ML20207K243
| ML20207K243 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 03/31/1988 |
| From: | Fineman C, Nalezny C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20207K246 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-8035, NUDOCS 8810040112 | |
| Download: ML20207K243 (24) | |
Text
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1 EGG-NTA-8035 t
i TECHNICAL EVALUATION REPORT TMI ACT!0N--NUREG-0737 (!!.D.1)
RELIEF AND SAFETY VALVE TESTING 4
YANKEE R0WE j
DOCKET No.50-029 i
I C. P. Fineman I
C. L. Nalezny i
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i March 1988 I
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Idaho National Engineering Laboratory i
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EG4G Idaho, Inc.
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i Prepared for the i
j U.S. Nuclear Regulatory Comission Washington. 0.C.
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Under DOE Contract No. DE-AC07-761001570 1
FIN No. A6492 l
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i ggi$61d>tlA XA AYy-
1 ABSTRACY 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 TM1 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 by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
Specifically, this report documents the review of the Yankee Rowe Licensee response to the requirements of NUREG-0578 and NUREG-0737. This review found the Licensee l
had provided an seceptable response, reconfirming that the General Design i
Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.
i FIN No. A6492--Evaluation of OR Licensing Actions-NUREG-0737, !!.0.1 i
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CONTENTS J
ABSTRACT..............................................................
ii 1.
INTRODUCTION.....................................................
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1.1 Background.................................................
1 1.2 General Design Criteria and NUREG Requirements.*............
1 2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM................
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PLANT SPECIFIC SUBMITTAL.....................................'....
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REVIEW AND EVALUATION............................................
7 4.1 Valves Tested..............................................
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4.2 Test Conditions............................................
8 4.3 Operability................................................
11 4.4 Piping and Support Evaluation..............................
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S.
EVALUATION
SUMMARY
19 6.
REFERENCES.......................................................
20 TABLE i
s 4.4.1 Fluid conditions and critical parameters assumed for the thermal-hydraulic analysis.................................
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TECHNICAL EVALUATION rep 0RT TMI ACT!0N--NUREG-0737 (!!.D.1) RELIEF AND SAFETY VALVE TESTING __ YANKEE R0WE DOCKET NO.50-029 1.
INTRODUCTION 1.1 Backaround Light water reactor experience has included a number of instances of improper performance of relief and safety valves installed in the primary coolant systems. The*e 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 i
failure of a power operated relief valve (PORV) to resent was a significant contributor to, e 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 recomend 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 Criterh 14, 15 and 30 of Appendix A to Part 50 of the Code of Federal Regulations, 10 CFR, are indeed satisfied.
1.2 General Desian Criteria and NUREG Recuirements 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 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.
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e To reconfirm 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 (OL), Office of Nuclear Reactor Regulation (NRR), to ALL OPERATING NUCLEAR POWER PLANTS. This requirement has since been incorocrated as Item !!.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements, which was issued for implementation on 0'tober 31, 1980.
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As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:
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Conduct testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
4 2.
Determine valve expected operating conditions through the use of analyses of accidents and anticipated operational occurrences
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referenced in Regulatory Guide 1.70, Rev. 2.
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Choose the single failures such that the dynamic forces on the safety and relief valves are maximized.
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Use the highest test pressure predicted by conysr.tional safety
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analysis procedures, j
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Include in the relief and safety valve qualification program the qualification of the associated control circuitry, j
6.
Provide test dat. for Nuclear Regulatory Commission (NRC) staff review and evaluation, including criteria for success or failure of valves tested.
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Submit a correlation or other evidence to substantiate that the valves tested in a generic test program demonstrate the j
functionability of as-installed primary relief and safety valves.
l This correlation must show that the test conditions used are J
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equivalent to expected operating and accident conditions as prescribed in the Finai Safety Analysis Report (FSAR). The effect of as-built relief and safety valve discharge piping on valve
- stability 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.
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2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM l
In response to the NUREG requirements previously listed, a group of utilities with PWRs requested the assistance of the Electric Powee Research Institute (EPRI) in developing and implementing a generic test program for pressurizer safety valves, power operated relief valves, block valves, and J
associated piping systems.
Yankee Atomic Electric Company,(YAEC), the owner
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of Yankee Rowe, was not one of the utilities sponsoring the EPRI Valve Test l
Program. The submittals for Yankee Rowe, however, made use of the EPRI data. The results of the program, which are contained in a series'of I
reports, were transmitted to the NRC by Reference 3.
The applicability of j
these reports is discussed below.
EPR! developed a plan (Reference 4) for testing PWR safety, relief, and block valves under conditions which bound actual plant operating 4
j conditions.
EPRI, through the valve manufacturers, identified the valves j
used in the overpressure protection systems of the participating utilities and representative valves were selected for testing. These valves ine.luded 4
a sufficient number of the variakle characteristics so that their testing f
would adequately demonstrate the performance of the valves used by utilities i
(Reference 5).
EPRI, through the Nuclear Steam Supply System (NSFS) vendors, evaluated the FSARs of the participating utilities and arrived at a j
test matrix which bounded the plant transients for which over pressure protection would be required (Reference 6).
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EPR! contracted with Westinghouse to produce a report on the inlet l
fluid conditions for pressurizer safety and relief valves in Westinghouse designed plants (Reference 7).
Although Yankee Rowe was designed by l
Westinghouse, Westinghouse indicated thir plant was not covered by the inlet
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conditions presented in Reference 7.
YAEC presented a plant specific inlet l
conditions report as discussed later.
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Several test series were sponsored by EPR!.
PORVs and block valves l
were tested at the Duke Power Company Marshall Steam Station located in j
Terrell, North Carolina. Additional PORV tests were conducted at the Wyle
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Laboratori9s Test Facility located in Norco. California.
Safety Valves were tested at the Combustion Engineering Company Kressinger Development Labord-l tory, which is located in Windsor, Ccnnecticut.
The results of the relief and l
safety valve tests are reported in Reference 8.
The results of the block l
valve tests are reported in Reference 9.
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The primary objective of the EPRI/CE Valve Test Program was to test l
each of the various types of primary system safety valves used in PWRs for l
the full range of fluid conditions under which they may be required to j
operate.
The conditions selected for test (based on analysis) were limited to steam, subcooled water, and steam to water transition. Additional l
objectives were to (1) obtain valve capacity data. (2) assess hydraulic and l
structural effects of associated piping on valve operability, and (3) obtain
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piping response data that could ultimately be used for verifying analytical j
piping mooels.
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Transmittal of the test results reets the requirements of Item 6 of Secticn 1.2 to provide test date to the NRC.
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PLANT SPECIFIC SUBMITTAL i
A preliminary assessment of the adequacy of the overpressure protection system was submitted by YAEC on March 30, 1982 (Reference 10). An assessment of the pressurizer safety valves, PORV, block valve, and piping was transmitted July 1, 1982 (Reference 11).
Plant specific valve inlet conditions were provided August 1, 1982 (Reference 12).
Additional 1
information on test conditions justification and valve oportbility was j
provided December 28, 1982 (Reference 13).
In a letter dated Aprt) 1, 1983 YAEC transmitted their final safety and relief valve piping analysis (keference 14). YAEC's final safety valve operability report was submitted f
April 2, 1984 (Reference 15). A request for additional information I
(Reference 16) was submitted to YAEC by the NRC on July 16, 1985.
YAEC responded to this request on November 22, 1985 (Reference 17).
In Reference 18 YAEC provided an Overpressure Protection Report on the new j
Dresser 31719A safety valves installed at Yankee Rowe. This report provided i
new valve inlet conditions for the safety valves at Yankee Rowe.
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request for additional information was sent to YAEC on July 23, 1987 l
j; (Reference 19) to which YAEC responded on October 27, 1987 (Reference 20).
l Three final questions were transmitted to YAEC on November 23, 1987
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(Reference 21) to which the Licensee responded on January 8, 1988 l
(Reference 22).
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The response of the overpressure protection system to Anticipated Transients Without Scram (ATWS) and the operation of the system during feed 4
j and bleed decay heat removal are not considerec. in this review.
Neither the l
Licensee nor the NRC have evaluated the performance of the system for thest I
events.
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4.
REVIEW AND EVALUATION 4.1 Valves Tested Yankee Rowe utilizes two safety valves, one PORV, and one PORV block valve in the overpressure protection system. The safety valves are Dresser 31719A valves.
The plant safety valves are mounted,on a short (1.17 ft) inlet pipe designed to prevent a water seal from forming.
The valves have staggered set pressures of 2500 and 2575 psia. The PORV is a l
Dresser 31533-VX-30 solenoid actuated pilot operated valve with a bore diameter of 7/8 in. The Yankee Rowe PORV has a long inlet pipe but designer l
to prevent a water seal from forming.
The block vtive is a 2 in. Pacific l
wedge gate valve equipped with a Limitorque SMA 00 10 operator.
The Dresser 31719A valve was not one of the valves tested by EPR!. The 31719A valve is smaller than either of the two Dresser valves tested by EPRI, the 31739A and 31709NA.
It 1s closest in size to the 31739A valve, l
the smaller of the two valves tested.
The 31719A valve differs from the test valve in the size of the inlet and outlet flanges and in the orifice size.
These differences do not affect valve operability. These considerations, and the fact all 0.' esser valves are similar in configuration and design philosophy, indicate the test valve is representative of the Yankee Rowe valves.
l The Dresser PORV installed at Yankee Rowe is of the dash 2 (31533-VX 30-2) design with a 7/8 in, bore diameter.
The valve tested by EPRI was a dash 2 (31533 VX 30 2) design with a 1 5/16 in, bore size.
The bore size difference only affects capacity not operability. The Yankee Rowe l
PORV with the dash 2 internals was shown to close satisfactorily at low pressure (less than 100 psig). The test valve, therefore, is considered an adeouate representation of the in plant valve.
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The block valve is a 2 in. Pacific wedge di.tc gate valve, with a l
Limitorque SMA-C M 0 operator. This valve was not one of the designs tested j
by EPRI. To verify the adequacy of block valve at Yankee Rowe, the Licensee cited in-situ testing performed on the plant specific block valve.
This j
testing is adequats to demon:trate valve operability.
q lased on the above, the valves tested are censidered to be apolicsble j
to the in plant valves at Yankee Rowe and to have fulfilled that part of the
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criteria of items 1 and 7 as identified in Sescion 1.2 regarding applicability of test valves.
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i 4.2 Test Conditions l
The valve inlet fluid conditions that bound the overpressure transients f
at Yan ee Rowe were provided by YAEC in References 12 and 18. The J
transisnts considered in this report include FSAR, extended high pressure
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inject ion (HP!), and low temperature overpressurization events.
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Fs r th SRVs only steam discharge was calculated for FSAR type transler.*,.
The peak pressure was 2665 psia and the maximum pressurization 1
rate was 3.5 psi /s (Reference 18).
A maximum backeressure of 450 psia is i
j d.veloped the sRv outlet (Referene. 17). YAEC stated in Reference 15 the i
plant valve aQusting rings will be set to 31 (upper) 24 (middle),
l and +3 (lower). These positions are relative to the level position.
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Four steam tests with the Dresser 31739A valve were run with ring
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settings of -48, -40, +11.
These were tests 316, 318, 320 and 322. These
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tests were run on the short inlet piping configuration, which is I
j representative of the plant configuration.
YAEC, in Reference 17 showed j
that with these ring settings the test valve geometry is similar to the plant valves. The lower ring in both the plant and test valves was set to l
provide a valve opening without chatter and full lift on valve opening. The middle ring setting was selectea to place the ring between 0.048 inches
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(plant valve) and 0.060 inches (test valve) below the seat.
The upper ring I
t was set to expose the vent holes in the valve guide one quarter of their i
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I diameter. Of these four tests, two (320 and 322) are directly applicable to Yankee Rowe because the peak pressure, pressurization rate, and backpressure in these tests were greater than 2667 psia, 311 psi /s, and 609 psia, respectively.
These conditions bound those expected at the plant.
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l Review of the Yankee Rowe inlet conditions reports (Re'arences 12 j
and 18) showed that water did not reach the valve during FS,AR transients or l'
an extended high probsure injection (HP!) event.
YAEC presented a discussion in Reference 17 that showed the valve inlet conditions for the feedwater line break, which is the limiting FSAR transient for water 4
discharge in many Westinghouse plants, were bounded by the loss-of-load steam only inlet conditions. The cutoff head for the Yankee Rowe HP! pumps j
is below the SRV setpoint so that an extended HP! event would not challenge l
the safe:y valves.
There was a concern that the extended safety valve blowdown (blowdown greater than 5%) observed during the EPRI tests could result in the I
pressurizer level increasing to the safety valve inlet.
YAEC, in
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Reference 18, reviewed the loss-of-load (LOLD) transient assuming 15%
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blowdown. The LOLD was chosen because it provided the design btsis for sizing pressurizer safety valves. The 15% blowdown is conservative since l
t the largest blowdown observed in the appitcable EPR! tests was 11.1%. This l
review showed the pressurizer level did not reach the inlet to the safety valves.
Thus, the steam inlet condition was maintained.
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The Dresser PORV at Yankee Rowe is mounted on a lor ) inlet pipe without l
loop seals. The peak oressure and pressurization rato for the PORVs during l
FSAR type transients are the same as the safety valves, 2665 psia and 56.5 psi /s, espectively. The maximum backpressure for the PORVs is 426 psia (Reference 20).
j The test valve was subject to fifteen steam tests.
In the steam tests, the peak pressure ranged from 2435 to 2505 psia.
Backpressures ranged from 170 to 760 psia. The testing of the Dresser PORV wu performed at peak pressures below that indicated in References 12 and 18 for Yankee Rowe l
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during an FSAR transient (2435 to 2505 psia versus 2665 psia).
Reference 6 J
I stated that the valve inlet pressure is considered to Have a potential for j
affecting PORV operation only during opening or closing.
Since the Dresser j
valve opens quickly (less than 0.5 seconds), the pressure increase during f
the valve opening cycle is minimal (approximately 28.3 psia increase based i
on the maximum pressurization rate of 56.5 psi /s). Testing at the Yankee l
j Rowe setoressure (2365 psia) or slightly above is, thereftre; considered f
adequate and the test conditions representative of the plant conditions.
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As with the safety valves, References 12 and 18 indicated that water I
did not reach the PORV during FSAR transients or an extended HP! event.
The cutoff head for the Yankee Rowe HP! pumps is below the PORV setpoint so that j
an extended HP! event would not (nallenge the PORVs.
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i The PORVs are used for low temperature overpressure (LTOP) protection at Yankee Rowe.
For low temperature overpressure protection, the valve is J
required to pass steam at pressures up to 539 psia, steam to water I
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transition, and liquid at pressures up to 539 psia with temperatures up to l
475CF.
The peak pressures noted above are based on analyses that assumed j
the pressurizer was liquid full (Reference 17).
The prssence of a steam l
bubble in the pressurizer would limit the peak pressure when the PORY opened l
on steam, but this condition was not specifically analyzed. Thus, the peak l
pressure during steam discharge was bounded using the liquid full analyses, j
Steam discharge conditions are considered to be adequately represented by
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the high pressure tests discussed above.
In addition results from low I
pressure steam tests by Dresser Industries, the valve manufacturer, were provided as part of the Calvert Cliffs submittal (Reference 23).
Steam to water transition is also considered to be adequately represented by the high pressure transition test 21 OR-85/W. Water discharge during a LTCP transient u represented by the low pressure (-690 psta) water tests with fluid tuperatures ranging from 112 to 4590F.
The Pacific block valve at Yankee Rowe was qualified using the results of in situ testing of the plant specific block valve.
In Reference 22 YAEC j
stated that the block valve at Yankee Rowe wat cycled open and closed at j
normal system pressure witn the DORV open.
The block valve, however, is 1
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required to open and close over a range of steam and water conditions.
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required torque to open or close the valve depends almost entirely on the f
differential pressure across the valve disk and is rather insensitive to the trementum loading. Therefore, the required torque is nearly the same for water or steam and nearly independent of the flow. The full pressure steam tests, therefore, are adequate to demonstrate operability of the valve for the required steam / water conditions.
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The test sequences and analyses described above, demonstrating that the l
test conditions bounded the conditions for the plant valves, verify'that j
ltems 2 and 4 of Section 1.2 were u t, in that conditions for the
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operational occurrences were determined and the highest predicted pressures were chosen for the test.
The part of Item 7, which raquires showing that the test conditions are equivalent to conditions prescribed in the FSAR, was I
also met.
I 4.3 Valve Operability 1
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j The Dresser 31719A safety valves at Yankee Rowe are required to operate with steam inlet conditions only.
The EPRI test program tested the representative Dresser valve for the required range of conditions. During F!AR transients the pCRV is required to pass only steam.
The PORV is used for LT0p protection and in this mode may be required to pass steam, steam to water transition, and water.
The test vaive was subjected to the required conditions.
The block valve is also required to operate for steam and licuid flow conditions.
The plant specific valve was tasted in situ at normal system pressure with the PORV open, which is adequate to demonstrate valve operability for the required steam / water conditions.
1 For the applicable safety valve tests (320 and 322), the test valve opened at 2530 and 2580 psia (less than +3.2% of the nominal set pressure),
had stable behavior, and closed with less than 11.1% blowdown.
In test 322, the valve achieved only 78% of rated lift but passed 113% of rated flow at j
3% accumulation.
In test 320 the valve achieved only 44% of rated lift and l
passed only 64% of rated flow at 3% accumulation.
This was attributed to 1
1 the very high backpressure (866 psia) in the test.
The backpresswr; in
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test 322 was 609 psia, which still bounds the plant back*ressure of 11
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l 450 psia, and the valve passed 113% of rated flow.
This indicates the valve was able to perform its safety function of opening, relieving pressure, and l
closing.
1 A maximum bending moment of 8640 ft-lb was applied to the 31739A valve discharge flanga during tests 320 and 322 without impairing valve operation. This bounds the maximum expected bending moment of 1361 ft-lb at i
the plant (see Reference 20).
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 safety valves at Yankse Rowe are mounted on a 2-in inlet pipe, 1.17 ft long. This length is bounded by the EPRI short inlet pipe configuration (length of 6.83 ft, including 1.33 ft of 2.5 in. pipe). The test configuration included a venturi and a reducer.
The plant valves should be as stable as the test valve.
As noted above, the valve blowdown for the 31739A valve during the applicable tests was less than 11.1T,.
A YAEC analysis for a Yankee Rowe LOLD with 15% blowdowa 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 test results discussed above, demonstration of safety l
valve operability is considered adequate.
i The Dresser p0RV opened and closed on demand for all nonloop seal i
tests.
Inspection of the valve after testing at the Marshall Steam Station showed the bellows had several welds partially fail.
The failure did not affect vah e performance and the manufacturer concluded the failure did not have a potential impa$t on valve performance.
The bellows was replaced and did not fail during any of the additional test series.
Results of tests done by Oresser Industries on a PORV similar to the one at Yankee Rowe were provided as part of the Calvert Cliffs, Units 1 and 2, submittal (Reference J3).
This data showed the PORV opened and 12
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d closed without fallnre at pressures ranging from 65 paia to 1979 psia, i
l During other tests, the minimum pressure achieved without leakage was l
90 psia. This test data indicates the valve will operate acceptably with
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l low pressure steam conditions.
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A bending moment of 2125 ft-1b was induced on the discharp flange of l
the test valve without impairing operability.
The maximum bending moment I
j' calculated for the Yankee Rowe PORV is less than 1087 ft-15. The EPRI tests, therefore, bot.ie tne :.aacked plant condition, 1
i Tie Yankee Rowe PORV is a pilot operated valve that usas system l
pressure 29 hyld the disk tight against the seat. At one point Dresser
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Industries recomended the block valve be closed at system pressures below I
l 1000 ps!c to avoid steam wirecutting of the PORV disk and seat.
Testing by Oresser later showed the 1000 psig pressure limit to be overly conservative I
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i and that the PORV as designed was qualified to system pressures 'f o
100 psig.
telow 100 psig the deadweight of the lever on the pilot valve was j
sufficient to kego the pilot valve open.
Dresser recomends, if the plant is to operate at pressures below 100 psig, that heavier springs be used j
under the main and pilot disks to ensure closure. Without the heavier I
springs recomended by Oresser, the PORV might leak at system pressures below 100 psig. However, Dresser stated that if a PORV has not experienced i
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leakage at low pressure the heavier spring would not be necessary.
In Reference 20. YAEC stated the PORV is used during start-up and shutdown for low temocrature overpressure protection, but the PORV has not had a leakage
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problem. Therefore. VAEC does not intend to install the heavier springs in i
the valve.
This response is adequate to resolve the conceens raised by
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I lased on the valve performance during EPRI ttsts, under t,he full range of expected inlet conditions, and based on plant cperating experience without the heavier springs, the demonstration of relief valve noorability is considered adequate.
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1 NUREG 0737 !!.0.1 requires qualification of associated control j
circuitry as part of the safety / relief valve qualification.
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Reference 20, YAEC stated the PORV is not required to perform a safety l
function to mitigate the effects of any design basis event in a harsh environment.
YAEC also stated, in Reference 17, that except for the PORV soienoid, the control circuitry is outside of the containment and is not l
exposed to a harsh environment.
The PORV solenoid is not qualified for a l
hars' environment, but f tilure of the solenoid cannot result in a spurious l
operation of the PORV nor can it prevent an open valve from closing., Based on the data provided by YAEC, it is concluded that the PORV circuitry meets the qualification requirements of NUREG-0737, Item !!.0.1.
The PORV block valve must be capable of closint iver 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.
In Reference 22, YAEC noted the plant block valve was successfully opened and closed at normal system pressure with the PORV open and with an operator that had a maximum torque output estimated at 50 ft-1b.
YAEC noted this corresponds to the calculated required closing torque.
As a result of the EpRI testing, VAEC reevaluated the required closing torque and modified the motor operato ro increase the torcue available.
Following this modification the closing torque was increased by 15't..
The new closing to*que is thus more than 15'i, higher than the maximum torque the operator could deliver when the valve was successfully tested in-situ early in plant life.
This indicates that the bicek valves at the plant should operate satisfactorily under all operating conditions.
The presentation above, demonstrating that the valves operated satisfactorily, verifies that the portion of item 1 of Section 1.2 that requires conducting tet.ts to qualify the valves and that part of Item 7 requiring that the effect of discharge piping on operability be consicered war. met.
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3 4.4 pioins and Suonort Evaluation I
i In the piping and support evaluation, the piping stresses were analyzed l
for 5.%e requirements of the ANS! B31.1 Code, 1977 Edition.
The pipe f
supports were analyzed for the requirements of the Al$C Code, Seventh or j
Eighth Edition, or manufacturer's allowable loads.
The load combinationr.
and acceptance criteria for the Yankee Rome piping and supports were
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reviewed and approved by the NRC staff as part of the Syste'matic Evaluation Program, Topic !!!-6 (Reference 20).
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j Four transient conditions were analyzed. These conditions are shown in j
j Table 4.4.1, which was taken from Reference 20.
The peak pressure and pressurization rate used in the thermal hyaraulic analyses were 2505 psia and 34.2 psi /s. These were based on the inlet conditions report submitted l
j August 1, 1982.
In the overpressure Protection Report submitted I
May 9, 1985, the peak pressure and pressurization rate were revised to 2665 psia and 56.5 psi /s.
In Reference 20, YAEC stated that, due to the higher peak pressure and greater than rated flows measured in tests 4
\\
performed by YAEC, the flow rate used in the original thermal hydraulic 4
1 analysis should have been 30% larger.
In Reference 22, the Licensee stated that the analysis was rerun with a 30'4 increase in the steam flow. This resulted in an average 30% increase in the fluid transient forces on the
)
piping and supports.
YAEC factored the 30% increase into the comparison of the calculated versus allowable loads and stresses for the piping and supports.
The thermal hydraulic analysis was performed with Stone & Webster's I
programs $TEHAM and WATA!R.
STEHAM was used to analyze the high pressure transients. $TTHAM calculates the transient fluid dynamic forcin) functions
)
acting on the pipe segments due to valve discharge. The code computes j
thermal hydraulic variables using the method of characteristics.
Forces on I
piping segments are computed by integrating the rate of change of the fluid momentum within a control volume.
For open pipe segments, discharge
]
blowdenn forces are included.
Time steps are selected internally based on I
input segment lengths and the instantaneous sound speed. Using this b
i 15 l
i I
TAh! 4.4.1 I
fl.ula creeltenes aae celtic 4L panAmstser usso ros tus amatises GutJ ttfrJ fit.r.J tiu 3 t
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.,e si ite.ee. ir
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's.1 r
fe.e (1)
(Il (l)
(1) tiew sete/ vel,e ll.6M lb/hr etes.
130.004 lb/br elew 130.040 lb/hr stee.
132.000 lb/hr weteg 44 3608 Pole et lett pets et litt pe6s et 6il PSIC (I)
(1)
(4)
.e 3.
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+
velve set 160s peig 36sg peng 34al p.la 645 pua eteesere
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(3)
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(3) rest Pressere 2694 pess 14'.4 pegg 1696 pets g3).3 peig III (33
(*)
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% 3 res/see 36.3 pet /see 33,3 pe,fece 84Le (3) l (33 (1)
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,3,,,,,,
(33 (1)
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siee.
sie..
j essenerte L 43 4.987 8.981 I
Cwitenent vel,e less l
Coefteesent 34 l
eeleel Velve bore Dieaster 6 7/8 te.
- 8. lent Valve sees se 0.914 sene: leambete sa ( ) se es(erence peerless to IIAC by TAIC, 1
l J.
TW Fl= eene
- d br aes eles e Sleekerte Ceef fielest of 0.e ter keter Fl.w (est l.r.
. Nile eeltvlesen bt the Prettee estus the field Conj 6el.mo iny,46 We6er e coecettes results with (Pel feste. tie, este Celeulated should be esm et,ettve.
tw ee E fest e
ll.f4 34 treasure e 4M.rala fescorenwee o eW*r Bete 84eestes e & l/l6" Flew este e 363.e00 lbett.
tentee lowe LT0f Festewre e Soc.P$la toneeteterse)%'t beve 84. aster e 1/4" Ft.. este e tiew gene s (Dete Sleesterf ya to.
..a 132.000 the/hr w
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"?/s e 0.34 e 261.600 e 91. WO the./hr ti e l)! A lbe /he c m teva?)vs l
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l e
1 1
I approach, time stepa on the order of 1 ms were used in the STEHAM analyses, The code was verified by Stone & Webster using manual calculation j
techniques. WATA!R was used to analyze tne 70P transient. WATAIR
]
calculates the one-dimensional transient flow field response and flow induced fore 19g functions in a piping system.
The code uses a Runge Kutta integration method to integrate the governing two phase fluid flow l
j equations.
Time steps of 1 ms were used in the WATAIR analysis. Adequate information on Stone & Webster's verification of STEHAM and'WATA!R was provided by YAEC in Reference 20.
I i
j
$TEHAM and WATA!R models of the Yankee Rowe pr.ssurizer safety and l
l reiief vaive piping w.re develop.d. The criticai input param.t.rs for STEHAM and WATAIR were reviewed and found acceptable.
Valve opening times i
were 0.15 s for the safety valves and 0.1 s for the PORV. These times are representative of those measured during EPRI tests. Node spacing was I
generally smaller than that used in the verification $TEHAM analysis.
Hand calculations were used to verify WATAIR so the nodalization used in the l
plant enalysis could not be checked against the v6cification analysis, but
{
the Yankee Rowe model is consistent with other plant analyses that used f
j WATAIR. Th6 Yankee Rowe submittal (Reference 17) indicated that the safety
)
and relief valve flows used in the thermal hydraulic smalysis were based on j
the ASME rated flow values (90% of the theoretical flow rate).
Originally,
)
no adjustments to the rated flows were made to account for the fact that the j
Dresser 31719A safety valve used at Yankee Rowe passed in exsess of 128% of
{
the rated flow in tests by YAEO (Reference 20). As noted above, however, j
YAEC accounted for the increased f1tw and higher peak pressure by increasing j
tne eiiculas.d fiuid tranuent ioads by 30s (R.forence 20). YAEC verified 1
that this load increase was sufficient to account for the difference between I
the analysis and test flow rates by redoing the thermal hydraulic analysis j
with a 30% increase in steam flow (Reference 22).
The structural analysis was performed using $ tone & Webster Engineering 1
Corporation's (5WRC) version of NVp!pt.
This is a linear elastic piping structural analysis program widely used in industry which was fully verified for pipe stress analysis.
The NUp!pE code was benchmarked by the NRC in l
197g as eart of a five piant review conducted by SwEC.
i 17 l
i
The cutoff frequency used in the piping analysis was 400 Hz (cutoff modes 144) with a time step of 0.001 seconds.
Valves and actuators are modeled using rigid elements with lumped weights at the. actuator center of gravity. Typically, the structural model contained at least three mass points between restraints active in the same direction.
A damping factor of 1% was used.
Seismic response spectra were based on the more severe NRC spectra developed for the Systematic Evaluation Program. Values for support and nozzle stiffnsss (provided in Reference 20) were based on generic values obtained from YAEC's Engineering Mechanics Division (Reference 24). The generic values were based on engineering judgement, experience, and the evaluation of common supports. The structural analysic is considered adequate.
In References 17 and 20, the Licensee provided data to show the all pipe stresses were less than code allowables.
The pipe stresses still met code allowables when the fluid transient loads were increased by 30%.
l The pipe supports were analyzed using hand calculations and the i
computer codes STAAD-!!!, GTSTRUDL, and BASEPLATE II.
Adequate verification of GTSTRUDL and BASEPLATE I: were provided in Arkansas Power & il dit's submit *.;l for Arkansas Nuclear One, Units 1 and 2 (Reference 25).
Information on the verification of STAAO-!!! was not provided even though two supports were ana1> zed using STAAD-I!!.
In Reference 22, however, YAEC stated that the two supports nere reanalyzed using GTSTRUOL.
Safety factors for anchor bolts were based on IE Bulletin 79 02.
Peferences 20 ano 22 included a summary of the ratio of calculated support load to the allowable load.
This summary showed the ratio for the supports and the anchor bolts was less than one in all cases (the maximum ratio was 0.99) even when the fluid transient loads were increased by 30%,
as discussed above.
The analysis discussed above, demonstrated that a bounding case was chosen for the piping configuration, indicates Item 3 of Section 1.2 was met.
Review of the piping and supports analyses indicates Item 8 was met.
4 18
5.
EVALUATION
SUMMARY
The Licensee for Yankee Rowe provided an acceptable response to the requirements of NUREG-0737, reconfirming that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met with.2 gard to the safety valves and PORVs.
The rationale for this conclusion is given below.
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 valves tested functioned correctly and safely for all steam and water discharge events specified in the test program that were applicable to Yankee Rowe 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 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, and supports) were 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 associated components demonstrated that this equipment was constructed in accordance with high quality standards, meeting General Design Criterion No. 30.
19
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 Stfety 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 J. A. Kay, YAEC, to D. M. Crutchfield, NRC, "TM! Item II.D.1, Safety and Relief Valves," March 30, 1982.
11.
Letter J. A. Kay, YAEC, to D. M. Crutchfield, NRC, "TMI It6m II.D.1, Safety and Relief Valves," July 1, 1982.
12.
Letter J. A. Kay, YAEC, to D. M. Crutchfield, NRC, "TMI Item 11.0.1, Safety and Relief Valves," August 1, 1982.
13.
Letter J. A. Kay YAEC, to D. M. Crutchfield, NRC, "TMI Item !!.0.1, Safety and Relief Valves," December 28, 1982.
14.
Letter J. A. Kay, YAEC, to 0. M. Crutchfield, NRC, "TMI Item !!.0.1, Safety and Relief Valves," April 1, 1983.
15.
Letter J. A. Kay, YAEC, to D. M. Crutchfield, NRC, "TM! Item 11.0.1, Safety and Relief Valves," April 2, 1984 16.
Letter J. A. Zwolinski, NRC, to G. Papanic YAEC, "Request for Additional Information on NUREG-0737, Item !!.0.1, Performance Testing of Relief and Safety Valves," July 16, 1985.
20
17.
Letter G. Papanic, ilr., YAEC, to J. A. Zwolinski, NRC, "TMI Item II.D.1, Safety and Relief Valves," November 22, 1985.
18.
Letter J. A. Kay, YAEC, to J. A. Zwolinski, NRC, "Supplemental Information for Proposed Change 184-Safety Valve Setpoint Tolerance,"
May 9, 1985.
4 19.
NRC ltr. to YAEC, July 23, 1987.
Second request for information.
20.
Letter G. Papanic, Jr., YAEC, to USNRC Document Control Desk, "TMI Item II.D.1 Safety and Relief Valves," October 27, 1987.
j 21.
Telecopy M. B. Fairtile, NRC, to G. Papanic YAEC, Additional questions on Yankee Atomic Electric Company's NUREG-0737, Item II.D.1, submittal for Yankee Rowe, November 23, 1967.
22.
Letter G. Papanic, Jr., YAEC, to USNRC Document Control Desk, "TMI Item II.D.1 Safety and Relief Valves," January 8, 1988.
23.
Letter R. S. Huffman, Dresser, to R. J. Quinn, Baltimore Gas and Electric, "Baltimore Gas and Electric, Calvert Cliffs Units 1 and 2, Power Operated Relief Valves, CE PO 9903304 and 9903305,"
August 12, 1985.
24.
Engineering Mechanics Division, "Stiffness Representation of Supports, Anchors, and Restraints for Pipe Stress Analysis and Support Design,"
EMD 80-2.
25.
Letter J. T. Enos, AP&L, to NRC Document Control Oesk, "Arkansas Nuclear One - Unit 2, NUREG-0737, Item II.D.1, Performance Testing of Safety and Relief Valves," 2CAN068702, June 5, 1987.
21
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