ML17258A653
| ML17258A653 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/09/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TASK-***, TASK-RR LSO5-82-03-054, LSO5-82-3-54, NUDOCS 8203170618 | |
| Download: ML17258A653 (27) | |
Text
if' Y
March 9, 1982 Docket No. 50-244 LS05 03-054 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 5 Electric Corp.
89 East Avenue Rochester, New York 14649
Dear Mr. Maier:
SUBJECT:
INTEGRATED ASSESSMENT MEETING AT GINNA REt HiVpi3 MAP, 17 l982~
g ggg kmliTNTKB"NM
[ggiITL':MBZfffbi TiOC Enclosed is the listing of all topics with identified differences from licensing criteria (Enclosure
- 1) and a brief summary of the actual identified differences (Enclosure 2).
These are the topics which we plan to discuss with your staff during the March 10 - 12, 1982 meeting.
As previously agreed we plan to limit the discussion of III-2, Wind and Tornado and III-4.A, Tornado Missiles to your latest submittals.
Enclosures:
As stated Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No.
5 Division of Licensing 5~of
/i os~ hsF.(
boa'.
g]a/Gy, cc w/enclosures:
See next page 8203i70hi8 82030~ l PDR ADOCK 05000244 P~DR OFFICE/
SURNAME/
DATEf
.KP,P.........
AWanj:bl 3 Pg/82
~P!
EJ
~
0 ~ ~ ~ 0 ~ ~ ~ ~ J ~
~
S EP.B WRussel 1
3/$ (82 GC Q lla
...3/'.Xl.8.2.......
/.82.
AD,'S DL.......
GL nas 3/8 /82
~ 0
~ ~
~ 0
~ ~ ~ ~ ~ ~ ~ ~ 0 ~
NRC FORM 318 {10-80) NRCM 0240 OFFlClAL RECORD COPY USGPO: 1981~960
/y, c.
gr C, Ginna Docket Ho.
5D-k4A Rev. 2/8/82 Mr. John E. Maier CC Harry H. Yoigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N.
W.
Suite 1100 Washington, D. C.. 20036 Mr. Michael Slade 12 Trailwood Circle Rochester, Hew York 14618 Ezra Bialik
'Assistant Attorney General Environmental Protection Bureau New York Sta'te Department of Law 2 World Trade Center Hew York, New York 10047 Resident Ihspector R. E; Ginna Plant c/o U. S.
HRC 1503 Lake Road
- Ontario, Hew York 14519 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 U. S. Environmental. Protection Agency Region II Office
. ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 Supervisor of the Tawn of Ontario 107 Ridge Road West
- Ontario, New York 14519 Dr.
Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory. Commission Washington, D. C.
20555.
Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.
C.
20555
~.:-
GI NNA rncioSure I TOPICS WHICH DO NOT MEET CURRENT CRITERIA OR E UIVALENT TOPIC NO.
II-3.B II-3.8.1 II-3.C II-4.D III-1 III-2 III-3.A III-3.C III-4.A III-4.C III-5. A III-5.B III-6 III-7.A III-7.B III-S.A
'.Y-5 IV-6
- Y-10.A TITLE Exclusion Area Authority and Control Severe Weather Phenomena Flooding Potential and Protection Requirements Capability of Operating Plant to Cope With Design Basis Flooding Conditions Safety-related Water Supply I.Ultimate Heat Sink (UHS}]
Stability of Slopes Classification of Structures, Systems and Components Wind and Tornado Loadings Effects of High Water Lev'el on Structures Inservice Inspection of Water Control Structures Tornado Missiles Internally Generated Missiles Effects of Pipe Break on Structures, Systems and Components Inside Containment Pipe Break Outside Containment Seismic Design Considerations Inservice Inspection, Including Prestressed Concrete Containment with Either Grouted or Ungrouted Tendons Design
- Codes, Design Criteria, and Loading Combinations Loose Parts Monitoring and Core Barrel Vibration Program Reactor Coolant Pressure Boundary
{RCPB) Leakage, Detection Reactor Yessel Integrity RHR Heat Exchanger Tube Failures' TOPIC NO.
V-1 0. B VI-2. D TITLE I
RHR =Reliabi'1 ity" Mass and Energy Release for Possible Pipe Break Inside Containment YI-3 VI-4 (Systems)
(Electrical)
VI-7.8.
VIII-3. B IX-3 IX-5 Containment Pressure and Heat Removal.Capability Containment Isolation'S F Swi tchover from Injection to Recircul ati on Mode DC Power System Bus Voltage Monitoring and Annunciation'tation Service and Cooling Water Systems Ventilation Systems IX-6 Fire Protection
GINNA TOPIC NO.
II-1.A Difference Summar
'ITLE Exclusion Area Authority. and.Control The Exclusion Area Boundary (EAB) has been
- changed, as submitted by RGEE letter dated June 26, 1981.
This change is potentially significant enough,to warrant a change to the Ginna Technical Specifications to in-corporate the new exclusion area boundary map.
TOPIC NO.
II,-2.A Difference Summar TITLE Severe Weather Phenomena 10 CFR 50 (GDC 2), requires that the plant be designed to withstand the effects of natural phenomena.
The combined snow load for structural
'apability assessment at Ginna is 100 lb/ft2.
Various safety related buildings were not constructed to withstand such a load.
1 TOPI Ci'O.
TITLE II-3.B Difference Summar Flooding Potential and Protection Requirements 10 CFR 50 (GDC 2),
as implemented by Standard Review Plan (SRP) 2.4.10 and Regulatory Guide (RG) 1.59 prescribes that the plant have adequate flood protection'.
The water, levels produced by a Probable Maximum Flood (PMF) on Deer Creek would cause water to pond 8'bove grade'on the north side.
TOPIC NO.
TITLE II-3.8.1 Difference Summar Capability of Operating Plants to Cope With Design Basis Flooding Conditions 10 CFR 50 (GDC 2),
as implemented by SRP 2.4.10 prescribes that the.
plant have adequate flood protection.
The plant has no existing plans or technical specifications '(TS).that relate to flooding from external sources.
TOPIC NO.
II-3. C Difference Summar TITLE Safety-Related Water Su*pply t.Ultimate Heat. Sink (UHS)j 10 CFR 50 (GDC 2), as implemented by SRP 2.4.10 prescribes that the plant have adequate flood protection.
An occurence of the Probable Maximum Flood on Deer Creek would inundate both the service water and circulating water pumps.
TOPIC NO.
II-4.D Difference Summar TITLE Stability of Slopes, 10 CFR 50 (GDC 2), as. implemented by SRP 2.5.5 prescribes that the
~
plant be adequately protected against failure of natural'r man-made slopes.
The failure of the onsite 'slopes would effect safety-related structures.
TOP.I C'O.
TITLE III-1 Difference Summar guality Group Classification of Structures, Systems'nd Components 10 CFR 50 (GDC I), as implemented by Regulatory Guide 1.26, requires that structures, systems and components important to safety be designed, fabri-
- cated, erected and tested to quality standards commensurate with the importance of the safety functions to be performed.
The following.are deviations from current requirements:
1)
Category C joints of vessels which would currently be classified by ASME Section III, 1977 as Class 2 or 3 but built to ASME Section III, 1965 as Class C do not satisfy current radiography requirements 2)
The regenerative heat exchanger and the excess letdown heat exchanger do not satisfy current radiography r'equirements because they are Class A
. vessels built 'to Class' requirements.
TOPIC NO.
III-2 Difference Summar Mind and Tornado Loading 10 CFR '50 (GDC 2), as implemented by Standard Review Plan Sections 3.3.1 and 3.3.2 and Regulatory Guide 1.76 and 1.117 requires that the plant be designed to withstand the effects of natural phenomena
~
The existing design and construction of structures important to safety for wind and tornado loadings does not meet current licensing criteria of remaining within Standard Review Plan stress limits.
TOPIC NO.
III-3.A Difference Summa TITLE Effects of High Water on Structures,;,
as implemented by SRP 2.4.12 prescribes that the plant Pe designed for groundwater problems.
External lateral loads were not considered in the design of the containment because of an external ring wall around the containment, however there is no pro-vision:to assure the groundwater loads will always be non-existent.
Also, seismic Category I structures, systems and equipment were not designed for flood due to Deer Creek.
TOPIC NO.
III-3.C TITLE Inservice Inspection of Water Control Structures Difference Summar 10 CFR 50 (GDC 45), as implemented by Regulatory Guide 1.127 requires that the cooling water.system shall be designed to permit appropriate periodic inspection of important components to ensure the integrity and capability of the system. 'he following are necessary for compliance with the intent of Regulatory Guide 1. 127:
1)
The inspection program now underway at Ginna should be formalized so that standard report forms are submitted by competent and qualified inspectors to be reviewed by qualified engineers.
2)
The licensee should develop a checklist for discharge canal inspections, including their frequency.
3)
The Deer Creek basin should be formally recognized as a water control structure and inspected accordingly on.an.annual: basis.and follewing I severe rains which cause flooding.
(a)
The Inservice Inspection Program for Deer Creek should be supple-mented by adding:
clogging of culverts. by debris, slump conditions,.
soil creep, and bed load movement.
(b)
The wooded area downstream of the Visitors Center should be cleaned
~ out to initially establish adequate water conveyance during floods and a baseline for future inspection and maintenance.
4)
The Licensee should compile a comprehensive file of engineering drawings for safety-related water control structures to establish immediate post-'onstruction conditions.
5)
The'routine inspection frequency is acceptable, but's'pecial. i'nspections also must be performed after extreme events such as floods and seiches which may jeopardize the integrity of water control structures.
The formal inspection program to be'initiated at the R.
E.
Ginna Plant should incorporate such special inspections.
6)
The Licensee should develop a f'ormal inspection program for water control structures that will result in the development of a comprehensive file of appropriate inspection reports.
7)
The Licensee's monitoring program to be developed for the revetment must be approved by the NRC.
TOPIC NO.
TITLE Tornado Nissil es Difference Sumiar 10 CFR 50 (GDC 2),
as implemented by. Regulatory Guide 1.117 prescribes struc-
- tures, systems and components that should be designed to withstand the effects
.of a tornado,,
including tornado missiles, without loss of capability to perform their'afety function.
The following safety-related structures, systems and components were found to not be protected from tornado missiles:
1)
Component Cooling System 2)
Refueling Water Storage Tank
3)
Electrical Busses 14, 17 and 18.
4)
Service Water System 5)
Diesel Generators and their Fuel Supply 6)
Relay Room 7)
Hain Steam Line B
8)
The Top'Surface of the Spent Fuel Pool is open and, therefore, the internals are exposed 9)
Boric Acid Tanks TOPIC HO.
TITLE III-AC Internally Generated Missiles Difference Summar
)
as implemented by SRP Section 3.5.1.1 and 3.5.1.2 prescribes that structures, systems and components important to safety be designed to withstand the affects of internally generated missiles inside or outside of containment.
The following are deviations or open items that have been identified:
1)
An evaluation of the piping a'nd components associated with the ECCS accumulators with respect to miss4le generation and protection has not been completed.
2)
An evaluation of the effects of missile generation along the CVCS"let-down line inside containment has not been completed.
3)
An evaluation of the potential effects of an unrestrained valve opera-tor associated with the steam generator blowdown system on safety re-'ated components and systems has not been completed.
4)
The refueling water storage tank is inadequately protected from missiles.
TOPIC NO.
III-5.'
Difference Summar TITLE Effects of 'Pipe Break on Structures, Systems and Components Inside Containment
~
r 4 10 CFR 50 (GDC 4), as implemented by SRP 3.6.2 prescribes that structures,
'systems and components important to safety be designed against the dynam-ic and environmental effects of postulated pipe ruptures.
The following are deviations from review guidelines that have been identified:
The-first open item was concerned with the general.assumptions. of this topic assessment was that a check valve in an incoming line would prevent primary system blowdown in the event of a pipe break upstream of the valve.
This is true provided the check valve closes.
- Adequate assurance must be demonstrated that these normally open check valves will fulfill.
their assumed isolation function.
2)
For the "A" accumulator line a mechanistic evaluation was performed.
The stresses in this line were all below the criteria, so breaks were postu-lated at terminal ends and at the loop compartment where no adverse interactions would occur.
The second point is located just on the reactor side of the (normally locked open) motor-operated valve.
At this location no adverse pipe whip inter-actions will occur.
If remedial measures to provide this protection can be shown to be impractical, fracture mechanics evaluations can be performed to'stablish that conditions that could lead to a double-ended rupture do not exist as discussed in the guidance provided in the Attachment to Enclosure 3.
The effect of a break in the two inch accumulator level taps on nearby instr~ment circuits is still under review by the licensee.
3) 4)
5)
For the pressurizer surge line, since some. jets could affect safety-related equipment, analyses similar to those described in item 4 above should be provided.
For the letdown line, licensee evaluation of the'ffects on cables and cable trays is continuing.
Ad'equate protection for instrumentation. should be provided.
The situation for the steam generator blowdown lines is similar to item 7
.for the instrumentation.
'With respect to the fan coolers,'this size break is not limiting with respect to containment pressure/temperature reduction capability.
The containment spray system would be available for containment cooling.
As.'for item 7 above, final resolution will occur after the effects on the cable trays are evaluated.
6)
Pipe breaks were not postulated in the primary loop-on the basis-of %he work done under TAP A-2.
We concur with this approach.
However,'-the SEP branch intends to evaluate the effects on safety-related equipment of jet loads resulting from the crack sizes associated with these analyses...
TOPIC HO.
III-5.B TITLE Pipe Break Outside Containment Difference Summar 10 CFR 50 (GDC 4), as implemented by SRP 3.6.1, 3.6.2,.BTP MEB 3-1 and BTP ASB 3-1, requires in part that structures, systems and components important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures.
The following are deviations from review guidelines that have been identified:
1)
Because high and moderate energy line breaks in the screen house could damage the power supplies to all servicewater pumps, the licensee'must provide protection for these power supplies in accordance with Standard geview Plan 3.6.1 consistent with the service water system modifications which must be performed in connection with other ongoing SEP reviews and the fire protection review.
l provide protection for these power supplies in accordance with Standard
. Review Plan 3.6.1 consistent-with the service water -system:modifications
'which must'be performed in'onnection with'othe} ongoing SEP reviews-and '
the fire protection review.
TOPIC NO.
III-6 Difference Summar TITLE Seismic Design Consideration The requirements of 10 CFR 50 (GDC 2) and'10 CFR 100, Appendix A'as imple-mented by Regulatory Guides 1.26, 1.29, 1.60, 1.61, 1..92, 1.122 and SRP 2.5,-
3.7, 3.8,
- 3. 9, 3.10 prescribe structures, systems and components that should be designed to withstand the effects of.a postulated earthquake with'out loss of capacity to perform their safety function.
The evaluation results are summarized below:
1)
The structures were found capable of withstanding, the, postulated seismic event except two sets of steel bracings located in auxiliary and turbine building for which modifications are required.
2)
ESW Pump Operability 'is an open item.
3)
RWS Tank and other safety related tanks are open items.
4)
Control room electrical panel structural integrity is an open item.
5)
The functional integrity of electrical equipment is being evaluated by, testing through SEP Owners Group program.
6)
(}ualification of electrical cable trays is being evaluated by testing through SEP Owners Group program.
TOPIC NO.
III-7. A TITLE Inservice Inspection, Including"Prestressed Concrete Containments with Either Grouted or Ungrouted Tendons Difference Summar Regulatory Guide 1.35, Revision 2 as interpreted in "the Standa'rd Technical Specifications requires that the licensee have an inspection program that will detect any structurally significant deterioration of Category I structures in order that the structures will be capable of performing their necessary func-tions.
The following are deviations between the tendon surveillance program at Ginna and Regulatory Guide 1.35, Revision 2:
- ,+
c 1)
The acceptable lift-offrequirement does not meet current criteria because the existing Technical Specification at.Ginna require that the average of the 14 tendon stresses be greater than a 'value constant. with time..:--Cur- ",.,"'"
~. rent criteria requires that, each tendon fall within acceptance limits 'that vary with time.
2)
Tendons which are found to be unacceptable are not handled as required in Section 7 of Regulatory Guide 1.35, Revision 2.
- 3) 'egulatory Guide 1.35, Revision 2 requires inspections and mechanical
.tests'e performed on one unstressed wire per tendon per inspection.
4)
Ginna should include in its inspection report wire breakage and filler grease.
TOPIC NO.
III-7.B TITLE l ~ lq
~
Design
- Codes, Design Criteria and Loading Combinat'ions
'ifference Summar 10 CFR 50 (GDC 1, 2 and 4), as interpreted by. Standard Review Plan 3.8, required
'the plant to be designed and contructed to various design codes, criteria, loads and lo'ad combinations.
The following are areas where differences exist between the plant-design and current licensing criteria.
1)
Code changes have been identified in the following structural elements:
{See table next page from SEP Topic III-7.B issued 12/30/81.)
~ 2)
Load and Load Combinations 3)
A thermal discontinuity exist in the liner plate at the point where the insulation stops.
This will cause high thermal stresses in the'iner during pos'tulated LOCA temperatures and could result in the liner buckling and-fai]ing.
. TOPIC NO.
Difference Summar TITLE Loose Parts'onitoring and Core Barrel Uibration Program The requirements of 10 CFR 50 (GDC 13),
as implemented by Reguilatory Guide 1.133, Revision 1, and SRP Section 4.4 prescribe a loose parts monitoring program for the primary system of light-water-cooled reactors.
Ginna does not have a loose parts monitoring program that meets. the criteria of Regulatory Guide 1.133.
0
-1.0-Structural Elements to be Code Chan e Affectin These Elements Examined
'ew Code,
. 'ld-Cod'e-Members Desi ned to rate AISC 1980 AISC 1963 in an Inelastic Re ime Spacing of lateral bracing 2 9
~ 2.8 Short Brackets and Corbels having a shear span-to-depth ratio of unity or less Shear Walls used as a
primary load-carrying member ACI 349-76 ll.13 ACI 349-76
- 11. 16 ACI 318-63 ACI 318-63 Precast Concrete Structural
- Elements, where shear is not a member of diagonal tension ACI 349-76
- 11. 15 ACI 318-63 Concrete Regions Sub 'ect to High Tem eratures ACI 349-76 ACI 318-63 Titne-dependent and po'sition-dependent temperature variations Appendix A Columns with S liced Reinforcement subject to stress reversals; fy in compression to 1/2 fy in tension Steel Embedments used to transmit load to concrete ACI 349-76 7.10.3 ACI 349-76 Appendix B ACI 318-63 805 ACI 318-63 Containment and Other
- Elements, transmi'ttina In-. l.ane shear BaPV Code Section III, Div. 2, 1960 CC-3421.5 ACI 318"63 Region of shell carrying concentrated forces normal to the shell surface (see case study 13 for details)
BaPV Code,Section III, Div. 2i 1980 CC-3421.6 ACI 318-63 1707
Structural Elements to be
~
Examined Beams a.
Composite Beams New Code AISC 1980 Old Code AISC 1963 Code Chan e Affectin These Elements 1.'hear connectors in composite beams 1.11.4 1.11.4 2.
Composite beams or girders vith formed steel deck 1.11.5 b.
Hybrid Girders Stress in flange
- 1. 10. 6
'1. 10.6 Com ression Elements AISC 1980 AISC 1963 With vidth-to-thickness ratio higher than speci-fied in 1.9.1.2 1.9.1.2 and Appendix C
1.9.1 Tension Meeker s AISC 1980 AISC 1963 When load is transmitted by bolts or rivets 1.14.2.2 Connections a.
Beam.ends vith top flange.
, coped, if subject to shear AISC 1980 1.5. 1.2.2 AISC 1963 b.
Connections carrying moment or restrained member connection 1'.15.5.2 1.15.5.3
- 1. 15.5.4
~Double dash
(
) indicates that no provisions vere provided in the older'ode.
TOPIC NO.
I TITLE I
Reactor Coolant Pressure. Boundary Leakage Detection Review Criteria; 10 CFR 50 '{GDC 2 and 30),
as implemented by SRP 5.2.5 and Regulatory Guide 1.45 requires the measurement of leakage from the reactor coolant pressure boundary
{RCPB) to the containment and interfacing systems and states de-sign criteria for the systems employed for such.
For systems employed for measurement of leakage from. the RCPB to the con-tainment,'Regulatory Guide 1.45 states that:
- 1) system should be an air-borne particulate radioactivity monitor that is SSE qualified, 2) a minimum..
of two others should be pr esent which are'BE qualified, and 3) all systems
. should have a sensitivity to detect leakage of 1
gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Those employed for measurement of intersystem leakage should.include sensors for things such's radioactivity, flow, level, pressure, temperature, etc.
and be OBE qualified.
All the above systems should 1) have alarms and indica-tors in the main control
- room, 2) be readily testable;and-calibrated, during normal operation, and have their availability in the technical specifications.
Difference Summar The following summarizes the deviations from review guidelines that have been identified:
1)
Although all of the recomended tyPes of leakage detection systems for measurement of leakage from the reactor coolant pressure boundary
'o the containment have been incorporated in the facility, the systems do not meet all of the sensitivity, operability or surveillance criteria.
2)
Information concerning the leakage detection systems for the detection of inter-system reactor coolant pressure boundary leakage is incomplete.
Therefore, we cannot determine the extent to which Regulatory Guide 1; 45 i,s met.
3}
Standard Techni'cal Specification 3/4.4.6 and the corresponding sur-veillance requirements concerning the operability of the reactor coolant pressure boundary to the containment leakage detection systems
'hould be added to the R.
E, Ginna Technical Specifications.
- Also, the current basis for Ginna Technical Specification 3.1.5.*3 and FSAR should be revised to state that the sensitivities of the reactor coolant pressure boundary to containment leakage detection systems.
Information concerning the use of the primary coolant system inventory balance leak rate sensitivity and time required to achieve sensitivity
's incomplete.
Therefore, we cannot determine the contribution of
'his technique to the overall leak detection sensitivity.
'I I
TOPIC NO.
V-6 Difference Summar TITLE Reactor Yessel Integrity 10 CFR 50 (GDC 31), as implemented by 10 CFR Appendix 6 prescribes specific minimum fracture toughness requirements for ferritic materials of pressure retainihg components.
The Ginna vessel materials are expected to fall below current fracture toughn'ess requir ements at about ll to 13 EFPY.
. TOPI C NO.
TITLE Y-10. A RHR Heat Exchanger Tube Failures ifference Summar SRP 9.2.'1 requires that the service water system include the capability for detection and control of radioactive leakage into and out of the system and prevent accidental releases to the environment.
The Service Water System does not have a radiation detector.
TOPIC NO.
Y-10. B Difference Summar TITLE RHR Reliability 10 CFR 50 (GDC 19 and 34),
'as implemented by SRP 5.4.7, BTP RSB 5-1 and Regu-latory Guide 1.139, require that the plant can be taken from normal operating conditions to cold shutdown using only safety-"grade
- systems, assuming a single failure and utilizing either onsite or offs5te power through the use of suit-able procedures.,
The Ginna plant has safety-grade plant systems capable of safe shutdown under these conditions;
- however, the plant'operating procedures rely upon other non-safety grade systems and: do not specify how the cooldown would be accomplished by the operator in the event of failures in 'non-safety grade systems.
TOPIC NO.
VI-2.D-TITLE s<
Mass.and Energy Release for Poss'fble'ipe Break,,
Inside Containment VI-3 Difference Summar Containment Pressure and Heat Removal Capability, "10 CFR 50 (GDC 50),
as implemented by SRP Section 6.2.1, require. that the containment structure and the containment heat removal system be designed so that the structure can accommodate, with sufficient margin, the calcu-
. lated pressure and temperature conditions.resulting from a pipe break.
For the MSLB accident, the peak calculated pressure was 85.8 psia.
This exceeded the containment design pressure by ll PSI.
TOPIC NO.
YI-4 TITLE Containment Isolation Systems
~
Difference Summar 1)
,The isolation valving arrangements do not meet the requirements of 10 CFR 50 (GDC 55 or 56),
as implemented by SRP 6.2.4 from the stan'-
'oint of valve location for penetrations 1, 5, 9, 10, ll, 'l4, 15, 16, 18, 19, 20,'1, 30, 35, 40, 43, 44, and 45.
Also, the licensee should discuss the unique characteristics of the valves closest to the containment to terminate valve shaft or bon'net, seal
- leakage, or the provisions in the plant for control of leakage.
2)
The isolation valving arrangements do not meet the requirements of 10 CFR 50 (GDC 55 or 56) as implemented by SRP 6.2.4 from the stand-point of valve number for penetrations 4, 15, 16, and 39.
Lines 4 and 6,
- however, were found to meet the GDC on some other deffned basis.
3)
The isolation valving'arrangements differ from the explicit.requirements of 10 CFR 50 (GDC 55, 56 and 57) as implemented by SRP 6.2.4 from the standpoint of valve type by using a check valve outside containment for penetrations 15, 17 and 18.
With regard to line 17, the containment isolation provisions were found:.
acceptable on some other defined basis without considering the simple check valves outside containment.
4)
The. isolation provided does not meet the requirements of 10.CFR 50 (GDC 55, 56 and
- 57) as implemented by SRP 6.2.4 from the standpoint of valve actuation for,penetrations 1, 2, 3, 5, 6, 7, 9, 10, ll, 17, 18,'1, 24, 25, 26, 31, 32, 40, 41, and 42.
.The actuation provisions for valves in
. lines 5, 6, and 17 were found to meet the GDC on some other defined basis.
5)
It should be noted that other lines contain open, local manual valves that are identified" as containment isolation. valves.
These valves were ignored-where other identified isolation va1ves were Vound Co...-
satisfy the GDC.
The isolation barriers differ from the explicit requirements of 10 CFR 50 (GDC 55, 56 and
- 57) as implemented by SRP 6.2.4: from the standpoint that blind flanges are used as containment isolation bar-ri,ers for penetrations 12, 46 ahd 47.
.6) 7)
8)
For containment isolation configurations having a blind flange. inside containment the blind flange can be an acceptable isolation barrier in lieu of an isolation valve, if leak testing provisions are made.
The licensee should address this.
It is not known to what extent pipe caps were used on test, vent and drain connections located be-tween isolation valves or in closed systems which constitute an isolation barrier.
Nevertheless, pipe caps are not.,suitable isolation barriers for containment isolation and should be replaced with locked closed manual valves or a blind flange that is leak testable (SRP 6.2.4, Item II.3).
The licensee should address this.
Certain penetrations have been provided with remote manual isolation
- valves, which is acceptable.'owever, provisions should be made to allow the operator in the main control room to know when to isolate f'1 uid systems equipped with remote manual isolation valves (SRP 6.2:4, Item II.11).
Lines 2 and 18 identify a pressure regulator valve as a containment isolation valve.
The licensee should provide additional information regarding the design and performance characteristics of the valve controls.
In essence, the actuation provisions for a valve of this type must satisfy the requirements for an automatic isolation valve.
10 CFR 50 (GDC 57) as implemented by SRP 6.2.4 was.used to judge the acceptability of the isolation provisions for line 31, 32, 41, and 42 (service water system lines) since a closed system was identified in-si'de containment.
The licensee should verify that this portion of the system is of safety grade design to assure that the use of GDC. 57 is appropriate.
- Also, as noted in Item 4, automatic isolation valves should be provided, unless it can be satisfactorily argued that the lines are essential.
- 9) 'he licensee should provide design information on the equipment hatch and personnel airlock, including the leak testability of the hatch gasket and airlock door seals.
Also, test/instrument "lines penetrating the airlock should be 'identified and the containment isolation provis-sions for them should be, justified.
- 56) as implemented in SRP 6.2.4 specifies that automatic isolation valves should,.upon loss of actuating power,,take,
=
'the position that provides greater safety.
The-position of =an'.isola-tion valve for normal and shutdown operating conditions, and'post-accident condition, depends on the fluid system function.
In the event of power failure to a valve operator, the valve. position should be con-..
sistent with the line function.
In this regard, separate power'up-
'lies for isolation valves in series may be required to assure the isolation of non-essential lines.
The licensee should provide the information on the position of isolation valves, whether or"not 'the line is essential and the isolation signals (including parameters sensed to actuate the signals) for each isolation valve Furthermore, the licensee should discuss and justify the acceptability of these isolation valve characteristics.
TOPIC NO.
TITLE YI-7.B ESF Switchover From Injection to Recirculation Mode Difference Summar 1)
Item 19 of SRP Section 6.3 states that the complete sequence of ECCS operation from injection to long term core cooling (recirculation) should be examined to see that a minimum of manual action is required, and that where manual action is needed a sufficient time (greater than 20 minutes is available for the operator to'respond.
The current Ginna proc'edures for switchover from injection to recirculation do not meet current NRC criteria for operator actions.
2)
Branch Technical Positions ISCB 20 has not been satisfied because of the short time (ll minutes) that is available for the operator to detect and correct a failure to follow procedures and his reliance on a single alarm to alert him to such an error.
TOPIC NO.
YIII-.3. B Difference Summar TITLE DC Power System Bus Voltage Monitoring and Annunciation 10 CFR 50.55a (h) as implemented by SRP 8.3.2 and Regulatory Guide 1.47 requires that the dc power system be monitored to the extent that it is shown ready to perform its intended function.
The Ginna control room has no indication of battery current, charger output current, charger output voltage, battery high discharge rate, bus under/over voltage, or battery or charger breaker/fuse status.:
~
~
~
TOPIC NO.
IX-3 Difference Summar TITLE Station Service and Cooling Water Systems'0 CFR 50 (GDC 44).
as implemented by SRP 9.2.1 and SRP 9.2.2 requires a
system to transfer heat -from structures, systems and componets important to safety to an ultimate heat si nk.'he CCW make-up system.is non-seismic.
Also, the technical specifications allow the plant to be operated.with only two out'of four service pumps which, since two pumps are needed to. handle post-acci'dent heat loads, renders the system vulnerable to a single failure.
TOPIC NO.
TITLE Ventilation Systems IX-5
~ '
Difference Summar.
as implemented by Standard Review Plan 9.4.5 requires that
.the plant include a means to suitably control ".the release of radioactive mater-ials i'n gaseous and liquid effluents.
Current criteria requires that the capagility exist to direct ventilation air from areas of low radioactivity to areas of progressively higher radioactivity.
There are two scenarios which'could possibly violate this requirement, both of which occur with the"main exhaust fans shut-down when offsite power is not available and the plant is opera.ing on emer-gency diesel power.
TOPIC NO.
IX-6 TITLE Fire Protection
'Difference Summar 10 CFR 50 (GDC 3),
as implemented by 10 CFR 50.48 and Appendix R requires that structures, systems and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability'nd effect of fires.
Ginna cannot reach cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, as required by Appendix R, in zone ABRH, since a fire there could cause the loss of both RHR pumps.