ML18018A779
| ML18018A779 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/30/1983 |
| From: | Mcduffie M CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| LAP-83-447, NUDOCS 8310070120 | |
| Download: ML18018A779 (52) | |
Text
REGULATORY FORMATION DISTR1BUTION SY N (RIOS)
AC~ESSPt)N NBR;8310070120 DOC+DATE! '83/09/30 NOTARIZED: NO DOCKET FACIL:50-440 Shearon Harris Nuclear Power PlantE Uni,t 1i Carolina 05000400 50 401 Shearon "
Harris Nu'clear Power Plantr Unit 2< Carol'ina 05000401 AUTH, NAME AUTH()R AFFILIATION NCDUFFIEiN ~ A, Car ol ina,Power 5 Light Co, ItECIP ~ NAME RECIPIRN f AFFILIATION DENTONrHOR ~ Office of Nuclear 'Reactor Regulationi Director
SUBJECT:
Forwards response to request for addi inforincluding response to Rev 1,to ."Basis for Auxiliary Feedwater Sys Flow Requirementsr" in support of safety review. Response iqcludes info re main steam turbine valve testing issue, DISTRIBUTION CODE:
TITLE: Licensing BOO IS COPIES RECEIVED:LTR Submittal: PSAR/FSAR Amdts 8,
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CRT, Carolina Power 8 Light Company SERIAL: LAP-83-447 SEP 801S83 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION
Dear Mr. Denton:
Carolina Power & Light Company hereby transmits one original and forty copies of additional information requested by the NRC as part of the safety review of the Shearon Harris Nuclear Power Plant. Some of the enclosed responses relate to new issues. The cover sheet of the attachment summarizes the related Open Item or the new issue addressed in the attachment along with the corresponding review branch and reviewer for each response.
We will be providing responses to other requests for additional information shortly.
Yours very truly, M. A. McDuffie Senior Vice President Nuclear Generation FXT/ccc (8010FXT)
Enclosure cct Mr. B. C. Buckley (NRC) Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP) Dr. Phyllis Lotchin Mr. J. P. O'Reilly (NRC-RII) Mr. John D. Runkle lir. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Daniel F. Read (CHANGE/ELP) Mr. G. 0. Bright (ASLB)
Mr. R. P. Gruber (NCUC) Dr. J. H. Carpenter (ASLB)
Chapel Hill Public Library Mr. J. L. Kelley (ASLB)
Wake County Public Library 8310070120 830930 pDR Aoocx asaooooa i A PDR Q 411 Fayetteville Street o P. O. Box 1551 o Raleigh, N. C. 27602
~
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LIST OF OPEN ITEMS/NEW ISSUES, REVIEW BRANCH AND REVIEWER Auxiliary System Branch/N. Wagner Open Item 225(4)
Instrumentation and Control Systems Branch/H. Li Open Items 78, 80, 86 Mechanical Engineering Branch/D. Terao Open Item 286 Power 'Systems Branch/E. Tomlinson Open Items 118, 126, S. R. Question 430.60, Main Steam Turbine Valve Testing Issue
Auxiliary System Branch/N. Wagner Open Item 225(4)
I Shearon Harris Nuclear Power Plant Draft SER Open Item 225(4)
(Supplemental Information)
Provide a response to the NRC's March 10, 1980 letter to near-term operating license applicants concerning the AElS design (NUREG-0737, Item II.E.l.l). The response should include the following'.
(4) A discussion of the design basis for the ASS flow requirements and verification that the AFWS will met these requirements (see Enclosure 2 of the March 10, 1980 letter).
Response
In a letter dated August 12, 1983, Carolina Power & Light Company responded to this item noting that additional information would be available later. Attached please find Revision 1, to our response entitled "Basis for Auxiliary Feedwater System Flow Requirements" which incorporates the additional information provided by Westinghouse.
(7 929NEClcv)
CAROLINA POWER & LIGHT COiMPANY SHEARON HARRIS NUCLEAR POWER PLANT UNITS 1 AND 2 RESPONSE TO ENCLOSURE 2 OF NRC LETTER OF MARCH 10, 1980 "BASIS FOR AUXILIARY FEEDWATER SYSTEM FLOW REQUIREMENTS" REVISION 1
'I I I
, Question 1 A. Identify the plant transient&and accident conditions considered in establishing ASS flow requirements including the following, events:
- 1. 'Loss of main feed (LMFW) 2~ LMFV with loss of off-site AC power
- 3. LMFW with loss of, on-site and off-site AC power 4~ Plant cooldown
- 5. Turbine trip with and without bypass
- 6. Main steam isolation valve closure
- 7. Main feedline break
- 8. Main steamline break
- 9. Small break LOCA
- 10. Other transient or accident conditions not listed above.
N
.B. Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.
The acceptance criteria should address plant limits, such as:
- 3. RCS cooling rate limit to avoid excessive coolant shrinkage
- 4. Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool down the primary system.
Res onse to 1A The Auxiliary Feedwater System serves's a backup system for supplying feedwater to the secondary side of the steam generators at times when the feedwater system is not available, thereby maintaining the heat sink capabilities of the steam generatorS, As an Engineered Safeguards System, the Auxiliary Feedwater System is directly relied upon to prevent core damage and system overpressuri"ation in the event of transients, such as loss of normal feedwater or a secondary system pipe rupture and to provide a means for plant cooldown following any plant transient.
Following a reactor trip, decay heat is dissipated by'vaporating water in the steam generators and venting the generated steam either to the condensers through the steam dump or to the'tmosphere through the steam generator safety valves or the power-operated relief valves. Steam generator water inventory must be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal px".. =ss. The water level is maintained under these circumstances by the Auxiliary Feedwater System, which delivers an emergency water supply to the steam generators. The Auxiliary Feedwater System must be capable of functioning for extended periods allowing time either to restore normal feedwater flow or to proceed with an orderly cooldown of the plant to the reactor coolant temperature where, the Residual Heat Removal System can assume the burden of decay heat removal. The Auxiliary Feedwater System flow and the emergency water supply capacity must be sufficient
'o remove core decay
~
heat,
~
reactor coolant pump heat, and sensible heat during the plant cooldown. The Auxiliary Feedwater System can also be used to maintain the steam generator water levels above the tubes following "a'LOCA. Xn the, latter function, the water head in the steam generators serves as a barrier to prevent leakage of fission produces from the Reactor Coolant System into the secondary plant.
DESIGN CONDITIONS The reactor plant conditions, which impose safety-related performance requirements on the design of the Auxiliary Feedwater System, are as follows for the Shearon Harris Nuclear Power Plant Units 1 and 2:
Loss of main feedwatez transient with off-site power available
- without off-site power available Secondary system pipe ruptures
- feedline rupture
- steamline rupture
- Loss of coolant accident (LOCA)
- Loss of all AC power
- Cooldown Loss of Main Feed'water Transients The design loss of main feedwater transients are those caused by:
Interruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system Loss of'ff-site power or blackout with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater tzansients are characterized by a rapid reduction in steam generator water levels which results in a reactor trip, a turbine trip, and auxiliary feedwater actuation by the protection system logic. Following reactor trip from high power, the power quickly falls to decay heat levels. The water levels continue to decrease progressively uncovering the steam generator tubes as decay heat is transferred and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety or powez~perated relief valves to the atmosphere. The reactor coolant temperature increases as the residual'heat in excess of that dissipated through the steam generators is absorbed. Mith increased temperature, the volume of reactor coolant expands and begins filling the pressurizer.
Qithout the addition of sufficient emergency feedwater, further expansion will result in water being discharged through the pressurizer safety and relief valves. If the temperature rise and the resulting volumetric
I I
~ ~
'xpansion of the primary coolant are permitted to continue, then 1) pressurizer safety valve capacities may be exceeded causing overpressuri-zation of the Reactor Coolant System and/or 2) the continuing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventuall'y in core uncovexing, loss of natural circulation, and core damage. If such a situation were ever to occur, the Emergency Core Cooling System would be ineffectual because the ry coolant pr imary coo system pressure exceeds the shut-off head of the safety injection system pumps, the nitrogen overpressure in the accumulat ator tanks, and the design pressuxe of the Residual Heat Removal Loop.
Hence, the timely introduction of sufficient emergency feedwater is necessary to ax'xest the decrease in the steam generator water levels to reverse the rise in reactor coolant temperature, to prevent the pressurizer from filling to a water solid condition, and eventually to establish stable hot standby conditions. Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satisfactori y coxrected.
The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equipment. Th e 1 oss o po~er f p to the electric-driven condenser circulating water pumps results in a loss of condenser vacuum and condensex dump valves. Hence, .steam formed by decay heat is relieved through the steam generator safety valves or the power-operated relief valves. The calculated transients are similar for both the loss of main feedwater and the blackout except that reactor coolant pump heat input is not a consideration in the blackout transient following loss of power to the reactor coolant pump bus.
Seconda S stem"Pi e Ru tures The feedwater line rupture accident not only'esults in the loss flow to the steam generatoxs, but also results in the complete of'eedwater 1 d of one steam generator within a short time if the rupture should occur downstream of the last nonreturn valve in the main feedwater i di id 1 steam generator. Another significant result of a feedline rupture may be the spilling of auxiliary feedwater from tthee faulted steam generator. t This could result in the pumping of a disproportionately i
large fract on o f th e total emergency feedwater flow to the faulted referentiall steam generator and out the break because the system preferent a y pumps water to the lowest pressure steam generator rather than to the effective steam generators which are at relatively high pressure. The i
system d es gn mus t allow for terminating, limiting, or minimizing that ti frac on of emergency feedwater flow which is delivered to a au te loop in order to ensure that sufficient flow vill be delivere d to the remaining effective steam generator(s). The concerns are similar for the main feedwater line rupture as those explained for the loss of main feedwater transients.
,Hain steamline rupture accident conditions are characterized initially b lant cooldown, and for breaks inside containment, by increasing containment pressure and temperature. Auxiliary feedwater is no t needed during the early phase of the transient,. but flow to the faulted loop will contribute to the release of mass and energy to containment. Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop. Eventually, however, the Reactor Coolant System will heat up again and emergency feedwater flow ll be fred to be delivered to the unfaulted loops but at somewhat lower rates than for the loss of feedwater transients described p e Provisions must be made in the design of the Auxiliary Feedwater System limit control or terminate the auxiliary feedwater flow to the faulted loop as necessary in. order to prevent containment overp ressuriza- s tion foll@wing a steamline break inside containment and to ensure the minimum flow to the remaining unfaulted loops.
Loss-of-Coolant Accident (LOCA)
The loss of coolant accidents do not impose on the emergency feedwater t
system ny flow any o requirements in addition to those required by the other res e in this response.
acc id entss addressed The following description o small LOCA is provided here for the sake of completeness to explaainn thee role 'of the Auxiliary Feedwater Sy'tem in this transient.
4 Small LCCAs are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume. The principal contr ibu tion on from the Auxiliary Feedwater System following such small LOCAs is basically the same as the system s function during hot sh ll 1
f win a spurious safety injection signal which trips the reactor.
Maintaining a water level inventory in the secondary side of e generators provides a heat sink for removing decay heat and establishes the capaabilitty for or p roviding a buoyancy head for natural circulation.
The Auxiliary Feedwater System may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the reactor to a cold shutdown condition.
Cooldown Th e coo ldo wn function performed by the Emergency Feedwater System is a i
partia 1 one s nce the reactor coolant system is reduced from norma 1 zero load oa ternempeeratures to a hot leg temperature of approximate tel y 350'F. The latter is the maximum temperature recommended for placing n the Residual Heat Removal System (RHRS) into service. The RHR system completes the cooldown to cold shutdown conditions.
Cooldown may be required following expected transients following an accident suc as a ma n feedline break, or during a normal cooldown prior to refueling or performing reactor plant maintenance. ce. If the tri ed following extended operation at rated power level, the EFS is capable of delivering sufficient emergency ee wa e 4
remove decay h eat an d reac tor o coolant pump (RCP) heat following reactor trip .while maintaining the steam generator (SG) water level. Followin o ow ng transients or acc id en t s, the recommended cooldown rate is consistent with expecte d nee d s an d at the same time does not impose additiona 1 requirements on t h, e, capac itieses of the emergency feedwater pumps considering a single failure. In any event, the process consists of being aablee to dissipate plant 1 sens ible heat ea in addition to the decay heat produced by the reactor. core.
Loss of all AC Power The loss of all AC power is postulated as resulting from accident conditions wherein not only on-site and off-site AC power is lost but a 1 so AC emergency e power is lost as an assumed common mode failure.
B tt ry power for operation of protection circuits is assumed availaable.
e.
The impact on the Auxiliary Peedwater System is the necessity forr providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown until AC power is restored.
Res onse to 1B Table lB-1 summarizes the criteria, which are the general design bases for each event discussed in the response to Question 1A above. Specific assumptions used in the analyses to verify that the design bases are met are discussed in response to Question 2.
The primary function of the Auxiliary Feedwater System is to provide sufHcient heat removal capability for heatup accidents following reactor trip to remove the decay heat generated by the core and prevent sys tern overpressurization. Other plant protection systems are designed to meet, short-term or pretrip fuel failure criteria. The effects oof excessive coolant shrinkage are bounded by the analysis of the rupture of a main, steam pipe transient. The maximum flow requirements determined by other bases are incorporated into this analysis resulting in no additional flow requirements..
~
TABI.F. 1B-1 CRITERIA FOR AUXILIARY FEEDNATER SYSTEM DESIGN BASIS CONDITIONS CONDITION OR TRANSIENT CLASSIFICATION CRITERIA* ADDITIONAL DESIGN CRITERIA Loss of main feedwater Condition II Peak RCS pressure not to exceed design No water released through the pressure. No adverse effects on the pressurizer power-operated relief core. valves or safety valves.
Station Blackout Condition II Same as LMFM. Same as LMFM.
Steamline Rupture Condition IV DNB design basis is met assuming most reactive RCCA stuck in fully withdrawn position. 10CFR100 dose limits not exceeded.
Feedline Rupture Condition IV RCS design pressure not exceeded. Core does not uncover.
10CFR100 dose limits not exceeded.
Loss of all AC power N/A Note 1 Same as blackout assuming turbine driven pump operation.
I.oss of Coolant Condition III 10CFR100 dose limits 10CFR50 PCT limits Condition IV (Same as for Condition III)
Cool down N/A 100'F/Ilr.
556'F to 350'F
- Ref:, ANSX N18.2 (Thi information provided for thos ransients performed in the FSAR).
Viote 1: Although this transient establishes the basis for auxiliary feedwater pump powered by a diverse power source, this is not evaluated relative to typical criteria since multiple failures must be assumed to postulate this transient.
estion 2 Describe the analyses and assumptions and corresponding technical justification used with plant condition considered in lA above including:
A. Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.
B. Time delay from initiating event to reactor trip.
C. Plant parameter(s) which initiates AFWS flow and time delay between initiating event and introduction of APWS flow into steam generator(s).
D. Minimum steam generator water level when initiating event occurs.
E. Initial steam generator water inventory and depletion rate before and after AFWS flow commences identify reactor decay heat rate used.
F. Maximum pressure at which steam is released from steam generator(s) and against which the AFW pump must develop sufficient head.
G. Minimum number of steam generators that must receive AFW flow; e.g., one out of two'wo out of fours H. RC flow conditi.on continued operation of RC pumps or natural circulation.
I. Maximum AFW inlet temperature.
Following a postulated steam or feedline break, time delay assumed to isolate break and direct AFW flow to intact steam generator(s). AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level. Also identify credit taken for primary system heat removal due to blowdown.
K. Volume and maximum temperature of water in main feed lines between steam generator(s) and AFWS connection to main feed li,ne.
L. Operating condition of steam generator normal blowdown following initiating event.
M. Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
N. Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.
Response to 2:
The limiting transients which define the AFWS performance requirements for Shearon Harris are as follows:
Loss of main feedwater (station blackout)
Rupture of main feedwater pipe Rupture of a main steam pipe inside containment Plant Cooldown Loss of Main Feedwater (Blackout)
A loss of feedwater assuming a loss of power to the reactor coolant pumps was performed in FSAR Section 15.2.6 for the purpose of showing that for a station blackout transient, two auxiliary feedwater pumps delivering flow to three steam generators does not result in filling the pressurizer. Furthermore, the peak RCS pressure remains below the criterion for Condition lI transients and no fuel failure occurs (refer to Table 1B-l). Table 2-1 summarizes the assumptions used in this analysis. The transient analysis begins at the time of reactor trip. This can be done because the trip occurs on a steam generator level signal; hence, the core power, temperatures, and steam generator level at time of reactor trip do not depend on the event sequence prior to trip. Although the time from the loss of feedwater until the reactor trip occurs cannot be determined from this analysis, this delay is expected to be 20 to 30 seconds. The analysis assumes that the plant is initially operating at 102% of the Engineered Safeguards Design (ESD) rating shown on the table, a very conservative assumption in defining decay heat and stored energy in the RCS. The reactor is assumed to be tripped on low-low steam generator level allowing for level uncertainty.
Rupture of Main Feedwater Pipe:
The double-ended rupture of a main feedwater pipe downstream of the main feedwater line check valve is analyzed in FSAR Section 15.2.8.- Table 2-1 summarizes the assumptions used in this analysis. Reactor trip is assumed to be actuated by low-low level in the affected steam generator. This conservative assumption maximizes the stored heat prior to reactor trip and minimizes the ability of the steam generator to remove heat from the RCS following reactor trip due to a conservatively small total steam generator inventory. As in the loss of normal feedwater analysis, the initial power rating was assumed to be 102% of the ESD rating. The Shearon Harris auxiliary feedwater design is assumed to supply a total of 430 gpm to the two intact steam generators, including allowance for feeding the affected steam generator. The criteria listed in Table 1B-1 are met.
(7929NEClcv)
Blloture of a ifain Steam Pipe Inside Containment Because the steamline break transient is a cooldown, the AFS is not needed to remove heat in the short term. Furthermore, addit'n oi excessive emergency feedwater to the faulted steam generator will affect the peak containment pressure following a steamline break inside containment. This transient is performed at four power levels for several break sizes. Emergency feedwater is assumed to be initiated at the time of the break independent of system actuation signals. The maximum flow is used for this analysis. Table 2-1 summarizes the assumptions used in this analysis. The criteria stated in Table 1B-1 are met.
This transient establishes the maximum allowaole emergency feedwater flow rate to a single-faulted steam generator assuming all pumps operating; establishes the basis for runout protection, if needed; and establishes layout requirements so that the flow requirements may be met considering the worst single failure.
Plant Cooldown
>hximum and minimum flow requirements from the previously discussed transients meet the flow requirements of plant cooldown. This operation, however, defines the basis for tankage size based on the required cooldown duration, maximum decay heat input, and maximum stored heat in the system. As previously discussed. in Response 1A, the emergency feedwater system partially cools the system to the point where the KHRS may complete the cooldown; i.e.,
350'F in the RCS. Table 2-1 shows the assumptions used to determine the cooldown heat capacity of the emergency feedwater system.
The cooldown is assumed to commence at 102% or engineered safeguards design power; and maximum trip delays and decay heat source terms are assumed when the reactor is tripped. Primary metal, primary water, secondary system metal, and secondary system water are all included in the stored heat to be removed by the AFS ~ See Table 2-2 for the items constituting the sensible heat stored in the ifSSS.
This operation is analyzed to establish minimum tank size requirements for emergency feedwater fluid source which is normally aligned.
{ 7 929ilEC1 cv)
Summary of Assumptions Used in AFWS Design Verification Analyses Loss of Feedwater Ilain Steamline Break Transient (station blackout) Cooldown Hain Feedline Break (containment)
- a. Hax reactor power 102>> of ESO rating 2910 Mitt 102'S of ESD rating 0, 30, 70, 102K of (102" of 2910 IIMt) (102>> of 2910 HWt) rated (percent of 2785 IIMt)
- b. Time delay from 46.6 sec** 24.5 sec" variable trip sec*'A event to Rx
- c. AFWS actuation sig- Low Low SG level Low Low SG level/ Assumed immediate1y nal/time delay for 1 minute 1 minute AFMS flow
- e. initial SG inventory 110,000 ibm/SG 106,000 1 txn/SG. 99,959 ibm/SG (Faulted) 0% - 163000 ibm/SG 09,699 ibm/SC (intact) 30% - 150000 ibm /SG 70% - 133000 ibm/SG 1021, - 120000 ibm/SG Rate of change before See Attached tl/A See Attached Figure 2 tt/A 6 after AFWS actuation Figure Decay heat AHS + 20% AHS 4 20"m AIIS + 20% AtIS + 20K
- f. AFM pump design 1236 psia 1236 psia 1236 psia H/A pressure
- g. Hinimum I of SGs 2 of 3 2of3 H/A
- which must receive AFW flow
- h. RC pump status Tripped at reactor One pump All operating All operating trip operating
- i. Maximum AFW 120'F 120 "F 120oF Equal to main FM temperature temperature
- j. Operator action none 30 min. Hone - Automatic AFM isolation at 200 sec.
- k. AFM purge volume/ 511.5 ft3/at tlone 160 6 ft3/at tlone and temperature FW temperature FW temperature
- 1. tlormal blowdown none assumed none assumed none assumed none assumed
- m. Sensible heat see coo)down>> Table 2-2 see cooldown* H/A
- n. Time at standby/time see cooldown 2 hr.'4 hrs see cooldown H/A to coo ldown to RHR 0, AFM flow rate 400 gpm &4'ariable 430 <Jpm 1500 gpm tlo thick metal heat transfer assumed. "'wo seconds d lay tiipe assumed from I'.eactor Trip Signal to Rod Motion.
1 I aes Applies to gtation Blackout, flow requirement for Loss of Feedwater is 475 gpm.
SVi'PNMARY OF SENSIBLE HEAT SOURCES Primary water sources (initially at engineered safeguards design power temperature and inventory)
RCS fluid
- Pressurizer fluid (liquid and vapor)
Primary metal sources (initially at engineered safeguards design power temperature)
Reactor coolant piping, pumps, and reactor vessel
~ Pressurizer Steam generator tube metal and tube sheet Steam generator metal below tube sheet Reactor vessel internals Secondary water sources (initially at engineered safeguards design power temperature and inventory)
Steam generator fluid (liquid and vapor)
Secondary metal sources (initially at engineered safeguards design power temperature)
All steam generator metal above tube sheet, excluding tubes uestion 3 Verify that the A'ref pumps in your plant will supply the necessary flow to the steam generators as determined by Items 1 and 2 above considering a single failure. Identify the margin in sizing the pump flow to allow for pump recirculation flow, seal leakage, and pump wear.
Response to 3 The AFM pumps in the Shearon Harris AFWS design will supply the required flow to the steam generators considering single failure. The two 100X capacity motor-driven pumps are sized for 450 gpm each and are capable of delivering 490'pm. 'he turbine-driven pump (200K capacity) is sized to deliver 900 gpm. Thus, for Condition IV events, the AFS has the capability of supplying 200% of the required flow even with a failure of the largest pump.
120vOO.
100000.
80000.
60000.
40000.
20000.
0.
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D D
D CV D~
DC DC
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CLI TtHE (SEC}
SHA~ROiV HARRIS NO. 1 STATION BLACKOUT (FIGURE 1)
120000.
~Loop 1(Faulted SG) 100000.
80000.
60000.
/ Loop 2 5 3 (Intact SG's) 40000.
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SHEARON HARRIS NO. 1 NIN FEEDLINE HRE.~K (FIGURE 2)
Instrumentation and Control Systems Branch/H. Li Open Items 78, 80, 86
Shearon Harris Nuclear Power Plant Draft SER Open Item 78 Supplemental Information Supplemental information provided per discussions at NRC (ICSB) meeting on August 18, 1983.
Response
See attached marked up FSAR Page 8.3.1-40 to be included in a future SHNPP PSAR amendment.
(8002NECccc)
SHNPP FSAR (a) Electrical penetzations are located in two areas which are separated from each other by a minimum distance of approximately 40 ft. between their boundaries. These areas are also separated by structural walls.
(b) When two channel redundancy is involved (i.e., divisions A and B), redundant cables are routed through penetrations in each of the two penetration areas. When three or four channel redundancy is involved (i.e., reactor protection system), pairs of channels share the same penetration areas; however, each channel of the pair is separated by a minimum distance of ten ft.
Safety related instrumentation channel pairs are routed through the same penetration area as their respective redundant division (A or B).
(c) All penetrations, Class TE and non-Class IE, are separated from each other by a minimum center-to-center distance of approximately four ft. horizontally and four ft. vertically.
The minimum distance between conductors of adjacent penetrations is approximately 28 in.
- 6) Separation Criteria for Class IE Instrument Cabinets, Sensors; and Sensor-to-Process Connections.
Discussion of these-separation criteria and methods is given in Chapter 7.
- 7) Separation Criteria for Actuated Equipment Locations of Class IE actuated equipment, such as pump drive motors and valve operating motors are normally dictated by the locations of the driven equipment.
The resultant locations of this equipment are reviewed to ensure that separation of redundant Class IE actuated equipment is acceptable.
8) are designed to meet IEEE-Standar 279-1971 including redundancy, separation, and single failure criteria (see detailed description in Section 7.2.1.1.2).
These circuits are designated as safety related and identified similar to the reactor protection system channels as described in Section 8.3.1.3. Separation of these circuits is maintained from other reactor trip circuits by routing each of these circuits independently in a separate conduit from the actuating device to the Reactor Protection System cabinet.
c) Hain Control Boards and auxiliary equipment panel have maintained the separation criteria by use of metal enclosures for Class 1E circuits within the board. Class lE devices are mounted in metal enclosures (module cans),
and wirings from these devices are enclosed in flexible metallic conduits and/or enc1.osed sheet metal wireways which are train dedicated inside the boards. Non-Class lE circuits are separated from Class 1E circuits by the use of internal wiring runs and exit points which are physically separated from those for Class 1E circuits.
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<-<<ZW-v@p o4 F48(48 r6~ ) Les~ yW eEEgkv4M4) 8.3.1-40 Amendment No. 9
Shearon Harris Nuclear Power Plant Draft SER Open Item 80 Supplemental Information Supplemental information provided per discussions at NRC meeting (ICSB) on August 18, 1983.
Response
The following marked up FSAR pages and inserts will be incorporated in a future SHNPP FSAR amendment.
Insert 80-A (page 7.3.1-10)
Feedwater system can be effectively isolated from the MCB manually (without actuating Manual Safety Injection) by tripping the main feedwater pumps or closing the Steam Generators Feedwater Isolation Valves FW-V26, FW-V27, FW-V28, FW-V123, FW-V124, and FW-V125.
Insert 80-B (page 7.3.1-12)
Auxiliary feedwater can be effectively isolated manually (without actuating the Manual Safety Injection) by the operator from the MCB by tripping the pumps or isolating the Steam Generator Valves AFW-V10, AFW-19, AFW-V23 when AFW motor operated pumps are in operation or the Steam Generator Valves AFW-V116, AFW-V117, and AFW-V118 when AFW turbine pump is in operation.
Insert 80-C (pa e 7.3.1-14)
In addition, the Fuel Handling Building can be effectively isolated manually by individual components at any time from the Control Room (Auxiliary Equipment Panel). The normal supply for the FHB can be isolated by tripping the running supply and exhaust fans which in turn shut the isolation dampers. The Spent Fuel Pool Pump Room ventilation can be isolated by tripping the running supply Fan AH-17. The FHB loading area can be isolated by shutting the inlet and outlet dampers D35 and D37.
Insert 80-D (pa e 7.3.1-14)
The RAB can be effectively isolated manually (without actuating the manual SI) from the Auxiliary Equipment Panel. The RAB switchgear rooms can be isolated by tripping the running supply Fans AH-12 or AH-13. The Normal Exhaust System can be isolated by tripping the exhaust Fans E17, E18, E19, and E20. The electrical protection room can be isolated by tripping the Supply Fan AH-16.
Insert 80-E (pa e 7.3.1-24)
The control zoom can be effectively isolated manually (without actuating the manual SI) by closing the normal supply air inlet damper CZ-B1SA or C2-B2SB from the Main Control Board.
(8003NECccc)
SHitPP FSAR c) MSIS Main Steam Isolation Signal d) S Safety Injection Actuation Signal e) CVI Containment Ventilation Isolation S;gnal f) M Remote Manual From Control Room Table 6.2.4-1 identifies the lines to be isolated and the signal that initiates isolation.
Power operated containment isolation valves have limit switches which indicate the status of the valve in the Control Room.
The Mi ain Steam Isolation System is automatically energized by an SSPS supplied signal. Refer to Figure 7.3.1-1 Sheet 2 for eve its which result in the MSIS.
For the main steam line isolation valves, the MSL'S is transmitted through Ebasco suppl.ied logic shown on Figure 7.3.1-7. "his logic prevents inadvertent opening of the main steam isolation ~alve in the presence of the MSIS. Refer to Section 7.7 for a description of control of other valves in the main steam lines.
'Four class 1E radiation monitors are used for the containment ventilation isolation signal (CVIS). Two monitors, RC-3561A and RC-3561C, are for train A
.and two moni.tors, RC-3561B and RC-3156D, are for train B. The CVIS is generated by a 2 out of 4 (2/4) logic. To obtain a train A 2/4 logic, the output of train B monitor signals are physically and electrically isolated from the system A and the output of train A monitor signals are 'solated from system B for the trai'n B logic. The functional Logics are shown on revised FSAR Figures 7.2.1-1 Sheet 8 and 7.3.1-1 Sheet 2. For additional information regarding the containment radiation monitors ref'.r to Section 12.3.4.1.8.1.
As shown on Figures 7.2.1-.1 Sheet 8 and 7.3.1-1 Sheet 2, Containment Ventilation Isolation Signal (CVIS) is generated by either containment high radiation (2/4 logic) or by the containment isolation Phase A (T) signal.
Upon receipt of a high radiation signal, the retentive memory relay in the circuit wilL energize and generate the CVIS signal. To assure that a "T" signal will not be blocked after resetting the r lay, the high radiation signal is present for only a duration of second (single shot). Although the 1
high radiati,on condition may still be present after resetting the signal, the single shot feature blocks the radiation input signal to the CVIS output relay. Thus, a "T" signal is still capable of activating the CVIS after it has been reset. The high radiation condition must be corrected before the single shot is itself reset. Resetting of the GVlS signal, by itself, will not change the position of actuated components. The resetting of the retentive memory logic can be done, through the CVIS signal reset switch from the Control Room.
The isolation of the Feedwater System can be accomplished manually through the Safety Injection (SI) manual actuation control switch from the Main Control Board. How .ver, individual components of the Feedw'ater System can be manually isolated at any time from the MCB. The FMIS & SI signal must be reset prior Xnl$8Cf 8~-+
7.3.1-10 Amendment 4'o. 9
SHNPP FSAR Upon any automatic initiation of the auxiliary feedwater pumps (motor or turbine driven), the steam generator blowdown and sampling valves are closed.
One of the two normally open auxiliary feedwater recirculation (bypass) valves (3AF-V187SA-1 or 3AF-V188SB-1) will be automatically closed when its respective motor driven auxiliary feedwater pump is running and a loss of the opposite power train on the 6.9kV emergency bus is sensed for more than twenty (20) seconds. Control switches are provided on the Main Control Board and Auxiliary Control Panel to allow remote manual operation of the recirculation valves.
Once initiated by an automatic signal, an auxiliary feedwater pump may be stopped or restarted manually by the operator.
A comparison of the differential pressures between all steam generators is made and used in conjunction with the MSIS to determine a steam line break and isolate the faulty steam generator from auxiliary feedwater. Both the steam generator (steamline) pressure "mismatch logic and MSIS are results of two out of three logics and are combined in a coincidence logic to develop a steam generator available signal (SGAS).
A SGAS w'ill perform the following actions:
a) Open the auxiliary feedwater regulating and pump discharge valves to
,the intact steam generator.
b) Close the auxiliary feedwater regulating and pump discharge valves to the faulty steam generator.
During operation, the auxiliary feedwater regulating valves to the intact .
steam generators are opened o ow steam generator level.
~7 $ 0-B Manual initiation and operation ot the Auxiliary Feedwater System is provided in the Control .Room. In addition, manual control switches for the auxiliary feedwater pumps, auxiliary feedwater isolation and regulating and recirculation valves are provided on the auxiliary control panel (ACP).
Manual auxiliary feedwater flow is provided by opening the auxiliary feedwater isolation valves and controlling the auxiliary feedwater flow, through the auxiliary feedwater regulating and recirculation valves, by means of manual controls on the main control board (MCB) or ACP. Process indication, alarm and status instrumentation is provided to enable the MCB or ACP operator to evaluate system performance and detect malfunctions.
The auxiliary feedwater pumps discharge pressure is maintained by regulating valves at the discharge of the auxiliary feedwater pump such that the valves cannot be operated beyond the pump low~pressure setpoint (runout protection).
B ass and Interlocks Auxiliary feedwater pumps lA-SA and 1B-SB incorporate pressure switches into the final actuation control circuitry precluding automatic (safety signal) and Amendment No. 9 7.3.}-12
SHNPP FSAR See Figure 7.3.1-13 and Section 8.3.1.1.2.13 for a detailed description of the fuel handling building emergency exhaust unit.
The conditions and signals that cause automatic initiation of the various exhaust units are as follows:
a) Following a designed basis 'accident (DBA)
The SSPS generated safety injection signal(s) causes the following:
- 1) Reactor auxiliary building (RAB) emergency exhaust Sans start (via the load sequencer)
- 2) Reactor auxiliary building isolation dampers close, (isolating the NNS portions of the ventilation from the safety portions) .
The RAB Emergency Exhaust System has two completely redundant systems which operate similarly.
b) Following a Fuel Handling Accident in the Fuel Handling Building:
On a signal of high radiation on the fuel handling operating floor, a fuel handling building isolation signal causes the following:
- 1) FHB emergency exhaust fans start
- 2) Fuel handling building isolation dampers close to isolate the normal supply and exhaust ventilation subsystems The FHB Emergency Exhaust System has two completely redundant systems which operate similarly.
F I gang.in~b.used kg n Kcoaf oneht Kasi4. T e FHB system is completely isolated upon radiation monitoring signal actuation.
wive~< 8d-C The RAB ventilation isolation is part of the component isolation during a manual safety injection actuation. Manual actuation of SI will also include the isolation of the RAB ventilation system. Manual reset of the SI from the MCB will automatically reset the RAB ventilation isolation signal.
~~S8R7- 80 -0 B asses Interlocks and Sequencing Electric heating coil SA control circuit is interlocked with the fan SA starter and its built-in air flow and temperature switches. The coil is energized and remains energized when: .g) the fan is started, 2) the flow velocity through the coil has exceeded a predetermined level as established by the coil manufacturer, and 3) the temperature of air leaving the coil remains below a level as established by the coil manufacturer. The coil electric power is modulated by a SCR drive based on the .outlet temperature of the charcoal filters. The fan inlet and outlet and filter train inlet valves are interlocked with the fan A-SA starter. The valves are opened when the fan starts. The heater'"coil SB control circuit and associated valves are similarly interlocked with the B-SB fan starter.
7.3.1-1'mendment No. 9
SHNPP FSAR pressurization of the control room envelope, emergency air cleaning, emergency air makeup and pressurization.
During normal plant operation, one air handling unit and one normal exhaust fan are requixed to operate. The normal outside air intake and exhaust valves are opened. All emergency air intake valves are closed.
Following a control room isolation signal (CRI), the normal outside air intake and exhaust valves are automatically closed, the normal exhaust fan is stopped, and emergency filtration units start. The operator may shut down the air cleaning unit to the air conditioning unit not in service. At the discretion, fresh air may be admitted to the system. control'perator's Initiatin Circuits and Lo ic A CRI signal is activated by any of the following signals:
a) Safety injection actuation signal b) High radiation signal generated by redundant radiation detectors at the normal air intake.
c) High chlorine signal generated by redundant chlorine detectors at the normal air intake.
d) High chlorine signal generated by redundant chlorine detectors at the chlorine storage area.
The, isolation of the 'control room can be accomplished manually through the Safety Injection (SI) manual actuation control switch from the MCB. However, individual components of the control room isolation can be manually operated from the MCB with the exception that SI and CRI signals must be reset prior to the equipment changeover. from an actuated mode to a normal mode. Separate control switches (Train A and Train B) are provided on the MCB for the resetting .of the SI and CRI signal.
Each air handling unit is normally operated by a control switch in the Control Room. Each of the normal outside air intake and exhaust valves is operated by a control switch in the Control Room.
Each normal exhaust fan is operated by a control switch in the Control Room. The exhaust isolation dampers are in parallel interlock with both normal exhaust fans. The dampers close when both exhaust fans stop and are opened when any of the two exhaust fans start. After an exhaust fan is started, its outlet flow control damper is automatically controlled by a pressure differential signal'o maintain a constant positive pressure in the control room envelope relative to the surroundings.
Each of the emergency air filtration unit fans is operated by control switches in the Control Room. When a fan is started, its filter train vol'ves are opened and filter train electric heating coil is energized.
Redundant class 1E radiation monitors are utilized for the Control Room Isolation system and are located in the normal outside air intake duct. These 7.3.1-24 Amendment No. 9.
'SHNPP FSAR TABLE 7.3 '-3 (Continued)
No. of Actuating No. of Channels Logic No Functional Unit Channels ~TO Tri Location~**
Containment Ventilation Isolation (Cont'd)
- c. Containment spray See item 2(a) of Table 7.3.1-2 actuation
- d. Manual oy(pefi dnr/1 dv&)
- 5. Control Room Isolation
- a. Safety injection See items 1(a) through 1(e) of Table 7.3 1-2
- b. Control room air intake radiation
- c. Control room air intake chlorine concentration
- e. Manual (component level) rr
- 6. FHB Ventilation Isolation
- a. -
Spent fuel pool radiation 2
- h. /Aa~viL (mgt ouC~r t eVC~')
7 ~ RAB Ventilation Isolation
- a. Safety injection See items 1(a) through 1(e) of Table 7.3.1-2
- b. Manual (component level)
- 8. Auxiliary Feedwater Isolation
- a. Differential 2/Steamline 2/Steamline 3 pressure between indicating that steamlines coincident its pressure with MSIV closure is low in comparison signal to other 2 lines
- b. Manual (component level)
- Actuating Logic Location Code 1 ~ Inside Containment 2~ Outside Containment Reactor Auxiliary Building 3~ Control Room 4~ Fuel fandling Building
- 5. Yard Area 7.3.1-44 Amendment No. 9
Shearon Harris Nuclear Power Plant Draft SER Open Item f186 Sup lemental Information In response to a Staff request, the applicant provided a list of non-Class lE control signals that are used as inputs to Class 1E control circuits.
The applicant stated that all non-Class lE signals are through an isolation device before signals are input to Class 1E control circuits.
Schematic drawings related to these circuits to verify the adequacy of the isolation scheme have been requested. This matter is subject to further staff review.
(The following statement is being provided in response to NRC-ICSB request made at the meeting held in Bethesda, MD on August 18, 1983. The various drawings requested by the staff were previously provided.)
Response
Although Non-Class 1E control signals provide input to Class 1E control circuits, a failure of the non-Class 1E components will not effect the proper safety operation of the Class 1E control circuits. The signal inputs from the non-Class 1E devices which feed the Class 1E circuits are through isolation devices and will be overriden by the Class 1E portion of the safety circuits.
(8001NEClcv)
Mechanical Engineering Branch/D. Terao Open Xtem 286
Shearon Harris Nuclear Power Plant Draft SER Open Item 286 Supplemental Information The applicant must provide the program for the leak rate testing of those valves that provide an interface between reactor coolant pressure boundary and low pressure systems.
Response
CP&L has reviewed the staff's criteria (based on proposed change to Standard Review Plan 3.9.6 Appendix A) for determining which valves are to be leak tested as part of the ASM< Section XI program. The attached list reflects all those valves which meet that criteria.
Each valve on the list has been reviewed in light of actual operating conditions and the following approved criteria and technical studies:
a) WASH-1400, Appendix V, pages V-43, V-44 b) 'UREG-0677 Probability of Intersystem LOCA c) EPRI-NP262 PWR Sensitivity to Alterations in the Interfacing Systems LOCA Those valves which CP6L has determined do not require testing and the justification for their exclusion have been indicated by reference to attached notes. The attached tables reflect revisions to the response provided in a letter to the NRC dated 8-12-83. The revision deletes certain valves which were drain valves and not the interfaces between the reactor coolant system and lower pressure closed systems. The notes were also amended to amplify the test frequencies to be used and the acceptance criteria for a leak test.
(
- Drawing Test Allowable Valve (location) Test Frequency Leakage Number Description (2165-) Yes/No Justification (Note 4) (GPH)
SI-V544 Accumulator Injection G809 (D5)
SI-V545 Accumulator Injection G809 (G5)
SI-V546 Accumulator Injection G809 (J5)
SI-V547 Accumulator Injection G809 (D3)
Sl-V548 Accumulator Injection G809 (G3)
SI-V549 f Accumulator Injection G809 (J3) 1 RH-'V500 RHR Suction G824 (I3) 5 (Note 5)
RH-V501 RHR Suction G824 (I4) 5 (Note 5)
RH-V502 RHR Suction G824 (L3) 5 (Note 5)
RH-V503 RHR Suction G824 (L4) 5 (Note 5)
SI-V510 Low Head Injec. (Hot Leg) G808 (C11) 1 SI-V511 Low Head Injec. (Hot Leg) G808 (C11)
SI-V512 High/Low Head Injec. (Hot Leg) G808 (C17) Note 1 SI-V513 High/Low Head Injec. (Hot Leg) G808 (C17) Note 1 SI-V507 High/Low Head Injection G808 (B3) Note 2 SI-V508 High/Low Head Injection G808 (B3) Note 2 SI-V509 High/Low Head Injection G808 (C3) Note 2 SI-V580 Low Head Injection G810 (C3) Note 2 SI-V581 Low Head Injection G810 (E3) Note 2 SI-V584 Low Head Injection G810 (C1)
SI-V585 Low Head Injection G810 (D1)
SI-V586 Low Head Injection G810 (E1)
SI-V17 BIT Injection (Cold Leg) G808 (D3) Note 3 SI-V23 BIT Injection (Cold Leg) G808 (D4) Note 3 SI-V29 BIT Injection (Cold Leg) G808 (D5) Note 3
- Drawing Test J'llowable Valve (location) Test Frequency Leakage Number Description (2165-) Yes/No Justification (Note 4) (GPM)
SI-U63 Alternate injection (Cold Leg) G808 (D6) N Note 3 SI-V69 Alternate Injection (Cold Leg) G808 (D7) Note 3 SI-V75 Alternate Injection (Cold Leg) G808 (D8) Note 3 SI-V39 High Leg Safety Injec. (Hot Leg) G808 (D12) Note 3 SI-V45 High Leg Safety Injec. (Hot Leg) G808 (D13') Note 3 SI-V51- High Leg Safety Injec. (Hot Leg) G808 (D14) Note 3 SI-V84 High Leg Safety Injec. (Hot Leg) G808 (D14) Note 3 SI-V90 High Leg Safety Injec. (Hot Leg) G808 (D15) Note 3 SI-V96 High Leg Safety Injec. (Hot Leg) G808 (D16) Note 3 SI-V514 High Leg Safety Injec. (Hot Leg) G808 (D18) Note 3 CS-V505 CVCS Normal Charging G803 (C3) Note 3 CS-V504 CVCS Normal Charging G803 (C3) Note 3 CS-V506 CVCS Alternate Charging G803 (83) Note 3 CS-'V507 CVCS Alternate Charging G803 (83) Note 3 CS-V711 CVCS Pressurizer Spray G803 (D3) Note 3 CS-V70 CVCS Pressurizer Spray G803 (D3) Note 3
G800 5.1.2-1 G809 6.3.2-2 G803 9.3.4-1 G810 6.3.2-3 G808 6.3.2-1 G824 5.4.7-1 (7994SALpgp)
Note 1 These valves (SI-V512, V513) are in series with SI-V510 and SI-V511 which will be leak tested and SI-V587 which is a normally closed HOV. CP&I has determined that these valves do not require testing based o'n the following analysis:
- a. Failure of SI-512 or V513 will cause Interfacing Systems Loca (ISL) only if V510 or V511 respectively and V587 also fail.
- b. Because V587 is a motor operated valve that is stroke tested at cold shutdown and verified closed, it will only contribute to ISL by a rupture failure.
- c. Leak testing does not affect the rupture failure probability of the MOV.
- d. Check valves V510 and V511 are leak tested (in accordance with Note 4, Frequency C) during startup after RCS has reached full pressure. This precludes a leak failure contribution to ISL, since check valve position remains constant during operation.
Using the above conditions, there are two sets of rupture-rupture failure sequences (V510 and V587, V511 and V587). Using MASH-1400 failure rates and assuming 1-year test interval, the failure probability is 1.5 x 10 /reactor year.
Note 2 For the series of three check valves providing isolation between the RCS and RHR, CP&L will test SI-V580 and V581 instead of V507, V508, and V509 and will test SI-V584, V585, and V586.
Note 3 The valves noted do not require leak testing. These valves provide isolation between the RCS and the CVS/safety injection (SI) system which is rated at the same or higher pressure. The safety injection pumps are the same pumps used for CVCS charging. An SI/charging pump is constantly running during normal operation providing a constant pressure source at higher than RCS pressure to the pump discharge check valves on the other SI/charging pumps. Excessive leakage would be detected during normal operation.
Note 4 Test frequencies are as follows:.
C Prior to entering Node 3:
- a. If the valve has been disturbed because 'of flow in the line; or
- b. at least once every nine months; or
- c. following maintenance, repair or replacement work.
R Prior to entering Node 4:
- a. At least once every eighteen months; or
- b. following maintenance, repair or replacement work.
Note 5 Testing of these valves shall be accomplished by pressurizing the line between the valves and noting the amount of water required to maintain pressure (make-up method). The total leakage shall be charged to one valve, i.e., the total leakage through each pair shall be limited to 5 GPN.
(7994SALpgp)
Power Systems Branch/E. Tomlinson Open Items 118, 126, S. R. Question 430.60, Main Steam Turbine Ualve Testing Issue
Shearon Harris Nuclear Power Plant DSER Open Item No. 118 (Question 430.36)
Supplemental Informaiton Provide a description of the Humphrey valve and ejector devices and their function as shown on FSAR Figure 9.5.4-2.
Response
The diesel engine fuel oil system as shown on FSAR Figure 9.5.4.2, Amendment No. 5 is divided into three subsystems as follows:
The fuel supply subsystem is composed of a fuel pump suction side duplex strainer, a constant speed engine driven positive displacement supply pump, a fuel pump discharge side duplex filter, supply header piping to conduct fuel to the fuel injection pumps and nozzles provided for each individual cylinder. The engine driven supply pump is sized to provide approximately 2 /2 times the required fuel for combustion at 110 percent load, in order to insure adequate injection pump fill, and to cool the injection pump. Fuel oil is positive pressure to the supply header from which the supplied under injection pumps takes suction. Uninjected fuel is returned to the day tank.
- 2. The fuel return subsystem is composed of, supply header, circulating header, piping for fuel oil return to the day tank and a pressure regulating valve.
The fuel return subsystem conducts uninjected fuel to the day tank through two paths. The uninjected fuel is returned to the supply header where excess fuel is returned to the day tank through a pressure regulating valve at the day tank via the fuel oil return connection. The pressure regulating valve in conjunction with the engine-driven supply pump delivered quantity, fuel used for combustion, and bypassed fuel, controls the supply pressure within an acceptable range required for proper system operation.
The second pathway involves the return of fuel oil which has been directed to the circulating header. The injection pump primary (or fill) cavity has a 1/16 inch diameter orificed fitting installed.
The fuel that is bypassed via the 1/16 inch injector orifice provides cooling of the injector pump and is collected in the circulating header which is maintained at a positive pressure. The oil pressure in the circulating header is slightly less than the supply header but at a sufficient positive pressure for return to the day tank via the fuel oil bypass outlet connection.
3~ The fuel leakage collection subsystem is composed of piping from the supply and circulating headers, a drip header, ejector, and a two way, two position hydraulic, pressure activated valve (Humphrey valve). The injection pumps and injector nozzles are mechanical devices and have no positive sealing arrangement other than very close metallic fits. During system operation a very small portion
of fuel is expected to leak past these parts under the very high pressures of injection. The fuel leakage could cause damage, in time, if it was crankcase.
allowed In order to to drain into and collect in the engine avoid this potential source of damage the engine is equipped with a fuel leakage collection subsystem which directs leaking (dripping) oil from the injector pump and injectors to a drip header. Since the day tank is located above the elevation of the engine driven pump to insure flooded pump auction, some motive force is required to remove and return the drip fuel to the
.day tank. The Shearon Harris diesel engine fuel oil system utilizes the motive force of the lar'ge quantity of by-passed fuel to return the drip fuel to the elevated tank. By connecting the supply header to the power port of a small high reliability, no-moving part ejectox, the supply side of a two way-two position hydraulic pressure activated positive shutoff valve (Humphrey valve) and drip headers to the suction port of the ejector via the positive shutoff valve (Humphrey valve), dripped fuel oil is removed and returned to the elevated day tank.
The ejector system acts like a jet pump providing suction for dxawing drip oil from the drip header through the Humphrey valve.
The ejector operates as high pressure oil is admitted to the power port of the ejector. This high oil pxessure is also used by the Humphrey valve which overcomes the "spring force" of the valve diaphram thereby opening the valve and admitting oil from the drip header to the suction port of the ejector. In the absence of a sufficient differential positive pressure across the valve diaphram the Humphrey valve remains closed and no oil is drawn from the drip header.
The use of an ejector eliminates the need for a separate electrical pumping system with its associated tank, level sensing equipment, motor contxols, and cabling. The positive shutoff Humphrey valve insures that in the event of a return line obstruction, bypassed fuel cannot backflow into the drip headers. The use of a check valve in the drip fuel line upstream of the suction port of the positive pressure actuated shutoff valve (Humphrey) provides additional assurance that if one device fails (Humphrey valve) the other (check valve) will protect. Even if the drip header should back up the engine will continue to operate since a vent in the drip system would divert the oil and prevent the fuel from entering the crankcase.
(7987NECtda)
Shearon Harris Nuclear Power Plant Draft SER Open Item No. 126 Su lemental Information The response to NRC Question 420.64 is acceptable except the last portion. If the vent line is non-safety it should be at least if seismically supported. The NRC asked how will the system respond there was a failure of the vent line.
(Note: This response has been revised per discussion with the NRC.)
Response
The crankcase vent line from the diesel engine is non-safety class and non-seismic. Rupture or breakage of the subject line will not impede engine capability to start or operate as required. In case the crankcase vent line fails, the vent doors provided on the diesel engine shall open at 0.7 psig without jeopardizing engine performance. The Diesel Generator Building Ventilation System, as described in Section 9.4.5 and as shown in Figure 9.4.5-2, utilizes a once-through arrangement. The flow rate will exceed 50,000 cfm during diesel engine operation. The oil vapor released from the crankcase vent consists primarily of combustion products (i.e. engine exhaust) which escape past the piston rings carrying over'a small quantity of fuel and lubricating oil. Although a quantitative estimate of oil vapor carry-over is not available, experience indicates that such carry-over will not be excessive. The diesel generator room air exhaust rate during engine operation is sufficiently large to prevent the concentration of oil vapors released into the room through the vent doors from reaching combustible limits.
See attached marked-up FSAR Section 9.5.8.3.
(8009NECtda)
Shearon Harris Nuclear Power Plant FSAR Question 430.60 Supplemental Information In Section 9.5.8.3, you discuss low barometric pressures as a consequence of a tornado. Expand your FSAR to include a discussion of the lowest barometric pressure anticipated for the geographic area of the Harris site. Discuss its impact on diesel generator performance, especially as regards the rapid fluctuations associated with tornados.
(Note: This response has been amended per discussions with the NRC to include a reference to the described tests.)
Response
As stated in FSAR Section 2.3.1.2.1 the SHNPP site lies in Region I for Design Basis Tornado. For a Region I Design Basis Tornado the pressure drop is 3.0 psi and the rate of pressure drop is 2.0 psi/sec.
The diesel generator building is designed to withstand the 3 psi pressure drop in 1.5 seconds (rate 2 pos/sec) as a result of tornado. Interior walls are designed for 2 psi differential pressure.
Transamerica-DeLaval, the manufacturer of the Shearon Harris diesel generator, has evaluated under and over pressure effects on Transamerica-DeLaval diesel engine and generator set performance (Reference 1). The results of this. study indicate no adverse effect on the performance of the diesel generator set due to a pressure drop of 3 psi in 1.5 seconds.
Reference 1. Department of the Army (DA-49-129-ENG-542) dated March 1970 "Shock Tube Test of Gas Turbine and Diesel Engine Performance and Comparison with Computer Simulation".
(8008NECtda)
Shearon Harris Nuclear Power Plant Main Turbine Steam Valve issue Test Turbine Steam Valves weekly according to Standard Technical Specifications or provide justification for monthly testing.
Response
Turbine valve testing will be in accordance with SRP 10.2. The Draft Technical Specifications for SHNPP, FSAR Section 16. 2, include weekly exercising of turbine valves and monthly visual observation of this testing.
(7'988NECtda)
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