ML20195G556

TMI Action--NUREG-0737 (II.D.1) Relief & Safety Valve Testing,Maine Yankee Docket, Technical Evaluation Rept

ML20195G556
Person / Time
Site:	Maine Yankee
Issue date:	09/30/1987
From:	Fineman C, Nalezny C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20195G560	List: ML20195G556
References
CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-7861, NUDOCS 8710150408
Download: ML20195G556 (25)

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EGG-NTA-7861 TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1)-

RELIEF AND SAFETY VALVE TESTING MAINE YANKEE DOCKET N0. 50-309 C. P. Fineman C.L. Nalezny April 1988 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6492

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I 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 l

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CONTENTS

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J ABSTRACT ............................................................. 11

1. INTRODUCTION .................................................... I 1.1 Background ................................................. 1 1.2 General Design Criteria and NUREG Requirements ............. 'l

2. PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM ............... 4

3. PLANT SPECIFIC SUBMITTAL ........................................ 6

4. REVIEW AND EVALUATION ........................................... 7 4.1 Valves Tested .............................................. 7 4.2 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

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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 l

force which prepared NUREG-0578 (Reference 1) and, subsequently, NUREG-0737 l

(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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\ 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).

l EPRf contracted with Combustion Engineering (CE) to produce a report on the inlet fluid conditions for pressurizer safety and relief valves in CE l

! 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 5

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.t' December 30, 1982 MYAPCo sumarized their responses to the NUREG-0737 q

requirements (Reference 13). Additional safety valve information and an y I

evaluation of the safety valve and PORV piping was submitted April 4, 1983 h l;

(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 t

4.1 Valves Tested L

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 8

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. 1 valves at. Maine Yankee, not installing the heavier springs is considered l 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, and t

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-00-15 1

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 l 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 11

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. l 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, 12

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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'i rely on the differential pressure across the valve disk and is 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

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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 p occurrencesLwere determined and the highest predicted pressures were chosen 9

.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.

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 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 l

l 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 16

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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 we ~ described.in Section 4.1, the EPRI tests with the 3 in. Anchor Darling valve and Rotork 30-NA1 operator and the full flow in situ test on the in plant h Ahchor ' Darling-valves are. adequate to demonstrate the operability of the in 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 and q,

measurements verified that the valve closed leak tight (Reference 16),

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.NUREG-0737 'II.D.1 requires qualification of. associated control circuitry

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. 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 b L

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 18

j 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 l 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  :

, maximum measured flow rates for the Dresser PORV under similar inlet f 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 20 l

t i

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.o 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, 21

a. O i,  !

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 events specified in the test pro s  :. that were applicable to Maine Yankee and that I the pressure bour.dary component design criteria were not exceeded. Analysis I and review of both the test resuits and the Licensee justifications indicated l 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 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 Rel_ief Valve Test Program Valve Selection / Justification Report, EPRI NP-2292, December 1982.

6. EPRI PWR Safety and Relief Valve 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-2628-5R, 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 Evaluation of Relief Valve Operation, March 30, 1982,

11. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safety and Relief Valve Operation, June 30, 1982.

12. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safety and Relief Valve Operation, August 5, 1982.

13. Letter J. H. Garrity, MYAPCo, to R. A. Clark, NRC, Evaluation of Safety and Relief Valve Operation, December 30, 1982.

14. Letter J. H. Garrity, MYAPCo. to R. A. Clark, NRC, Evaluation of Safety and Relief Valve Operation, April 4,1983.

l 15. Letter J. R. Miller, NRC, to J B. Randazza, MYAPCo, "Request for Additional Infonnation on NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves," February 11, 1985.

16. Letter G. D. Whittier, MYAPCo, to J. R. Miller, NRC, "Response to Request for Additional Information on Relief and Safety Valve Testing (NUREG-0737, Item II.D.1)," May 31 1985.

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