ML16138A719
| ML16138A719 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/14/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A718 | List: |
| References | |
| NUDOCS 9105290287 | |
| Download: ML16138A719 (7) | |
Text
pkREG;,jl UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 SAFETY-EVALUATION-BY-THE-FFICE-OF-NUCLEAR-REACTOR-REGULATIOC RELATED-TO-AMENDMENT-NG, -188 TO-FACILITY-OPERATING-LICENSE-DPR-38 AMENDMENT-NO,-188TO.-FACILITY-OPERATING LICENSE-DPR-47 AMENDMENT-NO,-1 85TO-FACILITY-OPERATING-LICENSE-DPR-55 DUKE-POWER-COMPANY OCONEE-NUCLEAR-STATION,-UNITS-1g-2,-AND-3 DOCKET-NOS,-59-269,-50-270 1-AND-50-287
1.0 INTRODUCTION
By letter dated November 15, 1989, as supplemented March 28 and August 29, 1990, Duke Power Company (the licensee) submitted a request for changes to the Oconee Technical Specifications (TS).
The requested changes would revise the Pressure/Temperature (P/T) Limits and the Low Temperature Overpressure Protection (LTOP) system operability requirements in Section 3.1 of the TSs and revise the associated Bases. The March 28 and August 29, 1990, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
The changes to P/T limits are in response to Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect On Plant Operation."
The current P/T limit curves are based on an assumed design basis neutron fluence through 15 effective full power years (EFPYs).
The proposed P/T limit curves were developed based on the guidance of Regulatory Guide (RG) 1.99, Rev. 2, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations," and they are applicable for a period of up to 15 EFPYs.
Prompted by the P/T Limit revisions, changes to the low temperature overpressure protection (LTOP) system requirements (TS 3.1.9.2) have been proposed to assure proper protection to the reactor vessel.
2.0 EVALUATION The NRC staff has evaluated the proposed changes to the P/T limits utilizing the following NRC regulations and guidance: Appendices G and H of 10 CFR Part 50, ASTM Standards and the ASME Code, 10 CFR 50.36(c)(2), RG 1.99, Rev. 2, Standard Review Plan (SRP) Section 5.3.2, and GL 88-11. Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements-for reactor vessel materials in accordance with the ASME Code and 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).
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-2 Appendix H of 10 CFR Part 50 states testing requirements for reactor beltline material in surveillance capsules and provides a connection with ASTM Standards.
GL 88-11 established the NRC requirement that the licensee use the guidance in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel material.
10 CFR 50.36(c)(2) states the limiting conditions for operation and SRP Section 5.3.2 provides the NRC staff position on P/T limits.
OCONEE UNIT 1 - P/T Limits The NRC staff has evaluated the effect of neutron irradiation embrittlement on each beltline material in the Oconee 1 reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2. The staff has determined that the material with the highest ART at 15 EFPY was weld SA-1585 with 0.21% copper (Cu), 0.59% nickel (Ni), and an initial RTndt of -12'F.
The licensee has removed four surveillance capsules from Oconee 1. The results from capsules F, E, A and C were published in Babcock and Wilcox (BAW) reports BAW-1421, Rev. 1, BAW-1436; BAW-1837, and BAW-2050, respectively. Surveillance capsule F contained Charpy impact and tensile specimens from plates. Surveillance capsules E, A and C contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and heat affected zone (HAZ) metal.
For the limiting beltline material, weld SA-1585, the staff calculated the ART to be 184.1 0F at 1/4T (T = reactor vessel beltline thickness) and 1 00 F for 3/4T at 15 EFPY. 2 The staff used a neutron fluence of 3.26E18 n/cm at 1/4T and 1.18E18 n/cm at 3/4T. The ART was determined by Section 1 of RG 1.99, Rev. 2, because the limiting weld material was not included in the surveillance capsules.
The licensee calculated the ARTs of 184 0F and 140 0 F at 1/4T and 3/4T, respectively.
Substituting the ARTs of 184'F and 140 0F 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.
OCONEE UNIT 2 - P/T Limits The NRC staff has determined that the material with the highest ART at 15 EFPY was the circumferential weld between the lower shell nozzle shell belt and upper shell (WF-25) with 0.31% Cu, 0.59% Ni, and an initial RT of -60 F.
ndt The licensee has removed three surveillance capsules from Oconee 2. The results from capsules C, A, and E in Oconee 2 were published in BAW reports BAW-1437, BAW-1699, and BAW-2051, respectively. All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metajL For the limiting beltline material, weld WF-154, the staff calculated the ART to be 191 0 F t 1/4T and 138.3 0 F at 3/41. The staff used a neutron fluence of 3.01E18 n/cm at 1/4T and 1.09E18 n/cm at 3/4T. The ART was determined by Section 1 of RG 1.99, Rev. 2, because the limiting material was not included in the surveillance capsules.
-3 The licensee selected the weld, WF-25, as the limiting material and calculated the ARTs of 193 0F and 139 0F at 1/4T and 3/4T, respectively. The licensee's ARTs are conservative; therefore, they are acceptable. Substituting the ARTs of 193 0 F and 139 0 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.
OCONEE UNIT 3 - P/T Limits The NRC staff has determined that the material with the highest ART at 15 EFPY at 1/4T was the nozzle belt to upper shell circumferential weld (WF-200) with 0.26% Cu, 0.64% Ni, and an initial RT of 20aF. At 3/4T, the material with the highest ART at 15 EFPY was the up09 to lower shell circumferential weld (WF-70) with 0.35% Cu, 0.59% Ni, and an initial RTndt of 200F.
The licensee has removed two surveillance capsules from Oconee 3. The results from capsules A and B were published in BAW reports BAW-1438 and BAW-1697, respectively. All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline materials,:welds WF-200 and WF-70, the staff calculated the-ART to be 202.2 0F at 1/4T a d 170.1 0 F for 3/4T at 15 FPY. The staff used a neutron fluence of 3.18E18 n/cm at 1/4T and 1.15E18 n/cm at 3/4T. The ART was determined by Section 1 of RG 1.99, Rev. 2, because the limiting material was not included in the surveillance capsules.
The licensee selected welds WF-67 and WF-70 as the limiting materials at the 1/4T and 3/4T locations, respectively, and calculated the ARTS of 195 0F at 1/4T and 171 0 F at 3/4T. The difference between the staff's ART and the licensee's ART at the 1/4T location (202oF vs. 195 0F) is because the licensee used a lower chemistry factor in the ART calculation. Substituting the ARTs of 202.2 0F and 171 0F 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. The licensee's P/T limits have sufficient safety margin such that with a lower ART the limits satisfy the Appendix G calculation.
OCONEE UNITS 1, 2 AND 3 - CLOSURE FLANGE LIMITS AND CHARPY USE AT EOL In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes 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 0 F for normal operation and by 90aF for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 600F, the staff has determined that the proposed P/T limits for Oconee Units 1, 2 and 3 satisfy Section IV.2 of Appendix G.
-4 Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb. Using the method in RG 1.99, Rev. 2, the staff predicted that the Charpy USE of weld metal SA-1430 and SA-1493 for Unit 1, SA-154 for Unit 2, and WF-200, WF-67 and WF-70 for Unit 3, will be below 50 ft-lb at the end of life. The licensee has joined the Babcock and Wilcox Owners Group (B&WOG) that is investigating the effect of low Charpy USE on fracture toughness of the reactor vessel beltline materials. The staff will withhold evaluation of the low Charpy USE until the B&WOG submits its findings on this issue for NRC staff review.
OCONEE UNITS 1, 2 AND 3 - LTOP In light of the above revisions to the P/T curves, the LTOP system has been reviewed to assure proper protection of the reactor vessel.
As a result of the review of the LTOP system, changes have been proposed to TS 3.1.9.2 to resolve conflicts which were identified in the current TSs concerning inadvertent actuation of the High Pressure Injection (HPI) system.
LTOP is provided by the power operated relief valve (PORV) on the pressurizer.
The PORV is set at a pressure low enough to prevent violation of Appendix G heatup and cooldown curves should a reactor coolant system (RCS) pressure transient occur during low temperature operations. The licensee, in its November 15, 1989, submittal provided the results of considerations of the most limiting overpressure transients in determining the necessary TS changes for LTOP.
Currently, at Oconee, LTOP is provided by two methods, at least one of which must be operable when RCS cold leg temperature is less than or equal to 3250 F.
One method requires that both train A and train B of the HPI system be disabled.
The other method consists of the PORV with a lift setting of less than or equal to 500 psig, pressurizer level less than or equal to 260 inches, and RCS pressure less than 400 psig. Because only one of the two LTOP methods is.
required to be operable in the existing TSs the current TSs are less conservative than intended, and operation is being controlled administratively to assure that the vessel is protected from LTOP events.
The licensee's proposed LTOP TSs require that two trains of LTOP be operable.
The first train is comprised of a PORV with a reduced lift setting. This train must be operable when RCS cold leg temperature is less than or equal to 325 0F and either RCS pressure is greater than or equal to 100 psig or HPI pumps are running. The second train is comprised of the controls which assure that 10 minutes are available for operator action to mitigate the consequences of an
[TOP event. The following controls are included in the second LTOP train:
- a. RCS pressure is limited to less than or equal to 350 psig for Units 1 and 2, and lessthan or equal to 345 psig for Unit 3, below an RCS temperature of 220*F.
- b. Deactivating train A and B of HPI.
- c. Deactivating both core flood tanks.
-5
- d. Pressurizer level shall be controlled such that 10 minutes are available for operator action to mitigate an LTOP event.
- e. Makeup flow shall be restricted such that 10 minutes are available for operator action to mitigate an LTOP event.
- f. Alarms shall be provided such that 10 minutes are available for operator action to mitigate an LTOP event.
The licensee has committed to include specific limiting values of maximum and minimum pressurizer level and maximum makeup flow rate in the Selected Licensee Commitments Manual which is found in Chapter 16 of the Final Safety Analysis Report.
The licensee has provided an evaluation and results of analyses to justify the proposal in its November 15, 1989 submittal.
The evaluations assumed initiation of the most limiting mass addition and heat addition transients plus the most limiting single failure, failure of the pressurizer PORV. No credit was taken for operator action until 10 minutes after the event is recognized.
The most limiting mass addition transients were analyzed, and the results of these analyses are as follows. Erroneous actuation of the HPI system or the core flood discharge valves has been precluded by requiring that both the A and B trains of the HPI system and the core flood injection function be deactivated while RCS temperature is below 325 0 F. It has been determined that erroneous addition of nitrogen to the pressurizer results in a maximum pressure of only 150 psig and therefore is not a limiting transient. The last mass addition LTOP event analyzed was RCS makeup fails full open. This scenario was found to cause RCS pressure to increase above the new P/T limits in less than 10 minutes, when an initial pressurizer level of 220 inches is assumed. The most limiting makeup flow rate providing at least 10 minutes for operator action was calculated for each of the Units, and the flow indication uncertainty was subtracted to determine the maximum allowable flow rate with RCS temperature below 325 0F. The makeup flow path will be restricted to the maximum allowable flow rate during low temperature operations.
Also included in the evaluation were the following limiting heat addition transients:
loss of decay heat removal during startup and shutdown, all pressurizer heaters erroneously energized, and start of a reactor coolant pump (RCP) with stored thermal energy in the RCS. The loss of decay heat removal scenario was analyzed using conservative assumptions. One or more alarms will actuate on loss of decay heat removal, therefore, operator awareness is assumed at the beginning of the transient. It was determined that at least 10 minutes are available between the start of the transient and the time at which the pressurizer reaches the P/T limits. The initial conditions of this analysis determine tWhmaximum allowable pressurizer pressure and level for operations with the RCS temperature below 220 0F, and these maximum values are 350 psig and 220 inches, respectively. For the scenario of all pressurizer heaters erroneously energized, it was found that the rate of pressurization was inversely proportional to the initial pressurizer water level.
As a result, a limit will be placed on
-6 the minimum allowable pressurizer level, such that 10 minutes will be available for operator action to mitigate the event. Finally, the start of an RCP with stored thermal energy in the RCS was analyzed. It was concluded that this transient will not cause the RCS pressure to exceed the P/T limits at any temperature.
These analyses were performed using the RETRAN-02 MD003 computer code.
RETRAN-02 M00003 has been generically approved by the NRC staff on October 19, 1988. Currently Duke Power Company's application for the use of RETRAN-02 MOD003 is under review by the staff. The staff considers that reasonable assurance exists that the results of the licensee's analysis using RETRAN-02 MOD003 supports the proposed TSs on LTOP. However, if any concerns should arise during the staff review of Duke Power's RETRAN-02 MOD003 application, the staff may require a reassessment of the proposed TS changes.
The staff has also reviewed the proposed Bases for TS 3.1.2.9 and finds the discussion correctly describes the proposed LTOP features and is acceptable.
3.0
SUMMARY
The NRC staff concludes that the proposed P/T limits for the RCS for heatup, cooldown, leak test, and criticality are valid through 15 EFPYs because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2, to calculate the ART. Hence, the proposed P/T limits may be incorporated into the Oconee Unit 1, 2 and 3 TSs.
The staff has determined that some beltline welds will have Charpy upper shelf energy below 50 ft-lb at end of life in all three Oconee units. The licensee has joined the B&WOG that is investigating the effects of low Charpy USE on fracture toughness of the reactor vessel beltline materials. The staff will withhold the conclusion on the matter until the staff has reviewed the B&WOG's report on the low USE in Oconee Units 1, 2 and 3.
Also, the staff concludes that the licensee proposed Technical Specification 3.1.9.2 are acceptable to support the updated Pressure-Temperature limits applicable for a period of up to 15 EFPY.
4.0 STATE.-CONSULTATION In accordance with the commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENIVIRONMENTAL--CONSI-DERATION The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no
-7 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 proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (56 FR 4862 on February 6, 1991).
Accordingly, the amendments meet 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 amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: John Tsao, EMCB/DET Amy Almond, SRXB/DST Frank Rinaldi, PDII-3/DRP Leonard Wiens, PDII-3/DRP