ML20083L228
| ML20083L228 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/08/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20083L224 | List: |
| References | |
| NUDOCS 9505180206 | |
| Download: ML20083L228 (6) | |
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u UNITED STATES
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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20565-0001
.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.106 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS. INC.
WATERFORD STEAM ELECTRIC STATION. UNIT 3 DOCKET NO. 50-382
1.0 INTRODUCTION
By application dated December 14, 1993, as supplemented by letter dated March 3, 1995, Entergy Operations, Inc. (the licensee), submitted a request for changes to the Waterford Steam Electric Station, Unit 3, Technical Specifications (TSs). The requested changes would revise pressure-temperature (P-T) limits in the Waterford Unit 3 TSs. The licensee revised the P-T limits for the unit and requested to extend the applicable period of the P-T limits from a current eight effective full-power years (EFPY) to 15 EFPY. The requested changes will also remove the reactor vessel material specimen withdrawal schedule from the TSs.
In addition, the licensee has requested to revise the low temperature overpressure protection enable temperature value and the related TSs.
2.0 BACKGROUND
Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to state TSs to be included as part of the license. The Commission's regulatory requirements related to the content of TSs are set forth in 10 CFR 50.36. That regulation requires that the TSs include items in five specific categories, including (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. However, the regulation does not specify the particular requirements to be included in a plant's TSs.
The Commission has provided guidance for the contents of TSs in its " Final Policy Statement on Technical Specifications Improvements for NucFefer Pwer Reactors" (" Final Policy Statement"), 58 FR 39132 (July 22,1993), in which I
the Commission indicated that compliance with the Final Policy Statement satisfies Section 182a of the Act.
In particular, the Commission indicated that certain items could be relocated from the TSs to licensee-controlled documents, consistent with the standard enunciated in Port 7and General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).
In that l
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case, the Atomic Safety and Licensing Appeal Board indicated that " technical specifications are to be reserved for those matters as to.which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."
Consistent with this approach, the Final Policy Statement identified four criteria to be used in determining whether a particular matter is required to be included in the TSs, as follows:
(1). installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to.the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient i
that either assumes the failure of or presents a challenge to the integrity of' a fission product barrier; (4) a structure, system, or component which operatingexperienceorprobabilisticsa[etyassessmenthasshowntobe significant to public health and safety.
As a result, existing TSs requirements which fall within_ or satisfy any of the criteria in the Final
' Policy Statement must be retained in the TSs, while those TS requirements which do not fall within or satisfy these criteria may be relocated to other, licensee-controlled documents.
3.0 EVALUATION A. Reactor Coolant System P-T Limits The staff evaluates the P-T limits based on the following NRC regulations and guidance: Appendix G to 10 CFR Part 50; Generic Letters (GLs) 88-11 and
.l 92-01; Regulatory Guide (RG) 1.99, Rev. 2; and Standard Review Plan (SRP)
Section 5.3.2.
Appendix G to 10 CFR Part 50 requires that P-T limits for the reactor vessel must be at least as conservative as those obtained by i
Appendix G to Section III of the ASME Code. GL 88-11 requires that licensees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron i
irradiation by calculating adjusted reference temperature -(ART) of reactor vessel materials.
The ART is defined as the sum of initial nil-ductility caused by neutron irradiation, and a 1) of the material, the increase in RT.,
transition reference temperature (RT margin to account for uncertainties in 1
The Commission recently promulgated a proposed change to 650.36, pursuant to which the rule would be amended to codify and incorporate these criteria (59 FR 48180, September 20,1994).
The Commission's Final Policy l
Statement specified that LCOs for Reactor Core Isolation Cooling, Isolation Condenser, Residual Heat Removal, Standby Liquid Control, and Recirculation Pump Trip are included in the TS under Criterion 4 (58 FR 39132). The Commission has solicited public comments on the scope of Criterion 4, in the pending rulemaking.
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the prediction method.
The increase in RT is calculated from the product of a chemistry factor and a fluence factor,.n The chemistry factor is dependent upon.the amount of copper and nickel in the vessel material. GL 92-01 requires licensees to submit reactor vessel materials data, which the staff will use in the review of the P-T limits submittals.
SRP 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified in Appendix G to Section III of the ASME Code.
For the Waterford Unit 3 reactor vessel, the licensee determined that lower shell plate M-1004-2, is the limiting material at the 1/4T location and 3/4T locations (T = the thickness of the reactor vessel beltline). The licensee calculated an ART of 65.4*F at the 1/4T location and 54.0*F at the 3/4T location. The calculation variables are shown in the attached table.
The staff verified the copper and nickel contents and initial RT of plate M-1004-2withrespecttotheNRCreactorvesselmaterialsubmittIdbythe licensee in response to GL 92-01. The staff also reviewed the fluence value used by the licensee in the revision of the P/T curves which is discussed 3
below.
The proposed revisions were based on a ABB Combustion Engineering report C-MECH-ER-021, Rev.00 which was attachment C to December 14, 1993 letter.
This report used the fluence values from surveillance capsule report BAW-2177,
" Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station, Unit No. 3," for the evaluation of capsule W-97.
The W-97 capsule report was submitted to the NRC by licensee's letter dated November 25, 1992.
Capsule W-97 was irradiated in Cycles 1 through 4 for 4.44 EFPY of operation, at a location 7 degrees off a major horizontal axis. The analytical methodology utilized the DOT-IV two dimensional code in (r,6) geometry with ENDF/B-IV based cross sections. AP angular scattering and a S angular j
quadrature approximations were used.3 Pin wise neutron sources were used which a
were provided by the licensee. However, the capsule report normalized the calculated values to the measured dosimeter activity, but no attempt was made to determine the associated uncertainty.
In addition, the initial requested extension was about 16 EFPY, (which is much longer than the average extension request) based on one capsule measurement for which the uncertainty was not estimated. The staff requested the licensee to justify the extension in terms of the potential uncertainty by letter dated January 11, 1995. However, in lieu of estimating the uncertainty the licensee elected to lower the period of applicability to 15 EFPY, while the fluence estimate remained unchanged i.e.,
for 20 EFPY.
The estimated fluence at 20 EFPY is about 25 percent higher than the estimated fluence at 15 EFPY.
The 25 percent compensate for the potential error due to the normalization of the estimate to the measured dosimetry.
We find the proposed period of performance conservative and thus, acceptable.
The staff performed an independent calculation of the ART for the reactor vessel materials using RG 1.99, Revision 2.
Based on the staff's calculation of the ARTS values, the staff is in agreement with the licensee's calculations.
3i Substituting the limiting ARTS in the attached table into equations in SRP 5.3.2, the staff verified that the proposed P-T limits for heatup, cooldown, criticality, and inservice hydrostatic test satisfy the requirements in Paragraphs IV.A.2 & IV.A.3 of Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload 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 and leak tests.
Based on the flange RT # of 20*F, the staff hasdeterminedthattheproposedP-TlimitshavesatisIiedtherequirementfor the closure flange region during normal operation, hydrostatic pressure test and leak test.
Based on above discussion, the staff concludes that the proposed P-T limits for heatup, cooldown, inservice hydrostatic test, and criticality are valid for 15 EFPY because the limits conform to the requirements of Appendix G of 10 CFR Part 50 and GL 88-11. Hence, the proposed P-T limits may be incorporated in the Waterford Unit 3 TSs.
Changes to the text of the TSs which related to P-T limits were editorial in nature, and found to be acceptable alterations.
B. Removal of Reactor Vessel Material Specimen Withdrawal Schedule from TSs As a part of the amendment, the licensee proposed to remove the surveillance capsule withdrawal schedule from the Waterford Unit 3 TSs.
In 1991, the staff issued Generic Letter 91-01 to allow the removal of the withdrawal schedule from the TSs provided that the withdrawal schedule is relocated in a controlled document such as UFSAR and the schedule has satisfied Appendix H to 10 CFR Part 50. The licensee stated that the withdrawal schedule is currently in the Waterford UFSAR. The staff has verified that the withdrawal schedule in the UFSAR satisfies Appendix H to 10 CFR Part 50; therefore, the staff concludes that the removal of the surveillance capsule withdrawal schedule from TSs is acceptable.
On this basis, the staff concludes that these requirements do not need to be controlled by TSs, and changes to the surveillance capsule withdrawal schedule, which will be described in the UFSAR, will adequately be controlled by Appendix H to 10 CFR Part 50. The staff has concluded, therefore, that relocation of the surveillance capsule withdrawal schedule described above is acceptable because (1) their inclusion in TSs is not specifically required by 10 CFR 50.36 or other regulations, (2) the withdrawal schedule requirements are not required to avert an immediate threat to the public health and safety, and (3) changes to these withdrawal schedule requirements will require prior NRC approval in accordance with Appendix H to Part 50.
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C.
Change In Low Temperature Overpressure Protection Enable Temperature The licensee has elected to change the low temperature overpressure protection (LTOP) enable temperature from 285* F to 272* F.
The calculations were performed such that the new temperature will protect the Appendix G limits.
In addition, it complies with Reactor Systems Branch Technical Position 5-2, l
Rev. 3, in that the. enable temperature is defined as the water temperature i
90* F at the beltline corresponding to a metal temperature of at least RT,g +ix location (1/4t or 3/4t) that is controlling the Append G limit. Therefore, we find the new enable temperature of 272* F to be acceptable.
. The changes in TSs 3.4.1.3, 3.4.1.4, 3.4.8.3 and Bases, B 3/4 4-10 paragraph 4, involve the change of the LTOP enable temperature from 285* F to 272* F.
Because the enable temperature change has been found acceptable we find the proposed TS changes both necessary and acceptable.
3.0 STATE CONSULTATION
l In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 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 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 occupational radiation exposure. The Commission has previously issued a pro-posed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (59 FR 2867).
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.
5.0 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.
Attachment:
Table Principal Contributors:
L. Lois C. Fairbanks Date:
May 8, 1995 4
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b 1 Table ADJUSTED REFERENCE TEMPERATURE FOR REACTOR VESSEL LIMITING MATERIALS FOR PRESSURE-TEMPERATURE LIMITS 15 EFFECTIVE FULL POWER YEARS 4
WATERFORD UNIT 3 Limiting Material:
Cu Ni Fluence ART,7 Initial Margin ART RT.7 Lower Shell Plate M-1004-2 Waterford Unit 3 (1/4)T
.03
.58 1.36E19 21.7 22 21.7
-65.4 (3/4)T
.03
.58 0.48E19 16.0 22 16.0 54.0 t
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