ML20127H136
| ML20127H136 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/15/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127H135 | List: |
| References | |
| NUDOCS 9301220216 | |
| Download: ML20127H136 (6) | |
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[gmagDo UNITED STATES l
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3 NUCLE AR REGULATORY COMMISSION nj WASHINGTON,0. C. 20555
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$AFETY EVALVATION BY THE OFFICE OF NVCLEAR REACTOR REGULATION MLATED TO AMENDMENT 169 TO FACIllTY OPERATING LICENSE NO. DPR-Qf DV0VESNE LIGHT COMPANY OHIO EDIS0N COMPANY Pf1tiSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 D_0GET N0. E0-334 0
1.0 J tR0fj)U ION The reactor coolant system (RCS) pressure-temperature (P/T) limits during plant heat-up and cooldown for Beaver Valley Power Station, Unit No. 1 are specified in Technical Specification (TS) Figures 3.4-2 and 3.4-3.
The requirements for the low temperature overpressure protection (LTOP) system are specified in TS 3.4.9.3.
An associated restriction on the start-up of a reactor coolant pump is specified in TS 3.4.1.6 and is consistent with-the assumptions used in the analysis supporting the LTOP setpoint.
By letter dated July 15, 1991 (Ref. 1), September 30, 1991 (Ref. 2),
December 13,1991 (Ref 3), and November 2,1992 (Ref. 4), Duquesne Light Company proposed amendments to the P/T curves in TS Figures 3.4-2 (for heatup) and 3.4-3 (for cooldown) and changes in the LTOP system setpoints. The proposed changes replace the current heatup and cooldown curves applicable to 9.5 effective full power years (EFPY) of operation, with curves applicable to 16 EFPY.
The new curves were developed based on the guidance provided in Revision 2 of Regulatory Guide 1.99.
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations," recommends RG 1.99, Rev. 2, be used in calculating P/T limits, unless the use of different methods can be justified.
The proposed changes to TS 3.4.1.f add a more stringent limitation by requiring both conditions 1 and 2 of that TS to apply, rather than just one of them.
In addition, the action statement in TS 3.4.1.6 proposes to clarify the action to be consistent with the stated LC0 conditions 1 and 2.
The supplemental letters provided minor revisions to the application but they do not change the initial proposed no significant hazards consideration determination and do not expand the scope of the original Federal Reaister notice.
2.0 EVALVATIQL{
Materials Considerations To evaluate the P/T limits, the staff uses the following NRC regulations and guidance:
Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Rev. 2; NVREG-0800 Standard Review Plan (SRP) Section 5.3.2; and Generic Letter (GL) 88-11.
9301220216 930115 PDR ADOCK 05000334
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e Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular,10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications.
The P/T limits are among the limiting conditions of operation in the Technical Specifications for all commercial nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits.
An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materialt in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50.
Appendix H, in turn, refers to ASTM Standards.
These tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy Upper Shelf Energy (USE). GL 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev.
2, to predict the effect of neutron irradiation on reactor vessel materials.
This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor-vessel. Appendix H refers to the ASTM St_andards which, in turn, reg'uire that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Beaver Valley Unit. reactor vessel.
The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 16 EFPY at 1/4T (T = reactor vessel beltline _ thickness) was the lower shel'1 plate 86903-1 with 0.20% copper (Cu), 0.54% nickel (Ni), and an initial RT,
of 27'F.
The material with the highest ART at 16 EFPY at 3/4T was the intermediate shell plate (B6607-2) with 0.14% Cu, 0.62% Ni, and an initial RT, of 73*F.
The licensee has removed three surveillance capsules-(V, U, and W) from Beaver Valley Unit 1 (References 5, 6, 7, 8, 9).
The results from capsule V were published in Westinghouse Report WCAP-9860; the results from capsule U in WCAP-10867; and the results from capsule W in WCAP-12005.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
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' For the limiting beltline material at 1/4T, plate B6903-1, the staff calculated the ART at 16 EFPY to be 226.0'F.
For the 3/4T limiting material, plate B6607-2, the staff calculated }he ART at 16 EFPY to be 188.7'F.
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staff used a fluence of 1.32E19 n/cm for 1/4T and 5.llE18 n/cm for 3/4T.
The ART of plate B6903-1 was determined by Section 2 of RG 1.99, Rev. 2, because that material was in the surveillance capsules.
The ART for plate B6607-2 was calculated per Section 1 of RG 1.99, Rev. 2, because it was not in the surveillance capsules.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 224*F at 16 EFPY at 1/4T and 188'F at 3/4T for the same limiting plate materials.
The staff judges that a difference of 2*F between the licensee's ART of 224*F and the staff's ART of 226'F at 1/4T is acceptable.
Substituting the ART of 226*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes
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P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.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 reference temperature of 60*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
LTOP Considerations In the current TS, when the temperature of one or more of the non-isolated RCS cold legs is less than or equal to 292*F, LTOP is provided by either of the I
pressurizer power operated relief valves (PORVs) with a lif t setting of less than or equal to 444 psig or an RCS vent greater than or equal to 3.14 square inches.
The operability of a PORV or an RCS vent of greater than or equal to 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS cold leg temperature is less than or equal to 292 F.
Each PORV has adequate relieving capacity to protect the RCS against P/T limits when the transient is limited to either (1) the start of an idle RCP with secondary water temperature in the steam generator less than or equal to 25'F above the RCS cold leg temperature or (2) the mass addition transient due to one centrifugal charging pump in operation with maximum charging flow into the RCS, assuming the letdown line is isolated.
By letter dated July 15, 1991, the licensee proposed a change in the PORY setting from 444 psig to 432 psig. The new PORV setting is based on the results of an analysis corresponding to the 16 EFPY P/T curves with a correction of 4 psig to address a non-comervatism identified in the new steam y
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. generator tube plugging analysis.
The licensee also proposed changing the LTOP enable temperature from 292 F to 314 F.
However, after reviewing tbt
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information submitted by the licensee, the staff cencluded that the newly proposed enab'le teinperature was not sufficiently conservative since the temperature difference between the bulk fluid and the vessel metal was not factored into the licen2en's analysis, in response to a staff request to consider this effect, the licensee, in its letter dated Noveraber 2.1992, e
proposed a revised enable temperature of 329 F.
This enable temperatura was calculated by the licensee's contractor, Westinghouse, as described in ti.pical r
re. port fDR1-SRPt0-266(92) " Methodology for calculating Enable Temperature set Foint for Beaver Valley Unit 1," dated Septeinber 1992.
The staff has reviewed the Westinghouse methodology and concluded that it leads to the most y
conserYative enable temperQture WhjCh is:
the reactor coolant fluid temperature corresponding to vessel that is controlling in the 10 CFkg+90*F at the beltline location metal temperature of at least RT Part 50 Appendix G fracture toughr.ess limit calculations.
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1he licensee's proposed changes in tethr.ical specifications 3.4.1,6, 3.4.9.3 and associated bases sections reflect the TS changes discussed above.
The k
staf f finds that the changes are based on applicable regulatory guidance in Standard Review Plan 5.2.2 (Revision 2), are reasonably conservative and are acceptable.
5 in addition, the staff concludes that the proposed P/T limits for the reactor 1
coolant system for heatup, cooldown, leek test, and criticality are valid i
through 16 EFPY because the limits canfor.n to the requirements of Appendices G and H of 10 CFR Part 50, The proposed limitt, also satisfy Generic letter 88-11 because the licensee used the method in RG 1.99 Rev. 2, to calculate the ART. Hence, the ;>roposed P/T limits may be incorporated into the Beaver Valley Unit 1 Technical Specifications.
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- 3. O STATE CONSIDERAT MN
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In accordance with the Commission's regul.tiors, the pennsylvania State official was notified of the proposed issuance of the amendme:nt.
The State official had no comments.
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.MVJ@ff!1EH1 Al CONSIDFJLAJ10B j
E This amendment changes a requirement with respect to installation or use of a p_-
facility component locateo within the restricted crca as defined in 10 CFR P&rt 20.
The NRC staff has determined that the amendment involves no significant increase in the aracunts, and no significant change in the types, of any effluents that may be re? eased of fsite, and that there is no significant increase in individual or cumulat ive occupational radiation exposure. The Conmission has previcusly issued a proposed finding that the 6
cmenament involves no significant hazards consideration, and there has been no public comment on such finding (56 FR 57394).
Accordingly, the amendment
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- meets the eligibility criteria for categorical exclusion set forth in 10 CHI 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or F
environmental assessment need be prepared in connection with the issuance of
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the amendment.
5.0 C.QNCLUSION The Commission has concluded, based on the considerations discus. sed Sove-l 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_ Conenission's regulations, l
and (3) the issuance of the amendment will not be inimical to the connon l-defense and security or to the health and safety of the public.
Principal Contributors:
J. Tsao l
C. Liang Date: January 15, 1993 1
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Letter from J. D. Sieber (0QlCo) to USNRC Document Control Desk,
Subject:
Beaver Valley Power Station, Unit 1, Proposed Operating License Change Request 191 (TAC No. 80559), July 15,1991.
2.
Letter from J. D. Sieber (Duquesne Light Company) to NRC, " Beaver Valley Power Station, Unit'No. 1 Docket No. 50-334, License No. DPR-66, Proposed Operating License Change Request No. 191 correction," dated September 30, 1991.
3.
Letter fenm J. D. Sieber (DLC) to NRC, " Beaver Valley Power Station, Unit No. J. Docket No. 50-334, License No. DPR-66, Response to information Request dated November 14, 1991," dated December 13, 1991.
4.
Letter from J. D. Sieber (Duquesne L-ight Company) to NRC, " Beaver Valley-Power Station, Unit Nc. 1, Docket No. 50-334, License No. DPR-66, Proposed Operating License Change Request No. 191,. Revision 1," dated November 2, 1992.
5.
Letter from J. D. Sieber (DL) to USNRC Document Control Desk,
Subject:
Beaver Valley Power Station, Unit 1, Reactor Vessel Capsule W Test Results Report (WCAP-12005), January 24, 1989.
6.
R. S. Boggs et al, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit Reactor Vessel Radiation Surveillance Program, WCAP-10867," Westinghouse Electric Corporation, September, 1985 7.
Letter from J. J. Carey (DL) to S. A.:Varga-(USNRC),-subject:- Beaver Valley Power. Station, Unit 1, Reactor. Vessel Irradiation Specimen Test Report (WCAP-9860), October 15, 1981.
8.
Letter from C. N.-Dunn (DL) to R. W. Peid (USNRC), subject: Beaver Valley Power Station, Unit 1, Reactor Vessel Material ~ Surveillance Program, July 21, 1977.
9.
. Beaver. Valley Final Safety Analysis Report 6
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