ML20116H651
| ML20116H651 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/05/1996 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20116H620 | List: |
| References | |
| NUDOCS 9608120167 | |
| Download: ML20116H651 (5) | |
Text
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l ATTACHMENT 2 LIMERICK GENERATING STATION llNIT 2 DOCKET NO. 50-333 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 96-18-2 LIST OF AFFECTED PAGES UNIT 2 2-1 B 2-1 9608120167 960805 PDR ADOCK 05000353 P
4
\\
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of -RATED THERMAL POWER with the actor vessel steam dome pressure less than 785 psig or core flow less than 10% ci rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 255 of RATED THERMAL POWER and the reactor vessel.
steam dome pressure less than 785 psig er cors flow less than 10% of rated flow, be in at'least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
THERMAL POWER, Hioh Pressure and Hich Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not less than + 0P or two recirculation loop operation and shall not be less than for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
IO I,1&
With MCPR less than for two recirculation loop operation or less than l
for single recirculation loop operation and the reactor vessel steam dome pressore greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measure'd in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam done, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply w.ith the requirements of Specification 6.7.1.
LIMERICK - UNIT 2 2-1 Amendment No.14 M4f 0 8195 W
w r
r
2.1 SAFETY LIMITS I
BASE $
2.0 INTRODUCTION
l.10 1%
- j. IL The fuel cladding, reacto j pressure vesse and primary system piping e the principle barriers to the release of radioac,1ve materials to the environs Safety Limits are established' to protect the integrity of these barriers dyring normal plant operations and a'nticipated trans ients. The fuel cladding intd rity Safety Limit is set such tha nofueldamagejiscalculatedtooccurifthe imit is not violated.
Because fu 1 damage is not pirectly observable, a step-ba k each is used to establis ;a Safety Limit such that the MCPR is not less than
(.10 for two recirculation lobo opgration an#@Yfor single recirculation oop operation.
MCPR greater than%977or two recirculation loop operation and E!B_
for single recirculation loop operation represents a conservative margin relative to the conditions required to kaintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or eracking. Although some corrosion or use rel?ted cracking may occur during the life of the cladding, fission product migrat~on from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER. Low Pressure or low Flow The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or cere flows less than 10% of rated flow.
Therefore, the fuel cladding integrity Safety Limit is established by other This is done by establishing a limiting condition on core THERMAL POWER means.
with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the c' ore pressure drop at low power and flows will alyays be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 psi driving head will be 1
greater than 28 x 10 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this j
flow is approximately 3.35 MWt. -With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
j LIMERICK -' UNIT 2 8 2-1 Amendment No. 14 MW 0 6 M
1 ATTACHMENT 3 LIMERICK GENERATING STATION UNIT 2 DOCKET NO. 50-353 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST l
NO. 96-18-2 Letter, R. M. Butrovich (GE) to H. J. Diamond (PECO Energy),
" Limerick Unit 2 Safety Limit MCPR Revision? dated July 7,1996
GENuclect Energy RichardM Buttonch
-i 1
- . v-er.
,,_.1,*-
=:5.,.
- a.
July 7,1996 5'
i RMB:96-145 Mr. R J. Diamond, Director Fuel & Services Division PECO NUCLEAR 965 Chesterbrook Boulevard Wayne, PA 19087-5691
SUBJECT:
Limerick Unit 2 Safety Limit MCPR Revision
REFERENCE:
- 1. Letter, R. M. Butrovich to R J. Diamond, " Generic gel 1 Safety Limit MCPR Calculation", April 2,1996
- 2. Letter, R. M. Butrovich to R J. Diamond, " Limerick Unit 2 Safety Limit MCPR", May, 21,1996.
3
- 3. Letter, R. M. Butrovich to R J. Diamond, " Limerick 2 Reload 3 (Cycle 4)
SRLR f.evision 1", June 14,1996.
Dear Hugh:
Reference I advised PECO Nuclear of discoveries related to the methodology used by GE to calculate SLMCPR that indicated the generic SLMCPR may not always yield the most conservative result. GE performed a plant unique evaluation for Limerick Unit 2 and reported the resuhs to PECO Nuclear in reference 2. 'Ibe SLMCPR for Limerick Unit 2 Cycle 4 went from 1.07 (Gell Generic) to 1.10 (Cycle specific). The single loop operation SLMCPR is 1.11. Reference 3 provided a revised Supplemental Reload Licensing Report for Limerick Unit 2 Cycle 4. PECO Nuclear should submit a technical speci6 cation revision indicating the revised cycle speciSc SLMCPR.
Ifyou have any questions, please give me a call.
Very t
- yours,
,/
HAN M. Batrovich t
h
--