ML20042E944
| ML20042E944 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/01/1990 |
| From: | Sears P Office of Nuclear Reactor Regulation |
| To: | Papanic G YANKEE ATOMIC ELECTRIC CO. |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 TAC-71570, NUDOCS 9005040308 | |
| Download: ML20042E944 (5) | |
Text
4 e trog'o c
'g UNITED STATES O
['>>/
NUCLE AR REGULATORY COMMISSION n
5 WAsHINotoN, D. C. 20sb5
\\'v j*
Dock No.50-029 0
Mr. George Papanic, Jr.
Senior Project Engineer - Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398
Dear Mr. Papanic:
SUBJECT:
YANKEE ROWE REACTOR VESSEL INTEGRITY (TAC 71570)
The purpose of this letter is to inform you of staff concerns related to Yankee Rowe reactor vessel integrity.
As a result of the information that you provided at the meeting on March 15, 1990, the staff has several concerns which they would like to discuss with you.
A meeting has been arranged on Wednesday, May 9,1990 at 1:00 p.m.
Based on the reported unitradiated Charpy Upper Shelf Energy (USE) of 78 ft-lb for the beltline upper plate and using the methods described in Branch Technical Position MTEB 5-2 in SRP 5.3.2 and Regulatory Guide (RG) 1.99, Rev. 2, the Charpy USE in the transverse direction for the plate on January 1990 could be as low as 35.5 ft-lb.
Appendix G,10 CFR 50 requires beltline materials to have Charpy USE no less than 50 ft-lb.
This appendix indicates that in the event that the 50 ft-lb requirements cannot be met, the reactor vessel may continue to operate provided all of the following requirements are satisfied:
1.
A volumetric examination of 100 percent of the beltline materials that do not satisfy the requirements-of Section V.B of this appendix is made and any flaws characterized according to Section XI of the ASME Code and as otherwise specified by the Director Office of Nuclear Reactor Regulation.
2.
Additional evidence of the fracture toughness of.the beltline materials after exposure to neutron irradiation is to be obtained from results of supplemental fracture toughness tests.
3.
An analysis is performed that conservatively demonstrates, making appro-priate allowances for all uncertainties, the existence of equivalent margins of safety for continued operation.
The surveillance test results from the Yankee Rowe plate and the reference plate indicates that the increase in reference temperature resulting from neutron radiation may be greater for the Yankee Rowe reactor vessel beltline materials than that predicted by Table 2 in Regulatory Guide 1.99, Rev. 2.
The chemistry factors from Table 2 in RG 1.99, Rev. 2 for the Yankee Rowe surveillance plate and the reference surveillance plate are 90.1 and 107, i
respectively.
The chemistry factors for these plates using a sum of square method for fitting the surveillance data are 221.8 for the Yankee Rowe surveil-lance plate and 187.1 for the reference surveillance plate.
The greater the chemistry factor, the greater the increase in reference temperature.
Since the chemistry factors for the surveillance materials are substantially greater than the values in the regulatory guide, the reference temperature for the beltline h$
9
$O e
g
Mr. George Papanic, Jr. lower plate and circumferential weld could be greater than the PTS screening criteria in 10 CFR 50.61, prior to expiration of the Yankee Rowe license.
We would like to discuss your plans; (a) for complying with Appendix G,10 CFR 50, particularly a means for vessel inspection, and (b) for accurately determining the reference temperatures for the beltline lower plate and circumferential weld.
In addition, we would like to inform you of the criteria to be used in your analysis of the Yankee Rowe reactor vessel material, which is below 50 ft-lb.
The criteria to be used is contained in the Attachment.
These criteria are the ones sent to us by the Chairman of the ASME Subgroup Evaluation / Standards on November 20, 1989, with one modification, which we believe will appear in the Code.
The pressures are given in terms of P-ace, the " accumulated pressure,"
a term that comes from Section III, paragraphs NB-7311 and NB-7410, which require that the set pressure for expected system pressure transient conditions shall be such that the " total accumulated pressure" shall be no more than 10%
above design pressure.
By basing the criteria on P-ace, we allow a plant to take advantage of reduced operating pressure provided the pressure relief devices are set to correspondingly lower pressures.
For the nominal case, P-design is 2500 psig, P-operating is 2250 psig, and P-ace is at its maximum value of 1.10 P-design or 2750 psig.
The factors of 1.15 and 1.25 P-ace for the two criteria would call for pressures of 3162 and 3437 psig, respectively.
These are maximum pressures for P-design equal to 2500 psig.
The actual values will depend on the value of accumulated pressure given in the Overpressure Protection Report (NB-7200).
The NRC staff intends to prepare some guidance defining the credibility of J-R curves that are representative of the vessel material toughness, and defining what is meant by a " conservative bound J-Aa data" in the last sen-tence of the attached criteria statement.
Until that guidance is prepared, you should also be advised to consult with the staff as you proceed.
For additional information, please call C. Y. Cheng, Chief of the Materials and Chemical Engineering Branch.
Sincer ly, I,s Patric M. Sears, Project Manager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
As stated cc:
See Next Page DISTRIBUTION:
DocketJFile?
NRCPOR Local PDRs PDI-3 r/f MRushbrook PSears OGC (info)
ACRS (10)
Plant file RWessman
- SEE PREVIOUS %0NCURREN NAME :MRus@btool.
0FC :LA T
- PM: PDI-3*
- 0:DET*
- PDI-3*
i
- PSears/bc
- JRichardson
- RWessman :
DATE :05/1 /90
- 05/\\ /90
- 04/27/90
- 05/01/90 :
OFFICIAL RECORD COPY - Document Name:
YR REACTOR VESSEL INTEGRITY
Mr. George Papanic, Jr. I We would like to discuss your plans; (a) for complying with Appendix G, 10 CFR 50, particularly a means for vessel inspection, and (b) for accurately determining the reference temperatures for the beltline lower plate and circumferential weld.
In addition, we would like to inform you of the criteria to be used in your l
I analysis of the Yankee Rowe reactor vessel material, which is below 50 f t-lb.
l The criteria to be used is contained in the Attachment.
These criteria are the ones sent to us by the Chairman of the ASME Subgroup, Evaluation / Standards on November 20, 1989, with one modification, which we believe will appear in the Code.
The pressures are given in terms of P-acc, the " accumulated pressure,"
a term that comes from Section III, paragraphs NB-7311 and NB-7410, which require that the set pressure for expected system pressure transient conditions shall be such that the " total accumulated pressure" shall be no more than 10%
above design pressure.
By basing the criteria on P-ace, we allow a plant to take advantage of reduced operating pressure provided the pressure relief devices are set to correspondingly lower pressures.
t For the nominal case, P-design is 2500 psig P-operating is 2250 psig, and P-ace is at its maximum value of 1.10 P-design or 2750 psig.
The factors of 1.15 and 1.25 P-ace for the two criteria would call for pressures of 3162 and 3437 psig, respectively.
These are maximum pressures for P-design equal to 2500 psig.
The actual values will depend on the value of accumulated pressure given in the Overpressure Protection Report (NB-7200).
The NRC staff intends to prepare some guidance defining the credibility of J-R curves that are representative of the vessel material toughness, and defining what is meant by a " conservative bound J-aa data" in the last sen-tence of the attached criteria statement.
Until that guidance is prepared, you should also be advised to consult with the staff as you proceed.
For additional information, please call C. Y. Cheng, Chief of the Materials and Chemical Engineering Branch.
l Sincerely, Patrick M. Sears, Project Manager Project Directorate I-3 Division of Reactor Projects - I/II l
Office of Nuclear Reactor Regulation
Attachment:
As stated cc:
J. E. Richardson DISTRIBUTION:
Docket File NRC& Local PDRs C. Y. Cheng PDI-3 r/f MRushbrook PSears B. A. Boger OGC (info)
ACRS (10)
Plant file S. A. Varga RWessman i
W. D. Travers L. B. Marsh B. J. Elliot See next page
%D 9
0FC :LAh
- PM PDI-3 kDDE k
- PDI-k NAME :MRus
- PSears/bc
- JRicha ison
- RWessman :
- 04/)@R REACTOR VESSEL INTEGRITY DATE :04/4 0
- 04/J/90 3
- 0f/ / /90 :
OFFICIAL ECORD COPY - Document Name:
Y
y Mr. George Papanic, Jr. cc:
J. E. Richardson C. Y. Cheng B. A. Boger S. A. Varga W. D. Travers L. B. Marsh B. J. Elliot Dr. Andrew C. Kadak, President and Chief Operating Officer Yankee Atomic Electric Company 580 Msin Street Bolton, Massachusetts 01740-1398
+
Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street Boston, Massachusetts 02110 Mr. T. K. Henderson Acting Plant Superintendent Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 6
Resident Inspector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission Post Office Box 28 Monroe Bridge, Massachusetts 01350 t
Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street 7th Floor Boston, Massachusetts 02111 Mr. George Sterzinger Commissioner Vermont Department of Public Service 120 State Street, 3rd Floor Montpelier, Vermont 05602 Ms. Jane M. Grant Senior Engineer - PLEX Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398
L i
a *.
g ATTACHMENT i
i LOW UPPER-SHELF ENERGY (USE) CRITERIA 1
Level A and B Conditions i
For a postulated semi-ellipitical surface flaw with a/t = 0.25 and with an aspect ratio of 6 to 1 surface length, 1, to flow depth, a, and oriented along
[
the weld of concern, two criteria must be satisfied as described below.
Smaller flaw sizes may be used on an individual basis if a smaller size of the above postulated flaw can be justified.
1.
The crack driving force must be shown to be less than the material toughness as given below:
<0 applied 0.1 (1) where J is the J-intergral value calculated for the postulated applied flaw under pressure and thermal loading where the assumed pressure is 1.15 I
times accumulated pressure, and thermal loading using the plant specified l
heatup and cooldown conditions.
The parameter J is the J-intergral 0.1 characteristic of the material resistance to ductile tearing (J material)'
as usually denoted by a J -R curve test, at a crack extension of 0.1 inch.
g 2.
The flaw must be stable under crack growth as given below:
dJapplied < dJmaterial (2) da da i
at i
dapplied
- 0 material where J is calculated for the postulated flaw under pressure and thermal applied loading for all service level A and B conditions where the assumed pressure is 1.25 times accumulated pressure, P-ace, and thermal loading, as defined above.
i The J-intergral resistance versus crack growth curve shall reflect a conservative L
bound to J-Aa data representative of the vessel material under evaluation.
I i