ML20058G376

Responses to NRC 820708 Request for Addl Info Re Safety Relief Valve Testing.Certificate of Svc Encl

ML20058G376
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	07/29/1982
From:	Irwin D HUNTON & WILLIAMS, LONG ISLAND LIGHTING CO.
To:
References
ISSUANCES-OL, NUDOCS 8208030301
Download: ML20058G376 (19)

Text

r-DOCKETED LILCO, July 29, 1982sNRC

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION '

Before the Atomic Safety and Licensing Board In the Matter of )

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LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)

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(Shoreham Nuclear Power Station, )

Unit 1) )

RESPONSE OF LONG ISLAND LIGHTING COMPANY TO NRC REGULATORY STAFF QUESTIONS OF JULY 8, 1982 RELATIVE TO SRV TESTING Long Island Lighting Company has received a letter, dated July 8, 1982, from the NRC Regulatory Staff involving six ques-tions relating to the testing of Safety Relief Valves for the Shoreham Station. The covering letter, as amplified by the oral testimony of Regulatory Staff witnesses, indicated that the Staff felt that it needed more information in the area described in the six questions attached to the letter in order to complete its review of SRV testing for Shoreham. The following submittal contains LILCO's response to the six questions.

Respectfully Submitted LONG ISLAND LIGHTING COMPANY j

8208030301 820729 PDR ADOCK 05000322 Donald P. Irwin G PDR Hunton & Williams Pont Office Box 1535 707 East Main Street Richmond, Virginia 23219 DATED: July 29, 1982 -

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1. Q. The test program utilized a " rams head" discharge pipe configuration. Shoreham utilizes a " tee" quencher configuration at the end of the discharge line. Des-cribe the discharge pipe configuration used at Shoreham and compare the anticipated loads on valve internals in the Shoreham configuration to the measured loads in the test program , Discuss the impact of any diff-erances in loads on valve operability.

A. The safety / relief valve discharge piping configuration at Shoreham utilizes a " tee" quencher at the discharge pipe exit. The average length of the 11 SRV dis-charge lines (SRVDL) is 137' and the submergence length in the suppression pool is approximately 13'.

The SRV test program utilized a ramshead at the dis-charge pipe exit, a pipe length of 112' and a sub-mergence length of approximately 13'. Loads on valve internals during the test program are larger than loads on valve internals in the Shoreham configuration for the following reasons:

1. No dynamic mechanical load originating at the " tee" quencher is transmitted to the valve in the Shoreham configuration because there is at least one anchor i

i point between the valve and the tee quencher.

2. The first length of the segment of piping downstream of the SRV in the test facility was twice that of l Shoreham piping, thereby resulting in a bounding dynamic mechanical load on the valve in the test t

program.

3. Dynamic hydraulic loads (backpressure) are experienced by the valve internals in the Shoreham configuration.

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The backpressure loads may be either (1) transient backpressures occurring during valve actuation, or (ii) steady-state backpressures occurring during steady-state flow following valve actuation.

(a) The key parameters affecting the transient back-pressures are the fluid inertia in the submerged SRVDL and the SRVDL air volume. Transient backpressure increases with line submergence and decreases with air volume. The transient backpressure in the test program was maximized by utilizing a submergence of 13', not less than Shoreham, and a pipe length of 112' which is less than Shoreham.

(b) The steady-state backpressure in the test program was maximized by utilizing an orifice plate in the SRVDL above the water level and before the ramshead. The orifice was sized to produce a backpressure greater than that calculated for any of the Shoreham SRVDL's.

The differences in the line configuration between the Shoreham plant and the test program as discussed above result in the loads on the valve internals for the test facility which bound the actual Shoreham loads. An addi-tional consideration in the selection of the ramshead for the test facility was to allow more direct measure-

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ment of the thrust load in the final pipe segment.

Utilization of a " tee" quencher in the test program would have required quencher supports that would unnecessarily obscure accurate measurement of the pipe thrust loads.

For the reasons stated above, differences between the SRVDL configurations in Shoreham and the test facility will not have any adverse effect on SRV operability at Shoreham relative to the test facility.

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2. Q. The test configtration utilized no spring hangers as pipe supports. Plant specific configurations do use spring hangers in conjunction with snubber and rigid supports. Describe the safety relief valve pipe, sup-ports used at Shoreham and compare the anticipated loads on valve internals fortthe Shoreham pipe supports to the measured loads in the test program. Describe the impact of any dif ferences in loads on valve operability.

A. The Shoreham safety-relief valve discharge lines (S RVDL's) are supported by a combination of snubbers, rigid sup-ports, and spring hangers. The locations of snubbers and rigid supports at Shoreham are such that the loca-tion of such supports in the BWR generic test facility is prototypical, i.e., in each case (Shoreham and the test facility) there are supports near each change' of direction in the pipe routing. Additionally, each SRVDL at Shoreham has only one or two spring hangers, all of which are located in the drywell. The spring hangers, snubbers, and rigid supports were designed to accommodate combinations of loads resulting from piping dead Weight, thermal conditions, seismic and suppression pool hydrodynamic events, and a high pressure steam discharge transient.

The dynamic load effects on the piping and supports of the test facility due to the water discharge event (the alternate shutdown cooling mode) were found to be sig-nificantly lower than corresponding loads resulting from the high pressure mteam discharge event. As stated in a

NEDE-24988-P, this finding is considered generic to all BWR's since the test facility was designed to be proto-typical of the features pertinent to this issue. Fur-thermore, analysis of a typical Shoreham SRVDL con-figuration has confirmed the applicability of this conclusion to Shoreham.

During the water discharge transient there will be sig-nificantly lower dynamic loads acting on the snubbers and rigid supports than during the steam discharge transient. This will more than offset the small increase in the dead load on these supports due to the weight of the water. Therefore, design adequacy of the snub-bers and rigid supports is assured as they are designed for the larger steam discharge transient loads.

This question addresses the design adequacy of the spring I

hangers with respect to the increased dead load due to l the weight of the water during the liquid discharge i

transient. As was discussed with respect to snubbers and rigid supports, the dynamic loads resulting from liquid discharge during the alternate shutdown cooling mode of operation are significantly lower than those from the high pressure steam discharge. It is believed I

that sufficient margin exists in the Shoreham piping l

system design to adequately offset the increased dead l

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load on the spring hangers in an unpinned condition due to a water filled condition. Nevertheless, stress analyses are being performed to confirm this assumption 4

regarding the increased deadweight loads for all SRVDL spring hangers. It should be noted that the effect of dead load weight does not affect the ability of SRVs to open to establish the alternate shutdown cooling path since the loads occur only after valve opening.

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3. Q. Report NEDE-24988-P did not identify any vdive functional deficiencies or anomalies encountered during the test program. Describe the' impact on valve safety function of any valve functional deficiencies or anomalies en-countered during the program.

> A. No functional deficiencies or anomalies of the safety relief or relief valves, not only for Target Rock two-stage valves but also for all other types of valves tested,wereexperienceddu\ingth'etestingbyWyle Laboratories for compliance with the alternate shutdown croling mode requirement. Al1 the valves subjected to test runs, valid and invalid, opefed and closed without .t loss of pressure integrity or damage. Anomalies en-countered during the test program were all due to failures of test facility instrumentation, equipment, data acquisition equipment, or deviation from the approved test procedure. N The test specification for each valve required six valid runs. Under the test procedure, any anomaly caused the test run to be judged invalid. In testing l for the Target Rock two-stage SRV, only one anomaly of any sort occurred: on water test run No. 302,,the test system GN2 regulator fai3 d, cenulting in a t'est which did not comply with c1 re adural test require-ments. The Wyle Laboratories test log sheet for the l Target Rock two-stage valve tests is attached.

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Each Wyle test report for the respective valves identi-fies each test run performed and documents whether or not the test run is valid or invalid and states the reason for considering the run invalid. No anomaly encountered during the required test program affects any valve safety or operability function.

All valid test runs are identified in Table 2.2-1 of NEDE-24988-P. The data presented in Table 4.2-1 for each valve were obtained from the Table 2.2-1 test runs and were based upon the selection criteria of:

(a) Presenting the maximum representative loading in-formation obtained from the steam run data, (b) Presenting the maximum representative water loading information obtained from the 150 F subcooled water test data, (c) Presenting the data on the only test run performed for the 500 F subcooled water test condition.

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OPERABILITY TEST REPORT FOR TARGET ROCK 6X10 SRV FOR LOW PRESSURE WATER TESTS FOR GENERAL ELECTRIC COMPANY GENERAL $ ELECTRIC NUCLEAR ENERGY BUSINESS GROUP 8-Sa62 W Q ED DATE 9 7- I

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• VPF NO.

ll0 A 7O TRANSMITTA'L NO.

175 Curtner Avenue San Jose, California

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-9b- PAGE NO. 8 TEST REPORT NO. 17476-04 TABLE I TEST LOG FOR SRV TR-1 Test Test Load Line Test No. Media Configuration Date Remarks 301 Steam I 3/17/81 Acceptable 302 Water 1 3/17/81 GN2 Regulator failed.

Data not acceptable.

303 Water i 3/17/81 Acceptable 304 Steam 1 3/17/81 Acceptable 305 Water i 3/18/81 Acceptable 306 Steam I 3/18/81 Acceptable 307 Water I 3/18/81 Acceptable 308 Water i 3/18/81 special test at elevated temperature and low pres-sure requested by G.E.

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l WYLE LABORATORIES Huntsville Facility

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4. Q. The purpose of the test program was to determine valve performance under conditions anticipated to be encountered in the plants. Describe the events and anticipated con-ditions at Shoreham for which the valves are required to operate and compare these plant conditions to the condi-tions in the test program. Describe the plant features assumed in the event evaluations used to scope the test program and compare them to plant features at Shoreham. For example, describe high level trips to prevent water from entering the steam lines under high pressure operating conditions as assumed in the test event and compare them to trips used at Shoreham.

A. The purpose of the test program was to determine valve performance, under conditions anticipated to be en-countered in the plants, which could result in liquid or two phase flow through the valves. The alternate shutdown cooling mode is the only anticipated event which is expected to result in liquid at the valve inlet.

Consequently, this was the event simulated in the SRV test program. This conclusion and the test results applicable to Shoreham are discussed below. The alter-nate shutdown cooling mode has been described in the response to NRC question 5.

The SRV inlet fluid conditions tested in the BWR Owners' Group SRV test program, as documented in NEDE-24988-P, are representative of the fluid conditions expected to occur in the alternate shutdown cooling mode of opera-tion at Shoreham. These fluid conditions at the SRV i inlet are 15 P to 500 F subcooled liquid at 20 psig to 250 psig.

The BWR Owners' Group, in their enclosure to the September 17, 1980 letter from D. B. Waters to R. H.

Vollmer, identified thirteen events which could result in liquid or two phase SRV inlet flow. These events were identified by evaluating the initiating events described in Reg. Guide 1.70, Rev. 2, with and without the additional conservatism of a single active component failure or operator error postulated with the event sequence. Of these thirteen events, only eight are applicable to the Shoreham plant because of its design and specific plant configuration. For these eight events, the Shoreham specific features, such as trip logic, power supplies, instrument line configuration, alarms and operator actions, have been compared to the base case l analysis presented in the BWR Owners' Group September 17, 1980 submittal and subsequent discussions with the NRC Staff. This comparison has demonstrated that in each case, the base case analysis is applicable to Shoreham in that the base case assumptions are applicable.

For example, the base case analysis for the reactor level 8 failure /HPCI overfill event included a level 8 trip scheme with two out of two logic, two variable in-strument legs and one power supply inputting to one HPCI turbine trip mechanism with one turbine stop valve. This scheme is the same as the Shoreham design.

As discussed above, the Shoreham plant features are represented in the base case analysis performed in the BWR Owners' Group evaluation. This evaluation c6ncluded that the alternate shutdown cooling mode is the only expected operating event involving liquid or two phase flow and therefore requires testing. The alter-nate shutdown cooling mode fluid conditions tested in the BWR Owners' Group test program accurately bound the Shoreham plant specific fluid conditions expected for this event.

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5. Q. The valves are likely to be extensively cycled in a con-trolled depressurization mode in a plant specific appli-cation. Was this mode simulated in the test program?

What is the effect of this valve cycling on valve per-formar.ce and probability of the valve to fail open or to fail closed?

A. The BWR safety / relief valve (SRV) operability test program was designed to simulate the alternate shutdown cooling mode, which is the only expected liquid dis-charge event for Shoreham. The sequence of events leading to the alternate shutdown cooling mode is given below.

Following normal reactor shutdown, the reactor operator depressurizes the reactor vessel by opening the turbine bypass valves and removing heat through the main condenser. If the main condenser is unavailable, the operator could depressurize the reactor vessel by using the SRV's to discharge steam to the suppression pool. If SRV operation is required, the operator cycles the valves in order to assure that the cooldown rate is maintained within the technical specification limit of 100o F per hour.

This would require on the order of 1-10 cycles of the SRV. When the vessel is depressurized, the operator initiates normal shutdown cooling by use of the RHR system. If that system is unavailable because the valve on the RHR shutdown cooling suction line fails to open, the operator initiates the alternate shutdown cooling mode.

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For alternate shutdown cooling, the operator opens one SRV and initiates either an RHR or core spray pump utiliz-ing the suppression pool as the suction source. The re-actor vessel is filled such that water ~is allowed to flow into the main steam lines and out of the SRV and back to the suppression pool. Cooling of the sys-tem is provided by use of an RUR heat exchanger, As a result, an alternate cooling mode is maintained.

In order to assure continuous long term heat removal, the SRV is kept open and no cycling of the valve is performed. In order to control the reactor vessel cooldown rate, the operator is instructed to throttle the injection valve into the vessel. Consequently, no cycling of the SRV is required for the alternate shutdown cooling mode, and no cycling of the SRV was performed for the generic BWR SRV operability test program.

The ability of the Shoreham SRV to be extensively cycled for steam discharge conditions has been confirmed during steam discharge qualification testing of the valve by the valve vendor. This qualification testing for the Target Rock two-stage valve used in Shoreham has been previously identified in the Shoreham response to NRC question 212.51. Based on the qualification

testing of the SRV's, the cycling of the valves in a controlled depressurization mode for steam discharge conditions will not adversely affect valve performance and the probability of the valve to fail open or closed is extremely low, i

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6. Q. Describe how the values of valve C 's in report NEDE-24988-P will be used at Shoreham. Show that the methodoingy used in the test program to determine the valv6 c y will be consistent with the application at Shoreham.

A. The flow coefficient, C ,y for the Target Rock 6 x 10 two-stage pilot operated safety relief valve (SRV) utilized in Shoreham was determined in the generic SRV test program (NEDE-24988-P) . The average flow t

coefficient calculated from the test results for the l Target Rock two-stage valve, model 7567F, is reported l

l in Table 5.2-1 of NEDE-24988-P. This test value has been used by LILCO to confirm that the liquid discharge

. flow capacity of the Shoreham SRVs will be sufficient i

i to remove core decay heat when injecting into the reactor pressure vessel (RPV) in the alternate shutdown cooling mode. The Cy value determined in the SRV test demonstrates that the Shoreham SRVs are capable of return-ing the flow injected by the RHR or CS pump to the suppression pool.

If the operator were to place the shoreham plant in the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following parameters: RHR or CS flow rate, reactor vessel pressure and reactor vessel temperature.

The flow coefficient for the Target Rock two-stage valve reported in NEDE-24988-P was determined from the SRV

t flow rate when the valvo inlet was pressurized to approximately 250 psig. The valve flow rate was measured with the supply line flow venturi upstream of the steam chest. The C y for the valve was calculated using the nominal measured pressure differential between the valve inlet (steam chest) and 3' downstream of the valve and the corresponding measured flowrate. Furthermore, the test conditions and test configuration were repre-sentative of Shoreham plant conditions for the alternate l shutdown cooling mode, e.g. pressure upstream of the valve, fluid temperature, friction losses and liquid flowrate. Therefore the reported cy values are appropri-ate for application to the Shoreham plant.

In the Matter of LONG ISLAND LIGHTING COMPANY (Shorehan Nuclear Power Station, Unit 1)

Docket No. 50-322 (OL)

I certify that copies of the " Response of Long -

Island Lighting Company to NRC Regulatory Staff Questions of July 8, 1982 Relative to SRV Testing" were served upon the following by hand on July 29, 1982, as indicated by an asterisk:

Lawrence Brenner, Esq.* Bernard M. Bordenick, Esq.*

Administrative Judge David A. Repka, Esq.

Atomic Safety and Licensing U.S. Nuclear Regulatory Board Panel Commission U.S. Nuclear Regulatory 1717 H Street, N.W.

Commission Washington, D.C. 20555 Washington, D.C. 20555 Herbert H. Brown, Esq.*

Dr. Peter A. Morris

• Lawrence Coe Lanpher, Esq.

Administrative Judge Karla J. Letsche, Esq.

Atomic Safety and Licensing Kirkpatrick, Lockhart, Hill Board Panel Christopher & Phillips U. S. Nuclear Regulatory 1900 M Street, N.W.

Commission Washington, D.C. 20036 Washington, D.C. 20555 Secretary of the Commission Dr. James H. Carpenter

• U.S. Nuclear Regulatory Administrative Judge Commission Atomic Safety and Licensing 1717 H Street, N.W.

Board Panel Washington, D.C. 20555 U. S. Nuclear Regulatory Commission Atomic Safety and Licensing Washington, D.C. 20555 Appeal Board Panel U.S. Nuclear Regulatory Walter H. Jordan

• Commission Administrative Judge 1717 H Street, N.W.

Atomic Safety and Licensing Washington, D.C. 20555 Board Panel U.S. Nuclear Regulatory Mr. Mark W. Goldsmith Commission Energy Research Group Washington, D.C. 20555 400-1 Totten Pond Road Waltham, Massachusetts 02154

6 David J. Gilmartin, Esq. Ralph Shapiro, Esq.

Attn: Patricia A. Dempsey, Esq. Cammer and Shapiro, P.C.

County Attorney 9 East 40th Street Suffolk County Department of Law New York, N. Y. 10016 Veterans Memorial Highway Hauppauge, New York 11787 Matthew J. Kelly, Esq.'

State of New York MHB Technical Associates Department of Public Service 1723 Hamilton Avenue 3 Empire State Plaza Suite k Albany, New York 12223 San Jose, California 95125 Mr. Jay Dunkleberger Stephen B. Latham, Esq. New York State Energy Office Twomey, Latham & Shea Agency Building 2 33 West Second Street Empire State Plaza P. O. Box 398 Albany, New York 12223 Riverhead, New York 11901 Howard L. Blau, Esq.

217 Newbridge Road Hicksville, New York 11801

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Donald P. Irwin Hunton & Williams Post Office Box 1535 Richmond, Virginia 23212 DATED: July 29, 1982

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