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Application for Amend of DPR-44 & DPR-56 Changing Tech Specs Re Spent Fuel Discharge Rates & Proposing Mods to Control Rod Drive Sys.Certificate of Svc Encl
	ML19291B167
Person / Time
Site:	Peach Bottom
Issue date:	08/23/1979
From:	Daltroff S; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML19291B168	List: ML19291B167; ML19291B170;
References
	NUDOCS 7908290571
	Download: ML19291B167 (21)
	v • d • e

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMIIISSION In the Matter of :

Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY : 50-278 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 & DPR-56 Edward G. Bauer, Jr.

Eugene J. Bradley 2301 Market Street Philadelphia, Pennsylvania 19101 Attorneys for Philadelphia Electric Company 7 9082 90 5~/l . '2 l l'"

, BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of :

Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY : 50-278 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 & DPR-56 Philadelphia Electric Company, Licensee under Facility Operating Licenses DPR-44 and DPR-56 for Peach Bottom Atomic Power Station Unit Nos. 2 and 3 respectively, hereby requests that the Technical Specifications incorporated in Appendix A of the Operating Licenses be amended by revising certain sections as described below.

2012 16"

. 4 Change I - Spent Fuel Discharge Rates The Licensee proposes revisions to Technical Specification, section 3.10.E, that change the limiting conditions for the discharge of spent fuel from the reactor vessel to the Spent Fuel Pool. The revised limits would permit a spent fuel discharge rate based on the elapsed time following the reactor shutdown, the quantity of fuel to be discharged, river water temperature, and the discharge cycle (number). A description of the safety analysis is included in section 3.10.E BASES on page 230a. The safety analysis, using data and assumptions previously submitted on January 19, 1978, in support of the installation of high density racks in the Peach Bottom spent fuel pools, was performed to develop the naximum permissible discharge rates shown in the proposed figures 3.10.E.1 through 9.

The changes are indicated by the vertical bar in the m4rgin of the attached revised pages 11, 228, and by adding pages 228a through 228j and 230a.

Change II - Control Rod Drive System In a letter dated October 21, 1977, concerning Peach Bottom Atomic Power Station Unit 3 Control Rod Drive (CRD) system o p e r a '. i on , the Division of Operating Reactors, inter alia, requested that the Philadelphia Electric Company submit proposed plans for permanent modification of the Unit 3 CRD system. The modification would eliminate the CRD return line to the reactor vessel. On January 5, 1978, Philadelphia Electric Company .

170 -

2012 -

1 submitted proposed plans to cut and cap the CRD return line which included removal of the two check valves. At this time the

. Philadelphia Electric Company additionally proposes to provide an alternate CRD return path by re-routing the return to the Reactor Water Cleanup System return line between the inboard and outboard isolation valves. The new CRD return line will be provided with a locked closed manual outboard isolation valve as parmitted by 10 CFR 50, Appendix A, Criterion 55, paragraph 2. Inboard isolation is provided by the existing isolation check valve on the feedwater supply line. Accordingly, the Licensee requests that Table 3.7.1 of the Technical Specifications be revised as indicated by the vertical bar in the margin of attached page 180 to include these valves as primary containment isolation valves.

Due to recommendations from the General Electric Company that the stagnant CRD return line be removed as soon as practicable, it is planned to cut and cap the CRD return line on Unit 3 during the Fall 1979 refueling outage. Due to long design and material lead times, the rerouting of the CRD return line cannot be properly implemented during this Unit 3 refueling outage. The Licensee concurs with the Cencral Electric Company's engineering position, as stated in the March 19, 1979, letter from General Electric Company to ti.e NRC concerning Control Rod Drive Return Line Removal, that the capability of the CRD system to provide water to the reactor vessel is not significantly affected by removal of the CRD return line. However, the Licensee intends to expedite the rerouting of the CRD return line as soon as practicable to be implemented no later than the end of 2012 n71

the next scheduled Unit 3 refueling outage. The Unit 2 entire modification is planned for Spring 1980 during the refueling

. outage.

Since the proposed changes to the Technical Specification do not involve a significant hazards consideration, pursuant to 10 CFR 170.22, Philadelphia Electric Company, for fee purposes, proposes that the Application for Amendment for Unit No. 2 be considered a Class III Amendment and that the Application for Amendment for Unit 3 be considered a Class I Amendment.

The Plant Operation Review Committee and the Operation and Safety Review Committee have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question, or a significant hazards consideration and will not endanger the health and safety of the public.

Respectfully submitted, P HIL AD ELPilI A ELECTRIC COMPANY

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COMMONWEALTil 0F PENNSYLVANIA :

ss.

COUNTY OF P i!IL AD EL Pili A :

S. L. Daltroff, being first duly sworn, deposes and says:

That he is Vice President of Philadelphia Electric Company, the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating Licenses and knows the contents the re of ; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

/

\Y, - f i; I /4 Subscribed and sworn to R2 before me this M 3 day of *

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t y Public (L

ELI BETH H. BOYED Notary P lac. Phila., Phila. ,0.

M ' Co slon Expires Jan.3q 193 ,,

2012 1_/5

CERTIFICATE OF SERVICE I certify that service of the foregoing Application was made upon the Board of Supervisors, Peach Bottom Township, York County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Albert R. Steele, Chairman of the Board of Supervisors, R. D. No. 1, Delta, Pennsylvania 17314; upon the Board of Supervisors, Fulton Towaship, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to George K. Brinton, Chairman of the Board of Supervisors, Peach Bottom, Pennsylvania 17563; and upon the Board of Supervisors, Drumore Township, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Wilmer P. Bolton, Chairman of the Board of Supervisors, R. D. No. 1, Holtwood, Pennsylvania 17532; all this 27th day of August, 1979.

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f ugene /J . Bradley f Attorney for Philadelphia Electric Company 2012 171

yaf.E OF COMTFMTp (Cont' d.)

mm

D'T Pace No.

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S U P.V EI L L AN CF.

LIMITIMC CONDITIONS FOR OPEPATION !ECUIPIMENTS 3 *, 6 PRIMARY SYSTEM DCUMDARY 4.6 143 A. The rmal and Pressuri::ation Limitations '

A 143 B. Ccolant Chemis try B 115 C. Ccolant Leak age C 146 D. Safe ty and Relief Valves D 147 E. Je t Pumps- E 148 T. Jet Pump Flcv Mismatch P 148 G. Structural Integrity G 149 3.7 COMTAINMEMT SYSTEMS 4.7 165 A. P rimary Containment A 165 L. Standby Gas Treatment Sys tem D L75 C. Secondary Containm2n t C . 176 D. P rimary Con tain:r.c n t Isolation Valves D 177 3.8 RADIOACTIVE MATERIALS 4.8 203 A. General A 203 B. Liquid Ef fluents B 204

. C. Airborne Effluents C 206 D Mechanical Vacuum Pump D 209 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 217 A. Auxiliary Electrical Equipment A 217 D. Operatica with Inope rable Eiuipment 3 219 C. Emergency Service Water Sys tem C 221 3.10 COr@ 4.10 225 A. Re f ue ling .aterlocks A 225 D. Core Mar.itoring B 227 C. Spent Tucl Pool Water Lovel C 228

0. Ileavy Loads Over Spent Fuel b 228 E. Spent Fuel I)ischarge Rat

[ 228 l 3.11 ADDITICNAL SAFETY RELATED PLANT CAPABILITIEL 4.11 231 A. Main Control P.com Ventilation A 233 B. ' Alternate Heat Sink Facility a 234 C. Emergency Shutdown Ccatrol Panel C 234 3.12 RIVER LCVEL 4.12 237 A. High River Water Level A 237 B. Lou River Water Level 3 237 C. Lcyc1 Ins trumentation C 238 3.13 MISCELLANEOUS RADI0 ACTIVE MATERIALS SOURCES 4.13 240a

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PBAPS LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 3.10.B (Cont' d) 4.10.B (Cont' d. )

1. The SRM shall be inserted to the normal operating level. (UE .: of special moveable, dunking type detectors during initial fuel loading and major ccre altera-tions in place of normal detec-tors is permissible as long as the detector is connected to the -

normal (SRM circuit) .

2. The SRM shall have a minimum of 3 cps with all rods fully in-serted in the core.

in the core.

C. Spent Fuel Pool Water Level C. Spent Fuel Pool Water Level Whenever irradiated fuel is Whenever irradiate 1 fuel is stored in the spent fuel pool, stored in the spent fuel water level shall be pool, the water level maintained at or above 8 1/2' shall be recorded daily.

above the top of the fuel.

D. Heavy Loads Over Spent Fuel Loads in excess of 1000 lbs (excluding the rigging and transport vehicle) sha?.1 be prohibited from travel over fuel assemblies in the spent fuel storage pool.

E. Spent Fuel Discharge Rates E. Spent Fuel Discharge Rates In order to maintain SFP Whenever spent fuel is being temperature less than or discharged, the following items equal to 1500F, the rate of shall be recorded: a) the -

discharge of spent fuel discharge (cycle) number ,

assemblies from the reactor b) the number of fuel assemblies to the spent fuel pool shall -

to be discharged, c) the time not exceed the maximum from reactor shutdown to first permissible discharge rate fuel assembly discharged, in ,

obtained from Figures hours, d) the temperature of 3.10.E.1 through 9, whenever the river water shall be the spent fuel pool cooling determined at least at system is in use. No limit- the beginning of the ataions are required whenever discharge and once per day the Residual Heat Removal thereafter (if it is System is arranged for spent desired to correct the fuel pool cooling. maximum permissible discharge rate for the effect of the temperature of the cooling (river) water for the SFP heat exchanger) .

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Piqiaren 3.10.M.1 through 9 INTRt)CTIOt_1g 1'.) Using the discharge (cycle) number, select the corresponding figure from Figures 3.10.E.1 through 9.
2) Select the one curve (of the four in the figure selected in item 1 above) which corresponds to the discharge size next highest to (or coual to) the total number of assemblies to be moved to the SFP during the outage.
~~3) Using the time from reactor shutdown to first fuel assembly discharged (ordinate) which is achieved during the outage, read the coordinate maximum permissible discharge rate~~
~~from the curve selected in item 2) above. This maximum permissible discharge rate shall be used as a limit during the outage, in order to maintain SFP temperature less than or equal to 1500F.~~
4) If it is desired to correct the maximum permissible discharge rate for cooling (river) water temperatures, determine the river water temperature at time of first discharge and every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. Using the appropriate correction factor, multiply the maximum permissible discharge rate obtained in item (3) above by the correction factor for that 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. The corrected maximum permissible discharge rate shall be used as a limit during the corresponding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months segment of the outage. Correction factors are as follows:
~~a) 710F - 800F cooling water: 1.5 b) 610F - 700F: 2.5 c) 600F or lower: 4.0~~
5) The maximum permissible rate operates cumulatively; that is, if the permissible rate for a particular discharge is 100 ausenblies per day and only 50 assemblies are discharged the first day, then 150 may be discharged the second day.
This is the maximum rate at which fuel assemblics in the discharge indicated above may be moved to the SFP in order to limit SFP temperatures to 1500F or less, for the four maximum discharge sizes shown in each figure, as a function of the time from when the reactor is placed in shutdown mode until the first fuel assembly in discharged, assuming that a) two SFP cooling trains are in operation and b) cooling (river) water temperature to the SFP heat exchangers is 900F.
i n <'
+
201e2
-228j-
PBAPS 3.10 nASES E. Maximum Permincible Rate for Spent Fuel Discharge
. In order to limit SFP peak bulk temperatures during a spent fuel discharge to 1500F or less, it is necessary to limit the rate of input of heat from spent fuel assemblies. The heat input is a strong function of a) the number of spent fuel assemblies in a particular discharge, b) the time after reactor shutdown mode at which fuel assemblies are discharged, and c) the rate at which the discharged fuel assemblies are moved to the spent fuel pool. The heat input is also a weak function of the exposure of the assemblies being discharged, and of number, exposure, and age of fuel assemblies in prior discharges. The parametric analysis whose results are shown in Figures 3.10.E.1-9 accounts for both the weak and strong factors, and in addition accounts for variations in the strong factors. By limiting the daily discharge rate to a conservative maximum as a function of the number of assemblies in the discharge and the time from shutdown mode to first assembly discharged, adequate control of the heat input to the spent fuel pool is maintained based on easily obtained operating data. Also, since the temperature of the cooling water to the SFP cooling system heat exchangers significantly affects the ability of the system to remove heat, correction factors have been included for temperature of the river water.
The fuel decay energy release rates were evaluated in accordance with the NRC Standard Review Plan, Section 9. 2.5, " Ultimate Heat Sink." Fuel assemblies in the discharges were assumed to have an average exposure of 40,000 MWD /MTU. The number of spent fuel assemblies assumed to be in the SFP prior to each discharge is noted in the corresponding figure. Two (of three) SFP cooling pumps / heat exchangera were assumed to be in operation, with 400,000 lbs/hr of cooling water to the shell side of each heat exchanger at the calculated bulk pool temperature.
The data resulting from this analysis is applicable when the spent fuel pool cooling system is used to cool the SFP during discharges of the maximum sizes shown in each figure. If the Residual Heat Removal (RHR) system is used for spent fuel pool cooling (such as during a full-core discharge) , no limitations on discharge rates are necessary in order to maintain SFP temperatures at or below 1500F.
' 7 2012 230a
TAJLF 3.7.1 (cont ' d. )
PPIMAPY coy *A f *rwr*r* IsctATro* V A t'.* t; Numier of Power waxirun 4c* ion on Opera *ed Valves opera t ir.q ::arnal Initia*ina Group Valve I,tentification Inboard . Outtoard Time ( sec. ) position ."isnal NA f eedwater check valves 2* 2* NA O P roce9s NA control Pod Hydraulic Feturn valve l' NT C Lockei closed NA control rod hydraulic return to 1* NA O ProceSM check valves 2D 0xygen Analyzer System 14 NA O GF NA standby liquid control systen 1* 1* NA C Process
, check valves
~
eo 2B FHRS shutdown cooling suction 1 1 32 C Sc cm isolation valves s
2B PHPS shutdown cooling 1 24 C FC injection isolation valves 2B RHRS Peactor Vessel Head Spray 1 1 30 c Sc isolation valves 2C Feedwater Flush Valves 2 50 e sc 4 MPCIS steam line isolation valves 1 1 20 0 GC 5 EcIts steam line isolation valves 1 1 15 o oc 2A Peactor' dater cleanup system 1 1 30 o Ge isolatiot valves 2A Peactor m er cleanup system 1 20 0 GC return isolation valve
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Talves not power crarmted *
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