ML20126F848
| ML20126F848 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/18/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126F844 | List: |
| References | |
| NUDOCS 9212310173 | |
| Download: ML20126F848 (6) | |
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UNITED STATES 3
.i NUCLEAR REGULATORY COMMISSION WASHINoTON, D.C. KE6 o\\**.*/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 45 10 FACillTY OPERATING LICENSE NO. NPF-58 THE CLEVELAND ELECTRIC ILLUMINATING COMPANY. ET AL.
PERRY NUCLEAR POWER PLANT. UNIT N0. 1 DOCKET NO. 50-440
1.0 INTRODUCTION
By letter dated September 14, 1990, as supplemented December 20, 1991, the Cleveland Electric illuminating Company (the licensee) requested changes to the pressure / temperature (P/T) limits of Section 3/4.4.6 of the Technical Specifications (TSs) for the Perry Nuclear Power Plant (PNPP), Unit 1.
The proposed revised limits were calculated in accordance with the methods described in NRC Regulatory Guide (RG) 1.99, Revision 2, as recommended by the staff in Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations."
Additionally, in a letter dated March 15, 1991, the licensee requested the relocation of Table 4.4.6.1.3-1, " Reactor Vessel Material Surveillance Program-Withdrawal Schedule," from the TSs to the PNPP, Unit 1, Updated Safety Analysis Report (USAR).
The additional request was submitted in accordance with GL 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications."
In a separate letter dated March 15, 1991, the licensee requested staff approval of a revised surveillance capsule withdrawal schedule, as required by 10 CFR Part 50, Appendix H, to be included in the licensee's USAR, Section 5.3.1.6.1.
A " Notice of Consideration of Issuance of Amendment to Facility Operating License and Proposed No Significant Hazards Consideration Determination and Opportunity for a Hearing" for the proposed action was published in the Federal Reaister on July 24, 1991 (56 FR 33961).
The additional information provided by the licensee's December 20, 1991, submittal did not change the initial no significant hazards consideration determination.
Petitioners, Ohio Citizens for Responsible Energy (OCRE) and Susan L. Hiatt, requested a hearing in response to the July 24, 1991, notice of opportunity-for hearing.
In their request for hearing, the petitioners stated their agreement with the Staff's "no significant hazards" finding and agreed that 9212310173 921218 DR ADOCK 0500 0
the amendment was merely an administrative matter.
By Memorandum and Order dated March 18, 1992, the Atomic Safety and Licensing Board (ASLB) denied the petition to intervene and request for a hearing after it ruled that the petitioners had failed to establish standing to intervene in this license amendment proceeding, On April 2,1992, the petitioners filed a Notice of Appeal of the ASLB decision that is awaiting resolution by the Commission.
2.0 EVALVATION The pressure / temperature (P/T) limits are established to provide acceptable-margins for the operation of the reactor coolant system during heatup, cooldown, criticality and hydrotest conditions.
The proposed P/T limits would be valid for 8 effective full power years (EFPY). To evaluate the accepubility of proposed P/T limits, the staff uses the following NRC regula: ions and guidance:
10 CFR Part 50, Appendices G and H (and by refereice, the American Society of Testing and Materials (ASTM) Standards and the American Society of Mechanical Engineers (ASME) Code); 10 CFR 50.36(c)(2);
RG 1.99, Revision 2; Standard Review Plan (SRP) Section 5.3.2; and GL 88-11.
As required by 10 CFR 50.36(c)(2). Technical Specifications containing limiting conditions for operation (LCO) are established for important plant-parameters.
For the PNPP, Unit 1, TS 3/4.4.6, " Pressure / Tem)erature Limits,"
contains an LC0 for the reactor coolant system that limits tie rates of change of temperature and pressure to values consistent with the fracture toughness requirements of the ASME Code-and Appendix G to 10 CFR Part 50.
Changes in the values of these limits are necessary because the fracture toughness properties of ferritic materials in the reactor vessel change as a function of reactor operating time.
For this reason, Appendix H to 10 CFR Part 50 requires licensees to establish a surveillance program to periodically withdraw and examine surveillance capsules containing reactor vessel material specimens from the reactor vessel. These capsules must be installed in.the vessel prior to initial startup and must contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor vessel beltline. Appendices G and H of 10 CFR Par.t 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 calculating appropriate 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, requires that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, which references the ASTM Standards, describes tests to define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in-reference temperature. Appendix G also requires licensees to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf
3 energy (USE). GL 88-11 requested licensees-to use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation on reactor vessel materials.
That regulatory 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, in addition to the requirements of Appendix H to 10 CFR Part 50, the PNPP, Unit 1, TSs have included a surveillance requirement, TS 4.4.6.1.3, that requires the removal and examination of the irradiated specimens of reactor vessel materials.
Table 4.4.6.1.3-1 identifies the material specimen and specifies the schedule for removal of each specimen. ut 91-01 recognizes that the TS duplicates the controls on changes to the schedule that have been established by Appendix H.
Because this duplication'is unneccesary, GL 91-01 allows for the removal of the TS schedule as a line-item improvement.
2.1 GL 88-11 CONSIDERATIONS The staff has evaluated the effect of neutron irradiation embrittlement on each beltline material in the PNPP, Unit 1, reactor vessel. The amount of embrittlement was calculated in accordance with RG 1.99, Revision 2.
The staff has determined that the material specimen with the highest ART at 8 EFPY is plate C2557-1, with 0.06% copper (Cu), 0.61% nickel (Ni), and an initial reference temperature (RT of 10'F. To date, the licensee has not removed any surveillance. capsules /r)om the PNPP, Unit 1, reactor vessel.- All g
surveillance capsules contain Charpy impact specimens and tensile specimens made from base metal, weld metal, and heat-affected-zone (HAZ) metal.
For the limiting beltline material (plate C2557-1), the staff calculated the ART to be 39.9'F at 1/4T (T - reactor vessel beltline thickness) and 30.4*F 2
for 3/4T at 8 EFPY, Jhe staff assumed a neutron fluence of 9.3E17 n/cm at 1/4T, and 4.5E17 n/cm at 3/4T.
The ART was determined using Section 1 of RG-1 99, Revision 2, since no surveillance capsules have yet been removed from the Perry reactor vessel.
The licensee's calculated ART for 8 EFPY is 37'F, about 3*F in the non-conservative direction. This difference is not significant, as the licensee compensated for it by using a more conservative method for calculating'the thermal stress intensity factor.
The staff independently confirmed the licensee's calculated final vessel temperatures at several representative pressures and found the results to be acceptable. Substituting the ART of-39.9'F into the equations given 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.
During plant startups and shutdowns, the licensee measures reactor coolant temperatures to verify compliance with the vessel P/T limits.
This is conservative for the cooldowns, but potentially non-conservative for startups.
The licensee stated in their response to NRC's Request for Additional
s i,
Information (RAI) that the heatup follows the thermodynamic saturated water / steam line, which is-much more conservative than the P/T limits (see the attached figure in Ref. 5). The staff accepts this argument and agrees that the proposed P/T limits showing calculated crack tip temperatures need not be-revised to accurately reflect the temperature at the location where temperature measurements are made, 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 riosure flange materials.
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' for hydrostatic pressure tests and leak tests.
Paragraph IV.A.3 of Appendix G states "[a]n-exception may be made for boiling water reactor vessels when water level is within the normal range for power operation and the pressure is_less than 20 percent of the pre-service system hydrostatic test pressure.
In-this case the minimum permissible temperature is 60'F (33*C) above the reference temperature of the closure flange regions that are highly stressed by.the bolt preload." Based on the flange reference -temperature of 10*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb. The material with the lowest initial USE is plate C2557-1 with 75 ft-lb.
Using the method in RG 1.99, Rev. 2, the staff predicted that the USE at 1/4 T at 32 EFPY will be 63.2 ft-lb.
Since this is greater than the required 50 ft-lb, it is acceptable.
2.2 GL 91-01 CONSIDERATIONS The removal of the schedule for withdrawing material from the TS-will eliminate the necessity of a license amendment to make changes to the reactor vessel specimen testing schedule, Table 4.4.6.1.3-1..Nonetheless,Section II.B.3 of Appendix H to 10-CFR Part 50 requires the submittal of a.3roposed withdrawal schedule for material specimens to the NRC and approval
)y the NRC before implementation.
Therefore, adequate regulatory controls exist to control changes to this schedule without the necessity of subjecting it to the license amendment process required if the schedule -is. included in the TS.
The licensee has provided a committment to include this schedule in the next revision of the USAR.
In addition, the licensee will include any subsequent NRC-approved revisions to-this schedule in an update of the USAR. The inclusion of the withdrawal schedule in the USAR provides a source for this information that is readily available as a reference for NRC inspectors and other staff use.
Finally, the surveillance requirements for removing material specimens and the bases section for this specification remain consistent with 10 CFR Part 50, Appendices G and H.
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' The staff concludes that the pro)osed P/T limits for the reactor coolant system for heatup, cooldown, lea < test, and criticality are valid through 8 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
The proposed P/1 limits also satisfy Generic letter 88 because the method in RG 1.99, Rev. 2 was used to calculate the ART.
Therefore, the proposed P/T limits may be incorporated into the PNPP, Unit 1, TS.
Additionally, the licensee has proposed a change to TS 4.4.6.1.2 that is consistent with the guidance provided in GL 91-01 for the removal of TS Table 4.4.4.1.3-1 from the TS.
The NRC staff has reviewed this matter and finds that the proposed changes to the TS for the PHPP, Unit 1, are acceptable.
Furthermore, the staff approves of the licensee's proposed specimen withdrawal schedule to be included in Section 5.3.1.6.1 of the next USAR update.
3.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATI@
The staff has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10 CFR 50.92(c),
a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment (s) would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.
Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.
Involve a significant reduction in a margin of safety.
The no significant hazards consideration determination was conducted independently for both GL 88-11 and GL 91-01 considerations as follows.
GL 88-11 The proposed revisions to the TSs do not involve a significant increase in the probability or consequences of a previously evaluated accident because-the changet remain consistent with the provisions of 10 CFR Part 50, Appendices G and H, and are in accordance with guidance provided in GL 88-11.
The changes will result in equivalent or more conservative limits on reactor vessel pressure as a function of temperature.
The proposed changes do not create the possibility of a new or different kind of accident as the proposed changes impose equivalent or more conservative limits and do not involve any new modes of operation.
The proposed changes do not involve a significant reduction in the margin of safety because the new P/T limits result in greater margins to nil ductile failure under plant operating and test conditions.
, 'GL 91-01 The proposed changes do not involve a significant increase in the probability or consequences of a previously evaluated accident because the proposed change remains consistent with 10 CFR Part 50, Appendices G and H, and its references.- The movement of the specimen withdrawal table from the TS to the USAR is only an administrative change. The withdrawal schedule is not impacted and must receive NRC-approval before it can be changed.
The proposed changes do not create the possibility of a new or_different kind of accident from any accident previously evaluated because no new modes of operation are allowed as a consequence of the changes.
The proposed changes do not involve a significant reduction in the margin of safety because the reactor vessel material surveillance program is not changed; this purely administrative change merely changes the location of the schedule.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the instal-lation or use of a facility component located within the restricted area as defincd in 10 CFR Part 20 or a change to a surveillance requirement. The 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 made a final-determination that this amendment involves no significant hazards consid-eration. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set-forth in 10 CFR 51.22(c)(g). This amendment also i_nvolves changes in recordkeeping, reporting or administrative procedures or requirements. Accordingly, with respect to these items, the amendments meet the eligibility criteria for categorical exclusion-set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
6.0 CONCLUSION
The staff 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: S Sheng T. Dunning M. Webb Date: December 18, 1992
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