ML20210P923
| ML20210P923 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/30/1986 |
| From: | Jaffe D Office of Nuclear Reactor Regulation |
| To: | Opeka J NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-2.D.1, TASK-TM NUDOCS 8610060861 | |
| Download: ML20210P923 (9) | |
Text
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SEP 3 01986 Distributio Docket No. 50-336 "DocketFile_7]
DJaffe NRC PDR PKreutzer Local PDR ACRS(10)
P8D8 Reading Mr. John F. Opeka, Senior Vice President OGC-Bethesda Nuclear Engineering and Operations EJordan Northeast Nuclear Energy Company JPartlow P. O. Box 270 NThompson Hartford, Connecticut 06141-0270 Gray File 3.5a
Dear Mr. Opeka:
We are in the process of reviewing information which you have provided in response to TMI Action Item II.D.1, " Relief and Safety Valve Testing".
In order that we may continue our review, we request that you respond to the enclosed questions within 60 days following receipt of this letter.
l This request for information affects fewer than 10 respondents; therefore, l
OMB clearance is not required under P.L.96-511.
Sincerely, M/
D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B
Enclosure:
As stated cc: w/ enclosure See next page I
I PCD-B:
PB PBD-8:
1 PJ0redutzer DJa ch AThadani 9g/86 9/q/86 9/ /86 u2.neh PSRS W /F(,
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Mr. John F. Opeka Millstone Nuclear Power Station b
Northeast Nuclear Energy Company Unit No. 2 cc:
Gerald Garfield, Esq.
Mr. Wayne D. Romberg Day, Berry & Howard Superintendent Counselors at Law Millstone Nuclear Power Station City Place
'P. O. Box 128 l
Hartford, Connecticut 06103-3499 Waterford, Connecticut 06385 Regional Administrator, Region I Mr. Edward J. Mroczka U.S. Nuclear Regulatory Commission Vice President, Nuclear Operations Office of Executive Director for Northeast Nuclear Energy Company Operations P. O. Box 270 631 Park Avenue Hartford, Connecticut 06141-0270 King of Prussia, Pennsylvania 19406 Mr. Charles Brinkman, Manager Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.
7910 Woodmont Avenue Bethesda, Maryland 20814 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Utilities Service Company ATTN: Mr. Richard R. Laudenat, Manager Generation Facilities Licensing Post Offic'e Box 270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Radiation Control Unit Department of Environmental Protection State Office Building Hartford, Connecticut 06106 Mr. Theodore Rebelowski U.S. NRC P. O. Box 615 Waterford, Connecticut 06385-0615 Office of Policy & Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106
r ADDITIONAL QUESTIONS ON THE MILLSTONE, UNIT 2, NUREG-0737, ITEM II.D.1, SUBMITTAL 1.
The plant block valve has a pressure sealed bonnet (the PS in B9-45PS-13MS) while the test valve had a vertical stem, bolted bonnet (the 48 in B10-30548-13MS). Discuss the differences in the valves due to these different designs and discuss what impact these differences may have on valve operability.
2.
In response to the question on the block valve / operator combination at Millstone-2 NNECO stated that, based on calculations performed by Velan and Limitorque, it was concluded that the block valve / operator arrangement was adequate to ensure proper operation.
It is the staff position that it is not adequate to base such conclusions solcly on manufacturer's calculations. The problems encountered with Westinghouse gate valves on closing, which were traced to the calculations used to size the valve operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination. NNECO should provide test data to
. demonstrate the SMB-000-5 operator at Millstone-2 is capable'of providing adequate torque to close the block valve, or alternatively to set the torque switch at a setting that will produce a closing torque equal to the lowest level that successfully diosed the test valve ( A torque of 82 ft-lbs successfully closed the 3 in. Velan Model 810-30548-13MS test valve) 3.
Provide the maximum expected backpressure and bending moment for the Millstone-2 PORVs.
4.
Dresser, Ind., in March 1976, recommended to Metropolitan Edison Co.
that the PORV block valve be closed at pressures below 1000 psig to I
prevent steam wirecutting of the PORV seat and disk.
In a subsequent j
letter to GPU Nuclear (January 1984) it was recommended that heavier i
springs be used in the main and pilot valves to help keep the valves
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tight at low pressures. Testing by Dresser later showed the 1000 psig pressure limit to be overly conservative and that the PORV as designed l
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was qualified to system pressures of 100 psig.
Below 100 psig the deadweight of the lever on the pilot valve was sufficient to keep the pilot valve open. Dresser recommends that heavier springs be used under the main and pilot disks to ensure closure if the plant is to operate at pressures below 100 psig. Without the heavier springs recommended by Dresser, the PORV should not be used at system pressures below 100 psig.
From the information received so far it is not clear that NNECO has installed the heavier springs in the PORVs at Millstone-2.
If the heavier springs have been installed, state when the heavier springs were or will be installed consistent with the Dresser recommendations since the minimum pressure is below 100 psig.
5.
The Dresser PORV tested by EPRI experienced a problem with delayed closure in loop seal tests. The closure delays ranged from 2 see with a hot (3210F) loop seal to 70 sec with a 1000F loop seal. The 3210F loop seal test is representative of the Millstone-2 I
configuration with 3250F loop seals.
Have the Millstone-2 loop seal temperatures been verified by field measurements? If not, the staff position is that_such measurements be made to verify the assumed temperature is correct. CEN-213 stated that, because the delay in closing during the hot loop seal test was only 2 sec, the operability of thq Millstone-2'FORV is acceptable. The staff position is that further information is needed in order to accept the Dresser PORV/ loop seal configuration at Millstona-2 as operable. This is because no reason has been given to explain why the valve stuck open. Also, as seen from the cold loop seal tests, the potential exists for the valve to delay closing for a significantly longer period of time. Provide l
information to show the Dresser PORV at Millstone-2 will operate as designed. Also show the problems which led to the. delayed closures with the Dresser PORV in the EPRI tests have been resolved.
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6.
Testing with the Dresser PORV only included inlet conditions where the loop seal temperature was less than the system fluid (eg., 3210F loop seal followed by 6500F steam). When operating in a low-temperature overpressure protection mode, the PORV could experience a thermal transient where the loop seal temperature was greater than the system fluid (eg., 3250F loop seal followed by 1000F water).
Provide data to demonstrate whether the Dresser PORV will operate as designed with these inlet conditions. Provide information to show the Dresser PORV delayed closure problems observed in the high pressure loop seal tests will not occur when the valve undergoes this type of thermal transient.
7.
Your response to our request for information dated January 11, 1985 stated the design basis at Millstone-2 does not take credit for operation of the PORVs and, therefore, the PORV control circuitry will not be qualifiediunder 10 CFR 50.49.
It was also stated the PORV control circuitry at Millstone-2 was designed to withstand a pressure of 69 psig, 107 rad, temperature of 2000F, and 90% relative
' humidity. This infcrmation, by itself, is not sufficient to demonstrate qualification of the control circuitry at Millstone-2 under NUREG-0737.
In order to demonstrate the Millstone-2 control circuitry i
is qualified, the information provided must be compared to the environment the control circuits will be exposed to. To allow a complete review of the qualification of the control circuitry for the PORV under NUREG-7037, provide the following:
A.
Provide a list of all PORV control circuitry needed to mitigate NUREG-0737 transients such as the following:
1.
Switchgear 2.
Motor control centers 3.
Valve operators and solenoid valves 4.
Motors 5.
Logic eouipment 6.
Cable 7.
Connectors 3
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8.
Sensors (pressure, pressure differential, temperature, flow and level, neutron, and other radiation) 9.
Limit switches
- 10. Heaters
- 11. Fans
- 12. Control boards 13.
Instrument racks and panels
- 14. Electric penetrations
- 15. Splices
- 16. Terminal blocks B.
For each item of equipment identified in 1, provide the following:
1.
Type (functional designation) 2.
Manufacturer 3.
Manufacturer's type number and model number 4.
Plant ID/ tag number and location C.
For each item of equipment listed in above, provide the environmental envelope, as a function of time, that includes all extreme parameters, both maximum and minimum values, expected to occur during NUREG-0737 transients, including post-accident conditions.
D.
For each item of equipment identified above, state the actual qualification envelope simulated during testing (defining the duration of the environment and the margin in excess of the design requirements).
If any method other than type testing was used for qualification, identify the method and define the equivalent
" qualification envelope" so derived.
E.
Provide a summary of test results that demonstrates the adequacy of the qualification program.
If any analysis is used for qualification, justification of all analysis assumptions must be provided.
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F.
Identify the qualification documents that contain detailed supporting information, including test data, for items D and E.
8.
The load combinations used to qualify the safety valve Class 1 and Class 2 piping and supports did not consider an emergency condition.
Justify not using an emergency condition or analyze the piping and supports using this type of load. Also, the upset condition for the Class 1 piping did not include a deadweight component in the equation.
Provide your reasoning for not including this effect in the upset load combination or reanalyze the upset condition including the effect of deadweight. Finally, several stress limits where missing from Appendix III in your answer to our request for information Provide this information.
9.
Provide more information on the verification of STEHAM and WATAIR by Stone & Webster.
Provide comparisons of the results for STEHAM and WATAIR calculations and EPRI/CE data to verify these codes are appropriate tools to evaluate piping discharge transients.
10.
Insufficient detail was received on the key parameters used in the STEHAM and WATAIR thermal-hydraulic analyses. Provide node diagrams of the thermal hydraulic models. Provide information on the node spacing, time step size, valve flow area and opening time, and choked flow locations used in the analyses. Discuss the rationale for their selection.
Since the Dresser 31739A valve passed in excess of 118% of rated flow, justify use of the ASME rated flow in the thermal-hydraulic analyses or provide the results of thermal-hydraulic and structural analyses which account for the larger flows seen in the tests. Compare the rated and test measured flow rstes for the M111 stone-2 PORVs.
If the test flows exceed the rated flow, provide the same information for the PORVs that was requested above for the safety valves.
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- 11. Justification for use of a cutoff frequency of 50 Hz is required. The submittal stated that a cutoff frequency of 50 Hz was used in the piping analysis. The cutoff frequency of 50 Hz appears to be too low for piping analysis based on EG&G Idaho, Inc. experience. Most piping analyses use a cutoff frequency of 100 Hz.
In a valve discharge, the fluid forces on the piping have frequencies as high as 100-120 Hz. The safety valve inlet pressure oscillations have frequencies as high as 180-260 Hz. The solution time step is dependent upon the expected frequency range of the solution, which is a function of the frequency of the driving force, and the frequency response of the structural system. A minimum number of time steps are required to accurately calculate the dynamic response of the structural system.
Since at least 8 time steps are required to define a single cycle, the minimym time step when the driving force or expected structural response has a frequency of 120 Hz should be about 0.001 s.
Large time steps can only be used if the expected frequency range of the solution is lower, or less accurate results are acceptable.
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- 12. The piping loads for the M111 stone-2 SRV discharge piping were analyzed j
assuming 100% steam quality.
Operation of the pressurizer spray system l
during an overpressure transient would add moisture to the steam volume l
at the top of the pressurizer. This moisture could add to the piping l
loads during a safety valve discharge.
It would also affect the expected valve inlet conditions at the plant. Provide a discussion on whether this effect was considered in determining valve inlet conditions and in the analysis doim to select the transient producing l
the maximum loads on the safety valve discharge piping.
- 13. The letter from M111 stone-2 on June 11, 1985 stated the structural l
analysis of the safety valve discharge piping and supports had been I
completed. This analysis showed the support and the piping stresses to l
be less than their allowables.
Information on the safety valve and l
PORV inlet piping and PORV discharge piping was not included. To complete our review, for both piping and supports, provide a table comparing the calculated and allowable stress for tne most highly loaded locations.
This information should be provided for the inlet l
and discharge piping for the safety valves and the PORVs.
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- 14. Pressurizer nozzle loads during safety valve and PORV discharge were not discussed.
Compare the calculated and allowable loads for the pressurizer nozzles.
- 15. The Combustion Engineering (CE) inlet conditions report listed the FSAR transients and accidents for each plant which result in a peak pressure greater than the safety valve setpoint.
For some plants this list
. included the feedwater line break (FWLB), but for other plants the FWLB was not included. Millstone 2 was a plant that did not include the FWLB in its list of transients and accidents that challenge the safety valves.
From the CE report it was not clear whether the FWLB was missing because the accident did not challenge the safety valves or because Millstone 2 was licensed prior to the issuance of Regulatory Guide 1.70, Rev. 2 and, therefore, the FWLB was not analyzed as part of Millstone 2's design basis. Discuss why the FWLB was not listed for MillstoneUnit2l If the FWLB was not listed for the second reason discussed above, it is the staff position that the Millstone 2 submittal is incomplete.
Item II.D.1 in NUREG-0737 specifically requires that PORVs and safety valves be qualified for fluid conditions resulting from transients and accidents referenced in Regulatory Guide 1.70, Rev. 2.
The FWLB is specifically defined in Regulatory Guide.1.70, Rev. 2.
Additionally, from the staff review of other plant-specific responses to Item II.D.1, it is clear that for may plants the FWLB accident is the limiting case for providing high
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pressure liquid to the safety valves, a fluid for which they were not specifically designed originally. This is exactly the type of concern that NUREG-0737, II.D.1, was established to address.
In accordance with the requirements of the NUREG, we require that information be provided to demonstrate that the PORVs and safety valves will function as required to assist in safe shutdown cf the plant and will not experience any degradation that would inhibit safe plant shutdown if exposed to the FWLB.
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