ML20136G576
| ML20136G576 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/14/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20136G571 | List: |
| References | |
| NUDOCS 9703180098 | |
| Download: ML20136G576 (7) | |
Text
_
anco p
4 UNITED STATES l
,j NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 20065 4001
/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.188 T0 i
FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT N0. 1 4
DOCKET N0. 50-313
1.0 INTRODUCTION
By letter dated November 26, 1996, Entergy Operations, Inc. (the licensee) submitted a request to amend the pressure-temperature (P-T) limit curves in the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1).
Additional information was submitted by letters dated December 17, 1996 and March 4, 1997. The current P-T limit curves are valid for a service period of 15 effective full power years (EFPY). The current service period will end in March 1997. The amendment was intended to extend the validity of the AN0-1 P-T limit curves to 32 EFPY.
4 However, based on an initial NRC staff review of the submittals, a discrepancy was identified in the methodology used by the licensee for determining the standard deviation when calculating the reference temperature. To incorporate the more conservative conclusior.s that resulted from using the NRC staff proposed standard deviation, the licensee revised the TS amendment request to decrease the validity of the P-T curves from 32 EFPY to 31 EFPY. The revised amendment request, changing the curve validity to 31 EFPY, was transmitted by letter dated March 10, 1997.
The licensee's letters sent subsequent to the November 26, 1996, request to amend the P-T curves were letters transmitting clarifying information and details related to the methodologies for performing calculations. They did not change the initial proposed no significant hazards determination.
In conjunction with the requested amendment, the licensee requested an exemption from certain requirements of 10 CFR 50.60. The requested exemption was granted on March 12, 1997. The exemption permits the licensee to use the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Case N-514 to determine the safety margins associated with low temperature overpressure protection in lieu of the safety margins required by 10 CFR Part 50, Appendix G.
97031soo9e 97ogg4 PDR ADOCK 05000313 P
. - = -
j
- 4 i
j The staff evaluates the P-T limits based on the following NRC regulations and guidance:
10 CFR Part 50, Appendix G; Generic Letter (GL) 88-11; GL 92-01, l
Revision 1 (Rev.1); GL 92-01, Rev.1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2 (Rev. 2); and Standard Review Plan (SRP), Section 5.3.2.
GL 88-11 advised licensees that the staff would use RG 1.99, Rev. 2 to review P-T i
limit curves. RG 1.99, Rev. 2 contains methodologies for determining the j
increase in transition temperature and the decrease in upper-shelf energy j
(USE) resulting from neutron radiation. GL 92-01, Rev. 1,-requested that licensees submit their reactor pressure vessel (RPV) data for their plants to j
the staff for review.
GL 92-01, Rev. 1, Supplement 1, requested that i
licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. These data are used by the staff as the basis for d
the staff's review of P-T limit curves, and " the basis for the staff's i
review of pressurized thermal shock (PTS) assessments (10 CFR 50.61 assessments). Appendix G to 10 CFR Part 50 requires that P-T limit curves for the-RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical i
Engineers Boiler and Pressure Vessel (ASME) Code.
I' SRP 5.3.2 provides an acceptable method of calculating the P-T limits for i
ferritic materials in the beltline of the RPV based on the linear elastic
}
fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor K, which is a function of the stress state and flaw configuration. The imethods of Appendix G postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress. This flaw-is postulated to have a depth that is equal to one-fourth of the RPV beltline thickness and a length equal to 1.5 times the RPV beltline thickness.
1he critical locations in the RPV beltline region for calculating heatup and cooldown P-T limit curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the depth of the maximum postulated flaw, initiated and grown from the inside and outside surfaces of the RPV, respectively.
The Appendix G ASME Code methodology requires that licensees determine the adjusted reference temperature (ART or RT The ART is defined as the sum of the initial (unirradiated) reference t.1).emperature (initial RT
, the mean value of the adjustment in reference temperature caused by irraEa) tion and 7)luence factor., and a margin (M) term. The ART is a product of a chemistry factor (ARTaf The chemistry facIo'r is dependent upon the rount of copper and nickel in the material and may be determined from tfoles in RG 1.99, Rev. 2 or from surveillance data. The fluence factor is depend " t upon the neutron fluence at the maximum postulated flaw depth. The margin O rm is dependent upon whether the initial RT 7 is a plant-specific.or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Rev. 2 or surveillance data. The margin term is used to account for uncertainties in the values of initial RT copper and nickel contents, fluence and calculational procedures.
RGT.,99,Rev.2describesthe methodology to be used in calculating the margin term.
. 2.0 EVALUATION 2.1 LICENSEE EVALUATION OF P-T CURVES The licensee's P-T limit curves were calculated using an ART of 212*F for the 1/4T location for the limiting RPV beltline material, the upper to lower shell circumferential weld WF-112. Weld WF-112 was fabricated by Babcock & Wilcox (B&W) using a submerged arc process, Linde 80 flux and with heat number 406L44 weld wire. The ART was the sum of an initial RT.1 of -5*F, a margin value of 1
67'F and a ART,7 of 150'F. The initial RT is a generic value for B&W fabricated submerged arc welds with Linde M flux.
(This value has a standard deviation of 19.7 F). The ART was calculated using surveillance data from theB&WOwnersGroup(B&WOG)1"nlegratedSurveillanceProgram. Themarginterm was calculated using a standard deviation for the initial RT,7 of 19.7 F and a standard deviation for the ART., of 27*F.
The surveillance data used to calculate the ART were from surveillance weld materialsirradiatedinsurveillancecapsulesofANO-1,Oconee-1,RanchoSeco, B&WOG and Point Beach-2. All these surveillance welds were fabricated by B&W using the submerged arc process, Linde 80 flux, and using the same heat number i
of weld wire as used in the limiting weld in the ANO-1 RPV beltline. The reported copper content for the surveillance welds is 0.28 wt% for the ANO-1 weld, 0.32 wt% for the Oconee-1 and B&WOG welds, 0.31 wt% for the Rancho Seco weld, and 0.25 wt% for the Point Beach-2 weld. All the surveillance welds contained 0.59 wt% Ni. The licensee indicates that weld WF-112 in the ANO-1 reactor vessel beltline has a best estimate chemistry of 0.31 wt% Cu and 0.59 wt% Ni. The licensee calculated the chemistry factor from the surveillance weld data using the ratio precedure specified in RG 1.99, Revision 2, Position 2.1.
This procedure specifies that the measured values of ART be adjusted by multiplying the values by the ratio of the chemistry factorior the RPV weld to that for the surveillance welds. Then, using the adjusted ART values and their corresponding fluence, the chemistry factor was calculatId by multiplying each adjusted ART, by the corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors. Using these calculated / normalized :iata, the chemistry factor for the limiting ANO-1 RPV weld was calculated to be 185.6*F. The licensee determined that the standard deviation of the difference between the data was 26.FF (27af rounded to the nearest *g the best fit af the adjusted adjusted ART data and the curve representin i
i F). The licensee used this value of standard deviation in its calculation of the margin term.
The margin term is 67'F when it is calculated using the methodology in RG 1.99, Rev. 2, a standard deviation for the initial RT 7 of 19.7"F, and a standard deviation for the ART 7 of 27*F.
I I
1 I
i 4-2.2 STAFF EVALUATION OF P-T CURVES i
The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the reactor vessel of ANO-1. The amount of irradiation embr$ttlement was calculated in accordance with RG 1.99, Rev. 2.
The staff l
confirmed that the material with the highest ART at the 1/4T location at 31 EFPY for ANO-1 is the upper shell to lower shell circumferential weld i
WF-112.
To calculate the ART at the 1/4T location, the staff used the same values 'for initial RT and MT margin is., as the licensee; but used a margin value of i
68.5*F.
Thisvalueof7 calculated using the methodology in RG 1.99, Rev. 2, a standard deviation for the initial RT,7 of 19.7'F and a standard deviation for the MT,7 of 28*F.
is 28*F RG 1.99,-Rev. 2 indicates that the standard deviation for the MT* he standard F
when surveillance data are not credible. RG 1.99, Rev. 2 permits t deviation for the MT,7 to be reduced to 14"F when data are credible.
If data are to be considered credible, RG 1.99, Rev. 2 indicates that the scatter of the MT,7 values about the best-fit line, as described in Regulatory l
Position 2.1, should normally be less than 28*F for welds. There are 14 data i
points from the B&WOG Integrated Surveillance Program for welds that were fabricated using heat number 406L44 weld wire. The difference between the l
adjusted measured data and the curve representing the best fit of the adjusted l
data exceeded 28'F for four of the adjusted data points.
Hence, in accordance with RG 1.99, Rev. 2, the data is not credible and the standard deviation for the MT,7 should be 28'F.
l The staff does not believe the standard deviation for the MT, that is recommended in RG 1.99, Rev. 2 should be reduced based on the small number of l
data points (14) in the licensee's evaluation. The standard deviation in RG 1.99, Rev. 2 is a more meaningful value, since it was based on analysis of all l
surveillance weld data available at the time of the development of the RG.
Therefore, the staff has determined that by summing the initial RT,7 of -5'F, i
the margin value of 68.5'F and the MT of 150'F, the ART should be 213.5 F.
j SubstitutingtheARTof213.5'FforAS1intoequationsinSRP5.3.2,the staff determined that the proposed P-T limits for heatup, cooldown, j
hydrotest, and criticality meet the beltline material requirements in Appendix G of 10 CFR Part 50 for 31 EFPY.
(
In addition to beltline materials, Appendix G of 10 CFR Part 50 also impcses j
P-T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.A.2 of Appendix G states that when the pressure i
exceeds 20% of the pre-service system hydrostatic test pressure, the i
temperature of the closure flange regions highly stressed by the bolt preload i
must exceed the reference temperature of the material in those regions by at least 120*F for normal operation and by 90*F for hydrostatic pressure tests l
and leak tests. Based on the flange reference temperature of 60'F for ANO-1, the staff has determined that the proposed P-T limits satisfy the i
requirements in Section IV.A.2 of Appendix G.
i ij i
1 i
5
~e.
.,.-.n n
i*
I i !
1 2.3 EVALUATION OF FLUENCE j
l The proposed values of the fluence at the inside surface of the pressure j
vessel were developed by Framatome Technologies Incorporated (FTI). The methodology is based on radiation transport calculations for which the uncertainty range was estimated experimentally.
Plant specific measurements l
were used only to evaluate the range of uncertainty rather than to adjust the results of plant specific calculations. However, a benchmarking experiment j
was used to adjust the energy group values in the benchmarking calculation.
i These adjustments relate to the methodology and are: plant independent, dosimeter location independent and dosimeter type independent.
4 l
The calculational methodology included the following features:
Neutron Sources:
Power distribution was included pin by pin and the effect of burnup on the neutron sources was calculated.
Geometrical Model: Neutron transport was performed using two dimensional discrete ordinates transport in (r,0) and (r,z) geometry.
l Macroscopic Cross Sections: The BUGLE-93 cross section library was used which is based on the ENDF/B-VI data.
I Two Dimensional Transport: The DORT code was used with a P Legendre 3
polynomial expansion for scattering and an S for the quadrature expansion.
A total of 48 directions were used in a 1/8, core configuration.
C/M Ratios: Comparison of calculated to measured (C/M) quantities was made on dosimeter activities. The fluence was judged to be correct based on statistical comparisons of the measured. dosimeter activities to the corresponding calculated activities.
Three Dimensional Results: The (r,6) and (r,z) results were synthesized to produce an (r,8,z) distribution.
Best Estimate Fluence: A calculational bias was determined using a statistical combination of the calculated dosimeter activities and the corresponding measured activities. The bias was given in terms of energy group' adjustment constants, but the overall effect is about 5%.
For AN0-1 the results were compared to the benchmark bias and it was found that there was no significant bias associated with this analysis beyond that identified in the cavity dosimetry program.
We find the methodology described above, the results of the benchmarking and the ANO-1 results to be acceptable, because they conform to the staff's recommendations for fluence calculation which are included in the draft Regulatory Guide on pressure vessel fluence.
3.0 TECHNICAL CONCLUSION The staff concludes that the proposed P-T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality satisfy the requirements in Appendix G to Section XI of the ASME Code and Appendix G of 10 CFR Part 50 for 31 EFPY. The' proposed P-T limits also satisfy GL 88-11 because the method in RG 1.99, Rev. 2 was used to calculate the ART. Hence, the proposed P-T limits may be incorporated into the ANO-1 TSs.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR l
Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and 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 occupat fonal radiation exposure. The Commission has previously issued a proposeu finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 4346). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 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 the amendment.
6.0 CONCLUSION
The Commission has 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
(1) Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
(2) NUREG-0800, Standard Review Plan, Section 5.3.2:
" Pressure-Temperature Limits."
(3) Code of Federal Regulations, litle 10, Part 50, Appendix G, " Fracture Toughness Requirements."
(4) Generic Letter 88-11, "NRC Position c., Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," July 12, 1989.
(5) ASME Boiler and Pressure Vessel Code,Section XI, Appendix G for Nuclear Power Plant Components, Division 1, " Protection Against Non-ductile Failure."
1 i (6) November 26,1996,. letter from C. Randy Hutchinson, (Entergy Operations, Inc.) to USNRC Document Control Desk, subject:
" Arkansas Nuclear One, Unit 1 -Proposed Technical Specification Change To The Reactor Coolant System Pressure And Temperature Curves."
a (7) December 17, 1996, letter from D. C. Mims, (Entergy Operations, Inc.)
to USNRC Document Control Desk, subject:
" Arkansas Nuclear One, Unit 1
- Additional Calculations Supporting Unit 1 Pressure and Temperature Limit TS Change Request."
(8) March 10, 1997, letter from D. C. Mims, (Entergy Operations, Inc.) to USNRC Document Control Desk, subject:
" Arkansas Nuclear One, Unit 1 -
1 Technical Specification Change To The Reactor Coolant System Pressure And Temperature Curves."
l Principal Contributors: Meena Khanna Lambros Lois Date: March 14, 1997 k
i l
a
!