ML20247G767
| ML20247G767 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/15/1989 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20247G771 | List: |
| References | |
| NUDOCS 8905310066 | |
| Download: ML20247G767 (15) | |
Text
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-10 CFR 50.90, g
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PHILADELPHIA ELECTRIC COMPANY-
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41 2301 MARKET STREET J
P.0, BOX 8699 PHILADELPHIA, PA.19101 (215)841-4000 i
May 15, 1989 Docket No. 50-277' License No. DPR-44 U.S. Nuclear Regulatory Commission ATTN: Document Control Desak Washington, D.
C.
20555
SUBJECT:
Peach Bottom Atomic Power Station, Unit 2 Technical Specifications Change Request
Dear Sir:
. Philadelphia Electric Company hereby submits Technical Specifications Change Request No. 88-07, in accordance with 10 CFR 50.90, requesting an amendment to the Peach Hottom Unit 2 Technical Specifications-(Appendix A) of Facility Operating License No..DPR-44.
Information supporting this Change Requ2st is' contained in to this letter, and the proposed replacement Technical Specifications pages are contained in Attactment 2.
The Company requests Technical Specification changes to modify the pressure-temperature limits for the reactor vessel.
If you have any questions regardinn this matter, please feel frce to contact us.
Very truly yours,
.0. Y,f.
p G.
A. Hunger, Jr Director l
Licensing Section Nuclear Support Division I
- Enclosure Affidavit
. Attachments 1 and 2
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T. Russell',' Administrator, Region I, T. P.. Johnson, USNRC' Senior Resident Inspector 1
R. E. Martin,cUSNRC PBAPS Project Manager.
'i T. M.. Gerusky, Director, PA' Bureau of F. ideological Protection I
T..E. Magette, State of Maryland J. Urbr' DC)marva Power:
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J, Public Service Electric & Gas H '.
C. Sch;.0;.an, Atlantic Electric i
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V COMMONWEALTH OF PENNSYLVANIA ss.
COUNTY'OF PHILADELPHIA
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J. Kowalski, being first duly sworn, deposes and says:
That he is Vice President of Philadelphia E;ectric Company, the i
Applicant herein; that he has read the enclosed request for amendment of Peach Bottom Unit 2 Facility Operating License No. DPR-44 (Change Request 88-07) and knows the contents thereof; and that the statements and matters set'forth therein are true and correct to the best of his knowledge, information and belief, bdW.
t Vice President Subscribed and sworn to before me this /fMday of-M 1989.
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Notary Public NOTMl!/1 SEAL SUSAN C, THonNTON. Nerv PuMe City of Philadel,Wa. Fida. C'can'y My Commbsion Expires Ma/ 4,1992
l' ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 Docket No. 50-277
. License No. DPR-44 TECHNICAL SPECIFICATIONS CHANGE REQUEST NUMBER 88-07
" Revision of Reactor Coolant Pressure Boundary Material Pressure-Temperature Limits" 1
Page 1 of 12
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Docket No. 50-277 l
r Licsnse No. DPR-44
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Philadelphia Electric Company, Licensee under Facility Operating. License DPR-44 for the Peach Bottom Atomic' Power Station Unit No.
2, requests that the Technical Specifications contained in Appendix A of the Operating License be amended by revising pages iva, j
143, 144, 152, 152a, 164, 164a, 164b and 164c (contained in Attachment
=
i 2).
The Figures on pages 164, 164a and 164b are to be replaced with new figures on the same pages.
The Figure on page 164c is being i
deleted and the page is being left blank.
Revisions are indicated I
with a vertical bar in the page margins.
These changes reflect the results of material analyses conducted as part of the reactor coolant pressure boundary material surveillance program pursuant to 10 CPR 50, Appendix G and Appendix H.
The requested changes will alter the reactor vessel pressure-temperature operating limits.
Miscellaneous administrative changes are also proposed.
INTRODUCTORY TECHNICAL DISCUSSION A surveillance capsule was removed from the Peach Bottom Atomic Power Station Unit 2 reactor vessel at the end of Fuel Cycle 7 (removed in May 1987).
The capsule contained flux wires for neutron fluence measurement, and Charpy and tensile test specimens for material property evaluation.
A combination of flux wire testing and computer 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).
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'f Docket No. 50-277 License No. DPR-44 The' irradiation effects, adjusted for consideration of the neutron flux wire specimen test results, wera projected in accordance with the guidance in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of' Reactor Vessel Materials", to conditions for~32
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effective full power years (EFPY) of operation.
The 32 EFPY conditions are predicted to be less severe than the limits that would j
require vessel thermal annealing.
Pressure-temperature operating limits valid to 32 EFPY were developed in accordance with the July 1983 requirements of 10 CPR 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
.q results were not used to develop the operating limits; they will be used after the second 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.
I J
i The surveillance capsule withdrawal and test results discussed above were the subject of a technical report submitted to a
the NRC on May 13, 1988 (GE Nuclear Energy, SASR 88-24).
Based on the results of the test specimen analyses, Licensee requests several changes to the Technical Specifications which are discussed separately below.
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Docket No. 50-277 License No. DPR-44 DESCRIPTION OF PROPOSED CHANGES The Category 1 changes are technical in nature and involve the reactor vessel pressure-temperature limits.
The category 2 changes are purely' administrative.
Category 1 Changes:
A.
Licensee proposes to replace the pressure-temperature limit curves in Figures 3.6.1, 3.6.2 and 3.6.3 (pages 164, 164a and 164b, respectively) with new curves which are based on the neutron flux surveillance specimen test results.
The new curves represent less restrictive operating limits than the current curves, but will still provide sufficient margin to prevent brittle fracture of reactor coolant pressure boundary material.
These curves are valid to 32 EPPY.
B.
Licensee proposes to delete Figure 3.6.4 (page 164c) which provides information on estimating the shift in nil-ductility transition temperature (RTNDT) relative to fluence.
This figure was for information only and did not establish any
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Technical Specification requirement.
Because the non-beltline regions of the vessel have been determined to be more limiting than the beltline region and non-beltline materials receive too little fluence to cause any shift in RTNDT, Figure 3.6.4 is not needed.
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Docket No. 50-277 jV License No. DPR-44 t
C.
Licensee proposes to reduce the minimum' temperature of the
. vessel head flange and vessel head at which the head bolting studs may be under tension (Specification 3.6.A.3).-
Currently, the temperature must be greater than 100 F;. we propose that 0
0 the temperature be greater than 70 F.
This change is recommended by the reactor-vessel supplier and is consistent with Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5), NUREG-0123, Rev.
3.
The current 0
100 limit is overly restrictive.
The ASME Code to which the vessel was built
- only required a 70 F (RTNDT 0
+600) limit, and the current Code permits an even lower limit.
Category 2 Changes:
A.
Licensee proposes to reword Specification 3.6.A.3 (page 144) to more accurately describe the vessel materials 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..."
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- ASME Boiler & Pressure Vessel Code,Section III, its t
interpretations, and applicable requirements including 1965 Winter Addendum for Class A vessels as defined therein.
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Docket No. 50-277 L
. g' License No. DPR-44 B.
Licensee proposes to replace the reference in Specification 14.6.A.2 (page 144) from." neutron flux specimens" to
" surveillance' specimens", which is the more common term.
The
. capsules contain specimens'for material property evaluation in i
addition to the " neutron flux" wires.
C.
Licensee proposes several administrative changes to reflect the fact that a' surveillance capsule has been removed and that specimens have been tested (page 144), delete references to Figure 3.6.4, which is proposed for elimination (pages iva, 144), add Figure 3.6.5 to the list of figures on page iva (because it was erroneously not added to the table previously
'when the figure was added to the Technical Specifications),
improve the format at the bottom of page 144 where the specimen removal schedule is provided, and correct minor typographical errors on page 143.
Licensee also proposes that the Bases of Specifications 3.6.A and 4.6.A (pages 152, 152a) be revised to provide current information about the surveillance program, and that several editorial corrections and improvements be made.
SAFETY ASSESSMENT Category 1 Changes:
Section 4.2 of the Final Safety Analysis Report (FSAR) states that a safety design basis of the reactor vessel and appurtenances is to " withstand adverse combinations of loadings and forces resulting from operation under abnormal and accident conditions."
Section 4.2 6 of 12
Dockat No. 50-277 License No. DPR-44 of the FSAR states that another safety design basis is to " minimize the possibility of brittle fracture failure of the nuclear system process barrier."
The revised thermal and pressurization 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.
Section 4.2 of the Updated FSAR states that (a) the initial ductile-brittle transition temperature of materials used in the reactor vessel are known; (b) expected shifts in transition temperature during design service life were determined and employed in the reactor vessel design; and (c) operation margins to be observed with regard to the transition temperature are designated for each mode of operation (designated by Figures 3.6.1, 3.6.2, and 3.6.3).
The original analysis of the reactor vessel material specimens i
in conjunction with the surveillance specimen program ensures that the reactor pressure boundary will behave in a non-brittle manner during plant testing, startup, and operation.
The revised pressure / temperature limit curves were conservatively generated in I
accordance with the fracture toughness requirements of 10 CFR 50,
(
l Appendix G, as supplemented by Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.
The proposed minimum allowable temperature at which the head bolting studs may be under tension is also in accordance with 10 CFR 50, Appendix G as supplemented by Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.
The RTNDT used to evaluate the new pressure / temperature limits for 7 of 12
' Docket No. 50-277
,. ;;+L License No. DPR-44 V
. the: beltline material was based on Regulatory Guide 1.99,. Revision 2, l
. hich is~the11atest guidance on RTNDT determinations.
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Category 2 Changes:
l The Category 2 changes are purely administrative because they do not impact' plant equipment or systems, plant operations'or testing,.or j
h
-management oversight.. These administrative changes will-improve.the a!
Technical Specifications by correcting typographical errors,--updating and improving terminology and information, and deleting unnecessary material.
Therefore,.these changes are of no safety significance.~
i SIGNIFICANT HAZARDS CONSIDERATION 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:
l
.1)
'The proposed revisions do not involve a significant increase in the probability or consequences of an accident previously evaluated 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 integrity j
will be maintained as assumed in the safety design analyses.
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i Docket No. 50-277 License No. DPR-44 i
.The RTNDT used to evaluate the new pressute/ temperature limits for the beltline material was based on the guidance in Regulatory Guide 1.99, Revision 2, which is the latest guidance on RTNDT determinations.
The revised pressure / temperature limit curves were conservatively generated in accordance with the fracture toughness requirements of 10 CFR 50, Appendir G, as supplemented by Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.
The proposed minimum allowable temperature at which head bolting studs may be under tension is also-in accordance with 10 CPR 50, Appendix G, as supplemented by Appendix G to Section III of the ASME Boiler and Pressure l
. Vessel Code.
Removal of Figure 3.6.4 is of no safety significance because it
.was for information only and is no longer appropriate.
ii)
-The proposed revisions do not create the possibility of a new or different kind of accident from any accident previously evaluated because the revised thermal and pressurization limits 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.
The proposed changes reduce the conservative margin that was incorporated into the development of the current limits.
The current limits have been shown by review of material characteristics, and by more recent and more accurate tests and analyses to be overly restrictive.
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p a4 Docket No. 50-277' gJ Licente No. DPR-44 I
The'pr6 posed revisions do not involve a significant reduction in i.
L a margin of' safety because the proposed pressure-temperature limits.still; provide sufficient saftty margin..The revised pressure / temperature limits, although less restrictive than the current limits, were established in accordance with current regulations and the latest regulatory guidance on RT NDT determinations.
Thus, the proposed changes merely reduce overconservative limits to acceptable limits.
Because operation will be within-these limits, the reactor vessel materials will behave in a non-brittle manner, thus, preserving the original safetyfdesign bases.
Category 2 Changes:
The NRC has provided guidance concerning the application of the standards for determing whether license amendments involve significant
. hazards considerations by providing examples in 51 FR 7751.. An I
example-(Example i) of a change that involves no significant 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:
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Docket No. 50-277 L
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- License No. DPR-44 p.
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The proposed revisions do not involve a significant increase in the probability or consequence of an accident previously 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)
The proposed revisions do not create the possibility of a new or different kind of accident from any accident previously evaluated because these changes are purely administrative and do not affect the plant.
Therefore, these changes cannot create the possibility of any accident.
iiii)
The proposed revisions do not involve a significant reduction in a margin of safety because the changes do not affect any safety related activity'or equipment.
These changea 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.
ENVIRONMENTAL IMPACT
-An environmental impact assessment is not required for the changes requested by this Application because the requested changes conform to the criteria for " actions eligible for categorical exclusion" as specified in 10 CPR 51.22(c)(9).
The requested changes have been shown by this Application not to adversely affect the 11 of 12
p Docket No. 50-277 License No. DPR-44 objective of the primary coolant pressure boundary to act as a radioactive material barrier.
The Application involves no significant hazards consideration as demonstrated in the preceding sections.
The Application involves no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and there w!.11 be no significant increase in individual or cumulative occupational radiation exposure.
CONCLUSION 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 significant hazards considerations and will not endanger the health and safety of the public.
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