ML18065A277
| ML18065A277 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/13/1995 |
| From: | Gamberoni M NRC (Affiliation Not Assigned) |
| To: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 9511160251 | |
| Download: ML18065A277 (5) | |
Text
- ~-.*' *Mr. Richard W. Smedle~
Manager Licensing Palisades Plant November 13, ~5 1
27780 Blue Star Memorial Highway Covert, MI 49043
SUBJECT:
PALISADES PLANT - SAFETY ASSESSMENT OF COOLDOWN,EVENTS IDENTIFIED DURING THE 1994 DET
Dear Mr. Smedley:
In March and April 1994, NRC's Diagnostic Evaluation Team (DET) identified a number of operating events in which the cooldown rate of the primary coolant system (PCS) was exceeded and the PCS reached temperatures below the 70°F design limit with the reactor vessel head bolts fully tensioned.
In its June 15, 1994, letter, NRC forwarded to Consumers Powers Company the 1994 DET report which enumerated all the significant findings including the above-mentioned items.
The staff performed a detailed review of the licensee's engineering evaluation.
The staff and licensee analyses indicate that during the events, the materials in the Palisades reactor vessel met the fracture toughness requirements in Appendix G of 10 CFR Part 50 and 10 CFR 50.60 and met the safety margins required in minimizing the probability of brittle fracture (nonductile failure) of the PCS pressure boundary.
The staff's safety assessment is enclosed.
If you have further questions, please contact me at (301) 415-3024.
Enclosure:
Safety Assessment cc w/encl:
See next page DISTRIBUII.ON.:
Sincerely,
~
- 0. ~/ r-Marsha Gamberoni, Project Manager Project Directorate 111-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation
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Mr. R;chard W. Smedley Consumers Power Company cc:
Mr. Thomas J. Pa 1 mfaano
- Plant General Manager Pali sades Pl ant.
27780 Blue Star Memorial Highway Covert, M;chigan 49043 Mr. Robert A. Fenech Vice President, Nuclear Operations Pal;sades Plant 27780 Blue Star Memorial Highway Covert, Mich;gan 49043*
M. I. Miller, Esquire Sidley & Austin 54th Floor One First National Plaza Chicago,.Illinois 60603
.Mr. Thomas A. McNish Vice President & Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michig~n 49201 Judd L. Bacon~ Esquire *
.Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission-801 Warrenville Road Lisle; Illinois 60532-4351 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor R~~m 1 - Capitol Building Lansing, Michigan 48913 U.S. Nuclear Regulatory Commission Resident Inspector's Office Palisades Plant 27782 Blue Star Memorial Highway.
Covert, Michigan 49043 Palisades Plant Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. 0. Box 30195 Lansing, Michigan 48909 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.
Washington DC 20037 Michigan Department of Attorney General Special Litigation Division 630 Law Building P.O. Box 30212 Lansing, Michigan 48909 Sqitembcr 1995
SAFETY ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONSUMERS POWER COMPANY MATERIALS INTEGRITY SECTION MATERIALS AND CHEMICAL ENGINEERING BRANCH DIVISION OF ENGINEERING
1.0 INTRODUCTION
In an attachment to a letter dated June 15, 1994, an NRC diagnostic evaluation team (DET) id~ntified two occasions on which the primary coolant system (PCS) was cooled to below the temperature limit of 70°F with the reactor vessel head bolts fully tensioned and several occasions on which the PCS cooldown rate was exceeded while entering shutdown cooling. The licensee documented its evaluation of the former events in Palisades engineering analysis (EA)
EA-D-PAL-94-170-01 and of the latter events in Palisades EAs EA-E-PAL-93-029U-Ol and EA-E-PAL-93-029A-Ol.
These EAs concluded that none of the violations exceeded the pressure-temperature limits based on American Society of Mechanical Engineers (ASME) Appendix G calculation procedures and that the safety margins have been maintained in minimizing the probability of brittle fracture (nonductile failure) of the PCS pressure boundary.
The regulatory requirements for fracture prevention measures for light-water nuclear power reactors for normal operation are specified in 10 CFR 50.60.
10 CFR S0.60 requires that light-water nuclear power reactors meet the fracture toughness requirements in Appendix G to 10 CFR Part SO.
Appendix G to 10 CFR Part SO requires when the core is not critical, the pressure-temperature limits for the reactor vessel must be as conservative as those obtained by following the.methods of analysis ~nd the required margins of safety of Appendix G of the ASME Code.
Appendix G of the ASME Code requires
- that pressure-temperature 1 imits be calculated (a) using a safety factor of 2 on the principal membrane (pressure) stresses, (b) assuming a flaw at the surface with a depth of one-quarter of the vessel wall thickness (l/4t) and a length rif 6 times its depth, and (c) using a conservative fracture toughness (K1r) that is based on the lower bound of static, dynamic, and crack arrest fracture toughness tests on material similar to the Palisades reactor vessel material.
To satisfy the requirements of Appendix G of the ASHE Code, the K1r fracture toughness must be greater than the sum of two times the membrane stress intensity factor (2K1m) and the thermal gradient stress intensity factor (Kt) at the tip of a postulated l/4t surface flaw.
In addition, Appendix ~ of the ASME Code recommends that when the flange and adjacent shell regions are stressed by the full intended bolt preload and by pressure not exceeding 20 percent of the preoperational system hydrostatic test pressure, minimummelal temperature in the stressed.region should be at least the initial RTNDr temperature for the material fo the stressed regions plus any effects of irradiation at the stressed region.
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EN CLO.SURE->
- 2. 0 EVALUATION 2.1 Licensee's Evaluation The Palisades Technical Specifications {Amendment 117) specify a cooldown rate limit of 20°F at PCS temperatures less than 180°F.
Between August 1988 and September 1993 the Palisades reactor vessel experienced 11 codldown transients that violated the technical specification cooldown limit of 20°F/hr.
For these 11 transients, the PCS cooldown rates varied from 27°F/hr to 57°F/hr, the PCS temperatures varied from 80°F to 170°F, and the PCS pressure varied from 225 psig to 270 psig.
The licensee calculated PCS pressure limits for various PCS cooldown rate~ and temperatures.
The PCS pressure limits were calculated to meet the requirements of. Appendix G of the ASME Code.
Based on these calculations, the licensee determined that the PCS pressure limit was greater than the PCS pressure at the reactor vessel beltline for all 11 cooldown transients.. Therefore, these calculations indicate that the 11 cooldown transients met the requirements of Appendix G of the ASME Code.
The 70°F bolt-tensioning.requirement was a limit specified by Combustion Engineering.
The regulatory requirement related to bolt tensioning is the ASME Code Appendix G requirement that the flange and adjacent shell regions stressed by the full intended bolt.prelOad,have a minimum metal temperature in the stre~sed region of at least*-the initial RTNDT temperature for the material in the stressed regions plus any effects of irradiation at the stressed region.
The licensee stated that the vessel flange and head flange that are the highly stressed regions during bolt tensioning have initial reference temperatures of 60°F.
The vessel and head flange are far enough from the reactor'~ore to result in a negligible amount of neutron embrittlement.
Hence, the RTNoT of the vessel and head flanges at the time of the bolt tensioning is 60°F.
Since the reported PCS temperatures were only 3-5°F below the 70°F requirement, the vessel and head flanges met the requirements of Appendix G of the ASME Code.
2.2 Staff's Evaluation The licensee's evaluation of the 11 temperature transients that exceeded the 20°F/hr cooldown rate indicated that the limiting transient is the January 31, 1989, event.
In this event the lowest PCS temperature was 85°F and the PCS pressure was 245 psig.
The PCS temperature cooled from 250°F to 142°F in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> {15°F/hr) and cooled from 142°F to 85°F in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> {57°F/hr).
The rapid cooldown of 57°F/hr was caused by increased flow from the shutdown coolin~
loop when the primary loop pumps were shut off. This cooldown rate of 57 F/hr was a conservative assumption since the PCS cooled down more slowly before the primary loop pumps were shut off. The licensee also included measurement uncertain~ies of 30 psi for the pressure and 5°F for the temperature in its analysis.
Taking measurement uncertainties into account, the staff peiformed an independent calculation to determine whether the limiting transient complied with the requirements of Appendix G of the ASME Code.
In the safety evaluation associated with Amendment 163 to the Palisades Technical Specifications, the staff identified the limiting material with respect to A-2
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pressure-temperature limits as the beltline circumferential weld, which was fabricated using weld wire heat no. 27204. This heat of material has a best-
- estimate copper of 0.208% and best estimate nickel of 1.00I, which res~lts in a chemistry factor of 228.
Using (a) a neutron fluence of 1.8E19 n/cm, (b) the method documented in Regulatory Guide 1.99, Revision 2, to calculate the adjusted reference temperature, (c) a cooldown rate of 57°F, and (d) a lo~st PCS temperature of 80°F (5°F below the lowest PCS temperature during the event to account for temperature uncertainty.), Ahe K.r fracture toughness at the tip of a postulated \\f.4t flaw is 28.3 ksi-in1
- lite staff calculated the K1t value as 12.7 ksi-in.,2 for the Palisades reactor vessel wit~ a 57°F/hr cooldown rate. The staff calculated the 2K value as 15.4 ksi-in11 for the Palisades reactor vessel at a pressure of 2~ psig (this 'pressure is 30 psi greater than the PCS pressure during the event to account for pressure
- uncertainty). Since the sum of 2K1111 and K11 is less than Kir* the cool down event met the requirements of Appendix G or the ASME Code.
3.0 CONCLUSION
S
- 1. The licensee's evaluation indicates that the reactor vessel materials met.
the requirements of Appendix G of the ASME Code during the cooldown events
- and the bolt-tensioning events.
- 2. The.staff's calculations confirm that the limiting cooldewni e~vent met the requirements of Appendix G of the ASHE Code.
3 *. Th:e staff and licensee analyses indicate that during *the events, the materials in the Palisades reactor vessel met the. fractllfn! toughness
.requi*rements in Appendix G of 10 CFR Part 50 and 10.CFR 50.60 and met the safety margins required in minimizing the probability of brittle fracture (nonductile failure)*of the* PCS pressure boundary.
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