ML20206S208
| ML20206S208 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/15/1986 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GL-84-09, GL-84-9, NUDOCS 8609220120 | |
| Download: ML20206S208 (37) | |
Text
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t i Mem.
UNITED STATES y g NUCLEAR REGULATORY COMMISSION WASHINGTON,D C.20555 c -
September 15, 1986
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Docket No. 50-219 LICENSEES: GPU Nuclear Corporation Jersey Central Power and Light Company FACILITY: Oyster Creek Nuclear Generatino Station
SUBJECT:
APRIL 10, 1986, MEETING WITH GPU NUCLEAR CORPORATION (GPUN) TO DISCUSS THE REQUESTED CANCELLATION OF THE UPGRADE OF THE CONTAI NITROGEN PURGE / VENT SYSTEM (TAC 59829) l On Thursday, April 10, 1986, a meeting was held at NRC, Bethesda, Maryland with GPUN, the licensee, on the licensee's request to cancel its comitment to In addition to this, the upgrade its containment nitrogen purge / vent system.
staff asked the licensee to discuss its submittals on the following subjects:
' Generic Letter (GL) 84-09, Recombiner Capability Requirements of 10 CFR 50.44(c)(3)(ii), and the licensee's request dated October 31, 1985, to cancel its commitments to install local suppression pool water temperature monitoring and forced mixing of the pool water. The staff had questions on the licensee's submittals on these sub.iects which are included in Attachment 3.
The staff previously presented the questions on GL 84-09 to the licensee in a phone conference call on March 21, 1986.
Attachment 1 is the list of individuals that attended the meetino. Attachment 2 is the material handed out by the licensee in the meeting for its presenta-tion. Attachment 3 is the material handed out by the staff. These are dis-cussed further in these meeting minutes. The following is a summary of the significant items discussed and the actions, if any, taken or proposed.
Attachment 2 has the following material presented by the licensee: (1) aoenda j
for the meeting on combustible gas control, (2) drawing of.the Oyster Creek i
containment purge and vent system, (3) slides for the presentation on the Boil-l ing Water Reactor Owner's Group (BWROG) NED0-22155 Report, (4) hydrogen / oxygen l flammability limits (5) Generic letter 84-09, "Recombiner Capability Require-ments of 10 CFR 50.44 (c)(3)(ii)", (6) SECY-83-292, " Applicability of Recombiner Capability Requirements of Revised 10 CFR 50.44 to BWR Licensee's t
with Mark I Containments," and (7) the licensee's draft response to the staff's request for' additional information on GL 84-09.
(1) request l
Attachment 3 has the following material presented to the licensee:
for additional information on the licensee's submittals on Generic letter l
84-09, (2) the staff's request for additional information on the licensee's request to cancel its suppression pool modifications dated October 31, 1985.
In its presentation, the licensee referred to and included tables and figures from BWROG NE00-22155 Report, " Generation and Mitigation of Combustible Gas Mixtures in Inerted RWR Mark I Containments," dated June 1982. This report documents analyses of oxygen gen 2 ration in inerted BWR Mark I containments for 8609220120 860915 PDR ADOCK 05000219 P PDR
a range of transient and accident events. These analyses were performed to evaluate the existing combustible gas control capability in these containments.
The purpose gas to determine if the existing inerted containments adequately controlled combustible gas concentrations to below acceptable limits without requiring the use of hydrogen recombiners or containment purging. The report concludes that Mark I containments including Oyster Creek can keep peak oxygen concentrations below combustible gas limits without reouiring hydrogen recom-biners or containment purging. This report is available in the public document room (NRC accession number 8208180027). The report was submitted to NRC by the BWR Owners' Group by letter dated August 12, 1982. This report addresses hydrogen generation rates only for some accident scenarios.
1.0 Suppression Pool Modifications By letter dated October 31, 1985, the licensee proposed to cancel two modifica-tions to provide (1) local suppression (torus) pool water temperature monitor-ing at the electromatic relief valve quenchers, or discharge headers,This in the was pool, and (2) forced water mixing in the vicinity of the quenchers. 22, 1986.
previously discussed in the meeting summaries dated March 14 and May The staff's questions on these modifications are in Attachment 3. -The staff explained that it has a concern that cancelling these modifications would (1) raise the maximum allowed torus pool temperature, and (2) affect the torus pool meeting the acceptance criteria on safety relief valve discharges in Appendix A to NUREG-0661. The higher pool temperatures would affect the net positive suction head (NPSH) on pumps drawing from the torus, e.g., the core spray system pumps. The licensee has stated these pumps would be the first pumps affected by higher pool temperatures.
The licensee stated that there are water temperature limits on the pool in the Technical Specifications (TS). These are based, as explained in the Bases of the TS, on the electromatic relief valve discharges to the pool. The licensee explained that its proposal to cancel the above pool modifications is notmay These limits a be request to raise the water temperature limits on the pool.
changed only by a request to change the TS and would require review of the NPSH for the pumps drawing from the pool.
The staff stated that the licensee should refer to the staff's two Safety Evaluations (SE) o,n the Mark I Containment Long Term Program which are dated January 13, 1984. The SE do not state that the above torus modifications are required for the torus bulk water temperature monitoring or to meet the These acceptance criteria for acceptance criteria in Appendix A to NUREG-0661.
local temperature in the vicinity of the quenchers require the torus pool water temperature to be less than 200 F. It also has a maximum local-to-bulk water temperature of 40 F based'on the Monticello pool tests results.
The staff requested that the licensee state that there are no concerns with The NPSH for the core spray pumps with the pool temperature limits in the TS.
staff stated that the TS seem to allow a pool temperature as high as 110 F.
The licensee stated that the Emergency Operating Procedures (E0P) require the operators to take action at temperatures of 95 F not at 110*F.
The staff also requested a confirmation that the Monticello pool tests without the residual heat rcmoval (RHR) pump operation apply to Oyster Creek, that calculations on pool temperature do not take credit for this pump operation and
s a 3
that without these two pool modifications the pool still meets the acceptance criteria in NUREG-0661.
2.0 Generic fetter (GL) 84-09 The staff raised cuestions and the licensee provided preliminary responses regarding GL 84-09. The actual response was submitted fomally to the staff in the licensee's letter dated June 16, 1986.
The staff stated that the licensee must meet the three criteria stated in GL 84-09. The licensee must eliminate all sources of oxygen as far as practical in normal and post accident conditions. This includes the combustible pas control system itself.
3.0 Cancel Uporade of Containment Nitrogen Purge / Vent System The licensee has reouested to cancel its commitment to upgrade the containment nitrogen purge / vent system. This request was in the licensee's letter dated September 24, 1985. In response to the requirements of item 2.1.5a of NUREG-0578, the licensee submitted its letter of June 10, 1980 and provided a brief description of the planned modifications to improve the reliability The of the containment penetrations for the nitrogen purge and vent system.
licensee stated that Oyster Creek used its nomal containment vent and purge As system for post accident venting and purging of the containment atmosphere.
a result of this operation, the licensee committed to modifying the system to be safety grade / single failure proof for both operation of the system and isolation of containment.
The nitrcgen purge / vent system is the system used to inert the containment for power operation and to de-inert the containment The TSfor entry into requires, thecompletion after containment of for required surveillances, for example.
the startup test program and demonstration of plant electrical output, that the primary. containment atmosphere shall be reduced with nitrogen gas to less than 4.0% oxygen within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch is placed in the run mode. Primary containment deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a-scheduled shutdown.
This is TS 3.5.A.6 in Section 3.5, Primary Containment, of the TS. The inerted containment and the nitrogen purae/ vent system could be used, if needed, as a means to control combustible gases generated in a loss-of-coolant accident (LOCA). If the concentration of the combustible gases in containment became too high, the nitrogen purge / vent system could be used to purge or pressurize the containment with nitrocen. Either would reduce the concentra-tions of oxygen and hydro. gen in containment.
The agenda for the meeting on the licensee's reouest to not upgrade its The slides for the licensee's nitrogen purge / vent system is in Attachment 2.The basis for the licensee's presen-presentation are also in Attachment 2. Report, Generation and Mitigation of Combustibl tation was the BWROG NED0-22155 Gas Mixtures in Inerted BWR Mark I Containments, dated June 1982.
and, therefore, applies to more than The presentation is based on NED0-22155 enveloped the Mark I containments at 14 di
.iust Oyster Creek. NED0-22155 sites.
4 The licensee explained that a safety-grade containment vent and purge system is not necessary for Oyster Creek. The existing Mark I containment design is adequate.for combustible control without a recombiner (i.e., GL 84-09) and withoutventingthecontainmenttoreducetheconcentrationofcombustible gases. The licensee explained that it controls on oxygen inside containment and the radiolysis of water inside containment is the only sour'ce of oxygen inside containment. The licensee discussed the assumed rate of generation of oxygen in radiolysis (i.e., G value or molecules of oxygen per 100 ev of energy deposited in water) used in NED0-22155. These G values are less than the guide-line values in Table 1 of Regulatory Guide (RG) 1.7, " Control of Combustible Gas Concentrations in Containment following A Loss-of-Coolant Accident."
The staff stated that it's the staff's position that the G values for oxygen and hydrogen generation during radiolysis in Table 1 of RG 1.7 should be used.
This is until sufficient data justifying other G values is submitted to the staff. The staff has previously reviewed the G value data in NED0-22155. It concluded that this data could be used as a basis for the staff to issue GL 84-09 but was not acceptable for use in combustible gas calculations for the design basis LOCA.
The licensee explained that a determination of the most limiting event at Oyster Creek was made. It was based on the design basis accidents, small break accidents, isolation events and normal shutdown. The limiting event or transient was a hypothetical isolation event including the termination of boiling within the containment within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the LOCA. The basis of this assumption was
, discussed.
The oxygen concentration as a function of time after the LOCA was calculated. See the l
This is the concentration in the drywell and torus of the containment.
! figure in Attachment 2. The oxygen concentration history for 7 cases were calculated and presented in NED0-22155. The figure given in Attachment 2 is case 2 of NED0-22155. The licensee stated that the figure for case 2 shows the 5% oxygen limit in RG 1.7 would not be exceeded until approximately 6 years after the LOCA. The figure goes out to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> or 42 days. The licensee explained that the figure is conservative for Oyster Creek for se'veral reasons given in the slides for its presentation.
I Based on its presentation, the licensee concluded that Oyster Creek would not need to vent or pressurize the containment for 6 years after an accident. It also concluded that, therefore, a safety-grade vent and purge system was not needed for Oyster Creek.
The staff reiterated its position on the use of RG 1.7 in combustible gas calculations for the containment following an accident. It stated that it would issue a position on the licensee's request to cancel the commitment to This position would address the upgrade the containment vent and purge system.The letter with the staff's position licensee's presentation in this meeting.
would address the question of the modifications being done in the current Cycle >
i 11 Refueling (Cycle 11R) outage. The staff stated that a problem with the l licensee's presentation in.this meeting was that the results presented from NED0-22155 enveloped several plants and were not specific to Oyster Creek and the licensee's generation rates do not include all design basis scenarios.
1
e 5
The staff's position on the licensee's request was issued on May 5, 1986. A copy of this letter is included in Attachment 3.
$~acR'N.- ( , Project Manager BWR Project inrectorate #1 Division of BWR Licensing Attachments:
- 1. List of Attendees
- 2. Licensee's handout
- 3. Staff's handout cc: R. Bernero G. Lainas G. Hulman Distribution Docket File NRC PDR Local PDR BWD1 Reading 0GC-BETH (For info only)
J. Zwolinski J. Donohew C. Jamerson J. Kudrick P. Hearn F. Witt E. Jordan B. Grimes ACRS (10) h b
0FC : DBL:BWDf1 : 13L ' Del :D - Del : DBL:BWD#1 : : :
_____:__________[_ _
NAME:CJamersong 'ew :JKu k JZwolinski DATE:$/fl/86 '
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i 0FFICIAL RECORD COPY
i c p- -_ ,.4 Distribution for Meeting Sunnary Dated: September 15, 1986 Facility: OysterCreekNuclearGeneratingStation*
< Dock'etiF11e? (50.-219)y NRC PDR Local PDR BWD1 Reading R. Rernero J. Zwolinski J. Donohew C. Jamerson OGC-BETH (For info only)
E. Jordan B. Grimes '
ACRS (10)
G. Lainas' G. Hulman J. Kudrick P. Hearn F. Witt OC File
- Copies sent to persons on facility service list
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t% g ATTACHMENT 1 6 MEETING WITH GPU NUCLEAR CORPORATION (GPUN)
TO DISCUSS CANCELLATION OF THE NITROGEN PURGE / VENT SYSTEM UPGRADE APRIL 10, 1986 NAME ORGANIZATION J. Donohew NRC/NRR/ DBL J. Kudrick NRC/NRRC/ DBL M. Laggart GPUN N. Trikouros GPUN J. Lackenmayer GPUN R. Tarantino. GPUN P. Hearn NRC/NRR/ DBL F. Witt NRC/NRR/ DBL G. Hulman NRC/NRR/ DBL J. Torbeck General Electric
- s ATTACHMENT 2
- 1. Agenda Wor meeting.
- 2. Drawing of the containment purge and vent system.
- 3. Slides for presentation on the BWROG NED0-22155 Report.
- 6. SECY-83-292.
- 7. Draft response to GL 84-09 request for additional information.
4
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1
- a OYSTER CREEK NUCLEAR GENERATING STATION COMBUSTIBLE GAS CONTROL GPUN - NRC MEETING APRIL 10, 1986 6
PURPOSE:
Inform NRC of GPUN position in order to facilitate staff review of GPUN's request for cancellation of vent end purge system up-grade.
HISTORY: J. D. Lachenmayer (GPUN) o November 27, 1978: 10 CFR 50.44 became effective o June 10, 1980: In response to NUREG 0578, GPUN committed to provide an operationally single failure proof vent and purge system o April 9, 1981: GPUN provided additional information o December 2, 1981: Rule amended to require recombiners o August 12, 1982: BWROG provided NRC NED0-22155
" Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containments" o May 8, 1984: NRC issued Generic Letter 84-09 o July 13,1984: GPUN provided response to Generic Letter 84-09 o August 14, 1985: GPUN provided response to NRC's request for additional information relative to Generic Letter 84-09 o September 24, 1985: GPUN requested cancellation of vent and purge system upgrade
,, 4 AGENDA L
BWROG REPORT: N. Trikouros (GPUN)
J. Torbeck (GE) o Specific applicability to Oyster Creek o Report conservatisms GPUN CONCLUSIONS: M. W. Laggart (GPUN) o G. L. 84-09 supports determination that Hydrogen Recombiners are not needed at Oyster Creek o Oyster Creek combustible gas control is a passive oxygen based approach with no venting required o BWROG Report supports determination that venting is not required for combustible gas control at Oyster Creek.
o 10 CFR 50.44(g) addresses the acceptability of non-safety grade purge systems.
o Non-safety grade venting system available i
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WHY SAFETY GRADE CONTAINMENT VENT AND PURGE IS NOT NECESSARY FOR OYSTER CREEK REVIEW 0F NED0-22155
" GENERATION AND MITIGATION OF COMBUSTIBLE GAS MIXTURES IN INERTED BWR MARK I CONTAINMENTS" APPLICATION TO OYSTER CREEK 6
JET 4/86
PURPOSE OF REPORT t
0 EXISTING INERTED MARK 1 CONTAINMENT DESIGN IS ADEQUATE FOR HYDROGEN CONTROL 0 NO RECOMBINER REQUIRED 0 NO VENTING OF CONTAlfiMENT REQUIRED 9
.i JET 4/86
GENERAL THEORY t
0 INERT CONTAINMEl1T d
PRODUCTION OF HYDR 0 GEN ONLY NOT A CONCERN METAL-WATER REACTION NO PROBLEM CONTROLLING ON OXYGEN 0 RAD 10 LYSIS IS ONLY SdVRCE OF OXYGEN
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APPROACH 6
0 BASES FOR OXYGEN SOURCE RATE G(02 ) FOR BO~ILING WATER BASED ON OPERATINO PL EXPERIENCE AND TEST DATA (0.1 MOLECULES PER 100 EV)eR@ IM (0,2 E)
G(02) IS MUCH LESS FOR SUBC00 LED WATER FOR CONDITIONS OF INTEREST (0 TO 0.003 MOLECULES PER 100 EV) bb } .7 (0.26)
EVALUATED CONTAINMENT OXYGEN CONCENTRATION U 0
CONTAINMENT CODE ME)D- 22 )Ecf .g ((@2)
USE ABOVE ASSUMPTIONS FOR2G(0 )
ASSUME INERT CONTAINMENT USE LIMITING PLANT (CORE POWER TO CONTAINMENT VOLUME RATIO)
USE LIMITING EVENT (MAXIMUM CORE BOILING TIME
- 0. SHOW THAT PEAK CONTAINMENT OXYGEN CONCENTRATI BELOW R.G. 1.7 LIMIT (5% VOL)
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1 APPROACH PERFORM SENSITIVITY STUDY TO EXAMINE OTHER EFFECTS o NITROGEN CAD REPRESSURIZATION 50% CONTAINMENT DESIGN PRESSURE 90% c0NTAINMENT DESIGN PRESSURE o CONTAINMENT LEAKAGE o NON-BOILING SOURCE TERM O
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6 LIMITING PLANT DETERMINATI0ti 0 KEY PARAMETER IS CORE POWER TO CONTAINMENT VOLUME RATIO 0 GIVES HIGHEST CONTAINMENT 02 CONCENTRATION O PEACH BOTTOM AND BROWNS FERRY ARE LIMITING l
JET 4/86
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TABLE 1-1
, CORE POWER TO DRYWELL VOLUME RATIOS
' 102%
EN T NITROGEN POWER DRYWELL CAD LEVEL VOLUME SYSTEM (MWT ) (FT3) DRY. Vol.
PLANT YES 3359 159000 .0211 BROWNS FERRY 1,2,3 YES 2485 166000 .0150 BRUNSWICK 1,3 2429 132000 .0184 NO COOPER 2578 158000 .0163 DRESDEN 2,3 NO 1625 144000 .0113 YES-DUANE ARNOLD YES 2485 150000 .0165 FITZPATRICK 2485 146000 .0170 YES HA1CH-1 2051 147000 .0140 NO MILLSTONE 1 1887 180000 .0105 YES NINE MILE POINT 1 1971 180000 .0109 YES 0YSTER CREEK 3359 159000 .0211 OTTOM 2-3 YES ,
PEACH 2038 147000 .0138 NO PILGRIM 2561 158000 .0162 QUAD CITIES 1,2 NO 1625 134000 .0121 NO VERMONT YANKEE JET 4/86
LIMITING EVENT DETERMINATION 6
0 BASED ON ANALYSIS OF DESIGN BASIS AND DEGRADED FROM NED0-24708 0 LIMITING EVENT HAS LONGEST CORE BOILING TIME 0 EVENTS ANALYZED DESIGN BASIS ACCIDENTS SMALL BREAK ACCIDENTS (POST-TMI)
ISOLATION EVENTS (Low PRESSURE SYSTEMS AVAIL NORMAL SHUTDOWN i
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' JET 4/86
N. O LIMITING TRANSIENT t
0 HYP0THETICAL ISOLATION EVENT 0 NO OPERATOR ACTION FOR 7 HOURS 0 RPV DEPRESSURIZATION FROM 7 TO 10 HOURS 0
W RHR SHUTDOWN AllGNMENT FROM 10 TO 12 HOURS 0 BOILING TERMINATED IN 12 HOURS BOUNDING EVENT FOR 02 GENERATED 1 aatartce gt Loc /
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BASIS FOR LIMITING B0lllllG CONDITIONS 6
0 MAXIMUM INTEGRATED 02 GENERATED EXTENT OF BULLING REGION
- BOILING TIME RATE OF GENERATION 0 EVENTS EXAMINED LOCA SPECTRUM (DBA)
TRANSIENT SPECTRUM (ISOLATION E'/ENT) 100% BOILING FOR 12 HOURS ,
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l GAS CONCENTRATION ANALYSIS i
0 PURPOSE:
DETERMINE OXYGEN CONCENTRATION HISTORY IN AND WETWELL AIRSPACES FOR EACH OF 7 CASE SIGNIFICANT INPUT PARAMETERS ARE VARIED.
O APPROACH:
USE COMPUTER PROGRAMS WHICH PERFORM MASS AND ENERGY BALANCE EQUATION INTEGRATIONS FOR NITROGEN, OXYGEN, HYDROGEN AND STEAM IN EACH AIRSPACE.
MODEL METAL-WAT$R REACTION, RADIOLYSIS, NITROGEN CONTAINMENT AIR DILUTION SYSTEMS, CONTAINMENT LEAKAGE, VENT FLOW, VACUUM BREAKER FLOW AND RHR SYSTEM OPERATION.
SHORT TERM TRANSIENT ANALYSIS USES COMPLEX COMPUTER PROGRAM.
LONG TERM ANALYSIS USES SIMPLE PROGRAM WR THIS PURPOSE.
JET 4/86
TABLE 3-1 i
SENSITIVITY ANALYSIS CASES RADIOLYTIc 02 YlELD (MOL/100 EV) _
PRESSURIZATION CONTAINMENT CORE CORE WITH N2 CAD SENS. LEAKAGE (PSIG)
. CASE (% PER DAY)_
BASE 0 0.10 0.0 0.0 ANALYSIS 0.1 0.003 0.003 30 0.10 0.10 1
0.003 0 0.10 0.003 2 0.0 30 0.10 0.003 0.003 3 0.0 30 0.10 - 0.10 0.003 4 0.10 62 0.25 0.25 0.003 5 0.10 30 0.25' O.25** 0.25 6 0.0 '
0 0.1 0.0 O.0 ;
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OXVGEN GENERATION ANALYSIS CASE 2 2- '
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CONSERVATISMS IN ANALYSIS 6
0 ALSUMED FRACTION OF GAMMA DECAY ENERGY ABSORBED WAS THE CONSERVATIVELY HIGH REGULATORY GUIDE 1 0.10.
O DECAY HEAT GENERATION RATE BASED ON CURVE FITT WHICH OVERPREDICT ANS 5.1 STANDARD CURVES BY 2 AND 4X107 SECONDS.
0 USE OF MAXIMUM TECHNICAL SPECIFICATION ALLOWAB INITIAL 0XYGEN CONCENTRATION.
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0 USE OF REGULATORY GUIDE 1.7 ASSUMPTIONS FOR RELEA FRACTIONS OF FISSION PRODUCTS TO REACTOR COOLANT SUPPRESSION P00L. ,
0 NEGLECTING OTHER SOURCES OF OXYGEN DEPLETION SUC OXIDATION.
0 USE OF 102% CORE THERMAL POWER, O FAILURE OF ONE RHR LOOP WHICH REDUCES THE AM WHICH CAN BE ADDED BEFORE THE LIMITING PRESSU 0 USE OF'A CONTAINMENT LEAKAGE WHICH IS BELOW M DATA.
O ASSUMPTION OF-THE HIGHEST ALLOWABLE INITIAL TEMPERATURE WHICH LIMITS THE AMOUNT OF NITRO JET 4/86
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NhiTIONAL CONSERVATISMS FOR OYSTER CREEK LjM G' 4Ab 7AGROWs ftpgtt )-
0 CORE POWER LESS THAN REFERENCE PLANT 1971 MWT <3359 MW 7
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0 DRYWELL VOLUME GREATER THAN REFERENCE PLANT 180000 FT3 >159000 FT3
'0 EDUlVALENT BOILING TIME FOR LIMITING LOCA LESS T
. USED IN ANALYSIS- ,
100% F0'R.8 HR 4100% FOR 12 HR 0 MEASURED CONTAINMENT LEAKAGE GREATER THAN M IN ANALYSIS t' -
0.204% PER DAY >0.1% PER DAY s
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CONCLUSION i
o VERY CONSERVATIV$ ANALYSIS (CASE 2 - NO CONTAINMENT LEAKAGE) SHOWS NO VENTING OR N 2 PRESSURIZATION IS REQUIRED FOR 6 YEARS SAFETY GRADE VENT AND PURGE IS NOT NECESSARY FOR OYSTER CREEK i
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' UNITED STATES
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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 8, 1984 ALL LICEkSEES OF SPERATING REACTORS Gentlemen:
SUBJECT:
RECOMBINER CAPABILITY REQUIREMENTS OF 10 CFR 50.44 (c)(3)(ii)
(GENERIC LETTER NO. 84-09)
On December 2, 1981, the NRC amended 650.44 of its regulations by addition of the provisions in 550.44(c)(3). One of these provisions requires licensees of those light water reactors (both BWRs and PWRs) that rely upon -
purge /repressurization systems as the primary means of hydrogen control to provide a recombiner capability by the end of the first scheduled '
outage after July 5,1982, of sufficient duration to pemit the required modifications. Those plants for which notices of hearing on applications for construction pemits were ,
not pemitted by 10 CFR e) 50.44(published on or after November to rely on purge /repressurization systems 5,197 I as the primary means for hydrogen control. Therefore, these plants are not affected by the requirement for recombiner capability; the licensees; ~
of these plants are being furnished a copy of this generic letter for infomation only.
After adoption of the amended rule, and as a result of the new inerting requirement in 650.44(c)(3), the BWR Mark I Owners Group (incorporating i studies perfomed by Northeast Nuclear Energy Company) undertook a l substantial program to demonstrate that the Mark I plants potentially affected by the recombiner capability requirements of the rule do not need to rely on use of the safety grade purge /repressurization system required
, g by the original 10 CFR 50.44 rule as the primary means of hydrogan control.
7 xtensive review and independent studies by the NRC staff supported the E
findings of the Mark I Owners Group program. (This letter does not address PWRs because the inerting requirement of $50.44(c)(3) does not apply to PWRs and, therefore, the licensees of PWRs are not likely to be able to make a comparable demonstration.) .
The Comission has determined that a Mark I BWR plant will be found to not rely upon purge /repressurization systems as the primary means of hydrogen control, if certain technical criteria are satisfied. To avoid any misunderstanding, we wish to make clear that a plant that has a " safety grade" purge /repressurization system designed to confom with the general requirements of Criteria 41, 42 and 43 of Appendix A of 10 CFR Part 50 and
, installed in accordance with 150.44(f) or 150.44(g) must continue to have
~_'
that system, even though it may be determined with respect to 550.44(c)(3) that the plant does not rely on that system as the primary means for
! hydrogen control; thus, a decision on recombiner capability does not affect
- the requirements oT 150.44(f) and 650.44(g) for the " safety grade" purge /
' repressurization system.
% :::::::: ~) '
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I' ForthosejnertedMarkIBWRcontainments(forwhichnoticesonthe construction permits were published before November 5, 1970) that do not rely upon safety grade purge /repressurization systems as the primary means of hydrogen control, the Commission detennination, cited above is applicable provided the following criteria are met:
- 1) The plant has technical specifications (limiting conditions for operation) requiring that, when the containment is required to be inerted, the containment atmosphere be less than four percent oxygen, and
- 2) The plant has only nitrogen or recycled containment atmosphere for -
use in all pneumatic control systems within containment, and
- 3) There are no potential sources of oxygen in containment other than that resulting from radiolysis of the reactor coolant. Consideration of po-tential sources of inleakage of air and oxygen into containment should include consideration of not only nonnal plant operating conditions but also postulated loss-of-coolant-accident conditions. These potential sources of inleakage should include instrument air systems, service g air systems, MSIV leakage control systems, purge lines, penetrations pressurized with air and inflatable door seals.
Accordingly, any Mark I BWR owner which has concluded that a recombiner capability is not required for its facility is requested to submit a response to this letter within 45 days. Each submittal should indicate the applicability of the generic studies submitted by the Mark I Owners Group to the licensee's facility and include additional information relative to the three criteria cited above to enable the staff to make a comparable decision.
This request for information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985.
Comments on burden and duplication may be directed to the Office of
~
Management and Budget. Reports Management Room 3208, New Executive Office i
Building, Washington, D. C. 20503.
Sincerely,
(
s 41 Darrell G. Eisenhut kirector Division of Licensing Office of Nuclear Reactor Regulation
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POLICY ISSUE July 19. 1983 (NEGATIVE CONSENT) SECY-83-292 ;
For: The Commissioners ,
From: William J. Dircks i Executive Director for Operations j j
Subject:
APPLICABILITY OF RECOMBINER CAPABILITY REQUIREMENTS j OF REY 15ED 10 CFR 50.44 TO BWR LICENSEES WITH MARK 1 CONTAINMENTS
Purpose:
To infom the Commission that the staff plans to ' notify all licensees of operating light water reactors that the recombiner requirements of 10 CFR 50.44 are not l j
applicable to those licensees that can show that a they do not rely upon installed safety grade purge i repressurization systems as the primary means of l
hydrogen control, and that the staff is prepared to accept such a demonstration from BWR licensees with inerted Mark I containments upon a showing that certain technical criteria are saticfied. ,
i Discussien: On December 2,1981, the NRC amended its regulations f applicable to light water reactors to require, among -
other things, a hydrogen recombiner capability to reduce the likelihood of the need to deliberately vent radioactive gases from the containment following de-graded core accidents. The revised 10 CFR 50.44 requires licensees of those light water reactors (both .
BWRs and pWRs) that rely upon purge /repressurization ,
systems as the primary means of hydrogen control (approximately 54 plants) to provide a recombiner capability by the end of the first scheduled cutage af ter July 5,1982, of sufficient duration to permit ,:
the required modifications. The more recent plants, *.!$
Contact:
R. Butler 49-29142 c )*
V. Rooney
- ' O,g2 .
49-28286 q -
9 - - '"
6 1.e., those for which notices of hearing on applica-tions for construction permits occurred on or af ter November 5,1970, are not pennitted by 10 CFR 50.44 to rely on purge /repressurization systems as the primary means for hydrogen control. Therefore, the more recent plants (approximately 10) are not affected by the requirement for a recombiner capability. Among the BWR plants, they include those with Mark II and Mark III containments and a few of those with the Mark I containment.
Af ter adoption of the rule, the Mark I Owners Group ,
undertook a substantial program to demonstrate that -
the 21 Mark I plants potentially affected by the recombiner requirements of the rule do not need to rely on use of the safety grade purge /repressuriza-tion system required by the original 10 CFR 50.44 for preventing the combustion of hydrogen. The recombiner capability required by the new rule would not have to be be provided if the demonstration were successful.
The licensee for Millstone I, a BWR plant with a Mark I containment, made separate and substantially identical submittals to those made by the Mark I ,
Owners Group. (The Millstone I licensee is also a member of the Mark I Owners Group. When we refer to the Mark I owners in the following paragraphs we intend to include the Millstone I licensee.)
The owners of the 31 PWR plants potentially af fected by the new rule have not undertaken a comparable program. Since PWR plants are not operated with an inerted containment, they are not likely to be successful in demonstrating adequate hydrogen control without reliance on the installed purge /repressuriza-
' tion system. However, this will depend on the magnitude of the other plant specific sources of hydrogen, e.g., corrosion of zine based paints, galvanized steel and of aluminum.
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To deal with the hydrogen releases associated with i i degraded core accidents, all Mark I BWR plants are ;
required to have inerted containments, which means i that, in essence, the oxygen concentration must be less than four percent by volume. The lower j flammability limit for oxygen in a hydrogen-air J
mixture is five percent by volume. 5 il The objective of the Mark 1 Owners Group program was to show that since post-accident oxygen sources J are minimal, there will be no need to purge /
repressurize an inerted containment after a degraded l 4
core accident to prevent the combustion of the hydrogen. $
The Mark I owners asserted that if they converted their instrument air system inside containment to I
one that operates on nitrogen (or recycled containment ,
atmosphere), then the only post-accident source of g
oxygen would be radiolysis of the coolant. i The related recent examination by the Mark 1 jl:
Owners Group of the radiolysis source tera j!
for oxygcn production in a Mark I containment y shows that the conservatively high rates of producing i oxygen by radiolysis assumed in Regulatory Guide i 1.7 (Revision 2) " Control of Combustible Gas i Concentrations in containment following a loss of .
Coolant Accident" can exist only when there is either: )
t (1) bulk boiling of the coolant in the reactor core; f or (2) a substantial release (i.e., greater than about three percent) of the total inventory of the fission product iodines to the coolant at a time when the hydrogen concentration in solution with the coolant is minimal (i.e., less than 2.9 x 10-5 moles / liter). ,
Beyond this, the Owners Group analyses demonstrate i that bulk boiling of the coolant must occur for more 3 than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before a flammable hydrogen-oxygen ?
mixture in the containment would result. Such a f period of time for bulk boiling would exceed that Further, predicted during design basis accidents.
if there is a substantial release of fission proccc:
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% The Mark I Owners examined both the design basis accidents and the degraded core accidents in terms of the above considerations. They found that while degraded core accidents lead to large releases of iodines and protracted bulk boiling of the coolant in the core region, the corresponding large releases of hydrogen from the resulting zirconium-water reaction would tend to suppress net radiolysis by back-reaction effects and would serve to reduce the oxygen concentra-tion due to its diluent ef fect. For design basis accidents, they found that bulk boiling would terminate af ter about 12-hours following onset of the accident and that the amount of iodines released to the coolant under this condition would be substantially less than three percent of the total inventory. This means that the radiolysis is a significant source of oxygen only during the first 12-hours following onset of the accident. Their analysis of this scenario showed that the oxygen concentration stays below the lower flammability limit of five percent during this period of time.
The staff and its consultants have reviewed the analyses submitted by the Mark I owners and conclude that:
- 1) Radiolysis is the only significant post-LCCA source of oxygen inside containment, if oxygen is removed from the working fluid for the instrument air system in containment, and if oxygen is not produced by any other source.
- 2) Except for certain low probability degraded core accident scenarios, bulk bciling of the reactor coolant is not expected beyond 12-hours following onset of the LOCA and the iodine release fraction is expected to be substantially less than three percent of total inventory.
- 3) The oxygen concentration in containment, starting from a pre-LOCA value of four percent and assuming the conditions prescribed in Item 2, above, will remain below five percent throughout the course of the accident.
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_4 todines to the coolant (as would be characteristic
' of degraded core accidents which would involve a large zirconium water reaction) there also would be significant hydrogen release to the containment.
This would result in a hydrogen concentration in solution with the coolant well above the minimal value that would sustain the high radiolysis rate assumed in the Regulatory Guide.
Until now, the staff had been requiring use of the non-mechanistic conservative assumption that 50% of the total iodine inventory is in solution with the c'oolant for all postulated accidents for all facilities. For these particular applications, the Mark I Owners Group has asserted and the staf f agrees that such use of the conservatively non-mechanistic assumptions of the Regulatory Guide is not warranted for all accidents.
The recombiner capability requirement was imposed to remove the need (based on assumptions of the Regulatory Guide) to intentionally increase offsite accident doses as part of the ef forts to prevent a post-accident combustible hydrogen mixture in containment. Hydrogen combustion in containment could be prevented under these conservative assumptions by venting it to the atmosphere via the existing purge /repressurization system. However, use of this system following an accident could involve a small increase in the total radiation dose released to the environment. If the assumptions of the Regu-
. latory Guide were used, recombiner capability would be required to prevent hydrogen combustion without release l of radiation to the environment. Nevertheless, since l the increase in offsite dose associated with use of the purge /repressurization system at Mark I B'4R plants, even when using the very conservative source tenn assumptions of releasing 100% of the noble gases and l 50% of the iodine to the containment, amounts to no more than about 60 rems thyroid and 0.2 rem whole-body, I, and since the cost of the hydrogen recombiner capa- -
bility is about 53 million per plant, we do no: ,
believe it warranted in these cases to require use of ;
the conservative prescriptions given in the Regulatory 1 Guide for all accident scenarios. l i
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- 4) Regardless of the hydrogen concentration in the
( containment, combustion cannot occur if the oxygen concentration remains below five percent.
- 5) Therefore, except for certain low probability degraded core accident scenarios, the Mark I BWRs need not rely on the purge /repressurization system as the primary means of hydrogen control.
For the window of low probability scenarios where use of the purge /repressurization system is needed to prevent combustible mixtures, the use of the purge /repressurization system would be acceptable because the associated total dose is well below 10 CFR Part 100 guidelines.
- 6) The cost of the recombiner capability, if implemented, will range between two and four million dollars per unit or around $60 million for the approximately 20 units af fected. This is substantially higher than the cost of $100,000 per unit considered by the staf f during rulemaking.
The staf f concludes, on the bases detailed above, that since the radiolysis source tems, except for certain low probability scenarios, are substantially lower than previously considered for these particular applications, and since the cost of the recombiner capability is substantially higher than previously considered, the recombiner capability should not be required for all Mark I BWR plants.
Based on the foregoing considerations, the staf f plans to notify all licensees of operating light water reactors that the recombiner capability ~
requirements of 10 CFR 50.44 are not applicable to those licensees that can show that they do not rely on safety grade purge /repressurization systems as the primary means of hydrogen control, and that the staff is prepared to accept such a demonstration
k from BWRs with inerted Mark I containments if the licensee demonstrates to the satisfaction of the staf f that the following criteria are met:
- 1) There are no potential sources of oxygen in the containment other than radiolysis of the reactor coolant; e.g., the instrument air systems inside containment must use either nitrogen or recycled containment atmosphere; and
- 2) The plant Technical Specifications contain a limiting condition for operation of four percent for oxygen concentration in the containment when the inert condition is required.
To avoid any misunderstanding, it should be noted that the licensees of Mark I BWR plants that do not have installed safety-grade recombiner capability, I or do not plan to install such capability, must continue to have installed safety-grade purge or purge /repressuri-zation systems as appropriate to the requirements of 10 CFR 50.44 (f) and (g). Conversely, if the licensees have installed safety grade recombiners, then safety-grade purge /repressurization systems are not required. Similarly, the Hark 1 and Mark
!! BWR plants for which notices of hearing on applica-tions for construction pemits occurred cn or af ter November 5,1970, must continue to have installed safety grade recombiners as called for in 10 CFR 50.44 (e). This will pemit the ef fective control of combustible gases following a LOCA without involving any significant releases of radioactivity frcm the containment even for the less likely accident sequences.
In addition to the above considerations concerning
, technical issues related to 10 CFR 50.44, related administrative questions concerning this rule have arisen. Several licensees have requested schedular l exemptions from 550.44 to provide time to confirm I
the plant-specific applicability of the Owners Group l analysis or other analyses that purport to show that safety-grade recombiner capability need not be provided. ;
i
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In addition, the licensee for Monticello ordered two safety grade recombiners in 1979 and is proceeding towards recombiner installation, but requested an exemption from the schedular requirements of the rule until December 31, 1983 in order to provide sufficient I time for engineering, procurement, and installation.
The staff plans to infonn all affected licensees that it is prepared to agree that the recombiner I capability requirements of 10 CFR 50.44 are not applicable to BWRs with inerted Mark I containments .
if the licensee demonstrates to the satisfaction of j the staf f that:
- 1) There are not potential sources of oxygen in the containment other than radiolysis of the reactor .
coolant; e.g., the instrument air systems inside containment must use either nitrogen or recycled containment atmosphere; and
- 2) The plant Technical Specifications contain a limiting condition for operation of four per- ::
cent for oxygen concentration in the contain- :
ment when the inert condition is required. j In addition, the staf f plans to grant the schedular relief related to this issue that has been requested by '
the several licensees (as discussed above), upon a showing of good cause.
The staff intends to proceed with this course of .
action within 10 working days from the date of this paper, unless instructed otherwise by the Commission. .
/ 0 S/
s William Di J./ rcks Executive Of rector for Operations i
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l SECY NOTE: In the absence of instructions to the contrary, )
6 SECY will notify the staff on Thursday, August 4, 1983 that the Commission, by negative consent, l assents to the action proposed in this paper. j l
DISTRIBUTION:
Commissioners OGC OPE OCA OIA OPA ~
REGIONAL OFFICES '
EDO ELD SECY l
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, )M P i RESPONSE TO 0YSTER CREEK GL84-09 M 901) C /3 REQUEST FOR ADDITIONAL INFORMATION
- l. Backup air supply to the N2 system will be automatically isolated when conIainmentisolationoccursforaLOCA.
- Identify components relied upon to isolate the backup air supply for the N2 system and verify they are safety grade.
RESPONSE
The components for isolating tne instrument air / nitrogen system are V-6-393 cnd V-6-395. V-6-395 is an air operated valve located outside the drywell which closes upon any automatic MSIV isolation signal (low-low Rx water level, steam line high radiation, steam line Dreak or steam line low pressure) from the reactor protection system. V-6-393 is a check valve located inside the drywell and is in series with V-6-395. V-6-395 will also f ail closed on loss of air or loss of electric power. These valves are designed to ASME III criteria; V-ti-395 as Nuclear Class IE and V-6-393 as Nuclear Class II. V-6-395 and the piping through and including V-6-393 were designed to meet seismic criteria Class I.
- Indicate if the response is valid for tne situation of a failed nitrogen system.
RESPONSE
Since tne MSIVs are' air-to-open valves, a failure of the nitrogen system would otherwise cause the MSIVs to close, initiating a reactor scram and cnallenging the integrity of the safety shutdown systems. By switching over to instrument air upon low nitrogen header pressure, a spurious reactor trip is avoided. The MSIVs do not rely solely on the instrument air / nitrogen supply for closure. They can close on either actuator spring The instrument air / nitrogen force or pressure storea in the accumulators.
system operates at normally 100 psig and is not needed once the MSIVs are closed upon a signal from the reactor protection system.
The switch over to instrument air upon nitrogen system failure wnile still maintaining atmosphere control and reactor isolation capability is considered a valid system response. ..
. \- --
l FI ffOhf 1
- .Uescribe the inspection and testing program to assure the operaDility I l
of these components.
RESPONSE
i Procedures 312, " Reactor Containment' Integrity ano Atmosphere Control" and 201.1, "Approacn to Critical" have prerequisites that all automatic containment isolation valves are operaole or are secured in the closed l position, reset.drywell isolation signal and complete a containment system integrity valve check-off list. The check list needs to be initialed, verified and approved to ensure the system's operability.
Operability of V-6-395 is tested in accordance with station procedure 652.3.001, " Instrument Air Isolation Surveillance Test", and 10CFR50 Appenaix J testing is performed in accordance with station procedure 665.5.006, " Local Leak Rate Tests".
I l
l 2. Service air ano breatning air systems are not connected to the drywell -
during power operation.
- Describe the administrative controls and/or interlocks used to prevent these systems from adding oxygen to the containment during power I operation.
RESPONSE
l There is'no permanent piping for tne service air and breathing air systems in the drywell. Hoses are used when these systems are needed for work in tne drywell during non-power operations. Prior to Approach to Criticality j (Procedure 201.1), at least one door in the airlock has to be closed and sealed, and there is a check-out list for loose parts, etc. Therefore, it is not possible that the service air and/or breathing air hoses remain connected into the drywell during power operation.
- - _ , . - ____.,.._.__7 - _ . . _ -,_ , ,_, , , ,. . , , , , _ _ ,
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- 3. The.TIP purge system (which may use N2 or air) uses nitrogen during power ope /ation. Describe the administrative controls and/or interlocks.
Verify tney are safety graae. Describe the inspection and testing programs to assure the reliability of these components.
RESPONSE
Steps for containment inerting in Procedure 312 requires the operator to open and tag the nitrogen supply to the TIP indexer V-23-162 and close and tag the air supply V-6-1321 in the TIP drive area. A check-off list (Figure 312-4) has to be checked, verified and approved by three individuals. This ensures that only nitrogen will be used in the TIP purge system during power operation.
9 5
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ATTACHMENT 3
- 1. Request for additional information on GL 84-09.
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- 2. Request for additional information on licensee's request to
- cancell two suppression pool modifications.
- 3. Staff's letter of May 5,1986. on the licensee's request of .
September 24, 1985.
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REQUEST FOR ADDITIONAL INFORMATION OYSTER CREEK NUCLEAR PLANT, UNIT 1 i .
DOCKET NO. 50-219; TAC NO. 58018
' PLANT SYSTEMS BRANCH By a letter dated August 14, 1985 the licensee responded to the staff's l
request for additional information (RAI) dated April 29, 1985. The following l
RAI is based on the licenste's August 14 submittal:
- 1. In the response to question 1, you stated that the backup air supply to the Nitrogen System will be automatically isolated when the primary containment isolation occurs for the design basis Loss-of-Coolant Accident (LOCA). Identify the components relied upon to isolate the backup air supply from the Nitrogen Systen and verify that they are safety grade. Also, indicate if the response is valid for the situation of a failed-nitrogen system. Describe the inspection and testing progam '
employed to assure the operability of these components.
- 2. In the response to question 2 you stated that the service air i and breathing air systems are not connected to the drywell during power
- operation. Furthermore, you stated the TIP purge system which j may use nitrogen or air, uses nitrogen during power operation.
Describe the administrative controls and/or interlocks used to prevent these systems from adding oxygen to the containment during power operation. Identify the components relied upon to isolate these potential oxygen contributors from the containment, and verify that these j components are safety grade. Describe the inspection and testing program '
employed to assure the reliability of these components.
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REQUEST FOR ADDITIONAL INFORMATION 0YSTER CREEK NilCLEAR PLANT DOCKET N0. 50-219; TAC NO. 60152, 60153 .
By letter dated October 31, 1985, the licensee requested cancellation of two modifications that were to be completed in the torus suppressinn pool during the Cycle 11 R outage. The fo11'owing infomation is needed by the staff to complete its evaluation:
- 1. Is .the Acceptance Criteria on safety relief valve discharq<n in Appendix A to NUREG-0661, Mark i Containment Long-Tem Prograa, applicable to these modifications and does Oyster Creek meet the Acceptance Criteria without the modifications. This includes the use of the Monticello test data to obtain the 43'F as the maximum local-to-bulk torus water temperature difference for Dyster Creek. Nas anything changed such that Oyster Creek is outside the assumptions made ,
i by the staff's two safety evaluations dated January 13, 1984, on the Mark I Containment Long Term Program.
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. . t SUPPRESSION POOL THERMAL MIXING MODIFICATIONS REGUEST FOR INFORMATION To provide the basis for postponing and eventually cancelling the requirements for suppression pool temperature monitoring local to the quenchers and the thermal mixing modifications for the suppression pool the following additional infonmation is needed:
- 1. Will the suppression pool temperature limits be relaxed, tightened or remain the same with the elimination of the SRV discharge temperature limit?
- 2. What are the temperature limits needed to provide adecuate NPSH for the core spray pumps under applicable accident conditions, or if not core spray pump NPSH, the temperature limits for the next most restrictive suporession pool limit?
4 3. What is your basis for determining the adequacy of the existing temperature monitoring system in providing information concerning the local temperatures to the core spray pump suction header?
- 4. What is your basis for determining the adequacy of existina 1
suppression pool mixing under applicable accident conditions in providino .uppression pool temperatures local to core i
spray pump suction header which are indicative of data
! obtained from the existing temperature monitoring system?
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UNITED STATES
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- n. NUCLEAR REGULATORY COMMISSION wAs nwoTom.o.c.aoses s -I e
May 5, 1986
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i Docket No. 50-2 9 Mr. P. B. Fiedler Vice President and Director
- Oyster Creek Nuclear Generating Station Post Office Box 388 *
- Forked River, New Jersey 08731
Dear Mr. Fiedler:
SUBJECT:
MEETING OF APRIL 10, 1986, ON REOUESTED CANCELLATION OF NITROGEN
- PURGE / VENT SYSTEM (TAC 59829) .
1 - Re: Dyster Creek Nuclear Generating Station l
In the meeting of April 10, 1986, the staff and GPU Nuclear personnel (the f
licensee) discussed the followinq Oyster Creek ifcensing actions before
' the staff: (1) cancellation of two torus pool modifications, (2) Generic Letter 84-09 dated May 8,1984, and (31 cancellation of upgradina the
( . nitrogen purge / vent system. These are licensing actions associated with the Cycle 11 Refueling (Cycle 11R) outage which began April 11, 1986, and is scheduled to end in October 1986. The expectation was that these actions
' would be resolved before the commencement of the current outace.
The staff requested information on the torus pool mcdifications and Generic Letter 84-09, in the April 10, 1986, meeting in order to complete its j
evaluation. The requested information is an enclosure to this letter.
Two important conclusions were drawn at the April 10, 1986 meeting: (1) the control on oxygen in the inerted containment at Oyster Creek during the
! design basis Loss-of-Coolant Accident (LOCA) is not sufficient in itself to '
be the safety-grade combustible gas control system (CGCS) required per i
10CFR50.44(glassumingthesourcetermsofRegulatoryGuide(RG)1.7and i
(2) the initial suppression pool temperature used in the LOCA analysis may be j
less than the pool temperature limits in the Oyster Creek Technical i Specifications positive suction head (TS)(NPSHandformay),
the corefor sprayspecific pumps transients, result during a LOCA. A in loss of n l question on item (2) is also in the enclosure to this letter.
The staff's position is that RG 1.7 is predicated on defense-in-depth and.
because of LOCA combustible cas uncertainties, is the appropriate source tems to be used in LOCA analyses. With your inerted containment at less than or equal 4% oxygen and fo11cwing the guidelines of RG 1.7, the oxygen concentration in the containment will be above 5% oxygen within 30 days of a It is also the staff's position that a safety gra'de CGCS
- (. design basis LOCA.
in addition to an inerted containment is required to meet the requirements
(' of 10 CFR 50.44(g). It is our understanding that you believe RG 1.7 is 3
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". the inerted overly conservative and, using the source terms of NE00-22155, The containment is sufficient in itself to meet 10 CFR Part 50.44(q). The staff source terms in this NEDO report have been reviewed by the staff.
concluded in GL 84-09 that, relative to the need for recombiner capability, the NEDO report provided a basis to accept the BWR Mark I Owner's/ Group position that Mark I plants do not rely on the use of the safety Thegrade staff ource repressurization as the primary means o' combustible gas control.
concluded that both the inerted containment and the puroe/repressurizatio were the means of combustible gas control.still concluded that R.
We reouire that you commit to provide a safety-grade CGCS (e.g., a contain purge and/or repressurization system) by the restart from the Cycle 12R ou to meet 10 CFR 50.44(g).
It is also the staff's position that the initial suporession pool temperature
. for a LOCA shall be consistent with the most conservative cool temperature Usino this analysis allowed by the TS for any extended period of time.there The core should system such as the core spray system during a design basis LOCA.
spray system is relied upon in accident analyses to protect the core dur This staff position was discussed with the licensee durino the a LOCA.
April 10, 1986, meeting and by telephone on April 11, 1986.
(' You should also address the questions enclosed in this letter.
For the question on the initial suppression pool temperature durino a design ba LOCA, you should provide a schedule for completing any calculations ne You should also provide any compensatory measures, to answer the question.if needed, to account for the LOCA calculations not appropriate suppression pool limits in the TS.These limits would be required prior to limits based on the LOCA analyses.
the restart from the Cycle 11R outage.
Becausa these issues are important and are now cl We request your response to this letter addressing the above issues by May 15, 1986.
The reporting and/or recordkeepino requirements co under P.L.96-511.
b Sincer ly, J .
John A. Zwolinski, Director BWR ro.iect Directorate #1 Division of BWR Licensina
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Enclosures:
Requests for additional information i
cc w/ enclosures:
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REQUEST FOR ADDITIONAL INFORMATION OYSTER CREEK NUCLEAR PLANT i{
DOCKET NO. 50-219; TAC N0. 58018 L
, j By a letter dated August 14, 1985 the licensee responded to the staff's 29, 1985. The following l request for additional infomation. (RAI) dated April RAI is based on the licensee's August 14 submitt21:
- 1. In the response to question 1, you stated that the backup air supply to the Nitrogen System will be automatically isolated when the primary containment isolation occurs for the design basis Loss-of-Coo Identify the components relied upon to isolate the Accident (LOCA).
l backup air supply from the Nitrogen System and verify that they are
~ safety grade. Also, indicata if the response is valid for the situation of a failed-nitrogen system. Describe the inspection and testing progam employed to assure the operability of these components.
2.
In the response to question 2 you stated that the service air
( and breathing air systems are not connected to the drywell during power operation. Furthermore, you stated the TIP purge system which 1
may use nitrogen or air, uses nitrogen during power operation.
Describe the administrative controls and/or interlocks used to prevent l these systems from adding oxygen to the containment during power l
operation. Identify the components relied upon to isolate these potentialoxygencontributorsfromthecontainment,.indverifythatthese Describe the inspection and testing program components are safety grade.
i employed to assure the reliability of these components.
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s REQUEST FOR ADDITIONAL INFORMATION OYSTER CREEK NUCLEAR PLANT 4
DOCKET NO. 50-219; TAC NO. 60152, 60153 By letter dated October 31, 1985, the licensee reouested cancellation of two modifications that were to be completed in the torus suppression pool The followino information is needed by the during the Cycle 11 R outage.
staff t'o complete its evaluation: .
- 1. Is the Acceptance Criteria on safety relief valve discharges in Appendix A to NUREG-0661, Mark I Containment leng. Term Procram, applicable to these modifications and does Oyster Creek meet the Acceptance Criteria without the modifications. This includes the use of the Monticello test data to obtain the 43'F as the maximum Pas local-to-bulk torus water temperature difference for Oyster Creek.
anythino changed such that Oyster Creek is outside the assumptions v.ade by the staff's two safety evaluations dated January 13, 1c84, on the Mark 1 Containment Long Tem Prooram.
In the April 10, 1986, meetino on the above technical issues, tne licensee the initial suppression pool temperature used in the LOCA ma.y stated that he less than the most conservative cool temperature allowed in the Technical For the staff to under-Srecification (TS) for any extended period of time.
stand what should be the appropriate TS on the suporession pool temperature, j
you should provide the following information requested 'in the meeting:
i 1.
Discuss the highest torus suppression pool temperature assumed in the Loss-of-Coolant Accident (LOCA) analysis and net positive suction head
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The statements made by the licensee (NPSH) of the core spray pumps.
are that these pumos would be the first safety pumps which would Was the initial experience loss of NPSH as the pool temperature rises.
- LOCA pool temperature based on the suppression pool temperature limits The initial temperature ,
in the station Technical Specifications (TS). -
should be the most conservative limits allowed by the TS for any extended period of time.
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7- I Althouoh thg followino detailed information was not reoues 1986, meetino, your discussion on suopression pool temperature a the core spray pumps should include the following:
2.
Describe the core spray pump suction header locations in the suppres-sion pool in relation to the relief valve (RV) quenchers, to the suppression pool bulk temperature monitors, and the pool water l 3.
Discuss the representativeness of the suppression pool temperature monitors for measuring the temperature at the core spray suction in the pool durino RV blowdown, 4
Discuss any assumptions made in the calculation of the temperature core spray suction in the pool that take credit for pool mixino durin RV blowdown to the torus, and .
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S.
What is the maximum temperature at the core spray suction header 4,
during RV blowdown in the LOCA, and what are your bases?
j 6.
Compare your assumptions in your calculations for NPSH to the Regulatory Position in Regulatory Guide 1.1, Net Positive S for Emergency Core Cooling and Containment Heat Removal dated November 1970. .
References to your Updated FSAR'as part of your response are a h
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0yster Creek Nuclear Generati ng Station s r cc: '
Resident Inspector Mr. Ernest L. Blake, Jr. .' ,
Shaw, Pittman, Potts and Trewbridge c/o U.S. NRC 1800 M Street, N.W. Post Office Box 445 Washington, D.C. 20036 Forked River, New Jersey 08731 J.6. Liberman, Esquire Commissioner Bishop, Liberman', Cook, et al. New Jersey Department of Energy 1155 Avenue of the Americas 101 Conteerce Street New York, New York 10036 Newark, New Jersey 07102 Mr. David M. Scott, Acting Chief Regional Administrator, Region I Bureau of Nuclear Engineering U.S. Nuclear Regulatory Commission Department of Environmental Protection 631 Park Avenue CN 411 King of Prussia, Pennsylvania 19406 Trenton, New Jersey 08625-BWR Licensing Manager HE. P. B. Fiedler GPU Nuclear Vice President & Director 100 Interpace Parkway Oyster Creek Nuclear Generating Parsippany, New Jersey 07054 Station 4: Post Office Box 388 Deputy Attorney General. Forked River, New Jersey 08731 State of New Jersey
- Department of Law and Public Safety
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s 36 West' State Street - CN 112 Trenton, New Jersey 08625 e
- Mayor Lacey Township 818 West Lacey Road Forked River, New Jersey 08731 Licensing Manager ,
' Oyster Creek Nuclear Generating Station Mail Stop: . Site Emergency Bldg.
P. O. Box 388
- Forked River, New Jersey 08731
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