ML18087A602

Submits Addl Info to Support 821005 Request for Amend to Licenses DPR-70 & DPR-75.Info Includes Analysis Results Demonstrating Motivation for Tech Spec Change & Effects on Core Limits & Axial Offset DNB Limits

ML18087A602
Person / Time
Site:	Salem
Issue date:	11/18/1982
From:	Liden E Public Service Enterprise Group
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 8212010155
Download: ML18087A602 (8)

Text

OPS~G Public Service Electric and Gas Company P.O. Box E Hancocks Bridge, New Jersey 08038 Nuclear Department November 18, 1982 Director of Nuclear Reactor Regulation u.s. Nuclear Regulatory Commission 7920 Norfold Avenue Bethesda, MD 20014 Attention: Steven A. Varga, Chief Operating Reactors Branch 1 Division of Licensing Gentlemen:

SUPPORTING INFORMATION FOR AMENDMENT REQUEST TO FACILITY OPERATING LICENSE DPR-70 AND DPR-75 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NO. 50-272 AND 50-311 Attached is additional information to support our Request for Amendment in Facility Operating Licenses DPR-70 and DPR-75 dated October 5, 1982. This information includes analysis results which demonstrate the motivation for introducing the F~H Technical Specification change request as well as the effects on core limits and axial offset DNB limits due to such a change.

very truly yours,

,,.,.!%etc~~

EY. A. Liden

/"Manager - Nuclear Licensing

~ and Regulation

• CC: Leif Norrholm NRC Senior Resident Inspector William Ross NRC Licensing Project Manager Attachment 8212010-£55' 021 i i --

• 8 -\

PDR ADOCK 05000272 p PDR The Energy People 95-2169 (10M) 8-82

ATTACHMENT 1 Additional Information in Support of FNH Technical Specification Change Request dated October 5, 1982 NFU2/2 03/5

Page 1 ANALYSIS FOR A REVISED F~H LIMIT AT LESS THAN RATED THERMAL POWER Historically, increasing the allowable FNH with de-creasing power has been permitted by all previously approved Westinghouse designs. Th~ increase is permitted by the DNB protection setpoints and allows for radial power distribu-tion changes with rod insertion to the insertion limit.

The results of the Salem Units F~H*Tech Spec limit analysis indicate that the limit may be modified by changing the limit slope from 0.2 to 0.3 at reduced power, resulting in the following relationship:

F~H = 1.55 [1.0 + .3 (l.;_P)]

where P =fraction of rated thermal* power. Note that the

• only change from the current FNH Tech Spec is the mu 1 t i p 1 i er on the qua n t it y * ( 1-P ) fr om 0

• 2 to 0

• 3

• No ch an g e is made in the F~H limit at full power.

Figures 1 and. 2 show the results of a conservative calcu-lation of the enthalpy rise hot channel factor, Ftlff, as a function of power level for Salem Unit 1 Cycle 5 and Salem Unit 2. Cycle 2 respectively.

The calculations were performed using the current Technical Specification rod insertion limits for Salem Unit 1 and a proposed rod insertion limit for Salem Unit 2 as shown in Figure 3. Salem Unit 1 Cycle 5 contains 52 feed assemblies, while Salem Unit 2 Cycle 2 has been designed with 72 feed assemblies as part of a transition to eighteen month cycles. The Salem Unit 2 FNH analysis is more typical of eighteen month cycle behavior since eighteen month cycles planned for both units require between 72 to 80 feed assemblies.

The F~ Technical Specification modification is recommended for Salem Uni ts 1. and 2 to allow optimization of the core loading pattern by minimizing restrictions on the F~H at low power. These changes will also minimize tne probability of making rod insertion limit.changes in the future to satisfy peaking factor criteria at low power with the control rod banks at the insertion limit.

NFU2/2 04/5

Page 2 Core limits and axial offset limits for an increased all6wable F~H at._reduced power levels were determined for Salem Units 1 and 2. The core limits at 1775 and 2000 psia remain unchanged from the current limits. At 2250 and 2400 psia, the proposed core limits are slightly more limiting below 100%

power. A new set of core limits, which apply to both Salem Units, is attached. The core limits have these minimal changes because at most conditions below full power, the restriction that the average enthalpy at the vessel exit be less than the enthalpy of saturated liquid is more limiting than DNB considerations. -This vessel exit enthalpy limit is not core peaking factor dependent. Current reactor core safety limit setpoints bound core limits obtained using an increased F~H at reduced power and_therefore no change to the K1, K2, K3, K4, Ks, or K6 factors is required in the Technical Specifications of either unit as a result of the new core and axial offset limits.

Therefore, no changes to the overpower and overtemperature 6T setpoints are necessary and hence no accident reanalysis is required. Since current f (6I) function bounds the new axial offset limits, there is no requirement to change the f ( hl) function i~ the Technical Specification of either unit.

These modifications were made in WCAP-9500 and approved by the NRC with the aqditional technical supporting information supplied by NS-TMA-2323, (letter from Anderson to Miller, dated October 24, 1980).

NFU2/2 05/5

-WESTINGHOUSE PR_O~RIETARY c~ss* .

FIGURE 1 SAL EM UN IT 1 . ( P-S E) . CY CL E 5 CONSERVATIVE CALCULATION OF ENTHALPY RISE FACTOR WITH POWER LEVEL AND TECHNICAL SPECIFICATIONS LIMITS 2.l' ~~~K~Ei*v~: .=:.~=!=.~..  ::=._t;:_:::_~. -~* -~-F~~~il~-~--~-.~--~-*ii~~

=..  ;;::_.

~-~~~~~~~~~~~~~~~~~~~~~~~!

~

~ e - calculated points with 2.0

= 8% uncertainty included: : --

(::' ' ** .. " ".

1.9 - ........~.-------~+--~--+-"-+"----+-"-'---'-<....-,- .,----...* ~ *

':> ~1......-*--'-' >-'--'-+"---+-....._._.__,

. ",.' ' . . )( _.. " ..... ~~--

0 I-:-

<:>~

'\..

1.8

' I ' '

• --- ' ' ' ' * ' 1 ' *

1. 7 ' .' '.

1.6 . .' .

' t ' I<.,

1.5 . '.

1.4 1.0 .75 .50 .25 0 Fraction of Rated Power

FIGURE 2 SALEM UNIT"2 CYCLE 2 .

CONSERVATIVE CALCULATION OF ENTHALPY RISE FACTOR WITH POWER. LEVEL BASED ON ROD INSERTION LIMIT IN FIGURE 3 2.1 --

calculated points with **..:.,_

8% uncertainty included

= L 2 0 1 9

'+-,.

' I

\'\-

1.6  ??

~

~ ~~~~

1. 5 1.4

1. 0 *8 .6 *4 ** 2 0 Fraction of Raied Power

FIGURE 3 SALEM UNIT 2 CYCLE 2 ROD INSERTION LIMITS USED FOR F~H ANALYSIS 228

-i 200 ,

160 z

0 H 120

~

H -,

Cf.l 0

P-i z

~

p:::i 80 ,

• -+ '-*~

40 '.,

0 ci *2 *4 .6 .8 1.0 FRAC~ION OF RATED PO~ER

....s.-

.:660 1-

-0 QJ ..c

+J C'I l'CI ....

~:::c:

4- x 0 :::I' 640 ...-

~LL..

"'0

.... 0 c:

s.-

4- +J

~

0  :::I*

QJ LL.. *=

o* EZ 620  :!+  :::I E QJ Ci") . =

.... C'I x c:

> .. ro re re ~ 0:::

I-ro s...

QJ C/')

u 600 0

.;..io 3:

0::: ~

-0 QJ QJ

....E ..c....,

.µ

.1~'-i-'-'. ....J .0 580 ~ .. : ;...,! *

• s... s...

QJ QJ 3: 3:

0 0

~~

rc rc 560 .-,.

E E s... S-QJ QJ

..c ..c I- I-0 .2 .4 .6 .8

• 1. 0 1. 2 Fraction of Rated Thennal Power FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SALEM _;, UN IT 1 2-2