ML20033A837
| ML20033A837 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 10/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0528, NUREG-0528-S02, NUREG-528, NUREG-528-S2, NUDOCS 8111300013 | |
| Download: ML20033A837 (72) | |
Text
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NUREG-0528 Supplement No. 2 Safety Evaluation Report related to the operation of Wm. H. Zimmer Nuclear -Power Station, Unit No.1 Docket No. 50-358 gEO2hp'j Cincinnati Gas and Electric Company q%
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U.S. Nuclear Regulatory Fg Commission Office of Nuclear Reactor Regulation October 1981
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be avai!able from one of the following sources:
t.
The NRC Public Document Room,1717 H Street., N.W.
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NUREG-0528 Supplement No. 2 Safety Evaluation Report related to the operation of Wm. H. Zimmer Nuclear Power Station, Unit No.1 Docket No. 50-358 Cincinnati Gas and Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1981
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TABLE OF CONTENTS PAGE 1
INTRODUCTION AND GENERAL DISCUSSION................
1-1 1.1 Introduction........................
1-1
- 1. 8 Summary of Outstanding Issues..
1-1
- 1. 9 Summary of Confirmatory Items................
1-1 1.11 NUREG-0737 " Clarification of TMI Action Plan Requirements".......................
1-2 1.12 -Operating License Conditions.....
1-3 1.13 Other Matters Resulting From the Staff Opdated Review....
1-3 2
SITE CHARACTERISTICS.......................
2-1
- 2. 2 Nearby Industrial, Transportation and Military Facilities........................
2-1 2.3 Meteorology.........................
2-1 3
DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS......
3-1 3.7 Seismic Design.........
3-1 3.8 Design of Seismic Category I Structures.....
3-1 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment.
3-2 3.11 Environmental Design of Mechanical and Electrical Equipment...
3-3 4
REACTOR..............................
4-1 4.2 Fuel System Design.....................
4-1 4.3 Nuclear Design.......
4-3 4.6 Functional Design of Reactivity Control Systems.......
4-3 5
REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS...........
5-1 5.2 Integrity of Reactor Coolant Pressure Boundary.......
5-1 6
ENGINEERED SAFETY FEATURES....................
6-1 6.2 Containment Systems.....................
6-1 6.3 Emergency Core Cooling System...
6-1 6.4 Habitability Systems....................
6-2 7
INSTRUMENTATION AND CONTROLS...................
7-1 7.1 General Information.....................
7-1 7.3 Engineered Safety Feature Systems..............
7-1 Zimmer SSER #2 i
---v.
u TABLE 0F CONTENTS ~(Continued)
PAGE
- 7. 5 Safety-Related Display; Inst'rumentation..
7-1 J7. 7.
Control' Systems Not Required for Safety...........
7-2 8
ELECTRIC' POWER..........................
~8-1
~ 8.1 - Introduction........................
8-1 l9 1 AUXILIARY SYSTEMS.............
9............
9.1 ~ Fuel Storage and Handling..................
9-1.
'9.5' Fire Protection Systems................... 1 9.6
' Diesel Generator Systems..................
'9-2
.12-
' RADIATION PROTECTION.......................
12-1 12.3-Health Physics' Program.
12-1 13 CONDUCT OF OPERATIONS.......................
13-1 13.7 Industrial Security.....................
13-1 22 TMI-2 REQUIREMENTS........................
22-1 22.2 TMI-Action Plan Requirements for Applicants for Operating. Licenses.....................
22 23 CONCLUSIONS......................
23.....
APPENDICES A.
CHRONOLOGY............................
A-1 D.
ERRATA..........................
D-1 E.
BIBLIOGRAPHY......
E-1 F.
NRC STAFF CONTRIBUTORS AND CONSULTANTS..............
F-1 LIST OF TABLES Table 2-2 LPZ Relative Concentrations.
2...............
- Zimmer SSER #2 li.
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introductio_n Our Safety Evaluation Report for the Zimmer Nuclear Power Station, Unit 1 (NUREG-0528), Dock" No. 50-358, was issued January 1979.
Supplement 1 to that report was i._ed June 1981.
This supplement (Supplement 2) addresses the status of outstanding and confirmatory issues that remained in our review at the time Supplement 1 was issued.
Each issue is addressed in the appropri-ate subsection which is numbered to correspond to the subsection in which it is discussed in Supplement 1.
The NRC Project Manager assigned to the Operating License application for Zimmer is Irving A. Peltier.
Mr. Peltier may be contacted by calling (301) 492-7038 or writing:
I. A. Peltier Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 This Safety Evaluation Report Supplement is a product of the NRC staff and its consultants.
NRC staff members and consultants who were principal contributors to this report are identified in Appendix F.
1.8 Summary of Outstanding Issues The resolution of the outstanding issues listed in subsection 1.8 of NUREG-0528, Supplement 1, are discussed in the corresponding subsection of tnis supplement.
One issue which remains outstanding for this supplement to NUREG-0528 is listed below with the appropriate subsection reference.
This issue will be resolved prior to a decision to issue an operating license.
Issue Subsection Containment Leakage Testing.
6.2.6
- 1. 9 Summary of Confirmatory Items The implementation status of the confirmatory items listed in subsectior. 1.9 of NUREG-0528, Supplemant 1, are discussed in the corresponding subsections of this supplement.
Issues which still require confirmation at the time of this supplement are listed below with the appropriate subsection references.
If implementation of some staf f requirements are not confirmed prior to a decision to issue an operating license, the issues remaining to be confirmed may be made a condition of the operating license.
Zimmer SSER #2 1-1
Issues Subsections Toxic Chemicals......................
2.2.1, 6.4.2 Design of Seismic Category I Structures........
3.8 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment.
3.10 Environmental Design of Mechanical and Electrical Equipment.
3.11 i-Seismic and LOCA Loading.................
4.2.3 Scram Discharge System..................
4.6.2 Physical-Separation of Associated Cables.........
7.1. 3 Concern ~that Common-Electric Power Sources or Sensor
-Malfunction May Cause Multiple Control System Failures.
7.7.3 Concern Regarding High Energy Line Breaks and Consequential Control System Failures..........
7.7.3 1.11 NUREG-0737 " Clarification of TMI Action Plan Requirements" Listed below are the NUREG-0737 requirements for Zimmer which need further consideration by the staff and applicant in order to achieve confirmation of-full implementation on the Wm. H. Zimmer Nuclear Power Station.
We will report further on these matters in a future supplement to this report.
Additional Item information Confirmation (NUREG-0737)
Short title from applicant by the staff I.C.5 Procedures for Feedback of Oper-ating Experience to Plant Staff...
X X
II.E.4.1 Dedicated Hydrogen Penetration...
X X
II.F.1 Containment High-Range Radiation Monitor.....
X X
II.F.2 Incore Thermocouples........
X II.K.3 Common Water Level Reference.
X X
III.A.1.1 Upgrade Emergency Preparedness...
X X
III.A.2 Improving Licensee Emergency Preparedness-Long Term..
X X
111.D.3.4 Control Room Habitability......
X X
Zimmer SSER #2 1-2
1.121 Operating License Conditions The following' requirements will be made a condition of the operating license.
Requirement Subsections
~
Implementation of NUREG-0577 Guidelines..........
5.2.3-
. Implementation of NUREG-0803 Guidelines..........
6.3.4 Degraded Grid:Vultage...................
8.1.2 Protection of Reactor Containment Electrical Penetrations..
8.1.2 Station Blackout Events..................
6.1. 2 Cuntrol of Heavy Loads...
9.1.4 Diesel Generator Reliability...............
9.6 Proper Safety Features Functioning............
II.K.1-5 Restart of Reactor Core Isolation Cooling System..
II.K.1-22 1.13 Other Matters Resulting From the Staff Updated Review Occasionally'the applicant will update the Final Safety Analysis Report in areas that previously had been reviewed by the staff.
Usually this updating results from changes in the facility or conditions relative to safety which have changed since the original submittal.
The staff reviews these changes for. implication regarding its completed safety evaluation and reports the results in a supplement to the original safety evaluation report.
Subsec-tions 2.3.3, 2.3.4, 2.3.6, 4.3.2, and 12.3 present the results of the staff's review of revised information in the Safety Analysis Report.
Zimmer SSER #2
.1-3
2 '5ITE CHARACTERISTICS 2.2 Nearby Industrial, Trar.sportation and Military Facilities 2.2.1 Transportation of Toxic Chemicals The applicant has implemented a hazardous cargo truck survey on U.S. Route 52 in the vicinity of the Zimmer plant site and has provided the staff with an interim report in Amendment 127, Revision 76, to the Final Safety Analysis Report.
The survey is planned to be completed late in 1981.
The applicant has ccmmitted to provide appropriate protection for the Zimmer control room if the results of the survey show that the probability of a toxic chemical release (for which the control room does not have protection), which could jeopardize the control room, exceeds 10 6 The staff will review the results of the survey and the probabilistic analysis when they are completed.
2.3 Meteorology 2.3.3 Onsite Meteorological Measurements Program Two meteorological towers are in use at the site.
The principal tower, 200-ft.
tall is situated in the Ohio River valley slightly over one mile north of the plant cooling tower.
On this tower, wind speed, and wind direction are measured at the 30- and 200-ft. levels.
Temperature, turbulence and dewpoint are measured at the 30-ft. level with temperature difference measured between the 30- and 200-ft. levels.
The second tower is 150-ft. tall on a hill top slightly over a mile northeast of the plant cooling tower.
The base of this tower is about 400-ft. above that of the valley tower.
Instrumentation on this 500-ft. tower measures wind speed and direction at the 50 and 150-ft. levels in addition to temperature, dewpoint and turbulence at the 150-ft. level.
The measurements from both towers are transmitted to the plant control room and technical support center as well as being remotely interrogable from offsite.
Information responsive to the meteorological requirements of NUREG-0654 Rev.1 Appendix 2 will be reviewed and evaluated in a separate emergency preparedness document.
Note:
Table 2-1 in NUREG-0528 may be deleted.
2.3.4 Short-term (Accident) Dif fusion Estimates Short-term accidental gaseous releases at ground level were evaluated.
Onsite 10 meter wind data from 1972-1974 was used with the direction dependent model described in Regulatory Guide 1.145, " Atmospheric Dispersion Models fnr Potential Accident Consequence Assessments at Nuclear Power Plants," for relative concentration (X/Q) calculation at the 250 meter exclusion radius and the 4826 meter outer boundary of the low population zone (LPZ).
Zimmer SSER #2 2-1 1
f I-
- The Q-2 hour X/Q that would be expected to be exceeded no more than five percent
- of the time is 7.1 x 10.a 3
sec/m at the 250 m exclusion radius northwest of the-plant.~
L At the LPZ, values of X/Q are given in Table 2-2.
Table 2-2 LPZ relative concentrations l
a
' Time X/Q (sec/m )
0-8 hours 9.2 x 10 5 l
8-24 hours 6.4 x 10 5 l
1-4 days 2.8 x 10 5 l-l 4-30 days 9.5 x 10 5 2.3.6 Conclusions l
l We conclude tiie applicant has provided sufficient meteorological information responsive to 10 CFR Part 100.10 in the Final Safety Analysis Report to deter-mine the plant's suitability for operation under the meteorological regime at the site.
I l
l l
Zimmer SSER #2 2-2=
'3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.7 -Seismic' Design 3.7.2 Seismic System and Subsystem Analysis Soil - Structure Interaction Analysis In Supplement 1-to the SER (NUREG-0528) we stated that the soil damping values used in.the analysis were appreciably higher than expected and the applicant was requested to make re-arelysis using lower soil damping values.
The applicant has made the.re analysis using a soil damping value of 5% and submitted the results of the comparative study which indicated that with the_ exception of two slabs where minor differences exist between the two spectra the-Zimmer design spectra envelop those obtained using the visco-elastic half-space method.
On the basis of our review of this information provided by the applicant, we--
consider this issue on soil-structure interaction analysis is satisfactorily j
resolved.
3.8 Design of Seismic Category I Structures Pool Dynamic Loads The applicant is continuing the assessment for structures such as the reactor supports, drywell floor, drywell floor columns and the containment structure for which margin factors are low using the final pool dynamic loads.
The effort is planned for completion by December 31, 1981.
The staff will review the results of this confirmatory effort when it is completed.
Masonry Wall In Section 3.8.2 of NUREG-0528, Supplement 1 (Zimmer SER) several criticisms of the applicant's design bases for masonry walls were listed.
By Amendment 127, Revision 76, to the Final Safety Analysis Report, the applicant aoreed to take the following actions:
"All Class I masonry walls at Zimmer will be strengthened as required before the fuel loading date.
The masonry walls will be reassessed to the NRC's Structural Engineering Branch Interim Criteria for Safety Related Masonry Wall Evaluation Revision 1, dated July 1981.
The masonry wall reassessment includes the following items:
(A) The 1.67 factor on the National Concrete Masonry Association allowable stresses will not (be) used. Allowable stresses for different loading conditions as provided in'"_SEB Interim Criteria for Safety Related' Masonry Wall Evaluation, Revision 1, dated July 1981," have been used with the following modification:
Zimmer SSER #2 3-1
1.
Design Assessment Report Loading Combination Table (Table 6.1-1) for reactor building and Final Safety Analysis Report Table 3.8-10 for auxiliary building have been used for the design of concrete masonry walls.
(B) Allowable stresses for Type N mortar have been used for the reassessment of blockwalls, even though masonry core samples taken in 1978 substantiated mortar strength equivalent to Type M mortar.
(C) Blockwalls with thermal gradients will be analyzed to include the effect of end restraints and fixes provided accordingly.
(D) To check the local stresses on the block, the mortar joints around the block only are considered to resist the load on individual blocks.
(E) All concrete masonry walls set on concrete curbs have been designed to span horizontally for out of plane loads.
Therefore, no shear transfer is required at the interface of the masonry wall and curb.
(F) Use of cracked section properties when the calculated stress is larger than the allowable stress but smaller than the rupture modulus is not appropriate.
Moment of intertia of the gross section should be used unless the section cracks under the applied load.
However, the strength of the section has been calculated on the basis of the cracked section whenever the allowable stress is exceeded."
The commitment to perform the above actions, within the time frame indicated, constitutes satisfactory resolution of the previously noted differences with the staff's position.
Accomplishment of these actions will thus assure that the masonry walls at Zimmer will not compromise any safety related components or systems.
3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment The Seismic Qualification Review Team (SQRT), consisting of Engineers from the Equipment Qualification Branch (EQB) and the Idaho National Engineering Laboratory (INEL, EG&G), conducted a site visit to Wm. H. Zimmer Nuclear Power Station at Moscow, Ohio, on June 2 to 5, 1981.
The purpose of the visit was two-folded:
(1) to perform a plant site review of the seismic and dynamic qualification methods, procedures, and results for selected safety-related mechanical and electrical equipment and their supporting structures, (2) to observe the field installation of the equipment in order to verify and validate equipment modeling employed in the qualification program.
The applicant has described the equipment qualification program in Sections 3.9 and 3.10 of the Final Safety Analysis Report, consisting of dynamic testing and analysis, used to confirm the ability of seismic Category I mechanical and electrical (includes instrumentation, control and electrical) equipment and their supports, to function properly during and after the safe shutdown earthquake (SSE) specified for the plant.
The applicant has also described the program for the combined seismic and hydrodynamic vibratory loads associated with the MARK II containment suppression pool.
Zimmer SSER #2 3-2 l
l In instances'where components have been qualified by testing or analysis to other than current standards such as Institute of Electrical and Electronics Engineers Standard, 344-1975, " Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear. Power Generating Stations," and Regulatory Guides 1.92, '! Combining Modal Responses and Spatial Components in Seismic Response Analysis," and 1.100, " Seismic Qualification of Electrical Equipment for Nuclear Power Plants," or where equipment-is af fected by and was not qualif:ed for the suppression pool hydrodynamic loads, the applicant has undertaken a reevaluation and requalification program.
The plant site review was performed to determine the extent to which the qualification of equipment, as installed in Zimmer, meets the current licensing
. criteria as described in the Standard Review Plan (SRP) Sections 3.9.2 and 3.10.
Prior to the site visit, the SQRT reviewed the equipment seismic qualification information contained in the pertinent Final Safety Analysis Report sections and the reports referenced therein.
A representative sample of Seismic Category I mechanical and electrical equipment, including both Nuclear Steam Supply System and Balance of Plant scopes, were selected for the plant site review.
The review consisted of field observations of the actual equipment configuration and its installation, followed by the review of the corresponding test and/or analysis documents.
Brief technical discussicns were held during the review sessions to provide SQRT's feedback to the applicant on the equipment qualification.
The results of field observations and the review of the qualification reports and pertinent documents were summarized for each piece of equipment evaluated.
The plant site review identified the need to provide additional information on certain generic issues as well as to clarify the details uf the qualification for some specific pieces of equipment.
The applicant has committed to submit additional information and clarification for a followup review.
Subsequently, on July 21, 1981 the applicant sent to NRC a post-audit submittal.
The review of applicant's post-audit submittal of July 21, 1981 resulted in the resolution of a number of concerns.
The review of the applicant's implementation of the equipment qualification program is continuing and the applicant is required to resolve all outstanding items as identified prior to licensing.
3.11 Environmental Design of Mechanical and Electrical Equipment In December 1979, the staff issued guidance for the environmental qualification of safety-related electrical equipment (NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment").
By letters dated October 11, 1979 and February 19 and 21, 1980, the staff requested Cincinnati Gas and Electric Co. to review the environmental qualification documentation for each item of safety-related electrical equipment which could be exposed to a harsh environment so as to identify the degree which the' associated environmental qualification program complies with the staff's position as described in this NUREG.
Further, where there are deviations, the applicant must commit to corrective action (regulation, replacement, relocation, etc.) consistent with the requirements to establish qualification.
If fuel F
Zimmer SSER #2 3-3
r
~
loading occurs before the deadline, justification for operation until the corrective actions are completed must be provided.
The Commission Memorandum and Order (rLI-80-21, dated May 23, 1980) directs the staff to couplete its review of' environmental qualification including the publication of Safety Evaluation Reports for all Operating Reactors.
In addition, this order directs that by no later than June 30, 1982, all electrical equipment in operating reactors subject to this review be in compliance with NUREG-0588 or the Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors.
The applicant submitted qualification information by letter dated January 7, 1980 and a subsequent amendment by letter dated August 20, 1981.
The staff is currently reviewing this information for completeness.
Upon completion of our review of the information, the staff will conduct an audit and prepare the environmental qualification safety evaluation report.
Zimmer SSER #2 3-4
4 REACTOR 4.2 Fuel System Design 4.2.3 Design Evaluation Fission Gas Release In 1976 we questioned (Ref. 1) tha validity of fission gas release calculation in-most fuel performance codes including GEGAP-III for burnups greater than 20,000 mwd /tu.
General Electric was informed of this concern and was provided with a method (Ref. 2) of correcting gas release calculations for burnups greater than 20,000 mwd /tU.
Although a reanalysis was not specifically performed for the Zimmer fuel, an 8x8 reanalysis (Ref. 3) performed for early reflooding plants reportedly bounds the Zimmer case.
In the generic reanalysis, fuel rod internal pressure was shown to remain below system pressure for rod peak burnups below 40,000 mwd /t.
This conclusion remains unchanged for the newer prepressurized fuel design as well (Ref. 4).
The generic reanalysis did, however, result in higher initial stored energy and rupture pressure in the loss-of-coolant accident (LOCA) analysis.
In Reference 5 and 6, General Electric Company requested that credit for recently approved emergency core cooiing system evaluation model changes be used to offset any operating penalties due to high burnup fission gas release.
This proposal was found acceptable (Ref. 7) provided the generic analysis was found applicable to Zimmer.
In Revision 76 of the Zimmer FSAR, the applicant stated that the generic analysis is applicable to the Zimmer design.
We, therefore, consider the issue of enhanced fission gas release at high burnup to be satisfactorily resolved for Zimmer.
Ballooning and Rupture A generic sensitivity study of fuel rod cladding ballooning and rupture phenomena during a loss-of-coolant accident (LOCA) was submitted by General Electric Company (GE)-(letter from R. H. Buchholz to L. S. Rubenstein, dated May 15, 1981) in response to a condition in the approval (letter from R. L. Tedesco to G. G. Sherwood, dated February 4, 1981) of the General Electric Company ECCS Evaluation Model and a subsequent submittal from G. G. Sherwood to L. S.
Rubew+,ein, dated August 14, 1981, addressed the cladding strain analysis, and included a sensitivity analysis of the effect of the degree of rupture strain on the calculated peak cladding temperature.
As reported in the initial generic study, General Elect ric Company assessed the boiling water reactor (BWR) emergency core cooling system (ECCS) sensitivity to rupture temperature by using three rupture temperature models:
(1) the GE CHASTE model, (2) the NUREG-0630 model, and (3) a proposed GE model termed the adjusted model.
For the Zimmer type of 8x8 fuel design, CE found that the use of the adjusted model, which may be the best of the three models and which is in fact a combina-tion of the CHASTE and NUREG models, gave a maximum impact on peak cladding temperature (PCT) of 5 10 F.
We concluded that the use of this adjusted model was acceptable.
Zimmer SSER #2 4-1
'l With regard to the BWR ECCS sensitivity to burst strain, the General Electric Company submittals assessed the impact of using a burst strain model that bounds the burst strain model given in NUREG-0630.
However, prior to performing this comparison, the bounding strain model was appreciably reduced by axially averaging the cladding strain.
Two reduction factors were used to effect this averaging process; one was 2.8 (for fuel bundle interior rods) and the other was 4.1 (for fuel bundle peripheral rods). 'While some averaging is appropriate for the whole-bundle analysis performed for Zimmer, there is still some uncertainty as to the appropriateness of the factors used by General Electric-Company.
However, the possibility of a future increase in calculated PCT resulting from a modification in the reduction factors for averaging the cladding strain notwithstanding, the calculated LOCA PCT for Zimmer is 1821 F, which is lower than the 2200 F licensing limit and which provides ample margin to accommodate those uncertainties in the generic sensitivity study.
Therefore, we conclude that the ballooning and rupture issue has been resolved for Zimmer Unit 1.
Seismic and LOCA Loadings The staff is pursuing this confirmatory matter on a generic basis.
Final confirmation of a satisfactory resolution will be reported in a future supplement prior to the issuance of an operating license.
Channel Box Deflection Channel box bulge is a result of in reactor creep, which if allowed to proceed unchecked, could eventually create an interference between the control blade and the fuel channels.
In a General Electric Company (GE) generic report, NEDE-21354, on channel box mechanical design and deflection, GE describes a channel lifetime prediction method and a backup recommendation for periodic channel measurements which consist of settling friction tests.
The settling friction tests would provide an exact control rod drive friction versus position profile by measuring the hydraulic pressure under the drive piston as the drive " settles" to a latch position.
In Revision 76 to the FSAR, the applicant proposed several steps that would be taken to mitigate the consequences of channel bowing, including the following.
(1) Prior to beginning a new operating cycle, control red drive friction tests shall be performed for those core cells exceeding specified general guidelines or containing fuel channels with exposures greater than 30,000 Mwd /T (associated fuel bundle exposures).
(2) In lieu of friction testing, fuel channel deflection measurements may be used to justify use of fuel channels exceeding 30,000 mwd /T exposure for a maximum of four additional operating cycles.
(3) In the future, analytical channel lifetime prediction methods, benchmarked and backed up by periodic deflection measurements of a sample of the highest duty fuel channels, may be used to ensure clearance between control rod blades and fuel channels without additional testing.
Zimmer -SSER #2 -
'4-2
In addition, the applicant has described some administrative guidelines that would be followed concerning channel location, exposure, and residence time.
While the review of-the GE generic report, NE00-21354, has not been completed, we agree with the applicant that the proposed actions should mitigate the consequences of channel bowing and that they provide reasonable assurance that channel bowing will not be a problem in Zimmer Unit 1.
Should our continuing generic review of this phenomenon (and the GE report) reveal that a modifi-cation to the proposed steps is necessary, the licensee will be informed to take appropriate action.
4.3 Nuclear Design 4.3.2 Design Description Vessel Irradiation Neutron fluences at the reactor vessel were calculated for the Zimmer size plant using a one-dimensional discrete ordinates code.
Continuous reactor operation at full power for 40 years was assumed and the reactor mid plane flux was calculated.
The calculations incorporated fission distributions prepared from core physics data as an initial fixed source distribution.
Anisotropic scattering effects were included outside the core and the resultant vessel fluence was 5.3 x 1019 neutrons per square centimeter for neutrons having energies greater than one million electron volts.
This value includes an azimuthal peaking factor of 1.4 and an additional factor of two which was included to ensure conservatism.
We conclude from our review that the calculated value of vessel fluence noted above is acceptable.
4.6 Functional Design of Reactivity Control System 4.6.2 Control Rod System Scram Discharge System By letter dated July 30, 1981, the applicant responded to the staff's December 1, 1980 report, "BWR Scram Discharge System Safety Evaluation," as clarified by the staff's letter dated March 30, 1981.
The staff is presently reviewing confor-mance of the Zimmer scram discharge volume design with its generic report.
We will provide our conclusions in a future supplement to NUREG-0528.
Zimmer SSER #2 4-3
5 REACTOR-COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary
- 5. 2.1 Design of Reactor Coolant Pressure Boundary Components Feedwater and Control Rod Return Line Nozzle Cracking Leaks and cracks in the heat-affected zones (HAZ) of welds that join austenitic stainless steel piping and associated components in boiling water reactors (BWR) have been observed over the past several years.
As a result of the detection of cracks in BWR components, two Generic Technical Activities were identified, A-10, "BWR Nozzle Cracking," and A-42, " Pipe Cracks in Boiling Water Reactors,"
as Unresolved Safety Issues, and Task Groups were formed and chartered by the Commission to. study and recommend resolution of the problems.
NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," was issued in April 1980, for resolution of Generic Technical Activ-ity A-10.
The applicant's response to the provisions of NUREG-0619 was evaluated and accepted by the. staff.
Valving out of the control rod drive line is accept-able provided an administrative control is established for this valve.
Because this line is made of carbon steel, the inspection requirements of NUREG-0619 are not applicable. We find the applicant's response to NUREG-0619 and corrective actions acceptable.
5.2.3 Reactor Coolant Pressure Boundary Materials Stainless Steel Pipe Cracking NUREG-0313, Revision 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," was issued in October 1979 for resolution of Generic Technical Activity A-42.
The applicant's response to the provisions of NUREG-0313, Revision 1, was contained in letter dated July 7, 1981.
Regular grade type 304 stainless steel has been replaced to the extent practical with type 304L or 316L stainless steel.
The applicant conforms to the materials selection and processing guidelines of NUREG-0313, Revision 1.
Inspection requirements in the Technical Specifications will be in compliance with the requirements of NUREG-0313, Revision 1.
We find the applicant's response to NUREG-0313, Revision 1, and corrective actions acceptable.
Fabrication and Processing of Ferritic Materials in Amendment 127, Revision 76, the applicant responded to NUREG-0577 and stated that future commitments regarding this matter are dependent on NRC-industry agreement on an acceptable program.
Implementation of such a program will be made a condition of-the operating license.
-Zimmer SSER #2 5-1
5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing Inservice Testing of Pumps and Valves In Supplement 1 to NUREG-0528 we stated that the applicant's response to Question 212.58 was unacceptable and that the pressure isolation valvas for the low pressure and high prewJre core 3 pray, residual heat removal, and the reactor core isolation cooling systems must be subject to certain leakage testing requirements.
Since our last report, the applicant has committed (Amendment 127, Revision 76, to the Final Safety Analysis Report) to modify the pump and valve testing program and to comply with the technical specifica-tions to meet.all of our requirements for leakage testing of the above valves.
Specifically the applicant has committed to categorize the pressure isolation valves as A or A/C and to meet the leakage testing requirements of ASME Section XI.
Also, the applicant has agreed to use 1.0 gallon per minute as the leakage limit and has correctly listed the valves to be tested in l
Table 5.2-20 of the Final Safety Analysis Report.
We find these commitments acceptable.
We conclude that the app;icant's commitments and the technical specifications i
to periodically leak test the pressure isolation valves between the reactor coolant system and low pressure systems will provide reasonable assurance that the design pressure of the low pressure systems will not be exceeded, and thus reduce the probability of an occurrence of an intersystem loss-of-coolant accident.
Criterion 55 of the General Design Criteria of Appendix A of 1
10 CFR 50 partially considers this matter.
Zimmer SSER #1 5-2
F 1
6 ENGINEERED SAFETY FEATURES 6.2-Containment Systems 6.2.6 Containment Leakage Testing The applicant is-currently. reviewing its emergency core cooling systems (ECCS) to demonstrate that in the event of loss of diesel power to the main pumps and booster pumps, a water seal can be maintained to the injection valves for 30 days. -If the 30 day water seal to the ECCS injection cannot be demonstrated, then the valves must be tested with air rather than hydrostatically tested with water.
The resolution of this item will be reported on in a supplement to this report.
6.3 Emergency Core Cooling Systems 6.3.4 Performance Evaluation Residual Heat Removal System (Operator Action)
If a crack in the residual heat removal common suction line outside primary containment is postulated to occur during shutdown cooling, reactor vessel water level would decrease in Level 3 causing isolation of the line.
Reactor pressure would rise to the oHV setpoint as a result of the isolation, thus causing safety / relief valves to open and reclose.
If all high pressure systems were assumed to be unavailable (plant is shutdown), and only low pressure core spray and one low pressure coolant injection were available, manual opening of safety / relief valves would be required to depressurize the vessel.
The applicant has evaluated the effects of the crack and analyzed the above scenario for Zimmer to show how much time the operator has to depressurize the vessel so that low pressure core injection systems can restore the reactor vessel level.
Results from the analysis showed that more than 20 minutes are available for the operator to do,.ressurize the vessel under the postulated conditions.
This is acceptable to us.
Scram System Pipe Break NUREG-0803 provides guidance for an acceptable plant - specific resolution of the issues related to this concern.
This guidance is grouped into three major areas:
(1) Piping Integrity - Licensees and applicants are to verify proper scram discharge valve (SDV) piping installation by as-built inspection and propose an inservice inspection program of the SDV system which meets the requirements of ASME Section XI for Class 2 piping.
(2) Mitigation Capability - Licensees and applicants are to implement revised emergency procedures for pipe break in the scram system.
Zimmer SSER #2 6-1
,i (3)- Environmental Qualif! cation - Licensees and' applicants are-to identify equipment needed to:
(s) detect an SDV: system breaks.and
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(b) mitigat'e the consequences'of such a break and propose a program for qualifying such equipment-(if not-environmentally qualified)..The Cincinnati Gas and Electric Company will be asked to provide commit-ments and schedules for implementation addressing these areas in their 120 day response to NUREG-0803.
. 6.4 Habitability Systems y
6,4.2 ' Toxic Gas Protection See subsection -2.2.1 of this supplement for the status 'on this inatter.
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Physical Separation of Associated Circuits ix. he applic
.ontinuing a 100% analysis of physical separation of associated p W'circbits anm glans,to' complete this effort by November of 1981.
In Amendment 127, Revision 76,/ o tnepinal Safety Analysis Report, the applicant provided the t
' - e aT l staff with a statJs rep 9rt and justification for any deviations from Regulatory W
Guide 1.75.
We will report on any unacceptable deviations from Regulatory
-, Guide1.75 in a future supplement to NUREG-0528.
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7.3 Engineered Safety Feature Systems 7.3.3 Specific Findings Y
Loss of Safety Function Aftet Reset 3
Aswasdonefor,operatingreactorsthroughIEBulletin80-06,werequested
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Fpplicant has committed to modify the reset design for those cases where
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changes in state of safety equipment upon rest of engineered safety feature
- (ESF) signals were identified.
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Any problems discovered during the tests will be corrected or
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On this basis, we conclude that use of ESF
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' q 5. 3 Specific Findings tLoss'of Power to Instruments and Control Systems s
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[\\ n' Wd requested that the applicant review the adequacy of emergency operating procedures to be used by control room operators to attain safe shutdown upon T~
3 loss of'any class 1E or non-class IE buses supplying power to safety or non-safety-rdlated instruments and to control systems (This issue was addressed foraperatirit reactors through I&E Bulletin 79-27).
The applicant has demon-
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strated that sufficient equipment for safe shutdown would remain available L
subsequ'ent ta lossiof any 1E or non-1E electrical bus.
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'i' committed 1.o conduct a failure effects analysis of such buses and to modify g
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cmergency operating procedures dealing with the resulting plant conditions where necessary. We believe that this commitment constitutes an acceptable resolution f
i of this issue.
7.7~ Control Systems Not Required For Safety 7.7.3 Specific rindings l
Control System Failures With regard to the effects of control system failures or malfunctions, the analyses reported in Chapter 15 of the Final Safety Analysis Report are intended to demonstrate the adequacy of safety systems in mitigating anticipated opera-
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tional occurrences ar.d accidents, including those related to control systems.
Based on the conservative assumptions made in defining tnese " design bases" events and the det iled review of the analyses by the staff, it is likely that they adequately bound tne conse pences of single control system failures.
To provide anurance that the Chapter 15 analyses adequately bound events initiated by a single credible failure or malfunction, we asked the applicant to provide additional information.
The applicant has committed to conduct a review to identify any power sources or sensors which provide power or signals to two or more control systems, and to demonstrate that failures or malfunctions of these power sources or sensors will not result in consequences outside the bounds of the Chapter 15 analyses or beyond the capability of operators or safety systems. We will report on this confirmatory matter in a future supplement to NUREG-0528.
H3hEnergy.LineBreaks' We have also requested a review by the applicant to determine whether the
. harsh environments associated with high energy line breaks might cause control system malfunctions and result in consequences more severe than those of Chapter 15 analyses or beyond the capability of operators or safety systems.-
The applicant has committed to conduct such a review. We will report on this confirmatory matter in a future supplement to NUREG-0528.
p Zimmar SSER #2 7-2
8 ELECTRIC POWER 8.1 Introduction 8.1.2 General Findings Degraded Grid Voltage The applicant has reanalyzed the voltage levels on the Class 1E buses, and in Amendment 127, Revision 76 to the Final Safety Analysis Report, dated August 1981 provided the results of his analysis.
The relay trip setpoint for the required second level undervoltage protection will be increased to approx-imately 3800 volts.
In addition, the applicant has changed the time interval for actuatior, of the alarm on the second level undervoltage relay from zero seconds to time interval greater than a motor starting transient (variable range of 0 to 20 seconds) and trip offsite sources in 5 minutes.
The proposed second level undervoltage scheme and relay setpoints will protect the Class 1E equipment from operation under sustained degraded grid voltage conditions within the expected range of grid voltage limits.
The proposed modifications conform to part (a) of position 1 on degraded grid voltage.
We find this acceptable and therefore consider this item resolved.
Because of the long delivery time of the equipment, the applicant states that this modification will not be implemented until the first refueling outage.
We find this acceptable. We will condition the operating license upon the satisfactory implementation of this modification per design, prior to restart following the first refueling outage.
Protection of Reactor Containment Electrical Penetrations The applicant in Amendment 127, Revision '/6 to the Final Safety Analysis Report, dated August 1981 provided the maximum fault current versus time profiles and time current characteristics (12Rt) of the penetration conductors.
In addition, the applicant has submitted drawings showing proposed designs for the protection of electrical penetrations.
Except for the 6.9-kV reactor coolant pumps penetrations, all other protective devices will be fault current actuated and will require no external power source.
This includes the 480-volt drawout case type breaker which receives tripping power from the fault via current transformers.
The 6.9-kV penetrations are the only items where control power for primary and backup protection devices will be provided from independent Class 1E DC sources, i.e., powered-from DC buses 10 and IE.
This confcrms with Regulatory Guide 1.63,
" Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants," and is acceptable.
We consider this item closed.
Zimmer SSER #2 8-1
~We.will condition the operating license upon the satisfactory implementation of the modifications as stated above, prior to restart following the first refueling outage.
5,tation Blackout Events The applicant is proceeding with response to generic letter 81-04, " Emergency Procedures and Training for Station Blackout Events", and has reported the status in Amendment 127, Revision 76,.to the Final Safety Analysis Report.
Interim emergency procedures and operator training for blackout events will be implemented prior to fuel loading.
Other requirements which may result from the staff's generic review (see A-44, Appendix C, NUREG-0528, Supplement 1) will be made an appropriate condition of the operating license.
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0 9 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.4. Fuel Handling Systems Control of Heavy Loads In Supplement No.1 to NUREG-0528 we stated that our review of load handling operations at nuclear power plants had been completed and that a report describ-ing the results of our review had been issued as NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." We further stated that the applicant was requested by letter dated December 22, 1980 to review the recommenaations con-tained in NUREG-0612 to determine the extent to which the guidelines are presently satisfied at the Zimmer facility, and to identify the changes and modifications that would be required in order to fully satisfy these guidelines.
Implementation of certain interim actions were also required by the Deceraber 22, 1980 letter. to that letter identified a number of measures dealing with safe load paths, procedures, operator training and crane inspections, testing and maintenance.
The applicant has committed to implement these interim actions prior to the final implementation of NUREG-0612 guidelines and prior to the receipt of their operating license.
This commitment is acceptable.
The appli-cant will submit the results of the review against NUREG-0612 guidelines at a later date.
Based on our review, we conclude that the fuel handling system is in conformance with the requirements of General Design Criteria 2 and 61 as they relate to its protection against natural phenomena and safe fuel handling and the guidelines of Regulatory Guides 1.13, " Spent Fuel Storage Facility Design Basis," and 1.29,
" Seismic Design Classification," with respect to overhead crane interlocks and maintaining plant safety in a seismic event and is, therefore, acceptable.
We further conclude that implementation of the interim actions of NUREG-0612 prior to final implementation of NUREG-0612 guidelines and prior to receipt of the operating license provides reasonable assurance of safe handling of heavy loads until NUREG-0612 can be fully implemented and is, therefore, acceptable.
95
,F_ ire Protection System Appendix R Requirements In sections "I Introduction" and "VII Conclusion" of the staff's " Fire Protec-tion Review," Appendix E to NUREG-0528, Supplement 1, we indicated that although the Fire Protection Program of the Wm. H..Zimmer plant, with the proposed improvements, met the guidelines contained in Appendix A to BTP ASB 9.5-1, the applicant had not committed to meet the requirements of Appendix R to 10 CFR Part 50 or provide equivalent protection.
By letter dated October 5, 1981, the applicant has committed to implement the technical requirements of Appendix R to 10 CFR Part 50.
Based on this commitment, we conclude that the Wm. H. Zimmer Nuclear Power Station fire protection program will meet the intent of all the requirements Zimmer SSER #2 9-1
. 43 of Appendix R to 10 CFR Part 50 when the committed modifications have been com-pleted, meets the guidelines of Appendix A to BTP ASB 9.5-1, ar'd meets the
_jy f 9 requirements of General Design Criterion 3, Appendix A, 10 CFR Dart 50, and, therefore, is acceptable.
9.6 Diesel Generator Systems In Amendment 127, Revision 76, to the Final Safety Analysis Report, the applicant reported the status of implementation of the guidelines presented in NUREG/
CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability." Full implementation of the staff's position is required prior to startup following c
the first refueling outage and will be made a condi'. ion of the operating license.
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12 RADIATION PROTECTION 12.3 Health Physics Program In Supplement I to NUREG-0528, we stated that Radiation Work Permits (RWPs) would be approved by the Rad-Chem Supervisor.
The applicant has stated that RWPs may also be approved by individuals designated by the Rad-Chem Supervisor.
It is our position that approvals of the radiation protection aspects of RWPs may be given only by individuals who meet the minimum qualifications for radia-tion protection technician or supervisor as specified in ANSI 18.1.
We, there-fore, find it acceptable for the Rad-Chem Supervisor to designate other indi-viduals to approve RWPs.
In Supplement 1 to NUREG-0528, we stated that plant personnel would be provided neutron badges for dosimetry at specified dose rates or doses.
The applicant has r.ow specified that personnel neutron dosimetry will be conducted in accord-ance with the recommendations of Regulatory Guide 8.14 (Rev. 1), " Personnel Neutron Desimeters." We find this acceptable.
Zimmer SSER #2 12-1
i 13 CONDUCT OF OPERATIONS 13.7 Industrial Security Safeguards Contingency Plan The revised pages to the Zimmer Nuclear Power Station Safeguards Contingency Plan submitted by the Cincinnati Gas & Electric Company letter of July 2, 1981 have been reviewed and found acceptable.
Our acceptance is based upon a review of the plant against the requirements of Section 73.55(h) and Appendix C of 10 CFR Part 73.
It was found that this plan adequately contains all of the elements of the material required for a plan to be acceptable.
It has been determined that the plan:
(1) Sets forth decisions and actions satisfying the stated objectives of contingency plans.
(2)
Identified data, criteria, procedures ana mechanisms to carry out these decisions and actions, and (3) Specifies individuals, groups or organizational entities responsible for each such decision and action.
The licensee shall fally implement and maintain in effect all provisions of the Commission-approved Safeguards Contingency Plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p).
The approved Con-tingency Plan, which was submitted pursuant to the authority of 10 CFR 73.40(d) is identified as "Zimmer Nuclear Power Station Safeguards Contingency Plan" dated March 23, 1979 as revised by pages dated May 27, 1981 and July 2,1981.
The Contingency Plan shall be fully implemented upon receipt of the operating license.
General Training and Qualification Plan The page changes to the Guard Training and Qualification Plan for the Zimmer Nuclear Station which was submitted by the Cincinnati Gas & Electric Company letter of May 27, 1981 have been reviewed.
Tt.e plan as revised is now found acceptable.
Our acceptance is based upon a review of the plan against the requirements of Section 73.55(b) and general criteria for security personnel set forth in Appendix B of 10 CFR Part 73.
It was found that the plan contains all of the elements required for it to be acceptable.
The plan:
(1) Establishes a security organization including trained and equipped guards to protect the facility against radiological sabotage, and (2) Outlines the processes by which guards, watchmen, armed response personnel and other members of the security organization will be selected, trained, equipped, tested, qualified and requalified.
Zimmer SSER #2 13-1
The licensee shall follow all provisions of the NRC approved Guard Training and Qualification Plan, including amendments and changes made pursuant to 10 CFR 50.54(p).
The approved Guard Training & Qualification Plan is identi-fied as "Zimmer Nuclear Station Security Training and Qualification Plan" dated March 13, 1981, as revised by pages dated May 27, 1981 and July 2, 1981.
The applicant's security plan is being withheld from public disclosure in accordance with Section 2.790(d) of 10 CFR Part 2.
l Zimmer SSER #2 13-2
22 THI-2 REQUIREt1ENTS 22.2 TMI Action Plan Requirements for Applicants for Operating Licenses I.
Operational Safety I.C.5 Procedures for Feedback of Operating Experience to Plant Staff
^c The applicant reported in Amendment 127, Revision 76, to the Initial Safety Analysis Report that development of procedures for feedback of operating experience to plant staff are under development pending final reorganization of the corporate structure for the nuclear facility.
It is anticipated that these procedures will be completed by January 1982.
We will review and report on these procedures when they are completed.
II.
Siting and Design II.B.2 Plani. Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Postaccident Operation In Supplement 1 to NUREG-0528, we stated that we found the review acceptable subject to receipt of postaccident dose rate maps.
As specified in NUREG-0737, the applicant has submitted postaccident dose rate maps for vital and poten-tially occupied areas and for essential access paths between the occupied areas.
These maps provide the postaccident dose rates for these areas at various times following an accident and are acceptable.
II.B.3 Postaccident Sampling Capability Evaluation In Revisions 73 (May 1981) and 76 (August 1981) to the Final Safety Analysis Report, the applicant submitted the necessary P& ids and the additional informa-tion needed to complete our review.
The applicant states that the postaccident sampling system provides the capability of taking either a liquid or containment atmosphere sample within 45 minutes under degraded core conditions without excessive exposure to personnel.
This meets the II.B.3 requirement for obtaining a sample within one hour.
Our review of the system drawings and of the plant layout drawings showing the areas through which the sample technician must pass in order to take a sample indicates that the 45-minute estimate of sampling time is reasonable.
Provision-nave been made for obtaining atmosphere (air) samples from the drywell, suppression chamber, and secondary containment.
Provisions have been made to obtain liquid samples from the jet pump instrumen-tation nozzles, the residual heat retoval heat exchanger discharge line, the suppression pool, drywell floor drain sump, and the drywell equipment drain sump.
Zimmer SSER #2 22-1
_-____.___.___m_________________.____.___
The sampling system includes provisions for online liquid analysis of pH, conductivity, dissolved hydrogen, and dissolved oxygen.
Capability is provided to obtain diluted or undiluted depressurized degassed liquid samples and also to obtain the associated off gas samples for analytical determination of the gaseous component derived from the liquid samples.
Chloride analysis will be provided by :n offsite contractor to be selected at a later date.
Conclusion The applicant has provided additional information to satisfy the requirements of II.B.3, Postaccident Sampling Capability and has provided c pability for liquid and gaseous sampling and analysis for postaccident conditions which meet the criteria contained in II.B.3, and the staff finds these provisions acceptable.
II.B.7 Analysis of Hydrogen Control II.B.8 Rulemaking Proceeding on Degraded-Core Accidents Discussion and Conclusions The Zimmer application has been informed that we are proceeding with the implementation of the staff's requirement regarding containment inerting.
With regard to Zimmer, our position is that inerting is necessary.
The proposed Interim Rule has been published for public comments, and is nearing completion as an effective rule.
In a letter dated March 24, 1981, the applicant stated the intention to inert the primary containment structure for Zimmer Unit 1.
The applicant has provided a description of the approach to be used to inert the primary contair. ment.
A feed and bleed systerr, directs nitrogen into the drywell or suppression pool using the primary containment and suppression pool purge (VQ) system.
This purge system was reviewed and found acceptable in the staff's Safety Evaluation Report of January 1979 for the Zimmer plant.
On this basis, our current requirements for this item are satisfied.
II.E.4.1 Dedicated Hydrogen Penetrations The staff is awaiting additional information from the applicant in order to complete its confirmatory review of this matter.
We will report our conclusions in a future supplement to NUREG-0528.
II.E.4.2 Containment Isolation Dependability Discussion and Conclusions The following discussion summarizes the applicant's response to this item and our evaluation.
Zimmer SSER #2 22-2
m (1) Diversity in carameters
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The following Table II.E.4.2-1 indicates the signals at Zimmer to initiate valve closure.
Valves which receive two or more of these signals will satisfy the diversity requirement.
2 Table II.E.4.2-1 Signals to initiate valve closure Signal Description
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A Reactor vessel Low (Level 3) Water Level 1
B Reactor vessel Low-(Level 2) Water Level C
Main Stea:nline High Radiation D
Main Steamline High Flow E
Main Steamline Low Pressure F
Main Steamline Tunnel Leak Detection (High Temperature or High A Temperature)
G Shutdown Cooling Reactor Pressure Hign
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Condenser Vacuum Low K
Reactor Water Clean High Differential Flow L
Drywell Pressure High M
Reactor Building Exhaust Plenum High Radiation N
RWCU Leak Detection (High Temperature, High a Temperature or Differential Flow)
P High Steamline Pressure Q
Low Dilution Flow or High Leakage Flow R
RHR/RCIC Combined Steamline High Differential Pressure (High Flow)
S RHR Equipment Area Leak Detection (High Temperature or High a Temperature)
T RHR Shutdown Cooling Flow High 1
U Reactor Vessel Low (Level 1) Water Level Zimmer SSER #2 22-3
Table II.E.4.2-1 (continued)
Signal Description V
RCIC System (1) Steam Tunnel Leak Detection (High Temperature or High Temperature)
(2) RCIC Equipment Area Leak Detection (High Temperature or High a Temperature)
(3) High Steam Flow (4)
Low Steamline Pressure (5) High Turbine Exhaust Pressure W
High Temperature at Outlet of Cleanup Non-Regenerative Heat Exchanger or Standby Liquid Control System Actuated X
Close Through Electrical Interlocks with Other Valves or Pump Motors Z
Refueling Floor Exhaust Radiation - High RM Remote Man,ual From Control Room Zimmer SSER #2 22-4
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Report on the Zimmer Plant of January 1979 and found them to be acceptable.
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(2) Essential and Nonessential Systems The applicant has performed an evaluation of essential and nonessential
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systems for the purposes of containment isolation and identified these Tf, systems for the purpose of containment isolation by penetration number in
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The containment isolation system is designed to prevent the release of radioactive material to the environs after an accident, while ensuring that systems important for postaccident mitigation are operatioral.
Isolation is provided on the following levels:
(a) Nonessential systems
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These systems are not required for postaccident mitigation and are 7.
isolated automatically upon receipt of one or more of the signals in Table II.E.4.2-1.
The majority of the nonessential lines receive at M
least two of the signals listed in Table II.E.4.2-1 and therefore,
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meet the diversity requirement.
Those nonessential lines that do not meet the diversity requirements and justification for finding the proposed isolation provisions acceptable are discussed below:
e (i) Service air, cycled condensate - these lines are provided with 7 -
locked closed valves; y,
(ii) Hydrogen gas control valve and RHR heat exchanger A valves, RCIP pump suction valve, RCIC minimum flow valve, RHR steam
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(IE51F076) valve - these valves are normally closed.
(iii) Recirculation seal purge supply - these lines are equipped with
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check valves to prevent flow out of the containment.
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(iv) HPCS, CPCS, RHR relief valves and relief vents - these valves are normally closed and actuate on high pressure; after the
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pressure transient the valves reclose.
(v) CRD return - isolation is provided by series connection of two j
check valves and a remote manual valve.
(vi) TIP system - these valves are normally cicsed except when the TlP probe is inserted.
The ball valve closes when the probe exits, and the shear valve is closed if the probe fails to
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exit.
(vii) Standby Liquid Control System - isolation is provided by a series of check valves.
(viii) Feedwater system - isolation is provided by check valves inside and outside containment.
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(ix) RCIC discharge valves - isolation is provided by a normally closed valve and a series of check valves.
(x) RCIC vacuum pump discharge and turbine exhaust - isolation is j*~
provided by a check valve and a remote manual valve connected j,,,
in series.
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Since these lines are vital to mitigate the consequences of an accident, the basis upon which these lines are designed is found to be comm n-i surate with the safety importance of isolating the lines.
They include:
(i) Lines that are provided with remote manual isolation capability
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to permit isolation from the main control room, if necessary, including the low pressure coolant spray suction, high pressure coolant spray suction, residual heat removal suction (low pressure coolant injection), reactor building closed cooling water, RHR drywell spray, RHR minimum flow lines and HPCS
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(ii) Lines that are provided with check valves to prevent reverse
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flow in case of a line break in addition to valves that have 9
remote manual isolation capability, including:
low pressure
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coolant injection, high pressure coolant spray, residual heat
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1 (iii) Lines that are provided with automatic isolation signals in n- %(
addition to remote manual capability, including:
residual heat removal wet well sprays "A" and "B", residual heat removal "A"
,.p heat exchanger condenser, low pressure core spray minimum flow
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lines and test lines, residual heat removal, shutdewn suction
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and return lines.
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J[I (iv) Other lines that have special provisions that are not described above, including the control rod drive hydraulic insert and withdrawal lines (manual valves).
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(3) Isolation of nonessential systems i
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As discussed previously, all nonessential systems that provide a possible path out of the primary containment were found to be either isolated by
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diverse isolation signals, by check valves which would prevent flow out
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of the containment, by locked closed valves, by manual valves which are normally closed during reactor operation, or as in the case of instrument
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lines by closed piping systems.
In the case of small diameter instrument lines which penetrate the containment at Zimmer, the design meets the
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2 Reactor Containment," by using manual shutoff valves, and excess flow check
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valves on instrument connected to the primary system and automatic isolation
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Zimmer SSER #2 22-6
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l In conclusion, each line penetrating the containment has been reviewed by the applicant to ensure that:
(a) isolation of the line was based on its need to be in service postaccident; and (b) each containment isolation
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valve received the proper isolation signal.
(4) Resetting of Containment Isolation Signals
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The applicant indicated that its review revealed several cases in which primary containment isolation is removed by the resetting of a contain-ment isolation signal.
The applicant stated that the e" trol of these valves will be modified so that each individual valve remains closed when the isolation logic is reset until the control switch is operated to open a particular valve.
(5) Containment Setpoint Pressure The applicant indicated that the containment setpoint pressure that initiates containment isolation is set to the minimum value compatible with normal operating conditions.
The containment isolation setpoint pressure for Zimmer is 1.69 psig (drywell pressure).
The applicant also indicated i
that under normal operating conditions fluctuations in the atmospheric barometric pressure as well as heat inputs from such sources as pumps can result in containment pressure increases on the order of 1 psi.
Consequently, the applicant feels that the isolation setpoint of 1.69 psig provides adequate margin above the maximum expected operating pressure, and that this margin has proved to be a suitable value to minimize the possibility
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of spurious containment isolation.
At the same time it is felt that such a low value provides a very sensitive and positive means of detecting and protecting against breaks and leaks in the reactor coolant system.
We have reviewed the applicant's containment setpoint pressure value and justification for this value and find it acceptable.
(6) Purge Valves The staf f's Safety Evaluation Report of January 1979 on the Zimmer plant included a review of the design of the containment purge system based upon the cirteria specified in BTP CSB 64, " Containment Purging During Normal Plant Operation." The purge system satisfied GDC 54 and 56 and was found acceptable.
m This acceptance is in accordance with the position on purge valve opercbility criteria as contained in the TMI Action Plan.
(7) Closure of Purge and Vent Valves on High Radiation Signal The containment purge and vent isolation valves must close on high radiation signal.
The applicant indicated that the purge and vent isola-tion valves receive signals to close on reactor building exhaust plenum high radiation (M) and refueling floor exhaust radiation high (Z) in addition to high drywell pressure (L) and low reactor vessel water level (B).
These signals satisfy the staff's criteria.
Zimmer SSER #2 22-7
Based on our review, we conclude that the applicant is in compliance with the requirements for containment isolation dependability as required by Item II.E.4.2.
II.F.1 Additional Accident-Monitoring Instrumentation ATTACHMENT 1, Ncble Gas Effluent Monitoring Requirements NUREG-0737 (November 1980)Section II.F.1-1 establishes criteria for the monitoring of radioactive noble gas effluents which could potentially occur as the result of accidents at nuclear power plants.
Evaluation In Revision 73 to the Final Safety Analysis Report, the applicant submitted details of his provisions to meet the requirements of Section II.F.1-1.
Monitoring will be provided by two noble gas effluent monitors for the main plant exhaust stack and by one monitor for the standby gas treatment system vent stack.
During normal operation, exhaust gases from the main stack will be monitored by an Eberline SPING-3 normal range noble gas effluent m7nitor; the standby gas treatment system (SGTS) vent stack is not used during normal operation.
Eberline AXM-1 accident range monitors are being provided for both the main plant stack and the SGTS vent stack; the AXM-1 monitors will cover the range from approximately 10 4 uCi/cc up to 10s uCi/cc (Xe-133).
1he SPING-3 and AXM-1 overlap in the area from 10 4 uCi/cc to 103 uCi/cc.
Since the Zimmer main stack services all the plant building ventilation exhaust systems (except the SGTS) which may potentially contain radioactive gases as the result of an accident, there is no requirement for additional accident range monitors within the building ventilation syst af the separate plant buildings.
The accident range AXM-1 monitors are designed t lass 1E requirements and it is anticipated that the monitors will be fm 4alified to IEEE 323-1974 when vendor type testing is completed.
NUREG-0737 #a.ification requirements specify only that the instruments shall perform the intended function in the environment to which they will be exposed during accidents.
Qualification to IEEE 323-1974 is not required by NUREG-0737 but such qualification is implied by Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," with an implementation date or June 1983; qualification to IEEE 323-1974 will preclude subsequent changes to meet Regulatory Guide 1.97.
Conclusion In Revision 73 to the Final Safety Analysis Report, the applicant provided information to satisfy the requirements for II.F.1-1, Noble Gas Effluent Monitoring.
The applicant has provided extended range noble gas effluent monitors with an upper range of 105 uCi/cc.
Separate monitors are provided for the main plant exhaust stack and for the Standby Gas Treatment System vent stack.
Zimmer SSER #2 22-8
T :.;
' ; g. ~
J The monitors meet or exceed the criteria contained in NUREG-0737,Section II.F.1-1, and the staff finds the monitoring provisions to be acceptable.
'L ATTACHMENT 2, Sampling and Analysis of Plant Effluents Requirements e
i NUREG-0737 (November 1980)Section II.F.1-2 establishes criteria for the sampling and analysis of radiciodines and particulates in plant effluents 1.
which could potentially occur as the result of accidents at nuclear power F"* N "
P.-.
plants.
.. \\
Evaluation In Revision 73 to the Final Safety Analysis Report, the applicant has submitted details of his provisions to meet the requirements of Section II.F.1-2.
Sampling for normal operational levels of radioiodines and particulates will J'
be accomplished us hg the particulate filters and radioiodine adsorber compo-D
- 1. -
nents of the Eberline SPING-3 monitor to be installed in the main plant exhaust stack.
The SPING-3 has the installed capacity for collecting and simultaneously
- n.. = V i detecting and measuring radioactive particulates and radioiodines in plant effluents at concentrations of from approximately 10 7 uCi/cc to approximately 10 3 uCi/cc.
n?'
U N-For accident-level particulate and radioicdine concentrations in plant efflu-c ants, the applicant is installing Eberline AXM-1 monitors in both the main plant 3
s'ack and the standby gas treatment system (SGTS) vent.
In addition to its cM no;1e gas monitoring function, the AXM-1 is designed to accommodate particulate
. i - n.,
a d radioiodine samples of up to 30 minutes' sample accumulation at an effluent p'
concentration of 100 uCi/cc.
The AXM-1 has provisions for " quick disconnect" of the accumulated sample; the sampler's lead shield, weighing about 100 pounds, T
is removed with the sample, and both are transported to the analytical facility for measurement and nuclide identification.
A replacement shield and sampling 1.'
- 4 assembly can be installed for additional sampling capability.
p-t 9:
Conclusion 7 } ;-
In Revision 73 to the Final Safety Analysis Report, the applicant provided 1,.'
~
information to satisfy the requirements for NUREG-0737, II.F.1-2, Sampling i
s and Analysis of Plant Effluents.
The applicant has provided the capability a
for sampling and analysis of particulate and radioiodine samples of plant
,. 4,
effluents at concentrations up to 100 uCi/cc for 30 minutes.
Sampling t-capability is provided for both the main plant stack and the SGTS vent.
y
. - +
The sampling and analytical capability provided by the applicant meets or
.-$If exceeds the criteria contained in Section II.F.1-2 and the staff finds the nonitoring provisions to be acceptable.
-._ n z..
ATTACHMENT 3, Containment High-Range Radiation Monitor
-( f[
The applicant has not submitted sufficiert final design information on the installation of the containment high-range radiation monitor for the staff to Q-completc its review. We will rcport on this confirmatory matter in a future
.14.
supplement to NUREG-0528.
7x su e
Zimmer SSER #2 22-9 f: y ;,
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m_____________
I
ATTACHMENT 5, Containment Water Level Monitor Discussion and Conclusions Zimmer will have two additional channels for measuring suppression pool water level.
These instruments will measure water level over a 21 f t. 4 in. range from approximately 5 feet above low water level to 16 f t. 4 in. below low water level.
This extends the low end of the range to approximately 1 foot below the centerline of the ECCS pump suction.
One channel will be continuously displayed on an indicating recorder in the control room.
The signal from the other channel will be supplied to the Technical Support Center computer and will be recorded in the computer system as well as displayed on demand on CTRs in the control room and the Technical Support Center.
The monitors will meet the minimum accuracy of +5% of the monitoring range, which we judge to be acceptable.
We conclude, therefore. that Zimmer complies with the provisions of Item II.F.1,.
II.F.2 Instrumentation for Detection of Inadequate Core Cooling The applicant has adopted the Licensing Review Group and BWR Owners' Group position that incore thermocouples are not required to monitor the approach to inadequate core cooling.
We will report resolution of this matter in a future supplement to NUREG-0528.
II.K.1 IE Bulletins on Measures to Mitigate Small-Break LCCAs and Loss-of-Feedwater Accidents Final resolution of Item 5 and Item 22 of this matter is required prior to fuel loading and will become a condition of the operating license.
(See NUREG-0528 Supplement 1.)
II.K.3 Final Recommendations of Bulletin; and Orders Task Force II.K.3.13 RCIC Automatic Restart Discussion and Conclusion Initially the applicant had proposed a delay of installation to the first refueling outage based upon the forecast of equipment availability.
We required further justification for this delay or a commitment for installation consistent with the schedule in NUREG-0737 (four months prior to the opcrating license).
The applicant in Final Safety Analysis Report Amendment No. 127 dated August 31, 1981 committed to an earnest effort for the installation consistent with qualified equipment availability. With this commitment, we conclude that the applicant will meet the requirements of this item.
II.K.3.15 Modify Break Detection Logic to Prevent Spurious Isolation of the RCIC Systems Discussion and Conclusion Even though the applicant has agreed to modify the break detection logic, initially the applicant had not addressed the schedule requirements of this item as specified by NUREG-0737.
Zimmer SSER #2 22-10
.3
~
v The applicant, in Final Safety Analysis Report Amendment No. 127 dated August 31, 12 1981, committed to an earnest effort for the installation of modifications c.
consistent with qualified equipment availability. With this commitment we conclude that the applicant will meet the requirements of this item.
fj II.K.3.18 Modification of ADS Logic Discussion and Conclusions The applicant is a participant in the BWR Owners Group study on this item.
The applicant has chosen Option 2 of BWR Owner's Group recommendation.
The t.,,
second option is to eliminate the high drywell pressure trip from the current logic sequence.
The automatic depressurization system (ADS) sequence would then be activated on low reactor water level only.
The remainder of the sequence remains unchanged.
The effect of high drywell pressure on other f
safety systems, such as reactor scram and the emergency core cooling system (ECCS) that initiates on high drywell pressure, is unchanged.
Modifications are to be implemented consistent with the schedule in NUREG-0737, i.e., to
- k J 4
install during the first refueling outage.
e II.K.3.24 Confirm Adequacy of Space Cooling for Reactor Core Isolation Cooling
. ,. [
and High Pressure Coolant, Injection Systems c
Long-term operation of the RCIC and HPCI systems may require space cooling to J.
maintain the pump room temperatures within allowable limits. We required (NUREG-0737) the applicant to verify the acceptability of the consequences of a complete loss of alternating current (AC) power.
As specified in NUREG-0737, vs the RCIC and HPCI systems should be designed to withstand a complete loss of i,.
offsite AC power to their support systems including coolers for at least two hours.
The Zimmer HPCS (rather than HPCI) and RCIC systems are designed to safety-related systems, and as such, will be designed to operate independently of offsite power.
These two systems will be serviced by a safety-related cooling system, i
which, in turn, will be independent of offsite power.
Since the HPCS and RCIC and their support cooling system will not be affected by a loss of offsite power, we conclude that the requirements of TMI Task Action Plan Item II.K.3.24 are met.
II.K.3.27 Provide Common Reference Level for Vessel Level Instrumentation In order to comply with the staf f's position on this matter, in Final Safety Analysis Report Amendment 127, Revision 76, the applicant agreed to modify the vessel level instrumentation to provide a common reference level in accordance c.
with human factors recommendations.
The modification will be completed prior
.i e*
to fuel load.
The staff will review the modification and report on this matter in a future supplement to NUREG-0528.
l II.K.3.44 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure Discussion and Conclusions
[
In a letter dated May 1, 1981, from J. D. Flynn (CG&E) to H. Denton (NRC), the applicant has provided information discussing an evaluation performed by the
- ~
Zimmer SSER #2 22-11 s
BWR Owners Group.
The applicant has stated that the study results show that adequate core cooling is maintaineo for any transient with the worst single failure.
The bounding event for Zimmer was stated to be the loss of feedwater transient with concurrent failure of the high pressure emergency core cooling system.
In this case, the core always remained covered.
The applicant also referenced studies involving a stuck open safety / relief valve in addition to the worst transient and worst single failure.
The results indicated that the core remained covered and adequate core cooling was available during the course of the transient.
The applicant has committed to verify that Zimmer is bounded by the General Electric Company generic analysis. We required the applicant to provide a summary of operator actions required to accomplish hot shutdown during the worst case event.
In final Safety Analysis Report Amendment No. 127 dated August 31, 1981, the applicant gave a summary of operator actions by referring to Station Emergency Procedures and Guidelines.
Staff review of Station Emergency Procedures is discussed under I.C.1 and I.C.8 of section 22 in NUREG-0528, Supplement 1.
Based on the results of the Owners Group Study and their applicability to Zimmer, we find the applicant's response acceptable for this item.
III.
Emergency Preparations and Radiation Protection III.A.1.1 Upgrade Emergency Preparedness III.A.2 Improving Licensee Emergency Preparation - Long Term In Final Safety Analysis Report Amendment 127, Revision 76, the applicant provided a status report on items III.A.1.1 and III.A.2.
The staff is continuing to work with the applicant toward final review and resolution of these matters and will report its conclusions in a future supplement to NUREG-0528.
III.D.3.4 Control Room Habitability See subsection 2.2.1 of this supplement for the status of final resolution of this matter.
Zimmer SSER #2 22-12
23 CONCLUSIONS Based on our evaluation of the application as set forth in NUREG-0528 and in its supplements (Supplements 1 and 2), we are able to affirm the conclusions presented in Section 22.0 of NUREG-0528.
1 Zimmer SSER #2 23-1
i APPENDIX A CHRONOLOGY (Continued from NUREG-0528, Supplement 1)
(Major Safety Review Correspondence).
March 30, 1981 Letter (Generic) to applicant (81-08) concerning BVR scram discharge.
March 30, 1981 Letter from applicant forwarding information requested by NRC.
April 1, 1981 Letter from applicant transmitting the 1980 Annual Financial Report.
April 2, 1981 Letter from applicant estimating monthly cost to maintain facility in inactive status while awaiting full power license.
April 3, 1981 Letter from applicant transmitting Amendment 116 to OL application consisting of Revision 17 to fire protection evaluation.
April 9, 1981 Letter from applicant disputing the NRC caseload forecast panel's projected completion date.
April 9, 1981 Letter to applicant transmitting " Control Room Design Review Audit Report."
April 10, 1981 Letter from applicant transmitting Amendment No. 14 to the
" Mark II Containment Design Assessment Report."
April 14, 1981 Letter from applicant responding to Generic Letter 81-10 concerning upgraded cmergency response facilities.
April 21, 1981 Letter to applicant transmitting NUREG-0487, Suppl #2.
April 22, 1981 Letter from applicant responding to NUREG-0737.
April 22, 1981 Letter from applicant transmitting a reanalysis of required transients using approved ODYN code.
April 23, 1981 Letter to applicant requesting additional information to complete reactor system review.
April 23, 1981 Letter to applicant transmitting comments on safeguards contingency plans.
April 24, 1981 Letter from applicant transmitting chapter mods for incor-poration into May revision of FSAR.
Zimmer SSER #2 A-1
=
_v dY APPENDIX A (continued)
-~
April 24, 1981 Letter to applicant transmitting comments regarding the emergency plan.
April 24, 1981 Letter to applicant transmitting " Safety Concerns Associated with Pipe Break in BWR Scram System."
y, April 24, 1981 Letter to applicant requesting additional information regarding guard training and qualification plan.
April 27, 1981 Letter from applicant transmitting information corcerning preservice inspection and testing of snubbers.
=:
April 28, 1981 Letter from applicant transmitting information to be included in Chapter 1 of design assessment report in 11ay 1981.
April 28, 1981 Letter to applicant requesting information on instrumentation and control.
April 28, 1981 Letter from applicant transmitting Revision 12 to the 1
industrial security plan.
-J April 29, 1981 Letter from applicant forwarding " Definition of Associated Cables," and " Comparison and justification of ZPS Design VS Reg Guide 1.75."
April 29, 1981 Letter from applicant transmitting responses to NRC JE Question 251.4.
April 29, 1981 Letter from applicant transmitting supplemental information to previous response to NRC question regarding soil-structure interaction and Reg Guide 1.60.
April 29, 1981 Letter from applicant transmitting supplementary material regarding fuel transfer shielding.
April 29, 1981 Letter from applicant responding to letter requesting additional information to Table 3.2.1.
~
April 29, 1981 Letter from applicant confirming telecon regarding NUREG-0737.
April 30, 1981 Letter from applicant transmitting revisions to Task Action Plan Items I.A.1.3 and I.C.2 to Section 4.7, Item F and Section 5.2.
April 30, 1981 Letter from applicant transmitting supplemental SER informa-tion regarding in plant radiation monitoring and containment pressure boundary fracture toughness.
April 30, 1981 Letter from applicant transmitting a marked-up copy of FSAR Page Q423.7-12 indicating NRC request for modification re snubber testing.
c Zitamer SSER #2 A-2
APPENDIX A (continued)
April 30, 1981 Letter from applicant advising that they will participate in safety relief valve surveillance program.
April 30, 1981 Letter from applicant transmitting a marked-up FSAR Page 13.1-8.
April 30, 1981 Letter from applicant transmitting additional information to be incorporated into Chapter 8 regarding degraded grid voltage.
April 30, 1981 Letter from applicant forwarding revised position regarding Item II.K.3.15 of NUREG-0737.
April 30, 1981 Letter from applicant forwarding a response to letter regarding the Reactor Systems Branch issues.
May 1, 1981 Letter from applicant concerning Restart of Core Spray Systems.
May 1, 1981 Letter from applicant transmitting revised writeups of NUREG-0737.
May 1, 1981 Letter from applicant concerning revisions to NUREG-0737.
May 1, 1981 Letter from applicant concerning Criteria for Radiological Plans.
May 1, 1981 Letter from applicant concerning Silt Prevention.
May 4, 1981 Letter to applicant concerning Qualification of Inspection, Examination, and Testing and Audit Personnel (Generic Letter 81-01).
May 4, 1981 Letter from applicant concerning NRC Question 212.79.
May 5, 1981 Letter to applicant concerning Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on January 29, 1981 (Generic Letter No. 81-22).
May 7, 1981 Letter from applicant concerning Comments on NUREG-0619.
May 8, 1981 Letter from applicant concerning Action Plan Item II.D.3.
May 11, 1981 Letter from applicant concerning revised pages to FSAR Table 3.2-1.
May 13, 1981 Letter from applicant concerning Control of Heavy Loads.
May 13, 1981 Letter to applicant concerning a special team appraisals of the emergency preparedness program.
Zimmer SSER #2 A-3
APPENDIX A (continued)
May 15, 1981 Letter from applicant concerning Reactor Feedwater Copper Content.
May 15, 1981 Letter from applicant concerning Soil Structure Interaction.
May 19, 1981 Letter from applicant concerning Task Action Item II.F.1.
May 28, 1981 Letter from applicant concerning Mark II Containment Design Assessment Report (DAR), Amendment 15.
May 22, 1981 Letter from applicant transmitting Amendment 120 consisting of Revision 73 to the FSAR.
May 27, 1981 Letter from applicant transmitting Amendment 121 consisting of Revision 13 to the Industrial Security Plan.
June 2-5, 1981 Representatives from NRC, EG&E & CG&E meet at the site to review and audit the equipment qualification program.
(Summary issued)
June 3, 1981 Letter to applicant concerning a meeting July 7-10, 1981 to discuss Environmental Qualification of Safety-Related Electrical Equipment.
June 4, 1981 Letter to applicant transmitting Supplement No. I to the Zimmer SER (20 copies).
June 10, 1981 Letter to applicant concerning Potential Loss of NPSH in BWR ECCS Suction Piping.
June 11, 1981 Letter to applicant concerning implementation of revised Guard Training and Qualification Plan.
June 12, 1981 Letter to applicant issuing Supplement 1 to the Zimmer SER (20 copies).
June 12, 1981 Letter from applicant transmitting Amendment 122 to the FSAR consisting of Revision 74 to the FSAR.
June 12, 1981 Letter from applicant transmitting the ZPS-1 Emergency Response Facility Conceptual Design.
June 15, 1981 Letter from applicant concerning pipe breaks in the BWR Scram System.
June 10, 1981 Letter to applicant concerning NUREG-0763.
June 15, 1981 Letter to applicant (generic Letter 81-25) Upgraded Emergency Plans.
Zimmer SSER #2 A-4
APPENDIX A (continued)
June 24, 1981 Letter from applicant concerning supplemental information in response to NRC letter of December 22, 1980 regarding control of heavy loads.
June 26, 1981 Letter to applicant concerning inservice testing of pamps and valves.
June 30, 1981 Letter from applicant concerning Mark II Containment Design Assessment Report (DAR), Amendment 16.
July 1, 1981 Letter to applicant concerning Steam Generator Overfill (generic letter 81-16).
July 1, 1981 Letter from applicant concerning Preliminary Safety / Relief Valve (S/RV) Operability Test Results.
July 2, 1981 Letter from applicant concerning Revisions to Service List.
July 2, 1981 Letter from applicant concerning Amendment 124 - Revision 14 to the Industrial Security Plan.
July 7, 1981 Letter from applicant concerning BWR Coolant Pressure Boundary Piping (Generic Task A-42).
July 9, 1981 Letter to applicant concerning (Generic Letter 81-27)
Privacy and Proprietary Material in Emergency Plans.
July 6, 1981 Letter to applicant (Generic 81-23A) concerning INP0 Evaluation Reports.
July 14-16, 1981 Representatives from NRC & CG&E meet at the Zimmer Plant site in Moscow, Ohio to discuss the inservice testing program for pumps and valves (summary issued)
July 20, 1981 Letter from applicant concerning fuel load date.
July 20, 1981 Letter to applicant concerning Prompt Notification in the Event of an Emergency.
July 22, 1981 Letter to applicant concerning Final Safety Evaluation Supplement for Zimmer Operating License Review.
July 30, 1981 Letter from applicant concerning BWR Scram Discharge System Safety Evaluation.
July 31, 1981 Letter from applicant transmitting Amendment 125 - Revision 75 to the Zimmer FSAR.
July 31, 1981 Letter to applicant concerning Contingency Plan.
Zimmer SSER #2 A-5
,~
APPENDIX A (continued)
July 31, 1981 Letter to applicant concerning Steam Generator Overfill' (generic letter 81-28) Formerly issued July 1,1981 as t*
August 4, 1981 Letter to applicant concerning Guard Training and Qualifica-'
tion Plan.
August 7, 1981 Letter to applicant concerning Simulator Examinations (Generic Letter 81-29).
August 7, 1981 Letter from applicant transmitting the Certificatt of Service for FSAR Amendment 125.
s -
August 7, 1981 Letter from applicant conceraing Steam Generator Overfill.
August 21, 1981 Letter to applicant concerning Fission Gas Release Analysis.
August 24, 1981 Letter from applicant concerning Response to NRC Generi,c Letter 81 Emergency Procedures & Training for Station Blackout Events.
l August 24, 1981 Letter to applicant concerning Long Term Operability of Deep Draft Pumps.
September 11, 1981 Letter from applicant transmitting a Certificate of Service for FSAR Amentment 127.
September 1, 1981 Letter from applicant concerning a typographical error in their letter, dated August 18, 1981.
September 2, 1981 Letter from applicant concerning the fuel load date for Zimmer (July 1982).
September 24, 1981 Letter to applicant transmitting NUREG-0808, " Mark II Containment Program Load Evaluation and Acceptance Criteria."
Zimmer SSER #2 A-6
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'i
d' APPENDIX D ERRATA (For Supplement 1) y
'PAGE 2-4
[ Lines 8 & 10] Change "20.06g/R2" to "s0.06g/R2n Change "20.02g" to "*0.02g"
^ 2-4
[Line 7 from bottom] Add to end of sentence "which includes about 4 ft wave action."
'\\ -12
[Line 8 from bottom] Change " miles" to " mils" 5
22-36
[Line 1] Change "F" to "<"
i 22-49
[Line 16] Change " material" to "wetwall" 22-50
[ Lines 8 & 9] Delete "(A copy... Attachment 1.)"
22-51
[ Lines 38 & 39] Change " valve" to "value"
?
22-53
[ Lines 1 & 2] Change " valve" to "value" 22-69
[Line 7] Change first "outside" to "inside" C-9
[Next to last line] Change " reacting" to " reaching" C-9
[Last paragraph] Delete "The staff's review.
. materials toughness,"
and replace with "The staff has completed its review of the Zimmer reactor vessel materials toughness and" D-3
[Line 4-8] Change "N05-1.5" to "s0.5-1.5" D-3
[Line 4-14] Change "first" to " third" and second "J" to ">"
D-3
[Line 15-10] Change second "F" to "<"
"WK" "AK" D-3
[Line 15-14] Change second K
K 2
E-6
[Line 33] Change "0.30 gpm/ft2" to "0.30 gpm/ft n Zimmer SSER #2 0-1
L I-APPENDIX E BIBLIOGRAPHY 2.3 Meteorology 1.
NUREG-0654, Rev. 1 " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" U.S. NRC/ FEMA, Washington, D.C., 11/80.
2.
Regulatory Guide 1.145 " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" U.S. NRC Washington, D.C., 8/79.
4.2 Fuel System Design 1.
D. F. Ross (NRL) letter to G. Sherwood (GE), November 23, 1976.
2.
R. O. Meyer, C. E. Beyer, and J. C. Voglewede, " Fission Gas Release Froin Fuel at High Burnup," NUREG-0418, March 1978.
3.
G. Sherwood (GE letter to D. F. Ross (NRC), December 22, 1976.
4.
R. B. Elkins, " Fuel Rod Prepressurization - Amendment 1," General Electric Report NE00-23786-1 May 1978.
5.
R. E. Engel (GE) letter to T. A. Ippolito (NRC), May 6, 1981.
6.
R. E. Engel (GE) letter to T. A. Ippolito (NRC), May 28, 1981.
7.
L S. Rubenstein (NRC) memorandum for T. M. Novak (NRC), " Extension of General Electric Emergency Core Cooling System Performance Limits,"
June 25, 1981.
l I
Zimmer SSER #2 E-1
'l i
APPENDIX F NRC STAFF CONTRIBUTORS AND CONSULTANTS This Safety Evaluation Report is a product of the NRC staff and their consultants.
The following NRC staff members were principal contributors to this report.
A list'of consultants follows the list of staff.
Name Title Review Branch Joseph Levine Meteorologist Accident Evaluation Chen P. Tan Structural Engineer Structural Engineering Owen 0. Rothberg Structural Engineer Structural Engineering Tsun-Yung Chang Sr. Mechanical Engineer Equipment Qualification Marylee Martin Slosson Equipment Qualification Equipment Qualification Engineer Michael Tokar Sr. Reactor Fuels Engineer Core Performance (Materials)
Walter L. Brooks Sr. Reactor Physicist Core Performance Max Bolotsky Materials Engineer Materials Engineering David E. Smith Materials Engineer Materials Engineering Anthony J. Cappucci Jr. Mechanical Engineer Mechanical Engineering Lawrence C. Ruth Containment Systems Containment Systems Engineer George Thomas Reactor Engineer Reactor Systems Bill M. Morris Stetion Lerder, Technical Instrumentation and Review Section, CRBR Control Systems Program Office Narinder K. Trehan Reactor Systems Engineer Power Systems (Electrical)
-Fred Clemenson Sr. Auxiliary Systems Auxiliary Systems Engineer
~ Gregory A. Harrison Fire Protection Engineer Chemical Engineering Charles S. Hinson Health Physicist Radiological Assessment George W. McCorkle Chief, Physical Security Physical Security Licensing Branch Licensing Phillip G. Stoddart Nuclear Engineer Effluent Treatment Systems Consultants Idaho National Engineering Laboratory (INEL, EG&G)
Zimmer SSER #2
- F-1
I l
)
I l
NRC r OaM 335 1, REPORT NUMBER (Ass <sned or DDCJ U.S. NUCLEAR REGULATORY COMM.SSION NUREG-0528 BIBLIOGRAPHIC DATA SHEET Supplement No. 2 4 TlTLE AND SU8itTLE (Add Votume No., st appropressel 2.(Leave bim k/
Safety Evaluation Report related to the operation of 1
' Wm.- H. Zimer Nuclear Power Station, Unit No.1
- 3. RECIPIENT *S ACCESSION NO.
I AUlHORISI
- 5. DATE HEPORT COMPLETED l YEAR
' MONTH October 1981 9 PERFORMING ORGAN 17ATION NAME AND MAJLING ADDRESS (Include le Codel DATE REPORT ISSUED MONTH l YEAR Office of Nuclear Reactor Regulation October 1981 U.S. Nuclear Regulatory Commission
- s.,1,,ve u,, ;
dashington, DC 20555
- 8. (Leave Nanki
- 17. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include le Code /
10 PROJECT / TASK / WORK UNIT NO.
Same as 9, above.
- 11. CONTRACT NO.
13 TYPE OF REPORT PE RIOD COVE RE D (trictusere dates)
Safety Evaluation Report - Technical June 1M1 - October 1981
'S StJPPt_EMENTARY NOTES
- 14. (Leave atma)
Docket No. 50-358 IG ABSTH ACT 200 words or lesst
.The Safety Evaluation Report for the Zimmer Nuclear Power Station, Unit I was issued in January 1979. At the time of issuance there were two outstandinra issues.
Supplement No.1, issued in June 1981 discusses the resolution of these issues and the concerns of the Acvisory Committee on Reactor Safeguards, which issued a favorable report on March 13, 1979. This Supplement discusses subsequent outstanding issues since June 1981. The review of this plant will continue until the unit is operating.
The Ziuner Station is located in Washington Township, Clermont County, Ohio.
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