ML19350F071
| ML19350F071 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/12/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-09-03, TASK-9-3, TASK-RR LSO5-81-06-050, LSO5-81-6-50, NUDOCS 8106240163 | |
| Download: ML19350F071 (14) | |
Text
_ _ _ _
June 12,1981 M
Docket No. 50-219 e
LS05-81-06-050 2
Mr. I. R. Finfrock, Jr.
5 2
Vice President - Jersey Central
\\N% IO84 Power & Light Company Post Office Box 288 K'
Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
FORWARDING DRAFT EVALUATION OF SEP TOPIC IX-3, STATION SERVICE AND COOLING WATER SYSTEMS, OYSTER CREEK Enclosed is a copy of our draft evaluation of SEP Topic IX-3, Station Service and Cooling Water Systems. This assessment compares your facility, as des-cribed in Docket Nc. 50-219, with the criteria currently used by the regulatory staff for licensing new facilities. Please infom us if your as-built facility differs from the licensing basis assumed in our assessment. Coments are re-quired within 30 days of the receipt of this letter so that they maybe included in our final report.
This evaluation will be a basic input to the integrated safety assessraent for your facility unless you identify changes needed to reflect the as-built con-ditions at your facility. This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are mod-ified before the integrated asaessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
Draft SEP Topic (I-3 cc w/ enclosure:
See next page 810 6 2 4 O/lM
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Docket No. E0-219 LS05-81-06-050 Mr. I. R. Finfrock, Jr.
'! ice President - Jersey Central Power & Light Coutpany Post Office Box 388 Forked River, New Jersey 0
Dear Mr. Finfrock:
SUBJECT:
FORWARDING DRAFT EVALUATION OF SEP TOPIC IX-3, STATIO:1 SERVICE AND COOLING WATER SYSTEMS, 0YSTER CREEK Enclosed is a copy of our draft evaluation of SEP Topic IX-3, Station Service and Cooling Water Systems. This assessment compares your facility, as des-cribed in Docket No. 50-219, with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment.
Comments are re-quired within 30 days of the receipt of this letter so that they maybe included in our final report.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built con-
- ditions 3t your facility. This assessment may be revised in the future if your facility design is changed or if MRC criteria relating to this subject are mod-ified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chi Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
Draft SEP Topic XI-3 cc w/ enclosure:
See next page l
Mr. I. R. Finf rock, J r.
Cc G. F. Trowbridge, Esquire Gene Fisher Shaw, Pittman, Pott 3 and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628-GPU Service Corporation ATTN: Mr. E. G. Wallace Commissioner Licensing Mar.nger New Jersey Department of Energy 260 Cherry Hill Road 101 Commerce Street Parsippany, New Jersey 07054 Newark, New Jersey 07102 Natural Resources Defense Council Plant Superintendent 91715th Street, N. W.
Oyster Creek Nuclear Generating Washington, D. C.
20006 Station P. O. Box 388 Forked River, New Jersey 08731 Steven P. Russo, Esquire 248 Washington Street Resident Inspector P. O. Box 1060 c/o U. S. NRC Tons River, New Jersey U8753 P. O. Box 445 Forked River, New Jersey 08731 Joseph W. Ferraro, J r., Esquire Deputy Attorney General Director, Criteria and Standards State of New Jersey Division Office of Radiation Progrars Department of Law and Public Safety '
(ANR-460) 1100 Raymond Boulevard Newark, New Jersey 07012 U. S. Environmental Protection Agency Ocean County Library Washington, D. C.
20460 Brick Township Branch 401 Chamoers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region II Office Mayor ATTN: EIS COORDINATOR Lacey Township 26 Fsderal Plaza P. O. Box 475 New York, New York 10007 Forked River, New Jersey 08731 Commissioner Department of Public Utilities State of New Jersey 101 Connerce Street Newark, New Jersey 07102
)
~
SEP REVIEW OF STATION SERVICE AND CCOLING WATER SYSTEMS TOPIC IX-3 FOR.THE C
OYSTER CPEEK NUCLEAR POWER PLANT e
I.
INTRODUCTION The safety objective of Topic IX-3 is to assure that the cooling water systems have the capability, with adequate margin, to meet design object-ives and, in particular, to assure that:
systems,are provided with adequate physical separation such that a.
there are no adverse interactions among those systems under any mode of operation; b.
sufficient cooling water inventory has been provided or that adequate provisions for makeup are available; c.
tank overflow cannot be released to the environment without monitor-ing and unless the level of radioactivity is within acceptable limits; d.
Vital equipment necessary for achieving a controlled and safe shutdown is not flooded due to the failure of the main condenser circulating water system.
II. REVIEW ' CRITERIA The current criteria and guidelines used to determine if the plant systems meet the topic safety objective are those provided in Standard Review plan (SRP) Sections 9.2.1, "Statio.n Service Water System," and 9.2.2, " Reactor Aux.iliary Cooling Water Systems."
In determining if plant design conforms to a' safety objective, use is made, where possible, of applicable portions of previous staff reviews.
For example. to verify that safety objective e.,
identified above, is met, staff review of the generic issue " Flood of Equip-ment.Important to Safety" is examined to assure that circulating water sys-tem failures are evaluated.
III.
R_E, LATED SAFETY TOPICS AND INTERFACES
~
Only those portions of the SRp which are not' covered by other reviews are applied in this review.
Therefore, this report does not address the foi-lowing areas which are reviewed under other SEp topics or staff generic reviews:
~
III Protection from missiles, pipe whip, jet impingement, and fires III-5 (physical separation),
3 IX-6.
IX-5 Ventilation Systems, 111-1 Quality and seismic classification, IV-1.A N-1 loop operation, III-7 Inservice inspection and testing, VI-10.A O
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ff2.
Adverse environmental phenomena (tornadoes, floods, etc.'),
VI-LO.B Sharing of systems.
f Topic IX-5 is dependent on this topic infonnation for its complstion.
IV.
REVIEW GUIDELINES In determining which systems to evaluste under this topic, the staff usec the definition of " systems important to safety" provided in Reference 1.
The definitten states ' systems important to safety are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary,I (2) the capability to shut down the reactor and maintain it in a safe cor -
dition, or (3) the capability to prevent, or mitigate the consequences of accidents that could result in potential offsite exposures co. parable to the guidelines of 10 CFR Part 100, " Reactor Site Criteria." This defini-ion was used to determine which systems or portions of systems we' e r
"esscntial." Systems or portions of systens which perform functions im-portant tn safety were considered to be essential.
It should be noted that this topic may require reevaluation if future SEP reviews of design basis events identify addtional coolir.g water systecs that are important to safety.
V.
EVALUATION The systems reviewed under this topic are the Turbir.e Building Closed Cooling Water System, Reactor Building Closed C:olir.g Water System, Ser-vice Water System, and Emergency Servi:e Water System.
It should be noted that the diesel ger.eratcrs for or. site snergency ' power have self-contained cooling.
Turbine Building Closed Cooling Water System
^
The Turbine Building Closed Cooling Water Systen (TE:CW) is a closed loo; system with two 50% capacity heat exchangers and thrse 505 capacity pumps The system contains a surge tank which provides net positive sucticn heac for the pumps and surge volume to accoicodate system fluid expansion and contracticn. The TBCCW system cools the fellowing equipment:
I 1.
Control room air conditioning unit i
2.
Air compressors (3) 3.
Air compressor after coolers (3)
I Reactor Coolant Pressure Boundary is defined in 10 CFR Part 50 a 50.2'(v).
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Reactor pumps ' motor generator ' oil coolers (5)
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Reactor feedwater pump oil coolers (3) 6.
Condensate pump motor coolers (3) 7.- Turbine lube oil coolers (2) 8.
Mechanical Condenser vacuum pump cooler 9.
Circulating water vacuum priming pump coolers (2)
- 10.. Generator stator, cooling heat exchangers(2)
- 11. Generator Hydrogen. coolers (6) 12.
Generato.r bus heat exchanger
- 13. Generator hydrogen seal-cil unit (presently disconnected) 14.. Office. building chiller flakeup water for the TBCCW system is supplied automatically from the de-nineralized water system through an air-operated valve controlled by a surge tank' level control system.' The system uses a chromate rust inhib-
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itor.
Under normal operating conditions, almost all the heat loads on the sys-tem are in operation. However, with the possible exception of control room air conditioning,:the TBCCW system is not required to perform any post-accident functions. The need for control room air conditioning will be determined under SEF Topic IX-5, " Ventilation Systems".
- The plant air systems are not required for safe plant shutdown or post-accident operation unless the control room air conditioning is required.
T3CCW flow may be lost to the remaining compcnents and systems supplied by the TBCCW system under both normal ano post-accident conditions; and, although operator action may be necessary to restore flow to continue plant operation, the consequences are of little safety concern.
Based on our review of the TBCCW system (subject to the findings of-the additional SEP reviews noted above), we have oetermined that the TBCCW system and plant air systems are not important to safety as defined in
~
Reference 1.
Reactor Building Closed Cooling Water System
~
The Reactor Building Closed Cooling Water System (RBCCW) is a closed icop system with two 100% capacity pumps and two 100% capacity heat exchangers.
The system also contains a surge tank which provides net pasitive suction head for the pumps and a surge volume to acccamodate system fluic thermal expansion and contraction. RBCCW pumps 1-1 and 1-2 receive power from 450 volt substations l A2 and 1B2, respectively. The RSCCW system cools the following equipnent:
1.
Cleanup system component::
a.
recire pumps (2) b.
auxiliary pump
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Drywell components:
a.
drywell cooling units (5) b.
reactor recirculation pwnps and motors (5) drywell equipment drain. tank cooling coil c.
3.
Reactor building. equipment drain tanks 4.
Radwaste building components:
a.
waste collector tank b.
recyclecoolerforprecoatpumps(2)
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concentrated waste storage tank
'd. ' waste concentrator components
- 1) condenser
- 2) condensate cooler 5.
Tunnel recirculation fans 6.
Shutdcwn cooling system:
a) beat exchangers(3) b) pumps (3) 7.
Fuel pool heat'exchangers(3) 8.
Control rod drive pumps thrust bearing and gea; box coolers The R9CCW system is not required to perform any post-accident functions.
The' system may be used to provide backup functicns for certain systems which are used for post-accident or safe shutdown functions.
No credit is taken for the containment drywell coolers in accident
. analyses for Oyster Creek and they are not necessary for safe plant shutdown. Therefore, RBCCW flow to the coolers is not essential.
The shutdown cooling system is not required to bring the plant to a safe i
shutdown or to achieve cold shutdown conditions (Referer.ce 2).
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l d by the RBCCW system (listed above) can RBCCW flow to the equipment coo e be lost 'under both normal _'an'd post-accident conditions; and, although operator action may be necessary to restore flow tc continue plant oper-ation,-the cUnsequences are of little safety concern.
Based on our review of the RBCCW system, subject to the findings of the.
SEP recirculation pump seizure DBE, we have determined that the RSCCW system is not important to safety as defined in Reference.1.
Service Water System The Service Water System (SWS) consists of-two pumps which take water from the plant intake structure and circulate it througn the RBCCW heat exchangers and, via strainers, through the Circulating Water pump bear-The SWS can also fre used to supply flow to the TBCCW heat exchang-
-ers if their normal source of cooling water, the Circulating Water System,'
ings.
is not op'erating. ~
Secause the Circulating Water System is not required for safe shutdown or post-accident operation and if the RECCW and TBCCW sy in Reference 1.
Emeroency Service Water System The Emergency Service Water System (ESWS) consists of four pumps (located ~
on the Circulating Water Intake Structure), two sets of two Containment
- Scray heat exchangers (located in the reactor building, NE and SE cor;ners, 23' elev.), and associated piping and valves.
!"'ectrical power for the pumps is supplied by 4160v ' Emergency Switchgear IC (pumps 52A, 523) and 4160v Emergency Switchgear 10 (pumps 52C, 520).
Following During normal plant operation, the ESWS is not in oper& tion.
and accident, the ESWS is actuated automatically on simultaneous signals On automatic of drywell high pressure and low-low reactor water level.
actuation, one pump in each loop.of containment spray and ESWS pump (See i
If any of the two con-Figure 1.) receives an automatic start signal.
tainment spray or two ESWS pumps fails to start it is annunciated in the'
[.
The operator can then start the manually operated pump i
control room.
that is in parallel with the failed pump (Reference 8).
The staff reviewed the heat removal requirements of the ESWS during post-The accidents considered were the loss-of coolant accident operation.
accident (LOCA) a M the steam line break (SLB) inside centainment becau these events result in the greatest potential accidsnt heat loads on the t
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Section VI-7 of Reference 3 discusse,s ESWS operation following a ESWS.
Energy is removed from the containment atmosphere by the contaia-
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ment spr1y system.
Energy is removed from the reactor core by the Core Spray (CS) system.2 The heat removed from the containment and/or the reactor core _is deposited in the' water. in the containment absorption cham-ber (torus), and.from there it is transferred to the ESWS through the containment spray /ESWS heat exchangers (Figure 1).
The minimum combination of -containment-spray and. ESNS pumps available oc-curs after a' postulated loss of offsite power-and the single failure of one of the emergency diesel. generators.
This minimum combination of equipment is two containment spray pumps and two ESWS pumps. The :ooling capability of this combination was analyzed in section XIII-2 of Refer-ence 3 (using the system design parameters renroduced in Table 1 of this report) for a reactor power level of 1860 MWt. This combination of pumps -
was determined to be adequate to acccamodate post-LOCA heat input to containment for plant operation at 1930 liMt, the current licensed full.
power level, in Reference 4.
During the initial period of energy release to containment.the heat is absorbed in various heat sinks inside containment --mainly the volume of water stored in the torus. The torus water is capable of. absorbing all the released energy during the initial blowdown phase of the LOCA_(sporox 60 seconds) with the water temperature increasing to about 130*F.3 The licensee's analysis 'shows that the post-blowdown energy release rate is equalled by the ESWS heat removal rate with' two containrant spray pumps and two ESWS pumps operating at approximately one hcur after,the. start of the accident. The staff's calculation of ESWS capability for the two containment spray /two ESUS pump case shows that the heat removal capacit;.
is sufficient to accomodate post-LOCA-heat loads at approximately twenty minutes after the start of the accident with a torus temoerature of 148':
and an AHS 5.1 reactor core decay heat rate of apprcximately 118E6' BTU /hr The most severe temperature transient in the drywell is caused by a small break discharging steam only which does not cause reactor system depressurization or automatic operation of reactor trip or ECCS. This SLB event, or a small LOCA which does not, lead to automatic ' initiation of reactor trip or ECCS, both depend upon operator acticn to initiate an orderly shutdown and cooldown of the plant to limit the quantity of energy released to the containment. The difference between a small LOCA or SLB and the large LOCA analyzed in Reference 3 from the stand-point of containment cooling, is the energy produced by the reactor t
4 2The ECCS functions following an accident are evaluated in the SEP l
Design Basis Event review of the LOCA.
i t
3The SEP t 411 reevaluate the post-accident energy balance in containma under Topic VI-2.D, " Mass and Energy Release for Postulated Pipe Bres Inside. Containment.
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. core which is available to the containment. The large LOCA results in a quick reactor shutdown (trip), and the energy produced in the core fotlowing trip is core decay heat which proceeds via the break to the containment.
For a very small LOCA or SLB, the energy available to the containment includes both the core decay heat and the energy which en-ters containment prior to the operatcr initiated reactor trip.
In Reference 5, the licensee has described the operator actions requir-ed to mitigate the very small LOCA or SLB. Plant emergency procedures direct the control room operatcr to manually initiate containment spray if a very small pipe-break causes containment pressure to exceed 10 psi and automatic spray actuation has not occured.
Since a reactor trip oc curs automatically at ? psig drywell pressure, the energy available tc the containmer.t after 2 psig has been reached is core decay heat.
The staff has c'alculated, using Reference 6, that a 0.0.2 ft2' SLB would
^
. add approximately 15E6 BTU to containmert prior to 10 psig.being reach-ed inside'containm'ent.
This energy would raise the torus water temper-
-ature about 3*F.
Since a reactor trip occurs at 2 psig in the drywell, the 1 mount of energy added to containment prior to reactor trip would raise the tcrus temperature much less than 3*F.
Therefore, from the standpoint of containment cooling, the effects of a small break LOCA ar approximately equivalent to a large break LOCA; and core decay heat is the dominant energy source in both cases.
Isolation of leaking ESWS components is acccmplished by securing the pumps and shutting the motor operated heat exchanger discharge valve ir the affected loop.
Leakage from the ESWS could result in floodi'ng con-ditions in the reactor building, turbine building, or intake structure.
The performance of an operating ESWS loop would not be significantly imoaired by a moderate energy line crack because the leakage rate calet lated in accordance wito Reference 7 (350 gpm) is a small fraction of loop flow rate (6,000 gpm).
ESWS flow rate is not indicated in the control room.
To determine prc-per ESWS performance, the control room operator has indication of the l
differential pressure from the containment sp~ ay side to the ESWS side r
of each containment spray heat exchanger (dPT IP05A, B, C, D and dPI IP06A,-B,C,D).
This is adequate indication of ESWS performance.
In-direct indication of ESWS performance can be obtained by containment spray loop temperature readings available in the control rcom.
Detection of radioactive leakage through the ESWS heat exchangers from the containment spray system is accomplished by a radiation monitor in the ESWS discharge line from each heat exchanger.
Isolation of a leak-ing heat exchanger is accomplished by shutting the motor operatsj valve in the diccharge line from the heat exchangers.
4 or a review of the effects of flooding, see SEP Topic III-5.B, " Pipe F
Breaks.0utside Containment" O
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Isolation of the ESWS from the 1 1/2" and 2" diameter SWS lines and t'h 3"'chlorjnation system line which. connect to the ESWS is accomplished by means of manual valves. The 2" diameter SWS line also contains check valves which would seat to prevent flow from the ESWS to the SWS.
A r' view of Licensee Event Re' ports shows no recurring problems with operation or maintenance of the ESWS.
CONCLUSION Based on our review of'the service and cooling water systems for Oyster Creek, we have concluded that the essential system and function is:
Emergency -Service Water System:
Torus heat removal.
We have' determined that the design of the above systems is in conform-ance with current regulatory guidelines and with Genera. Design Criterion (GDC) 44 regarding capability and redundancy of the essential functions of.
the systems. The systems also m.eet the requirements of GD0 45 and 46 re-garding system design to permit periodic inspection and testing.
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TABLE 1.
SYSTEM DESIGN PARAMETERS
- System / Reference Parameters
- Coiitainment Spray 4 pumps - 3,000 gpm each (Ref.~ 3, Section VI-7) 4 ESWS heat exchangers - 42.5E5 BTU /hr each (with 3,000 gpm :entainment spray @ 130*F and 3,000 gpm ESWS
@ 85'F)
Emergency Service 4 pumps - 3,000 gpm each Water (Ref. 3 Secticn VI-7)
IN Treismic-and Quality Group Classification data is provided in SEP Tepic III-1 O
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_ REFERENCES 1.
-Regulatory Guide 1.105, Systems Setpoints.
2.
SEP Review of Sa~fe Shutdown Systems for the Oyster Creek Nuclear Pcwer Plant (SEP Topics VII-3, V-10.8, V-11.A, V-ll.B. X).
3.
Oyster Creek Nuclear Power Plant Facility Description and Safety Analysis Report, transmitted with Amendment No. 3 by JCP & L letter
- 0. Rees to P. Morris, dated January 25, 1967.
4 License Amendment No. 3, to Provisional Operating License No. DPR-16 for the Oyster Creek Nuclear Plant dated November 5,1971.
5.
JCP & L letter, R. Sims to P. Morris, dated December 31, 1970, transmitting Amendment 65 to the Application for Construction Permit and Operating License for Oyster Creek Nuclear Power Plant.
6.
JCP & L letter, R; Sims to P. Morris, dated September S,1970, transmitting Amendment 62 to the Application for Ccnstruction Permit and Operating License for Oyster Creek Nuclear Power Plant.
7.
Eranch Technical Pocition ME3 3-1, appended to Standard Review Plan 3.6.2.
8.
SEP Technical Evaluation Topic VI-10.A, " Testing of Reactor Trip System 18 ar.d Engineered Safety Features" Oyster Creek, Docket No. 50-219, May 1981, E. W. Roberts l
i.
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