ML20081D102
| ML20081D102 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/07/1983 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20079F427 | List:
|
| References | |
| FOIA-84-35 NUDOCS 8310310445 | |
| Download: ML20081D102 (49) | |
Text
- _.. -
W D7 19 93 MEMORANDUM FOR:
Thomas M. Novak, Assistant Director i
for Licensing, DL l
FROM:
R. Wayne Houston, Assistant Director i
for Reactor Safety, DSI
SUBJECT:
SHEARON HARRIS 1 & 2 - SAFETY
- EVALUATION-REPORT (SER)
Plant Name:
Shearon Harris Units 1 and 2 Docket Nos.:
50-400/401 Licensing Stage:
OL Responsible Branch:
Licensing Branch No. 3 Project Marager:
B. Buckley DSI Branch Involved:
Reactor Systems Branch Review Status:
Open Issues - Section 5.2.2.2,.15.3.3/15.3.4, 15.4.6 und 15.6.3
. The Reactor Sy. stems Bra.nch 1as cogleted redew of Sections 5.2.2, 1
5.4.7, 6.3 and 15 of the Shearon Harris Units 1 and 2; through amendment
Resolution of the 'open issues sumarized below will be provided in a later supplement to this SER:
Open Issue SER Section Issue Summary Overpressure Protection 5.2.2.2 The applicant will provide During Low Temperature PORV setpoint values to con-Operation firm that the Appendix G requirements are met.
Reactor Coolant -Pump 15.3.3/15.3.4 The applicant's analysis has Rotor Seizure / Shaft determined that fuel rods in Break DNB during this event will not fail and therefore, no adverse radiological consequences are expected.
CONTACT:
E. Marinos
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2 Thomas H. Novak SCT g 7 19 83 Open Issue SER Section Issue Summary The fuel failure aspects of the analysis have been referred to CPB for review.
Inadvertent Baron 15.4.6 The time available for oper-I Dilution During ator action to teminate Cold Shut-Down boren dilution is inadequate.
The applicant is presently considering an increase of the shutdown margin, in order to increase'the time available for operator action.
Steam Generator Tube 15.6.3 The applicant has not Rupture provir'ed sufficient documen-tation to support the analysis that the primary to secondary le~ak will be teminated in 30 minutes.
1
... =
2 griginal signog.i.
R. Wayne Houston,'Assistaht Director for Reactor Safety Division cf Systems Integration
Enclosure:
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s 5 REACTOR COOLANT SYSTEM i
5.1 Sumary Descriotion 5.2 Integrity of Reactor Coolant Pressure Boundarv
- 5. 7. 1 Compliance with Code and Code Cases
- 5. 2. 2 Overpressure Protection
~
Overpressure protection for Shearon Harris Units 1 and 2 has been reviewed in accordance with SRP 5.2.2 (NUREG-0800).
A review of each of the areas listed in the Areas of Review portion of the SRP was performed according to the SRP Review Procedures.
Conformance with the acceptance criteria, except as noted below, formed the basis for concluding that the design of the facility for
' overpressure protection is acceptable.
' ~ Uverp' reisu~re pr~ote'ctioii 'for tr octor coolaN'prIssurs bo'undary (RCPBTis d' provided by three safety and um pcwer-operated rel.ief valves in combination
=
with the reactor protection system, and by operating procedures.
The ccmbina-tion of these festures provices overpressurization protectio.. as required by GDC 15, the ASME Boiler and Pressure Vessel Code,Section III, and 10 CFR 50' Appendix C.
The above requiraments ensure RCPB overpressure protection for both power operation and low-temperature operation (startup and s' utdown).
h The following is a discussion of both modes, of overpressure protection.
5.2.2.1 Overpressure Protection During Power Operation For this mode, the pressurizer power-operated relief valves (PORVs), are sized to limit system pressure to a value not exceeding the safety valve setpoint (2485 psig) to rinimize challenges tio the safety valves.
The pressurizer spray system is designed to maintain the reactor coolant system pressure below the' power-operated relief valve setpoint of 2335 psig during a step reduction in power level of up to 10%.
The PORVs limit the pressurizer pressure to a 5-1 HARRIS SER INPUT SEC 5
~
value below the high pressure reactor trip setpoint of 2385 psig for all design anticipated transients up to and' including the design basis 50% step load reduction with steam dump to the condensers.
Credit is taken only for safety valves in analyiing operational transients and faulted condit'lons.
Each pressurizer safety valve is spring-loaded and has a relieving capacity of 380,000 pounds mass per hour of saturated steam at 2485 psig.
The combined capacity of two of these three safety valves is adequate to. prevent the" pressurizer pressure from. exceeding the ASME Boiler and Pressure Vessel Code, Section'III limit of 110% design pressure following the worst reactor coolant systems pressure transient, identified to'be a 100% load
{
rejection resulting from a turbine trip with concurrent loss of main feedwater.
I This event was analyzed with no credit taken for operation of reactor coolant system PORVs, main steamline atmospheric steam dump valves, condenser steam dump system, pressurizer level control system, and pressurizer spray system.
SRP 5.2.2 requires that the applicant demonstrate adequate relief protection
~ ~fs provTded, ass'uming 't'h'e reactor" trip is' iniIiIteI b the~second safety grade
~
- signal from the reactor protection system.
In the analysis the applicant has taken cre'iit for a high pressurizer pressure trip (the first s2fety grade trip signal from the reactor protection system). The evaluation is supported by a generic sensitivity study of required safety valve flow rate versus trip parameter presented in WCAP-7769.
The applicant provided additional informa-tion to confirm that the primary safety valves a.re sufficiently sized to f
accommodate a reactor trip on the second safety grade trip signal.
The above analyses were performed using the LOFTRAN Code, a digital simulation that includes point neutron kinetics, reactor coolant system including the reactor. vessel, hot leg, primary side of the steam generator, cold leg, secon-dary side of tha steam generator, pressurizer, and pressurizer surge line.
This code has been reviewed by the staff and found acceptable.
The applicant has provided assurance that the secondary safety valves can provide the required minimum capacity assuming a single failure of one safety 5-2 HARRIS SER INPUT SEC 5
i m
i I
i valve per loop.
The staff has reviewed the assumed values of temperature and pressure, together with their assumed instrument and control errors that were
' used for the overpressure protection system design bases, and found them I
l acceptable.
The safety valves are designed in accordance with ASME Code,Section III, and periodic testing and inspection are performed in accordance with Section XI of this' code.
In FSAR Chapter 14, the applicant has described the preoperational test program, which includes testing of the pressure-relieving devices dis-cussed in this'SER section, and has indicated that these tests would be con-ducted in full compliance with the intent of RG 1.68.
Additionally, Items II.D.1 and II.D.3 of HUREG-0737 requires performance testing of relief and safety 1
i valves and their position indications.
Conformance of this item is addressed
'n Section i
i of this SER.
5.2.2.2 Overpressure Protection During Low-Temperature Operation f
' The ' RP~requirei that thie ove'rpressure protectiYn Iyste5i d'uring low-tempirature ' '
S
'~
- operation of the' plant shall be designed in accordance with the requirements I
of BTP RSB 5-2.
The low-temperature overpressure protection is primarily provided by the pres-surizer PuRVs (two of the three PORVs are used) with automatically adjusted opening setpoints (to be specified later) that vary as a functicn of reactor l
coolant temperatures.
The reactor coolant. temperature measurements will be l
auctioneered to obtain the icwest value.
This temperature is translated into a PORV setpoint curve that will be below the maximum allowable system pressure set forth'in 10 CFR 50, Appendix G.
If the measured reactor coolant pressure approaches the PORV setpoint curve within.a certain limit, an alarm is sounded
-in the control room indicating a pressurization transient.
If the reactor coolant pressure continues to increase, the PORVs are opened to relieve system pressure.
The applicant has performed low-temperature overpressure transient analyses to determine the maximum pressure for the postulated worst case mass input (650 gpm) and heat input events.
The mass input transient analysis was 5-3 HARRIS SER INPUT SE@ 3
o s
9 I
. performed assuming the inadvertent actuation of a safety injection pump, which, in combination with other misoperation, pressurizes the RCS.
The heat input analysis was performed for~an incorrect reactor coolant pump start assuming that the RCS water solid at the initiation of the event and that a 50'F mismatch existed between the RCS (250'F) and the secondary side of the steam generator (300*F).
Results concluded that the allowable limits will not I
be exceeded.
The applicant will provide PORV setpoint values at a later date and the staff will report its evaluation in a supplement to this SER.
The PORVs and associdted block valves are required to have safety grade emergency power, supplies in accordance with Item II.G.1 of NUREG-0737.'
Section
- of this SER provides-a discussion of Shearon Harris's compliance with this requirement.
As a backup to the low-temperature overpressure protectio'n system, each inlet line to the residual heat removal system (RMRS) is equipped with a pressure suction relief valve with a capacity of 900 g<pm each at,a setpoint pressure of s.....
450 psig.
The relieving capacity of each valve is adequate to relieve the
- ' c'ombined flow of 'the three centrifugal charging pumps, which 'are the high pres-
~~'
sure safety infection pumps of the emergency core cooling system (ECCS).
RHR suction relief valves provide overpressure protection efter the RHR is put into operation and the RHR suction isclation valvec are open at ECS pressure less than 425 psig.
Also, operating procedures require that below 1000 psig and 425'F, the cperator lock out the accumulator isolation valves in the closed position and the charging pumps that are not required to be operable.
The applicant has similar restrictions on reactor coolant pump (RCP) operation.
This prevents accidental accumulator discharge or inadvertent charging pump or RCP operation under these conditions. The preoperational and initial startup test program of the overpressure protection system complies with RG 1.68.
Subject to the resolution of the atorementioned PORV'setpoint value, the staff -
concludes that the overpressure protection system for bott) normal and low-temperature conditions will be acceptable and meets the relevant requirements
-*Section numbers to be'provided later.
i p
5-4
- - - -HARRIS SER ItGUT SEC; 5_..i__
of GDC 15 and 31.
The Appen~ ix G to 10 CFR 50 requirements will be confirmed d
when the PORV setpoint values are reviewed.
This conclusion is based on the following:
The overpressure protection system prevents overpressurization of the RCPB under the most severe transients. and limits the reactor pressure during normal operational transients.
Overpressurization protection is provided by three safety valves.
These valves discharge to the pressurizer relisf tank through a common header from,the pressuri'er.
The safety and power-operated relief z
valves in the primary system--in conjunction with the steam generator safety and atmospheric steam dump valves in the secondary system, and the reactor protection system--will protect the primary system against overpressure in the
, event of a complete loss of hatt sink.
The peak primary system pressure following the worst transient is limited to the ASME Code ellowable value (110% of the design pressure) with no credit taken for nonsafety grade relief systems.
The Shearon Harris plant was assumed
- * ~~ to b'e o 'eratin'g'at'de' sigh conditions (102% of'r'a'tEd power)~and the reacf.or is '
^
10 shut down by a high pressurizer pressure trip signal.
The calculated pressure is less than 110% of design pressure.
Overpressure protection during low-temperature operation of the plant is pro-vided by two (of three) PORVs and RHR suction relief valves in conjunction with administrative controls.
The applicant has met GDC 15 and 31 and Apoendix G requirements are expected to be met when the PORV setpoints are established.
In addition, the applicant has incorporated into his design the recommendations of Task Action Plan Items 11.0.1 and 11.D.3 of NUREG-0718' and NUREG-0737.
5.4.7 Residual Heat Removal System I
FSAR Section 5.4.7 has been reviewed in SRP 5.4.7 (NUREG-0800).
A review of t
each of the areas listed in the Areas of Review portion of the SRP was performed 5-5 HARRIS SER INPUT SEC 5
according to the SRP Review Procedures.
Conformance.with the acceptance criteria has formed the basis for concluding that the design of the facility for residual heat removal is acceptable.
The residual heat removal system (RHRS) is designed to remove heat from the reactor enolant system after the system temperature and pressure have been -
reduced'to approximately 350'F and 425 psig, respectively.
The RHR system is capable of reducing the reactor coolant temperature to the cold shutdown condition and maintain this temperature until' the plant is started up again.
.The RHRS operates in the following modes:
(1) Emergency Core Coolino System (ECCS), Injection Mode Functions in conjunction with the high head portion of the ECCS to provide injection of borated water from the refueling water storage tank (RWST) into the RCS cold legs during the injection chase following a loss-of-coolant e.-
. =-
(2)
Emeroency Core Coolina System, Recirculation Mode Provides long-term cooling during the recirculation phase following a LOCA.
This function is accomplished by alicning the RHRS to take fluid from the containment sump, cool it by circulation through the RHR heat exchangers, and supply it to the cold legs of RCS.
During this mode of operation, the RHRS discharge flow may be aligned to the suction of the charging pumps to prcvide water s'upplies for high head recirculation.
Flow paths are also available for hot-leg injection during long-term recirculation mode to prevent boron precipi-tation i.n the reactor core.
(3)
Refueling
-Used to transfer _ refueling water between the refueling cavity and the refueling water storage tank at the beginning and end of the refueling operations.
5-6
-, ~.
- - - - -- - - -HARRIS SER INPUT SEC 5 ' ~ ~- ~
(4)
Cold Shutdown Removes RCS decay heat and maintains cold shutdown conditions.
(5)
Startup Connected to the chemical and volume control system (CVCS) via the low pressure letdown line to control reactor coolant pressure.
DesigndataforthekHRSare (1)
Pressure:
600 psig
'(2)
Temperature:
400*F (3)
Pump capacity:
3750 gpm.
'(4)
Numbe'r of7ndipin'deht trains:- 2
- ~ ~-
The RCS cooldown time with one RHR train from initial conditions of 400 psig and 350*F to 212*F is approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
The two RHR trains are independent in action and powered by separate essential power supplies to provide redundancy.
5.4.7.1 Functionai Requirements
.As required by SRP Section 5.4.7, the RHRS for Shearon Harris must etet GDC 1 through b.
GDC 1 through 4 are covered in Sections
, and
- of this SER, respectively.
GDC 5 is met because components are not shired between units.
- Section numbers to be provided later 5-7 HARRIS SER INPUT SEC 5
~
Re'dundancy in the RHRS is provided by two independent trains for each unit.
Leak detection for the RHRS is discussed in Section
- of this SER.
Isolation
- valve and power supply' redundancy are di: cussed under separate topics in this s' action'.
The staff has reviewed the description of the RHRS and the piping and instrumentation diagrams to verify that the system can be operated with or without offsite power and assuming a single failure.
The two RHR pumps are connected to separate buses that can be powered by separate diesel generators in the event of loss of offsite power.
SRP 5.4.7 requires that the RHRS must be operable from the control room in ac-cordance with GDC 19.
Limited manual actions are permitted outside the control room assuming a single failure, if justified.
The Shearon Harris RHR system is designed to be fully operable from the control room. To ensure emergency core cooling system readiness and to protect RHR pumps, valve positions and pump running status indications are provided in the contr'o1 room.
The cooldown time of 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> with one RHR train is acceptable. With the
's" tat'ed4-h'ourYimefo[c'ooldo'wnfromstandbytoRHliconditions,theShearon
'~
~
~
N#
Harris plant can 'be brought to cold shutdown within a reasonable period of time with or without offsite power, as specified in SRP 5.4.7.
- o. 4. 7. 2 RHRS Isolation Requirements The RHRS valving arrangement is designed to provide adequate protection to the RHRS from overpressurization when the reactor coolant system is at high pres-sure operation.
i There are two separate and redundant motor-operated isolation valves (MOVs) t between each-of the two RHR pump suction lines and the RCS hot legs.
These valves ~ are separately and ' independently interlocked to prevent valve opening
~
g until the RCS pressure falls to below 425 psig.
If the valves are open, they are separately and independently interlocked to close when the RCS pressure "Section numbers to be provided later 5-8 HARRIS SER INPUT SEC 5
rises above 750 psig.
Each one of_the four RHR suction HOVs is aligned to a separate motor control center.
One MOV in each suction line is powered from a separate power train. That is, the loss of one power train will prevent open-ing of both suction lines and establishing ncrmal shutdown cooling.
Should his situation develop, RCS cooling via the steam system can be resumed until power is regained to the failed power train or manual action is taken.
Further discussion of this valve configuration is addressed in Section 3 of this SER.
In response to the staff request for additional information with regard to a potential problem of, losing shutdown cooling during reactor maintenance evalua-tions and the alarms and indications available to alert the operator to this svent, the applicant indicated in a letter dated September 30, 1982, that flow and pressure indications downstream of the pump would provide indication that ficw had decreased. When this situation is identified, the affected train would be isolated and heat removal would be accomplished by the redundant train.. Procedures will be developed to address the provision of alternate source of cooling should this event occur.
The staff concluded that the applicant response is acceptable.
a-s...
JR There are two check valves and a normally open MOV on. each RHR discharge line.,'
The two check valves protect the system frem the RCS pressure during normal clant operation.
The applicant has provided design features to permit leak testing of each check valve separately during plant operation to fulfill the' staff requirements for high/ low pressure isolation with two check valves.
This testing is further addressed in Section i of this SER.
S.4.7.3 RHR Pressure Relief Requirements Overpressure protection of the RHRS is provided by four relief valves, one on each of the suction and discharge lines..Each suction line relief valve has a capacity of 900 gpm at 450 psig, which is sufficient to discharge the flow from.three charging pumps at the relief valve setpoint.
Each discharge line from the RHRS to the RCS is equipped with a pressure relief valve to relieve
'Section numbers to be provided later l
5-9 HARRIS SER INPUT SEC 5 l
v the maximum possible back-leakage through the valves separating the RHRS from the RCS.
Each valve has a relief flow capacity of 20 gpm at a set pressure of 600 psig.
The fluid discharged by the suction side relief valves is collected in the pressurizer relief tank.
The fluid discharged by tht. discharge side relief valves is collected in the recycle holdup tank of the boron recycle system. These relief valves are adequate to protect the residual heat removal system from overpressurization.
In response to a staff request tb discuss procedures available to the operator for responding to the lifting of an RHR relief valve, the applicant stated in a letter dated September 30, 1982, that specific procedures will be developed to diagnose a lifted relief valve and isolate the affected train.
The staff finds this response acceptable.
The applicant has further committed, in
'esponse to the staff's request, to provide a relief system for the RHR ~ suction r
lines.
The provision will be implemented in a later revision to Section 5.4.7.2.4 of the FSAR. We find this commitment acceptable.
~ S.4.7.4-* RHR Pdmp ~Pr6te'ction -
^ ' - -
~
Each of the two RHR pumps has a miniflow bypass line to prevent overheating in
-the event'of inadeque.te pump flow.
A valve locatad in each miniflow line is regulated by a signal from the flow transmitters located in each pump discharge header.
The control valves open whe's tha RHR purp discharge flow is less than approximately 500 gpm and close when the flow exceeds approximately 1000 gpm.
A pressure sensor in each pump discharge header provides a signal for an indicator in the control room.
A high pressure alarm is also actuated by the pressure sensor.
The FSAR identifies the indications upon losing component cooling water to RHR pumps and. discusses the frequency of testing the RHR i
f miniflow lines.
5-10 HARRIS SER INPUT SEC 5
-- = -. -
-==
a--
5.4.7,5 Tests, Operational Procedures, and Support Systems l
The plant preoperational and startup test program provides for demonstrating the operation of the residual heat removal system in conformance with RG 1.68, as specified in SRP 5.4.7, Paragraph III.12.
The staff has reviewed the component cooling water system to ensure that suf-ficient cooling capability is available to the RHRS heat exchangers.
The
. acceptability of this. cooling ca'pacity and its conformance to GDC 44, 45, and 46 are discussed in Section
- of this SER.
The applicant states that the RHRS is housed in a structure that is designed to withstand tornadoes, floods, and seismic phenomena.
This area is addressed further in Section
- of this SER.
The RHRS capability to withstand pipe whip inside containment as required by GDC 4 and RG 1.46 is discussed in Section
- of this SER.
Protection against
' " pipi'ng'~ failure's~ outs [de of co'ntai6 ment'in accNila'nTe ~with TaDC 4 is discu~ssed '
~
- in Section
- of this SER.
(
All RHR lines, including instrument lines, have containment isolation features; their satisfaction of the requirements of GDC 56, 57, and RG 1.11 is discussed in Section
- of this SER.
t The applicant, according to the requirement,s of SRP 5.4.7, Paragraph II.D.1, has demonstrated that suitable plant systems and procedures are available to place the plant in a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.
' *Section numbers to be provided later 5-11 HARRIS SER INPUT SEC 5
p.
y 5.4.7.6 Conclusions l
I 4
The RHR function is accomplished in two phases:
the initial cooldown phase l
and the RHRS operation phase.
In the event of loss of offsite power, the initial phase of cooldown is accomplished by use of the auxiliary feedwater system and tne atmospheric dump valves.
This equipment is used to reduce the reactor coolant system temperature and pressure to values the permit operation of the RHRS. The review of the initial cooldown phase is discussed in Section *
' of this SER. The review of the RMRS operational phase is discussed below.
The RHRS removes core decay heat and provides long-term cooling following the initial phase of reactor cooldown.
The scope of' review of the RHRS included piping and instrumentation diagrams, equipment layout drawings, failure modes i
and effects analysis, and design performance specifications for essential com-ponents.
The review has included the applicant's proposed design criteria and design bases for the RHRS and his analysis of the adequacy of'these criteria and bases and the conformance of the design to these criteria --4 bases.
s.-
A_
- R Except for the above noted unresolved issues, the staff,conclecus -hat the design of the RHRS is acceptable and meets the relevant requirements of GDC 2,
~
5, 19, and 34.
This conclusion is based on the following:
(1) ~ The applicant has met GDC 2 with respect to Position C.2 of RG 1.29 concerning the seismic design of systems, structures, and components whose failure could cause an unacceptable reduction in the capability of the RHRS.
(2) The applicant has met the requirements of GDC 5 with respect to sharing of structures, systems, and components by demonstrating that such sharing does not significantly impair the ability of the RHRS to perform its safety function including, in the event of an accident to one unit, an orderly shutdown and cooldown of the remaining units.
- Section' numbers to'be provided later 5-12 HARRIS SER INPUT SEC 5
(3) Except as noted above, the applicant has met GDC 19 with respect to the main control room requirements for normal operations and shutdown and GDC 34 which specifies requirements for the residual heat removal system by meeting the regulatory position in BTP RSB 5-1.
The staff reviews of the following Task Action Plan items are addressed in Section i of this SER:
(1)~ Task Action Plan Item II.E.5.2 of NUREG-0660 as it relates to systems
~
capability and reliability of shutdown heat removal systems under various transients.
(2) Task Action Plan Item II.E.3.3 of NUREG-0660 as it relates to a coordinated study of shutdown heat removal requirements.
(3) Task Action Plan Item III.D.1.1 of NUREG-0737 as it relates to primary coclant sources outside of containment.
w-.
i 5-13 HARRIS SER INPUT SEC 5
v 6.3 Emeroency Core Coolino System The staff has reviewed'the Harris emergency. core cooling system (ECCS) in accorda'nce with SRP 5.3 (NUREG-0800).
Each of the areas listed in the Areas of Review portion of the SRP was reviewed according to the SRP Review Procedures.
Conformance with the acceptance criteria, except as noted below, formed the basis for concluding that the design of the facility for emergency core cooling is acceptable.
As specified in the SRP, the design of the ECCS was reviewed to determine that it is capable of performing all of the functions required by the design bases.
The ECCS is designed to provide core cooling as well as additional shutdown capability for accidents that result in significant depressurization of the reactor coolant system (RCS).
These accidents incl."de mechanical failure of the RCS piping up to and including the double ended ereak of the largest pipe, rupture of a control rod drive, spurious relief valve operation in the primary and secondary fluid systems, and breaks in the steam piping.
w.-
-u l-~
~,u principal bases for 'the staff's acceptance of this system are conformance I
to 10 CFR 50.46 and Appendix K to 10 CFR 50, and GDC 2, 5, 17, 27, 35, 36, and 37.
The applicant states that the requirements will be met even with minimum engineered safeguards available, such as the loss of one amergency power bus, with offsite power unavailable.
l 6.3.1 ' System Design As specified in SRP 6.3.1.2, the design of the ECCS was reviewed to determine that it is capable of performing all of the functions required by the design basec.
The-ECCS design is based on the availability of a minimum of two ac-cumulators, one charging pump, and one RHR pump, together with associated valves and piping.
Following a postulated LOCA, passive (accumulators) and active (injection. pumps and associated valves) systems wiil operate.
After 5-1 HARRIS SER SEC 6
the water inventory in the RWST has been depleted, long-term recirculation will be provided by taking suction from the containment sump and discharging to the RCS cold and/or hot legs.
The low pressure passive accumulator system consists of three. pressure vessels partially filled with borated water and pressurized with. nitrogen gas to approximately 650 psia.
Fluid level, boron concentration, and nitrogen pressure can'be remotely monitored and adjusted in each tank. When RCS pressure is lower than accumulator tank pressure, borated water is injected through the RCS cold legs.
The high-head injection system consists of three centrifugal charging pumps that provide high pressure injection of boric acid solution into the RCS.
The l
high hs.ad pumps are aligned to take suction from the RWST for the injection phase of their operation.
Low-head injection is accomplished by two RHR pump subsystems taking suction from the RWST during the short-term ECCS injection phsae and from the containment sump during long-term ECCS. recirculation.
The RWST nominal water inventory is 487,500 gal of 2000 ppm borated W e'-
1o l
' niaintaiN the RVTST Vtadei 'above' the ' temperature"of' bIron prelEipitation and
- freezing, the applicant has provided the RWST with a. heating system.
In response to the staff request for additional information related to the i
RWST and the failure cf nonseismic piping in lines connected to the RWST, the applicant stated in a letter dated Septemoer 30, 1982 that the pipe in question I
l is a seismically designed and analyzed pipe; failure is thus not expected.
This response is acceptable to the staff.
As specified in SRP 6.3, Paragraph II, the ECCS is initiated either manually or automatically on (1) low pressurizer pressure, (E) high containment pressure, or (3) low pressure in any steamline.
This meets the requirements of GDC 20.
The ECCS may also be manually actuated, monitored, and, controlled fron: the control room as required by GDC 19.
The ECCS is supplemented by instrumenta-tion that will enable the operator to monitor and control the ECCS equipment following a LOCA so that adequate core cooling may be maintained.
6-2' HARRIS SER SEC 6
~ - - -
o y
As specified in SRP 6.3, Paragraph III.3, the available net positive suction i
head (NPSH) for all the pumps in the ECCS, centrifugal charging, and RHR pumps
~
has been shown to provide adequate margin by calculations performed to meet the saf~ety intent of RG 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps."
As required in SRP 6.3, Paragraph III.11, the valve arrangement on the ECCS discharge lines has been reviewed with respect to adequate isolation between the RCS and the low pressure ECCT.
In some lines, this isolation is provided l
by two check valves in series with a normally clos;.d isolation valve (high head injection discharge, and low head injection discharge to the hog legs).
Other discharge lines have only two check valves in series.
This arrangement is acceptable because periodic leak detection across each check valve is performed during plant operation.
Test lines are provided for periodic leakage checks of reactor coolant past the check valves forming the reactor coolant system pressure boundaries.
This is discussed further in Section i of this I.
SER.
w....
n Containment isolation features for all ECCS lines, including instrument lines, the requirements of GDC 56 and RG 1.11, " Instrument 1.ines Penetrating Primary I
Reactor Containment," are discussed in Section i of this SER.
The applicant has provided additional information to address the potential debris ir. side containment (including thermal insulation and unqualified paint)~
that may inhibit ECCS pcrformance during the recirculation mode and the effects of a postulated high-energy line break in the vicinicy of the sump.
The effects of primary coolant sources outside containment (NUREG-0737, Item III.D.1)arediscussedinSectioniofthisSER.
L During normal operation, the ECCS lines will be maintained in a filled condi-tion.
Suitable vents are provided and a dinistrative procedures will require
- Section numbers to be provided later.
6-3 HARRIS SER SEC 6
that ECCS lines be returned to a filled condition following events such as maintenance that require draining of any of the lines.
Maintaining these lines in a filled condition will minimize the likelihood of water hammer occurring during system startup.
The safety injection lines are protected from intersystem leakage by relief valves in both suction header and discharge lines.
Intersystem leakage detec-tion is described in Section
, pump systems.
1 As specified in SRP 6.3, Paragraph II.B, no ECCS components are shared between I
units, which meets the requirements of GDC 5.
6.3.2 Evaluation of Single Failures As specified in SRP 6.3, Paragraph II, the staff has reviewed the system description and piping and instrumentation diagrams to verify that sufficient
- ' core ~couling wT1115e'pYovided duri'ng the' initTer bijection" phase with an~d
.r without the availability cf offsite power, assuming a. single failure.
The
' cold leg accumulators have normally open motor-operated isolation valves in their discharge lines.
One accumulator is attached to each of the RCS cold legs.
These isolation valves will have control power removed to preclude inadvertent valve movement that could result 'n degraded accumulator
-performance.
Two active injection systems are available,'each with two pumps.
The pumps in each system are connected to separate power buses and are powered from separate diesel generators in the event of loss of offsite power, as required by GDC 17.
Thus, at least one pump in each injection train would be actuated.
The high-
)p head injection systems contain parallel valves in the suction and discharge lines, thus ensuring operability of one train even~if one valve fails to open.
The low head injection systems are normally aligned so that valve actuation is not required during th,e injection phase.
- Section numbers to be provided.
6-4 HARRIS SER SEC 6
_i__________.____.____________._____
o y
I The applicant has proposed a partially automatic system with operator action required to complete switchover of the low head system from the injection to 5
the recirculation mode.' The automatic function of the system opens the RHR pump suction valves from the. containment sump, with operator action required to isolate the RWST.
Several valves thu would have to be actuated during the switchover are interlocked to other components to prevent out-of-sequence operation.
No operator action is required for 20 minutes after a large-break LOCA.
The staff has reviewed the plant's capability for hot-leg injection during the recirculation phase to preclude excessive buildup of boron concentration'in the pressure vessel.
The' staff has concluded that there is sufficient redun-dancy in injection lines and pumps to ensure adequate hot leg injection after i8 hours of cold leg injection. This meets the requirements of SRP Section 6.3, Paragraph III.6.
~
During the long-term recirculation cooling phase of ECCS, operation, leak
" detection"isreq'uiiedfto'iden'tifypassiveECCIfaiTures' cit'sideofcontainment,
~~
~
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- such as pump seal failures.
The applicant has provided a system of water-level monitors and radiation detectors located in each compartment that contains ECCS components.
With this system, the limiting leak (assumed to be 50 gpm) would be detected and isolated within 30 minutes. The applicant has calculated that tha total leakage in 30 minutes would not compromise long-term cooling.
Leak rates of less than 50 gpm Nould result in scenarios in which the detection (alarm) time would be longer, but the time available for operator response would also be longer.
The staff finds the system acceptable because it pro-vides a means of identifying and isolating a passive failure in the ECCS out-side of containment.
Flooding of ECCS components inside containment following a LOCA has been evaluated.
No ECCS LOCA.related instruments or valve operators will be flooded following a postulated accident.
All electrically operated valves in the ECCS required to be functional during and following a LOCA-are located outside con-tainment.
All other electrical equipment in the ECCS that is required during 6-5 HARRIS SER SEC 6
~-
-~
post-LOCA is either located outside containment or above the maximum calculated water level inside containment.
Based on its review of the design features and with satisfactory resolution of confirmatory items discussed above, the staff concludes that the ECCS complies with the single-failure criterion of GDC' 35.
I 6.3.3 Qualification of Emergency Core Cooling System The ECCS. design to seismic Category I requirements, in compliance with RG 1.29, and its housing in structures designed to withstand a safe-shutdown earthquake and other natural phenomena, as required by GDC 2, are discussed in Section*
of this SER.
The equipment design to Quality Group B, in compliance sith RG 1.26, is discussed in Section ^ of this SER.
l The ECCS protection against missiles inside and outside containment by the design of suitable reinforced concrete barriers which include reinforced ~
~ 4onc' ret'e' ialls ~a'nd'si'a}is'(codformance to 'GDC TI'is' disc 6ssTd in Section '_, of
~
s
.3 this SER.
The protection of the ECCS from pipe whip..inside and cutside of containment is discussed in Section
- of this SER.
The active components of th-ECCS designed to function ander the most severe-duty leads, including safe-shutdown earthquake,-are discussed in Section
- of this SER.
The ECCS design to permit periodic inspection in accordance with ASME Code,Section XI, which constitutes compliance with GDC 36, is discussed in Section
- of this SER.
This meets the requirements set forth in SRP 6.3, Paragraph III.23.c.
l l
The ECCS incorporates two subsystems that, serve other functions.
The RHRS provides for decay heat removal during reactor shutdown; at other times the RHRS is aligned for ECCS operation.
The centrifugal charging pumps are utilized I
for maintaining the required volume and water chemistry of primary fluid in
- Section numbers to be provided 6-6 HARRIS SER SEC 6 y
-r_-__------
_ _ _ _ - -.. - - - _ - - - -. ~
=-
..~u
- - ~ ~
the RCS.
On an ECCS actuation signal, the system is aligned to ECCS operation,
~
and the CVCS function is isolated.
In neither case (RHR or centrifugal charging) does the normal system use impair its capability to function as an
. integral portion of the ECCS.
6.3.4 Testing l
l The applicant has committed to demonstrate the operability of thi ECCS by sub-
~
jecting all components to preoperational and periodic testings, as requir'ed by RG 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled l
Power Reactors," and 1.79, "Preoperational Testing of Emergency Core Cooling System for Pressurized Water Reactors," and GDC 37.
6.3,4.1 Preoperational Tests One of these tests is to verify system actuation; namely, the operability of all ECCS valves initiated by the safety injection signal, the operability of s_. -
all safeguard pump circuitry down through the pump breaker control circuits,.
- and the proper operation of all valve interlocks.
Another test is to check the coid leg accumulator system and 'njection line to verify that the lines are free of obstructions and that the accumulator check valves and isolation valves operate correctly.
The applicant will perform a low pressure blowdown of each accumulator to confirm the line is clear and check the operation of the check valves.
The applicant will use the results of the p'reoperational tests to evaluate tho hydraulic and mechanical performance cf ECCS pumps delivering through she flow paths for emergency core cooling.
The pumps will be cperated under both mini-flow (through test lines) and full-flow (through the actual piping) conditionst "Secti,on numbers to be provided 6-7 HARRIS SER SEC @
~=.i..z--
i By measuring the flow in each pipe, the applicant will make the adjustments necessary to ensure that no one branch has an unacceptably low or high resist-As part of the ECCS verification, the applicant will analyze the results ance.
to ensure there is sufficient total line resistance to prevent excessive runout of the pumps and adequate NPSH under the most limiting system alignment and RCS pressure.
The applicant will veHfy that the maximum flowrate from the test.results confirms the maximum flowrate used in the NPSH calculations under the most limiting conditions and will also confirm that the minimum
~
acceptable flow used ein the LOCA analysis are met by the measured total pump flow and a rela' tive flow between the branch lines.
The applicant has provided a commitment to perform scale model testing of the containment recirculation sumps to verify vortex control and acceptable pres-sure drops across the screens and intake.
This commitment is acceptable to th'e staff.
The staff evaluation of the applicant's test result will be reported in a subsequent version of this SER.
i b
P "~The staTf "concTu'de's thit'the 'preoperational tIsY p'rograii cInforms to the-
[
l
- recommendations of RGs 1.68 and 1.79 and is acceptable pending successful F
completion of the program.
Additional discussion.of the preoperational test program is in Section "_ of this SER.
6.3.4.2 Periodic Component Tests Routine periodic testing of the ECCS components and all necessary support l
systems at power will be performed.
Valves that actuate after a LOCA are l
operated through a complex cycle.
Pumps are operated individually in this test on their miniflow lines except the charging pumps, which are tested by their normal charging function.
The applicant has stated that these tests will be performed in accordance with ASME Code,Section XI.
- Section numbers to be provided.
E-8 HARRIS _SER $EC @
..______m_..a-----
6.3.5 Performance Evaluation The ECCS has been designed to deliver fluid to the RCS to limit the fuel maximum cladding temperature:following transients and accidents that require ECCS actuation.
The ECCS is also designed to remove the decay and sensible heat during the recirculation mode.
10 CFR 50.46 lists the acceptance criteria for an ECCS.
These criteria include the following:
(1) The calculated maximum fueTcladding temperature does not exceed 2200F, (2) The calculated total oxidation of the cladding does not exceed 0:17.. times the total cladding thickness before oxidation.
(3). The calculated total amount of hydrogen generated from the chemical l
reaction of the cladding with water or steam does not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surround-i ini t'he pTe'ntii voYuine, w'ere fo react.'
~
(4) Calculated changes in coro geometry are such that the core remains amenable to cooling.
(5). After any calculated successful initial operation of the ECCS, the calculated core temperature is maintained at an acceptable low value and-decay heat is removed for the extended period of time required by the long-lived radioactivity remaining in the core.
In addition, 10 CFR 50.46 states
- ECCS cooling performance shall be calculated in accordance with an acceptable model, and shall be calculated for a number of postulated
. loss of-coolant accidents.
Appendix K to 10 CFR 50, ECCS Evaluation Models, sets forth certain required and acceptable features of evaluation mod'els.
6-9 HARRIS SER SEC 6
6.3.5.1 Large-Break LOCA The applicant has examined a spt::trum of large breaks in RCS piping, and these analyses indicate that the most limiting event is a cold leg double ended guillotine break with a Moody discharge coefficient of 0.4.
The applicant took credit for one train of active ECCS components and two of the three accumulators in the analysis.
In the large-break analysis, the worst casc break was assumed which resulted in decreasing RCS pressure.
ECCS was assumed to be initiated by low pressurifiiir pressure trip.
The analysis results demon-strated.that adequate core cooling is provided assuming the worst single failure, with no credit taken for nonsafety grade equipment.
The staff has requested additional information concerning the adequacy of the ECCS during
' shutdown /startup situations when portions of the ECCS are isolated to verify compliance with SRP 6.3.22.e.
The staff evaluation of this issue will be discussed in a subsequent version of the SER.
The large-break LOCA evaluation model utilized in this analysis is described Tn WCAPr9220.
This do' del was' approved by' NRCil'etYer'from"J. F. 'Stolz, TRC to' T. M. Anderson, Westinghouse, dated April 29,1978) and is used in large-break,'
- ~-
LOCA analyses for Westinghcuse plants.
Concerns expressed in NUREG-0630 about the conservatism of fuel-cladding swelling and rupture models used in LOCA analyses have been addressed by the applicant.
Containment parameters are chosen to minimize containment pressure so that core reflood calculations are conservative., Fuel rod initial conditions are chosen to maximize clad temperature and oxidation.
Calculations of core geometry are carried out past the point where temperatures are decreasing.
The most limiting break with respect to peak clad temperature is the double-ended guillotine break in the p' ump discharge leg with a CD = 0.4.
The peak clad temocrature is 2181F, which is below the 2200F limit of 10 CFR 50.46.
I The limiting local and core wide clad oxidation values calculated by the applicant were 7.5% and less than 0.3%, respectively.
6,10 HARRIS SER SEC 6
~
The amount of bypass flow into the upper head region has been predicted by the applicants to be sufficient to maintain the upper head region at cold leg tem-peratures.
Similar calculations performed for other types of Westinghouse I
plants have agreed with measured values of upper head temperatures.
The applicant has supplied the sensitivity study confirming the upper head region temperature in a plant of the Shcaron Harris class to be equal to the cold-leg temperatures.
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~~
6.3.5.2 Small-Break LOCA The LOCA sensitivity studies determined the limiting small break to be"less that a 10-in.-diameter rupture of the RCS cold leg.
A range of small break analyses was presented that established the limiting break size.
The analysis of this break has shown that the'high-head portion of the EC?.S. together with accumulators, prcvidas sufficient core flooding to keep the calculated peak clad temperature less than that calculated for a large break and below the limits of 10 CFR.50.46..
w....
e 2R The applicant has submitted analyses for a spectrum of small-break LOCA anal-
="
yses (3-in., 4-in., 6-in.).
These idertify that the 3-in, break is the limit-ing small break; the calculated peak cladding temperatere is 1808*F, the local metal water reactor is 4.16%, and the core-wide oxidation is less than 0.3%.
In response to the staff request to verify that the 3-in. break is the limiting small break for peak temperature and to provide the sensitivity study on which verification is based, the applicant stated (in a letter dated September 30, 1982) that a complete spectrum of small ' ceak loss-of-coolant accidents was examined in WCAP-9600, " Report on Small Break Accidents for Westinghouse NSSS." The studies in that report indicated that maximum peak clad temperature l.
(PCT) occurred for the 3-in. break; thus the PCT does increase ks break size decreases for the FSAR cases, but then decreases as break sizes decrease below-3 in. Because WCAP-9600 has been approved by the staff, the applicant's response is acceptable.
6-11 HARRIS SER SEC 6
N In response to the staff's request for additional information (FSAR Q 440.120)",
th'e applicant stated (in a letter dated September 30, 1982) that at 1000 psi, power is locked out from the accumulator isolation valves and from the non-operating charging pumps, and that the high containment pressure safety injec-tion actuation logic cannot be blocked.
LOCAs during startup and cooldown l
have been evaluated to determine the effects of the unavailability of the ac-l cumulators.
Analysis results show that with only one high head safety injec-tion (HHSI) pump, two RHR pumps, and no accumulator dis' charge, peak clad temperature would reach only abo ~ut 1100*F.
With only one HH5I pump, one RHR l
pump, and no accumula' or discharge, the peak clad temperature reaches only t
about 1700'F. This is significantly below the Appendix K requirement.
The staff concludes that this response is acceptable.
For small LOCAs (less than approximately 2-in. in diameter), the applicant stated that the containment kighpressuresetpointmaynotbereachedandtheoperatorwouldhavetorely orr the observation of the safety-related indications to initiate safety injec-tion manually.
Topical report WCAP-9600 indicates that the operator would have 1400 seconds to initiate safat:. injection manually in tha casa of a small LOCA
~
N..:. Vhich is 's'uffiliien't t'ide'.
~~
' '~
~
'The applicant hes analyzed the performance of the ECCS in accordance with the criteria set forth in 10 CFR 50.46 and Appendix K to 10 CFR 50.
The staff has reviewed the applicant's evaluation, and concludes that it is acceptable.
6.3.6 Conclusions The ECCS includes the piping, valves, pumps, heat exchangers, instrumentation, and controls used to transfer heat from the core after a LOCA.
The scope of f
review of.the ECCS for the Shearon Harris plant included piping and instrumen-tation diagrams, equipment layo'ut drawings, failure modes and effects analyses, l
and design specifications for essential components.
The review included the applicant's proposed design criteria and design bases for the ECCS and the manner in which the design conforms to these criteria and bases.
6-12 HARRIS SER SEC 6
.. -. - -.. - ~. ~..
:=
l The staff concludes that the design of the ECCS is acceptable and meets the requirements of GDC 2, 5,17, 27, 35, 36, and 37.
This conclusion is based on the.following:
(1) The applicant has met the requirements of GDC 2 with regard to the seismic design of nonsafety systems or portions thereof that could have an adverse effect on ECCS by meeting Position C.2 of RG 1.29.
(2) The applicant has met the requirements of GDC 5 with respect to sharing of structures, systems, and components by demonstrating that such sharing does not significantly impair the ability of the ECCS to perform its safety function including, in the event of an accident to one unit, an orderly shutdown and cooldown of the remaining units.
(3) The applicant has met the requirements of GDC 17 with respect to proviaing sufficient capacity and capability ensure that (a) specified acceptable fuel design limits and design conditions cf the RCPB are not exceeded as a result of ant [cIp~ated' operational occuNncIs anii (b') the core is P
cooled and vital functions are maintained in the event.of postulated accidents.
l (4) The applicant has met the requirements of GDC 27 with regard to providing combined reactivity control system capability to ensure that under postu-lated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained and the applicant's design meets the guidelines of RG 1.47.
(5) The applicant has met the requirements of GDC 35 to provide abundant cooling capability for ECCS by providing redundant safety grade systems that meet the recommendations of RG 1.1.
(6) The applicant has met the requirements of GDC 36 with respect to the d,esign of ECCS to permit appropriate periodic inspection of important
.omponents of the system.
S-13 HARRIS SER SEC 6
...,. ~.
(7) The applicant has met the requirements of GDC 37 with respect to designing the ECCS to permit testing of the operability of the system throughout the life of the plant, including the full operational sequence that i
brings the system into operation.
(8) The applicant has provided an analysis of the proposed ECCS relative to the acceptance criteria of 10 CFR 50.46, and Appendix K to demonstrate that the ECCS designs for peak cladding temperatu'rs, maximum hydrogen generation, cool,able core giometry, and long-term cooling are.in accord-l
~
ance with'the acceptable evaluation model.
I
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.1-l I
I I
6-14 HARRIS SER SEC 6
i 15 ACCIDENT ANALYSES The accident analyses for Shearon Harris Units 1 and 2 have been reviewed in l
accordance with SRP 15 (NUREG-0800).
Conformance with the acceptance criteria, except as noted for each of the sections, formed the basis for concluding that the design of the facility for each of the areas reviewed was found to be acceptable.
~
In accordance with SRP 15.1.1, Paragraph I, the applicant evaluated the ability of the Shearo.1 Harris station to withstand anticipated operational occurrences and a broad spectrum of postulated accidents without undue hazard to the.
health and safety of the public. The results of these analyses are used to l
show conformance with GDC 10 and 15.
For each event analyzed, the worst operating conditions a'nd the most limiting single failure were assumed, and credit was taken for minimum engineered safeguards response.
Parameters specific to individual events were conserva-
~ iive'ly 's'e1ecte'd'. 'Two typ'es o'f ' events were anal)zeI: '
~
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s (1) those incidents that might be expected to occur during the lifetime of the reactor (2) those incidents not expected to occur that have the potential to result in significant radioactive material release (accidents)
The nuclear feedback coefficients were conservatively chosen to produce the most adverse core response.
The reactivity insertion curve, used to represent the control rod insertion, accounts for a stuck rod, it is in accordance with GDC 26.
For transients and accidents, the applicant utilized a method that conserva-tively bounds the, consequences of the event by accounting for fabrication and operating uncertainties directly in the calculations.
DNBRs were calculated using.the W-3 correlttion with a modified spacer factor "R," with a minimum DNBR of 1.3 used as the. threshold for fuel failure.
15-1 HARRIS SER SEC 15
. :.: _.=
The applicant accounts for varistions in initial conditions by making the fol-lowing assumptions' as ap;.ropriate for the event being considered:
core power, 2785 MWt, +2%
average reactor vessel temperature (T,yg), 587.5 5.8"F pressure (at pressurizer), 2250
- 30 psi r
The staff concludes trie assumptions for initial conditions are acceptable because they are conservatively applied to produce the most adver.ce effects.
For transients and accidents used to verify the ESF design, the applicant has utilized the safeguards power design value of 291014Wt.
The applicant has also analyzed several events expected to occur one or more times in the life of the plant. A number ?f transients can be expected to occur with moderate frequercy as a result of equipment calfunctions or operator error in' the cCurs' ~ o"f refuel'ing ahd power opirr'aticn during' the plant liYetime'.
~
e r Specific events were reviewed to ensure conformance with the acceptance cri-teria provided in.the SRP.
The acceptance criteria for transients of nioderate frequency in the SRP include the followinc:
(1)
Pressure in the reactor ccol' ant and mai,n steam systems should be maintained below 110% of design values (Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code).
(2)
Fuel clad integrity shall be maintained by ensuring that the minimum DNBR will remain above the 95/95 DNBR limit for PWRs.
(The 95/95 criterion discussed in Section 4.4 of this SER provides a 95% probability, at a 95%
confidence level, that no fuel rod in the core experiences a DNB.)
15-2 HARRIS SER SEC 15-x--
o (3) An incident of moderate frequency shoula not generate a more serious plant condition without other faults occurring independently.
(4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel ele-
}
ment cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod inserta-bility is maintained.
Conformance with SRP ' acceptance criteria consititutes compliance with GDC 10, 15, and 26 of Appendix A to 10 CFR 50.
See Section of this SER for.a discussion of auxiliary feedwater system conformance to TMI Action Plan Item II.E.1.1 and Section for a discussion of compliance with TMI Action Plan Item II.E.1.2.
The transients analyzed are protected by the following reactor crips:
P
'(1) *ponr ranyi hTgh'is0 tron' flux"
- -~ '--
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1 l
(2) high p'ressure (3) low pressurk l
(4) overpower AT (5) overtemperature AT (6) low coolant flow (7) pump undervoltage/underfrequency (8) low steam generator wrter level (9) h'igH steam generator water level 15-3 HARRIS SER SEC 15
Time delays to trip, calculated for each trip signal, are incluaed in the
~
analyses.
See Section of this SER for a discussion of the staff review of reactivity control system functional design.
All of the transients that are expected to occur with moderate frequency can be grouped according to the following plant' process disturbances:
undercooling transients, increased cooling transients, changes in coolant inventory, decrease in reactor coolant flow rate, and changes in core reactivity.
Design-basis accidents have been e, valuated separately and are discussed at the end of this section of the SER.
15/1 Increase in Heat Removal by the Secondary System 1
The applicant's analysis of events that produced increased heat removal by the secondary system is addressed in the following paragraphs.
L 15.1.1 Decrease in Feedwater Temperature
\\
.0 See Section 15.1.4.
l 15.1.2 Increase in Feedwater Flow See Section 15.1.4.
25.1.3 Increase in Steam Flow See Section 15.1.4.
15.1.4 Inadvertent Opening of a Steam Gen.erator Relief Valva or Safety Valve The transient that is.most limiting of these with respect to fual performance l-is the inadvertent cpening of the steam generator relief or safety valve.
The increase steam demand causes a reactor power increase which results in a reactor trip due to high neutron flux or overpower aT signals.
The con-inued steam 15-4 HAREIS SER SEC 15
demand through the open valve will cause additional cooldown and additional positive reactivity insertion to the primary coolant system due to negative temperature coefficient.
The safety injection system (SIS) will inject highly concentrated boric acid from the boron injection tank into the primary coolant
. system on either two out of three pressurizer low pressure signals, two out of three Hi-1 containment pressure signals, 'or two out of three low steamline pressure signals in any one loop.
This ensures the reactor will remain shut-down with any subsequent cooldown.
The normal steam generator feedwater will be. isolated automatically upon STS initiation, and then an orderly cooldown would be effected.
The transient is terminated and only safety grade equipment has been utilized.
DNB dcas not occur during this transient The transient that is most limiting of these with respect to the peak pressure i
i's the increase in feedwater flow transient.
The applicant has calculated a peak prassure of 2230 psia during this transient.
15.1.5 Steamline Rupture e.
,. u.
p 2 The applicant has submitted analyses of postulated steamline breaks that show l
no fuel failures attributed to the accident.
These result are similar to those obtained for previously reviewed Westinghouse three-loop plants.
A postulated double ended rupture at hot standby power was analyzed as the worst case.
The applicant referenced WCAP-9226 as justification for this selection.
WCAP-9226 is curre.itly under review by the staff.
The doubled-cnded rupture would cause the reactor to increase in power due to the decrease in reactor coolant. temperature. The reactor would be tripped by either reactor l
overpower AT or by the actuation of the SIS.
The SIS will be actuated by any of the following:
two out of three low pressurizer pressure signals; two out of three Hi-1 containment pressure signals; or two out of three Hi-1 contain-ment pressure signals or two out of three low steamline pressure signals in any one loop.
The transient is terminated using only safety grade equipment.
The injection of highly borated water ensures the reactor is maintained in a shut-down c'ondition.
i
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15-5 HARRIS SER SEC 15
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o The staff concludes that the consequences of postulated steamline breaks meet the relevant requirements set forth in the GDC 27, 18, 31, and 35 regarding control rod insertability and core coolability and TMI Action Plan Items.
This conclusion is based upon the following:
(1) The applicant has met the requirements of GDC 27 and 28 by demonstrating that the resultant fuel damage was limited such that control rod inserta-bility would be maintained, and that no loss of core cooling capability resulted.
The minimum DNBlfaxperienced by any fuel rod was >1.30, result-ing.in 0%~of the rods experiencing cladding perforation.
(2) The applicant has met the requirements of GDC 31 with resM et to demon-strating the integrity of the primary system boundary to withstand the postulated a:cident.
(3) The applicant has met the requirements of GDC 35 with respect to demon-strating the adequacy of the emergency cooling systems to provide abundant
"*^
core coo 1Tng and IFeactivity control'(via"$5'r'o'n' infiictTon).
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(4) The analyses and effects of steamline break accidents inside and outside containment, during various modes of operation with and without offsite power, have been reviewed and were evitluated using a mathematical model' that has Laen previously reviewed nd found acceptable by the staff.
(5) The parameters used as input to this model were reviewed and found to be suitably conservative.
(6)
The applicant has met the requirements of Task Action Plan Items II.E.1 II.K.2.1, and 'II.E.1.2 with respect t.o demonstrating the adequacy of the auxiliary.feedwater design to remove decay heat following steam system
[
piping failures.
15-6 HARRIS SER SEC 15
l (7) The applicant has met the requirements of Task Action Plan Items II.K.3.25 and II.K.3.40 with respect to demonstrating the integrity and operation of the reacte" oolant pumps to withstand the postulated accident.
(8) The applicant has met the reouirements of Task Action Plan Item II.K.3.5 with respect to the operation and tripping of the reactor coolant pumps.
The assumptions used are conservative and consistent with the general resolution of Item II.K.3.5.
15.2 Decreases in He'at Removal by the Secondary System The applicant's analyses of events that result in a decrease in heat removal by the secondary system are presented below.
15.2.1 Loss of External Load See Section 15.2.7.
l ae u..
l 40
- ~ 15. 2. 2 Turbine Trip See Section 15.2.7.
15.2.3 Loss of Condenser Vacuum See Section 15.2.7.
l 15.2.4 Inadvertent Closure of Main Steam Isolation Valves See Section 15.2.7.
i 15.2.5 -Steam Pressure Regulator Failure See Section 15.2.7.
15-7 HARRIS SER SEC 15
15.2.6 Loss of Nonemergency Power to the Station Auxiliaries See Section 15.2.7.
15.2.7 Loss of Normal Feedwater Flow Plant transients that result in an unplanned decrease in heat removal by the secondary system that might be expected to occur with moderate frequency are identified in the above list.
All these postulated transients have been reviewed.
It was found that the most limiting. event in 'this g'oup of events r
in regard to the maximum pressure within the reactor coolant and main steam systems was the tut-bine trip at full power without credit taken for the pres-
, surizer spray, power-oparated relief valves (PORVs), or steam dump.
The reactor is tripped on the high pressurizer pressure signal and the peak pres-sure during the transient is 2560 pria, well below the ASME requirements for maximum pressure to be limited to 110% of design pressure.
"~The ipplicint Kat'e's lii FSAR Sectfon 15.2'.'6 tFai"tYe niost ' limiting event in fr regard to fuel performance is the loss of nonemergency ac power to the station,'
l
~ auxiliaries transient.
In this transient, the loss of offsite power is closely followed by a turbine trip and reactor trip.
The reactor trip is assumed 'to come froni low-low steam generator level which is the second safety grade trip.
I The emergency feedwater system is automatically started and two pumps are as-sumed to be feeding all three steam generators.
Because only safety grade equipment is used to mitigate the events, the primary system pressurizer relief valves are assumed not to function; therefore all residual heat must be removed through the steam generator safety valves.
The first 4 seconds after the loss of power to the reactor coolant pumps will ~1osely resemble a simulation of the complete loss of flow event (discussed in.Section 15.2.3 of this SER).
DNBR remains above 1.30 throughout the transient.
15-8 HARRIS SER SEC 15
v 15.2.8 Feedwater System Pipe Break The applicant has provided a feedwater line break analysis for Shearon Harris
{
using assumptions that would minimize secondary system heet removal capability, maximize heat addition to the primary system coolant, and maximize the calcu-lated primary system pressure.
A double ended rupture of the largest feedwater line was assumed, as well as failure of the turbine-driven auxiliary feedwater pump to start and supply emergency feedwater to the steam generator.
~
The system code used 'to perform these analyses is LOFTRAN (discussed in SER Section
).
The an.alysis assumed that the most restrictive single failure of the auxiliary feedwater system, emergency feedwater flow is supplied to two intact steam generators by the motor-driven auxiliary feedwater pumps only.
Thi3 is sufficient feedwater flow to adequately remove the residual heat after reactor shutdown.
The use of only safety grade equipment will mitigate this accident.
No fuel damage was calculated to occur, and the peak calculated pressurizer pressure was approximately 2500 psia.
The staff f'-is these results
- ' - ' to be within the regulFed limits. -
c
.1' 15.3 Decreases in Reactor Coolant Flow Rate 15.3.1/15.3.2 Loss of Forced Reactor Coolant Flow, Including Trip of Pump and Flow Controller Halfunctions The applicant has analy7ed the total loss of forced reactor coolant flow event that bounds partial loss of forced reactor coolant flow.
This event is reviewed using the review proe 'ures and accaptance criteria set forth in SRP 15.3.1 and 15.3.2.
The loss of offsite pows
-.sulting loss of all forced coolant flow through the reactor core causes an increase in the average coolant temperature and a decrease in the margin to DNB.
The reactor is tripped from an undervoltage trip monitoring the RCP power supply, and a minimum DNBR of 1.30 is reachet 3.4 seconds into the transient.
The maximum calculated RCS pressure is 2350,
t during the transient.
15-9 HARRIS SER SEC 15
15.3. 3/15.3.4 Reactor Coolant Pump Rotor Seizure and Shaft Break At the staff's request the applicant has reanalyzed these events assuming loss of offsite power to the unaffected RCPs.
The results of the analysis indicated that fuel rods in DNB during the event will not fail and therefore,, the radio-logical consequences would be acceptable.
However, the staff's acceptance criteria in the SRP Section 15.3.3 require that fuel rods in DNB be considered failed for determining radiological consequences.
.The applicant's evaluation of the fuel performance during these events has been referred to the Core Performance Branch for review.
Therefore, we will report the results of that review at a later supplement to this SER.
15.4 Reactivity and Power Distribution Anomalies In the following sections, the staff addresses the applicant's evaluation of events that result in reactivity and power distribution anomalies.
a-e.
.r 15.4.4/15.4.5 Sta-tup o? an Inactive Reactor Coolant. Pump at an Incorrect Temperature In FSAR Section 15.4.4, the applicant provides the results of an analysis for startup of an inactive reactor coolant pump event.
This event is reviewed using the review proce. ores and acceptance criteria set forth in SRP 15.4.4.
During.the first part of the transient, the increase in core flos with cold water results in an ir. crease in nuclear power and a decrease in core average temperature.
Reactivity addition for the inactive loop startup event is the result' of the decrease in core ' inlet water tempercture.
This transient was evaluated by the applicant using a mathematical model that has been reviewed and found acceptable to the staff.
The maximum calculated RCS pressure is 2340 psia and the minimum DNBR is above 1.3 throughout the transient.
15-10 RARRIS SER SEC 15 a..-
15~4.6 Inadvertent Baron Dilution FSAR Section 15.4.6 covers chemical and volume control system (CVCS) malfunc-tions that result in a decrease in the boron concentration in the reactor coolant.
The applicant has not met the requirements of SRP 15.4.6 which requ'res that at least 15 m'inutes be available from the time the operator is made aware of an unplanned boron dilution event to the time a loss of shutdown margin occurs during cold shutdown.
The staff requested that the applicant provide supplemental information in the FSAR section to show how he intends to comply with the SRP.
We will report our evaluation on a later supplement to this SER.
15.5 Increases in Reactor Coolant System Inventory 4
The applicant's analyses of events that result in increase in the primary system inventory are addressed below.
....s...
..1 InadvertentOperationoftheEmergency'C'o'r'eCoofing'SystemDuring
~
Power Operation See Section 15.5.2-l 15.5.2 CYCS Malfunction that Increases Reactor Coolant Inventory ECCS operation could be initiated by a spurious signal or operstor error.
Two cases were examined, one in which reactor trip occurs simultaneously as a re-sult of the safety injection signal (the Shearon Harris plant has an automatic reactor trip on SI signal), and the other in which reactor trips occur later i
in the transient because of low reactor coolant system (RCS) pressure.
The reactor pressure decreases during initial phase of the transient and reaches a peak pressure of 2350 psia at ISO seconds into the transient.
The DNBR never drops below its initial value for both cases of this transient.
All of these transier.ts are terminated by use of safety grade systems only.
The applicant's evaluation of the CVCS malfunction event is presented in FSAR Section.15.4.6 and the staff evaluation is in Section 15.4.6 of t~,is SER.
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15.6 Decreases in Reactor Coolant Inventory The applicant has analyzed the following event that results in a decrease in reactor coolant inventory.
- 15. E.1 Inadvertent Opening of a. Pressurizer Safety or Relief Valve In, FSAR Section 15.6.1, the applicant provides the results of an analysis for inadvertent opening of a pressurTzer safety valve.
During this event, nuclear power is. maintained at the initial value until reactor trip occurs on low pressurizer pressure.
The DNBR decreases initially, but increases rapidly followir.c the trip.
The minimum DNBR of 1.51 occurred at 27 seconds into the transient.
The RCS pressure decreases throughout the transient.
15:6.3 Steam Generator Tube Rupture In response to the staff's request, the applicant has provided additional data
" ~ ~Yega'rdiNg the Ty'stims' Eesponsi'and radiciogicaY'c'o'ffsequsnce's after a staam C' generator tube rupture accident. This information, however, did not support L
the isolation time of the affected steam generator at 30 minutes.
Upon receipt of additional information, the staff will complete the review of the radiologi-cal consequences of this accident, and complete Table 15.5.
15.6.5 LOCA In FSAR Section 15.6.5, the applicant has analyzed the double ended cold leg l-guillotine (DECLG) as the most limiting large-break LOCA.
The analysis is done l
using three.di& rent flow coefficients.
The results of these show that the DECLG with a Moody break discharge coefficient of 0.4 is the worst case.
In this analysis, peak clad temperature reached is 2181*F.
For the small-break LOCA the applicant has determined that a cold leg rupture of less'than 10-in.
diameter is the most limiting.
The analysis was performed for 3-in., 4-in. and 6-in. diameter breaks.
The results show that the 3-in. diameter break is the worst case, and it results in a peak clad temperature of 180MF.
Both of these
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Table 15.5 Assumptions used for the calculations of the radiological conse-i quences of.a postulated steam generator tube rupture accident" l'.
The rupture is a double-ended guillotine break, resulting in an l
average leakage of lb/sec from the reactor coolant system to the steam generator secondary side.
l
~2.
Concentration of iodine (as dose equivalent I-131) in primary coolant at start of accident:
60 pCi/g (technical specifications for maximum concentration allowed during a short time), and 1.0 pCi/g (equilibrium technical specification l
limit).
(Two different cases.)
3.
Secondary coolant activity at start of accident:
0.1 pCi/g l
dose equivalent I-131, technical specification limit.
4.
Average ratio of iodine mass concentration in the steam to that in the secondary side water, for both the affected and unaffected (but i
l 1eaking slightly) steam generators:
0.1.
5.
Steam release from p.ffected generator over a 0-2 hour period:
lb.
6.
Ts~oTation of affected steam generator at minutes.
7.
Primary to secondary leak rate of to ea'ch of the three unaffected steam generators.
-8.
Iodine release rate from fuel increases by a factor of 500 over the equilibrium release rate.
9.
Condenser vse-if Tost at-time of scram,1rticut-11-mi'nutes after the-l rupture.
=.
- The blanks in this table will be filled in when the open issue is resolved accidents are terminated by SIS and ECCS operations.
Only safety grade equip-ment is utilized to mitigate the accident.
The staff concludes that the LOCA analysis resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary is acceptable and meets the relevant requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50, GDC'35, and 10 CFR 100.
t The applicant has performed analyses of the performance of the ECCS in accord-ance with the Commission's regulations (10 CFR 50.46 and Appendix K to 10 CFR 50).
The analyses considered a spectrum of postulated break sizes and locations and were performed with an evaluation model that has been previously reviewed and 15-13 HARRIS SER SEC 15
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approved by the staff, as described in NUREG-0390.
The results of the analyses show that the ECCS satisfy the following criteria:
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(1) The calculated maximum fuel rod cladding temperature does not exceed 2200*F.
]
(2) The calculated maximum local oxidation of the cladding does not exceed 17% of the total cladding thic'kness before oxidation.
(3) The calculated otal amount of hydrogen generated from the chemical reaction of the cladding with water or steam does not exceed 1% of the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4) Calculated changes in core geometry are 1,uch that the core remains amenable to cooling.
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.- (5) After any calculated successful initial operatiop of the ECCS, the cal-culated core temperature is maintained at an acceptably low value and decay heat is removed for the extended period of time required by the long-lived radioactivity.
(6) The applicant has met the requirements of TMI Action Plan Items II.E.2.3, II.K.3.5, II.K.3.25, II.K.3.30, and II,.K.3.31.
l The radiological consequence meets 10 CFR 100 requirements for the postulated L
spectrum of LOCAs which were evaluated from the viewpoint of site acceptability.
I l
For the purposes of this analysis, large fractions of the fission products I
were assumed to be released from the core even though these releases would be precluded by the performance of the ECCS.
The staff concludes that the calculated performance of the ECCS following a postulated LOCA accident and the conservatively calculated radiological 15-14 HARRIS SER SEC 15 v
consequences of such an accident conform to the Commission's regulations and to applicable regulatory guides and staff technicial positions, and the ECCS performance is considered acceptable for the postulated accident.
15.9 TMI Action Plan Recuirements 15.9.1 II.B.1 Reactor Coolant System Vents The design for the reacter coolaht system vents is described in the THI Appendix of the FSAR and the applicant's submittal of August 11, 1983.
The applicant in its August 11, 1983 submittal, at the staff's request, has identified.the design criteria, location,' and failure modes and' effects analysis for these
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vents.
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The Reactor Systems aspects of the reactor coolant system vents have been I
i reviewed and we conclude that they meet the requirements set forth in NUREG-0737 and they are therefore acceptable.
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2'- Acceptability of the qualification rekuirements, instrumentation and control and operating precedures for these vents is addressed in sections of this SER.
15.9.2 II.K.1.5 Review ESF Valve Positions, Controls, and Related Test and i
Maintenance Procedures To Assure Proper ESF Functioning In rssponse to the above requirement, the applicant has stated that proper ESF functioning will be verified through completion of the applicable portions of the start-up test progran prior to fuel load.
Based on the above, we conclude that the applicant's commitments meets the schedular guidelines of this item and is acceptable.
- Section numbers to be provided later 15-15 HARRIS SER SEC 15
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.a 15.9.3 II.K.1.10 Review and Modify Procedures for Removing ESF from Service to' Assure Operability Status is known The applicant is presently preparing procedures which will govern the pre-maintenance and post-maintenance safety-related system status.
In order to remove safety-related equipment from service, two operations documents will be initiated.
One is an Operations Work Procedure (OWP) which will define the testi.ng prior to~and during maintenance and it will define l
test requirements.
Double verification line-ups will be provided on the OWP for " maintenance positions" and " return to normal positions." The second I
operations document is the Equipment Inoperable Record (EIR) which will log the equipment out of service and will keep a running total of any time limita-tions that apply.
The EIR will be updated at the beginnir.g of each shift as part of tne shift foreman's turnover procedure.
We find the applicants proposed procedures for removing ESF from service-
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~15.9.4 II.K.2.13 Thermal Mechanical Report Effect of High Pressure Injection on Vessel Intagrity for Small-Break LOCA with no Auxiliary l
Feedwater Staff review of this item will be covered in NRC unresolved safet9 issue A-49
" Pressurized Thermal Shock."
15.9.5 II.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients n
Westinghouse has performed a study which addresses the potential for void formation in Westinghouse 'Jesigned NSSS during natural circulation cooldown/
1
.depressurization transients.
This study has been submitted to the NRC by the Westinghouse Owners Group. 'This Westingheuse report is being reviewed by the a
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staff.
Shearon Harris plant will be required to modify the operating procedures, if required, after the sf.aff completes its evaluation of the Westinghouse topic report.
15.9.6 II.K.2.19 Sequential Auxiliary Flow Analysis Sequential auxiliary feedwater flow analytical requirements is only of concern to once-through steam generator designs.
Since Westinghouse utilizes inverted U-tube steam generator designs, ~ requirements set forth by Item II.K.2.1S are not applicable to Shearon Harris Plant.
15.9.7 II.K.3.1 Installation and Testing of Automatic PORV Isolation System As a response to II.K.3.2, the applicant referenced a generic Westinghous::
Owner's Group submittal.
Should staff generic review of 'this material conclude otherwise, NRC will require further consideration of the modification on Shearon Harris.
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15.9.8 II.K.3.3' Reporting SV and 00RV Challenges and Failures In response to the above requirsment, the applicant stated that any failure of n relief or safety valve will be reported to NRC in accordance with plant
" Licensee Event Report" procedures.
All challenges to a relief or safety valve will be documented in the olant annual report.
Based on the above, we l
conclude that the Shearon Harris procedures meet the requirement of this item and are acceptable.
1 1
15.9.9 II.K.3.5 Automatic Trip of RCDs During LOCA In response to this requirement. the app 1'icant stated that Westinghouse per-i formed an analysis of delayed RCP trip during LOCA.
This analysis is documented and is the basis for the Westinghouse position on RCP trip (i.e., automatic RCP trip is not necessary because sufficient time is available for manual tripping of the RCPs).
.... =. -. -...
I, The staff will issue requirements for RCP trip criteria in the near future for the applicant to implement.
15.9.10 II.K.3.10 Proposed Anticipatory Trip Modification The applicant has not proposed any modification to its standard anticipatory trip, fherefore, no TMI action plan requirements are imposed.
~
15.9.11 II.K.3.17 Report on Odfages of ECCS Before issuing the operating license, the applicant will be required to submit to NRC a procedure for collecting and submitting information concerning ECCS outage.
The staff will report its evaluation of this item in a supplement to this SER'.
15.9. 12 II.K.3.25 Effect of Loss of AC Power on RCP Seals
""*' ' ' In resp 6nse to 'thi's' rk(uireneht, the app 1'icanf* stated 'that-in the event of los's
~
21' of offsite power, the RCP motor is de-energized, the diesel generators are auto-matica11y started, and either seal injection flow or component cooling water to the thermal barrier heat exchanger is automatically restored within seconds.
Either.of these cooling supplies is adequate to provide seal cooling and pre-vent seal failure due to loss of seal cooling during a loss of offsite power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
l Based on the above, the staff concludes that the applicant's design meets the requirement of this item and is acceptabic.
15.9.13 II.K.3.30 Revised small-Break LO.CA Methods To Show Compliance with l
In response to this requirement, the applicant stated that Westinghouse has submitted a new small-break' evaluation model to NRC.
The staff's review of
~
this submittal is ongoing.
However, the st=ff has determined that the present l
15-18 HARRIS SER SEC 15
_ - _ _ _ _ _,. =. - -. - ~. - - - =
model is in compliance with Appendix K.
Should changes be required as a result of II.K.3.30 review, reanalysis will be accompl'ished as part of II.K.3.31 review.
15.9.14 II.K.3.31 Plant-Specific Calculations To Show Compliance with 10 CFR 50.46 In response to the above requirement, the applicant stated that he will submit the necessary specific calculations to show compliance with 10 CFR 50.46 either within 1 year of NRC approval of the Westinghouse model or before fuel load, whichever occurs later.
The staff concludes that this commitment meets the schedule requirement of this item and is acceptable.
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15-19 HARRIS SER SEC 15
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model is in compliance with Appendix K.
Should changes be required as a result of II.K.3.30 review, reanalysis will be accomplished as part of II.K.3.31 review.
15.9.14 II.K.3.31 Plant-Specific Calculations To Show Compliance with 10 CFR 50.46 In response to the above requirement, the applicant stated that he will submit I
the necessary specific calculations to show compliance with 10 CFR 50.46 either within 1 year of NRC approval of the Westinghouse model or before fuel load, whichever occurs later.
The staff concludes that this commitment meets the 1
schedule requirement of this item and is acceptable.
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15-19 HARRIS SER SEC 15