ML19262A430
| ML19262A430 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/28/1977 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19260A170 | List: |
| References | |
| NUDOCS 7910290778 | |
| Download: ML19262A430 (13) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION IN THE MATx u 0F DOCEET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical 3pecification Change Request No.
53 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, en the date given belov, been filed with the U. S. Nuclear Regulatory Ccesission and been served en the chief executives of Londonderry Township, Dauphin County, Penn;ylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:
Mr. Weldon B. Arehart Fr. Harry 3. Reese, Jr.
Beard of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Chur::h Read Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 MEIROPOLITAN EDISON COMPANY 37 /s/ R. C. Arnold Vice President E
Dated: '
1480 279 49102907 7
METROPOLITAN EDISON COMPANY JERSEY CENTRAL PO'JER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR- 50 Docket No. 50-289 Technical Specification Change Req 3est No. 53 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.
METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President Sworn and subscribed to me this 28th day of April , 1977
/s/ L. L. Lavfer Notary Public 1480 280
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Suecification Chance Request No. 51.1 The licensee requests that the attached revised page replace page 6-1 of the existing Technical Specifications, Apcendix A.
Reascns for Proposed Chance Technical Specification 6.1.1.a.5 requires the Unit Superintendent to forward a copy of evaluations conducted pursuant to Technical Specifications 6.1.1.a.2 (proposed changes to procedures, equipment, or systems which constitute a change of the facility or procedures as described in the FSAR) and 6.1.1.a.3 (proposed tests and experiments not described in the FSAR) to the Manager-Generation Engineering so that the Corporate Technical Support Staff (CTSS) can verify that an unreviewed scfety question was not involved.
Normally, the change requiring evaluation is originated by the unit staff. However, such changes may be originated by the CTSS and a written safety evaluation transmitted to the station. Accordingly, should the Unit Superintendent concur with the CTSS safety evaluation, no second review by the CTSS should be required before implementing the change if the specific CTSS evaluation is used and its require-ments invoked. The revised Technical Specification 6.1.1.a.5 vould not then require the Unit Superintendent to forward copies of such change / modification evaluations to the Manager-Generation Engineering in those cases when the CTSS has already completed and documented their evaluation >f the change / modification.
Safety Analysis Justifying Change This proposec . Technical Specification change does not involve any unreviewed safety question in that it only serves to avoid duplication of effort by the CTSS and reduce the time needed to put applicable changes into effect. In no way is the intent of the original specifi-cation violated since the CTSS vould still be required to evaluate the subject changes and the Unit Superintendent's concurrence is still required.
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6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1.a. The Unit Superintendent shall be responsible for the overall safety of plant operations and shall ensure that:
- 1. All proposed changes to procedures, equipment, or syste=s are evaluated to determine if they constitute a change to the facility or procedures as described in the Final Safety Analysis Report.
- 2. All proposed changes to procedures, equipment, or systems which constitute a change of the f acility or procedures as described in the Final Safety Analysis Report are evaluated to determine that they do not involve an unreviewed safety question as defined in paragraph 5).59 (c), Part 50, Title 10, Code of Federal Regulations.
3 All proposed tests and experiments , not described in the Final Safety Analysis Report, are evalua:ed to determine that they do not involve an unreviewed safety question as defined in paragraph 50 59 (c), Part 50, Title 10, Code of Federal Regulations.
4 Records are kept: a)' of changes to procedures, equipment or systemr completed under the provisions of paragraph 50.59 (b),
Part 50, Title 10, Code of Federal Regulations; b) of tests and experiments conducted in accordance with those provisions; and c) of the written safety evaluation used as a basis for determining that such changes, tests and experiments do not involve an unreviewed safety question.
- 5. Copies of evaluations conducted pursuant to 6.1.1.a.2 and 6.1.1.a.3 above are forwarded to the Plant Operations Review Committee, the General Office Review Board Secretary and the Manager-Generation Engineering. I:7 these evaluations were previously made by Generation Engineering, they need not be forwarded to the Manager-Generation Engineering.
- b. The Unit Superintendent shall have the authority to:
- 1. Make . determination that proposed changes to procedures, equipment, or systems do not involve a change to the procedures or facility as described in the Final Safety Analysis Report.
- 2. Make a preliminary determination that proposed changes to procedures, equipment or systems as described in the Final Safety Analysis Report, or that proposed tests or experiments not described in the Final Safety Analysis Report do not constitute an unreviewed safety question; however, such a determination must be based upon formal vritten evaluation.
3 Direct the Plant Operations Review Committee to review:
- a. Evaluations cf proposed changes to procedures, equipment or systems; 6-1 1480 282
Three Mile Island Nuclerir Station Unit 1 (TMI-1)
Operating. License No. DPR-50 Docket No. 50-289 TECHNICAL SPECIFICATION CHANCE REQUEST NO. 53.2 The licensee requests that tne attached revised rage replace page 3-35 of the existing Technical Specifications.
REASON FOR PROPOSED CHANGE Technical Specification 3.5 2.; presently states that, "A power map shall be taken to verify the expected ',over distribution at periodic intervals of approximately 10 full power days using the incore instrumentation detection system". However, the intent of this requirement is to determine if any anomaly exist in the e-pe power distribution and not to verify expected power distributions. This re.ision, therefore, changes the language of Technical Specification 3.5 2 7 to reflect the intent of this requirement. In addition, following Standard Technical Specificatious guidelines, the interval between successive power mappings is being changed from 10 FPD to 10-30 FFD.
SAFETY ANALYSIS JUSTIFYING CHANGE The subject change does not involve or constitute an unreviewed safety question in that it only serves to reflect the actual intent of Technical Specification 3.5.2 7 Periodic core power mappings vould sti31 be required to assure that the power distribution is within acceptable limits. Further, the change in the interval between power mappings is bounded by the requirements of the Standard Technical Specifications and does not, therefcre, involve an unreviewed safety question.
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3 5.2 5 Control Rod Ps.itions:
- a. Operating rod group overlap shall not exceed 25 percentt 5 percent, between two sequential groups except for physics tests.
- b. Position limits are specified for regulating and axial power shaping control rods. Except for physics tests or exercising control rods, the regulating control rod insertion /vithdreval limits are specified on Figures 3 5-2A, 3 5-2B, and 3 5-2C for four pump operation and Figures 3 5-2D, 3 5-2E, and 3.5-2F for three or two purp operation. Also excepting physics tests or exercising control rods, the axial power shaping control rod insertion /vithdrawal limits are specified on Figures 3.5-2K, 3.5-2L, and 3 5-2M. If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.
- c. Except for physics tests, power shall not be increased above the power level cutoff (See Figures 3 5-EA, 3 5-2B and 3 5-2C) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability,
- d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above h0 percent of rated power. Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shall be taken to maintain operation within the envelope defined by Figures 3.5-2G, 3 5-2H and 3 5-2I. If the imbalance is not within the envelope defined by Figures 3.5-2G, 3 5-2H and 3 5-2I corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.
- e. Safety rod limits are given in 3.1.3 5 3 5.2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
3 5.2 7 A power map shall be taken at periodic intervals of 10 to 30 full power days using the incore instrumentation detection system, to verify that the power distribution is within the limits shown in Figure 3.5-2I.
3 ees 1480 284 The power-imbalance envelope defined in Figures 3 5-2G, 3 5-2H, and 3.5-2I is based on LOCA analyses which have defined the maxi =um linear heat rate (see Figuze 3 5-2J) such that the maximum clad temperature vill not exceed the Final Acceptance Criteria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the withdrawal /insertien limits as defined by 3-35
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Srecification Change Recuest No. 53.3 ,
The licensee requests that the attached changed pages replace pages b-9 and h-10 of the existing Technical Specifications, Appendix A.
Reasons For Pronosed Chance The reason for this revision is to correct the nomenclature that describes the analysis used to determine radioactive contamination of the secondary coolant due to primary to secondary leakage at TMI-1. The "15 minute gross degassed beta gatsa activity" analysis is typical of that used for primary coolant determinations. In that application, minimum sensitivity is not as significant a factor and consistent, reliable results can be obtained from as little as 1 ml of sample. This procedure requires that the sample be obtained, prepared, and counted in approximately 15 minutes.
In order to obtain the sensitivity required by the Technical Specifications (less than 2x10-OpCi/ml) in the determination of radioactivity in the secondary coolant, a volume of at least 100 ml of secondary coolant must be concentrated by evaporation. This procedure requires at least 100 minutes.
The standardized Technical Specifications identify a determination of " gross activity" for determining primary to secondary system leakage in the secondary coolant system. This nemenclature is being applied to the TMI-1 Technical Specifications.
Safety Analysis Justifying Change This proposed change does not involve any unreviewed safety questions in that the intent of the existing Technical Specification for activity analysis of the secondary ecolant will not be changed with this revision. Sampling frequency and analytical sensitivity will not be : hanged.
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TABLE h.1-3 MINIMUM SA?eLING FREQUENCY Item Check Freauency
- 1. Reactor Coolant a. Radio-Chemical Analysis (l'
, Monthly
- b. E determination (2) Sesiannually
- c. 15 Min. Gross Degassed 5 times / week when Tavg Beta-Gamma Activity (1) is greater than 200 F
- d. Tritium Radioactivity Monthly
- e. Chemistry (C1, F and 02) 5 times / week when Tavg is greater than 200 F
- f. Boron Concentration 2 times / week
- 2. Berated Water Storage Boron Concentration Weekly and after each Tank Water Sample makeup when reactor coolant system pressure is greater than 300 psig or Tav is greater than 2000F
- 3. Core Flooding Tank Boron Concentration Monthly and after each Water Sample makeup when RCS pressure is greater than 700 psig
- h. Spent Fuel Pool Boron Concentration Monthly and after each Water Sample makeup 5 Secondary coolant a. Gross Activity Weekly when reactor coolant system pressure is greater s than 300 psig or Tav is greater than 200 F
- b. Iodine Analysis (3)
- 6. Boric Acid Mix Tank Boron Concentration Twice weekly or Reclaimed Boric Acid Tank 1480 286
- 10. Sodium Hydroxide Concentration Quarterly and after Tank each makeup h-9
TABLE 4.1-3 (Continued)
Item Check Frecuency
- 11. Sodium Thiosulphate Concentration Quarterly and aftar Tank each makeup
- 12. Condenser Partitlen 131 I Partition Factor Once if primary /
Factor secondary leakage developes, i.e. Gross Beta-Ga=ma on secondary side of OTSG is greater than l! x 10-8 micro curies per cc and evidence of fission products is present (1) When radioactivity level is greater than 10 percent of the limits of Specification 3.1.4, the t ampling frequency shall be increased to a minimum of 5 times per veel.
(2) Ei determination vill be started when the 15 minute gross degassed. beta-gamma activity analysis indicates greater than 10 pCi/ml and vill be redetermined each 10 pCi/ml increase in the 15 minute gross degassed beta-gn=ma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95 percent of radionuclides in reactor coolant with half lives of >30 minutes.
(3) When the gross activity increases by a factor of two above background, '
an iodine analysis vill be made and performed thereafter when the gross activity increases by 10 percent.
1480 287 h-10
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Chance Request No. 53.h The licensee requests that the attached revised page replace page 4-36f of the existing Technical Specifications, Appendix A.
Reason for Fronosed Change Technical Specification h.h.2.2.D, which is part of the original document of the TMI-l Technical Specifications, requires the submittal of the ring girder inspection results "within 30 days after the completion of each ring girder surveillance inspection." However, the recently approved Technical Specifica-tion 6.9.3.B (Amendment No. 11; February 26, 1976) requires the submittal of this report within three months after the performance of the inspection.
The purpose of this change request .4s to make Technical Specification 4.4.2.2.D consistent with the reporting require . ents of Technical Specification 6.9.3.B.
Safety Analysis Justifvine Chance This proposed Technical Specification change does not involve any unreviewed safety questions in that it only serves to revise a reporting requirement.
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(c) Inspection During the Ring Girder Surveillance:
(1) The concrete adjacent to all dome tendon bearing plates shall be visually inspected. A vritten description of the crack appearing in the concrete adjacent to the bearing plates shall be made. The description shall give the number and direction of visible crac)3 and the maximum and minimum dimensions of crack vidths.
(2) If any concrete cracks open vider than 0.010 inch, the inspector shall i= mediately notify the Engineer for an immediate evaluation and resolution.
(3) The data recorded during each ring girder surveillance shall be made available for comparison with the follow-ing surveillance inspections.
D. Submission of Inscection Results A detailed report shall be filed with the Unit'd States Nuclear Regulatory Commission within three months after ':he completion of each ring girder surveillance inspection.
1480 289 k-36f
Three Mile Island Neelear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Recuest No. 53.5 The licensee requests that the attached changed page replace page 3-27 of the existing Technical Specifications, Appendix A.
Reasons for Proposed Change The reason for this revision is to alloc the use of the shutdown bypass switch associated with each reactor protection channel during reactor power operation for testing and maintenance purposes. Technical Specification 4.1.1, Table h.1-1, Item 8 requires the monthly testing of the high reactor coolant pressure channel to verify that trip action is as designed. The channel shutdown bypass switch must then be placed in the bypass position to verify that a trip signal from the shutdown bypass bistable vill trip the channel.
Safety Analysis Justifvine Change The proposed change does not involve any unreviewed safety questions in that it only includes a provision in the Technical Specifications that vill permit the testing or preventive maintenance of the reactor protection channels shutdown bypass bistable and thus insure its proper operation. Further, the manual bypass of reactor protection channels for on-line testing or maintenance on only one channel at a time is allowed by the Technical Specifications. The use of a single channel shutdown bypass switch after the required channel bypass constitutes then an operation bounded by present requirements as long as the required number of channels are in operation.
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,=
35 INSTRUMENTATION SYSTEMS 3 5.1 OPERATIONAL SAFETY INSTRUMENTATION Applicability Applies to unit instrumentation and control systems.
Objective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.
Snecifications 3 5 1.1 The reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5-1, Column 'A' and 'B' are met.
3 5.1.2 For on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel vill be used to lock the reactor trip module in the untripped state as indicated by a light. Only one channel shall be Iceked in this untripped state at any one time. Unit operation at rated power shall be permitted to continue with Table 3.5-1, Column "A". Only one channel bypass key shall be kept in the control rocm.
3 5 1.3 In the event the number of protection channels operable falls below the limit given under Table 3 5-1, Column "A", operation shall be limited as specified in Colu=n "C".
3.5 1.h The key operated shut + a bypass switch associated with each reactor protection channel s ." ' not be used during reactor power operation g except for required mn atenance or testing. l 3.5.1 5 During startup when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade.
3.5 1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the failed trip device shall be manually removed within 30 minutes. The condition vill be corrected. The remaining trip device shall be tested within eight hours. If the condition is not corrected and the re=aining trip devices are not tested within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the reactor shall be placed in the hot shutdown condition within an additional h hours.
Bases Every reasonable effort vill be made to maintain all sefety instrumentation in operation. A startup is not permitted unless three power range neutron instru-ment channels and two channels each of the folleving are operable: "c 3-27 1480 29 ar ,