ML090720815
| ML090720815 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/30/1976 |
| From: | Office of Nuclear Reactor Regulation, Virginia Electric & Power Co (VEPCO) |
| To: | |
| References | |
| FOIA-2024-000060 NUREG-0053 S01 | |
| Download: ML090720815 (32) | |
Text
SUPPI:.EMENT ~O.
TO THE SAFETY EVALUATION-REPORT BY THE OFFICE OF NUClEAR REACTOR REGUI:.ATION U. S. NUCI:.EAR REGUI:.ATORY COMMtSStON IN THE MATTER OF VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION-UNITS 1 AND z DOCKET NOS.
50~338 AND 50-33~
NUREG-0053. SUPP.
June 30. 1976
1.0 2.0 6.0 7.0 8.0 TABtE-SF-eSNTENTS INTRODUCTION AND GENERAL DISCUSSION
1.1 INTRODUCTION
SITE CHARACTERISTICS 2.3 METEOROLOGY 2.3.3 Onsite Meteorological Measurements Program ENGINEERED SAFETY FEATURES 6.2 6.3 6.4 CONTAINMENT SYSTEMS 6.2. I Containment Functional Design EMERGENCY CORE CooLi NG SYSTEM 6.3.3 6.3.4 Performance Evaluation Tests and Inspections CONTROL ROOM HAB ITAB I LlTY SYSTEMS INSTRUMENTATION AND CONTROLS
- 7. I 7.2 7.3 GENERAL REACTOR TRIP SYSTEM 7.2.1 Reactor Trip System Actuation Logic ENGINEERED SAFETY FEATURES ACTUATION AND CONTROL SYSTEMS 7.3.2 7.3.4 7.3.5 Engineered Safety Features ~ctuation System Containment Depressurization System Changeover From Inspection to Recirculation Mode ELECTRICAL POWER SYSTEMS 8.4 PHYSICAL INDEPENDENCE OF ELECTRICAL, INSTRUMENTATION, AND CONTROL SYSTEMS
22.0 CONCLUSION
S PAGE I-I I-I 2-1 2-1 2-1 6-1 6-1 6-1 6-2 6-2 6-4 6-5 7-1 7-1 7-1 7-1 7-1 7-1 7-2 7-2 8-1 8-1 22-1
APPENDIX A APPENDIX B APPENDteES CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW ERRATA TO THE SAFETY EVALUATION REPORT FOR THE NORTH ANNA POWER STATION, UNITS I AND 2 PAGE A-I 8-2
1.0 iNTRODUCTION AND GENERAl-DISCUSStON
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 I and 2 (North Anna facility, North Anna plants, facility or plant).
The application was filed by Virginia Electric and Power Company (applicant).
The Safety Evaluation Report identified certain matters as either requ~r'ng additional information from the applicant or as still being reviewed, and stated that these matters would be covered in a subsequent report.
The purpose of this supplement 6s to update our Safety Evaluat~on Report by providing (I) our evaluation of additional information subm~tted by the applicant since the ~ssuance of the Safety Evaluation Report; (2) information regarding the current status of matters that were stil I under review and (3) additional information for those sections of the Safety Eval uation Report where further narrative or changes are in order.
Each section of this suppiement ~s 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.
Appendfx A is a cont~nuation of the chronology of our princ~pai actions related to the processing of the application. and Appendix B is a listing of errata to the Safety Evaluation Report.
A summary of the remaining outstanding ~ssues is presented ~n Section 22,0 of this supplement.
I-I
2.0 SITE CHARAeTERISTleS 2.3 METEOROLOGY 2.3.3 Onsite-Meteorotogicai Measurements Program We stated in the Safety Evaluation Report that the applicant had recently submitted an additional year of onsite meteorological data for the period covering April 1974 to April 1975.
We also stated that we were evaluating the data and would report our findings in a subsequent report.
We have now completed our review of this second year of meteorological data.
The applicant provided joint frequency distributions of wind speed and direction from the 35-foot level by atmospheric stability (defined by the vertical temperature gradient between 35 feet and 150 feet) in the format suggested in Regulatory Guide 1.23.
Wind speed and direction were determined from the low-threshold wind sensor on the satellite tower.
Data recovery for this period was 92 percent.
We have used these data to perform a confirmatory evaluation of relative concentration values previously calculated using onsite data collected from September 16, 1971 through September 15, 1972.
We have compared relative concentration values calculated using each of the availaDle data sets. For the relative concentration value for the two-hour time period, which is exceeded five percent of the time, at the exclusion distance of 1350 meters, there was no difference between the 2-1
two calculated values.
For the relative concentration values for the various time periods at the low popuiat~on zone boundary distance of 9656 meters~ the relative concentration values calcula"ted using data for the period April 1974 through April 1975 were generally only about ten percent higher than those presented in the Safety Eval uation Report.
These val ues are as follows:
Time Period 0-2 hours 0-8 hours 8-24 hours 1-4 days Rei ative* Concentration (seconds'per"cobic"meter)
-5 4.2 x 10
-5 1.8 x 10
-5 1.1 x 10
-6 4.6 x 10 v " 10
-6 1.2 x 10 On this basis, we conclude that the rela-rhe concentraNon values sentative est&mates of the atmospherijc diffusion characteristics at the North Anna site.
However. we d~d reevaluate the doses presented in Table 15.2 of the Safety Evaluation Report with these slightly higher concentrat~on values and have determined that those values presented in Table 15.2 remain unchanged and the conclusions of Section 15.0 remain as stated.
We cons~der this matter resolved.
6.2 6.2.1 6.0 ENGtNEEREB SAFETY FEAT~RES C0NTA I NMENl SYSTE1o'1S Containment Functional Desi9!1, We stated in the Safety Evaluation Report that we required and the appl~cant agreed to verify the structural integrity of the lower reactor cavity and the lower steam generator subcompartments and the supports for components located in these compartments using our calculated peak pressures, or to provide additional infor-mat
~n support of their peak differential pressure calculat We also stated that the resolution of this matter will be discussed in a subsequent report.
In Amendment 53 of the Final Safety Anaiysis the applicant supplied additional information for the analysis of the lower steam generator subcompartment.
Using this additional information we performed an inde-pendent analysis to confirm the applicant's calculated peak pressure.
With respect to the lower reactor cavity and component supports iocated tn the cavity, the applicant has verified the structural integrity by using our calculated peak differential pressure and the equations and methods outlined in Section 3 of the Final Safety Analysis Report, which we found acceptable.
On "rhe basi s of our rev iew, we have concl uded that the designs of the containment steam generator subcompartment and the lower reactor cavity are acceptable and we consider this issue resolved.
6-i
6.3 EMERGENCY CORE COOtlNG SYSTEM 6.3.3 Performance Evaluation We stated in the Safety Evaluation Report that following a postulated loss-of-coolant accident the initial cold leg injection must be followed by either alternate hot and cold leg injection with the time period between them sufficiently short to prevent a boric acid buildup above 23.5 weight percent, or by simultaneous hot and cold leg injection.
We also stated that we wil I require the applicant to modify their procedures accordingly and would report the resolution of this issue in a subsequent report.
In Amendment 53 of the Final Safety Analysis Report, the applicant has stated that wr4tten emergency operating procedures, to be established prior to fuel loading, wil I require switchover from cold leg to hot leg recirculation 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after a postulated loss-of-coolant accident and following the initial switchover, hot and cold leg injection wll I be alternated at intervals not to exceed 27-1/2 hours.
On the basjs of the applicant's commitment in the Final Safety Analysis Report, we consider this issue resolved.
With respect to the operation of the emergency core cooling system, the applicant in Amendment 53 of the Final Safety Analysis Report stated that one of the three charging pumps will always be locked out of service.
This proposed mode of operation is an assumption in the minimum net positive suction head calculation for a single low head safety injection pump drawing water from the sump.
This proposed operation is also consistent with the loss-of-coolant accident analysis which takes credit for flow from only one charging pump for smal I breaks.
6-2
We find the applicant!s proposed operation with only two charging pumps available to be acceptable because it is consistent with the design basis accident safety analyses.
The technical specifications will reflect this requirement in the limiting conditions of oepration.
In the Safety Evaluation Report we requested that the applicant identify al I valves with electrical operators whose failure could result in total loss of a system function and provide design modifications to preclude the loss of the system function.
The applicant has provided a modified design which includes the disconnecting of electrical power to those valves identified as the result of the above review.
We required that the applicant utilize Electrical, Instrumentation and Control Systems Branch Technical Position 18, "Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves", which is included in Appendix 7A of the Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG-75/087) in implementing their design modification.
We verified the implementation of the design modi-fication during the rev6ew of additional drawings and during our plant visit and have found"it acceptable and therefore consider this matter resolved.
As stated in the Safety Evaluation Report we requested that the applicant identify al I of the electrical equipment, and components, safety and non-safety, that may be submerged as the consequence of a postulated loss-of-coolant accident and the resultant effects on plant safety.
6-3
The applicant has provided a description of the equtpment that may be submerged and the effect all plant safety; however# they have not identiHed the specific equipment involved.
We will require the appiicant to provide a list of specific electrical equipment that may be submerged as the result of a postu!ated ioss-ot-coolant accident condition.
We will report the results of our eval uation of the Ust of the specific submerged equipment and effects on plant safety in a future supplement to the North Anna Power Station, Units i and 2 Safety Evaiuat~on Report.
6.3.4 Tests and-tnspections We stated tn Section 6.3.4 and 14.0 of the Safety Evaluation Report that (I) the applicant has not provfded adequate documentation on the maximum and minimum How criter-ja for each of the pumps and a basis for these criteria, (2) the applican+ has not provided adequate information on how he will use the testing data to demonstrate adequate net posit1ve suction head margin for the low head safety injection pumps when they are drawing water from the containment sump; and (3) we wil! report the results of our evaluation concerning this aspect of the preoperational test program in a subsequent report.
In Amendments 41 and 49 of the Ftna4 Safety Analysis Report the applicant has specif~ed minimum and maximum tlow requirements for the low head safety injection pumps in order to satisfy net pos~tive suction head and emergency core cooling criteria. Preoperational tests wil' be performed to demonstrate acceptabie performance of these pumps under the most 6-4
adverse design conditions.
Since the low head safety injection pumps will not be tested from the sump, the applicant stated in Amendment 53 of the Final Safety Analysis Report that procedures will be developed to correlate the results of the suction testing of the recirculation spray pump and flow testing of the low head safety injection pump from the refueling water storage tank to the calculated values for flow and net positive suction head when the low head safety injection pumps are taking suction from the containment sump.
We wil I review these procedures for demonstrating adequate net positive suction head for the low head safety injection pumps prior to conducting the relevant preoperational tests.
We will report the results of our evaluation in a future supplement to the Safety Evaluation Report.
6.4 CONTROL-ROOM-HABtTABtLtTY-SYSTEMS We stated in Sections 6.4 and 14.0 of the Safety Evaluation Report that we will review the procedures for preoperational testing of control room leakage and we would report the results of our evaluation in a subsequent report.
In Amendment 53 of the Final Safety Analysis Report the applicant has committed to preoperation and periodic tests of the control room to assure that inleakage of the contaminated air will not occur during an emergency.
In addition, the applicant has provided a description of the test procedures.
We have reviewed this information and have concluded that the applicant's testing procedures of the control room to assure that inleakage of the contaminated air will not occur during an emergency are acceptable.
Therefore, we consider this matter resolved.
6-5
1~O INSTR~MENTATteN-AND-eeNTRelS 7.1 GENERAL We stated in the Safety Evaluation Report that we had not completed our review of the instrumentation and control system logic and schematic diagrams.
We also stated that our final plant visit to review the imple-mentation of the design criteria was an outstanding item.
We have now completed our final plant visit and our review of the instrumentation and control system logic and schematic diagrams.
The results of this review are discussed in the following paragraphs.
7.2 REAeTeR TRIP SYSTEM 7.2.1 Reactor Trip System Actuation-logic During our review we examined those aspects of the reactor trip system that initiate, monitor, control, bypass and test the reactor trip system including selected logic diagrams and detailed schematics.
We have also verified the design implementation and adequacy of the physical separation of the safety and non-safety cabling and wiring during our plant visit and found it acceptable.
Therefore, we conclude that the reactor trip system is acceptable and we consider this issue resolved.
7.3 ENGINEERED'SAFETY'FEAT~RES-AeT~ATteN-AND-eeNTRel'SYSTEMS 7.3.2 Engineered-Safety-Features-Actaation-System Our review examined those aspects of the protection systems that initiate, control, monitor, test, and bypass the engineered safety feature systems and their auxiliary supporting systems.
As indicated in Section 7.1 of 7-1
this report, we have reviewed selected drawings and verified the implementation of the designs during our plant visit. Therefore, we conclude that the engineered safety feature actuation logic system is acceptable and we consider this issue resolved.
7.3.4 Containment Depressurization System We
~ave verified the implementation of the modified design of the containment depressurization system as, described in Section 7.3.4 of the Safety Evaluation Report, during the review of additional drawings and our plant visit and have found it acceptable.
Therefore, we consider this issue resolved.
7.3.5 Changeover from Injection to"RecirculationMode As stated in Section 7.3.5 of the Safety Evaluation Report, the applicant has revised the design of the emergency core cooling system and containment spray systems to provide automatic changeover from the injection mode of operation to the recirculation mode.
We have reviewed the details for the instrumentation and controls for this automatic function during our review of additional drawings a"t our plant visit to verify that no single fai lure of the inter10cks in one train of equipment can adversely affect the other train of equipment.
In addition, we have verified the testing provisions for this automatic function.
We have concluded that the automatic changeover from injection to recircu-lation and the associated test provisions are acceptable.
Therefore, we consider this issue resolved.
7-2
8.0 EtEeTRteAL POWER SYSTEMS 8.4 PHYSICAL INBEPENBENEE OF ElEETRtEAt, INSTR~ME~IJON, AND eONTROl SYSTEMS I n the Safety Eval uation Report we stated that we had rev iewed the criteria and procedures for providing physical independence of safety related circuits and equipment and would verify their implementation during our plant visit.
We have since conducted our paint visit and have reviewed the implementation of these cr"iteda and procedures and have cone! uded that the implementation is acceptable.
We.!rherefore consider this matter resolved.
8-1
22.0 eGNet~StGNS In Section 22.0 of the Safety Evaluation Report we stated that several items as set forth in Section 1.7 of the Safety Evaluation Report were still outstanding, and that satisfactory resolution of these items would be required before operating licenses for North Anna Power Station Units I and 2 could be issued.
A number of these have been resolved, as reported in this supplement. The remaining items which must be resolved and their present status are summarized below.
(I) The applicant has provided his reevaluation of the probable maximum flood analysis.
Our evaluation of this information has not been completed.
(Safety Eval uation Report Sections 2.4.2 and 3.4)
(2)
The applicant has provided his evaluation of foundation settlement for the service water reservoir and pump house.
The applicant has also provided an assessment of settlement potential for other seismic Category I structures.
Our evaluation of this information has not been completed.
(Safety Evaluation Report Section 2.6.2)
(3)
The applicant has provided our requested information regarding the dynamic analysis 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) 22-1
(4)
The applicant has not yet submitted information regarding the pre-operational tests of the recirculation mode of operation for the low head safety injection pumps.
(Safety Evaluation Report Sections 6.304 and 14.0 and Section 6.3.4 of this report)
(5)
The test program results to demonstrate that adequate electrical isolation exists between the safety related and non-safety related portions of the 7300 series process analog system have not yet been submitted.
We wil I verify its implementation and report our conclusions in a future supplement to the Safety Analysis Report.
(Safety Evaluation Report Section 7.2.2)
(6)
The applicant has not yet provided the results of an analysis to demonstrate that the reactor can be brought from a hot standby condition to a safe shutdown condition in the event that the residual heat release valve sticks open or if a pipe ruptures downstream of the stop check valves.
(Safety Evaluation Report Section 10.2)
(7)
The applicant has not yet provided his reevaluation of the plant design to demonstrate compliance with the new Appendix I to 10 CFR Part 50.
(Safety Evaluation Report Section 11.1) 22-2
(8)
The applicant has supplied ali the requested information with regard to the Emergency Plan to demonstrate that it meets Append'x E to 10 CFR Part 50 except for several letters of agreement with agencies within Spotsylvan~a County. Virghlia.
Our evaluation of this infor-mation has not been completed.
(Safety Evaluat~on Report Sectton 13.3)
(9)
The applicant has not yet provided adequate Information relating to the staffing and organizational respons~billtles of personnel managing and directing the preoperational test program.
(Safety Eval ua-r~on Report Seci'ion 14.0)
( 10)
The app! kant has stated in Amendment 53 to the Nor-rh Anna Power S*rai*~on Units I and 2 F~nal Safety Anal ')lsi s Report that he & s considering modifkat~ons to his aw<il ~ary feedwater system
~n order to extend *~he t&me required for operator action to 30 minutes in the event of a feedwai"er line break.
He has cammit*~ed to provid~ng a revised analysis of the feedwater line break when the modified design has been completed.
We wll! review these analyses when they are subm~tted and report our evaluation in a future supplement to the Safety Evaluation Report.
(Safety Evaluation Report Section 15.3)
( II)
We have not yet camp ieted our eval uation of one year of mkrosef smk data for the site area, (Safety Evaluation Report Section 2.5) 22-3
(12)
The applicant must provide additional information on the seismic and environmental qualification of seismic Category I jnstrumentation and electrical equipment.
(Safety Evaluation Report Section 3.10)
(13)
We have not completed our evaluation of the steam generator and reactor coolant pump supports.
(Safety Evaluation Report Section 5.4.2)
(14)
We have not completed our evaluation of the mass and energy release rates for a postulated main steam line break accident.
(Safety Evaluation Report Section 6.2.1)
(15)
The applicant must provide additional information concernfng all components and electrical equipment that may be submerged following a postulated loss-of-coolant accident.
(Safety Evaluation Report Section 6.3.3 and Section 6.3.3 of this report)
(16)
We have not completed our evaluation of the applkant's financial qualifications to operate the facility.
(Safety Evaluation Report Section 20.0)
(17)
We wil I ver'fy the acceptability of the service water reservoir for two-unit operation after we have evaluated the results of the initial operational testing program for North Anna Power Station Unit I.
(Safety Eval uation Report Section 2.4.3) 22-4
Subject to satisfactory resolution of the outstanding matters described above, the conclusions as stated in Section 22 of the North Anna Power Station, Units I and 2 Safety Evaluation Report remain unchanged.
22-5
APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REV!EW June 4, 1976 June 9, 1976 June 9, 1976 June II, 1976 June 141 1976 June 15, 1976 June 16, 1976 Subm~ttai of Amendment No. 52.
Representatj ves fl-om VEPCO & NRC meet to discuss outstanding issues.
Order issued by AS&LB.
Sun Shi piS petHion to intervene is granted and made a party to the proceeding.
VEPCO letter transmitting a response to Part 5 of the Staff's Comment 5.67 and a rev~sion of FSAR Table 5A.9-2.
VEPCO requested w$thhoiding from public disclosure.
Representatives from VEPCO & NRC meet to discuss foundation des~gn as ~t relates to the North Anna
- Station, ACRS issues notke of Subcommiti"ee meeting to be held on July 7,1976 in WashingTon, D. C.
Submittal of Amendment No. 53.
A-I
PAGE 1-8 1-8 2-13 APPENDIX B ERRATA TO THE SAFETY EVALUATION FOR THE NORTH ANNA POWER STATION; UNITS AND 2 LINE 21 28 17 Delete ilinstrumen+a.on" and add
"~n str* umentat ion Ii Delete "preoperational" and add
"~n itia I operationa I" Delete "preoperat&onal and add "initial operationai'j B-1