ML18054A755
| ML18054A755 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/31/1988 |
| From: | Finaman C, Nalezny C, Pace N EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML18054A756 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-8333, NUDOCS 8905240021 | |
| Download: ML18054A755 (26) | |
Text
TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING PALISADES DOCKET NO. 50-255 C. P. Fineman C. L. Nalezny N. E. Pace December 1988 EGG-NTA-8333 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6492 I
ABSTRACT Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system.
As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TM! Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. This report documents the review of these programs and their results by the Nuclear Regulatory Commission (NRC) a"nd their consultant, EG&G Idaho, Inc.
Specifically, this report documents the review of the Palisades Licensee response to the requirements of NUREG-0578 and NUREG-0737.
This review found the Licensee has not provided an acceptable response and, thus, has not reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.
FIN No. A6492--Evaluation of OR Licensing Actions-NUREG-0737, II.D.1 ii
CONTENTS ABSTRACT..............................................................
i i
- 1.
INTRODUCTION.....................................................
1 1.1 Background.................................................
1 1.2 General Design Criteria and NUREG Requirements.............
1
- 2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM................
4
- 3.
PLANT SPECIFIC SUBMITTAL.........................................
6
- 4.
REVIEW AND EVALUATION............................................
7 4.1 Valves Tested..............................................
7 4.2 Test Conditions............................................
8 4.3 Valve Operability..........................................
10 4.4 Piping and Support Evaluation..............................
13
- 5.
EVALUATION
SUMMARY
18 5.1 NUREG-0737 Items Fully Resolved............................
18 5.2 NUREG-0737 Items Not Fully Resolved........................
19
- 6.
REFERENCES.......................................................
22 TABLE 4.3.1 Valve pefformance data for the Dresser 31739A valve 12 iii
~
TECHNICAL EVALUATION REPORT TM! ACTION--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING PALISADES DOCKET NO. 50-255
- 1.
INTRODUCTION
1.1 Background
Light water reactor experience has included a number of instances of improper performance of relief and safety valves installed in the primary coolant systems.
There were instances of valves opening below set pressure, valves opening above set pressure, and valves failing to open or reseat.
From these past instances of improper valve performance, it is not known whether they occurred because of a limited qualification of the valve or because of basic unreliability of the valve design.
It is known that the failure of a power operated relief valve (PORV) to reseat was a significant contributor to the Three Mile Island (TMI-2) sequence of events. These facts led the task force which prepared NUREG-0578 (Reference 1) and, subsequently, NUREG-0737 (Reference 2) to recommend that programs be developed and executed which would reexamine the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. These programs were deemed necessary to reconfirm that the General Design Criteria 14, 15, and 30 of Appendix A to Part SO of the Code of Federal Regulations, 10 CFR, are' indeed satisfied.-
1.2 General Design Criteria and NUREG Requirements General Design Criteria 14, 15, and 30 require that (1) the reactor primary coolant pressure boundary be designed, fabricated, and tested so as to have extremely low probability of abnormal leakage, (2) the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions are not 1
exceeded during normal operation or anticipated transient events, and (3) the components which are part of the reactor coolant pressure boundary shall be constructed to the highest quality standards practical.
To re~onfirm the integrity of overpressure protection systems and thereby assure that the General Design Criteria are met, the NUREG-0578 position was issued as a requirement in a letter dated September 13, 1979, by the Division of Licensing (DL), Office of Nuclear Reactor Regulation (NRR), to ALL OPERATING NUCLEAR POWER PLANTS.
This requirement has since been incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements, which was issued for.implementation on October 31, 1980.
As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:
- 1.
Conduct testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
- 2.
Determine valve expected operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Rev. 2.
- 3.
Choose the single failures such that the dynamic forces on the
- safety and relief valves are maximized.
- 4.
Use the highest test pressure predicted by conventional safety analysis procedures.
- 5.
Include in the relief and safety valve qualification program the qualification of the associated control circuitry.
5; Provide test data for Nuclear Regulatory Commission (NRC) staff review and evaluation, including criteria for success or failure of valves tested.
2
- 7.
Submit a correlation or other evidence to substantiate that the valves tested in a generic test program demonstrate the functionability.of as-installed primary relief and safety valves.
This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR).
The effect of as-built relief and safety valve discharge piping on valve operability must be considered.
- 8.
Qualify the plant specific safety and relief valve piping and supports by comparing to test data and/or performing appropriate analysis.
3
- 2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM In response* to the NUREG requirements previously listed, a group of utilities with PWRs requested the assistance of the Electric Power Research Institute (EPRI) in developing and implementing a generic test program for pressurizer safety valves, power operated relief valves, block valves, and associated piping systems.
Consumers Power Company, the owner of Palisades, was one of the utilities sponsoring the EPRI Valve Test Program.
The results of the program, which are contained in a series of reports, were transmitted to the NRC by Reference 3.
The applicability of these reports is discussed below.
EPRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves under conditions which bound actual plant operating conditions.
EPRI, through the valve manufacturers, identified the valves used in the overpressure protection systems of the participating utilities and representative valves were selected for testing. These valves included a sufficient number of the variable characteristics so that their testing would adequately demonstrate the performance of the valves used by utilities (Reference 5).
EPRI, through the Nuclear Steam Supply System (NSSS) vendors, evaluated the FSARs of the participating utilities and arrived at a test matrix which bounded the plant transients for which over pressure protection would be required (Reference 6).
EPRI contracted with Combustion Engineering (CE) to produce a report on the inlet fluid conditions for pressurizer safety and relief valves in CE designed plants (Reference 7). Since Palisades was designed by CE, this report is ~elevant to this evaluation.
Several test series were sponsored by EPRI.
PORVs and block valves were tested at the Duke Power Company Marshall Steam Station located in Terrell, North Carolina. Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, California. Safety valves were tested at the Combustion Engineering Company, Kressinger Development Laboratory, which is located in Windsor, Connecticut.
The results of the relief and safety valve tests are reported in Reference e The results of the block valve tests are reported in Reference 9.
4
The primary objective of the EPRl/CE Valve Test Program was to test each of the various types of primary system safety valves used in PWRs for the full range of fluid conditions under which they may be required to operate.
The conditions selected for test (based on analysis) were limited to steam, subcooled water, and steam to water transition. Additional objectives were to (1) obtain valve capacity data, (2) assess hydraulic and structural effects of associated piping on valve operability, and (3) obtain piping response data that could ultimately be used for verifying analytical piping models.
Transmittal of the test results meets the requirements of Item 6 of Section 1.2 to provide test data to the NRC for the safety valves.
The Licensee has not justified that the EPRI test results are applicable to the new PORVs and PORV block valves (see Section 4.1) to be installed or provided other test data for review.
Therefore, Item 6 of Section 1.2 was not met for the new PORVs and PORV block valves.
5
- 3.
PLANT SPECIFIC SUBMITTAL A preliminary assessment of the adequacy of the overpressure protection system was submitted by Consumers Power on April 5, 1982 (Reference 10).
An initial assessment of the overpressure protection system was transmitted June 30, 1982 (Reference 11).
Consumers Power submitted their final assessment of the safety valves, PORVs, and block valves and the safety valve and PORV piping system on December 30, 1982 (Reference 12).
A request for additional information (Reference 13) was submitted to Consumers Power by the NRC on August 6, 1985.
They responded to this request on December 30, 1985 (Reference 14).
A second request for information was sent to Consumers Power on October 20, 1987 (Reference 15).
The Licensee responded to this request in several parts, March 29, 1988 (Reference 16),
May 25, 1988 (Reference 17), and September 14, 1988 (Reference 18).
The response of the overpressure protection system to Anticipated Transients Without Scram (ATWS) and the operation of the system during feed and bleed decay heat removal are not considered in this review.
Neither the Licensee nor. the NRC have evaluated the performance of the system for these events.
6
- 4.
REVIEW AND EVALUATION 4.1 Valves Tested Palisades utilizes three safety valves, two PORVs, and two PORV block valves in the overpressure protection system.
The safety valves are Dresser Model 31739A valves.
The plant safety valves are mounted directly on a pressurizer nozzle.
The safety valves have staggered setpoints of 2500, 2540, and 2580 psia.
The Dresser Model 31739A safety valve was one of the valves tested by EPRI.
The plant and test valves are identical except the plant valves were modified to restrict valve stem lift to -76% of full rated lift (0.34 in.
for the plant valves versus 0.45 in. for the test valve). This difference only affects valve capacity, not operability. Thus, the test valve is considered to adequately represent the plant valves.
The PORVs and block valves originally installed at Palisades are Dresser 31533VX-1 solenoid actuated pilot operated valves with a bore diameter of 1-5/8 in. and Walworth, Model 5262 PSCF BM, 2.5-in. gate valves with Limitorque'SMB-000-5 operators. Palisades does not have loop seals upstream of the PORVs.
However, in Reference 18, Consumers Power stated it is going to replace PORVs and block valves.
The new PORVs selected are two 4-in. Target Rock, Model 100, Y-type globe valves.
The new block valves selected are two 4-in. Rockwell wedge gate valves, Figure No. 12011.
The information supplied by Consumers Power in Reference 18 was not sufficient to evaluate the new valves to be installed at Palisades against the requirements of NUREG-0737, Item II.0.1. Therefore, this report will only evaluate the Palisades safety valves and the safety valve discharge piping.
The additional information Consumers Power needs to supply on the new PORVs and PORV block valves in order to complete the NUREG-0737, Item II.0.1, evaluation is listed in Section 5.2.
7
Based on the above, the safety valve tested is considered to be applicable to the in-plant valves at Palisades and to have fulfilled that part of the criteria of Items 1 and 7 as identified in Section 1.2 regarding applicability of test valves.
The information needed to complete the evaluation, of the new PORVs and block valves is listed in Section 5.2.
4.2 Test Conditions The valve inlet fluid conditions that bound the overpressure transients for CE designed PWR plants were identified in Reference 7.
The transients considered in this report include FSAR, extended high pressure injection (HPI), and, cold overpressurization events.
In addition, information on the inlet conditions for the feedwater line break at Palisades was presented in Reference 16.
For the safety valves, only steam discharge was calculated for FSAR type transients. The peak pressure was 2520 psia and the maximum pressurization rate was 45 psi/s. A maximum backpressure of 397 psia is developed at the safety valve outlet (Reference 19).
Palisades has the safety valves mounted directly on a pressurizer nozzle.
Consumers Power stated in Reference 12 the plant valve adjusting rings were set to -45 (upper), -30 (middle), and -2 (lower) in October 1981 based on Dresser's recommendations.
These positions are relative to the level position.
No EPRI tests were performed with ring settings of -45, -30, -2. Six steam tests with the Dresser 31739A valve were run with ring settings of
-48, -40, +11.
These were tests 316, 318, 320, 322, 1018, and 1104a.
Five of these tests (316, 318, 320, 322, and 1104a) were run on the short inlet piping configuration which is representative of the Palisades configuration.
One test (1018) was run on the loop seal configuration, but without the loop seal water.
Of these tests, only four are applicable to the Palisades valves because the valve lift achieved at 3% accumulation was
-BO%, or less, of rated lift. These were tests 320, 322, 1018, and 1104a.
Lift during these tests is representative of full lift for the Palisades valves where the lift is mechanically limited to -76% of the design lift. These tests had peak pressure from 2657 to 2720 psia and 8
)
pressurization rates greater than 308 psi/s. The peak backpressure in the tests ranged from 570 to 866 psia. These conditions bound those expected at Palisades.
One steam test (1012) was run with ring setting of -48,.-40, +3.
This test was run on the long inlet configuration with the loop seal drained.
This bounds the Palisades configuration with the safety valves mounted on the pressurizer nozzle.
The valve was at 72% of rated lift at 3%
accumulation, which is representative of full lift for the Palisades valves.
The test parameters were peak pressure, 2665 psia, pressurization rate, 309 psi/s, and backpressure, 564 psia. These conditions bound those at the plant.
In addition to the tests with the -40 middle ring setting, one steam test (314) was run with a middle ring setting of -20.
The exact lower ring setti~g for this test is not known, but the lower ring setting was somewhire between -13 and +11.
The valve achieved 79% of rated lift, which is representative of full lift for the Palisade~ valves.
The test parameters were peak pressure, 2680 psia, pressurization rate, 333 psi/s, and backpressure, 177 psia. These conditions, except for the backpressure, bound those at the plant. The effect of the lower backpressure in this test will be discussed in the section on operability.
Review of the CE inlet conditions report (Reference 7) showed that water did not reach the valve during FSAR transients or an extended high pressure injection (HP!) event.
The cutoff head for the Palisades HPI pumps is below the safety valve setpoint so that an extended HPI event would not challenge ~he safety valves.
In addition, the Licensee in Reference 16, discussed the results of feedwater line break analyses for Palisades that indicated the pressurizer did not fill during this accident so that steam inlet conditions were maintained.
There was a concern that the extended safety yalve blowdown (blowdown greater than 5%) observed during the EPRI tests could result in the pressurizer level increasing to the safety valve inlet.
CE, in Reference 19, analyzed the loss-of-load (LOLD) transient assuming 20%
9
)
blowdown.
Other conservative assumptions were also made to maximize pressurizer level swell. The LOLD was chosen because it provided the design basis for sizing pressurizer safety valves.
The 20% blowdown is conservative since the blowdown observed in the applicable EPRI tests ranged from 7 to 12.2%.
This analysis showed the pressurizer level did not reach the inlet to the safety valves. Thus, the steam inlet condition was maintained.
The test sequences and analyses described above, demonstrating that the test conditions bounded the conditions for the plant safety valves, verify that Items 2 and 4 of Section 1.2 were met, in that conditions for the operational occurrences were determined and the highest predicted pressures were chosen for the test. The part of Item 7, which requires showing the test conditions were equivalent to conditions prescribed in the FSAR, was also met.
The information needed to complete the evaluation of the new PORVs and block valves is listed in Section 5.2.
4.3 Valve Operability As discussed in the previous section, the Dresser 31739A safety valves at Palisades are required to operate with st~am inlet conditions only.
The EPRI test program tested the Dresser valve for the required range of conditions.
For the tests with the -48, -40, +11 ring settings (320, 322, 1018, and 1104a), the test valve opened at pressures from 2455 to 2580 psia (-1.8% to
+3.2% of t~e nominal set pressure), had stable behavior, and closed with 7 to 12.2% blowdown.
In tests 322, 1018, and 1104 the valve achieved from 77 to 80% of rated lift and passed from 107 to 113% of rated flow at 3%
accumulation.
In test 320, the 31739A valve achieved only 44% of rated lift and 64% of rated flow at 3% accumulation.
This was attributed to the very high backpressure (866 psia) in the test. The Palisades safety valve backpressure of 397 psia is bounded by the backpressure in tests 322 (609 psia) and 1018 (570 psia) where greater than rated flow was measured.
This indicates the valve was able to perform its safety function of opening, relieving pressure, and closing.
10
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In test 1012, with the -48, -40, +3 ring setting, the valve opened at 2490 psia, (-0.4% of the set pressure), had stable behavior, and closed with 10.7% blowdown.
The valve achieved 72% of rated lift at 3% accumulation and passed 99% of rated flow.
In test 314, with the -48, -20, -13 to +11 ring setting, the valve opened at 2537 psia, (+1.5% of the set pressure), had stable behavior, and closed with 7.8% blowdown.
The valve achieved 79% of rated lift at 3%
accumulation and passed 110% of rated flow.
As noted in Section 4.2, the backpressure for test 314 was less than the expected backpressure at Palisades.
To try determine the effect of backpressure and middle ring setting on the Dresser 31739A valve performance data from the tests in Table 4.3.1 below was reviewed.
This table shows the
-valve performance characteristics for the Dresser valve in tests that bound the Palisades middle ring setting and backpressure.
This table shows that lowering the middle ring setting 20 notches (from -20 to -40 or from -40 to
-60) with about the same backpressure (approximately 200 or 600 psia) resulted in a 30% increase in valve lift and a 9 to 12% increase in flow.
With the same middle ring setting, -40, an increase in the backpressure of 400 psia (from 200 to 600 psia) resulted in a decrease in the valve lift of 30% and a decrease of 9% in the flow.
Using this data to interpolate for a middle ring setting change of 10 notches, shows that lowering the middle*
ring 10 notches would increase the valve lift by 15% and increase the flow 6%.
Similarly for a backpressure increase of 200 psia, the valve lift would decrease 15% and the flow decrease by 5%.
Therefore, as shown in Table 4.3.1, for the Palisades valves where the middle ring is set at -30 and a peak*backpressure of 397 psia is expected, the valves should achieve approximately 78% of rated lift, which is equivalent to full lift at Palisades, and pass 110% of rated flow.
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TABLE 4.3.1.
VALVE PERFORMANCE DATA FOR THE DRESSER 31739A VALVE Test No.
Middle Ring Backpressure Percent of Percent of Setting (psi a)
Rated Lift Rated Flow 314
-20 177 79 110 316
-40 195 110 122 322
-40 609 78 113 324
-60 664 110 122 Palisades.
-30 397 78 110 (estimated)
Because the ring settings used with the Palisades safety valves were not tested by EPRI, Consumers Power Company took a similar approach to determining expected valve performance.
Consumers Power correlated the effect of ring settings on valve blowdown, lift, and flow rate (see Reference 14). Using a least squares linear regression (LSLR) technique data from applicable EPRI tests was correlated for% rated lift versus%
rate flow, ring setting versus% ~ated lift, and ring setting versus%
blowdown.
Using these correlations, Consumers Power determined the 31739A valve, with -45, -30, -2 ring settings, should produce 8% blowdown, 55% of rated lift, and 80% of rate flow.
Because the valve rated flow is approximately 298,000 lbm/h, 80% is 238,000 lbm/h.
This is 104% of the Palisades-FSAR required flow of 230,000 lbm/h.
These values are bounded by the estimated valve performance discussed above, which indicates the valves at Palisades should provide adequate overpressure protection.
A maximum bending moment of 241,738 in-lb was applied to the 31739A valve discharge flange without impairing valve operation. This bounds the maximum expected bending moment of 139,500 in-lb at the plant (see Reference 19).
12
')
For a test to be an adequate demonstration of safety valve stability, the test inlet piping pressure drop should exceed the plant pressure drop.
The measured stagnation pressure drop for the applicable tests was greater than 185 psid.
The plant specific stagnation pressure drop was calculated to be 158 psid.
In addition, the safety valves at Palisades are mounted directly on a pressurizer nozzle, resulting in the smallest pressure drop possible.
These considerations indicate the plant valves should be as stable as the test valves.
As noted above, the valve blowdown for the 31739A valve during the applicable tests ranged from 7 to 12.2%.
A CE analysis for a Palisades LOLD with 20% blowdown showed that the pressurizer level would not reach the safety valve inlet. This bounds the blowdown observed in the tests. Also, the hot leg remained subcooled during the LOLD analysis with the extended blowdown indicating adequate core cooling was maintained.
The presentation above, demonstrating that the safety valve tested operated satisfactorily, verifies that the portion of Item 1 of Section 1.2 that requires conducting tests to qualify the safety valves and the part of Item 7 requiring that the effect of discharge piping on operability be considered were met.
The information needed to complete the evaluation of the new PORVs and block val~es is listed in Section 5.2.
4.4 Piping and Support Evaluation In the p1p1ng and support evaluation, the piping between the welds on the flanges upstream of the three safety valves to the weld at the pressurizer relief tank nozzle were analyzed for the requirements of the USAS B31.l Code, 1967.
To calculate stresses, the primary stress incorporated a 0.75i factor, as introduced into ANSI B31.1 Code, 1973.
In addition, the allowables were 2.4 Su or 1.1 Sy (whichever is greater), for the primary stresses from normal loads and SSE.
The load combinations analyzed were originally based on the Palisades Final Safety Analysis Report (FSAR).
Since the original stress report was submitted December 30, 1982, the utility stated, in Reference 17, that the analysis was revised so that load combinations from both the Palisades FSAR and EPRI (Reference 22) were used.
13
In Reference 16, the Licensee stated the downstream pipe supports were analyzed for the_ requirements of the Palisades FSAR Update.
The FSAR Update methodology is consistent with that employed in IE Bulletins 79-02/79-14 implementation with regard to anchor bolts, baseplates, welds, and steel support components.
Vendor catalog data applicable to the load combination of concern was used for the load allowables.
Use of load combinations anci allowables from the plant FSAR is cohsidered acceptable.
One transient was analyzed. This case had the two lower setpoint safety valves lift at their respective setpoints of 2500 and 2540 psia.
The third safety valve and the PORVs remained closed.
The valve inlet condition was saturated steam.
Although the peak pressurizer pressure at Palisades (2520 psia) is below the setpoint of the two safety valves with the highest setpoints, the pressure ramp rate was maintained in the analysis until the valve with the higher setpoint opened.
The PORVs are used for low pressure overpressure protection at Palisades.
In Reference 17, the Licensee provided the results of an analysis of LTOP conditions for subcooled and saturated water discharge.
Simple hand calculations were used to try and bound the piping forces for a LTOP transient.
However, these conditions were based on the PORVs originally installed at Palisades.
As noted in Section 4.1, Consumers Power has committed to replacing the PORVs at Palisades. Therefore, the results of the LTOP analysis provided in Reference 17 no longer apply.
- However, analysis of the PORV discharge piping for LTOP conditions is not considered by the NRC staff to be in the scope of the Item II.D.1 review.
Therefore, this issue will be reviewed separately, at a later time, and will not impact the NUREG-0737, Item II.D.l, review for Palisades.
The thermal-hydraulic analysis was performed with RELAPS/MODl.
RELAPS/MODl was shown to be a suitable tool for performing the thermal-hydraulic analysis of valve discharge transients in Reference 23.
RELAP5 output is input to REFORC for calculation of piping loads.
REFORC was verified to accurately calculate these types of loads (Reference 17).
14
A RELAP5 model representing the piping from the pressurizer to the quench tank was assembled.
The model representing the lift of the-two safety valves contained 232 volumes and 242 junctions.
The piping system was modeled so that the short segments (less than S feet) of pipe immediately downstream of the valves were modeled with eight to ten volumes.
Farther down the line, larger nodes were used to minimize overall model size. The Licensee stated in Reference 17 that the use of the larger nodes in the downstream piping was verified by a sensitivity study in which the results of a calculation using the larger nodes were compared to the results of a calculation using approximately 0.5 ft nodes.
This study indicated the results were not significantly affected by the use of the larger nodes.
Also, a time of 2 x*10-4 s was used in the analysis (Reference 17).
The safety valves were assumed to open in 9.33 ms.
The safety valve opening time was based on the minimum pop time measured in the EPRI tests (12 ms) adjusted for the fact that Palisades valves are limited to 78% of full lift. The flow rate for the safety valves was based on EPRI test data.
In test 1104a, a flow of 318,690 lbm/h was measured at 3%
accumulation and 77% lift (77% lift is equivalent to full lift for the Palisades safety valves where lift is mechanically limited to 0.35 in.
compared to a full lift of 0.45 in.). Because the 318,690 lbm/h (1.39 times the rated flow rate of the Palisades safety valves) is larger than the calculated flow at 77% lift, it was used in the analysis. This approach would account for ASME derating and potential errors in the flow rate.
Thus, the important parameters used in the analysis are acceptable.
The structural analysis was performed using the EDS Nuclear program SUPERPIPE.
SUPERPIPE performs static, dynamic response spectra, and transient dynamic analysis.
SUPERPIPE was verified by extensive benchmarking against other piping analysis programs.
In the structural analysis, the damping factor was 1%.
The lumped masses were spaced so that mass points were placed at all elbows, valves, tees, and discontinuities in pipe routing.
In addition, at least one mass point was located between any two supports in the same global direction.
15
The damping and mass point spacing were chosen so that frequencies up to 150 Hz were accurately calculated.
The direct integration technique was used with SUPERPIPE and the integration time step was 0.001 s. This time step can accurately calculate frequencies up to approximately 125 Hz.
Although this frequency is smaller than that used to determine the mass point spacing and the damping, it is still 'considered adequate.
Also, in Reference 17, Consumers Power noted that its structural analysis consultant, Impell, performed a time step size sensitivity study and found that time steps of 0.001 and 0.0005 s gave similar results.
As noted above, the original structural analysis submitted by Consumers Power used'load combinations based on the Palisades FSAR.
Since then the analysis was revised to include those recommended by EPRI in Reference 22.
The revised analysis found that the piping system met the EPRI acceptance criteria for all EPRI load combinations.
Palisades FSAR criteria for the piping were violated downstream of the safety valve during a safety valve discharge.
The FSAR criteria, 1.2 Sh, for the downstream piping is normally applied to the safety related upstream piping.
The piping system met the EPRI criteria because the allowable used by EPRI, 1.8 Sh (50%
larger than the FSAR allowable), is more appropriate for the downstream piping. Therefore, the piping system was not modified to conform to the FSAR criteria. With respect to Item II.D.1 of NUREG-0737, meeting the EPRI acceptance requirements is acceptable, so use of the higher stress limit is acceptable.
In Reference 14, the Licensee stated that only one support was modified based on the structural analysis: This was one of the frame guides downstream of safety valve RV1040.
The support was modified by adding plate to close the channel sections. This increased the support's bending section substantially. Additional information on the support analysis was provided in Reference 16.
For the supports downstream of the valves, an additional four supports required modifications.
One support was already modified, and the remaining three supports will be modified during the next refueling outage.
16
The basis for the entire pressurizer design, including the nozzles, is CENC-1114 - Analytical Report for Consumers Power Pressurizer, 1969.
The design criteria for the pressurizer was per ASME Section III and certain code cases.
The design specification included fourteen load combinations to include the effects of 200 safety valve operations.
The original pressurizer nozzle analyses were performed using detailed interaction analyses which were not intended to be revised for incorporation of new loads.
The results of those analyses indicated a maximum primary plus secondary stress intensity range of 16.9KSI for the relief valve nozzle and 40.SKSI for the safety valve nozzle.
The design limit was 69.9KSI.
The maximum fatigue usage factor was determined to be 0.122.
The analysis concluded that thermal gradient stresses due to gross structural discontinuities are of no significance.
To supplement the original pressurizer nozzle design analysis, a simple pipe analysis was performed consistent with EPRI load combinations and acceptance criteria. All of the EPRI requirements were met.
In addition, the stress components of this evaluation were reviewed in terms of the maximum principal stress differences of CENC-1114.
It was concluded that allowables for primary plus secondary stress intensity range could include the added valve discharge loads and still meet 3Sm.
Fatigue usage is essentially unaffected.
The loading environment for the quench tank nozzle is simpler than that for the pressurizer. Analysis of the nozzle pipe indicated conformance to EPRI and FSAR requirements.
The nozzle/saddle weld was also evaluated and this weld met FSAR design requirements.
The analysis discussed above, demonstrating that a bounding case was chosen for the safety valve piping configuration, verifies Item 3 of Section 1.2 was met.
The analysis of the safety valve piping and support system verifies Item 8 was met.
The information needed to complete the evaluation of the new PORV inlet/outlet piping is listed in Section 5.2.
17
-*l
- 5.
EVALUATION
SUMMARY
The Licensee for Palisades has not provided an acceptable response to the requirements of NUREG-0737.
Therefore it has not reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.
The rationale for this conclusion is given below.
5.1 NUREG-0737 Items Fully Resolved Based on the information provided by the Licensee, the requirements of Item II.D.l of NUREG-0737 were partially met.
This includes Items 1 to 8 of Section 1.2 for the safety valves and safety valve discharge piping.
The Licensee participated in the development and execution of an acceptable relief and safety valve test program to qualify the operability of prototypical valves and to demonstrate that their operation would not invalidate the integrity of the associated equipment and piping.
The subsequent tests were successfully completed under operating conditions which~ by analysis, bound the most probable maximum forces expected from anticipated design basis events.
The test results showed that the safety valve tested functioned correctly and safely for all steam discharge events specified in the test program that were applicable to Palisades and that the pressure boundary component design criteria were not exceeded.
Analysis and review of both the test results and the Licensee justifications indicated the performance of the test valves and piping can be directly extended to the in-plant safety valves and safety valve discharge piping.
The plant specific safety valve discharge piping also was shown by analysis to be acceptable.
Therefore, the prototypical tests and the successful performance of the valves and associated components demonstrated that this equipment was constructed in accordance with high quality standards, meeting General Design Criterion No. 30.
18
5.2 NUREG-0737 Items Not Fully Resolved Based on the information provided by the Licensee, the following requirements of Item II.D.1 of NUREG-0737 as identified in Section 1.2 were not met for the new PORVs, PORV block valves, and PORV inlet/outlet piping.
Item 1:
Item 1, which requires conducting tests to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents, was not met.
This is because Consumers Power has not shown that tests on valves that are representative of the new valves were completed.
Item 2:
Item 2, which requires that valve expected operating conditions be determined through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Rev. 2, was not met.
This is because the Licensee did not provide any information to demonstrate that representative valves were tested under conditions based on the regulatory guide transients.
Item 3:
Item 3, which requires the forces on the safety and relief valves to be maximized, was not met for the PORV inlet/outlet discharge piping. This because the new PORVs to be installed at Palisades have a flow rate that is 3.4 times larger than the old PORVs for which the PORV piping was analyzed.
Item 4:
- The part of Item 4 that requires the highest predicted pressure be chosen for the tests was not met for the new PORVs or PORV block-valves because information was not supplied to demonstrate this requirement was met for the new valves.
Item 5:
Item 5, which requires the qualification of the associated control circuitry, was not met.
Consumers Power has not demonstrated the control circuitry for the new PORV is environmentally qualified for the conditions possible during the accidents and transient where the PORV will operate.
19 J
Item 6:
Item 6, which requires that the Licensee provide test data for NRC staff review and evaluation, was not met.
The Licensee has not justified that the EPRI test results are applicable to the new valves*
to be installed or provided other test data for review.
Item 7:
The following parts of Item 7 were not met.
- a.
The part of Item 7 that requires the Licensee to submit a correlation or other evidence to substantiate the valves tested in a generic test program demonstrate the functionability of as-installed primary relief and safety valves was not met for the new PORVs or PORV block valves to be installed at Palisades.
The Licensee did not present sufficient information to conclude the tested valves demonstrate the operability of the valves to be installed at Palisades.
- b.
The part of Item 7, which requires showing that the test conditions are equivalent to those prescribed in the FSAR, was not met for the new PORVs and PORV block valves.
The Licensee did not compare the test inlet conditions for the valves that would be representative of the new Palisades valves to the conditions expected at the plant. Therefore, it cannot be assured that the test conditions bound the conditions expected at Palisades.
- c.
That part of Item 7 that requires consideration of the effect of as-built discharge piping on PORV operability was not met.
The maximum calculated bending moment on the Palisades PORVs was not compared to the maximum tested bending moment to ensure the calculated bending moment is bounded by the test moment.
Item 8:
Item 8, which requires qualification of the piping and supports, was not met for the PORV inlet/outlet piping and downstream of where the PORV line connects to the common header.
With the new PORVs to be installed at Palisades this portion of the piping must be reanalyzed to show it is qualified for the new loads.
Because the 20.
e,./ n, flow rate of the new PORVs exceeds the flow rate of the Dresser 31739A safety valves at Palisades by a factor of 1.75, the Licensee should confirm the loads on the safety valve discharge piping are still bounded by a safety valve discharge.
In providing the information needed to meet the items listed above, the Licensee should provide the information at a level of detail that is consistent with the information supplied in the original Palisades submittal and in the responses to the NRC requests for additional information.
Therefore, the. Licensee has not.demonstrated by testing and analysis that the reactor primary coolant pressure boundary will have a low probability of abnormal leakage (General Design Criterion No. 14) and that the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) were designed with a sufficient margin so that design conditions are not exceeded during safety valve/PORV events (General Design Criterion No. 15).
21
- 6.
REFERENCES
- 1.
TMI-Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.
- 2.
Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980.
- 3.
R. C. Youngdahl ltr. to H. D. Denton, Submittal of PWR Valve Test Report, EPRI NP-2628-SR, December 1982.
- 4.
EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July 1980.
- 5.
EPRI PWR Safety and Relief Valve Test Program Valve Selection/Justification Report, EPRI NP-2292, December 1982.
- 6.
EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, December 1982.
- 7.
Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Combustion Engineering-Design Plants, EPRI NP-2318, December 1982.
- 8.
EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.
- 9.
EPRI/Marshall Electric Motor Operated Block Valve, EPRI NP-2514-LD, July 1982.
- 10.
Letter D. J. VandeWalle, Consumers Power Co., to D. M. Crutchfield, NRC, "Palisades Plant, TMI Action Plan Item II.D.l, Relief and Safety Valve Testing," April 5, 1982.
- 11.
Letter B. D. Johnson, Consumers Power Co., to D. M. Crutchfield, NRC, "Palisades Plant, Response to NRC Letter Dated September 29, 1981, Subject, NUREG-0737, Item II.D.l.A, 11 June 30, 1982.
- 12.
Letter B. D. Johnson, Consumers Power Co., to D. M. Crutchfield, NRC,
Palisades Plant, NUREG-0737, Item II.D.1, Relief and Safety Valve Test Requ~rements, 11 December 30, 1982.
- 13.
Letter J. A. Zwolinski, NRC, to D. J. VandeWalle, Consumers Power Co.,
"Request for Additional Information to Support Safety Evaluation of Utility Responses to NUREG-0737, TMI Action Item II.D.1,"
August 6, 1985.
- 14.
Letter J. L. Kuemin, Consumers Power to, to Director, NRR, NRC, "Palisades Plant, TMI Action Plan Item II.D.1, Performance Testing of Relief and Safety Valves, Additional Information," December 30, 1985.
- 15.
Letter NRC to Consumers Power, "Request for Additional Information on NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves," October 20, 1987.
22
.)
~-
~-~
- 16.
Letter, R. W. Smedley, Consumers Power, to U. S. Nuclear Regulatory Commission, Document Control Desk, "Docket No. 50-255 - License DPR-20
- Palisades Plant - NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves (TAC NO. 44604) - Additional Information,"
March 29, 1988.
- 17.
Letter, R. W.* Smedley, Consumers Power, to U. S. Nuclear Regulatory Commission, Document Control Desk, "Docket No. 50-255 - License DPR-20
- Palisades Plant - NUREG-0737, Item II.D.l, Performance Testing of Relief and Safety Valves (TAC NO. 44604) - Additional Information, 11 May 25, 1988.
- 18.
Letter, R. W. Smedley, Consumers Power, to U. S. Nuclear Regulatory Commission, Document Control Desk, 11Docket No. 50-255 - License DPR-20
- Palisades Plant - NUREG-0737, II.D.l, Performance Testing of Relief and Safety Valves (TAC NO. 44604) - Additional Information, 11 September 14, 1988.
- 19.
Summary Report on the Operability of Pressurizer Safety Valves in CE Designed Plants, CEN-227, December 1982.
- 20.
Summary Report on the Operability of Power Operated Relief Valves in CE Designed Plants, CEN-213, June 1982.
- 21.
Letter R. S. Huffman, Dresser Inc., to R. J. Quinn, BG&E, 11Baltimore Gas and Electric Co., Calvert Cliffs Units 1 and 2, Power Operated Relief Valves CE P09903304 and P09903305, 11 August 12, 1985.
- 22.
EPRI PWR Safety and Relief Valve Test Program Guide for Application of Valve Test Program Results to Plant-Specific Evaluations, Rev. 2, Interim Report, July 1982.
- 23.
Application of RELAP5/MOD1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads, EPRI-2479, December 1982.
23