ML20236A487
| ML20236A487 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/03/1989 |
| From: | Frantz S FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#189-8238 OLA-4, NUDOCS 8903170278 | |
| Download: ML20236A487 (51) | |
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD DOCKf bR i< A4,H .' - In the Matter of )
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FLORIDA POWER & LIGHT ) Docket Nos. 50-250 OLA - 4 COMPANY ) 50-251 OLA - 4
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(Turkey Point Plant, ) (P/T Limits) Units 3 and 4) ) LICENSEE'S RESPONSE TO PETITIONERS' AMENDED REQUEST FOR A HEARING A10 PETITION FOR LEAVE TO INTERVENE I. Introduction On February 17, 1989, the Center for Nuclear Responsibility, Inc. (" Center") and Joette Lorion (collectively ref9rred to as " Petitioners") filed " Petitioners' Amended Request for Hearing and Petition for Leave to Intervene." (" Amended Petition"). The Amended Petition sets forth proposed contentions pertaining to the issuance of certain amendments to the operating licenses for the Florida Power & Light Company's ("FPL" or
" Licensee") Turkey Point Units 3 and 4. The amendments, Nos. 134 and 128 respectively, issued on January 10, 1989, authorize the modification of the Turkey Point Technical Specifications to incorporate revised pressure / temperature (P/T) limit curves 1
applicable up to 20 Effective Full Power Years ("EFPY") of service 8903170278 890303 0 ADOCK 0500 PDR o 9 f
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life. 1/- The. Licensee hereby submits-its response to the Amended Petition. II. Standing of'the Petitioners The Licensee is not contesting the standing ~of the Petitioners to intervene in this proceeding. III. Petitioners' Contentions . A. Admissibility of Petitioners' Proposed Contention 1 Contention 1 states as follows: That the Nuclear Regulatory Commission Staff's Final Determination of No Significant Hazards Consideration issued on January 10, 1989 in support of license amendment nos. 134 and 128 issued to allow FPL to revise the pressure / temperature limits for Turkey Point nuclear units 3 and 4 respectively, is based on incomplete, faulty and non-conservative data, is'in error, and should be reviewed by this Atomic Safety and Licensing Board in order to protect the public health and safety from a loss of pressure vessel integrity and subsequent meltdown. The Licensee objects to this contention. Petitioners are challenging the NRC Staff's final "no significant hazards consideration" determination. This challenge is beyond the jurisdiction of the Licensing Board, and contrary to Commission regulations which provide that: 1/ See Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards' Consideration (January 10, 1989). l L_A-mmmm__.m_m.__m-_---__._.___m__ _ _ _ _ _ _ _ _ - _
f No petition or other request for review of or hearing on the staff's significant hazards.
. consideration determination will be entertained by the Commission. The staff's determination.is final, subject only to the Commission's discretion, on its own initiative, to review the determination.
See 10 CFR S 50.58(b)(6) (1988). As a result,. challenges related 1 to the NRC Staff's final "no significant hazards consideration" determination are not cognizable by licensing boards. See Pacific . Gas and Electric Co. (Diablo Canyon-Nuclear Power Plant, Units 1 and 2), CLI-86-12, 24 NRC 1, 4 (1986), rev'd in part on other grounds, San Luis Obispo Mothers for Peace v. NRC, 799 F.2d 1268 l (9th Cir. 1986); Florida Power & Light Co. (St. Lucie Nuclear Power Plant, Unit 1), LBP-88-10A, 27 NRC 452, 456-57 (1988); Vermont Yankee Nuclear Power Corp. (Vermont Yankee Nuclear Power Station), LBP-87-17, 25 NRC 838, 844 (1987). In fact, the Licensing Board in this proceeding recognized that matters pertaining to the NRC Staff's'"no significant hazards consideration" determination are beyond the Board's jurisdiction. See Order (January 19, 1989), p.2. Thus, Contention 1 is inconsistent with the Board's previous order in this proceeding. < Therefore, Contention 1 should be rejected because it raises an issue that is beyond the jurisdiction of the Licensing Board. l
3 8' i i s B.. Admissibility of1 Petitioners' Proposed ;
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Contentions 2 and 3 .t i
- 1. Applicable Standards' In addition to the standards governing the admissibility of contentions in all types of licensing proceedings, further limitations.are imposed upon the admissibility of contentions in operating license amendment proceedings due to the limited nature of such proceedings. For example, in license amendment proceedings, only thore contentions which fa '. within the scope of issues set out in the Federal Register notice of opportunity for hearing may be admitted for litigation. Commonwealth Edison Co.
l (Zion Station, Units 1 and 2),-ALAB-616, 12 NRC 419, 426 (1980); Portland General Electric Co. (Trojan Nuclear' Plant), ALAB-534, 9 NRC 287, 289 n.6 (1979); Public Service Co. of Indiana (Marble Hill Nuclear Generating Station, Units 1 and 2), ALAB-316, 3 NRC 167, 170-71 (1976); Florida Power s Light Co. (St. Lucie Nuclear Power Plant, Unit 1), LBP-88-10A, 27 NRC 452, 455 (1988). Moreover, contentions in a license amendment proceeding may not be used as a vehicle for reconsideration of issues which were previously considered by the NRC in earlier license proceedings, I absent any impact of the amendment on the issues in question. See Florida Power & Light Co. (St. Lucie Nuclear Power Plant, Unit 1), LBP-88-10A, 27 NRC 452, 466 (1988). As the licensing board stated in an amendment proceeding authorizing continued operation of a ) plant pending correction of certain design deficiencies:
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.i We are not authorized to examine matters that l
-were explored at'the: construction: permit or
' ' operating. license' stages, nor can we expand-c the issues.beyond-those related to the design-l deficiencies that resulted in the notice.of L hearing;which described the issues we are . _ empowered to' consider. Portland. General' Electric Co. (Trojan Nuclear Plant), LBP-78-40,-
- 8. NRC 717, 745 (1978)- (footnote omitted), af f'd ALAB-534, 9 NRC 287.(1979). Numerous other-cases have.also: applied the' principle that an amendmentfproceeding-is not an appropriate. forum for re-examining previous NRC-determinations that are not affected by the .
amendment. See, p.g., Florida Edwer & Light Co. (Turkey Point Nuclear. Generating Plant, Units 3 and 4), LBP-85-36, 22 NRC $90, 598-99 (1985) (contention related to the effects.of hurricanes and tornados on a spent fuel pool is inadmissible in'a spent fuel pool expansion amendment-proceeding, because such effects were considered in the Safety Evaluation at the operating license stage); Consumers Power Co. (Big Rock Point Nuclear Plant), LBP-80-4, ll.NRC 117, 127 (1980) (contention related to licensee's-financial ability to care for an expanded spent fuel pool is inadmissible in a spent fuel pool expansion amendment proceeding,. f because the licensee's financial qualifications were considered in granting the construction permit and operating license); Portland General Electric Co._ (Trcjan Nuclear Plant), LBP-78-32, 8 NRC 413, 415-16 n.1 (1978) (contention related to ability of a spent fuel pool to withstand earthquakes is inadmissible in a spent fuel pool expansion amendment proceeding, because the contention challenged e______-___
l." ( L vi 5-l We are not authorized to examine' matters that were explored at the construction permit or operating license stages, nor can we expand-the issues beyond those related to the design
. deficiencies that resulted in'the notice of hearing which_ described the issues we are empowered _to consider. 1 Portland General Electric Co. (Trojan Nuclear Plant), LBP-78-40,-
8 NRC 717, 745 (1978)-(footnote omitted), aff'd ALAB-534, 9'NRC. 287 (1979). Numerous other cases have also applied.the principle that an amendment proceeding is not an appropriate forum for re-examining previous NRC determinations that are not affected by the amendment. See, e.g., Florida Power & Light Co. (Turkey Point Nuclear Generating Plant, Units 3 and 4), LBP-85-36, 22 NRC 590, 598-99 (1985) (contention related to the effects of hurricanes and tornados on a spent fuel pool is inadmissible in a spent fuel pool expansion amendment proceeding, because such effects were considered in the Safety Evaluation at the operating license stage); Consumers Power Co. (Big Rock Point Nuclear Plant),.LBP-80-4, 11 NRb 117, 127 (1980) (contention related to licensee's financial ability to care for an expanded spent fuel pool is I inadmissible in a spent fuel pool expansion amendment proceeding, because the licensee's financial qualifications were considered in granting the construction permit and operating license); Portland General Electric Co. (Trojan Nuclear Plant), LBP-78-32, 8 NRC 413, 415-16 n.1 (1978) (contention related to ability of a spent fuel ; pool to withstand earthquakes is inadmissible in a spent fuel pool expansion amendment proceeding, because the contention challenged l
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th'e.safefshutdown' earthquake for the entire-facility). Conten-tions which seek to go beyond these bounds are objectionable and q inadmissible. -{
- 1 As! demonstrated'below, proposed Contentions 2 and'3 in, P' 'the' Amended Petition doLnot meet these requirements and; therefor'e.
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.should be rejected.
l 2; Contention 2 Contention ~2 states as follows:
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That the revised. temperature / pressure limits ~ q
'l that have been set for Turkey Point Unit 4 are
+ non-conservative and will cause that reactor- j 1
' unit to exceed the requirements of General Design Criterion 31 of. Appendix A to 10 CFR Part 50,.which requires that the reactor, coolant pressure boundary be designed with a 1 sufficient margin to ensure that', whenL stressed under operating, maintenance,-
testing,-and postulated. accident conditions, (1) the. boundary-behaves in a non-brittle' manner and (2) the probability of a rapidly propagating fracture is minimized. Petitioners. contend that,the new pressure /f temperature limits could cause the reactor vessel to exceed these requirements because the' Licensee has based its calculation of the predicted RTNDT for' Unit 4 partly on surveillance capsule V test results from Turkey Point Unit 3 rather than predicting the RTNDT for Unit 4 based on Unit 4 capsule V surveillance capsule data--a practice which is not scientific, not valid, and could cause the : Unit 4' reactor to behave in a brittle manner l which would make the chances of a pressure vessel failure and resultant meltdown more 4 likely. Petitioners contend that predictions of RTNDT and pressure / temperature l'imits derived from the shift in nil-ductility transfer should be based only on plant- i specific Unit 4 data, especially in light of ,
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+ the.factLthatLthe:only testsfever.: performed on
' Unit 14 weld specimens _ demonstrated:that.the
.weld materialiin the Unit-4' vessel.was'30%
more brittleEthan'that of.UnitL3'. Because Unit 4's weld material is more embrittled, - Petitioners contend that:the-FPL_ Integrated' Surveillance program'doessnotimeet thel Requirements of 10;CFR Appendix GLParts V.A
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and V.Bi!and 10 CFR Appendix H,? including " AppendixlH Parts IIC and IIIB.- Finally, Petitioners contend that'the surveillance capsule 1V-for. Unit-4 should ber testedLto establish the new' pressure / temperature limits i 1 and should the_ testing indicate that the p"
~RTNDT.for Unit-4 has passed the 300-degree s Fahrenheit? screening criterion, set'byDthe NRC, i
Unit 4Jshould be* shut down until itDis: 7 demonstrated that the Unit 4 reactor: pressure ^ vesselJean maintain its integrity-beyond this limit.. The Licensee objects to.the admission of Contention-2 on several grounds.- _First,.the Petitioners.are contending that-the," practice" of using surveillance capsules from one. reactor. vessel to help 1 predict the RTNDT for another reactor "is not scientific" and~is "not valid." Such a' contention constitutes an impermissible attack on the Commission's regulations in 10 CFR Part 50 Appendix H. Section II.C of Appendix H explicitly authorizes a licensee to utilize an integrated surveillance program "for a set of reactors that have similar design and. e
. operating features." Petitioners' claims that integrated surveillance programs are "not scientific" and "not valid" are Inconsistent with Section II.C of Appendix H and constitute an l attack upon the validity of this provision in the Commission's
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regulations. 2/ Therefore, this contentionishould be rejected, because contentions cha1*saging the validity of Commission regulations are inadmissible under 10 CFR S 2.758 (1988). See Carolina Power and Light Co. (She' aron Harris Nuclear Power Plant), ALAB-837, 23 NRC 525, 544, (1986); Kansas Gas and Electric Co. (Wolf Creek Generating Station, Unit 1), ALAB-784, 20'NRC 845, 846 The Petitioners have previously complained t' hat the use of an
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2/ integrated surveillance program is scientifically invalid, relying upon the letter by Dr. George Sih of Lehigh University that is attached to the Amended Petition. In response, the NRC Executive Director for Operations,' Victor Stello, stated that:- This contention is contrary to the basic > assumptions behind nuclear power industry and NRC use of rrdiation damage trend. curves (e.g., Regulatory Guide 1.99) derived-from analysis of many reactors. It is also contrary to NRC criteria for an integrated _ surveillance program (e.g., Appendix H, 10 CFR 30). Professor Sih's views were considered as part of the pressurized thermal shock studies in the 1981-82 time period and were rejected due to lack of technical. basis. Based upon the previously cited court decision and the technical basis in support of the integrated surveillance program for Turkey Point Units 3 and 4, Ms. Lorion's contentions have no merit.
. . . Contrary to Ms. Lorion's opinion, Turkey Point Units 3 and 4 are not operating in violation of NRC safety margins and their integrated reactor vessel material surveillance program conforms to NRC regulations and assures the best use of the available surveillance' capsules containing the critical weld materials for both units.
Letter from NRC Executive Director for Operations, Victor Stello, Jr., to U.S. Senator Lawton Chiles (Nov. 5, 1986), pp. 2-3 (emphasis added).
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(1984); Commonwealth Edison Co. (Byron Nuclear Power Station Units, 1 and 2), LBP-80-30 12 NRC 683, 692-93 (1980). 4 Second,.to the extent that Contention 2 is only challenging FPL's compliance with Appendix H (and not the validity of Appendix H itself), the contention is inadmissible because it seeks to re-examine a decision made in a previous license amendment proceeding. In early 1985, FPL applied for amendments to its operating licenses for Turkey Point in order to utilize an integrated surveillance program for the Turkey Point reactors. This application was duly noticed in the Federal Register (50 Fed. Reg. 9919 (March 12, 1985)), which stated that any interested person could request a hearing and submit a petition to intervene with respect to the requested amendments. No such requests or petitions were filed by Petitioners 3/ or any other person. On April 22, 1985, the NRC issued the requested amendments (No. 112 and No. 106, respectively) authorizing use of an integrated surveillance program for Turkey Point Units 3 and 4. These l amendments are enclosed for the convenience of the Board and the 3/ Petitioners have been active in opposing operation of Turkey Point since at least 1981, when they filed a letter asking the Commission to shut down Turkey Point due to Petitioners' concern regarding alleged reactor vessel embrittlement. See Florida Power and Light Co. (Turkey Point Plant, Unit 4), DD-81-21, 14 NRC 1078 (1981), dismissed, Lorion v. NRC, 712 F.2d 1472 (D.C. Cir. 1983), rev'd and remanded, Florida Power & Light Co. v. Lorion, 470 U.S. 729 (1985), aff'd on remand, Lorion v. NRC, 785 F.2d 1038 (D.C. Cir. 1986). For whatever reasons, Petitioners chose not to contest the issuance of the 1985 amendments approving the Turkey Point integrated surveillance program.
e. parties as Attachment A hereto. Contention 2 is a challenge to the validity of the integrated surveillance program that was previously approved by the NRC when it issued these amendments. Since nothing in the P/T limits amendments affects the integrated surveillance program, and since, as noted in Section III.B.1 above, a license amendment proceeding is not a proper forum for re-examination of determinations made by the NRC in earlier 4 proceedings (see Florida Power & Light Co., 27 NRC at 466), Contention 2 is inadmissible as a challenge to a previous NRC licensing decision and should be denied.
- 3. Contention 3 Contention 3 states as follows:
That the revised pressure / temperature limits that have been set for Units 3 and 4 are non-conservative and will not meet the requirements of General Design Criterion 31 of Appendix A to 10 CFR Part 50 which requires that the reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a non-brittle manner, and (2) the probability of a rapidly propagating fracture is minimized. Petitioners contend that the sufficient safety margin required by GDC 31 does not exist because the P/T limits for Units 3 and 4 were not based on the most limiting value of RTNDT as required by 10 CFR Part 50 Appendix G and H, for reactor vessel welds because the percentage of copper that was used in the RTNDT calculation is non-conservative in that it is lower than the percentage of copper that was used in previous surveillance test reports and lower than the percentage of copper quoted in many of the earlier FPL documents.
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Petitioners contend that the use of this non-conservative estimate of copper content means that the adjusted RTNDT is unrealistically low and that the current revised P/T limits are not restrictive enough to ensure that an adequate margin of safety against brittle fracture of the reactor vessel exists. This increases the possibility that. the reactor vessel (sic] for Unit 4 will behave in a brittle manner result.ing in a fracture of the vessel and subsequent meltdown of the reactor core. , Petitioners further contend that if a more conservative and accurate estimate of copper content was used to calculate the RTNDT, the P/T limits would be more restrictive and that in fact, there is a possibility that it could be discovered that the NRC screening criterion of 300-degree Farenheit has been reached and the Turkey Point Units 3 and 4 would have to be shut down because they do not meet the
. fracture toughness requirement of 10 CFR Part 50 Appendix G.
The Licensee objects to the admission of Contention 3. The contention (and the bases for the contention) alleges that a value of 0.26% was inappropriately used as the percentage value for copper in the Turkey Point reactor vessel-welds for the purpose of calculating the RTNDT. However, the value of 0.26% for the percentage of copper in these welds was approved in a Safety Evaluation issued by the NRC on April 26, 1984. For the convenience of the Board and the parties, a copy this Safety Evaluation is enclosed as Attachment B hereto. Contention 3 seeks to challenge the percentage value of copper that was approved by the NRC in this Safety Evaluation. Since nothing in the P/T limit amendments affects the percentage of copper in these welds, and __2______-______________ _-. 1
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? since a contention in a license amendment proceedi'ng may not seek to're-examine previous decizions made by the'NRC'(See Floridd
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Power &' Light Co.,-27 NRC at 466),-Contention'3 should be L rejected.'4/
- 4. Other~ Objections to Proposed Contentions'2 I and 3 Contentions 2 and.3 allege that Turkey Point Units 3 and-4 may not' satisfy the NRC's 300*F screening criterion'and should
-be shutdown. Presumably, Petitioners are referring to the 300*F screening 1 criterion in 10 CFR S 50.61 governing' protection.against
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. pressurized thermal shock (PTS) events. In the event the Licensing Board admits either Contention 2 or-3, the Licensing Board should strike any issues relating to whether or not the Turkey Point units satisfy the 300*F screening criterion'in 10 CFR
$ 50.61. This is not the proper forum to determine whether the screening criterion in 10 CFR S 50.61 is satisfied. First, such a determination is not within the scope of the notice of this 4/ The line of cases represented by Florida Power & Light Co.
-dealt with issues that had been previously subject to review by the NRC (either by the Staff in a Safety Evaluation or by a licensing. board in a' hearing) as part of a license . 1 proceeding. However, the reasoning in these cases applies equally whether the issues were previously subject to a license proceeding or, as here, were previously subject to review by~the NRC in issuing a safety evaluation outside'the context of a license proceeding. In both cases, the issue has been subject to a final determination by the NRC and is not affected by the amendment; therefore, the determination should not be open to reconsideration in'a later license proceeding. In both cases, 10 CFR S 2.206 would appear to be the appropriate' method of re-examining the continuing validity of the earlier action.
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- 13J-proceeding (53 Fed. Reg. 40,981, 40,988 (Oct. 19, 1988)), which relates solely to the adequacy of the Turkey Point P/T limit curves. See Commonwealth Edison Co. (Zion Station, Units 1 and 2), ALAB-616, 12 NRC 419, 426 (1980) (licensing board's jurisdiction limited by Commission's notice of hearing). .Second, the NRC has already determined that Turkey Point Units 3 and 4 satisfy the 300-degree Fahrenheit screening criterion set by the NRC, 5/ and nothing in the P/T limits amendments affects Turkey Point's compliance with this screening criterion. Consequently, Petitioners may not seek to re-examine this previous decision made by the NRC. See Florida-Power & Light Co., 27 NRC at 466. In any case, the Licensing Board has no authority to order the shutdown of Turkey Point Units 3 and 4, and at most can only deny the issuance of the license amendments. See Consumers Power Co.
(Palisades Nuclear Plant), LBP-79-20, 10 NRC 108, 123 (1979). l l 5/ See Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Projected Values of Material Properties , for Fracture Toughness Requirements for Protection Against ! Pressurized Thermal Shock Events, Florida Power and Light ' Company, Turkey Point Plant, Units 3 and 4 (March 11, 1987) (Attachment C hereto). 1 I
IV. Conclusion The Licensee is not contesting the standing of However, Petitioners Petitioners to intervene in this proceeding. Therefore, the have not proffered an admissible contention. Petitioners' request for a hearing and petition to intervene should be denied, and this proceeding should be dismissed. Respectfully submitted, 9M d V Harold F. Reis Steven P. Frantz Kenneth C. Manne Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036 (202) 955-6600 Co-Counsel for Florida Power & Light Company Co-Counsel John T. Butler Steel, Hector & Davis 4000 Southeast Financial Center Miami, Florida 33131 i (305) 577-2800 March 3, 1989
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. 00um . U+c UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 'gp g _g R: -
BEFORETHEATOMICSAFETYANDLICENSINGBOARDj"UCKf.3;,.g.g unu . , In the Matter of )
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FLORIDA POWER & LIGHT ) Docket Nos. 50-250 OLA - 4 COMPANY ) 50-251 OLA - 4
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(Turkey Point Plant, ) (P/T Limits) , Units 3 and 4) ) CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Response to Petitioners' Amended Request for a Hearing and Petition for Leave to Intervene" with Respect to License Amendments for Pressure / Temperature Limits in the above-captioned proceeding together with attachments were served on the following by deposit in the United States mail, first class, properly stamped and addressed, on the date shown below. B. Paul Cotter, Chairman Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Glenn O. Bright Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Jerry Harbour Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission i Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Office of Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 2
j' .- ., [; -l l. lL." - 2-- u Attention:' Chief, Docketing and Service'Section-(Original plus two copies) Joette Lorion, Director. l Center-for Nuclear _ Responsibility. , i 7210 Red Road'#217 .j Miami, Florida 33143 .-
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'.Janice Moore Office of General. Counsel U.S. Nuclear Regulatory Commission
- 1. Washington, D.C. -20555
-John T. Butler Steel, Hector.& Davis 4000 Southeast Financial' Center Miami,. Florida 33131 Dated this.3rd day of March.1989.
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Kenneth C. Mann'e Newman & Holtzinger, P.C. F 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036
sauw ATTACHMENT A - UNITED STATES , y NUCLEAR REGULATORY COMMISSION
; p, WASHWGTON, D C. 20555 j
'%,,,,,/ April 22, 1985 Docket Nos O ggq p q um ims Mr. J. W. Williams, Jr., Vice President m umuu u Newmali & Holtzinger Nuclear Energy Department Florida Power and Light Company y Post Office Box 14000 Juno Beach, Florida 33408
Dear Mr. Williams:
The Commission has issued the enclosed Amendment No.112 to Facility Operating License No. DPR-31 and Amendment No.106 to Facility Operating License No. DPR-41 for the Turkey Point Plant Units Nos. 3 and 4 respectively. The amendments consist of changes to the Technical Specifications in response to your application transmitted by letters dated February 8, 1985 and March 6, 1985. These amendments revise the Technical Specifications to provide consistency in identification of the surveillance specimen capsules in the Technical Specifications and the actual surveillance specimen capsules. The surveillance specimen examination schedule is also modified to provide better information in accordance with the current regulations. The proposed changes combine the existing Reactor Materials Surveillance Program into a single integrated program which conforms to the requirements of 10 CFR 50, Appendices G and H. We have discussed concerns and actions necessary regarding future core designs and in-cavity dosimetry in Section III of our Safety Evaluation provided in support of the amendments. Section II.C of 10 CFR 50 Appendix H, which was revised on July 26, 1983, permits an integrated surveillance program provided it meets the criteria specified and is approved by the Director, Office of Nuclear Reactor Regulation. We have indicated in our Safety Evaluation that the integrated surveillance program for the Turkey Point Plant permitted by the enclosed amendments meet the criteria specified in 10 CFR 50, Appendix H II.C. The . Director, Office of Nuclear Reactor Regulation, has approved the enclosed amendments which authorize an integrated surveillance program at the Turkey Point Plant in accordance with the requirements of 10 CFR 50, Appendix H II.C.
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.i Mr. Williams . April 22,1985 l A copy of the related Safety Evaluation is enclo' sed. A Notice of Issuance will be included in the Comission's next regular. monthly -
Federal Reaister notice. 4 Sincerely,. , T-_ Daniel G. Mcdonald, Jr., Project Manager i Operating Reactors Branch #1 Division of Licensing- / l
Enclosures:
- 1. Amendment No.112to DPR-31
- 2. Amendment No.106 to DPR-41
- 3. Safety Evaluation-cc: w/ enclosures '
See next page. h se _ _ _ - . - _ . . _ - - , _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _.__._.___._____-_-___-_.--__._.mm. _ _ _ _ - _ .
q . I, p - L J. W. Williams, Jr.- ' Florida Power and Light Company , l }
> cc: Harold F. Reis,- Esquire ' Administrator Department:of Environmental Newman and Holtzinger,~P.C.' ,
1615 L Street, N.W. Regulation Wa shington,' DC 20036 Power Plant Siting Section
< State of Florida L 'Mr. Jack Shreve 2600 Blair Stone Road Office of the Public Counsel Tallahassee, Florida 32301
. Room 4. Holland Building-
' Tallahassee. Florida ~ 32304 Dr. J.: Nelson Grace-Regional Administrator, Region II
. Norman.A. Coll, Esquire U.S Nuclear Regulatory Comission Suite 2900 Steel, Hector and Davis 4000 Southeast Financial 101 Marietta Street Center Atlanta, GA~'30303 Miami, Florida - 33131-2398:
Martin H. Hodder, Esquire
.1131 N.E. 86th Street Mr. Ken N. Harris, Vice President Miami, Florida ' 33138 Turkey Point Nuclear Plant Florida Power and Light Company Joette Lorion P.O. Box 029100 7269 SW 54 Avenue Miami, Florida 33102 Miami, Florida 33143 Mr. M. R. Stierheim Mr. Chris J. Baker,. Plant Manager County Manager of Metropolitan Turkey Point Nuclear Plant Dade County Florida Power and Light Company Miami, Florida 33130 P.O. Box 029100 Miami, Florida 33102 Resident Inspector .
Turkey Point Nuclear Generating Station ' Attorney General U.S. Nuclear Regulatory Comission ' Department of Legal Affairs Post Office Box 57-1185 The Capitol Miami, Florida 33257-1185 Tallahassee, Florida 32304 Mr. Allan~ Schubert, Manager Public health Physicist Department of Health and Rehabilitative Services 1323 Winewood Blvd. Tallahassee, Florida 32301 Intergovernmental Coordination and Review Office of Planning & Budget Executivs Office of the Governor The Capitol Building Tallahassee, Florida 32301
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FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT PLANT UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE-Amendment No.112 Licensj No. DPR-31 ,
- 1. The Nuclear Regulatory Comission (the Comission) has found that:
i A. The application for amendment by Florida Power and Light Company I (the licensee) dated February 8, 1985 and March 6, 1985, complies with the standards and requirements of the Atomic Energy Act of.1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. .There is reasonable assurance (1) 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 comon 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 changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License .
No. DPR-31 is hereby amended to read as follows: ; l l I
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i i (B) Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No. 112, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical
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Specifications. .
- 3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.
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FOR THE NUC1. EAR REGULATORY. COMMISSION Harold R. Denton, Director
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Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 22,1985 I
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FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT PLANT UNIT NO. 4 AMENDMENT TO FACILITY OPERATING LICENSE Amendynt No.106 License No. DPR-41
- 1. The Nuclear Regulatory Comission (the Comission) has found that:
A. The application for. amendment by Florida Power and Light Company. (the. licensee) dated February 8, 1985 and March 6, 1985, complies with the standards and requirements of the Atomic j Energy Act of 1954, as amended (the Act) and the Commission's i rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is 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 ef the Comission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license-amendment, and paragraph 3.B of Facility Operating License No. DPR-41 is hereby amended to read as follows:
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(B) Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.106, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective imediately and shall be/
implemented within 60 days of. issuance. FOR THE NUCLEAR REGULATORY COMMISSION tl Harold R. Denton, Director Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 22,1985 '
t i i ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 112 FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. 106 FACILITY OPERATING LICENSE NO. DPR-41 DOCKET NO. 50-250 AND 50-251
' Revise Appendix A as follows:
Remove Pa'ges Insert Pagdi iii iii iv iv
- Table 4.2.1 (cont'd) Table 4.2.1~(cont'd)
------------ 4.20-1 B3.1-3 B3.1-3 84.2-12 B4.2-12 B4.2-13 B4.2-13
------------ B4.20-1
*Last page of T&ble m
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'1 TABLE OF CONTENT 5 (Continued)
Instino. A T P. ass. 4.10 Aux'iliary Feedwater System 4.10 4.11 Reactivity. Anomalies - - 4.11-1
-4.12 Environmental Radiation Survey 4.12-1 4.13 Radioactive Materials Sources Surveillance 4.13 4.14 Snubbers .4.14-1 4.15 Fire Protection Systems 4.15-1 4.16 _ Overpressure Mitigating System .
4.16-1 4.17 Reactor Coolant System Pressure Isolation Valves 4.17-1 4.18. Safety Related Systems Flowpath 4.18-1 4.19 Reactor Coolant Vent System 4.19-1 4.20 Reactor Materials Surveillance Program / 4.20-l' l
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5.0 DESIGN FEATURES 3.1-1 5.1 Site . 5.1-1 5.2 - Reactor 5.2-1 5.3 - Containment 5.3 1 5.4 Fuel Storage 5.4 1 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Facility Staff Qualifications 6-5
' 6.4 Training ..
6-5 6.5. Review and Audit . 6-6 6.6- Reportable Event Action 6-14 6.7 Safety Limit Viohtlon 6-14 6.8 Procedures 6-14 6.9 Reporting Requirements 6-16 6.10 Record Retention 6-27 6.11 Radiation Protection Program 6-29 6.12 High Radiation Area 6-29 6.13 Post Accident Sampling 6-30 6.14 Systems Integrity 6-30 6.15 lodine Wnitoring 6-30 l 6.16 Back-up Methods for Determining Su5 cooling Wrgin 6-30 6.17 Process Control Program (PCP) 6-31 6.18 Offsite Dose Calculation Manual (ODCM) 6-31 B2.1 Bases for Safety Limit, Reactor Core B2.1-1 B2.2 Bases for Safety Limit, Reactor Coolant System Pressure B2.2-1 52.3 Bases for Limiting Safety System Settings, Protective Instrumentation B2.3-1 B 3.0 Bases for Limiting Conditions for Operation, Applicability B3.0-1 B 3.1 Bases for Limiting Conditions for Operation, Reactor Coolant System B3.1-1 B 3.2 Bases for Limiting Conditions for Operation, C6ntrol and Power Distribution Limits B 3.2-1 B 3.3 Bases for Limiting Conditions for Operation, Containment B3.3-1 B 3.4 Bases for Limiting Conditions for Operation, Engineered Safety Features B 3.4-1 I2 and
-lii. Amendment Nos.
' TABLE OF CONTENTS (Continued) isE11so. M East.
83.5 ' Bases for Limiting Conditions for Operation, ' ' Instrumentation B3.5-1 B 3.6 Bases for Limiting Conditions for Operation, Chemical and Volume Control System B3.6-1
' B 3.7 Bases for Limiting Conditbns for Operation, B3.7-1 Electrical Systems ' .
53.8 Bases for Limiting Conditions for Operation, ' Steam and Power Conversion Systems B3.8-1 B3.9 Bases for Limiting Conditions for Operation, Radioactive Materials Release 53.9 1 B 3.10 . Bases for Limiting Conditions for Operation, j Refueling B 3.10-1 f
'B3.!! Bases for Limiting Conditions for Operation,.
Miscellaneous Radioactive Material Sources B3.11-1 B3.12 Bases for Limiting Conditions for Operation, Cask Handling _ 33.12-1 B3.13 Bases for Limiting Conditions for Operation, Snubbers B3.13-1 B3.14 Bases for Fire Protection System B3.14-1 B3.15 Bases for Limiting Conditions of Operation, , B3.15-1 overpressure Mitigating System j B4.1-1 R B4.1 Bases for Operationa15afety Review - . 84.2 Bases for Reactor Coolant System In-Service Inspection B4.2 B4.3 Bases for Rea: tor Coolant System Integrity B4.3-1 Bases for Containment Tests B4.4- 1 B4.4 B4.5 Bases for Safety injection Tests . B4.5 1 B4.6 Bases for Emergency Containment Cooling System Tests 54.6-1 ' B4.7 Bases for Emergency Containment Filtering and Post Accident Containment Venting Systems Tests B4.7-1 Bases for Emergency Power System Periodic Tests B4.1-1
- B4.1 B4.9 Bases for Main Steam Isolation Valve Tests 84.9-1 Bases for (Auxiliary Feedwater System Tests 54.10-1 B4.10 B4.11 Bases for Reactivity Anomalies B4.!!-!
Bases for Environmental Radiation Survey B4.12-1 B4.12 Bases for Fire Protection Systems B4.13-1 B4.13 Bases for Snubbers B4.14-1 B4.14 B4.15 Bases for Surveillance Requirements, Overpressure ) Mitigating System B 4.15--! Bases for System Fbw Path Verifications - B4.18-1 B4.18 Bases for Reactor Coolant Vent System B4.19-1 B4.19 Bases for Reactor Materials Surveillance Program B4.20-1 l
.B4.20 l
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REACTOR MATERIAL 51IRVEILLANCE PROGR AM
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4.20.1 The iollowing Irradiation Specimen' Schedule shall be followed: CAPSULE REMOVAL SCHEDULE fdgadt. 32Dil' h V 3 12 years V 4 24 yer.rs X 3 33 years
. X 4 Standby Capsules U, W, Y, and I for Units 3 and 4 are held in standby. ,
4.20.2 The above surveillance shall be conducted using the Tensile a/nd Che.roy V - Notch Test. 1 I 112 106 l 4.20 1 Amendment Nos, and . _ .
The reactor vessel materials have been tested to determine their initial RTNDT. )
- Adjusted reference temperatures, based upon the fluena and copper content of the material in question, ars then determined. The heatup and cooldown !!mit curves include the shif t in RTNOT at the end of the service period shown on the heatup and cooldown curves.
The actual shift in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation survelliance specimans installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples has a definite relationship to the spectra at the vessel if. side radius, the measured transition shift for a sample can be related with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ARTNOT determined from the surveillance.e capsule is different from the calculated ART NDT forpp equivalent capsule radiation exposure. The pressure-temperatwe !!mit lines shown for reactor criticality and for irservice leak and hydrostatic testing hr.ve been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in TS 4.20 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. The !!mitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. ( l 1 112 106 B3.1-3 Amendment Nos. and i
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Item 6.5 (Cateforv G-2) - Pressure-Retalnine Boltine The bolting subject to this examination will be the bonnet botting in valves three (3) Inches in size or greater. This bolting will be inspected in acordance with Section XI of the Code as shown in Table 4.21. Item 6.6 (Category K-1) - Intetrally Welded Sgg;gg, There are no integrally-welded supports on the valves subject to this examination. Item 6.7 (Catemory K-2)- Suonorts and Hanners ) The supports and hangers of the valves subject to this examination Wil be visually examined in accordance with Section XI of the Code as shown in Table l#2-1. M!SCELL ANEOUS INSPECTIONS Item 7.1 - Reactor Coolant Pume Flywheels The flywheels shall be visually examined at the first refueling. At the fourth refueling, the outside surf aces shall be examined by ultrasonic methods. These 1 l examinations scheduled are shown in Table 4.2-1. Item 7.2 - Deleted. 1 I I1 I I l l l 112 106 B4.2-12 Amendment Nos. and j _j
i Item 7.3 - Steam Generator Tube Insoection The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83 Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. In service inspection of steam generator tubing also provider a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
's The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in neg!!gible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in' stress corrosion racking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (orimary-to-secondary leakage = 1 gallon per minute, total). Cracks having a primary-to-secondary leakage less than this ilmit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of I gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this !!mit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging !!mit which, by the definition of Specification 4.2.5.4.a is 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing in-service insoection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.2.a prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a l requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. B4.2.13 Amendment Nos.II2 a nd 106
84.20 BA3U - REACTOR MATERIAL SURVEILLANCE PROGRAn4 Each Type I capsule contains 28 Charpy V-notch specimen:, ten Charpy specimens machined from each of the two shell forgings. The remaining eight Charoy specimens are machined from correlated monitor material. In addition, each Type I capsule contains four tensile specimens (two specimens from each ! the two shell forgings) and six WOL specimens (three specimens from each of the two shelling forgings). Dosimeters of copper, nickel, aluminum-cobolt, and cadmium-shielded aluminum-cobalt wire are secured in holes drilled in spacers at the top, middle, and . bottom of each Type I capsule. l
~ I Each Type !! capsule contains 32 Charpy V-notch specimens: eight specimens machined from one of the shell forgings, eight specimens of weld metal and eight specimens of MAZ metal, the remaining eight specimens are correlyion monitors.
In addition, each Type 11 capsule contains four tensile specimens add four WOL specimens: two tensile specimens and two TOL specimens from od of the shell forgings and the weld metal. Each Type II capsule contains a dosimeter block at the ! center of the capsule. Two cadmium-oxide-shleided capsules, containing the two isotopes uranium-238 and neptunium-237, are contained in the dosimeter block. The double containment afforded by the dosimeter assembly prevents loss and 1 contamination by the neptunium-237 and uranium-238 and their activation products. ( Each dosimeter block contains approximately 20 milligrams of neptunium-237 and 13 milligrams of uranium-238 contained in a 3/8-inch-OD sealed brass tube. Each tube is placed in a 1/2-inch-diameter hole in the dosimeter block (one neptunium-237 and one uran!um-238 tube per block), and the space around the tube is filled with i cadmium oxide. After placement of this material, each hole is blocked with twn 1/16-inch aluminum spacer discs and an outer 1/8-inch-steel cover disc, which is welded in place. Dosimeters of copper, nickel, aluminum-cobalt, and cadmium-shielded aluminum-cobalt are also secured in holes drilled in spacers located at the top, middle, and bottom of each Type 11 capsule. Caosule Tvoe Caosule identification I 5 11 V II T I U 11 X
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! Y I Z This program combines the Reactor Materials Surveillance Program into a single integrated program which conforms to the requirements of 10CFR30 Appendices G and M.
112 106 B4.20-1 Amendment Nos. and
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NUCLE AR REGULA*: ORY COMMISSION c
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.S. AFETY EV/.LUATION BY THE OFFICE OF NL' CLEAR REACTOR RELATED TO AMENDMENT NO.112 TO' FACILITY OPERATING LICENSE NO. D AND AMENDMENT l:0.106'TO FACILITY OPERATIrlG LICENSE NO. D_P,R-41:
FLORIDA POWER AND LIGHT COMPAN'/, TURKEY P0ltli UN"IT' N05. 3 Af!D 4 DOCKET NOS. 50-250 AND 50-251 I. Introduction
. in' a letter f rom J. W. Williams, Jr. to D. G. Eisenhut, dated February 8,. .
1985, Florida Power & Light Company requested that the Turkey Point Units
.No. 3 and 4 Technical Specifications be amended to combine the reactor
-vessel material surveillance program for these units into a single inte-grated surveillance program. Additional information concerning the pro- _l posed change was provided by the licensee in a letter from J. W. Willi us, Jr.
to S. A. Varga dated March 6, 1985. f A revised Appendix:H, 10 CFR 50 was published in the Federal Register on May 27, 1983 and became effective on July;26, 1983. Section' II.C of the 1 revised Appendix H permits an integrated surveillance program provided it is approved by the Director, Office of Nuclear Reactor Regulation. This-l section of Appendix H identifies the criteria to be used in evaluating i.h2 integrated surveillance program. The criteria are: .
- 1. There must be. substantial advantages to be gained, such as reduced power outages or reduced personnel exposure to radiation, as a dircct result of not requiring surveillance capsules in all reactors in the set.
- 2. The design and operating features of the reactors in the set must be sufficiently similar to permit accurate comparisons of the predicted amount of radiation damage as a function of total pcwer output.
- 3. There must be an adequate dosimetry program for each reactor.
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- 4. :.IThere must be'a, contingency' plan to assure that the. surveillance
- program for each reactor will- not'.be jeopardized by operatica at -
reducedpowerlevelor'bj/anexthndedoutageofanotherreactor from which data-are expected.-
- 5. No' reduction' in the requirements for number of ' materials-to be y q H
irradiated,1 specimen type, or number of specimens.per reactor. is I g pennitted, but' the amount'of testing may be' reduced if the . initial results t 0 # agree with predictions.
- 6. There.must be adequate. arrangement for data sharing between plants. ;
i II. Evaluation Each unit at Turkey Point began commercial operation with 8 s' surveillance
. capsules in each reactor vessel. Ten capsules contained forging material:
j and six capsules' contained weld metal, forging, and heat affected zone (liAZ) J materials. To date, two capsules _containing forging material and two.- capsules containing weld metal, forging,- and HAZ materials were irradiated, removed from.the vessel,-and tested. The test results from the surveillance material' indicate that the weld metal will sustain the most irradiation damage. Since, based on the initial test the weld metal is more susceptible to irradiation damage than the forging material, the licensee has proposed to retain the capsules with forging material as. standby specimens in the reactor vessel and test. only those capsules with weld metal, forging, and HAZ materials. Since fewer capsules will be withdrawn than originally' anticipated, the radiation exposure (ALARA) to plant
- personnel should be reduced.
The licensee's FSAR Volume 2 indicates that the materials and designs for the core, thermal shield, core barrel and vessel are the same for each unit at Tur, key Point. Since the neutron energy spectrum is a function of geometry, materials, and core loading, the relative neutron spectrum for both reactors should be equivalent for equivalent core loadings. The ar*
t x , l licensee indicates that fuel management and cycle lengths for both units have been similar. Thus neutron' spectra profiles at the peak fluence locations should be equivalent. ,
~The neutron fluente, which is used to predict radiation damage, is calcu-lated using PDQT power distribution data, and computer codes SORREL and D'OT 4.3. As built dimensions and individual material properties 4re incorporated into the DOT 4.3 models. Hence, using these codes, the licensee will be able to predict radiation damage as a function of power !
output for each unit. i Each vessel has both in-capsule and in-cavity dosimetry, which will be j j used to verify the neutron spectra and the codes that were used to predict neutron fluence as a function of power output. Since each plant has its , own capsules and both plants are capable of independently predicting and r monitoring radiation damage as a function of power output, the program will not be significantly jeopardized by operation at reduced power levels or by an extended outage of either plant. Based on the intial test, the limiting material for each unit is weld material, which is identified as SA 1101. This material is in each capsule that will be irradiated and tested. Capsules that have been deleted from surveillance testing do not contain the limiting material and will ha retained as standby specimens in the reactor vessel. Since the amount of limiting material in the surveillance program has not chnaged, the number of useful surveillance specimens available for testing has not changed. Both units have common management and the surveillance program will be f j managed by their Nuclear Energy Department. Therefore, there should be i adequate data sharing. l 1
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!!1. Findings .
1.
- We have concluded based on the details in Section II of this: Safety Evaluation. that the. integrated surveillance program meets.the evaluation criteria specified in 10 CFR 50, Appendix H II.C. If' future core designs
. are significantly different than those documented by the)icensee, the licensee must explain the effect that the changes-have on neutron irradiation danage and the surveillance capsule withdrawal schedule.
- 2. In-cavity dosimetry testing should continue in order to reduce pro-jected uncertainties in neutron fluence. If these test results provide an ef fective method of monitoring vessel neutron fluence,
- the in-cavity dosimetry should be-incorporated into the integrated surveillance program.
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Ji ,IV. Environmental Consideration These amendments involve changes-in the installation or use of the facilities components located within the restricted areas as defined in 10-CFR 20 and.in surveillance requirements. The staff has determined that these amendments involve no significant increase in the amounts, WId'no significant change-in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative-occupational radiation exposure. . The Commission has previously issued a proposed finding that'these amendments involve no significant hazards consideration and there has been no public comment on such finding. Accordingly, these amendments ' meet the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental- impact statement or environmental assessment need be prepared-in connection with the issuance of these amendments. Y. Conclusion
-We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance-with the Commission's regulations and the issuance of these amendments will not
' be inimical to the common defense and security or to the health and safety of the public. .
Dated: April 22, 1985 Principal Contributors: B. Elliot
' ATTACHMENT B -
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION wasmNMC N,0. C. 20655
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' April 26,1984
' Docket Nos. 50-250 and 50-251 Mr. J. W.. Williams, Jr., Vice President Nu: lear Energy Department g- Florida Power and Light Company
. Post Office Box 14000
~ Juno Beach, Florida- 33408
Dear Mr. Williams:
SUBJECT:
EVALUATION OF REACTOR VESSEL MATERIALS DATA FOR TURKEY 7 POINT PLANT UNITS 3 AND 4 REACTOR VESSELS Sy letter dated February 10, 1984, you provided a report which included.a larger data base of information on the chemical composition of the reactor vessel welds .for the Turkey Point, Units 3 and 4, reactor vessels than .was . As a result of your evaluation of ! previously available to the NRC staff. should be t10*F,without any standard the data, you concluded that the RT deviation,CoppercontentequaltoON6%andNickelcontentequalto0.60%. The Component Integrity'Section, Materials Engineering Branch, Division of Engineering has reviewed the report provided in the submittal referenced e above. The staff has concluded, based on the enclosed Safety Evaluation, that the data base supports the above values and they are acceptable for screening criteria calculations for Pressurized Thermal Shock considerations for the Turkey Point Units 3 and 4 reactor vessels. , Sincerely, p i , te a'rga, C f Operating Reactors nch #1 Divisien of Licensing
Enclosure:
. As stated cc w/ enclosure:
See next page
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J. W. Williams Jr. Turkey Point Plants Florida Power and Light Company - Units 3 and 4 cc: Harold,F. Reis' Equire Administrator' Lowenstein, Newman, Reis and Axelrad Department of Environmental 1025 Conne~cticut Avenue, N.W. Regulation Suite 1214 Power Plant Siting Section Washington, DC' 20036 State of Florida 2600 Blair Stone Road Bureau.of Intergovernmental Relations Tallahassee, Florida 32301 660 Apalachee Parkway Tallahassee, Florida 33130 James P. O'Reilly Regional Administrator, Region II Norman A. Coll, Esquire . U.S. Nuclear Regulatory Commission Steel, Hector and Davis 101 Marietta Street, Suite 3100 1400 Southeast First National Atlanta, GA 30303 . Bank Building Miami, Florida 33131 Martin H. Hodder Esquire
- 1131 N.E. 86th Street Mr. Henry Yaeger, . Plant Manager Miami, Florida 33138 Turkey Point Plant .
Florida' Power and Light Company Joette Lorion P.O.' Box 013100 7269 SW 54 Avenue Miami, Florida 33101 Miami, Florida 33143 Mr. M. R. Stierheim County Manager of Metropolitan Dade County Miami, Florida 33130 Resident inspector . Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Commission Post Office' Box 1207 Homestead, Florida 33030 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.W.
- Atlanta, GA 30308 Mr. Jack Shreve l' Office of the Public Counsel -
Room 4. Holland Building Tallahassee, Florida 32304 4
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L ', " [',ba taig'sz,, _ UNITED STATES lT8 j( . p. NUCLEAR REGULATORY COMMISSION ' [ _g. - y WASHINGTON. D. C. 20555 l; f..S/p... SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FLORIDA' POWER AND LIGHT-COMPANY TURKEY POINT UNIT NOS. 3 AND 4 DOCKET NOS. 50-250 AND 50-251 t-I. Introduction , In the reactor a ssels for Turkey Point Units 3 and 4, made by the Babcock and Wilcox Comp.ny, the critical beltline circumferential welds were both
~
icentified as weld SA-1101. It was made using Page copper coated weld wire,; heat number 71249 and Linde 80 flux lot number 8445. The original-report on copper content from B & W, 0.21% copper, was disregarded in thet Pressurized Thermal snock (PTS) review, because those old values had been proven to be low in many cases. Instead, a value of 0.32% Cu was used, which was the average of 5 measurements on broken irradiated Charpy bars by Westinghouse. These had been reported 1o the NRC by FPL letter of Jan. 21, 1982 in their "150 day report" on PTS. , II. Evaluation , Letter L-84-31 from FPL dated Feb. 10, 1984 presented the results of a j total of 51 measurements of copper content, most of which were obtained i from B & W following the release of proprietary data in July,1983 and published as BAW 1799, "B & W 177-FA Reactor Vessel Beltline Weld Chemistry Study". The letter from FPL recommends t' hat the mean of the 51 values - 0.26%~Cu (standard deviation of 0.04%) - be used in futui e analyses. Similarly, there were 41 measured. values of nick,el content , with a,mean of 0.60% and a standard deviation of 0.04%.
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In reviews of this kind, our practice has been to consider that the copper content is determined by the weld wire heat number and to use
. best estimate values of copper and nickel content in entering our tables .
for calculation of shift. This practice is being put in writing in Revision 2 of Reg. Guide 1.99, which goes on to state that the best estimate is the mean of the me,asured values for the weld wire heat ; number when these are available. Thus, the procedure proposed by FPL is satisf actory, provided all 51 values are of equal weight. In addition to the 5 measurements reported by Westinghouse from surveillance Nine measure-work, there are two significant groups of data from BAW 1799. ments made on weld SA 1101 (Wire heat No. 71249), obtained from a nozzle dropout, yielded values ranging from 0.15 to 0.23% Cu, average of 0.18% Cu. Twenty-six measurements made on weld SA 1769 (Wire Heat No. 71249, but a . different weld flux lot) gave a range 0.24 to 0.34% Cu, average of 0.28% Cu. An explanation for the difference, solicited from A. Lowe of B & W, is that J it may reflect a difference in the amount of copper plating applied to (Copper plating is dif ferent redraw bar lots from Wire Heat Number 71249. inch diameter bars prior to applied while the-material is in the form of drawing the wire). Or, the difference may reflect some difference in weld procedures used for the surveillance weld, from which the higher values I J came, and the nozzle shell course longitudinal weld from which the lower copper values came. 1
t l t n It is disconcerting to find two populations of copper content having means of 0.18 and 0.28% Cu represented by one weld wire heat number. In deciding what value to use for the welds in Turkey Point Units 3 and 4 the choice is between a grand average (0.26% Cu), or the average for the higher of the two populations (0.28% Cu.), or the value used in the PTS work (0.32% Cu). To put the decision in perspective, from Table I of proposed Reg. Guide 1.99 Rev. 2, we find that at these nickel and copper levels, 0.01% Cu is equivalent to about 4.0*F change in RT NDT at a fluence of slightly over 19 n/cm2 (E >/Hev), the current fluence level for these plants. 1 X 10 Therefore by reducing the best estimate value of copper content from 0.32 to 0.26 we have reduced the calculated value of RT NDT by about 24 F. For comparison, the margin added to the nean, per the provision of proposed Revision 2 is 56*F. III. Conclusion , To be consistent with the practice of using the mean of the measured values for the weld wire heat number, as written in Revision 2 of Reg. Guide 1.99, the staff accepts the mean value of 0.26% Cu. The corresponding nickel content is 0.60%. FPl. also provides some measured values of initial reference temperature for weld SA 1101, obtained from an EPRI report. Following ASME The staff accepts Code rules, the initial RT NDT was found to be +10*F. this value. Date: April 26, 1984 Principal Contributor: P. Randall
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I, , 3 5.E C ATTACHMENT:CL =~ bNITED STATES -
$. 1g NUCLEAR REGULATORY COMMISSION
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- 4' o ' / WASHINGTON, D. C. 20555 g v f- ' March 11, 1987 Docket Nos. 50-250 and 50-251 7 MNg g7 -
M _ MAR 171987 Mr. C. O. Woody, Group Vice President Nuclear Energy Department i [@b
'Newman & Holtzinger, P.C.
Q Florida Power and Light Company Post Office Box 14000 Juno' Beach, Florida 33408
Dear Mr. Woody:
Subject:
Projected Values of Material Properties For-Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events - Turkey Point Plant, Units 3 and 4
Reference:
TAC Numbers 59992 and 59993 By letter dated January 23 you provided your response,to the Pressurized Thermal Shock (PTS ! 10 CFR 50.61 for the Turkey Point Plant,' Units 3 and 4. The staff, with the assistance of our contractor Brookhaven National Laboratory (BNL), have reviewed your submittals and performed confirmatory calculations. Based on our review and confit'matory calculations, we have determined that the material properties of the reactor vessels beltline materials,'the projected fluence at the inner surface of the reactor vessels at the expiration date of the licenses and the calculated RT ~ (April 27, 2007) tobeacceptable.PTlhecalculatedRTat the expiration date of the licenses both the licensee's and our confirmatory calculations, is well below the be,ening criterion of ~ 300 F for the limiting circumferential weld material at the expiration date of the licenses and is therefore in conformance with the PTS Rule. The details of our evaluation enclosed and the basis for our conclusions are included in the Safety Evaluation. The PTS Rule requires that the projected assessment of the RT must be updatedwheneverchangesincoreloadings,surveillancemeasubentsorother information change in the(including projected changes values. in capacity factor) indicate a significant This ensures that you will track the accumulated fluence for the limiting beltline materials throughout the life of the plant to verify that your assumptions remain valid. In this regard, we request that you submit a re-evaluation of the RTPTS and comparison of v
? L. . f.:]- :. p. Mr. C. 0.~ Woddy e 2 the predicted value in-any future Pressure Temperature submittals which a submitted'as Units. required by'10 CFR 50, Appendix G, for.each of the Turkey This concludes our ac'tions related to the' above TAC numbers. Sincerely, p V Daniel G. Mcdonald, Senior Project Manager PWR Project Directorate #2 Division of PWR Licensing-A
Enclosures:
As stated cc w/ enclosures: See next'page ( b-lt 1
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s Mr. C. O. Woody Florida Power and Light Company Turkey Point Plant cc: I Harold F. Reis, Esquire Newman and Holtzinger, P.C. Administrator. 1615 L Street, N.W. Dep&rtment of Environmental Regulation l Washington, DC 20036 ' Power Flant Siting Section Mr. Jack Shreve State of Florida Office of the Public Counsel 2600 Blair Stone Road Room 4, holland Building Tallahassee, Florida 32301 Tallahassee, Florida 3E304 Regional Administrator, Region II Norman A. Coll, Esquire U.S. Nuclear Regulatory Commission Steel, Hector and Davis Suite 2900 4000 Southeast Financial 101 Marietta Street Center Atlanta, Georgia 30323 Miami, Florida 33131-2398 Martin H. Hodder, Esquire Mr. C. M. Wethy, Vice President 1131 NE, 86th Street Turkey Point Nuclear Plant Miami, Florida 33138 Florida Power and Light Company P.O. Box 029100 Miami, Florida 33102 Joette Lorion 7269 SW, 54 Avenue Miami, Florida 33143 Mr. M. R. Stierheim County Manager of Metropolitan Mr. Chris J. Baker, Plant Manager I Dade County Turkey Point Nuclear Plant v ' Miami, Florida 33130 Florida Power and Light Company P.O. Box 029100 Resident Inspector Miami, Florida 33102 U.S. Nuclear Regulatory Commission Turkey Point Nuclear Generating Station Attorney General Post Office Box 57-1185 Department of Legal Affairs Miami, Florida 33257-1185 The Capitol Mr. Allan Schubert, Manager Tallahassee, Florida 32304 Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd. Tallahassee, Florida 32301 Intergovernmental Coordination and Review - Office of Plenning & Budget Executive Office of the Governor The Capitol Building Tallahassee, Florida 32301
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gWto f3 / ugk UNITED STATES 8 3 El NUCLEAR REGULATORY COMMISSION y . ,E wAssiscTow. o. c. 20sss
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING PROJECTED VALUES OF MATERIAL PROPERTIES FOR FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS FLORIDA POWER AND LIGHT COMPANY TURKEY POINT PLANT, UNITS 3 AND'4 I. Introduction As required by 10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock" (PTS Rule) which was published in the Federal Register on July 23, 1985, the licensee for each operating pressurized water surface) reactor "shall of reactor submit vessel projected beltline materials values by g of RT@ng(at thetime values from the inner vessel of submittal to the expiration date of the operating license. The assessment ! must specify the bases for the projection including the assumptions regarding I core loading patterns. This assessment must be submitted by January 23,.1986, and must be updated whenever changes in core loadings, surveillance measurements c or other information indicate a significant change in projected values." ' I By letters dated January 23, 1986, and supplemented on June 5 and July 7, 1986, the Florida Power and Light Company submitted information on the material properties and the fast neutron fluence (E > 1.0 MeV) on the inside surface of the reactor pressure vessel, in compliance with the requirements of 10 CFR 50.61 for the Turkey Point Plant, Units 3 and 4. The RT and fluence values were projected to April 27, 2007, which is the expiratiofi Tbate of both licensees. II. Evaluation of The Material Aspects The controlling beltline material from the standpoint of PTS susceptibility was identified to be intermediate-to-lower girth weld SA-1101 (weld wire heat number 71249) for both unit 3 and unit 4. The material properties of the controlling material and the associated margin and chemistry factor were reported to be: Utility Submittal . Staff Evaluation Cu (copper content, %) 0.26 0.26 Ni (nickel content, %) 0.60 0.60 I (Initial RT NDT, ) +10 +10
L~- I m m r t'
~r Utility Submittal p Staff Evaluation M (Margin, *F) --
48 CF (Chemistry' Factor, 'F) -- 166.8'- The controlling material has been properly identified. The justifications for the copper and nickel contents and thec initial RT are given by-reference to a submittal dated February 10,1984,whibTwas accepted by the staff on April 26, 1984 (S.A. Vu ga to J.W. Williams of FPL). The justifications meet our criteria for PTS submittals. The margin has been deriveo from consideration of the bases for these values, following the PTS Rule, Section 50.61 of 10 CFR Part 50. Assuming that the reported values of fluence are correct,' shown above. Equation 1 of the PTS rule governs, and the chemistry factor is as III. Evaluation'of the Fluence Aspects-
~
Early studies.of the PTS issue for the Turkey Point plants indicated that (a) the controlling beltline material is the intermediate-to-lower circum-ferential weld SA-1101 and (b) a flux reduction ~ factor of about 4.5 should
.be effected for both plants to prevent them from reaching the 10 CFR 50.61 screening licenses). criteria before April 2007 (i.e., the expiration ~of their operating To this end the licensee implemented a flux reduction scheme based on the use of part-length absorber rods . located on the assemblies on the core flats. The purpose of this review was to evaluate the effectiveness of the flux reduction measures and to evaluate the projected estimate of the peak azimuthal fluence at the end of the c'urrent license on the lower circum-ferential welds.
. The licensee's determination of the fast flux at the lower circumferential weld is based on the DDT 4.3 discrete ordinates transport code in (r,0) geometry.
The calculations employ a nuclear data library based on the 47-neutron group BUGLE-80 (ENDF/B-IV) library, and an Sa-P 3 angular decomposition. The neutron source is obtained from PDQ-7 generated pin-wise,' cycle-specific power distri-butions. The presence of plutonium is accounted for by a mixed U+Pu core neutron
. source normalization factor. The fast (E > 1.0 MeV) flux at the lower circum-ferential weld is then given by:
$ weld = $DDT(r=PV inner surface, 0) P (z weld elevation) i.e., the DDT 4.3 (r,0) result is multiplied by the relative axial power at the elevation of the weld (from a NODE-P calculation) to provide an estimate of the three-dimensional fast flux at that location.
The basic elements of the Brookhaven National Laboratocy (BNL), our contractor's, approach for determining the fast flux at the peak wal'1 location on the lower circumferential pressure vessel welds are summarized below: i
3-
' 1.
Forward and/or adjoint fixed source calculations are performed with 00T-4.3 in (r,0) and (r,z) geometries in order to determine the contributions of selected assemblies and axial zones to the E > 1.0 MeV
' flux peakatazimuthal the lower.circumferential location). welds, near the core major axis (the.
2.' The DOT calculations employ a 16 neutron group library derived from the ENDF/B-IV based on 100-neutron group EPR library and an Sg-P3 angular decomposition. 3. Cycle-specific source data provided by the licensee are used in conjunction with the 00T-4.3 results to synthesize the three-dimensional flux. Only assembly averaged sources are considered, and the neglected pin-wise power distributions from an earlierarestudy. accounted for via a generic adjustment factor determined 4. An exposure correction is applied on an assembly basis and includes the
.effect of plutonium dependent on both the source normalization and the energy-source spectrum.
Results for resulting the present values of RT and projected end of license fast fluences, and at the inner surface of the lower circumferential. weldnearthecoremaj$NaxisaregiveninTables1and2forTurkeyPoint Units.3 and 4, respectively. The four BNL results quantif exposure (Cases 1 vs. 2 and 3 vs. 4), and the licensee vs.ythe the effects BNL of approaches for estimating the three-dimensional flux at the limiting location. For Unit 3: (1) the exposure effect is worth 3.5% and 7% at present and e E0L conditions, respectively; (2) the axial treatment underestimates the present fluence by ~2% and the EOL fluence by ~10%; and (3) the difference between the licensee and BNL Case 1 results is <~3%. For Unit 4, the exposure dependent results show a similar trend relative to the cases with no exposure correction, and the different axial treatments have a smaller effect (<4%). However, comparison of the licensee results and those from Case 1 show an ~12% discrepancy (vs. <~3% for Unit 3). It is significant that, even though the BNL results for the fluence (Case 4) are RT higher than those obtained by the licensee, the resultant values for weNik,are still well below the NRC screenirq criterion of 300*F for circumferential with end of license values of 271 F and 276*F for Units 3 and 4, respectively. in a RT Therefore, we conclude that the proposed flux reduction results ' PTS which meets the 10 CFR 50261 criterion and is acceptable. IV. Conclusion , e Both the licensee's and our confirmatory calculations are well below the screening criteria for the limiting material at the expiration date of tne licenses. The licensee has calculated a RT of 236 F and 233*F for Units 3 and 4, respectively. AsstatedintheevalbkionportionofthisSafety Evaluation, the staff's confirmatory calculations are higher with a RT of 271*F and 276"F for Units 3 and 4 respectively, for the limiting weld material to April 27, 2007, which is the expiration date of both licenses. circherential l l l-
We therefore conclude that the Turkey Point Units 3 and 4 pressure vessels meet
.the toughness requirements of 10 CFR 50.61 for operation to the end of their current licenses provided that future fuel loadings continue to use the special assemblies for the reduction of the fast neutron fluence to the lower circumferential welds.
In order for.the staff to confirm the' licensee's projected estimated RT throughoutthelifeoftheTurkeyPointPlant, Units 3and4 operating and comparison
$Ihenses, the licensee is required to submit a re-evaluation of the RT [kals which are tothepredictedvaluewithfuturePressure-Temperaturesubmk required by 10 CFR 50, Appendix G.
Date: 1 Principa1' Contributors: P. N. Randall L. Lois t I 1
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TABLE'1 ; Present and Projected E0L Fluence (>1.0 MeV) and RTPTS for ' Turkey Point Unit-3 Present
' Case Fluence (l) RTPTS(2)
End-of-Licens'e Fluence (1)- RTPTS (2) 1 BNL-FP&L Axial- ) Treatment l
- 1. Zero Exposure 1.31 237 2.10 -262
- 2. Exposure Corrected 1.35 239 2.25 266 BNL 3-D Synthesis i 3. Zero Exposure 1.33 238 2.31 267
- 4. Exposure Corrected 1.37 240 2.47 271 FP&L 1.27 236 2.15 263 2
(1) Fluence -(>1.0 MeV) x 10-18 n/cm i (2) RTPTS from Eqn 1 of 10CFR 50.61 i l
.- 1 l
l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ . a . _ _ _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _
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f: t .,
-9 TABLE 2 L
'Present and Projected E0L' Fluence (>1.0 MeV) and:
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RTPTS for. Turkey Point Unit-4 End-of-Licens e L Present(l) Fluence (1) RTPTS I' Case Fluence RTPTS(2) BNL-FP&L' Axial Treatment
- 1. ~ Zero Exposure 1.33- 238 2.40 266 F 274
.2. Exposure Corrected . .1.39 ~240 2.60 BNL-3-D Synthesis.
238 -2.48 271.- o
- 3. Zero Exposure 1.32
- 4. Exposure Corrected 1.39 240 2.70L 276 1.19 233 2.16 263 F P& L ...
2 (1) Flu'ence (>1.0 MeV) x 10-" n/cm (2) RTPTS from Eqn.1 of 10CFR 50.61 _. -I k
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