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l UNITED STATES NUCLEAR REGULATORY COMMISSION l
W ASHINCTON. D. C. 20$5 5 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR RECUTATTON SUPPORTING PROPOSED AMENDMENT TO. PROVISIONAL OPERATING LICENSE DPR-16 AND PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS SUPPRESSION POOL NATER TEMPERATURE LIMITS JERSEY CENTRAL POWER 6 LIGHT COMPANY OYSTER CREEK kUCLEAR GENERATING STATION DOCKET NO. 50-219 Introduction By let'er dated February 15, 1975 to Jersey Central Power 6 Light Company, t
the Nuclear Regulatory Commission (NRC) requested that the licensee among other things, develop operating procedures and proposed changes to the Technical Specifications to preclude reaching elevated temperatures of the torus pool water and to provide for inspection of the torus as
. appropriate to identify any damage in the event of an extended relief valve operation.
By Ictter dated April 1,1975 Jersey Central submitte'd a response which stated that the present. Technical Specifications provide adequate limits for the suppression chamber water temperature, thus the licensee proposed.no change to the Technical Specifications.
j This response from the licensee was found unacceptable; and, as a result, the NRC staff prepared appropriate technical specification changes to revise the suppression pool water temperature limits for Oyster Creek Nuclear Cencrating Station.
By letter dated
- July 16, 1975, the NRC staff advised the licensee of its intent to initiate steps to issue these technical specification changes unless the licensee objected in writing.
By Ictter dated August 8, 1975, the licensee provided comments on the technical specification changes proposed by the NRC staff.
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Subsequent discussions between the NRC staff and the licensee resulted in a few mutually acceptabic changes to the Technical Specifications proposed by the NRC staff.
The licensee stated it would accept the proposed Technical Specifications with these changes.
Discussion Oyster Creek is a boiling water reactor (BWR) which is housed in a Mark I primary containment. The Mark I primary containment is a pressure suppression type of primary containment that consists of a drywell and a suppression chamber (also referred to as the torus).
The suppression chamber, or torus, contains a pool of water and is designed to suppress
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the pressure dufing a, postulated loss-of-coolant accident (LOCA) by condensing the steam released from the'rcactor primary system. The reactor _ system energy released by relief valve operation during operatitig transients also is released into the pool of water in the torus.
Experiences at various MlR plants with hhrk I containments have shown that damage to the torus structure can occur from tuo phenomena associated with relief valve operations.
Damage can result from the forces excrted
.on the structure' then, on first opening the relief, valves, steam and i
the air within the vent arc discharged into the torus water.
This-phenomenon is referred to as steam vent clearing., The sei:end source of potential structura) d.tmage stens from the vibrations rhich acccmpany extended relief valve discharge into the torus water if the pool water is at clevated temperatures. This offect is known as the steam quenching vibration phenomenon.
A.
Steam Vent Clearine Phenodenen h'ith regard to the steam vent c1 caring phenomenon, ue are actively reviewing this generic problem and in our letter dated February 15, 1975 we also requested the licensee to provide inforantion to demonstrate that the torus structure of the primary containcent will maintnin its integrity throughout tha anticipated life of the faci lity.
In it:i respon:sc dated Ap,ril 1,197. the licerisee stated-that it was invcstinating this matter and the results of the investigation would he subaitted to us on a schedule consistent with the timing which we proposed for licensee repsouse.
Ih rmise of the apparent sicw progression of the hateria) fatigue assrciated with the steam vent clearjno, phenomenon, we have concluded thtt there is no inm.ediate potential hazard resulting from this type of phenonenon; nevertheless, surveillance and revicw action en this matter by the NRC staff will continue in duo course during this year.
B.
Steam Quenching Vibration Phenomenon s
The steam quenching vibration phenomenon became a concern as a -
result of occurrences at two European reactors.
h'ith torus pool water temperatures increased in excess of 170F due to prolonged steam quenching from relief valve operation,-hydrodyncaic fluid -
vibrations occu~rred with subsequent moderate to high relief valve j
flow rates.
These fluid vibrations produced largo dynamic loads on the torus structure and extensive' damage to torus internal structures.
If allowed to continue, the dynamic loads could have resulted in l
structural damage to the toriis itself due to material fatigue.
l Thus, the reported occurrences of the steam quenching vibration i
phenomenon at the two European reactors; indicate that actual or incipient failure of the torus can occur from such an event.
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3-Such failure would he expected to involire cracking of the torus wall and loss of containment integrity.
Morcover, if a LOCA occurred simultaneously with or after such an cycnt, the consequences could be excessive radiological doses to the public.
In comparison with the steam vent clearing phenomenon, the potential risk associated with the steam ' quenching vibration phenomenon (1) reficcts the fact that a generally smaller safety marginl/ cxists between the present license requirements on supprossion pool temperature limits and the point at which damage could begin and (2) is more immediate.
Evaluation The existing Technical Specifications for Oyster Cicch limit the tolus pool temperature to 200F.
This temperature limit has been reduced to 95F to provide SF temperature difference between a ser: m requircent discussed below and provisions for performing necessaly surveillance.
The temperature of 95F assurcs that the pool water has th; capability to perforn as a constantly availchle heat-sinh with a reasonable operating temperature that can be maintained by use of heat exchangers whose secondary cooling water (the service cooling uater) is expected to remain belou 95F.
While this 95F limit prov. ides norm 1 operat J r. :
1 ficxibility, short-tem teoperatures permitted by operating procedures exceed the normal power operatin;, ta perature limit, but acco:mtedates the heat release resulting fro::. bnorual operation, r.uch as relief relve malfunctior., while still riaintaining the required hear-sJnk (absorpt ion) capacity of the pool water needed for the ps tulated frCA conditiom.
llowever, in view of the potentici risk associnted witb the steat quenchihb vibration phenomenon, it is necessary to modify thu ten.perature J
limits now in the license Technical Specifientions.
This actioc. r -
as discussed in 7ur Februaly 15, 1973 lettcr, first suggested by General Ulcetric Company (GE) who had earlier inforr.:ed us of the stean quenching vibration occurrences at a meeting on Noverber 1, 1974 and l
provided related information by letters to us dated November 7, and December 20, 1974. The Deccaber 20 letter stated that GE had in orned c
all of its customers with operating DER facilities and Mark I co'...ain-ments of the phenomenon and included in those communications GE's recommended interim operating temperature linits and proposed cperating procedures to minimize the probability of-cucountering the damaging regime of the steam quenching vibration phenomenon.
Our impicmentation of the GE recommended procedures and' temperature limits via changes in the Technical Specifications are evaluated in the following paragraphs:
- 1/ The difference, in pool water temperature, between.the license limit (s)'
and the temperaturo at which structural damage might occur is the safety-
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margia availabic to protect against the effects of the phenomenon discussed.
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a.
The new short-term limit applicable to all conditions requires that the reactor be scrammed if the torus pool water temperaturo reaches 110F. This requirement to scram at 110F provides additional assurance that-the torus temperature will remain below the 170F temperature related to potential damage to that torus, b.
For specific requirements associated with surveillance testing, i.e., testing of, relief valves, the water temperature shall not exceed 10F above the normal power operation limit.
This new limit
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during surveillance testing of. relief valves provides additional operating flexibility while still maintaining a maximum heat-sink capacity.
The current limits in the Technical Specifications make no provision for these requirements.
c.
For reactor isolation conditions, the new temperature limit is 120F, above which temperature the reactor vessel is to be de-pressurized.
This new limit of 120F assures pool capacity for absbrption of heat released to the torus while avoiding undesirable reactor vessel cooldown transients. Upon reaching 120" the reactor is placed in the cold, shutdown condition at the fastest rate consistent with the technical specifications on reactor pressure vessel cocidown rates.
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Conclusion We have concluded, based on the consideration discussed above that:
(1) there is reasonable assurance that the health and. safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
OCT e 3 73 w
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