ML19317G499
| ML19317G499 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/30/1971 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19317G500 | List: |
| References | |
| NUDOCS 8003180782 | |
| Download: ML19317G499 (19) | |
Text
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OPERATING LICENSE REVIEW PLAN FOR CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO.
50-302
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INTRODUCTION Florida Power Corporation (FPC) has submitted as Amendment No.11 to its application, the Final Safety Analysis Report
'in support of an application for a ' license to operate the "rystal_ River Unit 3 Nuclear Generating Plant (Docket No.
50-302). The Crystal River Nuclear Power Plantemploys for the reactor unit a pressurized water reactor nuclear steam supply system furnished by Babcock and Wilcox Company (36W).
De applicant requests a license to operate the plant at a core power level of 2452 MWt which is the same power icvel as specified at the construction pomit phase. All core performance data is based on this power level.
An ulti:aate core pcwer level of 2544 MWt is anticipated and the assessment of fissien product releases and radiation exposures associated with the design basis accident has assumed the ultimate power Icvel.
Re.4733 acre site is located in Citrus County, Florids, approxitaately 75 miles north of Tampa, Florida. The nearest population center is Gainsville, Florida, about 55 miles northeast of the site. Two fossil fired plants are currently operated en this site. Another nuclear unit is scheduled for construction and operation at this site also.
The B5W reactor core and steam supply system are very similar -
to the design of the Oconeo 1, nree Mile Island 1, Arkansas 1, and Rancho Seco reactors. The experience gained in our reviews of these plants will be reflected in our review of the Crystal River Plant.
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II OPERATING' LICENSE REVIEW PLAN ne designation of the group having responsibility for review and preparation' of comments, questions and report sections is indicated in parenthesis beside each paragraph.
In those instances where more than one group is designated, each group will be responsible for the specified review subject in those areas of their special competence. Stbject titles coincide to corresponding parts of the FSAR. We general design criteria referred to are those published in the FEDERAL REGISTER July 7,1971.
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' - 1.0 Introduction and Susunarv_
Review and~ evaluate the overall station (FWR #4) 1.1 design and compare with recent1'/ licensed facilities.
Review and evaluate the quality assurance (FWR #4) 1.2 program for plant operation to assure compliance with AEC criteria.
Review principal plant design for confore (FWR f4/DRS) 1.3 ance with the General Design Criteria for Nuclear Power Plants, Appendix A 1
of 10 CFR 50.
Review and evaluate the status of research (PWR f4/DRS) 1.4 and development programs which affect the planned operation of the f acility.
Raview and evaluate for acceptability those (PWR f4/DRS) 1.5 topical reports referred to fu support of the Crysta2 River design.
(FWR #4) 1.6 Review those portions of the final design affected by the cencerns expressed by the.
ACRS in their previous revisv of similar facilities.
Review the extent of foreign f abrict. tion of (PWR #_4) 1.7 materials and cctcponents and deternine the acceptability of such fabrication.
Review and evaluate the codes and quality (DRS) 1.8 classifications of all safety related structures, systems and components.
(DRS) 1.9 Review and evaluate the seismic spectra selected for dynamic analysis and resultant seismic design and analysis of Class I (seismic) structures, componenta, and equipment.
Review and evaluate instrumentation for (DRS) 1.10 earthquakes as indicated in AEC Safety Qaide 12.
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'(DES) 1.11 Review and evaluate the capability of structures, cc:aponents, and piping to withstand the ef fects of the design basis tornado.
(PWR #4) 1.12 Review and evaluate possible interactions-between the nuclear unit and the two fossil fired plants.
2.0' Site and Environment (SRS)'
2.1-Review and evaluate the acceptability of site boundaries, the exclusion area, the Icv population =ene, and the population center distance considering the population distribution data provided in the FSAR.
Check the population distribution against census tract data.
(SRS) 2.2 Review and evaluata site setmorological data and select the dif fusien coefficients that are used for routine. releases and for the accident analvsis as specified in Safety Guide 4.
(SRS) 2.3 Revies and evalucte the radiolotical environ-l
- t. ental rcnitering prcgrm, natural phenecena l effects, and anticipated radioactivity l
releases for conformance with proposed Appendin I to 10 CFR Part 50 and to identify potential mechanisns for accurulation and/or - j reconcentratlens of radioactivity.
-(DRS) 2.4 Review and evaluate site ground water conditions with regard to normal and in-advertent releases of radioactive liquids.
(BRS)
. 2. 5 Review and evaluate maximum flood level estimates considering the latest hurricane data.
(DRS) 2.6 Determine that the actual geologic findings during construction do not invalidate the asstened conditions which formed the basis of '
the accepted grouting program.
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(DRS)-
2.7 Detarmine that the actual grouting
-asasures performed during construction conformed to the grouting program accepted for Construction Permit issuance.
3.0 Reactor Review and evaluate significant diff erences (DRS) 3,1 in the nuclear design from previously approved facilities.
Review and evaluate any dif ferences in the (DRS) 3.2 fuel system from similar B&W product line
- reactors, i.e., Three Mile Island. Oconce.
(DRS) 3.3 Review and evaluate significant diff erences in the ther:nal and hydraulic design from previously approved f acilities.
(DRS) 3.4 Review and evalusta the adequacy of the analytical methods used to calculate core thernal and hydraulic design ch:racteristics of the punt. Ascertain the conservatism of the computational codes involved in multidimensional analysis of power distribu-tien, particularly power mis-natch.
Make une of prior evaluations, as c:plicable, on Oconce and ".1I-1.
(DRS) 3.5 Review and evaluate the design of core internals, their capability to withstand blowdown forces and the elimination of unwanted vibratien during normal operation making use of prior evaluations on Oconee and TMI-1.
(PWR f4) 3.6 Review and evaluate techniques for detection of failed fuel elements. Evaluate the adequc of the action to be taken upon detection of f ailed fuel.
(DRS) 3,7-Review and evaluate the analysis of the reactor, the reactor vessel, and the internals to-withstand stresses incosed by earthouake loads, hke use of nrier evalu-etions on Oconee and TFI-1.
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(DRS) 3.8 Review a evaluate the potential consequences '
of improper fuel loading within the fuel assembly and within the core.
(DRS) 3.9 Review and evaluate the intended uss and performance of the oroposed core instr.--enta-tion (e.g., inceres, excores, and thermo-couples) and the adequacy of the related proposed Technical Specifications.
(DRS) 3.10 Review and evaluate the potential for and consequences of a single rod withdrawal (uncontrolled).
4.0 Reactor Coolant SvJtem 4.1 Review and evaluate all significant dif-ferences from previously approved systeris pertaining to:
(DRS) 4.1.1 The reactor vessel design, fabrication and
- installation, especially in the a: plication of Section III and VIII ASME Boiler 2nd Pressure Vessel Code, electroclag welding, and any special problems associated with f abrication.
(URS) 4.1.2 The reacter coolant system includins; puros, valves, pressurizer, steam generators, and piping.
(DRS) 4.1.3 The pressure relief ::ystem including the bases for sizin;; the pressure relievin; devices and associated systems.
(FWR f4) 4.2 Review and evaluate adequacy of primary system leak detection methods, and the response to be taken upon. detection of leaks.
(DRS) 4.3 Review anc evaluate caeability of the reactor coolant system including ECCS for in-nervice inspection.cnformance with Section XI, ASME Boiler and Pressure vessel Code,' January 1970.-
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- (IRS) 4.4 Review 'and evaluate the reactor pressure
. vessel for fast neutron fluence and corre-sponding NDT. Note B&W topical report on.
this subject.
(DRS) 4.5 Review and evaluate primary system components relative to sensitization during fabrication.
~(DRS)
. 4.6 Review and evaluate the adequacy of missik and~ pipe whip protection for all vital systems.
(DRS)'
4.7.
Review and evaluate adequacy of design, in-service inspection procedtites. and quality control measures for primary coolant pump flywheels.
(DRS) 4.8 Review and evaluate the material surveillance program for.the reactor pressure vess el for conformance with the proposed requiret:ents of: Aprendix H to 10 CFR 50.
.DFS)-
4.9 Review and evaluata the 'abilhv to detect
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loose -parts and to monitor vibrations in the reac:or coolant systen.
Incitae review of plans for' ronitoring start-up vibration to establish initial vibration signatures for the critical coceenents and the system.
(DRS) 4.10 Review and evaluate the fracture toughness data of ferritic-=aterials of the reactor coolant prassura boundary.
(Consider the proposed requirements of Appendix C to 10 CPR 50.)
(DRS) 4.11 Review and evaluatt the safety and operating limits (i.e., heatup and cooldre.n rates, maximum pressure,. and minimum operating conditions) to be included in the Technical Specifications.
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5.0-
- Cont =4-nt svetens And other Snecial Structures 5.1 Review and evalusta all significant design differences from previously approved facilities; specifically, pertaining to the following:
(DRS) 5.1.1 Containment structural design including the capability of the containment to accept thermal stresses and differential pressures calculated from the postulated LOCA.
(DRS) 5.1.2 Materials testing for installed equipment and building materials to assure appropriate code techniques were employed. Consider reinforcing semel, tendons, steel liner, concrete, Cadweld splices, scalants, insulation, vessels, valves and puces.
(Include Safety Guide 10).
(DRS) 5.1.3 Penetration design and w.sthods for assuring contain:nent leak tightness at various tiscs in plant life.
(DRS)
'5.1.4
-The design and i=clementation of isel.ition criteria for air lacks, pipe penetracians,
instrunentation lines, and electrical penetrations.
(Include Snfetv Gaide 11).
(DRS) 5.2 Review and evaluate the calculations of peak pressure in the centainment following the postulated LOCA.
(7WR f4) 5.3 Review and evaluate post-LOCA conditions and their.long-term ef fect on the containment.
Include hydrogen build-up, potential metal-water reaction, contamination, and corrosive-properties 'of containment materials.
(Safety
. Guide 7).
(DRS).
5.4 Review and evalusts capability cf all
, Class I (seismic) structures to vithstand -
- aGeets without' loss of integrity f rom:
missiles and jet forces generated inside the structures, tornado-born missiles, and hurricane v.nds.
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'(PWR #4) 5.5 In conjunction with site analysia, review and evaluate the acceptability of the postulated containment leak rate at the containment design pressure and at lower
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pressure, and review and evaluate the proposed leak rate testing program.
(DRS) 5.6 Review and evaluate adequacy and appropriate-naas of the structural design to meet the criteria established during the CP review.
(DRS) 5.7 Review and evaluate the adequacy of the proposed tendon surveillance program.
f (323) 5.8 Raview and evaluate the consecuences of dropping a fuel cask in the spent fuel pit or other i=portant areas.
(DRS) 5.9 Revicv and evaluate the design adequacy of safety-related tanks and comonents from effects of natural phenomna cuch as tornado.
missiles and hurricane windu.
6.0
' Engineered Safer-Features (ESF)
(DRS)_
6.1 Revias and evaluate the performance of the emergency core cooling system (ECCS) for conformance with the policy statad in the Interim Policy Guide dated June 19, 1971.
(Include also Safety Guide 2 l
1 (PWR #4) 6.2 Review and evaluate the function of the Residual Heat Removal System motor operated valves.
(Include Position Paper - Motor Operated Valves RHR).
(PWR #4) 6.3 Review and evaluate the long-term core cooling capability following a IDCA, including the adequacy of water supply for
~ naintaining long-term cooling capability.
(Include Position Paper - Ultimate Heat Sink.)
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J (DRS) 6.4 Review and evalusta the ability of the containment spray and fan cooling systems to depressurize the containment and limit
. subsequent repressurization.
(PWR #4) 6.5 Raview and evaluate the safety injection tank isolation valve design.
(PWR #4) 6.6 Review and evaluate the compatibility of ESF with the accident envirorment.
( N 1 #4) 6.7 Review and evaluate design margins on adequacy of ont positive suction head for ECCS pumps I containment spray pumps.
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(N1 #4) 6.8 Review and evaluate the design of ESF in conformance with criteria for " single failures" including " passive" compenent designations in safety f eature dasign through to the ultimate heat sink.
(PVR f4) 6.9 Review and evaluace the capability for and adequacy of proposed in-aarvice testing and inspection of ESF connonents.
(DRS) 6.10 Review and evaluate ESF systems for ability
. to detect passive failurac during operation.
(WR #4) 6.11 Review and evaluate the extent tha: " field run" piping is to be used and the manner in which the final runs are to be checked against design predictions.
(PWR #4) 6.12 Review and evaluate the syste::rs provided to limit hydrogen buildup subsequent to a LOCA.
(Safety Guide 7).
(PWR 74/SRS) 6.13 Review and evaluate the function of the ESF equip sent in the post accident recirculation mode including leak detection, passive f ail-ure, and usintenance.
~(W R #4/33S) 6.14 Review and evaluate the equipment in the fuel pool aren that mitigates the conse-quences of 3 fuel handling accident.
11 7.0 Yantrumentation and Control' (DRS)
'7.1 Review and evaluate significant differences from previously reviewed plants of control systems for the actuation and control of engineered safety features and process safety systems considering conformance with IEEE 279, including diversity in systems actuation and ability of the systems to withstand possible natural phenomena and.
post accident environment.
(DRS) 7.2 Review and evaluate any significant dif fer-ences from previously acproved f acilities for instrumentatica and control sys tems associ-ated with process radiation monitors, refueling interlocks, area radiation menitora, site radiation monitors, and containment remote ronitoring instrument systems.
Eval-ucte capability of these syste=s to provide necessarv surveillance and control to assura conformance with radiation exposure and radioactivity release limitation of 10 CR Part 20 and 10 CFR Part 100.
C'C) 7.3 Reviev and evaluate electrical and control schematics related to energency power, reactor protection sys tem. ESF e,yscean, and containnent isolation and atacaphere centrol syste=s.
(PWR #4) 7.4 Review and evaluato. the adequacy of the control room and au: ciliary control station for reactor operations under normal and abnor:nal conditions.
(DRS) 7.5 Review and evaluate the adequacy of the proposed testing of the control and instrumentation system circuitry.
(DRS) 7.6 Review and evaluats the potential for and
'censequences of cocemen ecde f ailures in the reactor protection system.
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(DES) 7.7 Review and evaluate the consequences of possible single f ailures in the reactor protection system.
(DES) 7.8 Review and evaluate the ability of safety-related instrumentation to withstand the post-accident environnent in the containment.
(PWR #4) 7.9 Review and evaluate the adequacy of instrumentation for monitoring the course of an accident (meters, recorders, parameters). )
- (DRS) 7.10 Review and evaluate adequacy of auxiliary equipment to permit hot shutdown and potential capability of cold shutdown in the event the control recm is evacuated.
(DRS) 7.11 Review and evalusta the turbine qverspeed control instrumentation for conformance j
vith IEEE-2/9 require =ents.
~(?%*R #4) 7.12 Review and evaluate co ::munications cystems.
8.0
. Electriesl Svste=s (DRS) 8.1 Review and evaluate all significane differences free previously designed facilities of offsite, onsite and d.c.
power systems with particular a::tantien to the acequacy of supplying never to protection and safer-ins trumentation and equipment.
(Include Safetv Guide 6.)
(DRS) 8.2-Review and evaluate the plant's interaction with the external grid system and the relative independence of transmission lines.
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(DRS) 8.3' Review and evaluate the onsite power system with respect to the switchyard bus.and breaker arrangement, its connections and interlocks, and system independence.
(DRS) 8.4 Reviev and evaluate.the plant's conformance to the criteria for auxiliary electrical power nystems.
Consider also the evitching circuits for lead shedding from the emergency buses.
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(DES)
- 8. 5 Review and evaluate the design car, ability of the emergency diesel generators for
' confonnance with Safety Guide 9.-
a (DRS)
- 8. 6 Review and evaluete the conformity of Class IE electrical systems with the criteria in IEEE 297. Consider also the comments in the memo of R. L. Ferguson.
October 8,1969, "IEEE-ANS Standards Program-Report #9" with enclosure titled:
" Obtaining Power from.the Transmission Network for Nuclear Fueled Generating Stations."
(DRS) 8.7 Review and evaluate the imnlementation of a separation criteria, to be established, for cables and penetrations cf the reactor cro-tection, engineered safety and f
emergency syste ns.
9.0 Auxildarv and Emergency Svetems (Pk"R #4) 9.1 Review and evaluate fuel storage and handling for confor=anca vith Safety Guice_13 including linar materiala,
corrosion potential, protection fre=
missiles, effects of dropping fuel cask, handling equirnent operations and interlocks.
(Include ?osition Facer on fuel easP design.)
(PWR #4) 9.2 Review and evaluate irradiated fuel oit cooling and cleanup systema including water
- purity control, normal and maximum cooling capacity, level and vater drain control and pool leakaga effects.
(FWR #4) 9.3 Review and evaluate chemical and volume control system, boron recovery system, residual heat removal system and component cooling water syntem.
(PWR f 4).
9.4 Review and evaluate station fire protection including areas of automatic coverage and emergency scuer backup. Evaluation should include unsafe conditions the system might create.
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9.5 Review and evaluate safety aspects of station instrument and service air systems including emergency power sources.
(PWR f 4) 9.6.
Review and evaluate service water system including-redundancy, emergency power sources and. the effects of system failure.
(PWR #4) 9.7 Review and evaluate ventilation systems for the Contairunent Building, Turbine. Building, and Auxiliary Building, and ' the monitoring and isolation capabilities of the Centrol Room and Fuel Building ventilation systems with emphacia on design edecuacy under accident conditions (e.g., fire. LOCA, and Post-IDCA H
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( W R f4) 9.8 Review cnd evaluate equipment and floor drainage syste:ns includinz reliability of Icvel instrte:entation and alarms, lower-
- linics of leaic detection and surveillance of activity of drains.
10.0 S ters and Power conversion System
-(WR # 4) '
10.1 Paview and evaluata affecta of turbine and generator :: rips with an:1 without turbine by-pans operating.
- (NR #4)
>10.2 Review and evaluate rein s team hypaan system.:spability witn regard to load rejection and acceptance.
-_( W' R f'4).
10.3 Review and evaluate auxiliary stear systes.
- (PWR #4) 10.4 Review and evaluate condensate and feedvater system capacity for residual heat remov:tl' during emergency conditions.
. (WR f 4/0RS).
10.5 Review and evaluate potential for turbine-generator missiles including degree of protection provided for Class IL (seismic) equipment.
(Consider DRS-AEC afforts on i
it robr.bility of this event.)
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(FWR #4) 10.6' Review and evaluate main steam line isolation valve adequacy including require-ments for vendor test and field tests.
11.0
. Radioactive Waste and Radiation Protection (SRS) 11.1 Review and evaluate design and capability of the radioactive waste system to collect, confine, process, dispose, and scuitor radwa-ts within the limitations of Title 10, CFR Part 20, under normal and abnormal operating conditions. Evaluate conformance with the intent of proposed changes of 10 CFR Parts 20 and 50 ( Appendix I) concerning radiation exposures and releases of radioactive materials to unrestricted areas.
(Consider DREP ef forts on Cost-Benefit evaluation.)
~ (S RS) 11.2 Review and ' evaluate the es timated nornal releases including purgings and possible additional means for reducing planned or accidental release of radwaste to unrestricted areas.
(include Assessment of Cost vs. Benefits.)
FWR #4 11.3 Review and evaluate the ventilation system for the control room regarding isolation and/or filtering capability during a design basis acciden*
(SRS) 11.4 Review and evaluate die protection of FPC personnel against radiation during either normal or abnormal plant operating conditions.
12.0 Conduct '4f operations
-12.1 Review and evaluate the following aspects of the station operation:
(PWR f 4) 12.1.1.
Applicant's corporate and plant organization and responsibilities.
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0 (NR 4) 12.1.2 Technical qualification of operating.
. personnel and support staff.
. (FUR #4) 12.1.3 Selection and training of plant. personnel.
(Safety Guide 8)
(PWR #4)
.12.1.4 B&W and its contractors' relationship to the applicant's organization, responsibilities for training plant personnel and conducting tests prior to commercial operation.
(PWR f 4)
'12.1.5 Preparation and maintenance of records.
(PWR f4) 12.1.6 Emergency plans (reference Appendix E to 10 CFR 50).
-(PWR #4) 12.1.7 Review,- approval, authorization and control of procedures and (.hangas thereto.
(PWR f 4) 12.1.8 Industrial security progrm..
(PWR #4) 12.1.9 Plant audit provisions.
13.0 Initial Tests and Oeeration 1
13.1 Revies cad evaluate tne following:
(PWR #4) 13.1.1 Pre-operational tast pro;; ram.
(PWR 54)
-13.1.2 Initial criticality test progra::2.
(?WR #4) 13.1.3' Posteriticalitf test pro:; ram.
14.0 Safety Analvsis
-(PWR 4/SRS) 14.1 Review the accident analysis assumptions.
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m (SM) 14.2 Re-evaluate the radiological consequences resulting from the following accidents using the final site LPZ distance and current DRL meteorological dispersion curves:
a.
Loss of electric power b.
Steam line failure c.
Steam generator tube failure d.
Fuel handling accident a.
Control rod ejection accident f.
Letdown line rupture 3
Loss-of-coolant accident h.
Design basis accident.
(SRS) 14.3 review and e taluate the iodine removal capability of the filters and che:sical spray additive in the contaircent and the filters in the fuel storage area.
15.0 Technical _snacifications (FW #4, DLS, SRS) 15.1 Revica the proposed technical specifica-tions for conformance with 10 CTR 50.36.
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REVISED OPERATING LICENSE REVIEW S0!EDl'LE - TEGINICAL - CRYSTAL RIVER UNIT 3 TITLE DATE 1.
Ap plic at io n s ubmi t ted..................... February 8,1971 2.
Initial meeting with applicant............'..... September,1971
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3.
Initial draft requests for additional information PWR #4.... October 18, 1971 4.
Technical peeting with applicant.
November 11, 1971
'5.
Formal request for additional information to DPL Hansgement.
Decembet 3, 1971 6.
Formal request for additional information to applicant.
December 20, 1971 7..
Interim report draft to RP Branch.......
January 17, 1972 8.
Interim report to DRL Management.
January 31, 1972 9.
Response from applicant.
February 7,1972 10.
Second draft requests for additional information to RP Branch. March 20,1972
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11.
Fonnal request for additional information to DFL Management.
April 3, 1972 M
12.
Formal request for additional information to applicant.
April 12. 1972 Q
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. TITIE-DATE
~ 13. Responses from applicant.
.........May 10, 1972
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- 14.' Draft sections of ACRS report to RP Er:mch............ July 3,1972 -
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- 15. ACRS Reperc to DRL Management.
.. July,78, 1972
- 16. Transatt ACRS Report.
............ Augu s t 18, 1972-
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- 17. ACRS Meeting.
....Septe d er 1972 N
- 18. Safety Evaluation.
...oetober 1972 t
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