ML20070W023

Application for Amends to Licenses DPR-44 & DPR-56, Consisting of Tech Spec Change Request 90-17,modifying pressure-temp Limits for Reactor Vessels

ML20070W023
Person / Time
Site:	Peach Bottom
Issue date:	03/28/1991
From:	Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20070W025	List: ML20070W023 ML20070W027
References
NUDOCS 9104120286
Download: ML20070W023 (13)

Text

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- PIHLADELPillA EloECTRIC COMPANfU UI" l NUCLEAR OROUP HEADQUARTERS l j 955 65 CilESTERDROOK BIND.

i WAYNE, PA 19087.$691 (21$) 64MOOO

NUCt.LAR ENGINEERINO & $ERVICES DCPARTMENT March 28, 1991 I

Docket Hos. 50-277 50-278 License Hos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission Attn Document control Desk Washington, DC 20555

SUBJECT:

Peach Bottom Atomic Power Station, Units 2 and 3 Technical Specifications Change Request Dear Sir Philadelphia Electric Company hereby. submits Technical Specifications Chango Roquest No. 90-17, in accordance with 10 CPR 50.90, requesting an amendment to the Peach Bottom Units 2 and 3 Technical Specifications (Appendix A) of-Facility Operating License Nos. DPR-44 and DPR-56. Infornmtion supporting this Change Request t is contained in Attachment 1 to this letter, and the proposed replacement Technical Specifications pages are contained in '

Attachment 2.

The company requests Technical Specifications changes to

modify the pressure-temperature limits for the reactor vessels. .

If you have any questions regarding this matter, please

! feel free to contact us.

l very truly yours lAr m cl4 G. J. eck, Manager Licensing Section Nuclear Engineering.& Services

Enclosure:

Affidavit Attachments 1 and 2 cc T. T. Martin, Administrator, Region I, USNRC J.J. Lyash, JSNRC' Senior Resident Inspector, PB L g\\.

\T 9104120286 910328 4

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. ,, l COMMONWEALTl! OF PENNSYLVANIA:

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COUNTY Or Cl! ESTER t D. H. Smith, being first duly swer.n, deposos and sayst That he is Senior Vice President-Huclear of Philadelphia Electric Company, the Applicant herein; that he has read the enclosed request for amendment of Peach Bottom Units 2 and 3 racility Operating License Nos. DPR 44 and DPR-56 (Change Roquest 90-17) and knows the contents thereofr and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief, a

Senior Vice President-Nuclear I

l Subscribed and sworn to before me thise2/^$ay of March 1991.

I f $ow.i _. 0. hl&> 4.

Notary Public

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5 ATTACHMENT 1 PEACil BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Hos. 50-277 50-278 License Nos. DPR-44 {

DPR-56 1

TECHNICAL SPECIFICATIONS CilANGE REQUEST NUMBER 90 " Revision of Pressuro Temperature Limits for the Reactor Vesseln" 11 Pages l

8 a

Docket Nos. 50-277

. 50-278 License Non, DPR-44 DPR-56 Philadelphia Electric Company, Licensee under Facility operating Licenses DPR-44 and DPR-56 for the Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3 requests that the Technical Specifications contained in Appendix A of the Operating License be amenoad by revising the following pages of Unit 3: iv, iva, 143, 144, 152, 152a, 164, 164a. 164b and 164c and, the following pages of Unit 2: 143, 144 and 164a. Revisions are indicated with ,

a vertical bar in the page margins.

This amendment reflects the results of material analyses conducted as part of the reactor coolant pressure l boundary materini surveillance program pursuant to 10 CFR 50, '

Appendix 0 and Appendix !!. The requestod changes wjll alter the reactor vessel pressure-temperature operating limitt, for Unit 3. .

Additionally, a curve for the bottom head limits i s b3ing added I to the PBAPS Units 2 and 3 Technical Specificatione.

Also included in this amendment in the proposed removal of the withdrawal schedule for the reactor vessel material specimens in accordance with guidance provided in Generic Letter 91-01 (" Removal of the Schedule for the Withdrawal of Reactor Vessel Material specimens from Technical Specifications").

Miscellaneous administrative changes are also proposed.

.Introductort.Teghnical_ Discussion R e v i s i o n_ t o_th e_P_r e s su r e - Te mp e r a_tu.r e_Cu rv e s A surveillance capsule was removed from the Peach Bottom Atomic Power Station Unit 3 reactor vesse) at the end of Fuel Cycle 7 (removed in June 1989). The capsule contained flux wires for neutron fluence measurement, and Charpy and tensile test specimens for material propecty evaluation. A combination of flux wire testing and compute analysis was used to establish the vessel peak flux location and magnitude. Charpy V-Notch impact testing and uniaxial tensile testing were performed to establish the material properties of the irradiated vessel beltline (core region).

The irradiation effects were projected in accordance with the guidance in Regulatory Guide 1.99, Revision 2,

" Radiation Embrittlement of Reactor Vessel Materials", to conditions for 32 ef fective full power years (EFPY) of operation.

The 32 EFH conditions are predicted to be less severe than the limits that would require vessel thermal annealing.

Pressure-temperature operating limits valid to 32 EFPY were 2 of 11

Docket Nos. 50-277

. 50-278 License Nos. DPR-44 DPR-56 developed in accordance with the July 1983 requirements of 10 CFR 50 Appendix G. The irradiation shift in nil-ductility transition temperature was accounted for in accordance with the guidance in Regulatory Guide 1.99, Revision 2. As recommended by the Regulatory Guide, the material property test results were not used to develop the operating limits: they will be used after the cecond set of specimens are tested. The results of the analyses show that the non-beltline limits are more severe than the beltline limits, even including predicted 32 EFPY shift.

The surveillance capsule withdrawal and test results discussed above were the subject of a technical report submitted to the NRC on June 27, 1990 (GE Nuclear Energy, SASR 90-50).

Based on the results of the test specimen analyses, Licensee requests several changes to the Technical Specifications which are discussed separately in the "Descriptivn of Proposed Changes".

The proposed changes to the thermal and pressurization limitations displayed in the new Unit 3 Figures 3.6.1, 3.6.2, and 3.6,3 were developed considering the most limiting conditions of the discontinuity regions and the irradiated beltline region in order to bound all operating conditions. The limiting regions of the vessel affecting the curve's shapes are the feedwater nozzles, bottom head and closure flange regions. Since the temperature of the bottom head can lag behind the rest of the vessel under certain non-nuclear heatup/cooldown situations, a second curve (BBH) for the bottom head limits, has been added to Figure 3.6.2 of Unita 2 and 3. Curve B on Figure 3.6.2 will be used for the feedwater nozzle and vessel flange limits whereas Curve B will be used for the bottom head Control Rod Drive (CRD) pNIe'tration limits. The predicted irradiation shifts for the beltline materials are lok enough that the beltline is not predicted to be limiting through 32 EFPY of operation.

The Unit 2 bottom head limits contained in curve B UU for Figure 3.6.2 referenced above are based on the Unit 2 surveillance capsule withdrawal and test results discussed in a technical report submitted to the NRC on May 13, 1988-(GE Nuclear Energy. SASR 88-24).

Removal of the Schedule for the Withdrawal of Reactor Vessel Material Spe_cinions_(Generic Letter 91-01)_

Included in this amendment request is a proposed administrative change to remove the schedule for withdrawal of reactor vessel material specimens from the Technical Specifications.

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Docket Nos. 50-277

. 50-270 License Nos. DPR 44 DPR-56

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'The 1 LAPS Unita 2 and 3 ourveillance requirements specify the withdrawM schedule for the reactor vessel material specimene. Racent.ly, the NRC approved a request to remove this t.chedule froro the Technical Specifications of the Joseph M.

Farley Nuclear Plant. The HRC has determined that the placement of this schedule in the Techn$ cal Specifications duplicates the control 1 on changes to this schedule that have been established by 10 CFR 50, Appendix H. Therefore, the staff concluded that, because this duplication is unnecessary, the removal of this Technical Specification schedule as a line-item improvement is .

consistent with the Commission Policy Statement on Technien1 I Specification Improvements.

This change is being done in accordance with the l guidance provided in Generic Letter 91-01 (" Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications").

This administrative change is discussed below as a Category 2 change. l DESCRIPTIO LOF_ PROP _OSED_ CHANGES The Category 1 changes are technical in nature and involve the reactor vessel pressure-temperature limits. The Category 2 changes are administrative.

C a.t.e go ry_1_ Change s1 A. Licensee proposes to replace the Unit 3 pressure-temperature limit curves in Figures 3.6.1, 3.6.2 and 3.6.3 (pagne 164, 164a, and 164b, respectively) with new curves which are based on the Unit 3 neutron flux surveillance specimen test results. The new curves for Unit 3 represent less restrictive operating limits than the current curves, but will stil.t provide sufficient margin to prevent brittle fracture of reactor coolant pressure boundary material.

These curves are valid to 32 EFPY, Also included on Figure 3.6.2 (page 164a) of Units 2 and 3 is a new second curve. B g. for the bottom head limits.

Since the temperature of Ehe bottom head can lag behind the rest of the vessel under certain non-nuclear heatup/cooldown situations, this second curve is being added to the Technical Specifications.

l B. Licensee proposes to delete Unit 3 Figure 3.6.4 (page 164c) which provides information on estimating-the shift in nil-ductility transition temperature-(RTNDT) relative to 4 of 11

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. -.. i Docket Nos. 50-277 50-378 License Nos. DPR-44 ,

DPR _ __ _-_ _ . _ _ . . . _ - _ _ . _ - - .

l fluence. This figure was for information only and did not.

establish any Technical Spacification requirement, i

1 C. Licensee proposes to reduce the Unit 3 minimum temperature of the vessel head flange and vessel head at which the head

bolting studs may be under tension (as stated in Specification 3.6.A 3, page 144). Cugrently. ths temperature must be " greater than 100 F"4 w"e propose that This change is -

the temperature must be " greater than 70 F .

recommended by the reactor vessel-supplier and is consistent l

with Standard Technical Specifications for General Electric

Boiling Watgr Reactors (BWR/5) NUREG-0123 Rev. 3. The I current 100 F limit is overly _ restrictive. Thg ASME Code to whigh

+60 ) the limit, vessel was built

• only required a.70 F (RTandthecurrentCode-permits '

limit.

_Cate_gorL 2_ Changes 2

A. Licensee proposes to delete Figure 3.6.4 -(" Transition Temperature Shift vs. Fluence") from the Unit 3 " List of Figures" (page iv). Additionally. Licensee proposes to delete the content of Unit 3 page iva which is duplicated at the bottom of page iv.

B. Licensee proposes to correct a sentence in Unit'3 3 Specification 3.6.A.2 by adding a comma. The proposed Specification states: - "The reactor vessel shall not be pressurized during heatup by.non-nuclear means, during.

cooldown following nuclear shut down or during low level =

physics tests..."

C. Licensee proposes to correct the " Applicability" portion of Unit 3 Specification 4.6 (page 143).. Currently, this ,

Specification makes reference to the " reactor cooling system". The-proposed Specification would. change this j wording to the " reactor coolant system".-

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1 ASME Boiler & Pressure Vessel Code,Section III, its; interpretations, and applicable requirements including 1965 Winter Addendum for Class-A vessels as defined therein.-

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Docket Mos. 50-277 50-278 License Nos. DPR-44 DPR-56 D. Licensee proposes to reword Units 2 and 3 Specification 4.6.A.2 (page 143) to more accurately describe the test specimens installed in the reactor vessel. Currently, this Specification states: " Test specimens of the reactor vessel base, weld and heat effected zone metal subjected to the highest fluence of greater than 1 Mev neutrons shall be installed..." The proposed Specification states: " Test specimene of the reactor vessel base, weld and heat affected zone metal were installed..." The existing words could be interpreted to mean that the installed specimens represent vessel locations that receive only the highest fluence and neutron energies greater than 1 Mev. In fact, the installed specimens represent vessel locations that receive a variety of neturon encrgies and fluence levels.

2. Licensee proposes to delete the last paragraph of Unit 3 Specification 3.6.A.2 (page 144) which states that Figures 3.6.1, 3.6.2, and 3.6.3 will be updated prior to nine (9) effective full power yeart. of operation. This Technical Specification amendment is updating Figures 3.6.1, 3.6.2 and 3.6.3 prior to nino ef fective full power years of operation.

Therefore, this Specification is no longer needed.

F. Licensee proposes to reword Unit 3 Specification 3.6.A.3 (page 144) to more accurately describe the vessel materinia and appurtenances involved. Currently, this Specification states "... the temperature of the vessel head flange and the head is..." The proposed Specification states "...the temperatures of the closure flanges and adjacent vessel and head materials are..."

O. Licensee proposes to replace the reference in the Unit 3 Specification 4.6.A.2 (page 144) from " neutron flux specimens" to " surveillance specimens", which in the more common term. The capsules contain specimens for material property evaluation in addition to the " neutron flux" wires.

H. Licensee proposes to remove from the Units 2 and 3 Specification 4.6.A.2 (page 144) the specimen withdrawal-schedule (identified as a footnote to Specification 4.6.A.2) and insert the proposed words "in accordance with 10 CFT 50, Appendix H". This change is in accordance with the guidance provided in Generic Letter 91-01 (" Removal of the Schedule for the Withdrawal of Reactor Vessel Materini Specimens for Technical Specifications").

I. Licensee proposes to add to Unit 3 Specification 4.6.A.2 (page 144) the words "and irradiation embrittlement" which clarify the purpose of testing the surveillance specimens.

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' - Docket Nos. 50-377  !

50-278 l License Hon. DPR-44 DPR-56

! J. Licensee proposes to remove from Unit 3 Specification-

! 4.6.A.2 (page 144) the words "... for Figure 3.6.4 " and i replace them with "... for Figures 3.6.1, 3.6.2, and 3.6.3,

! and the figures shall be updated based on the results."

i This change clarifies that the testing of the surveillance specimens is used to determine and update Figures 3.6.1.,

3.6.2, and 3.6.3.

i l K. Licensee also proposes that the Bases of Unit 3 Specifications 3.6.A and 4.6.A (pages 152, 152a) be revised i to provide current information nbout the surveillanceL

program. ,

i AhEEIY ASSESSMENT J CategoIY_1_9hangesi Section 4.2 of the Final Safety Analysis Report (FSAR) states that the safety design bmses of the reactor vessel and.

appurtenances are to " withstand adverse combinations of loadings- i l and fet:es resulting from operation under abnormn1 and accident e i conditions" and to " minimize the possibility of brittle fracturr,  !

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failure of the nuclear system process barrier." The revised thermal and pressurize. tion limits will not compromise these safety objectives because. they were developed in accordance with NRC Regulations and the latest NRC guidance, which-do support these safety objectives.  ;

The original analysis of the reactor vesselLmaterial-

• specimens in conjunction with the surveillance specimen program i ensures that the reactor pressure boundary will behave'in a non-brittle manner during-plant testing, startup. and opert. tion.  ;

The revised pressure-temperature limit curves were conservatively  ;

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i generated in accordance with the fracture-toughness requirements of 10 CFR 50, Appendix 0, as supplemented by Appendix G-to Section XI of the ASME Boiler and Pressure Vessel Code,- The proposed minimum allowable temperature-at which the heed bolting .

stude may be under tension is also--inLaccordance'with.10 CFR 50, . - -

Appendix G.as supplemented-by Appendix G to Section XI of the. (

ASME Boiler and Pressure Vessel Code. - The10F used to evaluate- l thenewpressure-temperaturelimitsforthe.bU9Ilinematerialwas-based on Regulatory Guide 1.99, Revision 2, which is the latest

guidance on RT determinations.  !

NDT l

The proposed changes to:theLthermal and pressurization limitations displayed in new Figures 3.6.1, 3.6i2, and 3.6.3 for -

Unit 3 were-developed considering the most limiting conditions of-L the discontinuity regions and the irradiated beltline region-in 7 of 11 m s-

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Docket Hos. 50-277 50-278 License Hos. DPR-44 i DPR-56

order to bound all operating conditions. The limiting regions of the vessel affecting the curve's shapes are the feedwater nozzles, bottom head and closure finnge regions. Since the temperature of the bottom head can lag behind the rest of the vessel under certain non-nuclear heatup/cooldown situations, Figure 3.6.2 for.both units contains a second curve, (B for the bottom head limits. Curve B on Figure 3.6.2 will b$g)u sed for the feedwater nozzle and vessel flange limits whereas Curve B

! will be used for the bottom head CRD penetration limits. CurhN i B was generated in accordance with the fracture toughness i rhhuiremuntsof10CFR50,.AppendixG, as supplemented by Appendix j G to Section XI of the ASME Boiler and Pressure Vessel Code. The

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predicted irradiation shifts for the beltline materials are low-enough that the beltline is not predicted to be limiting through 32 EFPY of operation.  :

The new curve for the bottom head limits will also meet the safety design basis of the reactor vessel _and appurtenances as cited previously in Section 4.2 of the Final Safety Analysis Report.

Category _2_Changest The Category 2 changes are administrative because'they do not impact plant equipment or systems, plant _ operations, or

! testing. These administrative changes will-improve:the Technical Specifications by correcting typographical-errors,-updating and improving terminology and information,-and deleting unnecessary material.

The removal of the schedule for the withdrawal of the reactor vessel material specimens would be considered the removal of unnecessary material. As stated in Generic Letter 91-01, the removal from the Technical Specifications of the; schedule for the withdrawal of reactor vessel material surveillance specimens'will not result in any loss of regulatory control because changes to this schedule are controlled by the requirements of-Appendix H to 10 CFR Part 50. In addition.--to ensure that the surveillance

-specimens are withdrawn at the proper time, theisurveillance requirements in the Technical Specification on pressure and temperature limits indicate that the: specimens shall be removed and examined to determine changes in their material preperties, as required by Appendix H. As stated in the Generic Letter, a request for a license amendment to remove this1 table from the Technical Specifications cay be made based-upon this guidance.

As also stated-in the Generic Letter, "the-licensee should commit to maintain the NRC-approved version of the specimen withdrawal

, schedule in the UFSAR." The PBAPS.UFSAR will1be revised to l incorporate the specimen withdrawal schedule.

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Docket Nos. 50-27'i 50-278 License Nos. DPR-44 DPR-56 Therefore, the above administrative changes are of no safety significance.

No SIGNIFICANT HAZARDS CONSIDERATIO}[ DETERMINATIONS

. Category _1 Changes:

The Category 1 changes requested herein do not involve a significant hazards consideration based on the foregoing Safety Assessment for the following reasons:

1) The_propAsesi_reyisions do not i_nvolve a sign $ficant_ increase _

in the probability _or conseglences of an_accideAt_ prey _ iou _ sly eyajuated because the revised thermal and pressurization limits prohibit conditions where brittle fracture of reactor vessel materials is possible. Consequently, there will be no increase in the probability or consequences of previously evaluated accidents since the primary coolant pressure boundary _.tegrity will be maintained as assumed in the safety design analyses.

The RT used to evaluate the new Unit 3 pressuYSI t emperature limits for the beltline material and the Units 2 and 3 bottom head limits was based on the guidance in Regulatory Guide 1.99, Revision 2, which is the latest guidance on RT determinations. The revised Unit 3 pressure-temperature Y9 bit curves and bottom head curve for Units 2 and 3 were conservatively 'Jenerated in accordance with the fracture toughness requirements of 10 CFR 50, Appendix G, as supplemented by Appendix G to Section XI of the ASME Boiler and Pressure Yessel Code. The proposed Unit 3 minimum allowable temperature at which head bolting studs may be under tension is also in accordance with 10 CFR 50, Appendix G, as supplemented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code.

Removal of Figure 3.6.4 is of no safety significance because it was for information only and is no longer appropriate.

11) The_pr_op_osesi_ rey _isi_ons do not create the_n.ossibi_lify of a new or different kind of_ accident from any_acc.ident previousl_y__ evaluated because the revised Unit 3 thermal and pressurization limits and the addition of the Units 2 and 3 bottom head curve do not create any new kind of operating mode or introduce any new potential failure mode.

Conditions where brittle fracture of-primary coolant pressure boundary materials is possible will be avoided by use of the revised and new curves.

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Docket Noo. 50-377 50-278 License Nos. DPR-44 twn. 56 iii) The_ proposed _reyisions_do_not_inyolve_a_pfgnificant

.r_educti_on._1.n_a_margA n_gf _ sa f e_ty because the proposed ,

pressure-temperature limits provide sufficient nafety margin. The revised Unit 3 pressure-temperature limits and the new Unite 2 and 3 head curves, were established in accordance with current regulations and the latest regulatory guidance on RT determinations. Although there issomereductioninsafehTmargin. operation within the new limits will ensure that the reactor vesnel materials will behave in a non-brittle manner and will remain conservative in that the original safety design bases will be procerved.

. Category _2 Changest The NRC provided guidance concerning the application of the standards for determining whether license amendments involve significant hazards considerations by providing examples in 51 FR 7751. Jn example (Example 1) of a change that involves no sianificant hazards considerations is "a purely administrative change to technical specifications for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature." The Category 2 changes requested herein conform to this example and do not involve a significant hazards consideration based on the foregoing Safety Assessment for the following reasons:

1) T_he_pr.oposed_teyisi_qns _ do noLinvolve a signifi_g_ ant _ ing. tease i n th e,.pr ohab i_11t.y_o r_c_ols equ e n c e s _qf a n a c c i d e n t_p.r.eyi ou s l y evaluated because they do not affect operations, equipment, or any safety-related activity. Thus, these administrative changes cannot affect the probability or_ consequences of any accident.

ii) .Tle_p.r op.o.s e d _tey i s i.on s_d o_no tt_g.r e a t e_t h e_ p_o s s i b i lity_ o f_ a new or_different kind of accident _from_any_apcident.

RreX19RB11_._eYAaluated because these changes are purely administrative and do not affect the plant. Therefore, these changes cannot create the possibility of any accident, iii) The proposed o royisions do not imyolve_a_significant reduction in a maggi.n of__ safe.).y because the changes do not affect any safety related activity or equipment. These changes are purely administrative in nature and increase the probability that the Technical Specifications are correctly interpreted by adding clarifying information.' deleting inappropriate information, and correcting errors. Thus, these changes cannot reduce any margin of safety.

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Docket Nom. 50-277 50-278 License Hon. DPR-44 DPR-56

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E!NIRQtMENTAL._IMPACLASSESSMENT An environmental impact ansonsmont is not required for the changes requested by this Application because the requested changes conform to the criteria for "actionn eligible for categorical exclusion" an specified in 10 CFR 51.22(c)(9). The requested changes have been shown by this Application not to adversely affect the objective of the primar" coolant pressure boundary to act ne a radioactive materia.1 barvier. The Application involves no significant hazards consideration as demonstrated in the preceding sectionn. The Application involves no nignificant change in the types or nignificant 2ncrease in the amounts of any effluents that may be released offeite, and there will be no significant increase in individual or cumulative occupational radiation exposure.

copcwatou The Plant operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question and will not endanger the henith ant' safety of the public.

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