ML17292B070
| ML17292B070 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/26/1997 |
| From: | Colburn T NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9710060449 | |
| Download: ML17292B070 (47) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 September 26, 1997 go-59 7 LICENSEE:
Washington Public Power Supply System (WPPSS)
FACILITY:
Washington Nuclear Project No.
2 (WNP-2)
SUBJECT:
MEETING
SUMMARY
On August 27,
- 1997, Washington Public Power Supply System (WPPSS or the licensee) representatives met with members of the NRC staff to discuss the licensee's planned WNP-2 Final Safety Analysis Report (FSAR) upgrade program.
Also, in attendance as an observer was a representative from Nebraska Public Power District representing Cooper Nuclear Station.
The licensee provided the staff with a discussion of the scope and progress of the WNP-2 FSAR upgrade.
Details of the upgrade effort were presented to the staff and the licensee acknowledged that the program included the elimination of information the licensee felt was duplicative or contained unnecessary detail.
The licensee believed the final product would comply with Regulatory Guide 1.70 with some exceptions.
The licensee also provided examples of revised information.
The licensee indicated its schedule to complete the effort and provide the NRC with a complete updated FSAR was approximately March 1998.
The staff stated from the outset that it would listen to the licensee's presentation but was not in a position at this time to approve or deny what the licensee was proposing.
The staff has been tasked with providing a
framework to the Commission for considering deletions of information from FSARs by December 1997.
The staff informed the licensee that information taken out of FSARs prior to the issuance of Commission guidance would be done at the licensee's own risk.
The licensee acknowledged the staff's position.
The licensee summarized its position with respect to its program and the meeting adjourned.
97i0060449 970926 PDR ADQCK 05000397 P
A copy of'the licensee's slides is attached.
Also attached is a list of attendees.
Original Signed By Timothy G. Colburn, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-397 Attachments:
1.
Meeting Slides 2.
Attendance List cc w/atts:
See next page DISTRIBUTION
'~Docket File'UBLIC ACRS
- PGwynn, RIV JRHall (w/atts.
1 and 2)
PDIV-2 Reading OGC TColburn FAkstulewicz DISTRIBUTION:
(w/att. 1)
SCollins (SJCl)
WBateman (WHB)
FMiraglia (FJH)
EPeyton (ESP)
RZimmerman (RPZ)
EAdensam (FGA1)
DOCUMENT NAME:
WNP827HT.SUM OFC PDIV-2/PH PDIV-2/LA NRR:PGE NAME TCotburn:ye EPeey n
FAkstulewi DATE 9/2@97 9/8/97 A
R 9/4P/97 A copy of the licensee's slides is attached.
Also'attached is a list of
, attendees.
Docket No. 50-397 Attachments:
1.
Meeting Slides 2.
Attendance List cc w/atts:
See next page Timothy G.
olburn, Senior Project Manager Project Directorate IV-2 Di'vision of Reactor Projects III/IV Office of Nuclear Reactor Regulation
cc w/encls:
Hr. Greg 0. Smith (Hail Drop 927M)
WNP-2 Plant General Manager Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352-0968 Hr. Albert E. Houncer (Mail Drop 1396)
Chief Counsel Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352-0968 Hr. Frederick S. Adair, Chairman Energy Facility Site Evaluation Council P. 0.
Box 43172 Olympia, Washington 98504-3172 Mr. David A. Swank (Mail Drop PE20)
- Manager, Regulatory Affairs Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352-0968 Hr. Paul Inserra (Mail Drop PE20)
- Manager, Licensing Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower 8 Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Chairman Benton County Board of Commissioners P.O.
Box 69
- Prosser, Washington 99350-0190 Hr. Rodney L. Webring (Hai 1 Drop PE08)
Vice President, Operations Support/PIO Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352 Hr. J.
V. Parrish Chief Executive Officer Washington Public Power Supply System P.O.
Box 968 (Mail Drop 1023)
- Richland, Washington 99352-0968 Hr. Scott Boynton. Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O.
Box 69 Richland.
Washington 99352-0968 Mr. Perry D. Robinson, Esq.
Winston 8 Strawn 1400 L Street, N.W.
DC 20005-3502
Attachment 1
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
.FSAR UPGRADE PROGRAM AUGUST 27, 1997 HANDOUTMATERIALUSED AT THE MEETING
NRC/WPPSS Meeting August 27, 1997 0900-1100 (Proposed)
NRC Headquarters on WNP-2 FSAR Upgrade Program Agenda 0900 Introduction 0905 WNP-2 FSAR Upgrade Program Scope and Progress 0910 Compliance withRegulatory Guide 1.70 A. Elimination ofDuplicate Information B. Deletions of Detail C. Archiving of Information D. Archiving by Regulatory Guide Exceptions 1000 Examples ofRevised Information Potential Impacts and Questions 1045 Schedule for WNP-2 Submittal 1050 Summary WPPSS Attendees:
D.W. Coleman - Manager, Regulatory AfFairs (Acting)
P.J. Inserra - Manager, Licensing J.C. Gearhart - FSAR Upgrade Project Manager
Progress WNP-2 FSAR = Over 30 Volumes and over 8000 pages, divided into 176 sections.
58 sections have been evaluated, revised and issued for comment.
18 sections have been reviewed by all reviewers and either approved or approved with speci6c items requiring resolution.
Scheduled work is running about 4 weeks behind due to delays &om the contractor or WNP-2 reviewers.
This has no impact on the projected Upgraded FSAR submittal to the NRC (50.71(e) update in March of 1998).
I ~
l jl
FSAR Content Guidelines
- 1. Is the current content required'? Ifyes, verify and correct as necessary.
Commitment: RG 1.70, R2; SER; Specific Commitment in Correspondence.
Recommended:
RG 1.70, R3 or SRP (Current NUIKG0800).
- 2. Ifthe current "ontent is not required by Commitment or Management agreement on recommendation, consider archiving or deletion.
- 3. Archived material is deleted Rom the FSAR. Itwillno longer be referred to.during future 10CFR50.59 change reviews.
Candidate material for archiving should* meet the following:
- a. The material is not required, by commitment or management decision, and bl. The material is not subject to change by projected future plant, program or procedure changes, or b2. The material is at a level ofdetail that projected changes to it over plant life could not impact tne abilityofrequired SSCs to perform their design function, and
- c. The material must not contain the sole summary ofinformation needed to establish the basis for plant design or operational parameters.
~ The exception to the above is the case where the material is required by commitment, however it is historical in nature. In these cases we may elect to archive the material by taking specific exception to the commitment (e.g. RG 1.70, R2) and documenting the exception in the FSAR. This exception willbe performed using 10CFR50.59 and will generally result in a safety evaluation.
- 4. Archived vs. Deleted:
Archived material willbe removed &om the FSAR, however in its place the FSAR willhave a briefdescription ofwhat the material was and where it can be found. Deleted material willbe removed without any remaining indication. In both cases the License Document Change Notice, Screening for Licensing Basis Impact and, if necessary, the Safety Evaluation willclearly indicate the basis for removal ofthe information &om the FSAR.
WNP-2 USAR evision 5/22/97 (draft)
/Q p Q 5~~~m Dcscwipklo~
The main turbine is a tandem-compound unit, consisting of one double-flow high pressure turbine and three double-flow low pressure turbines (Figure 10.3-1), running at 1800 rpm with 47 in.eh last-stage blades.
Exhaust steam fmm the high pressure nubine passes thmugh twu moisture separator/reheaters (two stage reheat) before entering the low pressure turbine inlets.
The exhaust steam from the three low pressure turbines is condensed in the main condenser.
i ree hase c cle 25 V l ri I
ni ted at 1 2 ihd I
RVA s
te i
i and o ted o mi imize e hazard from fi rex lo ions di A
dix F s
10.2-2
which electrical load may be increased or decreased with and without reactor control rod motion or steam bypass);
and design codes to be applied.
10.2.2 Descriotion A description of the turbine-generator equipment, including moisture separation, use of extraction steam for feedwater heating, and control functions that could influence operation of the reactor coolant system, should be provided as well as drawings.
The turbine-generator-overspeed control system should be described in detail, including redundancy of
- controls, type of control utilized, overspeed setpoints, and valve actions required for each setpoint.
10.2.3 Turbine Disk Inte rit The failure of a turbine disk or rotor might produce a high-energy missile that could damage a safety-related component.
This section should provide information to demonstrate the integrity of turbine disks and rotors.
10.2.3.1 Materials Selection.
This section should include materials specifications, fabrication history, and chemical analysis of the disk and rotor forgings.
Particular attention should be paid to items affecting fracture toughness and metallurgical stability.
The mechanical properties of the disk material such as yield strength and fracture toughness should be listed.
The methods of obtaining these properties should be described.
10.2.3.2 Fracture Toughness.
The criteria used to ensure protection against brittle failure of low-pressure turbine disks should be described.
include detailed information on ductile-brittle transition temperature (NDT or PATT) and minimum operating temperature.
H' fracture mechanics approach is used, the analytical method and the key assumptions made should be described.
10.2.3.3 Hi h-Temperature Pro erties.
Provide the stress-rupture properties of the high-pressure rotor material and describe the method for obtaining these properties.
10.2.3.4 Turbine Disk Desi n.
Provide the following design infor-mation for low-pressure disks and high-pressure rotors:
1.
The tangential stress due to centrifugal loads, interference fit, and thermal gradie.rs at the bore region at noz a
speed and design overspeed.
2.
The maximum tangential and radial stresses and their locat."on.
10.2.3.5 Preservice
- ssoection.
Descr"'be the preservice inspection procedures and acceptance
-"'teria to demonstrate the initial integrity of the disks and rotors.
10-2
WNP-2 USAR
~ 4
++~1~ ~+D FEEDufh7GR QV$7em NRT&2pgt S Revision 5/9/97 (draft) 10.3.6.1 re To ne Impact tests in accordance with the size limitations specified in ASME Code Section III, Class 1, are performed on all ASME Code Section III, Class 1, main steam and feedwater materials, as well as Class 2 main steam system materials for all pressure retaining ferritic steel parts.
The tests are conducted at a temperature of45'F or lower in accordance with NB or NC-2310 of the Summer 1972 or Winter 1973 Addendum of ASME Code Section III, as applicable.
10.3.6.2 a
lecti n and Fabrica 'o All m teri u
in the steam and feedwater vstems are included in A ndix I to Section m f the ASME Boiler and Pressure Ve el
&PV e
The requirements for welding the main steam piping from the reactor to the turbine generator are in accordance with ASME Section III, 1971 Fdition through the Winter 1973 Addenda. The welding requirements for other steam and feedwater piping are in accordance with ANSI B31.1, October 1973 (see Section 3.2).
- 10. 3.5 Water Chemistry (PWR)
The effect oz the water chemistry chosen on the radioactive iodine partition coezzicients in the steam generator and air ejector should be discussed.
Detailed information on the secondary-side water chemistry, includ-ing methods of treatment for corrosion control and proposed specification limits should beprovided.
Discuss methods for monitoring and controlling water chemistry.
10.3.6 Steam and Feedwater S stem Materials This section should provide the information indicated below on the materials used for CLass 2 and 3 components.
- 10. 3. 6.1 Fracture Tou hness.
Indi,cate the degree of compliance with the test aathods and acceptance criteria of the ASME Code Section IIIin Articles NC-2300 and
~i 2300 for fracture toughness for ferritic materials used in Class 2 and 3 components.
10.3.6.2 Materials Selection and Fabri. cation.
Information on materials selection and fabrication methods used for Class 2 and 3 compo-nents should include the following:
1.
For any material not included in Appendix I to Section IIIof the ASME Code, provide the data ca1led for under Appendix IV for approval of new materials.
The use of such materials should be justified.
2.
For austenitic stainless steel components, the degree to which the recommendations of Regulatory Guide 1.44, "Control of the Use of Sensitized Stainless Steel;" Regulatory Guide 1.36, "Nonmetallic Thermal Insulation for Austenitic Stainless Steel;" and Regulatory Guide 1.31, "Control of Stainless Steel Welding," are followed should be indicated.
Justification zor any deviations from the procedures shown in these guides should be provided.
3.
For all Class 2 and 3 components, information on the cleaning and handling oz such components should be provided.
The degree to which the recommendations of Regulatory Guide l. 37, "Quality Assurance Require-ments zor Cleaning of Fluid Systems and Assoc'ated Components of Water-Cooled Nuclear ower Plants,"
and AQSI N45.2.1-73, "Cleaning of Fluid Systems and Associated Components for Nuclear Plants," are followed should be 'ndicated.
Justification
=or any deviations rom the position in these documents should be pqovided.
4.
Ind'cate wnether the preheat temperatures used for welding low-allov steel are in accordance with Regulatory Guide 1.50, "Control oz Preheat Temperature zor Welding oz Low-Alloy Steel."
Justizication for any ceviations
=rom the procedures snown 'n th's guide should be provided.
10-4
WNP-2 B 2 AMENDMENT NO.
51 August 1996 control switches or the electric motor driven operation of the air compressors on U and 1B diesel generators are on the local diesel engine cont=ol board.
These control switches permit on-auto-off operation.
A selector switch permits selection of either compressor unction as the primarv pressu ization compressor.
d,res'
'egqcrxaP
~
pressure switches in either air receiver bank automaticall sta the selected compressor when. the receiver pressure If the selected compressor fails to operate cannot hol system pressure, a separate low pressure alarm switch is rov'ded for each bank of air receivers and is set to alarm at on a local panel and in the main control room.
When the ec er pressure decays to a lower pressure, the back-up air compressor starts.
The HPCS Starting Air system has two separate air supply trains.
One supplied by a diesel driven compressor and the other by an electric motor dr'ven compressor.
The compressor discharge piping is cross connected.
Both air receivers charge if either compressor operates.
A check valve on each receiver inlet isolates one train from the other.
The compressors are controlled automatically by pressure switches on their associated air receiver.
The compressor's low pressure setpoint ensures that the compressor starts to maintain the air receiver pressure at.a sufficient amount to start the engine the required. number of times.
The diesel driven compressor shuts down prior to clearing the opposite train receiver low pressure alarm.
The air receivers are equipped with safety relief valves set at, the receiver design pressure.
The major system components are located adjacent to the diesel generator skid.
For each diesel generators (1A and 1B), two separate air cooled compressors discharge through common piping to two banks of four 32 cu. ft. air receivers which are connected in parallel.
Each bank of air receivers has the capability of a minimum of five engine starts.
Each bank is connected through separate piping to a pair of air start motors on each engine.
9.5-29
environmental desiga conditions, and the plans by whica additional oil
'may be procured, if required.
9.5.5 Diesel Generator Coolin Rater S stem The design bases for the cooling water system should be provf.ded and should include a discussion of the ability to meet the single-failure criterion.
A description of the cooliag water system, including drawings 9
should be provided.
9.6.6 Diesel Generator gtartin 9 stem The design bases for the starting system, including required system
- capacity, should be provided and should include a discussion of the abiLity to meet the single-failure cziterioa.
A description of the starting syststb intluding drauings, should be provided 9.5.7 Diesel Generator Lubrication S stem The design bases for the lubrication system should be provided aad should include a discussion of the abi1ity to meet the single-fai1ure criterion.
A description of the lubrication system, iacluding drawings 9
should be provided.
9.5.8 Diesel Generator Combustion Air Intake and Exhaust S stem 9.5.8.1 Desi Bases.
This section should provide the design bases for the diesel generator combustion air intake and exhaust
- system, includ-ing the bases for protection from the effects of natural phenomena, mis-
- siles, and contaminatiag substances as related to the faci1ity site,
- systems, and equipment and the capability of the system to meet minimum safety requirements assuming a single failure.
Seismic and quality group classifications should be provf.ded in Section 3.2 aad referenced in this section.
9.5. 8.2 S stem Descri tion.
A complete description of the system should be provided, including system drawings detailing component.
redun-
- dancy, where required, and showing the location of system equipment, in the facility aad the relationship to site systems or components that could affect the system.
9.5.8.3 Safety Evaluation.
Analyses should be provided to deaaa-strate that the minimum quantity and oxygen content requirements for intake combustion air will be met considering such effects as rec'zcula-tion or diesel combustion products, accidental re ease oz gases stored in the vicinity of the'diesel iatakes, restziction of inlet airflow, intake of such particulates as airborne dust, and 'ow baromet ic pressure.
The results of failure mode and effects analyses to ensure minimum requirements should be provided.
If system degradation could result from the consequences oi missiles or failures of nigh-or moderate-energy 9-16
I WNP-2 USAR Revision 5/9/97 (draft) 10.3 M IN TEAM PPLY SY TEM 10.3.1 DESIGN BASES The main steam supply system is designed for the following conditions:
a'.
Deliver steam from the reactor to the turbine generator from warmup to 105% of rated load, b.
Provide steam for the second-stage reheaters and steam-jet air ejectors, c.
Bypass steam to the main condenser during startup and in the event steam requirements of the turbine generator are less than that produced by the reac'tor, d.
Provide steam to the gland seal steam evaporator during startup, low load operation, and shutdown, e.
Provide steam'to drive reactor feedwater pumps during startup and low load operation~ GIld Provide steam to the offgas preheaters.
The design pressure and temperature of the main steam piping is 1250 psig and 575'F.
The main steam lines are designed to include accesses to permit inservice inspection and testing (refer to Sections 5.2.4 and 6.6).
~* idLd',i i ~dl 2-d.-,
Vabt~ item 2. Nuclear Bc'der System, and item 43. Power Conversion System.
The environmental design bases for the main steam supply system are contained in Section 3.11.
10.3.2 SYSTEM DESCRIPTION dd ddld d
i d
i di d,'-l d~
. The main steam line piping consists of four 30-in. (26-in. in reactor building) I.D. lines extending irom the reactor pressure vessel to the main steam header located upstream oi the.zwine s;op and control valves.
This header placement ensures assufes-a positive means oi bypassing steam via the turbine bypass svstem during transient conditions and startup.
~
tl
~
~
4
~
4h&~Branch lines from the main steam line provide the steam requirements for the 10.3-:
10.2;3. 6 Inservice Ins ection.
The inservice inspection program for the turbine assembly and the inspections and tests of the main steam stop and control valves and the reheat stop and intercept valves should be described.
10.2.4 Evaluation An evaluation of the turbine-generator and xelated steam handling equipment should be provided.
This evaluation should include a summary discussion of the anticipated operating concentrations of radioactive contandnants in the system, radiation levels associated with the turbine components and resulting shielding requirements, and the extent of access contxol necessary based on radiation levels and shielding provided.
Details of the radiological evaluation should be provided in Chaptexs 11 and 12.
10.3 Main Steam Su 1
S stem The design bases for the main steam line piping from the steam generator, in the case of an indirect cycle plant, or from the outboard isolation valve, in the case of a direct cycle plant, should be provided and should include performance requirements, environmental design bases, inservice inspection requirements, and design codes to be applied.
Capability of the system to dump steam to the atmosphere, if required, should be discussed.
Steam lines to and from feedwater turbines shou1d be included in the descziptions.
A description of the main steam line piping, including drawings showing interconnected piping, should be provided.
10.3.3 Evaluation An evaluation of the design of the main steam line piping should be provided and should include an analysis of the ability to withstand limiting environmental and accident conditions and provisions for permit-ting insezvice inspections to be performed.
Appropriate references shou1d be made to seismic classifications in Chapter 3 and to the analy-sis of postulated high-energy line failure in Section 3.6.
10.3.4 Ins ection and Testin Requirements The inspection and testing requirements of the main steam line piping should be described.
Describe the proposed requirements for preoperational and insezvice inspection. of steam line isolation valves or reference other sections of the SAR where these are described.
10-3
WNP-2 USAR REVISION 7/22/97 (draft) 2.4.6 PROBABLE TSUNAMIFLOODING The location ofthe WNP-2 site is in south~tral Washington and it is not adjacent to any coastal area. Itis not, therefore, vulnerable to tsunami flooding.
2.4.7 ICE EFFECTS Historically, the Columbia River has never experienced complete flow stoppage or significant flooding due to ice blockage.
Periodic ice blocking has caused reduced flows and limited flooding for only relatively short periods of time.
2.4-13
WNP-2 USAR REVISION 7/22/97 (draft) e potential for ice blockage or the combination ofblockage and flooding behind ice dams is so low as to be considered insignificant.
In any event, ice flooding willnot effect the capability to shut down the reactor in a safe and orderly manner.
Also, the daily fluctuating stage of the river at the intake location will discourage formation ofsheet ice as well as ice jams. Ice flows, should they occur, wiH normally pass over intake structure due to relatively high winter discharge in the river.
dditional histori di
'on n ice form tion and e ts vi ed n e F AR u h Amendment 52.
2.4.8 COOLING WATER CANALS AND RESERVOIRS
'12 I
4 reactor building i
. ~e-Qvo spray ponds located southeast of the cooldown and are the ultimate heat sink for normal reactor mergency cooling.
The spray ponds are
~41I 22 2.27 4
Id
'422
'scussed in Sections nt nd e
m fl 2.4-14
RG. ).70, R 2.
2.4. 6.3 Source Tsunami Vave Hei ht.
Provide estimates of the mmdmum tsunami wave height possible at each major local generating source considered and the maximum offshoxe deepwater tsunami height from distant generators.
Discuss the controlling generators for both locally and distantly generated tsunami.
2.4.6.4 Tsunami Hei ht Offshore.
Provide estimates of the tsunami height in deep water adjacent to the site, before bottom effects appreci-ably alter wave configuration, for each major generator.
2.4.6.5 H dro ra h and Harbor or Breakwater Influences on Tsun-uni.
Present the routing of the controlling tsunami, including breaking wave formation, bore formation, and any resonance effects (natural frequencies and successive wave effects) that result in the estimate of the maximum tsunami runup on each pertinent safety-reIated facility.
This should include a discussion both of the analysis used to txanslate tsunami waves from offshore generator locations,'or in deep water, to the site and of antecedent conditions.
Pxovide, where possible, verification of the techniques and coefficients used by reconstituting tsunami of record.
2.4.6.6 Effects on Safet -Related Facilities. Discuss the effects of the contx'oiling tsumuai on safety-reIated facilities and discuss the design criteria for the tsunami protection to be provided.
2.4.7 Ice Effects Describe potential icinp effects and design criteria for protecting safety-related facilities from the most severe ice jam flood, wind-driven ice ridges, or other ice-produced effects and forces that are reasonably possible and could affect safety-related faciIities with respect to adjacent
- streams, lakes, etc., for both high and low water levels.
Include the location and proximity of such facilities to the ice-generating mechanisms'escribe the regional ice and ice jam formation history with respect to water bodies.
2.4.8 Cool Mater Canals and Reservoirs Present the design bases for the capacity and the operating plan for safety-related cooling water canals and reservoirs (reference Section 2.4.11).
Discuss and provide bases for protecting the canals and reservoirs against wind waves, flow velocities (including a13.owance for fxeeboard),
and blockage and (where applicable) describe the ability to withstand a
probable maximum flood, surge, etc.
Discuss the emergency storage evacuation of reservoirs (low-level outlet and emergency spillway).
Describe verified runoff models (e.g.,
unit hydrographs),
flood routing, spillway design, and outlet protection.
2.4.9 Channel Diversions x
Discuss the potential for upstream diversion or rerouting of the source of cooling water (resulting from, zor example, river cutoffs, ice 2-19
I WNP-2 USAR g
REVISION 7/21/97 (draft)
- 5. Z.4. l. Syo kern PIoondkrS Sub>e<4 Q ZesIoeekron pcon7rriiuei) fh l~
rd I
i rrl) and has allowed the piping examination to be upgraded to conform to the requirements of the Summer 1975 Addenda to Section XI as far as practical.
The owner has developed an inservice inspection program coordinated with plant design, which complies with the intent of 10 CFR 50.55a to the maximum extent possible.
r The preservice examination was performed on Class 1 components and piping pursuant to the requirements of the 1974 Edition of the ASME BkPV Code,Section XI, including the Summer 1975 Addenda for both the RPV and associated piping, pumps, and valves. It is degcri~edde&hd in the WNP-2 Preservice Inspection Program Plan (Reference 5.2-6).
Preservice Ins ection Pro ram Plan submittal dates are li ted in the FSAR hrou h mendmen 52 for ea e of reference 7
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~
~
5.2-31
2.
Contml of welding.
Provide the following information relative to the control of welding of austenitic stainless steels for components of the RCPB:
a.
Sufficient information about the avoidance of hot cracking (fissuring) during weld fabrication.and assembly of austenitic stainless steel components of the RCPB to indicate whether the degree of weld integrity and quality will,be comparable to that obtainable by following the recommendations of Regulatory Guide 1.31, "Control of Stainless Steel Welding."
Describe the requirements regarding welding procedures and the amount, of and method of detezmihing delta ferrite in weld filler metals and in production welds.
b.
Sufficient information about electroslag welds in aus-tenitic stainless steel components of the RCPB to indicate whether the degree of weld integrity and quality wi11 be comparable to that obtain-able by following the recommendations of Regulatory Guide 1.34, "Control of Electroslag Weld Propezties."
Provide details on the contml of welding variables and the metallurgical tests required during procedure qualification and production welding.
c.
In regard to welding and weld repair during fabrication and assembly of austenitic stainless stee1 components of the RCPB, pro-vide sufficient details about welder qualification for areas of limited accessibility, requalification, and monitoring of production welding for adherence to welding qualification requirements to indicate whether the degree of weld integrity and quality will be comparable to that obtain-able by following the recommendations of Regulatory Guide 1.71, "Welder Qualification for Areas of Limited Accessibility."
3.
Provide sufficient information about the program for nondestructive exazxhxation of austenitic'stainless steel tubular products (pipe, tubing, f1anges, and fittings) for compo-
'ents of the RCPB to indicate whether detection of unacceptable defects (regardless of defect shape, orientation, or location in the pmduct) will be comparable to that obtainable by following the recommendations of Regulatory Guide 1.66, "Nondestructive Examination of Tubular Products."
5.2.4 Inservice Ins ection and Testin of Reactor Coolant Pressure Bo~~zo This section should discuss the insezvice inspection and testing program for the NRC QuaU.ty Group A components (ASME Boiler and Pressure Vessel Code,Section III, Class 1 components)
Provide sufficient detail to show that the inservice inspection program meets the require-ments of Section XI of the ASME Code.
Areas to be discussed should include:
l.
System boundary subject to inspection, including associated component supports, structures, and bolting, 5-7
2.
Arrangement of systems and components to provide accessibility, 3.
Exanination techniques and procedures, including any special techniques and procedures that might be used to meet the Code requirement, Inspection intervals, 5.
Inservf.ce inspection program categories and requirements, 6.
Evaluation of emuMxation results, 7.
System leakage and hydrostatic pressure tests.
In the PSAR, a detailed inservice inspection program including information on areas subject to emuuination, method of examination, and extent and frequency of examination should be provided in Chapter 16, "Technical Specifications."
5.2.5 Detection of Leaka e Throu Reactor Coolant Pressure Bounda The program should be described and sufficient leak detection system information should be furnished to indicate the extent to which the recoaxnendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"
have been followed.
Specifically, provide information that wi11 permit comparison with the regulatory positions of the guide, giving a detailed description of the systems employed, their sensitivity and response time, and the reli-ance placed on their proper functioning.
Also, the limiting leakage conditions that will be included in the Technical Specifications should be provided.
Identify the leakage detection systems which are designed to meet the sensitivity and response guidelines of Regulatory Guide 1.45.
Describe these systems as discussed in Section 7.5, "Safety-Related Display Instrumentation."
Also,,identify those systems that are used for alarm as an indirect indication of leakage and provide the design criteria.
Describe how signals from the various leakage detection systems are correlated to provide information to the plant operators on conditions of quantitative leakage flow rate.
Discuss the provisions for testing and calibration of the leak detection systems.
5-8
WNP-2 USAR REVISION 7/22/97 (draft) 2.4.13 GROUNDWATER 2.4.13.1 e cri an n i e e
Site undwater conditions are resented in ection 2.
4.6 d a historical discussion of ional and l undwater characteri tic a
ifers f rmation ur in d ite u w
rovided in theF AR throu h Amendment e
i n-i oun water evel of42 m l is based on the ro sed con c
n of e Ben Franklin Dam at RM 34 The roectwascanceled.
Watertableelevation tWNP-2is38 mslwithseasonalv
'ions ess than 1
7 7
2.4-20
WNP-2 USAR REVISION 7/22/97 (draft)
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n 0
'V 0
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2.4-21
WNP-2 USAR REVISION 7/22/97 (d?RA)
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2.4-22
%NP-2 USAR REVISION 7/22/97 (draft)
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2.4-23
WNP-2 USAR REVISION 7/22/97 (draft) s s
t
'Haec water supply wells are located on the WNP-2 si Two onsite wells draw fiom the unconfined aquifer in the Ringold Formation and a third well penetrates the confined aquifer in the underlying basalt flows. Normal water supply is from the river, and the deep well is maintained in the standby mode to provide supplemental makeup water for the potable and demineralized water system as needed.
The design is for a peak requirement of 250 gpm although average useege should he less than 20 gpm. The two shaHow wells were used during construction.
2.4-24
as related to existing or potential future water users.
Discuss the bases used to determine dilution factors, dispersion coefficients, flow velocities, travel times, sorption and pathways of liquid contanMants.
The locations and users of surface waters should be included in Section 2.4.1.2, and the release points should be identified in Section 11.2.3.
2.4.13 Groundwater All groundwater data should be presented in this section, in Section 2 5.4; ox in both and should be appropriately cross-referenced.
If the information is placed in both sections, the information in the two sections should be consistent.
2.4.13.1 Descri tion and Onsite Use.
Describe the regional and local groundwater aquifers, formations,
- sources, and sinks.
Describe the type of groundwater use, wells, pumps, storage facilities, and flow requirements of the plant. If groundwater is to be used as a safety-related source of water, the design basis protection from natura1 and accident phenomena should be compared with Regulatory Guide 1.27 guidelines and an indication should be given as to whether, and if so how, the guidelines have been followed; if not followed, the specific altexnative approaches used should be described.
Bases and sources of data should be adequateLy described.
2.4.13.2 Sources.
Descxibe present regional use and projected future use.
Tabulate existing users (amounts, water levels and elevations, locations, and drawdown).
Tabulate or illustrate the history of ground-,
water or piezometxic level fluctuations beneath and in the vicinity of the site.
Provide groundwater or piezometric contour maps of aquifers beneath and in the vicinity of the site to indicate flow directions and gradients; discuss the seasonal and long>>term variations of these aquifexs.
Indicate the range of values and the method of detemtuation for vertical and horizon-tal permeability and total and effective porosity (specific yield) for each relevant geologic formation beneath the site.
Discuss the potential for reversibility of groundwater flow resulting from local areas of pumping for both plant and nonplant use.
Describe the effects of present and projected groundwater use (wells) on gradients and groundwater or piezometric levels beneath the site.
Note any potential groundwater recharge area such as lakes or outcrops within the influence of the plant.
2.4.13.3 Accident, Effects.
Provide a conservative analysis of a postulated accidental release of liquid radioactive material at the site.
Evaluate (where applicable) the dispersion, ion-exchange, and dilution capability of the groundwater environment with respect to present and projected users.
Identify potentiaL pathways of contamination to nearby groundwater users and to springs,
- lakes, streams, etc.
Determine groundwater and radionuclide (if necessary) travel time to the nearest downgradient groundwater user or surface body of water.
Include all methods of calcu-lation, data sources,
- models, and parameters or coefficients used such as dispersion coefficients, dispersivity, distribution (sorption) coefficients, hydraulic gradients, and values of permeability, total and effective porosity, and bulk density along contaminant pathways.
2-22
Exceptions to RG 1.70, R2 WNP-2 complies with the guidance forRegulatory Guide 1.70, R2 except as indicated below.
General:
Certain information required by RG 1.70 becomes historical in nature after the granting ofan Operating License.
This information may include descriptions necessary to evaluate site acceptability, validate data analysis or conclusions, provide summaries or details which led to the selection ofa parametric values which were then modified by application ofengineering margins to establish design bases values, or summaries of submittal dates for various answers to questions for additional information during the pre-license review phase ofthe project.
Such information is historical in nature and does not contribute to the WNP-2 stafFs abilityto assess the impact ofproposed changes to plant design, configuration or operational practices.
Maintenance ofthis information in the FSAR is an unnecessary burden and detracts &omthe analysis necessary to properly evaluate proposed change to the plant or its operation.
This information has been evaluated and removed &omthe current FSAR through an archival process.
Archived information willnot be evaluated during future change analysis. Asummary ofthe information is provided in the FSAR section where it existed and a reference to its current location is also provided. The following, identified by RG section number, lists the specific FSAR content requirements inRG 1.70, R2, which have been removed under this exception.
2.4.13.1 This section requires a description ofregional and local groundwater aquifers formations, sources and sinks. It also indicates that bases and sources for the data should be adequately described.
WNP-2 provided detailed descriptions and data through amendment 52 to the FSAR. Detailed information on this subject has been removed &om the FSAR in amendment 53, as it is serves only historical purposes. Asummary ofthe pertinent conclusions ofthe details is all that willbe retained in this section ofthe FSAR.
~ +
WNP-2 TABLE OF CONTENTS (Continued) 2.4.14 TECHNICAL SPECIFICATIONS AND EMERGENCY OPERATION REQUIREMENTS 2 ' '5 REFERENCES e
AMENDMENT NO.
18 September 1981 Page 2 4-43 2.4-4 2.5
- GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 2-5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION 2.5.1.1 Regional Geology 2.5.1.1.1 Geologic History 2.'5.1.1.2 Provinces 2.5.1.1.2.1 Columbia Plateau 2.5-1 2.5-8 2.5-8 2 5-10 2.5-BROS tVE 2.5.1.1.2.2 Coast Range of Oregon and Wa 2.5.1.1.2.3 Puget Willamette Tro 2.5.1.1.2.4 Blue Mountains 2.5.1.1.2.5 High Lav ains Province 2.5.1.1.2.6 No em Rocky Mountains and Idaho Batholith Provinces I 2.5. 1... 7 Okanogan Highlands
.5.1.1.2.8 Cascade Mountains ngton 2.5<<15 2 5-15 2 5-'16 2.5-16 2'-16 2.5-17 2.5-19 I 2.5.1.2 Columbia Plateau Province and Site Geology 2.5 1.2.1 2 5 ~ 1 2
2 2 5 1.2.3 2'-1 '
3 1
2.5-1.2 '.2 2.5 '.2.3 '
Geologic History of Columbia Plateau Stratigraphy and Lithology Columbia Plateau Structural Geology Badger Mountain, Beezley Hills, Moses Stool Folds Kittitas Val enchman Hills 2-vi3.
2 5-26 2 5-26 2.5-38 2 '-66 2.5-66 2.5-67 2.5-68 pRcH >vi
TABLE OF CONTENTS (Continued)
AMENDMENT NO.
18 September 1981 Page 2.5.1.2.3.4 Manast:ash Ridge Hanson Creek 2.5. 1.2. 3. 5 Yakima Ridge 2.5.1.2.3.6 Ahtanum Ridge 2.5.1.2.3.7 Toppenish Ridge 2.5.1.2.3.8 Horse Heaven Hills 2.5.1.2.3.9 Columbia Hills 2.5.1.2.4 Pasco Basin Structural Geolo 2.5.1.2-.4.1 Saddle Mountains.
2.5.1.2.4.2 Umtanum Ridge 2.5.1.2.4.3 Gable Mountain 2.5.1.2.4.4 Rattlesnake - Mal ula Alignment 2.5. 1.2.4.4. 1 Rattlesnake ills and Rat.tlesnake Mountain iclines 2.5.1.2.4.4.2 Red Moun ain 2.5.1.2.4.4.3 Badge Mountain 2.5.1.2.4 '.4 N-11/M-Hill 2 5.1.2.4 4.5
-Hill 2.5.1.2.4.4 '
K-Hill 2.5.1.2.4.
.7 The Butte II 2.5.1.2
.4.8 Molly Hill 2.5..2.4.4.9 Kennewick, Horn Rapids, Cold Creek Lineaments t
.5.1.2.4.5 Wallula Fault Zone 2.5.1.2.5 Olympic Hallowa Lineament.
2.5.1.2.6 Hazards 2 viii 2.5-68 2.5 8
.5-70 2.5-70 2'-74 2'-75 2'-76 2.5-76 2.5-80 2.5-84 2'-88 2.5-89 2 5-90 2.5-91 2.5-91 2.5-92 2.5-92 2 '-92 2.5-93 2.5-93 2.5-95 2.5-98 2.5-101
WNP -2 AMENDMENT NO.
18 September 1981 5.1.1. 2 Provinces Th Pacific Northwest physiographic provinces are shown on Figu e 2.5-1.
The source of the province boundaries and desc
'ptions are taken from McKee (1972), Washington Public Power upply System (1977b),
and Rockwell (1979).
No distin tion is made between geologic and physiographic provinc due to their coincidence throughout the Pacific Northwes From the P ific Ocean eastward, the first major physiography feature is the Coast Range of Washington and Oregon.
The mountains extend northward from the Klamath Mountains of s uthern Oregon to the Strait of Juan de Puca.
East of these untains lies the Puget-Willamette
- Trough, a
series of topogr phic lowlands that extend parallel to the Coast Range from he Willamette River valley on the south to the Strait of Geor ia on the north.
East of the the Puget-Willamette Tr ugh are the Cascade Mountains.
The Cascade Mountains ex end from northern California to southern British Colu bia where they merge with the Coast Mountains.
East of th
- Cascades, the north-south grain of the regional physiograp y gives way to an east-west grain.
From north to south, the principal elements are the Okanogan Highlands, the Columbia P ateau, the Blue Mountains, and the High Lava Plains and Snake River Plain.
To the east and north in Idaho, western Mon ana, and British Columbia, the north to northwest regional ain returns in the form of the Northern Rocky Mountains.
A
'scussion of the tectonics of these provinces is contained i Section 2.5.2.2.1.
The WNP 1-2-4 site (Figure 2.5-1) lies in southeastern central Washington within the Colu ia Plateau province.
The site is situated near a north-s uth stretch of the Columbia River, the major watercours in the region.
We Pasco Basin contains the site and corn ises approximately 4,144 square km of undulating semiarid lain with low-lying hills, dunes, and intermittent streams.
The northern and southern boundaries of the Pasco Basin
( 'gures 2.5-4, 2.5-6a, 2.5-6b, and 2.5-6c) are defined by the Saddle Mountains and Rattlesnake Mountain, respect ely.
The easterly ends of Umtanum and Yakima Ridges ma k the western boundary of the basin.
To the east the basin erges into a vast expanse of dunes, dissected flatlands, and coulees northwest of the Snake River.
A detailed discuss n of the Columbia Plateau province is contained in Section
.5.1.2.
A detailed discussion of the Pasco Basin is contain in Section 2.5.1.2.4.
2 '-13
AMENDMENT NO.
18 September 1981 2 ~ 5.1.1.2.1 Columbia Plateau T e Columbia Plateau (a physiographic and geologic province) is bounded by the Blue Mountains and High Lava Plains on the so th, the northern Rocky Mountains-Idaho batholith on the
- eas, the Okanogan Highlands on the north, and the Cascade Moun ains Province on the west.
The Co umbia Plateau is drained by the Columbia River which flows w stward toward the Pacific Ocean.
The Snake River joins th Columbia River after draining the eastern Columbia Plateau a d parts of the adjoining provinces to the east and south.
Mo t of the Plateau (see Figure 2.5-7) has gentle topographic relief.
Exceptions to this gentle relief are the deep gor e of the Columbia River, the many steep-walled coulees north and east of the Columbia River, and the series of linear, gen rally west to northwest-trending, anticlinal ridges in the v'cinity of Yakima.
The Channeled Sca land of Washington covers the Columbia Plateau from Spoka e on the northeast to the Snake River on the south and to th Columbia River on the west.
The scabland topography as formed in Pleistocene time by the action of glacial mel waters and catastrophic floods due to breach ing of ice-damme lakes in western Montana.
The Columbia Plateau for ed between 16.5 and 6 m.y.b.p.
(Watkins and Baski, 1974; cKee and others, 1977) when large volumes of basalts were er ted from north-northwest trending linear vent system in northeastern Oregon and southeastern Washington (see igure 2.5N-2)(Waters, 1961; Taubeneck, 1970; Swanson and o hers, 1975; Fruchter and Baldwin,,1975; Price, 1977; Swa son and others, 1977).
The lavas of the Columbia Plateau over an area of approximately 202,018 square km an have an estimated volume of 170,894 cubic km (Figures 2.5-3 a
d 2.5N-2)
(Swanson and Wright, 1978).
The Columbia Plateau
's surrounded by topographically higher areas.
The cha cter of the pre-Tertiary rocks covered by basalt is isible only within highlands surrounding the plateau.
Individual basalt flows are voluminous, gen rally 8 to 25 cubic km, with a maximum known volume of 604 ubic km.
Flows range in thickness from a few inches to ore than 300 ft, with an average thickness of 90 to 120 t (Swanson and others, 1979a).
We thickest flows are inte reted as showing ponding in pre-basalt valleys, in structur ly 2 '-14
/
Exceptions to RG 1.'70, RX 2.5.1.1 This section on regional geology was addressed by WNP-2, through Amendment 52 to the FSAR, in specific discussions ofgeologic features which contain a mix of general and specific information describing and differentiating the various regional features.
Included in these discussions were tables ofdata refiecting investigative techniques, such as borehole analysis, trenching analysis and compilations ofscientific research in this area.
This information is historical in nature and has been removed Rom the FSAR by amendment 53. Sections archived Rom the FSAR via this exception include:
2.5.1.1.2 (all sections).
2.5.1.2 This section on site geology was addressed by WNP-2, through amendment 52 to the FSAR, with specific discussions of geologic features near and coincident with the WNP-2 site. Supporting these discussions were tables and maps ofdata used to characterize the features.
These specific discussions and data are historical in nature and have been removed &omthe FSAR by amendment 53, Sections 2.5.1.2.3.1 through 2.5.1.2.5 have been archived.
FSAR Content Revision Issues
- 1. Historical Information - What do we need to update per 50.71(e)?
Chapter 2 - Site Characteristics Geography and Demography Meteorology Hydrologic Engineering Geology, Seismology and Geotechnical Engineering Nearby Industrial, Transportation and MilitaryFacilities Chapter 11 - Radioactive Waste Management Source Terms Liquid Waste Management Solid Waste Management Chapter 14 - Initial Test Program II
- 2. Drawings and Figures - For plants which incorporated PEc ID into their FSARs, how can we remove drawing details which are non-safety significant from SAR space for 50.59 and 50.71(e) considerations?
- 3. Operating Plant Safety Analysis Report - What is really needed to ensure effective reviews and approvals ofchanges, induding NRC approvals?
- 4. State ofthe ArtApplications - When to incorporate in SAR and at what detail?
Substituted EPRI Chemistry Guidelines for GE BWR Chemistry Guidelines Substituted Condensate Suction Ion Chromatography sampling and Sodium HexaQouride Injection for Condensate HotweH Conductivity Sampling
Af P
Attachment 2
'EETING WITH WASHINGTON PUBLIC POWER SUPPLY SYSTEM FSAR UPGRADE PROGRAM ATTENDEES August 27.
1997 NRC T. Colburn W.
Bateman'.
Akstulewicz E.
HcKenna S. Hagruder Washin ton Public Power Su l
S stem D. Coleman P. Inserra J. Gearhart Nebraska Public Power District R.
Wenzl
L P