ML19290C376

SER Re Operation of Facility.Suppl 2

ML19290C376
Person / Time
Site:	North Anna
Issue date:	08/02/1976
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0053, NUREG-0053-S02, NUREG-53, NUREG-53-S2, NUDOCS 8001100805
Download: ML19290C376 (55)

Text

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Safety

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Evaluation Repori neo.i.,o,JoA;;a";

S related to operation of Office Nc ar North Anna Power Station Units 1 and 2 Docket No. 50-338 Virginia Electric and Power Company Augus'f,1976 Supplement No. 2 e

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Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $4.50; Microfiche $2.25

NUREG-0053, SUPP. 2 AUGUST 2, 1976 SUPPLEMENT NO. 2 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF VIRGINlA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION-UNITS I AND 2 00CKET NOS. 50-338 AND 50-339

TABLE OF CONTENTS PAGE I.0 INTRODUCTION AND GENERAL DISCUSSION 1-l

1.1 INTRODUCTION

l-1 2.0 SITE CHARACTERISTICS 2-1 2.4 HYDROLOGIC ENGINEERING 2-1 2.4.2 Flood Potential 2-1 2.4.4 Ground Water 2-2 2.5 GEOLOGY AND SEISMOLOGY 2-4 2.6 FOUNDATION ENGINEERING 2-11 2.6.2 Evaluation of Foundation Engineering 2-Il 2.6.3 Conclusions 2-15 3.0 DESIGN CRITERI A - STRUCTURES, SYSTEMS AND COMPONENTS 3-1 3.4 WATER LEVEL (FLOOD) DESIGN CRITERIA 3-1 5.0 REACTOR COOLANT SYSTEM 5-l 5.2 REACTOR COOLANT PRESSURE BOUNDARY 5-1 5.2.8 Overpressure Protection 5-1 6.0 ENGINEERED SAFETY FREATURES 6-1 6.3 EMERGENCY CORE COOLING SYSTEM 6-1 6.3.3 Performance Evaluation 6-1 6.3.3.1 Safety Related Electrical Equipment 6-1 6.3.3.2 Non-Safety Related Electrical Equipment 6-1 6.3.3.3 Electrical Penetration 6-2 6.3.3.4 Conclustons 6-2 10.0 STEAM AND POWER CONVERSION SYSTEM 10-1 10.2 MAIN STEAM SUPPLY SYSTEM 10-1 i

PAGE 11.0 RADIDACTIVE WASTE MANAGEMENT 11-1 11.1 SUNNARY DESCRIPTION ll-l 13.0 CONDUCT OF OPERATIONS 13-1 13.3 EMERGENCY PLANNING 13-1 14.0 INITIAL TESTS AND OPERATIONS 14-1

22.0 CONCLUSION

S 22-1 11

APPEND lCES PAGE APPENDIX A CONTINUATlON OF CHRON0 LOGY OF RADIOLOGICAL REVIEW A-l iii

LIST OF TABLES PAGE TABLE 2.1 GROUNDWATER LEVELS 2-3 TABLE II. I CALCULATED RELEASES OF RADI0 ACTIVE MATERI AL IN LIQUID EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS 1 AND 2 11-5 TABLE 11.2 CALCULATED RELEASES OF RADI0 ACTIVE MATERI AL IN LIQUsD EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS 3 AND 4 11-6 TABLE 11.3 CALCULATED RELEASES OF RADIOACTIVE MATERIAL IN GASEOUS EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS i AND 2 11-7 TABLE 11.4 CALCULATED RELEASES OF RADIOACTIVE MATERIAL IN GASEOUS EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS 3 AND 4 ll-8 TABLE 11.5 COMPARl5UN OF NORTH ANNA POWER STATION, UNITS NOS. I, 2, 3, AND 4, WITH APPENDIX l TO 10 CFR PART 50, SECTIONS ll.A, ll.B, AND ll.C (MAY 5, 1975) AND SECTIONS II.D, ANNEX

( SEPTEMBER 4, 1975) 11-11 iv

I.0 INTRODUCTION AND GENERAL DISCUSSION l.1 INTRODUCTION On June 4, 1976 the Nuclear Regulatory Commission (Commission) issued its Safety Evaluation Report regarding the application for licenses to operate the North Anna Power Station, Units I and 2 (North Anna facility, North Anna plants, f acility or plant). The application was filed by the Virginia Electric and Power Company (applicant).

Supplement No. I to the Safety Evaluation Report was issued on June 30, 1976. Supplement No. I to the Safety Evaluation Report documented the resolution of several outstanding items, and summarized the status of the remaining outstanding issues.

The purpose of this supplement is to update our Safety Evaluation Report (and Supplement No. I) by providing (l) our evaluation of additional Information submitted by the applicant since the issuance of Supplement No. I of the Safety Evaluation Report, (2) Information regarJing the current status of matters that were still under review and (3) additional information for those sections of the Safety Evaluation Report where further discussion or changes are in order.

Each section of this Supplement is numbered the same as the section of the Safety Evaluation Report, and is supplementary to and not in lieu of the discussion in the Safety Evaluation Report, except where I-l

specifically so noted. Appendix A is a conti.luation of the chronology of our principal actions related to the processing of the applic> tion.

A summary of the remaining outstanding issues is presented in Jection 22.0 of this supplement.

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2.0 SITE CHARACTERISTICS 2.4 HYDROLOGIC ENGINEERING 2.4.2 Flood Potential We stated in the Safety Evaluation Report that the applicant had recently advised us by letter, dated April 30, 1976, that it had been reevaluating the analysis for predicting the probable maximum flood and would submit a report on the results of this reanalysis by June 30, 1976.

We also stated that we would evaluate this report to-determine if our conclusions regarding protection against flood potentia are still valid and report our findings in a subsequent report, in Amendment 53 to the Final Safety Analysis Report the applicant submitted its revised probable maximum flood analysis. We have reviewed the applicant's revised probable maximum flood analysis and based on this review and our Independent analyses, we conclude that the applicant's revised estimate of the maximum Lake Anna water levei of 267.3 feet above mean sea level (including wave action) is conservative and meets the criteria outlined in Regulatory Guide 1.59

" Design Basis Floods for Nuclear Plants." This level is 3.7 feet Delow plant grade. We concur with the appileant that this level will have no safety impact on the plant, since service water is available from the service water reservoir, located 45 feet above plant grade.

The applicant also proposes to establish a technical specification to restrict facility operation when the water level in Lake Anna exceeds 2-1

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elevation 256 feet above mean sea level, to assure that pressure boundary f ailures within the circulating water system do not adversely af fect plant safety due to internal flooding.

We conclude that such a technical specification can be implemented to preclude flooding of safety related equipment. We therefore consider this issue resolved (see Section 3.4 of this report).

g 2.4.4 Groundwater Safety Anal c's Report, the applicant states in Amendment 54 to the Final i

that the dasign basis groundwater lovel for North Anna Power Station, Units I and 2 originally was elevation 256 feet mean sea level, but that as a result of recent rising groundwater levels, the maximum groundwater levels that could occur have been recomputed.

Table 2.1 lists specific locations and th* estimated maximum groundwater level that could occur at each location.

On the basis of our Independent evaluation, we conclude that these levels are conservative and therefore acceptable (See Section 3.4 of this rElort).

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TABLE 2.1 GROUNDWATER LEVELS Maximum Groundwater Level Location (feet above mean sea level)

Fuel Building 267 Auxillary Feedwater Pipe Tunnel 267 Auxillary Building 265 Main Steam Valve and Quench Spray 263 Pump House Unit i Main Steam Valve and Quench Spray 266 Pump House Unit 2 Safeguards Area Unit i 263 Safeguards Area Unit 2 267 Service Building 262 Buried Fuel Oil Tanks 270.5 waste Gas Decay Tank Enclosure 267 Contafrment Building Unit i 264 Containment Building Unit 2 266.5 2-3

2.5 GEOLOGY AND SEI Sh0 LOGY We stated in the Safety Evaluation Report that the applicant had filed a coport on the microselsmic data obtained with the Phase 2 network.

We also stated that we wert evaluating the data and would report our findings in a subsequent raport. We have now completed our review of this data, in May 1973, the applicant Informed us that a family of faults had been discovered in the foundations of the North Anna Power Station Units 3 and 4.

We immediately initiated an intensive review to determine whether these faults were capable within the meaning of Appendix A to 10 CFR Part 100. We decided early in our review that such a determination would require findings as to (1) whether the faults were capable as of the time Lake Anna, the cooling reservoir, was impounded and (2) If not, whether the presence of Lake Anna had reactivated the faults.

With respect to the latter consideration, af ter a thorough review of the pertinent professional literature and consultation with several scientists who were actively engaged in the study of artificial stimu-lation of earthquakes, we concluded that Lake Anna had not reactivated the faults. We based otr conclusion on the observations that no reservoir as shallow as Lake Anna located in a similar geologic environ-ment is known to have triggered an earthquake. We also determined that no earthquakes had been felt in the vicinity of the site since the impoundment of the lake, two years earlier.

Because both theory and 2-4

observation indicate that earthquakes triggered by reservoir loading tend to occur within a short time followlag impoundment, we considered the latter point to be a significant indication that Lake Anna had not triggered any eerthquake. Nevertheless, we concluded that confirmatory monitoring should be conducted.

We, therefore, required the applicant to implement a dense microearthquake monitoring network in order to verif y that there were no smali earthquakes occurring in assoclatIon with the faulting, in response to our requirements, the applicant proposed a two phase program. Phase I was to be essentIaliy a reconnaissance program utIlIzing five portable seismographs. These were to be located so as to best determine those basic characteristics of seismicity in the vicinity of the site which would be needed to design the denser Phase ll network.

With the one modification that seven rather than five Instruments be used in Phase 1, we accepted the program proposed by the applicant.

The seven station, Phase I network has been operational since January 21, 1974.

Installation of sixteen of the seventeen stations scheduled to comprise the Phase ll network was completed March 31, 1975.

Following equipment modification, check-out and calloration, these sixteen stations became fully operational April 10, 1975. The seventeenth has since been Installed.

During the monitoring period, we have required that VEPCO report the locations and magnitudes of all earthquakes having a magnitude greater than about 1.0 and close enough to the network that the difference 2-5

between the shear wave ano compressional wave travel times is less than five seconds.

The applicant has submitted a thorough analysis of the resulting microcarthquake data in a report titled " Summary Report on the In-Progress Seismic Monitcring Program at the North Anna Site, January 21, 1974 through May 1, 1976 " (May 1976 Report),it includes all locatable microearthquakes in addition to those larger than magnitude one and with the difference Detween the shear wave and compressional wave travel time less than five seconds as we had requested. Of the microearth-l quakes recorded only four were as large as approximately magnitude one within 10 kilometers of tha plant between January 1974 and May 1976.

Three have approximately the same epicenter. Both epicentral locations are close to the northeast boundary of Lake Anna.

Two other microearthquakes were located within 30 kilometers of the plant.

Extremely small microearthquakes (magnitude less than one) that could be located by the array, appear to cluster near the boundaries of Lake Anne.

Such microearthquake activity is not unexpected at new reservoirs,

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and can be detected only by sensitive monitoring equipment.

Because the microearthquakes are so small, not all seismographs are likely to record each earthquake, and consequently, focal plane mechanisms cannot be well determined, i f at all.

However, if clusters of micro-earthquakes are considered to have the same source, enough data points can be plotted to determine a combined focal plane solution.

Such solutions yield two alternative fault-strike, dip and slip-vector 2-6 4

o V

N.

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results. At places where such a determination was possible neither best-fit alternative solution agreed with the strike of faulting in the vicinity of the plant site or the north 35 east trend of Neuschel 's Lineament which is parallel to an extension of the 35 mile long Staf ford Fault Zone which approaches the site to within 25 miles from the northeast.

Micruearthquake activity in the North Anna Plant vicinity appears to be less than that predicted for all of central Virginia on the basis of the historic record for larger earthquakes and it poses no hazard in itself to the plant or dam impounding Lake Anna.

Operation of a network at the level required by the Nuclear Regulatory Commission has continued since December 1975.

Microcarthquake activity in the vicinity of Lake Anna appears bounded, or nearly so, to the northwest approximately coincident with an extrapolation of the plant site f aulting to the northeast, (see figures 7 and 30 of the May 1976 report).

There is no aspect of the data from the microearthquake network which indicates in any way that the microearthquakes recorded or that the faults through the plant site pose a hazard to the safe operation of the plant.

However, because there is a tenuous relationship between the limits of occurrence of microearthquakes and a boundary approximately defined by an extrapolation of the f aults in the vicinity of the plant site we require continued operation of the 17 station network for at least one year commencing August I, 1976, in order to 2-7 provide data 2-7

to determine if this relationship dissipates cr changes in such a way that additional action may be required. Appropriate reports regarding status of station operation (downtime, malfunctions, etc.) calibration and analysis of data are to be submitted by the utility to the Nuclear Regulatory Commission quarterly and a final report is to be submitted at the end of the monitoring period.

If this additional data indicates that additional action should be taken to insure safe operation of the f acility we will require that the applicant take any necessary actions.

In April 1976, the U. S. Geological Survey published two open file reports: " Detailed Investigations of a Coastal Plain-Piedmcnt Fault Contact in Northeastern Virginia," by Mixon & Newell. The structures described in these reports cansist of three en-echelon faults and a moncciine, which collectively compose the Stafford Fault Zone. They were mapped over a distance of 56 kilometers and displacements on the faults range from 15 to 60 meters. The closest known approach of this f aulting to the North Anna Site is approximately 25 miles.

The interpretation of the age of f aulting indicates that the major deformation occurred during Cretaceous and middle Tertiary with lessor movements extending into Late Tertiary or possibly Quaternary. Because the date of last movement was not adequately substantiated in these reports, the Nuclear Regulatory Commission requested Potomac Electric Power Company (PEPOD) to investigate further and determine what hazard, if any, these structures posed to PEPCO's proposed Douglas Point Nuclear Power Station.

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An investigation was carried out on PEP 00's behal f by Dames and Moore, incorporated. These findings were transmitted to the Nuclear Regulatory Commission on June 14, 1976, in a report entitled "Geoiogic Investigation of the Stafford Fault Zone." The report concluded:

"No evidence has been found during these studies to indicate surf ace or near surface movement along any of the investigated structures once within the past 35,000 years, or of a recurrent nature within the last 500,000 years. At present, no capable faults with which to associate the Stafford Fault Zone are known to exist within the central eastern

seaboard, in aadition, the zone demonstretes no recognizable' association with instrumental macrose i sm ic i ty.

Therefore, in consideration of the presently available infermation, it is concluded that neither the Stafford Fault Zone, nor any of its component structures, is capable according to NRC criteria as defined in Appendix A of 10 CFR 100."

The Nuclear Regulatory Commission has not yet taken a position on the age of last movement of the Stafford Fault Zone and is requesting additional evidence upon which to base a conclusion. We will report the results of our review in a subsequent supplement to the Safety Evaluation Report.

Mixon and Newell in their April 1976 report, suggest that the Stafford Fault System is related to Neuschel's lineament which is parallel to and on line with the Stafford Faults.

If so, the zone would be extended south an additional 25 to 30 miles and would pass within four or five miles of the North Anna site.

Neuschel 's l ineament was d iscovered by aeromagnetic and radiometric surveys. Although Neuschel postulated a basement f ault with the lineament, no evidence of f aulting was observed 2-9

v; hen the region was mapped by the U. S. Geological Survey.

This lineament has beer, described and considered previously in the North Anna f acility licensing proceeding (North Anna I and 2 show cause proceeding, LBP-74-49, June 27, 1974).

In that proceeding we concluded that the lineament, if it is a tectonic structure, developed during the Paleozoic and has no evidence of young movement or associated seismicity.

That conclusion is unchanged by subsequent data.

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2.6 FOUNDATION ENGINEERING 2.6.2 Evaluation of Foundation Engineering We stated in the Safety Evaluation Report that the applicant was conducting additional investigations (sampling, testing and analyses) to reassess the capability of the service water reservoir, dikes, and pump house to reliably perform their intended function, in addition to this work, the applicant had committed to provide additional information on the following items in an amendment to the Final Safety Analysis Report:

(1) A program to control groundwater levels under the pump house in order to reduce the potential for unexpected future settlement and to increase its resistance to seismic ef fects.

(2)

An assessment of foundation conditions and dike stability and a description of measures to be taken for remedial drainage features and stabilization or replacement, if warranted, to assure adequate resistance to seismic ef f ects.

(3) A program to monitor water inventory losses from the service water reservoir to assure that leakage rates from the reservoir do not reach levels that would affect the safe shutdown of the plant.

(4)

Results of a program to measure the settlement of all seismic Category I structures that has occurred since the start of construction, and a comparison of these measurements to predicted settlements and to allowable settlements.

(5) Results of a program to (a) inspect seismic Category I structures for settlement induced distress, particularly in areas of critical external penetrations such as safety related conduits and pipes; and (b) check settlement ef fects on the spray piping in the reservoir.

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We also stated that we would review the above information after it had been submitted by the applicant and would report our findings in a subsequent report.

With respect to item I, the applicant has committed to the installation of a system of well points to control groundwater levels under the service water pumphouse.

We find this an acceptable method of controlling groundwater level.

We have determined that the well point system does not need to be seismic Category I bel +;s) of the low permeability of the saprollte underlying the pumphouse. Before installation, the details of the system design will be required to be included in the Final Safety Analysis Report and reviewed and approved by the staf f to insure that it will fulfill its function without causing a loss of fines from the saprolite.

We will report the results of our review in a subsequent supplement to the Safety Evaluation Report.

With respect to item 2, the applicant has assessed the condition of the saprolite underlying critical sections of the dike which impounds the service water reservoir. The applicant has also performed laboratory tests on these materials and re-evaluated the stability of the dike.

The applicant has proposed a shallow trench drain to control groundwater levels below the dike.

The necessary reliability of a source of service water to safely shutdown the plant in the event of the design basis earthquake is based on the existence of the service water reservoir and Lake Anna.

These two sources are described in Section 2.4.1 of the Safety Evaluation Report. The 2-12

service water systems are described in Sections 9.2.1 and 9.2.3 of the Safety Evaluation Report. Our judgments regarding the reliability of the dikes impounding the service water reservoir are based on our review of the results of the applicant's investigations and, by inference, on the apparent reliability of an earth-fill dam which retains a second source of service water, Lake Anna.

There are two sources of service water for North Anna Power Station, Units I and 2.

We find that the dikes for the service water reservoir and North Anna dam together have an adoquate degree of stability, and resulting reliability, under the seismic ef fects of the proposed safe shutdown earthquake. Our review of the applicant's Information on

'he main dem on Lake Anna, and its foundations, Indicates that this dam has considerable seismic resistance and would likely survive the effects of the safe shutdown earthquake. We thus conclude that the foundations and earthworks features of these two service water sources combined have adequate reliability, under seismic conditions. We will require that the systems necessary to utilize the Lake Anna service water source be maintained and monitored to assure their functional reliability. This requirement will be included in the technical specifications.

With respect to item 3, the applicant has committed to a program to monitor, on a semi-annual basis, the water losses from the service water reservoir in order to assure that leakage rates will not affect the safe shutdown of the plant.

We find this acceptable.

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With respect to item 4, we asked the applicant to check the settlement and differential settlement of North Anna Power Station, Units I and 2, safety related structures because of the unexpected settlement of structures founded on saprofites.

The applicant reported the results of its measurements in Amendment 5' to the Final Safety Analysis Report. Aii reported settlements and differential settlement values were reasonable and within alloweble limits except for a few points along the crane rail at the northwest corner of the service building, an area occupied by the heating boiler room and a part of the emergency generator area.

The applicant has stated that there were no adverse affects to safety related equipment due to this settlement. We wil l review this matter and report the results of our review in a subsequent supplement to the Safety Evaluation Report.

The app l icant 's program for mon itoring f uture settlement wil l detect any settlement before hazardous conditions can develop.

With respect to item 5(a), the applicant has conducted an inspection of safety related structures to detect any signs of distress caused by differential settlement of these structures.

The applicant reported no signs of movements or distress to critical structures or pipes, although widened cracks along one wall of the service building were notea.

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With respect to item 5(b), the applicant has placed an extensive set of monuments to check the settlement of spray pond piping in the service water reservoir and has established criteria for acceptable settlement within the piping system.

We find this program adequate to detect the development of adverse pipe stresses.

2.6.3 Conclusions Subject to (1) our review and approval of the design of the system of well points and (2) resolution of the settlement of the northwest corner of service building, and based on information and commitments provided by the applicant, we find that the fouadations and earthworks for the North Anna Power Station, Units I and 2 are adequate to provide for the safe shutdown of this plant under the proposed normal and extreme environmental conditions. The proposed monitoring programs combined with adequate design and construction provides assurance of the continuing functional reliabili.'y of foundation features for this facility.

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3.0 DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS 3.4 WATER LEVEL (FLOOD) DESIGN CRITERIA We stated in the Safety Evaluation Report that (1) the applicant was r evaluating the analyses for predicting the probaDie maximum flood e

and would submit the results of that analysis, (2) we would determine the ef fects of this analysis on the design flood level and would assure that all seismic Category I structures are protected against the effects of the flood level, and (3) we would report our findings in a subsequent report.

As discussed in Section 2.4.2 of this report, the applicant has submitted a revised analysis for the probable maximum flood in Amendment 53 to the Final Safety Analysis Report, which predicts a maxirnum water level of 267.3 feet above mean sea level (including wave action). The finished plant grade is at 271 feet mean sea level. We have reviewed the effect of the revised probable maximum flood level on safety related structures and equipment. We have determined that the revised maximum water level does not af fect the design criteria of all the below grade structures and have concluded that all seismic Category I structures are protected against the ef fects of the revised ficod level.

The one area of concern related to safety related equipment involves a f ailure of the circe 8ating water system. Although the turbine building does not contain safety related equipment, there are three doors at the 3-1

bottom level of the turbine building which connect between It and the emergency switchgear rooms, control room air conditioning chiller rooms and the battery rooms. A f ailure of the system in the turbine building will automatically trip the circulating water pumps and inlet valves when the water level reaches one foot above the turbine building floor level.

The building design includes water tight barriers at these doors to an elevation of 257 feet mean sea level, or three feet above the turbine building floor.

The North Anna facIIIty technical specifications will have provisions for taking the f acIIIty out of service and closing the circulating water valves when the lake level exceeds 256 feet mean sea level. On the basis of our review, we find the facility design for protection of essential structures and equipment from the ef fects of the probable maximum flood acceptable.

We therefore consider this matter resolved.

As discussed in Section 2.4.4 of this report, the applicant has submitted a revised list of maximum groundwater levels that could occur at the North Anna facility.

In Amendment 54 to the Final Safety Analysis Report, the applicant submitted the results of a revised structural analysis reflecting these groundwater levels. We have reviewed the ef f ect of the revised groundwater levels on safety related structures and equipment.

Protection of structures is provided by water stops and subsurf ace drainings. We have determined that the results of a structural reanalysis for the revised groundwater levels indicate acceptable minimum factors of 3-2

safety against overturning and sliding in the event of a postulated safe shutdown earthquake.

Therefore, the stability of all seismic Category I structures will be assured in the event of a safe shutdown earthquake, in Amendment 42 to the Final Safety Analysis Report, the applicant has provided the results of an anal ysis which demonstrates that displacement of the underground diesel generation fuel tanks and piping due to groundwater buoyancy will not occur even for the most conservative assumption that the water table reaches grade elevation.

On the basis of our review, we find the facility design for protection of essential structures and equipment from the effects of groundwater acceptable.

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5.0 REACTOR COOLANT SYSTEM 5.2 REACTOR-COOLANT PRESSURE-BOUNDARY 5.2.8 Overpressure-Protection As a result of recent reviews, we have determined that several reported instances of reactor vessel overpressurization in Pressurized Water Reactors have occurred in which 10 CFR Part 50 Appendix G limitations have been exceeded. The majority of cases have occurred during cold shutdown in which tae reactor coolant system has been in water-solid conditions. We view each of these occurrences with concern and consider any violation of Appendix G to 10 CFR Part 50 limitations to be unde-sirable.

It is therefore considered appropriate that steps be taken to precl ude the possibility of f urther occurrences of reactor vessel overpressurization. The staff has discussed thir concern with the applicant as it relates to the Surry Power Station Units I and 2 (Docket Nos. 50-280 and 50-281). We intend to continue these discussions on the North Anna Power Station Units I and 2.

In addition we plan to review overpressurization incidents with Westinghouse Electric Corporation and the other pressurized water reactor vend >rs genericall y, and will report resolution of this item as it relates to th' North Anna Power Station Units I and 2 in a subsequent supplement to the Safety Evaluation Report.

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6.0 ENGINEERED SAFETY FEATURES 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.3 Per formance Evaluation We stated in Section 6.3.3 of Supplement No. I to the Safety Evaluation Report that we would require the applicant to provide a list of specific electrical equipment that may be submerged as the result of a postulated loss-of-coolant accident condition.

The applicant has provided a list of this electrical equipment.

For evaluation purposes, the equipment has been divided into three categories:

Safety Related Electrical Equipment, Non-Safety Related Electrical Equipment, and Electrical Penetrations associated with submerged equipment.

6.3.3.1 Safety Related Electrical Equipment For those cases where motor operated valves in the emergency core cooling system are required te operate following a postulated loss-of-coolant accident, the motor operators are located above the potential post-accident level.

The other safety related electrical equipment includes equipment not required to f unction following a postulated loss-of-coolant accident.

It includes the primary loop flow Instrumentation which is energized during normal operation. Protection against adverse conditions affecting the associated safety related 6-1

electrical equipment, is provided by Class IE protective devices including breakers, fuses, and the power supply's current limiting capabilities.

6.3.3.2 Non-Safety Related Electrical Eauipment The non-safety related equip

.nt, supplied from a Class IE electrical source which may get submerged, is isolated from the Class IE electrical source in one of the following manners:

(1)

The equipment is de-energized by the safety injection signal, or (2) The equipment has safety grade redundant current limiting and overcurrent protection devices.

6.3.3.3 Electrical Penetrations The applicant has reviewed the containment electrical penetrations which are associated with the potentially submerged equipment to ensure that the integrity of the penetration will be maintained for the postulated post-accident flooding conditions.

In all cases, either the potentlally submerged equipment is automatically de-energized or the maximum f ault currants are within the design ratings of the respective penetrations.

6.3.3.4 Conclusions We conclude that the applicant has adequately analyzed the ef fects on plant safety of potentially submerced electrical equipment and that the resu'ts of the evaluation are acceptable.

We therefore consider this matter resolved.

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10.0 STEAM AND POWER CONVERSION SYSTEM 10.2 MAIN STEAM SUPPLY SYSTEM We stated in the Safety Evaluation Report that we requested the applicant to analyze the consequences if the residual heat release control valve sticks open or if a pipe ruptures downstream of the stop-check valves.

In Amendment 53 to the Final Safety Analysis Report, the applicant stated that the three-Inch stop check valves would be administra-tively locked closed until an acceptable analysis is provided.

The stop check valves were seat leakage tesied in the reverse direction at maximum pressure differential and found acceptable.

A similar seat leakage test will be performed for pressure dif ferential in the normal direction with the valves closed, under full steam pressure conditions, and, if necessary, appropriate actions will be taken to reduce the leakage to the prescribed l im its.

Based on our review, we conclude that the main steam supply system is in conformance with the single f ailure criterion, the seismic recommendations of Regulatory Guide 1.29 " Seismic Design Classification" and main steam isolation valve closure time positions, and, is acceptable.

Therefore, we consider this matter resolved.

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10.7 TURBINE MISSILES The North Anna Power Station, Unit I and 2 turbine generators are arranged In a nonpeninsular orientation. This arrangement is such that low trajectory turbine missiles have the potential for directly impacting the auxillary building, containment, control room and other structures housing safety-related equipment.

The applicant's analysis Indicates that if a turbine failure occurs, as a result of destructive overspeed, there is about a 20 percent chance that any large, potentially damaging missile would strike (Impact) such structures. Using this strike probability and the historically observed destructive overspeed turbine f ailure rate of 10-4 per turbine year, we estimate that the upper limit probability for unacceptable damage by turbine missiles is about 2 x 10-5 per turbine year.

This is a demonstrably conservative estimate since it is based on the assumption that given a missile strike the proba-bility for structural penetration and unacceptable damage is unity.

It should also be noted that the appilcant predicts a much lower f ailure probability based on improvement in materials and design.

Whether a reduction in f ailure probability of two orders of magnitude, as compared with historically observed rates has in f act been achieved requires in our view confirmation by additional operating experience with more recent turbine designs and more detailed reviews of turbine missile risks.

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increased Inspection, maintenance and testing procedures will in our view substantially reduce the likelihood of major contributors to turbine f ailure - degraded valve performance and continued operation with flawed turbine materials. While the specific requirements for North Anna still remain under development 11 is our Intent that the applicant adhere to the turbine valve inspection, maintenance and testing procedures outlined in Section 10.2 of the Standard Review Plan

• and to the turbine inspection provisions Indicated in Section 10.2.3 of the Standard Review Plan.

The specific requirements will be included in the applicant's technical specifications.

We conclude that this requirement will provide protection from turbine missiles adequate to permit plant operation until we have completed our generic study of design criteria for turbine missiles.

When the results of this study are available, we will determine if additional protection is required beyond that offered by the proposed plant design as supplemented by the technical specification requirements identifled above.

At that time, appropriate action wilI be taken.

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG-75/087).

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e l1.0 RADI0 ACTIVE WASTE MANAGEMENT 11.1

SUMMARY

DESCdlPTION We stated In the Safety Evaluation Report that our detalled evaluation as to the capability of the proposed liquid and gaseous radioactive waste management systems to the requirements of Appendix I to 10 CFR Part 50 would be provided in a subsequent report when the informatIon was t

f filed by the applicant.

In amendments to the Final Safety Analysis Report the applicant has 1

provided the necessary Inf ormation, in Amendment 54 to the Final Safety Analysis Report, the applicant chose to comply with the Commission's September 9,1975* amendment to Appendix l.

This amendment provides for persons who have filed an application for construction permits for light-water-cooled nuclear power reactors which were docketed on or after January 2, 1971, and prior to June 4, 1976, the option of dispensing with the cost-benefit analysis required by Paragraph II.D of Appendix 1.

This option permits an applicant to design his radwaste management systems to satisf y the Guides on Design Objectives

}

for Light-Water-Cooled Nuclear Power Reactors proposed In the Concluding Statement of Position of the Regulatory Staf f in Docket RM-50-2, dated February 20, 1974**.

As Indicated in the Statement of Considerations r

• Title 10, CFR Part 50, Amendment to Paragraph II.D of Appendix 1, Federal Register, V. 40, p. 40918, September 4, 1975.

• • U. S. Atomic Energy Commission Concluding Statement of Position of the Regulatory Staf f (and its Attachment) - Public Rulemaking Hearing on:

Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criteria "As Low As Practicable" for Radioactive Mater!al in Light-Water-Cooled Nuclear Power Reactors, Docket Number RM-50-2, Washington, D.

C., February 20, 1974.

Il-l

included with the amendment, the Commission noted it Is un!!kely that f urther reductions to adloactive material releases would be warranted on a cost-benefit basis for light-water-cooled nuclear power reactors having radwaste systems and equipment determined to be acceptable under the proposed staff design objectives set forth in Docket RM-50-2.

In the case of plants whose applications for construction permits were filed prior to January 2, 1971', Appendix I does not provide specific guidance concerning the need for these plants to submit a detalled cost-benef f t analysis to determine conformance with Section ll.D of Appendix 1.

The staf f is in the process of determining whether plants for which construction applications were filed prior to January 2, 1971 can te treated in a manner similar to the treatment given plants doci.eted af ter January 2, 1971, including the cost-benefit option prov.ded by the Commission's September 4, 1975 amendment to Appendix 1.

U. d l l this determination has been made the staff's evaluation of the radwaste systems was performed to determine conformance with the criterla set forth in the Concluding Statement of Position of the Regulatory Staf f, Docket RM-50-2, in lleu of a detailed cost-benef f t analysis e required by Section l1.0.

The applicant's commitment was for the North Anna Power Station and will include Units I, 2, 3 and 4.

11-2

We have assessed this information and have extended our review to include North Anna Power Station Units 3 and 4 to demonstrate compliance with site-related criteria. Our evaluation was performed to determine if the proposed North Anna Power Station, Units I, 2, 3 and 4, meet the numerical design objectives specified in Sections ll.A, B, C and 0 of Appendix l of 10 CFR Part 50*.

The Safety Evaluation Report for North Anna Power Station, Units 3 and 4 was issued in December 1972 (Docket Nos. 50-404 and 50-405).

We have evaluated the radioactive waste management systems proposed for North Anna Power Station, Units I, 2, 3 and 4, to reduce the quantitles of radioactive materials released to the environment in liquid and gaseous effluents.

These systems have been previously described in Section 3.5 of the Final Environmental Statement, dated April 1973**, and in Section 11.0 of the Safety Evaluation Reports for North Anna Power Station, Units I and 2 and North Anna Power Station, Unit 3 and 4.

Based on more recent operating data applicable to the North Anna Power Station, Units I, 2, 3 and 4, and on changes in our calculational model, we have generated new liquid and gaseous source

" Title 10, CFR Part 50, Appendix I, Federal Register, V. 40, p. 19442, May 5, 1975.

• • Staf f of the U. S. Nuclear Regulatory Commission, " Final Env ironmental Statement Related to the Continuation of Construction and the Operation of North Anna Power Station, Unit Nos. I and 2, and the Construction of Unit Nos. 3 and 4," Docket Nos. 50-338, 50-339, 50-404, and 50-405, Washington, D.

C.,

April 1973.

11-3

terms to determine conformance with Appendix 1.

These values are different from those given in Tables 3.7, 3.8, and 3.11 of the Final Environmental Statement.

The new source terms, shown in Tables 11.1, 11.2., 11.3 and 16.4 of this report, were calculated using the models and methodology described in f4UREG-0017, " Calculation of Releases of Radloactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," Apr i l 1976.

These source terms were used to calculate the doses as descr ibed below. The dispersion of radlonuclides in and the deposition of radionuclides from the atmosphere were based on analyses performed by the s+aff for this evaluation.

The mathematical models used to perform the dose calculations are contained in Regulatory Guide 1.109, " Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of implementing Appendix 1," March 1976.

Included in our analysis are dose evaluations of three ef fluent categories: (1) pathways associated with liquid effluent releases to Lake Anna, (2) noble gases released to the atmosphere, and (3) pathways associated with radiolodines, particulates, carbon-14 and tritium released to the atmosphere, ll-4

TABLE 11.1 CALCULATED RELEASES OF RAD 10ACTI'sE MATERIAL IN LIQUID-EFFLUENTS. TOM NORTH ANNA POWER STATION; UNITS I-AND 2 Nuclide Curles Per-Year Nuclide Curies-Per Year Per Heactor Per Heactor Corrosion and Activation Products Tellurium 129 l.6(-4) a,b lodine 130 8.6(-4)

Chromium 51 3(-4)

Tell urium 13lm 2.9(-4)

Manganese 54 8(-5)

Tellurium 131 5(-5) fron 55 2.7(-4) lodine 131 l(-1) fron 59 2(-4)

Tellurium 132 3.l(-3)

Cobalt 58 2.7(-3) lodine 132 3.6(-2)

Cobalt 60 3.9(-4) lodine 133 1.5(-1)

Neptunium 239 1.8(-4) lodine 134 2.8(-3)

Cesium 134 4.7(-3)

Fission Products lodine 135 7.7(-2)

Cesium 136 2.4(-3)

Bromine 83 1.3(-3)

Cesium 137 3.6(-3)

Bromine 84 3(-5)

Barium 137m 3.l(-3)

Rubidium 86 2(-5)

Barium 140 3(-5)

Rubidium 88 2(-5)

Lanthanum 140 2(-5)

Strontium 89 7(-5)

Cerium 141 l(-5)

Strontium 91 5(-5)

Cerium 144 6(-5)

Yttrium 91m 3(-5)

Zirconium 95 2(-5)

All Others p(-5)

Niobium 95 3(-5)

Molybdenum 99 1.2(-2)

Total (except 4.l(-1)

Tritium)

Technetium 99m 9.8(-5)

Ruthenium 106 2(-5)

Tritium 5.8(+2)

Tellurium 127m 3(-5)

Tellurium 127 8(-5)

Tellurium 129m 2(-4)

-4 a = Exponential notation; l(-4)

I x 10

=

-5 b = Nuclides whose release rates are less than 10 Curies per year per reactor are not listed individually, but are in the category "All Others".

11-5

TABLE 11.2 CALCULATED RELEASES OF RADI0 ACTIVE MATERIAL IN LIQUID EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS 3 AND 4, Nuclide Curles Per Year Nuclide Curles Per Year Per Heactor Per Heactor Corrosion and Activation Products Tellurium 129 6(-5J a,b lodine 130 7(-5)

Chromium SI 1.2(-4)

Tellurium 13lm 5(-5)

Manganese 54 4(-5) lodine 131 5.4(-2)

Iron 55 1.l(-4)

Tellurium 132 1.2(-3)

Iron 59 7(-5) lodine 132 l.7(-3)

Cobalt 58 1.l(-3) lodine 133 f.7(-2)

Cobalt 60 2.l(-4) lodine 134 2(-5)

Zirconium 95 l(-5)

Cesium 134 2.5(-3)

Niobium 95 2(-5) lodine 135 4.6(-3)

Neptunium 239 7(-5)

Cesium 136 1.l(-3)

Cesium 137 2(-3)

Fission Products Barium 137m 1.7(-3)

Bromine 83 3(-5)

Barium 140 l(-5)

Strontium 89 3(-5)

Cesium 144 5(-5)

Strontium 91 2(-5)

Vitrium 9fm 1(-5)

All Others 6(-5)

Molybdenum 99 4.6(-2)

Technetium 99m 3(-2)

Total (except 1.6(-1)

Tritium)

Ruthenium 106 2(-5)

Tel l ur i um 127m l(-5)

Tritium 5.5(+2)

Tellurium 127 3(-5) fellurium 129m 9(-5)

-4 a = Exponential notation; l(-4) = 1 x 10

-5 b = Nuclidos whose release rates are less than 10 Curies per year per reactor are not listed individually, but are in the category "All Others",

ll-6

TABLE 11.3 CALCULATED RELEASES OF RADI0 ACTIVE-MATERIAL IN GASEOUS-EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS i AND 2 Curles Per Year Fer Reactor d

Reactor Auxiliary Turbine Air Decay Radlonuclide Building Building-Building Ejector Tanks Total Krypton 83m a

4 a

2 a

6 Krypton 85m a

17 a

il a

2.8(+1)

Krypton 85 a

a a

a 2(&2) 2(+2)

Krypton 87 a

1.2(+1) a 7

a 1.9(+1)

Krypton 88 I

3.5(+1) a 2.l(+1) a 5.7(+1)

Krypton 89 a

I a

a a

i Xenon 13lm 2

a a

a 2

4 Xenon 133m 5

7 a

4 a

16 Xenon 133 3.8(+2) 2.9(+2) a 1.8(+2) 4 8.5(+2)

Xenon 135m a

3 a

2 a

5 Xenon 135 5

3.8(+1) a 2.4(+1) a 6.7(+1)

Xenon 137 a

2 a

i a

3 Xenon 138 a

1(+1) a 6

a 1.6(+1) lodine 131 3.2(-4) 4.7(-3) 7.8(-4) 2.9(-2) a 3.5(-2) lodine 133 1.4(-4) 9.2(-3) 1.3(-3) 5.8(-2) a 6.9(-2)

Manganese 54 5.8(-5) 1.8(-4) c c

4.5(-5) 2.8(-4)

Iron 59 2(-5) 6(-5) c c

l.5(-5) 9.5(-5)

Cobalt 58 2(-4) 6(-4) c c

l.5(-4) 9.5(-4)

Cobalt 60 7(-5) 2.7(-4) c c

7(-5) 4.3(-4)

Strontium 89 3.3(-6) 1.3(-5) c c

3.3(-6) 2.l(-5)

Strontium 90 6i-7) 2.4(-6) c c

6(-7) 3.8(-6)

Cesium 134 4.5(-5) 1.8(-4) c c

4.5(-5) 2.8(-4)

Cesium 137 7.5(-5) 3(-4) c c

7.5(-5) 4.7(-4)

Tritium 5.8(+2) c c

c c

5.8(+2)

Carbon 14 I

a a

a 7

8 Argon 41 2.5(+1) c c

c c

2.5(+1) a = less than 1.0 Curies per year per reactor for noble gases and carbon-14,

-4 less than 10 Curies per year per reactor for iodine

-2 b = exponential notation; l.4(-2) - 1.4 x 10 c = less than I percent of total for this nuclide d = radionuclides not listed are released in quantitles less than those specified in notes a and c from all sources 11-7

TABLE 11.4 CALCULATED RELEASES OF RADI0 ACTIVE MATERIAL IN GASEQUS EFFLUENTS FROM NORTH ANNA POWER STATION, UNITS 3 AND 4 Curles Per-Year Per Reactor d

Reactor Auxiliary Turbine Air Decay Radionuclide-Building Building Building Ejector Tanks Total Krypton 85m i

2 a

I a

4 Krypton 85 9

a a

a 1.9(+2) 2(+2)

Krypton 88 2

3 a

2 a

7 Xenon 131m 1.4(+1) a a

a 2

1.6(+1)

Xenon 133m 2.2(+1) 2 a

i a

2.5(+1)

Xenon 133 2.4(+3) 1,3(+2) a 8.3(+1) 4 2.6(+3)

Xenon 135 9

5 a

3 a

1.7(+1) lodine 131 5(-3) 5.4(-3) 7.7(-4) 3.4(-2) a 4.5(-2) lodine 133 1.3(-3) 6.3(-3) 8.9(-4) 4(-2) a 4.8(-2)

Manganese 54 5.8(-3) 1.8(-4) c c

4.5(-5) 6(-3) f r on 59 2(-3) 6(-5) c c

l.5(-5) 2.l(-3)

Cobalt 58 2(-2) 6(-4) c c

l.5(-4) 2.l(-2)

Cobalt 60 9(-3) 2.7(-4) c c

7(-5) 9.3(-3)

Strontium 89 4.5(-4) 1.3(-5) c c

3.3(-6) 4.7(-4)

Strontium 90 7.9(-5) 2.4(-6) c c

6(-7) 8.2(-5)

Cesium 134 5.8(-3) 1.8(-4) c c

4.5(-5) 6(-3)

Cesium 137 f(-2) 3(-4) c c

7.5(-5) l(-2)

Tritium 5.5(+2) c c

c c

5.5(+2)

Carbon 14 I

a a

a 7

8 Argon 41 2.5(+1) c c

c c

2.5(+1) a = less than 1.0 Curies per year per reactor for noble gases and carbon-14,

-4 less than 10 Curles per year per reactor for iodine

-2 b = exponential notation; l.4(-2) = 1.4 x 10 c = less than I percent of total for this nuclide da radionuclides not listed are released in quantities less than those specified in notes a and c from all sources ll-8

The dose evaluation of pathways associated with liquid ef fluents was based on the maximum exposed individual.

The dietary and living habits for an adult individual incl uded the consumption of 21 kilograms per year of fish harvested in the immediate vicinity of the discharge from the cooling arm of the lake, consumption of 730 liters per year of water from the lake, and recreational use of the lake and its shoreline for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per year.

For an infant, the diet incl uded consumption of 510 liters per year of water from the lake. These pathways have been addressed in this evaluation utilizing the most recent NRC source terms and dose models.

For these dose assessments, the critical radionuclides (tr it ium, ces ium-134 and ces ium-137) were permitted to reconcentrate in the reservoir water as determined by expected flows, with no significant dilution of the liquid releases prior to downstream usc.

The maximum dose commitment (all four units operating) resulting from ingestion of water from the lake was estimated to be 0.095 millirem per year (total body) and 0.61 millirem per year (thyroid) for an infant.

As shown in Tables 11.1 and 11.2 of this report, the expected quantity of radioactive materials released in liquid effluents from Units I, 2, 3 and 4 will be less than 5 Curles per year per reactor (0.41 Curles per year per reactor for Units I and 2, 0.16 Curies per year per reactor for Units 3 and 4), excluding tritium and dissolved gases, in conformance with the amendment to Section ll.D.

The liquid effluents released frora Units I, 2, 3, and 4 will not result in an annual dose or dose commitment 11-9

to the total Dody or to any organ of an individual, in an unrestricted area from all pathways of exposure, in excess of 5 millirem (see Table ll.5 of this report).

The dose evaluation of noble gases released to the atmosphere included a calculation of beta and gamma air doses at the site boundary and total body and skin doses at the residence having the highest dose.

The maximum air doses at the site boundary were found at 0.85 miles north-northeast relative to North Anna Power Station, Unit 1.

The location of maximum total body and skin doses were determined to be at a residence I.09 miles south of the North Anna f acility (see TABLE 11.5 of tnis report).

The dose evaluation of pathways associated with radiolodine, parti-culates, carbon-44 and tritium released to the atmosphere was also based on the maximum exposed individual. One such individual is a child whose diet included the consumption of 520 killgrams per year of crops, 330 liters per year of milk, and 41 k!!ograms per year cf beef and poultry produced at the location of the milk cow having the highest calculated dose from these and two other pathways noted below.

This location is 1.86 miles east of the North Anna f acility.

Another such individual is a child whose diet includes the consumption of 520 kilograms per year of crops grown at the location of the residence having the highest calculated dose from this and two other pathways noted below. This location is 1.09 miles south of the North Anna ll-10

TABLE 11.5 COMPARIS0N OF NORTH ANNA POWER STATION, UNIT NOS. 1, 2, 3, AND 4, WITH APPENDIX I TO 10 CFR PART 50, SECTIONS II.A, II.B AND II.C MAY 5, 1975 a AND SECTION II.D. ANNEX SEPTEMBER 4, 1975 D

8 b

Calculated Calculated Appendix 1 Annex Doses

ses c

Criterion Design Objectives Design Objectives Unit Nos. 1 and 2 Unit w 3 and 4 Liquid Effluents Dose to total body from all pathways (infant) 3 mrem /yr/ unit 5 mrem /yr/ site 0.025 mrem /yr/ unit 0.023 mrem /yr/ unit Dose to any organ from all pathways (infant-thyroid) 10 mrem /yr/ unit 5 mrem /yr/ site 0.20 mrem /yr/ unit 0.11 mrem /yr/ unit d

Noble Gas Effluents Gamma dose in air 10 mrad /yr/ unit 10 mrad /yr/ site 0.10 mrad /yr/ unit 0.13 mrad /yr/ unit Beta dose in air 20 mrad /yr/ unit 20 mra:','yr/ site 0.14 mrad /yr/ unit 0.34 mrad /yr/ unit Dose to total body of an individual 5 mrem /yr/ unit 5 mrem /yr/ site 0.030 mrem /yr/ unit

,0.041 mrem /yr/ unit Dose to skin of an individual 15 mrem /yr/ unit 15 mrem /yr/ site 0.072 mrem /yr/ unit 0.12 mrem /yr/ unit Radiofodines and Other Radio-nuclides Released to the Atmospheree Dose to any organ from all pathways (infant-thyroid) 15 mrem /yr/ unit 15 mrem /yr/ site 0.78 mrem /yr/ unit 1,2 mrem /yr/ unit aFederal Register, V.40, p. 19442, May 5, 1975.

dLimited to noble gases only, b

8 Federal Register, V, 40, p. 40816. September 4, 1975, Carbon-14 and Tritium have been added to cDesign Objectives given on a site ba is. Therefore, this category.,

these design objectives apply to 4 units at the site.

mrem /yr/ unit - m1llirem per year per' unit mrem /yr/ unit - millirad per year per unit mrem /yr/ site - millirem per year per site mrad /yr/ site - millirad per year per site

facility.

These maximum exposed individuals were also exposed to inhaled radionuclides in this category, as well as those deposited on the ground at each of the locations described above (see Table 11.5 of this report).

Based on our evaluation of the gaseous radwaste management systems, the total quantity of radioactive materials released in gaseous effluents from the North Anna Power Station, Units I, 2, 3, and 4 will not result in an annual gamma air dose in excess of 10 millirads and a beta air dose in excess of 20 m!!!irads at every location near ground level, at or beyond the site boundary, which could be occupied by individuals (see Table 11.5 of this report).

As shown in Tables l1.3 and ll.4 of this report, the annual total quantity of iodine-131 released in gaseous ef fluents will be less than ! Curie per reactor (0.035 Curies per year per reactor for Units I and 2 and 0.045 Curies per year per reactor for Units 3 and 4) in conformance with the amendment to Section 11.0 and the annual total quantity of radiciodine and radioactive,particulates released in gaseous effluents from North Anna Power Station, Units I, 2, 3, and 4 will not result in an annual dose or dose commitment to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 millirem (see Table 11.5 of this report).

Our evaluation demonstrates that the radwaste treatment system proposed for North Anna Power Station, Units I, 2, 3, and 4 are 11-12

capable of maintaining releases of radioactive materials in effluents during normal operation such that the doses will not exceed the design objectives of Sections ll.A, B and C of Appendix l of 10 CFR Part 50.

Our evaluation also shows that the applicant's proposed design of North Anna Power Station, Units I, 2, 3, and 4 satisfies the design objectives set forth in RM-50-2 specified in the option provided by the Commission's September 4, 1975, amendment to Appendix I and, therefore, meets the requirements of Section ll.D of Appendix I of 10 CFR Part 50.

We conclude that the liquid and gaseous radwaste treatment systems will reduce radioactive materials in ef fluents to "as low as is reasonably achievable" levels in accordance with 10 CFR Part 50.34a and, therefore, are acceptable.

11-13

13.0 CONDUCT OF OPERATf0NS 13.3 EMERGENCY-PLANNING We stated in Section 22.0 of Supplement No. I to the Safety Evaluation Report that the applicant has supplied all the requested information with regard to the Emergency Plan to demonstrate that it meets Appendix E to 10 CFR Part 50 except for several letters of agreement with agencies within Spotsylvania County, Virginia.

In Amendment 54 to the Final Safety Analysis Report, the applicant has provided all the necessary documentation. On this basis, we have determined that the plan meets the requirements of Appendix E to 10 CFR Part 50 and provides an adequate basis for an acceptable state of emergency preparedness. We therefore consider this matter resolved.

13-1

14.0 INITIAL TESTS AND OPERATIONS We stated in the Safety Evaluation Report that in order for us to complete our evaluation relating to the staf fing and organizational responsibilities of personnel managing and directing the test program, additional Information had been requested from the applicant. We also stated that our evaluation of this aspect of the test program will be included In a subsequent report.

in Amendments 52 and 54 to the Final Safety Evaluation Report the applicant has presented the necessary Information. We have reviewed this information and conclude that the applicant's plans relative to assigned manpower, responsibilities and qualification requirements for personnel Implementing the startup test program are acceptable. We therefore consider this issue resolved.

14-1

22.0 CONCLUSION

S I n Section 22.0 of Supp lement No. I to the Safety Evaluation Report we stated that several items as set forth in Section 1.7 of the Safety Evaluation Report were still outstanding and that satisfactory resolution of these items would be required before operating licenses for North Anna Power Station, Units I and 2 could be issued. A number of these have Deen resolved, as reported in this supplement. The remaining items which must be resolved, and their present status, are summarized below. Resolution of each item will be discussed in a future supplement to the Safety Evaluation Report.

(1) Our review of the design of the system of well points for groundwater control and the resolution of the settlement of the northwest corner of the service building is not completed. (Section 2.6 of this report).

(2) The applicant has provided our requested information regarding the dynamic analyses of the ef fects of a postulated loss-of-coolant accident on fuel elements. Our evaluation of this information has not been completed.

(Safety Evaluation Report Section 4.2.4)

(3)

The applicant has submitted information regarding the preoperational tests of the recirculayton mode of operation for the low head safety injection pumps, ha have reviewed this information and have requested 22-1

additional information from the applicant which we have determined necessary for the completion of our review.

(Safety Evaluation Report Sections 6.3.4 and 14.0 and Section 6.3.4 of Supplement No.1)

(4)

The test program results to demonstrate that adequate electrical isolation exists between the safety related and non-safety related portions of the 7300 series process analog system have not yet been submitted.

(Safety Evaluation Report Section 7.2.2)

(5)

The applicant has stated in Amendment 53 to the North Anna Power Station, Units I and 2 Final Safety Analysis Report that it is considering modifications to the auxf ilary feedwater system in order to extend the time required for operator action to 30 minutes in the event of a feedwater line break.

In Amendment 54 the applicant has provided a revised analysis of the feedweater line break for a modified system design. We are reviewing these analyses.

(Safety Evaluation Report Section 15.3)

(6)

The applicant must provide additional Information on the seismic and environmental qualification of seismic Category I instrumentation and electrical equipment.

(Safety Evaluation Report Section 3.10) 22-2

(7) We have not completed our evaluation of the steam generator and reactor coolant pump supports.

(Safety Eval uation Report Section 5.4.2)

(8) We have not completed our evaluation of the mass and energy release rates for a postulated main steam line break accident.

(Safety Evaluation Report Section 6.2.1)

(9) We have not completed our evaluation of the applicant's financial qualifications to operate the facility.

(Safety Evaluation Report Section 20.0)

(10) We will verif y the acceptability of the service water reservoir for two-unit operation after we have evaluated the results of the initial operational testing program for North Anna Power Station Unit 1.

(Safety Evaluation Report Section 2.4.3)

(11) We plan to review overpressurization incidents of the reactor coolant system when in a water-solid concitton on a generic basis.

(Section 5.2.8 of this report)

(12) We have not yet taken a position on the age of last movement of the Staf ford Fault Zone and are seeking additional evidence upon which to base a conclusion.

(Section 2.5 of this report) 22-3

Subject to satisf actory resolut'on of the outstanding matters described above, the conclusions as stated in Section 22 of the North Anna Power Station, Units I and 2 Safety Evaluation Report remain unchanged.

22-4

APPENDlX-A CONTINUATION OF-CHRONOLOGY OF RADIOLOGICAL REVIEW June 16, 1976 Summery of meeting held on June 9, 1976 to discuss the outstanding issues in the North Anna Station.

June 21, 1976 Summary of meeting held on June 14, 1976 to discuss foundation of the service water reservoir and service water pump house.

June 23, 1976 Letter from VEP00 requesting withdrawal of their April 21,1976 transmittal considered proprietary June 23, 1976 Reconstitution of Atomic Safety and Licensing Appeal Board.

June 24, 1976 Order issued by Atomic Safety and Licensing Appeal Board. Appeal concerning the admittance of Sun Shipbuilding and Dry Dock Company as an intervenor.

June 29, 1976 Order issued by Atomic Safety and Licensing Board.

Sun Shipbuilding and Dry Dock Company will serve the Applicant a list of items which it wishes to have produced.

June 30, 1976 Order issued by Atomic Safety and Licensing Appeal Board.

Extension of time to and including July 8, 1976 within which the Staf f may file its brief.

June 30, 1976 Division of Project Management letter requesting additional information.

June 30, 1976 issuance of Supplement No. I to the Safety Evaluation Report July 1, 1976 Olvision of Project Management letter returning VEPCO's letter and enclosures, dated April 8,1976, in accordance with their request.

Jul y 2,1976 Order issued by Atomic Safety and Licensing Appeal Board. Oral Argument calendared for 9:00 a.m. on Monday, Jul y 19, 1976 at the East West Highway Building, Bethesda, Maryland.

July 2, 1976 VEPCO letter sending advance copies of pages that will appear in Amendment No. 54 to the Final Safety Analysis Report.

A-1

July 7, 1976 Division of Project Management letter advising that the schematics transmitted on May 18, 1973 are not considered as proprietary information.

July 9, 1976 Division of Project Management letter requesting additional information.

July 9, 1976 VEP00 letter advising they are experiencing minor delays and, therefore, may not be able to commence fuel loading before December 1976.

Jul y 12, 1976 Division of Project Management letter requesting additional information.

July 13,1976 Division of Project Management letter transmitting VEPC0 copies of Supplement No. I to the North Anna Safety Evaluation Report.

Jul y 14, 1976 Submittal of Amendment No. 54 to the Final Safety Analysis Report.

Jul y 21, 1976 VEPCO and NRC representatives meet in Bethesda, Maryland to discuss steam generator and reactor coolant pump supports.

A-2