ML20006D934

Application for Amend to License NPF-85,consisting of Tech Spec Change Request 89-02.Amend Allows Required Source Range Monitor Count Rate to Be Reduced While Ensuring That Design Level of Counting Certainty Maintained.Ge Rept Also Encl

ML20006D934
Person / Time
Site:	Limerick
Issue date:	01/29/1990
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19310C708	List: ML20006D934 ML20006D936
References
NUDOCS 9002150277
Download: ML20006D934 (10)

Text

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• 4 *< * - 10 CPR 50.90 1

i PHILADELPHIA ELECTRIC COMPANY ,

p; NUCLEAR GROUP HEADQUARTERS  !

d. 955 65 CHESTERBROOK BLVD. i N '

WAYNE. PA 19087 5691 .!

(at s) s40 sooo i,

I January 29, 1990 ,

Docket No. 50-353 '

License No. NPF-85 {

l U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 i

SUBJECT Limerick Generating Station, Unit.2 Technical Specifications Change Request

' Dear-Sir Philadelphia Electric Company hereby submits Unit 2 Technical Specifications Change Request No. 89-02, in accordance

.with 10 CPR 50.90, requesting an amendment to the Technical Specifications (Appendix A) of Operating License No. NPF-85.

t This submittal requests changes to the Technical L Specifications (TS) to allow the the required source range monitor (SRM) count rate to be reduced while ensuring that the design level of counting certainty is maintained at all times for the SRMs.

! Information supporting this Change Request is contained in Attachment 1 to this letter, and the proposed replacement pages are contained in Attachment 2. -Also enclosed is a document entitled,

! "SRM Count Rate vs. S/N Ratio for Philadelphia Electric Company Limerick Unit 2." This document contains information that is proprietary to General Electric Company (GE), and therefore GE

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requests that the document be withheld from public disclosure in i accordance with 10 CPR 2.790(a)(4). As required by 10 CFR 2.790(b)(1) an application and affidavit to withhold this document from public disclosure is also enclosed.

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d' pcumentCD;trclDe~k Jcnunty 26, 1990 -l Page 2 l L-l 1

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If you have any questions regarding this matter, please  !

contact us.

Very truly yours, I I

.A. nger, Jr i Director 4 Licensing Section Nuclear Services Department i

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t Attachmente '

Enclosures i

! cc: W. T. Russell, Administrator, Region I, USNRC '

T. J. Kenny, USNRC Senior Resident Inspector, LGS 1 T. M. Gerusky, PA Bureau of Radiation Protection j 1

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• UNIT 2 TECNICAL SPECIFICATION CHANGE REQUEST 89-02 L f i

COMMONWEALTH OF PENNSYLVANIA  :

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! ss. i COUNTY OF CHESTER  :

k D. R. Helwig, being first duly sworn, deposes and says  !

i That he is Vice President of Philadelphia Electric Company, I

the Applicant herein; that he has read the foregoing Application  :

for Amendment of Facility Operating Licenses to ensure the design level of counting certainty is maintained at all times for Source i

Range Monitors, and knows the contents thereof; and that the i

statements and matters set forth therein are true and correct to  :

the,best of his knowledge, information and belief.  ;

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Vice Pres it Subscribed and sworn to before me this d ay of e 1990.

IouvlA 4 0o;) '~N 0 '

Notary'Public '

NOTAR:AL SEAL ANGELA G. OLENGINSKl. Notary Pubic Wayne. Ches:er County ,

t My Commission Ex;.*es Au;l. 31.1992

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ATTACHMENT 1

,1 LIMERICK GENERATING STATION Docket No. 50-353 License No. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST Source Range Monitor Count Rate Requirement Revision Supporting Information for Changes - 7 pages I

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', Dockat No. 50-353 i i

Philadelphia Electric Company, Licensee under Facility )

Operating License NPP-85 for Limerick Generating Station (LGS) Unit l 2, hereby requests that the Technical Specifications (TS) contained I in Appendix A of the Operating License be amended as proposed herein l to allow the required source range monitor (SRM) count rate to be reduced while ensuring that the design level of counting certainty  !

is maintained at all times for the Source Range Monitors (SRMs).

The proposed changes are provided in Attachment 2 and are indicated by vertical bars in the margin of the pages viii, 3/4 3-60a, 3/4 3-60b, 3/4 3-88, and 3/4 9-4. We request the changes proposed herein to be effective upon issuance of the Amendment.

This Change Hequest provides a discussion and description of the proposed TS changes, a safety assessment of the proposed changes, information supporting a finding of no Significant Hazards Consideration, and information supporting an Environmental Assessment.

Discussion and Description of Changes The LGS Unit 2 TS presently allow reduction of the minimum SRM count rate required for control rod withdrawal or core alterations during initial fuel loading and startup from the normal three (3) counts per second (cps) to 0.7 cps as long as the signal-to-noise ratio is greater than or equal to two (2). In order to provide greater flexibility, a change to the TS la proposed to incorporate a graph of SRM minimum count rate versus signal-to-noise ratio such that reduction of the count rate required by TS (3 cps to 0.8 cps) is accompanied by a corresponding increase in the signal-to-noise ratio (2 to 30).

The SRM system consists of four identical neutron detection channels. Each channel contains a miniature in-core fission chamber, a pulse preamplifier, an electronics drawer, and reraote reading indicators. Each detector is equipped with a motor driven mechanism to allow retraction from the core at neutron flux levels above the SRM range.

The SRM system monitors thermal neutron flux over a range sufficient to observe core shutdown source level, approach to criticality, and overlap into the Intermediate Range Monitoring (IRM) system. The indicating range of the SRM may be extended by retracting the detectors from the core. The SRM system provides four ran channelg to of 10 negtron flux level information displayed over a .

10~getoof10- cps (approximately corresponding to 102 of rated reactor thermal power), and four channels of flux level rate of change information displayed as reactor period over a range of -100 to 410 seconds.

The SRM fission chambers are operated in the pulse counting mode-and produce discrete output pulses which represent the

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Dockot No. 50-353 '

composite effect of thermal neutron flux and gamma flux at the

-detector. Due to the nature of the detector, the pulses produced by thermal neutrons are of much greater magnitude than those produced 3 by gamma, although the number of gamma pulses may far exceed the i number of neutron pulses. An electronic circuit performs a 1 discrimination action based on the amplitude of these pulses, thus  !

providing an output signal proportional only to the neutron count  ;

rate. 1 The following changes to the TS are proposed.

1) Insert Figure 3.3.6-1 on page 3/4 3-60b, "SRM Count Rate versus Signal-to-Noise Ratio."

2) Revise Table 3.3.6-2 on Page 3/4 3-60a, TS Section 4.3.7.6 on  ;

Page 3/4 3-88, and TS Section 4.9.2 on Page 3/4 9-4, to permit i reduction of the minimum required SRM count rate below 3 cps "provided the source range monitor has an observed count rate and signal-to-noise ratio on or above the curve shown on Figure 3.3.6-1."

3) Revise the index on Page vill to include Figure 3.3.6-1.

Safety Assessment The necessity for maintaining a minimum count rate on the SRMs when operating at pre-critical and low power conditions is based on the most conservative evaluation which includes fresh fuel loaded in the initial fuel cycle with no neutron sources present. A multiplying medium with no neutrons present forms the basis for the accident scenario in which reactivity is gradually but inadvertently added until the medium is in a supercritical configuration. The introduction of some neutrons at this point would cause the core to undergo a sudden power burst, rather than a gradual startup, with no warning from the nuclear instrumentation. While this scenario is of concern when a reactor is loaded with fresh fuel, it is of less concern when loaded with irradiated fuel.

Irradiated fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, by photo fission, and by photo disintegration of deuterium naturally present in the moderator. The neutron production in irradiated fuel is normally great enough to meet the normal minimum SRM count rate of 3 cps for the duration of a refueling outage, including the subsequent reactor startup. Ilowever, there is a possibility that a minimum count rate requirement of 3 cps could not be satisfied following an extended reactor shutdown. Providing a requirement of at least 3 cps as measured begins at or by above the SRMthe assures that any initial value of transignt, should it occur, 10~ of rated reactor power with the reactor critical as used in analyses of transient cold reactor core conditions. A review of Chapter 15, " Accident L

., , Docket No. 50-353 i Analyses" (Section 15.4, " Reactivity and Power Distribution Anomolles")oftheLGSFgnalSafetyAnalysisReport confirmed the use of 10- of rated reactor power with (PSAR), has the  :

reactor critical in analyses of transient cold conditions.

Furthermore, the inadvertent criticality concerns evaluated in Chapter 15 (Section 15.4.1) take no credit for the SRMs, since the SRMs are used for indication only.

• Since 0.8 cps, the lowest coun '

correspondstoapproximately8x10'grateshownonFigure3.3.6-1, of rated reactor power, we -

have concluded that reducing the downscale setpoint in accordance  :

with Figure 3.3.6-1 will not invalidate the assumptions used in the transient analyses. Stipulating a signal-to-noise ratio in .

! accordance with Figure 3.3.6-1 assures that the SRMs are indeed responding to neutrons and the critical will be well above 10~geutron flux level with powerthe duereactor of rated reactor to subcritical multiplication. '

The curve presented in Figure 3.3.6-1 was derived by General .

Electric Company specifically for LGS Unit 2. The technical basis for this curve is presented in General Electric Company Report EDE-10-0489, "SRM Count Rate vs S/N Ratio for Philadelphia Electric Company, Limerick Unit 2," dated April 7, 1989. The curve ensures the same level of confidence at lower cps setpoints as is provided at the nominal 3 cps setpoint. The 3 cps setpoint is based on an assumed signal-to-noise ratio of two, which yields a statistical neutron monitoring confidence of 951 that the indicated signal is correct. At lower cps setpoints, a higher signal-to-noise ratio is required to maintain the same level of counting certainty. The  !

inverse relation between cps and signal-to-noise ratio depicted in Figure 3.3.6-1 ensures the 95% confidence level. '

Information supporting a Finding of No Significant Hazards consideration We have concluded that the proposed changes to the LGS

. Unit 2 TS, which ensure the design level of counting certainty is maintained at all times for the SRMs, do not constitute a significant hazards consideration. In support of this determination, an evaluation of each of the three standards set forth in 10 CPR 50.92 is provided below.

A. The proposed changes do not involve a significant increase in the probability or consequences of an accident

• l previously evaluated The proposed changes reduce the minimum SRM count rate required to permit core alterations during reactor refueling and withdrawal of control rods during reactor startup. The proposed count rate is still within the design range of the SRM and specifying a signal-to-noise i

Docket No. 50-353 ratio assures the SRMs are responding to thermal neutron flux. No hardware changes are required to the SRM system, therefore malfunction of an SRM will still produce the required control rod withdrawal blocks.

The applicable accident analyses related to the proposed change are those involving SRMs during reactor startup (control rod drop accident (RDA) and continuous control rod withdrawal) and those involving SRMs during reactor refueling (i.e., control rod removal error during refueling, a second control rod removal or withdrawal, fuel insertion with the control rod withdrawn, and a control rod removal without the fuel removed). No credit is taken for the SRMs in any of these accident analyses.

The RDA is the more limiting accident during reactor startup. Chapter 15 (Secti an initial condition of 10~gn 15.4.9) of the of rated FSAR reactor assumes power with the reactor critical in the analysis of transient cold reactor core conditions including ths RDA. Therefore, reduction of the minimum SRM count rate required to permit withdrawal of control rods as shown in Figure 3.3.6-1 etill ens 10~gres that criticality of rated will beAccordingly, reactor power. achieved well above this proposed change does not affect-the probability or consequences of previously evaluated accidents.

The nuclear characteristics of the core assure that the reactor is suberitical even at its most reactive condition with the most re' :tive control rod withdrawn during refueling. When the reactor mode switch is in the REPUEL position, only one control rod can be withdrawn. Selection of a second control rod initiates a control rod block,

thereby preventing the withdrawal of more than one control rod at1a time. Therefore, the refueling interlocks prevent any condition which could. lead to inadvertent criticality due to a contcol rod withdrawal error dcring Lcfueling.

Further, the mechanical design of the control rod, incorporating the velocity limiter, does not physically (

L permit the upward removal of the control rod without the l simultaneous or prior removal of the four adjacent fuel I assemblies, thus eliminating any hazardous condition. In l dddition, the refueling interlocks require that all control rods must be fully inserted before a fuel assembly may be inserted into the core. Therefore, based on the above evaluation, the proposed changes do not involve a i significant increase in the probability or consequences of an accident previously evaluated. 4 l .

L B. The proposed changes do not create the possibility of a new or difterent kind of accident from any accident previously evaluated.

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I No hardware modifications are required to implement this l proposed change. The design functions of the SRM system )

are not being changed. The only effect of this proposed l change is a reduction in the minimum count rate required for control rod withdrawal and reactor core alterations.

The reduced minimum count rate, which must be accompanied by a corresponding increase in signal-to-noise ratio, will  !

continue above10-goassurethatalltransientsofconcernbegin of rated reactor power, thereby maintaining the validity of the FSAR analysis. Accordingly, the  ;

proposed changes do not create the possibility of a new or different kind of accident from any accident previously  ;

evaluated.

C. The proposed changes do not involve a significant" reduction  !

In a margin of safety. t The normal requirement of at least 3 cpn assures that any  ;

transient, a valueof10~goulditoccur,beginsatorabovetheinitial of rated reactor power with the ri tor critical as used in analysis of transient cold rsactor core .

conditions. In fact, any observable neutron count rate on l the SRM is sufficient to ensure the analysis remains valid.  ;

Therefore, reduction of the minimum count rate from 3 cps .

to the values shown in Figure 3.3.6-1 will not reduce this 2 margin of sa orabove10-getybecauseanytransientwillstillbeginat of rated reactor power with the reactor r

Y critical. Further, the SRMs are not required to ensure the i

margin of safety as analyzed in Chapter 15 of the FSAR.

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L Information supporting an Environmental Assessment l An environmental assessment is not required for the changes E

proposed by this Change Request because the requested changes conform to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR 51.22(c)(9). The requested L changes will have no impact on the environment. This Change Request l does not involve a significant hazards consideration as discussed in

! the preceding section. This Change Request does not involve a r significant change in the types or significant increase in the ,

y amounts of any effluents that may be released offsite.

In addition, '

-this Change Request does not involve a significant increase in individual or cumulative occupational radiation exposure.

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• DockGt No. 50-353 Conclusion' ,

The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the TS and have ,

concluded t: sat they do not involve an unreviewed safety question or a significant hazards consideration, and will not endanger the r health and safety of the public.

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