ML19318C967
| ML19318C967 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/02/1980 |
| From: | Bartnoff S JERSEY CENTRAL POWER & LIGHT CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8007070113 | |
| Download: ML19318C967 (6) | |
Text
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Jersey Central Power & Light Company k= M( -
3 Mad son Avenue at Punch Bow! Road Momstcwn, New Jersey 07960 (201)455-8200 July 2, 1980 Director of Nuclear Reactor Regulation Nucicar Regulatory Commission Washington, D. C.
20555
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Core Spray System Our letter dated June 27, 1980 to you submitted an evaluation of two (2) possible cracks discovered in the Core Spray piping within the reactor vessel between the inlet no::le and the vessel shroud.
On June 30, 1980 there was a meeting between members of my staff and your staff to discuss Jersey Central Power 6 Light Company's evaluation.
The purpose of this letter is to document our response to questions raised at the June 30 meeting and to supplement our June 27 evaluation.
The attachments to this letter provide the response to the questions raised at the June 30, 1980 meeting.
In addition to the information provided in the attachments, the effect of cracks in the core spray piping located within reactor vessel between the inlet nozzle and the vessel shroud on the triple low (1o-10-1o) water level indication has been assessed and shows that the effects, if any, will be in the conservative direction.
A quantitative evaluation on Triple Low water level indication has been initiated and will be forwarded to you when it has been completed.
This report is expected to be completed by August 1,1980. Also, the technical report :q, porting Attachment I will be forwarded at that time.
During the next operating cycle, Jersey Central Power 6 Light Company will continue to work on a development of better inspection techniques for this' location in the vessel.
We have been active in the advancement of the
" state of the art" of-in-vessel inspection techniques and will continue to pursue improvements in-this area in order to provide more definitive information on these types of indications in the future if that is at all possible. We are currently developing tooling and procedures as well as alternate designs to permit replacement of the core spray spargers during 800707o y Jersey Centra! Power & Light Company is a Member of tr.e General PutAc Utaties Systern
, our 1981 refueling outage.
Some of the alternate designs involve changes to the core spray piping between the no::lc and the core shroud; the direct replacement option has not' required such changes. This sparger replace-ment project will be expanded to include the tooling and design work and procedures necessary to replace this in-vessel core spray piping at the next refueling outage.
Inspection methods and techniques available at the time'and proven to be effective under these circumstances will be utilized to reinspect this piping and permit a sound decision on the need for replacement.
4 Since the evaluation of all the cracks and possible cracks found on the Core Spray System within the reactor vessel and the repairs already performed on the core spray spargers shows that there is no significant change in safety margin from that of the original design, Jersey Central Power G Light Company requests approval for the return to operation of the Oyster Creek plant.
Very truly yours, Wtr
' Shepard Bartnoff
/
President Ir Enclosures (3) b t
ATTACifMENT 1 OYSTER CREEK SPRAY LINE BREAK SENSITIVITY STUDY Exxon Nuclear Company (ENC) was requested by General Public Utilities (GPU) to perform additional calculations for the Oyster Creek reactor in connection with postulated Core Spray System Failures. The Oyster Creek core spray system had been previously modified such that both core spray systems must be operable even under NRC single failure criteria.
Thus, at least one operable core spray system is assured for all postulated LOCA-ECCS breaks, except one,even assuming the additional failure of that portion of the Core Spray System within the Reactor Vessel.
The LOCA-ECCS break not appropriately treated by the latest ENC ECCS analysis, considering the additional failure of the core spray sparger or piping within the vessel, is the assumed break of the other core spray system line outside the reactor vessel. For this situation, the independent spray system would be lost, and the core spr3y coolant inventory would be delivered to the vessel either through the sprays or through the postulated crack, however, the appropriate spray distribution could not be assured.
A core spray line break analysis was performed by ENC for the latest Oyster Creek ECCS analysis and reported in XN-NF-77-55 Rev. 1.
The ENC WREM-Based Non-Jet Pump BWR ECCS Small Break Evaluation Model was used for this calculation.
This evaluation model computes core flow during the blowdown calculation (until time of rated spray) with the core spray inventory being added to the vessel but no spray cooling assumed.
Thus, the blowdown results remain valid even for the cracked sparger or piping situation. These results show that the high power core regions (both the hot channel and the average core) remained covered by the mixture level for the ' initial 500 seconds of the blowdown transient. The spray system was calculated to reach rated flow at 527 sec. After rated spray flow is calculated, the ENC model uses the NRC 10 CFR 50 Appendix K spray heat transfer coefficients and continues the calculation with the HUXY program.
A Peak Cladding Temperature (PCT) of 1335'F was calculated for the spray line break.
For the case with a cracked sparger or piping, the spray cooling cannot be assumed after the spray is calculated to reach rated flow.
However, since the spray line break is above the reactor core, the spray ECCS inventory will accumulate in the vessel, and terminate the transient by reflooding.
Since the blowdown results until the time of rated spray are valid, che water mass inventory in the system is known at the time of rated spray.
By extrapolation of the pressure vs. time curve, the break flow, ADS flow, and spray flows can be conservatively estimated.
A conservative net water accumulation rate can then be calculated and integrated until the water inventory reaches the core midplane.
Applica-tion of the appropriate NRC 10 CFR 50 Appendix K rSflood heat transfer 4
1
. coefficient of 25 BTU /HR-ft2
- F at this time will terminate the temperature _ transient. This calculation was performed for the Oyster Creek reactor spray line break. The core midplane was calculated to recover at 357 seconds after core rated spray or 884 seconds after the break. A HUXY calculation was performed assuming the core spray line break blowdown heat transfer and adiabatic conditions for the fuel assembly after rated core spray. The maximum cladding temperature from the HUXY results at the time of core midplane recovery are below 2100'F and the metal-water reactic.: is well below the 10 CFR 50.46 limits.
Thus, based on this conservative calculation, the Oyster Creek reactor can still meet 10 CFR 50.46 and Appendix K requirements for a core spray line break so long as the spray coolant inventory is added to the vessel.
9
ATTAONENT 2 PROBABLE INDICATION MECHANISM
'As discussed in mur first submittal on this matter, the cause of these indications cant at be positively identified. Even though we evaluated these indica : ions as cracks to insure that safety objectives could be met, there are several points which suggest that these. indications are merely surface marks from installation or manufacture and that they have existed for the life of the plant:
1.
These indications do not have the characteristics, shape, texture, and depth that are associated with cracks inspected under these conditions.
2.
None of the inspectors positively identified these indications as cracks.
In fact, one inspector was convinced they were not cracks but rather some sort of tool or installation marking. The others, in the absence of a better explanation, could not preclude that the indications were cracks and therefore identified them as "possible cracks".
3.
Extensive evaluation has not resulted in the identification of a mechanism which could cause a crack of this nature.
4.
While this piping is subjected to a variety of loadings during normal and emergency operation, the stresses are insufficient to initiate a crack or propagate a flaw into a crack.
.a ATTACINENT 3 SAFETY EVALUATION The Oyster Creek Emergency Core Cooling System is made up of two core spray systems cach of which is single-active failure proof. Each core spray system is provided with redundant emergency power sources, valves, pumps, etc. so that no active single failure can prevent it from distributing, by itself, sufficient core spray flow to fully justify use of the spray cooling heat transfer coefficients assumed in LOCA analyses. Therefore only one of the two core spray systems is necessary to mect the requirements of 10 CFR 50.46 with the exception of a core spray line break in which case the other core spray system is required to assure adequate core cooling.
Bae LOCA analysis of a core spray line break is provided in XN-NJ-77-55 Rev. 1.
The results show that the core remained covered by the mixture level for the initial 500 seconds of the blowdown transient.
It also shows that credit for spray cooling was assumed when the intact core spray system reached rated flow at 527 seconds, resulting in a peak cladding temperature (PCT) of 1335'F for this break. However, assuming that the intact core spray system fails to provide a satisfactory spray distribution over the core because of passive piping failures inside the reactor vessel, a conservative calculation has been performed which does not take credit for spray cooling, but accounts for water inventory addition for reflood.
Under these assumptions the maximum cladding temperature at the time of core midplane recovery is below 2100*F and the' metal-water reaction is well below the 10 CFR 50.46 limits.
Therefore the Oyster Creek Plant still meets 10CFR50.46 and Appendix K requirements even if one of the two core spray systems is degraded such that it is only able to provide water inventory addition to the reactor vessel, but not an adequate spray distribution over the core.
In view of the above discussion, it can be concluded that the two surface indications found on the in-vessel piping of Core Spray System II are acceptable even if these indications proved to be cracks.
Adequate core cooling will be available under all postulated LOCA conditions.
Therefore there is reasonable assurance that the health and safety of the general public will not be jeopardized by the operation of the Oyster Creek plant for the next fuel cycle.