ML20205T124
| ML20205T124 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20205T115 | List: |
| References | |
| NUDOCS 8704070218 | |
| Download: ML20205T124 (5) | |
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UNITED STATES s
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIDW SUPPORTING AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA P0tTR CORPORATION, ET AL.
CRYSTAL RIVER UNIT h0. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 INTRODUCTION By application dated February 17, 1986, as supplemented by letters dated November 19 and 25, 1986, and February 17, 1987, Florida Power. Corporation (FPC or the licensee) requested amendment to Facility Operating License No. DPR-7? for the Crystal River Unit No. 3 Nuclear Generating Plant (CR-3).
The proposed amendment would extend the expiration date of the license from September 25, 2008, to December 3, 2016.
7.0 DISCUSSION Section 103.c of the Atomic Energy Act of 1954 states that a license is to be issued for a specified period not to exceed 40 years.
10 CFR 50.51 specifies that each license will be issued for a fixed period of time not to exceed 40 years from the date of issuance. The currently licensed term for CR-3 is 40 years commencing with the issuance of the construction permit until September PS, 2008. Accounting for the time that was required for plant construction, this represents an effective operating license term of 3? years.
Consistent with Section 103.c of the Atomic Energy Act and Section 50.51 of the Commission's regulations, the licensee, by the February 17, 1986 1
application, seeks extension of the operating license term for CR-3 so the l
fixed period of the license would be from the date of the operating license issuance or until December 3, 20I6.
3.0 EVALUATION ALARA and Dose Assessment The following evaluation was conducted to assure that the licensee's "as low as reasonably achievable" (ALARA) measures and dose projections are applicable for the additional years of plant operation and are in accordance with 10 CFR Part 20,
" Standards For Protection Against Radiation" and Regulatory Guide 8.8 "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As low As Reasonably Achievable" (Revision 3).
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The licensee stated that operating and maintenance personnel will follow specific plans and procedures to ensure that ALAPA goals are achieved in the extended years of operation.
FPC anticipates that ALARA techniques will continue to improve through the use of robotics, remote tonlino, and improved methods of decontamination.
In 1981, FPC established goals and objectives for maintaining occupational radiation exposure within ALARA guidelines. FPC added additional measures and methods to reduce radiation exposures to plant staff such as reactor head shielding, letdown line shielding, pressurizer shielding, pre-fabricated tents, reactor coolant piping valve relocation, training mockups for high exposure jobs, and free standing shielding boxes.
FPC's commitment to state-of-the-art ALARA is contained in the Company Policy and Fuclear Operations Department Procedure which identifies the health physics organization. The licensee's radiation protection /ALARA program has been recognized by the NRC staff as adecuate overall in the Systematic Assessment of Licensee Performance (SALP) from 1981 to 1986 (Category ? rating).
Therefore, we find that FPC has an adequate health physics organization and radiation protection program, and that personnel are adequately trained in ALARA.cansiderations for the additional years of operation. We further conclud that the updated Final Safety Analysis Report (FSAR) for CP-3 (Radiation Protection) is in accordance with 10 CFP Part 20 and is consistent with t'te criteria of Regulatory Guide P.8.
Thus, we find the ALAPA program and reactices to be acceptable.
FPC provided tables specifying person-rem exposures at CR-3 by plant system independent of when these exposures were obtained (e.g., during normal operations, maintenance, repair, or refueling activities) and by whom (e.g.,
plant operations personnel, plant maintenance personnel, or contracter/ vendor personnel). We audited the licensee's dose assessment for the extended years (2009-2016) against the criteria of Standard Review Plan (SRP) Section 12.3.
The licensee based the estimate on nine years of operating experience, engineering judgment and on personnel exposure at CR-3 for the years 1977-1985.
FPC expects the additional years of operation of CR-3 to result in an average of 224 person-rem per year. Currently, operating Pressurized Water Reactors (PWRs) average more than 569 person-rem per unit annually (1980-1985) with some plants experiencing an average lifetime annual dose as high as 1300 person-rem. These averaae doses are based on widely varying yearly doses at PWRs.
The licensee estimated four additional refueling outages for the years 2009-7016. Barring major plant modifications which are not now contemplated, the total dose is predicted to be 179? person-rem. The predicted value is based on an assumed 84 person-rem for a non-outage year and 364 person-rem for a refueling outage year. Dose allowance for crud buildup will be offset by dose savings from a continually improving ALARA program.
It is expected that state-of-the-art technologies will be in use including some robotics, enhanced chemistry control and modern decontamination processes.
Based on the above, we conclude that the licensee's dose assessment is acceptable, and the CR-3 radiation protection program is adequate for ensuring that occupational radiation exposures will be maintained in accordance with ALARA guidelines and in compliance with 10 CFR Part 20 requirements.
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, Pressure Vessel Touchness The licensee, in response to the requirements of 10 CFR 50.61, " Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events,"
previously submitted information on the projected values of the RT to the expiration of their current operating license and for 40 calendar hrs of operation (3? effective full power years (EFPYs) at an 80% load factor).
RTp e is a calculated reference temperature which is used as a screening criterion.
As established in 10 CFR 50.61, the pressurized thermal shock (PTS) rule, this figure must be less than 300 F for circumferential weld material, the controlling material for CR-3.
In our evaluation of the licensee's previous submittal, enclosed with our letter of September 4,1986 to FPC, we found the estimated RT for 32 EFPYs (equivalent to 40 calendar years) to be 284.6. Sincethisisles$Tkhan300*F,thescreening i
criterion for the limiting material at expiration of the license, it meets the PTS rule and is acceptable.
The PTS rule requires that the projected assessment of the RT must be updated whenever changes in core loadings, surveillance measurements hbother infonnation (including changes in capacity factor) indicate a significant change in the projected values. This ensures that the licensees will track the fluence at the limiting beltline materials throughout the life of the plant to verify their assumptions.
In this regard, we requested FPC to submit a reevaluation of the RT and comparison to the predicted value with future Pressure-Temperatuhe submittals which are required by 10 CFR 50, Appendix G.
b Systems and Equipment The licensee states that:
... safety-related mechanical systems, equipment, and components considered will not he impacted by a 40 year operating lifetime. This does not imply that some mechanical system related eouipment and components will not wear out or need replacement during the plant operating lifetime.
However, existing serveillance and maintenance programs are sufficient to maintain or determine replacement of safety-related components.
Periodic inservice inspection and testing requirements have been.
incorporated into procedures to provide the added assurance that any
,I unanticipated degradation in systems or equipment will be identified and corrected in a timely manner.
Reactor vessel material and fluence analyses have shown that the expected cumulative neutron fluence on the reactor vessel will not limit the 40 year operating life. The Babcock A Wilcox Owners Group's Integrated. Reactor Vessel Surveillance Program and the planned Cavity Dosimetry Program shall provide a means for continuing to monitor the cumulative effects of the neutron exposure on the materials of the reactor vessel to satisfy the requirements of 10 CFR 50, Appendices G and H.
The analyses of the CP-3 plant specific surveillance capsules irradiated inside the redctor vessel of CR-3 will confirin that the predictions used in the analytical techniques for establishing operating limitations for the reactor vessels are conservative.
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-4 We conclude that safety-related structures, mechanical systems, equipment and components considered will not be impacted by a 40-year operating life-span.
This conclusion is based upon continuation of the periodic inservice inspection and testing requirements which have been incorporated into Crystal Piver Unit 3 Technical Specifications, programs and procedures, and assumes that any unanticipated degradation in structures, systems and components will be identified and corrected in a timely manner.
l'ith regard to electrical equipment, we have reviewed the licensee's submittel and agree with the licensee's conclusion that such equipment would not he adversely affected by a 40-year operational lifetime. Where failed or worn parts are replaced with new parts, no period after the first year or two is different from any other with respect to the expected rate of electrical component failures. The surveillance, maintenance, and replacement practices at CR-3, coupled with the equipment qualification requirements of 10 CFR 50.49, should prevent any increase in the probability of failure of two redundant safety trains in any emergency during a 40-year plant lifetime.
Aging analyses have been performed for all safety-related electrical equipment in accordance with 10 CFR 50.49, " Environmental Qualification of Electrical Eouipment Important to Safety for Nuclear Power Plants,"
identifyina qualified lifetimes for this equipment. These lifetimes will be incorporated into plant equipment maintenance and replacement practices to ensure that all safety-related electrical ecuipment remains qualified and available to perform its safety function regardless of the overall age of the plant.
Summary of Findings Based upon the above, we find that extension of the operating license for i
CR-3 to allow a 40-year service life is consistent with the safety analyses for CR-3 and that the Commission's previous safety findings are not changed.
Issues associated with plant systems and equipment, including aging and changes in fracture toughness properties of materials, have been addressed and are acceptable for 40 years of operation. The site continues to meet the guidelines of 10 CFR Part 100. Accordingly, we find the proposed change to the expiration date of the Crystal River Unit 3 Facility Operating License to be acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
An Environmental Assessment was issued on March 26, 1987, and Notice of Issuance of Environmental Assessment and Finding of No Significant Impact relating to the proposed extension of the Facility Operating License termination date for Crystal River Unit 3, was published in the Federal Register on Parch 31, 1987 (52 FR 10274).
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5.0 CONCLUSION
We have 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, and (?) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: Merch 31,1987 Principal Contributors:
J. Minns, C. Morris, R. Lipinski, H. Silver i
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