Text

Fowards Revised Responses to Acceptance Review Requests 212.14,212.76,212.104,223.38,223.72 & 223.73,re Level 8 Instrumentation,Reactor Pump Trip Electrical Schematics & Rod Sequence Control,To Be Included in Next FSAR Update
	ML20090F182
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	06/28/1983
From:	James Smith; LONG ISLAND LIGHTING CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	SNRC-924, NUDOCS 8307050091
	Download: ML20090F182 (81)
	v • d • e

,

O O

7 LONG ISLAND LIGHTING COM PANY kh SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD + WADING RIVER. N.Y.11792 Dkat Dial Number June 28, 1983 SNRC-924 Mr. Harold R.

Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Revised Responses to Acceptance Review Requests 212.14, 212.76, 212.104, 223.38, 223.72 and 223.73 Final Safety Analysis Report (FSAR)

Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

LILCO's efforts are directed at resolving all outstanding NRC Staff concerns associated with their acceptance review.

Therefore, we are providing as an enclosure to this letter, o

current and complete information to the above subject acceptance review requests.

These six requests fall into three subject areas:

1)

Level 8 instrumentation (212.104)

2) Recirculation Pump Trip electrical schematics (212.14, 212.76 and 223.73) and 3) Rod Sequence Control System electrical schematics (223.38 and 223.72).

These revised responses and the associated figures will be included in the next FSAR update and are provided herein to allow for a timely staff review.

Should you have any questions concerning the enclosed subject material do not hesitate to ct.11 this office.

Very truly yours, J.

L.

Smith Manager, Special Project S5oreham Nuclear Power Station

,Sj

)

GJG:bc 3

Enclosure cc:

J.

Higgins All Parties Listed in Attachment 1 8307050091 830628 PDR ADOCK 05000322 FC-8935.1 A

PDR m-

m ATTACHMENT 1 Lawrence Brenner, Esq.

Herbert H. Brown, Esq.

Administrative Judge Lawrence Coe Lanpher, Esq.

Atomic Safety and Licensing Karla J. Letsche, Esq.

Board Panel Kirkpatrick, Lockhart, Hill U.S. Nuclear Regulatory Commission Christoper & Phillips Washingten, D.C.

20555 8th Floor 1900 M Street, N.W.

Washington, D.C.

20036 Dr..' Peter A. Morris Administrative Judge Atomic Safety and Licensing Mr. Marc W. Goldsmith Board Panel Energy Research Group U.S. Nuclear Regulatory Commission 4001 Totten Pond Road W shington, D.C.

20555 Waltham, Massachusetts 02154 Dr. James H. Carpenter MHB Technical Associates Administrative Judge 1723 Hamilton Avenue Atomic Safety and Licensing Suita K

~

Board Panel U.S. Nucicar Regulatory Commission

. San Jose, California 95125 Washington, D.C.

20555 Stephen B. Latham, Esq.

Twomey, Latham & Shea Daniel F. Brown, Esq.

33 West Second Street Attorney P.O. Box 398 Atomic Safety and Licensing Riverhead, New York 11901 Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Ralph Shapiro, Esq.

Cammer and Shapiro, P.C.

9 East 40th Street Bernard M. Bordenick, Esq.

New York, New York 10016 David A.

Repka, Esq.

U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Matthew J. Kelly, Esq.

State of New York Department of Public Service James Dougherty Three Empire State Plaza 3045 Porter Street Albany, New York 12223 Washington, D.C.

20008 9

SNPS-1 FSAR Request 212.104 (RSP) (15.0):

In analyzing anticipated operational transients, the applicant has taken credit for plant operating equipment which has not been shown to be reliable as required by General Design Criterion 29.

The staff has discussed the application of this equipment generically with General Electric.

In these discussions General Electric has stated that the most limiting transient that takes credit for this equipment is the excess feedwater event.

Further, General Electric has stated that the only plant operating equipment that plays a significant role in mitigating this event is the turbine bypass system and the Level 8 high water level trip (closes turbine stop valves).

We will allow the use of the turbine bypass and Level 8 high water level trip systems in mitigating transients except for the turbine trip and generator load rejection without bypass transients which are currently minimum critical power ratio-limiting.

To assure an acceptable level of performance, it is the staff's position that this equipment be identified in the plant Technical Specifications with regard to availability, set points, and surveillance testing.

The applicant must submit his plan for implementing this requirement along with any system modifications that may be required to fulfill this requirement.

Response

In discussions between GE and the NRC on November 20 and 21, 1978, GE presented the results of transient analysis performed to design basis accident condition assumptions (i.e.,

equipment availability).

The analysis indicated that failure to give credit to the Level 8 turbine trip and the main turbine bypass system could result in a difference in the critical power ratios of 0.02 and 0.08, respectively.

Therefore, these postulated conditions could not result in unacceptable impacts on the health and safety of the public.

The Level 8 instrumentation will be subject to technical specification requirements associated with the feedwater system / main turbine trip system.

The proposed Shoreham technical specification addresses this concern in Section 3.3.9 of the limiting conditions for operation.

The turbine steam bypass system and stop valves are furnished with the main turbine generator and have exhibited high reliability on existing nuclear and fossil fueled operating units.

Normal operating procedures require that the valves be functionally exercised periodically in accordance with vendor recommendations.

This effort will ensure valve operability and l

provide adequate assurance that the valves will operate when required.

212-104 l

SNPS-1 FSAR i

l l

Request 212.14 (5.2.2.7.6 15.1.1.1) :

In order to assess the effectiveness of PRT, the following i

information, as it relates to the Shoreham plant, is required.

a.

Provide turbine trip transient analyses assuming scram on turbine stop

valve, position, recirculation pump trip, and no turbine bypass.

The results should include pressure, CPR, neutron flux, surface heat flux and fuel pellet temperature as a

function of time for the following conditions:

1) without

PkT, using design conservatism factors on void coefficients and scram reactivity curve.

2) without PRT, using expected operational factors on the scram reactivity curve and void coefficient (i.e.,

best estimate of actual factors at m;OC for an equilibrium core).

3) part 1) with PRT 4) part 2) with PRT b.

Provide generator trip transient analyses assuming scram on turbine control valve position, with bypass, without recirculation pump trip, and using design conservatism factors on the scram reactivity curve and the void coefficient (i.e.

more mild, more frequently expected transient).

The results should include

pressure, CPR, neutron

flux, surface heat flux and fuel pellet temperature for the following conditions:

1) with PRT i

2) without PRT c.

On currently operating plants, what is the relief valve operation frequency that has been experienced?

That is, what is the number of individual valves that will be expected to open per unit time of operation, counting an event that causes 3 valves to open as 3 instances of operation, etc?

d.

What is the expected increase in the part c) frequency with PRT for the Shoreham plant?

e.

What does available data show regarding failure

{

frequency of S/R valves to reseat properly following opening?

i f.

Is there any reason to expect that the S/R valves designed for the Shoreham plant will have a

different i

failure (to close) frequency?

212-14 Revision 7 - August 1977

SNPS-1 FSAR g.

Will the specified number of operating cycles for the lifetime of the Shoreham S/R valves be increased to account for the additional duty due to PRT operation.

If so, what is the increase both in number and as a

percent of total duty cycles?

h.

Will the specified lifetime number of rapid depressurization cycles for the reactor pressure vessel be increased to allow for improper PRT operations (i.e.

stuck open relief valves) ?

If so, by how many cycles?

1.

Will the specified lifetime total number of thermal cycles for the suppression pool be increased to account for both partial (proper PRT operation) and complete (improper PRT operation, stuck relief valve) blowdowns due to PRT?

If so, ny how many cycles?

j.

Will the specified number of dynamic pressure cycles for the suppression pool and associated piping be increased to account for PRT operation?

If so, by how many cycles?

k.

Provide analyses for all accidents for which the PRT system is expected to function.

Provide information to show whether criteria which must be met for each accident are still satisfied assuming the PRT functions designed.

Accidents considered should include LOCA as (steam line and recirculation line breaks) and ATWS.

1.

Provide of the PRTa summary table of advantages and disadvantages system For

example, our tentative and incomplete table included:

ADVANTAGES; decreased duty cycle (power, temperature,

pressure, etc.)

on

fuel, decreased duty cycle (pressure) on RPV assuming proper PRT operation.

DISADVANTAGES; increased plant complexity, increased duty cycle (cooldown transient for improper PRT operation),

increased duty cycle on suppression pool.

m.

Provide the results of comparative analyses (with and without PRT) of expected radiation exposures, both to plant operating personnel and

offsite, for normal operations, transients and accidents.

What scram reactivity (or related) experiments have been n.

run?

How do the results of any such tests compare l

quantitatively with calculational results from the

(

latest models?

Is there operating plant transient information which would be relevant to the assessments for the need for PRT?

If so, please identify and discuss.

212-14a Revision 7 - August 1977

l SNPS-1 FSAR

Response

PRT has been removed from the Shoreham plant design due to technical and licensing concerns.

As an alternative to

PRT, General Electric has designed a recirculation pump trip (RPT) feature to operate on turbine trip and generator load rejection transients.

A description of the recirculation pump trip system is contained in Section 7.6.1.3.

An analysis of conformance to Regulatory Guides and Industry Standards is found in Section 7.6.2.3.

The drawings which are applicable to the RPT system are specified in the response to Request 212.76.

212-14b

SNPS-1 FSAR Request 212.76(15.1):

Several transient analyses in this section take credit for the recirculation pump trip (RPT) function.

Description of these trip circuits is not presented in the FSAR.

Discuss operation of this system as it was assumed to function to mitigate the transients analyzed in Section 15.

Describe the signals or mechanisms that actuate the tripping devices and include setpoint values.

If the recirculation pump trip is initiated by tripping the M-G set field breaker or a generator output breaker (rather than the drive motor breaker),

the coastdown apparently would be more rapid than that presented in NEDO-10802.

Provide details of the analytical models and methods used to predict recirculation system behavior during transients which take credit for RPT.

Justify the time delay used for two pump trip.

Discuss plans for startup tests which will justify the analytical methods and verify that core flow during the coastdown is in agreement with that predicted.

A loss of off-site power (all grid connections) associated with a LOCA may cause a generator load rejection and RPT at the time of LOCA initiation.

With regard to a concern that installation of RPT may compromise the recirculation pump coastdown assumed in l

the LOCA analysis, show that RPT would not cause a coastdown of the recirculation pumps more rapidly than that assumed.

l

Response

1.

The RPT system design and functional requirements are described in Sections 7.6.1.3 and 7.6.2.3.

Figure 7.6.1-3 is I

a simplified functional control design which shows the actuation signals and the permissive functions for recirculation pump trip.

Figures 4.2.3-8A to -C, 5.1.2-1A to -C, 5.2.2-2A and -B, 7.7.1-1A to -C, 7.2.1-1A to -D, 7.3.1-7A to

-E, 7.3.1-12A to

-E, 7.7.1-2A to -G, and 7.7.1-5A to -E have been revised to include the RPT system.

2.

The RPT system trips the redundant line breakers which are located between the motor generator and the pump motor.

Thus, for events during which the RPT system is outlined, the coastdown characteristic of the pump, pump

motor, and the pump shaft is an important aspect of the transient.

A description of these characteristics follows:

Section 2.12 of NEDO-10802 includes the following equation:

vJ dn AT

=

30 ge 3t 212-76

~

SNPS-1 FSAR Request 223.73:

Supplement the description of the recirculation pump trip presented in Revision 5 to the FSAR with revised electrical schematics.

Response

Revised drawings to reflect the incorporation of recirculation pump trip are specified in the response to Request 212.76.

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_ 83 0 7 o s o o_o_, - 2,B_
e SNPS-1 FSAR Request 223.38 (7.0) (15.0)
The Rod Sequence Control System is assumed to function in Section 15.0 of the FSAR to mitigate or prevent several accidents and as such it must be designed in accordance with criteria equivalent to that of protection systems.
Provide the design basis description, drawings and all other information for this system.
~~Response~~
The rod sequence control system (RSCS) is described in Sections 7.7.1.12 and 7.7.2.12.
The RSCS is not relied upon for the safety action required for the control rod drop accident or_the continuous control rod withdrawal during startup accident.
The RSCS is designed as a
backup to procedural controls on the movement of control rods.
There are no specific regulatory requirements for the RSCS.
The RSCS is shown in the control rod drive hydraulic system FCD, Figures 7.7.1-2A through G.
The RSCS acts to prevent withdrawal of an out-of-sequence control rod, to prevent continuous control rod withdrawal errors during reactor
~~startup, and to minimize the core reactivity transient during a rod drop accident.~~
The consequences of a rod withdrawal error in the startup range were generically analyzed in NEDO-10527, demonstrating that the licensing basis criterion for fuel failure is still satisfied even when the RSCS fails to block rod withdrawal.
The safety action required for the control rod withdrawal accident (a reactor scram) is provided by the safety-related intermediate range monitor (IRM) subsystem of the neutron monitoring systems (NMS).
If the core flux scram trip setpoint is reached during a flux transient, the IRM or a second safety-related NMS scram
~~trip, supplied by the average power range monitor (APRM), can initiate a scram to terminate the core power transient.~~
Section 15A.l.12 has been revised to reflect the ultimate safety actions provided by the IRM or APRM.
223-38
SNPS-1 FSAR Request 223.72:
With respect to the rod sequence control system described in Revision 5 to the FSAR, provide the functional control diagrams and as-built drawings supported with explanations, as deemed necessary, to permit an independent evaluation of the system design depicted in the diagrams and drawings.
~~Response~~
The rod sequence control system (RSCS) is a backup to procedural controls governing the movement of control rods.
There are no specific regulatory requirements for the RSCS.
The purpose and description of the RSCS are found in Sections 7.7.1.12 and 7.7.2.12.
The RSCS is depicted in the control rod drive hydraulic system FCD, Figures 7.7.1-2 A through G.
The RSCS acts to prevent withdrawal of an out-of-sequence control rod, to prevent continuous control rod withdrawl errors during reactor
~~startup, and to minimize the core reactivity transient during a rod drop accident.~~
The consequences of a rod withdrawal error in the startup range were generically analyzed in NEDO-10527, demonstrating that the licensing basis criterion for fuel failure is still satisfied even when the RSCS fails to block rod withdrawal accident (a reactor scram) is provided by the safety-related intermediate range monitor (IRM) subsystem of the neutron monitoring systems (NMS).
If the core flux scram trip setpoint is reached during a flux transient, the IRM or a second safety-related NMS scram trip, supplied by the average power range monitor (APRM), can initiate a scram to terminate the core power transient.
223-72
t SNPS-1 FSAR hoists.
The platform operator can immediately determine whether the platform and hoists are responding to his local instructions and
~~can, in conjunction with the main control room operator, verify proper operation of each of the three categories of interlocks listed previously.~~
7.7.1.12 Rod Sequence Control System 7.7.1.12.1 System Identification l
The rod sequence control system is a subsystem of the reactor manual control system.
1 The purpose of the rod sequence control system (RSCS) is to reduce the consequences of the postulated rod drop
~~accident, to prevent withdrawal of an out-of-sequence control rod, to prevent continuous control rod withdrawal errors during startup.~~
The RCCS accomplishes this by restricting the patterns of control rods that can be established to predetermined sets.
7.7.1.12.2 Power Sources The RSCS will operate from the same instrument bus as the rod position information system (RPIS), the subsystem that is the primary data source for the RSCS.
The RSCS is designed so that it will apply rod movement inhibits to the rod drive control system (RDCS) in the event of loss of input power.
7.7.1.12.3 Equipment Design 1.
General Description The RSCS is a
subsystem of the reactor manual control system (RMCS).
It receives inputs from the RPIS and the RDCS, both of which are also subsystems of the RMCS, and from the operator directly.
The RSCS provides display outputs directly to the operator and rod movement interlocks to the RDCS.
The RSCS has five primary functional blocks plus buffering and interfacing hardware.
The five are as follows:
a.
rod pattern controller, b.
substitute position generator, c.
operator display, d.
tester, and e.
bypassed rod identifier.
Each of these is described in Section 7.7.1.12.3.4.
7.7-57 i
SNPS-1 FSAR refueling interlocks are not required for any postulated design basis accident or for safe shutdown.
Furthermore, the interlocks are required only for the refueling mode of plant operation.
~~However, the system does conform to the specific industry standard and regulatory guide listed on Fig.~~
7.1.1-2, while the requirements of 10CFR50 Appendix B are met in the manner set forth in Chapter 17.
7.7.2.12 Rod Sequence Control System 7.7.2.12.1 Conformance to General Functional Requirements The RSCS provides backup to procedural controls, by imposing restrictions on the movement of control rods to reduce the consequences of a
postulated rod-drop accident, to prevent the withdrawal of an out-of-sequence control rod and to prevent continuous rod withdrawal errors during startup.
Both the RWM and the RSCS are compatible and redundant to each other.
7.7.2.12.2 Conformance to Specific Regulatory Requirements There are no specific regulatory requirements for the RSCS.
The quality level of design and hardware of the RSCS is equivalent to that of a
single channel of a
protection system.
The RSCS neither interfaces with the RPS nor affects the ability of the reactor to respond to any requirement of the RPS.
7.7.2.13 Neutron Monitoring System Instrumentation and Controls 7.7.2.13.1 Source Range Monitor Subsystem 1.
General Functional Requirements Conformance The arrangement of the neutron sources and startup chambers in the reactor is shown on Fig. 7.7.1-11.
This arrangement produces at least three counts per sec in the SRM using the sensitivity noted in Section 7.7.1.13.1 and the design source strength at initial reactor start-up.
If the discriminator setting is adjusted to produce the specified sensitivity, the signal-to-noise count ratio is well above the 2:1 design basis for cold start-up.
If the multiplication of one section of the core increases to put that section of the reactor on a 20 see
~~period, the nearest SRM chamber shows an increase in count rate.~~
In general, at least one detector indicates the change in multiplication (Figs. 7.7.2-1 through 7.7.2-6).
7.7-80
I SNPS-1 FSAR Normal startup procedures ensure that withdrawal of control rods is distributed about the core to prevent excessive multiplication in any one section of the core.
~~Hence, each SRM chamber can respond in some degree.~~
4 i
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7.7-80a/b
SNPS-1 FSAR This unlikely set of circumstances makes possible the rapid removal of a control rod.
The dropping of the rod results in a
high local K in a small region of the core.
For large, loosely coupled cores, this would result in a
highly peaked power distribution and subsequent shutdown mechanisms.
Significant shifts in the spatial power generation would occur during the course of the excursion.
Therefore, the method of analysis must be capable of accounting for any possible effects of the power distribution shifts.
In order to limit the worth of the rod which could be dropped, the rod sequence control system (RSCS) is installed.
This system prevents the movement of an out-of-sequence rod in the 100 to 75 percent rod density range, and from the 75 percent rod density point to the preset power level the RSCS will only allow group i
notch mode rod withdrawals or insertions.
The 75 percent rod density configuration corresponds to the condition in which I
75 percent of the rods are fully withdrawn.
With the condition that no out-of-sequence rod may be moved, the postulated rod drop accident cannot result in peak enthalpies in excess of 280 calories /g for any possible plant operation or core exposure I
conditions.
Table 15.1.33-1 presents the parameters used in this analysis.
15.1.33.3 Accident Description The accident is defined as:
1.
The rod worth minimizer (RWM) does not function.
2.
The highest rod worth that can be developed at any time in core life under any operating conditions, by one inadvertent operator
~~error, drops from fully inserted position to fully withdrawn position.~~
3.
The rod drops at the maximum speed of 2.79 ft/sec.
4.
The scram time is 5
see to the 90 percent insertion point.
5.
The RSCS is assumed to function.
The sequence of events and the approximate times of occurrence are as follows:
Approximate Elapsed Event Time, seconds i
1.
Reactor is operating at 75 per-cent control rod density pattern 2.
RWM is not functioning 3.
Maximum worth control blade becomes decoupled 15.1-22
.i
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SNPS-1 FSAR 15A.l.ll.5 Radiological Consequences An evaluation of the radiological consequences was not made for this event since no radioactive material is released from the fuel.
j 15A.l.12 Continuous Rod Withdrawal During Reactor Startup i
15A.1.12.1 Identification of Causes While operating in the power, source, and/or intermediate range of operation, the reactor operator makes a procedural error and withdraws the maximum worth control rod continuously.
15A.l.12.2 Sequence of Events and Systems Operation control rod withdrawal errors are highly improbable in the startup power range.
The RSCS and the RWM prevent the operator from selecting and withdrawing an out-of-sequence control rod.
Continuous control rod withdrawal errors during reactor startup are precluded by the RSCS.
The RSCS prevents the withdrawal of I
an out-of-sequence control rod in the 100 to 75 percent control rod density range and limits rod movement to the banked position 8
(
mode of rod withdrawal from the 75 percent rod density to the-i desired power level.
Since only in-sequence control rods can be I
withdrawn in the 100 to 75 percent control rod density and control rods are withdrawn in the banked position mode from the i
j 75 percent control rod density point to the desired power level, i
there is no basis for the continuous control rod withdrawal error in.the startup power range.
See Section 15A.l.ll for description of continuous control rod withdrawal during power range operation.
The bank position mode of the RSCS is described in Reference 4.
No operator actions are required to preclude this event since the plant design as discussed above prevents its occurrence.
For any operation involved in a possible initiating failure (or i
error), the necessary safety actions are taken (rod blocks) prior to any possible subsequent single failure.
The probability-of initiating causes (or multiple errors) for this event alone is considered low enough to warrant its being categorized as an infrequent incident.
The probability of further development of this event is extremely low'because it is l
contingent upon the simultaneous failure of two redundant systems, the RSCS and RWM systems concurrent with high worth rod, out-of-sequence rod selection contrary to procedures, plus operator nonacknowledgment of continuous alarm annunciations prior to safety system-actuation.
15A-38 1
,~..._.~.,___m_._..
4 l
SNPS-1 FSAR Not withstanding the design provisions to preclude rod withdrawal, a special analysis is included (Section 15A.1.12.4) to shoW' that continuous withdrawal of an out-of-sequence rod in the startup range results in acceptable peak fuel enthalpies less than the licensing basis criterion.
15A.l.12.3 Core and System Performance The performance of the RSCS and RWM prevent erroneous selection and withdrawal of an out-of-sequence control rod, as described in Section 15A.l.12.2.
Thus, the core and system performance is not affected by such an operator error.
1 15A.l.12.4 Special Analysis The continuous control rod withdrawal analysis in the startup range was performed to demonstrate that the licensing basis criterion for fuel failure will not be exceeded when an out-of-sequence control rod is withdrawn at the maximum allowable normal drive speed.
The sequence and timing assumed in this special analysis is shown in Table 15A.l.12-1.
The rod sequence control system (RSCS) and the rod worth minimizer (RWM) constraints on rod sequences will prevent the continuous withdrawal of an out-of-sequence rod.
This analysis was performed to demonstrate that, even for the unlikely event j
where the RWM and RSCS fail to block the continuous withdrawal of an out-of-sequence rod, the licensing basis criterion for fuel failure is still satisfied.
The methods and design basis used for performing the detailed analysis for this event, are similar to those previously approved for the control rod drop accident (CRDA) (References 10, 11, and 12).
Additional simplied point model kinetics calculations were performed to evaluate the dependence of peak fuel enthalpy on the control blade worth.
For the detailed calculation, the-50 percent control rod density pattern was selected as the initial starting condition which is consistent with the approved design basis for the CRDA (References 10, 11, and'12).
The licensing basis criterion for fuel failure is the contained energy of a fuel pellet located in the peak power region of the core shall not exceed 170 cal /gm-UOs.
15A.1.12.4.1 Methods of Analysis Since the rod worth calculations using the approved design basis methods (References 10, 11, and 12) use three-dimensional
~~geometry, it is not practical to do a detailed analysis of this event parameterizing control rod worths.- Therefore, the methods of' analysis employed were to perform a detailed evaluation of 1'~~
this event for a typical BWR and control rod worth (1 percentak) 15A-38a
SNPS-1 FSAR and to use a point model calculation to evaluate the results over i
the expected ranges of out-of-sequence control rod worths.
The detailed calculations are performed to demonstrate (1) the consequences of this event over the expected power operating range and (2) the validity of the approximate point model calculation.
The point model calculation will demonstrate that the licensing criterion for fuel failure is easily satisfied over the range of expected out-of-sequence control rod worths.
These methods are described in more detail below.
The methods used to perform the detailed calculation are identical to those used to perform the design basis control rod j
drop accident with the following exceptions:
a.
The rod withdrawal rate is 3.6 fps rather than the blade j
drop velocity of 3.11 fps.
b.
Scram is initiated either by the IRM or 15 percent APRM j
scram in the startup range.
The IRM system is assumed j
to be in the worst bypass condition allowed by technical i
specifications.
c.
The blade being withdrawn inserts along with remaining drives at technical specification insertion rates upon initiation of scram signal.
Examination of a number of rod withdrawal transients in the low power startup range, using an R-2 model, has shown clearly that higher fuel enthalpy addition would result from transients starting at the 1 percent power level rather than from lower power levels.
The analysis further shows that for continuous rod withdrawal from these initial power levels (1 percent range) the APRM 15 percent power level scram is likely to be reached as soon as the degraded (worst bypass condition)
IRM scram.
Consequently, credit is taken for either the IRM or APRM 15 percent scram in meeting the consequences of this event.
The transients for this response were initiated at 1 percent of power and were performed using the 15 APRM scram.
An initial point kinetics calculation was run to determine the line to scram based on an APRM scram setpoint of 15 percent power and an initial power level of 1 percent.
From this time and the maximum allowable rod withdrawal speed, it is possible to show the degree of rod withdrawal before reinsertion due to the scram.
From this information Figure 15A.1.12-1 showing the modified effective reactivity shape, was constructed.
l
'he point model kinetics calculations use the same equations l
e,aployed in the adiabatic approximation described on page 4-1 of l
Reference 10.
The rod reactivity characteristics and scram j
resetivity functions are input identical to the adiabatic l
calculations, and the Doppler reactivity is input as a function 15A-38b
SNPS-1 FSAR of core average fuel enthalpy.
The Doppler resctivit?
feedback function input to the point model calculatio..s was derived from the detailed analysis of the 1.6 percent rod worch case described above.
This is a conservative assumption for higher rod worths since the power peaking and hence spatial Doppler feeaback cill be larger for higher rod worths.
As will be seen in the results section, maximum enthalpies resulted from cases initiated at 1 percent of rated power.
In this power range, the APRM will initiate scram at 15 percent of power; hence, the APRM 15 percent power scram was used for these calculations thereby e.iminating the need to perform the spatial analysis required for the IRM l
Scram.
All other inputs are consistent with the detailed I
transient calculation.
l The point model kinetics calculations results in core average l
enthalpies.
The peak enthalpies were calculated using the follcwing equations:
I where h
= ho + (P/A) T (h f -
ho):
l h
= Final peak fuel enthalphy:
ho = Initial fuel enghalpy:
I hf = Total peaking factor (radial peaking)
+ (axial peaking)
(local fuel pin peaking).
+
For these calculations, the (radial x axial) peaking factors as a function of rod worth were obtained from the calculations l
performed in Section 3.6 of Reference 11 and are shown in Fig. 15A.l.12-2.
It was conservatively assumed that no power flattening due to Doppler feedback occurred during the course of the transient.
15A.l.12.4.2 Results The reactivity insertion resulting from moving the control rod is shown in Fig. 15A.l.12-1 for the point kinetics calculations.
[
The core average power versus time and the global peaking factors from Section 3.6 of Reference 11 are shown in Figs. 15A.l.12-3 and 15A.1.12-2, respectively.
The results of the point kinetics calculation are summarized in Table 15A.l.12-2 along with the results of the detailed analysis.
i From Fig.
15A.l.12-3 and Table 15A.l.12-2, it is shown that the core average energy deposition is insensitive to control rod worth; therefore, the only change in peak enthalpy as a function of rod worth will result from differences in the global peaking which increases with rod worth.
Comparison of the global peaking factors shown in Fig. 15A.l.12-2 with the values used in the 15A-38c l
L
)
SNPS-1 FSAR detailed calculations demonstrates that the Reference 11 values are reasonable for their application in this study.
For all
~~cases, the peak fuel enthalpy is well below the licensing design criteria of 170 cal /gm.~~
Cases 4
and 5 of Table 15A.l.12-2 show that the point kinetics calculations give conservative results relative to the detailed evaluations.
The primary difference is that the global peaking will flatten during the transient due to Doppler feedback.
This is accounted for in the detailed calculation but the point kinetics calculations conservatively assumed that the peaking remains constant at its initial value.
The dif ferences in core average and peak enthalpy between Cases 1 and 5 are due to the fact that for Case 1, the scram was initiated by the 15 percent APRM scram setpoint, whereas, in Case 5, the scram was intiated by the IRM's.
As seen in Fig. 15A.l.12-4 this occurred at a
core average power of 21 percent.
Since the APRM trip point will be reached first, it is reasonable to take credit for the APRM scram.
15A.1.12.4.3 Conclusions From this study the following conclusions can be stated:
a.
The resultant peak fuel enthalpies due to the continuous withdrawal of an out-of-sequence rod in the startup range results in peak fuel enthalpies which are significantly less than the licensing basis criteria of 170 cal /gm.
b.
The point model calculations used to assess the sensitivity of peak enthalpy as a function of control rod worth are in good agreement with, and slightly conservative relative to the more detailed design basis model which is employed to evaluate the continuous rod withdrawal transient in the startup range.
15A.1.12.4 Barrier Performance An evaluation of the barrier performance was not made for this event because resultant peak fuel enthalpies, due to continuous rod withdrawal, are significantly less than the licensing basis criterion.
15A.l.12.5 Radiological Consequences An evaluation of the radiological consequences was not made for this event since no radioactive material is released from the fuel.
15A-38d
~-.. -..
SNPS-1 FSAR References in Appendix 15A 1.
P.W.
~~Marriot, et al., "The Loss of Coolant Accident and the Environment, A Probabilistic View," ASME 72-WA/NE-9.~~
l 2.
" General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application,"
NEDO-10958 and NEDE-10958, November 1973.
3.
R.
~~Linford,~~
" Analytical Methods of Plant Transient Evalua-tions for the General Electric Boiling Water Reactor,"
NEDO-10802, April 1973.
. C.J. Paone, " Banked Position Withdrawal Sequence," NEDO-21231, January 1977.
5.
R.B. Elkins, " Fuel Rod Prepressurization, Amendment 1," NEDO-23786-1, May 1978.
6.
E.D. Fuller, letter to O.D.
Parr, "NRC Request for Additional Information on Fuel Rod Prepressurization," June 8, 1978.
7.
E.D. Fuller, letter to O.D. Parr, "NRC Request for Additional Information on Fuel Rod Pressurization,"
{ sic}
August 14, 1978.
8.
R.B.
.Elkins,
" Fuel Rod Prepressurize. tion,"
NEDE-23786-P, March 1978.
9.
F.
~~Odar,~~
~~" Safety Evaluation for General Electric Topical Report: Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors", NEDO-24154, 1980.~~
~~10. C.J.~~
Paone et al,
" Rod Drop Accident Analysis for Large Boiling Water Reactors", NEDO-10527, March 1972.
~~11. R.C.~~
Stirn et al,
" Rod Drop Accident Analysis for Large Boiling Water Reactors," NEDO-10527, Supplement 1, July 1972.
~~12. R.C.~~
~~Stirn,~~
" Rod Drop Accident Analysis for Large Boiling Water Reactors Addendum No. 2, Exposed Cores,"
NEDO-10527, Supplement 2, January 1973.
15A-64
TABLE 15A.1.12-1 SEQUENCE OF EVENTS FOR CONTINUOUS ROD WITHDRAWAL DURING REACTOR STARTUP TIME (Sec)
EVENT 0
~~1. The reactor is critical and operating in the startup range.~~
>0
~~2. The operator selects and withdraws an out-of-sequence control rod at the maximum normal drive speed of 3.6 ips.~~
~~-4 sec~~
~~3. Both the RWM and the RSCS fail to block the selection (selection error) and continuous withdrawal (withdraw error) of the out-of-sequence rod.~~
~~4-8 sec~~
~~4. The reactor scram is initiated by the intermediate range monitor (IRM) system or the advantage power range monitor system (APRM).~~
~~5-9 sec~~
~~5. The prompt power burst is terminated by a combination of Doppler and/or scram feedback.~~
~~10 sec~~
~~6. The transient is finally terminated by the scram of all rods, including the control rod being withdrawn. Scram insertion times are assumed to be 5 seconds to 90 percent insertion.)~~
t 1 of 1 l
1
-,r.-
SNPS-1 FSAR TABLE 15A.1.12-2
~~SUMMARY~~
OF RESULTS FOR DETAILED AND POINT KINETICS EVALUATIONS OF CONTINUOUS ROD WITHDRAWAL l
IN THE STARTUP RANGE
)
Control Rod Case Worth (%Ak) h(cal /qm)
P/At*)
h(cal /gm) r 1
1.6 17.3 24.2 42.7 2
2.0 17.3 30.9 50.0 3
2.5 17.2 46.0 58.5 4
1.6
(*)
18.3 19.7
(*)
56.2 5
~~1. 6 ( * )~~
18.3 19.7 59.6 P/A = global peaking factor (Radial x Axial).
( *)
(*)
Detailed transient calculation.
All other data reported are for point kinetics calculations.
19.7 is the initial value.
For the detailed
( *)
The P/A
=
analysis this value will decrease during the course of the transient since the power shape will flatten due to Doppler feedback.
( *)
Point kinetics calculation with IRM initiated scram and3-D simulator global peaking.
l
{
l I cf 1
.,..-~,-__-.
_ __,_I
.026
.024
.022
.020
.018
.016
.014 CONTRO L ROD BEING PULLED d.o12
.010
.008
.006 t
.004 b
SCR AM INSERTS CONTROL ROD
.002 III 111 I ilI l
l l
I I
I I
o O
4 8
12 16 20 24 28 32 36 40 TIME (SECONDS) 1 FIG U R E 15 A.1.12-1 POINT KINETICS CONTROL ROD REACTIVITY INSERTION SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT
60 50 40 3
5 X<
w p
30 -
a E
N' P/A FROM 20
[ DETAILED AN ALYSIS 10 O
l.0 2.0 3.0 CONTROL ROD WORTH (%AK) f FIGURE 15 A.1.12-2 P/A Vs. ROD WOg NEDO-10527 SUPPLEMENT 1 AND DETAILED AN ALYSIS SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT
_, _.. _ -.... ~. ~.,,. _, _ _.. _ _ _.. _., _ _..
i i
f
/i li i
\\
t 10^
,I
\\
\\.
i
\\
2.5 % ROD WORTHNj
/
e E
/./\\.\\
\\.
E
~
//
\\\\
~
./
\\
/
E
/
\\
m
/./
\\
/
g
/ /
my
/ /
2.0% ROD WORTH f
/.
l.6% ROD WORTH 8
Y Th 10-2 I
I I
2.0 4.0 6.0 8.0 TlME( SECONDS)
FIGURE 15 A.I.12-3 CONTINUOUS RWE IN THE STARTUP RANGE CORE AVERAGE POWER Vs. TIME FOR 1.6%,2.0% AND 2.5% ROD WORTH'S (POINT MODEL KINETICS)
SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT
10~'
3 0
us E
~
8 s
i~
10-2 l
l l
l l
l 2
3 4
5 6
7 8
9 TIME (SECONDS)
ASSUMPTIONS:
~~1. 1.6 % ok R OD~~
~~2. O.3 fps WITHDRAWAL VELOCITY l~~
~~3. lRM SCRAM FOR WORST l~~
~~BYPASS CONDITION FIGURE 15 A. I.12-4~~
~~4. Po = 10-2 OF RATED CONTINUOUS CONTROL ROD~~
~~5. 19 6 PR T~~
~~NE TECH WITHDRAWAL FROM HOT STARTUP pEC CRAM R E~~
~~6. EXPOSURE = 0.0 GWD/T SHOREHAM NUCLEAR POWER STATION-UNITI FINAL S AFETY ANALYSIS REPORT~~
~
_ -

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