ML20246G211
| ML20246G211 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/06/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20246G215 | List: |
| References | |
| DPR-50-A-150 NUDOCS 8907140193 | |
| Download: ML20246G211 (22) | |
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NUCLE AR REGULATORY COMMISSION h[
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y FETROPOLITAl' EDIS0t! CCMPANY t'EP.SEY CENTP.AL POWER & LIGHT COMPANY PENNSYLVANIA' ELECTRIC COMPANY GPU NUCLEAR CORPORATION DCCKET NO. 50-289 THPEE !!ILE ISLAND !!UCLEAR STATION, UNIT NO.1 AME!!DFENT TO FACILITY OPERATING LICENSE Amendment No.150 License f:o. DPR-50 1.
The tsclear Regulatory Com.ission (the Commission) has found tr.:t:
A.
The application for amencment by GPU fluelear Corporation, et al.
(the licensec) dated April 28,15E9, as supplemented on May 18, 1989 and June 19, 1989, complies with the standards and requirements of the Atenic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter !;
B.
The facility will operate in conformity with the application, the previsions of the Act, and the rules and regulations of the Comission; C.
There is rcasonable assurance (i) that.the activities authorized by this amendment can be conducted without endangering the health and safety of the pt.blic, and (ii) that such activities will be conducted in cortpliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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Accordingly, the-. license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.c.(2) of Facility Operatincj License No. DPR-50 are hereby amended to. read as follows:
(2) Technical Specification The Technical Specifications contained in Appendix A, as revised through Amendment No.150, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance. to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
' o v, l h[T -
, John F. Stolz,/') Director
/'DivisionofReactorProjectsI/II Project Directorate I-4 Office of Nuclear Reactor Reralation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 6,1989
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ATTACHMENT:TO LICENSE AMENDMENT NO no n
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-FACILITY OPERATING LICENSE NO. DPR DOCKET NO. 50-289 l
Replace the following pages of the Facility Operating. License and the Appendix A Technical Specifications'with the attached pages.. The revised pages are ice ntified by amendment number and contain vertical ~ 1ines indicating the area a
oi. change.
Remove In:S r. t_
.. i -
i v
v vi vi vii-vii viii viii 1-7 1-7 2-3 2-3 3-33 3-33 3-34 3-34
'3-34a 3-34a 3-35 3 3-35a' 3-35a 3-36 3-36 3-36a 3-36a 3-38 3-38 6-19 6-19 6-19a.
6-19a 5-4 5-4 5-5 5-5
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-_-_--________-_-__-____-_---a
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e L[
_ TABLE OF CONTENTS t
Section h
Pace 2
TECHNICAL SPECIFICATIONS 1
DEFINITIONS 1-1 1,1
_ RATED POWER 1-1 1.2
_REACIOR OPERATING CONDITIONS l-1 1.2.1 Cold shutdown 1-1 1.2.2 Hot Shutdown 1-1 1.2.3 Reactor Critical 1-1 1.2.4 Hot Standby 1-1 1.2.5 Power Operation 1-1 1.2.6 Pefueling Shutdown 1-1 1.2.7 Refueling Operation 1-2 1.2.8 Refueling Interval 1-2 1.2.0 Startup 1-2 T vg L2.10 A
1-2 1.2.11 Heatup-Cooldown Mode 1-2 1.2.12 Station, Unit, Plant, and Facility 1-2 1.3 OPERABLE 1-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument channel 1-2 1.4.2 Reactor Protection System 1-2 1.4.3 Protection Channel 1-3 1.4.4 Reactor Protection System Logic 1-3 1.4.5 Engineered Safety Features System 1-3 1.4.6 Degree of Redundancy 1-3 1.5 INSTRUMENTATION SURVEILLANCE 1-3 1.5.1 Trip Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Channel Check 1-4 1.5.4 Channel Calibration 1-4 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-5 1.6.1 Quadrant Power Tilt 1-5 1.6.2 Axial Power Imbalance I-5 1.7 CONTAINMENT INTEGRITY 1-5 1.8 FIRE SUPRESSIUN WATER SYSTEM l-5 1.12 D05E EQUIVALENT I-131 1-6 1.13 SOURCE CHECK 1-6 1.14 SOLIDIFICATION 1-6 1.15 0FF51TE 005E CALCULATION MANUAL 1-6 1.16 PROCES5 CONTROL PROGRAM l-6 1.17 GA5EOUS RADWA5TE TREATMENT SYSTEM l-6 1.18 VENTILATION EXHAUST TREATMENT SYSTEM l-6 1.19 PURGE-PURGING l-7 1.20 VENTING l-7 1.21 REPORTABLE EVENT l-7 1.22 MEMBER (5) 0F THE PUBLIC 1-7 1.23 SUBSTANTIVE CHANGE 5 l-7 1.24 CORE OPERATING LIMITS REPORT l-7 i
Amendment No. Z J(, 32f, Jg,1P[,150
_3
<i, N
TABLE OF CONTENTS
{n'
(
Sectice Pace r
i 5
CES:GN FEATURES 5-1 5.1 SITE 5-1 5.2 ITUAINMENT 5-2 5.2.i REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 HEW FUEL STORAGE 5-6' 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6
ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 TECisNICAL REVIEW AND CONTROL 6-4 6.5.2 INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.6.4 INDEPENDENT ONSI1E SAFETY REVIEW GROUP 6-8 6.6 REPORTABLE EVENT ACTION 6-10 6.7 5AFETY LIMIT VIOLATION 6-10 6.8 PROCEDURE 5 6-11 6.9 REPORTING RE0VIREMENTS 6-12 6.9.1 ROUTINE REPORT 5 6-12 6.9.2 DELETED 6-14 s
6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-18 6.9.5 CORE OPERATING LIMITS REPORT 6 19 l
6.1.0 RECORD RETENTION 6-B 6.11 RADIATION PROTECTION PROGRAM 6-21 6.12 HIGH RADIATION AREA 6-21 6.13 PROCESS CONTROL PROGRAM 6-21 6.14 0FF5ITE D05E CALCULATION MANUAL (ODCM) 6-22 6.15 DELETED 6-22 6.16 PO5T ACCIDENT SAMPLING PROGRAMS _
6-22 NUREG 0737 (II.B.3, II.F.1.2) 6.17 MAJOR CHANGE 5 TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6-23
-v-Arnendment No. )(, ff, JI,
, 1 6, 150 O
LIST OF.-TABLES u
TABLE-
. TITLE' W,.
PAGE-f' 1.2
-Frequency Notation-1-8 1
2.3-1 Reactor Protection System Trip. Setting Limits 2-9
-3.1.6.1 Pressure Isolation Check Valves Between the Primary 3-15a Coolant System and LPIS-
- 3. 5-l '
Instruments Operat.ing Conditions' 3-29
-3.5-1A DELETED 3.5-2 Accident Monitoring Instruments 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d
-3.21-1 Radioactive Liquid Effluent Monitoring instrumentation 3-97 3.21-2 Radioactive Gaseous Process and Effluent 3-101.
.Mo,nitoring. In strumentat ion 3.23-.1-Radiological Environmental Monitoring Program 3-122*
3.23-2' Reporting Levels for Radioactivity Concentration 3-126.
in Environmental Samples 4.1-1 Instrument Surveillance Requirements 4-3 4.1-2 Minimum Equipment Test Frequency 4-8' 4.1-3 Minimum Sampling Frequency 4-9 4.1-4 Post Accident Monitoring Instrumentation 4-10a 4.19-1 Minimum Number of Steam Generators to be 4-84 InsbectedDuring.'InserviceInspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21 l Radioactive Liquid Effluent Monitoring 4-88 In'strumentation Surveillance Requirements 4.21-2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Li;Jid Waste Sampling & Analysis Program 4-96 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-102 4.23-1 Maximum Values for the Lower Limits of Detection (LLD) 4-118 v1
- i Amendment No. 54. 7E.106.1Dfi.1Mr. M1.1Mf.16 150
7 LIST OF FIGURES p'
Ficure Title 2.1-1 TMI-l Core Protection Safety Limit 2.1-2
'TMI-1 Core Protection Safety Limits 2.1-3 TMI-1 Core Protection Safety Bases 2.3-1 TMI-l Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Axial Power Imbalance, TMI-1 3. ~1 - 1 Reactor Coolant System Heatup/Cooldown Limitations (Applicable to 10 EFPY) 3.1-2 Reactoi C:olant System Inservice Leak and Hydrostatic Test Limitations (Applicable to-10 EFPY)-
3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H O 2
3.5-2A thru DELETED 3.5-2J vii Amendment Nos. )#,,W,
, J% 6, 5tf P[, Jf.lefI, MMT,126,12(
134. P,' 150
b,.
w-,
@f LIST OF FIGURES-
.s
[m;W
/
Ficure Title 3.5-2K thru DELETED 3.5-2M-
- 3. 5'- 1 Incore. Instrumentation Specification Axial Imbalance Indication TMI-1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, Tiil-1 3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from' Cask Loading Pit 4.17-1 Snubber Functional-fest - Sample Plan 2 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mile Radius 5-3 Site Boundary for Gaseous Effluents viii Amendment Nos. M M M M M,1 9.150
I a
1
- E ')
I 7.,
1.19 PLRGE - PURGING h
FURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating concitions in such a manner that re:lacement air or gas is recuired to purify the confinement.
1.2G VENTING l
VENTING is the controlled process of discharging air as gas from a f
confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided. Vent used in system name coes not imply a VENTING process.
1.21 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR part 50.
1.22 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not incluce employees of the GPU System, GPU contractors or vendors.
Also excluded from this category are persons who enter the site to
. service equipment or to make deliveries.
1.23 SUBSTANTIVE CHANGES SUBSTANTIVE CHANGES are those which affect the activities associated with a document or the document's meaning or intent.
Examples of non-substantive changes are: (1) correcting spelling; (2) adding (but not deleting) sign-off spaces; (3) blocking in notes, cautions, etc.;
(4) changes in corporate and personnel titles which do not reassign responsibilities and which are not referenced in the Appendix A Technical Specifications; and (5) changes in nomenclature or editorial changes which clearly do not change function, meaning or intent.
1.24 CORE OPERATING LIMITS REPORT The CORE OPERATIhG LIMlTS REPORT is a TMI-1 specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5.
Plant operation within these operating limits is addressed in individual specifications.
1-7 Amendment No. JE, 97, J.Af,150
ne
%,,a Lihe'specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three
[G<
pumps, and.one pump in each loop, respectively.
LJ The curve-of Figure 2.3-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3.
The curves of Figure 2.1-3 represent the conditions at which the DNBR limit is predicted at the maximum possible thermal
_ power for the number of reactor coolant pumps in operation or the local cuality at the point of minimum DNBR is equal to 22 percent, (B&W-2)(4), or 26 percent (BWC)(2) whichever condition is more restrictive.
The maximum thermal power for three pump operation is 89.3 percent due to a power level trip produced by the flux-flow ratio (74.7 per cent flow x 1.08 = 80.6 percent power) plus the maximum calibration and instrumentation error.
The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.
Using a local quality limit of 22 percent (B&W-2), or 26 percent (BWC) at the noint of minimum DNBR as a basis for curves 2 and 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
l The DNBR as ralculated by the B&W-2 or BWC correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher m.id is a function of the pressure.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left.of the curve would result in a DNBR greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22 percent (B&W-2), or 26 percent (BWC) for the particular reactor coolant pump situation. Curve 1 is more restrictive than~any.Other reactor coolant pump situation because any' pressure / temperature point above and to the left of this curve will be above and to the left of the other curves.
REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lyr.chburg, Virginia, AprTFTMS'^
(3) FSAR, Section 3.2.3.1.1.3 (4) FSAR, Section 3.2.3.1.1.10 l
2-3 Amendment No. % % FT, SC.lE JK, X150
F.
3;5.2 CONTPOL R0D GROUP AND POWER DISTRIBUTION LIMITS h
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Aeolicability This specification applies to power distribution and cperation of control rods during power operation.
Objective To asure an acceptable core power distribution during power operation-, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.
Specification 3.5.2.1 The available shutdown margin shall not be less than one ptrcent SK/K with the highest worth control rod fully wi thdra wn.
3.5.2.2 Operation with inoperable rods:
Operation with more than one inoperable road as a.
defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted.
b.
If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification Paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated'immediately to verify the existence of one percent ok/k hot shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the worth of the inoperable rod or until the regulating banks are fully withdrawn, whichever occurs first.
Simultaneously a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.
If within one hour of determination of an inoperable c.
rod as defined in Specification 4.7.1, it is not determined that a one percent ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the HOT SHUTDOWN condition until this l
margin is established.
i d.
Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
i If a control rod in the regulating or safety rod e.
groups is declared inoperable per 4.7.1.2, power shall be reduced to 60% of the thermal power allowable for the reactor coolant pump combination.
3-73 Amendment No. g.lsn
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- f. If a control red in the regulating or axial power shaping-b, groups is declared inoperable pcr Specification 4.7.1.2.,
operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.
- g. If the inoperable red in Paragraph "e"'above is in groups 5, 6, 7, or 8, the other rods in the group may be trimed to the same position. Normal operation of 100 percent of the thermal power. allowable for the reactor coolant pump combination may then continue provided that the rod that was declared inoperable is maintained within allowable-group average position limits in 3.5.2.5.
3.5.2.3 The worth of single inserted control rods during criticality is limited by the restriction of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
3.5.2.4 Quadrant Tilt:
- a. Except for physics tests, the cuadrant tilt, as determined using the full incore system (FIS), shall not exceed the values in the CORE OPERATING LIMITS REPORT.
The FIS is OPERABLE for monitoring quadrant tilt provided the number of valid symmetric string individual SPND signals in any one quadrant is not less than the limit in-the CORE OPERATING LIMITS REPORT.
- b. When'the full intore system is not OPERABLE and except for physics tests quadrant tilt as determined using the power range channels for each quadrant (out of core detector system)(OCD), shall not exceed the values in CORE OPERATING LIMITS REPORT.
q
- c. When neither detector system above is OPERABLE and, except i
for physics tests, quadrant tilt as determined using the minimum incere system (MIS), shall not exceed the values in the CORE OPERATING LIMITS REPORT.
l
- d. Except for physics tests, if quadrant tilt exceeds the tilt j
limit, allowable power shall be reduced 2 percent for each i
1 percent tilt in excess of the tilt limit.
For less than f
four pump operation, thermal power shall be reduced 2 percent below the thermal power allowable for the reactor l
coolant pump combination for each 1 percent tilt in excess of the tilt limit.
3
- e. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for
{
physics tests, or the following adjustments in setpoints l
and limits shall be made:
3-34 Amendment No. JT, JC )( JC f>( 36 p( I150
- j '
. 1.
The protection system reactor power / imbalance envelope trip setpoints shall be reduced 2 percent in power for.-
p;y each 1' percent tilt in' excess of the tilt limit, or
/M when thermal power,s equal to or less than 50% full-i power with' four reactor coolant. pumps running, set the nuclear overpower trip setpoint equal to or less than 60% full power.
2.
The control rod group withdrawal limits in the CORE OPERATING LIMITS REPORT shall be reduced 2 percent in 1
i power for each 1 percent tilt in excess of the. tilt limit..
3.
The operational imbalance limits in the CORE OPERATING LIMITS REPORT shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.
i
- f. Except for physics or diagnostic testing, if quadrant tilt is in excess of the tilt limit defined in the CORE OPERATING LIMITS REPORT and using the applicable detector system defined in 3.5.2.4.a. b, and c above, the reactor will be placed in the HOT SHUTDOWN condition. Diagnostic testing during power operation with a quadrant tilt is
. permitted provided that the thermal power allowable is restricted as stated in 3.5.2.4.d above, g.-Quadrant. tilt.shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.
3-34a Amendment No. pT, g, g, MF, pf, Ff, JM,126; JWT,150
3,E'2I5 teatrcl Roc pcsitiens:
a.
Operating roc grcup overlap shall not exceec 25 g
percent +5 percent, between two secuential groups except for physics tests.
c.
cosition limits are.specified for regulating control rods.
Except for physics tests or exercising control rces, the regulating control rod insertion / withdrawal ~
limits are specified in the CORE OPERATING LIMITS rep 0RT.
If any of these control. rod position limits are exceeced, corrective measures shall.be taken intediately to achieve an acceptable control rod position.
Acceptable control rod positions shall be attained within four hours, c.
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 Axial Power Imbalance:
Except for physics tests the axial power imbalance, as a.
cetermined using the full incere system (FIS), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.
The FIS is operable for monitoring axial power imcalance proviced the number of valid self powered neutron detector (SPND) signals in any one cuadrant is not less than the limit in the CORE' OPERATING LIMITS REPORT.
b.
When the full incere detector system is not OPERABLE and except for physics tests axial power imbalance, as determined using the power range channels (out of core detector system)(OCD), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.
c.
When neither detector system above is OPERABLE and, except for physics tests axial power imbalance, as determined using the minimum incere system (MIS), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.
d.
Except for physics tests if axial power imbalance exceeds the envelope, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shall be taken to maintain operation within the envelope.
l 3-35 Amendfrent No. p, J7, N, W, 39, W, Z IM, y[,150 L
y,
e.
If ar acceptacle axial ;cwer imbalance'is not acndevec h{,,
witnin four reurs, reactor power shall te recutec until imbalance limits are tret.
f.
Axial power imbalance shall be monitored on a minimum frecuency of ence every two hcurs during power operation abeve 40 percent of rated power.
3.5.2.8 A power map shall be taken at intervals not to exceed 30 l
effective full power days using the intore instrumentation detection system to verify the power distribution is within the limits shown in Figure 3.5-2M.
Bases The axial power imbalance, quadrant power tilt, and control red position limits are based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2M).
These limits are ceveloped in a manner that ensures the initial condition LOCA maximum linear neat rate will not cause the maximum clad temperature to exceec the Final Acceptance Criteria (10 CFR 50 Appendix K).
Operation outside of any one limit alone does not necessarily constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.
Each limit represents the boundary of operation that will preserve the Final Acceptance Criteria even if all three limits are at their maximum allowable values simultaneously. The effects of the gray APSRs are included in the limit development. Additional conservatism included in the limit development is introduced by application of:
- a. Nuclear uncertainty factors
- b. Thermal calibration uncertainty
- c. Fuel densification effects
- d. Hot rod manufacturing tolerance factors
- e. Postulated fuel rod bow effects
- f. peaking limits based on initial condition for Loss of Coolant Flow transients.
The incore instrumentation system uncertainties used to develop the axial power imbalance and quadrant tilt limits accounted for various combinations of invalid SPND signals.
If the number of valid SPND signals falls below that used in the uncertainty analysis, then another system shall be used for monitoring axial power irabalance and/or cuadrant tilt.
3-35a I
.wendment No. J(, J9'.38, 35, 50",126,126,142',150
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Tee 2:-5 percent Overlap'between successive control rod groups is h'
e1 6 el since the worth of'a rod is lower at the upper and lower eart of the streke.
Control rods are arranged in groups or banks gj, deHned as follows:
3 Group Function 1
Safety 2
Safety 3
Safety 4
Safety
'5 Regulating 6
Regulating 7
Regulating 8'
APSR (axial power shaping rod bank)
Control rod groups are withdrawn in sequence beginning with group 1.
-Groups 5,6 and 7 are overlapped 25 percent. The normal position at power is for group 7 to be partially inserted.
The rod position limits are based on the most limiting of the following three criteria:
ECCS power peaking, shutdown margin, and potential ejected rod worth.
As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time,: assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups will. not contain single rod worths' greater than: 0.65% ok/k at rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. 'A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod position limits at hot zero power.
A single inserted control rod worth 1.0% ak/k at beginning of life, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than 0.65% 4k/k ejected rod worth at rated power.
I The plant computer will scan for tilt and imbalance and will satisfy the technical specification requirements.
If the computer is out of service, than manual calculation for tilt above 15 percent power and imbalance above 40 percent power must be performed at least every two hours until the computer is returned to service.
3-36 Amendment No. W, 39, Jr, 40', 50~, 126, Mf,150
l r
l i
- w
- en::ir" Of t"e nuclear overpower trip setpoint to 60% full power J
?
k
.te-teernal pc.er is epual to er less than 50% full power maintains i
i
t
- .t, c:re protection and an coerability margin at reduced power j
sir 41er :: that at full power.
1 Curie; the physics testing program, the high flux trip setpoints are ac-4ristratively set as follows to assure an additional safety g
i marg n is provided:
i Test Power Test Setpoint 0
<5%
15 50%
i 40 50%
1 50 60%
75 85%
>75 105.1%
REFERENCES (1) FSAR, Section 3.2.2.1.2 i
(2) FSAR, Sectier 14.2.2.2 i
l 9
3-36a Amendment No. 39,12E.14,150 Il
'I 3.~.a
- 'GI :NSTRUPE'~AT:GN k olicability i
Mciies to tne operability of the intore instrumentation system.
Dbjective i
To specify the functional and operational requirements of the incore instrumentation system for the Minimum Incore System (MIS).
l Specificatico Above 80 percent of operating power determined by the reactor I
coolant pump combination, reference Table 2.3.1, at least 23 individusi incere detectors shall be OPERABLE to check gross core l
power distribution and to assist in the periodic calibration of the out-of-core detectors in regard to the core imbalance trip limits.
The detectors shall be arranged as follows and may be a part of both basic arrangemer,ts.
3.5.4.1 Axial Power Imbalance l
Three detectors in each of three strings shall lie in a.
the same axial plane with one plane in each axial core *
- half, b.
The axial planes in each core half shall be symmetrical about the core mid-plane.
The detectors shall not have radial symmetry.
c.
3.5.4.2.
Quadrant Tilt l
a.
Two sets nf four detectors shall lie in each core half.
Each set of four shall lie in the same axial plane.
The two sets in the same core half may lie in the same axial place.
b.
Detectors in the same plane shall have quarter core radial symmetry.
Bases A system of 52 incere flux detector assemblies with seven detectors per assembly has been provided primarily for fuel management purposes.
The system includes data display and record functions and is also used for out-of-core nuclear instrumentation calibration and 4
for core power distribution verification.
I
- a. The out-of-core instrumentation calibration includes:
1.
Calibration of the split detectors at initial reactor startup, during the power escalation program, and j
periodically thereafter.
3-38 Amendment No.150
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6.9.4.2.5 The Radioactive Effluent Release Reports shall include k
tne instrumentation not returned to GPERABLE status within 30 days per TS 3.21.1.b and TS 3.22.2.b.
E 6.9.4.3 The following information shall be included in the D
Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year.
6.9.4.3.1 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of.each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual sunrnary may be either in the form of an hour-by-hour listing of wind i
speed, wind direction, atmosphere stability, and precipitation (if measured) on magnetic cape, or in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability.
6.9.4.3.2 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.
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6.9.4.3.3 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an i
1 assessment of the radiation doses from radioactive liquid 1
and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the site boundary (Figuros 5-3 and 5-4) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).
6.9.4.3.4 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium fuel cycle sources including doses from primary effluent pathways and direct radiation for the previous 12 consecutive months to show conformance with 40 CFR 190 " Environmental Radiation Protection Standards for Nuclear Power Operation". Acceptable methods for calculating the dose contributions from Liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.
6.9.5 CORE OPERATING LIMITS REPORT 6.9.5.1 The core operating limits addressed by the individual Technical Specifications shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle.
6-19 A nendaent No. X. 7,125, IF, Wl,150
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~re.erelytical retrods usec to cetermine the core h(;
c:e-ating limits accressec by the indivicual Technical 5:ecificatiers shall be these previously reviewec and N
accrovec ey the NRC for use at TM!-1, specifically:
U) EAW-1:122A Rev. 1, "Norral Operating Controls,"
May 195c.
(2)
BAW-10116-A, " Assembly Calculations and Fitted Nuclear Data," May 1977.
(3) BAW-10117P-A, " Babcock & Wilcox Version of PDQ User's Manual," January 1977.
(C EAW-10115A, " Core Calculational Techniques and Procedures," December 1979.
(5)- BAW-10124A, " FLAME 3 - A Three-Dimensional Nodal Code for Calculating Core Reactivity and Power Distributions," August 1976.
(C) BAW-10125A, " Verification of Three-Dimensional FLAME Coce," August 1976.
(7) BAW-10152A, " NODDLE - A Multi-Dimensional Two-Group Reactor Simulator," June 1985.
(E) BAW-10119, " Power Peaking Nuclear Reliability Facters," June 1977.
6.0.5.3 The core operating limits shall be determined so that all acclicable limits (e.g. fuel thermal-mechanical limits, core tnermal-hycraulic limits, ECCS limits, nuclear limits sucn as snutcown margin, and transient / accident analysis lir,its) of the safety analysis are met.
6.9.5.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk witt copies to the Regional Administrator and Resicent inspector.
C, :';
r,E 3C RE?EUION 6.10.I' ine fo11 ewing records shall be retained for at least five years:
a.
Recorcs of normal station operation including power l
levels anc periocs of operation at each power level.
b.
Records of principal maintenance activities, including inspection, repairs, substitution, or replacement of principal items of equipment related to nuclear safety.
c.
All REPORTABLE EVENTS.
c.
Records of periodic checks, tests and calibrations.
Records of reactor physics tests and other special e.
tests related to nuclear sa'ety.
f.
Changes to procedures required by Specification 9
6.B.1.
g.
Recorcs of solid radioactive shipments.
6 19a Arencrent Nc. Z, JT, W,, W, g. 1 9, 150 l
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Acclicability l
I Applies to the design features of the reactor core and reactor cociant system.
Objective To define the significant design features of the reactor core and reactor coolant system.
Specification 5.3.1 REACTOR CORE 5.3.1.1 The reactor core is composed of slightly enriched uranium dioxide pellets contained in fuel rods. A fuel assembly contains 208 fuel rods arranged in a 15 by 15 lattice.
The details of the fu 1 TMI-l FSAR Chapter 3.9 ) assembly design are described in il 5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches.
active fuel height is defined in TMI-1 FSAR Chapter 3.(The 3i 5.3.1.3 The core average and individual batch enrichments for he present cycle are described in TMI-1 FSAR Chapter 3.(2 5.3.1.4 The control rod assemblies (CRA) and axial power shaping rod assemblies (APSRA) are distrih ted in the reactor core as shown in TMI-1 FSAR Chapter 3.L 1 The CRA and APSRA design data are also described in the FSAR.
5.3.1.5 The TMI-l core may contain burnable poison rod assemblies (BPRA) as described in TMI-l FSAR Chapter 3.(4) 5.3.1.5 Reload fuel assemblies and rods shall conform to design and evaluation data described in the FSAR and sht I not exceed an enrichment of 4.3 weight percent of U a,
2 5.3.2 REACTOR COOLANT SYSTEM 5.3.2.1 The reactor coolant system shall be c'esigned and constructed in accordance with code requirements.(4) 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, sht.ll be designed for a pressure of 2,500 psig and a temperature of 650 F.
The pressurizer and pressurizer surge line shall be designed for a tempera-ture of 670 F.(5) 5-4 bendment No l$ )M.150
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5.3.2.3 Tre redctor coolant system volume shall be less than 12.200 cubic feet.
RErERENCES (1) FSAR, Section 3.2.1 (2) FSAR, Section 3.2.2 (3) FSAR, Section 3.2.3 (4) FSAR, Section 3.2.4 (5) FSAR, Section 4.1.3 (6) FSAR, Section 4.1.2 1
i u< 5 l
Amendment tio.150 i