ML090720823

NUREG-0053, Suppl. 7, Safety Evaluation Report Related to Operation of North Anna Power Station, Units 1 and 2

ML090720823
Person / Time
Site:	North Anna
Issue date:	08/18/1977
From:	Office of Nuclear Reactor Regulation
To:
References
FOIA-2024-00060 NUREG-0053, Suppl. 7
Download: ML090720823 (84)

Text

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uppiement No" 7 NUR SuppL Noo 7

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N Regulatory Comm Office of Nuclear Reactor Reg~Jdation August 1911

SUPPLEMENT NO.7 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION - UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339 NUREG-0053 Supplement No. 7 August 18, 1977

TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DISCUSSION,........................................... 1-1 1.1 Introduction.............................................................. 1-1 2.0 SITE CHARACTERISTICS........................................................... 2-1 2.6 Foundation Engineering......................... "........................... 2-1 2.6.2 Evaluation of Foundation Engineering.............................. 2-1 3.0 DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS.......................... 3-1 3.9 Mechanical Systems and Components......................................... 3-1 3.9.4 Analysis Methods for Loss-of-Coolant Accident Loadings............ 3-1 4.0 REACTOR......................................................................... 4-1 4.2 Mechanical Design......................................................... 4-1 4.2.4 Design Analysis for Loss-of-Coolant Accident Loadings............* 4-1 5.0 REACTOR COOLANT SySTEM......................................................... 5-1 5.2 Reactor Coolant Pressure Boundary......................................... 5-1 5.2.8 Overpressure Protection........................................... 5-1 15.0 ACCIDENT ANALySIS.............................................................. 15-1 15.4 Radiological Consequences of Accidents.................................... 15-1 15.4.5 Fuel Handling Accident Inside Containment......................... 15-1 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS......................... 18-1 18.2 Operating License Review.................................................. 18-1 18.2.1 Verification of Westinghouse 17 X 17 Fuel Assembly Design......... 18-1 18.2.2 Steam Generator and Reactor Coolant Pump Supports................. 18-2 18.2.3 Anticipated Transients Without Scram.............................. 18-2 18.2.4 Generic Items..................................................... 18-2 18.2.5 Evaluation of Safety Factors for Safety-Related Systems During Safe Shutdown Earthquake Conditions............................. 18-3

18.2.6 Investigation of Construction Activities of North Anna Power Station Units 1 and 2........................................... 18-4 18.2.7 Evaluation of Asymmetric Loads on Pressure Vessel Structures...... 18-5 18.2.8 *Fire Protection System............................................ 18-5 18.2.9 Long-Term Seal Capability......................................... 18-6

21. 0 FINANCIAL PROTECTION AND INDEMNITY REqUIREMENTS................................ 21-1 21.3 Operating License......................................................... 21-1

22.0 CONCLUSION

S.................................................................... 22-1 1 i

APPENDIX A APPENDIX B APPENDIX C APPENDIX D APPENDICES PAGE CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW.................... A-1 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS' LETTER DATED OCTOBER 26, 1976..................................................... B-1 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS' LETTER DATED JANUARY 17, 1977..................................................... C-l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS - GENERIC MATTERS........... D-1 iii

LIST OF TABLES PAGE TABLE 15.8 RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS - FUEL HANDLING ACCIDENT INSIDE CONTAINMENT............................... 15-3 TABLE 15.9 ASSUMPTIONS USED IN THE ANALYSIS OF FUEL HANDLING ACCIDENT DOSES INSIDE CONTAINMENT........................................... 15-4 iv

1.0 INTRODUCTION

AND GENERAL DISCUSSION 1.1 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 1 and 2 (North Anna facility).

The application was filed by the Virginia Electric and Power Company (applicant).

Supplement No.1 to the Safety Evaluation Report was issued on June 30, 1976; Supplement No. 2 was issued on August 2, 1976; Supplement No. 3 was issued on September 15, 1976; Supplement No.4 was issued on December 8, 1976; Supplement No~ 5 was issued on December 29, 1976; and Supplement No.6 was issued on February 2, 1977.

Supplement Nos. 1, 2, 3, 4, 5 and 6 to the Safety Evaluation Report documented the resolution of several outstanding items, and summarized the status Df the remaining outstanding issues.

The purpose of this supplement is to update our Safety Evaluation Report (and Supple-ment Nos. 1, 2, 3, 4, 5 and 6) by providing (1) our responses to the comments of the Advisory Committee on Reactor Safeguards in its letters of October 26, 1976 and January 17, 1977, (2) our evaluation of additional information submitted by the applicant since the issuance of Supplement No.6 to the Safety Evaluation Report, (3) our evaluation or additional information for those sections of the Safety Evaluation Report where further discussion or changes are in order, and (4) information regarding the current status of matters that are still under review.*

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 and supplements thereto, except where specifically so noted.

Appendix A is a continuation of the chronology of our principal actions related to the processing of the application.

The letters of the Advisory Committee on Reactor Safeguards dated October 26, 1976 and January 17, 1977 are attached as Appendices Band C, respectively.

Appendix D addresses generic matters addressed in the Advisory Committee on Reactor Safeguards letter of October 26, 1976.

A summary of the remaining outstanding issue is presented in Section 22.0 of this supplement.

1-1

2.6 2.6.2 2.0 SITE CHARACTERISTICS Foundation Engineering Evaluation of Foundation Engineering We stated in Section 2.6.2 of Supplement No.2 to the Safety Evaluation Report that the applicant has committed to the installation of a system of well points to control groundwater levels under the service water pumphouse.

We further stated that before installation of this system, the details of the system design will be required to be included in the Final Safety Analysis Report and reviewed and approved by the staff to insure that it will fulfill its function without causing a loss of fines from the saprolite.

In a letter dated November 1, 1976, the applicant indicated that it would use an underdrain system to control groundwater levels under the service water pumphouse in lieu of a system of well points and also presented information concerning the design of this system.

As a result of our review of this information, we requested addi-tional information which we determined to be necessary to support the proposed design of the groundwater control system.

In a letter dated December 14, 1976, the applicant responded to our request for additional information concerning the underdrain.system for the North Anna Power Station, Units 1 and 2 service water pumphouse.

A general description of the ground-water control system and design conditions follow.

The founding elevation of the pumphouse is at 297 feet.

The groundwater control

• system to be installed beneath the pumphouse consists of 1.5 inch diameter polyvinyl chloride pipes placed in an essentially horizontal plane at an elevation of 274 feet.

The pipe runs will be about 16 feet apart, and the slotted portion of the pipes will extend beyond the limits of the foundation mat.

The drains are to be positioned beneath the existing pumphouse by drilling horizon-tally, pushing the polyvinyl chloride pipes into cased holes, and removing the casing.

Instruments to monitor the alignment and position of the installed drains should assure that they are placed within about a foot of their planned elevation.

On the basis of our review of the above information, we required that actual drain locations be determined and that the basis of the underdrain system design be docu-mented in the Final Safety Analysis Report.

We also required that the effluent be checked periodically to assure that excessive fine soil particles are not carried out of the saprolite by the groundwater.

The essential elements of the testing program 2-1

we require follow.

Samples of the effluent must be tested for turbidity and suspended solids at least once every six months and records of these tests be maintained.

If these tests show that the turbidity and suspended solids content of the effluent from the drains exceed ten rarts per million, a formal report should be made to us and the responsible North Anna Power Station safety office and operating organization.

An evaluation of the cause of the excessive turbidity and suspended solids in the effluent must be made and corrective action carried out.

In a letter dated March 4, 1977 and in Amendment No. 61 to the Final Safety Analysis Report, the applicant indicated that the effluent from the horizontal drains to be installed beneath the service water pumphouse for North Anna Power Station, Units 1 and 2 will be monitored for suspended solids and turbidity.

The North Anna Power Station, Units 1 and 2 Technical Specifications will require measurement of these parameters at least once every six months.

Also, in the event that either the solids content or the turbidity exceed 10 parts per million, a report will be sub-mitted to us for our review outlining possible causes and planned corrective action to mitigate the loss of fines from the soil.

An excessive loss of fines over a long period of time could eventually result in unsafe support conditions for the pump-house.

The monitoring of the above parameters will assure that the loss of fines will be discovered before an unsafe condition could develop.

On the basis of our review of the groundwater control system, we have determined that the design of the groundwater control system for the North Anna Power Station, Units 1 and 2 service water pump house is acceptable.

We have also concluded that the implementation of the technical specification discussed above will provide an adequate assurance of the long term support capability of the soil beneath the service water pumphouse.

In a letter dated April 15, 1977, the applicant stated that the installation of the groundwater control system under the service water pumphouse for North Anna Power Station, Units 1 and 2 will not be complete until September 1, 1977.

In lieu of the installed groundwater control system, the applicant has proposed a temporary technical specification wherein the plant would be shut down if the average groundwater elevation at four different locations exceeds 285 feet without specifically taking into account groundwater levels beneath the pumphouse.

Groundwater elevations would be monitored at least weekly.

The temporary specification would be deleted when the horizontal drain system is completed.

We have determined that, because the applicant's proposed temporary specification could allow the indefinite operation of Unit 1 without regard to groundwater levels occurring beneath the pumphouse, we will require that more specific criteria be used in the temporary specification.

For the temporary specification, we require that 2-2

four piezometers closest to the pumphouse be monitored for groundwater elevations at least weekly.

If the groundwater elevation in any two of these four piezometers exceeds 285 feet, Unit 1 will be shut down until the groundwater control system is completed and operational.

Tn addition, we intend to condition the operating license of Unit 1 to require that the plant be shut down December 31, 1977 if the groundwater control system is not completed and operational.

The technical specifications will include our requirements as discussed above.

On this basis, we consider this matter resolved.

An additional item of concern was made known to us in a letter dated December 7, 1976 to N. Mosley of the Office of Inspection and Enforcement from S. Brown of Virginia Electric Power Company.

In this letter, the applicant advised us of a deficiency relative to the maximum contact pressures in the soil beneath the mat foundation of the service water pumphouse.

The applicant's design criteria allowed a three thousand pounds per square foot loading, but the pumphouse is expected to exert nearly four thousand pounds per square foot on the saprolite during loading transients.

The applicant has reevaluated the soil bearing capacity of the service water pumphouse foundation and has determined on the basis of recent soils investigations and un-drained laboratory tests, that a minimum value of allowable bearing capacity is 4,200 pounds per square foot.

This allowable bearing value is based on the ultimate bearing capacity of the soil under the toe of the service water pumphouse of about twelve thousand pounds per square foot.

In Amendment No. 61 to the Final Safety Analysis Report, the applicant adequately documents the bases for increasing the allowable transient bearing load for the saprolite soils beneath the toe of the service water pumphouse for North Anna Power Station, Units 1 and 2.

On the basis of our review of the information presented in Amendment No. 61 to the Final Safety Analysis Report, we have determined that the allowable transient bearing value of 4,200 pounds per square foot is acceptable for the pumphouse foundation.

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3.9 3.9.4 3.0 DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS Mechanical Systems and Components Analysis Methods for Loss-of-Coolant Accident Loadings In Section 3.9.4 of the Safety Evaluation Report we stated that the design adequacy and structural integrity of the reactor pressure vessel support system would be verified for the loadings which result from a postulated loss-of-coolant accident.

The applicant has performed detailed time history dynamic analyses of the reactor pressure vessel support system when subjected to asymmetric internal and external pressure loadings and jet thrust reaction forces resulting from a postulated loss-of-coolant accident at the reactor pressure vessel nozzle/cold leg piping connection.

This was the controlling loss-of-coolant accident postulated for evaluating the design of the reactor pressure vessel support system.

The analysis was based on the assumption that the reactor pressure vessel support system remains linearly elastic during the def.ormation process.

Load interaction failure curves were developed which enveloped the individual loadings for the reactor pressure vessel support capscrews using elastic-plastic material properties of the capscrews.

The thermal hydraulics structure system dynamics computer code IIMULTIFLEX" was employed for the reactor pressure vessel support system analysis.

During the course of the review, the following areas in the MULTIFLEX code were identified by us as requiring modification:

(1) The core barrel model should be represented by ten mass points instead of five mass points.

(2) A revised radial transport distance through the lower plenum of the vessel should be used.

(3) A revised sonic veloc~ty calculation should be applied.

The applicant has considered these code modifications in determining the resulting loads and displacements on the reactor pressure vessel support system.

We have determined that the analytical and modeling procedures used for the reactor pressure vessel support system analysis are acceptable and give acceptable results as discussed below.

The results reported to us by the applicant indicate that the maximum internal forces acting on the supports resulting from the loss-of-coolant accident, safe 3-1

shutdown earthquake and deadweight fall within the load interaction failure curves envelope, and that plastic deformation would occur in a very small portion of the entire system.

This procedure, in addition to other conservatisms, is acceptable and falls within the context of Appendix F of Section III, Division 1 of the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers.

We therefore consider that the applicant has demonstrated that the structural integrity of the reactor pressure vessel will be maintained under these extreme loadings.

The results also demonstrate the structural integrity of reactor coolant loop piping, core support structures and other internals, safety systems and components as well as functionability of control rods are maintained under these loadings.

These analyses also verify that the integrity of the safety systems is assured during a loss-of-coolant accident and that the reactor can be safely shutdown and maintained in a safe condition.

On the basis of our evaluation, we have concluded that the reactor pressure vessel support system is acceptable and that the North Anna Power Station, Units 1 and 2 can be safely operated with respect to this matter.

3-2

4.2 4.2.4 4.0 REACTOR Mechanical Design Design Analysis for Loss-of-Coolant Accident Loadings We stated in the Safety Evaluation Report that when we receive the information required to complete our evaluation of the applicant's dynamic analysis of the fuel elements, we would report the results of the evaluation in a subsequent supplement.

The combination of a loss-of-coolant accident-induced load with seismic motion can cause both lateral loads that may crush the fuel pin spacer grid of a fuel assembly, and vertical loads that could produce structural instability of the fuel assembly column.

Maintaining a coolable geometry may be more difficult if the spacer grid crushes and rod-to-rod spacing is reduced.

An objective of the fuel assembly design is that the components are strong enough to resist such loadings.

Our review of fuel assembly integrity was initiated in 1974 when Westinghouse Electric Corporation submitted a licensing Topical Report entitled, IISafety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident,1I WCAP-8236, December 1974.

The report contained what was believed to be a bounding study, which would eliminate the need for plant-specific analyses.

During the North Anna Power Station, Units 1 and 2 review, however, Westinghouse discovered that there are addition~l loads on the reactor vessel support as well as on the fuel assemblies resulting from asymmetric loads (reactor internals and reactor cavity) when a pipe break is postulated inside the biological shield.

This issue is discussed in Amendments 43, 46, 53, 60 and 61 to the North Anna Power Station, Units 1 and 2 Final Safety Analysis Report.

The applicant used the MULTIFLEX computer code to compute the hydraulic loads during a postulated loss-of-coolant accident.

MULTIFLEX is a coupled hydrodynamic-structural interaction code.

We have evaluated the MULTIFLEX code and a topical report evaluation was issued June 17, 1977.

The evaluation of the MULTIFLEX code resulted in three changes being made to the code and to the modeling procedures used by Westinghouse for an acceptable licensing calculation. The MULTIFLEX analyses related to the fuel assembly integrity have been reviewed and we determined that these changes were incorporated in an acceptable manner.

The Westinghouse ~tudy of the fuel assembly response (WCAP-8236) consisted of analysis and tests. Fuel assembly damping, stiffness and natural frequencies, and spacer grid stiffness and damping characteristics were obtained from experiments.

Westinghouse 4-1

ran a scale model multi-beam lateral shake test to determine impact characteristics and to verify analytical methods.

They also verified the design of the fuel assembly by dropping it on a hard surface, thereby simulating an axial impact.

We found that the Westinghouse evaluation approach is a balanced combination of analysis and experiments and is based on sound engineering practice.

A similar conclusion was reached by our consultant (R. L. Grubb, "PWR Fuel Assembly Mechanical Response Analysis," INEL Report, RE-S-76-l64 September 1976 and Amendment 1, RE-E-77l03, January 1977).

However, it was our opinion that the evaluation of the design was not complete inasmuch as the appropriateness of the following items had not been adequately shown:

(1) Treating nonlinear system as linear (assembly stiffness and damping).

(2) The sloshing effect of the coolant on the fuel assembly during lateral excita-tion.

(3) The dynamic treatment of gap and impact.

Most importantly, even though many separate-effects experiments were performed, no full-scale lateral impact test was performed on a multifuel assembly configuration.

We determined that an additional safety margin above that recommended by Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code would compensate for the lack of full-scale tests and shortcomings 1, 2, and 3 above.

This additional margin is discussed below.

Originally, the applicant submitted a calculation of fuel assembly response for the North Anna Power Station, Units 1 and 2 in Amendments 43 and 46 to the Final Safety Analysis Report based on loads associated with conservative assumptions concerning postulated pipe break size and break opening time.

This analysis showed that there was an adequate margin in the thimble tube stress thus assuring control rod insertion; a similar conclusion could not be reached regarding the ability of the spacer grids to resist crushing.

In Amendment Nos. 53, 60, and 61 to the Final Safety Analysis Report, the applicant reevaluated the safety margin using more realistic assumptions in the analysis of the fuel assembly integrity.

The changes were:

(1) The seismic ground motion input was decreased from 0.20g to 0.12g.

(2) The break opening area was decreased from 144 square inches to 58 square inches based on the true system response.

(3) The break opening time was increased from 1 millisecond to 10 milliseconds (it is estimated to take 10 milliseconds to develop a 58 square inch flow area).

4-2

These changes were discussed in Amendment No. 60 to the Final Safety Analysis keport.

We found them acceptable because they represent the most realistic input and true system response for seismic and loss-of-coolant accident loadings, respectively.

The resultant safety margin in the grid crushing force was found to be 1.76 when compared with a 95 x 95 confidence level for the low side of some 40 grid crushing data.

The 1.76 safety margin is acceptable for the North Anna facility.

As an example,Section III, Appendix F of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, which is applicable to core support structures and pressure retaining components, requires a margin of 1.50 for a support member when it is subjected to a buckling load.

The additional margin for the North Anna fuel grid should bound such uncertainties as the linear calculation for a nonlinear system, the effect of fluid on dynamlc motion, and the lack of a full scale multi assembly impact test previously discussed.

It should be noted that there are other conservatisms inherent in the analysis.

The calculation demonstrated that only the peripheral fuel assemblies adjacent to the core barrel would experience high impact forces.

Most of the interior assemblies would see only a small force.

Those assemblies along the core barrel whose spacer grid may be subjected to a relatively high impact force operate at a low power level and therefore require less cooling.

As a result, there is a larger safety margin than the one calculated above.

During our earlier review of WCAP-8236 we found that the safety margin on structural stability of the control rod thimble tubes was adequate.

The additional asymmetric loss-of-coolant accident loading discussed here does not alter that conclusion since the thimble tube would be subjected primarily to axial loads and the asymmetric loss-of-coolant accident loading only affects the lateral load.

On the basis of our evaluation, we conclude that there is reasonable evidence that the 17 x 17 Westinghouse fuel assembly would withstand the combined effects of postulated seismic and loss-of-coolant accident load without impairing either coo1-able geometry or control rod insertion and, therefore, the fuel assembly design is acceptable.

4-3

5.2 5.2.8 5.0 'REACTOR COOLANT SYSTEM Reactor Coolant Pressure Boundary Overpressure Protection We stated in Section 5.2.8 of Supplement No.2 to the Safety Evaluation Report that we planned to reveiw overpressurization incidents with Westinghouse and other pressur-ized water reactors vendors generically, and would report resolution of this item as it relates to the North Anna Power Station Units 1 and 2 in a subsequent supplement to the Safety Evaluation Report.

Several instances of reactor vessel overpressurization have occurred in pressurized water reactors in which the technical specifications implementing Appendix G to ~O CFR Part 50 have been exceeded.

The majority of cases have occurred during cold shutdown while the primary system was in a water-solid condition.

The Virginia Electric and Power Company is a participant of a task group of utilities for the purpose of determining a solution to this issue.

In the interim, we have required that administrative controls be instituted to minimize the likelihood of these over-pressurization events while the plant is in a water-solid condition.

The applicant has proposed several modifications to its administrative procedures, design, and operator training as discussed in letters to us dated April 14, 1977 and June 24, 1977.

Included in these proposals are:

(1) Operator Training - During cold license training, operators will be briefed on the types of events that could cause overpressurization and the changes made in the procedures to minimize the probability of such events.

(2) Residual Heat Removal Relief Valve - The setpoint of the residual heat removal relief valve is 550 pounds per square inch gauge which is below the isolation pressure of the residual heat removal system.

The valve has a 900 gallons per minute relief capacity which is above the flow capacity of a single charging pump.

(3) Steam Bubble - A steam bubble will be formed in the pressurizer at 200 degrees Fahrenheit when the plant is being heated up and the bubble will be collapsed at 200 degrees Fahrenheit when the plant is cooled down.

This procedure will minimize the amount of time in a water-solid condition.

(4) Charging Pump - One charging pump will be operable at reactor coolant system temperatures below 200 degrees Fahrenheit which will limit the potential volumetric insurge.

If there is a path to the residual heat removal relief 5-1

valve, no overpressur;zationwould occur if the letdown line is inadvertently closed.

The other charging pumps will have power removed.

(5)

Letdown Line - The letdown heat exchanger bypass control valve (FCV-1605) will be mechanically blocked such that the maximum closure of the valve will not allow more than 50 pounds per square inch differential between the residual heat removal pumps discharge and the relief valve tap.point when running one pump on minimum recirculation flow.

(6) Reactor Coolant Pumps - The procedures will include a precaution*to verify residual heat removal suction valves are open prior to starting a reactor coolant system pump during water-solid operation.

To start a reactor coolant pump' while the reactor coolant system is in a water-solid condition, the operating procedures will require (a) special permiss~on from the operating supervisor, and (b) that the associated steam generator secondary side bulk water temperature is not greate~ than 10 degrees Fahrenheit above the reactor coolant system temperature.

When the reactor coolant system temperature is greater than seal water temperature the operating supervisor's permission will be required to start a reactor coolant pump.

A bubble must be formed in the pressurizer prior to starting more than one reactor coolant pump when the system is water-solid. It is preferable to only start a reactor coolant pump when there is a pressurizer bubble.

(7) Accumulators - The accumulator isolation valves will be closed and power locked out whenever the plant is on residual heat removal cooling.

(8) Alarm - An alarm has been installed utilizing an existing reactor coolant system pressurizer control channel which will be rescaled to provide a low as well as high pressure alarm.

The alarm will annunciate on the main control board (audio and visual) when the primary system pressure is greater than or equal to 450 pounds per square inch gauge.

(9) Residual Heat Removal Automatic Isolation - The setpoint for automatic residual heat removal isolation will be 660 pounds per square inch gauge.

(10) Power-Operated Relief Valve - One of the power-operated relief valves will be modified to have a dual setpoint.

The lower setpoint will open the valve at a pressurizer pressure of 475 pounds per square inch guage.

We have reviewed these proposed administrative and design changes and find them acceptable as an interim measure to minimize the likelihood fo a water-solid overpres-surization pending confirmation of their implementation by our Office of Inspection and Enforcement.

5-2

The potential effects of water-solid overpressurization for North Anna Units Nos. 1 and 2 have been reviewed.

Because of the minimal neutron damage suffered by the pressure vessel during its first operating cycle, we have concluded that no credible event could cause vessel rupture due to overpressurization during this period.

Because of the applicant's proposed administrative procedures and the pressure vessel fracture toughness, we have concluded that the reactor can operate for its first cycle with reasonable assurance that the health and safety of the public are protected.

The applicant is a member of a utility group that is developing a long-term solution to mitigate the consequences of pressure transients during water-solid operation.

The design modification being considered utilizes the power-operat~d relief valves to preclude violating Appendix G limits.

We will review the proposed design modification when the supporting analytical results are available to justify its effectiveness.

We require that an effective overpressure protection system, that meets our require-ments, be installed prior to the initiation of the second operating cycle.

5-3

15.4 15.4.5 15.0 ACCIDENT ANALYSIS Radiological Consequences of Accidents Fuel Handling Accident Inside Containment In a letter dated March 16, 1977 to the Virginia Electric and Power Company, we advised the applicant that we were in the process of evaluating a postulated refueling accident inside containment.

We also stated that based on our preliminary review, potential site boundary radiation exposures due to such an accident at the North Anna Power Station, Units 1 and 2 would be within the exposure guidelines of 10 CFR Part 100 even assuming no isolation of containment.

In order to confirm our results and determine if the acceptance criteria of Standard Review Plan 15.7.4 are met, and to document the factors involved in the evaluation, we requested that the applicant provide a detailed evaluation of the potential consequences of such an accident at the North Anna Power Station, Units 1 and 2.

In a letter dated June 11, 1977, the applicant described the plant systems that will be used to mitigate the consequences of a fuel handling accident inside containment.

The applicant plans to have a containment air recirculation and purging system in operation during refueling operations. A set of fan coolers located in the lower portion of the containment below the refueling cavity surface receives a total flow rate of 282,000 cubic feet per minute, and exhausts 22,000 cubic feet per minute to the environment through two containment isolation valves.

The remaining 260,000 cubic feet per minute is cooled and returned to the upper portions of the containment.

Three non-safety grade radiation monitors are located within the containment.

One monitor is located on the fuel handling crane and has a response time of less than one second.

Two other monitors are located so as to sample from the discharge of the recirculation coolers.

The transit time of the sample from the discharge of the recirculation cooler to the radiation monitors is about 30 seconds.

Detection of a high radiation signal by any of these detectors will automatically initiate closure of the containment isolation valves.

The containment isolation valves have a technical specification closure time of 20 seconds.

Both we and the applicant have analyzed the operation of these systems and features and have estimated the consequences given the occurrence of a fuel handling accident inside containment.

The applicant estimates the transit time of a radioactive release from the refueling cavity surface to the exhaust isolation valves to be 60 seconds.

Since this transit time is significantly longer than the sum of the response time of the monitor on the fuel handling crane plus valve closure time (estimated to be less 15-1

than 21 seconds), the applicant concludes that the isolation valves will close before any activity is released to the environment.

We concur in this conclusion.

The applicant has also evaluated the case where the fuel handling crane radiation monitor is presumed to be inoperative, but either one or both of the monitors which sample from the recirculation coolers are operable.

We regard it as possible, but unlikely, that one or more radiation monitors might be inoperable.

It should be noted that the technical specifications require that each of the purge and exhaust penetrations which provide direct access from the containment to the atmosphere be closed if any of the radiation monitors are inoperable.

In this case, the applicant assumed that any activity released from the refuelirg cavity would be somewhat mixed with the containment air before reaching the purge duct, and that a diluted activity release of 50 seconds occurred prior to containment isolation (30 seconds monitor response plus 20 seconds valve closure).

Although we believe that mixing of any activity will occur prior to arrival at the purge duct, we performed an independent analysis which conservatively assumed no mixing, but which accounted for the fact that the recirculation cooler allows less than 10 percent of the activity to be exhausted to the environment.

Our other assumptions are shown on Table 15.9, while the calculated radiological consequences are shown on Table 15.8.

The doses are seen to be well within the guideline values of 10 eFR Part 100, and are acceptable.

We conclude, based upon the above analyses, that the existing plant systems for the North Anna Power Station, Unit Nos. 1 and 2 will be effective in mitigating the consequences of a fuel handling accident inside containment and that the consequences of such an accident would be acceptably low.

15-2

Case I (Fuel Handling Crane Monitor Functions)

Case II (Fuel Handling Crane Monitor Inoperative)

TABLE 15.8 RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Exclusion Boundary Two-Hour Dose, Rem (1350 Meters)

Thyroid o

9 15-3 Whole Body o

<1 Low Population Zone Course of Accident Dose, Rem (6 Miles)

Thyroid o

<1 Whole Body o

<1

TABLE 15.9 ASSUMPTIONS USED IN THE ANALYSIS OF FUEL HANDLING ACCIDENT DOSES INSIDE CONTAINMENT Power Level Power Peaking Factor Shutdown time Number of Fuel Rods Assumed Failed Number of Fuel Rods in Core Fraction of Inventory in Failed Pins Released to Pool:

Noble Gases Iodines Fraction of Iodine in Pool Released from Pool Isolation Valve Closure Time Case I Case II Relative Concentrations, seconds per cubic meter 0-2 hours at 1350 meters 0-2 hours at 9654 meters 15-4 2900 Megawatts thermal

1. 65 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> 264 41,448 10 percent 10 percent percent 20 seconds 50 seconds 4.2 x 10-4

4. 1 x 10-5

lB.O REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS lB.2 Operating License Review lB. 2. 1 The Advisory Committee on Reactor Safeguards completed its review of the application for operating licenses for the North Anna Power Station, Units 1 and 2 at its January 6-B, 1977, meeting.

The Committee previously completed a partial review of the North Anna Power Station, Units 1 and 2 at,its October 14-16, 1976, meeting.

A copy of the Committee's letter dated October 26, 1976, concerning its partial review is attached as Appendix B.

A copy of the Committee's letter dated January 17, 1977, concerning its final review is attached as Appendix C.

We have considered the comments and recommendations made by the Advisory Committee on Reactor Safeguards.

The actions we have taken or plan to take in response to these comments and recom-mendations are described below.

Verification of Westinghouse 17 x 17 Fuel Assembly Design The Committee noted that a considerable portion of the Westinghouse 17 x 17 fuel assembly research and development programs had been completed, and had been evaluated and accepted by the NRC staff.

The Committee also stated that it wished to be kept informed on those matters still under review (see Advisory Committee on Reactor Safeguards' letter which is attached as Appendix B).

We stated in Section 4.4 of the Safety Evaluation Report that our review of departure from nucleate boiling data for assemblies with spacer spans of 22 inches, as presented in the Westinghouse Topical Report "Critical Heat Flux Testing of 17 x 17 Fuel Assembly Geometry with 22-inch Spacing," WCAP-B536, May 1975, was in progress.

Since the issuance of the Safety Evaluation Report, we have completed our review of this topical report and have determined that the information presented in this report is acceptable.

Our evaluation of this report is presented in a letter from John Stolz (NRC) to C. Eicheldinger (Westinghouse) dated February 11, 1977.

With the completion of the above item, our evaluation of the Westinghouse 17 x 17 fuel assembly verification tests is complete except for the Westinghouse 17 x 17 fuel assembly surveillance program (see Tables 4.1 and 4.2 of the Safety Evaluation Report).

With respect to the 17 x 17 fuel assembly surveillance program, the status of this program is the same as discussed in Sec.tion 4.2.2 of the Safety Evaluation Report with the exception that the two lead assemblies in Surry Unit 1 have completed their 1B-1

18.2.2 18.2.3 second fuel cycle and were discharged as planned.

Inspections of these assemblies after the first two cycles have revealed no anomalies.

We intend to keep the Committee informed regarding the Westinghouse 17 x 17 fuel assembly surveillance program.

Steam Generator and Reactor Coolant Pump Supports The Committee noted that two different steel specifications (ASTM A36-70a and ASTM A 572-70a) covered most of the material used for the steam generator and reactor coolant pump supports.

Toughness tests, not originally specified and not in relevant ASTM specifications, were made on those heats for which excess material was available.

The Committee also stated that "The toughness of the A 36 steel was good, but thn toughness of the A 572 steel was relatively poor at an operating temperature of 80°F.

The Applicant, therefore, proposes to operate so that all A 572 material is at 180°F or above.

He also plans periodic inspection of the A 572 members to the extent that they are accessible.

The Committee believes that increasing the operating temperature is an acceptable solution, but recommends that the operating temperature of the A 572 material be substantially above the proposed temperature.

The Committee believes also that it would be prudent not to permit pressurization of the primary system to sUbstantial levels while temperatures of the supports might be well below the operating temperature." The Committee further noted that the NRC staff has not completed its review of the repair of the steam generator and reactor coolant pump supports for Unit 2 (see the Advisory Committee on Reactor Safeguards' letter which is attached as Appendix B).

The resolution of these matters is discussed in Section 5.4.2 of Supplement No.3 to the Safety Evaluation Report and Section 5.4.2 of Supplement No.6 to the Safety Evaluation Report.

In those sections, we indicated in our conclusions that the operating limitations subsequently proposed by the applicant regarding temperature and pressure satisfy the concerns of the ACRS and are acceptable to the staff.

Anticipated Transients Without Scram The Committee recommends an early resolution of the matter of anticipated transients without scram for North Anna Units 1 and 2.

The Committee also stated that it wished to be kept informed (see the Advisory Committee on Reactor Safeguards' letter which is attached as Appendix B).

This matter is discussed in Section 7.2.4 of the Safety Evaluation Report.

We intend to keep the Committee informed regarding this matter.

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18.2.4 18.2.5 Generic Items The Advisory Committee on Reactor Safeguards noted that generic problems relating to light water reactors are discussed in the Committee's report, entitled, "Status of Generic Items Relating to Light Water Reactors:

Report No.4," dated April 16, 1976.

Those problems relevant to North Anna, Units 1 and 2, should be dealt with appropriately by the NRC staff and applicant as solutions are found.

The relevant items are 11-1, 2,3,4, 5, 6, 7, 9, 11; II A-l, 4, 5,6,7,8; II B-2; II C-l, 2, 3, 4, 5, 6, 7 (see the Advisory Committee on Reactor Safeguards' letter which is attached as Appendix B).

Appendix D to this report discusses the disposition and status of the indicated items.

Evaluation of Safety Factors for Safety-Related Systems During Safe Shutdown Earthquake Conditions The Committee noted that the applicant presented partial information concerning the calculated safety factors during safe shutdown earthquake conditions by some of the engineered safety features.

The Committee recommended that the NRC staff revi*ew this aspect of the design in detail and assure itself that significant margins exist in all systems required to accomplish safe shutdown of the reactors and continued shutdown heat removal, given a safe shutdown earthquake.

The Committee also believes that such an evaluation need not delay the start of operation of North Anna Power Station Units 1 and 2 and the Committee wishes to be kept informed (see the Advisory Committee on Reactor Safeguards letter which is attached as Appendix C).

We are presently conducting a seismic design review program for North Anna Units and 2.

This program includes the structures, systems and components which are required to accomplish safe shutdown of the plant and continued decay heat removal in the event of a safe shutdown earthquake.

We will also audit the design drawings, specifica-tions and design calculations for the above structures and systems, and the operability assurance program for active components thereof.

In conducting this evaluation, the general arrangement of these systems and components will be reviewed to determine sensitivity to seismic loadings.

The design adequacy and margin for selected components will be evaluated as a means of auditing the procedures employed.

For those items qualified by analysis, the stresses, strains, or deflections, as appropriate, will be evaluated in order to make a determination as to the contribution of the safe shutdown earthquake-produced loads to the overall loads and thus make a qualitative assessment of the indicated margins for seismically-produced loading.

Equipment and components, which have been tested instead of analyzed, will be similarly reviewed.

The basis for this review will be the seismic loading resulting from the design basis earthquake of 0.12 g for structures resting on competent rock and 0.18 9 for structures 18-3

18.2.6 resting on saprolite (weathered rock) as described in the Final Safety Analysis Report.

A 1 though our conc 1 us ions regardi ng the acceptabil ity of sei slili c design for North Anna Units 1 and 2 are as stated in the Safety Evaluation Report and its supplement, we will assure that additional safety margins exist as recommended by the ACRS.

Investigation of Construction Activities of North Anna Power Station Units 1 and 2 The Committee stated that the NRC staff has conducted and is continuing extensive investigation of construction activities of North Anna Units 1 and 2 and that these investigations have been separated into four phases.

The Committee also stated that the NRC staff has concluded that various items of noncompliance with NRC requirements have occurred and has defined a program to remedy the matter.

The Committee noted that it had had the benefit of a review and evaluation of this matter by its own consultant, who supports the adequacy of the NRC investigations and has made several recommendations, including one related to a program to ascertain that significant deficiencies do not exist in safety-related piping systems.

The Committee also noted that it concurs with the consu1tant' s recommendations and wishes to be kept informed regarding resolution of these recommendations (see Advisory Committee on Reactor Safeguards I letter which as attached is Appendix C).

On February 16, 1977, the Office of Inspection and Enforcement forwarded to the Virginia Electric and Power Company a series of questions and concerns raised by Mr. W. R. Gall, pertaining to Virginia Electric and Power Company's actions resulting from worker allegations and the Office of Inspection and Enforcement investigation findings.

Mr. Gall is a consultant to the Advisory Committee on Reactor Safeguards and stated his questions in a letter to Dr. David Okrent of the Aqvisory Committee on Reactor Safeguards dated January 3,1977.

In large measure, Mr. Gall IS concerns paralleled those identified in the Office of Inspection and Enforcement investigation and were subjects of corrective actions required as a result of the investigation.

Nevertheless, the Office of Inspection and Enforcement felt it was appropriate for Virginia Electric and Power Company to respond explicitly to each of Mr. Gallis concerns.

In a letter dated February 25, 1977, Virginia Electric and Power Company provided a response to each of the matters identified.

However, the Office of Inspec-tion and Enforcement notified Virginia Electric and Power Company that additional information was needed to supplement their response to five questions.

The additional information was subsequently provided by letter dated March 25, 1977.

One of Mr. Gall's questions dealt with the resolution of an investigation finding'"

related to questionable weld and/or radiograph quality acceptability for pipe

• Office of Inspection and Enforcement Investigation Report 50-338/76-28, 50~339/76-l6, North Anna Units 1 and 2, dated November 23, 1976, page 18, Item 12.

18-4

18.2.7 18.2.8 fabricated by Southwest Fabricating and Welding Company.

Part of the difficulty with the resolution of this matter has been the disagreement in interpretation of the radiographs in question.

Following Region II's review of earlier audits and their examination of the radiographs, the licensee agreed to cut out and examine several of the welds where interpretive disagreement existed. This effort resulted in con-firmation that some of the Southwest Fabricating and Welding Company shop welds indeed contained code-rejectable defects.

As a result of these findings, Virginia Electric and Power Company committed to perform a 100 percent audit of all welds where the radiograph evaluation sheets indicate weld profile problems of any kind.

The remaining welds (excluding those already covered) will be audited using an appropriate sample size.

Region II inspectors have reviewed the results of all the audits conducted to date, and the proposed corrective action.

The quality assurance procedures have been found acceptable.

The Office of Inspection and Enforcement will continue to monitor the correctiv~ actions in the field to assure that installed piping meets applicable requirements prior to making a finding of plant completion.

Based on discussions with Mr. Gall and the Advisory Committee on Reactor Safeguards, we believe that the answers and actions provided by the licensee have satisfied the concerns of Mr. Gall and the Advisory Committee on Reactor Safeguards.

Evaluation of Asymmetric Loads on Pressure Vessel Structures The Committee stated that the NRC staff has reported that the matter of asymmetric loads on pressure vessel structures is essentially resolved.

The Committee agrees that, subject to final evaluation by the NRC staff, this matter is in an acceptable status for-North Anna 1 and 2 (see Advisory Committee on Reactor Safeguards' letter which is attached as ~ppendix C).

The resolution of this matter is discussed in Section 3.9.4 of the Safety Evaluation Report and Sections 3.9.4 and 4.2.4 of this supplement.

Fire Protection System The Advisory Committee on Reactor Safeguards stated that the applicant is in the process of studying fire protection measures at the plant in accordance with the guidelines of Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976." The Committee indicated that the NRC staff had stated that North Anna 1 and 2 will be given priority in the evaluation of fire protection matters, and that most, if not all, improvements will be implemented prior to the start of 1~5

18.2.9 operation on the second fuel cycle.

The Committee further stated that they find this approach acceptable (see Advisory Committee on Reactor Safeguards' letter which is attached as Appendix C).

Subsequent to our evaluation of the North Anna Power Station Units 1 and 2 fire protection system reported in Section 9.5.1 of the North Anna Power Station Units and 2 Safety Evaluatlon Report, we issued revised fire protection guidelines, "Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1," dated August 23, 1976.

On Septemper 30, 1976, we transmitted Appendix A to Auxiliary and Power Conversions Systems Branch Technical Position to the applicant and requested performance of a fire hazards analysis and a reevaluation of the fire protection program, including a comparison with Appendix A.

On April 1, 1977, the applicant submitted the information requested in our letter, A site visit has been scheduled for the week of August 1, 1977 to review this matter.

We have given priority to evaluating this information to assure that most, if not all, improvements will be implemented prior to the start of operation on the second fuel cycle.

However, as stated in Section 9.5.1 of the North Anna Power Station Units 1 and 2 Safety Evaluation Report, we concluded that the present fire protection system desiqn is acceptable and therefore the plant can be safely operated prior to implementation of the improvements.

Long-Term Seal Capability The Committee noted that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on i nstrumentat i on and e 1 ectri ca 1 equ*ipment within containment which is susceptible to ingress of steam or water if the hermetic seals are either initially defective or should become defective as a result of damage or aging.

The Committee believes that appropriate test and maintenance procedures to assure continuous long-term seal capability should be developed (see Advisory Committee on Reactor Safeguards' letter which is attached as Appendix 0.

In response to the Committee's concern, the applicant, in a letter dated February 14, 1977, has provided its program to assure the continued service of the seals for all qualified Class IE transmitters located inside the containment.

The program includes implementation of the manufacturer's recommendations and a maintenance procedure to be utilized initially, and at the time of periodic calibrations.

The calibration interval as specified in the technical specification for such equipment does not exceed 18 months.

The program will include:

(1) Replacement of any seal broken onsite with a new seal.

(2) Installation of new seals according to manufacturer's procedures.

(3) Compliance with manufacturer's recommendation for the torque applied to certain equipment, such as covers, closures, and bolts.

18-6

(4) Replacement of all seals within their specified life as recommended by the manufacturer.

The applicant has stated its commitment to keep abreast of future generic information, such as results of testing to IEEE 323-74, for use in upgrading or revision to its procedures.

On the basis of our review, we conclude that the program adequately resolves the Committee's concerns in this regard and find it acceptable.

18-7

21.0 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS 21.3 Operating License The Safety Evaluation Report, dated June 4, 1976, addressed the financial protection and indemnity requirements for the North Anna Power Station, Units 1 and 2.

By publication in the Federal Register Volume 42, Number 74, April 18, 1977, 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," was amended to increase the amount of primary financial protection required for facilities having a rated capacity of 100 electrical megawatts or more from $125 million to $140 million, effective May 1, 1977.

On the basis of the above considerations and those identified in our Safety Evaluation Report we conclude that the presently applicable requirements of 10 CFR Part 140 have been satisfied and that prior to issuance of the operating license, the applicant will be required to comply with all of the provisions of 10 CFR Part 140 applicable to operating licenses, including those as to proof of financial protection in the requisite amount and as to execution of an appropriate indemnity agreement with the Commission.

21-1

22.0 CONCLUSION

S In Section 22.0 of Supplement No.6 to the Safety Evaluation Report, we stated that several items were still outstanding, and that satisfactory resolution of these items would be required before operating licenses for North Anna Power Station, Units 1 and 2 could be issued.

Three of these have been resolved, as reported in Sections 2.6.2, 4.2.4, and 5.2.8 of this supplement.

The remaining outstanding item which must be resolved and its present status is summarized below.

Resolution of this item will be discussed in a future supplement to the Safety Evaluation Report.

(1)

The applicant has provided additional information on the seismic and environ-mental qualification of seismic Category I instrumentation and electrical equipment.

Our evaluation of this information has not been completed (Section 3.10 of the Safety Evaluation Report and Section 3.10 of Supplement No.3 to the Safety Evaluation Report).

Subject to satisfactory resolution of the outstanding matters described above, the conclusions as stated in Section 22.0 of the North Anna Power Station, Units 1 and 2 Safety Evaluation Report are reaffirmed.

22-1

January 28, 1977 January 28, 1977 January 31, 1977 January 31, 1977 February 1, 1977 February 2, 1977 February 4, 1977 February 8, 1977 February 10, 1977 February 14, 1977 February 15, 1977 February 18, 1977 APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW Representatives from VEPCO & NRC meet in Bethesda, Md. to discuss the Technical Specification for Units 1 & 2.

VEPCO transmits supplemental affidavit to Westinghouse letter LP-592, dated January 14, 1977.

This affidavit applies to requests for withholding from public disclosure proprietary information involving the evaluation of seismic and LOCA effects on a fuel assembly.

Summary of Meeting on technical specifications held on January 28, 1977.

VEPCO letter concerning scale model testing concerning LHSI pumps.

Division of Project Management letter requesting additional information - instrumentation and controls.

Issuance of Supplement No. 6 to the North Anna Power Station, Units 1 & 2 Safety Evaluation Report.

VEPCO letter advising of changes to a report transmitted on December 10, 1976, Class IE Electrical Distribution System Evaluation for Degraded Voltage Conditions, North Anna, Unit 1.

Division of Project Management letter transmitting Amendment No. 6 to the North Anna, Units 1 & 2 Safety Evaluation Report to applicant.

VEPCO letter concerning ATWS.

Division of Project Management letter withholding Westinghouse information from public disclosure as proprietary.

VEPCO letter transmitting additional information on their request for extension of construction completion dates for North Anna, Units 1 & 2.

Division of Project Management letter concerning the withholding of information from public disclosure.

A-I

February 14, 1977 February 15, 1977 February 16, 1977 February 17, 1977 February 24, 1977 February 25, 1977 March 2, 1977 March 4, 1977 March 4, 1977 March 7, 1977 March 7, 1977 March 7, 1977 March 9, 1977 March 15, 1977 VEPCO letter concerning long term seal capability of instrument transmitters located in the containment.

VEPCO letter transmitted information as to the delay in the completion of North Anna Units 1 & 2.

VEPCO letter concerning the clearance between the Band C main steam safety valve manifolds of North Anna 1 & 2.

VEPCO letter concerning specific exceptions for the testing of pumps and valves required by ASME Code Class 1, 2 and 3.

Division of Project Management letter transmitting a copy of letter to ACRS from B. Rusche and reply from ACRS concerning the ACRS letter of January 17, 1977.

Division of Project Management letter concerning physical protection (signed by Ben C. Rusche).

Order issued by the AS&LB advising that Contentions 1 and 2 of Intervenor Arnold have been adopted as the subject of Board inquiry.

An evidentiary session beginning May 2, 1977 will present these contentions as evidence.

VEPCO letter advising that a specification for the measurement of the suspended solids and turbidity in the effluent from the horizontal drains beneath the service water pumphouse will be included in the Technical Specifications for North Anna.

VEPCO letter concerning an error in NRC Comment 10.24 response.

VEPCO letter correcting the setpoints for the seal water standpipe level alarms originally forwarded 12/23/76.

VEPCO letter concerning the liquid radwaste disposal system.

VEPCO letter advising that a response to Comment 3.73 will be submitted by March 31, 1977 and a response to Comment 3.74 by April 15, 1977.

VEPCO letter concerning expanding the spent fuel storage capability for North Anna, Units 1 & 2.

VEPCO transmits the 1976 Annual Report.

A-2

March 15, 1977 March 16, 1977 March 16, 1977 March 17, 1977 March 17, 1977 March 22, 1977 March 23, 1977 March 30, 1977 March 31, 1977 March 31, 1977 April 1, 1977 April 4, 1977 April 6, 1977 Division of Project Management letter concerning instrument trip setpoint values.

VEPCO letter advising they will reply to NRC overpressurization letter by March 31, 1977.

Division of Project Management letter concerning Fuel Handling Accident.

VEPCO letter transmits the reply to NRC Comment 5.80 (dynamic analyses of the potential effects of a postulated loss-of-coolant acci dent.

Division of Project Management letter transmitting a report entitled "The Undrained Cyclic Triaxial Response of a Saprolitic Soi 1. II Division of Project Management letter concerning Monitoring Releases from North Anna, Units 1 & 2.

VEPCO letter concerning specific exemptions for the testing of ASME Code Class I, II and III pumps and valves.

Order issued by the Atomic Safety and Licensing Board setting the date of the evidentiary hearing for May 2, 1977 at 9:30 a.m.,

City Hall, Charlottesville, Virginia.

VEPCO transmits the response to Comment 3.73 (seismic Category I, electrical and mechanical equipment).

VEPCO transmits Amendment No. 61 to the Final Safety Analysis Report.

This amendment consists of revised pages of additions and deletions.

VEPCO letter concerning fire protection program for the North Anna Power Station, Units 1 and 2.

VEPCO letter advising they will respond to the NRC's request for additional information regarding a potential refueling accident by April 22, 1977.

VEPCO letter concerning Class I pipe breaks.

A-3

April 8, 1977 April 14, 1977 April 14, 1977 Apri 1 15, 1977 April 15, 1977 April 19, 1977 April 22, 1977 April 25, 1977 Apri 1 27, 1977 May 4, 1977 May 6, 1977 May 13, 1977 May 16, 1977

. Division of Project Management letter requesting additional information concerning the reactor coolant system.

Order issued by the Atomic Safety and Licensing Board.

Sun Ship has until May 6, 1977 to serve affidavits stating its position.

VEPCO letter transmitting response to Staff Comment 5.81.

Division of Project Management letter to Westinghouse concerning withholding information from public disclosure - AW-76 North Anna, Units 1 & 2.

VEPCO letter concerning completion of the horizontal drain system beneath the service water pump house.

Division of Project Management letter transmitting Order extending construction completion dates for North Anna 1 & 2 and Staff Evaluation, Negative Declaration and Environmental Impact Appraisal.

Division of Project Management letter to all utilities concerning standard format for meteorological data on magnetic tape.

VEPCO letter concerning safety-related equipment temperatures during a main steam line break.

VEPCO letter transmitting responses to Division of Project Management request for additional information concerning the preservice inspection of North Anna Unit 1.

Division of Project Management letter concerning Intrusion Detection Systems Handbook (Regional Meetings on 10 CFR Section 73.55).

VEPCO letter requesting a 40 year operating license (from date of issuance of OL).

VEPCO letter transmitting a revised Final Safety Analysis Report Figure 11.2.2-3.

Division of Project Management letter requesting additional information concerning site characteristics.

A-4

May 17, 1977 May 17, 1977 May 18, 1977 May 20, 1977 May 24, 1977 May 25, 1977 May 25, 1977 May 26, 1977 May 27, 1977 May 27, 1977 June 9, 1977 June 2, 1977 June 6, 1977 Order issued by the Atomic Safety and Licensing Board ~etting forth date for the evidentiary hearing.

The resumption of the North Anna hearing will be held May 31, 1977 at 9:30 a.m. in the City Council Chambers, 2nd Floor of City Hall, 7th and Main Streets, Charlottesville, Va.

VEPCO transmits Certificate of Service for Amendment No. 62.

VEPCO letter concerning MULTIFLEX computer code.

Order issued by the Atomic Safety and Licensing Board.

Ruling on documents submitted by Mrs. Arnold will be reserved until the hearing on May 31, 1977.

VEPCO letter transmitting the Emergency Plan, Change No.1.

VEPCO letter transmitting information concerning instrument trip setpoi nts.

VEPCO letter concerning the physical security plan for North Anna.

VEPCO letter advising they anticipate providing NRC with informa-tion concerning stresses in service water piping due to the settle-ment of the service building by May 31, 1977.

VEPCO letter transmitting the response to Staff Comment 2.22 concerning service water piping due to the settlement of the service building.

VEPCO letter transmitting a report entitled, "Report on the Phase II Network Seismic Monitoring Program - December 3, 1976 Through April 2, 1977 - North Anna Power Station."

Division of Project Management letter requesting additional information - reactor coolant system.

VEPCO letter transmitting preliminary test data and test setup drawings that were presented at the meeting held on May 18, 1977 in Holden, Massachusetts with representatives from VEPCO, Stone &

Webster, Alden Research Laboratories, and NRC.

VEPCO letter transmitting additional data and the original LOCA test results for response to Comment 7.17.

A-5

June 7, 1977 June 9, 1977 June 16, 1977 June 21, 1977 June 21, 1977 June 22, 1977 June 22, 1977 June 22, 1977 June 24, 1977 June 24, 1977 June 24, 1977 June 29, 1977 June 29, 1977 VEPCO letter transmitting the information available for review to changes made to Section 3.8 of the Final Safety Analysis Report.

Additional information is being developed and will be transmitted when complete.

VEPCO letter transmitting revised pages to 3.8-31 and 3A.15-1 of Amendment No. 61 to the Final Safety Analysis Report.

VEPCO letter transmitting the results of a detailed evaluation of the potential consequences of a fuel handling accident inside containment.

VEPCO letter transmitting the results of an analysis of a main steam line break with the assumption of no moisture-entrainment in the steam released to the containment.

VEPCO letter transmitting reports from Dupont de Nemours & Company and Union Carbide on "Kapton" and Polysulfone."

VEPCO letter transmitting a supplement to the response to Staff Comment 6.135.

VEPCO letter concerning specific exemptions to Section XI of the ASME Code.

VEPCO letter transmitting Revision 4 to NA-TR-1002, "Safety Related Electrical Schematics."

VEPCO letter transmitting a revised response to NRC Staff Comment 5.81.

VEPCO letter concerning the preservice inspection program for North Anna Unit 2 and initial inservice inspection programs for North Anna Units 1 and 2.

VEPCO letter transmitting additional seismic qualification information from the vendors.

VEPCO letter concerning Quench Spray CQS) starting times used in the Final Safety Analysis Report for North Anna, Units 1 and 2.

VEPCO letter concerning cracks in the leveling mat in Units 3 and 4 and in the Visitor's center at North Anna.

A-6

June 30, 1977 July 1, 1977 July 8, 1977 July 12, 1974 July 14, 1977 July 20, 1977 VEPCO letter transmitting the response to NRC Staff Comment 3.74.

VEPCO letter transmitting the response to Regulatory Position 7.6 concerning temperature monitoring of areas containing Class IE equipment.

VEPCO transmits Amendment No. 63 to the Final Safety Analysis Report.

This Amendment consists of substitution and addition of pages.

DPM letter requesting additional information concerning the service water piping between the service building and the main steam valve house on Unit No.2.

DPM letter requesting additional information regarding the preservice and inservice inspection programs for North Anna Units 1 & 2.

Letter from the Advisory Committee on Reactor Safeguards concerning the Review of the North Anna Power Station, Units 1 and 2.

A-7

APPENDIX B UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. D. C. 20555 October 26, 197ft Honorable Marcus A. Rowden Chairman U.s. Nuclear Regulatory commission Washington, DC 20555 SUBJEcr:

REPORI' ON PARrIAL REVIEW OF NORm ANNA PCMER STATION UNITS 1 AND 2

Dear Mr. Rowden:

At its 198th meeting, October 14-16, 1976, the Advisory Committee on Reactor Safeguards completed a partial review of the application of the Virginia Electric and Power Company for authorization to operate the North Anna Power Station, Units 1 and 2. The project was previously considered at Subcommittee meetings in Washington, D.C., on July 7,1976, August 11, 1976, and October 13, 1976, and at the 196th meeting of the Committee on August 12-14, 1976. Tours of the facility were made by Subcommittee members on February 3, 1976 and May 27, 1976.

During its review, the Committee had the benefit of discussions with representatives and consultants of the Virginia Electric and Power Company, the Westing-house Electric Corporation, the Stone and Webster Engineering Corporation, the Sun Shipbuilding and Dry Dock Company, the North Anna Environmental Coalition, and the Nuclear Regulatory Commission (NRC) Staff. The Committee also had the benefit of the documents listed. The Committee discussed the application for a construction permit for the North Anna Power Station, Units 1 and 2, in its report of August 20, 1970. The Committee also discussed matters related to fault zones under or adjacent to th~ foundations of North Anna Power Station, Units 1, 2, 3, and 4 in its report of April 15, 1974.

The site is located on 1,075 acres on the shores of Lake Anna in Louisa County, Virginia, about 24 miles southwest of Fredericksburg, Virginia, and 40 miles north-northwest of Richmond, Virginia.

The Committee has not completed its review of North Anna Units 1 and 2 with regard to the following matters:

adequacy of seismic design basis and seismic design; loss-of-coolant accidents and emergency core cooling; quality assurance and contiol in on-site fabrication and installation; asymmetric loads on pressure vessel structures arising from certain postulated pipe breaks; and plans for upgrading protection against fires.

B-1

Honorable Marcus October 26, 1976 Also, in Supplement Noo 3 to the Safety Evaluation Report, the NRC Staff has identified several items to be resolved, and the Committee has a few remaining items relating to systems interactions on which it wishes further information.

An unexpected amount of settlement has been experienced by the service water pump house for the North Anna Units 1 and 2.

Some cracking of the pump house walls has resulted. The Applicant has examined the causes of the settlement and has made design changes, including the provision of flexible expansion coupling between the piping and the pump house to accommodate additional settlement. The NRC Staff is satisfied with the re-analysis of stresses and, except for review of the design of a system of well points for ground water control, believes the situation is currently acceptable.

Future settlement, which should be modest, will be monitored carefully in accordance with technical specifications to be prepared. The Committee concurs with the NRC Staff.

The Applicant has submitted a revised probable maximum flood analysis.

The NRC Staff has reviewed the analysis and found it acceptable with the inclusion of a technical specification to restrict facility operation when the water level in Lake Anna exceeds an elevation of 256 feet above mean sea level. The Committee concurs.

The North Anna Power Station, Units I and 2 will employ a l7x17 fuel assembly similar to that employed in Beaver Valley Unit 1. A consider-able portion of the Westinghouse research and development program on these assemblies has been completed, and has been evaluated and accepted by the NRC Staff. The Committee wishes to be kept informed on those matters still under review.

The steam-generator and reactor-coolant-purnp supports are constructed of heavy rolled steel shapes and thick plate. After delivery of these structures at the site, the Applicant found many weld defects and pro-ceeded to remove all welds and to reweld the supports. The Unit I steam-generator supports had been installed and were rewelded in place, which made it necessary to substitute peening for thermal stress relieving.

The Committee finds this procedure acceptable.

The Unit 2 supports were rewelded in the shop and thermally stress relieved. The NRC Staff has not completed its review of this unit.

Two different steel specifications (ASTM A36-70a and ASTM A572-70a) covered most of the material used for the supports. Toughness tests, not originally specified and not in the relevant ASTM specifications, were made on 8-2

Honor able Marcus October 26, 1976 those heats for which excess material was available. The toughness of the A36 steel was good, but the toughness of !Be A572 steel was relatively poor at an operating temperature of 80F. The Applicant, therefore, propoSes to operate so that all A572 material is at 1800F or above.

He also plans periodic inspection of the A572 members to the extent that they are accessible. The ComRdttee believes that increasing the operating temperature is an acceptable solution, but recommends that the operating temperature of the A572 material be substantially above the proposed temperature. The Comndttee believes also that it would be prudent not to permit pressurization of the primary system to substantial levels while temperatures of the supports might be well below the operating temperature.

The NRC Staff is satisfied with regard to the Emergency Plan, and the Applicant has made considerable progress in providing instrumentation to follow the course of an accident.

The Committee recommends an early resolution of the matter of anticipated transients without scram for North Anna Units 1 and 2. The Comndttee wishes to be kept informed.

Other generic problems relating to large water reactors are discussed in the Committee's report, entitled "Status of Generic Items Relating to Light water Reactors: Report No.4," dated April 16, 1976. Those problems ~elevant to North Anna, Units 1 and 2, should be dealt with appropriately by the NRC Staff and Applicant as solutions are found.

The relevant items are: II-I, 2, 3, 4, 5, 6, 7, 9, 11: I IA-l, 4, 5, 6, 7, 8: IIB-2; I IC-l, 2, 3, 4, 5, 6, 7.

The ACRS believes that, if due regard is given to the items mentioned and subject to satisfactory resolution of those matters still under re-view and to satisfactory completion of construction and pre-operational testing, there is reasonable assurance that the North Anna Power Station, Units 1 and 2 can be operated at power levels up to 2775 MW(t) without undue risk to the health and safety of the public. The Com-mittee will report in the future on those matters for which its review is not yet oorrplete.

Addi tional C01IIIIents by Dr. Spencer H. Bush are presented on the following page.

Sincerely yours, fdruk 9,/,~~

B-3 Dade W. Moeller Chairman

Honorable Marcus October 26, 1976 Additional Comments by Member Spencer H. Bush These additional comments are directed to what appears to be the NRC Staff's position regarding acceptance of operation with the North Anna, Units 1 and 2 steam-generator and reactor-coolant-pump supports at or below temperatures of l800F.

I find it difficult to accept system pres-surization to substantial levels while temperatures of the supports might be well below those suggested as "equilibrium", e.g., <1800F' temperature.

I do not consider it unreasonable to require that the minimum temperatures of the supports be at a level of 225-250Of, obtainable by methods such as electric "trace" heating. The combined benefits of operation in the elastic-plastic fracture mechanics regime, major increase in critical flaw size and minimization of fast fracture propagation, admittedly rep-resent conservatisms, but these conservatisrns can be achieved relatively easily with no apparent adverse degradation mechanisms.

Since we do not have complete impact or fracture mechanics data, equilibrating at 225-2500P prior to pressurizing fully is recognized as conservative, but is considered desirable.

8-4

Honorable Marcus October 26, 197h REFERENCES

1.

Final Safety Analysis Report, North Anna Power Station, Units 1 and 2, with Amendments 1 through 56.

2.

Safety Evaluation Report related to the operation of North Anna Power Station, Units 1 and 2, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, with Supplements 1, 2, and 3.

(NUREG-0053)

3.

Letter dated October 8, 1976, from Ernst Volgenau, Director, Office of Inspection and Enforcement, USNRC, to R.F. Fraley, Executive Director, ACRS,

Subject:

"Comments Regarding North Anna Nuclear Plant".

4.

North Anna Environmental Coalition (NAEC) letters dated August 17, 1976 and September 1, 1976 and NAEC Statement of August 11, 1976 continued on October 13, 1976.

5.

"Interim Report on the Examination of Core Sarrples from Reworked Steam Generator Supports of VEPCO, North Anna", William S. Pellini, April 8, 1976.

6.

"The Safety of Steam Generator Support Structures for North Anna, Units 1 and 2", J.D. Harrison and R.E. Dolby, for Sun Shipbuilding and Dry Dock Company, May 1976.

7.

"Additional Information found in VEPCO and Stone and Webster files",

3 pp., Sun Shipbuilding and Dry Dock Company.

8.

"The Safety of Steam Generator Support Structures for North Anna, Units 1 and 2", Sun Shipbuilding and Dry Dock Corrpany, May 20, 1976 i' with Appendix 1, plus a one-page "Final Note".

9.

"Book 1, Summary of Information on Core Samples Including Source, Dimensions", (with 30 pages of photographs), Sun Shipbuilding and Dry Dock Company, May 20, 1976.

10.

"Book 2, Photographic Documentation of Defects in Core Samples",

(with 30 pages of photographs) Sun Shipbuilding and Dry Dock Company, May 20, 1976 B-5

Honorable Marcus Octoher 26, 1976 REFERENCES (con't)

11.

"The Safety of Stermt (,,enA!"ator Support Structures for North Anna, Dnits 1 and 2'e by Sun Ship:>uilding and Dry Dock Corrpany, July 7, 1976.

12.

liThe Safety of Ste;;un Generator Support Structures for North Anna, Units 1 and 2" Statement before the ACRS by Sun Shipbuilding and Dry Dock Company, October 13 f 1976.

13. "Further Comments on the Safety of the North Anna Support Structures",

LD 22955/2, June 1976, J.D. Harrison and R.E. Dolby, the Welding Institute, (for Sun Shipbuilding and Dry Dock Ltd.).

14.

"Catalog of Brittle Failures of Bridges and Other Related Structures, and Brittle Failures of Other Items Recorded at Higher Temperatures",

VEPCO report to ACRS North Anna Subcommittee, October 13, 1976.

15. "Test Data for Materials in North Anna Units 1 and 2 Steam Generator and Reactor Coolant Pump Supports", VEPCO report to ACRS North Anna Subconmittee, October 13, 1976.

16.

"VEPCO North Anna Units 1 and 2 Support Structures, Discussion of Fracture Mechanics Studies Presented by Various Parties", H.T. Corten, October 1976.

17. "Repairs, Inspection and Quality Assurance, Steam Generator and Reactor Coolant Pump Repair Program", VEPCO report to ACRS North Anna Subcommittee, October 13, 1976.

B-6

APPENDIX C UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 January 17, 1977 Honorable Marcus A.. Rowden Chairman u.s. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

REl?ORl' 00 NORm ANNA POiER STATIOO, UNITS 1 AND 2

Dear Mr.. Rowden:

At its 20lst meeting, January 6-8, 1977, the Advisory COImIittee on Reac-tor Safeguards completed its review of the application of the Virginia Electric and Power Company for a license to operate North Anna Power Station, Units 1 & 2. This project was also considered during a Subcom-mittee meeting held in Washington, D.C., on January 5, 1977.. The Com-mittee previously completed a partial review of this project at its 198th meeting, October 14-16, 1976, as discussed in its report to you, dated October 26, 1976. During its review, the Comnittee had the benefit of discussions with representatives and consultants of the Virginia Electric and Power Company, the Westinghouse Electric Corporation, the Stone and Webster Engineering Corporation, and the Nuclear Regulatory Coomission (NRC) Staff. The Comnittee also had the benefit of the documents listed..

In its report of October 26, 1976, on North Anna, Units 1 &: 2, the ACRS had not conpleted its review of the adequacy of seismic design bases and seismic design; loss-of-coolant accidents and emergency core cooling; quality assurance and control of on-site fabrication and installation1 asymmetric loads on pressure vessel structures arising from certain pos-tulated pipe breaks; and plans for upgrading protection against fires.

The NRC Staff has now completed its review of the Stafford fault zone and concluded that the available geological and seismological information supports the conclusion that, the Stafford fault zone is not capable with-in the meaning of ~ix A to 10 CFR Part 100, and that the available information does not warrant any change in the previously approved seismic design bases for North Anna 1 and 2. Representatives of the u.s.

Geological Survey concurred that there exists no definitive information showing significant movement during the last million years and that the fault is not capable. Consultants to the ACRS concur with this interpre-tation. While they generally find the current design bases acceptable for C-1

Honorable Marcus January 17, 1977 the already constructed North Anna plants, they have recommended that, in view of the uncertainties of knowledge concerning the sources of earthquakes in the Eastern United States, a minimum safe shutdown earth-quake (SSE) of 0.2g acceleration should be utilized for new plants for which construction permit applications are submitted in the future.

The Applicant presented partial information concerning the calculated safety factors during safe shutdown earthquake conditions for some of the engineered safety features.

The Coomi ttee recommends that the NRC Staff review this aspect of the design in detail and assure itself that significant margins exist in all systems required to acconplish safe shutdown of the reactors and continued shutdown heat removal, given an SSE.

The Ccmmi ttee believes that such an evaluation need not delay the start of operation of North Anna 1 and 2. The Committee wishes to be kept informed.

The NRC Staff has now completed its review of emergency core cooling system performance and found it to be acceptable.

The Committee con-curs.

The NRC Staff has conducted and is continuing extensive investigation of (.~nstruction acti vi ties of North Anna Units 1 and 2. These investi-gat ions have been separated into four phases:

1. investigation of specific allegations made by three individuals of faulty construction practices; 2..

a detailed inspection of certain safety-r~lated piping not directly implicated in the original allegations but which was potentially subject to similar problems;

3. detailed monitoring of the nondestructive preservice baseline examination of selected"welds in safety-related piping by the Licensee and his contractors; and

4.

inspections of the performance of selected components in specific piping systems during the preoperational testing program.

The NRC Staff has concluded that various items of non-compliance with NRC requirements have occurred and has defined a program to remedy the matter.

The Coomi ttee has had the benefit of. a review and evaluation of this matter by its own consultant, who supports the adequacy of the NRC C-2

Honorable Marcus January 17, 1977 investigations and has made several recommendations, including one related to a program to ascertain that significant deficiencies do not exist in safety related piping systems e The ACRS concurs. The Committee wishes to be kept informed regarding resolution of these recom:nendations.

The NRC Staff has reported that the matter of asymmetric loads on pres-sure vessel structures is essentially resolved. The ACRS has had the benefit of meetings of an Ad Hoc Working Group on this general subject, in Toronto on August 5, 1976, and in Los Angeles an December 1, 1976.

The Committee agrees that, subject to final evaluation by the NBC Staff, this matter is in an acceptable status for North Anna 1 and 2.

The Applicant is in the process of studying fire protection measures at the plant in accordance with the guidelines of Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1. The NRC Staff has stated that, as a plant about to come into operation, North Anna 1 and 2 will be given priority in the evaluation of fire protection matters, and that most, if not all improvements will be implemented prior to the start of operation on the second fuel cycle. The Committee finds this approach to be acceptable.

The Committee notes that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrumentation and electrical equipment within containment which is susceptible to ingress of steam or water if the hermetic seals are either initially defective or should be-come defective as a result of damage or aging. The Committee believes that appropriate test and maintenance procedures to assure continuous long-term seal capability should be developed.

The ACRS believes that, if due regard is given to the items mentioned above and in its report of October 26, 1976, and subject to satisfactory completion of construction and preoperational testing, there is reason-able assurance 'that the North Anna Power Station, Units 1 and 2, can be operated at power levels up to 2775 MWt without undue risk to the health and'safety of the public.

C-3 2:J;t.el Y yq""""~II.AII"""""",,---

M. Bender Chairman

Honorable Marcus

Attachment:

Report of WeRe Gall, ACRS Consultant, dated January 3, 1977,

Subject:

Review of Allegations and Inspectors Findings as Reported in NRC In-vestigation Report 150-338/76-28, 50-339/76-16 North Anna, Units 1 and 2.

REFERENCES :

January 17, 1977

1. North Anna Power Station, Units 1 &:.2 Final Safety Analysis Report, with Amendments 1 through 60e

2. Safety Evaluation Report (NUREG-0053) related to operation of North Anna Power Station, Units 1 and 2, with Supplements 1 through 5.

3. Virginia Electric and Power Company (VEPOO) letter Serial No. 338 to Mr. Benard C. kusche, ONRR, NRC, dated November 24, 1976, on environmental testing of safety related instrumentation.

4..

VEPCO letter Serial No.. 350 to Mr" Benard C.. Rusche, ONRR, NRC, dated November 30, 1976, forwarding a document entitled, "Safety Related Equipment Temperature Transients During the Limiting Main Steam Line Break. III

5.

VEPCO letter Serial No.' 346 to Mr.. Benard C. Rusche, ONRRI" NIC, dated November 30, 1976, on measures considered for use at North Anna re overpressurization events.

6.

VEPCO letter Serial No. 316A, dated December 3, 1976,.re model testing of LBSI pumps.

7..

VEPCOletter Serial No. 298/102276, dated December 16, 1976, contain-ing information on LOCA effects on reactor fuel. (Westinghouse PRO-PRIETARY)..

8.

NRC letter of December 14, 1976, from D.. B. Vassallo to Dr. Dade W.

Moeller, Chairman, ACRS, subject "Staff Report - Assessment of the Stafford Fault Zone. w

9. NRC memo dated Decenber 2, 1976, from Dudley Thonpson and Boyce H. Grier to Ernst Vo1genau, I&E, subject, "Transmittal and Evaluation of In-vestigation Report, No. 50-338/76-28, 50-339/76 North Anna Units 1 and 2.1\\1

10. VEPCO letter Serial No. 371, dated December 9, 1976, forwarding a copy of VEPCO W s reply to E.. Vo1genau re I&:E Investigation Report Number 50-338/76-28 and 50-339/76-16.

11. NRC letter dated December 6, 1976 from E. Volgenau, I&E, to VEPCO Attn: Mr. T. Justin Moore, President referring to the I&E investi-gation of construction activities at North Anna 1 and 2 forwarding a ~otice of Vio1ationP, and a nNotice of Proposed Imposition of Civil Penalities. R C-4

Honorable Marcus January 17, 1977 REFERENCES (con It)

12. USNRC q IE Investigation Report 50-338/76-28, 50-339/76-16,

Subject:

"Investigation of alleged discrepancies in the construction and quality control program for piping installation at the North Anna Power Station."

13. VEPCO letter serial 390 to Dr. Dade W. Moeller, Chairman, ACRS, for-warding a copy of Mr. T. Justin Moore's letter of Decemer 23, 1976 to Dr. Ernst Volgenau re the North Anna investigation.

14.. VEPCO letter Serial No. 391, dated January 4, 19771' providing infor-mation re concerns related to auxiliary power and containment systems.

15. North Anna Environmental Coalition (NAEC) letter dated January 5, 1977, to Dr.. Dade W. Moeller and Dr.. David Okrent, ACRS, requesting that certain items be made a part of the record of the January 6-8, 1977 I ACRS meeting.

16 e NAEC letter dated January 7, 1977, to Dr.. Dade W. Moeller and Dr 0 David Okrent, ACRS, adding two additional items to the list submitted in the NAEC letter of January 5, 1977.

C-5

Dr. David Okrent Energy and Kinetics Department 5532 Boetler Hall School of Engineering and Applied Science University of California Los Angelas, CA 90024

Dear Dr. Okrent:

Oak Ridge National Laboratory P.O. Box X Oak Ridge, Tennessee 37830 Jcmuary 3, 1977

Subject:

Review of Allegations and Inspectors Findings as Reported in NRC Investigation Report #50-338/76-28, 50-339/76-16 North Anna, Units 1 and 2 The purpose of this memo is to transmit my conclusions and recommendations regarding the reported allegations and the inspectors' findings as reported in the NRC Investigation Report on North Anna Units 1 and 2.

It is my opinion that the investigation of specific allegations as covered in the report has been sufficiently thorough to provide an evaluati9n of probable validity of the allegations and their possible effect on the integrity of the system.

My comments are based on the study of their report supplemen~ed by two visits to the.plant site, discussions with Stone & Webster, and Vep~o staff members and with NRC Inspection and Enforcement staff in Bethesda, Marylar The report presents the inspectors' findings and explains the method of investi-gation upon 'tvhich their conclusions are based. It does not in all ~ases cover corrective actions that have been taken o~ that may be proposed as a result of the findings.

In some cases, an evaluation of the integrity of the affected systems will depend upon the corrective action that is proposed or taken.

I have-the. follmving specific items of concern:

1.

Cutting of Rebar Apparently the cutting of rebar became so prevalent that Stone & Webster themselves became concerned about it and initiated actions to curtail or control such cutting.

But prior to initiation of those actions, various methods of cutting rebar were used, some of which may be detrimental to the properties of the concrete and particularly the use of ca.rbon-arc, o~Jgen-flame cutting and welding rod processes which could provide high levels of heat input to the concrete.

The proposed analysis described in the licensee's response may be sufficient to establish the adequacy of the rebar but further evaluation may be necessary to determine if the concrete was damaged.

C-6

Dr. David Okrent 2

January 3, 1977

2.

Allegations Concerning Fake Anchor Bolts The interference between anchor bolts and rebar may be responsible for the faking of two anchor bolts which were reported in allegations B-7 and P-l,. It is also possible that some anchor bolts were cut to avoid cutting rebar which would result in the length of bolts being shorter than specified, thus affecting the strength of the anchor. It is my under-standing that ultrasonic measurements will be made to detect those bolts which were shortened.

I recommend that an evaluation be made by the licensee to determine the adequacy of any bolts which are found to be short.

3.

Welders Performing Welds Outside the Range of Their Qualifications It was established by the inspector that 30 Class 1 type welds in Units I and 2 were performed by welders qualified for thinner sections.

I believe all of these welds are identified in QC records and that all of the welds have been examined by radiography and found to be acceptable.

Tne welders were qualified on thinner sections than those cited. The acceptance of radiographed production welds as qualification weld~ may be a valid pro-cedure provided the initial weld performed outside the previously qualified thicbless range meets the requirements of QW-30l.4, QW-302.2, and QW-305,2 of Section IX of the ASME Code,is acceptable without weld repairs, and also provided the complete weld was performed by the same.welder.

One instance was observed during a visit to North Anna site in which a single welded joint in a primary coolant loop had the weld identification nlli~bers of twelve different welders.

I believe this would not be a satisfactory way to qualify any of the welders.

4.

32-Inch Main Steam Riser in Safety Valve Station The circumferential joints performed in the modification to the 32-inch

~ain steam riser have been evaluated and seem to be satisfactory except in the matter of mismatch.

Permissible mismatch al'lowed by the Code is 3/32 of an inch.

The inspector determined in at least one case a maximum of 5/16 inch mismatch.

This is a factor of 3 over that permitted by the Code and assuming that the stress in the longitudinal direction was judged satisfactory with the permissible misalignment, the affect of mismatch

~V'ould increase that stress by a factor of 3 in the case of 5/16 inch mismatch This pipe is probably subject to extreme axial compressive loads when the safety valves operate.

I have not found evidence that a failure of this pipe could not cause a pipe whip in the main steam. valve house which would react on the penetration of the containment wall sufficiently to breach the containment.

5.

Welding Electrodes There were two items concerning improper storage of welding electrodes and one concerning use of welding electrodes prior to receipt of material certifications for them. It would be difficult if not impossible to determin4 C-7

Dr. David Okrent 3

January 3, 1977 whether welding electrodes which had been stored overnight or over a shift outside of the required drying ovens have been used in welds or in which welds they may have been used.

Furthermore. detection of the effect of excessive moisture or other contaminates principally hydrogen. embrittle-ment, would be difficult to detect by means of radiography.

It would be desirable to establish that all electrodes held over were being kept for personal use.

The use of ~.,elding materials prior to receipt of proper documentation requires verification after the weld material lIas been used and could result in a determination that incorrect materials were used.

This verification should be made in all cases where this was done and corrective action taken tvhere necessary.

6.

Defective Shop Welds Two instances are reported in the inspector's findings - Items 2-C and 2-K in which noncorming shop welds in pipes performed by others were discovered by Stone & l~ebster quality control.

Corrective action is not indicated.

Of particular interest is the disposition of those Southwest Fabricating and Welding Company's pipe welds which were not included in the 1. 5% sample examined by Stone & Hebster QC and the applications in which they were used.

Approximately half of the 1.5% sample were found to be nonconforming and presumably were repaired.

Corrective action should be applied to all other,velds represent.ed by those samples to assure con-formance with quality requirements in the Class 2 system.

7.

Improper Identification of Haterials and Parts In the four reported incidents of improper identification of materials, it was possible to establish acceptability for the materials affected.

Can it be established with a reasonable degree of confidence that these reported instances are the only ones in which materials were improperly identified, or that all materials installed are in compliance with require-ments?

8.

Conclusions A.

I agree in general with the Evaluation of Findings enclosed with the transmittal of the report of the investigation.

Corrective action to correct the deficiencies in the quality assurance program must be augmented by actions to verify quality of construction already completec Phases 2, 3 and 4 of the investigation> I believe. were conceived for this purpose.

Phase 2 has been completed with some deficiencies yet to be resolved.

Phases 3 and 4 should form a basis for establishing the integrity of the system.

An effort should be made to establish that welding electrodes which were improperly stored were not used in welding of safety related systems, or that if used the effects ~vill not compromise safety.

C-8

Dr. David Okrent 4

January 3, 1977 B.

It is my opinion that all of the identified "quality of work" non-conformities can be corrected by corrective action.

Some of the quality control non-conformities affecting work that is already com-pleted cannot be corrected now, but the quality of the work affected may be verified by preoperational testing, and examination, and if deficient it can be corrected.

C.

The "unresolved items" listed in Part E of the report can also be resolved by appropriate corrective actions.

D.

The licensee's quality assurance program has. not functioned in accordance with established procedures and requirements in some cases.

This leads to con.cern about possible undetected non-conformances *. Phases 1 and 2 of the investigation constitute a thorough study of these possible deficiencies in the important safety related systems and it resulted in disclosure of some additional deficiencies which should be corrected.

E.

In my study of the report and my discussions with persons involved at the site, I have developed a number of detailed questions related to the allegations and the findings which are given in Attachment I.

F.

The al1egatio~s which were concluded to be unsubstantiated are reviewed in Attachment II to this letter and the substantiated allegations are reviewed in Attachment III.

WRG:mb cc: s. H. Bush J. C. Ebersole H. Etherington M. S. P1esset File -

RC Very truly yours, W. R. Gall C-9

Attachment I Questions

1.

What degree of conservatism is used in design of the supports which depend on anchor bolts?

2.

\\{hat action will be taken to establish that the length of anchor bolts is adequate?

3.

What action will be taken to establish the integrity of concrete affected by arc, or flame cutting of rebar?

4.

What action is proposed to verify adequacy of cadweldingperformed in non-conformance with requirements?

5. What corrective action will be taken on welds performed by welders out-side of their qualified thickness range?

6. If the main steam pipe fails outside of the containment, between the penetration and the stop valve, will containment be breached?

7.

What action will be taken to correct misalignment at welded joint in main steam riser?

8.

In determining acceptable thinning of pipe walls during grinding, is 0.875 x t (t = nominal thickness) used as an acceptable thickness?

n n

9.

~~at defects could be incurred as a result of lack of QC in-process surveillance?

10. Will the welds in the reactor coolant loops be examined by the ultrasonic method during pre-service testing?

11.

What action will be taken to ascertain whether improperly stored weld rods were used in production and may have affected quality of welds -

e-specially welds in the reactor coolant system?

12. What action will be taken on the Southwest Fabricating Company's welds which were not examined by S&W during their audits.

C-10

Attachment II Unsubstantiated Allegations A total of 58 allegations \\.;rere made by the allegors A, B, and C including the addit::onal allegations.

Of this total, 45 were found not to be substan-tiated for various reasons.

In the follm.;ring list allegations are grouped according to the reasons for w'hich they ~.;rere not substantiated.

Reason for Not Substantiating

1.

Allegations that were not substantiated because the investiga-tors examined the affected part or the records and found them to be in conformance with the requirements.

A-3, 8, 9, 11, 12, 17,21, B-1, 12, 13, 14, 16,18,20,21,22,24,25,27,31, and No. of Allegations P-6' 21 In the report on Allegations A-8 and B-24, the inspector reported that quality control Nonconformance and Disposition reports showed that quality control had identified and documented instances of welders welding beyond limit qualifications.

The report does not indicate what disposition was made of these occurrenceS.

Inspector-identified Item 2b (Appendix 3) cites as an-infraction the welding of more than 30 welds in Class 1 piping in Unit 2 by welders who were not qualified for the thickness of the pipe which they were welding.

2. Allegations which are shown by quality control records to have been identified and corrected in accordance with procedures.

A-6, 7, 13, 15, 18, B-6, 26, P-2, 4, and 5 10

3. Allegations that were true but either were not related to quality or were in accordance with procedures and requirements.

A-14, 16, 19, B-4, 17, 30, and P-3 7

4. Allegations that were found to be a mistak~ on the part of the allegor.

A-20 1

5. Allegations which were not related to quality whether true or not.

B-9, C-l 2

6. Allegations that could not be verified by interviews with personnel, review of records, or other means and were concluded to be not substantiated.

A-2, 10, B-lO and 15 4

These allegations having to do with falsification of records are the type which would be difficult to verify or disprove.

The investi-gators' conclusions on these items were based on review of available records and intervie~vs with persons on the job and, though not considered substantiated, some of these violations could have occurred either without the knowledge of those interviewed or without being recorded in the documents.

C-11

(continuation of Attachment II)

Allegations A-2, A-10, and B-15,.Jere that a welding inspector and a QC inspector signed off papers without fully reviewing the work, that no one cares about quality or checks the work being done by welders, and QC inspectors had craftsmen perform fit-up inspections for them.

Allegation B-IO 1;vas that a particular field weld was performed by a different welder than the one whose number was recorded as having performed the weld. It is my opinion that the conclusions drawn by the inspector are correct, but I believe it is possible for violations of this type to occur in such a way that substantiation is almost impossible.

However, examination of the completed work by nondestructive methods can be performed to show that the work is satisfactory.

C-12

Attachment III SubstaQtiated Allegations Of the 13 substantiated allegations, four allegations (B-3, B-8, B-9, and B-1I) dealt with incorrect identification numbers on materials for Class 3 systems.

However, traceability was established through the heat numbers and the materials were found to be acceptable.

Allegation A-I, cutting of concrete reinforcement steel ("rebar"), was substantiated and it was established that rebar was cut during the drilling of about 25% of the anchor bolt holes for anchor bolts for support$.

The Licensee's response to this finding indicates that an engineering analysis

";vill be made to es*tablish the adequacy of the concrete.structures. It is my opinion that corrective action can be taken to assure the adequacy of these structures.

Allegations A-4 and B-2 dealt with the 32-inch main steam risers to the safety valve headers.

A serious problem with this incident is the verification that the joint~ in one case at least, had a mis-match of 5/16 inch as compared to the Code ma~imum of 3/32 inch.

Bending stresses in the pipe wall as a result of such misalignment would be increased by a factor of approximately three due to this effect.

Corrective action should be taken in regard to this misalignment.

Allegation A-5concerns lack of in-process surveillance of piping work.

The Licensee's response indicates that the procedures cplled for this sur-veillance and that it was carried out in part. It is likely that an increase in quality control personnel would be required if this is implemented as it is supposed to be, which would tend to substantiate allegations A-12 and B-3l that there are too few QC personnel.

Allegations B-7 and P-I concerning fake anchor bolts in pipe supports are related to the cutting of rebar.

The difficulty in installing anchor bolts without interfering with rebar apparently has caused Some people to subvert the requirements by faking the anchor bolt installation. It is my opinion that action should be taken to check the length of all the anchor bolts used for supports of this type to establish that the lengths are in accordance with the requirements or that. deviations are permissible as established by.engin-eering verification.

Allegation B-5 refers to unrecorded welds which were made in 2-inch pipe in a Class 3 system for which corrective action has not been reported. It is my opinion that if the procedures described in the response from the Licensee are followed, any additional unrecorded welds of this type will be discovered and they should be examineQ to establish thei~ acceptability.

Allegation B-23, holding over welding rod.

This problem seems to be very difficult to control but it is important that uncontrolled electrodes not be used in pipe welds.

The Licensee's response to this allegation does not indicate that they plan to take any corrective action on this item.

C-13

(continuation of Attachment III)

Allegation B-32, improper storage of stainless steel and carbon steel pipe and valves. It was evident during my visit to the site that many items are stored throughout the plant awaiting installation.

Although this is probably only temporary storage it appears that damage could occur and dirt could be accumulated in some of the valve operators and controls which could effect their performance.

I believe corrective action is required on this item.

C-14

Dr. David Okrent 2

January 3>> 1977

2.

Allegations Concerning Fake Anchor Bolts The interference between anchor bolts and rebar may be responsible for the faking of two anchor bolts which were reported in allegations B-7 and P-l. It is also possible that some anchor bolts were cut to avoid cutting rebar which would result in the length of bolts being shorter than specified, thus affecting the strength of the anchor. It is my under-standing that ultrasonic measurements will be made to detect those bolts which '!:vere shortened.

I recommend that an evaluation be made by the licensee to determine the adequacy of any bolts which are found to be short.

3.

Welders Performing Welds Outside the Range of Their Qualifications It was established by the inspector that 30 Class 1 type welds in Units 1 and 2 were performed by welders qualified for thinner sections.

I believe all of these welds are identified in QC records and that all of the welds have been examined by radiography and found to be acceptable.

The welders were qualified on thinner sections than those cited.

The acceptance of radiographed production welds as qualification welds may be a valid pro-cedure provided the initial weld performed outside the previously qualified thickness range meets the requirements of QW-301.4, QW-302.2, and QW-305.2 of Section IX of the ASME Code,is acceptable without weld repairs, and also provided the complete weld was performed by the same welder.

One instance was observed during a visit to North Anna site in 'tvhich a single welded joint in a primary coolant loop had the weld identification numbers of twelve different welders.

I believe this would not be a satisfactory way to qualify any of the welders.

4.

32-Inch Main Steam Riser in Safety Valve Station The circumferential joints performed in the modification to the 32-inch main steam riser have been evaluated and seem to be satisfactory except in the matter of mismatch.

Permissible mismatch allowed by the Code is 3/32 of an inch.

The inspector determined in at least one case a maximum of 5/16 inch mismatch.

This is a factor of 3 over that permitted by the Code and assuming that the stress in the longitudinal direction was judged satisfactory with the permissible misalignment, the affect of mismatch would increase that stress by a factor of 3 in the case of 5/16 inch mismatch.

This pipe is probably subject to extreme axial compressive loads when the safety valves operate.

I have not found evidence that a failure of this pipe could not cause a pipe whip in the main steam valve house which would react on the penetration of the containment wall sufficiently to breach the containment.

5.

Welding Electrodes There were two items concerning improper storage of 'velding electrodes and one concerning use of welding electrodes prior to receipt of material certifications for them.

It would be difficult if not impossible to determine C-15

Dr. David Okrent 3

January 3, 1977 whether welding electrodes which had been stored overnight or over a shift outside of the required drying ovens have been used in welds or in which welds they may have been used.

Furthermore, detection of the effect of excessive moisture or other contaminates principally hydrogen embrittle-ment, would be difficult to detect by means of radiography. It would be desirable to establish that all electrodes held over were being kept for personal use.

The use of welding materials prior to receipt of proper documentation requires verification after the weld material has been used and could result in a determination that incorrect materials were used.

This verification should be made in all cases 1;vhere this was done and corrective action taken 'tvhere necessary.

6.

Defective Shop Welds Two instances are reported in the inspector's findings - Items 2-C and 2-K in which noncorming shop welds in pipes performed by others were discovered by Stone & l-1ebster quality control.

Corrective action is not indicated.

Of particular interest is the disposition of those Southwest Fabricating and Welding Company's pipe welds which were not included in the 1.5% sample examined by Stone & Webster QC and the applications in which they were used.

Approximately half of the 1.5% sample were found to be nonconforming and presumablY were repaired.

Corre~tive action should be applied to all other welds represented by those samples to assure con-formance with quality requirements in the Class 2 system.

7.

Improper Identification of Haterials and Parts In the four reported in~idents of improper identification of materials, it was possible to establish acceptability for the materials affected.

Can it be established with a reasonable degree of confidence that these reported instances are the only ones in which materials were improperly identified, or that all materials installed are in compliance with require-ments?

8.

Conclusions A.

I agree in general with the Evaluation of Findings enclosed with the transmittal of the report of the investigation.

Corrective action to correct the deficiencies in the quality assurance program must be augmented by actions to verify quality of construction already completed Phases 2, 3 and 4 of the investigation~ I believe, were conceived for this purpose.

Phase 2 has been completed \\vith some deficiencies yet to be resolved.

Phases 3 and 4 should form a basis for establishing the integrity of the system.

An effort should be made to establish that welding electrodes which were improperly stored were not used in welding of safety related systems, or that if used the effects will not compromise safety.

C-1.6

Dr. David Okrent 4

January 3~ 1977 B.

It is my opinion that all of the identified "quality of workll non-conformities can be corrected by corrective action.

Some of the quality control non-conformities affecting tvork that is already com-pleted cannot be corrected now, but the quality of the work affected may be verified by preoperational testing, and examination, and if deficient it can be corrected.

c.

The "unresolved itemsfl listed in Part E of the report can also be resolvec by appropriate corrective actions.

D.

The licenseeis quality assurance program has not functioned in accordanc~

with established procedures and requirements in some cases.

This leads to concern about possible undetected non-conformances.

Phases 1 and 2 of the investigation constitute a thorough study of these possible deficiencies in the important safety related systems and it resulted in disclosure of some additional deficiencies which should be corrected.

E.

In my study of the report and my discussions with persons involved at the site, I have developed a number of detailed questions related to the allegations and the findings which are given in Attachment I.

F.

The allegations which were concluded to be unsubstantiated are reviewed in Attachment II to this letter and the substantiated allegaticns are reviewed in Attachment III.

WRG:mb cc: s. H *. Bush J. C. Ebersole H. Etherington H. S. Plesset File -

RC Very truly yours,

w. R. Gall C-17

Attachment I Questions

1.

Hhat degree of conservatism is used in design of the supports which depend on anchor bolts?

2.

i~at action will be taken to establish that the length of anchor bolts is adeq ua te?

3. What action will be taken to establish the integrity of concrete affected by arc, or flame cutting of rebar?

4.

What action is proposed to verify adequacy of cadwelding performed in non-conformance with requirements?

5. What corrective action will be taken on welds performed by welders out-side of their qualified thickness range?

6. If the main steam pipe fails outside of the containment, between the penetration and the stop valve, will containment be breached?

7.

What action will be taken to correct misalignment at welded joint in main steam riser?

8.

In determining acceptable thinning of pipe walls during grinding, is 0.875 x t (t

"" nominal thickness) used as an acceptable thickness?

n n

9.

~~at defects could be incurred as a result of lack of QC in-process surveillance?

10. Will the welds in the reactor coolant loops be examined by the ultrasonic method during pre-service testing?

11.

What action will be taken to ascertain whether improperly stored weld rods were used in production and may have affected quality of welds -

especially welds in the reactor coolant system?

12.

What action will be taken on the Southwest Fabricating Company's welds which were not examined by S&W during their audits.

C-18

Attachment II Unsubstantiated Allegations A total of 58 allegations were made hy the allegors A, B, and C including the additional allegations.

Of this total, 45,,,ere found not to be substan-tiated for various reasons.

In the follo~"ing list allegations are grouped according to the reasons for ~vhich they \\Vere not substantiated.

Reason for Not Substantiating

1.

Allegations that were not substantiated because the investiga-tors examined the affected part or the records and found them to be in conformance with the requirements.

A-3, 8, 9, 11, 12, 17, 21, B-1, 12, 13, 14, 16, 18, 20, 21, 22, 24~ 25, 27, 31, and No. of Allegations P-6' 21 In the report on Allegations A-8 and B-24, the inspector reported that quality control Nonconformance and Disposition reports showed that quality control had identified and documented instances of welders welding beyond limit qualifications.

The report does not indicate what disposition was made of these occurrences.

Inspector-identified Item 2b (Appendix 3) cites as an infraction the welding of more than 30 welds in Class 1 piping in Unit 2 by welders who were not qualified for the thickness of the pipe vlhich they were welding.

2. Allegations which are shown by quality control records to have been identified and corrected in accordance with procedures.

A-6, 7, 13, 15, 18, B-6, 26, P-2, 4, and 5 10

3. Allegations that were true bue either were not related to quality or were in accordance with procedures and requirements.

A-14, 16, 19, B-4, 17, 30, and P-3 7

4. Allegations that. were found to be a mistake on the part of the al1egor.

A-20 1

5. Allegations which were not related to quality whether true or noto B-9, C-l 2

6. Allegations that could not be verified by interviews with personnel, review of records, or other means and were concluded to be not substantiated.

A-2, 10, B-lO and 15 4

These allegations having to do with falsification of records are the type which would be difficult to verify or disprove.

The investi-gators' conclusions on these items were based on review of available records and intervie~"s with persons on the job and, though not considered substantiated, some of these violations could have occurred either without the knowledge of those interviewed or without*

being recorded in the documents.

C-19

(continuation of Attachment II)

Allegations A-2, A-10, and B-15 ~vere that a welding inspector and a QC inspector signed off papers without fully reviewing the work, that no one cares about quality or checks the work being done by welders, and QC inspectors had craftsmen perform fit-up inspections for them.

Allegation B-10 was that a particular field weld was performed by a different welder than the one whose number was recorded as having performed the weld. It is my opinion that the conclusions drawn by the inspector are correct, but I believe it is possible for violations of this type to occur in such a ~ay that substantiation is almost impossible.

However, examination of the completed work by nondestructive methods can be performed to show that the work is satisfactory.

C-20

Attachment III Substantiated Allegations Of the 13 substantiated allegations, four allegations (B-3, B-8, B-9, and B-ll) dealt with incorrect identification numbers on materials for Class 3 systems.

However, traceability was established through the heat numbers and the materials were found to be acceptable.

Allegation A-I, cutting of concrete reinforcement steel ("rebartl ), was substantiated and it was established that rebar was cut during the drilling of about 25% of the anchor bolt holes for anchor bolts for supports. The Licensee's response to this f~nding indicates that an engineering analysis will be made to establish the adequacy of the concrete structures. It is my opinion that corrective action can be taken to assure the adequacy of these structures.

Allegations A-4 and B-2 dealt with the 32-inch main steam risers to the safety valve headers.

A serious problem with this incident is the verification that the joint, in one case at least, had a mis-match of 5/16 inch as compared to the Code maximum of 3/32 inch.

B~nding stresses in the pipe wa+l as a result of such misalignment would be increased by a factor of approximately three due to this effect.

Corrective action shou~d be taken in regard to this misalignment.

Allegation A-5*concerns lacf of in-process surveillance of piping work.

The Licensee's response indica~es that the procedures c~lled for this sur-veillance and that it was carifed out in part. It is likely that an increase in quality control personnel would be required if this is implemented as it is supposed to be, which would tend to substantiate allegations A-12 and B-31 that there are too few QC personnel.

Allegations B-7 and P-I concerning fake anchor bolts in pipe supports are related to the cutting of rebar.

The difficulty in installing anchor bolts without interfering with rebar apparently has caused some people to subvert the requirements by faking the anchor bolt installation. It is my opinion that action should be taken to check the length 0*£ all the anchor bolts used for supports of this type to establish that the lengths are in accordance with the r-equirements or that deviations are permissible as established by engin-eering verification.

Allegation B-5 refers to unrecorded welds which were made in 2-inch pipe in a Class 3 system for which corrective action has not been reported. It is my opillion that if the procedures described in the response from the Licensee are followed, any additional unrecorded welds of this type will be discovered and they should be examine4 to establish thei~ acceptability.

Allegation B-23, holding over welding rod.

This problem seems to be very difficult to control but it is important that uncontrolled electrodes not be used in pipe welds.

The Licensee's response to this allegation does not indicate that they plan to take any corrective action on this item.

C-21

(continuation of Attachment III)

Allegation B-32, improper storage of stainless steel and carbon steel pipe and valves *. It was evident during my visit to the site that many items are stored throughout the plant awaiting installation.

Although this is probably only temporary storage it appears that damage could occur and d~rt could be accumulated in some of the valve operators and controls which could effect their performance.

I believe corrective action is required on this item.

C-22

APPENDIX 0 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS-GENERIC MATTERS The Advisory Committee on Reactor Safeguards (Committee) periodically issues a report listing various generic items applicable to large light-water reactors.

These are items which we and the Committee, while finding present plant designs acceptable, believe have the potential of adding to the overall safety margin of nuclear power plants, and as such should be considered for application to the extent reasonable and practicable as solutions are found, recognizing that such solutions may occur after completion of the plant.

This is 'consistent with our continuing efforts toward reducing still further the already small risk to the public health and safety from nuclear power plants.

The most recent such report concerning these generic items was issued to the Commission February 24, 1977 in a letter from Committee Chairman M. Bender to Commission Chairman M. Rowden.

However, with respect to the North Anna Power Station, Units 1 and 2, the Committee in their letter of October 26, 1976 references the report that was issued on April 16, 1976 to Chairman Rowden, NRC from Chairman D. Moller, ACRS.

All of the discussions below relate to the Committee1s report of April 16, 1976.

The status of staff efforts leading to resolution of all these generic matters is contained in our Status Report on Generic Items periodically transmitted to the Committee.

The latest such Status Report is contained in a letter from B.

Rusche to Committee Chairman M. Bender dated January 31, 1977.

The Committee, in its letter dated October 26, 1976 (attached as Appendix B), on North Anna Power Station Units 1 and 2, identified which of these generic items it deems applicable to the North Anna Power Station Units 1 and 2.

For many of the items so identified, we have provided in the Safety Evaluation Report and supplements specific discussions particularizing for the North Anna facility the generic status given in the January 31, 1977 Status Report.

These items are listed below with the appropriate section numbers of the Safety Evaluation Report and supplements where such discussions are to be found.

The numbering corresponds to that in the April 16, 1976 report of the Committee.

For those items applicable to the North Anna Power Station Units 1 and 2 which have not yet progressed to where specific action can be initiated relevant to individual plants, our Status Report on Generic Items referred to above provides the appropriate information.

0-1

Group II - Resolution Pending (1) Turbine Missiles - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977.

However, as stated in Section 10.7 of Supplement No.2 to the Safety Evaluation Report we required increased maintenance and testing procedures which in our view reduce the likelihood of major contributors to turbine failures (Section 10.7 of Supplement No.2 to the Safety Evaluation Report).

(2)

Effective Operation of Containment Sprays in a LOCA - This item is resolved for the North Anna Power Station, Units 1 and 2 by use of sodium hydroxide additive to sprays (Section 6.2.3 of the Safety Evaluation Report).

(3) Possible Failure of Pressure Vessel Post - LOCA by Thermal Shock - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977.

(4) Instruments to Detect Fuel Failures - This item is resolved for the North Anna Power Station, Units 1 and 2 by the installation of a gross failed fuel monitor (Section 9.3.4 of the Safety Evaluation Report).

(5) Monitoring for Excessive Vibration of Loose Parts Inside the Pressure Vessel - This item is resolved for the North Anna Power Station, Units and 2 by the installation of a loose parts monitor (Section 5.5 of the Safety Evaluation Report).

(6)

Common Mode Failures - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977, and as indicated in Section 7.2.4 of the Safety Evaluation Report.

(7) Behavior of Reactor Fuel Under Abnormal Conditions - This item is under generic review as' indicated in our status report to ACRS dated January 31, 1977.

(9) The Advisibility of Seismic Scram - A seismic scram is not proposed for the North Anna Power Station, Units 1 and 2 and we will not require such a scram (see letter, dated May 19, 1977, from E. Case, Acting Director, Office of Nuclear Reactor Regulation, to Committee Chairman Bender; subject, "The Advisibi1ity of a Seismic Scram.").

(11) Instrumentation to Follow the Course of an Accident - This item is resolved for the North Anna Power Station Units 1 and 2 by compliance with our requirements (Section 7.5 of the Safety Evaluation Report).

0-2

Group II A - Resolution Pending - Items Since December 18, 1972 (1) Pressure in Containment Following a LOCA - This item is resolved for the North Anna Power Station, Units 1 and 2 by our confirmatory analysis which indicate reasonable agreement with the peak containment pressure calculated for the LOCA by the applicant (Section 6.2.1 of the Safety Evaluation Report).

(4) Rupture of High Pressure Lines Outside Containment - This item is resolved for the North Anna Power Station, Units 1 and 2 by compliance with the criteria specified in the letter from A. Giambusso (NRC) to the applicant dated December 12, 1972, "General Information Required for Consideration of the Effects of a Piping System Break Outside Containment" (Section 3.6.2 of the Safety Evaluation Report).

(5)

PWR Pump Overspeed During a LOCA - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977, and as indicated in Section 5.4.1 of the Safety Evaluation Report.

(6) Isolation of Low Pressure from High Pressure System - This item is resolved for the North Anna Power Station, Units 1 and 2 by compliance with our requirements (Section 7.6 of the Safety Evaluation Report).

(7) Steam Generator Tube Failures - This item is resolved for the North Anna Power Station by measures taken by the applicant to maintain secondary water chemistry within specified limits and imposition of operating limita-tions on primary-to-secondary leakage.

The applicant also has provisions to detect tube degradation should it occur (Section 5.2.7 of the Safety Evaluation Report).

(8)

ACRS/NRC Periodic lO-Year Review of all Power Reactors - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977.

GROUP II B - Resolution Pending - Items Added Since February 13, 1974 (2) Qualification of New Fuel Geometries ~ This item is not totally resolved for North Anna Power Station Units 1 and 2.

While our evaluation of the Westinghouse 17 x 17 fuel verification tests is complete, our evaluation of the fuel surveillance program has not been completed (Section 18.2.1 of this supplement).

D-3

GROUP II C - Resolution Pending - Items Added Since March 12, 1975 (1)

Locking Out of ECCS Power - Operated Valves - This item is reso1v~d for North Anna Power Station, Units 1 and 2 by the technical specifications which require lockout of power to appropriate valves (Section 6.3.2, 6.3.3 and 7.33 of the Safety Evaluation Repo.rt and Section 6.3.3 of Supplement No. 1 to the Safety Evaluation Report).

(2) Fire. Protection - The applicant has submitted a fire hazards analysis and a reevaluation of the fire protection program in accordance with Appendix A to the Auxiliary and Power Conversions Systems Branch Technical Position.

A site visit has been scheduled for the week of August 1, 1977 to review this matter.

We will assure that most, if not all improvements regarding fire protection will be implemented prior to start of operation on the second fuel cycle (Section 18.2.8 of this supplement).

(3) Design Features to Control Sabotage - This item is'under generic review ~s indicated in our status report to the ACRS dated January 31, 1977.

The security plan for the North Anna Power Station, Units 1 and 2 was orgina11y accepted on the basis of conformance with the Commission's regula-tions including 10 CFR 73.40 and Regulatory Guide 1.17, "Protection of Nuclear Power Plants Against Industrial Sabotage."

On February 29, 1977, the Commission published new requirements for the phys,ical protection of nuclear power plants against acts of sabotage (10 CFR 73.55).

The applicant has submitted a revised plan in accordance with the new regulations.

We are presently conducting a review of this plan and a site visit was held during the week of July 25, 1977, to discuss these matters with the applicant.

We intend to require the applicant to comply with these new requirements.

(4) Decontamination and Decommissioning of Reactors - This item is under generic review as indicated in our status report to ACRS dated January 31, 1977.

(5) Vessel Support Structure - This item is resolved for the North Anna Power Station Units 1 and 2, on the basis of our evaluation of the reactor pressure vessel support system (Section 3.9.4 and 4.2.4 of the Safety Evaluation Report and Section~ 3.9.4 and 4.2.4 of this supplement).

(6) Water Hammer - This item is under generic review as indicated in our status report to the ACRS dated January 31, 1977, and as indicated in Section 10.3 of the Safety Evaluation Report.

(7) Maintenance and Inspection of Plants - This item is under generic review as indicated in our status report to the ACRS dated January 31, 1977.

0-4

UNHED STATES NUCLEAR REGULATORY COMMiSSiON WASI1INGTON, D. C.

20555 OFFICiAl.. BUSINESS PENALTV FOR PRIVATE USE, noo POSTAGIE AND FEES PAiD UNI1EO ~TATES NUCLEAR REGULATORV LOMM:.SION