ML20011D850

Application for Amend to License DPR-16,consisting of Tech Spec Change Request 181 Eliminating Eight Main Steam Safety Valves W/Two Highest Set Points

ML20011D850
Person / Time
Site:	Oyster Creek
Issue date:	12/18/1989
From:	Fitzpatrick E GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20011D848	List: ML20011D847 ML20011D850
References
NUDOCS 9001020161
Download: ML20011D850 (9)

Text

'

(

ii CPU NUCLEAR CORPORATION'

=.

b.,

~

OYSTER CREEK NUCLEAR GENERATING STATION!

Provisional Operating License No. DPR-16

-)

i Technical Specification.

I Change Request No.'181 q

~

Docket No. 50-219 Applicant: submits, by this. Technical Specification Change Request No.181 to the Oyster Creek Nuclear Generating Station Technical Specifications, a i

change to pages 2.3-2, 2.3-6, and'4.3-1.

By

/

/

h. E Fitzpatrick 3

Vice President and Director-Oyster Creek Sworn and Subscribe to before me this

[8 day of

.g.,d ),

, 1989.

l

, ;.. ha.

>AL.Lo

' A Notary Sdblic.of NJ t

OlANA M. DeBLASIO' N06ARY PUBUC 0F NEW JERSEY Wikalub WM A * #~ 9 /

~

I l

l 9001020161 891219 PDR ADUCK 05000119

^

P PDC t

i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

)

)

Docket No. 50-219 J

GPU Nuclear Corporation

)

QERTIFICATE OF SERVICE

~

This is to certify that a copy of Technical Specification Change Request No.-

181 for Oyster Creak Nuclear Generating Station Technical Specifications,.

-filed with the U.S. Nuclear Regulatory Commission on December 16, 1969 and' has this day of December 18, 1969

, been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States _

mail, addressed as follows:

The Honorable Christopher-Connors Mayor of Lacey Township 818 West Lacey Road Forked River, NJ, 7

j mA By J

E.\\g'Pktzpatrick Vice President and Director Oyster Creek-5 e

S

f GPU Nuoleer Corporation fr,_

g-One Upper Pond Road

' I 1

Parsippany. New Jersey 07054 -

201-316-7000 TELEX 136-402 l

Writers Direct Dial Number.

December 18, 1989 Mr. Kent Tosch, Director Bureau of Nuclear Engineering Department of Environmental Protection CN415 Trenton, New Jersey- 08625 Dear Mr. Toscht Subjects Oyster Creek Nuclear Generating Station Provisional Operating License No. DPR-16:

Technical Specification Change Request No. 181 Pursuant to 10CFR50.91(b)(1), please find enclosed a copy of the subject document which was filed with the United States Nuclear Regulatory Commission en December 18, 1989.

Very truly yours, J

. E. Fitzpatrick Vice President and. Director Oyster Creek EEF/DJ/cjg

~{

Attachment l

1 GPU Nuclear Corporation is a subsidiary of General Pubhc Utilities Corporation

l' i

(

L i

GPU Nuclear Corporation Nu01Mr.

One t,'pper Pona Road -

Parsippany, New Jersey 07054 201 316-7000 TELEX 136-482 '-

Writers Direct Dial Number.

I

. December 18, 1989

.I The Honorable Christopher.Connors Mayor of Lacey Township 818 West Lacey Road Forked River, New Jersey 08731 c

Dear Mayor Connorst.

Enclosed herewith is one copy of Technical Specification Change Request No.-

?

181 for the Oyster Creek Nuclear Generation Station Operating License.

This document was filed with the United States Nuclear Regulatory Commission' on December 18, 1989 very truly yours,/

.E.

tzpatrick Vice President'and Director.

Oyster Creek j

EEF/DJ/cjg Attachment l.

l 1

l GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation 4

-OYSTER CREEK NUCLEAR GENERATIN3 STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO.-50-219 TECHNICAL ~ SPECIFICATION CHANGE REQUEST No. 181 Applicant hereby requests the Commission to change Appendix A to the.above captioned license as below, and pursuant to 10CFR50.91, an analysis concerning the determination of no significant hazards consideration is also presented:

1.0 SECTIONS TO BE CHANGED Sections 2.3 and 4.3.

2.0 EXTENT OF CHANGE Eliminate eight main steam safety valves (safety valves) by.taking credit-for.high flux reactor scram in the. safety analysis.

Sections 2.3.F and 4.3.E are revised to delete eight safety. valves with the two highest setpoints. The bases for Section 2.3 are revised to incorporate credit for reactor scram for safety valve, sizing and change total number of safety valves from sixteen to eight.

3.0 CHANGES REOUESTGQ The requested changes are shown on attached Technical Specification pages 2.3-2, 2.3-6 and 4.3-1.

4.0 DISCUSSION The purpose of this Technical Specification Change Request is to propose the elimination of eight safety valves with the two highest setpoints.

Appropriate safety analyses have been performed to demonstrate the acceptability of the reduction in the number'of safety valves. A reduction in safety valves would result in significant cost savings in maintenance and surveillance testing.

In addition, it is estimated that the deletion of eight safety valves would reduce exposure by 20 man-rem per outage.

The reactor pressure vessel (RPV) and the pressure relief system were designed in accordance with Section I, 1962 edition of the American r

Society of Mechanical Engineers (ASME) " Boiler and Pressure Vessel' Code".

Under the provision of Section I, code qualified safety valves must flimit the rise in the RPV pressure to less than the ASME code limit;, Previous

'i analyses performed to demonstrate compliance with the code requiretaunts did not take credit for reactor scram, electromatic relief' valves (EMRVs),

turbine bypass valves and the isolation. condensers.

To satisfy this requirement, Oyster Creek currently employs 16 steam safety velves.

AAI-2 i

Th3 curr nt v0rcion of th0 ASME cod 3, S!ction=I, Cllows crOdit frr

~;

-independent sensing devices that'stop'the. flow of fuel to the boiler.-

since the code is for fossil boilers, the analogy for a nuclear plant is that credit for an independent or diverse shutdown system such as flux-l

?

scram and recirculation pump trip (RPT), would perform the same function-of fuel stoppage, i.e. boiler shutdown.

Thus, credit could be taken for' their functioning in the overpressure protection analysis consistent with the. current interpretation'of the ASME code.

Further, the NRC acceptance (NRC Ltr. 10/31/88, Safety Evaluation) of the GPUN. reload license methodology does allow credit for RPT in the overpressure protection analysis.

In addition,-NUREG-0800, " Standard Review Plan", indicates that the safety

. valves should be designed with sufficient capacity to limit the pressure 1

to less than 110% of the reactor coolant pressure boundary (RCPB) design pressure (as specified by ASME Boiler and. Pressure Vessel Code, Section.

l III) during the most severe' abnormal operational transient with credit for a reactor scram. All BWR plants designed in accordance with Section III of the ASME Code currently take credit for high neutron flux scram for safety valve sizing.

The appropriate code limits are observed for the new configuration.

This system has no function during normal operation, and'it is anticipated that there is a low probability of safety valve actuation since overpressure is relieved by the isolation condeneers, the turbins bypass valves.and the EMRVs.

The safety analysis requirements ^^r oyster Creek have been reviewed in order to establish the analyses that are potentially'affected byythe reduction in the number of. safety valves.

In the safety analysis process, no credit is taken for the operation of the safety valves except for the ASME code overpressure protection analysis and the evaluation of anticipated transients without-scram (ATWS).

These events have been reanalyzed using the NRC approved methodology for oyster Creek with the exception below, to demonstrate compliance with the appropriate event acceptance limits.

i License Basis Analyses 4

The safety valvee at Oyster Creek are required to protect the primary coolant pressure boundary against overpressure, overpressure protection is provided by limiting peak pressure in the reactor vessel ta) 110% of design pressure and to 115% of design pressure for the recirculation piping. The RPV design pressure is 1250 peig which' requires the limit to be 1375 psig (1390 psia). 'The recirculation piping design pressure is 1200 psig, which results in a limit of 1380 psig (1395 psia).

~

For. oyster Creek, a main steam isolation valve (MSIV)-closure without-scram or credit for operation of the EMRVs, also known as:the safety valve sizing transient, is analyzed in the updated FSAR Chapter 15 to determine the adequacy of the safety valves to prevent vessel overpressurization.

This event has been demonstrat'ed as being limiting using the NRC approved

+

Oyster Creek safety analysis methodology.

In previous license basis analyses, this event analysis was used to cover both the code overpressure protection and ATWS analysis requirements.

For this analysis, the MSIVs are assumed to close in 3 seconds, all scram activations fail, and all-solenoid-operated relief valves (EMRVs), bypass valves and isolation condenser isolation valves are

- AAI-2 i

b c:sumed to fail. Cr:dit is t k:n fcr the RPD.

The v3id c 11 Cps 3 rCruits l

~

in a power increase followed by a pressure. increase which is limited only by the safety valves' and the nuclear characterictics of the core design.

l This transient was snalyzed as part of the cycle 12 reload and resulted in a peak pressure at the bottom of the. vessel of 1305 psia which is well l

below the vessel. pressure safety limit of 1390 psia (1375 psig).

i i

fa order to evaluate the impact of the reduction in the safety valves on tht license basis analysis requirements, it is necessary to evaluate two avente i s place of the safety valve sizing transient. These two events are the MSIV closure with high flux scram and the MSIV closure ATWS.

Previoac evaluations have demonstrated that the use of the MSIV closure as the initiating event bounds the spectrum of potential initiating events.

l for oyster Creek.

The analysis results for these two. events are described in more detail below.

MSIV Closure with Hioh Flux Scram (8 Safety Valyggi l

l The licensed cycle 12 reload model was used for this; analysis with the NRC j

approved RETRAN-02 Mod 4 code. A single change to the model was made to assure conservative results for peak pressure.

This involved increasing the rainout velocity in the upper downcomer volume from 3 feet per second to 2000 feet per second.

It was observed for this transient that when the level in the upper downcomer dropped below the separator drains, liquid was entrained in the steam region of the upper downcomer and subsequently,-

)

some of this liquid was carried over by RETRAN into the upper plenum i

volume. The use of a largs rainout-velocity in the upper downcomer volume

{

prevented carry over of liquid and resulted in higher peak pressures.

j For the reload transients, the level does not drop below the separators' drains and the 3 feet per second rainout velocity is conservative.

I The same assumptions as previously listed were used with the exception of allowing a high flux scram. For this_ analysis, a scram would.normally

~

occur on MSIV closure of 10%.

This anticipates the pressure and neutron flux transients which occur during normal or inadvertent valve closure.

i However, no credit is taken for th!s scram signal in the analyses. The reactor is assumed to be scrammed by the high flux-scram signal at a conservative setpoint of 120% as compared to the actual setpoint of 115.7%.

For this analysis, the eight safety valv6W (Banks l'and 2) are assumed to open on a high steam line pressure of 1240 psia (Bank 1-4 valves) and-1249 psia (Back 2-4 valves), and close on a low steam line pressure of 1190 psia (Bank 1-4 valves) and 1199 pela (Bank 2-4 valves).

It was datermined that eight uafety valves-(as opposed to 16 for the license I

basis analysis) would limit the peak pressure at the bottom of the vessel to 1370 psia, below the code limit of 1390 psia.

The pressure in the recirculation piping is 1377 psia which is within the 1395 psia code l

requirement for the pipir.g.

MSIV Closure ATWS (8 Safety Valves)

A MSIV closure ATus was evaluated to demonstrate that the results<for the L

posttiated ATWS event were acceptable considering the reduction of safety l

h AAI-2 i i

l v 1v03. Thio trcnti nt to en31y;ed with the came c:nditiing ca de:cribed j

above, using eight safety valves, but credit is taken for the EMRVs and i

for recirculation pump trip.

The poak pressure in the reactor vessel was i

determined to be 1297 psia. Based on these results, the ATWS analysis is less limiting.than the above flux scram case and does not have to be.

j reanalyzed for future reloads.

This analysis only addresses the j

overpressurization limits associated with an ATWS since the effects of l

other limits remain the same with eight or 16 safety valves.

5.0 DETERMINATION The proposed Technical Specification Change Request does not involve a l

significant hazards consideration for the reasons as stated below:

i 1.

Involve a significant increase in the probability or consequences of l

an accident previously evaluated:

The removal of eight safety valves will not increase the probability of occurrence of an accident previously evaluated in the SAR since i

the remaining safety valves remain unchanged. The only event initiator that involves a safety valve is a spurious valve opening.

The proposed reduction in valves will slightly reduce the probability of a spurious valvo opening. Thus, the probability of a valve opening is not increased.

In the safety analysis process, credit for the operation of the safety valves is only taken in the code overpressure protection and ATWS events.

These events have been reanalyzed using the approved I

oyster Creek license analysis methodology. With the reduced number of safety valves and no credit for the high flux scram, the peak calculated pressure due to these events previously reported in the Safety Analysis report would be increased.

However, with the l

proposed change to the design basis to take credit for the high flux scram, the appropriate event acceptance. limits are satisfied.

The activity will not significantly increase the~ probability of l

occurrence or consequence of a malfunction of equipment important to safety previously evaluated in the SAR based on a reliability analysis of RPT, EMRVs and.n.maining safety. valves. (8) vhich shows that the likelihood of reactot vescel overpressure due to an ATWS J

remains very small. Also, since there will be eight fewer' safety valves, the likelihood of an initiating event involving spurious opening of a safety valve is reduced.

I 2.

Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed activity does not create a possibility for an' accident or malfunction of a different type than any previously identified in

.the SAR since existing safety valves remain unchanged, and no systems are affected by this modification. Analyses demonatrate that all of the appropriate event acceptance limits have been satisfied for'the proposed new configuration.

t AAI-2

• I

l_

'Th3 Cight sOf0ty v01v s removed will be rep 1CT;;d with blind fitnges After l

-to' maintain the reactor coolant pressure boundary (RCPB).

. installation, initial service leak test will be performed, thus l

~

assuring the integrity of the RCPB.

3.

Involve a significant reduction in a margin of safety.

_l The margin of safety as_ presently defined in the basis for the-Technical Specifications does not take credit for high flux scram..

This Technical Specification Change Request proposes to take credit for high flux scram and then require only eight safety valves to mitigate the consequences of a MSIV closure transient.

]

For the purposes of this evaluation, the. margin of safety is defined f

as the margin between the safety limit and fission product barrier failure. Because the event does not exceed the event limit (1375 psig), the margin of safety is not reduced.

t f

6.0 IMPLEMENTATICM i'

It is requested that the amendment aEthorizing this change become effective for operating Cycle 13.

j 4

1

- 5 i

l

-i 4

}

7 i

AAI-2',

c e