ML090720820
| ML090720820 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| FOIA-2024-000060 NUREG-0053 S4 | |
| Download: ML090720820 (28) | |
Text
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1~";IIII;ltittll I~ellttl*t related to operation of orth Anna Power Station Units 1 and 2 Virginia Electric and Power Company Supplement No.4 NUREG-0053 Suppl. No.4 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339 December 1976
Ava il ab 1 e from National Technical Information Service Springfield, Virginia 22161 Price:
Printed Copy $3.50 ; Microfiche $3.00
SUPPLEMENT NO.4 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, SUPP. 4 December 8, 1976
TABLE OF CONTENTS
1.0 INTRODUCTION
AND GENERAL DISCUSSION.
1.1 Introduction.....
6.0 ENGINEERED SAFETY FEATURES......
6.3 Emergency Core Cooling System..
6.3.3 Performance Evaluation.
7.0 INSTRUMENTATION AND CONTROLS....
7.2 Reactor Trip System......
7.2.2 Process Analog System 10.0 STEAM AND POWER CONVERSION SYSTEM.
10.2 Main Steam Supply System.
15.0 ACCIDENT ANALYSES.
15.3 Accidents.
22.0 CONCLUSION
S...
t PAGE 1-1 1-1 6-1 6-1 6-1 7-1 7-1 7 -1 10-1 10-1 15-1 15-1 22-1
APPENDICES PAGE APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW........ A-l i i
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, and Supplement No.3 was issued on September 15, 1976. Supplement Nos. 1, 2 and 3 to the Safety Evaluation Report documented the resolution of several outstanding items, and summarized the status of the remaining outstanding issues.
The purpose of this supplement is to update our Safety Evaluation Report (and Supplement Nos. 1,2 and 3) by providing (1) our evaluation of additional information submitted by the applicant since the issuance of Supplement No.3 of the Safety Evaluation Report, and (2) 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, except where specifically so noted.
Appendix A is a continHation of the chronology of our principal actions related to the processing of the application.
A summary of the remaining outstanding issues is presented in Section 22.0 of this supplement.
1-1
6.3 6.3.3 6.0 ENGINEERED SAFETY FEATURES Emergency Core Cooling System Performance Evaluation In Section 6.3.3 of Supplement No.3 to the Safety Evaluation Report, we stated that we had 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 the requirements of Section 50.46 of 10 CFR Part 50.
This was necessary because in a letter dated August 13, 1976, Westinghouse Electric Corporation had reported to us that results of measurements made in an operating plant and additional 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 analyses performed for Westinghouse two, three and four loop plants. Westinghouse has performed sensitivity studies that show the peak clad temperature of the fuel element following a postulated loss-of-coolant accident increases for higher upper head temperatures.
As a result, we have required that the 10ss-of-coolant accident analysis for large breaks (most limiting) be reanalyzed conservatively with an upper head temperature corresponding to the hot leg temperature.
In a letter dated October 12, 1976, the applicant submitted a loss-of-coo1ant accident analysis for three large pipe ruptures using the hot leg temperature in the upper head. Westinghouse Topical Report WCAP-8853 (October 1976), "Westinghouse ECCS -
Three-loop Plant (17 x 17) Sensitivity Studies," was referenced by the applicant and included the appropriate generic break study that used the increased upper head temperature.
This generic study indicated that the double-ended cold leg guillotine (DEClG) was still the most limiting break for three-loop plants. These calculations satisfy the break spectrum requirements of Section 50.46 of 10 CFR Part 50.
These analyses were performed with the same approved Westinghouse evaluation model, geometries, containment conditions and criteria that were used in the previously approved North Anna Power Station Units 1 and 2 calculations as stated in Section 6.3.3 of our June 4, 1976 Safety Evaluation Report.
The new analyses identified the worst break as a double-ended cold leg break with a discharge coefficient (Moody multiplier) of 0.4. The calculated peak clad temperature of the fuel element was calculated to be 2181 degrees Fahrenheit, which is below the acceptable limit of 2200 degrees Fahrenheit as specified in Section 50.46 of 10 CFR Part 50.
In addition, the calculated maximum local metal-water reaction of 7.86 percent and a total core-wide metal-water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and one percent, respectively. These analyses were performed with a total peaking factor (Fq) of 2.32 at 102 percent of the rated 6-1
nuclear steam supply system power level of 2775 megawatts thermal. Also, loss of offsite power and loss of a low head safety injection pump were assumed in the analyses to provide additional factors of conservatism.
Based on this review and that described in Section 6.3.3 of our Safety Evaluation Report, we conclude that the emergency core cooling system performance conforms to the acceptance criteria in Section 50.46 of 10 CFR Part 50.
Therefore, we consider this matter resolved.
6-2
7.0 INSTRUMENTATION AND CONTROLS 7.2 Reactor Trip System 7.2.2 Process Analog System In Section 7.2.2 of the Safety Evaluation Report, we stated that we would continue our generic review of the Westinghouse 7300 Series Process Analog System, verify its implementation at the North Anna facility during a site visit, and evaluate the results of the test program.
We also stated that upon completion of this generic investigation we would determine that if the North Anna facility process analog system requires any modification, we would require such modification.
In a letter dated September 1,1976, Westinghouse Electric Corporation transmitted a report entitled "Westinghouse 7300 Series Process Control System Noise Tests" dated August 1976.
We have reviewed the test program and the test results discussed in this report and conclude that selected postulated electrical faults will not degrade the systems safety actions below acceptable levels.
Therefore, we find the system acceptable for the North Anna Power Station Units 1 and 2 facility.
In addition, we have reviewed the applicant's commitment to make the applicable modifications resulting from the solder splatter problems.
These modifications are outlined in the Westinghouse report for North Anna Power Station Units 1 and 2.
Our Office of Inspection and Enforcement will verify that these modifications have been implemented prior to making a decision concerning the issuance of operating licenses for North Anna Power Station Units l/and 2.
On the basis of the applicant's commitment and satisfactory implementation of these modifications, we conclude that this matter is resolved.
7-1
10.0 STEAM AND POWER CONVERSION SYSTEr~
10.2 Main Steam Supply System In Section 10.2 of Supplement No.2 to the Safety Evaluation Report, we identified our concern regarding the possibility of a slow blowdown of all three steam ~enerators caused either by failure of the residual heat release valve which is common to the three steam generators, or failure of the common header supplying the subject valve.
As discussed in Supplement No.2 to the Safety Evaluation Report, the applicant proposed administrative and system changes in Amendments 53 and 54 to the Final Safety Analysis Report which would preclude the occurrence of the malfunction postulated above.
We concluded in Supplement No.2 that the modified system design was acceptable.
Subsequently, in Section 10.2 of Supplement No.3 of the Safety Evaluation Report, we stated that as a result of our review of the modified auxiliary feedwater system design, we determined that rupture of the auxiliary feedwater turbine steam supply header could also cause a slow blowdown of the three steam generators.
The main steam supply system furnishes steam to the auxiliary feedwater system turbine by means of branch lines leading off the main steam lines, upstream of the main steam isolation valves.
The system design includes a three-inch branch line coming off each main steam line. Each three-inch line includes a normally open manual valve and a 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, provides steam to the turbine. A rupture of this header could result in a non-isolable slow blowdown of the three steam generators.
We stated that the applicant was informed that we would 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.
In Amendment 58 to the Final Safety Analysis Report, the applicant submitted the results of an analysis which covers each of the failures postulated above.
Two limiting cases were analyzed.
The first case assumes maximum cooldown; i.e.,
reactor hot, subcritical, no decay heat, and all three auxiliary feedwater pumps operating. The second case assumes maximum heat generation; i.e., reactor initially at 102 percent nominal power increasing to 109.7 percent of nominal power at which point it trips due to steam generator low-low level, conservative decay heat generation, and one motor driven auxiliary feedwater pump s~arts. For both cases the applicant concluded that the reactor remains subcritical following shutdown, adequate core cooling is maintained, and the site radiological consequences are within requirements of 10 CFR Part 100 and are acceptable.
We have reviewed the applicant's data and assumptions which were included in the analysis and have concluded that the results 10-1
were acceptable since core cooling is maintained, the reactor remains subcritical and the radiological consequences meet our requirements.
Based on our review, we have concluded that the main steam supply system design is acceptable and, therefore, we consider this matter resolved.
10-2
15.0 ACCIDENT ANALYSES 15.3 Accidents In Section 15.3 of the Safety Evaluation Report. we stated that the applicant had not analyzed the case of a postulated feedwater line break without the coincident loss of offsite power.
We further stated that we would require that the applicant analyze the rupture of a feedwater line without the coincident loss of offsite power and that our evaluation of this matter would be presented in a supplement to the Safety Evalua-tion Report.
In Amendments 53 and 54 to the Final Safety Analysis Report, the applicant has modi-fied the auxiliary feedwater system to mitigate the consequences of a postulated feedwater line break accident.
The modified system involved changes of system piping and valves to ensure that a minimum of 340 gallons per minute is delivered to one or more intact steam generators even with a single failure coupled with a high energy line failure.
The applicant has also reanalyzed the postulated feedwater line break with this modified auxiliary feedwater configuration and the results show that (1) the primary system pressure does not exceed 110 percent of the design pressure. (2) no operator action is required for at least 30 minutes after initiation of the postulated accident.
and (3) the water discharge flow rate through the safety relief valves is less than 10 percent of their capacity.
We have reviewed the applicant's data and assumptions which were included in its analysis and concur with the conclusions.
Based on our review we have concluded that adequate safety systems have been provided to mitigate the consequences of a postulated feedwater line break.
Therefore. we consider matters related to the consequences of a postulated feedwater line break to be resolved.
15-1
22.0 CONCLUSION
S In Section 22.0 of Supplement No.3 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 supple-ment.
The outstanding items which must be resolved and their present status are summarized below.
Resolution of each item will be discussed in a future supplement to the Safety Evaluation Report.
(1)
The design of the system of well points for groundwater control has recently been submitted.
We are reviewing this design (Section 2.6 of Supplement No. 2 to the "Safety Evaluation Report).
(2)
The applicant must provide additional information regarding the dynamic analyses of the effects of a postulated loss-of-coolant accident on fuel elements (Safety Evaluation Report Section 4.2.4).
(3)
The applicant has recently submitted additiona"l 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 applicant must provide additional information on the seismic and environ-mental qualification of seismic Category I instrumentation and electrical equip-ment (Safety Evaluation Report Section 3.10 and Section 3.10 of Supplement No.3).
(5)
The applicant must provide additional information on overpressurization of the reactor coolant system when in a water-solid condition (Section 5.2.8 of Supple-ment No.2 to the Safety Evaluation Report).
(6)
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).
(7)
The applicant must provide a reanalysis of the stress distribution in the spent fuel pool (Section 3.8.2 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 of the North Anna Power Station, Units 1 and 2 Safety Evaluation Report remain unchanged.
22-1
September 1, 1976 September 3, 1976 September 3, 1976 September 7, 1976 September 8, 1976 September 10, 1976 September 10, 1976 September 15, 1976 September 15, 1976 September 16, 1976 APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW Summary of August 27, 1976 Meeting on Outstanding Issues.
VEPCO letter transmitting additional information which should permit resolution of several remaining out-standing issues.
VEPCO letter concerning the Stafford Fault Zone.
Order issued by Acting Secretary of the Commission.
Oral Argument will be held on Tuesday, October 5, 1976 at 10:30 a.m.
in the Commissioner's Conference Room, 11th Floor, 1717 H Street.
Notice and Order issued by the AS&LB concerning a Prehearing Conference to be held at 2 p.m. on 9/23/76 at H Street, N.W.
Washington, D. C.
Division of Project t*1anagement letter requesting information concerning steam and power conversion system.
VEPCO transmits Amendment No. 57 to the FSAR.
This Amendment consists of revised and updated pages to be inserted into the FSAR.
Amendment 57 contains responses to Division of Project Management questions in letters, dated 8/26/76, 8/31/76 and 9/3/76.
Memo to ACRS transmitting 20 copies of Supplement No. 3 to the North Anna Safety Evaluation Report.
Issuance of Supplement No. 3 to the North Anna Safety Evaluation.
Division of Project Management letter transmitting Supplement No. 3 to the North Anna Safety Evaluation, Units 1 & 2 to applicant.
A-l
September 16, 1976 September 17, 1976 September 17, 1976 September 2D, 1976 September 27, 1976 September 28, 1976 September 29, 1976 September 29, 1976 September 29, 1976 September 30, 1976 September 30, 1976 October 8, 1976 VEPCO letter advising that the Westinghouse Electrical schematics transmitted originally on May 18, 1973 are no longer considered proprietary and therefore withdraw the request for withholding.
VEPCO letter replying to Division of Project Management letter of September 10, 1976 concerning the blowdown of all three steam generators.
VEPCO letter transmitting a report entitled "An Analysis and Safety Evaluation of Spent Fuel Shipping Cask Handling at North Anna Power Station, Units 1 and 2."
VEPCO letter transmitting the program "Environmental Qualifi-cation of Westinghouse NSSS Scope Safety-Related Instrumenta-tion for North Anna, Units 1 and 2."
VEPCO requested withholding as proprietary material.
VEPCO letter addressing responses to Comments 10.22 and 10.23 in o. Parr's letter of 9/10/76.
Division of Project Management letter transmitting two proof and review copies of the North Anna Station Unit 1 Tech.
Specs.
VEPCO letter transmitting a response to Comment 10.24 concerning the integrity and inspection of turbine disks.
VEPCO letter concerning balance of plant (BOP):
NRC & VEPCO representatives meet in Bethesda, Md. to discuss matters related to the environmental qualifications of equipment.
Division of Project Management letter concerning Fire Protection Evaluation.
Order issued by AS&LB.
Order rules on outstanding issues concerning discovery, prehearing conference, final discovery requests, evidentiary hearing & Mrs. Arnold appearing on her own behalf.
A-2
October 11, 1976 October 12, 1976 October 12, 1976 October 22, 1976 October 22, 1976 October 26, 1976 October 28, 1976 October 28, 1976 October 28, 1976 October 29, 1976 November 1, 1976 November 4, 1976 November 5, 1976 November 9, 1976 November 11, 1976 VEPCO transmits Amendment No. 58 to the FSAR.
This amendment consists of deletion of pages, substitution and addition of pages and an attached sheet showing Tabulation of Changes.
Summary of September 30, 1976 meeting to discuss matters related to the environmental qualification of equipment.
VEPCO letter concerning three-break LOCA-ECCS reanalysis.
Division of Project ~1anagement letter requesting additional information.
Summary of October 13, 1976 ACRS Subcommittee r~eeti ngs and October 14, 1976, ACRS Committee Meeting.
ACRS releases its Report on Partial Review of North Anna Power Station, Units 1 & 2.
NRC representatives make a site visit to North Anna, Unit 1 site.
Division of Project Management letter transmitting the ACRS Report on Partial Review for North Anna, Units 1 & 2.
VEPCO letter transmitting a report entitled "Preliminary Report -
Seismic Monitoring with the Phase II Network - North Anna Power Station - May 1, 1976 thru August 15, 1976."
VEPCO letter advising that a fire protection plan will be submitted on or before April 1,1977.
VEPCO letter concerning groundwater control beneath the service water pumphouse.
Summary of Site Visit to Discuss Testing of the Recirculation Mode of ECCS Operation (Site visit held on October 28, 1976.)
Representatives from VEPCO and NRC meet to discuss matters related to the North Anna Steam Generator Supports.
Summary of November 5, 1976 Meeting with Representatives of VEPCO.
The meeting centered around discussions relating to the Steam Generator and Reactor Coolant Pump Supports.
Representatives from VEPCO, Westinghouse and NRC meet to discuss questions submitted by the NRC relating to fuel element loading during a LOCA.
A-3
November 11, 1976 November 12, 1976 November 15, 1976 November 15, 1976 November 16, 1976 November 1976 November 18, 1976 November 19, 1976 November 22, 1976 November 24, 1976 November 24, 1976 VEPCO letter concerning a proposed test of the LHSI pumps.
Opinion issued by the Secretary of the Commission.
The licensee was fined $32,500 for material omissions related to material false statements.
Order for Hearing issued by the Atomic Safety and Licensing Board.
The Board will take evidence on stipulated contentions of Mrs. Geraldine Arnold on November 30, 1976 at City Hall, Charlottes-ville, Virginia.
Notice of Evidentiary Hearing issued by the Atomic Safety and Licensing Board.
The Evidentiary Safety Hearing is scheduled to commence on November 30, 1976 at 9 a.m. in City Council Chambers, Second Floor of City Hall, 7th and Main Streets, Charlottesville, Virginia.
Representatives from NRC and VEPCO meet in Bethesda, Md. to discuss safety concerns expressed by the Advisory Committee on Reactor Safeguards Subcommittee related to the Containment and Auxiliary Power Areas.
Division of Project Management lette~ to all utilities concerning security plans.
Summary of November 11, 1976 Meeting with VEPCO and Westinghouse representatives to Discuss Loading on Reactor Internal Components During a LOCA.
VEPCO letter concerning Steam Generator Supports.
Division of Project Management letter concerning a request for additional information concerning Reactor Coolant System.
Division of Project Management letter concerning a request for additional information - Groundwater Control System.
Summary of November 16, 1976 Meeting with VEPCO to Discuss Matters Related to Auxiliary Power and Containment Systems.
A-4
November 24, 1976 November 24, 1976 November 30, 1976 November 30, 1976 VEPCO letter transmitting Amendment No. 59 to the Final Safety Analysis Report.
This Amendment reformats the Emergency Plan.
VEPCO letter concerning installed transmitters, their function and the planned replacement models.
Division of Project Management letter concerning VEPCO's request for withholding information from public disclosure. The Division of Project Management requested additional information on the affidavit for Comment 5.71.
VEPCO 1 etter transmitti ng a Stone & Webster Report, "Safety Related Equipment Temperature Transients During the Limiting Main Steam Line Break."
A-5
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