ML20042D070

TMI Action-NUREG-0737 (II.D.1),Braidwood Units 1 & 2, Technical Evaluation Rept

ML20042D070
Person / Time
Site:	Braidwood
Issue date:	01/31/1988
From:	Geoffrey Miller, Nalezny C, Yuan C EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20042D072	List: ML20042D070
References
CON-FIN-D-6005, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-8027, NUDOCS 8808170184
Download: ML20042D070 (27)

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k TECHNICAL EVALUATION: REPORT TMI ACTION--NUREG-0737 (II.D.1)-

BRAIDWOOD UNITS 1 and 2 M

DOCKET NO. 50-456~, 50-457-r

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G. K. Miller C. Y. Yuan C. L. Nalezny

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C. P. Fineman l*

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Idaho National Engineering Laboratory-EG&G Idaho, Inc.'

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Prepared for the.

U.S. Nuclear Regulatory Commission

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Washington, D.C.

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Under DOE Contract No~. DE-AC07-761001570 FIN No. D6005 4

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ABSTRACT i

3 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

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' Learned Task Force Status Report and Short-Term Recomendations). and J

subsequently NUREG-0737 (Clarification of TMI Action Plan. Requirements) recommended that programs be' developed and completed which would reevaluate; j

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the functional. performance capabilities of Pressurized Water Reactor-(PWR)-

1 safety, relief, and block valves and which would verify the integri'ty of the piping systems for normal, transient, and accident (conditions. This-report documents the review of these programs by the Nuclear Regulatoryz Comission.

(NRC) and their consultant, EG&G Idaho, Inc.

Specifically, this review j

examined-the responseJof the Licensee for the Braidwood. Nuclear-Power

-Station, Units 11 and 2, to the requirements of NUREG-0578 and NUREG-0737 and~

finds that _ the' Licensee provided an acceptable response, reconfirming = that-

'I the General Design Criteria 14, 15, and 30 of Appendix.'A to-10 CFR 50 were met.

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FIN No. 06005--Evaluation of CW Licensing Actions-NUREG-0737, 11.D.1' 0

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. ABSTRACT.............................................................. ii g

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INTRODUCTION.....................................................

1 1.1 Background.................................................

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47 Genera'l Design Criteria and NUREG Requirements.............

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'i 1.2.

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. PWR OWNDtS' GROUP RELIEF AND SAFETY VALVE PROGRAM................

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3.

- PLANT SPECIFIC SUBMITTAL.....................................

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'4.

. REVI EW AND EV ALU AT I ON.. '.......................................... : 7 I

4.1 Valves Tested..............................................

7 br 4.2 Test Conditions............................................

7 4.2.1 FSAR Steam Transients..............................

8 4.2.2 FSAR Liquid Transients.............................

9 4.2.3-Extended High-Pressure Injection Event............

11 L.

.4.2.4 Low Temperature Overpressure ~ Transient............

11 L1 4.2.5 PORV Block Valve Fluid Conditions.................

12 4.2.6 Test Conditions Summary

............................-12 4.3 Operability...............................................

13 4.3.1 Safety'Va1ves....................-..................

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4.3.2-Power Operated Relief Va1ves'......................

16 4.3.3 Electric Control Circuitry.....................=...

16 4.3.4 PORV: Block Valves..................................

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4. 3. 5 '

Operability Summary...............................

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4.4 Piping and Support Evaluation.........................'....

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4.4.1 Thermal Hydraulic Analysis........................

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4.4.2'

. Stress Analysis...................................

20 4.4.3 Piping and Support Summary........................

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5.

EVALUATION

SUMMARY

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6., REFERENCES......................................................

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L 4.3.1 EPRI tests on Crosby HP-BP-86 6M6 safety valve................

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>a TECHNICAL EVALUATION rep 0RT T

TM1 ACTION--NUREG-0737 (!!.D.1)

BRAIDWOOD NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NO. 50-456. 50-457 s

1.

INTRODUCTION 1.1 Backaround Light water reactor experience has included a number of instan*ces of improper performance of relief and safety valves installed in the primary

. coolant system.- There were instances cf valves opening below set pressure, valves opening above set pressure, and valves failing to open or ressat.

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 a~ basic unreliability'of.the valve design.

It is known that the failure:of a power-operated relief valve (PORV)-to ressat was a significant l

contributor'to the Three Mile Island (TMI-2) sequence of events. These

-facts. led the task force which prepared NUREG-0578-(Reference 1) and,

--subsequently,-NUREG-0737'(Reference 2) to recommend that programs be developed and executed which would reexamine the functional performance-

-capabilities of: Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions.

These programs were deemed

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necessary to reconfirm that the General Design Criteria 14, 15, and 30 of-Appendix A to Part 50 of the Code of Federal Regulations,- 10 CFR, are indeed satisfied.

1.'2 General Design Criteria and NUREG Reauirements e

General-Design Criteria 14, 15, and 30 require that (a) the reactor primary coolant pressure-boundary be designed, fabricated, and-tested so as to have an: extremely low probability of abnormal leakage, (b) the reactor coolant system and associated auxiliary, control,'and protection systems be designed with sufficient margin to assure that the design conditions are 1

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a lnot exceeded during: norma 1' operation or anticipated transient events, and-(c) the components which are part of the reactor coolant pressure boundary i

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.shall be constructed toLthe highest quality standards practical.

To reconfirm the integrity of_ overpressure protection systems and l

thereby assure that the General Design. Criteria are met, the NUREG-0578 position was issued as a requirement in a letter datid September 13, 1979 by.

-the Division:of Licensing'(DL), Office of Nuclear Reactor Regulation (NRR),

to-ALL OPERATING NUCLEAR POWER PLANTS. This requirement has since',been j

incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements;(Reference 2), which was issued for implementation on October 31,.1980'.

As stated in the NUREG reports, each pressurized water

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reactor Licensee or Applicant shall:

1..

. Conduct testing to qualify reactor coolant system relief and s

safety valves under expected operating conditions for design basis transients and accidents.

2.

. Determine valve expected operating conditions through the use of.

- analyses of accidents and anticipated operational occurrences 4

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 pressures predicted by conventional safety j

analysis procedures.

5.

Include in the relief and safety valve qualification program the--

i qualification of the associated control circuitry.

6.

Provide test data 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 Jubstantiate that the J

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valves tested'in a generic test program demonstrate the l

functionability ef as-installed primary relief and safety valves.

l This correlation must show that the test conditions useds are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (F$AR). The effect.

l of as-built relief and safety valve discharge piping on valve.

i operability must be considered.

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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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PWR OWNERS' GROUP RELIEF AND CAFETY VN.VE 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 power operated relief valves, safety valves, block valves, and associated piping systems. The Commonwealth Edison Co. (Ceco), owner of the Braidwood Nuclear Power Station Units 1 and 2, was one of the utilities sponsoring the EPRI Valve Test Program. The results of the program,are contained in a group of reports which were transmitted to the NRC by Reference 3.

The applicability of these reports is discussed below.

EPRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves under conditions which bound actual plant operating i

conditions.

EPRI, through the valve manufacturers, identified the valves

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used in the overpressure protectic,n systems of the participating utilities.

Representative valves were selected for testing with a sufficient number of the variable characteristics that their testing would adequately demonstrate i

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 4

the plant transients for which overpressure protection would be required

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(Reference 6).

l EPRI contracted with Westinghouse Electric Corp. to produce 16 report on the inlet fluid conditions for pressurizer safety and relief valves in Westinghouse des'gned plants (Reference 7).

Since Braidwood, Units 1 and 2, l

were designed by Westinghouse 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 valves were l

tested at the Combustion Engineering Company,-Kressinger Development L

Laboratory, located in Windsor, Connecticut.

The results for the relief and safety valve tests are reported in Reference 8.

The results for the block valves tests are reported in Reference 9.

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The crimary objective of the EPRI/C-E Valve Test Program was to test each of the various types of primary system safety valves used in PWRs for i

g 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. Additisnal I

objectives were to (a) obtain valve capacity data, (b). assess hydraulic and structural effects of associated piping on valve operability, and (c) chtain-

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piping response data that could ultimately be used for verifying analytical f

piping models.

Transmittal of the test results meets the requirement of Item 6 of Section 1.2 to provice test data to the NRC.

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PLANT SPECIFIC SU8MITTAL.

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A preliminary plant specific evaluation of the adequacy of the i

overpressure protection system for Braidwood, Units 1 and 2, was submitted by Ceco to the s'J on October 26, 1982 (Reference 11). This was followed by l

a submittal of additional informatica regarding piping and support adecuacy'.

l on December 30, 1983 (Reference 12). A request fer additional information was submitted to Ceco by the NRC on March 26, 1987 to which Ceco responded on December 2, 1987 (Refer 3nces 13 and 14).

11 The response of the overpressure pr'otection system to Anticipated l

Transients Without Sceam (ATW3) 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.

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REVIEW AND EVALUATION

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u 4.1 Valves Tested 4

s Braidwood, Units 1 and 2, are four-loop PWRs designed by the Westinghouse Electric Co.

Each un*;t is equipped with three (3) safety

,x valves, two (2) PORVs, and two (2) PORV block valves in its overpressure protection system. The safety valves are 6-in. Crosby Model HB-BP-86, 6M6, e o spring loaded valves with loop seal internals. The design set pressure is 2485 psig and the rated steam flow capacity is 420,000 lbm/h.. The PORVs are 3 in. Coper-Vulcan Model D-100-160 globe valves with 316 SS stellited plugs and'17-4 PH cages.

The PORV opening set pressure is 2335 psig and the rated

' steam flow capacity is 210,000 lbm/h. The PORV block valves are 3-in.

Westinghouse Model.-3000GM88 gate valves with Limitorque SB-00-15 motor L

operators.

The inlet pipe to the safety valve includes a hot loop seal I

(2820F); the inlet to the PORV has a cold loop seal (1700F).

Safety valves, PORVs, and PORV block valves identical to those used at l

Braidwood, Units 1 and 2, were included in the EPRI tests.

Since there is no' difference between the valves tested and the valves installed at the plint,thetestresultsforthesevalvesaredirectlyapplicableto l.

Braidwood, Units 1 and 2.

Therefore, those parts of the criteria of Iter 6s 1 and 7 as identified in Section 1.2 of this report regarding applicability of the test valves are fulfilled.

4.2 Test Cenditions r

I-As stated above, Braidwood, Units 1 and 2, are four-loop PWRs designed by the Westinghouse Electric Corp. The valve inlet fluid conditions that L

, bound the overpressure transients for Westinghouse designed PWR Plants are L

' identified in Reference 7.

The transients considered in this report include FSAR, extended high pressure injection, and cold overpressurization events.

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The expected fluid conditions for each of these events and the applicable EPRI tests are discussed in this section.

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~4.2.1 FSAR Steam Transion,ts L

i For the Braidwood PWRs,-the, limiting FSAR' transients resulting in steam

' discharge through the' safety valves alone and in steam discharge through both the safety and relief valves are the loss of load event (for u.aximum pressurizer pressure) and the locked rotor event (for the maximum pressurization rata).-

In the case when the safety valves actuate alone, the maximum,

pressurizar pressure and maximum pressurization rate are predicted to be 2555 psia and 144 psi /s, respectively. The maximum developed backpressure in the outlet piping is $33 psia (Reference 14).

Hot loop seals are used at the safety valve inlet. The loop seal temperature at the inlet to the valve 1

is 2820F. The safety valves at Braidwood, Units 1 and 2, use manufacturer's recommended ring settings.

t EPRI tests representative of the valve inlet fluid conditions for the limiting transient were selected for the plant specific evaluation.

In selecting the EPRI tests, the safety valve ring settings were considered.

For steam flow conditions, four loop rial discharge tests (Test No. 929, 1406,1415,1419) were applicable to Braidwood, Units 1 and 2.

These tests' '

were performed with valve ring settings representative of the typical ring settings;used in Westinghouse PWRs including Braidwood. The ring settings used in these tests were (-71, -18) or (-77, -18).

These represent the upper and lower ring positions measured from the level position r'eferenced to the bottom of the disc ring.

Sinca both the test ring settings and the in-plant ring settings were determined by the valve manufacturer, the Crosby Valve and Gage Co., using the same methods and the same standard of

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perfgrmance,thesetwosetsofringsettingsareconsideredcomparableto each other.

The loop seal temperature measured in the tests ranged from 90 to 3600F at the valve inlet. The maximum pressurizer (tank 1) rressures were in the range of 2675 to 2760 psia and the pressurization rate was 90 to-360 psi /s.

The backpressures developed in the tests were 245 to 710 psia.

9 The above data show that the inlet fluid conditions and backpressures of these tests envelop the corresponding fluid data predicted for the Braidwood safety valves.

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When both the safety valves and PORVs are actuated, the maximum 1

pressurizer pressure is predicted to be 2532 psia and the maximum I

pressurization. rate is 130 psi /s.

In the EPRI tests on the Copes-Vulcan PORV, the maximum steam pressure at valve opening was 2715 psia, >$1ch bounds the predicted pressure at Braidwood.

A test simulating loop seal discharge was conducted at a pressure of 2725 psia with a water temperat'ure of 1340F at the valve inlet. The backpressure developed at the outlet of i

the PORVs is not an important consideration because the air operated PORVs used at the Braidwood plant are not sensitive to backpressure (Reference 6). Therefore, the EPRI test inlet fluid conditions for*the PORV with steam discharge are representative of the plant specific transient conditions.

4.2.2 FSAR Liouid Transients The limiting FSAR transient resulting in liquid discharge through the

'PORVs and safety valves is the main feedline break accident (Reference 7).

The submittal did not address the transient conditions that involve liquid discharge through safety valves and PORVs.

The Licensee stated that its decision not to evaluate the Braidwood I and 2 safety _ valves and PORVs for liquid discharge was based on a probabilistic risk study presented in I

Appendix A of Reference 11. This study concluded that liquid discharge in the feedline break, extended high pressure injection, or low temperature overpressurization event was very unlikely to occur. Therefore, liquid discharge throsgh the s6fety valves and PORVs was not considered.

However, the Westinghouse Valve Inlet Fluid Condition Repo.*t (Reference 7) stated that the main feedline pipe rupture event was classified as a Class IV licensing event. That is, one which was not expected to take place but was postulated because its consequences include the' potential for the release of a significant amount of radioactive material. Also, NUREG-0737 specifically requires the safety valves and PORVs be qualified for inlet fluid conditions resulting from translents and accidents referenced in Regulatory Guide 1.70, Rev. 2.

The feedwater line break is specifically defined in Regulatory Guide 1.70, Rev. 2.

From a 1

review of the feedwater line break analysis for Braidwood (see below), it is 4

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clear that the f+edwater line break is most likely to be the limiting transient for providing high pressure liquid to the safety valves, a fluid for which they were not originally designed. Therefore, in accordance with the NUREG requirements, the safety valves and PORVs should be qualified for inlet conditions typical of the feedline break event even though the probabilistic analysis showed the frequency of c.ccurrence is extremely low.

i The Braidwood feedline break analysis indicated that the safety valves i

f and PORVs opened on saturated steam at about seven minutes into the c

transient and steam to saturated liquid transition'would follow at thirteen minutes into the event (Appendix A, Reference 11).

It is apparent that liquid discharge through the safety valves and PORVs cannot be ruled out.

1 Therefore, valve operability will be reviewed using the feedline break data provided in Reference 7.

Reference 7 showed that, in a feedline break accident at Braidwood, the

'n.aximum pressure at the safety valve inlet during liquid discharge was calculated to be 2508 psia and the pressurization rate was 3.5 psi /s.

Fluid temperatures at the valve inlet range from 615 to 6350F and the' maximum liquid surge rate into the pressurizer is 569 gpm.

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In a feedline break accident resulting in safety vr.1ve actuation, water discharge is always preceded by cteam and steam to water transition flows.

l Among the EpRI tests performed on the 6M6 valve, Tests 931a and 931b were performed for loop seal / steam, steam to water transition, and water l'

discharge conditions. The valve ring settings and inlet ripe configuration

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used in those tests were comparabia to those of the in-plant safety valves.

L In Test No. 931a, the maximum inlet pressure was 2578 psia. The L

presWrization rate was 2.5 psi /s, the inlet fluid temperature was 1170F i

and"the tank fluid temperature was 6350F.

After the valve closed in Test 931a, the' system was allowed to repressurize and the valve cycled on approximately 640GT water (Test 931b).

Because the inlet temperature and l

pressure of the tests compare favorably with the predicted in-plant I-conditions, the results of these tests are applicable to the Braidwood safety valves.

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,*.s The expected fluid conditions at the safety valve inlet were based on a Westinghouse analysis that assumed the PORVs were not operable during the f

feedline break transient.

If the PORVs are operable, the same fluid conditions postulated for the safety valve inlet can also be expected at the s

PORV inlet (Referetce 6).

In the EPRI tests, high temperature water.

discharge and stoa.n to water transition tests were performed with the Copes Vulcan PORV.

In the water discharge test, Test No. 76-CV-316-2W. the maximum pressure at the ve.1ve inlet was 2535 psia and the te'mperature was 6470F.

In the transition test, Test No. 77-CV-316-75/W, the maximum inlet pressure was 2532 psia and the water temperature nas 6570F. The inlet fluid conditions for these tests compare well with the predicted maximum pressure and temperature of 2508 psia and 6350F for the Braidwood plant.

Therefore, these tests are adequate to represent the in-plant PORV performance in the feedline break event.

4.2.3 Extended High Pressure Injection Event The limiting extended high pressure injection event is the spurious actuation of the safety injection system at power (Reference ?).

For a four-loop plant, both the safety valves and PORVs will :

challenged.

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steam and water discharge are expected.

In this event, however, the safety valves or PORVs open on steam and liquid discharge would not be observed until the~ pressurizer becomes water solid. According to Reference 7, this would not occur until at least 20 minutes into the event which allows ample time for operator action.

Thus the potential for liquid discharge in extended HPI events can be disregarded.

4.2.4 Low Temperature Overpressurization Transient t

The P0. ' is used for low temperature overpressure protection (LTOP) durfng the 10.i temperature stages of reactor start-up and shutdown operations. The expected valve inlet conditions in the low temperature overpressure protection mode were given in Reference 14 as pressures ranging from 350 to 2450 psig and water temperatures ranging from 70 to 4500F, It is also possible for the PORV to actuate under high and low pressure steam conditions.

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For steam discharge through the PORV, the high pressure steam tests discussed above cover the low pressure steam conditions predicted for LTOP.

For water discharge conditions, there were two low pressure and low temperature water tests performed ein the Copes Vulcan PORV with stellited plug and 17-4 PH cage. The tests were conducted at an inlet pressure of 675 psia and water temperatures of 105 and 4420F, respectively.

In addition there was a high pressure water test with a pressure of 2535 psia and a water temperature of 6470F. These conditions are representative of those at Braidwood. Therefore, the EPRI tests can be used to evaluate the performanct of the Braidwood PORV during LTOP tradients.

4.2.5 PORV Block Valve Fluid Conditions The block valves at Braidwood are Westinghouse 3 in, gate valves, Model 3000GM88, with Limitorque SB-00-15 operators.

The block valves are required to operate ovt'r a range of fluid conditions (steam, steam-to-water, water) similar to those of the PORVs.

The 3-in. Westinghouse 3GM88 valve i

with. Limitorque SB-00-15 operator was subjected to full pressure steam conditions (to 2485 psia) in the EPRI-Marshall tests.

Later tests were performed by Westinghouse using subcooled water as a test fluid.

Tests were conducted at differential pressures across the valve discharge rangmg from 200 to 2600 psi and with flow rates ranging from 60 to 600 gpm. The i

Westinghouse tests showed that the torque required to operate the valve is

-almost entirely dependent on the differential pressure and is rather insensitive to momentum loadi N.

Thus operability of the valve 1,s nearly the su e for steam and liquid discharge, and the results of the EPRI-Marshall tests can be used to assess operabilny of the valve.

I 4.2,6 Test Conditions Summry The test sequences and analyses described above demonstrate that the test conditions bound the conditions for the plant valves.

They verify that Items 2 and 4 of Section 1.2 were met, in that conditions for the operational occurrences were determined and the highest predicted pressures were chosen for the test. The part of Itam 7, which requires showing that the. test conditions are equivalent to conditions prescribed in the FSAR, was also met.

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4.3 Ooerability 4.3.1 Safety Valves The EPRI tests reprasentative of the steam discharge condition for the Braidwood safety valves are the loop seal tests on the Crosby 6M6 valve, i

Test No.' 929, 1406, 1415, 1419.

In all these tests (except Test No. 1415),

the valve fluttered or chattered during loop seal discharge'and stabilized when steam flow started. The valve opened within 12% of the design set pressure and closed with 5.1 to 9.4% blowdown. Up to 111% of rated flow was achieved at 3% accumulation with valve lift positions at 92 to 94% of rated lift. These testa demonstrated that the valve performed its function in spite of the initial chatter during loop seal cischarge.

In Test 1419, the valve chattered on closing and the test was terminated after the valve was manually opened to stop the cha*.ter. This result does not indicate a valve closing problem for the Braidwood safety j

valve since an identical test (Test 1415) had already demonstrated that the r

i valve performed satisfactorily and exhibited no sign of instability. The closing chatter in Test 1419 may possibly be a result of the repeated actuation of the valve in loop seal and water discharge tests. As shown in Table 4.3.1, the 6M6 test valve was subjected to seventeen steam, water, and-transition testa.

In the first four or five tests, the valve fluttered and chattered during loop seal discharge but stabilized and closed successfully. After Test 913, there were four instances in which the test was terminated due to chattering on closing.

Galled guiding surfaces and damaged internal parts were found during each inspection and the damaged parts were refurbished or replaced before the next test started. The test results showed that the valve performed acceptably in the test following each repair, but that closing chatter recurred in a subsequent test.

' Test 1415 was performed imediately after valve maintenance and the valve performed stably. The next test (Test 1419) encountered chatter.in closing though i+. was a repeat of Test 1415. These results suggest that irtspection and maintenance are important to the continued operability of the valves.

The Licensee should develop a formal procedure requiring that the safety 13 1

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e TA8tt 4.3.1 ~ ' EpRI TESTS ON rWO58Y HC-Sp-86 liN6 SMETY VRIVE a.

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Test No.

Settine Test Type Stellity (gyse)

(ese) 1 903 1

Steam Stable 9

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908 1

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910 1

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4 Inspection / Repair 5

913 2

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914a,b,c 2'

L.S. Transtilen Teminated 9

tarys inspectlen/ Repair i

7-917 3

L.S.

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8 8

920 3

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Terminotad S

9 Inspectlen/ Repair 9

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ffater Terminated S

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Inspection / Repair 14 1405 4

L.S.

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Inspectlen/ Repair 15 1411 4

Stese Stele 0.76 9.37 Inspectlen/Wepair i

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Terminated 9

1.5 Inspectlen/ Repair c--chatter t

f/c--f1stter/chetter L.S.--leep seal-Ring setting--feer different sets of ring sett, legs esere tested. Actuel ring posi'. lens net shown.

Terminated--Test tominated after valve esos menen11y opened to step chettet.

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valves be inspected after each actuation and the procedure should be incorporated into the plant operating procedures or licensing documents such as the plant technical specifications.

4 Blowdown in these tests (5.1 to 9.4%) was in excess of the 5% value specified by the valve manufacturer and the ASME Code. Westinghouse performed an analysis, " Safety Valve Contingency Analysis in Support offthe EPRI Safety / Relief Valve Testing Program--Volume 3: Westinghouse Systems,"

EPRI NP-2047-LD, October 1981, on the effects of increased blowdown and concluded that no adverse effects on plant safer.y occurred in that the reactor core remained covered.

Therefore, the increased blowdown that l

occurred in the Crosby 6M6 steam tests is considered acceptable.

1 As discussed in Section 4.2.2, the limiting FSAR transient resulting in liquid discharge is the main feedline break accident.

Tests 931a and 931b with typical plant ring settings of (~71, -18) simulate the expected Braidwood feedwater line break conditions. Test 931a was a loop seal / steam / water transition test. The 6M6 valve initially opened, fluttered or chattered in a partial lift position during loop seal discherge, then popped open, stabilized on steam, and closed with a 12.7% blowdown.-

i

?

Test 931b was a saturated water test. The 6M6 valve opened on 6400F water, chattered, and then stabilized. The valve closed with 4.8%

blowdown.

For these tests the valve opened within -1% and +3% of the set pressure.

The maximum calculated surge rate at Braidwood, Units 1 and 2 during the feedline breet transient is 569 gpm. The 6M6 valve tested by EPRI passed 2355 gpm at 2415 psia and 6410F which is much hightr'than the predicted flow rate for Braidwood. The above results demonstrate that the Braidwood safety valves would be adequate to parform the required wattr relief ' unction.

Bending moments as high as 298,750 in-lb (Test 908) were induced on ths discharge flange of the Crosby 6M6 test valve, which had no advarse effect on valve performance.

Because this applied moment exceeds the maximum estimated bending moment of 177,460 in-lb for the Braidwood valves (Reference 14).the performance of the plant valves is also expected to be unaffected by bending moments imposed during discharge transients.

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j For the tests to be an adequate demonstration of safety valve stability,:the test inlet piping pressure drop should exceed the plant pressure drop.

The test inlet pressure drop for the Crosby 6M6 valve on the t

' loop seal configuration was'263 psid on opening and 181 psid on closing.

The values calculated for the Braidwood, Units 1 and 2, safety valves were i

235 and 120 psid for opening and closing, respectively (see Reference 14).

Therefore, the plant valves should be as stable as the test valves.

4.3.2 Power Operated Relief Valves l

The EPRI tests on the Copes-Vulcan PORV with 316 SS stellited plug and 17-4 PH cage demonstrated that the valve opened and closed on demand in loop seal /steem, steam, water, and steam to water transition conditions. The opening and closing time were within the 2.0 second opening and closing time normally required for Westinghouse PWRs. The lowest steam flow rate observed in the tests was 232,000 lb/h which exceeded the rated flow of 210,000 lb/h for'the Braidwood PORVs.

The predicted value of the maximum bending moment induced at the Braidwood PORV discharge flange was calculated to be 65,860 in-1b (Reference 14).

This exceeds the maximum tested bending moment of.

43,000 in-lb. Therefore, operability of the Braidwood PORVs with the t

maximum expected bending moment cannot be shown directly using the EPRI data.

However, Ceco stated in Reference 14 that Westinghouse qualified the Braidwood PORVs for these loads by analysis.

The maximum loading of' 65,860 in-lb results in stresses of 71% of the allowable value. Therefore.

the Copes-Vulcan PORVs at Braidwood are expected to operate with the higher bending moments. This is reasonable because the EPRI test bending moments represent the maximum value tested, not the maximum permissible bending

-moment.

[

4.3.3 Electric Control Circuitry 4

NUREG-0737 Item II.D.1 states that the control circuitry associated with the PORVs shall also be qualified for design basis accidents and transients. The Nuclear Regulatory Commission staff agreed that meeting the 16 2

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licensing requirements of 10 CFR 50.49 for this electrical equipment is satisfactory and that specific testing per the NUREG-0737 requirements is not necessary.

CECO included the PORV controls that would be located in a j

l.

harsh environment in the Braidwood 10 CFR 50.49 environmental qualification I

program (Reference 14) thereby ensuring that the control. circuitry will D.

function properly.

Ceco stated the NRC reviewed and approved the Braidweed j

g environmental qualification program (NUREG-1002, Supplement 2).

t 4.3.4 PORV Block Valves j

?

The PORV block valve must be capable of closing over a rante of steam and water conditions. As described in Section 4.2, results from tho high

]

pressure steam tests,can be used to evaluate valve operation over the full j

range of inlet conditions.

The tests on the Westinghouse 3GM88 valve with Limitorque SB-00-15 actuator showed that the valve successfully opened and closed on courand once the torque switch was set to its maximum.. The plant block valves were also modified by adjusting the torque switches to optimal values for opening and closing thrust, adjusting the pinion gear ratios, and f

changing the wiring from torque controlled to limit controlled stroking.

With these changes, the plant valves are expected to operate acceptably.

4.3.5 Operability Summary The above discussion, demonstrating that the valves operated satisfactorily, verifies that the part of item 1 of Section 1.2, which requires conducting tests to qualify the valves, and that part of Item 7, which requires the effect of discharge piping on operability be considered, t

were met provided the Licensee documents a formal precedure for the inspection of the safety valves as discussed in Section a.3.1.

Also, the licensing action for 10 CFR 50.49 satisfies Item 5 of Section 1.2.

P 4.4 Pioing and Support Evaluation This evaluation covers the piping and supports upstream and downstream of the sa.'ety valves and PORVs extending from the pressurizer nozzle to the pressurizer relief tank.

The piping was designed for deadweight, interr,a1 pressure, thermal expansion, earthquake, and safety and re*ief valve 17

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4, discharge conditions. The calculation of the time histories of hydraulic forces due to valve discharge, the method of structural ana', sis, and the load combinations and stress evaluation are discusser' below.

4.4.1 Thermal Hydraulic Analysis Pressurizer fluid conditions were selected for use in the thermal hydraulic analysis such that the calculated pipe discharge forces would bound the. forces for any of the FSAR, HP1, and low tmperature overpressuaization events including the single failure that would maximize I

the forces on the valve.

The safety valve and PORV discharge transients were analyzed as two separate events. This approach is acceptable, since the safety valves and PORVs have different set points and will not lift simultaneously. !.lso the.

sequential discharge of the PORVs and safety valves (i.e., PORV discharge followed by safety valve discharge) would not generate higher piping loads tnan the separate PORV and safety valve discharge events. Therefore, thr:

valve discharge conditions considered for the piping and support analyses

+

were the steam discharge condition resulting from the simultaneous actuation of all safety valves and the steam discharge condition resulting from the simultaneous actuation of all PORVs.

Because water seals are maintained upstream of the safety valves and PORVs, steam discharge conditions would generate the highest loads on the piping system when the water slug is expelled from the, loop seal and forced down the discharge piping.

Therefore, the selection of the steam dise5.arge cases as the limiting conditions for the evaluation of the piping loads is considered adequate.

f For this analysis, steam at a pressure of 2575 psia and enthalpy of 1130 Btu /lb was assumed to be discharged through the safety vrives.

A loop seal having an enthalpy. distribution based upon a temperature profile consistent with EPRI hot loop seal Test 917 was assumed to be present upstream of the safety valves. That is, a temperature of approximately i

3000F at the valve.inist and saturation temperature at the steam-water interface.

The safety valves were assumed to open linearly in 0.040 s.

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e, The initial conditions for steam discharge of the PORVs were: 2350 psia and enthalpy of 1162.4 Btu /lb.

Cold water seals (1700F) were assu'med upstream of the PORVs. The PORVs were assumee to open linearly in 1.0 s.

i The valve opening times used for the safety valves and P0kVs were greater than those measured in the EPRI tests. For the Crosby 6M6 valves used at Braidwood, measured pop times were on the order of 0.020 s.

For the Copes-Vulcan PORVs, the measured main disk opening time was'en the order of 0.40 s.

However, with Braidwood using loop seals upstream of both the safety valves anti PORVs, the opening times used in the analysis arer considered adequate because the peak piping loads are des to the passage of

)

the loop seals and valve characteristics such as opening time are less important in the calculation of these loads. This conclusion is supported by the analysi: of EPRI Test 917 presented in Reference 15. When a valve opening time of 0.090 s was used, compared to a measured pop time of 0.015 s, the measured forces were still adequately calculated.

I l

The thermal hydraulic analysis was performed using the Westinghouse h

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computer code, ITCHVALVE.

ITCHVALVE calculates the fluid parameters as a function of time.

The unbalanced force or wave-force in each piping segment is chiculated from the fluid properties obtained from the ITCHVALVE analysis using.another Westinghouse program, FORFUN.

The forcing functions on the l

piping system resulting from the fluid transients were obtained from these calculations.

L The adequacy of the ITCHVALVE/FORFUN programs for thermal hydraulic analyses was verified by comparing the analytical and test results for thermal hydraulic loadings in safety valve discharge piping for two EPRI tests (Test Nos. 938 and 917). The detailed comparisons of the ITCHVALVE predicted force time histories and the EPRI test results were presented in

]

Refefrence 14 and these comparisons are considered satisfactory.

The thermal hydraulic and stress analyses of the Braidwood safety valve and PORV piping and supports were performed by the Westinghouse Electric Co.

as a consultant to the Licensee. The typical Westinghouse analysis for such piping systems was fully reviewed in previous submittals for similar PWR plants such as Diablo Canyon Units 1 and 2 (Reference 16). The method of 19

, f ", *o s.

analysis used by Westinghouse including the analysis assumptions,'the structural modeling as well as the key parameters used in computer inputs such as the node spacing, calculation time interval, valve opening time, etc. was examined and found to be acceptable.

Because the Braidwood piping analysis followed the same method and procedure used in previous Westinghouse analyses, the analysis method is considered acceptable. The safety valve and PORV flow rates used in the analysis were greater than~120%

of the rated flow. The conservative factor contained in these flow rates is more than sufficient to account for the 10% derating of the safety, valve required by the ASME Code and includes allowance for uncertainties or errors.

4.4.2 Stress Analysis, The strectural responses of the piping system due to safety valve /PORV discharge transients were calculated using the modal superposition method.

The fluid force time histories generated from the FORFUN program were used as forcing functions on the structure.

The Westinghouse series of structural analysis programs, WESTDYN, FIXFM3, and WESTDYN2, were used to calculate the piping natural frequencies, mode shapes, nodal displacements, and the internal forces and support reactions.

The FIXFM3 code calculated I

the displacements at the structural node points using the forcing functions generated by FORFUN and the modal data from WESTDYN. The structural

-l displacements were then used by WESTDYN2 to compute the piping internal loads and support loads.

The WESTDYN' series of structural programs mentioned above was previously reviewed and approved by the NRC (Reference 17).

The adequacy of these programs for piping discharge analysis was further verified by comparing the solutiors generated by these programs with the EPRI safety valve test results (Reference 18).

l The important parameters in the structural analysis were reviewed and found acceptable.

The time step size was 0.0005 s.

Damping of 1% was used for the OBE and 2% for the SSE.

t. umped mass spacing was determined to ensure all appropriate mode shapes were accurately represented.

For the thermal hydraulic analysis, the cutoff frequency was greater than 333 Hz.

20

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'l-The piping upstream of the safety valves and PORVs was analyzed for the requirements of the ASME Code, Section !!!, 1977 Edition, Addenda through Summer 1979. Tnt downstream piping was analy:ed for the requirements of the l

ANSI B31.1 Code,1973 Edition, Sumer 1973 Addenda. The load combinations and stress limits used to evaluate the upstream and downstream piping are identical to those recomended by EPRI (Reference 19). The Licensee provided a comparison of the highest stresses in the piping against the applicable stress limits for the load combinations defined above.

All stresses were within allowable stress limits (Reference 14).

The upstream pipe supports were designed in accordance wi+h ASME Section !!!, tubsection NF, and the downstream supports were designed in accordance with ANSI B31.1 (Reference 14). The load combinations were consistent with the 1cto combinations in the EPRI suomittal Guide (Reference 19),, and all stresses were less tuan the code allowables.

In EPRI tests performed on the Crosby 6:46 safety-valve, pressure oscillations of 170-260 Hz occurred in the piping upstream of the safety valve as the loop seal water was discharged. This phenomenon was not accounted for,in the structural analysis of the system. The piping upstream of the safety valves in the EPRI tests was 8-in. Schedule 160 and bin.

Schedule XX while at Braidwood, Units 1 and 2, the piping is 6-in.

Schedule 160.

The test piping did not sustain any discernible damage during

[

the tests.

Thus, the plant piping is also not expected to be damaged during similar oscillations, and an analysis for these pressure oscillations is not necessary for this plant.

t 4.4.3 Pioino and Suoport Summary l

y The selection of a bounding case for the piping evaluation and the pipfngandsupportanalysisdemonstratethattherequirementsofItem3and Item 8 of Section 1.2 outlined in this report were met.

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EVALUATION Sul91ARY r>

The Licensee'for Braidwood, Units 1 and 2, provided an acceptable response to the requirements of NUREG-0737, and thereby reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.

P 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 designed to qualify,the operability of prototypical' valves and to demonstrate that their operation would not inva.11date the integrity of the associated equipment and piping.

The subsequent tests were successfully completed under operating conditions which by analysis bounded the most probable maximum forces expected from anticipated design basis events. The generic test results and piping analyses showed that the valves tested functiont.d correctly and safely for 4

all relevant steam discharge events specified in the test program and that the pressure boundary component design criteria were not exceeded.

Analysis and review of the test results and the Licensee's justifications indicated direct applicability of the prototypical valve snd valve perintmances to the in-plant valves and systems intended to be covered by the generic test program. The Licensee must document a formal procedure to inspect the safety valves each time they discherge the loop seal or water.

The plant specific piping was shown by an6.'cAis to meet code require.ments.

Thus, the requirements of Item !!.D.1 of NUREG-0737 were mej; 11 tams 1-8

't in Paragraph 1.2) and, thereby demonstrate by testing and enalysis, that the reactor primary coolant pressure boundary will have a low probability of abnsrmal leakage (General Design Criterion No. 14) and that the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) were designed with sufficient margin such that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).

Furthermore, 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 (General Design Criterion No. 30).

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REFERENCES 1.

TMI Lessons Learned Task Force Status Report and Short-Term Recomendations, NUREG-0578, Ju' y 1979.

2.

Clarification of TMI Action Plan Recuirements, NUREG-0737, November 1980.

i 3.

D. P. Hoffman, Consumers Power Co., letter to H. Denton, NRC,

" Transmittal of PWR Safety and Relief Valve Test Program Reports,"

j September 30, 1982.

l 4.

EPRI Plan for Performan:, Testino of PWR Safety and Relief Valves, July 1980.

5.

_EPRI PWR Safety and Relief Valve Test procram Valve,

Selection / Justification Report, EPRI NP-2292, January 1983.

6.

EPRI PWR Safety and Relief Valve Test Procram Test Condition Justification Report, EPRI NP-2460. January 1983.

7.

Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves

]

in westinnhouse-Desicned Plants, EPRI NP-2296, January 1983.

l 8.

iPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Repo_rt, EPRI NP-2628-SR, December 1982.

/

9.

R. C. Youngdahl, Consumtrs Power Co., letter to H. Denton, NRC, 6

" Submittal of PWR Valve Data Package," June 1, 1982.

10.

EPRI PWR Safety and Relief Valve Test Procram Guide for Ap>1ication of Valve Test Procram Results to Plant-Specific Evalu.ations, tevision 2 Interim Report, July 1982.

11.

T. R. Tramm. Commonwealth Edison, letter to H. R. Denton, NRC, " Byron i

Station Units 1 and 2, Braidwood Station Units 1 and 2, Pressurizer Safety and Reli,ef Valves," October 26, 1982.

12.

T. R. Tramm Commonwealth Edison, letter to H. R. Denton, NRC, " Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Pressurizer Safety and Relief Valves," December 30, 1983.

13.

J. A. Lw.> ens, NRC, letter to D. L. Farrar Commonwealth Edison,

" Request for Additional Information on NUREG-0737. Item II.D.1, l,

e Performance Testing of Safety and Relief Valves," March 31, 1987.

14.

K. A. Ainger, Commonwealth Edison, letter to USNRC Document Control Desk, " Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Pressurizer Relief and Safety Valves, NRC Docket Nos. 50-454/455 and l

50-456/457," December 2, 1987.

p 15.

Application of RILAPS/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic loads, EPRI-2479, December 1982.

l 23 s

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G. K. Miller et al.,' Technical-Evaluation Reoort TMI Action--NUREG-0737 16.

(II.D.1) Relief and Safety Valve Testino Diablo Canyon Units 1 and 2.

00cket No. 50-275. 50-323 EGG-RST-6972, July 1985.

]

17.

R. L. Tedesco, NRC letter to T. M. Anderson, Westinghouse Electric Co.,

i

" Acceptance for Referencing of Licensing Topical Report WCAP-8252, Revision 1," April 7, 1981.

18.

L. C. Smith and T. M. Adams, " Comparison of Analytically Determined Structural Solutions with EPRI Safety Valve Test Rasults," 4th National Congress on Pressure vessel and Pipino Technolony. Portland. Orecon.

R June 19-24. 1983, PVP-Volume 74, pp. 193-199.

1 19.

EPRI PWR Safety ?nd Relief Valve Test Procram Guide for Ao)11dation of l

Valve Test-Procram Results to Plant-Soecific Evaluations,,tevision 2,

, i Interim Report, July 1982.

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