ML18087A631
| ML18087A631 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/13/1983 |
| From: | Liden E Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8301200091 | |
| Download: ML18087A631 (4) | |
Text
PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department January 13, 1983 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Steven Varga, Chief Operating Reactors Branch #1 Division of Licensing CYCLE 5 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Gentlemen:
Salem Unit No. 1 has concluded cycle 5 refueling and is expected to startup in early February.
The cycle 5 reload core consisted of 52 new Westinghouse 17xl7 fuel assemblies enriched to 3.4 w/o as discussed in our letter of September 27, 1982.
This letter is to inform you of a change in loading pattern and PSE&G review of the new cycle 5 reload core.
During the cycle 4/5, 52 fuel assemblies were replaced with Region 7 fuel (fresh), and two fuel assemblies were replaced with one Region 3 assembly and one Region 1 assembly which were both discharged at the end of cycle 1.
The cycle 5 loading pat-tern is shown in Figure 1.
This differs from the pattern in our September 27th transmittal in location L3.
One Region 4 assembly was replaced by a Region 1 assembly of equivalent reactivity.
~**
A review was performed on the revised Salem No. 1 cycle 5 reload core which addressed those incidents analyzed and reported in the Salem FSAR that could potentially be affected by the fuel reload.
This review was performed in accordance with the Westing-house reload methodology as outlined in the March 1978 Westing-ACJ()f house topical report "Westinghouse Reload Safety Evaluation f'
Methodology" (WCAP-9272).
All cycle 5 assemblies are of the same mechanical, nuclear and thermal hydraulic design as the cycle 4 assemblies.
The total peaking factor envelope is the same for both cycle 4 and cycle 5.
Kinetic parameter values 8301200011 The Energy People 95-2168 (5M) 10-82
Steven Varga, Chief January 13, 1983 for cycle 5 remain within the bounds of the current limits.
Since all parameters for cycle 5 remain within the bounds of the current limits, no accident reanalysis was required.
However, the dropped RCCA event was analyzed according to the new dropped rod methodology described in Reference 1.
Results show that the DNB design basis is met for all dropped rod events initiated from full power so that the interim operating restrictions are no longer necessary.
- However, until formal NRC notification is received to remove them, the plant shall continue to operate under the interim restric-tions.
The locations of the Optimized Demonstration Fuel Assemblies are shown in the attachment.
Both are instrumented with a moveable incore flux detector and thermocouple.
The criteria with respect to F~H and FQ used to determine the location of the Optimized Assemblies are as follows:
- 1.
Demonstration assemblies are placed in the core such that the lead power fuel rods operate at least 6% lower in F~H than the maximum allowed design value for the standard assemblies.
- 2.
Demonstration assemblies are located such that they operate with FQ values at least 0.10 lower than the design value for stand-ard assemblies.
PSE&G has reviewed the bases of the reload analysis and the Westinghouse Reload Safety Evaluation (RSE) Report with West-inghouse.
The review of all incidents demonstrated that the results of all the postulated events are within allowable limits.
The reload safety evaluation demonstrated that Tech-nical Specification changes are not required for operation of Salem Unit 1 at rated thermal power during cycle 4.
Salem's Station Operations Review Committee has concluded that no unreviewed safety questions as defined by 10 CFR 50.59 are involved with this reload.
Therefore, based on this review, application for amendment to the Salem Unit 1 operating li-cense is not required.
The reload core design will be verified during the startup physics testing program.
This program will include, but not be limited to, the following tests:
r Steven Varga, Chief January 13, 1983
- 1.
Control rod drive tests and drop time
- 2.
Critical boron concentration measurements
- 3.
Control rod bank worth measurements
- 4.
Moderator temperature coefficient measurement
- 5.
Power coefficient measurement, and
- 6.
Startup power distribution measurements using the incore flux mapping system.
RTB:dgh Attachment Very truly yours, E. A. Liden, Manager - Nuclear Licensing & Regulation cc:
Leif Norrholm; Senior Resident Inspector n.-~. Fisher, Licensing Project Manager Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
- e SALEM l CYCLE.S Region 1 - Standard 17xl7 2.2S w/o Region 3 - Stanqard 17xl7 3.30 w/o Region 4 - Standard 17xl7 2.80 w/o Region 4A-Optimized 17xl7 2.80 w/o R
p N
M L
K.
J Region ~A - Standard 17xl7 Region SB ~ Standard 17xl7 Region 6
- Standard 17xl7 Region 7
- Standard 17xl7 H
G F
E D
C.
B 8
7 4
SB 4
6 SA 6
3 6
8 SA 6
4 SB 4
8 X
Region Y
Number of Burnable Poison Rods SS Secondary Source Rod~
2.80 W/G 3.41 w/o 3.40 w/o 3.40 w/o A
7