ML20010H127

Partial Response to FOIA Request for Documents Re Radiation Levels Which May Exist in Structures,Sys or Components of Nuclear Power Reactor Low Power Test Program.Forwards Documents Listed in App a

ML20010H127
Person / Time
Site:	North Anna
Issue date:	08/05/1981
From:	Felton J NRC OFFICE OF ADMINISTRATION (ADM)
To:	Lanpher L HILL, CHRISTOPHER & PHILLIPS
Shared Package
ML20010H128	List: ML20010H127
References
FOIA-81-270 NUDOCS 8109230707
Download: ML20010H127 (2)

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o UNITED STATES f 'g NUCLEAR REGULATORY COMMidSION

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fu) i Mr. Lawrence Coe Lanpher s

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1900 M Street, N.W.

IN RESPONSE REFER' ~

Washington, DC 20036 TO F01A-81-270

Dear Mr. Lanpher:

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This is in reply to your letter dated July 16, 1981, in which you requested, pt:rsuant to the Freedom of Information Act, copies of documents relating to radiation levels which exist or would exist in the structures, systems or components of a nuclear power reactor after completion of a low power (up to approximately 5 percent of full power) test program.

As'a. result of Mr. Frank W. Karas' telephone conversation with you from this office on July 16, 1981, Mr. Karas clarified the scope of the request by talking with members of your client's establishment, MHB Tecnnical Associates of San Jose, California.

It was agreed that our search for records would include (1) documents dated from March 28, 1979 to July 16, 1981, which relate to the Near Term Operating Licenses for North Anna Unit 2, Salem Unit 2, and Sequoyah Unit 1, and which provide the information requested in paragraphs one thru three of your letter dated July 16, 1981; and (2) Generic Reports from vendors and the NRC staff concerning (a) evaluation of low power testing, and (b) evaluation of testing programs in general.

In partial response to item one above, we have located the documents identified on Appendix A, and a copy of each is enclosed.

Our search for other documents which may be subject to your request is continuing, and we will communicate with you again when that search ef fort is completed.

Sincerely, j

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th J. 4 ton, Director-Di ision of Rules and Records

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Office of Administration

Enclosures:

As stated

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'J' Re: F0!A-81-270 APPENDIX A

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April.10,1980 ;

NUREG-0053, Supplement Po. 10, Docket No. 50-339,

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Safety Evaluation Report related to operation of-North Anna Power Station, Unit ?. Virginia Electric

~ Power Company.

2.

July-3, 1980 Letter, to: J. H.' Ferguson from: H. R. Denton,

Subject:

Issuance of Amendment No. I to Facility License No. NPF-7 North Anna Power Station, Unit No. 2'(50 pages).

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July 25, 1980

~ Memorandum, to: R. L. Tedesco from: V.-A. Moore

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Subject:

Transmittal. of North Anna 2 SER Imput for Special Low Power Test Program (6 pages).

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August 1980-NUREG-0053, Supplement No.11, Docket No. 50-339, Safety Evaluation Report related to operation of North Anna Power Station, Unit 2, Virginia Electric Power Company.

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DISTRIBUTION:

Docket File NRC/PDR l8.

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L. Swanson, OELD D. Eisenhut

-R. Purple R. Tedesco J. Youngblood L. Engle-K. Parrish

'M..Rushbrook L. Cobb, OIE OIE (5)

D. Biddle, MPA I. Bailey, (4)

R. Diggs, ADM R. Mattson S. Ilanauer R. Vollmer, DE D..Ross, SI B. Scharf (10)

I. Dinitz, UFB/DE J. Sal tzman, DE -

11. Bustoui, NMSS C. Miles M. Virgilio A. Dromerick NSIC TIC ASLAB ASLBP

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UNITED STATES 0

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NUCLEAR REGULATORY COMMISS!ON D

4 WASHINGTON, D. C. 20555 July 3, 1980 r

Docket No.:

50-339 Mr.

J. H. Ferguson Executive Vice President - Operations

' Virginia Electric and Power Company P. O. Box 26666 Richmond, Virginia 23261

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SUBJECT:

ISSUANCE OF AMENDMENT NO. 1 TO FACILITY LICENSE N0. NPF-7 NORTH ANNA POWER _

il0N, UNIT NO. 2 The Nuclear Regulatory Comission (the Commission) has issued Amendment No. I to Facility License NPF-7 (Enclosure 1) in accordance with your letter, dated June 18, 1980, requesting exceptions to the North Anna Power Station, Unit 2 Appendix A Technical Specifications which will permit you to perform the special low power test program identified in Conditions 0.6(b) and (c) of License NPF-7.

In discussions with representatives of your staff, it was agreed upon that an exception to conduct the low power test program would not be required for Technical Specifications 3.8.2.1 and 3.8.2.3.

In your letter of June 18, 1980, you also provided a safety analysis to support performing the low power test program. Operating procedures for conducting this program were provided by you in your letter, dated June 13, 1980.

We have reviewed the above information and have concluded that an exception to the

' Technical Specifications for conducting low power testing is acceptable and that VEPC0's tcst procedures for low power testing are acceptable and can be perfonned without posing an undue risk to the public. Our Safety Evaluation regarding this matter is presented in Enclosure 2.

Thh amendment authorizes the Virginia Electric and Power Company to conduct the special low power test program as defined in the Safety Evaluation and in Appendix A Technical Specification 8.13.

Enclosure No. 3 is a copy of the Federal Register Notice of Issuance of Amendment No. I to License No. NPF-7.

In your letter of May 30, 1980, you submitted emergency operating procedures with respect to Condition D.6(a) for the small break loss-of-coolant accident and inadequate core cooling. We have reviewed these emergency operating procedures and have concluded that they are acceptable for operation at power levels not exceeding five percent. Our evaluation of this matter is presented in Enclosure 2.

Our Office of Inspection and Enforcement has advised us that matters related to Condition D 6(d) of License NPF-7 and Items 7.3(1) through 7.3(3) cf Appendix A to the Technical Specifications related to operation above zero power have been satisfactorily completed. Therefore we consider these matters resolved.

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Based on the above we have determined that items required to be completed prior to conducting the low power test program for North Anna Power Station Unit 2 have been satisfactorily resolved, and therefore, operation at. power levels not exceed-ing five: percent is permitted.

In your letter of June 18, 1980, you also requested that the listing of safety i

related hydraulic' snubbers (Table 3.7-4 of Appendix A to the Technical Spe:ifications) be revised. We have not yet completed our review of this matter. Upon completion of our review we will advise you of the results of our evaluation.

Sincerely, N l.

Harold R. Denton, Director Office of Nuclear Reactor Regulation

Enclosures:

1.

Amendment No. I to NPF-7 with Technical Specification page change.

-2.

Safety Evaluation for Special Low Power Test Program 3.

_ Federal Register Notice cc w/ enclosures:

See next page

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1 Mr. J. ;i. Ferguson

'l Executive Vice President - Power Operations Virginia Electric A Power Company P. O. Box 26666 Richmond, Virginia 23261 cc:

,Mr. Anthony Gambaradello Clarence T. Kipps, Jr., Esq.

Office of the Attorney General 1700 Pennsylvania Avenue, N.W.

11 South 12th Street - Room 308 Washington, D. C.

20006 Richmond, Virginia 23219 Carroll J. Savage, Esq.

Richard M. Foster, Esq.

1700 Pennsylvania Avenue, N.W.

Musick, Williamson. Schwartz Washington, D. C.

20006 Leavenworth & Cope, P. C.

P. O. Box 4579 Mr. James C. Dunstan Boulder, Colorado 80306 State Corporation Commission Commonwealth of Virginia Michael W. Maupin, Esq.

Blandon Building Hunton, Williams, GTy & Gibson Richmond, Virginia 23209 P. O. Box 1535 Richcond, Virginia 23212 Alan S. Rosenthal, Esq.

Atomic Safety and Licensing Appeal Board Mrs. June Allen U.S. Nuclear Regulatory Commission 412 Owens Drive Washington, D. C.

20555 Huntsville, Alabama 35801 Michael C. Farrar, Esq.

Mr s. James Torson -

Atomic Safety and Licensing Appeal Board 501 Leroy U.S. Nuclear Regulatory Commission Socorro, New Mexico 87801 Washington, D. C.

20555 Mrs. Margaret Dietrich

.. Dr. John H. Buck Route 2, Box 568 Atomic Safety and Licensing Appeal Board Gordonsville, Virginia 22942 U.S. Nuclear Regulatory Conmission Washington, D. C.

20555 William H. Rodgers, Jr., Esq.

Georgetown University Law Center Atomic Safety and Licensing Board Panel 600 New Jersey Avenue, N. W.

U.S. Nuclear Regulatory Commission

' Washington, D. C.

20001 Washington, D. C.

20555 Mr. Peter S. Happ Mr. Michael S. Kidd Executive Vice President U.S. Nuclear Regulatory Commission Sun Shipping a Dry Dock Company P. O. Box 128 P. O. Box 540 Spotsivania, Virginia 22553 Chester, Pennsylvania 19013 Dr. Paul W. Purdom Mr. R. B. Briggs Department of Civil Engineering Associate Director Drexel University 110 Evans Lane Philadelphia, Pennsylvania Oak Ridge, Tennessee 37830 19104

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Kendal-at-Longwood Kennett Square, Pennsylvania 19348 Mr. Irwin R. Kroot Citizens Energy Forum p

P. O. Box 138 McLean, Virginia 22101

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James B. Dougherty, Esq.

Potomac Alliance 1416 S Street, N.W.

Washington, D. C.

20009 W. R. Cartwright, Station Manager P. O. Box 402 Mineral, Virginia 23117 W. L. Stewart, Manager P. O. Box 315 Surry, Virginia 23883 Mr. J. B. Jackson, Jr.

Commonwealth of Virginia Council on the Environment 903 9th Street Office Building Richmond, Virginia 23219 Mr. A.

D. Johnson, Chairman Board of Supervisors of Louisa County Trevillians, Virginia 23170 Mr. Bruce Blanchard Environmental Projects Review-

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Department of the Interior 18th and C Street, N. W.

Washington, D. C.

20240 U.S. Environmental Protection Agency ATTN:

Ms. Elizabeth V. Jankus-j Office of Environmental Review Room 2119M, A-104 401 M Street, S.

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Washington, D.

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20460 U.S. Environmental Protection Agency Region III Office Attn:

EIS Coordinator Curtis Building 6th and Walnut Streets Philadelphia, Pennsylvania 19106 9

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UNITED STATES E

NUCLEAR REGULA10RY COMMISSION o

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a VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 s

NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY LICENSE Amendment No. 1 License No. NPF-7

-1..The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment by the Virginia Electric and Power Company (licensee), dated June 18, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, (the Act) and the Commis, ion's regulations set forth in 30 CFR Chapter I; B.

The facility will operate in confonnity with the license, as amended, the provisions of the Act, and the regulations of the Commission;

.C.

There s reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's' regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is~ amended by the addition of Technical Specification 8.13 to Appendix A of the Technical Specifications.

This addition pennits the licensee to perform the special low power test program identified in Conditions D.6(b) and 0.6(c) of License NPF-7.

This license is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. I are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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This license emendment is effective as of the date of its issuance.

FOR Tile NUCLEAR REGULATORY COMMISSION h/WW Harold R. Denton, Director

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Office of Nuclear. Reactor Regulation

Attachment:

Page.8-3 to Technical

' Specification Appendix A Date of Issuance:

July 3, 1980 9

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I 8.8 Prior to startup following the first regularly ccheduled refueling outage, the Vir'ginia Electric and Power Company shall remove and inspect the insidre recirculation spray pumps and replace pump bearings if necessary.

A similar inspection shall be performed at least once every five years

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.g 8.9 Following completion of the radiation-thermal testing of the encapsulated saddle mate-ial used for shielding, the Virginia Electric and Power Company will evaluate the testing and provide the NRC with results of the evaluation.

8.10 Prior to the startup following the first regularly scheduled refueling outage, the. Virginia Electric and Power Company shall specify the details of the inspection program for guide thimble tubc wall wear.

8.11 Within sia months from issuance of this license, the Virginia Electric and Power Company shall submit a design for the backup overcurrent protection system by containment electrical penetrations for m r review and approval:

The backup system shall be ir. stalled and opere.t al prior to the startup following the first regularly scheduled refueling outage.

8.12 Within five years from the date of issuance of the operating license, the Virginia Electric and Power Company shall demonstrate that examination techniques allow reliable detection and evaluation of individual nozzle clad cracks should they grow larger than the acceptance standards contained in Section XI cf the American Society of Mechnical Engineers Boiler and Pressure Vessel Code.

8.13 for the conducting of the low power test program only, the licenree h'as been granted an exception from the requirements of those Technical Speci-fications identified in Table 6.1 of our Safety Evaluation Report dated, July 2,1980, related to the Special Low Power Test Program.

Amendment No. 1 NORTH ANNA - UNIT 2 8-3

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JULY 2, 1980 v

r SAFETY EVALUATION REPORT C

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0FFICE OF NUCLEAR REACTOR REGULATION O. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF i

l VIRGINIA ELECTP.IC AND POWER COMPANY SPECIAL LOW POWER TEST PROGRAM FOR i

i NORTH ANNA POWER STATION, UNIT N0. 2 p.

DOCKET NO. 50-339 l

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TABLE OF CONTENTS Page 1.0 ' SP EC I AL LOW POWER TEST PROGRAM.................................

1 1.1 Introduction..............................................

1 2.0 S A F ETY AS S ES S M E NT.............................................. 4 t

3.0 DELETION OF TEST 8 AND MODIFICATION OF TESTS 9A and 98......... 4 4.0 CO MB I N I N G T ES TS................................................

5 5.0 R EV I EW 0F T EST PROC EDUR ES......................................

6 6.0 EXCEPTIONS TO TECHNICAL SPECIFICATIONS......................... 8 6.1 Exceptions Involving Reactor Trip and Safety I n j ec t i o n ( S I ).......................................... 8 6.2 Other Exceptions to Technical Specifications.............

12 7.0 OP ERATI ONAL S AFETY CRITE RI A...................................

14 8.0 SAFETY EVALUATION.............................................

19 8.1 I n t ro d u c t i o n.............................................

19 8.2 Cooldown Transients......................................

24 8.3 Loss of Coolant Accidents (LOCA)......................... 27 3.4 Rod Wi thdrawal and Ejection..............................

28 8.4.1 Uncontrolle'd god Cluster Control Assembly Rod Wi thdrawal a t Power.........................

28 8.4.2 S_ ingle Rod Cluster Assembly Withdrawal -At Power...

30 8.4.3 Rup(ture of a Control Rod Dr'ive MechanismCRDM)...........................

31 8.S Dose Analysis............................................

32

9.0 ENVIRONMENTAL CONSIDERATION

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35

10.0 CONCLUSION

S...................................................

35 11.0 EMERGEt!CY OPFRATING PROCEDURES................................

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LIST OF TABLES p

t Page Table 6.1 EXCEPTIONS TO TECHNICAL SPECIFICATIONS FOR LOW POW ER T ES T P ROG RAM.....................................

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• b Table 7.1 OPERATIONAL S AFETY CRITERI A..............................

16 Table 8.1

SUMMARY

OF S AFETY EVALUATI ON.............................

21 Table 8.2 EVENTS B0UNDED BY FS AR RESULTS...........................

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h 1.0 SPECIAL LOW POWER TEST PROGRAM 1.1 Introriuction In Section I.G of Part II of Supplement No.10 to the Safety Evaluation Report for North Anna Power Station, Unit No. 2 we indicated that one of the activities proposed was to conduct a series of natural circulation tests at power levels up to five percent of normal full power.

The proposed test program was described in letters of February 8, 1980 and March 19, 1980, from Mr. Stallings to Mr. Varga.

The low power test program proposed by VEPC0 consisted of nine tests, eight of which involve natural circulation in the reactor coolant system at low power conditions, but at normal, or nearly normal, operating pressures and temperatures.

The specific tests proposed by VEPC0 were:

1.

Natural circulation test; 2.

Natural circulation with a simulated loss of offsite power; 3.

Natural circulation with loss of pressurizer heaters; 4.

Effect of secondary side isolation on natural circulation; 5.

Natural circulation at reduced pressures; 6.

Cooldown capability of the charging and letdown system; 7.

Sinulated loss of all onsite and offsite ac power; I

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Establishment of natural circulation from stagnant conditions;.

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Forced circulation cooldown (Part A) and boron mixing and cooldown (Part 8)

The proposed low power test program for VEPC0 was reviewed by the staff using the following five criteria:

1.

The tests should provide meaningful technical information beyond th'at obtained in the normal startup test program.

2.

The tests should provide supplemental operator training.

3.

The tests should not pose an undue risk to the public.

4.

The risk of damage to the nuclear plant during the test program should be low.

5.

The radiation levels that will exist after the low power test program is completed (including that from crud deposits)'must not preclude imple-mentation of requirements steming from the NRR Lessons Learned Task Force, Kemeny Comission, Rogovin Commission or Task Action Plan.

In a letter to the staff dated April 29, 1980, Westinghouse expressed concern with the conduct of two of the proposed. tests _(Test No._8 " Establishment of natural r

circulation from stagnant conditions" and Test 9B " Boron mixing and cooldown")

at plants other than Sequoyah. The reasons for their concern were: (1) special conditions required to conduct the tests and (2) little, benefit is to be derived from repeating the test since plant behavior should not be plant specific, whereas the difficulty of performing the test remains the same.

a By letter dated June 5,1980, VEPC0 requested deletion of Tests 8, 9A and

98. Subsequently, Test 9A was incorporated into Test 4.

VEPC0 also stated that in lieu of performing test 98 during the low power test prograir they would perform a similar test using decay heat insteail of performing it with the reactor critical. This test would be performed in conjunction with a planned test to demonstrate cold shutdown.

Use of decay heat eliminates many of the special conditions required for test 98, thus reducing the risks associated with performing this test.

On June 13,1980, VEPC0 submitted test procedures that had been approved by their safety comittee for the seven remaining tests. These seven tests were combined in four procedures to take advantage of established initial conditions. On June 18,1980, VEPC0 submitted the safety analysis and technical specification exceptions necessary to conduct these tests.

They also requested an amendment to the operating license to reflect the technical specification exceptions and indicated that Westinghouse has reviewed-and approved the safety analysis and technical specification exceptions. On June 24,1980, VEPC0 submitted changes to the test procedures that had also been approved by ths safety committee.

The purpose of this safety evaluation is to present the results of the staff review of the 3r~oposed special low power test program since approval by the staff is necessary for the conduct of the program.

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~2.0 SAFETV ASSESSMENT Tests 1, 3 and 5 Ifsted in Section 1.1 (Natural circulation, Natural circulation tith loss of pressurizer A te.s, and Natural circulation at reduced pressure) have been combined and designated as ST-8; tests 2 and 7 (Natural circulation L

uith a simulated loss of offsite ac power and Simulated Loss of all onsite and Offsite ac Power) have been combined into a single test designated at ST-9.

Test 9A has been incorporated into Test 4 designated as Test ST-11 (Effect of steam generator secondary side isolation on natural circulation).

Test 6 (Cooldown capability of the charging and letdown systera) is designated as ST-6.

Sections 3.0, 4.0 and 5.0 of this evaluation address (1) VEPC0's request to dele'.e tests 8 and 9A and 98, (2) combining the tests, and (3) the test procedures.

Sections 6.0, 7.0, and 8.0 of this evaluation address, (1) exceptions to the technical specifications, (2) operational safety criteria and (3) safety evaluation.

3.0 DELETION OF TEST 8, AND MODIFICATION OF. JESTS 9A AND 9B The desirability of conducting test 8 " Establishment of natural circulation from stagnant conditions, test 9A " Forced circulation coeldown" and test 9B " Boron

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mixing and cooldown" has been' discussed with the NSSS vendor, Westinghouse, and with YEPCO. As a result of these dir.cossions, VEPC0 in a letter dated June 5,1980, has requested that these tests be deleted or modified from the special test program.

VEPC0 stated that there is a significantly higher risk associated with performance of tests 8 and 98 as compared with the other tests because of the special test conditions required. VEPC0 also stated that Westinghouse agrees with this concern.

Since the purpose of Test 9A was to provide calibration data for reactor power measurements l

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over a range of cold leg coolant temperatures it was to be conducted as a J

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By combining test 9A with test 4 sufficient data 1

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We have considered the VEPC0 request to delete tests 8 and 98 and have emcluded

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that test 8 can be deleted and a similar test to 98lmay be performed using decay heat

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near the end of the startup test progrant for Unit ho. 2 for the following reasons:

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(1) there is a greater risk invo'ved ir. operating the plant under the conditions -

described in the tests, (2) there appears to be littie benefit to be derived from conducting these tests at more than one plant.

(The plant response to this test should not be plant specific and Westinghou'se and TVA have agreed to make i

the data collected from Sequoyah available to other applicants for training f

purposes.), (3) the Sequoyah operators hale received special training in per-

'i forming these tests, thus minimizing the risk at Seouoyah, (4) since it will take approximately six months for these test results to be fed back into simulator training programs for other plants, the relative, schedules of the near term

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operating license applicants is considered insignificant, and (5) VEPC0 will conduct a test to demonstrate boron mixing and cooldown capability on natural

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l circulation (similar to test 98) at the end of its startup test program. At that l

time there will be sufficient decay heat to perform the test with the reactor sub-critical.

The same training benefits will be derived as if the test were --

performed as.part of the low power test program because the test procedure will be close to operating conditions and relieves the operator of maintaining the reactor critical during test.

4.0 COMBINING TESTS We have reviewed the VEPC0 proposal to combine tests and have concluded that combining the tests will not compromise the test objectives with regard to

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training. Each of the first seven tests and test 9A originally p oposed are addressed discreetly in the four combined tests.

The principle reasons for

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combining the tests are to take a vantage of established initial conditions (e.g., reactor coolant pumps tripped and main feedwater isolated).

The changes will eliminate the time that would have been required to re-establish the initial conditions and could reduce the chance for operator error by not having to restart each test all over again. Thes,e changes will not affect the m

overall test results.

5.0 REVIFW 0F THE TEST PROCEDURES Wes;nghouse has reviewed the revised, combined test procedures and provided comments which VEPC0 has incorporat<.d.

The staff his reviewed the test procedures using the following criteria:

1.

The tests should provide meaningful te:hnical information beyond that obtair ed in the normal star tup test program.

2.

The tests should provide supplemental operator training.

2 3.

The tests should not pose an undue risk to the health and safety of the public.

4.

The risk of damage of the facility during the test program should be low.-

5.

The radiation levels that will exist after the low power test program is completed (including that from crud deposits) must not preclude implementation of. requirements from the NRR Lessons Learned Task Force, Kemony Comission, Rogovin Comission or Task Action Plan.

We have reviewed the procedures for the low power tests and conclud that they are acceptable based on the above criteria. However, the simulated i

loss of onsite and offsite ac power (portion of ST-9) does not fully meet criteria 1 and 2.

This test will provide information on decay heat

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removal with the steam driven auxiliary feedwater pump but using reactor power in lieu of decay heat. The auxiliary feedwater system configuration for this test will not be the same as the configuration which would exist in the event of a real L

loss of all ac power.

The normal lineup of the auxiliary feedwater system at North Anna Unit No. 2 consists of two motor driven and one steam turbine driven auxiliary feedwater pumps each providing auxiliary feedwater to one of the three steam generators.

In the event of loss of both onsite and offsite ac power, only the steam turbine driven pump would be available and consequently only one steam generator would receive auxiliary feedwater.

There is some concern that flow maldistribution in the core may occur and could result in power anomalies when bl$e reactor is used as the heat source.

Consequently, VEPC0 would prefer and we agree, not to conduct the test with only one steam generator renoving heat while simulating decay heat with reactor power.

The test procedure specifies that operators will proceed to the auxiliary feedwater pumphouse and using sou'nd power telephones, manually realign the auxiliary feedwater system to distribute the feedwater to all three steam generators and will manually control feedwater addition to each steam generator.

The operators in the control room will monitor steam generator levels and give instructions to the operators 'in the auxiliary feedwater pumphouse.

Although this procedure does not simulate an actual loss of all ac power, it will provide (1) some plant information on the capabilities of the auxiliary feedwater system, (2) operator experience in manually. throttling flow and (3) experience in training the operators to coordinate critical system realignments and control at remote locations of the plant.

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Based on our reStew of the test procedures, we conclude that the special low power i

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test program can be safely conducted as proposed at North Anna Power Station Unit Nd. 2.

We will witness selected portions of the special test as necessary to ensure' that the safety precautions and acceptance criteria are met.

7 6.0 EXCEPTIONS TO TECHNICAL SPECIFICATIONS Exceptions to a number of technical specification requirements for North Anna Unit No. 2 will-be made during the low power test program. Some exceptions are required because of operation with a critical reactor under conditions outside of the range allowed in the Technical Specifications (e.g. natural circulation conditions and Iow coolant temperatures and pressure).

Other' exceptions are required because some systems normally required to be operable will be rendered temporarily inoperable as part of the test program (e.g. simulated loss of offsite power and simulated loss of all ac power.)

The exceptions required are listed in Table 6.1 for each of tests in the Special Lower Power Test-Program and are discussed below.

6.1 Exceptions Involvino Reactor Trio and Safety Injection (SI)

The exceptions involving reactor trip and safety injection (T.S. 2.2.1, 3.3.1, 3.3.2)are:

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a.

The' 0ver-Temperature and Over-Power AT trip functions are based on reactoFcoolant' system (RCS) hot and cold leg temperatures obtained from resistance temperature detectors (RTD's) which are located in bypass manifolds. Under natural circulation conditions, the very low expected flows in the bypass manifolds could result in spurious

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g readings and inadvertent trips. Therefore, these trip functions will be bypassed.

During the Special low Power Test Program, the protection functions of these automatic trips will be performed by operator actions based on limiting values of system parameters and automatic trip. at reduced neutron flux setpoints.

b.

The setpoint for reactor trip on steam generator low level, which

~

has a nonnal setting of 21% of the narrow range span will be reduced to 5% of the narrow range span.

This reduction will be made to preven,t inadvertent scrams for tests where it may be difficult to maintain the margin between the nonnal operating level and the nonnal setpoint.

This trip provides margins for maintaining the secondary side heat sink.

The low decay heat resulting from the low power levels during the test

,progrcm permits reduction in the level setpoint.

Automatic Mety injection will be blocked to prevent inadvertent c.

safety injection at the low coolant flow rates expected in the Manual safety injecti* n initiation will be operable.

tesi program.

o In addition, any safety injection signal will provide a reactor trip and control room indication /alann.

For tests,3 and 5, the low pressurizer pressure safety injection signal which would cause reactor trip, is blocked to allow operation at low pressures.

During this period of operation, the pressurizer power operated relief block valve 'will be closed to remove the major credible' source of inadvertent depressurization, r

d.

Secondary pressure trip protection will.be modified in several ways.

The safety injection signal resulting from high steam ifne flow in two main steam lines coincident with either low-low Tavg or low steam line pressure in two main steam lines will be modified 6.

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),

by (a) blocking the low-low Tavg input and (b) setting the high steam line. flow setpoint to zero flow (i.e., b! stable in tripped position).

Reactor trip and' main ' steam isolation valve (MSIV) isolation i

..~..... -..

will then be actuated by low steam line pressure signals in any two steam I

lines to protect against steam line breaks downsteam of the steam line check valves.

For test 4 the setpoint for low steam line pressure, sill be 4

reduced from the normal value of 600 psig to about 500 psig to permit operation at primary coolant temperatures down to about 550*F.

i The reactor trip resulting from the SI signal caused by high differential pressure between steam lines will be disaMed.

This signal gives i

the normal protection against large steam line ruptures upstream of the steam line check valves. Manual action based on the operational safety criteria will be used for such breaks.

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TABLE 6.1 EXCEPTIONS TO TECHNICAL SPECIFICATIONS FOR LOW POWER TEST PROGRAM

=1.

TEST TECHNICAL ~ SPECIFICATION 1

2 3 T~ '5 6

7 2.1.1 Core Safety Limits X

X X

X' X

X X

2.2.1 Various Reactor Trips Overtemperature AT X

.X X

X X

X Overpower AT X

X X

X X

X

~

Steam Generator Level X

X X

X X

X X

3.1.1.4 Moderator Temperature Coefficient X

3.1.1.5 Minimum Temperature for l

Criticality X

3.3.1 Various Reactor Trips Overtemperature AT X

X X

X X

X Overpower AT X

X X

X X

X Steam Generator Level X

X X

X X

X X

3.3.2 Safety Injection - All l

automatic functions X

X X

X X

X X

3.4.4 Pressurizer X

X X

l l

3.7.1.2 Auxiliary Feedwater X

X 3.10.3

'Special TesE Exce.otion ~

Physics Tests X

l X -- Exceptions Required i

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. '.i 6.2 Other Exceptions to Technical Specifications

.s a.

T.S. 2.1.-1, "Reacto r Core Safety !imits," gives limits to the average reactor coolant temperature in terms of reactor power, RCS.

pressure and number of operable loops.

For. the natural circulation tests, this specification cannot be met simply because no reactor-

.-.s-coolant (RC) pumps would be running. Ilowever, the intent of the

...-.----.---n

' specifications with respect to clad temperature limits will be met by the planned operational limits on core exit temperature, average coolant temperature, loop AT and subcooling mar lin.

b.

T.S. 3.1.1.4, " Moderator Temperature Coefficient," limits the moderator temperature coefficient of reactivity to zero or i

negative values.

During some tests, this coefficient may be slightly positive, liowever, the isothermal temperature coefficient is expected to be zero to slightly negative.

The effect of moderator temperature coefficient of reactivity was considered in the safety analysis.

c.. The minimum temperature for criticality is limited to 541 F by T.S.

3.1.1.5," Minimum Temperature for Criticality," and to 531 F by T.S.

3.10.3, "Special Test Exceptions.- Physics Tests.

During Test 4 it is expected that the average reactor coolant temperature will drop below these limits.

VEPC0 has stated that operation with the 0

average reactor coolant temperatures as low as 500 F is acceptable l

assuming that:

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1.

Control Bank D is inserted no deeper than 100 steps withdrawn and.

2.

The Power Range Neutron Flux low setpoint and Intermediate Range Neutron Flux reactor trip setpoint are reduced from 25% rated thermal power (RTP) to 7% RTP.

These restrictions reduce the consequences of transients involving individual rod withdrawl or rod bank withdrawal by limiting reactivity insertion rates from inadvertent individual rod withdrawal or rod bank withdrawal, providing sufficient shutdown margins, maintaining the moderate temperature coefficient at near zero values and limiting the maximum power during power excursions.

-The trip setpoint of 7% RTP is based on a coolant temperature in 0

the reactor vessel downcomer region of about 545 F.

Operation at a lower coolant temperature in the downcomer region results in a reduced output of the ex-core detectors for a given core power.

Hence, for operation at lower coolant temperatures, reactor trip would occur at powers higher than 7% RTP.

This effect was included in the safety analysis by using a conservative estimat'e 'of 1% ~

reduction-in the ex-core detector reading per OF.

Prior to the start of test 4, a special test will be run to assure that the ~

~~

actual decrease in the ex-core detector reading is less than that used in the safety analyses.

. - ~

T.S. 3.4.4 requires operability of the pressurizer.

In tests 2, 3, 5, and 7 the pressurizer heaters will either be turned off or rendered inoperable as the' result of loss of power.

This

.lo'de of operation is found acceptable because pressure control can l

still be maintained by use of the auxiliary spray and pressurizer level control.

T.S. 3.7.1 requires operability of at least three independent steam generator auxiliary feedwater pumps.

During two tests simulating loss of offsite power anJ total loss of a'c power, the auxniary feedwater system will be rendered partially inoperable (motor drives pumps). The [ow decay heat allows sufficient time ($ 1/2 hour) for plant perst qnel to return ac power and regain steam gen. rator level.

7.0 OPERATIONAL SAFETY CRITERIA As the result of a safety evaluation of the Low Power Test Program at North Anna Unit 2, VEPCO has specified a set of ope ational safety criteria for test conditions (see Table 7.1) and for conditions requiring prompt operator initiation of reactor trip or safety injection or termination of test.

The safety criteria include:

a.

lioits o.imaximum core exit temperature, maximum loop AT for any loop, maximum coolant average temperature, and minimum subcooling.

These limits and operator actions are provided to ensure adequate margin to the saturatt temper 5ture and adequate core cooling.

b.

limits on the minimum steam generator water level to provide a sufficient secondary side heat sink.

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limits on the minimum pressurizer water level for heater coverage c.

and pressure control.

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d.

limits on maximum insertion of control band D to ainimize consequences of inadvertent rod withdrawal and maintain a small moderator temperature coefficient while providing sufficient margin for shutdown.

limits on the Power Range Neutron Flux low setpoint and Intermediate e.

Range Neutron Flex reactor trip setpoint to limit maximum power to low values following possible uncontrolled power increases.

f.

limits on containment pressure and unplanned or unexplained changes in pressurizer water level and pressure.

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Guidelines for All Tests r!.

a)

Primary System Sub-cooling (T Margin)

> 20%

sat b)

Steam Generator Water Level

> 33% Narrow Range Span c) -Pre'ssurizer Water Level (1) With RCPs running

> 22% Span (2) Natural Circulation

> Value when RCPs tripped 0

'd)

Loop AT 1 65 F 0

e)

T,yg 1 580 F 0

-f)

Core Exit Temperature (highest) 1 610 F g)

Power Range Neutron Flux Low Setpoint and Intermediate Range Neutron Flux Reactor Trip Setpoints 1 7% RTP h)

Control Bank D 100 steps withdrawn or higher 2.

Reactor Trip and Test Temination must occur if any of the following con-ditions are met:

a)

Frimary System Sub-cooling (T 14argin) 1 15 F sat b)

Steam Generator Water Level

< 5% Narrow Range Span or equivalent Wide Range Level c)

NIS Power Range, 2 channels

> 10% RTP d)

Pressurizer Water Level

< 17% Span or an unexplained decrease of more than 5% not concurrent with a T,yg change 0

e)

Any Loop AT

> 65 F 0

. f)

T

> 580 F avg 0

g). Core Exit -Temperature (highest)

> 610 F 1)

Uncontrolled rod motion

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TABLE 7.1-(Continued) 3.

Safety Injection must be manually initiated if any of the following conditions are met:

0 a) ' Primary System Sub-cooling (T Margin) 5,10 F sat b) Steam Generator Water Level

< 0% Narrow Range Span or equivalent wide range level l

.c) Containment Pressure

> 17 psia L

d)

Pressurizer Water Level

< 10% Span or an unexplained l

decrease of more than 10% not concurrent with a T change.

avg e)

Pressurizer Pressure Decreases by 200 psi or more l

in an unplanned or unexplained l

manner.

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The staff has been concerned with uncertainties in the core AT and RCS subcooling measurements under natural circulation flow conditions.

These uncertainties dru the result of uncertainties in the core exit thermocouple and loop resistance temperature detector readings.

The fiorth Anna subcooling meters use input from four hot leg RTD's and twenty core exit thermocouples.

For North Anna the concerns involve principally (a) possibic stratification in the hot and cold leg piptng, (b) thermowell heat loss effects and (c) long time constants for the hot and cold leg temperature measurements since the i

resistance temperature detectors are inserted in thermowells which have good thermal contact with the RCS pipt.ng. Uncertainties in the temperature measurements are di/ficult to predict since local flow and temperature patMns under natural circulation conditions are unknown. 5EPC0hasstathdthatthe results of Test 1 will be reviewed to determine the behav.ior of tbr :e temperature detectors.

The objective of this review, which will be completed prior to the start of the remaining natural circulation tests, is to evaluate the adequacy of these measurements under natural conditions with respect to the specified core AT and RCS subcooling limits.

Since cach of the two florth Anna subcooling meters uses the h.ighest of two RID's and ten core exit thermocouples, the uncertainties associated with the hot leg RTD's should not compromise the safety of these tests.

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t 8.0 SAFETY EVALUATION B.1 Introduction VEPCO submitted the results of a study of the safety effects of the special conditions of the Low Power Test Program, including the exceptions to the technical specifications, which lead to operating conditions that are outside the bounds of conditions assumed in the Final Safety Analysis Report (FSAR).

The effects of these conditions on the Condition II, III, and IV events treated in Chapter 15 of the FSAR were evaluated.

Condition II events, at Norst, shall result in a reactor trip with the plant being capable of return to operation. Condition II events shall not propagate to cause a more serious Condition III or IV event and are not expected to result in fuel rod failure or reactor coolant system over-pressurization; Condition III events are very infrequent faults which will be accommodated with the failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude immediate resumption of operation.

For infrequent. incidents, the plant should'be dEiiigned to

~

limit the release of radioactive material to assure that doses to persons offsite 'are limited to values which are a small fraction of 10 CFR Part 100 guideline values. A Condition III event shal: not generate a Condition IV event or result in loss of function of the reactor coolant system or containment barriers; Condition IV events are limiting design bases accidents which are not expected to occur, but are postulated because their consequences include a potential for the release of significant amounts of radioactive r;aterial, t

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. h re System design for Condition IV events will prevent a fission product release to the environment which would result in an undue risk to the health and safety of the public in excess of Ifmits established in 10 CFR Part 100. A condition IV event is :not to cause a consequential loss of required function of. systems needed to mitigate the consequences of the accident, such as the emergency core cooling system and the containment.

The results of the analyses of Condition II, III and IV events are categorized in Table 8.1 according to the followfng' evaluation bases.

i I

. ANALYSIS OF TEST RESULTS OF ANALYSIS Bounded by FSAR analysis results 1

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Reanalysis shows fuel clad integrity is

,i l-maintained 2

4 Operator action is required for protectir.,n 3

4-Probability of occurrence reduced by restrictions on operating conditions 4

Probability of occurrence reduced,by short-testing period only 5

_._.7 -.

Table 8.2 lists those events for which a qual'itative evaluation is sufficient

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to conclude that the consequences of the event for the low power ' test' program are bounded'bi~thi FSAR risultE"

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.SUPNARY OF SAFETY EVALUATION TRANSIENT TEST:

1 2

3 4

5 6

7

..RCCA Bank With., Subcritical 2, 4 i,4 2,4 2,4 2,4 1

2,4 RCCA Bank With., at power 4

4 4

4 4

1 4

RCCA Misalignment 1

1 1

1 1

1 1

Boron Dilution 1

1 1

1 1

1 1

Partial loss of Flow 1

1

-1 1

1 1

1 Start Inactive Loop 1

1 1

1 1

1 1

^

Loss of Load 1

1 1

1 1

1 1

Loss of feedwater 1

1 1

1 1

1 3

Loss Offsite Power 1

1 1

1 1

1 3

Excessive Feedwater 2

2 2

2 2

1 2

Excessive Load 2

2 2

2 2

1 2

RCS Depressurization 1

1 4

1 4

1 1

Steam Depressurization 1

1 1

1 1

1 1

Spurious Safety Injection 1

1 1

1 1

1 1

Small LOCA 3

3 3

3 3

3 3

Small Secondary Breaks 2,3 2,3 2,3 2,3 2,3 1

2,3 Single RCCA Withdrawal 4

4 4

4 4

1 4

Misloaded Fuel Assembly 1

1 1

1 1

1 1

Complete Loss of Flow 1

1 1

1 1

1 1

Waste Gas Decay Tank Brk.

1 1

1 1

1 1

1 Major LOCA 3

3 3

3 3

3 3

Major Secondary Break 2,3 2,3 2,3 2,3 2,3 1

2,3 S/G Tube Rupture 1

1 1

1 1

1 1

RCP Locked Rotor 1

1 1

1 1

1 1

Fuel Handling 1

.)

1 1

1 1

1 Ruptured CRDM 3,5 3,5 3,5 3,5 3,5 1

3,5

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r TABLE 8.2 EVENTS BdVNDED BY'FSAR RESULYS

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se el EVENT _.

REASON WHY CONSEQUENCf3 BOUNDED BY FSAR RCCA. Misalignment Decrease in power caused by dropped rod cluster control assembly (RCCA). No increase in probability or consequences caused by test condition.

i Uncontrolled Boron Dilution Low setpoint for neutron flux scram (7%)

Control rods not inserted to insertion limit Constant operator monitoring during tests.

Partial Loss of Coolant low power level NoW Startup of Inactive Reactor Small moderator reactivity coefficf rnts.

Low Coolant Loop power level during test. Low setpoint for neutron flux scram.

Loss of Offsite Power to Low power level.

Trip on low-low steam generator Station Auxiliaries (Station blackout) water level.

Low decay heat.

Loss of Normal feedwater Low power level.

Trip on low-low steam generator water level.

Low decay heat.

Loss of Load and/or Yurbine trip Low power level. Turbine not operating 4

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• 9 EVENT

' REASON Wily CONSE00ENCES B0UNDED BY FSAR Excessive Load Inciease Turbine not operating. Load control limited Incident to single steam dump valve or relief valves.

JS urious Operation of Actuation of safety injection by any source m-Safetylnjection System

'except. manual action disabled during tests.

Accidental Depressurization For FSAR analysis where transient starts at

_0f Main Steam System hot shutdown with worst RCCA stuck out of s

core, safety injection prevents return to criticality.

For tests, reactor remains subcritical down to room temperature without safety injection.

M sloaded Fuel Assembly low power level Complete Loss of Flow Low power level WasteGasDecaylankRupture low fission product inventory Single Reactor Coolant Pump Locked rates Low poutor' level

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Fuel llandling Acciden h Accident independent of low power test program conditions or low fission n_._

product inventtery.

Rod withdrawal from Test procedures require that RC pumps will be subcrit_ical condition operating before rods withdrawn from subcritical condition.

_ Steam Generator Tube Rupture Low radioactivity level in primary and secondary systems, h

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_. 8.2 Cooldo n Transients-a-

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Cooldown transients considered in the FSAR included (a) excessive increase in load, (b) accidental depressurizatien of the main steem system, (c) small secondary system breaks, (d) excessive heat removal due to feedwater system malfunctions, and (e) major secondary system breaks. With the exception of some types of breaks in the main steam lines, the consequences of these transients during the test ' program should be minor because of the low power levels, low neutron flux trip and small moderator temperature coefficient of reactivity.

The turbine will not be used during the tests and load control will be limited to operation of a single steam dump valve or the relief valves. A load increase or small steam pipe break equivalent to the opening of a single steam pressure relief valve, dump valve or safety valve would cause a small

(-4% RTP), increase ;n reactor power, assuming the bounding negative value of the moderator temperature coefficient for the beginning of life (Cycle 1 ).

Consequences of the event, Excessive lleat. Removal Due to Feedwater System Malfunctinns,are reduced during the test program because the main feedwater centrol valves will not be used when the reactor is at power or critical.

With flow restricted to the main feedwater bypass valve or auxiliary feedwater system, the maximum flow ru is about 15% of normal flow.

Analysis of the above types of transients indicates that the departure from f

nucleate boiling (DNB) criterion of the fSAR is met.

e 4

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' Automatic reactor trip and steam line isolation following postulated large steam line breaks which result in uniform depressurization of all loops is provided by low pressure signals from any i.c steam lines (normally requires coincident high steamline flow signal setpoint set to zero flow).

An example

' s a double-ended break in a main steamline outside of the check and isola i

valves. An analysis of this event indicated reactor trip about 15 seconds

~

after the break and no powe excursion.

The reactor remained subcritical after the trip.

For large steam line breaks upstream of one of the steamline check valves, automatic reactor trip normally would result from the SI signal on high differential pressure between steam lines.

However, this signal will be disabled for all. tests.

Isolation of the broken line for this case is provided by the non-return (floating disc type) valves which require no initiating signal. - Reactor trip would be required by operator action based on the operational safety criteria discussed previously.

Reactor trip could also occur at the Power Range Neutron Flue low setpoint.

However, since the nuclear instrumentation syst (NIS) detectors are not completely

~

qualified for steamline break conditions, this flux trip might be delayed or pre-

-vented.

An analysis of this event, usuming trip on the neutron flux slgnal, was_.

made for an initial power of 1% RTP, one steam generator isolated and a double-ended

~

break' upstream of the steam venturi.

The results indicated a reactor trip at i

about 104 seconds into the transient with a maximum core heat flux of about 5% of the full power value.

Transients for which credit was not taken for the neutron flux trip were not analyzed.

Since the Evaluation of such transients based upon calculations could lead to fuel d' mage, VEPC0 provided a conservative a

estimate of the two-hour' dose at the site boundary to bound the consequences of this event.

The source term inside containment, obtained using the conservative

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assumptions discussed in Section 8.5 was corrected for the reduction in dose due to containment. The results of this analysis show that the calculated two-

. hour siAe boundary thyroid ' dose would be 9.2 rem.

u.

For steam line breaks outside of containment, automatic protection is available and' the accident is bounded by the FSAR results because of the low fission products inventory and is acceptable to the Staff.

For steam line breaks inside of containment, corrective operation actions are needed.

Close operator waervision during the tests and corrective actions based on the operational i

safety criteria should be sufficient to prevent significant clad damage.

In addition, the. bounding dose analysis performed for the postulated accident,'

which assumed 100% clad failure and other conservatisms, indicate that the offsite dose would be acceptably small.

The consequences of a main feedline rupture would be bounded in the cooldown 1

direction by those for a major break in a main steamline break.

Because of low operating power levels and decay heat, the heatup aspects of a feedline rupture are bounded by the FSAR results. --

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t 8.3 Loss of Coolant Accidents (LOCA)

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The' probability of occurrence of a greak in the reactor coolant pressure boundary 3

during the Low Power' Test Program is very low because of the short tiir.e period

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t; involved (f.e. about 2-3 weeks). As the result of the low power level and short

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i.

operating history, the magnitude of clad temperature transients for a LOCA event i

during tne Low Power Test Program would be significantly less than that for the FSAR I

. event because of low decay heat and stored energy in the fuel.

In addition, the off-s site dose consequences are reduced' because of the low fission product inventory.

The system inventory and normal charging flow can provide short-term cooling for very i

- small breaks.

VEPC0 has estimated that for a postulated 2 inch break, the time to uncover the core would be at least one hour if there were no safety injection.

c For major breaks in the reactor coolant pressure boundary, the applicant has stated that, even without automatic safety injection, there is sufficient cooling water available to prevent overheating of the fuel rod cladding in the short-term.

For a 1

large break the system inventory and cold leg accumulators will have removed sufficiert energy to have filled the reactor vessel to the bottom of the nozzles.

i After sysNm depressurization the water in the reactor vessel is sufficient to keep the core covered for more than one hour.

As &c result of the low initial power levels of the test program, the decay heat j

'i which must be removed by the ECCS and the corresponding fuel rod surface heat fluxes are very low.

For example, assuming reactor operation at 5% power for 1 year prior to the LOCA, the decay heat at one hour after the LOCA would be only 2.5 MW.

At this time the maximum fuel rod surface heat flux would be less than 500 Btw/hr-ft2 and the water needed to be added to the vessel to match boiloff would be about 20 gpm.

i Because of the limited core operating history prior to and during the Special Low Power }

Test Program, the actual-decay heat load and corresponding surface heat fluxes and b

b coolant in makeup-requirements should be much less than the above values.

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-, The staff conclud'es that the above times are sufficient for thd ' operator to take manual action to initiate safety injection and align the system for long-term cooling <

8.4 Rod Withdrawal and Eiection 8.4.1 Uncontrolled Rod Cluster Control Assembly Rod Withdrawal at Power

- Analyses of uncontrolled rod withdrawal were perfonned assuming natural circulation, starting power of 1% and 5% of full power, and with all steam isolation valves open or two'of those closed. A range of reactivity insertion rates up to the maximum for two banks moving was assumed for cases with all steam

~

lines open, and up to the maximum for one bank moving for the cases with steam lines isolated.

Both maximum and minimum bounds on reactivity coefficients were investigated. Reactor trip was initiated at 10% nuclear power.

These assumptions conservatively bound the test conditions.

The analyses performed show that the rod bank withdrawal at power is a mild transient.

Because of the absence of the full complement of normal reactor trips, difficulty of calculating core hydraulic' behavior under test conditions, and the paucity of DNB data in the low ficw-high pressure regime of the tests, the potential for DNB has not been precluded in the applicant's analysit.

On the basis of the small amount of data and extrapolation of other data, the applicant concludes that DNB is not expected for any rod withdrawal event.

We have reviewed the data presented by Westinghouse and additional data by Babcock and_Wilcox and data from Bowring.

Based on our review of the data we conclude _that, at the. low flow rates associated with natural circulation, the critical heat flux will be caused by an annular film dryout rather than by a disturbance in a bubbly surface layer, as is usually the case with DNB.

In addition, we conclude that, at the low flow rates associated with natural.

circulation, annular film dryout will not occur until the fluid quality i

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reaches the 80% to 100% range.

It appears very unlikely that the fluid'

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quality would' approach this range for any of the rod withdrawal events.

~

Assuming that DNB occurs, however, VEPC0 has performed analyses of

~

the clad temperature for the RCCA bank withdrawal at power.

The high power range neutron flux trip setpoint is 7% for the test program.

To allow for calorimetric errors and normal system errors a trip setpoint is assumed to occur at 10% power.

For the worst case, which assumes a low initial downcomer coolant temperature, a trip was assumed to occur at 20% power.

The analyses show that the peak clad temperature would be well below 1800 F.

In fact, the peak clad temperature would be expected to be approximately 1200 F.

We agree that these results indicate a clad temperature excursion resulting in fuel damage is not likely-to occur, even if DNB is assumed.

In addition, the bounding dose analyses performed fo'r a hypothetical accident involvir.g 100% -lad failure and other conservatisms indicate that the offsite

doses would be accel tably small.

These analyses therefore include three levels of conservatism and the results are acceptable.

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.8.4.2_ Single Rod _ Cluster Control Assembly Withdrawal at Power el.

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.y This accident was.not analyzed by the licensee. Although the FSAR analysis is

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-not bounding for the test condition of natural circulation, the low probability of this accident, and the extra surveillance of the operator for uncontrolled control rod motion, power, and hot leg temperature are considered sufficient to eliminate the need for' consideration of the consequences of this accident.

In addition, the bounding dose analyses performed for a hypothetical accident involving 100% clad failure and other conservatisms indicate that.the calculated offsite doses would be acceptably small even if such an unlikely event were to occur.

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8.4.3 Rupture of-a Control Rod Drive Mechanism (CRDM)

-Limitation of_ operation of the reactor with control rod withdrawn (Bank D

i

- only inserted, to 100 steps withdrawn) make an ejected rod worth less than the delayed neutron fraction, which would result in a transient which is relatively mild compared to those analyzed in the FSAR. _We agree with the licensee's con-clusion that the consequences are not considered severe enough to warrant analysis of the transient.--

In addition, the bounding dose analyses performed for a hypothetical accident

. involving 100% clad failure ar.1 other conservatisms indicate that the off-site doses would be acceptably small.

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_8.5 Dose' Analysis.

VEPC0 presented the.results of' calculations of the two hour site k

' boundary doses 'resulting from a hypothetical accident during the 1.ow Power Test Program which would bound the consequences of Condition II type transients analyzed'in"the.FSAR.

The analysis was based on an accident with coincident loss' of condenser vacuum which did not involve a break in the primary coolant pressure boundary.

The assumptions made in the analysis include:

v 139 Mwt (5% power) 1.0 micro curie per gram dose-equivalent I-131 RCS activity (technical specification limit) 500 gallons per day (gpd) steam generator leak in each SG (technical specifi-cation limit) 100% clad damage and gap activity release 10% iodine / noble gas in gap space 100 DF in steam generators 500 iodine spike factor over steady state 509,000 lb. atmospheric steam dump over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.7 x 10-3 sec/m3 x/Q percentile value The results of the analysis show that the 'tWb hour site boundary doses

.would be 5 rem thyroid, 0.9_ rem total body and 0.4 rem to the skin.

The staff did not make independent calculations of the dose values because it believes VEPCO's calculated doses are conservative for the following reasons:

1) 100% of the fuel clad is assumed to fail.

'This assumption is conservative for the evaluation performed during a safety review.

Typical values for cladding failure are about 10 to 20 percent.

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Equilibrium radionuclide inventories for operation at 5% power were used to estimatelhe amount of' Ectivity in the core.

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This assumption would be conservative for the expected intermittent and shorter-tenn operation of the reactor prior to and during the North Anna low power tests.

3)

Maximum technical specification values for the primary coolant concentration of iodine plus an iodine spike as a result of the accident.

.This assumption is in addition to the already assumed source of 100%

cladding failure and therefore definitely maximizes the amount of iodine available for release or leakage to the secondary system.

4)

Condenser vacuum is lost.

This assumption is normally made for accidents occurring at 100% power.

Since the nuclear station is attached to the electrical grid and pre-sumably supplies a significant portion of the base load, a transient resulting in a turbine trip could cause the grid to become unstable l

with an increased potential for losing the electrical supply.

During the low power tests the North Anna Station will not be supplying any power to the grid.

Should the nuclear unit have a station transient, offsite l

power will probably continue as normal and condenser vacuum would not be lost.

5)

Maximum technical specification steam generator tube leakage is assumed.

Since there is always the possibility that even new tubes are defective, it is not possible to exclude steam generator tube leakage entirely.

However, past experience suggests that new steam generator tubes do not leak at the technical specification limit.

Therefore, a 1 9Allon per

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minute (gpm) leak rate would be conservative for.the new steam generators.

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6). Meteorology-is conservative.

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The value for the short term diffusion coefficient (X/Q=1.7x10 sec/m )

-is larger than the value used by the staff (X/Q=4.2x10-4 sec/m3 - Safety Evaluation Report value) for the consequence estimates contained in the staff safety evaluation. report. This adds conservatism to the calculation of the dose estimates.

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9.0 ENVIRONMENTAL CONSIDERATION

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' We have determined that.the amendment does not authorize a change in effluent g

i types, total amounts or an. increase'in design power level of 2900 MWt. The test i

program will not; result in any environmental impacts other than those evaluated in the Staff's Final, Environmental Statement since the test program is encempassed by the overall activity evaluated in the Final Environmental Statement.

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10.0 CONCLUSION

S The Low Power Test Program for North Anna Unit 2 involves seven tests at low power levels conducted over a short period of time and with a very low fission product inventory.

i On the' basis of the above considerations, the proposed operational safety criteria and the safety evaluations which include the effects of the exceptions to the_ Technical Specifical. ions and operation under natural circulation conditions, the staff concludes that the Low Power Test Program will not result in undue riskuto public health and safety and is acceptable.

Therefore, we have concluded based on the considerations discussed above, that:

(1) it does not involve a significant hazards consideration, (2) there is

-reasonable assurance that the health and safety of the public will not be endangered by i

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Loperationiinf the proposed manner, and (3) such activities will be i

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l Lissuance'of this amendment will not be inimical to the common defense land secuEity or to the he'lth and-safsty of the public. Also, we.

reaffirin our conclusions 'as 'otherwise stated in our Safety Evaluation f and its Supplements.

11.0 -EMERGENCY OPERATING PROCEDURES-1 In addition to our requirement that the special low power test program be approved' prior to operation above zero power, we stated in Section 1.C.1 of Part II of Supplement No.10 to the North Anna Power Station, Unit No. 2

Safety Evaluation Report that~ VEPC0 must also revise to our satisfaction emergency operating procedures related to the small break loss-of-coolant accident and' inadequate core cooling.

IIn a' letter dated fiay 30,1980, VEPC0 provided copies of emergency orocedures that had been revised to reflect the analysis of small break loss-of-coolant accidents and inadequate core cooling in accordance with license condition 2D(G)a.:andTaskActionP1an(NUREG-0660)itemI.C.I. The emergency procedures

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submitted by VEPC0 have been rev_tewe,d by tbe NSSSjuppl{er, Westinghouse, _ '. _ _ _ _.

. Electric Corporation, and changes recommended by Westinghouse have been

' incorporated in compliance with Task Action Plan item I.C.7(a).

The staff has reviewed VEPC0's emergency procedures and has reconmended some changes to VEPCO. VEPC0 has made the recommended changes and is continuing with safety committee approval of the changes and operator training.

The staff will-observe a simulation of the emergency conditions conducted by North Anna Unit No. 2 personnel and a walk-through of at least one emergency procedure in the North Anna Unit No. 2' control room. We have concluded that the emergency procedures are adequate to support operation

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up.to 5% power for' training duriNg low power testing.

Prior to coeration P^ '

above 5% power we will eva10 ate the results 'of the procedure walk-throughs and.~ ensure that.the licensee has made_any necessary procedural changes.

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UNITED STATES NdCLEAR REGULATORY COMMISSION DOCKET N0. 50-339 VIRGINIA ELECTRIC AND POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO LICENSE NPF-7 The U. S. Nuclear Regulatory Comission (the Comission) has issued Amendment No. I to Facility License NPF-7, issued to the Virginia Electric and Power Company (licensee), which added Technical Specification 8.13 to Appendix A of the Technical Specifications for operation of the North Anna Power Station, Unit No. 2 (the facility) located in Loui,a County, Virginia. The amendment is effective as of its date of issuance.

The amendment permits the licensee to conduct the special low power test progra n as presented in our Safety Evaluation,, dated July 2,1980.

The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations. The Comiss. ion has made appropriate findings as required by the Act and the Comission's regulations in 10 CFR Chapter I, which are set forth in the license amendment. The activity authorized by the amendment is encompassed by the overall action involving the proposed issuance of an operating license for which prior public notice was issued in the Federal Register on May 25, 1973 (38 F.R. 13772).

The Comission has detennined that the issuance of this amendment will not result in any environmental impacts other than those evaluated in the Final Environmental Statement since the activit,y authorized by the amendment is encompassed by the overall action evaluated in the Final Environmental Statement.

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For further details with respect to this action, see (1) the application

-for amendment, dated June 18,1980,(2) Amendment No. 1 to NDF-7, and (3) the Commission's related Safety Evaluation concerning a Special Low Power Test Progr.am and Emergency Operating Procedures.

All of these items. are available for public inspection at the Commission's Public Document Room,1717 H Street, N. W., Washington, D. C.

and at the local public document rooms in the Alderman Library, Manuscripts Department, University of Virginia, Charlottesville, Virginia 22901 and at the Office of the Board of Supervisors, Louisa County Courthouse, P. O. Box 27, Louisa, Virginia 23093. A copy of items 2 and 3 may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C.

20555, Attention: Director, Division of Licensing, Office of Nuclear Reactor Regulation.

Dated at Bethesda, Maryland, this 3rd day of July,1980.

FOR THE NUCLEAL RECULATORY COMMISSION 1,%AdMl B. J. Yot gblood, Chief

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Licensing Branch No. 1

Division of Licensing

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NUCLEAR RECULATORY COMMISSION f

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July 25, 1980 p

Docket No. 50-339

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MEliORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing.

Division of Licensing i

FROM:

Voss -A. lioore, Acting Deputy Director, Division of Iluman Factors Safety

SUBJECT:

TRANSMITTAL OF fl0RTH ANNA 2 SER INPUT FOR SPECIAL LOW POWER TEST PROGRAM The enclosed safety evaluation for Task I.G.1 is submitted for incorporation in the North Anna Unit 2 SER.

We have concluded that VEPCo has met all requirements of the staff position for this item.

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w Voss A. l1oore, Acting Deputy Director Division of lluman Factors Sa fety

Enclosure:

-As stated I

cc w/ enclosure:

D. Eisenhut S. llanauer A. DromeriplL, D. Zicmannt

g D. Fischer N. Anderson

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,e SAFETY EVALUATION REPORT NORTH ANNA UNIT 2 TASK I.G.1 TASK I.G.1 POSITION The TMI Task Action Plan states tnat applicants for operating licenses will perform a set of low power tests to increase the capability of shift crews and ensure training in plant evolutions and off-normal events. Near-term operating license facilities will be required to develop and implement

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intensified exercises during the low power testing programs, This may

' involve the repetition of startup tests on different shifts for training purposes.

DISCUSSION AND CONCLUSIONS By letters of April 2 and April 21, 1980, Virginia Electric and Power Company (VEPCo) submitted draft procedures for conducting nine tests similar to those proposed by TVA for Sequoyah Nuclear Plant Unit No.1.

The April 21 letter also transmitted the safety evaluation r.od training program for the tests, In a letter dated April 23, 1980, Westinghouse stated concerns with repeating two of the proposed tests, Startup from Stagnant Conditions and Boron Mixing and Cooldown, at plants other than Sequoyah.

By letter dated June 5,1980, VEPCo requested deletion of these two tests from its low power test program, however, a test similar to one of the tests would be performed at the end of the startup test program using decay heat.

On June 13,1980, VEPCo submitted test procedures that had been approved by their safety committee for the seven remaining tests (these seven tests were combined into four procedures).

On June 18, 1980, VEPCo submitted the safety analysis and technical specification exemption requests recessary to conduct.thp tests, On June 24, 1980, VEPCo submitted changes to the test procedures that had also been approved by the North Anna safety committee.

The special law power test program was reviewed and approval to conduct the tests was granted in Amendment 1 to the low power operating license.

The special low power test program, as approved' by the NRC, was conducted at North Anna Unit 2 starting on July 3,1980.

NRC staff representatives were present to observe each test the first time it was performed.

In addition an NRC Resident Inspector was present for all tests.

Special Tests 2-ST-6 (Cooldown Capability of the CVCS) and 2-ST-ll (Effect of Steam Generator Secondary Side Isolation on Natural Circulation) were each performed once.

Test 2-ST-8 (Natural Circulation Verification).

was performed five times and 2-ST-9 (Natural Circulation with Loss of Offsite Power and Loss of Offsite and Onsite AC Power) was performed four times. We have c.oncluded that VEPCo satisfied the requirement for operator training by having every licensed operator participate in at least one test a_nd observe two or more,

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b 2-o VEPCo has committed to submit a test report on the special low power test program which will provide a detailed evaluation of the test results. This evaluation is based on the preliminary test results and the observations made by the staff during conduct of the tests.

Sufficient preliminary data has been reviewed which together with direct observation of the tests makes it possible to determine if the staff requirements have been fully met.

A summary of the first. performance of each test follows.

.s 1.

Test ST-6 Cooldown Capability of the CVCS This test was conducted at zero power with one reactor coolant pump running and all three steam generators isolated.

Heat was removed from the primary system using the charging and letdown systems.

The maximum charging and letdown flow rate of approximately 120 gpm resulted in a Primary Ccolant System temperature decrease of about 2 F per hour.

The minimum flow rate of approximately 40 gpm resulted in a temperature increase of approximately 2.5 F per hour.

The plant responded as expected during this test and all test abjectives were met.

2.

Test ST-8 Natural Circulation Verificat(gn With the reactor at 3% power and heat being removed with all three steam generators, all three reactor coolant pumps were tripped and natural circulation was established.

Primary system pressure increased to 2310 psig where one PORV lifted.

The PORV rescated and the pressure was. controlled using auxiliary sprays.

The primary coolant

• system stabilized at approximately 30 AT.

Following this portion of the test, a core flux map was run in natural circulation for comparison with the zero power flux nap taken before the test.

In stable natural circulation the average core exit thermocouple reading was 581 *F.

The average of the three hot leg RTDs was 580.5*F.

A map of core exit thermocouple readings taken in natural circulation indicated that the core flow distribution did not change.

Location of the high and low thermocouple readings stayed the same.

There was a 6*F difference in high and low readings in natural circulation and a 3"F difference with forced flow.

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The second part of. Test ST-8 was to demonstrate that PCS saturation fL margin can be maintained without pressurizer heaters.

With natural circulation established, all pressurizer heaters were turned off.

Both auxiliary and main spray valves were closed to ensure that spraying would not influence pressure drop.

The depressurization rate over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period was approximately 38 psi per hour.

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temperature decreased at about 5 F per hour and PCS subcooling margin decreased by about 6 F per hour.

The last part of Test ST-8 was to determine the effect'of decreased

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subcooling margin on natural circulation and the effects of charging and steam flows on saturation margin.

System pressure was decreased using pressurizer sprays to a subcooling margin of approximately 30*F.

Saturation margin could not be substantially increased by changing steam' dump rate alone, Charging flow or pressurizer heaters were needed to increase saturation margins.

The plant responded as expected during this test.

Lifting of the PORV was a normal system response with the specified test conditions.

3.

Test ST-9 Natural Circulation with Loss of Offsite Power and Simulated Loss of All Offsite and Onsite AC Power This test was conducted at 1% reactor power.

offsite power the following actions were take,To simulate loss of n:

A.

tiotor driven auxiliary feedwater control valves were closed.

B.

Steam dump controllers were placed in manual control.

C.

Pressurizer backup heater groups 2'and 5 and control heaters group 3 were tripped and locked out.

D, All three reactor coolant pumps and the operating main feedwater pump were simultaneously tripped.

Approximately eight minutes after initiation of the test, a PORY lifted at 2305 psig.

Pressure was controlled using atmospheric steam dumps.

The PORV rescated and the pressure remained below the PORV setpoint. Auxiliary feedwater was initiated normally and stable natural circulation was achieved with cold leg temperature of 547 F and AT of 22 F.

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n To simulate loss of all AC, both motor driven auxiliary feca pumps were turned off.

The feed pump house exhaust fans were also turned off.

This test deviated from a real loss of all AC in that the steam turbine driven auxiliary feed pump was aligned to feed all three steam generators, The normal alignment at fiorth Anna is to have each of the three auxiliary feed pumps feeding one steam generator.

If all AC were lost, only one steam generator would receive feedwater.

This realignment did not preclude meeting test objectives, Operators were sent to the auxiliary feed pump house to realign the feedwater system to feed all three steam generators and to control feed flow by manually adjusting control valve position at the direction of the control room operator, Steam generator levels were initially 40% to 45%.

Levels dropped to about 35% and stabilized.

By manually controlling flow, levels were maintained between 35% and 43% throughout the test, Stable natural circulation was established and maintained at approximately 22' AT.

The plant responded as expected during this test.

Lifting of the PORV was a normal system response with the specified test conditions.

4.

Test ST-ll Effect of Steam Generator Secondary Side Isolation on flatural Circulation This test was initiated by tripping all three reactor coolant pumps with the reactor maintained at 1% power. After stable natural circulation was achieved with a AT of approximately 21*F, the B steam generator was isolated.

Following isolation of the B steam generator, AT in the isolated loop slowly decreased to about 9 F in one hour.

AT in the other loops was at 27 and 25*,

'At~this' time it was suspected that some leakage was occurring around the feedwater isolation valve and steam isolation valve.

Operators were dispatched to tighten the feedwater bypass valve and the other isolation valves.

The test then continued approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at which time the isolated loop had a AT of 3 F and was stable.

The plant responded as expected during this test.

It is concluded that the special low power tests conducted at tiorth Anna satisfy all requirements of Item I.G.1 of the Task Action Plan.

This conclusion is based on '.he following:

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All licensed operators received adequate training during the program by participating in at least one test and observing at least two or more.

2.

Meaningful technical information was obtained on plant response to a variety of abnormal conditions.

3 At all times during the tests, the plant was under complete control and responded as predicted.

4.

Acceptance criteria for each test as specified in the test procedures were met.

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