ML20236N485
Text
{:
~
ac n7 g
UDIffED STATES NUCMAR RECUMTORY COMMISSION
\\
massonoren,s.a. m m Octoter 16,1997 IIbRAPOUM TO: Jon R. Johnson, Direcsor D,,isen,Raedor Pnveas Region 5 FROM:
Frederick J. Hotafon, Director #\\
Pn$ect Directorate ll.3 CTWE}h DMalon of Reedor Projects-t/15 OfBce of NudestReactorReputation SUBKCT.'
NRR RESPONSE TO TIA 96017 CEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - INTERPRETATION OF TECHNICAL SPECIFICATION 3.2.4 ON QUADRANT POWER TILT RATIO (TAC NO. M97294)
By enamorandum dated October 16,19g6 Region il requested technical assistance regansng interpretaban of Sequoyah Technical Specincation (TS) 3.2.4, *0uedrant Power TR Rate,' or QPTR. Sped 6caey, a guestion was raised reganAng the Sequoyah Scansee's practice of h,e( the power range nuclear hatruments fosowing completion of a core aux map.
NrtR has reviewed Sedian 3.2.4, "Ous& ant Power TR Ratio (QPTR),* in the Watts Bar TS and the Westinghouse Standard TS (NUREG-1431, Revision 1 dated 4m95). The Watts Bar TS is based upon NUREG.1431. R requires (among other actions) that power be reduced by a 3%
tom rated thermai power for each 1% of QPTR > 1.00, and performance of an evaluation of the safety analyses and connrmsoon that resotts remain vald for the duration of operation under this condbon. A4er this eratusbon is completed (Raquired Action A.4), Required Action A.5 spedbes caltrabon of escore detectors to show zero QPTR.
The Bases escalon (B 3.2.4) of the Watts Bar TS and NUREG-1431 contah the fosowing escussiorc The peakhg factors F" and Fe(Z) are of primary importance h ensuring that the power detreution remains consistent with the initial condsons used h the safety anefyses. N.L..is SRs [survetance requirements] on F" and Fo(Z) within the Complebon Trne of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures that these primary inc5cators of power dstrtuhon are wthin their respecthe Emits. A Compiehon Trne of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> tales into considershon the rate at which peaidng factors are Ikafy to change, and the ame required to e=ha= the plant and perform a aux map. If these penedng tactors are not wnhin their smas, the Required Actons of thew Swvenances provide an appropriate response for the abnormai consten. fr ew QFTR romahs stee ks spedned Emit, the peaking factor survetances are regidred each 7 days thereafter to ev=Ma F" and Fo(Z) with changes h power cEstrtu5an. Reiseve#y smas changes are expected due to other bumup and menon redstrtuhon or correction of the cause for % the OPTR Emit.
l Athough F", and Fo(Z) are of primary importance as hitial condmons h the e
safety ana#yses, coer changes h the power cEstribubon may occur as the OPTR MC PLE CEMEB copy I
II 4 DEEE
- ,Illi l
9807150141 980624 gi f ER
-155 PDR
- d. $ehneen
2
tmt is encoeded and may have an impact on the veldry of the safety analysis. A change h me power eswoumon can an cs such reacsor penwneters as bank worms and pseung factors for rod malfunction accidents. When me QPTR escoeds as bmit, a does not necessar9y rnean a safety concem exists. It does mean het more is an mecation of a change h the gross rodal power estntwtion met requires an EC M and evalueton met is accomptshed by examining me hoore power estreucort. Specdca8y, me core peaWng factors and the quadrant set must be evaluated because they are the factors that best characsertre the core power esbeumon. This re evaluation is required to ensure that, before increasing THERMAL POWER to atxwe the bmit of Required Acalon A.1, he reacsor core condemns are consistent wtth the assumpbons in the safety analyses.
If me QPTR has exceeded the 1.02 emit and a re evaluation of the safety analysis is w':M and shows that safety requirements are met, the exco o detectors are recabbrated to show a sero QPTR prior to increasing THERMAL POWER to above the kmd of Required Acsion A.1. This is done to detect any subsequent signdcant changes in QPTR. [ NOTE: NUREG.1431 is bein0 changed to state that the QPT, not the QPTR, shodd be rezorped)
The Sequoyah TS affer from the Watts 8ar TS in a number of areas. First of as, the power reduchon amount when QPTR exceeds 1.02 is at least 3% power for each 1% QPTR in excess of 1.02 h the Sequoyah TS whereas the Watts Bar TS requires a power reduchon of at least 3%
for each 1% QPTR in excess of 1.00. The Sequoyah TS were changed in this area in Apnl 1997 in cor$uncton with conversion to Framstomo nucisar fust. Seconey, the Sequoyah TS requirements retsove to vertrying that QPTR is withh its Amit wnhin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are more vague i
than Watts Bar TS or NUREG 1431 in that specdc surveRances are not required for the nuclear anthalpy and heat flux hot channel peaking factors and a reassessment of the van &ty of the safety anafysis is not =g+"-riroquired. Fnacy, the Sequoyah TS contain no provisions for resetting the QPTR (recabbrating the encore nucisar hstruments) to 1.0 (i.e., normakzing).
However, the intent of the OPTR hstrument and associated alarm is to detect significant changes to the QPTR (beyond those that may occur earty in core Bfe as a resuit of xenon pr**wwi and bumup). The alarm at 1.02% also alerts plant operators to rnmor changes such that these changes cari be evaluated and core flux mapping can be performed, as necessary.
As long as Sequoyah site procedures speczfy evaluabons that meet the intent of NUREG-1431, normahnng the OPTR alarm is acceptable at Seqyoyah even though not specified in the TS.
if you have any queshons regardng the above determmabon, please contacs the Sequoyah Projecs Manager, Ron Heman, at (301) 415-2010.
Distrtubon:
Docket Nos. 50327 and 50-328 Docket File JUeberman original signed by F.Hebdon SQN Reading MBoyie (MLB4) cc: C. Heht, Region i PECS Brarx:h Chef JZwoEnski G. Grant, Region til JRoe TUu l
T. Gwynn, Region IV SRachards MShannon, Ril DOCUMENT NAME: GASQPAS7294.TIA
- See previous concurrences To recebo e espy er eus soownerit, hecame h en bar C=Capy afo smacrenerwarunneuro E= Copy wei j
amatfwnerwertsoeuro N = No copy omCE PM.PD2-3 C
LAPD2-3 C BC:SRXB D'PD2-3 D DR
/
MAME RHeman*
BClayton*
TCosins*
FHebdon*
JZwoGnski DATE Gr23/97 9/23f97 9f30/97 10/07/97 10/h7 OFFKMAl. RECORD COPY
.=
J Johnson 2-Smt is escoeded and may have an imped on the vektty of the safety anaPysis. A change h the power estntNeon can affed such reactor parameters as bank worths and peaung facsors for rod malfunction accidents. When the QPTR escoeds Rs tmt, R does not necesserfy mean a safety concem exists. It does mean that there is an becaton of a change h the gross ra6al power dstnbution that requires an hvestgabon and er Asabon that is wT+;shed by examining she incore power estrevoon. Specmeasy, the core puung facsors and the quadrant tot must be evaeusted because they are the fedors that best charecsertre the core power estreution. This re-evaluation is required to ensure that, before heressing THERMAL POWER to above the hmit of Required Achon A1, me reacsor core conebons are consistent with the assumptions in the safWy analyses.
E the QPTR has exceeded the 1.02 Emt and a re evaluation of the safety analysis is compaeted and shows that safety requirements are met, the excore detectors are recaterated to show a zero QPTR prior to increasing THERMAL POWER to above the emit c.* Required Acean A1. This is done to detect any subsequent sigrdficant changes in QPTR. [ NOTE: NUREG 1431 is behg changed to state that the QPT, not the QPTR, shodd be rozerood)
The Segsoyah TS effer from the Watts Bar TS h a number of areas. First of an, the power reduchon amount when QPTR exceeds 1.02 is at leest 3% power for each 1% OPTR in excess of 1.02 h the Sequoyah TS whereas the Watts Bar TS requires a power reduct>on of at leest 3%
for each 1% QPTR h excess of 1.00. The Sequoyah TS were changed in this area in April 1997 h coryuncnon with cortversion to Framatomo nuciear fust. Seconey, the Sequoyah TS requirements redaeve to venfytng that QPTR is withh ks Emit wrthh 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are more vague than Watts Bar TS or NUREG 1431 h that specific surveRances are not required ior the nuclear enthalpy and heat Sex hot channel peaung factors and a reassessment of the validity of the safety analysis is not spec 6cafy required. Fanasy, the Sequoyah TS contain no provisions for resetting the QPTR (recalerahng the excore nucisar instruments) to 1.0 (i.e., norma! Wag).
However, the blant of the QPTR hstrument and associated starm is to detect sigr9ficant changes to the QPTR (beyond those that may occur earty in core Efe as a resutt of xenon pr=W and bumup). The alarm at 1.02% also alerts plant operators to minor changes such that these changes can be evaluated and core flux mapping can be performed, as necessary.
As long as Sequoyah site proce&avs specify evalustaons that meet the intent of NUREG 1431, normakmng the QPTR alarm is acceptable at Seqyoyah even though not specified in the TS.
W you have any quesbons reg.ieg the above determination, please contad the Sequoyah Project Manager, Ron Heman, at (301) 415-2010.
Docket Nos. 50 327 and 5CM328 cx:: C. Hohl, Region I G. Grant, Region m T. Gwynn, Region fV
UNITED STATES NUCLEAR REGULATORY COMMISSION
=
r wAsmwoTow, n.c. amens. coot October 16, 1997 Mr. Roger O. Anderson, Director Nuclear Energy Engineering Northem States Power Company 414 Nicollet Mall Minneapolis, MN 55401 l
SUBJECT:
TECHNICAL SPECIFICATION INTERPRETATIONS FOR AUXlLIARY FEEDWATER AND SAFETY INJECTION SYSTEMS OPERABILITY - PRAIRIE t
ISLAND NUCLEAR GENERATING PLANT UNIT NOS.1 AND 2 l
(TAC Nos. M99666, M99667, M99669, and M99670)
Dear Mr. Anderson:
The NRC stan has reviewed Northem States Power Company's (NSP) requests for technical specification (TS) interpretations dated September 12 and September 15,1997. The questions pertained to safety injection (SI) system OPERABILITY during changing plant I
conditions and auxiliary feedwater (AFW) system OPERABILITY when the AFW pump selector switches are in MANUAL and SHUTDOWN AUTO positions and the pump discharge valves are throttled, respectively. The staffs interpretations of Prairie Island TS for the Si and AFW systems are enclosed.
The sta# agrees with NSP that the Si system is OPERABLE and there is no violation of the TS between reactor coolant system (RCS) temperatures of 310 *F and 200 *F with both SI control switches in the pullout position, and that the operator has a reasonable amount of time to perform the actions necessary to place the control switch in the pullout or AUTO l
position at the RCS tra isition temperatures of 310 *F and 200 *F. However, the sta#
requests that NSP st<bmit a TS change to eliminate any ambiguity or questions regarding OPERABILITY dung changing plant conditions.
The sta# also agrees with NSP that Prairie 161and (PI) TS 3.4.B.1.c allows the AFW pump discharge valves to be throttled under administrative control during startup and shutdown operations. However, the sts# does not agree with the NSP interpretation that in MODE 2 with the AFW pump selector switch in SHUTDOWN AUTO or MANUAL position which disables Function 7.d - automatic AFW pump start on main feedwater pump trip - that the i
I AFW system is OPERABLE. Since Function 7.d is required to be OPERABLE in MODES 1 and 2 in accordance with Pl TS Table TS 3.5-28, placing the AFW pump selector switch in the SHUTDOWN AUTO or MANUAL position would make funciton 7.d inoperable and require on'ry into the appropriate ACTIONS. The sta# requests that NSP submit a TS change to C 17JCQ.gQOO9l
(,
bh f
e
j,..
l s
R. O. Anderson October 16, 1997 Table TS 3.5-2B to change the APPLICABILITY from MODES 1 and 2 to MODES 1 and 2 t
above 2% power, as well as TS 3.4.B.1.c to explicitly state shutdown operations for the throttle valves.
Sincerely,
)
ORIGINAL SIGNED BY Beth A. Wetzel, Senior Project Manager Project Directorate ill-1 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
As stated cc w/ encl: See next page DISTRIBUTION:
Docket File l
l PUBLIC PD# 3-1 Reading EAdensam (EGA1)
OGC ACRS WBeckner RGiardina JMcCormick-Barger, R!ll DOCUMENT NAME: G:\\WPDOCS\\ PRAIRIE \\Pl99666.LTR To receive a copy of this document. 6ndicate in the box C= Copy w/o saammentleh E= Copy with attachment /endosse N = No copy
- No major changes to SE.
OmCE PM:PD31 E,LA:PD31 E
BC.TSB*
D:PD31 R
BWetzel:db[MlCJamersorchWBeckner JHannon (3 I
NAME DATE 16 //h/97 le /l4/97 //
SE 10//Y/97
@l' j/97 OFFICIAL RECORD COPY
Mr. Roger O. Anderson, Director Prairie Island Nuclear Generating Northem States Power Company Plant cc:
I J. E Silberg, Esquire Site Licensing j
Shaw, Pittman, Potts and Trowbridge Prairie Island Nuclear Generating j
2300 N Street, N. W.
Plant f
Washington DC 20037 Northem States Power Company 1717 Wakonade Drive East Plant Manager Welch, Minnesota 55089 Prairie Island Nuclear Generating
- Plant Tribal Council Northem States Power Company Prairie Island Indian Community 1717 Wakonade Drive East ATTN: Environmental Department Welch, Minnesota 55089 5636 Sturgeon Lake Road Welch, Minnesota 55089 Adonis A. Nebiett Assistan' Attomey General Office of the Attomey General 455 Minnesota Street Suite 900 St. Paul, Minnesota 55101-2127 U.S. Nuclear Regulatory Commission j
Resident inspectors Office 1719 Wakonsde Drive East i
Welch, Minnesota 55089-9642 Regional Administrator, Region til U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532 4351 Mr. Jeff Cole, Auditor / Treasurer Goodhue County Courthouse Box 408 Red Wing, Minnesota 55066 0408 Kris Sanda, Commissioner Department of Public Service 121 Seventh Place East Suite 200 i
St. Paul, Minnesota 55101 2145
- was
l
~
TECHNICAL SPECIFICATION (TS) INTERPRETATIONS FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NOS.1 AND 2 Safety injection System Operability Prairie island (PI) TS 3.3.A.1.c requires two safety injection (SI) pumps to be OPERABLE when the Reactor Coolant System tRCS) temperature is k310 *F. When the RCS t
I temperature is <310 *F, PI TS 3.3.A.3 and 3.3.A.4 determine Si pump OPERABILITY. PI TS 3.3.A.3 requires that at least one SI pump control switch in the control room be in the pullout position whenever the RCS temperature is <310 *F, while 3.3.A.4 requires both Si pump t
l
, control switches to be in the pullout position whenever the RCS temperature is <200 *F.
I in an effort to satisfy both of these requirements, the Pl TS have recently been interpreted to l
require plant operators to instantaneously place the one remaining Si pump control switch in pullout when the RCS cools to 200 *F. Likewise, as the RCS is heated, when the plant l
reaches 310 *F, the second SI pump control switch is required to be instantaneously placed l
in the
- AUTO" position. With the control switch in the ' AUTO" position, no SI switches can be l
in pullout or the heatup has to be stopped. Yet, the SI pump control switc5 cannot be placed in AUTO position prior to reaching 310 *F because this would violate the requirements of TS l
l 3.3.A.3 which requires at least one SI pump.ani.'ol switch in pullout when less than 310 *F.
l Since you cannot satisfy these requirements instantaneously, both of these circumstances were considered noncompliance with TS requirements that required reporting under 10 CFR j
l 50.73 as a licensee event report (LER 1-97 09).
For changing plant conditions at 200 *F, NSP has re-interpreted the TS to allow tioth Si pump control switches to be in pullout position. This is ba' sed on the interpretation that both i
TS 3.3.A.3 and 3.3.A.4 requirements can be met if the TS phrase "at least one' is interpreted to allow both Si pump control switches to be placed in pullout prior to cooling down to 200 *F. Thus, one switch will be placed in pullout when the RCS is below 310 *F until the temperature approaches 200 *F. At that time both switches will be placed in pullout. This maintains the plant in a safe condition since events that require operation of the Si system at low RCS temperatures occur slowly and allow ample time for the operator to manually place the Si system in service. The staff agrees that the phrase "at least one" will allow both Si r
control switches to be placed in the pullout position at any time the RCS temperature is
<310 *F.
For changing plant conditions as the RCS temperature reaches 310 *F, NSP interpreted TS 3.3.A.1.c and TS 3.3.A.3 to allow some reasonable time for operators to place the second Si pump control switch in the AUTO position. This reasonaWe time does not require definition since the plant is not in any imminent danger if the second SI pump control switch is not instantaneously placed in the AtJTO position. This is based on the fact that the temperature of 310 *F is a target temperature at which the operators take their actions to implement low temperature overpressure protectio'. However, it was not expected the operator actions n
would occur precisely at 310 *F since the instrumentation that provides the RCS temperature i
indication is not exact and this temperature was based on heatup and cooldown curves where a precise temperature was not calculated. Thus, it is not reasonable to expect operator action instantaneously at the precise temperature of 310 *F.
2 The sta# agrees with the licensee that the temperature is a target temperature at which actions need to be taken and that the actions cannot be taken instantaneously. However, the sta# would expect the operator actions to be taken immediately upon reaching 310 'F.
immediately is defined in the Standard TS (NUREGs 1430-1434) as actions that should be pursued without delay and in a controlled manner. Thus, the staff agrees with the licensee that a reasonable tirne limit for operator action is acceptable. However, the sta# requests that Northem States Power Company (NSP) submit a TS change to eliminate any ambiguity or questions regarding OPERABILITY during changing plant conditions.
Auxiliary Feedwater System Operability with Throttled Pump Discharge Valves.
Whenever the unit is critical or the RCS average temperature exceeds 350 *F, for single unit operation, TS 3.4.B.1.s requires the turt>ine-driven auxiliary feedwater (AFW) pump associated with the operating reactor and a motor driven AFW pump be OPERABLE: and for two unit operation, TS 3.4.B.1.b requires all four plant AFW pumps to be OPERABLE. TS 3.4.B.1.c requires all valves and piping associated with the AFW pumps to be OPERABLE whenever the unit is critical or the RCS average temperature exceeds 350 'F. This specification makes an exception that *during STARTUP OPERATION necessary changes may be made in motor-operated valve pocition." During normal plant startup and shutdown operation, the AFW pump discharge valves are throttled to control steam generator levels.
i NSP questions whether the AFW pump discharge valves are OPERABLE when they are throttled, particularly during shutdown operations. The staff has reviewed TS 3.4.B.1.c and l
concludes that the term *STARTUP OPERATION
- encompasses MODES 2 through 4 in the l
PI TS. However, based on the definition of STARTUP OPERATION in Section 1.0 of the Pl TS, the words *during STARTUP OPERATION
- apply only when the plant is going up in power (startup). For shutdown operations the Pl TS are not as explicit. However, the staff concludes that the definition of OPERABLE - OPERABILITY in Section 1.0, the Objective of TS 3.4 *sssure the capability cf removing decay heat from the reactor," and the allowance to change the throttle valves position on start-up and manual valve position during operations l
provide sufficient justification to conclude that the throttle valves can be throttled during shutdown operations. Thus, there is no question that the AFW pump discharge valves can be throttled as long as the throttling is done under direct administrative control, and that the valves and the system are considered OPERABLE. However, the staff requests that TS 3.4.B.1.c be modified to explicitly state that the motor. operated valves' position can be l
changed both on startup and shutdown.
Auxiliary Taedwater System Operability with AFW Pump Selector Switch in Shutdown Auto in MOE 2.
l AFW pump seisetor switches have thme positions, AUTO, MANUAL, and SHUTDOWN AUTO. When the selector FWnch is in the AUTO position, all required automatic AFW start i
l functions listed in Table TS 3.5-2B are enabled for the AFW pump. In the MANUAL position l
the AFW pump can be started or stopped by placing the associated pump control switch in i
the START or STOP positions, respectively. In the MANUAL position, no automatic AFW l
pump start functions apply to the pump. The SHUTDOWN AUTO position enables all automatic AFW start functions on the pumps except the automatic start for loss of both main feedwater pumps. Pl TS Table TS 3.5-2B requires the AFW system to automatically start for
_ _ _ _.. _ _ ~
e; 3
I Function 7.e (SI) during MODES 1,2,3, and/or 4; for Functions 7.b and 7.f during MODES 1,2, and 3; and for Functions 7.c and 7.d during MODES 1 and 2. Function 7.d is the trip of both main feedwater pumps.
During startup and shutdown operatio'n when there is a low feed demand, the normal method for providing steam generator feedwater is by using the motor <lriven AFW pump. This is required because et !aw power levels, extended operation of the main feedwater pumps in the recirculation mode would cause damage to the main feedwater pumps due to heating and i
vibration of the pumps. Thus, safe, prudent operation of the plant requires placement of the l
AFW pump selector switches in SHUTDOWN AUTO during plant startup and shutdown.
During startup operations, os the plant approaches 2% power, the motor-driven AFW pump is
' operating and both pump selector switches are in SHUTDOWN AUTO. The first main feedwater pump is started and once stable operation is confirmed, the motor <lriven AFW pump selector switch is briefly placed in MANUAL, the pump control switch is tumed to STOP to secure the pump, and then the AFW pump se! actor switch is set on AUTO. Likewise, during plant shutdown, once the plant is below 2% power one AFW pump selector switch is momentarily tumed to MANUAL, the pump is started, and then the switch is placed in the l
SHUTDOWN AUTO position and the other AFW pump selector switch is placed in l
SHUTDOWN AUTO. Then the last operating main feedwater pump is secured. During both l
startup and shutdown evaluations the AFW pump selector switches are temporarily in SHUTDOWN AUTO or MANUAL while a main feedwater pump is operating.
l l
Thus, NSP proposed that the PI TS Table TS 3.5-2B Functional Unit 7.d requirement for MODE 2 OPERABILITY be interpreted to only require automatic AFW pump start 'on loss of i
both main feedwater pumps when one or both main feedwater pumps are required for operation in MODE 2, and that AFW pumps be considered OPERABLE within the requirements of Table TS 3.5-2B, Functional Unit 7.d, for the brief period of time when the AFW pumps are operating in SHUTDOWN AUTO or MANUAL while the main feedwater I
l pumps are operating.
l The staff has reviewed the subject TS and concludes that the NSP interpretation is not correct. The TS require Functional Unit 7.d to be OPERABLE in MODE 2; placing the AFW pump selector switch in the SHUTDOWN AUTO or MANUAL positions makes this function (7.d), as well as the other AFW start functions for the manual position, inoperable. This then would require entry into the appropriate ACTIONS for the TS for inoperable instrumentation amilor AFW system. While the staff agrees that use of the AFW system during startup and shutdown to av0id main feedwater pump damage is required at Prairie Island, interpretating the TS to allow this type of operation is not correct. The TS should be revised to limit the l
i OPERABILITY of Functional Unit 7.d to MODE 2 above 2% power or other quantifiable limit.
i i
a
=
ear >