ML20010C457

Forwards Response to NRC Requests for Info Re NUREG-0737 Item II.K.3.28 & SER Open Items 11,36 & 44

ML20010C457
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	08/18/1981
From:	Mccaffrey B LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.28, TASK-TM SNRC-614, NUDOCS 8108200100
Download: ML20010C457 (24)

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• 7 LONG ISLAND LIGHTING COMPANY I. l SHOREHAM NUCLEAR POWER STATION g j b-mm,..%,a P.O. BOX 618, NORTH COUNTRY ROAD

• WADING RIVER, N.Y.11792 August 18, 1981 SNRC-614 Mr. Harold R. Denton, Director . m Office of Nuclear Reactor Regulation , , p M! D /,,

U.S. Nuclear Regulatory Commission  %

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Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322  %

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Dear Mr. Denton:

Enclosed herewith are fd# teen (15) copies, per your directive, of brLCO responses to specit : NRC concerns which were previously identified as requiring additional information to complete NRC review. Attachment A provides a list of the specific responses included.

If you require additional information or clarification, please do not hesitate to contact this office.

Very truly yours, bk N #

B. R. McCaffre Manager, Project Engineering Shoreham Nuclear Power Station RWG/mh Enclosures cc: J. Higgins ODI

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@108200100 01081 PDR ADOCK 050005 I-A PDR F C-893 5

SNRC-614 August 18, 1981 ATTACIIMENT A Additional and/or revised information is provided for the following:

1. NUREG-0737, Item II.K.3.28 - Study and Verify Qualification of of Accumulators on ADS Valves.

2. SER Open Item No. 44 - Reactor Water Level Measurement Error.

3. SER Open Item No. 11 - Supplemental ECCS Calculations with NUREG-0630 Model.

4. SER Open Item No. 36 - Containment Purge System i

fl . a SNPS-1 FSAR NUREG-0737, II.K.3.28 - STUDY AND VERIFY OUALIFICATION OF ACCUMULATORS ~

ON ADS VALVES (AMEMDED RESPONSE)

Item 1. As a minimum safety function requirement, 6 of the 7 ADS valves must cycle one time. The design basis of the short term accumulators is to provide 2 ADS valve actuations at 70 percent of peak drywell pressure, this value is equal to 33.6 psig. A 25 psig air pressure differential is re-quired to actuate the ADS pilot valves, therefore the min-imum required pressure in the short term accumulators is 58.6 psig. This val'ue translates to 5 ADS valve actuations if the drywell ir at atmospheric pressure, assuming iso-thermal expansioz, The longest time period before which the ADS function is required after an accident is approximately 230 seconds, an represented by a 0.07 ft2 recirculation discharge-break with IIPCI failure. Details of this event are provided in FSAR Section 6.3.3.7 and on Figure 6.3.3-32.

Item 2. The leakage limits for each short term accumulator system are such that greater than 58.6 psig is maintained after 230 seconds based on our response to Item 1. In order to provide margin in both time and pressure, acceptance crit-cria for ADS short term accumulator leak-tight integrity will be the retention of 70 psig minimum, after starting at normal pressure of 90 poig, af ter a time period of 10 min-utes. This period provides nearly 20 percent margin in pressure in the time period which is much longer than that required for ADS function.

Item 3. Periodic leak testing of each ADS short term accumulator system will be accomplished by tempora-ily connecting a pressure gauge to the accumulator, pre.3arizing the accum-ulator to 90 psig, venting the supply header, and checking accumulator pressure after 10 minutes. This test will verify that the acceptance criteria of Item 2 have been met.

Item 4. The Icak test of Item 3 shall be performed periodically on a schedule consistent with the Shoreham integrated leak rate test to be performed during reactor shutdown or refueling, but in no case at intervals greater than 3 years. The ina-bility of any short term accumulator system to pass the accep-tance criteria of Item 2 will require repair / modification and retest per Item 3.

Item 5. The seismic and environmenta] qualification criteria for Shoreham are described in FSAR Sections 3.10 and 3.11. In addition, comprehensive description and status reports for 8/18/81 SNRC-614

,J SNPS-1 FSAR NUREG-0737, II.K.3.28 Page 2 Item 5. both the seismic and environmental qualification program have been submitted by SNRC-575 dated May 28, 1981, and SNRC-576 dated May 27, 1981 respectively. As described therein, Class IE electrical equipment is qualified in ac-cordance with NUREG-0588 Category 2.

The ADS accumulator system is qualified in accorcance with the program outlined above and, therefore, meets the re-quirements of GDC 2 and 4.

Item 6. The applicant will perform a leak test prior to initial op-eration. Should the leakage exceed that established in Item 2, the system will be repaired / modified and retested as required.

8/18/81 SNRC-614

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SNPS-1 FSAR SER OPEN ITEM #44 - LEVEL MEASUREMENT ERROR Review of Reactor Water Level Measurement Instrumentation:

The cold reference leg reactor water level mecsurement design for Shoreham is illustrated in Figure 1. Reactor vessel water level is measured by differential pressure transmitters which measure the

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difference in static head between two columns of water. One column is a " cold" (ambient temperature) reference leg outside the reactor

' vessel; the other is the reactor water inside the reactor vessel.

The measured differential pressure is a function of reactor water level.

The cold reference leg is filled and maintained full of condensate by a condensing chamber at its top which continuously condenses reactor steam and drains excess condensate back to the reactor vessel through the upper level tap connection to the condensing chamber.

The upper vessel level tap connection is located in the steam zone above the normal water level inside the vessel. Thus, the reference leg presents a constant reference static head of water to the high pressure tap on the d/p transmitter. The low-pressure tap of the transmitter is piped to a lower-level tap on the reactor vessel which is located in the vater zone below the normal water level in the vessel. The low-pressure side of the transmitter thur senses the static head of water / steam inside the vessel above the lower vessel level tap. This head varies as a function of reactor water level above the tap and is the " variable leg" in the differential pressure measured by the transmitter. Lower taps for various instruments are located at various levels in the vessel water zone to accommodate both narrou- and wide-range level measurements (see Figure 2).

Typical reactor level indicators and recorders.are shown on Figure

3. This figure also shows the condensing chamber. Shoreham level instrumentation, including elevations and set points, is shown in Figure 4.

Problem

Description:

Small (e.g., . 01 f 2t ) and intermediate (e.g., .04 ft2) break accidents (LOCA's) that discharge steam into the drywell (at temperatures as high as 3400F) for an extended time period result in substantial heat-up of components / air in the drywell (including reactor water level sensing lines). If the reactor is subsequently depressurized below 118 psia, water in the reactor water level sensing lines located in the drywell will flash. .

' General Electric has conservatively evaluated many steam break'acci-dents and has determined that, for the worst case scenario (small break accident with ADS operation after 1800 seconds), flashing will result ft .

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in a loss of up to 20% of the water in the sensing lines. Water in the variable leg sensing line will be replenished by drain back from the reactor, while water in the reference leg sensing line will continue to be gradually depleted due to boil-off. If no operator action is taken, all of this water could, for the worst case, boil off after more than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the accident. Loss of water from the reference leg results in a sensed reactor water level that is higher than the actual reactor water level. Shoreham reactor water level instrumentation utilizes two reference legs for the narrow and wide-range level instrumentaiton. Utilizing instrumentation keyed to the longer leg (worst case), a level error of approximately 9.1' could occur. It should be noted that all reactor water level activated safety trips will occur since they would initiate before the reactor is depressurized below 118 psia.

Operator Actions and Conditions that Prevent and/or Eliminate Flashing /

Soll-Off:

Plashing/ Boil-off will not occur if:

a) The break discharges two-phase fluid only; b) The drywell achieves the higher temperatures before level is recovered such that the saturated liquid spilling out of t:.e break and cooling the steam lines and drywell environment terminates the heatup transient; c) The operator initiates drywell spray before the reactor is depressurized below 118 psia; d) The reactor pressure is maintained above 118 psia.

In addition, even if flashing / boil-off were to occur, it would not be a concern if the operator follows the emergency procedure guidelines (EPG) and maintains reactor level in the normal water level range.

Furthermore, the error due to flash,ing/ boil-off will be eliminated if:

a) The operator follows the EPG and takes action to refill the reference leg after reactor depressurization if the temperature near the reference leg has exceeded the reactor saturation temperature end cor tinues reactor injection until the temper-ature near the reference leg is below 2120 F; or b) The operator determines that a flashing / boil-off condition exists and takes corrective action to refill the reference leg. Indications available to the operator that indicate reference leg flashing / boil-off are:

1) erratic level indication

2) mismatch between narrca, wide and upset range level indi-cators and recorders (Note: Since EPG requires the oper-ator to monitor water level from multiple indications, he should be aware of level instrument mismatch and hence flashing / boil-off conditions.)

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Conclusion:

'Considering the limited number of events, operator errors and conservative; analysis assumptions described above, the probabil-ity of reference leg flashing / boil-off resulting in core uncovery is considered _ extremely low. Even if one assumes that the worst

. case scenario described above occurs, the operator would receive a level.2 alarm (keyed to the shorter. reference leg) approximately 54 minutes prior to initial core uncovery. If this were disre-garded, he would receive another level 2 alarm (keyed to the longer reference leg) approximately 12 minutes-prior to initial core uncovery.

Based on the above,-it is concluded that the Shoreham reactor water level measurement instrumentation is acceptable.

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,SER Open Item No. 11 -' Supplemental ECCS Calculitions With NUREG-0630 Model The issue of fission gas release at high burn-up is. addressed generically.in the letters, R.E. Ingel of G.E. to P.A. Ippolito,

-NRC, " Extension of ECCS Performance Limits", dated May 6 and 28, 1981. .The' position stated in those letters is applicable to the Shoreham plant. '

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8/18/81

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SNPS-1 FSAR Item No. 36 - Containment Purge System The drywell and suppression chamber vent purge valves areisix inches in size and 150 pound class..The governing code for design;is ASME Section III Code Class 2. The valves have a design pressure of 48 psig and temperature of 90 F. The ambient temperature specified for the valves' environment is 40 - 120 F. The valve body form _is globe type with quick opening trim. The valve actuator is a spring and diaphragm with a 2-inch stroke. Air to the actuator is 125 psig.

Valve actuator operates on 80 psig. In addition to ASME Section Class 2 design, the valves have also been seismically qualified (in-cluding Mark II loads) by calculation. Additional seismic evaluation was performed which required analysis to assure the function and structural integrity. The' valves have been analyzed for stress and deflection due to combined seismig loads (simultaneously vertical and horizontal), pressure, and maximum operator load, and found that valve function is unimpaired. In addition, the associated solenoid valves are environmentally qualified in accordance with the require-ments of NUREG-0588.

These valves were also stroke tested immediately following testi g for leakage, (hydrostatic test) . The bench test was performed in a test fixture which did not account for an actual system downstream piping configuration nor could a constant differential pressure be maintained across the valve during stroke test.

In addition, the following ana.lytical documentation will oe provided to demonstrate the performanc and reliability characteristics of the vent system isolation valyca: -

1. Report on a Copes-Vulcan valve actuator defle: tion test.

2. Report on the seismic test performed at Acton Laboratory on the.D-100-160 actuator.

3. Report containing the Copes-Vulcan analysis' showing that the valve and actuator supplied for Shoreham (S&W Valve IT46*AOV079A&B) will function under the design conditions prescribed in the S&W specification.- An analysis showing that valve internal components remain functional under these same design conditions will also be provided.

SNRC-614 8/18/81

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-SNRC-614 August 18, 1981

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ATTACHMENT A Additional and/or revised inforruation is provided for the following:

1. NUREG-0737, Item II.K.3.28 - Study and Verify Qualification of of Accumulators on ADS Valves.

'2. SER Open Item No. 44 - Reactor Water Level Measurement Error.

3. SER Open Item No. 11 - Supplemental ECCS Calculations with NUPEG-0630 Model.

4. SER Open Item No. 36 - Containment Purge System e

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.. :e SNPS-1 FSAR i

.NUREG-0737, II.K.3.28 - STUDY AND VERIFY OUALIFICATION OF ACCUMULATORS ON ADS VALVES (AMENDED RESPONSE)

Item 1. As a minimum safety function requirement, 6 of the 7 ADS valves must cycle'one time. The design basis of the short term accumulators is to provide 2 ADS valve actuations at 70 percent of.. peak'drywell pressure, this value is equal to 33.6 psig. . A 25 psig air pressure differential is re-quired to actuate the ADS pilot valves, therefore the min-imum required pressure in the short term accumulators is 58.6 psig. This value translates to 5 ADS Valve actuations if the drywell-is at atmospheric pressure, assuming iso-thermal expansion.

The longest time period before which th~e ADS function is required after an accident is approximately 230 seconds, as represented by a 0.07 ft2 recirculation discharge-break with IIPCI failure. Details of this event are provided in FSAR Section 6.3.3.7 and on Figure 6.3.3-32.

Item 2. The leakage limits for each short term accumulator system are such that greater than 58.6 psig is maintained after 230 seconds based on our response to Item 1. In order to provido. margin in both time and proc 3ure, acceptance crit-cria for ADS short term accumulator leak-tighc integrity will be the retention of 70 psig minimum, after starting at narmal pressure of 90 psig, after a time period of 10 min-utes. This period provides nearly 20 percent margin in.

pressure in the time period which is much longer than-that required for ADS function.

Item 3. Periodic leak testing of each ADS short term accumulator system will be accomplished by temporarily connecting a

. pressure gauge to the accumulator, pressurizing the accum-ulator to 90 psig, venting the supply header, and checking accumulator pressure after 10 minutes. This test will verify that the acceptance criteria of Item 2 have been met.

Item 4. The leak test of Item 3 shall be performed periodically on a schedule consistent with the Shoreham integrated leak rate test to be performed during reactor shutdown or refueling, but in no case at intervals' greater than 3 years. The ina-bility of any short term accumulator system to pass.the accep-tance criteria of Item 2 will require repair / modification and rotest per Item 3. o Item 5. The seismic.and environmental qualification criteria for Shoreham are described in FSAR Sections 3.10 and 3.11. In addition, comprehensive description and status reports for 8/18/81 SNRC-614

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. , SNPS-1 FSAR NUREG-0737, II.K.3.28 Page 2 Item 5. both the seismic and environmental qualification program have been submitted by SNRC-575 dated May 28, 1981, and SNRC-576. dated Mcy 27, 1981 respectively. As described therein, Class IE electrical equipment is qualified in ac-cordance with NUREG-0588 Category 2.

The ADS accumulator system is qualified in accordance with the program outlined above and, therefore, meets the re-quirements of GDC 2 and 4.

Item 6. The applicant will perform a leak test prior to initial op-eration. Should the leakage exceed that established in Item 2, the system will be repaired / modified and retested as required.

8/18/81 SNRC-614

SNpS-1 FSAR SER OPEN ITEM #44 - LEVEL MEASUREMENT ERROR Review of Reactor Water Level Measurement Instrumentation:

The cold reference leg reactor water level measurement design for Shoreham is illustrated in Figure 1. Reactor vessel water level is measured by differential pressure transmitters which measure the difference in static head betwe'n e two columns of water. One column is a " cold" (ambient temperature) reference leg outside the reactor vessel; the other is the reactor water inside the reactor vessel.

The measured differential pressure is a function of reactor water level.

The cold reference leg is filled and maintained full of condensate by a condensing chamber at its top which continuously condenses reactor steam and drains excess condensate back to the reactor vessel through the upper level tap connection to the condensing chamber.

The upper vessel level tap connection is located in the steam zone above the normal water level inside the vessel. Thus, the reference leg presents a constant reference static head of water to the high pressure tap on the d/p transmitter. The low-pressure tap of the transmitter is piped to a lower-level tap on the reactor vessel which is located in the water zone below the normal water level in the vessel. The low-pressure side of the transmitter thus senses the static head of water / steam inside the vessel above the lower vessel level tap. This head varies as a function of reactor water 1cvel above the tap and is the " variable leg" in the differential pressure measured by the transmitter. Lower taps for various instruments are '

located at various levels in the vessel water zone to accommodate both narrow- and wide-range level measurements (see Figure 2).

Typical reactor level indicators and recorders are shown on Figure

3. This figure also shows the condensing chamber. Shoreham level instrumentation, including elevations and set points, is shown in Figure 4.

Problem

Description:

Small (e.g., . 01 f 2t ) and intermediate (e.g., . 04 -f t2 ) break accidents (LOCA's) that discharge steam into the drywell (at temperatures as high as 3400F) for an extended tir..e period result in substantial heat-up of components / air in the drywell (including reactor water level sensing lines). If the reactor is subsequently depressurized below 118 psia, water in the reactor water level sensing lines located in the drywell will flash.

General Electric has conservatively evaluated many steam break acci-dents and has determiner1 that, for the worst case scenario (small break accident with ADS operation after 1800 seconds), flashing will result

SNPS-1 FSAR in a loss of up to 20% of the water in the sensing lines. Water in the variable leg sensing line will be replenished by drain back from the reactor, while water in the reference leg sensing line will continue to be gradually depleted due to boil-off. If no operator action is taken, all of this water could, for the worst case, boil off after more than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the accident. Loss of water from the reference leg results in a sensed reactor water level that is higher than the actual reactor water level. Shoreham reactor water level instrumentation utilizes two reference legs for the narrow and wide-range level instrumentaiton. Utilizing instrumentation keyed to the longer leg (worst case), a level error of approximately 9.l' could occur. It should be noted that all reactor water level activated safety trips will occur since they would initiate before the reactor is depressurized below 118 psia.

Operator Actions and Conditions that Prevent and/or Eliminate Flashing /

Boil-Off:

Flashing / Boil-off will not occur if:

a) The break discharges two-phase fluid only; b) The drywell achieves the higher temperatures before level is recovered such that the saturated liquid spilling out of the break and cooling the steam lines and drywell environment terminates the heatup transient; c) The operator initiates dryvell spray before the reactor is depressurized below 118 psia; d) The reactor pressure is maintained above 118 psia.

In addition, even if flashing / boil-off were to occur, it would not be a concern if the operator follows the emergency procedure guidelines (EPG) and maintains reactor level in the normal water level range.

Furthermore, the error due to flashing / boil-off will be eliminated if:

a) The operator follows the EPG and takes action to refill tne reference leg after reactor depressurization if the temperature near the reference leg has exceeded the reactor saturation temperature and continues reactor injection until the temper-ature near the reference leg is below 212 0 F; or b) The operator determines that a flashing / boil-off condition exists and takes corrective action to refill the reference leg. Indications available to the operator that indicate reference leg flashing / boil-off are:

1) erratic level indication

2) mismatch between narrow, wide and upset range level indi-cators and recorders (Note: Since EPG requires the oper-ator to monitor water level from multiple indications, he should be aware of level instrument mismatch and hence flashing / boil-of f conditions. )

SNPS-1 FSAR'

Conclusion:

Considering the limited number of events, operator errors and conservative analysis assumptions described above, the probabil-ity of reference leg flashing / boil-off resulting in core uncovery is considered extremely low. Even if one asrumes that the worst case scenario described above occurs, the operator would receive a level 2 alarm (keyed to the shorter reference leg) approximately 54 minutes prior to initial core uncovery. If this were disre-garded, he would receive another level 2 alarm (keyed to the longer-reference leg) approximately 12 minutes-prior to initial core uncovery.

Based on the above, it is concluded that the Shoreham reactor water level measurement instrumentation is acceptable.

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SNPS-l~FSAR SER Open Item No. 11 - Supplemental ECCS Calculations'With NUREG-0630 Model The issue of fission gas 1elease at high burn-up is addressed Jenerically in the letters, R.E. Ingel of G.E. to P.A. Ippolito, NRC,." Extension of ECCS Performance Limits", dated May 6 and 28, 1981. The position stated ir. those letters i's applicable to the Shoreham plant.

4 9

s SNRC-614 8/18/81

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SNPS-1 FSAR 1 Item No. 36 - Containment Purge System i

i The drywell and suppression chamber vent purge valves are.six inches

.in size and 150 pound class..The governing code for design is ASME Section III Code Class 2. The valves have a design pressu're of 48 psig and temperature of 90 F. Tht; ambient temperature specified for the valves' environment is 40 - 120 F. The valve body form is globe type with quick opening trim. The valve actuator is a spring and diaphragm with a 2-inch stroke. Air to the actuator is 125 psis, Valve actuator operates on 80 psig. In addition to ASME Section Class 2 design, the valves have also been seismically qualified (in-cluding Mark II loads) by calculation. Additional seismic evaluation was performed which required analysis to assure the function and structural integrity. The' valves have been analyzed for strens and deflection due to combined seismig loads (simultaneously vertical and horizontal), pressure, and maximum operator load, and found that valve function is unimpaired. In addition, the associ ated solenoid valves are environmentally qualified in accordance with the require-ments of NUREG-0588.

These valves were also stroke tested immediately following testing for leakage, (hydrostatic test) . The bench test was performed in a test fixture which did not account for an actual system downstream l piping configuration nor could a constant differential pressure be maintained across the valve during stroke test.

l In addition, the following analytical documentation will be provided to demonstrate the performance and reliability characteristics of the vent system isolation valves:

1. Report on a Copes-Vulcan valve actuator deflection test.

2. Report on the seismic test performed at Acton Laboratory on the D-100-160 actuator.

3. Report containing the Copes-Vulcan analysis showing that the valve and actuator supplied for Shoreham (S&W Valve 1T46*AOV079A&B) will function under the design conditions prescribed in the S&W specification.- An analysis showing that valve internal components remain functional under these same design conditions will also be provided.

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