ML033240446
| ML033240446 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/10/2003 |
| From: | Scarola J Progress Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP 03-126 | |
| Download: ML033240446 (34) | |
Text
tJOV 102003 ~~James Scarola Progress Energy OVI Vice President Harris Nuclear Plant Progress Energy Carolinas, Inc.
United States Nuclear Regulatory Commission SERIAL: HNP 03-126 ATTENTION: Document Control Desk IOCFR50.90 Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50400/LICENSE NO. NPF-63 SUPPLEMENTARY INFORMATION CONCERNING REQUEST FOR LICENSE AMENDMENT, TECHNICAL SPECIFICATION 5.6.3.d, TAC NUMBER MB7736 Ladies and Gentlemen:
On February 14, 2003, Progress Energy requested a license amendment for the Harris Nuclear Plant (HNP) to allow an increase in the decay heat load from fuel stored in Spent Fuel Pools C and D in Technical Specification 5.6.3.d. This purpose of this letter is to provide the NRC with information regarding Progress Energy's use of the ORIGEN2 code to calculate decay heat, respond to requests from the NRC received by e-mail on September 23, 2003, and to request additional changes to Technical Specification pages to accommodate an administrative update of the Technical Specification Index.
Attachment A provides the discussion concerning implementation of the ORIGEN2 code for calculating decay heat and the impact of this code on the subject license amendment request.
Secondly, Attachment B provides the response to questions asked by the reviewer concerning the subject license amendment request. Finally, Attachment C provides Technical Specification markup pages for an administrative update to the Index.
In accordance with 10 CFR 50.91(b)(1), HNP is providing the State of North Carolina with a copy of this letter.
Please contact John Caves, Licensing/Regulatory Programs Supervisor with any questions regarding this letter at (919) 362-3137.
Sincerely, (7
JS/rgh Attachments:
A: Change from Standard Review Plan 9.1.3 to ORIGEN-2 for Calculating Decay Heat B: Response to Reviewer's Questions C: Administrative Revision to the Technical Specification Index P.O.
Box 165 New Hill, NC 27562 loDl T > 919.352.2502 F > 919.362.2095
Jim Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge, and belief and the sources of his information are employees, contractors, and agents of Progress Energy Carolinas, Inc.
(alternately known as Carolina Power & Light Company).
Notary (Seal)
My commission Expires:
My ComIM Ex. 2-21-M. 3.
cc:
Mr. R. A. Musser, NRC Sr. Resident Inspector Ms. Beverly Hall, Section Chief N.C. DENR Mr. C. P. Patel, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator Attachment A Change from Standard Review Plan 9.1.3 to ORIGEN-2 for Calculating Decay Heat On February 14, 2003, Progress Energy requested a license amendment for the Harris Nuclear Plant (HNP) to allow an increase in the decay heat load from fuel stored in Spent Fuel Pools C and D in Technical Specification 5.6.3.d. The method used in this amendment request for calculating spent fuel decay heat was consistent with NRC Standard Review Plan 9.1.3 (SRP 9.1.3), as noted on page A1-7.
Since submittal of the February 14th license amendment request (LAR), HNP has implemented the ORIGEN2 code (version 2.1) for calculating decay heat using 10 CFR 50.59. This change in methodology was made to allow use of a single methodology for calculation of decay heat production for both the Progress Energy used fuel shipping activities and storage in spent fuel pools, as well as to gain flexibility by using the more precise ORIGEN2 code. The NRC previously approved the use of ORIGEN2 for calculating decay heat load in spent fuel pools in the Safety Evaluation Reports for license amendment No. 160 for the Virgil C. Summer Nuclear Station and license amendment No. 242 for the Duane Arnold Energy Center.
To confirm this change in methodology has no impact on this LAR, HNP performed ORIGEN2 calculations for the design heat load cases described in the LAR. The ORIGEN2 results demonstrated that the analytical values for heat loads presented in the LAR and in the responses to the reviewer's questions in Attachment B remain conservative with respect to the evaluation for the revised temperature limits included in the LAR. ORIGEN2 will be used in the future to account for the heat load of the actual inventories of the spent fuel pools.
HNP requests the following change be made to the LAR for clarity:
Replace Attachment 1, Page A1-7 text:
The SFPs A and B heat loads were calculated using a method that is consistent with NRC Standard Review Plan (SRP) 9.1.3.
With:
The methodology used by HNPfor calculation of SFP decay heat load is ORIGEN2. The SFP heat loads referenced in this LAR are analytical values that are conservative with respect to actual heat loads using the ORIGEN2 methodology.
-Al -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 1 In Attachment I of the LAR (Page Al-]), the licensee states that the projected end of plant life heat removal capacity required in spentfiel pools (SFPs) C and D is 15.63 MBTU7Ir. Explain how this projected end of life heat removal capacity was determined.
Response
The 15.63 MBTU/hr is a long range, bounding forecast of the heat load used in the evaluation of the spent fuel rack configuration. This forecast includes spent fuel from HNP, BNP and RNP. The LAR requests an increase of the existing 1.0 MBTU/hr limit to 7.0 MBTU/hr which, like the 1.0 MBTU/hr limit in the current technical specifications, is less than the long range bounding heat load forecast. The evaluation of the heat removal capacity of the Fuel Pool Cooling and Cleanup System (FPCCS) or the Component Cooling Water (CCW) system is not dependent on the 15.63 MBTU/hr.
HNP Technical Specifications 5.6.3.b and 5.6.3.c specify limits for the number of assemblies in SFP C and D; however the inventory of SFPs C and D will be controlled to limit the total heat load less than the proposed limit of 7.0 MBTU/hr.
-B1 -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 2 In Attachment I of the LAR (page Al -5), the licensee states that the impact of the higher SFP heat load on the performance of the CCW system was analyzed using bounding heat load values in the following calculations:
CCW supply temperature for each mode of CCW operation CCW perfonnance during a LOCA
- Reactor Coolant System (RCS) Cooldown time when on residual heat removal system (RHRS)
- Analysis of the ultimate heat sink (UHS) during a LOCA CCWflow balance for each mode of CCW operation The licensee further stated that with the exception of the RCS Cooldown time and the CCWflow balance, the current FSAR analyses include sufficient design margin to allow SFPs C and D heat load to be increased to 7.0 MBTU1hr. New FSAR analyses for RCS Cooldown time and CCWflow balance were prepared.
- a. Briefly describe the changes that create design margin which allow the increase to SFP C and D heat capacity, while ensuring other components remain adequately cooled.
Response for part a The plant changes that create the margin which allow the increase in the SFPs C and D heat load include:
a Raising the SFP bulk temperature limit to 150'F (from 140WF)
Decreasing the CCW flow from the SFPs A and B spent fuel heat exchangers and increasing the flow to the SFPs C and D spent fuel heat exchangers.
The changes do not alter the minimum CCW flow rates to other components on the CCW system. The higher spent fuel heat load causes CCW supply temperatures to increase by small amounts (<50F) for normal operation and refueling. These changes were evaluated and found to be acceptable.
- b. Point out the FSAR textfor the current FSAR analysis which includes sufficient design margin to allow SFPs C and D heat load to be increased (e.g., examples of design margin values), and for the new FSAR analyses addressing RCS cooldown time and CCWflow balance.
-B2-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Response to Part b:
The FSAR currently includes analyses and system description that bound performance required by the increase in the SFPs C and D heat load to 7.0 MBTU/hr. Specific examples include:
The composite spent fuel pool heat load on the CCW system concurrent with Post-LOCA recirculation is 25.31 MBTU/hr. Reference LAR Table A1-3 for SFPs A and B heat load of 18.31 MBTU/hr in addition to 7.0 MBTU/hr for SFPs C and D. The bounding composite heat load used in long term containment analysis is 39.0 MBTU/hr. The bounding composite SFP heat load is an input to the results presented in FSAR Figures 3.11.4-1, 3.11.4-2 and 3.11.6-1 for the long term Post-LOCA containment temperature and pressure.
The composite spent fuel pool heat load is part of the design heat load on the Ultimate Heat Sink. The composite spent fuel pool heat load is again 25.31 MBTU/hr. The bounding value used in the UHS analysis is 27.0 MBTU/hr as indicated in FSAR Table 9.2.1-12.
The CCW system has been evaluated for a maximum RCS cooldown supply temperature of 1250F as described in FSAR Section 9.2.2.1. As described in the response to part "a" above, the total CCW flow requirement remains unchanged. The CCW performance parameters are described in FSAR Table 9.2.2-1.
The RCS cooldown analysis inputs and results change as a result of the increase in the SFP heat load. The current inputs are contained in FSAR Table 5.4.7-1A. The current results are presented in Figures 5.4.7-3 and 5.4.7-4. The information in FSAR Section 5.4.7 will be updated after approval upon implementation of the increase in the allowed heat load in SFPs C and D.
-B3-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 3 Explain how the analyses of CCW supply temperature for each mode of CCW operation were used in calculating the SFP equilibrium temperature (Attachment I of the LAR Page A1-6). Provide a matrix showing CCW mode of operation versus CCW supply temperature versus SFP equilibrium temperature.
Response
Calculations were performed using the SFP heat exchanger design details, CCW supply temperature, CCW flow rate, and FPCCS flow rate and heat load. Features of the calculation include:
- Heat losses due to evaporation, conduction and convection were conservatively neglected.
- The CCW flow rate was less than nominal.
- Fouling factor of 0.0005 BTU/hr-ft2-F is used. The chemistry of both the CCW and the FPCCS are monitored and controlled.
SFP thermal inertia is neglected.
- The calculation solved for the lowest SFP inlet temperature that would result in the removal of the required heat load. The equilibrium temperature represents the bulk temperature of the water after it is heated by passing through the spent fuel storage racks.
A matrix of results is provided below in Table 1 and Table 2. The heat loads for this analysis are those presented in LAR Table A1-3.
Table 1 SFPs A and B Operating Condition CCW Supply Equilibrium SFP SFP Temperature Temperature (F)
Temperature (F)
Limit Criteria (fF)
Incore Shuffle 109.2 137.1 150 Full Core Offload See Note 1 150 150 Emergency Core Offload 108.5 134.1 150 Normal Operations 105 125.7 150 RCS Cooldown 125 148.0 150 LOCA Recirculation 125 146.6 160 Note 1 - The allowed CCW supply temperature is a function of the time after reactor shutdown. The curve is a function of time from reactor shutdown because the core decay heat decreases exponentially as a function of time. The function is presented as a table or curve and is incorporated into the administrative controls for refueling activities.
-B4-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Table 2 SFPs C and D Operating Condition CCW Supply Equilibrium SFP SFP Temperature Temperature F)
Temperature (F)
Limit (F)
Incore Shuffle 109.2 131.7 150 Full Core Offload 107.8 130.3 150 Emergency Core Offload 108.5 129.9 150 Normal Operations 105 123.1 150 RCS Cooldown 125 147.6 150 LOCA Recirculation 125 144.1 160
-B5-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 4 In Attachment I of the LAR (Pages Al-6 and Al -7), the licensee states the following:
Table Al -2 shows the previous analyzed heat loads and Table Al -3 shows the heat loads analyzed for this license amendment request... The analyzed heat load for SFPs A and B increased from 16.45 MBTU/7r to 18.31 MBTU/7r to allow for a refueling outage as short as 15 days and to provide additional heat storage capacity in the SFPs A and B.
The heat load increase in SFPs A and Bfor the emergency core offload case was due to a more conservative calculation for the decay heat for the discharged core used in that specific case.
For each of the operating conditions (incore shuffle, normal fidl core offload, etc.) shown in Table A 1-3, explain the assumed scenario which resulted in the heat loads (MBTU/hr) listed for SFPs A and B and SFPs C and D respectively. In particular, explain why SFP heat load values remained tie same or increased when compared to Table Al-2. With regard to the analyzed heat load for SFPs A and B, explain the rationale to allow for a refueling outage as short as 15 days and to provide for additional heat storage capacity in SFPs A an B. Describe the more conservative calculation of SFP decay heat for the discharge core used in the emergency core offload case.
Response
A comparison of the changes between Table A1-2 and A1-3 is provided below. As a basis for the comparison of individual cases, the following contributors to heat load are defined.
- 1. SFPs A and B "Base Heat Load" - SFPs A and B are maintained essentially full with empty spaces for prudent operating reserve. After each HNP refueling, the plant staff moves some PWR fuel to SFP C. This action restores prudent operating reserve capacity in SFPs A and B. The heat load of the fuel prior to the subsequent outage is the Base Heat Load. The Base Heat Load is calculated for the fuel previously discharged to SFPs A and B and projected HNP discharges in the future. This heat load contributor is used in each of the operating conditions.
- 2. Full Core Offload Heat Load - Space is normally reserved in SFPs A and B for full core offload during a refueling. The full core offload heat load is the heat load added by discharge of the entire core based on end of cycle decay heat. The cooling time varies for each of the Operating Conditions listed in LAR Table A 1-3.
- 3. Most Recent Discharge Batch Heat Load - During a refueling, the discharged fuel is normally placed in SFPs A and B. The Most Recent Discharge Batch Heat Load is the heat load of 65 discharged bundles. The cooling time for this contributor varies depending on the spent fuel heat load case.
-B6-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools The SFPs C and D heat load changes in each case from 1.0 MBTU/hr to 7.0 MBTU/hr.
The change allows additional storage in SFPs C and D.
Incore Shuffle - This heat load is the sum of the Base Heat Load and the Most Recent Discharge Batch Heat Load with a cooling time of 6 days. This value did not change between LAR Tables A1-2 (Existing Analysis) and LAR Table A1-3 (License Amendment Analysis).
Normal Full Core Offload - This heat load is the sum of the Base Heat Load and the Full Core Offload Heat Load with a cooling time of 6 days. This value remained unchanged between the LAR Table A1-2 and A1-3. As stated above in Note 1 to Table 1, a time-varying heat load is used to calculate the required CCW supply temperature to commence core offload. The SFPs A and B heat load during Normal Full Core Offload depends on the starting time and duration of the offload. SFPs A and B temperatures are administratively limited to a maximum of 1500F.
Emergency Core Offload - The heat load increases from 42.46 MBTU/hr (LAR Table A1-2) to 46.23 MBTU/hr (LAR Table A1-3). The Table A1-3 heat load is the sum of the Base Heat Load, the Most Recent Discharge Batch Heat Load at a cooling time of 36 days and the Full Core Offload Heat Load at end of cycle with a cooling time of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. The Table A1-2 heat load is the sum of the Base Heat Load, the Most Recent Discharge Batch Heat Load at 55 days, and the Full Core Offload Heat Load, after 30 effective full power days with a cooling time of 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />. These changes were made to eliminate an input of outage duration and to provide a conservative analytical upper bound on the decay heat of this case.
Normal Operation and RCS Cooldown - The heat load is conservatively selected to bound the sum of the Base Heat Load and the Most Recent Discharge Batch Heat Load for a 15 day outage. The FPCCS heat load on the CCW system increases as the outage duration decreases. The outage duration controls the heat load from the discharge batch. Use of a 15 day outage duration provides a conservatively high calculation of spent fuel pool heat loads used in conjunction with other design basis events.
-B7-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Question No. 5 In Attachment 1 of the LAR (Page Al-8), the licensee states the following:
7he makeup rates use the heat loads from Table Al -2 and use a makeup source temperature of 125 deg F. T7is assumed makeup source temperature is conservative because it bounds the Technical Specification 3. 7.5.b limit for the Ultimate Heat Sink or 94 deg F and the Technical Specification 3.4.5.d limit for the Refileling Water Storage Tank of 125 deg F. 77Te Emergency Service Water System, which takes waterfrom the Ultimate Heat Sink and the RWST are two possible sources of makeup to the SFP.
77ze total makeup requirements conservatively assume both FPCCS cooling systems are simultaneously impacted. 77e total makeup rates are within the makeup capabilities of systems available to makeup water to the SFP.
- a. Provide a matrix showing the makeup sources for the SFPs versus makeup source capability versus makeup source temperature. Identify the whether the makeup source is primary or backup.
- b.
In the above sentence, the makeup rates use the heat loads from Table Al -2..."
should Table Al-3 be reference instead? Also, in the sentence. "The total makeup requirements conservatively assume both FPCCS cooling subsystems are simultaneously impacted, clarify "simultaneously impacted".
Response
Part a:
Table 3 provides a matrix of makeup systems and their capacities.
The cooling subsystems of the FPCCS for HNP subsystems of the FPCCS for HNP consist of redundant safety-related trains for both the north and south pools. The systems are protected from externally generated missiles and are Safety Class 3 systems. The licensing basis of the system requires makeup capacity to offset evaporation from the pool surface at design temperatures.
Complete loss of SFP cooling is not a design basis event for HNP. As allowed by SRP III. 1.f, the HNP design includes a backup makeup system to add coolant to the spent fuel pools. As described in FSAR Section 9.1.3, the backup method uses a temporary connection between the Emergency Service Water (ESW) system and the FPCCS. This emergency backup system provides a flow rate greater than the evaporation rate from the pool and canal surfaces. The evaporation rate from the four SFPs and connecting canals is calculated to be less than 20 gpm at a water surface temperature of 160'F. The makeup rate from ESW is a minimum of 30 gpm.
-B8-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools For clarification, please make the following change to the LAR:
Replace Attachment 1, Page A1-8 last sentence text:
The total makeup rates listed are within te makeup capabilities of systems available to makeup water to the SFP.
With:
Th7e total makeup rate for the maximum boil off is within the capacity of either the Demineralized Water Storage Tank or the Refiueling Water Storage Tankflow path. Te backup Emergency Service Water method is capable of meeting the makeup rate for normal evaporation from the SFPs.
As presented in LAR Table Al-5 the total makeup rate requirement of the highest heat load case (Emergency Core Offload) is 101.8 gpm. Other makeup sources used in routine operation and the emergency backup source are described in Table 3.
Table 3 Matrix of SFP Makeup Sources Makeup Makeup Capacity Maximum Seismic Notes Source (gpm)
Temperature F) Category Demin Water Tank maximum capacity Storage Tank
>105 See Note 2 Non-500,000 gal.
(DWST) seismic No Tech Spec required minimum capacity.
Refueling Tech Spec minimum Water
> 105 125 Category capacity of 436,000 gal.
Storage Tank I
Portions of flow path are (RWST) non-safety related Reactor Tank maximum capacity Makeup 85,000 gal.
Water 53 See Note 2 Category No Tech Spec required Storage Tank I
minimum capacity.
(RMWST)
Portions of flow path are non-safety related.
Emergency Backup method.
Service 34 per train 94 Category Only one train can be System used at a time.
Note 2 - These tanks are outdoor tanks. FSAR Section 2.3.2.1.2 states that the highest recorded local ambient temperature is 107'F.
-B9-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Response to b:
The observation concerning the reference to LAR Table A1-2 is correct. The correct reference is LAR Table A1-3. Please make the following change to the LAR:
Replace Attachment 1, Page A1-8 second sentence text under SFP Makeup Rates:
T7he makeup rates use the heat loads from Table Al1-2 and use a makeup source temperature of 125F.
With:
Thle makeup rates use the heat loads from Table Al-3 and use a makeup source temperature of 125T With regard to the phrase "simultaneously impacted", the total required makeup rate in LAR Table A1-5 does not take credit for heat removal by any of the spent fuel pool cooling trains.
-1B10-
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 6 In Attachment 1 of the LAR (page Al-10), the licensee states the following:
Table Al -6(b) lists the analyzed time to heatup from the maximum pool bulk temperature for normal operations to 150 degrees F, 160 degrees F and 212 degrees F, respectively.
The pool heatup rates are based on the "Nonnal Operations" heat loads listed in Table Al -3. 7he corresponding data for the existing design is presented in Table AJ-6(a).
For Tables Al -6(a) and A 1-6(b), explain how the "Maximum Nonnal Operating temperatures"for tie SFPs were determined. Given that tie resulting pool heatup rates are based on the "Normal Operations" heat loads listed in Table Al -3, are the "Normal Operations" case results bounding for the other operating conditions (e.g. Incore shuffle, normalfidl core offload, etc.)? Explain.
Response
The response to Reviewer Question No. 3 describes the calculation of the SFP Equilibrium Temperatures. The respective Maximum Normal Operating Temperature values are 125.70F for SFPs A and B and 123.1F for SFPs C and D.
The purpose of LAR Table A1-6(b) is to illustrate the time available to restore FPCCS cooling following a LOCA. The values in Table A1-6(b) only apply to SFP cooling restoration following a LOCA. The values for the remaining cases are presented in Table 4 and Table 5.
Table 4 SFPs A and B Heatup Times Case Starting Heatup Time to Time to Time to Temperature F)
Rate 1500F 160F boiling Refer to Table 1 (0F/hr)
(Hours)
(Hours)
(Hours)
In-Core Shuffle 137.1 5.75 2.2 3.9 13.0 Normal Core Offload 150 10.56 0.0 0.9 5.8 Emergency Core Offload 134.1 11.98 1.3 2.1 6.5 Table 5 SFPs C and D Heatup Times Case Starting Heatup Time to Time to Time to Temperature (F)
Rate 150F 160WF boiling Refer to Table 2 (0F/hr)
(Hours)
(Hours)
(Hours)
In-Core Shuffle 131.7 2.54 7.2 11.1 31.6 Normal Core Offload 130.3 2.54 7.7 11.6 32.1 Emergency Core Offload 129.9 2.54 7.9 11.8 32.3
-1B11 -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 7 In Attachment I of the LAR (Page Al -11), the licensee states that the following:
Due to the heat loads in SFPs A and B, that pool complex is the limiting location. The values presented in Tables Al -6(a) and (b) contain several conservatisms in the inputs for the calculated heatup times. In particular:
7he SFP heat load is based on the beginning of core life 7he CCW supply temperatures are based on a SFP composite heat load which bounds the proposed composite heat load 77e performance of the FPCCS is conservatively modeled 7he water volumes assumed as part of the thermal inertia are conservatively low 77Te thermal mass of the fuel, fiel rack and SFP strncture is neglected Provide explanation for the above second, third andfourth conservatisms as follows: for the second, clarify what is meant by "composite heat load", for the third, how is FPCCS conservatively modeled? Andfor the third [fourth], how are water volumes conservatively low?
Response
In the second conservatism, the phrase "composite heat load" refers to total heat load from all four pools.
Concerning the third conservatism, the SFP starting temperatures are an input to LAR Table A1-6(b). The calculation of the performance of the FPCCS determines the starting temperature. Reviewer Question No. 3 describes the conservatism used in evaluating FPCCS performance.
Concerning the fourth conservatism, the water volumes are conservative because the calculation neglects the water volume of the Transfer Canals and the Main Transfer Canal that would normally be part of the volume subject to heat up. Including the volume of water in the canals would reduce the heat up rates. FSAR Figure 1.2.2-55 illustrates the relative size of these volumes. The gates between SFPs A and B and the Unit 1 Transfer Canal are inserted infrequently. The gates isolate the spent fuel pools from the transfer canals. The gate between SFP C and the north transfer canal is inserted infrequently. Pending the storage of fuel in SFP D, this pool is isolated.
-B12 -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 8 In Attachment l of the LAR (Page Al.-I), the licensee states the following:
The time available to perform the restoration of cooling to the SFPs after a LOCA is conservatively calculated and provides sufficient time for required operator actions to be implemented. The method used to restore forced cooling of the Spent Fuel Pool has not changed. Therefore, the increase in the SFPs C and D heat load results in acceptable time available for restoration for CCW to FPCCS for LOCA.
Describe the conservative calculation which demonstrates sufficient time to restore cooling to the SFPs after a LOCA. Briefly explain the method used to restore forced cooling to the SFPs (that has not changed).
Response
Table A 1-6(b) of the LAR presents the time available to restore SFP cooling. As described in the response to Reviewer Question No. 3 and Reviewer Question No. 7, there are conservatisms in both "Maximum Normal Operating Temperature" and the method used to calculate the heatup rates.
FSAR Section 9.1.3.3 describes the restoration of CCW to FPCCS following a LOCA.
-B13 -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 9 In terms of heat exchanger performance in determining SFP temperature, discuss any conservatism introduced to more closely approximate actual operating conditions (e.g.
heat exchange tube fouling which could impact heat transfer).
Response
The calculation uses a constant tube side (FPCCS) heat transfer coefficient. The heat transfer coefficient is based on fluid properties at 120'F. The tube side heat transfer coefficient increases approximately 10% with fluid properties at 150'F. The calculation neglected this benefit. Also, the calculation includes a fouling factor of 0.0005 BTU/hr-ft2 -F. The calculation applies the fouling factor to the inside and outside surface of the tubing. The chemistry of the fluids in CCW and FPCCS is monitored and controlled.
-B14 -
Attachment B Response to Reviewer's Questions License Amendment Request for change to Technical Specification 5.6.3.d Increase Heat Load of C and D Spent Fuel Pools Reviewer Ouestion No. 10 On page Al -9, SFP Design, of the submittal, it is stated that the SFP structure and liner were re-evaluatedfor a pool temperature of 1600! to account for the new LOCA acceptance criteria, and the results indicated that the liner did not exceed the allowable stress. However, the submittal did not state that the concrete and reinforcing steel in the pool structure did not exceed their allowable stresses. Licensee should make it clear that the concrete and reinforcing steel in the pool structure did not exceed their allowable stresses in its re-evaluation.
Response
The re-evaluation confirmed that the concrete and reinforcing steel in the pool structure are within their respective allowable stresses. For clarification, please make the following change to the LAR:
Replace Attachment 1, Page Al-9 last sentence text under SFP Design:
77e evaluation concluded that adequate design margin existed to allow for the higher liner temperature without exceeding allowable stresses.
With:
77Te evaluation concluded that adequate design margin exists for the SFP structure and liner to allow for the higher pool temperatures without exceeding allowable stresses.
-B15 -
Attachment C Administrative Revision to the Technical Specification Index
-C1 of 16-
INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE..........................
2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.................
2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIS - THREE LOOPS IN OPERATION.
. 2-2 WITH MEASURED RCS FLO 293,540 GPM x (1.0 + C)]
'bc~A>
'd ishd 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...........
2-1 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE.......................
B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE................
B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS..........
B 2-2 SHEARON HARRIS - UNIT 1 iii Amendment No. j
INDEX 3.0/4.0 BASES SECTION PAGE 3/4.0 APPLICABILITY.....
B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL B 3/4 1-1 3/4.1.2 BORATION SYSTEMS B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES............... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS.......
B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE......
B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FAC TOTO j (cR4c 2 AND NUCLEAR ENTHALPY RISE HOT CHANNETFC R..
B 3/4 2-2 FIGURE B 3/4.2-1 (DELETED)
B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATIO.....
B 3/4 2-6 3/4.2.5 DNB PARAMETERS.....
B 3/4 2-6 DQ1 3/4.3 INSTRUMENTATION
- 5. r 5vl tfA-c5t-I Ins.,+
3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......
B 3/4 3-1 3/4.3.3 3/4.3.4 MONITORING INSTRUMENTATION.
(DELETED)............
B 3/4 3-3 B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT 3/4.4.2 SAFETY VALVES..........
3/4.4.3 PRESSURIZER...........
3/4.4.4 RELIEF VALVES..........
3/4.4.5 STEAM GENERATORS 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7 CHEMISTRY............
3/4.4.8 SPECIFIC ACTIVITY........
3/4.4.9 PRESSURE/TEMPERATURE LIMITS.
CIRCULATION......
B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2
............ B 3/4 4-2{ j B 3/4 4-3 B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 Amendment No.(o SHEARON HARRIS - UNIT 1 xiii
INDEX BASES SECTION 3/4.7 PLANT SYSTEMS PAGE 3/4.7.1 3/4.7.2 3/4.7.3 3/4.7.4 3/4.7.5 3/4.7.6 3/4.7.7 3/4.7.8 3/4.7.9 3/4.7.10 3/4.7.11 3/4.7.12 3/4 7.13 TURBINE CYCLE.........................................
B STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION......... B COMPONENT COOLING WATER SYSTEM.......................... B EMERGENCY SERVICE WATER SYSTEM.......................... B ULTIMATE HEAT SINK...................................... B CONTROL ROOM EMERGENCY FILTRATION SYSTEM................ B REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM.....
B SNUBBERS................................................ B SEALED SOURCE CONTAMINATION............................. B (DELETED)............................................... B (DELETED)............................................... B (DELETED)............................................... B ESSENTIAL SERVICES CHILLED WATER SYSTEM................. B 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 7-1 7-2 7-3 7-3 7-3 7-3 A
7-3 7-4 7-5 7 6 7.6 f
7.6 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1. 3/4.8.2 AND 3/4.8.3 A.C. SOURCES, D.C. SOURCES. AND ONSITE POWER DISTRIBUTION...............................
3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.................
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.....................................
3/4.9.2 INSTRUMENTATION.........................................
3/4.9.3 (DELETED)...............................................
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.......................
3/4.9.5 (DELETED)...............................................
3/4.9.6 (DELETED)...............................................
3/4.9.7- (DELETED)...............................................
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...........
3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM................
3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND NEW AND SPENT FUEL POOLS........................................
3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM........
B 3/4 8-1 B 3/4 8-3 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 9.1 9-1 9-1 9-1 wq 9-2 9-2 9-2 9-2 B 3/4 9-3 w
B 3/4 9 G
le SHEARON HARRIS - UNIT 1 xv Amendment No. 02 I
INDEX BASES SECTION 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN................
3/4.10.2 GROUP HEIGHT, INSERTION. AND POWER DISTRIBUTION 3/4.10.3 PHYSICS TESTS.................
3/4.10.4 REACTOR COOLANT LOOPS.............
3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.....
PAGE
..... B 3/4 10-1 LIMITS. B 3/4 10-1
..... B 3/4 10-1
..... B 3/4 10-1
..... B 3/4 10-1 3/4.11 3/4.11 3/4.11 3/4.11 3/4.11 3/4.12 3/4.12 3/4.12.
3/4.12.
RADIOACTIVE EFFLUENTS
.1 LIQUID EFFLUENTS
.2 GASEOUS EFFLUENTS.
.3 (DELETED)........
.4 (DELETED)........
....... B 3/4 11-
....... B 3/4 111
....... B 3/4 11-
....... B 3/4 11-
.1
.2
.3
-RADIOLOGICAL (DELETED).
(DELETED).
(DELETED).
ENVIRONMENTAL MONITORING
......B 3/4 12-1 I.....
B 3/4 12-1 B 3/4 12-1 Amendment No.08 3v SHEARON HARRIS - UNIT 1 xvi
INDEX 5.0 DESIGN FEATURES SECTION 5.1 SITE 5.1.1 EXCLUSION AREA....................
5.1.2 LOW POPULATION ZONE..................
5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS......
FIGURE 5.1-1 EXCLUSION AREA...................
FIGURE 5.1-2 LOW POPULATION ZONE................
FIGURE 5.1-3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS............................
FIGURE 5.1-4 ROUTINE GASEOUS RADIOACTIVE EFFLUENT RELEASE POINTS PAGE
.... 5-1 5-1 5-1 5-2
.... 5-3
.... 5-4
.... 5-5 5.2 CONTAINMENT 5.2.1 CONFIGURATION.........................
5.2.2 DESIGN PRESSURE AND TEMPERATURE................
5-1 5-6 5.3 REACTOR CORE 5.3.1 5.3.2 FUEL ASSEMBLIES........................
CONTROL ROD ASSEMBLIES....................
5-6 5-6 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.
5.4.2 VOLUME.............
5-6 5-6 A A, r
5 -
k 5.5 METEOROLOGICAL TOWER LOCATION.
5.6 FUEL STORAGE 5.6.1 CRITICALITY...............
5.6.2 DRAINAGE................
5.6.3 CAPACITY................
FIGURE 5.6-1 BURNUP VERSUS ENRICHMENT FOR PWR FUEL 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT.....
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.
........... 5-7
........... 5-7
........... 5-7Kj
........... 5-7 o 5-8 Amendment No.
3 SHEARON HARRIS - UNIT 1 xvi i
INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION 6.1 RESPONSIBILITY PAGE 6-1 6.2 ORGANIZATION 6.2.1 ONSITE AND C 6.2.2 UNIT STAFF FIGURE 6.2-1 DELE1 FIGURE 6.2-2 DELE1 TABLE 6.2-1 MININ 6.2.3 DELETED..
6.2.4 SHIFT TECHNI 6-1 G IFFSITE 6-1 cal
............................ 6-1
-ED.......................... 6-3
-ED.......................... 6-4 1UM SHIFT CREW COMPOSITION 6-5
............................ 6-6 CAL ADVISOR.....................
6-6 6.3 DELETED.
6.4 TRAINING.
6-6 6-7 6.5 DELETED.
6-7 DdxV xviii Amendment No. C2 SHEARON HARRIS - UNIT 1
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION
....... 6-16 6.7 SAFETY LIMIT VIOLATION...........
6-16 6.8 PROCEDURES AND PROGRAMS..........
....... 6-16 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS..............
Startup Report Annual Reports..............
Annual Radiological Environmental Operating Annual Radioactive Effluent Release Report Monthly Operating Reports.........
Core Operating Limits Report 6.9.2 SPECIAL REPORTS..............
6.10 DELETED..................
............ 6-20
............ 6-20
............ 6-20 Report........ 6-21
............ 6-22
............ 6-23
...... 6-24 11
............ 6-24k
. 6-241 6.11 RADIATION PROTECTION PROGRAM 6.12 HIGH RADIATION AREA.....
6.13 PROCESS CONTROL PROGRAM (PCP)
................... 6-26
................... 6-26
................... 6-27 Amendment No. &
SHEARON HARRIS - UNIT 1 xix
Revised Technical Specification Pages
- C9 of 16 -
INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE
........................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.................
2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS - THREE LOOPS IN OPERATION.
2-2 WITH MEASURED RCS FLOW > [293,540 GPM x (1.0 + C)]0 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS........... 2-1 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..... 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE..
B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE................
B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS.......... B 2-2 SHEARON HARRIS - UNIT 1 iii Amendment No.
INDEX 3.0/4.0 BASES SECTION 3/4.0 APPLICABILITY..................
3/4.1 REACTIVITY CONTROL SYSTEMS PAGE
..... B 3/4 0-1 3/4.1.1 BORATION CONTROL................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.
B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES
.............. B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE.
B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.B 3/4 2-2a FIGURE B 3/4.2-1 (DELETED)..
B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATIO............... B 3/4 2-6 3/4.2.5 DNB PARAMETERS.
................... B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.3 3/4.3.4 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION................
MONITORING INSTRUMENTATION..........
(DELETED)...................
I
.... B 3/4 3-1
.... B 3/4 3-3
.... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 3/4.4.2 3/4.4.3 3/4.4.4 3/4.4.5 3/4.4.6 3/4.4.7 3/4.4.8 3/4.4.9 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION SAFETY VALVES................
PRESSURIZER.................
RELIEF VALVES................
STEAM GENERATORS..............
REACTOR COOLANT SYSTEM LEAKAGE.......
CHEMISTRY..................
SPECIFIC ACTIVITY..............
PRESSURE/TEMPERATURE LIMITS.........
..... B 3/4 4-1
..... B 3/4 4-1
..... B 3/4 4-2
..... B 3/4 4-2
..... B 3/4 4-2b
..... B 3/4 4-3
..... B 3/4 4-4
..... B 3/4 4-5
..... B 3/4 4-6 SHEARON HARRIS - UNIT 1 xiii Amendment No.
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE...........................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.........
B 3/4 7-2 3/4.7.3 COMPONENT COOLING WATER SYSTEM..........................
B 3/4 7-3 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM..........................
B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK......................................
B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM................
B 3/4 7-3 3/4.7.7 REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM.....
B 3/4 7-3a 3/4.7.8 SNUBBERS.............................................
B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION.............................
B 3/4 7-5 3/4.7.10 (DELETED).B 3/4 7-5 3/4.7.11 (DELETED).B 3/4 7-5 3/4.7.12 (DELETED).B 3/4 7-5 3/4 7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM.B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, AND 3/4.8.3 A.C. SOURCES, D.C. SOURCES. AND ONSITE POWER DISTRIBUTION................................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.....................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION.........................................
B 3/4 9-1 3/4.9.3 (DELETED).............................................
B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.......................
B 3/4 9-1 3/4.9.5 (DELETED).............................................
B 3/4 9-2 3/4.9.6 (DELETED).............................................
B 3/4 9-2 3/4.9.7 (DELETED).............................................
B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...........
B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM................
B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND NEW AND SPENT FUEL POOLS.B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM.B 3/4 9-4 SHEARON HARRIS - UNIT 1 xv Amendment No.
INDEX BASES SECTION 3/4.10 PAGE SPECIAL TEST EXCEPTIONS 3/4.10.
3/4.10.
3/4.10 3/4.10 3/4. 10.
3/4.11 3/4.11 3/4.11 3/4.11.
3/4.11.
.1
.2
.3
.4
.5 SHUTDOWN MARGIN....................
GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS PHYSICS TESTS.....................
REACTOR COOLANT LOOPS.................
POSITION INDICATION SYSTEM - SHUTDOWN.........
B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 RADIOACTIVE EFFLUENTS
.1 LIQUID EFFLUENTS
.2 GASEOUS EFFLUENTS.
.3 (DELETED).....
.4 (DELETED).....
.................. B 3/4 11-1
. B 3/4 11-1 I
.................. B 3/4 11-2
.................. B 3/4 11-2 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 3/4.12.2 3/4.12.3 (DELETED)
(DELETED)
(DELETED)
................. B 3/4 12-1
................. B 3/4 12-1
................. B 3/4 12-1 SHEARON HARRIS - UNIT 1 xvi Amendment No.
INDEX 5.0 DESIGN FEATURES SECTION 5.1 SITE 5.1.1 EXCLUSION AREA........................
5.1.2 LOW POPULATION ZONE......................
5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS FIGURE 5.1-1 EXCLUSION AREA FIGURE 5.1-2 LOW POPULATION ZONE.
FIGURE 5.1-3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.....................
FIGURE 5.1-4 ROUTINE GASEOUS RADIOACTIVE EFFLUENT RELEASE POINTS PAGE 5-1 5-1 5-1 5-2 5-3 5-4 5-5 5.2 CONTAINMENT 5.2.1 CONFIGURATION...........................
5.2.2 DESIGN PRESSURE AND TEMPERATURE 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES..........................
5.3.2 CONTROL ROD ASSEMBLIES......................
5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE..................
5.4.2 VOLUME..............................
5.5 METEOROLOGICAL TOWER LOCATION...................
5-1 5-6 5-6 5-6 5-6 5-6 5-6a I 5.6 FUEL STORAGE 5.6.1 CRITICALITY...................
5.6.2 DRAINAGE....................
5.6.3 CAPACITY....................
FIGURE 5.6-1 BURNUP VERSUS ENRICHMENT FOR PWR FUEL 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT.......
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.
......... 5-7
......... 5-7
......... 5-7
......... 5-7b
......... 5-7aI
......... 5-8 SHEARON HARRIS - UNIT 1 xvi i Amendment No.
INDEX 6.0 ADMINISTRATIVE CONTROLS SECTION 6.1 RESPONSIBILITY.............
PAGE
.............. 6-1 6.2 ORGANIZATION..............
6.2.1 ONSITE AND OFFSITE ORGANIZATION.
6.2.2 UNIT STAFF..............
FIGURE 6.2-1 DELETED............
FIGURE 6.2-2 DELETED............
TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION 6.2.3 DELETED................
6.2.4 SHIFT TECHNICAL ADVISOR........
6.3 DELETED................
6.4 TRAINING................
.............. 6-1
.............. 6-1
.............. 6-la
.............. 6-3
.............. 6-4
.............. 6-5
.............. 6-6
.............. 6-6
.............. 6-6
.............. 6-7 6.5 DELETED.........
SHEARON HARRIS - UNIT 1 6-7 xviii Amendment No.
INDEX ADMINISTRATIVE CONTROLS SECTION 6.6 REPORTABLE EVENT ACTION..........
6.7 SAFETY LIMIT VIOLATION...........
6.8 PROCEDURES AND PROGRAMS..........
6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS..............
Startup Report..............
Annual Reports Annual Radiological Environmental Operating Annual Radioactive Effluent Release Report Monthly Operating Reports.........
Core Operating Limits Report.......
6.9.2 SPECIAL REPORTS..............
6.10 DELETED..................
6.11 RADIATION PROTECTION PROGRAM.......
6.12 HIGH RADIATION AREA............
6.13 PROCESS CONTROL PROGRAM (PCP).......
PAGE 6-16 6-16 6-16 Report 6-20 6-20 6-20 6-21 6-22 6-23 6-24 6-24c 6-24c 6-26 6-26 6-27 SHEARON HARRIS - UNIT 1 xix Amendment No.