ML20038C048
| ML20038C048 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/05/1981 |
| From: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8112090476 | |
| Download: ML20038C048 (8) | |
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! L LU M ! N AilN G C O M P A N Y P.o. Box 5000 m CLEVELAND. OHIO 44101 e TELEPHONE (216) 622-9800 e ILLUMIN ATING BLDG.
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". Dalwyn iDavidson
, VICE PAESUENI SYSTE4 ENGINEERING AND CONSTRUCh0N Iiovenher 5,1981 Mr. Robert L. Tedesco Assistant Director for Licensing Division of Licensing U. S. tPac. Leas Regulatory Co:::tission Washington, L. C.
20555 Perry Ibaclear Power Plant Docket Iios. 50-440; 50-441 Response to Request for Additional Infor=ation -
Reactor Systems Branch
Dear Mr. Tedesco:
This letter and its attachment is subnitted to provide revised responses to the concerns identified in a October 19, 1981 phone conversation with Messrs. G. Thonas, Reactor Systems Branch re-viewer, and M. D. Houston, Project Manager.
It is our intention to incorporate these responses in a subsequent amendnent to our Final Safety Analysis Report.
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Dalwyn R. Eavidson
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440.3 Report on overpressure protection is required by the ASME code and is to serve as the basis for many of individual review steps during OL review.
(Ref. SRP 5.2.2)
Although the applicant provided some of the information re-quired by the overpressure protection report, the applicant has not provided the complete information. We require the applicant to submit the code specified report for staff review.
Response
The Perry Reactor Vessel Overpressure Protection Report, 22A6483 is based on the transient analysis method described in NED0-10802. Additionally, the NRC has requested another transient analysis using the newly approved ODYN. The attached report 22A6483, will be updated, if necessary, when the new analysis is completed; estimated before April 1982.
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440.6 Participation in GE safety relief valve surveillance program -
confirmation from the applicant is required on the applicant's participation in this surveillance program (NUREG-0512)
Response
Participation in a safety relief valve surveillance program is an issue currently being addressed by the Licensing Review Group, LRG II.
Perry as an LRG-ll plant is involved in the development and will participate in an SRV surveillance pro-gram when a mutually acceptable program is established.
REQUIREMENT 440.9 Reactor coolant system vents - TMI Item II.B.l.
Response is not acceptable.
Refer to BWR owner's group position and sub-mit plant specific response in detail as given by other BWR owners.
RESPCUSE The reactor coolant vent T.ine is located at the very top of ti reactor vessel as shown in the schematic (Figure 5.1-3).
This 2-inch line contains two safety-related Class lE moton operated valves (B21-FOO1 and B21*FOO2) that are operated from the control room. The location of this line permits it to vent the entire reactor coolant system normally connected to the reactor pressure vessel, with the exception of the reactor coolant isogation cooling (RCIC) head spray piping which comprises approximately 0.15 ft of volume above the elevation of the RPV. This small volume was considered in the original design of the RCIC system and is of no consequence to its operation.
In addition, since this vent line is part of the original design for the FNPP-units, it has already been considered in all the design-basis accident analysis contained elsewhere in the FSAR.
Each Perry Plant unit is provided with nineteen power-operated safety-grade relief valves which can be manually operated from the control room to vent the reactor pressure vessel. The point of connection to the vent lines (main streamlines) frcan near the top of the versel to these valves is such that accumulation of' gases above that point in the vessel will not affect natural accumulation of gases of the reactor core.
These power-operated relief valves satisfy the intent of the NRC position.
Information regarding the design, qualification, power source, etc., of these valves is provided in Subsection 5.2.2.
The BRR Owners' Group position is that the requirement of single-failure criteria for prevention of inadvertent actuation of these valves, and the requirement that power be removed during normal operati n, are not applicable to BWR's.
These valves serve an important function in mitigating the effects of transients, and provide ASME code overpressure protection. Therefore, the addition of a second " block" valve to the vent lines would result in a less safe design and a violation of the code. Moreover, the inadvertent opening of a relief valve in a BWR is a design-basis event and is a controllable i
l transient.
l
Pg. 2 of 5 In addition to these power-operated relief valves, the Perry Plant BWR/6's include various other means of high-point venting. Among these are:
a.
Normally closed reactor vessel head vent valves, operable from the control room, which discharge to the drywell; b.
Normally open reactor head vent line, which discharges to a main steamline; c.
Main steam-driven reactor core isolation Cooling (RCIC) system turbines, operable from the control room, which exhaust to the suppression pool; d.
Main steam-driven reactor feedwater pumps operable frca the control room, which exhaust to the plant condenser when not isolated. Condenser gases are continuously processed through the off-gas system.
Although the power-operated relief valves fully satisfy the intent of the venting requirement, these other means also provide protection against the accumulation of noncondensibles in the reactor pressure vessel.
Under most circumstances, no selection of vent path is necessary because the relief valves (as part of the automatic depressurization system), HPCS, and RCIC will function automatically in their designed modes to ensure adequate core cooling and provide continuous venting to the suppression pool.
Analyses of inventory-threatening events with very severe degradations of system performance have been conducted. These were submitted by GE for the BWR Owners' Group to the NRC Bulletins and Orders : Task Force on November 30, 1979 The fundamental conclusion of those studies was that if only one ECC system is injecting into the reactor, adequate core cooling would be provided and the production of large quantities of hydrogen was avoided. Therefore, it is not desirable to interfere with ECCS functions to prevent venting.
The small-break accident (SBA) guidelines emphasize the use of HPCS/RCIC as a first line of defense for inventory-threatening events which do not quickly depressurize the reactor. If these systems succeed in maintaining inventory, it is desirable to leave them in operation until the decision to proceed to cold shutdown is made. Thus the reactor will be vented via RCIC turbine steam being discharged to the suppression pool. Termination of this mode of venting could also terminate inventory makeup if the HPCS had failed also.
This would necessitate reactor depressurization via the SRV, which of course is another means of venting.
If the HPCS/RCIC are unable to maintain inventory, the SEA guidelines call for use of ADS or manual SRV actuation to depressurize the reactor so that the low-pressure core spray system can inject water. Thus, the reactor would be vented via the SRV to the suppression pool. Termination of this mode of venting is not recommended. It is preferable to remain unpressurized; however, if inventory makeup requires HPCS or RCIC restart, that can be accomplished manually by the operator. It is more desirable to establish and maintain core cooling than to evoid venting. IftheHPCS/RCICandsafety/
Pg. 3 of 5
- relief vntves are not operable (a very degraded and extremely unlikely case),
another emergency means of venting the reactor must be used. It is empha-sized, however, that such emergency venting would be in the. interes', of core cooling and therefore could be employed under Emergency Procedure Guidelines.
It is thus conclubd that there is no reason to interfere with ECCS ' operation to avoid venting. It is further. concluded that the Emergency Procedure
' Guidelines, by correctly specifying operator actions for HPCS, RCIC,~and SRV operation, also correctly specify operator actions to vent the reactor.
In the event of HPCS failure and continued ' vessel pressurization, the effect of noncondensibles in the RCIC turbine steam was evaluated for three cases:
1.
Continuous evolution of noncondensibles due to radiolysis; 2.
Quasi-continuous evolution of noncondensibles due to core heatup; 3
The presence of a quantity of noncondensibles in the reactor at-the time of HPCS/RCIC startup.
Case 1 is a normal operating mode for RCIC and is of no concern.
- For Case 2 to exist, the core must be uncovered. Such a condition requires multiple failures as shown in the degraded cooling. analyses. Core uncovery
- is prevented (or cladding heatup into the rapid oxidation range is prevented) when only one ECC system is operating. For a small pipe break or a loss of feedwater, which would allow.the reactor to remain at pressure, the HPCS and/or RCIC pumps would maintain inventory and there would be no substantial hydrogen production. If neither HPCS nor RCIC could maintain inventory, the reactor would be automatically or manually depressurized via safety / relief valves (or via the break, for larger breaks). The low pressure water injection system (LPCS) would then make up inventory. With the core covered nei.ther the rapid generation of noncondensibles nor their accumulation would be possible.
1 The performance of RCIC under Case 3 is of concern only if there has been a very substantial production of hydrogen due to core uncovery and there is a need to start the RCIC. This is extremely unlikely and an intolerable cir-cumstance, because it could arise only if the core were allowed to remain uncovered for a long period with the reactor at high pressure. Automatic j.
depressurization system operation and explicit operating instructions and the emergency operator guidelines are intended to preclude this. If the level has fallen with the reactor at high pressure, the vessel would be 3
depressur1 zed either automatically or manually to permit low-pressure hjection independent of RCIC performance.
i In the post-LOCA condition, it is possible to have noncondensible gases come out of solution while operating the residual heat removal (RHR) system.
These gases would accumulate at the top of the RER heat exchanger since this is a system high point' and an area of relatively low flow. Gases trapped here i
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Pg. 4 of 5 will be tented through a one-inch vent line with two safety-related Class lE motor-operated valves (E12-Fo73 and E12-F074) operated from the control room (as shown in Figure 5.4-13). As this vent line and associated valves are part of the original design, they have also been considered in the design-basis accident analysis contained elsewhere in the FSAR. To accommodate the continuous release of noncondensibles from the RHR Heat Exchanger when employed in the steam-condensing mode of long-term cooling, these remote vera valves on the heat exchsnger vent line are opened to discharges through a submerged line inte the suppression pool inside the drywell.
The result of a break in the SRV discharge piping, or any of the other pipe-lines for the systems enumerated above, would be the same as a small steam-line break. A complete steamline break is part of the Perry design basis, and smaller-size breaks have been shown to be of lesser severity. A number of reactor system blowdowns due to stuck-open relief valves (also equivalent to a small steamline break) have confirmed this in practice. Thus no new anal-yses are required to show conformance with 10 CFR 50.46.
Because the relief valves and RCIC will vent the reactor continuously, and because containment hydrogen calculations in normal safety analysis calcu-lations assume continuous venting, no special analyses are required to demonstrate "that the direct venting of noncondensible gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentration limits in containment."
Conclusion and Comparison with Requirements The conclusions fran this vent evaluation for FUPP are as follows:
a.
Reactor vessel head vent valves exist to relieve head pressure (at shutdown) to the drywell via remote operator action.
b.
The reactor vessel head can be vented during operating conditions via the SRV's to the suppression pool.
c.
The RCIC system provides an additional vent pathway to the suppression pool.
d.
The size of the vents is not a critical issue because BWR SRV's have substantial capacity, exceeding the iall power steaming rate of the nuclear coiler.
e.
The SRV's vert to the containment suppression pool, where discharged steam is condensed without causing a rapid con-tainmentpressure/temperaturetransient.
f.
The SRV's are not smaller than the NBC defined small LOCA.
Inadvertent actuation is a design-banis, event and a demon-strated controllable transient.
1
Pg. 5 of 5 g.
Inadvertent actuation is of course undesirable, but since the SRV's serve an important protective function, no steps such as removal of power during normr21 operation should be taken to prevent inadvertent actuation.
h.
Ar indication of SRV position is provided in the control room. Temperature sensors in the discharge lines confirm possible valve leaktse. This indication is being upgraded in accordance with NUREG-0588.
,j. Each SRV is seismically and Class lE qualified.
k.
Block valves are not required, so block valve qualifications are not applicable.
1.
No new 10 CFR 50.46 conformance calculations are required, because the vent provisions are part of the systems in the plant's original design and are covered by the original design bases.
m.
Plant procedures govern the operator's use of the relief mode for venting reactor pressure. These procedures are available for Regional NRC inspection at the PNPP plant.
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