ML20031A524
| ML20031A524 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/17/1981 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20031A525 | List: |
| References | |
| TASK-04-02, TASK-09-03, TASK-4-2, TASK-9-3, TASK-RR FYR-81-136, NUDOCS 8109230622 | |
| Download: ML20031A524 (13) | |
Text
.
YANKEE ATOMIC ELECTRIC C00PANY 2.C.2.1 1671 Worcester Road, Framingham, Massachusetts 01701
= 8 -236
,Yauxes September 17, 1981 United States Nuclear Regulatory Commission Washington, D. C. 20555 A t ten tion :
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing
Reference:
License No. DPR-3 (Docket No. 50-29)
Subject:
SEP Topic Assessment Completion (Topics IV-2 & IX-3)
Dear Sir Enclosed please find our assessment of the following topics:
IV Reactivity Control Sys tems, Including Functional Design and Protection Against Single Failures.
IX Station Service and Cooling Water Systema,.
We trus t you will find this information satisfactory; however, if you have any ques tions, please contact us.
Very truly yours,
'KEE ATOMI ELECTRIC COMPANY d
J. A. Y-
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f Senior j neer - Licensing O
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SEP Review of Topic IX-3 Station Frvice and Cooling Water Systems _
Yankee Rowe e
I.
INIRODUCTION The safety objective of Topic IX-3 is to assure that the cooling water systms have the capability, with adequate margin, to meet their design objectives.
II. REVIBJ CRITERIA The current criteria and guidelines used to determine if the plant systes meet the topic safety objectives are those provided in Standard Review Plan (SRP) Sections 9.2.1, " Station Service Water Systs," and 9.2.2, " Reactor Auxiliary Cooling Water Systems."
III. REIATED SAFEIY TOPIG AND INTERFACES
'Ihe scope of review for this topic was limited to avoid duplication of efforts since some aspects of the review were performed under related topics. The related topics and the subject matter are identified below.
Each of the related topic reports contains the acceptance criteria ard review Suidar.ce for its subject matter.
II-2. A - Severe Weather Ihenomena II-3.B.1 - Flooding of Equipnent III-3.B - Flooding of Equipment (Failure of Underdrain System)
VI-7.D - Flooding of Equipont (long-Term Passive Failures)
III-3.C - Inservice Inspection of Water Control Structures III-4.C - Int;ernally Generated Missiles III Mass and Energy Releases (liigh Energy Line Break)
VI-2.D - Mass and Energy Releases III Seismic Qualification III Environmental Qualification VI-7.C.1 - Independence of Onsite Power VII Systm s Required for Safe Shutdown VIII Diesel Generators IX Fuel Storage IX Fire Protection IV.
EVALUATION The syt.ws which were reviewed under this topic are the Component Cooling Syst m and the Service Water System. These are the only two cooling water systms at Yankee. They supply a l cooling water l
requirements.
IV.1 Component Cooling Syst m lhe Component Cooling Systm (CCS) removes heat from various plant syst ms and components and transfers this heat to the Service Water Systm. An intennediate cooling systs is utilized to insure that l l g
there are no radioactive releases to the environment via the service water syst s.
We heat loads on the systs are:
1.
Main Coolant Pumps 2.
Neutron Shield Tank Cooling Coils 3.
Sample Cooler
- 4..
Waste Evaporator and Heat Exchanr.ars 5.
Waste Gas Compressor and Heat Exdmngers 6.
Shutdown Cooling Syst s Heat Exchc ger 7.
Iow Pressure Surge Tank Cooling f,ptem Heat Exchanger 8.
Spent Fuel Pit Cooling Systm Heat Exchanger During normal plant operation, one punp and one heat exchanger are in operation with the second punp in standby.
Each punp and heat exchanger have 1007. capacity, therefore the system has cmplete redundancy in these major cmponents.
CG punp No.1 is powered from 2400 volt bus No. 3, and CCS punp No. 2 is powered from 2400 volt bus No. 2.
Each 2400 volt bus is normally powered by separate off-site power supplies. Rese punps are not required innediately post-accident; therefore, they are not powered by emergency power. However, the electrical system can be realigned to power them post-accident.
Each heat exchanger is designed to remove the decay heat removal capacity of the Shutdown Cooling System (SCS) at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after a plant shutdown, 16.0 x 106 BTU per hour. He heat load during normal operation is only 8.5 x 106 BTU per hour.
E erefore, the system design limit exceeds the normal operating loads.
During post-accident conditions, the only heat load which is essential is the shutdown cooling system heat exchanger. D e SCS system is required for plant cooldown below 2000F. % e SCS system is designed to be placed in service,when the main coolant system temperature has been reduced to less than 3300F and the pressure to less than 300 psig for normal phnt operation. During post-accident operation, the plant would be cooled down as low as possible using the steam generators or the post-IhCA recirculation system prior to placing the SCS system in operation. Unen it is placed in operation, the heat loads will be greatly reduced, and therefore, the CCS system will be well within design limits.
%e low Pressure Surge Tank Cooling System (LPSTCS) beat exchanger is identical to the SCS system heat exchanger. He two systes are designed so that each heat exchanger can be lined up to function as a backup for the other.
%e main coolant punps are an essential load for normal plant operation.
Rese pumps are of the canned-tnotor type, requiring cmponent cooling water to an external cooling jacket which removes stator and rotor heat from the internal cooling water. Operation of the pumps is not permitted when conponent cooling water is lost, and this requires a plant shutdown. This event has been proceduralized and the plant can be safely shutdown with a loss of main coolant punps. Also, main coolant mp operation is not required for plant response to any postulated accidents..
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o, Based on this review, the only safety-related function of the CCS system is the SCS system heat exchanger, and it is not required until several hours into the accident. Also, if the CCS system is not available, provisions have been made to permit connecting firehoses to the fire system or the service water system to supply the CCS system loads.
IV.2 Service Water System The Service Water System (SWS) consists of three 2500 gpm pumps which take a suction from a common intake well in the Circulating Water Pump House. The pumps discharge into a common 12" header which branches into two 12" supply headers. These two headers can be manually cross-connected so that any combination of pumps can supply the necessary loads. Table I contains a listing of all SWS loads.
During normal plant operation, two pumps are running and the third pump is in standby.
On a low pressure in r* a common discharge header, the standby pump will start automatically, and an alarm will sound in the control room. The power supplies for SWS pumps No. 1, 2, and 3 are the station 2400 volt buses No. 3, 1, and 2, respectively. The electrical system can be realigned to power any pump post-accident from emergency sources if required.
The majority of the loads supplied by the SWS system are required for normal plant operation, and are not required to operate to mitigate post-accident plant conditions. The major heat loads renuired for post-accident plant recovery are the CCS system heat exchangers. As discussed in the CCS section, the system is designed to remove the plant decay heat removal capacity of the SCS system at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after shutdown.
This requires a SWS flow rate of 2500 gpm, or one pump. Cooling water to the SCS pump is also supplied by the SWS system. The LPSTCS pump is identical to the SCS pump and can be lined up to function as a backup.
It is also cooled by the SWS system. These pumps only require a few gpm for cooling.
Since only one pump is required to meet these heat loads, the other two pumps function as backups for this essential system function.
If the SWS system is not available, the capability exists to supply cooling water to these heat loads via the on-site fire system and existing fire hose l
connections.
l Based on this review of the SWS system, ;he only essential loads on the l
system are the CCS system heat exchanger for removal of plant decay heat I
via the SCS system, and shutdown cooling pump cooling. These functions are not required until several hours following plant shutdown.
V.
CONCLUSION The design of the above system is in conformance with current regulatory guidelines and with General Design Criteria (GDC) 44 regarding capability and redundancy of the essential functions of the systems. The systems also meet the requirements of GDC 45 and 46 regarding system design to permit periodic inspection and testing.
! i l'
VI.
REFERENCES 1.
Yankee Rowe Final Hazards Sucunary Report, Volume 1, Rev. I dated
. July 1, 1980.
2.
Plant Procedure No. OP-3115, Loss of component cooling.
3.
Plant Procedure No. OP-3009, Loss of Service Water.
k.....
g TABLE 1 Service Water System Loads Primat:, "lant Component cooling heat exchangers vapor container coolers sample coolers charging pump cooling low pressure surge tank cooling pump cooling shutdown cooling pump cooling Secondary Plant traveling water screen washing water treating plant turbine oil cooling generator hydrogen cooling main transformer cooling air compressors cooling j.
boiler feed pumps cooling heater drain pumps cooling
' circulating water vacuum priming pump seal water sample coolers in turbine area j
i i
1 4
4. -
SEP Tepic IV-2:
Reactivity Control Systens, Including Functional Design and Protection Against Single Failures I.
INTRODUCTION The purpose of this evaluation is to ensure that the design basis for the Yankee reactivity control systens is consistent with analyses performed to verify that the protection systen neets General Design Criteria (GDC) 2 5.>
GDC 25 recuires that the reactor protection systen be designed to assure tl.at specified acceptable fuel design limits are not exceeded for any single nalfunction of the reactivity control systens, such as accidental withdrawal of control rods.
Reactivity control systens need not be single failure proof.
However, the protection systen nust be capable of assuring that acceptable fuel design limits are-not exceeded ir the event of a single failure in the reactivity control systens.
This report identifies and evaluates consequences of single failures in the electrical circuits of the reactivity control systens. Related topics not covered in this report but needed for the completion of this topic are SEP Topics XV-8 (" Control Rod Misoperation - Systen Malfunction or Operator Error *, and XV-10 ("Chenical and Volume Control Systen Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant") (Topic XV - 10 was submitted in Reference 3).
II. EVALUATION A.
Single Failure Analysis An evaluation of the effects of a single failure on the Yankee rod control systen has been performed. The results of this analysis are presented in Table 1.
Occurences of any of the listed events initiated as the result of a single failure will be annunciated by alam in the control roon. A description of these alarns and a description of the systen design features which limit rod malpositions and prevent inadvertent reactivity insertions is provided below.
B.
Design Features The following design features limit rod malpositions and limit reactivity insertion resulting fron a single failure:
1.
The rod noverent switch must be returned to the "0FF" position after each rod withdrawal step taken when generator load is more than 130 MWe.
Should a failure in this circuit occur, rod nalposition would be limited to one rod step (3/8 of an inch) (Reference 4).
2.
Rod selector lights on the rod position indication board indicate which rods or groups have been selected. Monitoring of the rod position lights gives an indication of rod positioning.
(Reference 4).
3.
The secondary rod position indication systen indicates the proper operation of the rod drive can, can notor, and speed reducer by sensing electrical current from the can drun and displaying the can novement in the control roon (Reference 4).
i 4.
The T-average circuit will insert the controlling rod group at six (6) inches per ninute should T-average exceed the circuit setpoint by approxinately 20F (Reference 5).
5.
The High Start-up Rate /High Neutron Flux Ale m functions are available to alert thc operator in time to pemit prompt operator actin in order to avoid attainnent of conditions that could produce scrau.
(Reference 6).
C.
Safety Analysis Events more limiting thal the single failure events cited above are evaluated in each core cycle reload analysis.
These events are:
(1) uncontrolled group rod withdrawal, (2) single rod drop, and (3) single rod eje
- i. Most recently the Core XV Performance Analysis Gscribes a nalyse 1enonstrating conpliance with applicable design criteria for these events.
In addition, evaluations were nade of Core XI single rod withdrawals confiming that DNB ratios exceeded 1.30 in all cases.
Sinilar results can be expected for Core XV based upon a comparison of respective peaking factors, scram worth and noderate tenperature coefficient. The nethods used for the Core XI scoping calculations of a single rod withdrawal event were approved by the NRC and are essentially the same netnods used for Core XV licensing calculations.
Protection against the above single failure events is provided by automatic scran on a high power trip signal, without creditine operator intervention to terminate the event in advance cf scran.
Dractor scrans are also initiated for high or low nain coolant pressure and high pressurizer level indications.
In all cases, the substantial steady state operating nargin to DNB, manifest in the design-value DNB ratios near 3.30, ensures that no violation of Condition II design criteria occurs. This criteria recuires that DNB ratios exceed 1.30 for these single failure events.
Furthemore, other than for surveillance testing required by Technical Specifications, individual control rods are not routinely withdrawn during steady state operation within the power range. Operator nanipulation of a single rod in violation of the Power Dependent Insertion Linit Curve, Technical Specifications Figure 3.1-1, is unlikely during operations in the power range.
During low power physics testing, usually conducted at approximately 2% ratcd power level, when rod worth i
neasurenents require individual rod control, the low power scran switch l
1s set to LOW (neutron flux trip reduc !d froni 108% to d 35%) and the high startup rate trip at 6 5.2 DPM is available to limit the consecuence of single failures in the rod drive centrol systen.
In contrast, analyses of group rod withdrawal aceidents and rod ejection accidents are allowed to progress to the high power trip condition (108%) for purposes of calculating limiting DNB istios.
L Analysis perfomed since Core XI using currently approved nethods demonstrates DNB ratios greater than 1.30 for the group rod withdrawal and single rod drop events. An Elam systen is designed to sense a dropped rod and to alert the operetor; however, DNB ratios will exceed 1.30 without the need for a direct reactor trip when a rod drops. The group rod drop event exhibits less liniting consequences than the single.-.
F.
3 -
rod drop because local feel power peaking is comparatively lower. These events are discussed in the Core XV Perfomance Analysis (Reference 2).
Rod ejections, which are a type of limiting single rod withdrawal event, were discussed in the SU Topic XV-12 submittal (Reference 3). The analysis of the limiting fault rod ejection events denonstrates that no fuel pin failures occur and no design criteria are exceeded due to large thenaal nargins manifest in the steady state design-value DNB catios near 4
3.30.
The _large nargin to DNB at steady-state conditions discussed in Reference
? ensares that the results for the less limiting single failure cyents will satisfy design criteria for Condition II coderate frequency events..
Both the group rod withdrawal and single rod drop analyses bound the consecuences of the single failure events for Topic IV-2, including control rod mispositioning and rods failing to move upon conman.f.-
D.
Operating Features 1
In addition to those design features described in Section II.B. there are certain alams, interlocks, operating procedures, and protection system actions which nust be considered in limiting the consequences-following a single failure within systems used for reactivity control.
(a) Alams -- The first indication of an event caused by one of the postulated single failures within the reactivity control systems will be an alam condition in the control room.
l The following alam conditions are annunciated in the control room:
i 1.
Dropped rod 2.
High Start-up rate 3.
High-Low T-average 4.
High Flux Level 5.
Power P,ange Loss of Power 6.
Can Motors Loss of Voltage j
7.
Reactor Scran i
(b)
Interlocks -
(1)
The rod movement circuit provides an interlock to limit rod movenent when generator load is above 130 awe.
(See Section II.B)
(2)
The rod movement circuit provides.an interlock to limit rod movenent when the start up rate exceeds 1.5 DPM at any power l evel.
i (c) Operating Procedures - Pr'ocedures have been reviewed to verify that adequate-provisions have been made to ensure that proper operator actions are taken in response to the postulated events. Operator.
actions in accordance with these procedures will prevent'the prolongation of any single failure event.
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(d) Protection systen actions - A reactor :ceda is an automatic plant protection systen action which can be activated hy the above single failures.
Movenent of tha controlling rod group by the T-average circuit is described in Section II.B.
III. CONCLUSION Based !pon bounding calculations performed by Yankee, fuel design limits are not exceeded for any of the postulated events caused by single failures within the reactivity control systens, and General Design Criteria 25 is net with regards to such failures within the reactivity control systens.
IV.
REFERENCES (1) Letter, USNRC to YAEC, dated April 24, 1981.
(2) Letter, YAEC to USNRC, Core XV Refueling, Enclosure D, 28 March 1981.
(3) Letter, YAEC to USNRC, FYR 81-95, June 30,1981.
(4) YAEC Drawing 517F076 (Enclosure A).
(5) YAEC Drawing - T-average circuit diagran (Enclosure B).
(6) YAEC Drawing 601J897 (Enclosure C).
1 1
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1.
i TABLE 1 SINGLE FAILURE ANALYSIS - ROD CONTROL SYSTEM A Single Rod SYSTEM EFFECT A Single Rod May Be An Entire Rod An Entire ROD UPON SINGLE A Single May Not Move Ina dver ten tly.
An Enti re Group May Not Rod Group All Rods CONTROL FAILURE Rod May When Commanded Moved Or Group Move When May Be May Be SYSTEM COMPONENT Drop To Malpositioned May Drop Commanded To Malpositioned Ma1 positioned a.
Sta tionary Gripper X
con tactor
(
b.
Stationary Gripper coil X
I control contactor c.
Moveable Gripper contactor X
l d.
Moveable Gripper coil.
X control contactor c.
Lif ting Coil contactor X
f.
Lif ting Coil coil control.
X con tactor g.
Coil fuse X
X X
l l
h.
Stationary Gripper coil X
l l
1.
Moveable Gripper coll' X
X l
- j. Lif ting Coil coil X
k.
Stationary Gripper variable X
resistor 1.
Stationary Gripper vsriable X
X resistor cutout relay m.
Rod select switch (S1 to X
X' S24) n.
Rod Movement switch (S27)
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