ML20087A019

Forwards Response to NRC 910701 Request for Addl Info on 891220 Application for Amend to License DPR-59,changing TS Re Safety Relief Valve Performance Requirements.Rev 1 to Technical Rept TR-7543-1 Also Encl

ML20087A019
Person / Time
Site:	FitzPatrick
Issue date:	01/03/1992
From:	Ralph Beedle POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20087A020	List: ML20087A019 ML20087A029
References
JPN-92-001, JPN-92-1, NUDOCS 9201080123
Download: ML20087A019 (9)

Text

_

123 Ma>n Street Wnite Plains, tevv York 10601 O , #e

, 91.' E810846 t

1. > NewYorkPower n. . o.-.

4# Authority ll% E='?""

January 3,1992 JPN 92 001 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1 137 Washington, D.C. 20555

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Request for Additionaliniormation Regarding Proposed Technical Specification Change Updated Safety Relief Valve Performance Requirements (JPTS-89-017)

References:

1. NRC letter, Brian C. McCabe to R. E. Boodle dated July 1,1991 regarding request for additionalinformation proposed technical spr,cification change for safety relief valve setpoints.

2. NYPA letter, J. C. Brons to USNRC dated December 20,1989 (JPN 89 084) regarding updated S/RV performance requirements (JPTS-89-017).

Dear Sir:

Attachment I is the Authority's answers to the four questions included with the reference letter. The questions concern a proposed technical specification change regarding safety relief valve (S/RV) performance requirements at the Authority's James A. FitzPatrick nuclear power piant.

Attachment 11 is a Teledyne Engineering Services report (TR-7543- 1, Revision 1) which supplements the Authority's answer to question 1. Telodyne's report evaluates the potential effects of a single S/RV setpoint on the torus shell, submerged structures, attached piping, T-quenchers and their supports, and the S/RV piping. In anticipation of the Authority's application to increase FitzPatrick's licensed power level by approximately four percent, the evaluation assumed a higher bounding setpoint.

The Authority's responses frequently refer to earlier reports and evaluations. Except for the report inc!uded as Attachment II, these reports have been sent to the NRC by the Authority or other organizations. The Authority will provide additional copies of reports referred to in these responses if the NRC staff cannot easily retrieve them.

mumm p m er em moon ,

(hh i  ;

FDn '

. _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ - _ - _ - _ - _ _ _ _ . = _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ . _ _ _ _ _ _ _ _ _- - _ _ _ _ - _ . _ _ - - ..

l'

.= o-

. The Authority included a General Electric report (NEDC-31697P) as part of it's technical specification amendment application (Reference 2). This report has been .

revised to include pert.nent information in the attached answers. The Authority will

- submit Revision 1 to NEDC 31697P separately because General Electric considers this report to be proprietary, if you have any questions, please contact J. A. Gray, Jr.

Very truly o rs,

. W.

Ralph E. Beedle u

Exocutive Vice President Nuclear Generation

- cc: Regional Administrator U.S Nuclear Regulatory Commission 475 Allendate Road King of Prussia, FA 19406 Office of the Resident inspector

. U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Brian C. McCabe Project Directorate 11 Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission Mail Stop 14 B2 ~

Washington, DC 20555 e

a

iL* - .. .

i g.

Attachment 1 to JPN 92-001 New York Power Authority

. James A. FitzPatrick Nuclear Power Plant

! NYPA RESPONSE TO NRC JULY 1,1991 REQUEST FOR ADDITIONAL L INFORMATION - PROPOSED TECHNICAL SPECIFICATION CH ANGE REGARDING SAFETY REUEF VALVE SETPOINTS FOR THE JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. 75561) -

Question 1 With a single setpoint for all eleven safety / relief valves (S/RVs), it is expected that all valves will actuate simultaneously. Certainly, with a single setpoint, this possibility cannot be precluded.-

With the current staggered S/RV setpoint, the FitzPatrick plant's unique analysis for piping and containment loads was based on asst aptions that a lesser number of S/RVs actuate

- simultaneously. Describe how the increase in loads from the simultaneous actuation of all valves were considered in the following:

(a) S/RV thrust loads on main steam line piping.

(b) Loads on S/RV discharge piping due to increased motion of main steam lines.

(c) - Water jet loads on submerged structures.

(d) Air bubble drag loads on submerged structures.

(e) Torus shellloads.

(f) Torus support loads.

(g) Torus attached piping loads.

Response 1 Simultaneous actuation of all eleven S/RVs is highly unlikely. However, the Authority's mport

- (NEDC-31967P, Reference 3) conservatively considered the increased loads that would result from the simultaneous actuation of all S/RVs. NEDC-31967P considered the reports and analyses identified below and concluded that a single nominal S/RV setpoint of 1110 psig will not r ssult in pipe loads and stresses in excess of those calculated previously.

FitzPatrick's main steam lines (MSLs) were analyzed assuming the simultaneous actuation of all S/RVs at 1140 psig (OPE 10-379, Reference 4). The analysis in OPE 10-379 considered the effects of increased MSL motion and the presence of attached piping. This report analyzed MSL C because, based upon an engineering evaluation, it is the most highly stressed MSL FitzPatrick's Mark i Plant Unique Analysis Reports (PUAR, References 5 and 6) calculated the effects oi a simultaneous actuation of all S/RVs at 1140 psig on submerged structures, torus shell,

' supports and attached piping. A subsequent analysis (TR-7543-1, Reference 7) performed as part

'of the FitzPatrick power uprate program demonstrated that the stresses calculated in the PUAR Page 1 I

l l Attachm nt 1 to JPN 92-001 are greater than those calculated for an 1110 psig S/RV relief set point. A copy of TR-7543-1 is included as Attachment 11.

Section 2.0 of tne PUAR for torus attached piping (PUAR/ TAP) discusses the S/RV piping l l analysis. Section 2.1 addresses applicable codes and standards; Section 2.2 describes the analysis methods used and Section 2.4 summarizes the results of the analyses.

Section 3.0 of the PUAR/ TAP discusses torus attached piping. Section 3.1 addresses applicable )

l codes and standards; Section 32 addresses torus attached piping loads; Section 3.3 describes i l the analysis methods used and Section 3.4 summarizes the resdts of the analyses.  !

l Sections 4.6 (" Containment Integrity") and 7.0 (* References *) of NEDC 3196P are cur:'ently being revised to include some of this information. The Authority will submit Revision 1 to ME00 3196P ,

as soon as it is available.

1 l

Page 2 L

Attachm:nt 1 to JPN 92-001 Question 2 In the report (NEOC-31697P) submitted January 16,1990, it is stated that the Operational Basis Earthquake (OBE) is combined with other loads to calculate the stresses in the particular structures. However, the plant load definition report (PLDR) and NUREG-0061 specify load combinations which also include the Safe Shutdown ".-thquake (SSE). How was the SSE included in the analysis of loads for the proposed f.fP setpoirits?

Response 2 Safe shutdown earthquake loads were combined with other loads to calculate stresses in the

. S/RV discharge lines. This point has been clarified Iri Revision 1 to NEDC-31697P.

l 1

l Page 3 l

I l

Attachment 1 to JPN 92 001 Ouustion 3

-In the May 25,1984 lotter from J. Silva, GE, to L Guaquil, NYPA, it is stated that because of an error in the computer code RVFOR, a study was performed to determine the effect on the originally analyzed loads. It was determined that a 40% to 50% reduction in maximum water clearing thrust loads are expected for a typical S/RV discharge line. However, the letter also

recommends that plant unique calculations be performed in order to quantify the reduction of Icads. Based on an examination of the force equation in error, it does not appear that the amount of reduction could be casily determined with a simp!c factor on the resulting forces. Please verify through more detailed analysis, that the additional loads due to the increased S/RV setpoints will not exceed those allowed for the S/RV discharge piping and other structures.

, Response 3 Potential Effects of Error The potential effects of this error have boon carefully studied. The error only affects the calculation of water clearing thrust loads and associated stiess loads on the submerged S/RV discharge line piping and causer, RVFOR to overestimate water clearing thrust loads.

The results of tests conducted at Monticello were compared with stress loads predic'ed by a version of RVFOR with the errot (see NEDO-237491, Referenc3 9) and a corrected version of RVFOR.

These comparisons show that Monticello's test results are consittent with the loads predicted by the corrected versions of RVFOR. They a!so show that RVFOR with the error overestimates the water clearing thrust loads for a typical BWR by as much as 50 percent.

Plant Unique Tests Appendix 1 to TR 5321 1 (RJerence 5) describes how tests conducted at PtzPatrick in April 1982 were used to verify the accuracy of the computer model. One hundred transducers were installed on the torus shall, tee-quencher and support, vent header support and downcomers. Empirical i factors were applied to stresses calculated by the RVFOR computer model to better correlate with test results, t

RVFOR Calculations for FitzPatrick Margins exist in the stresses calculated by RVFOR water clearing thrust loads on the T-quencher, quencher rapport and submerged S/RV discharge piping at FitzPatrick.

Teledyne conservatively assumed an S/RV setpoint of 1145 psig in their analyses. This is conservative because stresses calculated based on an 1145 psig S/RV setpoint will rtSult in higher stress levels than with an 1110 psig setpoint.

Page 4

Attachm:nt 1 to JPN 92-001 1

Based on Telodyne's engineering evaluations, thoro is greater than a 30 percent margin to the code stress allowables for the too quenchor, quencher supports and submerged S/RV piping.

See paget 90-91 of Reference 5.

Page 5

Attachment 1 to JPN 92401 a

Question 4 Discuss the troplications the singlo S/RV cotpoint and 2 D/RV out.of tervice proposal will have on plant response to a MSIV clot,.ro ATWS assuming that the scram falluto is due to a mechanical problem (1.o., ARI falls). Includo in your ditaussion, the offects this scenario will havo on ,

conV'iment and cora damcgo.

, Ret onso 4 fel'on 3.1.1 of NED924222 (Reference 10) summarlzos the results of transient ana!ysos, including predicted poak reactor pressure,in the ovent of an ATWS ovent. MSIV closure following a failure to scram causes st,mo of the most sovoro ATWS conditions.  !

NEDE 24222 concludos that tho most limiting caso of MSIV closurc with ARI failuro will result in a peak reactor vossol prt,ssure of 129G psig. This is loss than the ATWS limiting case of 1500 psig.

Soo Section 4.1 of Reference 10.

With two S/RVs out of sorvico, the ,1ominal rellof valvo capacity will bo reduced by approximately 18 percont. The peak vossol pressuro will increaso by 137 psl to approximately 1433 psig. This is also below the peak ATWS prossuro of 1500 psig. Soo Figuro 3.1.4.1.9 of Reforonce 10.

Based oi, Socilon 4 of NEDE 24222, there is substantial rr.argin in the plant's ability to cool the coro and safoly shut down the plant. The guldance in Appendix IV to Volumo 3 of NUREG44GO (Referenco 11) can be satisfied during the most sovoro ABVS ovents. Thoso conclusions ato appl! cable to FitzPatrick.

4 i

Pago6 l

1 l

I I

l Attachment 1 to JPN 92-001 c

R_oferoncos

1. NRC letter, B. C. McCabo to R. E. Doodle, dated July 1,1991 regarding roquest for additional information . proposod technical spocification change rogarding safoty roliof valvo sotpoints for the James A. FitzPatrick Nuclear Power Plant (TAC No. 755561),

2.

NYPA lotter, J. C. Brons to NRC dated December 20,1989 (JPN 89484) rogarding proposed changos to the technical specifications regarding updated S/RV performanco toquiroments and miscellaneous changos (JPTS 89-017). )

3. NYPA lotter, J. C. Brons to NRC dated January 16,1991 (JPN 90 010) togarding propriotary material supporting the proposed change to the technical specificathn regarding updated S/RV performanco toquirements and miscollanoous changos (JPTS-89-017). Includos copy of NEDC 31697P,

• Updated S/RV Por'ormance Roquiromonts ,

for the James A. FitzP.itrick Nuclear Power Plant," dated April 1989.

4. Gonoral Electric Report OPE 10-379, ' Evaluation of Main Steam Uno 'C' Piping Systom, prepared for the Power Authority of the Stato of Now York,' dated May 29,1979.

5. Totodyne Report TR 5321 1, Revision 1, " Plant Uniquo Analysis Report of the Torus Suppression Chamber for FitzPatrick,' dated September 25,1984.

6. Tolodyne Report TR 53212, Revision 1, ' Mark l Containment Program, Plant Unlaue Analysis Report of the Torus Attached Piping for FitzPatrick,' dated November,1984

7. Tolodyno Report TR 75431, Revision 1,

• James A. FitzPatrick Nuclear Power Plant ,

Evaluation of Supprossion Chambor and Main Etoam Safoty Roliof Unos for Simultaneous Actuation of All Safety Rolief Valvos Set at 1145 psig,' dated October 9, 1991.

8. Gonoral Electric Roport NEDO 21888, ' Mark i Containment Program Load Dofinition Roport,' Rovision 2, dated November 1988.

9. Gonoral Electric Report NEDO-237491, " Mark l Containment Program Comparison of Analytical Model for Computing S/RVOL Transient Pressures and Forcos to Monticello Data.* dated February 1979.

10. General Electric Poport NEDE 24222, ' Assessment of BWR of ATWS," Volume II, (NUREG-0406 attornate 3), dated Docomber 1979,

11. NUREG-0460, Volume 3, ' Anticipated Translonts Without Scram for Ught Water Reactors Staff Report,' dated Docomber 1978.

Pago 7

- - - . -