ML18094A308

Forwards Salem Units 1 & 2 10% Tube Plugging Large Break LOCA Bash Analysis.

ML18094A308
Person / Time
Site:	Salem
Issue date:	03/31/1989
From:	Miltenberger S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18094A309	List: ML18094A308 ML18094A310
References
NLR-N89052, NUDOCS 8904110156
Download: ML18094A308 (8)

Text

Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 Vice President and Chief Nuclear Officer MAR 3 1 1989 NLR-N89052 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REVISED LARGE BREAK LOCA ANALYSIS SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 Attached, pursuant to our commitment in PSE&G letter NLR-N88176, dated October 21, 1988, is a revised large break LOCA BASH analysis. The submittal of this report is based on steam generator inspections performed during the Salem Unit 2 fourth refueling outage.

During that outage, defective tubes were discovered in two steam generators. It was decided to plug the row 1 tubes in all four Salem Unit 2 steam generators as a precautionary measure. This resulted in 2.7% of the Salem Unit 2 steam generator tubes being plugged. Also, during refueling operations, a burnable poison rodl'et assembly hold down nut, a locking weld pin and a hand held gamma measurement probe with cable connector were inadvertently dropped into the reactor cavity of Salem Unit 2. Subsequent efforts to retrieve these items were unsuccessful. These objects were therefore evaluated as loose parts within the reactor cooling system (RCS). These changes in plant configuration could potentially affect the peak cladding temperature (PCT) during a large break loss-of-coolant-accident (LOCA).

For plants which have been licensed based on the 1978 Westinghouse large break LOCA model, NRC generic letter 86-16 required subsequent plant changes which affect the results of the model, to be reevaluated against the updated, approved model and submitted in accordance with 10 CFR 50.46(a) (1) (i). Therefore, PSE&G performed a formal reanalysis to confirm that Salem Unit 2 meets the applicable criteria of 10 CFR 50.46(b) based on the current plant configuration. This analysis also supports the Vantage 5H fuel License Change Request (LCR) which is currently under review by the NRC staff. This LCR supports the Unit 1 Cycle 9 core reload, currently scheduled to begin April 21, 1989.

8904110156 890331  : ':1 PDR ADOCK 05000272 '

P PDC

'*'

Document Control Desk 2 MAR 3 1 1989 NLR-N89052 Because of the long time frame required to perform the formal reanalysis with the new 1981 large break LOCA model, it was not possible to complete the required analysis prior to the scheduled Salem Unit 2 Cycle 5 startup. Consequently, a one time, temporary exemption from 10 CFR 50.46(a) (1) (i) was requested by PSE&G and granted by the NRC based on a conservative safety evaluation of the above plant changes (Reference 1). PSE&G committed to submit its formal reanalysis by March 31, 1989.

Attachment #1 contains a copy of the Westinghouse LOCA analysis report "Salem Units 1 and 2 10% Tube Plugging Large Break LOCA BASH Analysis", including marked up FSAR pages. Attachment #2 contains the Westinghouse loose parts evaluation of the effects of the above missing objects on the new large break LOCA analysis results for Salem Unit 2.

This LOCA analysis, in combination with the plant safety analysis report for VANTAGE 5H fuel submitted to the NRC previously, completes all the required accident analyses for both VANTAGE 5H and 17x17 STD fuel for both Salem Unit 1 and Salem Unit 2 for plugging up to 3.5% of the tubes in all steam generators.

Additionally, this revised large break LOCA analysis incorporated additional parameters, including 10% tube plugging, as described in Attachment 1. These additional parameters may be used to support future Licensing actions.

The limiting case for this large break LOCA analysis, which uses the Westinghouse 1981 LOCA model, was determined to be the Cd=0.4 size break assuming minimum safeguards safety injection. The resulting PCT was 2091 degrees Fahrenheit, well below the 10 CFR 50.46 limit of 2200 degrees. As discussed in Attachment #2 the Salem Unit 2 missing objects assumed to be within the reactor cooling system will not increase the calculated PCT and will not challenge any of the other 10 CFR 50.46 criteria for the Emergency Core Cooling system.

As demonstrated in Attachment 1 and 2, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR

• 50.46 for breaks up to and including the double ended severance of a reactor coolant pipe. There will be no transition core penalty for cycles with mixed STD and VANTAGE 5H (w/o Intermediate Flow Mixers) fuel. The missing objects assumed to be within the Salem Unit 2 reactor coolant system do not affect these conclusions.

If you have any questions, please do not hesitate to contact us.

Document Control Desk 3 MAR 3 1 1989 NLR-N89052

REFERENCES:

1) Letter from James c. Stone (NRC) to Steven E. Miltenberger (PSE&G), "Schedular Exemption from 10 CFR 50.46(a) (1) (i)

(TAC No. 69814) Re: Salem Generating Station, Unit 2 11 ,

dated November 1, 1988.

2) Letter from s. LaBruna (PSE&G) to NRC, "Request for Amendment Facility Operating License DPR-70 and DPR-75 Salem Generating Station Unit Nos. 1 and 2 Docket Nos. 50-272 and 50-311 11 , Re: use of Vantage 5H Hybrid Fuel, dated December 30, 1988.

Sincerely,

/J//f/~~

Attachments c Mr. J. c. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625

SECL-88-547, Rev. 1 Customer Reference No(s}.

Westinghouse Reference No(s).

(Change Control or RFQ As Applicable)

WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST

1) NUCLEAR PLANT (S)___.;:S=a.. .:. .;le=m.:.....=Un""""'i....:..t-=2=---------------

2) CHECK LIST APPLICABLE TO: Impact of having unrecovered loose (Subject of Change) parts in the reactor coolant system on the LOCA accidents

3) The written safety evaluation of the revised procedure, design change or modification required by IOCFRS0.59 has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.

Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.

CHECK LIST - PART A

• (3.1) Yes_!_ No_ A change to**the"plant as*described in the*FSAR?.

(3.2) Yes_ No_!_ A change to procedures as described in the FSAR?

(3.3) Yes __ No__L. A test: or experiment no:t described in the*FSAR?

(3.4) Yes_ No_!_. A change to the plant technical specifications (See :Note on Page::.l-~.Jr:>

Page 1 of 3

SECL-88-547, Rev. 1

4) CHECK LIST - Part B (Justification for Part B Answers must be included on Page 2.)'

(4.1) Yes~ No~ Will the probability of an accident previously evaluated in the FSAR be increased?

(4.2) Yes~ No~ Will the consequences of an accident previously evaluated in the FSAR be increased?

(4.3) Yes~ No~ May the possibility of an accident which is different than any already evaluated in the FSAR be created?

(4.4) Yes~ No~ Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

(4.5) Yes~ No~ Will the consequences of a malfunction of equipment importan~ to safety previously evaluated in the FSAR be increased?

(4.6) Yes~ No~ May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

(4.7) Yes~ No~ Will the margin of safety as defined in the bases to any technical specifications be reduced?

If the answers to any of the above questions are unknown, indicate under 5) REMARKS and explain below .

..

• -**

I.f*ttre:'*ans:wers' to:.':any~H1f:.*the:7above:":QUe'Stton:s"**i:n:* Part*A.'(3'..'4}::,.or *Part.:B.~*;*

cannot be answered in the negative, based on the written safety evaluation, the change review would require an application for license amendment as* required by lOCfRS0 ..59.{c:).. and..-:submitted. . to.;.the:.NRC .. - ->

pursuant to IOCFRS0.90.

5) REMARKS:

NONE Page 2 of 3

SECL-88-547, Rev. 1 The following summarizes the justification upon the written safety evaluation,(!) for answers given in Part A (3.4) and Part B of this Safety Evaluation Check List:

See Attachment A 1>Reference to Documents containing written safety evaluation:

NS-SAT-TSA-89-76 FOR FSAR UPDATE Section: Page(s): Table(s): _ __ Figure(s) : _ __

Reason for/Description of Change:

Prepared By (Nuclear Safety): ~~~~~::..--......wT~SAru. Date: WR9 Coordinated With Engineer(s *. Date: ~/afJ[

Coordinated With Engineer(s): TSA Date: 3~i/8J Coordinating Group Manager: -......._:~--i-.icAII Date: .3 5f6 'l Nuclear.,. Safety:.: Group.: Manager.*:"~

. '-+--+-__:._----o.li:::....---1-~.*..:.;TS=_A.:.A.., Date:>c?25'.,,4"~:, , , ,

Page 3 of 3

SECL-88-547, Rev. 1 LOCA BASES LARGE BREAK LOCA - FSAR CHAPTER 15.4.1 A new large break LOCA analysis has been performed for Salem Units 1 and 2 using the Westinghouse 1981 Evaluation Model with BASH to support operation with up to 103 uniform steam generator tube plugging with Westinghouse Standard 17xl7 and/or VANTAGE 5 Hybrid (VSH) fuel.

The limiting case for this analysis was determined to be the Cd=0.4 break assuming minimum safety injection flow, which resulted in a PCT of 2091°F.

To determine the effect of the missing objects on the large break LOCA analysis, an evaluation has been performed which considers locations of the objects within the RCS which could affect the large break LOCA PCT calculations. The core flow area assumed in the evaluation reflects the Salem Unit 2 core with Westinghouse 17xl7 standard fuel or V5H fuel.

The location which was determined to have the greatest potential to affect the large break LOCA transient was that which would block the reactor coolant flow through the active core fuel assemblies. The chemical and mechanical evaluation of these objects within the reactor vessel environment indicates that the plastic, rubber, aluminum and tin-lead alloy of the HP-290 probe will disperse into benign particles within the reactor coolant. The remaining metal components of the lost objects will remain intact; however, some deformation can be expected. An evaluation of the size of the remaining objects indicates that only the lock pin, the ground wire and the stainless steel central wire of the HP-290 probe have the potential to enter any of the fuel assembly subchannels. The remaining objects will not be able to pass through the fuel assembly bottom nozzles.

As in the previous evaluation of the effect of these loose parts on the current licensing basis large break LOCA analysis (78 Evaluation Model), a11 three of the above mentioned objects remaining-* in*- the*:* RCS are conse rv at ivety; assumed :.ta: :be~ *resident** s*imu ltaneaus.l y

• i n. *the -*.'" *. *..

hottest*core'* subchannel*~ , _ However~ due**'.*to**~the **lack *of explicit grid- * * ** *

• modeling in the 78 Evaluation Model, no consideration was given to the effect of the grids on the evaluation. On the other hand, the analysi.s ..wi.th .the .1981 ..EM .witlLBASH.explic.itly_models. the grids_ and ...

this effect is ..tncluded.:*here**.

• The grid dimensions are such that the loose parts .. co.uld. potentially.

• pass through the grid only if they are vertically oriented.

• Conversely, it is highly probable that the parts will become entrapped at a grid if they are in.the active fuel region. Thus, this evaluation assumes that if the loose parts are trapped in the active fuel region, they will be at the grid elevations.

l *- '

SECL-88-547, Rev. 1 Furthermore, the pieces were assumed to be oriented in a manner which creates the greatest flow blockage. Calculation of the cross sectional area of the missing pieces resulted in an area capable of blocking up to 36% of an assembly subchannel. A subchannel blockage of this magnitude was evaluated and found to potentially create a clad temperature increase on the order of 2S.4°F. This penalty, when-added to the PCT calculated at the gridded elevations, resulted in net peak clad temperatures ranging from 18IS°F to 2007°F. Thus, the Salem BASH analysis PCT of 209I°F, occurring at an elevation of 8.5 feet, remains the limiting PCT.

Since there is no increase in the maximum calculated PCT, there is no change to the maximum local Zirconium water reaction reported for the analysis, nor any challenges to the remaining IOCFR 50.46 acceptance criteria*-_

In addition, the following information is provided to further assure that there is no- increase in the risk to the health and safety of the public.

Recent development of a Best-Estimate large break LOCA model and test performed to determine the effects of fuel assembly flow blockage-have demonstrated that even large amounts of flow blockage (< 90%) result in a PCT benefit. The benefit is related _to breakup of the entr~ined water droplets which are present during a LOCA. However, current LOCA models developed in response to IOCFRS0.46 and Appendix K to IOCFRSO do not have the sophistication to model non-equilibrium effects and the presence of entrained water droplets during blowdown. Thus, sensitivity studies based on Appendix Kmodels result in a calculated increase in clad temperature at the blockage elevation. However, the expected location of the loose parts during* a LOCA would be at either the top or bottom of a grid, depending upon the flow direction. The local power is lower and heat transfer is much higher in the region around grids than calculated by the Westinghouse Evaluation Models.*

Credit for these effects would offset the penalty associated with the loose objects in the RCS.

Tfie>'evaluatton';--:based*,,upon:::tnerl98l:--Evaltiation**Model;-with*BASH for~-_;_)"

.

"_)_ *, Salem Unit 2, has demonstrated that the criteria of IO CFR 50.46 would -- -

be satisfied during power operation with the loose parts residing in the Reactor Coolant System. In addition, recent development of

• ~ *,

Best~Estimate models. and testing performed to determine the effect on cladding- temperatures as a result of flow blockage . providecadditional-assurance that LOCA criteria are maintained.