ML20249B833
| ML20249B833 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/19/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20249B832 | List: |
| References | |
| NUDOCS 9806240288 | |
| Download: ML20249B833 (3) | |
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UNITED STATE 8 S
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20086 4001 4*.,**
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RFI ATED TO AMENDMENT NO.175rO FACILITY OPERATING LICENSE NO. DPR-35 BOSTON EDISON ConnPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
By letter dated March 25,1998, Boston Edison Company (BECo or the licensee), proposed to amend Facility Operating License No. DPR-35 in accordance with 10 CFR 50.90. By letter dated April 8,1998, BECo requested that this amendment be reviewed under exigent circumstances.
The staff attempted to process this amendment on an exigent basis in accordance with 10 CFR q
50.91(a)(6)(1)(A). However, due to technicalissues which arose during the review of the
- amendment, additional time was required beyond the normal 30-day public notice period.
Therefore, this amendment is not being issued on an exigent basis. The proposed change modifies the Pilgrim Nuclear Power Station (PNPS) Technical Specification (TS) 3.6.A.1 and 4.6.A.1 as it pertains to Primary System Boundary, Thermal and Pressurizer Limitations, Surveillance Requirements, and Basis. Specifically, the licensee proposed to eliminate a 145'F temperature difference limit between the reactor vessel flange and adjacent vessel shell and eliminate its associated surveillance requirement contained in the TS. By letter dated April 27, 1998, the staff transmitted a request for additional information to the licensee. The licensee responded to the request for additional information in a letter dated May 5,1998. The May 5, 1998, letter provided clarifying information that did not change the intent of the initial proposed no significant hazards consideration determination.
2.0 DISCUSSION The PNPS reactor vessel has thermocouple installed on the lower flange and the adjacent vessel wall as depicted in Attachment D of the licensee's March 25,1998, letter. PNPS TS 4.6.A.1 requires the monitoring and recording of the thermocouple differer'tial temperature (DT)
' between the reactor vessel lower flange and the adjacent shell during heatup and cooldown. TS 3.6.A.1 further requires the themiocouple DT not to exceed 145'F as measured at the 4
thermocouple locations.. According to the licensee, the thermocouple DT is monitored to ensure that the licensing basis design criteria are not exceeded during normal startup and shutdown procedures. The licensee indicated that the accuracy of the reactor vessellower flange thermocouple had recently deteriorated and was no longer reliable. The licensee performed analyses to demonstrate that there is a correlation between the calculated thermocouple DT and the measured fluid heatup and cooldown rates recorded in the past, hence, making the thermocouple DT monitoring redundant.
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3.0 EVALUATION The lic?nsee identified two areas of compliance with the PNPS licensing basis associated with monitoog the thermocouple DT. These are compliance with the ASME Code Section lil design stress limits and compliance with 10 CFR 50 Appendix G fracture toughness requirements.
Compliance with the ASME Code Section ill design criteria is documented in Combustion Engineering (CE) Report CENC-1139, " Analytical Report for Pilgrim Reactor Vessel," 1971.
Section lli contains a requirement that the range of primary plus secondary stress intensity not exceed a SS,, limit (80 ksi). The maximum range of primary plus secondary stress intensity reported in CENC-1139 is 38 ksi for the 100'F/hr heatup transient. This evaluation considered loads from all sources: pressure, bolt load, and the heatup transient. The licensee also indicated that the maximum calculated thermocouple DT of f,3*F occurs during the cooldown transient and the calculated stress intensity range for the cooldown transient is lower. The maximum heatup/cooldown rate, controlled by plant TS, is limited to 100*F/hr when averaged l
over a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. Therefore, the licensee concluded that ample margin exists in meeting the i
ASME Code Section lli design criteria for primary plus secondary stress intensity as long as the heatup/cooldown rates are maintained within the 100*F/hr TS limit.
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The licensee also examined the temperature measurements recorded over several outages. The licensee compared the measured thermocouple DT during initial startup testing for a cooldown rate that approached 100'F/hr with the DT predicted by the analysis in CENC-1139. The measured thermocouple DT of 90'F was in reasonable agreement with the calculated value of 83'F cited above. On the basis of this comparison, the licenses further concluded that measurement of the heatup and cooldowr' rate is sufficient to ensure that the ASME Code Section ill limits for primary plus secondary stress intensity have not been exceeded.
The licensee was unable to determine the original basis for the TS thermocouple DT limit of 145'F. The licensee performed an evaluation, using the analytical technique contained in CENC-1139, to determine the maximum DT that could be allowed at the location of the thermocouple and still meet the ASME Code Section lli criteria for primary plus secondary stress intensity range. The licensee's calcuisted thermocouple DT corresponding to this stress intensity range was 149'F for the heatup and 166'F for the cooldown. Since stresses during the cooldown corresponding to a thermocouple DT of 166*F are still within the acceptable code limits, the licensee concluded that the TS requirement to limit the flange-to-shell DT to less than 145'F is unnecessary.
The staff agrees with the licensee's assessment that the therrpocouple DT monitoring is redundant. The licensee's requirement to monitor and maintain the heatup/cooldown rdes within the 100*F/hr TS limit provides an adequate basis to monitor compliance with the ASME Code Section 111 design limit for primary plus secondary stress intensity in the reactor vessel shell.
A second concem is the fracture toughness limits specified in 10 CFR 50 Appendix G. The licensee reviewed the basis for the TS regarding reactor vessel pressure-temperature (P-T) limits
- and concluded that the Section ll1 stress intensity requirement is more limiting than the Appendix G fracture toughness requirement for the flange-to-shell stresses. The licensee did not request any revision to current P-T limits, therefore, the calcu'.ation sheet provided by the licensee regarding the P-T curve for the closure flange region is considered to be for information only. In future P-T submittals, the licensee should discuss this information for the clesure flange'along
_with the P-T limits for the most limiting beltline material.
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3 The staff finds the safety limits provided by the 145'F thermocouple DT limit and associated monitoring requirement in Technical Specification 3.6.A.1 and 4.6.A.1 are redundant to the TS controlled heatup/cooldown rates and therefore, are unnecessary. The 145'F thermocouple DT limit and associated monitoring requirement in Technical Specification 3.6.A.1 and 4.6.A.1 are also bounded by the TS heatup/cooldown rates and therefore, this limit does not need to be in the TS. Based on the above, the staff concludes that this proposed TS change' is acceptable.
4.0 STATE CONSULTATION
in accordance with tf e Commission's regulations, the Massachusetts State Official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC 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 previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public j
comment on such finding (63 FR 23304). Accordingly, the amendment meets -
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 asr assment need be prepared in connection with the issuance of the amendment.
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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
J. Fair S. Sheng A. Wang Date: June 19, 1998 r
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