ML20055D723
| ML20055D723 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/27/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20055D718 | List: |
| References | |
| NUDOCS 9007090332 | |
| Download: ML20055D723 (6) | |
Text
pu:q
',1(
-k UNITED STATa8 g'
.~
L NUC.LE AR REGULATORY COMMISSION WASHINGTON, D. C. 20666
.'..../
1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING-AMENDMENT NO.125 TO FACILITY OPERATING LICENSE NO. DPR-29 COMMONWEALTH EDISON COMPANY A.E 10WA-1LLIN015 GAS AND ELECTRIC COMPANY QUAD CITIES NUCLEAR POWER STATION. UNIT 1 DOCKET NO. 50-254
1.0 INTRODUCTION
By letter _ dated May 19, 1990, CommonwealthEdisonCompany(thelicensee) submitted a request for-a temporary waiver of compitance and emergency Technical Specification change for Quad Cities Nuclear Power Station, Unit 1.
The proposed change to the Technical Specification excludes eight containment pathways from the requirement to perform Type C local leak rate tests in accordance with Appendix J to 10 CFR Part 50 until the next refueling-outage scheduled for October.1990. This proposed change was verbally requested on May 18, 1990 and verbally granted by the staff later on that same day. By letter dated May 22, 1990, the staff provided written approval of the temporary waiver of compliance.
2.0 DISCUS $10N AND EVALUATION in addition to not being Type C tested, the valves in the eight containment pathways were not properly included as part of the tested containment boundary during the last containment integrated leak rate test (Type A test). Thus, neither the sum of local leak rates (Type B and Type C) nor the overall integrated containment leak rate (Type A) is known. Because the valves cannot be tested with the plant at power, nor can the lines be isolated by other barriers with the plant at power, the requested change to the Technical Specifications was needed to allow continued plant operation.-
Even though the subject valves are untested, there are other valves, piping systems, anti other components which could act u barriers to containment leakage through the eight containment pathways under consideration. Although these components and systems may not be safety-grade or may only satisfy some but not all of the staff's standards for safety-grade equipment, they nevertheless provide some assurance that, during the limited time
)eriod until the next refueling outage, potential leakage through the eigit pathways would be limited or eliminated during an accident. These considerations,
.which are described below, provide sufficient assurance such that the staff's engineering judgement is that the plant may o>erate until the next refueling outage without undue risk to the public healti and safety.
9007090332 900627 DR ADOCK OSCC 4
1 The eight untested pathways are associated with five systems: clean demineralized water, core spray, standby liquid control, instrument air, and reactor building closed cooling water.
Each is discussed below.
(1) Clean Demineralized Water. Penetration X-20 This pathway is a single three-inch line that penetrates the containment.
The containment isolation valves are a check valve and a locked-closed manual
' valve in series outside of containment, in addition to these two containment isolation valves, there are two normally closed manual valves in series inside containment. ThereisalsoaclosedpipIngsystemoutsidecontainment.
The entire system is pressurized with water at about 100 psil during unit i
operation.
This. water could serve to seal any potential lescage through the l
valves during an accident and it also continuousiy demonstrates the integrity of the piping system. The system is supplied by multiple pumps feeding a i
common header taking suction from a 100,000 gallon storage tank. With these l
barriers in place, the water pressure would have to fail and containment atmosphere would have to leak past all four valves and then escape the closed system outside containment in order to reach the environment.
(2) Core Spray Systeni. Penetration X-16 A & B The Core Spray System is a low pressure emergency core cooling system which provides reactor coolant in the event of a loss of coolant accident (LOCA).
The system is designed to be pressurized with high pressure water (relative to Pa, peak containment accident pressure) which acts as a water seal for the containment isolation valves during post-accident conditions. The 10-inch injection lines which pass through the subject penetrations are each equipped with a check valve inside containment in addition to two containment isolation valves (remotely-operated gate valves) in series outside containment. This check valve is subject to reactor pressure during normal operation which gives an indication of its leak-tightness. There 4; also a closed piping system outside containment which would either cont,in containment leakage or route it to the suppression pool.
As with the clean demineralized water system, core spray injection would have to stop and water pressure be lost, containment atmosphere would have to leak past three valves, and then it would have to escape the closed system outside containment to reach the environ-ment.
(3) Standby Liquid Control (SBLC) System. Penetration X-47 The 11-inch SBLC line which penetrates containment contains a closed valve outside containment in series with two containment isolation check valves (one inside containment and one outside). The closed valve is an explosive
,.--,r..--
3 i
squib valve which consists of a solid metal cap which blocks the pathway unless actuated. The potential of a seat or packing leak, therefore, does not exist. The SBLC system is an Engineered Safety Feature and the squib valves are only actuated in the event that the control rod scram function fails and reactor power cannot be reduced using normal methods. The squib charge is then exploded, shearing off the metal cap and opening the valve.
The valve,-therefore, would not be actuated during a design basis LOCA.
i There is also a closed system outside containment. Thus, containment atmosphere would have to leak past two valves, through a solid piece of metal, and out of a closed system in order to reach the environment.
(4) Instrument Air to the Drywell and Torus: Penetration X-216 & X 22 The instrument air system penetrates containment by two lines. The line which penetrates the drywell is a one inch line and that which penetrates the torus is one-half inch.
Containment isolation is achieved in each line by one check valve inside containment and one check valve outside containment.
The penetrating lines are connected inside of containment to a closed piping system that does not interface with the drywell atmosphere. Outside of contain-ment, the lines are connected to a closed piping system that does not interface with the Reactor Building atmosphere.
During normal operation, the lines are pressurized with nitrogen at a pressure of approximately two times Pa.
i This pressurization may serve as a valve sealing system in the event of a leak.
DuringthelastIntegratedLeakRateTest(ILRT),theselineswereproperly depressurized and vented outside of containment. The closed piping inside of containment,lation valves were not adequately challenged.
however, was not vented to the containment; therefore, the containment iso The ILRT was successfully completed which provides assurance that either the inside piping system or the containment isolation valves were not leaking excessively.
With these provisions in place during a LOCA, the lines would have to depres-surire and containment atmosphere leak past two valves and penetrate two closed systems to reach the environment.
(5) Reactor Building Closed Cooling System (RBCCW):
penetration X-23 &
X-24 The RBCCW system consists of two eight-inch lines that penetrate containment.
The supply line containment isolation valves are a check valve inside and a remotely-operated manual gate valve outside containment. The return line containment isolation valves are two remotely operated valves, one inside and one outside of containment.
In addition to the two containment isolation valves on each line other barriers exist.
Insideofthecontainment,thepipingformsaclosed loop. Outside of containment, the piping is configured such that loop water seals are created. The system is filled with pressurized water during i
normal operation. The water serves as a seal for potentially leaky valves and as a system leakage detection system. Any through-wall water leaks would be detected either inside or outside of the containment through operational indicators (sumplevels,systempressures,tanklevels,etc.).
l
^
t
'e '
l
~4 The piping outside of containment is connected to a vented surge tank. This tank receives makeup that is supplied by multiple pumps connected to a common header, which provides suction from a 100,000 gallon storage tank. This configuration provides substantial assurance that the system would remain water filled in post accident conditions. Containment leakage, then, would have to enter the closed system inside containment, pass through two valves i'
and a loop set), and the system would have to be depressurized, before leakage could enter the environment.
In addition to the above, the licensee has performed a probabilistic risk assessment (PRA) to calculate the probability of the occurtence, in the next five months, of a loss-of-coolant accident in conjunction with certain additional failures (such as pipe ruptures outside containment) that the licensee states would allow containment leakage to be released to the environment through any of the eight containment pyhways. The licensee has calculated this prob-ability to be less than 1 x 10 The licensee believes that this is sufficiently low so that the plant can be operated for the next five months without signifi-cant increase in risk to the public health and safety during that period.
The staff has evaluated the licensee's submittal and concludes that operation until the next refueling outage with the subject valves untested will not add significantly to the risk to public health and safety, and therefore, the proposed Technical Specification change is acceptable.
3.0 FINDINGS OF EMERGENCY WARRANTING AN AMENDMENT WITHOUT NOTICE in the fourth quarter of 1989, the licensee assessed the leak rate testing program at Quad Cities and identified 29 containment pathways that were not being local leak rate tested.
The licensee determined the Appendix J, Type C testing was not required for these 29 pathways.
However, tie NRC staff questioned this determination and on May 18, 1990, concluded that these 29 pathways were, in fact, subject to Appendix J Type C testing. Once informed of the NRC's position on these pathways, the licensee requested a temporary waiver of compliance. The NRC staff orally granted this waiver on May 18, 1990.
By May 19, 1990, the licensee had tested those pathways that could be tested with the plant at power (21 pathways) and submitted a written request for a temporary waiver of compliance and an emergency technical specification amendment for the remaining eight pathways. The licensee requested that the testing of these eight pathways be deferred until the next refueling outage.
s
-~,
. =
.~-,,,.._.e
p i
j j
- i The NRC staff finds that the licensee acted as quickly as possible once informed of the requirement to test these pathways.
Furthermore, the staff i
finds that failure to grant the proposed changes in a timely manner would have required a shutdown of Quad Cities, Unit 1.
Accordingly, the staff concludes that the licensee has satisfied the requirements of 10 CFR 50.91(a)(5), and that a valid emergency exists.
4.0 FINAL. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The NRC staff reviewed the licensee's amendment application and determined, in accordance with the criteria of 50.92(c), that operation of Quad Cities.
Unit 1, according to the proposed amendment will not:
(1)
Involve a significant increase in the probability or consequences of
-an accident previously evaluated.
With respect to cn increase in the probability of previously evaluated accidents,tiating aspects of the events. leakage through the associated valv change ini With regard to the consequences of an accident previously evaluated, the continued operation in the existing configuration does not present a significant increase in the probability of a larger release of radioactivity than described in the FSAR.
For those credible events, the normal as-built designs function to inhibit potential release paths either as a sealed water-filled system or normally manually isolated system.
In addition, it has been calculated that for less credible events, the probability of increased risk over an additional five months of continued operation is not significant (less than 1.0E-07). The potential for an adverse change in consequences is not supported for accidents of reasonable probabilities but would require accident :;cenarios of very low probabilities. Thus it is concluded that continued operation willnotposeasignIficantincreaseinriskwithregardstoaccident l
l probabilities or consequences.
L (2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed amendment does not result in any physical plant changes during the period of interest.
Potential leakage of the valves in question would, at worst, affect the severity and not the type of accident.
(3)
Involve a significant reduction in the margin of safety.
1
?
p j
j a
, (
As described in the Technical Spacific6 tion Bases, dose calculations suggest that the accident les,: rate could be allowed to increase to about 2.6f. a day before the guideline thyroid dose value given in 10 CFR 100 would be exceeded.
However, 1.0f a day provides an adequate snargin of safety to assure the health and safety of the general public. Additional n,argin is achieved by establishing the i
allowable operational leak rate at 0.75 of the maximuni allowable leak rate.
As described in the attached docunentation and despite the lack of leak testing, substantial barriers to fission product
}
release are provided by the intact system piping and associated valves. These barriers provide mitigating capability such that the potential inipact on the margin of safety is insignificant. The Probabilistic Risk Assessment performed demonstrates that the probability of containnent function failure coincident with LOCA conditions is also acceptably small.
Accordir. gly, the Conciission has determined that this request does not involve
[
a significant hazards consideration.
5.0 STATE CONSULTATION
The State of Illinois was inforned by telephone June 7,1990, of the staff's final no significant hazards consideration deterrination and intent
,to issue a license amendment. The State contact had no concent.
6.0 ENVIRONMENTAL CONSIDERATION
i This amendnent involves a change to a requirenent with respect to the instal-16 tion or use of a facility component located within the restricted area as 6 fined in 10 CFR Part 20.
The staff has determined that the anendnent involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite and that there is no significantincreaseinindividualorcunulativeoccupationalradiationexposure.
The Connission has previously issued a proposed finding that this amendnent involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendo.6nt cetts the eligibility criteriaforcategoricalexclusionsetforthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environnental impact statenent or environinental assess.
nent need be prepared in connection with the issuance of this amendment.
7.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there h reescr.able assurance that the health and safety of the will not be end6ngered by operation in the proposed nienner, and (2) public such activities will be conducted in compliance with the Cormiission's regulations, and (3) the issuance of this anendnent will not be inimical to the concion defense and security or to the health and safety of the public.
Principal Contributors:
J. Pulsipher, L. 01shan Dated: June 27,1990
,