ML090720819
| ML090720819 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/30/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| FOIA-2024-000060 NUREG-0053, Suppl. 3 | |
| Download: ML090720819 (40) | |
Text
SUPPLEMENT NO.3 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF VIRGI~IA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION-UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339 NUREG-0053, SUPP. 3 SEPTEMBER 15,1976
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TABLE OF CONTENTS PAGE
- 1. 0 INTRODUCTION AND GENERAL DISCUSSION..........................................
1-1 1.1 Introduction............................................................
1-1 2.0 SITE CHARACTERISTICS.........................................................
2-1 2.4 Hydrologic Engineering..................................................
2-1 2.4.3 Low Water Considerations.........................................
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 3.8 Design of Seismic Category I Structures.................................
3-1 3.8.2 Other Seismic Category I Structures..............................
3-1 3.10 Seismic and Environmental Qualification of Seismic Category I Instrumentation and Electrical Equipment..........
3-1 4.0 REACTOR......................................................................
4-1 4.4 Thermal and Hydraulic Design............................................
4-1 5.0 REACTOR COOLANT SySTEM.......................................................
5-1 5.4 Component and Subsystem Design........................*.................
5-1 5.4.2 Steam Generator and Reactor Coolant Pump Supports................
5-1 6.0 ENGINEERED SAFETY FEATURES...................................................
6-1 6.2 Containment Systems.....................................................
6-1 6.2.1 Containment Functional Design....................................
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TABLE OF CONTENTS (Continued)
PAGE 6.3 Emergency Core Cooling System...........................................
6-2 6.3.3 Performance Evaluation...........................................
6-2 10.0 STEAM AND POWER CONVERSION SYSTEM............................................
10-1 10.2 Main Steam Supply System................................................
10-1 20.0 FINANCIAL QUALIFICATIONS............................ *........................
20-1 20.1 Introduction............................................................
20-1 20.2 Estimated Operating and Shutdown Costs..................................
20-1 20.3 Source of Funds.........................................................
20-2 20.4 Conclusion..............................................................
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22.0 CONCLUSION
S..................................................................
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APPENDICES PAGE APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW..................
A-1 APPENDIX B ERRATA TO SUPPLEMENT NO. 2 TO THE SAFETY EVALUATION REPORT FOR THE NORTH ANNA POWER STATION, UNITS 1 AND 2....................
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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 Vi rgi ni a El ectri c and Povler Company (appl i cant). Supplement No. 1 to the Safety Evaluation Report was issued on June 30, 1976 and Supplement No.2 was issued on August 2, 1976.
Supplements No.1 and No.2 to the Safety Evaluation Report document-ed the resolution of several outstanding items, and summarized the status of the remaining outstanding issues.
The purpose of this supplement is to update our Safety Evaluation Report (and Supple-ments No.1 and No.2) by providing (1) our evaluation of additional information submitted by the applicant since the issuance of Supplement No.2 of the Safety Evaluation Report, (2) information regarding the current status of matters that were still under review and (3) additional information for those sections of the Safety Evaluation Report where further discussion or changes are in order.
Each section of this supplement is numbered the same as the section of the Safety Evaluation Report, and is supplementary to and not in lieu of the discussion in the Safety Evaluation Report, except where specifically so noted.
Appendix A is a con-tinuation of the chronology of our principal actions related to the processing of the application, and Appendix B is a listing of errata to Supplement No.2 to the Safety Evaluation Report.
A summary of the remaining outstanding issues is presented in Section 22.0 of this supplement.
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2.0 SITE CHARACTERISTICS 2.4 Hydrologic Engineering 2.4.3 Low Water Considerations In the Safety Evaluation Report we stated that we would verify the acceptability of the reservoir for two-unit operation after we have evaluated the results of the operational program for Unit 1.
We also stated that in the event the results of this program do not verify the acceptability of the reservoir for two-unit operation, we would require that the design of the reservoir be modified so that a safe shutdown capability for both units can be assured prior to permitting two-unit operation.
Prior to a decision concerning the issuance of an operating license for Unit 2, we will review the results of the operational program and will take appropriate action at that time to assure the adequacy of the reservoir for two-unit operation.
2.6 Foundation Engineering 2.6.2 Evaluation of Foundation Engineering In the Safety Evaluation Report we stated that all reported settlements and dif-ferential settlements of Category I structures were reasonable and within allowable limits except for a few points along the crane rail at the northwest corner of the service building.
We also stated that we would review this matter and report the results of our review in a subsequent supplement to the Safety Evaluation Report.
We have reviewed the information provided by the applicant related to this matter and have also visited the North Anna facility to inspect and assess the structural damage, if any, caused by this settlement.
On the basis of our review of the applicable information and our site visit, we conclude that no safety related structure and equipment will be adversely affected due to this settlement and that no corrective action is necessary. Therefore, we consider this matter resolved.
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3.0 DESIGN CRITERIA-STRUCTURES, SYSTEMS, AND COMPONENTS 3.8 Design of Seismic Category I Structures 3.8.2 Other Seismic Category I Structures The applicant has recently advised us by letter, dated August 10, 1976 that a review of the structural design computations for the spent fuel pool give a preliminary indication that the completed spent fuel pool may not satisfy the requirements stated in Section 3.8.1.4 of the Final Safety Analysis Report relative to the allowable stresses under static plus thermal loadings, and static plus thermal plus seismic
. loadings.
The applicant has advised us that he is presently performing an extensive evaluation to fully define the state of stress in the entire spent fuel pool under its design loadings and that he will submit a final report on this matter on or before October 15, 1976. After we receive the report we will evaluate it to deter-mine if our conclusions regarding the adequacy of the spent fuel pool are still valid and report our findings in a supplement to the Safety Evaluation Report.
3.10 Seismic and Environmental Qualification of Seismic Category I Instrumentation and Electrical Equipment
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Our analysis of the containment peak temperature for a postulated main steam line break accident (see Section 6.2.1 of this supplement) has predicted a peak tempera-ture.greater than the temperatures for which the instrumentation and equipment that are required to operate foJlowing a postulated main steam line break were qualified.
Therefore, we have required the applicant to provide a qualification program that would assure that the required equipment will withstand this temperature environment.
The applicant has indicated in Amendment 56 to the Final Safety Analysis Report and, reiterated in a recent meeting that he is presently working with vendors of the instrumentation involved to institute necessary measures to adequately qualify the involved equipment.
When the applicant submits the appropriate test results and/or design modifications, we will report the results of our review regarding this matter in a supplement to the Safety Evaluation Report.
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4.0 REACTOR 4.4 Thermal and Hydraulic Design In Section 4.4 of the Safety Evaluation Report, we concluded that procedures for including rod bow in the departure from nucleate boiling calculations are acceptable, with the provision that the magnitude of the rod bowing be based on ongoing surveil-lance programs.
At a meeting held on August 9, 1976, Westinghouse Electric Corporation informed us that the penalty on departure from nucleate boiling when considering bowed fuel rods was underestimated under certain conditions.
Previous tests using only heated rods showed a penalty of eight percent in departure from nucleate boiling heat flux for rods bowed to contact. For this condition this value of the contact penalty could be predicted by an analytical model.
However, additional tests have been performed with a rod bowed to contact with an unheated rod (thimble tube) in the test array. These tests show a significant increase in the departure from nucleate boiling penalty over that predicted by either the previous tests or the available analytical models.
Therefore, for operating Westinghouse designed reactors, we will impose an interim rod bow departure from nucleate boiling penalty, to be used until the most recent data are submitted and reviewed by us.
This penalty is derived by extrapolating available test data and provides a conservative safety margin for these reactors which we consider acceptable. A similar penalty will be incorporated into the tech-nical specification limits for North Anna 1 and 2.
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5.4 5.4.2 5.0 REACTOR COOLANT SYSTEM Component and Subsystem Design Steam Generator and Reactor Coolant Pump Supports We stated in Section 5.4.2 of the Safety Evaluation Report, that on March 12, 1976, we were notified by the Sun Shipbuilding and Dry Dock Company that documentation existed indicating that activities ~ssociated with the construction, and subsequent repair of the steam generator and reactor coolant pump supports for North Anna Units 1 and 2 may have compromised the design safety of these supports.
We also stated that we would address this matter further in a subsequent report.
Each of the three primary coolant loops is comprised of a steam generator, a reactor coolant pump and piping connecting these components to the reactor vessel.
The steam generators and reactor coolant pumps are supported on frame-type welded structures.
I Hydraulic snubbers are provided. as lateral support. to permit slow horizontal move-
! ment of the components due to thermal expansion of the system while providing resis-I tance against the dynamic loadings that arE calculated to result from the postulated 11 oss-of-cool ant acci dent and safe shutdO\\~n earthquake.
The applicant has performed a system dynamic analysis to evaluate the design adequacy of the support structures in order to assure that the design criteria have been satisfied and that structural integrity is maintained under all design loading conditions.
For purposes of design. two extremely low probability events, the postulated loss-of-coolant accident and design basis earthquake were assumed to occur concurrently.
The supports were then required to carry the absolute sum of loads generated by the concurrent loss-of-coolant accident and safe shutdown earthquake in combination with normal steady state loads.
This combination of loads resulted in stresses that were within the conservative design stress limit of ninety percent of the material yield strength.
I The analyses cited above demonstrate that the basic structural design of the supports l -i s acceptab 1 e.
Typi ca 1 of a 11 such ana lyses, the mai ntenance of reasonable ductile I behavior of the steel members is prerequisite to the conclusions drawn.
Maintenance I of ductile behavior is assured in any material by a combination of basic metallurgi-
\\ cal conditions and operation at temperatures sufficiently high to result in ductile i \\rather than brittle material behavior.
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In those instances where a structure is not redundant (i.e., where failure of a single member or joint, such as the reactor pressure vessel, lead to unacceptable consequences) extraordinary measures, usually in terms of metallurgical control and operating temperature control, are called for to avoid any approach to brittle behavior.
In those instances where a structure is highly redundant (i.e.* failure of a number of members does not lead to unacceptable consequences) historical and suc-cessful engineering practice has often employed materials in the transition region somewhat below full ductile behavior of the material.
As a result of the various circumstances related below, we have focused particular attention on the review of brittle versus ductile behavior in the steam generator and reactor coolant pump supports at the North Anna facility.
The steam generator lower support structures are fabricated mainly from American Society for Testing and Materials (ASTM) A-36 material although some ASTM A-572 materi-al is used in the upper and middle level horizontal and vertical structural members.
The reactor coolant pump supports are fabricated from ASTM A-36 material with the exception of two structural members in the Unit 2, cubicle C support, where ASTM A-242, Type 1 material was substituted.
Notch-ductility requirements were not speci-fied in the design of the support structures.
The minimum service temperature was established at 80 degrees Fahrenheit.
Extensive amounts of weld cracking were observed in these structures after fabrica-tion and shipment to the plant site.
On October 5, 1973 Virginia Electric and Power Company informed us that all the welds on the steam generator and reactor coolant pump support structures were to be replaced.
The Unit 1 steam generator support structures were repaired in place at the North Anna facility site while the reactor coolant pump supports were shop-repaired.
Most of the repairs on Unit 2 supports were performed at the Surry facility site.
The repair program, including the quality assurance programs of the organizations involved were reviewed by the Commissions' Office of Inspection and Enforcement.
Magnetic particle, visual and ultrasonic inspection techniques were used during the weld repair.
Defects found as a result of these examinations were excavated and repaired.
The Unit 1 reactor coolant pump support structures and all the Unit 2 support structures were post-weld heat treated following the weld replacement while the Unit 1 steam generator support structures were peened to assist in relieving the weld stresses. At this time, the Commissions' Office of Inspection and Enforcement is satisfied that the repair of the Unit 1 steam generator and reactor coolant pump supports have been completed in accordance with approved procedures.
The Unit 2 steam generator and reactor coolant pump repairs are also complete.
However, our evaluation of the repair and the nondestructive examination report for Unit 2 is not complete.
If our evaluation of this information concludes that additional action by the applicant is necessary, we will require that such action be taken prior to the issuance of an operating license for Unit 2.
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Sun Shipbuilding and Dry Dock Company, the fabricator of the support structures for the steam generators and reactor coolant pumps, has made the following allegations concerning the safety of these support structures:
(1)
North Anna steam generator and pump supports are not adequate by reason of materials and design.
(2)
Redundancy has not been established.
(3)
The cracking problems are inherent to the design and steel selection.
(4)
North Anna support structures must have the same level of protection against brittle fracture as the reactor pressure vessel.
(5)
A detailed finite element analysis should be undertaken.
(6)
The possibility exists of a brittle fracture as a result of defects in the highly restrained welds in a structure built from steels with no fracture toughness requirements.
We have met with the applicant and with representatives of the Sun Shipbuilding and Dry Dock Company to discuss these allegations and to examine them in light of the various actions and analyses.
Our evaluation of the steam generator and reactor coolant pump supports has considered this information and is discussed below.
Since the time of the original support structure fabrication and during the course of the repair a significant amount of notch ductility testing has been performed on the support structure materials to determine whether they are susceptible to brittle fracture.
Nil ductility transition temperature as determined by drop weight tests and Charpy V impact test data have been obtained on approximately 50 percent of the total number of heats of ASTM A-36 material used in the Unit 1 and 2 steam generators and pump supports. This data shows the ASTM A-36 material to have adequate toughness at the minimum service temperature (80 degrees Fahrenheit) for these structures.
The maxi-mum nil ductility transition temperature measured was 40 degrees Fahrenheit and Charpy V impact tests on all heats of the ASTM A-36 material tested exhibited greater than 25 mils lateral expansion thus satisfying the fracture toughness criteria cur-rently established by Section III (NF) of the American Society of Mechanical Engineers 5-3
Boiler and Pressure Vessel Code.lI The remaining heats of ASTM A-36 material (for which toughness data is not available) were produced by the same steel mi 11 at ap-proximately the same time as the tested material, have similar chemical compositions and can be expected to have similar toughness.
Nil ductility transition temperature and notch toughness data were obtained also on approximately 50 percent of the heats of ASTM A-572 material which is used in the steam generator supports.
The transition temperature for this material was deter-mined to be 100 degrees Fahrenheit maximum, which is significantly higher than the transition temperature for the ASTM A-36 material, suggesting that brittle fracture could initiate in the ASTM A-572 material at the minimum service temperature (80 degrees Fahrenheit) and expected service loadings, provided that a sufficiently large defect was present at the initiation site.
Because of the potential for brittle fracture at the originally established minimum service temperature (80 degrees Fahrenheit) the applicant has elected (see Amendment 54 to the Final Safety Analysis Report) to increase the minimum service temperature of the steam generator supports by selectively removing some pipe insulation and installing an insulation blanket around the support structures. This arrangement assures that all of the ASTM A-572 structural members will be at a minimum service temperature of at least 180 degrees Fahrenheit while the facility is operating.
Heat balance studies submitted by the applicant show that the nominal operating temperatures of the ASTM A-572 material will be considerably higher than the estab-lished minimum and will range from 210 to 300 degrees Fahrenheit.
Based on our review of these analyses we believe this is a conservative estimate.
The applicant has also committed to install a temperature monitoring system which will display the steam generator lower support temperatures in the reactor control room.
The bottom members of the steam generator lower supports will not be insulated and are expected to operate at a temperature of approximately 90 to 125 degrees Fahrenheit.
The bottom members are all fabricated of ASTM A-36 material which has been shown to possess adequate toughness at these temperatures.
lISubsection NF of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code contains design, material, fabrication, inspection and examination standards for current construction.
This subsection was published in 1974, after the design, fabrication and initiation of repairs of the North Anna facility component supports.
Although subsection NF requirements are satisfied for certain material and design considerations, this should not be interpreted as implying overa 11 conformance to NF ru 1 es since it was not used as the so le bas i s for acceptance in this safety evaluation.
Specific exceptions to the NF rules are in the areas of postweld heat treatment and nondestructive examination criteria where alternate methods were employed in construction of the North Anna facility supports.
Overall acceptability of these structures is based on consideration of the measured material properties, calculated stress levels, evaluation of the specified fabrication and examination requirements and verification that these requirements have been satisfied.
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Increasing the minimum service temperature of the steam generator structures reduces the brittle fracture probability of the ASTM A-572 (and ASTM A-36) material to an extremely low lev~l. It is also noted that at the projected operating temperatures the current American Society of Mechanical Engineers Boiler and Pressure Vessel Code-Subsection NF fracture toughness requirements are satisfied for the ASTM A-572 materia"l.
Notch ductility tests on the ASTM A-572 material also showed that at least two heats of the material possessed relatively low energy absorption characteristics.
In-creasing the temperature of the supports will increase the material energy absorption capability. The longitudinal Charpy V notch energy absorption values for all material tested will exceed 25 foot-pounds at the projected service temperature.
Structural material of this type can be subject to lamellar tearing and evidence of this was observed in samples removed from the supports.
It is likely that such imp"erfections currently exist in the structures. Very limited testing of small (O.505-inch diameter) welded, short transverse specimens from support material vias performed by Sun 5hi pbuil di n9 and Dry Dock Company "and appeared to di sc lose low elongations for the ASTM A-572 material.
For several reasons failure is unlikely due to the reported imperfections or elongation values.
The short transverse yield strength of the material is greater than 60 thousand pounds per square inch (i.e., a factor of five greater than the maximum reported stress of 12 thousand pounds per square inch).
Moreover much significance can not be attached to short transverse elongation values for large structural shapes which have been determined using small cross section (0.2 square inches) specimens since the effect of relatively small lamellar inclusions can be grossly exaggerated in comparison with actual load bearing areas encountered in the fabricated structures. Additionally, some of the replaced joints were first buttered with weld filler to help prevent the occurrence of lamellar tearing immediately beneath the weld joint, This technique has been used successfully in similar applications.
The reactor cool ant pump support structures are fabri cated from ASnl A-36 materi a 1 with the exception of the support in cubicle C of Unit 2 in which ASTM A-242 Type 1 material was substituted for two structural members.
As discussed previously, the ASrM A-36 material is considered to possess adequate toughness for the intended service. Mechanical test data on the ASTM A-242 material was submitted by the ap-plicant on August 26, 1976.
Based on our review of this data in conjunction with the service temperature of these structures we consider this substitution to be acceptable.
There is currently no requirement for the inservice inspection of external supports for reactor coolant system vessels if the supports do not have a welded attachment to the reactor coolant system pressure boundary.
However, to further assure the integrity of the lQwer steam generator supports we will require that all accessible main member to main member welds joining the ASTM A-572 material be visually examined during each inservice inspection interval as defined in Section,XI of the American Society of 5-5
Mechanical Engineers Boiler and Pressure Vessel Code (no less than one third of the above welds in all supp6rts in all loops shall be examined during each 40 month period). In addition we will also require that these welds be visually examined following initial heat up and cool down during the preservice testing program.
To further study the question of redundancy and alternatives to elevated operating temperatures, we required the applicant to perform additional evaluations, supple-mental to the original analyses, to further demonstrate the redundant characteristics and the structural integrity of the support structures.
The additional evaluations follow:
(1)
One full interior corner of the frame structure (seven beam member) was eliminated from the applicant's mathematical model (simulating complete failure of these members) to evaluate load redistribution under simultaneous application of loads resulting from the design basis earthquake, loss-of-coolant accident and dead-weight on the remaining frame structure.
The resulting stresses in all remain-ing frame members of the steam generator and reactor coolant pump support struc-tures did not exceed the ninety percent of material yield strength design stress criteria, demonstrating satisfactory load redistribution.
(2)
The applicant performed elastic analysis of the support structure eliminating all ASTM-A-572 beam members in the mathematical model to demonstrate redundant characteristics of the support structures under deadweight loading.
It was determined that resulting stresses in the remaining support members were within the ninety percent of material yield strength design stress criteria, again demonstrating satisfactory load redistribution.
(3)
An analysis was performed in which the largest ASTM A-572 beam member was analytically removed and the simultaneous application of loads resulting from deadweight, the design basis earthquake and loss of coolant accident were applied.
It was determined that the remaining support members experienced redistributed stresses within the ninety percent of material yield strength design stress criteria.
(4)
The applicant has evaluated the alternative of replacing the ASTM A-572 material in the structure.
In the course of this evaluation the applicant has shown that available clearances are very small, that it would be extremely difficult to temporarily support the massive components and that the prospects for markedly improving the metallurgical properties of the frame structure and the condition of the welded joints are not good.
As a result of these findings the applicant has concluded that such replacement has the potential for deleteriously affec-ting the remaining structure and components of the primary coolant loop, thus having an adverse overall effect while offering no substantive increase in safety.
Based on our review considering the elevated operating temperatures to be employed, we concur in this conclusion reached by the applicant.
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Based on these analyses (items 1, 2 and 3 above) we have concluded that, under the most conservative combination of postulated loadings, both structural integrity and member redundancy of the steam generator and reactor coolant pump support structures have been demonstrated.
This conclusion is based on analyses that assume some fail-ures have occurred and demonstrate the structures are adequate with such failures present. This conclusion is valid irrespective of whether the structures are at elevated temperatures or at ambient temperatures.
To significantly decrease the probability of brittle fracture the applicant will maintain the support structure service temperature above a specified minimum value.
Since the original analyses did not include the thermal effects caused by such oper-ation, the applicant has performed an evaluation which adequately accounts for the additional loads which result from operation at elevated temperature.
The applicant selected the current criteria governing allowable stresses under such loadings as provided for in Subsection NF of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
We find this acceptable.
The effect of elevated temperature on yield strength and on the design criteria for allowable stress based on yield strength has been evaluated and was found to be acceptable.
The frame stress analysis which was performed by the applicant adequately defines member loads for purposes of design evaluation and we have therefore concluded that a detailed finite element analysis for this redundant frame-type, support structure would not add significantly to the information required to assess the adequacy of these structures for service.
no prevent lock-up of the hydraul i c snubbers install ed in the North Anna fad 1 ity I ; support structures under normal operation, features such as by-pass flow paths and separate pressure relief devices are employed.
These features and the successful operating history of hydraulic snubbers of the large size,employed at North Anna 1 facility provide confidence that lock-up is extremely unlikely. Should such an unlikely event occur it is equally unlikely that brittle fracture would occur at the operating temperatures involved.
However, it has been demonstrated that the failure of the largest beam member at the attachment point of the snubbers will not result in loss of support function.
Our evaluation of the reactor coolant pump and steam generator support structures has lead to the conclusion that:
(1) the acc~pted design stress criteria of ninety percent of the material yield strength under the conservative simultaneous design loadings of deadweight, safe shutdown earthquake and loss-of-ooolant accident has been achieved, (if these structures had been designed using the present Subsection NF design criteria which are acceptable for new construction, further design margin would be available since the allowable criteria in Subsection NF permits stresses up to one hundred and twenty percent of the material yield strength) thus effectively establishing the structural integrity of the supports under the postulated conserva-tive loading combinations; (2) acceptable redundancy of these structures has been 5-7
established including demonstration that the complete removal of the largest ASTM A-572 beam member under postulated accident loadings will not affect the frame struc-tural integrity, and that additional beam members may be removed, again without effect on structural integrity or the safety of the plant, under normal operating condition loadings and temperatures.
Both conclusions were reached without consider-ation of the effects of increasing the specified minimum service temperature. There-fore, in the very unlikely event that brittle fracture should occur in members with marginal fracture toughness properties, structural failure significantly affecting the safety of the plant will not occur; (3) provision of thermal insulation and a temperature monitoring and display system to maintain a specified minimum service temperature significantly reduces the probability for the occurrence of brittle fracture and therefore provides additional margin over that already available as discussed in (1) and (2).
[The notch ductility properties which have been determined by additional material testing (not required in the original construction) in all cases met current minimum values established in Subsection NF of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for those heats of material which were tested.]
On the basis of our review we have determined that the overall design and material of these structures is acceptable and failure is highly unlikely due to the design conservatisms provided by the redundancy of beam members; the conservative designa-tion of loadings and the method for load combinations; conservative levels of allow-able stress and the determination that computed stresses are within these levels; the inservice examination program for the structures; the adequate program used to effect weld repairs of the structures; the additional testing program to determine material properties; and the provisions for service operation at a specified minimum temper-ature. Therefore, we conclude that the steam generator and reactor coolant pump supports are acceptable for service at the North Anna Power Station, Units 1 and 2.
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6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design With respect to the containment pressure and temperature response for a spectrum of postulated main steam line break accidents, we stated in the Safety Evaluation Report that we were reviewing the applicant's model for calculating the mass and energy release rates and \\~ere performing an independent calculation of the maximum containment pressure and temperature using the CONTEMPT-24 computer code.
We also stated that we would report the results of our review in a subsequent report.
Following a postulated main steam line break inside the containment, steam will initially be dischal~ged from all three steam generators.
Flow from the steam generators in the unbroken loops will be terminated by closure of the isolation valves following the main steam line isolation signal.
Flow from the steam generator in the broken loop will continue until all the fluid is discharged.
Flashing liquid, as well as heat flow from the primary system, vlill cause the steam generator fluid level to rise following a rupture of the steam line.
If steam is formed within the secondary fluid faster than the steam removal rate, the two-phase level will rise within the steam generators, and flow through the broken steam pipe into the containment.
The amount of entrained liquid released from the break may be reduced by the action of the internal steam separators within each steam generator.
The maximum energy release to the containment will occur if the two-phase level remains below the exit pipe, so that only steam flows into the containment.
For this condition the maximum of primary system energy will be utilized in producing steam.
For a postulated steam line break accident, the applicant has calculated the mass and energy release to the containment using the LOCTIC code, which is described in amend-ments to the Final Safety Analysis Report.
The LOCTIC code calculates heat flow from the primary system into the steam generators as well as steam-water separation within each steam generator and entrained liquid carryover out the break.
The applicant has compared the energy release using the LOCTIC code with the experimental data contained in Battelle Northwest Laboratory Report BNWL 1463, "Coolant Blowdown Studies of a Reactor Simulator Vessel Containing a Perforated Sieve Plate Separator," dated February 1971 and General Electric Topical Report NEDO-10329, "Loss-of-Coolant Accident and Emergency Core Cooling Models for GE Boiling Water Reactors," dated April 1971.
In both cases the LOCTIC code approximately predicted the test data.
The action of the internal steam separators is taken into account by the LOCTIC code by assuming 100 percent steam separation. This assumption causes an additional 90 6-1
6.3 6.3.3 million British Thermal Units of energy to be added to the containment atmosphere.
We have therefore concluded that the applicant's method of calculating mass and energy releases following a steam line break is conservative since no entrainment is assumed.
The applicant has calculated the containment pressure and temperature response to various postulated main steam line break accidents assuming zero entrainment for the mass and energy release model and using the LaCTIC computer code with the assumption of no revaporization of the condensate formed on the heat sinks. The applicant has found, and we concur, that the worst case was a double-ended main steam line rupture upstream of the flow restrictor, at hot standby.
The applicant calculated a peak pressure of 43.6 pounds per square inch gauge, which is below the design pressure of 45 pounds per square inch gauge, and a peak atmospheric temperature of 442 degrees Fahrenheit.
Our confirmatory analysis was done using the CONTEMPT-LT MOD 26 computer code, assuming no revaporization of the condensate, and 'the app1icant's mass and energy release data.
Our results are in good agreement with the applicant's results.
We have examined the effect of the containment atmosphere temperature exceeding the design temperature on non-electrical equipment inside the containment and on the con-tainment structure itself.
On the basis of this examination we conclude that because of the large heat capacity of the structures and the relatively short period of time the containment atmosphere is above the containment design temperature of 280 degrees Fahrenheit, there would be no adverse effects on non-electrical equipment inside the containment and on the containment structure.
As discussed in Section 3.10 of this supplement, we are reviewing the effect of the resultant environment on electrical equipment inside containment in the event of a main steam line break and will report the results of our review in a supplement to the Safety Evaluation Report.
Based on our review of the material presented in Section 6.2.1 of the Final Safety Analysis report regarding the main steam line break accident analysis, we conclude that the containment functional design is acceptable. Therefore, we consider this matter resolved.
Emergency Core Cooling System Performance Evaluation In Section 6.3.3 of the Safety Evaluation Report, we had concluded that the emergency core cooling system analysis was in conformance with Appendix K to 10 CFR Part 50 and Section 50.46(b) of 10 CFR Part 50.
In a letter dated August 13, 1976, Westinghouse Electric Corporation has reported to us that measurements made in an operating plant and calculations have indicated that the temperature of the reactor coolant in the upper head region of the reactor vessel may be higher than the temperature which was assumed in the emergency core cooling system analysis for Westinghouse two, three and four loop plants. Using the higher tempera-ture for the upper head region in the analysis is expected to increase the calculated peak clad temperature when all the other inputs to the analysis remain unchanged.
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Therefore, this could affect our conclusions concerning the emergency core cooling system performance for Westinghouse reactors.
Since the analysis of the emergency core cooling system for the North Anna facility was also performed with this lower temperature, we have requested that the applicant perform a reanalysis of the North Anna facility to reaffirm that the emergency core cooling system design for this facility can still meet Appendix K to 10 CFR Part 50 and Section 50.46(b) of 10 CFR Part 50.
The applicant has informed us that he wiil submit this information in the near future.
When we receive the information to complete our evaluation of the applicant's reanal-ysis, we will report the results of the evaluation in a supplement to the Safety Evaluation Report.
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10.0 STEAM AND POWER CONVERSION SYSTEM 10.2 Main Steam Supply System In Section 10.2 of Supplement No. 2 to the Safety Evaluation Report we concluded that the main steam supply system is in conformance with the single failure criterion, the seismic recommendations of Regulatory Guide 1.29, "Seismic Design Classification," and main steam isolation valve closure time positions, and, is acceptable.
As a result of our review of the modified auxiliary feedwater system design, we have identified a common header, whose rupture could cause a slow blO\\~down of the three steam generators.
The main steam supply system furnishes steam to the auxiliary feedwater system turbine by means of branch lines off the main steam lines, upstream of the main steam isolation valves.
The present design includes a three inch branch line coming off each main steam line.
Each three inch line includes a manual valve and check valve.
The three lines are Joined in a header, from which two other three inch lines, each of which contains a remote air operated valve, transport the steam to the turbine. A rupture of this header could result in a non-isolable slow blowdown of the three steam generators.
The applicant has been informed that we require system modifications to prevent blowdown of more than one steam generator or demonstration by analysis that the postulated header rupture will not adversely affect the capability of the plant to attain a safe shutdown condition.
We will require that the applicant resolve this matter prior to a decision concerning the -j ssuance of an operati ng 1 i cense for North Anna Power Sta ti on Unit 1 and will report the resolution of this matter in a supplement to the Safety Evaluation Report.
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20.0 FINANCIAL QUALIFICATIONS 20.1 Introduction We stated in Section 20.0 of the Safety Evaluation Report that we were reviewing the applicant's financial qualifications to operate the North Anna Power Station Units 1 and 2 and maintain it in a safe shutdown condition.
We also stated that we would report the results of our evaluation in a subsequent report.
The Nuclear Regulatory Commission's regulations relating to the determination of financial qualifications of applicants for facility operating licenses appear in Section 50.33(f) and Appendix C of 10 CFR Part 50.
In accordance with these regulations, Virginia Electric and Power Company submitted operating cost estimates with its appli-cation as well as providing additional financial information at our request.
The following analysis summarizes our review of the financial information and addresses the Virginia Electric and Power Company's financial qualifications to operate North Anna Power Station, Units 1 and 2 and to permanently shut down the facility and maintain it in a safe shutdown condition, should that become necessary.
The Virginia Electric and Power Company is an investor-owned electric and gas utility serving most of Virginia and parts of North Carolina and West Virginia.
Electric sales account for approximately 96 percent of total operating revenues and gas sales, the balance.
Its customers include residential, commercial, industrial and wholesale users.
Operating revenues for the 12 months ended June 30, 1976 were 1.1 billion dollars and net income was 154.1 million dollars.
Invested capital at June 30, 1976 amounted to 3.7 billion dollars and consisted of 51.6 percent long-term debt, 15.3 percent preferred and preference stock, and 33.1 percent common equity.
The first mortgage bonds are rated A. upper medium grade, by both Moody's and Standard and Poor's.
20.2 Estimated Operating and Shutdown Costs For the purpose of estimating the units' annual operating costs, the applicant assumed that Unit 1 would begin operation in April 1977 and that Unit 2 would begin operation in November 1977.
The Virginia Electric and Power Company's estimates of the total annual cost of operating the units during each of the first five years of commercial operation are tabulated below.
The unit costs (mills per kilowatt hour) are based on a net electrical capacity of 907 megawatts electrical (per reactor) and on the follow-ing projected plant capacity factors:
Unit 1 - 1977 - 67 percent, 1978 - 58 percent, 1979 - 74 percent, 1980 - 74 percent, 1981 - 74 percent; and Unit 2 - 1977 - 73 percent, 1978 - 73 percent, 1979 - 69 percent, 1980 - 74 percent, 1981 - 74 percent.
The five-year average costs were calculated by annualizing the estimated costs for 1977 in combination with the annual estimates for 1978 through 1981.
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Unit 1 Unit 2 Total Cost Mi 11 s per Tota 1 Cost Mi 11 s per thousands kilowatt thousands kilowatt of dollars hour of dollars hour 1977
$ 85,243
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$ 13,339 13.8 1978 110,019 23.9 79,775 13.8 1979 115,069 19.6 77,529 14.1 1980 117,608 20.0 80,226 13.6 1981 121,126 20.6 82,797
- 14. I 5-year average 115,490 20.9 80,072 13.9 In estimating the costs of permanently shutting down the North Anna facility, the company assumed that the facil ity would be fully decommissioned and no longer used as a commercial nuclear power facility.
Expected decommissioning activities include processing, shipping and disposal of nuclear waste material, the sealing of plant components, and the reprocessing of nuclear fuel.
Decommissioning costs are estimated to total 27.0 million dollars.
The Virginia Electric and Power Company estimates the annual cost of maintenance after decommissioning to be 672,000 dollars.
Included in this estimate are the costs of a security force, surveillance, radiation monitoring and miscellaneous operating expenses.
20.3 Source of Funds The Virginia Electric and Power Company expects to cover all operating costs through the revenues generated from its system-wide sales of energy.
Current operating costs will be paid out of current operating revenues.
The estimated unit operating costs shown in Section 20.2, above, compare favorably with the Virginia Electric and Power Company's revenue experience.
Its average unit price for electricity sold during the year 1975 was 3.16 cents per kilowatt hour, well above the total estimated unit operating costs for the subject facility.
The company plans to meet the costs of decommissioning and subsequent maintenance also from operating revenues.
This would be supplemented by capital funding, if required.
In accordance with the regulations cited in Section 20.1, above, there must be reason-able assurance that the applicant can obtain the necessary funds to cover the estimated costs of the activities contemplated under the license.
This reasonable assurance standard must be viewed in light of the potentially long period of commercial utiliza-tion of the facility. Consequently, one must necessarily assume that there will be rational regulatory policies over this period with respect to the sett"ing of rates.
This implies that rates will be set to at least cover the cost of service, including the cost of capital.
Based on the preceding analysis, we have concluded that the Virginia Electric and Power Company is financially qualified to operate North Anna Power Station, Units and 2 and, if necessary, to permanently shut down the facility and maintain it in a safe shutdown condition.
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22.0 CONCLUSION
S In Section 22.0 of Supplement No. 2 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.
A number of these have been resolved, as reported in this supplement.
The outstanding items which must be resolved, including new items identified in this supplement, and their present status are summarized below.
Resolution of each item will be discussed in a future supplement to the Safety Evaluation Report.
(1)
Our review of the design of the system of well points for groundwater control is not completed (Section 2.6 of Supplement No.2 to the Safety Evaluation Report).
(2)
The applicant has provided our requested information regarding the dynamic analyses of the effects of a postulated loss-of-coolant accident on fuel elements.
Our evaluation of this information has not been completed (Safety Evaluation Report Section 4.2.4).
(3)
The applicant has recently submitted additional information regarding the pre-operational tests of the recirculation mode of operation for the low head safety injection pumps.
We are reviewing this information (Safety Evaluation Report Sections 6.3.4 and 14.0 and Section 6.3.4 of Supplement No.1).
(4)
The test program results to demonstrate that adequate electrical isolation exists between the safety related and non-safety related portions of the 7300 series process analog system have recently been submitted.
We are reviewing the test program results (Safety Evaluation Report Section 7.2.2).
(5)
The applicant has stated in Amendment 53 to the North Anna Power Station, Units and 2 Final Safety Analysis Report that it is considering modifications to the auxil i ary feedwater system in order to extend the time requi red for operator action to 30 minutes in the event of a feedwater line break.
In Amendment 54, the applicant has provided a revised analysis of the feedwater line break for a modified system design.
We are reviewing these analyses (Safety Evaluation Report Section 15.3).
(6)
The applicant must provide additional information on the seismic and environmental qualification of seismic Category I instrumentation and electrical equipment (Safety Evaluation Report Section 3.10 and Section 3.10 of this report).
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(7)
We plan to review overpressurization incidents of the reactor coolant system when in a water-solid condition on a generic basis (Section 5.2.8 of Supplement No.2 to the Safety Evaluation Report).
(8)
We have not yet taken a position on the age of last movement of the Stafford Fault Zone and are seeking additional evidence upon which to base a conclusion (Section 2.5 of Supplement No.2 to the Safety Evaluation Report).
(9)
The applicant must provide a reanalysis of the emergency core cooling system which will take into account the higher temperature of the upper head fluid (Section 6.5.3 of this report).
(10)
The applicant must provide a reanalysis of the stress distribution in the spent fuel pool (Section 3.8.2 of this report).
(11)
The applicant must either modify the steam supply system to the auxiliary feed-water turbine or demonstrate by analysis that a blowdown of the three steam generators in the event of a rupture of a common header will not adversely affect the capability of a safe shutdown (Section 10.2 of this report).
Subject to satisfactory resolution of the outstanding matters described above, the conclusions as stated in Section 22 of the North Anna Power Station, Units 1 and 2 Safety Evaluation Report remain unchanged.
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July 22, 1976 July 26, 1976 July 28, 1976 July 29, 1976 July 29,1976 July 30, 1976 August 2. 1976 August 3. 1976 August 4. 1976 August 5. 1976 APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW Division of Project Management letter concerning a warning alarm circuit in the solid state protection system.
Division of Project Management letter requesting additional information conc~rning Engineered Safety Features.
Submittal of Amendment No. 56 to the Final Safety Analysis Report.
Division of Project Management letter advising that the response to Part 5. as well as a revision to FSAR Table 5A.9-2 will be withheld from public disclosure as proprietary.
VEPCO letter transmitting additional test data obtained for the material used in North Anna steam generator supports.
Division of Project Management letter transmitting 20 copies of Supplement No. 2 to the North Anna Safety Evaluation Report.
VEPCO letter concerning reformatting the Final Safety Analysis Report for North Anna.
Summary of Meeting held on July 21. 1976. concerning Steam Generator and Reactor Coolant Pump Support Structures.
Division of Project Management letter requesting additional information.
VEPCO letter confirming telephone conversation concerning commercial operation date of North Anna.
Unit 2. from November 1977 to May 1978.
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August 9, 1976 August 11, 1976 August 12, 1976 August 12, 1976 August 16, 1976 August 24, 1976 August 26. 1976 August 26, 1976 August 26, 1976 August 27. 1976 August 30, 1976 August 31, 1976 Division of Project Management "letter advising of copy change requirements for SAR, Amendments and ER from 40 to 60 copies.
ACRS Subcommittee meets in Washington, D.C. to discuss the North Anna review.
VEPCO letter concerning the general warning alarm circuit in the solid state protection system.
ACRS Committee meets in Washington, D.C. to discuss the North Anna review.
VEPCO letter requesting an additional 30 days extension to justify withholding the electrical schematics submitted May 18, 1973.
(Withholding these electrical schematics was denied in DPM letter, dated July 7, 1976.)
Summary of the August 11,1976 ACRS Subcommittee meeting and the August 12, 1976 ACRS meeting.
Division of Project Management letter concern-ing the Impact of Stafford Fault Zone on the North Anna Power Station, Units 1 and 2.
VEPCO letter concerning a steel beam to be used in Unit 2 cubicle "c" reactor coolant pump support back weldment.
VEPCO letter' containing responses to questions contained ;n DPM letter dated August 4, 1976.
VEPca and NRC representatives meet in Bethesda, Md.
to discuss outstanding issues related to North Anna Station, Units 1 and 2.
Report on Laboratory Soil Testing at the North Anna Power Station Service Water Reservoir.
Memorandum and Order issued by the AS LAB deferring the appeal of Sun Shipbuilding and Dry Dock Company to intervene.
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Page APPENDIX B ERRATA TO SUPPLEMENT NO. 2 TO THE SAFETY EVALUATION REPORT FOR THE NORTH ANNA POWER STATION. UNITS 1 AND 2 Line 25 Add line 25 as follows:
"10.7 TURBINE MISSILES" 10-2 8-1