ML20205M695
| ML20205M695 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/27/1988 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TASK-2.D.1, TASK-TM TAC-44621, TAC-51421, NUDOCS 8811030229 | |
| Download: ML20205M695 (6) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _. _ _ _ _ _ _ _ _ _ _ _
t TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 7 7401 SN 1578 Lookout Place r
00T 271988 i
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 i
Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - ADDITIONAL INFORMATION FOR THI ACTION ITEM L
t II.D.I. PERFORMANCE TESTING OF REACTOR RELIEF AND SAFETY VALVES FOR SQN UNITS 1 AND 2
References:
1.
TVA letter from J. A. Domer to E. Adensam, NRC, dated July 10, 1985 2.
TVA letter to NRC dated June 24, 1988, "Sequoyah Nuclear Plant (SQN) - Performance Testing of Reactor Relief And i
Safety Valves" j
3.
NRC letter from Suzanne Black to S. A. White dated I"
August 18, 1988, "RAI for THI Action Item, II.D.1, Performance Testing of Reactor Relief and Safety Valves I
for Sequoyah, Units 1 and 2 (TAC 44621, 51421)"
i t
Enclosed is TVA's response to reference 3, which requested additional l
Information regardirg TMI Action Item II.D.1, "Performsnce Testing of i
Reactor Relief and Safety Valves for Sequoyah, Units 1 and 2."
This information is provided to enable NRC to complete review of the subject
[
item.
I Please direct questions concerning this issue to Don V. Goodin at
[
(615) 870-7734.
[
Very truly yours, l
TENNESSEE VALL Y AUTHORITY
/
I R. Giidley, H ger Nuclear Licensing and
[
Regulatory Affairs
[
l Enclosure cc:
See page 2 g
j OV i
8811030229 8s1027 ADOCK 050 g 7 DR f
An Equal Oppcrtunity Empicyer
---,- ~ ~ -
U.S. NucIvar Regulatory Commission 0Grf 271988 cc (Enclosure):
Hs. S. C. Black, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Hr. F. R. McCoy Assistant Director for Inspection Programs TVA Proj2 cts Division U.S. Nuclear Regulatory Commission Region II 101 Harletta Street, NH, Suite 2900 Atlar.ta Georgia 30323 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 1
Enclosure Addltional Information For THI Action Item II.D.1, Performance Testing of R1 actor Relief and
, Safety Valves for SQN Units 1 and 2 NRC Question No. I "Provide the thermal hydraulic and stress analyses of simultaneous S/RV actuation with no prior actuation of the power-operated relief valves (PORVs) for the case of steam discharge (Reference 1 implies that such analyses exist).
Indicate what the administrative procedures are for draining the loop seal."
TVA Response Several different scenarios were considered during the development of the forcing functions for the PORV relief and safety valve events.
Forcing functions were developed wtth and without N RV actuation in an effort to determine the vorst case.
The thermal hydraulle analysts of steam discharge for simultaneous pressurl:er safety valve actuation without prior actuation of the PORVs (and no water-filled loop seal) is case 6 of the analysis of record, TVA calculation TI-ANL-96, revision 2.
This analysis is available within TVA's Records Information Managemt:nt System (RIMS).
Loads were generateci and stress analyses performed for the case where the PORVs discharge before the relief and safety valve's ',1fting -this is case 5.
A comparison of the results for these two cases indicates very little difference in the magnitude of loads resulting from the discharge of the safety and relief valves. A similar analysts of these two conditions, performed by an independent contractor, also indicated that prior PORV opening does not significantly affect the loads resulting from the safety and relief valve discharge.
The unit I pressurizer safety and relief valve piping and supports were designed for a much more severe load case of a heated water loop seal discharge in comparison to steam-only loads.
The pipe support modifications required for this heated loop seal condition are installed. Coincident with the determination from the plant operating staff to operate the plant with the steam-trimmed safety valves, an additional analysts was performed for the piping and supports.
This analysis was based on case 5 results, which assume a PORV discharge occurring before the safety and relief valve discharge.
As previously stated, the results of the steam discharge cases with and without PORV at.tuation are comparable.
Further, the steam discharge case is much lower than the previously quallfled heated loop seal condition.
The piping supports for the unit 2 pressurl:er safety and relief valve piping were designed for the extrerne case of a cold water loop seal discharge.
The pipe support nodifications for this extreme cold loop seal condition have been installed, and the piping satisfied the Final Safety Analysis Report (FSAR) stress limits for the safety and relief valve discharge condition.
Der a decision was made to operate the plant with a drained loop and steam trimmed safety and relief valves, an earlier code of record analysts was reinstated.
This earlier analysis was performed using the forcing functions from case 5.
The earlier analysis was reviewed, and all open items were reconciled.
The earlier support loads were cccpared with the latest support loads for which the supports were quallfled.
The support margins were substantial when us',nr the steam discharge case Ivads.
In summary, the calculations of record do not include stress analyses for forcing functions from case 6.
However, prior actuation of the PORVs does not significantly affect'the loads resulting from the safety and relief valve discharge.
Further, the piping and supports for both units have been previously quallfled for higher transient loads than are generated by the steam discharge case.
The piping and supports are overdesigned for the current steam trim configuration. Additional analysis or reconciliation of other slightly different, but c'f comparable magnitude, steam discharge loids is not warranted.
Administrative proceou.:- require all valves in the drain lines on the inlet loops of the pressurizer safety valves to be loc %ed open.
System operating instruction 68.1 for the reactor coolant system requires double verification of the valves' position and locking before entering mode 4 operation.
Locking open of the loop seal drain valves allows any condensate in the pressurizer safety valve inlet piping to drain continuously back to the pressurizer through a tap located near the bottom of the pressurizer.
NRC Question No. 2 "In the TVA response to Question 3 of Raference 2, it is stated that 'In the unlikely event of a feedline break and the unsuccessful mitigation action by the operator, the safety valve is still capable of passing water as described above.'
If TVA cannot take credit for operator action, TVA should provide an analysis that shows the design can withstand the loads developed for this scenario."
TVA Response The FSAR analysis was based on the conservative assumption that no mitigating actions were taken and indicates that the pressurizer does not go solid for almost 10 minutes. Our submittal to NRC dated July 10, 1985, discussed a more recent analysis contained in Hestinghouse Electric Corporation calculation l
Hr.AP 10105, which indicates that the saturated liquid discharge is not expected for at least 20 minutes.
This 20-minute time period is sufficient for operatoi actions to mitigate the feedwater line break, which would arrest the pressure rise and prevent a continued surge of water into the pressurtzer even though credit for such actions was not taken in the FSAR analyses.
The operator actions include isolation of the faulted steam generator so that the automatic aualliary feedwater actuation would deliver water to the intact steam generators and isolation of safety injection to mitigate the overpressure event.
These actions are described in Emergency Instructions E-0, "Reactor Trip or Safety Injection"; E-1, "Loss of Reactor or Secondary Coolant"; E-2, "Faulted Steam Generator Isolation"; and ES-0.2, "S!
Termination."
The ability of the pressurizer safety valves to operate satisfactorily with water discharge was demonstrated in the Electric Power Research Institute Safety / Relief Valve Test Program.
The discharge piping has not been
I rigorously analyzed for the transition from steam to saturated liquid discharge.
The analyses used to qualify the piping on units 1 and 2 were steam discharge cases; and, as discussed in the answer to question 1, previous analyses have demonstrated that the present piping and hanger configurations are satisfactory for design conditions more severe than the current steam l
discharge case. Because the analyses that have been performed bound this l
case, specific analyses have not been performed to determine the loads because of the steam / liquid transition.
NRC Question No. 3 "Provide the thermal hydraulic and stress analysis of simultaneous PORV actuation in case of steam and cold water discharge.
The latter case is required to address the low temperature overpressure transient.
(Has the loop been removed in f.*ont of the PORV?)"
TVA Response The pressurizer PORVs and block valves are configured on a horizontal run of piping elevated above the pressurizer. As such, there is no location amenable for the collection water to form or sustain a loop seal. As such, blowdowns j
through the PORVs are limited to steam blowdowns, except in the unlikely event of a transition to 11guld rettef.
The fluid transient forces resulting from simultaneous actuation of the PORVs are provided in case 8 of the analysis of record. Case 8 is similar to, but bounded by, case 5, which considered PORV actuation before safety and re'lef valve discharge.
The resulting loads from case 5 were considered in the piping analysis.
The cold water discharge condition has not been specifically analyzed.
The possibility of this scenarlo arises from the use of the PORVs in the cold l
overpressure mitigation system (COMS). During reactor coolant system heatup, I
the PORVs serve as overpressure protection at low system temperature and pressure. SQN operating procedures call for a steam bubble to be drawn at very low temperature and pressure; therefore, the operation of the COMS when the pressurtzer is water solid would only permit suticooled,11guld discharge at low pressure.
The flow rates arising from COMS actuation are not of sufficient quantity to result in significant fluids transient loads. As a result, the loads because of such a discharge are bounded by the loads caused by the high-pressure steam discharge. Also, previous analyses and qualifications discussed in the response to question I show that both units have considerable design margin beyond the high-pressure, steam discharge only scenario.
NRC Question No. 4 "Describe the administrative controls that TVA has approved to ntnimize leakage through the S/RVs."
~
~4-i i
TVA Response f
I Technical specification (TS) surveillance requirement (SR) 4.4.6.2.1.d requires performance of a reactor coolant system water inventory balance at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Surveillance Instruction (SI) 137.2, "Reactor Coolant System Water Inventory," is the procedure used to comply with this TS t
SR.
If the Ilmit on the identified leakage as set forth in the plant TSs is exceeded, the plant personnel are required to reduce the leakage rate to less than the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or have the plant in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> l
and in COLD SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l SI-112. "Testing and Setting of Setpoint of Pressurizer Safety Valves," covers I
testing of the pressurizer safety valves to satisfy SRs 4.4.2 (applicable for modes 4 and 5), 4.4.3.1 (applicable for modes I, 2, 3), and 4.0.5.
This i
Instruction also fulfills the requirements for in-service performance testing I
)
of nuclear power plant pressure relief devices in accordance with American National Standards Institute /American Society of Hechanical Engineers i
i
)
OH-1-1981.
The instruction includes leak rate testing of the valves after I
a completion of the setpoint testing.
The acceptance criteria require the ItaK l
i rate to be less than or equal to 15 bubbles nitrogen per minute measured in l
l ac urdance with the vendor's test procedure. The valves may be refurbished to j
bring the valve performance to within the required limits. The SI is required
[
j to be performed on at least one of the installed pressurizer safety valves in
)
any 24-month period.
i Temperature indicators and acoustic monitors aie provided on the pressurizer l
)
safety valve discharge piping to aid the operators in deterrining which, if i
))
any, pressurizer safety valve is leaking.
These instruments annunciate in the
(
control room. However, no operator action is required unless there is an i
indication of excessive loss of reactor coolant system inventory.
Instrument Indications can be used to determine if detected leakage is from a pressurizer safety valve and which one.
Indications can also be used to select pressurizer safety valves for planned maintenance.
The instruments are vertfled to be operational periodically according to sis.
'l t
i i
l 3
i 1
l J
i I
i i
i i
l
- ^-
_