ML19261F144
| ML19261F144 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/25/1978 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| GQL-0565, GQL-565, NUDOCS 7910240803 | |
| Download: ML19261F144 (6) | |
Text
.
g 65 o, nf REGULATORY INFORMATION DISTRIEUTION SYSTEM (RIDS)
DISTRIEUTION FOR INCOMING MATERIAL 50-289 REC: GRIER E H ORG: HEREEIN J G DOCDATE: 04.<25/78 NPC METROPOL EDISON DATE RCVD: 05/01/79 COPIES PECEIVED DOCTYFE: LETTER NOTARIZED: NO LTR 1 ENOL 1
SUBJECT:
FORWARDING COFY OF REPT CHANGES TO SUGJECT FACILITY AS DESCRIBED IN THE TMI-1 FSAR, CONDUCTED WITHOUT PRIOR NRC VAL DURING 1977, IN ACCORDANCE WITH REQUIREMENTS OF.CCFR50.59(B) AND
SUMMARY
OF THE SAFETY EVALUATION ASSOCIATED WITH EACH CHANGE PLANT NAME: THREE MILE IILAND - UNIT 1 REVIEWER INITIAL:
XJM DISTRIEUTOR INITIAL: p
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DISTRIEUTION OF THIS MATERIAL IS AS FOLLOWS *****************+
ANNUAL & SEMI-ANNUAL OPERATING REPORTS (OL STAGE).
(DISTRIEUTION CODE AGO8)
FOR ACTION:
BR CHIEF REID**W/6 ENCL INTERNAL:
EGJ1LE** M NCL-D NRC PDR**W/ ENCL A & E**W/2 ENCL MIFC**W/2 ENCL HANAUER**W/ ENCL STELLO**W/ ENCL EISENHUT**W/ ENCL SHAO**W/ ENCL BAER**W/ ENCL BUTLER **W/ ENCL EEB**W/ ENCL P.
CHECK **W./ ENCL J.
COLLINS **W/ ENCL EXTERNAL:
LPDR'S HARRISEURG, PA**W/ ENCL NATL LAB ANL**W/ ENCL TIC **W/ ENCL NSIC**W/ ENCL ACRS CAT B**W/15 ENCL 1485 344 DISTRIEUTION:
LTR 41 ENCL 41 CONTROL NER:
781220080 S!ZE: IP+4F
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THE END
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METHOPOLITAN EDISON COMPANY
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POST OFFICE BOX 542 READING. PENNSYLVANIA 19603 TELEPHONE 215 -929 3601 April 25,1978 GQL 0565 Pr. 3. H. Grier, Director Office of Inspection & Inforcement, Region I U. S. Nuclear Regulatory Co =ission 631 Fark Avenue King of Prussia, Pennsylvania 19406
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Docket Ho. 50-289 Operating License No. DPR-50 In acco1 dance with the requi2 v.ents of 10 CFR 50 59(b), attached please find two copies of the report of changes to our facility, as described in the TMI-1 F.11.R, conducted without prior IEC approval during 1977. A su==ary of the safety evaluation associated with each chan6e is also included.
No tests or experinents pursuant to 10 CFR 50 59(a) were carried out during the year.
Sincerely, Signed J. G. E.
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J. G. Herbein Vice President-Generation J3H:DGM:dkf s(.,. '.
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Attachment:
Report of Facility Changes Conducted
,l, without Prior I2C Approval
- s s
cc: Director, Office of Inspection & Enforcement (40 copies) 5,. h U. S. Nuclear Regulatory Commissica V
c/o Distribution S rvicca Branch E 0. f.DU G,,x ashington, D. C.
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1.
Channe Mod $592: Re=ove Surveillance Spec. Holder Tube 2
Co=pletion Date: 3/1/77 3.
Description of Chan:re:
The change involved the re= oval of the three SurveilhneeSpecimen Holder Tubes (SSET) from the Core Support Assembly. The change was performed because of M ge incurred during the course of nor=al operation. Con-tinued use of the tubes '
not possible.
The Unit 1 specimens were not irradiated in the reactor vessel during Cycle II operation. It was anticipated that durin6 Cycle II a new tube would be designed for instalh tion prior to Cycle III.
(See Technical Specification Change Request 32)
An acceptable design could not be found and an integrated surveillance program was developed for implementation in Unit II in accordance with 10 CFR 50 Appendix H Paragraph II.C.4.
(See Technical Specification Change Request h0.)
Approval of this program was gained thri igh acceptance of the Cycle III Re-load Report.
(See Technical Specification Change Request 45.)
4.
Safety Ivaluation Su=ary:
The purpose of the surveillance program is to monitor changes in the fracture toughness properties of the reactor vessel material. Appendix E to 10 CER 50 outlines four objectives of the program, A site integrated reogram accomplishes those objectives.
There are no si nificant differences in tlc. fast neutron 6
flux spectrum, operating temperatures or other environmental conditions for the TI4I-l capsules in the T S-II core. It is, therefore, concluded that this change does not represent undue risk to the health and safety of the public.
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1.
Change Med #850; iC-R7:dA/B. Change Setpoint 2.
Comoletion Date: 6/13/77 3
Descriution of Chance:
This change concerned resetting of the Pressurizer Code Safety Valves to 2500 psig in accordance with Technical Specification Change Request #39 submitted to the IIRC. The reason for the change was to improve the capa-bility of the TNE-1 plant to withstand a less of electrical load (LOEL) frcm 100% power without tripping the reactor. The high pressure trip set-point was increased to 2h05 psig, and in conjunction with this, the safety valve setpoint was incressed to 2500 psig.
h.
Safety Evaluation Sum =ary:
The vorst case accident (i.e. Feedvater Line Break) was reanalyzed in light of the proposed changes. The results of this reanalysis indicated a peak RC system pressure of 273h psig. Based on this, it has been concluded that a high pressure trip and pressure code safety valve setting limit of 2h05 psig and 2500 psig, respectively, will maintain RCS pressure below the safety limit of 2750 psig for any design transient. This change, therefore, does not represent undue risk to the health and cafety of the public.
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1.
Change Med #852: Fu ual Tie to Panel ATA 2.
Completion Date:
5-6-77 3
Descriptien of Change:
A manual tie breaker was installed to permit panel APA to be supplied directly from Vital bus VBA. The normal supply is from VBA through an automatic transfer switch. The manual tie vill be used during planned maintenance or in the event of c failure of the static switch.
h.
Safety Evaluation Summary This change resulted in no unreviewed safety question nor created an unsafe condition in that the manual tie, as installed, added a backup power source to the ATA panel, thereby enhancing safe operation.
1485 548
1.
Procedure Revision:
0.P. 110h-29E Rev. T 2.
Ccmoletion Date:
5/06/77 3
Description of Revision:
The procedure revision provided guidance to permit bypassing the dilution-control rod interlock to allow system cleanup after reactor refueling.
h.
Safety Evaluation Surnary:
The dilution-control rod interlock terminates the dilution cycle, if the regulating rod group is inserted to the 755 vithdrawn position, to pre-vent inadvertent excessive dilution of the reactor coolant. Administrative controls, chemistry sa=ples, and the appropriate rign offh controlled the process so as to prevent any nuclear safety impact Or threaten the health
.nd safety of the public.
1485 349