ML20107F613
Text
/.
_.__..__,___.7 UNITED SrATES
, REGULATORY COfAPAISSION CLEAR W AsHINGToN, D. C. 20555 a
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CHANGES TO TECHNICAL SPECIFICATIONS OF' PROVISIONAL OPERATING LICENSE DPR-16 JERSEY CENTRAL PONER 6 LIGirr C0"PANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
...e Introduction Dy 1 citer dated February 15, 1975 to Jersey Central Power 6 Light Company, t'he Nuclear Regulatory Commission (NRC) requested that the licensee anong other things, develop operating procedures and proposed changes to the Technical Specifications to preclude reaching cicvated 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 letter dated April 1, J975 Jersey Central submitted a response which stated that the present Technical Specifications provide adequate limits for the spppression chauber water tenperature, thus the licensee proposed no change to the Technical Specifications.
For the reasons set forth in this evaluation, this response from the licensee was found unacceptabic.
Appropriate chan;',cs'to the Technical Specif.ications are needed to assure the proper operation and integrity of the pressure suppression primary containment system.
Discussion 1
Oyster Creek is a boiling water reactor (BNR) 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 the pressure during a postulated loss-of-coolant accident (LOCA) by condensing the steam released from the reactor primary system.
The i
reactor system energy released by relief valve operation during~ operating transients also is released into the pool of water in the torus.
Experiences at various Bh'R plants with Mark I Containments have shown that damage to the torus structure can occur from two phenomena associated with relief valve operations.
Damage can result from the forces exerted go ibe structure when, on first opening the relief valves, stco, and the
[
e nu
'f j
%a.#
)
DEKOK95-258 PDR
h
^
4 air within the vent arc discharged into the torus water. This phenomenon is referred to as steam vent c1 caring. The second source of' potential structural damage stems from the vibrations which accompany extended relief valve discharge into the torus water if the pool water is at elevated temperaturcs. This effect is known as the steam quenching vibration phenomenon.
A.
Steam' Vent Clearing Phenomenon 111th regard to the steam vent c1 caring phenomenon, we are actively reviewing this generic probica and in our Ictter dated February 15, 1975-we also requested the licensee to provide information to demonstrate that the torus structure of the primary containment will maintain its integrity throughout the anticipated life of the facility.
In its response dated April 1, 1975 the licensee stated that it was investigating this matter and the results of the investigation would be submitted to l
us on a schedule consistent with the timing which we proposed for licensec respanso.
Because of the apparent slow progression of the mat erial fatigue associated with the steam vent c1 caring phenomenon, we have concluded that there is no immediate potential hazard resulting from this type of phenomenon; neverthc1 css, surveillance and review action on this matter by the NRC staff will continue in due course during this year.
1 B.
Steam Quenching Vibration Phenomenon The steam quenching vibration phenomenon became a concern as a result of occurrences at two European reactors.
111th torus pool water temperatures increased 'in excess of 170F due to prolonged steam quenching from relief valve operation, hydrodynamic fluid vibrations occurred j
with subsequent uoderate to high relief valve flow rates.
These fluid vibrations produced large dynamic loads on the torus structure and extensive damage to torus internal structures.
If allowed to continue, the dynamic loads could have resulted in structural damage to.the torus itself, due to material fatigue.
Thus, the reported occurrences of the steam quenching vibration phenomenon at the two European reactors indicate that actual or incipient failure of the torus can occur from j
such an event.
Such failure would be expected to involve cracking of the torus wall and loss of containment integrity.
Morcover, if a LOCA occurred simultaneously with or after such an event, the consequences could be excessive radiological doses to the public.
Ir. comparison with the steam vent cicaring phenomenon, the potential risk associated with the steam quenching vibration phenomenon (1) reficcts the fact that a generally smaller safety margird/ cxists between the present license requirements on suppression pool temperature limits and the point at which damage could begin and (2) is more immediate.
1[ The difference, in pool water temperature, between the licenso limit (s) and the temperature at which' structural damage might occur is the safety margin availabic to protect against the effcets of the phenomenon discussed.
~
4
^=
^ -
I I
~* ' *
([l
.'3-
)
Evaluation The exi.hting Technical Specifications for Oyster Creek limit the torus pool temperaturo.to 100F.
This temperature limit has been reduced to 95F to provide SF temperature difference between a scram requirement discussed below and provisions for performing necessary survcllance.
The temperature of 95F assures that the pool water has the capability.to perform as a constantly avail-abic heat-sink with a reasonabic operating temperature that can.be maintained by use of heat exchangers whose sc.condary cooling water (the service cooling water) is expected to remain belou 95F.
While this 95F limits provides normal 1
operating ficxibility, short-term temperatures permitted by operating procedures cxceed the normal power. operating temperature limit, but accommodates the heat release resulting from abnormal operation, such as relief valve malfunction, while still maintaining the required heat-sink (absorption) capacity of the pool water needed for the postulated LOCA conditions.
Ilowever, in vicw of the potential risk associated with the steam quenching vibration phenomenon, it is necessary to modify the temperature limits now in the license Technical Specifications.
This action was, as discussed in our February 15, 1975 letter, first suggested by -the General Electric Company (GO who had earlier informed us of the steam quenching vibration occurrences at a meeting on Nover..ber 1, 1974 and provided related information by letters to us dated November 7, j
and December 20, 1974.
The December 20 letter stated th,t GE had informed all of its customers with operating Bim facilitics and Mark I containments
)
of the phenomenon and included in those communications GlPs recommended interim operating temperature limits and proposed operating procedures to mininize the probability of encountering the damaging regine of the steam quenching vibration phenomenon.
Our inplementation of the GE recommended procedures and temperature linits via changes in the Technical Sp,ccifications are evaluated in the following paragraphs:
a.
The new short-term limit applicable to all conditions requires that the reactor be scrammed if the tor'us pool water temperature reaches 110F.
This requirement to scram at 110F provides additional assurance that the torus temperature will remain below the 170F temperature related to potential dmnage to that torus.
i b.
For specific requirements associated with surveillance testing, 1.
c., testing of relief valves, the water ter.perature shall not exceed 10F above the normal power operation limit.
This new limit during surveillance testing of relief valves provides additional operating flexibility uhile still maintaining a maximum heat-sink' capacity.
The current limits in the Technical Specifications make no provision for these requirements.
e 4
O
~
...q p,
.... h For reactor isolation conditions, the now temperature' limit.is
.c.
, *~
120F,'above-which temperature the reactor vessel is to be depressurized.
This new lir.iit of 120F assures.poo.1 capacity for absorption of heat released to the torus while avoiding undesirabic reactor vessel cooldown transients.
Upon reaching'120F, the reactor is-placed in the cold, shutdown condition at the fastest rate consistent with'the technical specifications on reactor pressure vessel cooldown rates.
d.
In addition to the new limits on.temperaturn of the torus poo'l-water, the discussion in the-liasis includes a. summary of required 1
operator actions to be taken in the event of a relief valve malfunction.
- These operating actions are taken in order to avoid the development of tenperatures approaching the 170F threshold for potential damago by the steam quen,ching phenomenon.
Conclusion t'c have evaluated the GE recommendations'. consisting of new suppression pool temperature limits and operating procedures.
h'c conclude that thesc procedures and temperature limits discussed above are appropriate and are needed to assure that the containment function as designed in order to protect the public'hcalth and safety.
Dated:
g i o W5 e
6
' n 2
4
.,