 m and 850 m downwind, respectively, is suggested by lslltzer [3) to account for building turbulence at the source.

It ls our conclusion that the applicant's dilution factor of 7.6 x 10-4 sec/m3 is reasonably conservative for a total release time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Since the highest seasonal frequency of a wind from a particular 22¥> sector is 161., the assumption of a 301 frequency in such a sector for a 30 day period ls reasonably conservative for dose computations of the long-term release.

In summary, di ffusion conditions in the general area of the slte are expected to be somewhat poorer on the average as compared to other locations in the United States.

A check on the applicant's relative diffusion factors for 24-hour and 30-day periods show them to be conservative.

References

[1] Hosler, C.R., 1961: Low-level inversion frequenc, in the contiguous U. s., Monthly Weather Review, 89, pp.319-339.

[2] Slade,.D.

H. 1966:

Estimates of dispersion from pollutant releases of a few second to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration, ESSA Tech. Note 39-ARL-3.

Islitzer, N. F., 1965:

Aerodynamic effects of large reactor complexes upon atmospheric turbulence and diffusion, 100*12041, 15 pp.

INTRODUCTION I

APPENDIX C REVIEW OF THE GEOLOGY AND HYDROLOGY SECTIONS OF THE CAROLINA POWER AND LIGHT COMPANY H1 ' B, ROBINSON UNIT N01 2 (AEC DOCKET N01 50-261)

The site is in Darlington County* South Carolina* on the shore of Lake Robinson*

an impoundment of Black Creek about 37 miles upatream from its confluence with the Pee Dee River, Hydrologic questions at this site concern primarily the recycling of the flow of Black Creek* whose average annual flow is less than 20 percent of the coolant flow needs of the nuclear power plant alonJ, GEOLOGY The analysis of the geology of H, B, Robinson Unit No, 2 site presented in Atomic Energy Commission Docket No, 50-261 was reviewed and compared with the avail-able geological literature, Although it may be anticipated that earthquakes within the general *r*gion will continue to occur with approximately the same frequency and with approximately the same intensity with which they have been recorded during the past 100 years. there are no identifiable geologic structures which could be expected to localize earth-quake* in the immediate vicinity of the site, The fact that foundation support of the plant is to be on saturated* relatively incompetent* Coastal Plain sediments rather than on competent bedrock auggest the possibility of some degree of seismic amplification, Detailed correlation of laboratory test data with tratigraphic unit* as determined from individual bore hole* will be reqµired as a neces ary basis for fiBal computation of the support capability and proper engineerini design of the pile foundation for the plant, HYDROLOGY Water Stages The Safety Analysis Report shows a calculated flood of 39.000 cfs (cubic feet per second) based on probable maximum precipitation for the area.

The unit hydrograph on which the calculation is based is not shown; therefore* the computation cannot be checked* but the figure appears to be reasonably conservative in relation to known floods in the region, The spillways at Lake Robinson dam were desi~ed to pass a flood of 40.000 cfs at a stage of 221,67 feet* or about 3.3 feet below the grade elevation of the plant, Failure of the tainter gates which operate the spillways could lead to higher stages*

but it appears that any potential failures could be corrected during the approach of a flood prior to over-topping of the grade of the plant, Low stage of Black Creek at the site prior to construction of Lake Robinson dam was between the 160 and 170 feet contours as shown in Fig. 2-7 of the Safety Analysis

Report, Intake for the steam condenser is at about 210 feet, Re-cycling of Water The flow of Black Creek is small in comparison to the condenser cooling water requirements of a large electric power plant.and a dam creating Lake Robinson was constructed to permit recirculation of stored water.

The capacity of the lake is given as 1,35 billion cubic feet representing approximately a 10-day supply for the nuclear plant at 1.100 cfs, Average flow of Black Creek is conservatively estimated at 169 cfs* and minimum monthly flow at 21 cfs.

Records of nearby stream flow indicate that average annual flow is quite variable, For instance* during the drought in the southeastern United States that lasted from 1950 to 1957* the average annual flow of the Lynches River at Effingham was less than 60 percent of the long term mean during 1951* 1955* 1956* and 1957, Seasonal minimum flows occur usually in the period 0 July to October, The average residence time of water in the lake has been calculated to be 37 days during a wet period 1 January through March 1964 1 and 70 days during a somewhat drier period from October through December 1963 1 when flow in Black Creek averaged 181 cfs.

During dry summers* average residence time would increase to over 100 days, as it did during the period July through October 1963* and since the lake holds a 10-day supply of cooling water for the reactor* water would then be recycled on the average of 10 times or more, The Safety Analysis Report discusses accumulation of radionuclides in the lake as a result of recycling and derives a formula for a maximum release rate of radio-nuclides for a given flow rate of Black Creek.

The formula is mathematically correcto Figures of maximum permissible releases per day of selected radionuclides are given in Table 2-6.

Although these maximum permissible releases are intended to take dry periods into account 1 the figure given for cesium-137 could lead to concentrations higher than the limits set in 10 CFR 20 during extended drought periods. It may be necessary to reduce releases to take drought conditions into considerationo Sorption activities on particulates is not considered in the safety analysis report.

Radionuclides sorbed by particulates would probably settle out and accumulate in the lakee The rate of sorption and ac.cum.ulati.on would depend on various factors*

such as the type and concentration of suspended sediments in the water* the composi-tion of the lake bed, and of the discharge canal bed.

The significance of the potential accumulation of the sorbed radionuclide& should be assessed and taken into cc:msideration in setting release limits.

The recycling of water.. would also lead to a rise in water temperatures in Lake Robinson and would affect the inlet temperatures of the power plant's condenser cool-ing system.

An estimate of this rise under severe conditiona, such as an extended summer drought 1 should be made so that this factor can be taken into cori.aideration in the design of the plant's condenser cooling system, Black Creek is not used for public water supply downstream from the site.

Five miles downstream, the creek is again ponded hy a d1tm in Hartsville where the water is used industrially and m:1y Hlso h:-lve recrc.1tlonal uses,

APPENDIX D RER)RT ON THE SEISMICITY OF THE LAKE ROBINSON (HARTSVILLE), SOUTH CAROLINA AREA At the request of the Division of Reactor Licensing of the Atomic Energy Commission, the Seismology Division of the Coast and Geodetic Survey has eval-uated the seismicity of the are~ around the proposed reactor site near Lake Robinson (Hartsville), South 9arolina, and has reviewed the similar analysis made by the applicant in the "Carolina Power and Light Company, Preliminary Facility Description and Safety Analysis Report." The applicant's seismicity report is exhaustive and includes a complete listing of the earthquakes both distant or nearpy, which may have affected the proposed site. The attached list is a summary of those earthq~es which were reported by the applicant and which the Coast and Geodetic Survey believes are most significant for site evaluation.

In reviewing the applicant's seismicity report there are some seismicity statements that should be noted.

The Charleston earthquake of 1886 was reported by the applicant as intensity IX, but according to the well known U.S. Geological Survey report by M. L. Fuller this earthquake was rated intensity X, RF and MM.

The site is rated intensity VII MM in the USGS Report and not Vas stated in the applicant's report, pages 2-61. According to the seismic zone map the site is classified Zone 1 which includes earthquakes of minor damage (VII MM).

Consideration should also be given to the evaluation by Professor Charles F. Richter in bis Regionalization Map which shows a seismic potential of intensity VIII for the site.

Our estimate, based on the seismic history of the site, the adjacent seismic areas near Summerville and to the west of the site, and the geology of the site, is that during the lifetime of the facility, we believe that an MM intensity VII earthquake with accelerations of 0.2g (on dense underlying stratum), might occur and should be considered as the maximum potential earthquake.

U. s. Coast and Geodetic Survey Rockville, Maryland 20852 February 3, 1967

Date 1886 Aug 31 Oct 22 Nov 5 1912 Jun 12 1913 Jan 1 1914 Sep 22 1945 Jul 26 1959 Oct 26 Int. at site (Iv,v)

III*

III Lake Robinson (Hartsville), s. C.

Carolina Power and Light Company.

Felt Area Dist.

EEicentral Res;ion S9.. Miles Miles Charleston, s. c.

2,000,000 120 II 30,000 120 II 30,000 120 Summerville, S. C.

35,000 120 Union Co., S. C.

43,000 90 South Carolina 30,000 100 Murray Lake, s. c.

25,000 70 Near McBee, S. C.

4,8oo 15

-Estimated from max. int. and distance.

-Estimated as within total felt area listed.

-Estimated from nearby towns.

Max.

Int. at Int.

Site X

VII VI III*

VII III*

VII (IV)

VII (V)

V III*

VI III VI VI

APPENDIX E UNITED STATES DEPARTMENT OF THE INTERIOR FISH AND WILDLIFE SERVICE WASHINGTON 25, D. C.

Hr. Harold L. Price Director of Regulations U. s. Atomic Energy Commission Washington, D. C.

20545

Dear Mr. Price:

FEB 1 1 1987 IN REPI.V REFER TO:

This is in reply to your letter of August 3, 1966, requesting our comments on the application of the Carolina Power and Light Company for a construc-tion permit and facility license for its proposed H.B. Robinson Unit No. 2 nuclear power project, Lake Robinson, Darlington County, South Carolina, Docket No. 50-261.

The project would be located adjacent to H.B. Robinson Unit No. 1, an existing coal-fired powecylant, on the southwest shore of Lake Robinson about 4.5 miles northwest of Hartsville, South Carolina. A pressurized '

water reactor, designed for an initial output of 2,094 thermal megawatts :

and a net electrical output of 700 megawatts, would be used as a power source.

A radioactive waste disposal system and other facilities required for a complete and operable nuclear powecylant would be provided. Solid wastes would be prepared for off-site disposal; gaseous wastes would be releaned through the ventilation stack; and liquid wastes would be released to the

' condenser water discharge canal.

Condenser cooling water would be con-.

veyed to the plant through a conduit approximately 12 feet in diameter and discharged to the lake through a 4.3 mile canal.

Condenser cooling water

• would be circulated at a rate of about 482,100 gpm (1,074 cfs) through the nuclear plant and 87,000 gpm (194 cfs) through the coal-fired plant.

The applicant plans to initiate a radiological survey of the Lake Robinson environment about 12 months prior to plant operation. Samples of air, rain, lake water, and food materials would be analyzed for radio-activity.

T'ae nature and extent of the post-operational survey would be determined on the basis of results from the pre-operational studY.

Lake Robinson was constructed on Black Creek, a tributary of the Pee Dee River, by Carolina Power and Light Company to provide cooling watern for its thermoelectric generating facilities. The lake is about 4,000 feet wide and 7.5 miles long. At the time of construction, the company received a permit from the South Carolina Water Pollution Contr9l Authority requiring operation of the lake in such a manner that the released flow will be at least 1.47 times the flow recorded at the USGS gage, McBee, South Carolina, about 10 miles upstream from the dam.

In addition, the

maintenance of Class A (swamp) water quality standards, which are suitable for fish propagation, was required.

Two Howell-Bunger *aeration valves are located in the drun for use during periods of low flow and/or high water temperatures to provide dissolved oxygen and temperature control in the downstream area.

Fish species present in Lake Robinson and Black Creek include largemouth bass, bluecill, redbreast sunfish, other sunfishes, carp, catfishes, and suckers.

Sport fishing is moderate.

There is no significant commercial fishery in the reservoir or stream.

The application indicates that the release of radioactive wastes would not exceed maximum permissible limits prescribed in Title 10, Part 20, of the Code of Federal Regulations. Although these limits refer to maximum levels of radioactivity that can occur in drinking water for man without resulting in any known harmf'ul effects, operation within the limits may not always guarantee that fish and wildlife will be protected from adverse effects. If the concentration in the receiving water lrere the only con-sideration, maxim.um pennissible limits would be adequate criteria for determining the safe rate of discharge. However, radioisotopes of ma~r elements are concentrated and stored by organisms that require these elements ~or their normal metabolic activities.

Some organisms concen-trate and store radioisotopes of elements not normally required but which are chemically and physiologically similar to elements essential for metabolisu.

In both cases, the radionuclides are transferred from one organism to another through various levels of the food chain just as are the nonradioactive elements.

These transfers may result in further concentration of radionuclides and a wide dispersion from t he project area particularly by migratory fish, mammals, and birds.

In viei1 of the above, we believe that the pre-and post-operational radiological surveys planned by the applicant should include studies of the effects of radionuclides on selected organisms which require the waste elements or similar elements for their metabolic activities. These surveys should be planned in cooperation with the Fish and Wildlife Service and the South Carolina Wildlife Resources Department.

If the post-operational surveys establish that the release of radioactive effluent at levels permitted under Title 10, Part 20, Code of Federal Regulations, results in harmful concentrations of radioactivity in fish and wildlife, the data from the radiological surveys should serve as a guide to reduce the discharge of radioactivity to acceptable levels.

In view of the importance of the sport fishery of Lake Robinson and Black Creek, it is imperative that every possible effort be made to protect these valuable resources from radioactive contamination. Therefore, it is recommended that the Carolina Power and Light Company be required to:

2

One problem we foresee is the possible effects of increased water temper-ature on aquatic organisms. Additional thermal input to the reservoir would result from installation and operation of the proposed nuclear power plant.

Company engineers and consultants have made reservoir heat studies relating tocperation of the proposed nuclear plant. These studies indicate that the maximum temperature of discharged cooling water would be about 115°F. when the plant is generating at full capacity, stream flow is low, and surface water and air temperature are extremely high.

According to company estimates the area of surface water eX].Jected to exceed l00°F. will be less t han 1,000 acres under these extreme conditions.

The frequency with which these adverse conditions would occur is expected to be about 1 in 50 years. Under average summer weather conditions and plant operation, the area of surface water expected to exceed l00°F.

would be relatively limited.

The average summer reservoir surface temper-ature occuring 9 out of 10 years is expected to be about 92°F.

Summer stream temperatures 1,000 feet downstream from the dam would be increased by an average of 5°F., to 87°F., with a maximum increase to 950F.

Although large volumes of heated water discharged into an aquatic environ-ment may not be sufficient to cause mortality among the organisms present, subtle biological changes could occur causing long term changes in the envirorunent.

To mensure biological changes in aquatic orBanisms and long term changes in the environment, ecological surveys should be carried out prior to and foll0'\\*1ing plant operation so that comparative data will be available for analysis.

These surveys should be planned in cooperation with the Fish and Wildlife Service and the South Carolina Wildlife Resources Department. If the ecological surveys establish that the heated water discharged into Lake Robinson results in changes in the environment of the lake or Black Creek that are significantly detrimental to fish and wildlife, as determined by t he Fish and Wildlife Service or the South Carolina Wildlife Resources Department, corrective measure to reduce the temperature of the effluent should be taken.

In view of the Administration's policy to maintain, protect, and improve the quality of our environment and most particularly the water and air '

media, we request that the Commission urge the Carolina Power and Light Company to:

1. Cooperate with the Fish and Wildlife Service and the South Carolina Wildlife Resources Department and other interested State agencies in developing plans for ecological surveys; initiate these surveys at least two years before reactor operation; and continue them on a regular basis during-.operation or until it has been conclusively demonstrated that no significant adverse conditions exist.

4

l. Cooperate with the Fish and Wildlife Service, the South Carolina Wildlife Resources Department, and other interested State agencies in developing plans for radio-logical surveys.

2. Conduct or arrange for the conduct of pre-operational radiological surveys of selected organisms and of the environment by competent scientists knowledgeable in the fish and wildlife field to include, but not be limited to, the following:

a. Water and sediment samples which should be collected within 500 feet of the reactor effluent outfall (need be measured only for gamma radioactivity).

b. Selected plants and animals which should be collected as near the reactor effluent outfall as possible (should be analyzed for both beta and gamma radioactivity).

3, ~repare a report of the pre-operational radiological survey and provide 5 copies to the Secreta11' of the Interior for evaluation prior to project operation.

4.

Conduct radiological surveys, similar to those specified in recommendation 2 above, analyze the data, and prepare and submit reports every 3 months during the first year of reactor operation and every 6 months thereafter or until it has been conclusively demonstrated that no significant adverse conditions exist. Five copies of these reports shall be submitted to the Secretary of the Interior for distribution to the appropriate State and Federal agencies for evaluation.

5. Make such reasonable modifications of project structures and operations as may be ordered by the Atomic Energy Commission upon its own motion or upon the recommendation of the Secretary of the Interior or the South Carolina Wildlife Resources Department.

We understand it is the Commission's opinion that its regulatory authority over nuclear power plants involves only those hazards associated with radioactive materials. However, we recommend and urge that, before the permit is issued, thermal pollution and other detrimental effects to fish and wildlife which may resu""lt from plant construction and operation be called to the attention of the applicant.

We recommend further that the applicant be requested to discuss this matter with appropriate State conservation officials and the Fish and Wildlife Service and to develop measures to minimize these hazards.

3

2. Meet with Fish and Wildlife Service and State of South Carolina agencies at frequent intervals to discuss new plans and to evaluate results of existiog surveys.

3. Hake such modifications in project structure and operation as may be determined necessary as a result ot the surveys.

The opportunity for presenting our views on this project is appreciated.

SincereJ.¥ yours,

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APPEIIDIX F Report to AEC Regulatory Staff ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE H. B. ROBINSON UNIT NO. 2 CAROLINA POWER AND LIGHT CCMPANY

{Docket No. 50-261) by N. M. Newmark and W. J. Hall 4 Feburary 1967

Introduction ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE H.B. ROBINSON UNIT NO. 2 CAROLINA POWER AND LIGHT COMPANY by N. M. Newmark and W. J. Hall This report concerns the adequacy of the containment structures and components for the 2094 MWt Robinson Unit No. 2, for which application for a construction permit and operating license has been made to the United States Atomic Energy Commission (Docket No. 50-261) by the Carolina Power and Light Company.

The facility is tobe located about 5 miles WNW of Hartsville, South Carolina, and about 25 miles NW of Florence, South Carolina.

Unit No. 2 is located adjacent to Unit No. 1 of the Robinson plant, a coal fired steam power plant of 182 MWe (net) capacity on a site situated on Lake Robinson, a man-made 2250-acre lake.

Specifically, this report is concerned with the evaluation of the design criteria that determine the ability of the containment system to withstand a design earthquake acting sil1'llltaneously with other applicable loaqs forming the basis of the containment ciesign.

The facility also is to be designed to withstand a maximum earthquake simultaneously with other applicable loads to the extent of insuring safe shutdown conditions.

The seismic design criteria for Class I equipment and piping also are reviewed.

The report is based on information and criteria set forth in the Preliminary Facility Description and Safety Analysis Reports (PFDSAR), and amendments thereto as listed at the end of this report.

We have participated in discussions with the ARC Resulatory staff, in which many of the design criteria were discussed in detail.

Description of the Facility H.B. Robinson Unit No. 2 is described in the PFDSAR as a pressurized water reactor nuclear steam supply system designed for initial power output of 2094 MWt.

The Teactor coolant system consists of three cloaed reactor coolant loops conttected in parallel to the reactor vessel, each containing a reactor coolant pump and. a steam generator.

The r*actor vessel will have an inside diameter of about 13 ft, and will operate with a design pressure of 2235 psig, 0

a design temperature of 650 F, and is made of SA*:302 Grade Blow alloy steel, internally clad with austenitic stainless steel. In most respects the design is similar to that for Ginna (Brookwood), Turkey Point, ani Indian Point.

The reactor containment structure, which enclose* the reactor and steam generators, consists of a steel-lined concrete shell in the form of a vertical right cylinder with a hemispherical dome, and a flat base supported by means.of steel-shell piles. The cylindrical side walls measure approximately 126 ft. from the liner on the base to the spring line of the dome; the inside diameter of the cylindrical shell is about 130 ft. The wall thicknesses of the cylinder and dome will be 3 ft-6 in. and 2 ft-6 in., respectively.

flle ban ~ill consist of a 10 ft.

thicK structural concrete slab. The discontinuity of the spring line, due to the change in thickness between the dome and cylindrical wall, ls on the outer surface.

The dome of the containment structure will be made of reinforced concrete.

The cylinder walls will be concrete reinforced circumferentially and prestressed vertically. The base slab will be of reinforced concrete.

The welded steel liner will be 3/8 in. thick in the cylinder wall region, 1/2 in. thick in the bottom and in the dome.

The liner is to be constructed of

-.. carbon steel conforming to ASTM designation A442-64 Grade 60.

The bottom plates of the liner are to be welded directly to embedded steel members in the founda-tion slab.

In the vertical walls and the dome the liner will be anchored to the concrete s\\lell by means of '1' -shaped anchor studs fusion welded to the liner plate so that it forms an integral part of the entire composite structuct 4nder all loadings.

As described in Amendment No. 3, the vertical prestressing system which has been chosen for post-tensioning of the Robinson containment structure will consist of 1-3/8 in. diameter high strength steel bars (supplied by the Stress Steel Corporation) grouped into tendons consisting of six bars per tendon.

These tendons will be placed within heavy wall 6 in. galvanized steel pipe sheaths, and will be located on the center line of the wall at a spacing of approximately 3 ft. around the periphery of the containment vessel.

The reinforcing steel to be used in the dome and base slab of the containment vessel will conform to ASTM A432-65.

The reinforcing steel to be used in the circumferential direction in the cylindrical portion of the containment vessel will be eilther ASTM Al5 or A408--Intermediate Grade.

The information on the geology at the site presented in Section 2.9 of the PFDSAR indicates that at the site the Piedmont crystalline basement rock is overlain with approximately 460 ft of unconsolidated coastal plain sediment.

These sedi-ments are comprised of about 30 ft. of surface alluvium lying over 430 fto of the Middendorf formations, which are made up of sands, silty and sandy clay,

sandstone, and audstone.

Compressional wave velocities are noted** 17,500 pfs in the basement rock, 7200 fps in the Middendorf, and 1500 fps in the top 30 ft. of alluvium.

Pile supports are to be used for all major structures.

Personnel and equipment access hatches are provided to permit access to the facility.

In addition there are a number of other penetrations for piping and electrical conduits.

Sources of Stresses in Containment Structure and Type 1 Components The containment structure is to be designed for the following loads; dead load of the structure; live load (including snow loads, construction loads 9 ice loads, and equipment loads); accident pressure load, associated with loss-of-coolant accident, of about 42 psig; test pressure load of 48.3 psig (115 percent of the 111aximum expected design pressure); internal negative pressure of 2 psig; uplift arising from buoyant forces crested by displacement of ground water by the structure (ground water elevation is assumed to be at grade level);

wind loading corresponding to 30 psf corresponding to a wind velocity of 85 mph; also the structure will be analyzed for tornado loadings (assumed wind velocity cf 300 mph with an *atlioaphetic pressure drop oJ l 'psi) u d**~*ribed on page Ao(7)-l of Aml!ndmeat No. 1 3 and on pa~e A(7)-l of A1111nd*11t No,. 7.;. thaTaal load arising from a maximum temperature gradient through the concrete shell and mat, based on the maximum design temperature; and seismic loa41 as described next.

'lbe PFDSAR (see page 2-63 of Volume 1 for example) recoaaends a design earthquake having a maximum ground acceleration of 0.10 sand a hypothetical earthquake having a aaxi.. ground acceleration of 0.20 g.

Comments on Adequacy of Design Seismic Design Criteria In connection with the selection of the design earthquake, we agree with the approach adopted, namely that of a basic design for a "design" earthquake with the provision that safe shutdown can be made for an earthquake of larger intensity, called the maxinun credible earthquake.

Thus, for the "design" earthquake there shall be a margin of safety. of the plant in terms of its containment capability.

For the "maximum credible" earthquake (called the hypothetical earthquake by the. Jpplicant) provisions are made so that safe shutdown of the plant can be achieved.

Therefore, the equipment and safeguards required for safe shutdown would be capable of functioning under this hazard.

It is noted that the site is 120 miles north of Charleston, South Carolina, the site of one of the largest earthquakes to have occurred in the continental United States. It is also noted that the McBee earthquake of October 26, 1959, had an epicenter only 15 miles away from the site, with an intensity at the site of about MM v.

In view of these facts special consideration is required to define the earthquake design criteria. The applicant has presented evidence indicating that in the opinion of it experts no earthquake motion greater than intensity MM VII :ha*

• bccurred in the

• general uea

• and one need not consider maximum earthquake ground acceleration of greater than 0.20 g for the maximum credible earthquake hazard at the site.

As a result of their review of the seismicity and geology of the area 1 the u. s.

Coast and Geodetic Survey and u. s. Geological Survey accept the applicant's pro-posed value of a design earthquake of 0,10g and a maximum credible earthquake of 0,20g.

We believe that 1 for these earthquake loadings 1 the design is suitably conservative considering the damping factors used and the fact that under primary stress loadings no yielding is permitted to take place in structural components and equipment essential for safe shutdown capability.

We understand that it is the intent of the designer to limit primary stresses in Class I components to less than the yield point of the material.

It is noted on page 2-63 of Volume 1 of the PFDSAR that the vertical acceleration shall be taken as two-thirds of the horizontal acceleration 1 which appears acceptable to us.

The spectra shown in Figs. 2-16 and 2-17 of Volume 1 of the PFDSAR and again as Figs, 1 and 2 of Appendix B of Volume 3 of the PFDSAR are patterned along the lines of the El Centro earthquake and after the California response spectra of Dr. G. w. Housner of the California Institute of Technology.

The shape of the spectra for the particular O,lg acceleration appears acceptable to us for this location and ground conditions.

With regard to the damping values to be used with the seismic design 1 it is indicated on page 5-38 (Volume 2

• PFDSAR) that the "dynamic analysis will be made on an idealized structure of three lumped masses and weightless elastic columns acting as spring restraints with 2% of critical viscous damping 0 "

In answer to Question VIII.A.(20) in Amendment l, it is noted that a critical damping of 2 percent will be used for the dome, cylinder, and base slab of the containment structure. It was indicated by the applicant that a damping value of 51,critical will be used for these parts of the facility for the maxitrum credible earthquake.

On page A-1 of Appendix A, Volume 3, PFDSAR, ls given a table of damping coefficients.

We believe the values given in this table are reasonable, and concur in the values of 2% and 51 for the containment structure for the design earthquake and the maximum credible earthquake, respectively.

The applicant has indicated that the stresses arising from the dead load, live load, thermal loads, seismic loads, etc. will be added directly in arriving at the analysis of the structure.

We consider this to be acceptable.

On page 2-65 and following of Volume l of the PFDSAR it is noted that there is an earth dam at the Robinson plant site and further that it has been analyzed through two different approaches for earthquake effects.

In one case it is noted that the factor of safety is about 1.08 for the "hypothetical" earthquake.

In answer to Question VIII.A.(11) it is noted that the failure of the dam is not considered credible, atd, further, that if the dam fails there will be no adverse structural effect on the containment or other structures important to plant safety as a result of lowering the lake surface or flooding.

We have reviewed the dynamic analysis of the dam and believe that there ~is

• an adequate margin of safety for the effe<;ts* of an earthquake l!ven *. somewhat gr~et in intensity th~ -tbe appli~ant'a lbypothetical eaTthqvaRe,of 0,20g, The containment structure is to be supported on piles.

A discussion of the design of the piles is presented in Section 5 of the PFDSAR and is amplified further in answer to Question VIII.A.(21) of Amendment 3. It is noted that the piles will be designed to function by a combination of point bearing and friction in the hard silty-clay layer at approximately 50 ft. below grade, and that no reliance will be placed on the upper soils for vertical pile support.

Some studies have been carried out concerning the possibility of liquefaction at the site, and these results appear to be negative. It is noted on page 5*23 of Volume 2 of the PFDSAR that "all piles will be capable of resisting wind and earthquake forces." In accordance with discussions presented elsewhere concerning the dynamic analyses, we understand that the analysis will consider rocking on the foundation which will have an effect upon the piles, as well as total lateral shear and vertical shears carried into the piles directly from the surround-1ng medium.

The applicant has indicated how the piles are expected to be attached to the foundation slab, and will use steel-shell piles which appear to have the requisite resistance to lateral displacements.

We believe the proposed pile foundation for the structure is adequat~.

General Design Approach The load factor design approach is described in page 5-16 of Volume 2 of the PFDSAR and amplification is given in Amendment 3.

We are in agreement with the approach outlined, however, in answer to Question VIII.A.(2) on page A(2)-2 it ia noted: "Elastic design procedures and analyses will be used in design of all elements of the r.ontainment structure.

Under factored load conditions portions of ~he hoop reinforcing steel will be allowed to yield so long as the average hoop steel strain remains within the yield stress.

For design basis acci-dent with no increased loads the steel stresses will' t-e below the yield stress."

This would indicate that under the factored 1.5 times design basis pressure loading a small amount of yielding in hoop-reinforcing steel would occur.

However, the design ls such that under the combined loadings of the design basis internal pressure of 42 pslg and the 0.2 g earthquake no yielding will occur.

We understand that in local areas of discontlnuties, such as the large opening, detailed analysis will be made and no yielding will be allowed for all factored load conditions.

We believe this design approach will result in an ample margin of safety.

The shear design is discussed in some detail in Section 5 of the PFDSAR and is elaborated on in Amendment 3.

In terms of tangential shear, it ls noted on page 5-31 of the PFDSAR that "the liner wi 11 be used to reaiJit shear loads insofar as principal stress analysis and shear deformation allow.

The anchorages are designed to assure composite action.

This participation in the side walls will be minimal because of the low shear strains experienced by the steel elements. "

In Amendment 2 it is noted tn answer to Question VIII.A.(12)(c) that the tangential shear in the dome will be resisted by the liner which will be proportioned so as to maintain the principal stresses under factored loads within the criteria established in tbl PPDSAJl, It is further noted that tangential shear in the cylinder will be resisted by the concrete in the cylinder wall, which will have a capability to resist all or most of this shear in accordance with ACI 318-63.

Any shear which cannot be resisted by concrete will be resisted by diagonal rein-forcing bars spiralling around the containment wall and extending beyond the height to which they are required.

This approach appears satisfactory to us o The pro-visions for radial shear appear appropriate.

Our main concern involves the employ-tnent of the doweling action to carry the vertical shears, as outlined in detail in answer to Question VIII.A.(12)(a). It is very difficult to calculate rationally the doweling action that can be expected, and to assure that no overstressing or gross deformations will occur.

This is particularly true in this type of facility*

where there is a liner for which integrity must be maintained to prevent leakage.

It is noted, for example, that it has been estimated that the principal shearing stresses will be about 16 9000 psi in the steel dowels.

Such an analysis must necessarily involve consideration of the bearing areas over which the doweling takes place, the axial and lateral deformations, the liner participation, etc.

It is our belief that, whereas some doweling action can be counted on for taking shear, in those locations where the shearing action becomes predominant, the doweling action alone should not be counted on for shear resistance, The applicant per-formed a rather detailed analysis (Amendment No. 7) and concluded that the system can handle the shears without distress of the liner or otherwise impairing the integrity of the system, We agree that the analysis appears to demonstrate that the structure will function as intended under the applicant's hypothetical earth-quake of 0,20g, In answer to Question VIII,Ao(3)* it is indicated that under the factored load conditions no vertical membrane tension will be permitted on the concrete.

We concur with this approach,

11 -

Cranes The matter of cranes deserves special attention to be sure that these cannot fall during an acci4ent or earthquake and thereby impair the ability of the structure for safe shutdown.

Some co!lll\\ents concerning the crane design are gtven in answer to Question VIII.A.(19), where it is indicated that a dynamic analysis will be made to determine the loadings for the crane c0tr.pone~ts and special provisions made for rail clamps and trolley clamps to insure stability of the crane.

We are in agreement with this approach, and only wish to emphasize that care must be taken to insure that the crane and its trolley cannot be displaced during an earthquake or other accident loading and thereby impair the ability for safe shutdown.

Penetrations The design of the large penetrations ls described briefly in Section S of tte PFDSAR and also receives further attention in answer to Questions VIII.A.(13), and with regard to instrumentation, in VIII.E.(l)(a).

We note that developments have proceeded regarding the poEsible methods of handling these penetrations and that a finite element analysis will be made by the Franklin Instlt~te Research Laboratories for these openings.

We concur with this approach and also assume, although this is not described in detail, that secondary effects arising from local bending, thermal effects, and other loadings will be considered either through modifications in the finite element analysis procedure or in the employment of other rational procedures for estimating these effects.

One necessary aspect*of providing proof of the integrity of these penetrations lies in the measurement program that is made concurrently with the proof testing of the containment vessel.

This instrumentation program is to be determined based on

• 12 -

the results of the Franklin Institute Analysis.

We should like to review this pro-gram prior to the measurements; and we understand that these measurements will be made not only on the stiffened ring section itself but in the transition zone be~ween the vessel and the reinforced ring section to insure that there is no gross distress or deformation in this transition region.

Piping and Other Type I Components The only cooment we can find in detail that concerns the design for seismic effects for piping i1 given on page 1-29 of Volume 1 of the PFDSAR, wherein it is stated:

"All piping, components and supporting structures of the reactor and safety related systems are designed to withstand any seismic *disturbance predictable for the site.

These criteria specify that there will be no loss of function of such equipment in the event of the assumed hypothetical ground acceleration acting in the horizontaJ and vertical directions simultaneously." We concur in this approach in general pro-vided primary stresses remain below yield including the 0.2g earthquake stresses as primary stresses.

Provisions will be made for the inertial forces arising from the earthquake excitations and the jet forces arising from breaks or ruptures or other forces arising from accident conditions in such a manner that the integrity of the containment structure is not violated and to insure that the liner and isolation valves at the piping penetrations, etc., are maintained in operable condition to in-sure leak-tightness of the containment structure.

CONCLUSIONS On the basis of the information presented, and in keeping with the design goal of providing serviceable structures and components with a reserve of strength and ductility, we believe the *design criteria outlined for the containment vessel and Class I equipment can provide and adequate margin of safety for seismic resistance.

APPENDIX G REPORT ON STRUCTURAL DESIGN OF THE H. B0 ROBINSON UNIT NO, 2 CONTAINMENT Certain features of the proposed H.B. Robinson containment design are novel*

and we have emphasized these characteristics in our review.

In this report, we will present our evaluation of the adequacy of the design presented in the PFDSAR and amendments thereto, This material represents a comprehensive document on which we have based our belief that the proposed design is acceptable.

1,0 Comparison In most aspects* the proposed containment design is very similar to the Ginna containment design which we have previously reviewed and determined to be accept-able.

Significant differences between the H. B, Robinson containment design and that developed for Ginna are as follows:

(1)

The Ginna design relies on rock anchors to resist vertical (longitudinal) forces from pressure loading whereas in the Robinson desi~ the internal pressure is entirely self-contained.

(2)

The Ginna containment is founded on bedrock while the Robinson containment has a pile foundation.

(3)

The Robinson containment employs a hinged sump detail whereas for the r,1nna design there was no requirement for such a detail.

(4)

The Ginna containment uses a wire/buttonhead prestressing system (BBRV) for its vertical prestress whereas the Robinson containment employs a stress steel rod system.

(5)

The Ginna containment has provisions for in-service surveillance of its pre-stressing tendons whereas the Robinson containment, by virtue of its anchorage details and grouting, does not have internal capability for periodic inspec-tion of tendons.

(6)

The Ginna containment uses a wax corrosion inhibitor for tendon corrosion protection whereas in the Robinson design cement grout is used 0 (7)

The Ginna design employs a neoprene hinge detail at the base-cylinder juncture whereas in the Robinson design a fixed detail is proposed, 2.0 Principal Design Features The proposed design* consisting of a concrete cylindrical shell with hemisphe-rical dome and flat base slab* relies on a prestressing tendon system to carry the longitudinal loads in the shell.

The structural loads in the dome* flat slab base.

shell hoop direction* and local flexural loads longitudinally in the shell are carried by mild steel reinforcing, A welded steel liner provides the vapour-tight barrier and is connected to the concrete wall using stud connectors.

The cylindrical wall portion of the liner is insulated, 3o0 Design Criteria and Analysis The applicant has proposed the use of a load factor* limit design approach for his design analysiso The loads 1 load factors 1 and loading combinations are essen-tially the same as those proposed for previous prestressed and reinforced concrete applications which we believe are conservative, 4.0 Large Opening Design The applicant acknowledges the complexities associated with arriving at an optimum opening design, Furthermore 1 the anisotropy associated with his partially preatressed (longitudinally) and partially reinforced (hoop direction) design is a complication to an already difficult design problem.

The Franklin Institute has been engaged to make a computerized finite element analysis of the design.

Since such an analytical technique enables the analyst to account in detail for the variation in structural stiffness and to predict detailed stress patterns around the lara* openings in a preciae UDA*r* the analysis of the opening design and 1 where required* modification of the proposed opening design through use of this technique provides added assurance of achievement of an adequate design, Based on the results of this design analysis* development of a meaningful confirmatory instrumentation program for the opening will be developed, We believe that this combination of analysis and thorough instrumentation during the proof test pro-vides assurance that the functional performance objectives for the large openings will be achieved.

s.o Shear Design Reliance is being placed on the provisions of ACI 318-63 for the computation of allowable shear stress across the unreinforced section; and where these shear stresses are in excess of code allowables. reinforcing steel is added, Membrane stress is being kept compressive across the net section up to the maximum factored load conditions to provide increased assurance that the imposed shears will be carried by the concrete section.

In addition* because of the existence of some degree of uncertainty in the application of the semi-empirical code provisions for a structure of this type and loading* the principal stresses are being critically examined in areas of high lateral and radial shear, This procedure for design against radial shears is considered acceptable, Noteworthy* also* is the designer's detail using radial bent bars instead of radial horizontal stirrups.

In view of the tendency for tensile cracking to be oriented horizontally* this type of rein-forcement is considered particularly desirable.

With regard to longitudinal shear, the designer has proposed a design detail that relies on dowel transfer of these stresses by the hoop reinforcing across concrete longitudinal cracks. It is considered that the mechanism proposed for this shear transfer is a logical one.

Test data which has formed the basis for the designer's assumptions has been furnished.

The applicant has additionally studied liner distress incident to development of longitudinal shear through dowel action and the structural failure point implicit in carrying longitudinal shear in this manner.

The results of his study indicate that liner strains incident to shear transfer in this manner will be acceptable and that failure in longitudinal shear is not a controling failure mode.

6.0 Seismic Design The applicant proposes the use of a three-lumped mass approximation of the structure and incorporation of a rocking mode in his dynamic analysis.

Using this analytical model, an analysis based on the response spectrum approach will be accomplished.

Several details of the designer's approach including selection of damping values, combination of flexural and rocking modes, and the mass/stiffness value selections have been reviewed in detail, We believe the damping value and mass/stiffness value selections and the means of combining flexural and rocking modes are conservative.

7o0 Penetrations The penetrations are designed to the criteria and standards of Section III, ASME Boiler and Pressure Vessel Code.

The liner is thickened in the region of the penetrations and anchorage is provided close to the penetration/liner junct ure to preclude failure at the leakage barrier (the weld) through load transfer from t he piping.

Consideration has been given to mechanical and thermal fatigue as well as static loading on the liner through piping system penetrations.

Cooling coils are provided for hot process piping penetrations to protect the surrounding concrete structure, These provisions are considered acceptable, 8,0 Liner Design It is the stated intent of the designer to maintain the liner design stress below yield under accident conditions.

In view of the secondary nature of liner compressive stresse such a criterion is conservative, To ensure composite action between the liner and its backing concrete* "L" shaped stud connectors are employed 0 The performance of these stud/liner connections under fatigue loading and the capability of the studs to act as a semi-flexible connection (and* in so doing*

redistribute excess loading to its neighbors* hence preventing a propagation type failure) have been supported by referenced test data.

The stud spacing has* using a conservative analytical model* been designed so that liner buckling will be precluded, 9,0 Materials The applicant will use Type II cement of controlled alkalinity and composition.

Calcium chloride will be excluded, and the ACI-recommended mix design procedures are specified, Aggregate will conform with the generally applicable ASTM aggregate specifications.

The reinforcing steel and prestressing bars are being procured in accordance with ASTM standards.

In addition, user testing of each heat of both reinforcing steel and prestressing bars is proposed.

The tendon and hardware performance capabilities under static and dynamic loads have been examined.

The stress steel bars to be utilized possess a high degree of ductility.

The coupling and anchorage hardware are capable of withstand-ing static and dynamic loads far in excess of design requirements.

In addition, the following program will be followed:

(1) fatigue testing at the minimum stress level in the bars under design loading* (2) full scale prototype testing of the stacked anchorage assembly* and (3) verification of the relaxation characteristics of the stress steel bars.

With respect to the liner material* the applicant proposes the use of a liner steel (A442) with good weldability* ductility* and controlled NDT propertieso Considerable attention has been devoted by the applicant to corrosion con-siderations, The selection of a quality tendon sheathing* attention to stress steel rod corrosion prior to prestressing 1 and close attention to grouting specification.

and procedures is planned, For the reinforcing steel conservative concrete cover requirements are being

followed, The cathodic corrosion potential, likewise, has been studied extensively and site soil resistivity surveys conducted.

It has been concluded that cathodic protection is not necessary, The provisions outlined by the applicant with regard to materials selection are considered to be acceptable, lOoO Construction and Quality Control User check testing of the materials of construction is proposed 1 and frequent control testing by cylinder strength and slump tests for concrete will be per-formed.

Careful attention is indicated for the bonding between construction lifts.

With regard to structural steel splicing, a high standard of performance and frequent testing of splices will be implementedo With regard to construction inspection, an organization independent of the construction crew and possessing the authority to stop construction as necessary has been established, The inspection group possesses experienced engineering personnel in close liaison with the design organization; and in addition~ an expert in prestressing has been engaged to monitor the prestressing phase of the construction.

We believe that the program for quality control is adequately organized and staffed to provide assurance that high quality construction will be achieved.

11,0 Testing and Surveillance With regard to selection of proof test pressure, the designer has elected to use the 115 percent of the design basis accident pressure, now becoming a rather common choice, In support of the adequacy of this proof test pressure level, he has shown that the predominant liner strains are more tensile under test loading than those stresses calculated to exist under design basis accident loading.

Inasmuch as increased leakage has typically been associated with the opening of leakage paths at high tensile stresses, the over-all (integrated) leak rate per-formance of the vessel under test pressure loadi.ngs at desi~ basis accident pressure is expected to be conservative with regard to its related performance under design basis accident conditions.

Likewise, the applicant has established that the stresses in the major structural elements of the containment under his selected proof test pressure loading closely approximate those in the structure under design basis accident conditions.

In addition* we unde~stand that the test pressure selection is well above the pressure level (81 percent of design pressure) required for the structure to perform in an analogous manner (i.e., the structure is predicted to be in the cracked regime) to that expected under design basis accident conditions.

Furthermore, it has been identified that the selected level of proof test pressure is compatible with the strain and deformation requirements of the proof test instrumentation.

The applicant's proof test is considered adequately justified, and we believe it is acceptable.

An instrumentation program has been outlined for the containment.

This pro-gram includes measurements of the deformation of the structure in essential regions.

provides for examination of the cracking pattern, provides verification of the hinge action of the reactor sump and includes measurement of liner membrane strains.

In addition, to relate this instrumentation program back to the design analysis, the applicant has indicated that strain and deformation predictions will be made after the detailed design is completed but before the proof test, The predictions will, in turn, be compared with measured values during the proof test as a check that the structure is performing as intended, With regard to in-service surveillance* the designer's use of stacked anchor-ages and grouted tendons precludes direct surveillance for loss of tendon prestress through relaxation* creep* or corrosion.

Inasmuch as such surveillance is not pro-vided and since proper performance of the structure depends on maintenance of pre-stress (longitudinally)* it is considered prudent that some periodic over-all proof test surveillance* though perhaps limited in pressure and infrequent* be pro-vided.

The design as established will permit such surveillance; however* the applicant continues to express his desire not to perform periodic pressure tests.

We understand* however* that small test capsules consisting of grouted tendons will be fabricated and destructively tested for corrosion at various periods during the life of the plant.

We believe that the design is acceptable since it does not preclude periodic testing.

12o0 Conclusions The applicant has presented in detail his design intent with regard to the containment.

In areas such as the analytical technique for analysis of the opening design* in the use of a hinge detail for the containment sump* in the reliance on dowel action for longitudinal shear resistance and in the use of stacked tendon anchorages* the design is somewhat novel.

Based on our review and evaluation of these novel features, the containment design, in general* and the construction pro-cedures* we have concluded that the structure will be adequate £or its intended purpose and will be capable of meeting its functional performance requirements.

APPENDIX H CHRONOLOGY -

REGULATORY REVIEW OF THE

~ARO!.INA POWER AND LIC,HT COMPANY APPLICATION (1)

July 12

• 1966 (2)

September 2, 1966 (3)

September 20-21, 1966 (4)

October 10, 1966 (5)

October 28, 1966 (6)

November 28, 1966 (7)

December 1, 1966 (8)

December 1, 1966 (9)

December 12, 1966 (10)

December 13, 1966 (11)

December 20. 1966 (12)

January la 1967 (13)

January 12, 1967 Submittal of Preliminary Facility Description and Safety Analysis Report (PFDSAR).

Applicant and regulatory staff met to discuss con-tainment design, Applicant and re~ulatory staff met to discuss all aspects of plant design, Submittal of Amendment No. 1, a complete revision of Section 5 (r.ontainment System.~) of the PFDSAR, List of regulatory staff questions issued, concerning all aspects of the plant desim, Submittal of Amendment No. 2, a partial reply to DRL questions issued October 28, Submittal of Amendment ~o, 3, a renly to all of n'R.T.

questions concernin~ the containment design.

Submittal of Amendment No. 4, a reply to the remainder of DRL questions issued October 28, Applicant and regulatory staff met to discuss answers contained in Amendments No. 2 and No, 3, First Advisory Committee on Reactor Safeguards (ACRS)

Sub-committee meeting with the applicant and n'R.L (at the site) primarily to discuss containment desi~.

Applicant and regulatory staff met to discuss the con-cainment design and answers contained in Amendment No. 4 concerning *safeguards systems and ~ccident analyses.

Submittal of Amendment No. 5, clarification of answers submitted in Amendments No, 2, No, 3, and No, 4, First ACRS meeting with the applicant and the regulatory staff,

(14)

January 27, 1967 (15)

January 31, 1967 (16)

February 1* 1967 (17)

February 7* 1967

( 18)

February 8

• 196 7 APPENDIX H (continued)

Submittal of Amendment No. 6, ~oncerning.seismicity of the site.

Applicant and DRL met to discuss,. information included in Amendments No. 5 and No. 6.

Second ACRS Sub-committee meeting with the applicant and the regulatory staff.

~ubmittal of Amendment No. 7 concerning containment shear design and final clarification of information submitted in response to DRL questions o*f October 28 9 1966.

Second ACRS meeting with appli*cant and the regulatory staff.

TABLE I TABLE II TABLE III TABLE IV APPENDIX I COMPARISON OF H, B. ROBINSON REACTOR CHARACTERISTICS WITH THOSE OF TURKEY POINT AND OTHER RECENT PWRs lNDEX TO TABLES Includes a summary of similarities and differences between H, B. Robinson and Turkey Point.

Includes a comparison of the thermal-hydraulic parameters of H.B. Robinson with other PWRs.

Includes a comparison of nuclear design characteristics.

Includes a comparison of reactor safety features of recent PWRs.

TABLE I SIMILARITIES AND DIFFERENCES BETWEEN H, B1 ROBINSON AND TURKEY POINT UNITS Component Reactor Static and Kinetic Nuclear Ch aract.e ris ti.cs Thermal-Hydraulic Design Mechanical Design Instrumentation Coolant Loops Containment System Engineered Safety Features High and Low Pressure Safety Injecti.on and Accumulators Containment Spray Containment Coolers Comparison Essentially Identical Essentially Identical Essentially Identical Essentially Identical Identical Significantly Different Identical Different Different Remarks' The components and design parameters are identical with the exception of a few slight differences which are of no significance to the safety evaluation of the systemso There may be some small differences, depending on the preference of the applicantso Certain features of this prestressed concrete design ar.e different, The tendons are grouted 0 and the building is only vertically pre-

stressed, Capacities are identical.

Capacity is smaller due to larger containment volume.

Also, incorporates iodine removal system.

Same finned cooling units used for normal operations are used for accident conditions, Turkey Point uses plain tube emergency cooling units.

Parameter Power (MW)

Nominal Pressure (psir)

Average Core b T (° F)

Coolant Volume (ft 3)

DNB Ratio at Nominal Conditions Minimum DNBR for Design Transients TABLE II COMPARISON OF THE THERMAL-HYDRAULIC PA'RAMETF:RS H, B.

Robinson 2094 2250 59 9800 1.85 1.30 Turkey Point 2097 2250 59 9800 1.85 Indian Point 112 2748 2250 57 12.209 Ginna 1300 2250 54 6238 Average Mass Velocity

-2.35 x 106 2.35 x 106 2.56 x 106 2.43 x 106 (lb/hr-ft2)

Average Heat Flux x 10-6 O.1642 O.1642 O.1756 O.1518 (Btu/hr-ft2)

Maximum Heat Flux x 10-6 0 05336 O.5336 005708 O.5175 (Btu/hr-ft2)

Linear Heat s.3/17.3/19.4 5.3/17.3/19,4 s.7/18.5/20.7 4.9/16.7/18, 7 Generation Rate Average/Maximum/Maximum at Overpower (Kw/ft)

Maximum Fuel 4070/4270 4070 I 42 70 4200/4400 3920/4150 Center Temperature 100% Power/Overpower (°F)

TABLE III COMPARISON OF NUCLEAR DESlr,N CHARACTERISTICS Parameter Core Equivalent Diameter (in.)

Core Height (in.)

Cold Water /U. Ratio No. of Fuel Assemblies No. U02 Rods per Assembly K eff. (Beginning of Life)

Cold, Clean, Zero Power Hot* Clean, Zero Power Boron Cone. to reduce K eff.

to 0.90 with no rods inserted, clean* cold/hot (ppm)

Moderator Temperature Coefficient (8 k/k/°F)

H. B.

Robinson 119.5 144 3.48 157 204 1.275 1.225 2300/2500

+1 X 10-4 to

-3 X 10-4 Turkey Point 119.5 144 3.48 157 204 1.275 1.225 2300/2500

+1 X 10-4 to

-3 X 10-4 Indian Point 112 133. 7 144 3.48 193 204 1.275 1.225 3400/3500

+1 X 10-4 to

-3 X 10-4 r.inna 96.5 144 3.32 120 179 1.255 1.210 2500/2350

+1 X 10-4 to

-3 X 10-4

system Emergency feedwater supply Water sources 1

, L.E IV COMPARISO~ OF REACTOR SAFETY FEATURES H. B8 Robinson 1-100% capacity steam driven turbine pump 2-50% capacity pumps operating from diesel. power.

Turkey Point 1-100% capacity turbine pump plus shared back.up pump Indian Point II 2-100% + capacity steam driven turbine feedwater pumps in normal use plus 1 auxiliary steam driven turbine feedwater pumps Condensate storage Condensate storage.

Condensate storage tank* service water tank tanke plus one-system or deep million gallon tank Ginna 1-100% capacity steam driven turbine pump 2-507. capacity pumps operating from diesel power Condensate storage tank and firewater system wells on hi_llside

_,;;...,...,.._""'30_,,.....,..,.""""...,..==-==-=,cxa;,~--=--

- ~"""""--""""'.....,.""3-..:,o;,o--------------------

Emergency : core CQ~li~g equip~nt Containment Volume Containment spray system Heat removal capacity (both pumps) BTU/hr Add-mixture for iodine removal 1-1200 ft 3 1~1200 ft3 1-1200 ft3 accumulator for accumulator fer accumulator for each loop each loop each loop 3-300 gpm high head safety injection pumps 3-300 gpm high head pumps (shared system) 3-400 gpm high head pumps 1-1750 ft 3 accumulator for each loop 3-400 gpm high head pumps 2-3000 gpm core 2-3000 gpm core 2-4000 gpm core 2-2000 gpm core 4~ l~_ge_ J?.20 s

-,.:.,..- ~--- - q_e.a..c)=_*_ll.=.;)J!:..._e

• =-<0-=D"'-' _

_u-'C..

m=

• -0.....:S='----'

- =-=-

-==*=-*.=::c-.....;*a:..** d_e__,_;~u=-ge_.. l?,..._U...,,!!!P--8_.a..a-=ro... -........,,__................. d_e_lJ!_,..ge ___ t>:..U

.... 111l>_.. _s ___ _

2 0 1 X 106 ft3 1,55 X 106 ft3 2 0 6 X 106 ft 3 0,97 X 106 ft) 2-880 gpm pumps 2-1320 gprn pumps 2-2600 gpm pumps 2-1200 gpm pump*

144 X 106 240 X 106 360 X 106 200 X 106 Sodium thiosulfate None Sodium thioaulfate Sodium thiosulfate

system Fan-cooler system Normally operating Cooling coil water supply Heat removal capacity (total)

Charcoal filters Service water system Loads on system after accident Radiation monitor downstream of fan-coolers component cooling system Service during MCA Emergency Power source H, B1 Robinson 4-85.000 cfm units Yes Service water 144 x 106 BTU/hr None 2 Separate systems with capability to isolate any break from rest of system Component cooling heat exchangers and fan-coolers Yes Single line 3 Pumps 2 Heat exchangers Residual heat exchangers only 2 Diesels approx, 2 MW(e) each TABLE IV (continued)

Turkey Point 3-81.000 cfm units No Component cooling water 180 x 106 BTU/hr None 2 Separate systems Component cooling heat exchangers only None needed except in component cooling loop Single line 2 Pumps 2 Heat exchangers Residual heat exchangers plus fan-coolers 2-approx, 2,5 'MW(e) each Indian Point II 5-65 0000 cfm units Yes Service water 360 X 106 Yes 2 Separate systems similar to Robinson Component cooling heat exchangers and fan-coolers Yes Single line 3 Pumps 2 Heat exchangers Residual heat exchangers only 3-approx, 2,0 'MW(e) each Ginna 4-38,000 cfm units Yes Service water 200 X 106 Yes on 2 of the 4 units 2 Separate systems similar to Robinson Component cooling heat exchangers and fan-coolers Yes Single line 3 Pumps 2 Heat exchangers Residual heat exchangers only 2-approx, 2 MW(e) each

Uc"tNst Au-r-i,, o I n O,,ITY FltE Cop*

May 18, 1970 SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENSING U.S. ATOMIC ENERGY COMMISSION IN THE MATIER OF CAROLINA POWER AND LIGHT COMPANY H.B. ROBINSON UNIT NO. 2 DOCKET NO. 50-261

1.0 2.0 3.0 TABLE OF CONTENTS INTRODUCTION SITE AND ENVIRONMENT 2.1 Site Description 2.2 Meteorology 2.3 Hydrology 2.4 Geology and Sei811ology 2.5 Environmental Radiation Monitoring 2.6 Radioactive Waste Disposal FACILITY DESIGN 3.1 Reactor Design 3.2 Reactor Coolant System 3.3 Containment Structure 3.4 Design of Class I Structures and Components 3.5 F.mergency Core Cooling System 3.6 Containment Cooling Systems 3.7 Protection System Instrumentation 3.8 Emergency Power System

3. 9 Auxiliary Systems 4.0 ACCIDENT ANALYSES 5.0 EMERGENCY PLANNING 6.0 CONDUCT OF OPERATIONS i

7.0 TECHNICAL SPECIFICATIONS 8.0 REPORT OF ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9.0 C01'MON DEFENSE AND SECURITY 10.0 FINANCIAL QUALIFICATIONS

11.0 CONCLUSION

S APPENDIX A - Report of Advisory Comnittee on Reactor Safeguards APPENDIX B - AEC Regulatory Staff's Evaluation of Financial Qualifications ii

1.0 INTRODUCTION

The Carolina Power and Light Company (CP&L, applicant) has filed an application for all necessary AEC licenses to construct and operate the H.B. Robinson Unit No. 2 (Robinson, facility) reactor plant at its site in Darlington County, South Carolina.

In Amendment No. 8, dated November 22, 1968, the applicant filed its final Safety Analyses Report (FSAR) in connection with the application for an operating license which would authorize operation of the facility at steady state power levels up to 2200 megawatts (thermal). The facility, which will utilize a Westinghouse pressurized water reactor (PWR), has been under construction since issuance of a construction permit on April 13, 1967.

It is located in Darlington County, in the northeastern section of South Carolina, at a site near the southwestern corner of Lake Robinson.

Unit No. 1, which is a fossil-fueled power plant, is also located at this site.

The information submitted in the FSAR was subsequently supplemented by Amendments 9 through 21 of the application. Our safety evaluation of the design of the facility has been based on Amendments Nos. 8 through 21 of the application. All of these documents except Amendment No. 20 are available for public inspection at the Atomic Energy Colllllission's Public Document Room at 1717 H Street, N.W., Washington, D. C.

Amendment No. 20 which contains information on the startup organization, including the names of individuals involved has been withheld from public disclosure pursuant to 10 CFR § 2.790(b) at the request of the applicant.

The substance of this information, without reference to names, is provided in Amendment No. 21 which has been placed in the Public Document Room.

In the course of the review we have held numerous meetings with the applicant to discuss and clarify the technical material submitted.

In addition to our review, the Advisory Comittee on Reactor Safeguards (ACRS) reviewed the application and met with both us and the applicant to discuss the facility.

The report of the Comaittee, dated April 16, 1970, is attached as Appendix A to this safety evaluation.

In our review of the application we (a) performed an evaluation of site-related features; (b) compared the design and safety features of this plant with those of plants previously reviewed, and where possible, relied on previous evaluations of like systems and components, and structures without performing a separate evaluation; (c) assured that the design features and the treatment of safety matters were consistent with current regulatory criteria and policies, and that the applicant adequately addressed concerns which have been identified by the ACRS in previous reviews; and (d) identified and evaluated those design features and safety matters which have been changed since our review of the preliminary design in connection with the issuance of a construction permit for this plant, or which are unique to the Robinson plant.

We have reviewed the capability of the plant engineered safety features and the radiological consequences of various postulated accidents at both the ultimate power level of 2300 MWt and the rated power level of 2200 MWt.

The licensed power level will be 2200 MWt.

Our evaluation of the performance of the emergency core cooling system (ECCS) indicates that in the unlikely event a design basis loss-of-coolant accident were to occur in operation at the 2300 MWt power level the resulting calculated peak fuel clad tempera-ture would be in a range in which the chemical reaction between the clad and the coolant water would be rapid and autocatalytic. The ability of the ECCS to arrest the clad temperature rise under these conditions has not yet been demonstrated experimentally.

As discussed in Section 3.5 of this Safety Evaluation, for a loss-of-coolant accident occurring at the rated power level of 2200 MWt, the performance of the ECCS would limit calculated peak fuel clad temperatures to lower values, which we have found acceptable on the basis of presently available information on the performance of zircaloy cladding at high temperatures.

Based upon our evaluation of the facility as presented in sub-sequent sections, we have concluded that the H.B. Robinson Unit No. 2 can be operated at a power level up to 2200 MWt without undue risk to the health and safety of the public.

However, initial plant operation will be lilllited to a maximum power level of 5 MWt until additional analyses of certain Class 1 piping and equipment have been completed and we have reviewed the results of these analyses. A discussion of the analyses required is presented in Section 3.4 of this Safety Evaluation.

2.0 SITE AND ENVIRONMENT 2.1 Site Description The H.B. Robinson site is located in Darlington County, in the northeastern section of South Carolina.

The plant is situated in the southwestern corner of Lake Robinson, a man-made lake which was created to provide cooling water for thermal-electric-generating units at the site. Agriculture is the predominant activity in the innediate environs of the site. The nearest town of signifi-cant size is Hartsville, which is about five miles from the site, and which has a population of about 8,700 people.

Florence, South Carolina, about 25 miles southeast, is the nearest population center with about 25,000 people.

The nearest site boundary, 1400 feet south of the plant, determines the minimum radius of the exclu-sion zone.

The applicant considers the low population zone to extend to a distance of 4.5 miles.

The outer boundary of this zone excludes the town of Hartsville.

We have used the 1400-foot exclusion zone radius and the 4.5-mile low population zone distance in our evaluations of the potential consequences of postulated accidents.

2.2 Meteorology The H. B. Robinson site is located in the southeastern section of the United States, an area characterized by relatively unfavorable dilution climates.

The applicant has presented data obtained from one year of accumulated meteorological measurements taken at the site which include wind speed, direction, and persistence, and atmospheric stability distribution.

We and our meteorology consultant, the Environmental Science Services Administration, have reviewed the data presented by the applicant and conclude that the data are adequate to provide a basis for establishing limits on the routine release of radioactive effluents from the site; and that the data confirm the adequacy of the meteorological assumptions used in our accident analyses.

2.3 Hydrology We evaluated the requirements for flood protection of the facility prior to issuance of a construction permit and during our current review.

The applicant has performed an analysis of potential flooding at the site which could be caused by the probable maximum hurricane.

Lake Robinson lies in a potential hurricane area, which according to past records, may be subjected to a maximum rainfall of 20 inches in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

We and our consultants, the U.S. Geological Survey, reviewed the applicant's analysis and agree that flooding of the facility is improbable since the plant grade is above the maximum lake level which can be maintained by Lake Robinson dam and its appurtenant structures.

2.4 Geology and Seismology Our consultant, the U.S. Geological Survey, studied the geological aspects of the site during our review of this facility prior to issuance of a construction permit.

They concluded, and we agree, that the geology of the site provides an adequate founding medium for the facility building and structures.

No information was developed during excavation or construction which has changed this conclusion.

The applicant's seismic design bases specify that (a) for a maximum gromid acceleration of O.lg, resultant stress levels for critical components, equipment and structures necessary to ensure a safe and orderly shutdown (designated as Class I) will not exceed the values allowed by the applicable codes and (b) for ground accelerations of 0.2g, there will be no loss of function of critical structures and components necessary to ensure a safe and orderly shutdown of the plant.

Based upon the report provided at the construction permit stage by our seismic consultant, the U.S. Coast and Geodetic Survey, ve have concluded that these design basis accelerations are acceptable.

The growd acceleration of O.lg corresponds to the Operational Basis Earthquake (OBE),

whereas a ground acceleration of 0.2g corresponds to the Design Basis Earthquake (DBE).

Two strong-motion seismographs have been installed to record data related to ground motion in the event of a seismic disturbance at or near this site. One recorder is located adjacent to the con-tainment building, on the foundation mat, and the other is located approximately 100 yards from the containment.

These data would be employed in the subsequent evaluation of the effects of the seismic event on future safe operation of the facility. A plan for evaluating the effects of an earthquake on the facility using the data from the seismographs is being developed by the applicant.

The applicant will advise us of the details of the plan as they become available.

We are also developing guides for the use of seismic measuring instrumentation in evaluating the effects of earthquakes on nuclear facilities.

2.5 Environmental Radiation Monitoring The principal requirements for the applicant's environmental radi-ation monitoring program are listed in the Technical Specifications.

The applicant will collect samples of surface and ground water, bottom sedi.lllents, soil, fish, air, and benthic plants and animals in the lake; and will make measurements of atmospheric radiation levels.

Reconnendations from our consultants, the Fish and Wildlife Service of the U.S. Department of the Interior, have been incorporated into the applicant's environmental radiation monitoring program.

We conclude that the applicant's program will be adequate for monitoring the radiological aspects of plant operation on the environs and assessing the health and safety aspects of the release of radioactivity to the environment from the operation of the plant.

2.6 Waste Disposal Small quantities of radioactive liquid waste will be released routinely to the condenser circulating water where they will be diluted and discharged to Lake Robinson.

Since the rate of flow through the lake is relatively small compared with the rate of circulation through the plant, the applicant has evaluated the dilu-ting capabilities of the lake in order to establish the rate at which radionuclide& may be released.

The Technical Specifications limit the discharges to less than 26 millicuriesper day of mixed radio-nuclides and 10.5 curies per day of tritium. These release rates would result in concentrations in the circulating water and in the lake that are well below the concentration limits of 10 CFR Part 20, Appendix B, Table II, Column II.

We conclude that they are acceptable.

Release of gaseous waste containing radioactivity will be limited by Technical Specifications such that annual average concentrations at the minimum exclusion radius of 430 meters will not exceed the limits of 10 CFR Part 20, Appendix B, Table II, Column I.

For halogens and particulates with half lives greater than 8 days, these limits are further reduced by a factor of 700 to take into account potential reconcentration effects for these nuclides.

Technical Specifications further require that radioactive waste discharges from the facility be maintained as low as practicable.

3.0 FACILITY DESIGN The following sections briefly describe the design of those systems and features of the Robinson plant that are important to its safe operation.

The reactor is a closed cycle, pressurized, light water moderated and cooled reactor.

The nuclear steam supply system for the Robinson plant was designed and manufactured by Westinghouse under a turnkey contract with the applicant.

Westinghouse contracted with Ebasco Services for architectural engineering services and construction.

The principal design features, materials of construction and architectural arrangement of various systems of the Robinson facility are the same as those of the R.E. Ginna facility (Docket No. 50-244) which has been licensed for operation by the Comnission.

For these features we have relied on our previous review of the Ginna application in reaching conclusions regarding the adequacy of the H.B. Robinson plant.

3.1 Reactor Design 3.1.1 Core Thermal and Hydraulic Design We have evaluated the adequacy of the core thermal and hydraulic design for steady state plant operations and for anticipated plant transients.

The design criteria for the prevention of fuel damage are (a) the departure from nucleate boiling (DNB) ratio (defined by the Westinghouse W-3 correlation) shall not be less than 1.3 during normal plant operation, or as a result of anticipated transients; and (b) no fuel melting shall occur during normal operation, including anticipated transients.

The applicant's analysis of anticipated transients, including loss-of-flow and secondary system transients, indicate that the DNB ratio will be greater than 1.3 for operation at a power rating of 2200 MWt.

The lowest DNB ratio was calculated for the case of simultaneous loss of electrical power to the three reactor coolant pUlllps, which resulted in a DNB ratio of 1.31.

In addition no fuel melting is predicted by the applicant t o occur for steady state operation and anticipated transients.

Based on our evaluation of the results of the analysis, we con-clude that the reactor core thermal and hydraulic design is acceptable at the rated power of 2200 MWt.

3.1.2 Nuclear Design The nuclear design of the Robinson reactor incorporates the principal features of the Ginna reactor.

These include the use of part-length control rods, burnable poisons, a chemical neutron absorber and zircaloy cladding.

On the basis of our previous review of these features for the Ginna reactor we conclude that they are acceptable for this facility.

Unlike Ginna, the Robinson reactor will use fuel rods pre-pressurized with helium.

Two-thirds of the initial core loading for the Robinson reactor will consist of these pre-pressurized fuel rods.

The appli-cant intends that subsequent core loadings will consist entirely of pre-pressurized rods.

The applicant states that pre-pressurization of the fuel rods will increase the fatigue life of the fuel cladding by minimizing the contact between the fuel and the cladding, and by partially offsetting the effect of the external pressure of the reactor coolant which causes the clad to creep inward upon the fuel pellets.

In addition, the improved heat transfer ~haracteristics of helium result in reduced fuel temperatures at operating conditions. The mechanical and thermal-hydraulic design bases of the pressurized fuel rods are the same as those for nonpressurized rods that were used for Ginna.

The pressurized fuel rods have been under investiagion by Westinghouse for approximately five years.

These investigations include out-of-pile and in-pile experimental programs and analytical studies.

Internally pressurized fuel rods are currently being irradiated in other reactors; however, the number of these fuel rods is relatively small and the results of the investigations of the in-pile experimental programs are not yet available.

We have reviewed the results of the applicant's analytical studies and the proposed programs and we conclude that the use of pre-pressurized rods is acceptable and should reduce the potential for fuel failures due to cladding fatigue under normal operating conditions.

However, we have concluded that some of the Robinson fuel should be examined after irradiation to a high burnup.

The Technical Specifications include a program for inspection of irradiated fuel to provide assurance that unexpected fuel deterioration does not occur and that the pre-pressurized rods perform as anticipated.

Studies have been conducted to assure that core power distributions can be maintained within design limits. The spatial stability of the xenon distribution and resultant core power peaking abnormalities have been investigated by Westinghouse with the conclusions that (a) the core is stable against radial, diametral, azimuthal, and quadrant-to-quadrant xenon spatial oscillations, and (b) the core may be unstable against axial xenon spatial oscillations. Part-length control rods are provided to prevent unacceptable axial power distributions.

Furthermore, the out-of-core power range nuclear instrumentation provides an automatic trip set point reduction on detection of an excessive axial power imbalance.

During the start-up program, a test will be performed to verify that diametral stability exists. Additional measures are available to assure such stability and will be taken if necessary.

Provision is also made to display (at the control console) power tilts in the horizontal plane.

A channel of the area radiation monitoring system is provided for the early detection of failed fuel in the reactor.

A program to develop more precise instruments for the prompt detection of failed fuel is currently being conducted by Westinghouse.

The applicant is following these development efforts and will evaluate the effectiveness of such instruments when it becomes available.

We conclude that the nuclear design of the Robinson facility including the use of pressurized fuel rods, is acceptable.

3.2 Reactor Coolant System The reactor primary coolant system, including the pressure vessel and all pressure piping is designed for a pressure of 2485 psig and a temperature of 650°F.

All material in contact with the pri-mary coolant is stainless steel (or Inconel in the steam generators).

The reactor coolant system has been designed to withstand the normal loads of mechanical, hydraulic and thermal origin plus loads resulting from anticipated transients and from the OBE and DBE, based on the stress limits of Section III of the ASME boiler and pressure vessel code.

The reactor coolant system piping has been designed to the ANSI B31.l-1955 code for power piping, including the requirements of all code cases applicable to nuclear piping.

These design criteria are acceptable for the Robinson plant.

In addition to the inspection requirements of the code for the reactor coolant piping, the applicant will conduct additional ultrasonic and dye penetrant inspections of forged fittings and pipe, and radiographic and dye penetrant inspections of all pipe welds.

This program upgrades the inspection of the Robinson reactor coolant piping to essentially that required for Class I (functional) systems under the ANSI B31. 7 nuclear power piping code.

We find the inspection program to be acceptable.

An in-service inspection program for the reactor cool ant system is described in the Technical Specifications.

The applicant will review the program with us after five years of reactor operation, and modify it as necessary based on experience gained during operation.

At that time, we will require the applicant to per-form such inspections of components outside the reactor coolant pressure boundary as deemed necessary to provide continuing assurance of structural integrity.

We conclude that the in-service inspection program, with provision for continuing review, is acceptable for this plant.

The reactor internals are designed to withstand the normal opera-tional loads in conjllllction with either the loss-of-coolant accident blowdown loads or the loads which would result from the design basis earthquake.

The applicant has submitted a general description of the techniques used to study and analyze the structural response of the vessel internals under blowdown and seismic excitation.

The results of the study indicate that the maximum deflections and stresses in the critical structures are below values which would prevent adequate core cooling or reactor shutdown.

An analysis was performed which indicated that the reactor internals are designed to withstand flow-induced vibrations.

In addition, cold and hot vibration measurements will be made during the start-up test period to verify this capability.

We have concluded that the reactor coolant system for the Robinson plant is acceptable.

3.3 Containment Structure 3.3.1 Containment Structural Design The containment structure is a stePl-lined concrete shell in the form of a vertical right cylinder with a hemispherical dome and a flat base supported by means of piles. The Robinson con-tainment is the second coumercial reactor containment structure which is prestressed only in the vertical direction in the cylinder walls and conventionally reinforced in the hoop direction.

The R. E. Ginna plant, the only other such design, is presently beginning operations.

The most significant structural difference between the two containments is that Ginna uses a greased-wire tendon prestressing system coupled with a base hinge and rock anchors, while H.B. Robinson utilizes grouted steel bar tendons, with no base hinge or rock anchors.

The Robinson containment structure has a free volume of 2.1 million cubic feet and a design pressure of 42 psig.

The structure consists of side walls measuring 126 feet from the liner on the base to the spring line of the dome and an inside diameter of 130 feet.

The side walls of the cylinder are 3 1/2-ft thick and the dome is 2 1/2-ft thick.

The radius of the inside surface of the dome is equal to the inside radius of the cylinder, i.e., the discontinuity at the spring line due to the change in thickness is on the outer surface.

The base of the containment is a ten-foot-thick concrete slab.

The base liner is installed on top of the structural slab and covered with two feet of concrete.

In a manner similar to other plants, the applicant has calculated the pressure transients which would occur in the containment following loss-of-coolant accidents assuming various sizes of primary coolant system pipe breaks.

Break sizes up to and including the double-ended severance of the largest reactor coolant pipe result in a calculated peak containment pressure of about 38 psig.

The 11%

margin between the calculated peak accident pressure and the containment design pressure of 42 psig is acceptable.

The structural design of this containment is such that it acts as a series of rectangular, vertical prestressed wall panels.

~ongitudinal (i.e., vertical) shear between these panels is carried by the dowel action of the hoop reinforcing across vertical cracks.

The applicant has determined that the liner could carry 30% of the total longitudinal shear and remain within yield stresses.

The liner is not bonded to the base mat and cannot be relied upon at the base for transfer of shear to the foundation.

Therefore, it would not be possible with this design to carry all the shears solely in the liner.

We have reviewed the resulting stresses and conclude that they are acceptable.

Radial shears are being carried by concrete and radial bent bars.

Membrane stress is being kept compressive across the net wall section to provide assurance that the concrete will contribute to the radial shear capacity.

We consider this design to be acceptable.

During our evaluation of the construction permit application, we reviewed and accepted the tendon system (six 1 3/8-inch-diameter Stressteel bars per tendon, anchored with Howlett Grip Nuts).

The foundation pile design was reviewed at both the construction permit stage and the operating license stage of review and we determined that its design is acceptable.

We conclude that the structural design aspects of the containment are acceptable.

3.3.2 Containment Structure Testing and Surveillance Initial proof-testing of the containment structure will consist of a preoperational structural test at 48.3 psig, which is 115% of the design pressure, and an integrated leak rate test at the

• design pressure of 42 psig, followed by a sensitive leak test of penetration and liner seam welds.

The allowable maximum test leak rate has been set at 0.08% of the contained air weight per day.

We have concurred in the applicant's preoperational testing program.

The applicant has designed and installed a containment penetration pressurization system and a containment isolation valve seal water system.

Because these systems provide the capability to perform partial leak tests of the containment structure, the applicant did not propose to conduct any additional integrated leak rate tests during the service life of the containment. It was the applicant's intention to periodically test the channels covering the liner seam welds and to check individual penetration seals on the premise that if leakage were to occur these would be the most probable leakage sources.

Although we agree that the welds and penetrations are the most likely sources of leakage, we concluded that the con-tainment should be subjected to periodic, integrated leak tests as has been required for all previously licensed facilities because of the uncertainty in accurately identifying all small, undetected leakage paths.

Therefore, we will require periodic in-service con-tainment integrated leakage tests, as well as local tests.

We have included in the Technical Specifications the requirement for periodic integrated leak tests of the containment.

We will review the results of these local and integrated leak tests and will consider appro-priate modifications to the testing program during service life.

For confirmation of continued containment structural integrity during service life, the applicant proposes to test two sample tendons which have been placed on a wall contiguous to the containment wall.

These part-length tendons were placed at the same time and in the same general physical environment as the functional tendons.

One sample tendon will be removed five years after startup and checked for deterioration by a laboratory.

The second tendon will be inspected 25 years after startup.

We have also concluded that additional periodic pressure testing should be performed in conjunction with an in-service inspection program which is based on a ten year frequency of inspection.

Such tests are required by the Technical Specifications.

The above program provides appropriate means for determining the continued reliability of the containment structure and its leak tight capability.

3.4 Design of Class I Structures and Components The applicant has categorized as "Class I" those structures (e.g.*

containment or auxiliary building), and those components (e.g.,

reactor vessel and internals, and the emergency core cooling system),

whose failure could cause a significant release of radioactivity or which are vital to a safe shutdown of the facility and the removal of decay heat.

We have reviewed the applicant's classification of structures and components and we conclude that all structures and components have been classified appropriately.

In evaluating the design of the Class I structures, systems, and components, our seismic design consultant, J. A. Blume and Associates, Engineers, considered the following general aspects:

1.

The geology and nature of the subsoils;

2.

Design loads and load combinations; and

3.

The seismic design parameters and methods of analysis.

We and our consultant have found these design methods and criteria to be generally acceptable.

However, as stated in the ACRS letter, the applicant is reviewing the seismic design calculations and will supplement the seismic analysis of Class I piping and some Class I components with additional evaluations to demonstrate their adequacy.

This additional evaluation will include (a) an analysis of the piping stresses at the most critical Class I piping locations and the corres-ponding allowable stresses; (b) an analysis of the most critical Class I piping support loads, including maximum seismic loads; (c) a description of the type and location of each support analyzed, its design and ultimate capacity, and the seismic amplification factor associated with the location of the support in the building; and (d) an evaluation of these results.

The applicant will perform a dynamic analysis of the reactor coolant system and other Class I piping that will be selected on the basis of stress criteria described in the applicant's letter of April 6, 1970. Until this information is submitted and we have determined that it is satisfactory, reactor power will be limited to 5 MWt.

The containment structure and the auxiliary building are designed to withstand the effects of wind loading and the potential missiles resulting from a tornado with the equivalent of 300-mile-per-hour winds with a concurrent pressure drop of 3 psi.

The refueling building superstructure was not initially designed to withstand tornado loading, but was subsequently evaluated for these loads.

The applicant concluded that the roofing and siding would blow off, but that the structural frame would not collapse, and no missiles with sufficient energy to damage spent fuel elements or Class I structures and equipment would be generated by the refueling building roofing and siding.

The externally located primary water storage tank and diesel fuel oil tank which are required for the safe shutdown of the plant were not initially designed to withstand tornadic wind loads and do not have tornado-protected backup sources.

The applicant has revised the design of these tanks such that they are now capable of withstanding tornado wind loads.

The installation of the additional structural members will be completed prior to fuel loading.

We conclude that the plant is adequately protected against the effects of tornadoes.

On the basis of our review and that of our seismic design consultant, we conclude that the Class I structures and components of the H.B.

Robinson plant should adequately resist all applicable loads and are acceptable.

Potential missiles may be generated from the failure of the turbine of the turbine-generator.

The applicant has located and protected Class I equipment to reduce the potential for missile damage to equip-ment which would affect the capability to maintain containment integrity or to shut down the plant safely.

As indicated in the ACRS letter, the applicant stated that a second completely independent turbine speed control system would be provided to reduce the probability of an overspeed condition which might result in turbine failure.

This system will be designed to the requirements of the IEEE-279 criteria for protection systems.

The Technical Specifications require periodic testing of the overspeed devices to assure operability, and periodic inspections of the turbine to detect possible flaws.

We conclude that the applicant has made appropriate provisions to reduce the probability of a destructive turbine missile from being generated and affecting Class I equipment.

3.5 Emergency Core Cooling Systems The emergency core cooling systems for this plant are the same as those provided for the Ginna plant.

The major equipment consists of (a) three 50% capacity safety injection system pumps, (b) two 100% capacity residual heat removal pumps, (c) a boron injection tank, and (d) three 50% capacity accumulators.

The system pro-vides capability to inject borated cooling water rapidly into the core in the event of a loss-of-coolant accident.

The applicant's analysis of the performance of these systems for the loss-of-coolant accident is based on an assumed peak linear power generation of 19.l kW per foot, which corresponds to 102%

of the ultimate power rating of 2300 MWt.

For the case of an accident initiated by a double-ended break in cold leg coolant piping, a maximum temperature of the fuel clad of 2450°F was predicted, using previously accepted calculational models.

Based on the present state of knowledge of the behavior of the zircaloy clad at high tem-peratures and the results of experimental programs underway, we conclude that if the clad were to reach this temperature the threshold for significant zircaloy-water reaction would be exceeded.

Further, the ability of the cooling system to arrest clad temperature rise at this level has not yet been demonstrated experimentally, and extrapolation of current data into this high temperature region is uncertain.

At a peak linear power of 18.3 kW per foot, corresponding to 102%

of the rated power level of 2200 MWt, the predicted maximum clad temperature based on the previously approved model would be only 2280°F which is comparable to that we have approved in recent construction permit applications.

For operation at the licensed power level of 2200 MWt, we conclude that for all pipe breaks up to and including the instantaneous double-ended rupture of the largest reactor coolant pipe, the Robinson emergency core cooling systems will (a) limit the peak clad temperature to well below the clad melting temperature, (b) limit the fuel clad-water reaction to less than 1% of the total clad mass, (c) terminate the clad temperature transient before the geometry necessary for cooling is lost and before the clad is so embrittled as to fail upon quenching, and (d) reduce the core temperature and then maintain core and coolant temperature levels in a subcooled condition until accident recovery operations can be accomplished.

We conclude that the emergency core cooling systems are acceptable and will provide adequate protection for the loss-of-coolant accident at the licensed power level of 2200 MWt.

3.6 Containment Cooling System To control the containment pressure and temperature following a loss-of-coolant accident, two independent heat removal systems are provided.

Each system acting alone at its rated capacity can prevent overpressurization of the containment structure.

The two syste11lS are the containment spray system and the fan cooling system, and both are similar to those provided in the Ginna plant and other recently reviewed pressurized water reactors.

The containment spray system consists of two 50% capacity spray pumps and headers capable of limiting the containment post-accident pressure to below design pressure.

The applicant has proposed the use of sodium hydroxide and boric acid as additives to the spray.

We have reviewed the use of these chemical spray additives in terms of their iodine removal capabilities and in addition have evaluated the chemical capability of the spray solution with other reactor components.

As a result of our review, we conclude that the alkaline spray solution should effectively remove iodine from the containment atmosphere and that corrosion of other materials used in the contairuuent does not introduce a significant hazard.

We therefore find the system acceptable.

The containment fan cooling system provides complete redundancy to the containment spray system for heat removal from the containment atmosphere during post-accident conditions.

Four 25% capacity fan coolers are provided.

Since the fan coolers are located within containment, they must be capable of operating in the post-accident environment.

Westinghouse is conducting a component environmental test program to demonstrate this capability.

Our review of these tests, including the heat removal capability of the heat exchangers, environmental and radiation testing of the fan cooler motor, valve motor operators and electric cabling indicates that these components will satisfactorily function in the accident environment.

We conclude that there is reasonable assurance that this system will operate as proposed during a loss-of-coolant accident.

3.7 Protection System Instrumentation The protection system instrumentation for the Robinson plant is the same as that installed at the Ginna plant.

The adequacy of the protection system instrumentation was evaluated by comparison with the Conmission's proposed General Design Criteria published on July 11, 1967, and the Proposed IEEE Criteria for Nuclear Power Plant Protection Systems (IEEE-279), dated August 28, 1968.

On the basis of our review of the schematic diagrams, we conclude that the Reactor Trip System, the Containment Spray System, the Fan Cooler System, the Containment Isolation System, the Feedwater Isolation System, and the Steamline Isolation System are acceptable.

The basic design has been reviewed extensively in the past; con-sequently, the following sections of this evaluation discuss only those items which are unique to Robinson, for which new information has been received or which have remained as continuing areas of concern during this and prior reviews of similar plants.

3.7.1 Safety Injection System Instrumentation One deficiency was observed in the instrumentation for the safety injection system.

This system involves three safety injection pumps, two of which are respectively assigned to one of the two emergency buses.

The third is a comnon pump which is automatically connected to the preferred bus, or to the alternate bus should the preferred bus fail.

Our review indicated that the interlock system which

• prevents simultaneous closure of both co111Don PlDIIP breakers could be compromised by a single failure.

Such a failure would result in the momentary paralleling of the two independent diesel generators, which might lead to loss of all onsite power.

The applicant revised the design to correct this problem.

We have reviewed the revised design and it is acceptable.

3.7.2 Cable Installations We have reviewed the applicant's cable installation criteria relating to the preservation of the independence of redundant channels by means of separation and to the prevention of cable fires through de-rating and proper cable tray loading.

We have found these criteria and the resultant cable layout to be acceptable.

3.7.3 Engineered Safety Feature Manual Actuation Panel The control switches for manual actuation of the engineered safety features are mounted on a panel in the control room.

Since there is no electrical isolation between these switches and the automatic portions of the circuitry for the engineered safety features, it is important that redundant components and wiring at the panel be kept physically separate.

We have reviewed the layout of the panel and conclude that the design is acceptable.

3.7.4 Environmental Testing Westinghouse has conducted an environmental test program for the instrumentation and controls located inside containment which must actuate following a loss-of-coolant accident in the accident environ-ment.

We have reviewed the results of this testing progra* and conclude that the essential instrUDlentation and controls should function properly in the accident environment.

3.7.5 Seismic Testing We have reviewed the applicant's program for evaluating the seismic response of the protection system instrumentation and associated electrical equipment.

On the basis of our review, we find that adequate testing has been performed on the nuclear instrumentation, switchgear, and process system instrumentation.

3.7.6 Testing of Engineered Safety Feature Initiating Circuits We have reviewed the capability for testing engineered safety feature circuits during reactor operation. Resistance tests of the master and slave relay coils is the only method of routinely determining their operability.

The circuits upstream of these relays can be partially tested during operation.

During plant shutdown, the circuits can be tested completely by coincident tripping of instrument channels and the consequent operation of the master and slave relays and the entire downstream initiating system.

We have concluded that this testing capability is acceptable for the Robinson plant.

3.7.7 Testing of the Reactor Coolant System Temperature Detectors The reactor coolant system temperature sensors (resistance temperature detectors) are installed in coolant loop bypass lines.

Based on previous operating experience, Westinghouse has modified their desi~n for the utilization of resistance temperature detectors in bypass lines to permit replacement of any defective temperature elements while the plant is in the hot shutdown condition without requiring the draining of all the reactor coolant loops.

Bypass loops for the hot leg resistance temperature detectors are provided across the steam generators and across the reactor coolant pumps for the cold leg resistance temperature detectors.

A conmon flow circuit with alarm and status light for each reactor coolant bypass loop is provided in the control room.

Local flow indications are provided inside containment.

Westinghouse will perform tests to determine the adequacy of the design of the bypass loops.

These tests will determine the accuracy, repro-ducibility, and tiae response of the temperature measurements.

The tests will be performed on the H.B. Robinson plant.

We conclude that the successful performance of these tests will verify the adequacy of the bypass loop design.

3.7.8 Multiple Setpoints, Overpower -- Overtemperature Channels The overpower and overtemperature channels measure reactor coolant loop temperature difference and automatically vary the setpoints as a function of several other parameters, such as pressure and neutron flux.

When the plant is operated with one loop out of service, the physical characteristics of the reactor system require overtemperature setpoints which are more restrictive than for full flow if all other conditions are the same.

The applicant proposed to make these instrument adjustments manually and stated that only five adjust-ments were required.

We have studied the applicant's proposed design and conclude that if more restrictive setpoints are required for safety, the requirement for manual adjustment is not in accordance with Section 4.15 of IEEE-279 which requires a positive means of assuring that the more restrictive setpoints are used.

Section 4.15 further requires that the devices used to prevent improper use of less restrictive setpoints shall be considered a part of the protection system and shall be designed in accordance with the single failure criterion.

The applicant does not propose to revise the design of the over-temperature instrumentation to include provisions to provide that the removal of a pump from service would automatically act on the protection system such that the more restrictive setpoints are placed in force.

In its report, the ACRS indicated that unless design limits would not be exceeded with fewer than three loops in operation, such operation should be prohibited.

We have concluded that operation with fewer than three loops in service can be permitted only at a reduced power level so that the effects of anticipated transients would not result in fuel failures.

Operating limits in the Technical Specifications and automatic power trips will prohibit partial pump operation unless power is reduced to a level at which resetting of the overtemperature setpoints is not required.

3.7.9 Systematic Failures As stated in the ACRS letter studies are underway by the applicant of means of preventing comnon failure modes from negating scram action and of design features to make tolerable the consequences of failure to scram during anticipated transients.

These matters will be reviewed by us as the results of these studies are submitted.

3.8 Emergency Power System 3.8.1 Offsite Power Offsite power is available from nine transmission lines which converge at the site via several rights-of-way.

Four lines terminate at the unit No. 2 230kV switchyard; the remaining lines terminate at the unit No. l llOkV switchyard.

The switchyards are connected by a 300 MVA autotransformer.

A single startup transformer carries the power to the plant emergency buses. Stability studies by the applicant indicate that the sudden loss of the plant generating capacity at full power will not result in loss of the external grid.

Our review indicates that the only portion of the offsite system vulner-able to the random failure of a single component (as opposed to a cascading black-out), is the single startup transformer.

A failure of this transformer during normal reactor operation would not cause an accident.

However, should a loss-of-coolant accident occur following transformer failure, the engineered safety features would be dependent on the redundant diesel generators for power until disconnect links could be removed from the station generator to permit the backfeeding of offsite power to the auxiliary electrical system through the station main transformer.

This would require about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to accomplish.

We find this design to be acceptable because of the inherent reliability of such transformers (approximately 50 to 100 years between failures),

• and since additional operating restrictions can be imposed should the transformer be out of service.

The Technical Specifications require that, at the time of transformer failure, the applicant notify the AEC of the precautions that are being taken and of plans for the restoration of incoming power.

Based on our review, we conclude that the offsite power system is acceptable.

3.8.2 Onsite Power Onsite power is supplied by two redundant, independent diesel generator sets connected in a split-bus configuration; i.e., there is no automatic closure of tie breakers between the redundant buses to which the generators are connected.

The generators are rated for continuous operation at 2500kW and for operation at 2750kW for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any 24-hour period.

Each set is started automatically on a safety injection signal or upon undervoltage on its corresponding 480-volt emergency bus.

They are housed in separate rooms in the reactor auxilary building.

Our review indicated that the two starting circuits were not independent since the starting of either diesel would, after a prescribed time interval, be interrupted by successful operation of the other.

The purpose of the interlocks was to conserve starting energy.

We concluded that the possibility of a comuon mode failure more than offset this advantage.

The applicant has revised these interlocks so that the starting circuits are now independent.

The revised design is acceptable.

The maximum design engineered safety feature load per generator is 2493kW which will occur only if high containment pressure is experienced during the high head recirculation phase, thus initiating the containment spray pumps.

Otherwise, the design loads will never exceed 2329kW (during the injection phase). If the 2493kW load is required, this demand will be within the continuous rating and 257kW below the 2-hour rating.

This margin is acceptable.

Onsite fuel storage capacity is sufficient for a minimum of seven days operation of required safety feature loads.

This capacity is acceptable.

Our review of the de system indicated that it is compatible with the split-bus ac system:

the load shedding, load connecting, and supply breakers at the emergency buses are each controlled from one of the two redundant de buses.

One battery and charger are assigned to each de bus.

There are no automatic cross-ties between redundant de circuits, a provision which prevents a fault in one circuit from being automatically propagated to the other.

The two de systems are,

- therefore, electrically independent and, in conjunction with the diesel generators, provide a completely split onsite power system.

The battery banks and chargers, however, are located in the same room.

Because of the potential for a coUIJIOn mode failure to the de system, we required additional protection to assure independence of the battery sets. The applicant has revised his design to include the installation of a steel barrier between the battery banks to achieve an acceptable degree of physical independence.

The barrier is 5 feet high and is open at the top, bottom, and each end to facilitate ventilation.

We have concluded that the applicant's design is acceptable for this plant since it provides redundant de sources, electrical independence (split-bus) and adequate physical protection of each battery against fires and missiles originating in the vicinity of the other battery.

3.9 Auxiliary Systems The auxiliary systems necessary to assure safe plant shutdown include (1) the chemical and volume control system, (2) the residual heat removal system, (3) the auxiliary coolant system, and (4) the service water system.

The systems necessary to assure adequate cooling for spent fuel include (1) the spent fuel pit fuel cooling system, (2) the fuel handling system, and (3) the service water system.

These systems are the same as those we reviewed and found acceptable for the Ginna plant.

3.9.1 Chemical and Volume Control System The chemical and volume control system (a) adjusts the concentration of boric acid for reactivity control, (b) maintains the proper reactor coolant inventory and water quality for corrosion and radioactivity control, and (c) provides the required seal water flow to the reactor coolant pumps.

The amount of boric acid to be added to the core for reactivity control is determined by the operator.

The addition of unborated water as a result of operator error could result in an inadvertent reactivity increase.

The applicant has analyzed the possibility of inadvertent dilution during refueling, reactor startup and power operations.

The applicant's analysis indicated that there is ample time for the operator to recognize this inadvertent dilution and take corrective action.

Continuous monitoring of the reactor coolant boron concentration is desirable.

However, a suitable device has not been completely developed.

Tests of experimental devices are being conducted by Westinghouse.

The applicant is following these tests closely and will evaluate the feasibility of installing such a monitor when developed.

Our review of the safety-related aspects of the design of the system indicates that the design conforms to current practice and we find it acceptable.

3.9.2 Residual Heat Removal System The residual heat removal system removes core decay heat after the reactor has been shut down and cooled below 350°F.

The system also provides emergency cooling in the event of a loss-of-coolant accident.

On the basis of our review of the system design and functional requirements, we conclude that it is acceptable.

3.9.3 Auxiliary Coolant System The auxilary coolant system is a closed system that transfers heat to the service water system from the chemical and volume control system, residual heat removal system and other systems during normal plant operations.

In the event of a loss-of-coolant accident, the system also provides cooling to the residual heat removal system and the emergency core cooling system pumps.

The system is designed with redm1dancy in heat exchangers and pumps; however, it is not desi~ned against nonisolable failures of comnon headers.

Since the system provides cooling to the high head safety injection pumps, the residual decay heat removal pumps and the con-tainment spray pumps, such a failure following an accident might affect the operation of these pumps.

The Technical Specifications require periodic hydrostatic testing of the comnon headers to further reduce the probability of a failure occurring in these headers.

We conclude that the sysLem is acceptable.

3.9.4 Spent Fuel Pit Cooling System The spent fuel pit cooling system removes heat from the stored spent fuel elements.

The capacity of the system is sufficient to remove the heat generated by storage of one and one-third reactor cores.

The piping is arranged so that a failure of any one pipe will not drain the water in the pool below the top of the fuel elements.

Normally the water temperature is maintained at 120°F, but for the maximum heat generation rate, the equilibrium temperature would be 134°F.

If one and one-third cores are in storage and the system fails, the water temperature would rise to 180°F in 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />.

This indicates that there is sufficient time in which to correct any failure or provide alternate cooling before excessive fuel temperatures are reached.

We conclude that the spent fuel cooling system is acceptable.

3.9.5 Fuel Handling System The fuel handling system is designed to transfer spent fuel to the storage pool, and to transport and provide storage for new fuel.

The base slab of the spent fuel pit is not designed to withstand the dropping of a fuel cask.

Such an event could cause failure of the slab with a resulting loss of water in the spent fuel pit. The spent fuel pit lifting crane has been provided with an additional auxiliary safety device consisting of safety cables and wedging devices which will prevent the cask from falling in the event of failure of the main hoist cables.

The crane rails extend into the refueling building only far enough to position the crane above the fuel cask handling area.

The cask will not be handled above the spent fuel storage racks.

We have concluded that this design is acceptable for this plant.

3.9.6 Service Water System The service water system supplies cooling water to the auxiliary coolant system, diesel generators, containment coolers, compartment coolers housing the residual heat removal equipment, auxiliary feed pumps and some power conversion equipment.

The system is designed to Class I seismic criteria with the exception of that part serving the power conversion equipment located in the turbine building.

That part of the system not designed to Class I criteria can be remotely isolated with redundant isolation valves.

Red\\U\\dancy is provided for both active and passive components for accident conditions.

The applicant has evaluated the abi1ity to shut down the plant in the event this system should not be available as a result of a failure of the Lake Robinson dam.

The applicant's analysis shows that:

(a) heat can be removed from the containment by circulating outside air through containment with the ventilation system, (b) reactor decay heat can be removed by operation of the steam driven auxiliary feed pump without external seal cooling water, (c) sufficient condensate storage tank capacity and well water is available to cool down the plant, and (d) the safety injection pumps can be operated for the time required to borate the system without seal cooling water.

The critical time for berating is 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown because of the xenon decay time.

We conclude that the system is acceptable and that there is capa-bility to maintain the core in a safe condition without the service water system in the event of a failure of Lake Robinson dam.

4.0 Accident Analyses The applicant has submitted analyses of the consequences of various transients and postulated accidents.

We have independently evaluated the consequences of those accidents which could result in significant offsite doses.

We have evaluated (1) the control rod ejection accident, (2) the steam generator tube rupture accident, (3) the steam line break accident, (4) the failure of a waste gas decay tank, (5) the fuel handling accident, and (6) the loss-of-coolant accident in the same manner as on previous applications.

Our review of the applicant's analysis of the rod ejection accident was based on a comparison with similar analyses and parametric studies performed for the Ginna plant and other reactors.

The calculated peak fuel enthalpy values are less than our present estimate of the threshold for significant fuel failure or pressure generation.

We have concluded that, based on our present knowledge of fuel failure mechanisms, the results for this reactor indicate adequate margins of safety for the rod ejection accident.

We have determined that the Technical Specification limits on specific activity in the primary and secondary coolants will limit the calculated potential doses which would result from the failure of a steam generator tube and from a steam-line-break accident to significantly less than the guideline values of 10 CFR Part 100.

We calculated the doses which would result from the instantaneous failure of a waste gas decay tank which contains the noble gases from one primary coolant volume consistent with the Technical Specification limits on the radioactivity contained in the waste gas decay tanks.

The calculated doses are well below the guideline values of Part 100.

The results of our evaluations of the loss-of-coolant accident, and the fuel handling accident, are summarized in the following sections.

4.1 Loss-of-Coolant Accident In calculating the consequences of the design basis loss-of-coolant accident, we assumed that the fractions of the total core fission product inventory released from the core are those suggested in the AEC Technical Information Document 14844 "Calculation of Distance Factors for Power and Test Reactor Sites," i.e., 100% of the noble gases, 50% of the halogens and 1% of the solids. In addition, 50% of the halogens released from the core are assumed to plate out onto internal surfaces of the containment building or onto internal core parts.

Of the iodine available for leakage, 10% was assumed to be in the form of organic iodides and 5% was assumed to be in particulate form.

A containment spray removal constant for inorganic iodine of 2.7 per hour was assumed.

We assumed that organic iodides and particulates were not removed by the spray.

We assumed a design containment leak rate of 0.1% per day (neglecting penetration pressurization systems and containment isolation valve seal water system) for the first day, and one-half of this value thereafter.

We assumed groWld release with Pasquill Type F meteorological diffusion parameters, and a wind speed of 1 meter per second.

For prior opera-tion at the rated power level of 2200 MWt, we calculate a potential two-hour thyroid dose at the site boundary of 280 Rem and a whole body dose of 4 Rem.

At the outer boundary of the low population zone, we calculate a thyroid dose of 35 Rem and a whole body dose of less than one Rem for the course of the accident.

The calculated radiological doses at the ultimate thermal power level of 2300 MWt are approximately 5% higher, but are still within the ~uide-line values of Part 100.

However, as discussed previously, we have not conclude<l that operation at 2300 MWt is acceptable.

In addition to the radiolo~ical consequences of an assumed loss-of-coolant accident, the potential consequences of radiolytic decomposition of water have been considered.

Such decomposition would result in the production of gaseous hydrogen and oxygen in the containment atmosphere.

If sufficient hydrogen and oxygen are produced by such a reaction, it is possible that a flammable mixture could be attained in the contain-ment that if ignited would introduce an additional source of energy into the containment system.

An analysis(!) has been made in support of a proposal to purge the con-tainment atmosphere to control the concentration of hydrogen evolved following a loss-of-coolant accident.

The analysis considers the production of hydrogen by:

a zirconium-water reaction involving 2%

of the fuel cladding, radiolysis of the coolant, and the corrosion of aluminum components by the hydroxide spray.

The results indicate that with intermittent daily pur~ing initiated when the hydrogen concentration in the containment reaches 3 percent by volume, the predicted offsite doses resulting from the purging would be within 10 CFR Part 20 limits on release of radioactive effluent to the environment.

Our more conservative evaluation predicts higher calculated doses, although they are well within 10 CFR Part 100 guideline values.

The proposed purge system would not be desi~ned to Class I seismic standards beyond the existing Class I containment isolation valves.

Use of the instrument air compressors is proposed in order to raise the containment pressure to approximately 2 psig to facilitate a (1) WCAP-7372, Control of the Hydrogen Concentration Following a Loss-of-Coolant Accident by Containment venting for the H.B. Robinson Plant, November 1969 low, controlled, purge flow rate (about 240 cfm/day) through the purge exhaust system.

A filter assembly is proposed for the purge system but no credit is claimed for iodine filtration in the applicant's analyses of the purge doses.

On the basis of the foregoing, it appears that with suitable iodine filters installed in the purge exhaust, the incremental offsite dose resulting from purging could be limited to a small fraction of the 10 CFR 100 guidelines.

We have stated to the applicant that his efforts to investigate ways and means for post accident hydrogen control should not be restricted to purging.

We will continue our evaluation of additional data as they become available and will require the applicant to take such action as deemed necessary to control the concentration of hydrogen in the contairunent.

4.2 Fuel Handling Accident A fuel handling accident may occur as a result of dropping a fuel assembly onto the floor of the spent fuel pit or fuel transfer canal.

We have evaluated the refueling accident assuming that (a) all 204 rods in a dropped bundle are damaged, (b) the accident occurs 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> after shutdown, (c) 20% of the noble gases and 10% of the iodines in the dropped bundle are released, (d) 90% of the released iodines are retained in the refueling water, and (e) the same meteorological conditions exist as were assumed for the loss-of-coolant accident.

The resultant calculated doses exceed the guideline values of Part 100 for the short term thyroid dose at the site boundary for a fuel handling accident occurring in the spent fuel pit.

We are currently reviewing information provided recently by Westinghouse in connection with our Point Beach Plant evaluation regarding assumptions on gap activity and retention of iodine in the water.

The applicant has agreed that irradiated fuel will not be handled outside the containment until the differences in our assumptions are resolved and we can agree that the radiological doses that result from the accident are well within the guideline values of Part 100.

Since the differences in model assumptions also influence the ~alculated consequences of fuel handling accidents occurring within containment, i.e., dropping a fuel assembly onto the floor of the fuel transfer canal, the applicant has further agreed to maintain the isolation valves in the ventilation systems closed if inspection of irradiated fuel is required within containment, prior to the resolution of this matter.

5.0 Emergency Planning The applicant has prepared a site emergency plan for dealing with incidents involving radioactive materials, fires and other con-tingencies.

Immediate response to abnormal situations will be provided by an Emergency Team composed of on-duty shift personnel.

It is intended that action of this ~roup will include: protection of personnel by evacuation and rescue, termination of the incident, establishment of control over affected areas, protection of facility equipment and preliminary data collection.

Support for the Emergency Team will be available from the Control Group which is composed of senior plant managerial and supervisory personnel and their associated staffs.

The Control Group will direct and advise primarily during the post-incident phase of the emergency.

This group may also be supplemented by members of outside support agencies as determined by the nature and severity of the emergency.

In the event of an emergency which involves areas beyond the juris-diction of the applicant, arrangements have been made to obtain the assistance of various local, state and federal agencies, These support groups will establish road blocks, perform radiation monitoring work and institute other applicable protective measures, Medical support arrangements have been made with several physicians and with Byerly Hospital for the treatment of injured and/or con-taminated personnel.

Other medical assistance and transportation arrangements are available from the Hartsville and Lake Robinson Rescue Squads.

We conclude that the applicant's Emergency Plan is acceptable.

6.0 Conduct of Operations The applicant proposes operation of Unit No. 2 in conjunction with the existing fossil station, Unit No. 1, located adjacent to the nuclear plant.

The present Unit No. 1 staff of 47 personnel will be augmented by an additional 36 members to provide a total staff of 83 personnel for combined plant operation.

This staff is functionally divided into three groups reporting to the plant superintendent.

The operations group consists of approximately 40 personnel who perform duties in an operations and service section, For normal operation, the applicant proposed five five-man shifts for dual unit staffing from a co11111on control room.

Each shift would con-sist of a Foreman, Control Operator and Auxiliary Operator assigned to each unit.

The Shift Foreman, who is responsible for operation of both units would be qualified as a Senior Licensed Operator.

The Control Operator assi~ned to Unit No. 2 would hold an operator license, Of these five men, only three per shift could be assumed to be qualified and available for nuclear ~lant operations. This crew size represents the smallest yet proposed for operation of a plant of this rating.

Our analysis of this proposal, including discussions with the applicant and taking into account the advantages of the compact control room layout, indicated the need for an additional licensed operator for assignment to Unit No. 2 at all times and one additional auxiliary operator during cold plant start-ups.

The additional operators are needed to provide adequate shift manpower to handle abnormal operating conditions and other station emergencies, and to minimize the necessity of the Foreman being required to per-form operator functions.

We have taken into account the special design features provided by the applicant to reduce crew size requirements in reaching our conclusion that a 4-man crew would be acceptable for Robinson Unit No. 2.

In addition, for all reactor start-ups, except for recovery from unscheduled reactor shutdowns for which the cause has been clearly identified and corrected, an additional shift member is to be provided.

The above requirements are reflected by appropriate Technical Specifications.

This matter will be reviewed again after an adequate period of satisfactory operation and a careful assessment of the crew size required for emergencies.

Our review of the qualifications of the plant staff indicates that of the nine persons filling the positions of Plant Superintendent, Operations Supervisor and Shift Foreman, only one, a Shift Foreman, has had prior nuclear operating experience, although all received training at the Saxton nuclear facility.

The applicant's operating staff has undergone both formal and on-the-job training programs and will be subject in the near future to AEC operator licensing examinations.

As indicated above, a nucleus of operating experience exists within those personnel who will fill control operator positions.

However, operating experience possessed by personnel in supervisory and management positions is limited.

During plant start-up and power escalation, additional operating experience will be obtained by the plant staff under the direction of vendor personnel.

We conclude that subject to satisfactory licensing of operator personnel and with the successful completion of plant start-up and power escalation, the applicant's operating staff will be qualified to operate Robinson Unit No. 2.

Over-all plant maintenance activities will be performed by a force of approximately 28 personnel divided into two sections, each headed by a foreman reporting to the Maintenance Supervisor.

Activities relating to maintenance of mechanical and electrical systems will be performed by nine mechanics and five helpers.

A Foreman and nine Instrumentation and Control Technicians will perform calibration and maintenance work on the plant instru-mentation systems.

Although not highly experienced in nuclear plant maintenance, the key maintenance personnel have been actively participating in check-out and preoperational test operations and

will gain further experience during the start-up program with the assistance of the vendor's specialists.

We conclude that the plant maintenance capability is acceptable.

An on-site F.ngineering Group consisting of nine members is proposed to provide technical support and radiation protection services for the plant operating staff. Radiation protection activities will be performed by a Foreman and five technicians who will report to the F.ngineering Supervisor.

Technical support personnel assigned to the plant staff consist of an F.ngineering Supervisor and two Senior F.ngineers.

In response to our request, the applicant will provide an additional resident individual at the plant site who will have a minimum of a Bachelor's degree in engineering or the physical sciences and two years' experience in such areas as reactor physics, core measurements, core heat transfer, and core physics testing experience. Additional technical support will be provided by

~e Senior Physicist and Senior Reactor Engineer, both of whom are assigned to the general office staff in Raleigh, North Carolina.

Other support will be available from outside consultants, as deemed necessary.

We conclude that personnel with sufficient technical experience will be available to identify complex problem areas and provide onsite technical support.

Technical support activities associated with all Carolina Power and Light Company nuclear plants (Robinson Unit No. 2 and Brunswick Units land 2) are provided by a corporate level Technical Services Staff. This support group consists of a director and seven staff positions in radiation control, radiochemistry, chemistry, reactor engineering, instrumentation and controls, environmental and civil engineering.

A senior physicist, assigned to the manager of nuclear generation, will monitor reactor performance, supervise start-up of new cores and direct periodic physics tests and refueling operations.

Although not an assigned member of the Technical Services staff, this individual is available for staff support in fuel management and core performance evaluations. All refueling operations will be performed under the direction of the Senior Physicist who will be detached from the Technical Services Staff and report directly to the Plant Superintendent. Consultants have been engaged to supplement in-house efforts in areas of their respective specialties.

In addition, Westinghouse technical assistance is available for other specialized fields of interest.

Westinghouse will participate in the start-up and initial operation of the plant and will continue to make available technical support to the Robinson staff throughout the operating lifetime of the facility.

On these bases, we conclude that adequate engineering capability will be available to support the applicant's operating staff.

The applicant has formed two connnittees for review, evaluation and audit of the Robinson plant.

An on-site plant safety committee has been established to provide regular review and evaluation of station operating practices which could affect safety or present potential safety hazards.

During start-up and initial operation, Westinghouse field-service start-up personnel will be represented on the committee.

An off-site company nuclear safety committee has been formed to ftmction as an independent technical advisory group to senior management.

As a group, the members will collectively provide expertise in nuclear engineering reactor physics, health physics and power plant operation.

We have reviewed the composition, reporting functions, and over-all competency of the applicant's review and audit committees and find them acceptable.

Based on the above considerations, we conclude that the applicant is technically qualified to operate the plant and has established

~ffective means for continuing review, evaluation, and improvement of plant operational safety.

7.0 Technical Specifications The applicant's proposed Technical Specifications were presented in Amendment 19.

We have reviewed these proposed Technical Specifica-tions in detail and have held numerous meetings with the applicant to discuss their contents.

Modifications to the proposed Technical Specifications submitted by the applicant were made to more clearly describe the allowable conditions for plant operation.

The Technical Specifications as formally approved by the regulatory staff are appended to the proposed operating license and a copy may be examined in the Commission's Public Document Room.

Included are sections covering safety limits and limiting safety system settings, limiting conditions for operation, surveillance requirements, design features and administrative controls.

Based upon our review, we conclude that normal plant operation within the limits of the Technical Specifica-tions will not result in potential off-site exposures in excess of 10 CFR Part 20 limits and that means are provided for keeping the release of radioactivity from the facility within ranges that may be considered as low as practicable. Furthermore, the limiting conditions of operation and surveillance requirements will assure that necessary engineered safety features to mitigate the consequences of unlikely accidents will be available.

8,0 Report of Advisory Committee on Reactor Safeguards The ACRS reported on the suitability of constructing the H.B. Robinson Plant at the proposed site in a letter dated February 17, 1967.

We consider that the applicant has been responsive to the recommendations of the ACRS as indicated in this letter, and we conclude that the matters raised have been satisfactorily resolved during the design and construction of H.B. Robinson Unit 2.

L The ACRS completed its review of the application for an operating license for H.B. Robinson Unit 2 during its 120th meeting held April 9-11, 1970.

A copy of the ACRS letter, dated April 16, 1970, is attached as Appendix A.

In its letter, the ACRS made several recol'IIDendations and noted several items.

These items have been considered in our evaluation and include:

seismic design analysis (Section 3.4), reactor operation in the event of an earthquake (Section 2.4), operating crew size (Section 6.0),

surveillance of pre-pressurized fuel (Section 3.1.2), containment proof testing and lea rate testing (Section 3.3.2), partial loop operation (Section 3.7.7), turbine missiles (Section 3.4), failed fuel monitor (Section 3.1) and the boron concentration monitor (Section 3.9.1).

Studies are being continued by the applicant and Westinghouse on means for preventing common mode failures and of features to make tolerable the consequences of fail ure to scram during anticipated transients (Section 3.7.9), the control of hydrogen in the containment (Section 4.2.1), and instrumentation to monitor the course of events following an accident.

These and other recommendations identified in previous ACRS reports will be implemented by the applicant as suitable approaches are developed.

The ACRS concluded in its letter that if due regard is given to the items mentioned above, H. B. Robinson Unit No. 2 can be operated at power levels up to 2200 MWt without undue risk to the health and safety of the public.

9.0 Comnon Defense and Security The application reflects that the activities to be conducted would be within the jurisdiction of the United States and that all of the directors and principal officers of the applicant are United States citzens.

The applicant is not owned, dominated or controlled by an alien, a foreign corporation of a foreign government.

The activities to be conducted do not involve any restricted data, but the applicant has a~reed to safeguard any such data whi '!: =ight become involved in accordance with the requirements of 10 CFR Pa rt 50.

The applicant will rely upon obtaining fuel as it is needed from sources of supply available for civilian purposes, so that no diversion r f special nuclear material for military purposes, is involved.

For these reasons and in the absence of any information to the contrary, we have found that the activities to be performed will not be inimical to the common defense and security.

10.0 Financial Qualifications We have reviewed the financial information presented in Amendment No. 16 dated February 18, 1970.

The funds necessary to meet the estimated average annual cost of $13 million for the first five years of operation will come from operating revenues.

Information contained in Amendment No. 16 indicates that these revenues will be ample to cover the estimated cost of operating this facility for five years as well as cover the estimated $6.5 million cost of permanently shutting down the facility and the additional costs of maintaining it in a safe condition, should such actions become necessary.

On the basis of our review of this information, we have concluded that Carolina Power and Light Company is financially qualified to engar,e in the activities which would be authorized by the provisional operating license.

Attached as Appendix B is a more detailed analysis of the financial qualifications of Carolina Power and Light Company.

11.0 CONCLUSION

Based on our evaluation of the application as set forth above, we have concluded that:

a.

Carolina Power & Light Company (the applicant) has submitted to the Commission all technical information required by Provisional Construction Permit No. CPPR-26, the Atomic Energy Act of 1954, as amended (the Act), and the rules and regulations of the Commission to complete the application for a construction permit and facility license dated July 12, 1966, as amended by Amendment Nos. 8 through 21 (the applica-tion);

b.

The construction of H.B. Robinson Unit 2 has proceeded, and there is reasonable assurance that it will be completed in conformity with Provisional Construction Permit No. CPPR-26, the application, as amended, the provisions of the Act, and the rules and regulations of the Connnission;

c.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regu-lations of the Commission;

d.

There is reasonable assurance (i) that the facility can be operated at steady state power levels up to 5 MWt and that, upon satisfactory completion of the seismic analysis of certain Class I piping and equipment as noted in Section 3.4, the facility can be operated at steady state power levels up to 2200 megawatts thermal, in accordance with the license without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission;

e.

The applicant is technically and financially qualified to engage in the activities authorized by the facility operating license, in accordance with the rules and regulations of the Comnission; and

f.

The issuance of the facility operating license will not be inimical to the common defense and security or to the health and safety of the public.

Before a facility operating license will be issued to the applicant, the facility must be completed in conformity with Provisional Con-struction Permit No. CPPR-26, the application, the provisions of the Act and the rules and regulations of the Corrrnission.

Completeness of construction will be assured by a Cot11Dission inspection of the facility prior to the issuance of the license.

The applicant, prior to issuance of the license, will be required to furnish to the Commission proof that financial protection has been obtained to satisfy the requirements of 10 CFR Part 140 of the Commis-sion's regulations.

May 18, 1970 Date Peter A. Morris, Director Division of Reactor Licensing APPENDIX A ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITCO STATES ATOMIC ENERGY COMMISSION WASHINGTON, 0.C. :!0545 Honorable Glenn T. Seaborg Chairman U. s. Atomic Energy Connnission Washington, D. c.

20545 April 16, 1970

Subject:

REPORT ON H.B. ROBINSON UNIT NO. 2

Dear Dr. Seaborg:

During its 120th meeting, April 9-11, 1970, the Advisory Committee on Reactor Safeguards completed its review of the application by the Carolina Power and Light Company for a license to operate the H.B. Robinson Unit No. 2 at power levels up to 2200 MWt.

During this review the project was considered at Subcommittee meetings held on January 21, 1970 at the plant site and on March 26, 1970 in Washington, D. C.

In the course of these meetings, the Committee had the benefit of discussion with representatives and consultants of the Carolina Power and Light Company, Westinghouse Electric Corporation, Ebasco Services Incorporated, and the AEC Regulatory Staff.

The Committee also had the benefit of the documents listed.

The Committee reported to you on the construction of this plant in its letter dated February 17, 1967.

The H.B. Robinson site is in northeastern South Carolina about 56 miles from Columbia, South Carolina and consists of more than 5,000 acres in-cluding Lake Robinson.

The minimum exclusion radius is 1400 feet and the nearest population center with more than 25,000 residents is Florence, South Carolina, approximately 25 miles to the southeast.

The nuclear steam supply system for the H.B. Robinson Unit No. 2 is the first of the three-loop Westinghouse line to be reviewed f~r operation.

The design features are similar to those of the Ginna plant, previously dis~ussed in the Committee's report to you dated May 15, 1969.

Th~ applicant is reviewing his seismic design calculations.

The r,!sults of this analysis and any corrective actions required should be reviewed by the Regulatory Staff prior to operation above 5 MWt.

Further study is required of the bases and means whereby decisions con-cerning reactor operation will be made in the event of an earthquake in the region of the site. This matter should be resolved in a manner satis-factory to the Regulatory Staff.

-.33-H~norable Glenn April 16, 1970 ThP npplicnnt proposes to operate Robinson Unit No. l (cnnl-firrd), and Robinson Unit No. 2 (nucll'nr) from one contt'l)} t"l)Om wi.th n cn*w of five, consisting of a (orem.'.ln (licl'nsed senior opt>rator), n lic-cnscu operator at the nuclear unit console,.'.ln unlicensed oper.'.ltor at the coal-fired unit console, and two auxiliary operators, one (licensed) responsible for the nuclear unit and the other for the coal-fired unit. It is the opinion of the Committee that the crew size proposed by the applicant for the nuclear unit is insufficient for safety during initial operation but might be found sufficient after an adequate period of satisfactory operation and a careful assessment of the crew size required for emergencies.

The applicant is using a partial loading of helium "pre-pressurized" fuel rods.

The Committee believes that some surveillance of the Robinson fuel at high burnup is appropriate, with regard to assuring the ability of fuel elements to maintain their integrity while undergoing anticipated operational transients near the end-of-life.

The applicant plans to conduct containment proof testing and leak rate test-ing, prior to initial operation.

Subsequently, he proposes leak rate test-ing only of each seam and penetration of the containment.

The Committee believes that periodic integrated leak rate tests should be performed until the Regulatory Staff is satisfied that the methods provided by the applicant assure the required leak tightness of the containment.

The Committee rec-ommends that further study be made of possible means to assure the continued structural integrity of the containment throughout the life of the reactor.

The applicant is currently studying the consequences of plant operation with less than three loops in service. Until it can be shown that no design limits are exceeded or that trip points will be reliably reset by automatic action, power operation with less than three loops in service should be prohibited.

The applicant stated that he would provide a second completely independent turbine speed control system designed to meet nuclear protection system criteria of redundancy, separation, and reliability to reduce the probability of an overspeed condition.

In addition, protection is to be provided in app~opriate areas against damage in the unlikely event of large missiles arising from failure of the turbine rotor or discs.

This matter should be res~lved in a manner satisfactory to the Regulatory Staff prior to or early in the operation of this plant.

As methods for continuous monitoring of boron concentration and a more de-finitive determination of gross failure of a fuel element are developed, consideration should be given to their implementation in this plant.

- Honorable Glenn April 16, 1970 Studies by the applicant are underway on the following problems identified in previous reports of the Committee:

(a)

A study of means of preventing common failure modes from negating scram action and of design features to make tolerable the con-sequence of failures to scram during anticipated transients.

(b)

Review of development of systems to control the buildup of hydrogen in the containment and of instrumentation to monitor the course of events in the unlikely event of a loss-of-coolant accident.

As solutions to these problems develop and are evaluated by the Regulatory Staff, appropriate action should be proposed and taken by the applicant on a reasonable time scale.

The proposed action should be reviewed by the ACRS.

Other problems relating to large water reactors which have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports should be dealt with appropriately by the Regulatory Staff and the applicant as suitable approaches are developed.

The Advisory Committee on Reactor Safeguards believes that, if due regard is given to the items mentioned above, and subject to satisfactory comple-tion of construction and pre-operational testing, there is reasonable assurance that the H.B. Robinson Unit No. 2 can be operated at power levels up to 2200 MWt without undue risk to the health and safety of the public.

References attached Sincerely yours,

~ l t ____

~ _

.... ~

t,..

--.1s __,0

~-, ~---..DJ-.._ -

Joseph M. Hendrie Chairman Honorable Glenn April 16, 19 70 References

1)

Carolin;i Power & Light Comp:my lt*tter dalf*lt July 16, 1Q68 tsmtg Rl'port on Incidence of CotTl)Slon on l'rt>strcssi.ng SteC'l TC'1Hlons

2)

Carolina Powc>r & Li~ht Company letter dated December 8, 1969 tsmtg Containment Design Report

3)

Carolina Power & Light Company letter dated February 18, 1970 - Responding to Fish and Wildlife Service comments on Proposed Environmental Monitoring Program

4)

Carolina Power & Light Company letter dated April 6, 1970 - Identifying the Program to develop and document the additional seismic analysis for Class I equipment and piping

5)

Amendment No. 8 to License Application (Final Safety Analysis Report-Volumes 1 > 2 and 3) dated November 20, 1968

6)

Amendment No. 9 to License Application (designated FSAR Amendment No. 1) dated September 4, 1969

7)

Amendment No. 10 to License Application (designated FSAR Amendment No. 2) dated October 27 1 1969

8)

Amendment No. 11 to License Application (designated FSAR Amendment No. 3) dated December 2, 1969

9)

Amendment No. 12 to License Application (designated FSAR Amendment No. 4) dated December 15, 1969

10)

Amendment No. 13 to License Application (designated FSAR Amendment No. 5) dated December 15, 1969

11)

Amendment No. 14 to License Application (designated FSAR Amendment No. 6) dated January 23, 1970

12)

Amendment No. 15 to License Application (designated FSAR Amendment No. 7) dated February 6, 1970

13)

Amendment No. 17 to License Application (designated FSAR Amendment No. 8) dated February 24, 1970

14)

Amendment No. 18 to License Application (designated FSAR Amendment No. 9) dated February 27, 1970

15)

Amendment No. 19 to License Application (designated FSAR Amendment No. 10) dated March 18, 1970

16)

Amendment No. 20 to License Application (designated FSAR Amendment No. 11)

March 24, 1970 APPENDIX B AEC REGULATORY STAFF'S EVALUATION OF THE FINANCIAL QUALIFICATIONS OF THE CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 This is with regard to Carolina Power and Light Company's request for a license to operate the H.B. Robinson Unit No. 2 nuclear facility.

In the hearing held in the spring of 1967 concerning the issuance of a pemit to construct a nuclear reactor to be known as the H.B.

Robinson Unit No. 2 and to be located at the applicant's site near Hartsville, South Carolina, we testified that Carolina Power and Light was financially qualified to construct the proposed facility and to assume responsibility for payment of charges for special nuclear material to be leased from the Commission for the first core.

Such testimony is incorporated herein by reference.

We have reviewed the financial information presented in Amendment No. 16 dated February 18, 1970, including the unaudited "Interim Financial Statements-December 31, 1969," and the Company's Annual Reports for 1968 and prior years.

Based on this information it is our opinion that Carolina Power and Light is financially qualified to operate the facility and to undertake the proposed activities.

The H.B. Robinson Unit No. 2 will be used to augment the applicant's present electrical generating capacity.

Accordingly, the total cost to operate the unit will come from operating revenues.

These costs are presently estimated by the applicant to total about $65,000,000 for the first five years of operations, or on the average of

$13,000,000 annually, and cover expenses for fuel, operations, main-tenance and insurance.

In addition, the cost of shutting down the H.B. Robinson facility and maintaining it in a safe condition, tf and when such an event becomes necessary, is estimated not to exceed 6.5 million dollars, based on shutdown measures comparable to those authorized for Hallam and on current dollar values.

The sources of revenue for a public utility such as Carolina Power and Light are characteristically stable and rates for the electricity produced are regulated by the respective states. Therefore, it is reasonable to assume that, aside from the si~nificant net earnings to be anticipated from the other generating facilities in the

~37-applicant's system, the revenues produced by the successful operation of this facility will be ample to cover the above-stated estimated cost of operating this reactor, as well as the safe shutdown and maintenance of the unit if such should become necessary.

Carolina Power and Light is well financed and has significant resources at its command.

Its current Dun and Bradstreet credit rating is AaAl and Moody's Investors Service rates the company's first mortgage bonds as Aa(high).

The pertinent financial data and ratios, computed from the 1969 interim statements and previous Annual Reports, reflect significant growth and indicate a sound financial position: the ratios are well in line with the corresponding composite ratios for the electric utility industry.

A copy of our financial analysis reflecting these ratios and pertinent financial data is attached.

. CAROLINA POWER AND LIGHT CXJMPANY Docket No. 50-261 Financial Analysis tong-term debt Utility plant (net)

Ratio - debt to fixed plant Utility plant (net)

Capitalization Ratio of net plant to capitalization Stockholders' equity Total assets Proprietary ratio Earnings available to common equity Common equity Rate of return on common equity Net income Stockholders' equity y

(dollar* in millions)

Calendar Year Ended Dec. 31 1969 1968 1967 309.0

$ 309.0

$ 269.0 673.5 594.9 537.3

.46

.52

.50 673.5 594.9 537.3 595.2 558.3 510.2 1.13 1.07 l.05 286.2 249.3 241.2 718.6 629.8 570.9

.40

.40

.42 24.4 23.0 22.8 226.8 189.9 181.8 10.ax 12.11 12.51 Rate of earnings on stockholders' investment 27.4 286.2 9.6l 26.0 249.3 10.441 25.0 241.2 10.41 Net income before interest Liabilities and capital Rate of earnings on total investment Net income before interest Interest on long-term debt No. of times fixed charges earned Net income Total revenues Net income ratio (percentage)

Operating expenses (including taxes)

Operating revenues Operating ratio Retatned earnings Earnings per share of Comnon 1969 41.4 718.6 5.81 41.4 14.5 2.9 27.4 187.1 14.6l 145.6 187.1

.78 62..5

$2.05 Capitalization as of 12/31:

u,ng-term debt

.t\\mount 1 of Total Preferred stock Common stock Tot~l Moody's Bond Ratings:

First Mortgage Dun and Bradstreet Credit Rating

!/ Unaudited interim statements.

$309.0 59.4 226.8

$595.2 51.9'1.

10.0 38.l 100.0%

Aa AaAl 37.1 629.8

5. 9-X.

37.1 12.3 3.0 26.0 170.0 15.3t 132.9 170.0

.78 55.5 32.9 570.9 5.St 32.9 9.7 3.4 25.0 146.8 11.oi 113.8 146.7

.78 48.7

$1.98

$1.91 1968 Amount t of Total

$309.0 59.4 189.9

$558.3 55.4'Z 10.6 34.0 100.07.