ML19262A871
| ML19262A871 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/06/1979 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 LAC-6680, NUDOCS 7912110230 | |
| Download: ML19262A871 (5) | |
Text
e DatIItWSND POWEIt COOPEItatTIVE Ea Crone, Olkonan 54601 December 6, 1979 In reply, please refer to LAC-6680 DOCKET NO. 50-409 U.
S. Nuclear Regulatory Commission ATTN:
Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation Washington, D.
C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 THREE-MILE ISLAND LESSONS LEARNED -
SHORT TERM REQUIREMENTS
References:
(1)
NRC Letter, Denton to All Operating Nuclear Power Plants, dated October 30, 1979.
(2)
NRC Letter, Eisenhut to All Operating Nuclear Power Plants, dated September 13, 1979.
(3)
DPC Letter, Linder to Eisenhut, LAC-6616, dated November 5, 1979.
Gentleren:
In response to your letter (Reference 1) which requested descriptions and justifications for differences from the staff's "short tern" requirements resulting from the NRC staff investigations of the TMI accident as presented in NUREG 0578, your letter (Reference 2), we are presenting the ollowing information in support of our earlier response (Reference 5).
Item 2.1. 2_ - Perfon'ance Testing for BWR and PWR Relief and Safety Valves There are three Crosby Type HCC-6S, 3" x 6" safety valves installed at LACBWR designed to meet the reactor primary system overpressurization requirements of the ASME Nuclear and Boilar and Pressure Vessel Codes.
The valves are direct spring-loaded popping type valves and are either fully seated or when actuated, fully open.
The valves are installed on the horizontal run of the shutdown condenser steam line and are exposed only to saturated steam conditions.
A steam trap is installed in the horiziontal run of the steam line to the shutdown condenser in the proximity of the safety valves to ensure the removal of condensed
~
steam during reactor operation.
Each valve exhausts through individual short run (approximately 6 feet) of 6"-8" diameter pipe directly into 1531 1E5 791211025/7-
U.
S.
Nuclear Regulatory Commission LAC-6680 ATTN:
Mr. Harold R.
Denton, Director December 6, 1979 Office of Nuclear Reactor Regulation the containment building.
Therefore, downstream restriction in this short run of pipe is minimal with either steam or two-r: lase flow conditions.
The relieving capacity with one operational valve of 294,612 lbm/hr is sufficient to limit the system pressure to less than 110% of the vessel design pressure during an Abnormal Operational Transient wit' the highest pressure.
During a normal most limiting pressure transient, the reactor pressure does not reach the lowest safety valve set ~oint (1390 psig).
Therefore, safety valve operation does not occur during Normal Operational Transients.
The safety valves are bench-tested at a frequency that is more con-servative than the testing frequency recctmended in the ASME Code Section XI.
An average of at least one valve has been tested each refueling outage and in some cases, all valves have been functionally tested during an outage.
Set-pressure relief tests are routinely performed with nitrogen at the plant site (testing requires removal of the valves from the primary system during cold shutdown). Live steam verification tests have also been conducted at the manufacturer's facility and inde-pendently at a testing laboratory.
When the velves are relief tested at the plant site, the set-pressures are adjusted to correct for the small difference (less than 1.7%)
between the nitrogen test conditions and steam conditions.
The correction cerm was established by independently performing live steam tests at simulated reactor operating temperature and pressure conditions.
During steam testing, the valves were instrumented to monitor temperature popping pressure and reset pressure.
Approximately 35 tests representing an estimated 338 pops have been performed over a 10-year period of steam and nitrogen testing.
Experience has shown that though the valves may weep clightly after 1._v i ng been in service and subsequently bench-lifted, there is no evidence that the valves have failed to seat as designed.
It has been demonstrated from numerous relief tests that the LACBWR safety valves will perform their design function in the remote event of a postulated overpressurization transient.
Item 2.1.3.a - Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs AND BWRs Indication of LACBWR's safety valve positions is already provided in the Control Room. 1531 ia6
U.
S. Nuclear Regulatory Commission LAC-6680 ATTN:
Mr. Harold R.
Denton, Director December 6, 1979 Office of Nuclear Reactor Regulation Two alarms for three valves indicate when a safety valve is open by alarming when the thermoccaple at the discharge of the valve reads 250 F.
When the valve is closed, the thermocouple reads approximately containment ambient temperature.
When it opens, the steam passing through it raises the temperature at the discharge thermocouple, thus causing a " Low Set Reactor Safety Valve Open" or The "High Set Reactor Safety Valve Open" alarm in the Control Rocm.
alarm setpoint of 250 F. represents a level of discrimination between a small break LOCA atmosphere in the Contaimment Building and the flow of primary steam from the open safety valve.
Item 2.1.3.b - Instrumentation for Detection of Inadeauate Core Cooling Procedures to address the use of instrumentation other than water level will be prepared by 1/1/80.
Item 2.1.4 - Containment Isolation The LACBWR automatic isolation of non-essential systems has already been addressed in answer to IE Bulletin 79-08, " Events Relevant to BWR's Identified During Three Mile Island Incident."
Please reference LAC-6262 (Linder to Keppler, 5/2/79), LAC-6319 (Linder to Keppler, 5/29/79), LAC-6474 (Linder to Ziemann, 8/16/79), and
" Supplemental Information Related to IEB 79-08" (Shimshak to Shea, 9/25/79).
The ventilation system facility change discussed in LAC-6474 is planned to be instituted during the spring 1980 refueling outage.
Control Penetrations Items 2.1.5.a - Dedicated H2
& 2.1.5.c - Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant The safety analyris and full term license application for the La Crosse Boiling Water Reactor (Letter LAC-2788, Madgett to Giambusso, dated 10/9/74) identified on Page 338 that no requirement exists for hydrogen recombination at this facility.
The use of stainless steel clad results in very low hydrogen accumulation.
Item 2.1.8.b.1 - Interim Procedures for Quantifying High Level Accidental Radioactivity Releases The La Crosse Boiling Water Reactor is currently equipped with 4
noble gas monitoring equipment capable of 5 x 10 pCi/cc (Xe-133).
The lower power density of the La Crosse facility combined with the
~
much smaller core size and fuel inventory indicate that our existing 153)
\\57 U.
S.
Nuclear Regulatory Commission LAC-6680 ATTN:
Mr. Harold R.
Denton, Director December 6, 1979 Office of Nuclear Reactor Regulation equipment is more than adequate to yield the same measurement capability at 5 x 10" pCi/cc (Xe-133) as larger facilities achieve with a 1 x 10 5 pCi/cc (Xe-133).
We, therefore, intend to utilize existing equipment.
Item 2.1.8.b.3 - Hiah Range Containment Radiation Monitor High range monitors will be installed.
Item 2.1.9 - Transient and Accident Analysis Containment Pressure Monitor The redundant LACBWR Containment Building pressure monitors arc capable of indicating in the Control Room two times the maximum pressure resulting from the design basis accident as calculated for the SAR.
This capability is adequate.
Both monitors are environ-mentally qualified to function in a design banis event.
As the LACBWR is equipped with redundant vacuum breakers on the Containrent Building, capability of reading a vacuum is not necessary in this installation.
The environmental qualifications of these pronqure monitors (37-35-301 and 37-35-302) were reported te the Nuclear Regulatory Commission in LAC-5520, dated October 26, 1978 (Letter, Madgett to Ziemann).
Containment Water Level Monitors The LACBWR's redundant containment water level trancmitters (37-42-310 and 37-42-302) cover 41 feet of the Containment Building height, including tha full height of the reactor vessel.
The LACBWR does not have a suppression pool as referenced in the staff's position; however, LACBWR's instrumentation more than meets the staff's require-ment on Containment Building water level indication.
The environmental qualifications of these water level indicators were reported to the NRC in LAC-4420, dated 10/26/ 78, letter, Madgett to Ziemann.
Containment Hydrogen Indication The SAR and the full term license application for the LACBWR (Letter LAC-27b2 Madgett to Giambusso, dated 10/9/74) identified on Page 338 that no potantial for significant hydrogen generation exists at this facility.
The use of stainless steel clad fuel results in a very Icw hydrogen accumulation in the event of an incident.
Therefore, contin-uous indicaticn of hydrogen concentration in the containment atmosphere is not necesst.ry at LACBWR.
.teactor Coolant System Venting The LACBWR's Manual Depressurization System already provides redundant adequate venting of the primary coolant systems.
The reactor emergency flooding vent valves are manually operated from the Control Room.}} }30 4-
o U. S. Nuclear Regulatory Commission LAC-6680 ATTN. !!r. Harold R. Denton, Director December 6, 1979 Office of Nuclear Reactor Regulation Sufficient venting is also provided for the shutdown condenser. Indication of the position (open/close) of these vent valves is proviCed in the Control Room. The ef2ilities of these venting systemi have been demonstrated in past operation of the f acility. Item 2.2.1.b - Shift Technical Advisor We intead to utilize individuals who are familiar with the facility who are currently employed on the plant staff. These individuals will include either - 1) Degreed engineers 2) Former licensed Senior Operators 3) Current licensed Senior Operators not in direct supervision of Shift Supervision with college level engineering courses. 4) Other technical supervisory personnel capable of obtaining a Senior Operator License. 5) Degreed engineers of technical consultants. Item 2.2.2.b - On-Site Technical Center An on-site Technical Support Center will be installed in the new Administration Building at LACBWR. The habitability requirements will be reviewed consistent with existing structural capability. If you have any questions regarding this matter, please contact us. Very truly yours, DAIRYLAND POWER COOPERATIVE l g, s ~ Agf, .x Frank Linder, General Menager FL:JDP:af cc: J. Keppler, Reg. Dir., NRC-DRO III f}} k3 3 - }}