ML20151B772
| ML20151B772 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/04/1988 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20151B770 | List: |
| References | |
| NUDOCS 8804110240 | |
| Download: ML20151B772 (5) | |
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SAFETY EVALUATION BY THE OFFICE OF huCLEAR REACTOR REGULATION SUPPORTING AMENDHENT NO.117 TO FACILITY OPERATING LICENSE N0. DPR-71 AND AMENDMENT N0.147 TO FACILITY OPERATING LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324
1.0 INTRODUCTION
In letters to the NRC dated March 5, 1986 and December 17, 1987, Carolina Power & Light Company requested a revision to Table 4.4.6.1.3-1 of the Technical Specifications for Brunswick Steam Electric Plant.
Units 1 and 2 (Brunswick Units 1 and 2). The schedule for removing the reactor vessel material surveillance capsules is specified in Table 4.4.6.1.3-1.
Carolina Power & Light Company (the licensee) proposed to change the schedule for removing the surveillance capsules.
In the initial submittal of March 5,1986, the licensee proposed an integrated schedule for Brunswick Units 1 and 2 whereby a capsule would be removed alternately from each reactor after 8, 10, 13 and 15 effective full power years (EFPY) of operation.
This se1edule was proposed in lieu of the schedule in the present technical specifications which calls for the removal of a capsule from eae.h reactor after 10 and 30 years of operation. The licensee proposed to remove the first and third capsules from Brunswick Unit I after 8 and 13 years of EFPY operation and the second and fourth capsules from Brunswick Unit 2 after 10 and 15 years of EFPY of operation.
Following discussions with the staff, the licensee modified their proposal in a December 17, 1987 submittal.
In the latest submittal, the initial capsule would be removed from Brunswick Unit 1 after 8 effective full power years (EFPY) oper6 tion and the initial capsule would be removed from Brunswick Unit 2 after 10 EFPY.
The schedule for the removal of each Unit's second rnd third capsule would be proposed after the results of the first capsule had been evaluated.
In addition, the licensee proposed to insert, in paragraph 4.4.6.1.3 of the Surveillance Requirements in the Technical Specifica-tions, a statement that "The cumulative effective full power years shall be determined at least once per 18 months."
The stated purpose of the proposed change is to revise the reactor vessel surveillance capsula removal schedule to achieve compliance with the provisions of Appendix H. 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements," and ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Paragraph II.B.1 of Appendix H, 10 CFR Part 50, requires that the test procedures and reporting requirements for the reactor vessel material surveillance program meet the requirements of ASTM E185-82 to the extant practical.
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The capsule withdrawal schedule is to cemit monitorino of the fracture l
touchness properties of the vessel materials to the end of life (E0L) of the reactor.
Table 1 of ASTM E185-8? lists the recommended number of caosules and the withdrawal schedule for three ranges of shifts in the predicted nil-ductility transition temperature. Where the shift at the inside sur' ace of the vessel is eoual to or less than 100'F (37.8"C), the program would consist of three capsules which are withdrawn after 6 and 15 EFPY and at the EOL operation.
The first capsule would be withdrawn either after S EFPY operation or after the predicted shift of all encapsulated meterial is about 50 F (IP'C1, whichever arrives first.
Paragraph II.R.3 of Appendix H,10 CFR Part 50, requires that the proposed capsule withdrawal schedule be submitted to the Director, Office of Nuclear Reacter Regulation, with technical justification for approval prior to implementation.
2.0 EVALUATION The Safety Evaluation for the Rrunswick Units 1 and 2. November 1973, Section 5.2.7, states that the reactor vessel materials surveillance program was acceptable in regard to the number of capsules, the number and type of specimens, and the retention of archive material.
Detailed information on the eheapsulated raterials for the prog.*am are recorded in General Electric Company Reports, NED0-24161 and NEDO-24157.
The limiting raterials in the vessels are identified as beltline plates:
plate 201 in Brunswick, Unit 1, containing 0.15% Cu, 0.54% Hi and 0.012% P, and plate 351 in Rrunswick Unit 2. containing 0.19% Cu, 0.58% Ni and 0.013% P.
The reactor vessels were purchased prior to the issuance of Appendix G, 10 CFR Part 50, and the pressure bounda'y raterials were qualified by drop weicht test for the plates end Charpy V-Notch test for the weld metal.
The requirenants for the plates were NDT of 10'F or less and for the welds was a Charpy V-Notch energy of 30 ft-lb at 10'F.
Full impact curves were not obtained on the pressure boundary materials, and the upper-shelf energy levels were not reported, fheEOL(3?EFPY) fluence at the surface was gstimated at 1.98 x 10 n/cmr and at the i
1/4t position at 1.42 x 10 n/cm2 for both reactors.
The values were calculated by Westinghouse Electric Corporation from dosimetry measure-ments.
The EOL transition temperature shift at the reactor vessel surface was calculated using methodology uescribed in Regulatory Guide 1.99, Revisions 1 and ?.
For Brunswick, Unit 1, the transition temperature shift was 57.8'F and 58.8'F, for Revisions 1 and ? respectively; while i
for Brunswick, Unit 2, the transition temperature shift was 77.0'F and 77.0'F, respectively.
These surface transition temperature shifts correspond to an adjusted reference temperature of 67.8'F and 102.8'F for Brunswick Unit 1 and 87.4*F and 121.0'F for Brunswick Unit 2.
3 The transition temperature shift at the 1/4t position was calculated for both reactors after 8 and 10 EFPY operation.
The 1/4t calculated values using Regulatory Guide 1.99 Revisions 1 and 2 are tabulated as follows:
EFPY Fluence Transition Temperatu*e Shift, 'F.
Operation n/ cme Regulatory Guide 1.99, Regulatory Guide 1.99, Revision 1 Pevision 2 Unit 1 Unit 2 Unit 1 Unit 2 17 8
2.54 x 10 20.6 27.8 21.4 28.0 17 10 3.18 x 10 23.1 31.1 24.6 32.2 17 13 4.13 X 10 26.3 35.5 27.8 36.4 17 15 4.76 x 10 28.3 38.1 29.9 39.2 18 49.0 65.9 52.4 68.6 E0L(32) 1,42 x 10 The fluence at the surveillance capsules in both reactors lags the fluence at the inside surface of the reactor vessels by a factor of 0.56.
The estimated fluence values in the tabulation for the 1/4t positions are slightly less than the values currently estimated by the licensee for the capsules. Thelicgseeestimatesthatthecapsulesw{ receive a fluence of 2.79 x 10 n/cm2 after 8 EFPY and 3.49 x 10 n/cm3 after 10 EFPY operation. ASTM E185-82 recommends that the surveillance capsules lead the fluence received at the surface of the reactor vessel by a factor from one to three. The withdrawal of the first capsule in a surveillance program should be scheduled early in the reactor vessel life in order to verify the initial response predictions of the surveillance material to the actual thermal and radiation environment of the reactors. The removal of the first capsule from Brunswick Unit 1 after 8 EFPY and from Brunswick Unit 2 after 10 EFPY operation is expected to permit verification i of the adequacy and conservatism of the reactor vessels' pressure / tempera-l ture operational limits. The first capsule withoraNal schedule complies with the requirements of ASTM E185-82 to the extent practical. The requirements of Appendix H of 10 CFR Part 50 are such that af ter each capsule withdrawal, the test results must be the subject of a sumary technical report to be submitted as specified in 10 CFR Part 50.4 within one year after capsule withdrawal unless an extension is granted by the Director, Office of Nuclear Reactor Regulation. The report must include the data required by ASTM E 185-82, as specified in paragraph II.B.1 of Appendix H, and the results of all fracture j toughness tests conducted on the beltline materials in the irradiated and unirradiated conditions. If a change in the Technical Specifications is required, either in the pressure-temperature limits or in the operating procedures required to meet the limits, the expectsd date for submittal of the revised Technical Specifications must be provided with the report. I i
- The withdrawal schedule for the second and third capsules in Table 1 of ASTM E185-82 is fluence related. The second capsule should be withdrawn when the accumulated fluence corresponds to the approximate EOL fluence at the reactor vessel inner wall location, or after 15 EFPY operation, whichever arrives first. After 15 EFPY operation in Brunswick Units 1 and 2 the surveillance capsule is estimated to receive approximately 25% of the E0L vessel surface fluence. Upon withdrawal, the third capsule should have received not less than the peak EOL inside surface fluence nor more than twice that value. As a result of the severe lag in capsule fluence in Brunswick Units 1 and 2, the licensee has proposed that the withdrawal schedule for the second and third capsules be deferred pending the results of the analysis and evaluation of the test specimens from the first capsules withdrawn from the respective reactors after 8 and 10 EFPY operation. The review of the test results from the initial capsule withdrawal will allow the licensee and the staff to determine when the second capsule withdrawal for Unit 1 and for Unit 2 is appropriate. EVALUATION
SUMMARY
i The purpose of the materials surveillance program required by Appendix l H,10 CFR Part 50, is to monitor changes in the fracture toughness i properties of ferritic materials in the beltline region of nuclear power reactors resulting from the neutron irradiation and thermal environment. The proposed changes to the Technical Specifications of Brunswick Units 1 and 2, revising the withdrawal schedule of sur-veillance capsules shown in Table 4.4.6.1.3-1, meet, to the extent practical, the provisions of Appendix H 10 CFR Part 50, and ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Further, the staff believes that the licensee's withdrawal schedule + should be reviewed again after the fracture toughness data of materials from the first surveillance capsules from both Brunswick Unit I and 2 are known and analyzed. Consideration should be given to the dosimetry measurements made in each 18 month period with a deterinination of the i fluences at the surveillance capsule and the wall of the reactor vessel.
3.0 ENVIRONMENTAL CONSIDERATION
S The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined j in 10 CFR Part 20 and change surveillance requirements. The staff has deterTnined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite; and that there should be no significant increase in individual or cumulative occupational radiation exposure.
t , The Commission has previously publisheo a proposed finding that the amendments involve no significant hazards consideration, and there nas been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR951.22(c)(9). Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
4.0 CONCLUSION
The Commission made a proposed determination that these amendments involve no significant hazards consideration which was initially published in a May 21, 1986 Federal Register Notice (51 FR 18677), and then again on February 24, 1988 (53 FR 5487) and consulted with the State of North Carolina. No public comments were receiveo, and the State of North Carolina did not have any comments. 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, and (2) such activities will be conducted in compliance with the Commission's regulations, and 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 Contributor: Felix Litton Dated: April 4,1988 I ..}}