ML20010E204
| ML20010E204 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/27/1981 |
| From: | Phyllis Clark JERSEY CENTRAL POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20010E202 | List: |
| References | |
| NUDOCS 8109030198 | |
| Download: ML20010E204 (18) | |
Text
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1 JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 1
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Technical Specification Change Request No. 98 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No. 98 to the Oyster Creek Nuclear Generating Station Technical Specifications, a change to Specifications Sections 2.2. and 2.3.
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) 'l c- t-l Philip R. Clark Vice President - Nuclear i Jersey Central Power & Light Co.
l Executive Vice President -
GPU Nuclear Sworn and subscribed to before me this y? day of( , < , 1981.
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PHYLLI3 A. l'A3!J NOTARY PULLIC C7 NEW llR5:Y fay Comm:ssion Empires Aug. 16,1984 8109030198 810827 PDR ADOCK 05000219 p PDR
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF )
) DOCKET NUMBER 50-219 '
JERSEY CENTRAL POWER 6 LIGHT COMPANY )
CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 98 for the Oyster Creek Nuclear Generating Station Technical
.7,pecif icat ions , filed with the U. S. Nuclear Regulctory Commission on August 27,1981, has this 27 day of August, 1981 , heen served on the Mayor of Lacey Township, Ocean County, New Jersey by der,osit in the United States mail addressed as follows:
The Honorable Jorge A. Rod Mayor of Lacey Township 818 West Lacey Road Forked River, N. J. 08731
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1 / i BY [ - ' ! L' M C i -
Philip R. Clark Vice President - Nuclear Jersey Central Power & Light Co.
Executive Vice President -
GPU Nuclear DATED: August 27, 1981 1
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Jersey Central Power & Light Company L.f ,{. J ,b .g - a / f""'f
.l [q Mad. son Avenue at Punch Boal Road LA ~~~u Mornstown Nea Jersey 07960 (201)455 8200 August.27, 1981 The lionorable Jorge A. Rod Mayor of Lacey Township 818 West Lacey Road Forked River, N. J. 08731
Dear Mayor Rod:
Enclosed herewith is one copy of Technical Specification Change Request No. 98 for the Oyster Creek Nuclear Generating -
Station Technical Specifications.
These documents were filed with the U. S. Nuclear Regulatory Commission on August 27, 1981.
Very truly yours, pg l, jij' L LO ij qQ /,!c [, .
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Philip R.. Clark Vice President - Nuclear Jersey Central Power & Light Co.
Executive Vice President -
GPU Nuclear i
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{ Enclosures Jersey Central Power & LK;ht Company is a Member of the General PutAc Uti!.es System
Jersey Central Power & Light Company Oyster Creek Nuclear Generating S+ation (D,ocket No. 50-219)
Provisional Operating License DPR-16
- 1. Section to be changed:
- a. Section 2.2 bases
- b. Section 2.3
- c. Section 2.3 bases
- 2. Extent of changes:
- a. Section 2.2 remove reference and results of a superceded analvsis and eliminate the 25 psi margin to safety valves set point require-ment for the curbine trip without bypass transient.
- b. Section 2.3 Specification for reactor high pressure relief valve Initiation is changed from less than or equal to 1070 psig to 2 valves less than or equal to 1070 psig and 3 valves less than or equal to 1090 psig.
- c. Section 2.3 eliminated reference to peak pressure for turbine trip with failure of bypass to 1188 psig (25 psig margin to safety ,
valves).
- 3. Change requested:
The requested change is on the attached revised Tecnnical Specification pages 2.2-1, 2.3-2 and 2.3-5.
- 4. Discussion:
See attached report "Satety Evaluation f c.- the increase in the Oyster Creek Relief Valve Set ro;nts", dated May 14, 1980.
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SAFETY EVALUATIO:: FOR Tile INCREASE 1" TIIE OYSTER CREEK RELIEF VAI,VE SET Pol.NTS PREPARED BY R. R. FURIA C ** b APPROVED: R. _ B . I.ee Superviser Oyster Creek Fuels APPROVED: f f'
/ G. R. Bond Manager Nuclear Fuels
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APPROVED: y- c . 6
,R. W. Keaten Director Systems Engineering May 14, 1980
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INTRODUCTION The current Oyster Creek Technical Specifications require relief valves set 3
points to lift at pressure less than or equal to 1085 psia. This value would provide a 50 psi separation between the set points and the normal operating pressure of 1035 psia.
l A previous study, which was sent to the NRC in reference 1, was conducted to evaluate the structural response of the suppression chamber (torus) to the vent clearing effects during ERV discharge transients. One of the recommenda-tions f ro n this study, which was im;11 enen ted to nit igate the initial lift of the electrom.agnetic relief valves (ERV's) into a cold pipe, was the lowering of the set point of one ERV in each of the two discharge lines to 1065 psia.
This limited the nur.ber of initial lifts into a conmon discharge header to one, clearing the non-condensible gases from the pipe at a slower rate which reduces stresses. The remaining ERV's in each header would discharge into a pipe in which steam is flowing and would cause no abnormal stresses. Based upon the maximum pressurization rate of 100 psi per second which occurs during the Turbine Trip withaut hypann transient, a 20 psi margin was required between i
the two ERV set points to insure a 0.2 second time diffe ence in ERV lifts to cicar noncondensiM e:, froa. the discharge pipe during blowdown. The 20 psi re-duction in the set points on two valves resulted in narrowing the margin to normal operating pressures from 50 to 30 psi.
A requirement from TMI-2 lessons learned (reference 2) was to investigate ways to reduce the number of challenges to relief valves. One of the methods sug-gested to accomplish this goal was increasing set points. GPU Nuclear perforced an analysis to determine is the Oyster Creek ERV set points could be raised o provide margin similar to that which existed prior to lowering the set points on 2 valves. In addition, the 20 psi separation was maintained to reduce torus stresses.
An amendment to the Oyster Creek technical specification is required to change
- the set points fron 5 relief valves at less than or equal to 1085 psia to two L valves at less than or equal to 1985 psia and three valves at less than or equal to 1105 psia. Pressurization transients which result in relief valve actuation were reanalyzed with the higher set points and evaluated against de-sign criteria specified in the technical specification. The results of these analyses support such a change to the technical specifications. '
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ANAIXSIS OF PRESSURIZATION TRANSIENTS Pressurization transients were analyzed for a full core of ESC SNS l fuel usin",the RC Plant Transient Analysis Code PTSRWR2 (referenc<
3). The transients were initiated from a full power level of 1930 tid . End-of-cycle kinetics parameters and a bounding scram curve were used in the analysis. Conservative multipliers of 1.25 on the void reactivity w ificient and 0.9 on the doppler reactivity were used, consist <ut with previous analyses (References 6, 7 and 8).
The criteria used for accepting the results of a pressurination transient are 1) a peak pressure of less than 1227 psia (minimum safety valve (SV) set pointi for normal operational transients,and 110% of reactor coolant pressure boendary design pressure for the most severe abnorr.al operational trassient with reactor scram (TTWonP).
- 2) a chanc.e in critical power ratio (A CPR) of less than 0.15. The 4 CPR i :- to insure that the previous]) calculated minimum CpR for an Oyster Creek transient (HoJ Withdrawal Crror) is not penetrated.
The pre: .u r e criteria is a change from the current technical speci-fication bases. The technical specification bases includes a 25 psi margin to SV's for the TTWODP transient. Because the actuation of a SV is not a safety concern, the 25 psi margin requirement to SV set point is being elininated in the technical specification change re-quest. The justification for removal of the 25 psi limit is provided in reference 9.
A sunary of the results is presented in TTble 1, and description of the transients is provided below.
Turhine Trir A turbine trip is the primary turbine protection nechanisn and is initiated whenever various turbine or reactor system malfunctions threaten turhine operation. The turbine trip initiates closure of the turbine stop valves and opens the bypass valves. Upon initiation of the trip, pressure rises, opening the first two ERV's at 1.4 seconds.
All 5 ERV's are open by 2.0 seconds and are closed by 6.0 seconds. A peak pressure of 1129.3 psia occurs at 2.0 seconds which is well below the SV set points. The transient response is sheen in Figures la and Ib.
Turbine Trip Without Bypass The relief-valve sizing transient (TTKOBp) demonstrates the adequacy of the relief valves in keeping pressure below the safety valves set points.
The failure of the bypass valves to open following a turbine trip results in a rapid pressurization, opening the first two relitf valves at 1.2
[
I seconds and all five valves by 1.5 seconds. The valves are all closed I at 10.5 seconds. A peak pressure of 1201 psia is reached at 3.6 necends, j
'26 psi below the lowest safety valve lift set point.. Pressure in the ;,
i done reaches 1204.9 psia and 1215.3 at the core midplane, which is below design limits. The isolation condenser is actuated at 1065 psia; this 9 does not affect peak pressure, but assures that the system will not re-pressurize following closure of the relief valves. The transient re-sponse is shown in rigures 2a and 2b. ,
_' tai._n Stean Line Isolatfen Valve Closure The transient resultinp, f rom a 3 second closure of the main steam line isolation valves is shown in Figures 3a and 3h. Two relief valves open at 3.7 seconds rid al1 five are open by 4.0 seconds, remaining open to 14.9 seconds. A peak pressure of 1135.4 psia occurs at 4.5 seconds which is well below SV set points. The analysis shows that the relief valves would reeren at 18 seconds. Operation of the isolation condenser, for which no credit is taken in this analysis, would prevent subsequent lifts.
Loss of Electrical Load The Loss-of-Electrical load transient exhibits the same characteristics as the Turbine Trip transient. Ilowev e r , since the steam flow to the turbine is initially reduced by action of the throttle valve rather than sudden closure of the stop valve, the pressure and power transients are milder, and a detailed analysis of this transient was not performed.
Loss of Auxiliary Power Loss of auxiliary power causes loss of condenser cooling water, trip of feedwater pumps, trip of recirculation punps and turbine trip. It is assumed (4) that the bypass will remain open for about 1.5 seconds. The typass valves trip shut when the main condenser vacuum reaches 10 inches IIg. Reactor operating experience has shown that vacuum does not drop below the 10 inch set point until after 1.5 seconds. This period of steam bypass significantly reduces the magnitude of the reactor power transient following the turbine valve closure, and the ensuing pressures :
transient is less severe tLin that obtained in the Turbine Trip Without !
Bypass Transient. Hence, this transient is not analyzed in detail. [
Loss of Main Condenser Vacuum ,
I' The sudden loss of main condenser vacuum causes a simultaneous scram, turbine trip, and bypass valve closure. This transient is identical ,
to the turbin< trip without bypass transient which is discussed above. !
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RELIEF VALVE ".0D_EL The relief valve model in PTSiniR-2 represented the five valves conbined as one valve with a single set point. The separation of the relief valve set points ceuld not be analyzed "i thout nod i fying the program. Therefore, the latest version of PTSBWR-2 (5), which includes an isolation condenser model, vparates the relief valves into two groups containing two and three valves, respectively. The model change uas made such that the two groups of valves with equal set points would previde identical resultn to the combined five-valve model. The group with two relief val"ves were assigned the lower set points to represent the one ERV in each header lifting first.
EFFECT OF :'.0 DEL CllANGE 0:: PREVIOUS ANALYSES The relief valve model in PTSIM:R2 is bypassed by the basic calculations until activated when pressure reaches the ERV set point s. Thus the model change would not alter the results of transients other than those having ERV actuation. Those transients with ERV actuation were re-evaluated as part of the current analysis. P re'c i ous analyses are given in references 6, 7 and S.
EFFFCT OF ERV SET POINT INCREASE ON CYCLE 9 OPERATT0]:
The limiting pressurization transient (t" Sine trip without bypass) was analyzed for the current cycle 9 operatte". The 2nd of cycle (EOC) 9 kinetics parameters were used with the censervative multipliers described earlier. The EOC 9 scram curve was used with a 0.8 multiplier. The peak pressure was calculated to be 1193.6 psia and a d CPR of 0.0251. These values are within the requi red saf et;. nargins and would not alter the con-clusions reached in the cycle 9 safety evaluation (Reference 10).
EFFECT OF INCREASED CRV SET POINTS ON LOCA ANALYSIS The limiting LOCA analysis (11) for Oyster Creek is a large break which does not result in ERV actuation. Therefore the results of that analysis will not change with increased ERV set points. The most limiting LOCA analvsis which lifts ERV's is the 1.0 sq. ft. break size. The increased set points would have an increase in peak clad temperature of less than 3.00F. This would not change the limiting LOCA size or PCT location nor result in exceeding the 2200 F limit for PCT.
EFFECT OF INCREASED ERV SET POINTS ON TORUS STRUCTURAL RESPONSE An analysis was performed to evaluate the affect of the proposed set point change on the hydrodynamic load ong the south and north header relief valve discharge pipes. The rest .ts showed that the maximum hydrodynanic load increased by 0.81% during the blowdown transient.
This small increase in hydrodynamic load will have a negligible impact on the torus structure, and is much less than the original design with five valves at the same point.
I CONCLUS IO_N_S_
, The resetting of the ERV set points .20 psi higher increase the peak pressure of the limiting pressurization transient by 4.6 psi while increasing the margin het,"en ERV set points and normal operat ing i pressure to 50 psi. The .ncrease in margin of 20 psi providet the margin that existed before the set points were lowered. This added ;
nargin will nininize the probability of inadvertent relief valve aCluation.
! Based on these considerations and others previously ment ioned, it is ceneluded that the proposed amendment will not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a signficant decrease in a safety i na rg;in . Nor will it increase the possiblity for an accident or mal-function of a different type than any evaluated previously.
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RESUI.TS OF PTSII'.JR2 TRX;SIENT ANAIXSIS_
2ERV's at 1085, 3 at 1105 psia 5 ERV't, at 1085 psia
! TPE;SIENT PEAR PR!:SSUl:E (PSIA) 4 CPR*** PEAK PRFSSUI:I: ACPR r
- 1129.3 0.0 1121.0 0.0 TTWolP** 1201.0 0.0752 1196.4 0.068 MSLIVC* 1135.4 0.0 1134.8 0.0
- I.initing Pressure =
1227.0 PSIA (minimum safety valve setpoint) i **l,init ing Pressure =
110?, of reactor coolant pressure b o und a r:. design pressure
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- I. i n i t. i n g A CPil = 0.1i l
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REFERENCES:
- 1. Let ter t o George Lear (NRC) f rom I . R. Finf rock (JCP&l.), Report on Steam Vent Clearing Phenomenon, January 10, 1977.
- 2. NUREG-0737, " Clarification of TMI Action Plan Requirements"
! October 31, 1980 1
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- 3. XN-74-6 Revision, " Plant T rans i en t Simulation Code for Boiling Water Reactors, January 1975.
. 4. Letter to NRC f rom JCPSL Co. , dated May 8, 1975.
- 5. Letter to NRC fron JCP6L Co., " Transient of May 2, 1979", May 19, l 1979.
i C. XN-74-41, Revision 2, " Plant Transient Analysis of the Oyster Creek liWR with Exxon Nuclear 7X7 l*02 Iuel Assemblics, January 1975.
- 7. XN-7'.-41 Revision 2. "Pla a Transient Analysis of the Oyster Creek BWR with Exxon Nuclear SXS l'02 Fuel Assemblies, January '975.
- 8. X%75-51, Add i t ional Plant Transient Analyses of the Oyster Creek BWR with I:xxon Nuclear Fuel Assen.blies, September 1975.
- 9. I.etter to NRC (0.D.Part) from CE (J. F. Quirk), " General Electric Licensing Topical Report NEDE-240ll-P.A, 'Ceneric Reload Fuel Application,' Appendix D, Record Submittal" dated February 28, 1979.
- 10. Reload luformation a d Safety Evaluation Report for Oyster Creek Cycle 9 Reload, by R B. Lee, dated May 15, 1980.
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- 11. XN-75-55 Revision 2, Exxon Nuclectr Company, the Exxon Nuclear Company WREM-Based NJP-BWR ECCS Evaluation Model and Application to the Oyster Creek Plant, A u;',u s t 1976 9
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