ML18094B234
| ML18094B234 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/09/1990 |
| From: | Crimmins T Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NLR-N89223, NUDOCS 9001180402 | |
| Download: ML18094B234 (5) | |
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Public Service Electric and Gas Company Thomas M. Crimmins, Jr.
Public. Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4700 Vice President - Nuclear Engineering
(
i JAN 0 9 1990 NLR-N89223 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR ADDITIONAL INFORMATION NUREG 0737 ITEM II.D.1 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 The purpose of this letter is to provide additional responses to questions 10 and 13 for Unit 2 and question 13 for Unit 1 for the NRC request for additional information letter dated January 4, 1989 regarding NUREG 0737, Item II.D.1. provides the final response for question 13 for Unit 1 thus completing the Unit 1 responses.
The evaluations presented therein have determined that minor modifications will be made to strengthen the Unit 1 piping supports downstream of the anchor at elevation 131'4".
Modifications will be performed during the 1992 Refueling Outage.
As stated previously, the thermal-hydraulic inputs for Unit 2 are unavailable.
PSE&G will reconstruct the Unit 2 thermal-hydraulic analysis and evaluate the Unit 2 piping downstream of elevation 131'.
This review will be completed by June 30, 1990.
Any modifications found to be necessary will be performed during the Unit 2 Fall 1991 refueling outage.
Should you have any questions, please feel free to contact us.
Attachment
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PDR ACOCK 05000272 P
Document Control Desk NLR-N89223 c
Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector 2
Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, _NJ 08625 IJAN O 9 1990
NLR-N89223 ATTACHMENT 1 Question 10 -
Thermal Hydraulic Analysis Inputs In the description of the thermal hydraulic analysis of the safety valve and PORV discharge conditions, the Impell stress report only defined the valve opening pressure and pressurization rate used in the analysis (Reference 2).
It did not give details on the important input parameters used in the analysis.
Provide information on the peak pressure developed at the safety valve and PORV inlets and other key parameters used in the computer calculations such as piping model node spacing, computation time interval, choked flow locations, etc. so that the review of the thermal hydraulic analysis can be completed.
Question 13 - Portion of Discharge Piping Not Analyzed The stresses in the piping downstream from the pipe anchor (Anchor C-PRA-144) at elevation 131 ft. 4 in. to the pressurizer relief tank were not addressed in the Impell stress reports.
The Licensee contended that this portion of the piping was isolated from the Reactor Coolant Pressure Boundary (RCPB) by the pipe anchor, and the piping was dispensable and could be excluded from the analysis.
Although the piping downstream of the anchor at elevation 131 ft. 4 in. is outside of RCPB, the failure of this piping may still affect valve operability.
For instance, if the failure of the downstream pipes results in excessive pipe deformation, the discharge of flow may be restricted to such an extent that the valve can no longer function properly.
If an abrupt rupture of the pipe occurs, the pipe whip may cause severe damage to the upstream piping and event the valve itself.
Therefore, the piping from the anchor at elevation 131 ft. 4 in.
to the discharge tank cannot be ignored.
Provide an analysis of the piping and supports downstream of the anchor to ensure that the integrity of the piping within the RCPB and the operability of the valves will not be adversely affected by the downstream piping.
RESPONSE
A bounding analysis was conducted for Unit 1 to provide assurance that the PORV discharge piping from the anchor at elevation 131' 4 11 to the discharge tank will perform its design basis function and not compromise adjacent safety related equipment in the event of a PORV and Safety Valve discharge.
The analysis consisted of a review of both piping and support loadings.
As previously discussed in our response of August 1, 1989, piping stresses remain within code allowable limits for "Emergency Condition" (Level C).
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NLR-N89223 The load cases for the support analysis include deadweight, thermal, seismic, and RVA (Rapid Valve Actuation).
The RVA has 2 cases, all steam and all water discharge.
The steam case was selected as governing as its resultant loadings were approximately double the "all water" case.
The results of the support analysis indicate that the Unit 1 PORV discharge piping supports downstream of the elevation 131' 4 11 anchor do remain intact and permit the supported piping to perform its design basis function without compromise of adjacent equipment.
The stresses observed in a limited number of support components do exceed Code Faulted Condition criteria (Level D);
however, the use of collapse load criteria based upon material ultimate strengths do indicate maintenance of both physical and functional continuity.
The maximum Unit 1 support load sustained is 40,785 lbf. acting on tandem snubbers whose combined faulted allowable is 47,200 lbf.
The maximum load on a single snubber is 21,063 lbf.,
compared to a faulted allowable of 23,600 lbf.
Supplemental steel, welds, concrete inserts, and similar components were evaluated and, although exceeding faulted load criteria in a limited number of cases, do show ability to maintain functionality with respect to collapse load criteria.
Due to the conservatisms inherent in the analysis, as discussed in a.
previous submittal, and the extremely short temporal duration of the peak loadings, the current design is assessed to be acceptable for the near term.
Minor modifications will be made during the 1992 Unit 1 refueling outage to bring these latter miscellaneous support components up to Level D criteria, thus providing a more balanced design.
Should a PORV and Safety Valve actuation occur in the interim, a physical inspection of the piping and supports will be made, with any indicated repairs being performed.
Our previous response dated August 1, 1989, indicated that with the formation/relocation of the Nuclear Department, magnetic media files documenting the thermal hydraulic (TH) analysis inputs for the Unit 2 PORV discharge piping are unretrievable; only the outputs of the TH runs are available, which were used as inputs to the loading/stress analyses.
In the aforementioned response, it was proposed to perform a physical dimensional similarity analysis between Unit 1 and Unit 2 with the intent of providing a degree of assurance that because the units' piping was dimensionally similar, the similarity of the TH code output which generated the "downstream" loadings was reasonable.
The results of this analysis indicate that the Unit 2 PORV/Safety Valve piping upstream of the anchor is not dimensionally similar to Unit l's piping when considered in light of solid mechanics parameters such as space coordinates, directional changes, Page 2 of 3
NLR-N89223 support locations, flexibility, etc.
Similarity may exist between the units, however, when viewed in light of parameters critical to TH excitation of the downstream piping, i.e. total path length, conduit volume, etc. may indeed have sufficient similarity to provide assurance of the reasonableness of downstream loading.
This has not yet been shown, however, configuration of the piping below elevation 131 is very similar between Unit 1 and Unit 2 and it is PSE&G's belief that the loadings experienced in Unit 2 would be of the same order of magnitude as Unit 1 piping, given similar excitation.
Rather than pursue the upstream similarity approach, PSE&G ~ill reconstruct the thermal hydraulic analysis for Unit 2 using the same assumptions used in the Unit 1 analysis.
This analysis will be used to confirm the piping and support loadings in Unit 2 piping downstream of elevation 131.
The analysis will be performed by June 30, 1990.
Any piping modifications necessitated by the results of this analysis will be performed during the Unit 2 Fall 1991 refueling outage.
Based on the results found in Unit 1, PSE&G believes the Unit 2 piping and supports to be adequate pending completion of this work.
Should Salem Unit 2 experience a safety/relief valve lift in the interim, the piping will be inspected and any repairs performed as required.
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