EGG-NTA-7861 TECHNICAL EVALUATION REPORT RELIEF AND …
Transcript of EGG-NTA-7861 TECHNICAL EVALUATION REPORT RELIEF AND …
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EGG-NTA-7861
TECHNICAL EVALUATION REPORTTMI ACTION--NUREG-0737 (II.D.1)-RELIEF AND SAFETY VALVE TESTING
MAINE YANKEEDOCKET N0. 50-309
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C. P. FinemanC.L. Nalezny
April 1988
Idaho National Engineering LaboratoryEG&G Idaho, Inc.
Idaho Falls, Idaho 83415
Prepared for theU.S. Nuclear Regulatory Commission
Washington, D.C. 20555Under DOE Contract No. DE-AC07-76ID01570
FIN No. A6492
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ABSTRACT
Light water reactors have experienced a number of occurrences of improper
performan'ce 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 TMI 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 woult! verify the integrity of the piping sistems
for normal, transient, and accident conditions. This report documents the
review of these programs by the Nuclear Regulatory Commission (NRC) and their
consultant, EG5G Idaho, Inc. Specifically, this report documents the review
of the Main Yankee Licensee response to the requirements of NUREG-0578 and
NUREG-0737. This review found the Licensee had provided an accaptable
response, reconfirming 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
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CONTENTS.
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ABSTRACT ............................................................. 11
1. INTRODUCTION .................................................... I
1.1 Background ................................................. 1
1.2 General Design Criteria and NUREG Requirements ............. 'l2. 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 Te s t . C o n d i t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
4.3 Operability ................................................ 12
4.4 Piping and Support Evaluation .............................. 15,
x .5. EVALUATION SUMMARY .............................................. 19
6. REFERdNCES ...................................................... 20
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TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1) RELIEF AND
SAFETY VALVE TESTING MAINE YANKEE DOCKET N0. 50-309
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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 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 fcilure 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 program 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 50 of the Code of i
! Federal Regulations, 10 CFR, are indeed satisfied.
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1.2 General Design Criteria and NUREG Reauirements
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 i
system and associated auxiliary, control, and protection systems be designed
wit.h sufficient margin to assure tilat the design conditions are not 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 reconfirm the integrity of overpressure protection systems and thereby
assure that tha 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 Pegulation (NRR), to ALL
OPERATING NUCLEAR POWER PLANTS. This requirement has since been incorporated
as Item II.D 1 of NUREC-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.
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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
k safety and relief valves are maximized.
4. Use the highest test pressure predicted by conventional safety
analysis procedures.
5. Include in the relief and sa'ety valve qualification program the
qualification of the associateo control circuitry.
6. Provide test data for Nuclear Regulattey Commission (NRC) staff
review and evaluation, including criteria for success or failure of
valves tested.
7. Submit a correlation or other evidence to substantiate that the
valves tested in a generic test program demonstrate the
function ability 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 oppropriate
analysis.
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q-\ 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 scfety valves, power operated relief valves,1 block valves, and
associated piping systems. Maine Yankee Atonic Power Company (MYAPCo), the
owner of Maine Yankee, 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 FWR safety, relief, and
block valves under conditions which bound actuel plant operating conditions.
EPRI, through the valve manufacturers, identified the valves used in the
everpressure protection system 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).
| EPRf 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 Maine Yankee was designed by CE, this
report is relevant to this evaluation.
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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 relief valves (SRVs) 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 8. The results of the
block valve tests are reported in Reference 9.
The primary objective o' the EPRI/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.
3. PLANT SPECIFIC SUBMITTAL
A preliminary assessment of the adequacy of the overpressure protection
system was submitted by MYAPCo on March 30, 1982 (Reference 10). An
evaluation of safety and relief valve operability was transmitted June 30,
1982 (Reference 11). In a letter dated August 5, 1082 MYAPCo submitted their
plant specific evaluation of valve inlet conditions (Reference 12). On
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December 30, 1982 MYAPCo sumarized their responses to the NUREG-0737q
requirements (Reference 13). Additional safety valve information and an yI
evaluation of the safety valve and PORV piping was submitted April 4, 1983 h|;
(Reference 14). A request for additional information (RAll was submitted by
the NRC on February 11,1985 (Refer'ence 15). MYAPCo responded to this request
on May 31, 1985 (Reference 16). A second RAI was submitted by the NRC on
December 31, 1986 (Reference 17), to which MYAPCo responded on March 31, 1987
(Reference 18) and August 13, 1987 (Reference 19).
The rcsponse 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.
d. REVIEW AND EVALUATION
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4.1 Valves Tested
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Maine Yankee utilizes three : safety valves, two PORVs, and two PORV block
valves in the overpressure protection system. The safety valve.c are Dresser
31709Ka valves. The safety valves have staggered setpoints of 2500, 2525, and
2550 psia. The plant safety valve inlet configuration consists of a long
inlet pipe but designed so that a water seal does not form at the valve
inlet. The PORVs are Dresser 31533-VX-30 solenoid actuated pilot operated
valves with a bore diameter of 1-5/16 in. Maine Yankee has a long inlet pipe
without a loop seal upstream of the PORVs. The block valves are 2-1/2 in, i
Anchor Darling gate valves with Limitorque SMB-00-15 operators.
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The Dresser 31709KA valve was not one of the valves tested by EPRI. The
31709KA valve uses a smaller orifice than either of the two dresser valves
tested b'y EPRI. It is closest in size to the 31739A valve, the smaller of the
two test valves. The most important difference only affects valve capacity,
not operability. These considerations, and the fact that all Oresser valves
are similar in configuration and design philoscphy, indicated the test valves
are representative of those at Maine Yankee.
The Dresser PORVs installed at Maine Yankee are of the dash 1
(31533-VX-30-1) design with a 1-5/16 in, bore diameter. .The valve tested by
EPRI was a dash 2 (31533-VX-30-2) design with the same bore size. The dash 2
design resulted from a need to improve the seat tightness and included
modifications to the internals, body, and inlet flange. The body and flange
modifi::ations were not of a nature that would affect operability. The Maine
Yankee valves have not been modified to incorporate the internals of the dash
2 design. No time table has been established by MYAPCo to convert their
valves to the dash 2 intervals. This is MYAPCo's policy because Dresser
Industries informed them the replacement internals have no effect on valve
operability (Reference 16). MYAPCo will ultimately convert their PORVs to the
dash 2 internals since they are they only type of replacement parts Dresser
now Scils but not until current stocks of replacerrent parts run out. Also,
Dresser Industries recomends that heavier springs be used under the main and
pilot disks to ensure closure at pressures below 100 psig. Because leakage
has not been a problem at low pressure, MYAPCo has not installed the heavier
springs at Maine Yankee. Based on this plant experience with the Dresser
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valves at. Maine Yankee, not installing the heavier springs is considered |
acceptable. At full system pressure, the spring force is small relative to i
the force from'the system pressure, so that not using the heavier springs does ,
not affect valve operability. Based on the statement by Dresser and the fact
S the PORVs will ultimately be converted to the dash 2 design, the test valve is
considered an adequate representative of the in-plant valves.i
The Anchor Darling block valves at Maine Yankee are 2-1/2 in, solid wedge !
gate valves, Model 4701-8300-21, with Limitroque SMB-00-15 operators. The
valves originally had SMB-00-10 operators but MYAPCo installed larger'
operators in 1984. The Anchor Darling test valve was a 3 in double disk gate
valve, Model 5J-1512, with a Rotork 30-NA1 operator. While the plant valves
are similar in operation to those used in the EPRI Test Program, the j
capability of the valve / operator combination in the plant specific
configuration. Following the Marshall Steam Station Tests, Maine Yankee, upon
consultation with the valve manufacturer (Anchor Darling), modified the
Limitorque SMB-00-10 operators used at that time to increase the maximum
closing torque delivered by the motor operators. After that modificatior, andt
prior to start-up following a refueling, an in situ test of the block valve's
ability to close against full steam flow was performed on July 9,1981.
During the 1984 Refueling Outage, the modified Limitorque SMB-00-10 operators
were replaced by environmentally qualified SMB-00-15 operators. The SMB-00-151
operators have gearing identical to the operacors used in the 1981 test in
order to provide the same closure force as the tested configuration (Reference
16). Therefore, the in plant valves are adequately represented by tested |
valves.
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l Based on the above, the valves tested are considered to be applicable to
the in-plant valves at Maine Yankee and to have fulfilled that part of the
criteria of Items 1 and 7 us ider.tified in Section 1.2 regarding applicability
of test valves.
A 4.2 Test Condition
The valve inlet fluid conditions that bound the overpressure transients
for CE designed PWR plants are identified in Reference 7. A plant specific
evaluation of valve inlet conditions for Maine Yankee was submitted in
Reference 12. The transients considered in these reports include FSAR,
extended high pressure injection (HPI), and low temperature overpressurization
events.
For the SRVs only steam discharge was calculated for FSAR type
transients. The peak pressure was 2589 psia and the maximum pressurization
rate was 63.1 psi /sec. A maximum backpressure of 457 psia is developed at the
SRV outlet (Reference 19). Maine Yankee has the SRVs i E nteo on a long inlet
pipe without a loop seal. MYAPCo stated in Reference 14 the plant valve
adjusting rings will be set at -48 (upper), -68 (middle), and 0 (lower).
These positions are relative to the level position.
The Dresser 31709KA valve at Maine Yankee was not one of the valves
tested by EPRI. MYAPCo, as part of a redesigning of the safety valve inlet
pipirg, used the COUPLE code developed by Continuum Dynamics, Inc. (CDI),to
determine safety valve ring settings and analyze valve operability. Reference
20 showed this to be a valid approach to determining plant valve ring .tttings
and operability. The valve was analyzed with COUPLE for steam flow coniitions
at 2575 psia.
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Review of the MYAPCo inlet conditions report (Reference 12) showed that
water did not reach the valve during FSAR transients but could reach the valve
inlet during an extended high pressure injection (HPI) event. The cutoff head
for the Maine Yankee HPI pumps is above the SRV setpoint and, therefore, the
extended I:PI event challenges the safety valves. But the analysis showed that
for the inadvertent actuation of the HPI at power, which is the limiting<
extended HPI evert at Maine Yar.kee, approximately 27 minutes are raquired to
fill the pressurizer. This allows ample time for the operator to terminate
the transient and prevent liquid flow through the safety valves.
There was a concern that the extended valve blowdown (blowdown greater
than 5%) observed during the EPRI tests could result in the pressurizer level
increasing to the safety valve inlet. CE, in a report transmitted with !
Reference 13, analyzed the loss-of-load (LOLD) transient assuming 20%
blowdown. Other conservative assumptions were also made to maximize
pressurizer level swell. The LOLD was chosen because it provided the design
basis for sizing the pressurizer safety valves. The 20% blowdown is
conservative since the blowdown calculated with COUPLE was 13.5% for all
valves discharging and 17% for one valve discharging. This analysis showed
the pressurizer level did not reach the inlet to the safety valves. Thus, the
steam inlet condition was maintained.
The two Dresser PORVs at Maine Yankee t.re mointed on a long inlet pipe
without a water seal. The peak pressure and pressurization rate for the PORVs
during FSAR type transients are the same cs the safety valves, P589 psia and !
63.1 psi /sec, respectively. The maximum backpressure for the PORVs was not
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provided by MYAPCo but it is expected to be bounded by the backpressure
calculated for the safety valves, 457 psia, because the analysis used to
determine the safety valve backpressure assumed all three safety valves and
both PORVs were oper and flowing.
The test valve was subject to fifteen steam tests. In the steam tests,
the peak pressure ranged from 2435 psia to 2505 psia. Backpressures ranged
from 170 psia to 760 psia. The testing of the Dresser PORV was performed at
peak pressures below that indicated in Reference 12 for Maine Yankee during an
FSAR transient (2435 to 2505 psia versus 2589 psia). Reference 6 stated that
the valve inlet pressure is considered to have a potential for affecting PORV
operation'only during opening or closing. Since the Dresser valve opens
quickly(lessthan5.5 seconds),thepressureincreaseduringthevalve
opening cycle is minimal (approximately 31.5 psia increase based on the
maximum pressurization rate of 63.1 psi /sec). Testing at the Maine Yankee
setpressure (2400 psia) or slightly above is, therefore, considered adequate
and the test conditions representative of the plant conditions.
As with the safety valves, Reference 12 indicated that water did not
reach the PORV during FSAR transients but would during an extended HPI event.
As noted above, the pressurizer would take approximately 27 minutes to fill,
allowing plenty of time for the operator to terminate the transient before the
PORY passed water.
The PORVs are used for low temperature overpressure (LTOP) protection at
Maine Yankee. For low temperature overpressure protection, the valve is
required to pass steam at pressures up to 564 psia, steam to water transition,
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and liqui'd at pressures up to 564 psia with temperatures ranging from 100*F to
479'F (Reference 21). The peak pressures noted above are based on analyses
that assumed the pressurizer was liquid full (Reference 12). The presence of'
a steam. bubble in the' pressurizer would limit the peak pressure when the PORY
opened on steam but this condition was not specifically analyzed. Thus, peak
pressure during steam discharge was bounded using the liquid full analyses.,
The steam discharge conditions are considered to be adequately represented by'
the high pressure tests discussed above. In addition results from low
pressure steam tests by Dresser Industries, the valve manufacturer, were
provided as part of the Calvert Cliffs submittal (Reference 22). Steam to
water transition is also considered to be adequately represented by the high
pressure transition test, 21/DR-85/W. Water discharge during a LTOP transient
is represented by the low pressure (-690 psia) water tests with fluid
temperatures ranging from 112*F to 459 F.
The block valve is required to open and close over a range of steam and
water conditions. The required torque to open or close the valve depends
almost ent' rely on the differential pressure across the valve disk and isi
rather insensitive to the momentum loading and, therefore, is nearly the same
for water or steam and nearly independent of the flow. Full pressure steam
tests, therefore, are adequate to demonstrate operability of the valve for the
required steam and water conditions. A full flow steam test was performed on
the plant valve during a refueling on July 9,1981 with the pressurizer
pressure at 2248 psig and the pressurizer level at 64%. The Marshall tests
also included an Anchor Darling 3" double disk gate valve with Rotork 16-NA1
and 30-NA1 operators which is similar to the Maine Yankee block valves. These
tests were full pressure steam (2445 psia) tests.
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<TheLtest sequences and analyses descHbed above, demonstrating that the. . *w ..
Etest' conditions bounded the conditions for the plant valves, verify that Items
2 and ~.4 of-:Section'1.2 were met, in:that conditions for the operational,
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p occurrencesLwere determined and the highest predicted pressures were chosen9
.for the~ test. The part of Item 7, which requires showing that the test
conditions.are equivelent to conditions prescribed in the FSAR, is also met.,
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4.3 Valve Operability
As discussed in the previous section, the Dresser 31709XA safety valves'
-c-at Maine Yankee are required to operate with steam inlet _ conditions only. The
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COUPLE analysis used to evaluate the Maine Yankee safety valves analyzed their
operability fer the required range of conditions. During FSAR transients the
PORVs are required to only pass steam. PORVs are used for LTOP protection and
in this mode may be required to pass steam, steam to water transition, an.d
water. . The test valve was subjected to the required conditions. The block
valves are also required to operate for steam and liquid flow conditions.
These _ valves were subjected to full pressure steam tests, the results of which
also apply to liquid flow.
In the COUPLE analysis the valve was assumed to open at 2575 psia.
During the calculation the valve had stable behavior and closed with 13.6%
' blowdown when all valves discharged and 17% blowdown when one valve'
discharged. The valve achieved 100% of rated lift and passed 100% of rated
flow at 3% accumulation. This indicates the valve was able to perform its
safety function of opening, relieving pressure, and closing.
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Bending moments of up to 20,144 ft-lb were applied to the 31739A valve
discharge flange during EPRI testing without impairing valve operation. -The
31739A valve was the smallest Dresser valve tested. This bounds the maximum
expected bending moment of 4071 ft-lb at the plant (see Reference 16).
For a test to be an adequate demonstration of safety valve stability, the
test inlet piping pressure drop should exceed the plant pressure drop. MYAPCo
provided a comparison of the plant calculated pressure drop to those measured
during the EPRI test program. On valve opening, the plant pressure drop is
238 psid compared to the EPRI test pressure drop of 643 psid. On closure, the
plant pressure rise is 72 psid compared to the EPRI test pressure rise of 150
psid. Therefore, the plant valves should be as stable as the test valves.
As noted above, the valve blowdown for the 31709KA valve during the
COUPLE analysis was 13.6% to 17%. A CE analysis for a Maine Yankee 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.
Based on the information discussed above, demonstration of safety valve
operability is considered adequate.
The Dresser PORV opened and closed on demand for all nonloop seal tests.
Inspection of the valve after testing at the Marshall Steam Station showed the
bellows had teveral welds partially fail. The failure did not affect valve|
| performance and the manufacturer concluded the failure did not have a
potential impact on valve performance. The bellows was replaced and did not
fail during any of the additional test series.
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The-results of tests done by Dresser Industries on a PORV similar to the
one at Main Yankee were provided as part of the Calvert Cliffs, Units 1, and
2, submittal. This data showed the PORV opened and closed on saturated steam
without failure at pressures ranging from 65 psia to 1979 psia. There was no
apparent leakage after closing in any of these tests. During other tests, the
minimum pressure achieved without leakage was 90 psia. This data indicates
the valve will operate acceptably with low pressure steam conditions.
A bending moment of 2125 ft-lb was induced on the discharge flange of the
test valve without impairing operability. The maximum bending moment
calculated for the Maine Yankee PORVs is less than 1534 ft-lb. The EPRI
tests, therefore, bound the expected plant condition.
The Maine Yankee PORVs are pilot operated valves that use system pressure
to hold the disk tight against the seat. At one point Dresser Industries
recommended the block valve be closed at system pressures below 1000 psig to
avoid steam wirecutting of the PORV disk and seat. Testing by Dresser later
showed the 1000 psig pressure limit to be overly conservative and that the
PORV as designed 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, if the plant is to operate at pressures
below 100 psig, that heavier springs be used under the the main and pilot
disks to ensure closure. It was also recomended by Dresser that the PORV
should not be used at system pressures below 100 psig without the heavier
springs. However, in Reference 16, the utility stated that they have not
experienced any problems with PORV leakage at low system pressures, not
installing the heavier springs is considered acceptable. The utility also
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stated that, because the only replacement parts Dresser now sells include the
heavier springs, they will.ul*.imately be used in the Maine Yankee PORVS.
Based on the valve performance during EPRI and Dresser tests, under the
full range of expected inlet con'ditions, the demonstration of relief valve
operability is considered adequate.
The.PORV block valve must be capable of closing over a range of steam and
. water conditions. ,As described in Section 4.2, high pressure steam tests are
adequate to bound operation over the full range of inlet conditions. As
~ described.in Section 4.1, the EPRI tests with the 3 in. Anchor Darling valvewe
and Rotork 30-NA1 operator and the full flow in situ test on the in plant
Ahchor ' Darling-valves are. adequate to demonstrate the operability of the inh
plant block valves. During the EPRI tests, the valve fully opened and closed
during all the applicable. tests with the Rotork 3-NA1 operator. During the
-situ test at the plant, the valve. closure time was less than 14 ceconds ar.d
,- measurements' verified that the valve time was less than 14 seconds andq,
[9 measurements verified that the valve closed leak tight (Reference 16),s ,
, | .NUREG-0737 'II.D.1 requires qualification of. associated control circuitry
|. as part of the safety / relief valve qualification. In Reference 19, however,
W; < MYAPCo stated the PORV control circuits at the Maine Yankee plant are
[ considered to be non-nuclear safety. This is because no credit is taken for
them in any accident analysis and because if a PORV opens 'and cannot be
closed, the PORV block valves, which are environmentally qualified pursuant to
10 CFr 50.49, will be used to isolate the event. Therefore, it can be'
concluded that the Maine Yankee PORY control circuitry meets the requirerents
of NUREG-0737, Item II.D.1.
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The presentation above demonstrates that the valves operated
satisfactorily and verifies that the part of Item 1 of Section 1.2 that
requires conducting tests to qualify the valves and that part of Item 7 that
requires the effect of discharge piping on operability be considered were met.
4.4 Piping and Support Evaluation
In the piping and support evaluation, the safety valve and PORY piping
between the pressurizer nozzles and the pressurizer relief tank were analyzed
for the requirements of the ANSI B31.1 Code, 1977 Edition. The load
combinations and acceptance criteria were equivalent to those proposed by EPRI
in Reference 23, although the acceptance criteria were slightly more
conservative. The supports were qualified to the requirements of the AISC bL
Code, Seventh Edition.
Fire transient conditions were analyzed. These conditions are shown in
Table 4.4.1, which was taken from Reference 16. The forces generated from
these conditions bound those from all other conditions expected at the plant.
The thermal-hydraulic analysis was performed with Stone & Webster's
programs STEHAM, WATAIR, and WATSLUG. STEHAM was used to analyze the high
pressure steam transients. STEHAM calculates the transient fluid dynamic
forcing functions acting on the pipe segments due to valve discharge. The
code computes thermal-hydraulic variables using the method of
characteri s tics. Forces on piping segments are computed by integrating the
rate of change of the fluid momentum within a centrol volume. For open pipe j
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segments, discharge blowdown forces are included. Time steps are selected
internally based on input segment lengths and the instantaneous sound speed.
The code was verified by Stone & Webster by comparing STEHAM result.s to those|
obtained with RELAPS/"001 and EPRI test results (both obtained from Reference
24). WATSLUG was used with STEHAM in the thermal-hydraulic analysis of the
water discharge portion of the extended HPI event. WATSLUG calculates the
forcing functions for a piping system during a water slug discharge event.
The code uses rigid body motion to describe the the water slug and an ideal
gas representation and rigid column theory to describe the steam or air to
track the water-steam or water-air interface. WATSLUG was verified in the
same manner as STEHAM. WATAIR was used to analyze the LTOP transient. WATAIR
calculates the one-dimensional transient flow field response and flow induced
forcing functions in a piping system. The code uses a Runge-Kutta integration
,nethod to integrate the governing two-phase fluid flow equations. Hand
calculations were used to verify WATAIR and the comparison of the hand
calculated values and the WATAIR results were in excellent agreement.
STEHAM, WATAIR, and WATSLUG models of the Maine Yankee presr rizer safety
and relief valve piping were developed. The critical input parameters for
STEHAM, WATAIR, and WATSLUG were reviewed and found acceptable. Valve opening
times were 0.015 s for the safety valves and 0.06 to 0.10 s for the PORVs.
These are representative of the opening times measured in the EPRI tests.
Time steps on the order of 1 ms were used in all the analyses. This is the
same order of magnitude time step used in the verification problems. Choked
flow was detected by the programs, as appropriate, at crea changes.
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The flow rates used for the PORVs and safety valves in the analysis were
also reviewed. The flow rates used for the PORVs are representative of the :
f,, maximum measured flow rates for the Dresser PORV under similar inlet
cor.ditions. .The flow rate used for the Dresser safety valves at Maine Yankee, !!
028,000 lbm/h, was slightly lower than would be expected based on the overall
performance e ne Di sser valves in the EPRI tests. In Reference 14, MYAPCo
noted the stamped rated flow for the 31709KA valves at Maine Yankee was
216,000 lbm/h. In the EPRI tests, the Dresser 31739A valve passed in excess
of 118% of rated flow and in Test 1008 the 31739A valve passed 111% of rated
flow (in Test 1008 the middle ring setting for the 31739A valve was -80 and
this was the ring setting used to scale the middle ring setting for the '
31709KA valves at Maine Yankee). Therefore, a minimum safety valve flow rate
of 240,000 lbm/h (111% of the 31709KA rated ficw) would have been more
appropriate to use in the thermal-hydraulic analysis. However, the
thermal-hydraulic analysis is still considered adequate for several reasons
noted by the utility in Reference 19. First, because Maine Yankee does not
use loop seals upstream of the safety valves, the forces and stresses
resulting from the safety valve discharge are not major contributors to the
system's high stress points. Second, the seismic accelerations used in the
piping analysis were very conservative and reducing the seismic accelcrations.
to either FSAR values or the NUREG-0098 value would counterbalance the,
increased loads due to higher safety valve flows. Therefore, the
thermal-hydraulic analysis is considered adequate.
The structural analysis was performed using Stone & Webster Engineering
Corporation's (SWEC) version of NUPIPE. This is a linear elastic piping
i structural analysis program widely used in industry which is fully verified
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for pipe stress analysis. The NUPIPE code was benchmarked by the NRC in 1979 :
as part of a five plant review conducted by SWEC. !
The key structural analysis parameters of lumped mass spacing,
integration time step, cutoff frequenc.y, and damping are adequate. Lumped
mass spacing was selected so the model contained at least three mass points
between restraints active in the same direction. Integration time steps were
0.001 s and the Licensee stated that a cutoff frequency of 400 Hz was used.
Althougo the stated cutoff frequency was 400 Hz, the tine step used in
analysis, 0.001 s, would only allow frequencies up to approximately 100 Hz to
be accurately calculated. However, this is still considered adequate. A
damping factor of 1% was used.
The results of the piping analysis showed the pipe stresses in the safety
valve and PORV inlet and outlet lines were less than their allowables.
Analysis of the supports showed some 19 locations where the loads or stresses
exceeded the allowable. Where this was true, appropriate modifications were
made. Modifications resulted from the new safety valve inlet piping, the
newlycalculatedsupportloads(deadicad, thermal, seismic,andvalve
discharge), and resolution of IE Bulletin 79-02. With the modification to the
support system, all stresses and loads in the piping and support systems are
within their allowables. MYAPCo stated the necessary modifications to the
support system were made during the 1984 refueling outage.
The discussion above demonstrates that a bounding case was chosen for the
piping configuration and verifies Item 3 of Section 1.2 was met. The analysi:
of the piping and support system verifies Item 8 wat met.
5. EVALUATION
The Licensee for Maine Yankee provided an acceptable response to the,
requirements of NUREG-0737, reconfirming that the General Design Criteria 14,
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15, and 30 of Appendix A to 10 CFR 50 were met with regard to the safety
valves and PORVs. The rationale for this conclusion is given below.
The Licensee participated in the developmcnt and execution of an
acceptabic 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 corditions which,
by analysis, bound the most probable maximum forces expected from anticipated [design basis events. The test results showed that the valves tested
functioned correctly and saf , for 'lli steam and water discharge eventsIspecified in the test pro :. that were applicable to Maine Yankee and thats
the pressure bour.dary component design criteria were not exceeded. Analysis I
and review of both the test resuits and the Licensee justifications indicated |
the perfonnance of the prototypical valves and piping can be directly extended
to the in-plant valves and piping. The plant specific piping also was shown by
analysis to be acceptable. '
Thus, the requirements of Item II.D.1 of NUREG-0737 were met (Items 1-8
in Paragraph 1.2) and, thereby, ensure that the reactor primary coolant
pressure boundary will have a low probability of abnormal leakage (General
Design Criterion No. 14). In addition, the reactor primary coolant pressure
boundary and its associated components (piping, valves, ard supports) were j
designed with a sufficient margin so that design conditions are not exceeded
during relief / safety valve events (General Design Criterion No. 15). Further,
the prototypical tests and the successful performance of the valves and
dssociated components demonstrated that this equipment was constructed in h
accordance with hQh quality standards, meeting General Design Criterion No.
30.
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6. REFERENCES,
1. TMI-Lessons Learced Task Force Status Report and Short-Term -
Recommerdations, NUREG-0737, November 1980.
2. Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980.
3. R. C. Youngdahl letter to H. D. Denton, Submittal of PWR Valve TestReport, EPRI NP-2628-SR, December 1982.
4. EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July1980.
5. EPRI PWR Safety and Rel_ief Valve Test Program ValveSelection / Justification Report, EPRI NP-2292, December 1982.
6. EPRI PWR Safety and Relief Valve Program Test ConditionJustification Report, EPRI NP-2460, December 1982.
7. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valvesin Combustion Engineering-Design Plants, EPRI NP-2318. December 1982. '
8. EPRI PWR Safety and Relief Test Program Safety and Relief Valve TestReport, EPRI NP-2628-5R, December 1982.
9. EPRI/ Marshall Electric Motor Operated Block Valve, EPRI NP-2514-LD,July 1982.
-10. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Preliminary Evaluationof Relief Valve Operation, March 30, 1982,
11. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safetyand Relief Valve Operation, June 30, 1982.
12. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safetyand Relief Valve Operation, August 5, 1982.
13. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safetyand Relief Valve Operation, December 30, 1982.
14. Letter J. H. Garrity, MYAPCo. to R. A. Clark, NRC, Evaluation of Safetyand Relief Valve Operation, April 4,1983.
| 15. Letter J. R. Miller, NRC, to J B. Randazza, MYAPCo, "Request forAdditional Infonnation on NUREG-0737, Item II.D.1, Performance Testing ofRelief and Safety Valves," February 11, 1985.
16. Letter G. D. Whittier, MYAPCo, to J. R. Miller, NRC, "Response to Requestfor Additional Information on Relief and Safety Valve Testing(NUREG-0737, Item II.D.1)," May 31 1985.
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