ML20127F665

Forwards Responses to NRC Questions 1-11 & 14 Re Pressurizer Safety & Relief Valves,Per .Responses to Questions 12 & 13 Will Be Provided by 851001

ML20127F665
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	06/18/1985
From:	Leblond P COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 0269K, 269K, NUDOCS 8506250222
Download: ML20127F665 (17)

Text

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Commonwealth Edison On3 First Nation 11 Plaza. Chicigo. Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 June 18, 1985

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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Zion Nuclear Power Station Units 1 and 2 NUREG-0737, Item II.D.1 NRC Docket Nos. 50-295 and 50-304

Reference:

February 19, 1985 letter from S. A. Varga to D. L. Farrar.

Dear Mr. Denton:

The referenced letter contained 14 questions concerning Zion's pressurizer safety and relief valves. This submittal transmits Commonwealth Edison Company's responses to question numbers 1 through 11 and 14. The responses to questions 12 and 13 will be provided by October 1, 1985, as discussed with J. Norris of your office on March 22, 1985.

If you have any further questions regarding this matter, please contact this office.

Very truly yours, fP. C. LeBlond Nuclear Licensing Administrator 1m 1

Attachment cc: NRC Resident Inspector - Zion J. Norris - NRR G. Wright - State of Ill.

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ATTACHMENT COMMONWEALTH EDISON COPANY Zion Station Units 1 and 2 NUREG 0737 Item II.D.1 Submittal Concerning Plant Specific Valve Operability Assessments 0269K

NRC Question 1:

The Westinghouse valve inlet fluid conditions report identifies Zion 1 and 2 as plants not being covered by the report with respect to the cold overpressurization event. However, the Zion 1 and 2 submittal states that the cold overpressure inlet fluid conditions for the 4-loop reference plant as provided in the Westinghouse report represents the Zion 1 and 2 cold overpressure inlet fluid conditions. Provide additional detail and discussion on how the Zion 1 and 2 cold overpressure transient and PORV. inlet conditions were determined to be represented by the Westinghouse inlet fluid conditions report.

Response to Question #1:

As stated in Reference 1, the fluid inlet conditions tested are comparable to _the range of conditions achievable at Zion in cold overpressurization events. In Reference 1, the Zion PORV setpoints in cold shutdown are conservatively assumed to be 500 psig, in contrast to the Technical Specification limit of 435 psig. The maximum liquid pressure that can be reached in a postulated cold overpressurization event is 519 psig as reported in Reference 1. The maximum liquid pressure in the EPRI test was reported as 600 psig which is comparable to the Zial condition of 519 psig.

NRC Question 2:

In valve operability discussions on cold overpressurization transients, the submittal identifies conditions for water and steam discharge transients. Since no low pressure steam tests were performed for the PORVs, provide justification that the high pressure steam tests demonstrate operability for the low pressure steam case for both opening and closing of the PORVs.

Response to Question #2:

The EPRI test matrix was selected to envelope the most severe conditions to denonstrate valve operability. This matrix was developed jointly by the valve manufacturers and the utilities. After thorough review and discussion with NRC staff, it was agreed that this test matrix did envelop the most severe conditions for valve operability and the tests were performed. ,

As stated in Reference 1, the Zion pressurizer power operated relief valves are Copes-vulcan 2 inch Model D-100-160 type, with 17-4PH cage and 316SS stellite clad plug. Because this is a direct-acting valve, it is not expected that the primary fluid conditions will significantly affect either the valve-opening or valve-closing time. This is clearly demonstrated by the insignificant variation in main disc opening or closing times for water in Table VIII-3 of Reference 2. The variation in main disc opening or closing time measured in the EPRI tests is indeed small when compared to the PORV cpening time of 2.5 seconds assumed in the Zion cold overpressurization system design (which conservatively exceeds the 0.66 second opening time observeo as a maximum in all the EPRI tests) or the PORV system design closure time of 5 seconds for this same postulated event (which conservatively exceeds the 1.24 second closure time observed as a maximum in the EPRI tests).

Furthermore, a cold overpressurization event represents the most likely source for 11 auld S/RV discharge in which the liquid may not be preceded by any steam discharge. This postulated event was selected because subcooled liquid is present in the pressurizer and S/RV discharge may assume maximum mass flow rates and hence result in waterhammer and high downstream piping loads. Valve operability requirements and downstream piping loads for low pressure steam flow conditions are clearly much less severe than for the case of subcooled liquid discharge.

NRC Question 3:

Results from the EPRI tests on the Crosby safety valves indicate that the test blowdowns exceeded the design value of 5% for both the "as installed" and " lowered" ring settings. If the blowdowns expected for Zion 1 and 2 also exceed 5%, the higher blowdowns could cause a rise in pressurizer water level such that water may reach the safety valve inlet line and result in a steam-water flow situation. Also the pressure might be sufficiently decreased such that adequate cooling might not be achieved for decay heat removal. Discuss these consequences of higher blowdowns if increased blowdowns are expected.

Response

As stated in Reference 1, the Crosby 6M6 valve always closed within 10%

of the design actuation pressure. The manufacturer's (Crosby) original blowdown ring adjustments provided 10% or less blowdown. The Zion plant specific blowdown ring adjustments have never been varied for service from the manufacturer's recommended positions. These ring settings are verified correct by procedure during setpoint verification testing conducted during each refueling outage. Therefore, the same blowdown performance is predicted for Zion as that demonstrated in the EPRI q tests.

The consequences of higner blowdown have been evaluated. Typically, i there are seven transients at Zion for which SRV actuation is necessary to protect the primary system and each of these transients may experience a higher blowdown. The transients considered are:

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1. Loss-of-Load;

2. Loss-of-Offsite Power;

3. Loss-of-Normal Feedwater;

4. Rod Withdrawal at Power;

5. Locked Rotor;

6. Rod Ejection; and

7. Feedline Break Based upon the results provided by Westinghouse in WCAP-10105 for the limiting overpressurization event for four-loop plants, the loss-of-load (turbine trip) event was selected to evaluate the consequences of a higher blowoown in the Zion Station. The thermal hydraulic program RETRAbO2 M00002 was used to model Zion, and the model was verified (benchmarked) by comparison with the FSAR turbine trip analysis. The same SRV opening characteristics used by Westinghouse in their analysis were incorporated into the Zion model, and blowdown values of 5%,10%

and 15% were evaluated by examining the same 4 cases that are in the Zion FSAR Chapter 14 analyses. The results of this study demonstrated that increasing blowdown effectively reduced value cycling and that blowdown values of 15% would not result in operational concerns.

Therefore, it is concluded that an increase of pressurizer safety valve blowdown setpoint to 15% does not result in plant operational concerns.

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tEC Question 4:

The Westinghouse Inlet Fluid Conditions Report stated that liquid discharge through both the safety and relief valves is predicted for an FSAR feedline break event. The Westinghouse report provided expected peak pressure and pressurization rates for specific plants having an FSAR feedline break analysis. The Zion 1 and 2 plant was not included in this list 01 plants having such an FSAR analysis. The submittal did not discuss the feedline break event. The NUREG-0737, however, requires analysis of accidents and occurrences referenced in Regulatory Guide 1.70, Revision 2, and one of the accidents so required is the feedline break. Provide a discussion on the feedwater line break event for Zion 1 and 2 identifying the fluid pressure, pressurization rate, fluid temperature, valve flow rate, and time duration for the event. Assure that the fluid condition was enveloped in the EPRI tests and demonstrate operability of the safety and relief valves for this event. Further, assure that the feedline break event was considered in analysis of the safety / relief valve piping system.

Response

This postulated event is very unlikely to occur at Zion Station. The only situation in which this event may produce liquid discharge from the pressurizer is if a large pipe break occurs between the check valve and the steam generator, thereby causing a loss of heat sink in the affected loop and there is no operator action to terminate the transient.

The FSAR transient response for Byron /Braidwood is shown in Fjgure 15-2.4 of the FSAR. This event is not analyzed in the Zion FSAR, because it is not a design transient required for NRC licensing. However, the results for Zion are not expected to be significantly different from those presented for Byron /Braidwood. Following the injection of cold ECCS water, the pressurizer pressure and level both initially drop due to the negative surge rates caused by primary loop water shrinkage.

After about 5 minutes, the pressure again rises due to reduced heat -

removal through the steam generators. Eventually saturated steam is ,

discharged through the S/RVs at about 7 minutes into the transient followed by a transition to saturated liquid discharge after 13 minutes into the event. The safety relief valves thus serve as a heat (and mass) sink to stabilize the transient until their final closure at 20 minutes.

In the Byron FSAR analysis, no credit is taken for the operation of the PORVs, which would open at 2335 psig and discharge saturated steam for pressure relief. One PORV would be adequate to balance the cold-water input from both safety injection pumps and thus control the primary loop pressure to 2335 psig, so that the spring-loaded safety valves would never open. It is possible, however, that the PORVs may eventually discharge some saturated liquid after a transition from saturated steam discharge if there is no operator action.

No credit is taken for operator action in this postulated event, although shutting down the charging pumps would at any time ouring the 20 minute transient effectively terminate the pressure increase leading to high pressure liquid discharge. Operating procedures exist which require such actions to be taken once the pressurizer pressure has stabilized and begun its increasing trend.

The initiation frequency of this event is very small, because it involves a large break in a relatively short stretch of piping between the check valve and steam generator. Based upon the WASH-1400 pipe failure probability of 1.0E-6/yr- section, a conservative initiation frequency is 1 x 10-6 events / reactor-year. The PORV system failure probability is 5.4E-4 for Zion, (see Appendix A of Reference 1), -

resulting in a very small event frequency which would be even smaller should operator action be included. Compared to the other postulated accident events, the feedwater line break is thus insignificant as an initiator for high pressure liquid discharge through safety relief valves. Further, such liquid discharge is always preceded by saturated steam oischarge, so that chatter instability is precluded, and the transient loads to discharge piping would be substantially mitigated.

Therefore, the EPRI tests which included subcooled water, steam and transition conditions (i.e. steam to water) clearly enveloped all fluid  ;

conditions predicted to occur at Zion Units 1 and 2 for a postulated feedwater line break.

g NRC Question 5:

Overpressure transients cause the pressurizer sprays to activate which adds moisture to the steam volume. When the safety or relief valves open they would then pass a steam-water mixture. Since the safety inlet piping utilizes loop seals, it has been concluded that this condition has been enveloped by the water discharge case for the safety valves.

The submittal did not identify if loop seals are used upstream of the PORVs. The Science Applications, Inc. (SAI) report for Zion Units 1 and 2 states that PORV liquid discharge is possible for the extended high pressure injection event. Zion Units 1 and 2 should provide a discussion for one of the following: (1) was the steam water discharge case considered, or (2) was a solid water discharge case considered in establishment of maximum loads for the PORV discharge piping.

Response to Question #5:

A discussion relating to steam / water or solid water discharge is not relevant for Zion Station. All of the FSAR transient analyses, for which Zion is licensed, do not result in any liquid discharge from the PORVs.

The SAI discussion of the " extended high pressure injection event" represents a low probability event (3.6E-0 events / reactor-year) beyond Zion's design basis, utilizing extremely pessimistic assumptions such as no operator action. No further analysis is warranted for such a postulated event.

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_8-NRC Question 6:

The EPRI Inlet Fluid Justification Report suggested a method for demonstrating safety valve stability. This method compares the total inlet piping pressure drop for the in- plant safety valves and piping to the applicable EPRI test safety valve and piping combinations. The total inlet piping pressure drop is composed of a frictional and acoustic wave component evaluated under steam conditions. The Zion 1 and 2 plant submittal did not provide pressure drop calculations or any other methods to demonstrate safety valve stability. Provide the necessary documentation and discussion demonstrating stability for the Zion 1 and 2 plant safety valves at the expected inlet conditions, ring settings and inlet piping configuration.

Response _:_

Stability analysis of S/RV chatter is complicated by two considerations. First, important chatter modes involve full- stroke motion with sudden stopping of the valve stem face. A valve dynamic model must include nonlinear response characteristics to account for such " pegging" (or stroke limitation) and thus cannot be even qualitatively represented by a linear raadel involving small and smooth motions. Therefore, the analysis must be carried out in the time domain (as opposed to the Laplace Transform domain, for example) using a nonlinear computer model.

Second, a chatter instability occurs occause of a mutual interaction -

between the fluid and the valve face. Thus, the valve dynamics and

- fluid behaviour cannot be treated separately and a fully coupled model is required to analyze the phenomena.

An approach based upon discrete wave propagation was selected to determine safety valve stability for the Zion Station. In this approach, only the " local" pressure, or instantaneous upstream pressure at the valve face, is of interest in determining the valve acceleration. As the valve face moves, the available discharge flow area changes, in turn changing the inlet velocity to the valve. By approximating the valve motion by a sequence of steps, the inlet fluid velocity can likewise be represented as a series of step functions in time. Each step change in inlet velocity generates a small pressure wave. The small incremental square wave is propagated down the pipe where it is reflected from the pressurizer and eventually returns to the valve face to influence the local pressure there, but at a time delayed by 2L/c, where L is the pipe length and c is the acoustic speed. This delay time corresponds to the transit time required for an emitted pressure wave to reflect off the pressurizer boundary and return to the valve.

By computing the magnitude of the pressure waves due to valve motion, tracking their progress and attenuation as they propagate to the pressurizer and back to the valve, and evaluating their eventual contribution to the local pressure at the delayed time, we can fully account for the fluid-valve interactions without solving the partial differential equations governing the detailed transient pressure distribution. This has the additional advantages that several valve characteristics and fluid conditions can be evaluatad for the plant specific geometry. This method of computer solut' .ss programmed and calibrated to EPRI/CE test #903. Comparable test 't n EPRI/CE 900 and 1400 series have been shown to give results consit. with test #903.

This computer model was used to evaluate several different cases when saturated steam was discharged fran the pressurizer. The response of the Crosby 6M6 Safety / Relief valve and inlet pipe loop seal configuration to saturated steam discharge was analyzed for the Zion Station. The blowdown stability limit was determined for a range of pressurizer pressures. The results of this evaluation show that a blowdown setting of 5% will result in stable valve performance for saturated steam discharge during all expected overpressure events. Therefore, this analysis has demonstrated the stability of the Zion Units 1 and 2 safety valves at the expected fluid inlet conditions, ring settings and inlet piping configuration.

NRC Question 7:

The Westinghouse inlet fluid conditions report stated that liquid flow could exist through the PORV for the FSAR feedline break event and the extended high pressure injection event. Liquid PORV flow is also predicted for the cold overpressurization event. These same flow conditions will also exist for the Block Valve. The EPRI/ Marshall Block Valve Report did not test the block valves with fluid media other than steam. The Westinghouse Gate Valve Closure Testing Program did include tests with water; however, the information presented in the report did not provide specific test results. Since it is conceivable that the EMOV could be expected to operate with liquid flows, discuss EMOV block valve operability with expected liquid flow conditions and provide specific test data.

Response

As stated in Reference 1, we are in full agreement with ti e Generic PWR utilities position and do not see any need for further tet ting of Block Valves. The PWR utilities' position was thoroughly discussed with NRC staff at a meeting held on July 17, 1981. Substantial agre3 ment was reached at this meeting regarding resolution of the testirg requirements for block valves. At Zion this block valve is not used for: shutting off flow or leakage from the pressurizer during normal operation. The operability of block valves is not a safety issue. The PORV's installed and utilized at the Zion Station have been tested in the EPRI Owners Group program for the most severe bounding fluid inlet conditions achievable at Zion. The test results clearly show that these PORV's are operable under all severe inlet fluid conditions tested and that seat leakages never exceeded 0.0042 GPM. Operability of block valves does not significantly affect the probability of a small break LOCA.

Furthermore, a similar 2.5 inch Velan gate valve was in service at TMI during the accident of 1G79. This valve was cycled 30 times at 2000 psig and was still able to shut-off the flow from the pressurizer (See velan Valve Corporation, Technical Nuclear Safety Report No.150, December, 1979). Therefore, we do not believe that further analysis of block valve operability with liquid flows or further testing under such conditions is warranted.

NRC Question 8:

Bending moments are induced on the safety valves and PORVs during the time they are required to operate because of discharge loads and thermal expansion of the pressurizer tank and inlet piping. Make a comparison between the predicted plant moments with the moments applied to the tested valves to demonstrate that the operability of the valves will not be impaired.

Response

A quantitative evaluation of the bending moments induced on the safety valves and PORVs during operation because of discharge loads and thermal expansion will be provided in our response to NRC Question 13.

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NRC Question 9:

The Westinghouse Valve Inlet Fluid Conditions Report states that liquid discharge could be expected through the safety valves for both the feedline break and extended high pressure injection events. The EPRI 6M6 test safety valve experienced some chatter and flutter while discharging liquid at certain ring settings. Testing was terminated after observing chattering to minimize valve damage. Inspection revealed some valve damage which was presumably caused by the valve chatter and flutter. Liquid discharge for Zion 1 and 2 may conceivably occur for longer periods of time than the EPRI testing. Thus longer periods of valve chattering may cause severe valve damage. Discuss the implications this may have on operability and reliability of the Zion 1 and 2 safety valves. Identify any actions that will be taken to inspect for valve damage following safety valve lift events.

Response

As stated in Reference 1, there were no observations in the EPRI tests of structural damage to the 6M6 that impeded its ability to function repeatedly. The same or better results are expected at Zion because:

(1) Subsequent actuations are expected to take place only with saturated steam fluid inlet conditions for Zion plant specific situations. This is clearly less severe than repeated loop seal discharges, such as those that EPRI experienced by conducting multiple loop seal tests without overhauling the 6M6 between tests.

(2) Despite high frequency cycling (other than loop seal discharge) in subcooled water tests ano tests such as 920 and 1419, no damage sufficient to freeze the 6M6 in a stuck-open or stuck-closed position occurred during the duration of the tests. Although these tests are not directly applicable to Zion, they do demonstrate that the 6M6 is capable of withstanding a highly significant amount of structural distress beyond What is anticipated in the Zion application.

Furthermore, Appendix A of Reference 1 clearly demonstrates that the only fluid inlet condition for the Zion Station consist of discharge of the loop seal water immediately upstream of the valve, followed by saturated steam for all such events that may resJlt in safety valve actuation. The Westinghouse Valve Inlet Fluid Coquitions Report is based upon assumed generic plant characteristics end is not applicable to the Zion Station for this situation. Therefore, no additional actions are required to inspect for valve damage other than the existing maintenance procedures and existing schedule of inspection.

NRC Question 10:

NUREG-0737 Item II.O.1 requires that the plant-specific PORv control circuitry be qualified for design-basis transients and accidents.

Please provide information which demonstrates that this requirement has been fulfilled.

Response to Question #10 Subsequent to the issuance of NUREG-0737, the NRC's requirements for environmental qualification of electrical equipment were modifieo in 10 CFR 50.49 (48 FR 2733, January 21, 1983). Because the PORVs are not safety-related and are not relied upon to mitigate design-basis accidents, technically they do not fall within the scope of 10 CFR 50.49.

Nevertheless, all of the electrical components in the PORV control circuitry that can be exposed to a harsh environment (solenoid valves, limit switches, cable, terminal blocks, electrical penetrations) have been included in Zion's Environmental Qualification Program, and are fully qualified. (Reference 3) 1

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NRC Question 11:

The Zion Units 1 and 2 plant safety valves are Crosby 6M6 and were tested by EPRI. EPRI testing of the 6M6 was performed at various ring settings. The submittal did not provide the present ring settings but stated the rings were set at the factory recommended settings. If the plant current ring settings were not used in the EPRI tests, the results may not be directly applicable to the Zion Units 1 and 2 safety valves.

Identify the Zion Units 1 and 2 safety valve ring settings. If the plant specific ring settings were not tested by EPRI, explain how the expected values for flow capacity, blowdown, and the resulting back pressure corresponding to the plant-specific ring settings were extrapolated or calculated from the EPRI test data. Identify these values so determined and evaluate the effects of these values on the behaviur of the safety valves.

Response

As stated in Question 3, the manufacturer's original blowdown ring adjustments for both the Zion Station and the EPRI tests provided for 10% or less blowdown. Therefore, the same performance is expected for Zion as that demonstrated in the EPHI tests. Of course, similar flow capacity and back pressure characteristics are expected at Zion as Ucmonstrateo in the EPRI test results. The test results clearly demonstrate that the Crosby 6M6 valves used at Zion meet or exceed all design req 2irements.

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NRC Question 14:

According to results of EPRI tests, high frequency pressure oscillations of 170-260 Hz typically occur in the piping upsteam of the safety valve while loop seal water passes through the valve. An evaluation of this 1 phenomenon is documented in the Westinghouse Report WCAP 10105 and states that the acoustic pressures occurring prior to and during safety valve discharge are below the maximum permissible pressure. The study discussed in the Westinghouse report determined the maximum permissible pressure for the inlet piping and established the maximum allowable ',

bending moments for Level C Service Condition in the inlet piping based on the maximum transient pressure measured or calculated. While the internal pressures are lower than the maximum permissible pressure, the pressure oscillations could potentially excite high frequency vibration modes in the piping, creating bending moments in the inlet piping that should be combined with moments from other appropriate mechanical loads. Provide one of the following: (1) a comparison of the expected peak pressures and bending moments with the allowable valves reported in the WCAP report or (2) justification for other alternate allowable pressure and bending momants with a similar comparison with peak pressures and moments induced in the plant piping.

Response

Some high frequency pressure oscillations in the piping upstream of the safety valve were observed in the EPRI tests while the loop seal water passes through the valve. An evaluation of this phenomenon, documented in WCAP 10105, clearly shows that the acoustic pressures occurring prior to and during safety valve discharge are below the maximum permissable pressure and the maximum allowable bending noments for Service Level C.

The ASME code allowables are conservatively selected to account for uncertainties in the forcing functions, materials properties, dynamic effects, etc.. There is no ASME code or NRC licensing requirement to evaluate the potential for exciting high frequency vibration modes in the piping caused by internal pressure oscillations. There is no need for additional evaluation because of the following:

1. The pressure stresses and bending moment stresses are conservatively estimated, conservatively conbined with stresses from other loads, and then compared to conservative code allowable stresses. This procedure provides an adequate safety margin that accounts for uncertainties in the forcing functions of the loads.

2. Safety valve actuation is an extremely infrequent event at Zion because the PORV's are adequate to relieve the pressure before safety valve actuation can occur.

3. If a safety valve did actuate and discharge the loop seal water this transient is over within several hundred milliseccnds. _This short duration transient in which valve chatter may potentially occur is not likely to excite significant high frequency vibration modes in the p!. ping.

4. Finally, it must be observed that an evaluation of loop seal water discharge using the computer simulation model described in our response to Question 6 suggests that valve chatter may not occur at all for Zion station. If a small amount of internal valve damping (e.g. 6% of critical damping) exists, saturated liquid loop seal disc 6arge for a pressure ramp reaching 2600 psia occurs without valve chatter at a blowdown of 4.5%. Therefore, it is not prudent to perform either additional tests or analysis to provide additional evaluation of this phenomena because the Zion Station safety margin is not significantly affected.

References

1. Letter to D. G. Eisenhut, USNRC, from E. O. Swartz, CECO, " Zion Station Units 1 and 2, NUREG 0737, Item II.D.I, Plant Specific Submittal, NRC Docket Nos. 50-295/304", dated July 1, 1982.

2. "EPRI/Wyle Power-Operated Relief Valva Phase III Test Reports.

Volume 8: Summary of Phase III Testir3g of the Copes-Vulcan 316 W/ Stellite Plug and 17-4PH Cage Relief Valve". EPRI Np-2670-LD.

3. Zion EQ Report, May 19, 1983 letter from F. G. Lentine to H. R.

Denton.

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