ML20141H409
| ML20141H409 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/17/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141H407 | List: |
| References | |
| NUDOCS 9707230134 | |
| Download: ML20141H409 (6) | |
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4 UNITED STATES j
NUCLEAR REGULATORY COMMISSION
%*****/g WASHINGTON. D.C. 20066-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.198 AND 181 TO FACILITY OPERATING LICENSE N05. DPR-70 AND DPR-75 P_UBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY SALEM NUCLEAR GENERATING STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By letter dated June 18, 1996, as supplemented August 19, 1996, April 28, 1997, and June 11, 1997, the Public Service Electric & Gas Company (the licensee, PSE&G) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos. I and 2, Technical Specifications (TSs). The requested changes would change TS 5.2.2, " Design Pressure and Temperature," by adding design parameters for Main Steam Line Break (MSLB).
The MSLB analysis results in a higher containment air temperature than the current value in TS 5.2.2.
The August 19, 1996, April 28, 1997, and June 11, 1997, letters provided clarifying information that did not change the initial proposed no si_gnificant hazards consideration determination nor the Federal Reaister notice.
2.0 EVALUATION The licensee stated in its June 18, 1996, letter that recent calculations have determined that a maximum temperature of 351.3*F, concurrent with a pressure of 25 pounds per square inch gauge (psig), could exist following an MSLB. The temperature specified in TS 5.2.2 was 271*F.
(This discrepancy was reported in LER 272/95-016-0, dated August 18,1995.) The proposed change would eliminate the temperature of 271*F and replace it with, " Containment air temperatures up to 351.3*F are acceptable providing the containment pressure is in accordance with that described in the (Updated Final Safety Analysis Report) UFSAR."
The licensee reviewed the structural analysis of the containment concrete and liner plate, and the environmental qualification of equipment, to determine the significance of the increase in temperature.
9707230134 970717 PDR ADOCK 05000272 P
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5^ I 2.1 Structural Considerations L
i The major differences in the design of the concrete containments between the Salem units.and most other. nuclear plants are that (1) the Salem design used i
the factored load design for all load combinations including the structural integrity. test (SIT), which should be considered on the basis of the working i
stress design, and (2) the steel liner is considered to contribute to the
' tensile strength of the containment, which is generally not considered.
The yield strengths of the steel materials are:
60 ksi for reinforcing bars and 32 ksi for the liner. The compressive strength of concrete is 3,500 psi.
i The liner is attached to the concrete shell wall through the use of 1/2 in.
i diameter studs spaced at 15 inches horizontally and vertically.
The inside diameter of the containment is 140 feet, and the concrete wall thickness is 4.5 feet for the cylinder portion with a liner thickness of three-eighths of an inch. The hoop reinforcing bar area is 12.8 square inches per foot.
The combined effects of temperature and pressure on the reinforced concrete wall and the steel liner are analyzed for the loss-of-coolant accident (LOCA) i and MSLB conditions respectively, taking into account the interaction between the liner and the concrete for the load combination consisting of pressure and temperature only and the results of the analysis are as follows:
i LOCA:
rebar stress = 36.927 ksi, liner stress = 3.48 ksi (tension)
MSLB:
rebar stress = 27.964 ksi liner stress = 21.09 ksi-(compres.).
The staff performed the above calculation without using any load factor because the staff's. evaluating the leaktight integrity of the liner, not the structural integrity of the containment. The above results, as expected, confirm the general observation that the LOCA pressure controls the rebar stress and the MSLB temperature controls the compressive stress of the liner.
For a steel liner under compression, there is concern that the liner may buckle.
If it does buckle, the buckling may lead to cracking of the liner and the shearing off of the anchors, jeopardizing the leaktight integrity.
The licensee provided the information on the original design of the liner and liner anchor. system. The original liner analysis indicates the critical buckling stress for the 3/8-inch liner is more than twice the yield stress of 32 ksi-for the liner steel material.
Since the liner stress resulting from MSLB is only 21.09 ksi, buckling should not occur. However, due to possible differences in workmanship, plate thickness, yield stress, alignment, etc.,
the actual critical buckling load may be more or less than that calculated.
If in two adjacent panels, one panel has a lateral deflection and the other does not, the critical buckling load of the plate with the deflection will be lower than the plate without the deflection.
Such a condition gives rise to a shearing force across the anchor stud.
Shearing force was considered in the original design of the anchor stud by estimating the potential strain in the liner when it yields.
From the strain displacement of the liner, a
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4 calculation was made to determine if the deflection of the stud can i
accommodate the displacement without giving rise to a high shear stress in the 1
stud. The tension in the stud is estimated by assuming the stud will act as a lateral support to prevent the liner buckling when it is subjected to a compressive stress of 25 ksi. The lateral load is taken as 2% of the liner i
buckling load, which is assumed to be the compressive load in the liner due to 1
the 25 ksi compressive stress in the liner.
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PSE&G did not perform a systematic analysis of the. liner-anchor system and there are no specific design calculations for the original design.
It was analyzed piecemeal through a number of assumptions and judgements.
- However, i
PSE&G claimed:
"Three tensile and three shear tests were performed on test assemblies approximating as close as possible to the actual welded stud i
configuration used in the field. These test specimen studs pulled from the liner at 74,500-80,600 psi in tension and 62,600-67,000 psi in shear. Neither failure mode affected the leaktight integrity of the liner plate." The tests gave the ultimate capacities of the liner-anchor system and were not intended to validate the analysis for the LOCA condition or any other condition.
PSE&G i
used the test results to envelope those from its empirical analyses. The i
licensee committed to provide an analysis of the containment liner anchorage i
for the loading induced on the containment liner during an MSLB event to confirm the assumptions in the PSAR and the UFSAR.
The staff found this to be 1
l acceptable, and considers this to be a condition to the License.
R To reinforce its position that the steel liner was adequately designed, PSE&G l
l provided the staff with information about the original steel liner design as contained in the Preliminary Safety Analysis Report (PSAR).
The information.
i is summarized in a table which shows the hoop and the vertical stresses in the liner and.the resulting Marain of Canability for each load combination.
For tension, it is defined as the ratio between the yield stress to the maximum tensile stress component.
For compression, it is the ratio between the critical buckling stress for combined compression and the combined compressive stress.
From the tabulated results, we observed that the margin of capability is greater than 1.0.
For the adequacy of the stud anchors, the PSAR cited the tests done at the University of Illinois under the sponsorship of the manufacturer of the Nelson studs. There was extensive. speculative discussion i
about the behavior of the liner and its anchors but there is no detailed i
analysis to support what was discussed.
It is not clear how the stresses in the table were obtained. The licensee will address the stresses in the j
analysis mentioned above.
l
. Modifications required for the reactor coolant pump platform have been 1
completed on Unit 2.
Modifications required for the Unit I coolant pump j
platform and containment spray piping support are to be completed before tne restart.
PSE&G has provided a set of sketches showing how various g
j modifications are carried out to relieve the thermal stresses due to the increased temperature. The staff has reviewed the modifications and found them to be appropriate.
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The staff has reviewed the information provided by PSE&G for the Salem TS change with respect to the containment temperature resulting from MSLB which is higher than the temperature in the original TS resulting from a LOCA.
It i
is clear that LOCA pressure governs the structural integrity of the containment and MSLB temperature controls the steel liner design, i.e., the leaktight integrity of the containment.
In order to be leaktight, the liner should not be cracked as a result of the liner being subjected to bending and/or tensile stress.
Such a condition exists when a liner panel is under compression and buckles while the adjacent panels retain their original configuration, resulting in. a shear across the anchor with the potential of separating the liner from its anchor. The ultimate tension and shear forces obtained in tests represent the maximum values for the liner-anchor system used.
However, in the information provided there is no quantitative computation of the shear and tension force applied on the liner-anchor system due to the load condition involving the high temperature.
The licensee will provide this information in the analysis mentioned above.
In summary, we find the proposed changes acceptable in the structural area based on the following:
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- a. The structural integrity of the concrete containment is maintained because it is governed by the LOCA pressure loading, which is greater than the pressure loading of the MSLB.
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- b. Prior to restart, modifications will be completed on the reactor coolant i.
pump platforms (both units) and the containment spray piping supports (Unit 1 only).
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- c. Testing was performed to demonstrate the adequacy of the liner and liner-i anchor system.
The test results provide the staff with reasonable j
assurance that the liner will maintain its'leaktight integrity following j
the MSLB.
t Additionally, as mentioned above, by letter dated June 11, 1997, the licensee has committed to perform an evaluation of the containment liner anchorage by November-30, 1997, for the loading induced on the containment liner during an MSLB event to confirm the assumptions provided in the PSAR and the UFSAR.
Accordingly, an appropriate license condition will be included in Appendix C of the Operating Licenses for the Salem units.
2.2 -Environmental Qualification Supplement 4 to the staff's Safety Evaluation Report for Salem (NUREG-0517) accepted 350*F for use in equipment qualification for Unit 2.
In its June 18, 1996, letter, the licensee stated that the increase from 350*F to 351.3*F does not have a significant. impact on equipment qualification.
By telephone call, the staff requested verification that the equipment inside containment had been reviewed against the higher temperature.
By letter dated August 19, 1996, the licensee stated that it had reviewed the vendor equipment qualification data against the revised temperature profiles and that this
-. review supported the qualified status of the equipment and demonstrated that the operability of the equipment would not be jeopardized by the increase in containment. temperatures. The staff concludes that the small increase in temperature is generally of little or no consequence. Accordingly, the' proposed maximum temperature of 351.3F-is acceptable.
Further, the
. licensee's evaluation provides -adequate assurance that safety-related equipment will function as required during accident conditions.
However, as a separate initiative outside the scope of this evaluation, the staff may review the adequacy of the licensee's analytical methodology.
2.3 Information in UFSAR The proposed change refers to the UFSAR for some of the information regarding the design parameters of containment pressure and temperature.
(In response to-a staff verbal request, the licensee identified the specific sections of the UFSAR that will contain this information in its letter of August 19, 1996.) Changes to these design parameters are controlled by the requirements of 10 CFR 50.59.
Furthermore, these design parameters are related to existing TS Limiting Condition for Operations (LCOs) that establish acceptable requirements for ensuring that the performance of the containment and reactor coolant system is maintained and that any changes which may impact safety would receive prior NRC review and approval.
Since the features with a potential to impact safety are sufficiently addressed by the LCOs, and since the associated design features, if altered in accordance with 10 CFR 50.59, would not result in a significant impact on safety, the criteria of 10 CFR 50.36(c)(4) for including these design features in the TS are not met.
Therefore, the staff concludes that referring to the UFSAR for the information regarding these design features is acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined 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 is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment 'on such finding (61 FR 37302). Accordingly, the amendments meet 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 assessment need be prepared in connection with the issuance of the amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
C. P. Tan Date: July 17, 1997 1