ML20058N279
| ML20058N279 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/10/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20058N283 | List: |
| References | |
| NSD930928, NUDOCS 9312210313 | |
| Download: ML20058N279 (10) | |
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GENERAL OFFICE i
P.O BOX 499. COLUMBUS. NEBRASKA 6esu2-0499 I
Nebraska Public Power District
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l NSD930928 December 10, 1993 l
l U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C.
20555 Gentlemen i
Subject:
Proposed Change No. 119 to Technical Specifications Revision of Pressure - Temperature Limitation Curves Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46 In accordance with the applicable provisions specified in 10 CF2 50, the Nebraska Publ.4* Power District (District) requests that the Cooper Nuclear f
nical Specifications be revised as specified in the Station (CN s
attachment.
oposed changes revise the pressure vs. temperature t
operating limit curves for CNS based on the results of the testing and analysis performed as part of the District's reactor vessel material l
surveillance program required by 10 CFR 50 Appendix H.
l Accordingly, the attached contains a description of the proposed change, the l
attendant 10 CFR 50.92 evaluation, and the CNS Technical Specification pages revised by the institution of this change. This proposed change has been reviewed by the necessary Safety Review Committees and incorporates all amendments to the CNS Facility Operating License through Amendment 165 issued July 16, 1993.
By copy of this letter and attachment, the appropriate State of Nebraska of ficial is being notified in accordance with 10 CFR 50.91(b) (1). Copies to
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the NRC Region IV Office and the CNS Resident Inspector are also being sent in-accordance with 10 CFR 50.4 (b) (2).
should you have any questions or require any additional information, please contact me.
Sin erel,
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. Horn V
res.4. dent - Nuclear GRH/MJB 200037 Attachment cc:
H.R. Borchert Department of Health State of Nebraska NRC Regional Administrator Region IV Arlington, TX lh
[ 00 NRC Resident Inspector 9312210313 931210 i
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1 U. S. Nuclear Regulatory Commission Page 2 of 2 December 10, 1993 j
1 S'" ATE OF NEBRASKA) j
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NEMAHA COUNTY
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G. R. Horn, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this request on behalf of Nebraska Public Power District; and that the statem 'ts conta ned herein are true to the best of his knowledge and belief.
r AffA
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G. R. Horn
/o Subscribed in tny presence and.esorn to before me this day of MC -
1993 suunnumea W seat:
6 p jj SARf R.HANSEN K/ W f\\ c /2 # M Ny h D$ March 9.1997 NNARY PUBLIC 1
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4 Attachment to NSD930928 Page 1 of 7 PROPOSED CHANGE NO. 119 TO THE CNS TECHNICAL SPECIFICATIONS REVISION OF PRESSURE VS. TEMPERATURE OPERATION LIMITATION CURVES i
Revised Pares s
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132 147 154 155 156 I.
INTRODUCTION The Nebraska Public Power District (District) requests that the NRC approve the proposed changes to the Cooper Nuclear Station (CNS) Technical Specifications described below. The proposed changes revise the existing pressure-temperature operating limit curves (PT Curves) based on the results of the recent testing and analysis performed on CNS reactor vessel material specimens removed from the CNS vessel during the Reload 14, Cycle 15 Refueling outage.
Section 3.6.A of the CNS Technical Specifications,
" Thermal and Pressurization Limits," defines, through Figure Nos. 3.6.1.a. 3.6.1.b, and 3.6.2, the pressure and temperature boundaries within which CNS must be eperated to ensure adequate margin exists to prevent reactor vessel i
brittle fracture.
These PT Curves are generated, in part, based on the predictions of the change in Adjusted Reference Temperature (ART) of the reactor vessel limiting waterials a the reactor vessel neutron exposure increases with operating life.
The ART is the initial nil-ductility
- m. P us the expected shift in RTm due to the transition temperature, RT l
estimated neutron fluence received.
The current PT Curves are based on ARTS calculated in accordance with Regulatory Position 1.1 of Regulatory Guide 1.99, Revision 2, which provides a means of estimating reference temperature shift with less than two sets of plant specific Jurveillance data available. The proposed PT-Curves are based on ARTS calculated based on the guidance in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2, which adjusts the ARTS to account for the results of plant-specific curveillance data, where two or more sets of surveillance data are available.
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1 Att:chment to NSD930928 Page 2 of 7 The testing and analysis of thr; second set of CNS reactor vessel material surveillance specimens was recently completed.
The results of that testing and analyses is documented in GE Nuclear Energy Report No. GE-NE-523-159-12921' which we.s transmitted to the NRC by letter dated-February 25, 1993.2' The proposed PT Curves reflect the results of that survei Hance testing and analysis.
The following discussion describes the specific changes proposed, and the bases for those proposed changes.
Appendix A provides a mark-up of the proposed Technical Specifications changes (with the exception of the PT Curves), and Appendix B provides the revised Technical Specifications pages, including the revised PT Curves.
II.
DISCUSSION Regulatory Guide 1.99, Revision 2 provides a method acceptable to the NRC for predicting the effect of neutron radiation on reactor vessel materials required by Paragraph V.A of 10 CFR 50 Appendix G.
Because of the as scatter inherent to Charpy test data, Regulatory Guide 1.99, Revision 2 requires at least two sets of surveillance data be available before using the reactor-specific data to determine ART and the Charpy upper-shelf energy of reactor beltline materials. As a result of testing and analysis performed on the second set of CNS reactor vessel material specimens, the District now has two sets of reactor vessel material surveillance data for CNS.
Therefore, the District proposes to revise the CNS PT Curves to account for the results of its reactor vessel surveillance testing program which have been applied in accordance with NRC Regulatory Guide 1.99, Revision 2.
In accordance with the guidance of Regulatory Guide 1.99, Revision 2, Regulatory Position 2.1, the District adjusted the ART predictions based on a surveillance adjustment factor calculated in accordance with that guidance.
The result is a conservative estimate of ART for predicted reactor vessel fluence levels.
l The CNS Technical Specifications contain three PT Curves for operator use based on the corresponding application.
Figure 3.6 1.a provides the mir' mum vessel temperature vs. vessel pressure for non-nuclear heatup and for core cooldown following nuclear shutdown, and is valid through 32 i
EFPY. Figure 3.6.1.b provides tia minimum vessel temperature vs. vessel pressure for core operation (when the core is critical), and is also valid through 32 EFPY. Figure 3.6.2 provides the minimum vessel temperature vs.
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1.
GE Nuclear Energy Report No. GE-NE-523-159-1292, dated February, 1993,
" Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis."
2.
Imtter from G. R. Horn to NRC dated February 25, 1993, " Submittal of Reactor Vessel Surveillance Test Results."
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Attachment to-NSD930928-
'Page 3 of'7 l
vessel pressure for pressure t'ests such as that required by Section XI of
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the.ASME code. This figure provides five curves based on 15, 18, 21, 24, and 32 EFPY.
Five curves were. generated for Figure 3.6.2l to provide i
greater operational flexibility while performing system pressure ' tests,
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depending upon vessel exposure.
Each of these curves are' based on a
'i calculated ART for the limiting vessel material for a given reactor vessel.
fluence level, given in EFPY.
I Based on the new analysis, performed in accordance with the guidance of j
Regulatory Guide 1.99, Revision 2, the District has revised Figures 3.6.1.a. 3.6.1.b, and 3.6.2, and the corresponding Bases discussion. _
j The' specific changes are. described-below in Section III, " Description of Changes."
j III.
DESCRIPTION OF CHANGES i'
Page 132 - Section 3.6.A.2 is revised to clarify.that.this specification '
does not apply to inservice hydrostatic or pressure vesteel-l 1eak testing. ~Section 3.6.A.2 is also rearranged to_ provide' i
a more logical order of : discussion.
Additionally. section-3.6. A.3 is revised to change the Figure 3.6.2 curve i.eferences j
from 13, 18, & 21 EFPY to 15, 18, 21, ' 24 or 32 EFPY.. In 3
addition, "botton" is corrected to read:"botton."
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Page 147 - The 3/4.6 Bases section is revised to describe the basis for~
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l the revised PT Curves.
i Page 154 - Figure 3.6.1.a. " Minimum Temperature for Non-Nuclear Heatup or.
l Core Cooldown Following Nuclear Shutdown," - is - revised - to -
l reflect the new pressure vs. temperature requirements for this operational mode.
In addition, this figure's ? period of validity is revised from 21 to 32 EFPY, and the figure title i
is truncated to
" Minimum Temperature for Non-Nuclear
'1 Heatup/Cooldown.
l Page 155 - Figure 3.6.1.b,
" Minimum Temperature for Core -Operation-Includes 40*F Margin Required by 10CFR50 l
(Criticality)
Appendix G,"
is revised to reflect the new pressure. vs.
j temperature requirements for this operational mode, and the j
period of validity is revised from 21 to 32 EFPY.
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Additionally, the title for Figute ' 3.6.1.b is truncated - to j
read " Minimum Temperature for Core Critical Operation."
j Page 156 - Figure 3.6.2, " Minimum Temperature for Pressure Tests Such as Required by Section' XI,"
is revised to reflect-the new l
pressure vs. temperature requirements for this operational-i mode, and the period _of validity is changed from 13, 18,.& 21 EFPY to 15, 18, 21, 24, and 32 EFPY.
The title for Figure 3.6.2 is truncated to read'" Minimum Temperature for Pressure Tests."
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4 Attochment to NSD930928 Page 4 of 7 IV.
SIGNIFICANT HAZARDS DETERMINATION 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazards posed by the issuance of the amendment.
This evaluation is to be performed with respect to the criteria given in 10 CFR 50.92(c). The following analysis meets these requirements.
Evaluation of this Amendment with Respect to 10 CFR 50.92 The enclosed Technical Specifications change is judged to involve no significant hazards based on the following:
1.
Does the proposed change invrave a significant increase in the probability or consequences od an accident previously evaluated?
Evcluation The proposed revisions to the existing Cooper Nuclear Station (CNS)
Technical Specifications pressure vs. temperature operating limit significant increase in the curves (PT Curves) do not involve a probability or consequences of an accident previously evaluated.
The existing PT Curves, approved with Amendment No. 155 to the CNS operadng license, were developed based on Regulatory Guide 1.?9, Revisir n 2,
Regulatory Position 1.1, which provides a method acceptsble to the NRC for predicting Adjusted Reference Temperatures (ARTS) with less than two sets of surveillance data available. When the existing PT Curves were submitted and approved, only one set of reactor vessel surveillance data was available for CNS.
Since that time, the District has withdrawn, tested, and analyzed reactor vessel material samples to obtain a second set of surveillance data. Accordingly, the proposed PT Curves are based on ARTS for the limiting CNS reactor vessel materials which are derived from the results of the testing and analysis performed for the two sets of surveillance specimens withdrawn to date.
The predicted ARTS from which the proposed PT Curves are based have been calculated in accordance with the methodology dest.ribed in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2, which describes a means acceptable to the NRC for predicting ART shift when two or more sets of surveillance data are available.
The results of the testing and analysis of the second set of l
surveillance data, and the development of the proposed PT Curves are i
detailed in CE Nuclear Energy Report No. GE-NE-523-159-1292,21 3.
GE Nuclear Energy Report No. GE-NE-523-159-1292, dated February, 1993,
" Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis."
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f Attcchment to NSD930928 l
Page 5 of 7 j
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l which was transmitted to the NRC by letter dated _
February 25, 1993.F This testing and analy.= h resulted in predicted ARTS for the limiting CNS reactor vessel beltline materials as shown in the following table.
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1 EFFECmVE FULL FOWER YEARS AEUUSTED REFERENCE TERIFERATURE GEFPY)
(F) 15 80 i
t 1s os I
21 101 24 108 32 128 Based on the new ART predictions for the CNS limiting reactor vessel materials, new PT Curves have been generated. These include. Figure
.l 3.6.1.a.
" Minimum Temperature for Non-Nuclear Heatup/Cooldown,"
Figure 3.6.1.b, " Minimum Temperature for Core Critical Operation,"-
and Figure 3.6.2,
" Minimum Temperature for Pressure Tests."
Additionally, five separate curves are plotted in Figure 3.6.2 to provide operational flexibility when performing pressure tests, t
based on reactor vessel fluence level.
These curver, are based on fluence levels predicted for 15, 18, 21, 24, and 32 EFPY, with corresponding ARTS as shown in the above table.
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The proposed revision to the CNS PT Curves' is based on an NRC-accepted means of ensuring protection against brittle reactor vessel failure, and compliance with 10 CFR Appendix G will be maintained.
1 Therefore, this proposed change will not involve a significant increase in the prob.ibility or consequences of an accident j
previously evaluated, i
2.
Does the proposed change create the possibility for a new or different kind of accident from any accident previously evaluated?
Evaluation The proposed changes update the CHS reactor vessel pressure vs.
temperaturn operating limits to account for the results of the 4.
Letter from G. R. Horn to NRC dated February 25, 1993, " Submittal of Reactor Vessel Sutveillance Test Results."
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Attechment to 1
NSD930928 1
Page 6 of 7 1
testing and analysis of the second set of CNS reactor vecsel l
surveillance specimens. This analysis was performed in accordance with Regulatory Guide 1.99, Revision 2, and therefore, corresponds with the current NRC guidance.
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The proposed changes do not involve any plant design changes nor any new mode of operation.
These changes ensure compliance with the
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brittle fracture prevention requirements of 10 CFR 50 Appendix G, and therefore do not create the possibility for a new or different kind of accident from any accident previously evaluated.
3.
Does the proposed change create a significant reduction in the j
1 margin of safety?
Evaluation i
i The proposed changes to the CNS PT Curves do not create a significant reduction in the margin of safety. The proposed changes
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revise the existing CNS PT Curves in accordance with the. results of ;
the testing and analysis of the second set of reactor vessel surveillance specimens.
This analysis has been ' performed in accordance with the recommendations of _ Regulatory Guide.1.99,'
Revision 2, the current NRC guidance given to ensure compliance with -
10 CFR Appendix G.
The proposed revisions to the CNS PT - Curves represent a ' slight reduction in predicted ARTS from that assumed in the existing PT
- Curves, and therefore, a
slight reduction in temperature requirements for the PT Curves plotted in Figure 3.6.2, " Minimum Temperature for Pressure Tests."
This reduction is. due to improvements in the reactor vessel neutron flux distribution modeling which demonstrated an axial variance, and resulted in a.
reduction in the predicted fluence for the limiting reactor vessel beltline plate.
However, the amount of shift predicted, based on the reactor vessel surveillance material adjustment determined in accordance with the guidance of Regulatory Guide 1.99, Revision 2, is greater than that predicted using Regulatory Guide 1.99, Revision 2 methods without surveillance adjustment. Therefore, while the net change to the ARTS and accordingly, the PT Curves in Figure 3.6.2 has been a reduction in temperature requirements, adequate margin is added to these predictions as described in Regulatory Position 2 1 of Regulatory Guide 199, Revision 2.
Therefore, impleme<2tation of the proposed PT Curves in Figure 3.6.2 will not ropresent a significant reduction in the margin of safety.
The minimum reactor vessel temperature requirements for nea-nuclear heatup/cooldown and core critical operation given in Figures 3.6.1.a and 3.6.1.b, respectively, are actually higher than the existing PT Curves, as these two figures were developed to be valid through 32 EFPY.
Additionally, as stated above, these curves were developed based on the same ART predictions discussed previously, and
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Attcchment to NSD930928 Page 7 of 7 accordingly, are in accordance with the current NRC guidance.
Therefore, implementation of these PT Curves will not represent a significant reduction in the margin of safety.
Details of the testing and analysis discussed above is provided in GE Report No. GE-NE-523-159-1292 ' submitted to the NRC by letter 1
dated February 25, 1993.i' V.
CONCLUSION The District has evaluated the proposed changes described above against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has determined that this proposed change will n21 1) involve a significant increase in the probability or consequenceu of an accident previously evaluated, 2) create the possibility for a ne. or different kind of accident from any accident previously evaluated, or 3) create a significant reduction in the margin of safety.
Therefore, for the reasons detailed above, the District requests NRC approval of Proposed Change No. 119, s
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