ML19270H333
| ML19270H333 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/02/1979 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Eysymontt G, Guibert J, Thompson H NRC COMMISSION (OCM) |
| References | |
| FOIA-79-98 NUDOCS 7906260216 | |
| Download: ML19270H333 (12) | |
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Distributien w/enci-Central File D. Ret.
LWR 43 File D. Vassallo A. Orc =erick
- 0. Swansen, OELD
- 0. Darr M. Rusnorcok
."0TE TO: John Guibert, Technical Assistant to ComifdbqerKennedy Hugn Thompson, Technical Assistant tu Ccmissioner Bradford George Eysymontt, Technical Assistant td Corrnissioner Gilinsky George Sauter, Technical Assistant to'Comissioner Ahearne
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- 2. F. ' toss, Jr, Oe:uty Directer /livision of 3-0 ject "anagerent le 9 ave received a recuest frc-the accl ce tain esquire <'ents related to Accen:ic:gant recardi~ In execotion ::
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Enclosed is a draft SIR on :nese exerc*4 s 5.;c,': orth Anna 2.
'ie intend to issue a succlement wnich kn.es these exerations\\at One time an oceratina license is issued or ?; orth Anna Unit 2.
The exemotions recuested for "crth Anna Unit 2 related to accendices 3, H ', J cf 10 CFR 50 are similar to these granted for tne "c9uire.';uclear Station.
It is antici;:ated that Unit 2 will be ready f:r fuel leadinc by June 1979.
Mcwever, cur review cf flerth Anna Unit 2 is not cecolete, and it will be scre time before it is ccmoleted. Nonetheless, we would ap;reciate an ex;:editicus review of these exem::tions and ycur advisinc me by telechene (extensicn 27373) of any cc : ents you may have in this regard so that work can continue on the SER succlement.
Original SJgned By Roger S. Scyd D. F. Ross. Jr., Oecuty ;irecter
/M Ofvision of Project "ailagement m'
Office af.'iuclear Reac:cr.egulation Enclosure?
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VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT NO. 2 (OL)
DOCKET NUMBER 50-339 COMPLIANCE WITH APPENDICES G AND H, 10 CFR PART E0 MATERIALS ENGINEERING BRANCH MATERIALS INTEGRITY SECTION Comoliance with Accendices G and H, 10 CFR Part 50 The react 0r vessel for North Anna Unit No. 2 was manufactured by Rotterdam Dry Dock Comoany of Netherlands. The purchase order was issued on May 1, 1969. The vessel was. fabricated to 1968 Winter Addenda of the ASME Section III Soiler and Pressure Vessel Code. Since the ASME Code editions defined in 10 CFR 50.55a preceded the publication of Appendices G and H, 10 CFR Part 50, some of the fracture toughness tests for the ferritic materials in the primary coolant oressure boundary were not conducted to demonstrate explicit compliance with tre current requirements of Appendices G and H.
Virginia Electric and Power Company stated that the fracture toughness requirements of ?,ppendices G and H,10 CFR Part 50 were met for North Anna Unit No. 2 except for the specific requirements of Section IV.A.4 of Appendix G and Section II.C.2 of Appendix H.
Alternate methods for compliance with Appendices G and H,10 CFR Part 50 were proposed by Virginia Electric and Power Company and exemptions were requested from the identified requirements. VEPC0 also provided additional information in support of their methods of compliance with Appendix G.
We have concluded from our review of information submitted that exemptions to some of the specific requirements of Appendices G and H,10 CFR Part 50 are required, and we have determined that the identified exemptions are justified. The bases for justification are discussed in the subsequent paragraphs of this report.
Evaluation of Comoliance with Accendix G Based on our review of the applicant's submittal for compliance with tapendix G,10 CFR Part 50, we have determined that the requirements of Aapendix G have been met for North Anna Power Station, Unit No. 2 except
.for Section IV.A.4.
Section IV.A.4 of Accendix G recuires that a Charoy V-notch test program be conducted for the crimary coolant cressure bcuncary ferritic bolting exceeding one incn in diameter to demonstrate tnat the bolting material exhibits One minimum recuiremen:s of 25 mils iatarai exoansion and 2315 002 i
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45 ft.-lbs. impact energy at the lower of either the preload temperature or the lowest service temperature. The North Anna Unit No. 2 reactor vessel bolting material tests were performed in accordance with the ASME Code, 1968 Edition, including the Winter 1969 Addendum of Section III.
These codes have no lateral expansion measurement requirements and specify that an average impact energy of 30 ft.-lbs. be obtained at a temperature 60*F lower than either the hydrotest or the lowest service temperature.
The impact test results for North Anna Unit No. 2 indicate that the material impact energy exceeded all the ASME Code recuirements at a test temperature of 10*F.
The test results further show that at 10*F, which is more conservative than that required by Appendix G, the average imoact energy is approximately 42 ft.-lbs.
To provide assurance that the bolting material fracture toughness complies with the requirements of Appendix G, additional imcact testing was conducted at 32'F, 50*F, and 68'F.
The respective average impact energies at these three test temperatures was equal to or greater than 45 ft.-lbs.
We have reviewed the test data obtained to qualify A 540 Grade 324 bolting material used at North Anna Power Station Unit No. 2.
The test data consisted of Charpy V-notch energy values obtained at 10*F on 30 test specimens, representing two heats of steel. These heats had average Charpy impact energies of 41.0 and 43.0 ft.-lbs. at 10*F.
Scme tests also were conducted at 32 F, 50*F, and 68*F.
The average Charpy V-notch impact values were 44.7, 46.7 and 49.0 ft.-lbs. respectively.
We have also reviewed similar tests results for A 540 Grade B24 bolting material reported for several heats of steel in Electric Power Research Institute Report, EPRI NP-121, Volume II, Part One, April 1976. These data were reviewed to provide additional assurance that Charpy specimens having a minimum impact energy of 45 ft.-lbs. also had a minimum lateral expansion of 25 mils. Our review of these data indicated that soecimens with impact energies greater than 45 ft.-lbs. did have lateral expansions greater than 25 mils.
Based on our evaluation of the test data, we conclude that an exemption for the area of nonccmpliance of Appendix G is justified. Our conclusion is based on the following:
Appendix G requires the measurement of both lateral expansion and absorbed energy to provide additional assurance that the material has adecuate
' fracture tougnness. However, absorbed imoact energy and lateral expansion are very closely related criteria and orovide an almost identical indication of the material cuality and the toughness level. Consecuently, we have determined that the measurement of the absorbed energy, in accordance with the ASME Soiler and Pressure 7essel Code recuirements, is suf#icient :o 2
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t demonstrate acceptable fracture toughness properties. Added assurance of our conclusion is supported by our review of additional test data obtained from an EPRI research program conducted for similar bolting material. As indicated previously, some of the impact specimens tested at the ASME Code 10 F test temcerature had absorbed energies less than the 45 ft.-lbs.
required by Appendix G.
However, the 10*F test temperature specified by the ASME Code is more conservative than the test temperature required by Apcendix G.
To provide additional assurance that the bolting material has adequate fracture toughness some tests were conducted at higher temperatures representative of the Appendix G requirements. The results frem these tests indicate that the fracture toughness for the bolting material meet the Appendix G requirement of a5 ft.-lbs. impact energy at the Icwer of eitner the prelcad temperature or lowest service temcerature. The imoact tests performed according to the ASME Code requirement and the additional tests performed at higher temperatures are sufficient to indicate that the bolting materials were manufactured properly, are of acceptable quality and have adequate fracture tougnness to provide reasonable assurance that adequate safety margins aill be obtained and maintained during cperation as required by Appendix G.
We have evaluated the data presented in the FSAR and based on the results of our evaluation we have determined that sufficient information has been provided to demonstrate that the safety margins required by Apoendix G, 10 CFR Part 50 have been achieved.
Evaluation of Comoliance with Accendix H Based on our review of VEPCO's submittal for compliance with Appendix H, 10 CFR Part 50, we have determined that the requirements of Appendix H have been met for North Anna Unit No. 2, except for Section II.C.2.
Section II.C.2 of Appendix H was not complied with for Unit No. 2 to the
- extent that six of the eight specimen capsules are located in areas where the lead factor is less than one. The lead factor is the ratio of the neutron flux at the specimen capsule to the maximum neutron flux at the vessel inner surface. Consequently, the neutron flux received by the capsules will be less than that received by the inner surface of the reactor vessel. The purpose of the Appendix H limitation that the surveil-lance capsule lead factor be in the range frca one to three is to ensure that reduction in reactor vessel material tougnness resulting frcm neutron irradiation is monitored in advance ef actual vessel conditions and to minimize calculational uncertainties in extracolating the surveillance measurements frcm the specimens to the reactor vessel wall. As an alternative to maintaining the recuired lead factor at a fixed location
'/E?CO has suggested a schedule for rotation of tne scecimen caosules at different locations such that wnen a capsule is removed for testing it will have a lead factor greater than one.
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We have reviewed the alternate method proposed by VEFCC to achieve the re;uired lead factor and conclude that this alternative is equivalent to the Appendix H requirement and no inaccuracy will result from the capsule rotation. Adequate data are assured because the proposed capsule rotation will proside the required total accumulated neutron fluence without significant changes in temperature or neutron fluence rate relative to the vessel inner surface during the total time of irradiation.
Based on cur review and evaluation we conclude that an exemption from Section II.C.2 is justified because the alternate method proposed by VE?CO is equivalent to the Appendix H recuirem'ent and will provide adequate data to monitor the reduction in material fracture toughness and ensure that adequate safety margins are maintained during operation.
Our tecnnical evaluation has not identified any practical method by wnich tne existing North Anna Unit No. 2 reactor vessel can comply with the specific requirements of Section IV.A.4 of Appendix G and Secticn II.C.2 of Appendix H,10 CFR Part 50. Requiring compliance with tre identified specific requirements would delay the startup of the units due to the need to complete the following actions:
(1) retest the bolting materials to confirm compliance with Appendix G, and (2) relocate the installed material surveillance specimens.
Based on the foregoing, pursuant to 10 CFR Section 50.12, exemption to the specific requirements of Appendices G and H of 10 CFR Part 50 as discussed above is authori:ed by law and can be granted without endangering life or property or the common defense and security and is othemise in the public interest. We conclude that the public is served by not imposing certain provisions of Appendices G and H of 10 CFR Part 50 that have been deter-mined to be either impractical or would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Furthennore, we have determined that the granting of this exemption does not authori:e a change in effluent types or total amounts nor an increase in pcwer level and will not result in any significant environmental impact.
We have concluded that this exemption would be insignificant frem the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal, need not be prepared in connection with this action.
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Evaluation of Proposed Exemption from Apoendix J Requirements North Anna Power Station, Unit 2 In the Technical Specifications for North Anna Unit 2 the applicant describes its proposed leak testing procedure for the containment airlocks, and proposes an exemption from the associated requirements of Appendix J to 10 CFR Part 50. Based on our review, we find the crocesed leak testing procedures and the proposed exemotion to Appendix J acceptable. The rationale for our finding acceptable the applicant's proposed leak testing practices for the personnel airlocks and the proposed exemption from the associated requirements of Appendix J to 10 CFR 50, is discussed below.
Appendix J to 10 CFR 50 requires the containment personnel airlocks to be leak tested at six-month intervals and after each opening during such intervals (III.D.2). Appendix J further requires that the test be conducted at the peak calculated containment pressure related to the design basis accident; i.e., Pa, (III.B.2).
Considering that a full pressure airlock test is to be performed every six months, it is our judgment that testing airlocks within three days after each opening or after the initial opening in a series of openings, at Pa, will adequately demonstrate the continuing integrity of the airlock door seals such that the public health and safety will be 2315 006 P
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ensured. The effect on accident consequences of testing after each opening versus testing within three days of an opening is judged to be insignificant. Furthermore, if an airlock door seal is damaged, it will be manifested during testing at '
Pa.
This is an adequate demonstration of continuing airlock integrity for the period between the six-menth tests.
We find that leak testing an airlock in the manner described above is an acceptable alternative to the requirements of Appendix J.
Accordingly, the proposed exemption from the requirements of Acpendix J is acceptable.
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, Docket No. 50-338, page 1 Response to IE Bulletin 79-06A North Anna Power Station Unit No. I
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Actions taken or planned in response to each item of IE Bulletin 79-06A are as follows. Number sequence is the same as in the bulletin.
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A detailed review of this event, by the appropriate personnel has been 6.
completed. The station training group conducted this review with h:: --
Station Supervision and the Station Nuclear Safety and Operating E ;.;;.
Committee.
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Operational personnel were instructed in the specific concerns of item Ib in a briefing held on April 21, 1979.
Ic.
Station Supervision and Operations personnel have received a briefing on the Three Mile Island incident. This briefing was conducted by NRC personnel on April 21, 1979. Those individuals n'eeding this briefing who were not in attendance will receive this infor=ation as soon as possible.
2a. and b.
The pri=ary operator action required to prevent the formation of voids
,is to insure the proper initiation and continuing performance of the engineered safety features.
Present procedures require this verifica-tion.
Procedure changes to prevent premature or inappropriate shutdown of engineered safety features will be made as explained in our response to items 7a and b.
A procedure change to insure forced flow by reactor coolant pumps will be made as explained in our response to item 7c.
These procedure changes will be completed by May 4, 1979.
2c.
Procedural changes to provide additional guidance on enhancing core cooling in the event of void formation are under review at this time.
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3.
North Anna Unit No.1 uses pressurizer water level coincident with pressuri:er pressure for automatic initiation of safety injection. The low pressurizer level bistables for all rhree channels will be
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tripped, such that low pressuri:er pressure only will initiate safety injection.
During the performance of pressurizer pressure channel functional e
surveillance tests, all three level channel bistables will be returned 7.=
to normal.
In the event that one pressurizer pressure channel becemes inoperable, its associated level channel bistable will be returned to normal. This
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will provide a 1 of 2 low pressurizer pressure safety injection from the 2 remaining operable channels.
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A standing order has been issued requiring operators to manually initiate safety injection when the pressurizer pressure indication reaches the actuation setpoint whether or not the level indication has dropped to the actuation setpoint.
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7-4.
We have completed our review of containment isolation initiation design I"*""
and procedures and have determined that no changes are required.
5.
North Anna Unit No. I has automatic auxiliary feedvater initiation.
Per Technicial Specification 3/4.7.1.2, limiting conditions for opera-tion and surveillance requirements have been established to maintain operability of the system.
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Automatic initiation results from:
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Low-Low S/C level, 3.
Loss of Main Feedwater Pumps, 4.
6a.
We have reviewed the applicable procedures and have determined that
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no changes are needed for this item.
6b.
This item has been covered by a Standing Order.
The appropriate procedures will be revised by May 4, 1979.
7a. and b.
-The applicable procedures will be revised by May 4,1979, to prohibit overriding engineered safety features, unless continued operation of engineered safety features would result in unsafe conditions.
Specific-ally, emergency procedures will specify that if the high pressure
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injection system has been automatically actuated because of a low pressure conditions, it must remain in operation until either:
- 1) Both low pressure injection pumps are in operation and flowing for 20 minutes or longer, se a rate which would assure stable plant behavtor; or 55
- 2) The high pressure injection system has been in operation for 20 minutes and all hot and cold leg temperatures are at least 50
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degrees below the saturation temperature for the existing RCS pressure.
If 50 degrees subcooling cannot be maintained after high pressure injection cutoff, the high pressure injection should be reactivated. The degree of subcooling beyond 50 degrees F and the length of time high pressure injection is in operation shall be limited by the pressure / temperature considerations for reactor vessel integrity. Shutdown of the high pressure injection system prior to 20 minutes is permitted only when overpressurization of the reactor coolant system is eminent and provided the above listed margins to saturation temperature are maintained.
7c.
Operating procedures will be revised by May 4,1979 to specify that in E=
the event of high pressure injection initiation with reactor coolant r"
pumps (RCPs) operating,,at least two RCPs shall remain operating as 5 ---
long as the pumps are providing forced flow.
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A Standing Order has been issued which precautions against overreliance on pressurizer level indications and recommends examination of other
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plant parameters in assessing water inventory and plant conditions.
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Operator training will incorporate a review of this concern.
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8.
Periodic tests will be revised to address this concern. Additionally,
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administrative procedures for shift turnover already address this item raEi:
in that known unit conditions are reviewed. Maintenance Operating Procedures already address this concern.
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.~i 9a.
Two interlocks exist to prevent the transfer of radioactive gases when
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high radiation indication exists.
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- 1) Containment purge and exhaust is secured on high containment
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activity
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- 2) Containment vacuum pump operation is terminated on high activity in the Process Vent System.
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9b.
Transfer of potentially radioactive gases and liquids is prevented by the initiatien of Phase A containment isolation.
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9c.
Technical Specification 4.6.3.1.2.a provides for periodic testing of I
Phase A containment isolation. Procedures will be revised to insure that su f ficient liquid waste tank capacity is available prior to pumping the containment sump. Operating procedures will be revised to incorporate precautions in containment su=p pump operations.
10a. The concern is addressed by the Shift Supervisor in reviewing the
" Action Statement Status" log prior to releasing a piece of redundant equipment for testing or preventive maintenance. For corrective maintenance on engineered safety features equipment, the redundant equipment will be tested before removing from service the equipment needing maintenance. However, in cases where testing of the redundant equipment makes that equipment inoperable, it will not be tested.
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10b. Maintenance Operating Procedures already cover this item.
10c. With one exception, all Periodic Tests concerning ESF equipment require Shift Supervisor notification prior to commencement of and following
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completion of the test.
In the case of the exception, that test will be revised by May 4,1979, to include the notification requirement.
Our maintenance reporting system involves Shift Supervisor review prior
. co emintenance and following completion of maintenance.
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11.
Existing notification procedures will be revised to specify that the gi((~
NRC be notified within one hour of the time the reactor is not in a
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controlled or expected condition of operation. The procedure will
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include provisions for establishing and maintaining a continuous open
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channel of communication with the NRC.
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Docket No. 50-338, page 4 g;[
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12.
The existing equipment for removal of hydrogen from containment consists
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of two identical portable skid mounted hydrogen recombiners, two hydrogen analyzers, two purg<s blowers pad associated piping systems.
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Operating procedures presently exist to strip hydrogen from the primary coolant.
Additional operating modes and procedures for dealing with significant amounts of hydrogen gas are under review at this time.
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A May 3, 1979 g:.:
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Mr. James P O'Reilly, Director Serial No.
274A Office of Inspection and Enforcement P0/DLB:baw U. S. Nuclear Regulatory Comnission Docket Nos.:
50-338 Region II 50-339 101 Marietta Street, Suite 3100 License Nos.:
NPF-4 Atlanta, Georgia 30303 CPPR-78 g...
Subject:
TE Bulletin 79-06A North Anna Unit No. 2
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Dear Mr. O'Reilly:
_ to our letter of April 26, 1979 identified our actions taken
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or planned on North Anna Unit No.1 in response to IE Bulletin 79-06A. This is to infor :t you that all commitments made for North Anna Unit No.1 in response to the subject bulletin will also apply to North Anna Unit No. 2 and will be rr inplemented on Unit No. 2 prior to the issuance of the operating license.
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Vice Pre ident-Power Supp C
and P duction Operation U
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