ML20041E990
| ML20041E990 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/01/1982 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | Tallman W PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| References | |
| NUDOCS 8203160068 | |
| Download: ML20041E990 (59) | |
Text
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7 DISTRIBUTION:
Docket File 50-443/444 1
SE NSIC TERA Docket hos: 50-44a ACRS (16) dild 50-444 LB#3 Files 8
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i:r. ullliara C. Tallnan SHanauer s
f, Chairrian and Chief Executive Of ficer RVollmer
.Q, 'S Public Service Coupony of hew Ho:apshire Riiattson 4
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7 P. O. Box 330 HThompson
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- 1anchester, New Hanpshire OJ1u5 OELD t
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Dear c;r. Talluan:
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Friiraglia
Subject:
Request for Additional Information LWheeler Enclosed are requests for additional infornation fron Mechanical Engineering Branch (r:LB, 210.3 thru bb), Siting Analysis Branch (51w, 311.1), Eftluent Treatment Systems Branch (ETSB, t>40.la thru 40) and Containment syste"'is Branch (C5B, 400.6 thru 2J).
Your responses to the au and Cbti requests should be forwarded to the EC staff not later tnan April 2,1962. The MLB responses will be reviewed by the statf at a meetirig with Unitec Lngineers and constructors, Incorporated in Philaceiphia, Pennsylvania on April 20-23, l'382.
The ETSB responses will be revieuec at a c.ceting at the Seabrook site tentativt.ly scheduled for April 7, 1962.
It clarification of these requests is necessary, the Seabrook Project liona er, u
Mr. Louis L. Wheeler, is available to provide any adnitional inforr.ation required.
Mr. Wheeler's telephone nuuner is (301) 492-7794.
Sincerely,
' original signed by.
?.k$.* $.#-
J Frank J. Miraglia, Chid Licensing Branch !!o. 5 Division of Licensing x
Enclosure:
N As stated cc w/ enclosure:
See next pc9e 8203160068 820303 PDR ADOCK 05000443 A
PDR omca>..D..L..:.,L.B.. # 3 f
LWheeler
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J SEABROOK William C. Tallman Chairman and Chief Executive Officer Public Service Company of New Hampshire P. O. Box 330 Manchester, New Hampshire 03105 John.1. Ritscher, Esq.
E. Tupper Kinder Esq.
Popes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State House Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire The Honorable Arnold Wight P. O. Box 330 New Hampshire House of Representatives Manchester, New Hampshire 03105 Science, Technology and Energy Committee State House G. Sanborn Concord, New Hampshire 03301 U. S. NRC - Region I 631 Park Avenue Resident Inspector King of Prussia, Pennsylvania 19406 Seabrook Nuclear Power Station c/o V. S. Nuclear Regulatory Commission Ms. Elizabeth H. Weinhold P. O. Box 700 3 Godfrey Avenue Seabrook, New Hampshire 03874 Hampton, New Hampshire 03842 Mr. John DeVincentis, Project Manager Robert A. Backus Esq.
Yankee Atomic Electric Company O'Neill, Backus and Spielman 1671 Worcester Road 116 Lowell Street Farmingham, Massachusetts 01701 Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager Norman Ross, Esq.
United Engineers ant. Constructors 30 Francis Street 30 South 17th Street Brookline, Massachusetts 02146 Post Office Box 8223 Philadelphia, Pennsylvania 19101 Karin P. Sheldon, Esq.
Sheldon, Harmon & Weiss Mr. W. Wright Project Manager 1725 I Street, N. W.
Westinghouse Electric Corporation Washington, D. C.
20006 Post Office Box 355 Pittsburg, Pennsylvania 15230 Laurie Burt, Esq.
Office of the Assistant Attorney General Thomas Dignan, Esq.
Environmental Protection Division Ropes and Gray One Ashburton Place 225 Franklin Street Boston, Massachusetts 02108 Boston, Massachusetts 02110 D. Pierre G. Cameron, J r.. Esq.
General Counsel Public Service Company of New Hampshire e
P. O. Box 330 Manchester,- New Hampshire 03105 e-,n.
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ENCLOSURE o.
(
e 3.2 Seismic Classification 210.3 3.2.1, Table 3.2-2, Sheet 4 Justify the non-sei'smic classification of the containment recirculating filter system.
Show that its failure wiLL not impair either the fans or ductwork.
21'0.4 3.2.1, Table 3.2-2, Sheet 1 Explain note 9 as it a p p l i e's to the reactor coolant pump flywheel.
210.5 3.2.1, Page 3.2-1 Describe methods used to confirm the structural integri.ty of non-seismic Category I components whose failure or collapse could result in loss of function of seismic Category I equipment.
210.6 3.2.1, Table 3.2-2, Sheets 2'b a~nd 21 Why are the computer room system components and the primary auxiliary building dampers and ductwork not considered seismic Category I?
210.7 3.2.2, Table 3.2-2 I t. is the staff's position that certain systems important not identified in Regulatory Guide 1.26 should be classified Quality Group C,
or its equivalent.
A"mong these systems are:
diesel fuel oil storage and transfer system,.
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diesel engine cooling water system, diesel engine lubrication i
system, diesel engine starting system, and diesel engine combustion air intake and exhaust system.
Justify the absence of a quality group classification of portions of those systems listed below:
q Diesel Generator Fuel Oil Storage and Transfer System i
1.
Remaining on-engine equipment and piping.'
e Diesel Generato.r Cooling Water System 1.
Auxiliary Coolant Pump i
2.
Remaining on-engine equipment and piping.
Diesel Generator Starting System l'
Air compressor 2
2.
Remain'ing on-engine equipment and piping.
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Diesel Generator Lubrication System 1.
Auxiliary tube oil pump f
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.2.
Remaining on-engine equipment and piping.
)
I Diesel Generator Combustion Air Intake and Exhaust System i
1.
Piping 2.
Air intake filter I
3.
Exhaust silencer o
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,.- a 210.8 3.2.2.2, Table 3.2-2, Sheet 1 Explain your rationale for classifying the shel.l side of the reactor coolant pump thermal barrier heat exchanger as ASME Code Class 3 although the tube side'is Code Class 1.
3.2.2.2, Table 3.2-2, Sheet 20 210.9 Justify the absence of a quality group classification for the entire computer room air conditioning system.
210.10 3.2.2'.2, Table 3.2-2, Sheet 20
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Control room complex emergency cleanup filter s y s t e.m fans and filter unit have been given ANS safety classification Non-Nuclear Safety, and the ductwork no safety classification at all.
This system is considered important to safety.
Provide justification for your' classification.
.210.11 3.2.2.2, Table 3.2-2 The following ventilation systems that serve the control room or engineered safety feature rooms have portions of their systems
.l a c k i n g a qua.lity group classification.
Assign an, appropriate quality group classification or its equivalent or justify the nonclassification:
1.
Control room complex ventilation system ductw'ork.
2.
Fuel storage building ventilation system, ventilation fans, ductwork.
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r 210.12 3.2.h.2, Tabte 3.2-2, Sheets 29, 30 Explain the NNS ANS safety classification of the entire liquid and solid waste systems.
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a 3.6.2 Determination.of Rupture Locations and Dynamic l
Ef.fects Associated with the Postulated Rupture of l
Piping 4
210.13-3.6(B).2.1, Page 3.6(B)-6 Confirm that the "elastical'ly calculated basis" for loadings of operating p l a'n t conditions plus an operating basis earthquake i's the maximum stress as calculated'by
)
equation 9 in Paragraph.NB-3652 of the ASME Code, Section i
III.
j 210.-14 3.6(B).2.1, Page 3.6(B)-7 i
j Provide -drawings of all. postulated pipe breaks, showing i
the type of break, structural barriers, restraint locations, l
and constrained directio'ns in each restraint.
Also provide a table showing calculated stress intensities, cumulative usage factors, and primary plus secondary stress ranges for J
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each postulated break.
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,e 210.15 3.6(5).2.1, Page 3.6(B)-8 i.
Specify where pipe restraints or anchors have required welding to the outer surface of the pipe.
Provide details of the stress analysis performed as in the case of a riser cl' amp Lug.
- 210.16 3.6(B).2.1, Page 3.6(B)-8 Inservice inspection of break-exclusion piping must 9
include 100% volum,etric examination of all pipe welds.
Augment your' inservice inspection description to include r
this requirement at intervals shown in IWA-2400, ASME Code,Section XI.
210.17 3.6(B).2.3, Page 3.6(B)-14 Justify the 90% of yield stress criteria in p l a s t'i c restraint design.
Provide examples of analysis of such a design.
219.18 3.6(B).2.3, Page 3.6(B)-15 Provide a reference or further justification for the i
use of a maximum fiber strain of 50% of ultimate strain as i
i an adaquery requirement for the load carrying capacity of piping.
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l 210.19 3.6(B).2.3, Pa g e 3. 6 (B)-17 What strain-rate an'd strain-hardening effects you have included in plastic system analysis?
210.20 3.6(B).2.5, Page 3.6(B)-18 In order to complete our review, we must examine Appendix 3B, "Line. Designation T'abulation".
Provide a copy of this appendix.
]
210.21 3.6(N).2.1, Page 3. 6 ( N) -1 In the primary loop, what size breaks are postulated for the design of pipe whip restraints?
What size breaks are postulated in the primary loop for determination of compartment pressurization and asymmetric loads?
If breaks for either case are less than full size, provide justification.
210.22 3.6(N).2.3, Page 3.6(N)-7 Provide a copy of test results of pipe-to pipe impact.
Also provide test results that show whipping or bending of a stainless steel pipe does not cause the section to become a missile.
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210.23:
3.6(N).2.5, Page 3.6(N)-11 Review of this section shows that you have used a i
cumulative usage factor of D.2 for postuatted pipe rupture criteria.
Branch Technical Position MEB 3-1 specifies a i
cumulative usage factor of less than 0.1.
Provide a commitment to meet this criteria.
210.24 3.6(N).2.5, Figure 3.6(N)-2 i
1 In addition to showing postulated break locations, they must be identified as either circumferential or longitudinal.
Structural barriers,.if any, restrain locations and constrained directions must also be included i n order to complete our review.
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3.7.3 Seismic Subsystem Analysis 210.25 3.7.3.1,.Page 3.7(B)-11 What criteria is used to determine the number of degrees of freedom in your dynamic analysis?
210. 2fi '
3.7.3.1, Page 3.7(B)-11 Demonstrate that the equivalent static load method analysis you have used accounts for relative motion between all parts of support.
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3.9 Mechanical Systems and Components 3.9(B).1.1, Page 3.9(B)-1 21n,27 Are there any reactor' coolant pressure boundary, ASME Code Class 1 or CS components in BOP?
If so, provide or reference an. appropriate design transient list.
210.28 3.9(B).1.2, Page 3.9(B)-1 f40 REG-0800 requi re s that computer programs in analyses of seismic Category I Code and non-Code items have the following information provided to demonstrate their applicability and validity:
a.
The author, source, dated version and facility, b.
A description and the extent and limitation of its application.
c.
Solutions to a series of test problems which shall be demonstrated to be substantially similar to solutions
.obtained from any one of sources 1 through 4, and source 5:
1.
Hand calculatioqs f
2.
Analytical results published in the literature.
3.
Acceptable experimental tests.
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By an MEB acceptable similar program.
5.
The benchmark problems prescribed in Report fJUREG/
CR-1677, " Piping Benchmark Problems".
Demonstrate compliance with these r'equirements and provide summary comparisons for the cor.puter programs used in se'ismic Category I analyses.
210.29 3.9(B).1.4, Page 3.9(B)-6 Whbre is AISC criteria used in evaluation of faulted conditions?
Justify its use.
210.30 3.9(B).1.4, Page 3.9CB)-7
,This section does not address the criteria used to assure the f 'u n c t i o n a l capabi.lity of essential systems when they are cubjected to loads in excess of those for which Service Limit B limits are specified.
By essential systems are meant those ASME Class 1, 2 and 3 and any other piping systems which are necessary to shut down the plant following,
'o r to mitigate the consequences of an accident.
Provide such criteria.
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- 3. 9 (.8 ). 2.1, Page 3.9(B)-8 i
What are the acceptance limits for steady s t a t'e and transient vibration?
The program must include a list of different flow modes and
- a. list of selected locations for v.isual inspection and measurements..
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210.32 3.9(B).2.1, Page 3.9(B)-10 l
What Code-allowable stress limits are used'for I
acceptability of mot. ion due so dynamic effects?
, 210.33
- 3. 9 ( B ). 3.1, Page 3. 9 ( B ) -13 What piping systems are not designed according to A Sti E Section III?
What design criteria was used for
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' 210.34 3.9(B).3.3, Page 3.9(B)-22 i
Regulatory Guide 1.67 does not address closed systems or systems with a water slug.
How wa s 1.67 used for the t
installation and design of pr.ssure relief devices?
210.35 3.9(B).3.3, Page 3.9(B)-23 i
Was, Regulatory Guide 1.67 used to determine the spacing of the safety valves on the main steam lines?
21(1.36 3.9(B).3.3, Page 3.9(B)-24 Provide a schedule for completion of dynamic analyses t
i results.
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3.9(B).3.4, Page 3.9(B)-26 Provide your i n't e r p r e t a t i o n of jurisdictional boundaries as they p'ertain to NF supports.
Justify your position.
210.38 3.9(B).3.4, Page 3.9(B)-26 Provide an example of the analysis performed on ASME Code class 1, 2,
and 3 valve supports.
210.39 3.9(B).3.4, Page 3.9(B)-26 The design criteria used for mechanical equipment supports needs clarification.
Subsection NF, ASME Code,Section III is applicable to these supports.
Justify the use 6f AISC' allowable stresses and demonstrate that your design criteria satisfy the requirements of Subsection NF.
210.40 3.9(B).3.4, Page 3.9CB)-26 Provide design criteria for any snubbers.
210.41 3.9(B).6.
As. required by 10 CFR 50.55a(g), we request that you submit your preservice and initial.20_ month inservice testing program for pumps and. valves.
Attachment A provides a suggested format for this submittal and a discussion of information we require to justify any relief requests.
210.42 3.9(N).1.2, Page 3.9(N)-20 Provide references 1 and 2 for our review.
210.43 3.9(N).1.4, Page 3.9(N)-33 How is.the critical buckling strength for component i
supports determined?
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210.44 3.9(N).2.5, Pages 3.9(N)-38 to 44 Previous analysis for other nuclear plants have shown that certain reactor system components and their supports may be-subjected to previou, sly under-estimated asymmetric
' loads under the conditions that result from the postulation of ruptures of the reactor coolar.1 piping at various locations.
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'escribed the design of the reactor The applicant has d
internals for blowdown loads only.
The applicant should also provide information on asymmetric loads.
It is, therefore, necessary to reassess the capability of these reactor system components to assure that the calculated
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dynamic asymmetric loads resulting from these postulated pipe ruptures will be within the bounds necessary to provide high assurance that the reactor can be brought safety to a cold shutdown condition.
The reactor system components that require reassessment shall include:
a.
b.
Core supports and other reactor internals.
c.
Control rod drives.
d.
ECCS piping.that is attached to the primary coolant piping.
e.
Primary coolant piping.
f.
Reactor vessel supports.
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The-following information should be included in the FSAR about the effects of postulated asymmetric LOCA loads on the above mentioned reactor system components and the various cavity structures.
1.
Provide arrangement drawings of the reactor vessel
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support systems in sufficient detail.to show the geometry of all principal elements and materials of construction.
2.
'If a slant-specific analysis will not ~
submitted oe for your plant, provide supporting information to demonstrate that the generic plant analysis under
, consideration adequately bounds the postulated accidents 1
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at your " facility.
Include a comparison of the geometric, structural, mechanical, and thermal-hydraulic thbca,se similarities between your facility and analyzed.
Discuss the effects of any differences.
3.
Consider all postulated breaks in.the reactor coolant piping system, including the following locations:
Steam line nozzles to piping terminal ends.
a.
b.
Feedwater nozzle to piping terminal ends.
c.
Recirculation inlet and outlet nozzles t5 recircu-Lation piping terminal ends.
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Provide an a s 's e s s m e n t of the effects of asymmetric pressure differentials
- on the systems and c o aponent s listed.above in combination with all external loadings including safe shutdown earthquake loads and other faulted condition loads-for--the postulated breaks described above.
This assessment may utilize the
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following mechanistic effects as applicable:
c disp;tatement break areas.
a:
Limited b.
Fluid-struct,ure interaction.
I-c.
Actual time-dependent' forci,n~g function.
-d., Reactor' support stiffness.
Breakfopeni,ng times.
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If the results of the assessment on item 3 above i
indicate loads leading to inelastic action of these systems or displacement exceeding previous ' design
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limits, provide an evaluation of the inelastic behavior (including strain hardening) of the material used in the s y s t e m "d e s i g ~n and the effect of the load transmitted to the backup structures to which these systems are attached.
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- Blowdown jet forces at the location of rupture (reaction forces), transient differential pressures in the annular region between the component and the wall, and t r a n s i ~e n t
'I different'ial pressures across the core barrel within the
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reactor vessel.
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6.
For all analyses performed, included the method of analysis,.the structural and hydraulic computer codes
- employed, drawings of the models employed and comparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.
7.
Demonstrate that safety-related components will retain their structural integrity when su'bjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.
8.
Demonstrate the functional capability of any essential when subjected to the combined loads resulting piping from the loss-of-coolant accident and the safe shutdown earthquake.
210.45 3. 9 ( tJ ). 2. 5, Page 3. 9 ( tJ ) - 41 Your statement that the loading imposed by the SSE is ger.erally small compared to blowdown loadings implies that in certain cases you have neglected loads due to an SSE.
If this is true, provide analysis details justifying your doing so.
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210.46 3.9(N).4.3, Page 3.9(N)-60 The statement "The' stress' limits bre established not only to assure thatipeak stresses will not reach unacceptable values, but also limit the' amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials." needs clarification.
What are these stress limits and from what source were they obtained?
210.47 3.9(N).4.3, Page 3.9IN)-61 Provide assurance that deformation limits are sufficient to guarantee control rod drive system integrity and functioning after a dynamic event such as an OBE.
210.48 3.9(N).5.2, Page 69 The statement "The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the material." needs clarification.
What "are these stress limits and from what source were they obtained?
210.49 3.9(N).5.2, Page 3.9(N)-68 to 71 Subsection NG, ASME Code Section III should be referenced as the design criteria for all design analyses, not just for the design basis accident.
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210.50 3. 9 ( tJ ). 5. 4, Page 3. 9 ( fi) -
Verify that reactor' internals are designed in accordance with Standard Review Plan 3.9.3 " core Support Structures" or justify alternate. design criteria.
210.51 3. 9 ( rJ ). 5. 4, Page 3. 9 ( fJ ) -71 dhat are the stresses associated with the maximum c
deflections in Table 3. 9 ( fi) -17 ?
State the basis for these
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deflection limits.- Justify the lack of safety margin for radial outward deflecti.on.
'210.52 3. 9 ( rJ ). _2. 3, Table 3. 9 ( tJ ) - 1 Provide information on how the number'of occurrences of steady state fluctuations were determined.
How was the effect of transients listed ~i'n this table considered for BOP equipment?
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210.53 3.9.6.2 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coo'lant system (RCS) pressure.
There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high Pressure RCS and the low pressure systems.
The leak tight integrity of these valves must be ensured by periodic leak t e s t i,ng to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require correction action; i.e.,
shutdown or system isolation when the final approved leakage limits are not met.
Also surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.
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Periodic leak testing of each' pressure isolation valve 1
is required to be performed at least once per each refueling i
I outage, after valve maintenance prior to return to service, l
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1 and for systems rated at less than 50% of RCS design v.
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l pressure each time the valve has moved f' rom its fully closed l
l l-position unless justification is given.
The testing i
i interval should average to be approximately one year.
Leak l
i testing should also be pe'rformed after all disturbances to
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the, valves are complete, prior to reaching power operation i
i following a refueling outage, maintence, etc.
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The staff's present position on leak rate limiting i
conditions for operation must be equal to or less than 1 f
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ga.llon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the l
L redundant pressure isolation function and give an indication of valve degradation over a finite period of time.
l Significant increases over this limiting valve would be an indication of valve degradation from one test to another.
Leak rates higher than'1 GPM will be considered if the j
leak r a't e changes are below 1 GPM above the previous test leak rate or system design precludes. measuring 1 GPM with r-i suf*.icient accuracy.
These items will be reviewed on a case by case basis.
The Class 1 to Class 2 boundary will be considered the i
i isolation point which must be protected by redundant isolation valves.
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In cases where pressure isolation is provided by two valves, both will be independently leak tested.
When three or more valves provide isolation, only two of the valves need to be leak tested.
Provide a list of all pressure isolation valves included in,our testing program along with four sets of y
Piping and Instrument Diagrams which describe your reactor coolant system press.ure isolation valves.
Also discuss in detail how your leak testing program will conform to the above staff position.
210.54 It is the staff's position that all essential safety-related instrumentation lines should be included in the vibration monitoring program during pre-operational or start-up testing.
We require that either a visual or instrumented inspection (as appropriate) be conducted to identify any excessive vibration that will result in fatigue failure.
Provide a list of all safety-related small bore piping and instrumentation lines that will be included in the initial test vibration monitoring program.
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,210.55 Due to a long history of problems dealing wit.h inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that mainte' nance records for snubbers be documented as follows:
c Pre-service Examinat ion "A
pre-service examination should be made on all snubbers listed in Tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9.
This examination should be made after snubber installation but not more than six months prior "t o initial system pre-operational testing, and should as a minimum verify t h' e following:
1.
There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.
2.
The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifications.
3.
Snubbers are not seized, frozen or jammed.
4 Adequate swing clearance is provided to allow snubber movement.
5.
If applicable, fluid is to be recommended level and is not leaking from the snubber system.
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Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
If the period between the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1,
4, and 5 shall be performed.
Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
Pre-Operational Testing During pre-operational testing, snubber thermal m o v e m e n.: s for systems whose operating temperature exceeds 250 F should be verified as follows:
a.
During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
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b.
For those systems which do not attain operating ll*
temperature, vewrify via observation and/or calcualtion that the snubber will accommodate the projected thermal movement.
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Verify the snubber swing clearance at specified heatup and cooldown intervals.
Any discrepancies or inconsistencies shatl be evaluated for cause and corrected prior to proceeding to the next specified interval.
The above described operability program for snubbers shoutd be included and documented by the pre-service e
inspection and pre-operational test programs.
The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion.
This test. program should be specified in Chapter 14 of the FSAR.
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A V" C m - M A I
fiP.C STATT C00iEtiTS ON IfiSERVICE PUMP AND VALVE TEST
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RELIEF REQ' JESTS The liRC staff', after reviewing a number of pump and valve programs, has determined that further guidance might be helpful to the type and extent of informatien we feel is necessary to expedite t review of these programs.
We feel that the Licensee can, by incorporating these guidelines into each program submittal, reduce considerably th staff's review time and time spent by the Licensee in responding to liRC sthff requests for additional information.
The pump testing program should include all safety related* Class 1 2 and 3 purps which are installed in water ccoled nuclear power plan and which are provided with an emergency power source.
The valve testing program should include all the safety related val in the following systems excluding valves used for operating conveni only, such as manual vent, drain, instrument and test valves, and valv used for maintenance only.
. PUR High Pressure Injection System a.
b.
Low Pressure Injection System c.
Accumulator Systems d.
Containrent Spray System Primary and Secondary System Safety and Relief Valves e.
f.
Auxiliary reedwater Systems
'g.
Reactor Building Cooling System
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h.
Activa Conponents in Service Uater and Instrement Air Systems which are required to support safety system functions.
i.
Containment Isolztion Yalves require' to change position to isolate containment.
j.
Chemical.& Volume Control System k.
Other key components in Auxiliary Systems which are required to direc suoport plant shutdown or safety system fuh: tion.
- Safety related - necessary to safely shut down the plant and mitigate the censequences of an accident.
I
.~ 2 _
1.
Residual H:at Rcmoval System m.
" ~ '
~
c.
High Pressare Core Injection System b.
Low Pressure Core Injection System Residual Hect Removal System (Shutdcwn Cooling System) c.
d.
Emergency Condenser System (Isclacion Condenser System)
Low Pressure Core Spray System e.
f.
Contair. ment Sprey System 9
Safety, Relief, and Safety / Relief Valves
.8 h.
RCIC (Reactor Core Isolation Cooling) System i.
Containatnt Cooling System j.
Containment isolation valves required to change position to isolate containment
.E.
Standby liquid ccntrol system (Boren System)
, l '.
Autcastic Depressurization System (any pilot or control valves, esscciatt hydraulic or pneumatic systems, etc.)
Centrol Rod Drive Hydraulic System (" Scram" function) m.
n.
ther key components in Auxiliary Systems which are requi, red to directly sucport plant shutdo.tn or safety cystem function.
~
o.
Reactor Ccolant System Jnservice Pum? ar.d Valve Testinc Procram, I.
Information reqaired for URC Staff Review cf the Pump and Valve Testing Program A.
Three sets of FoID's, which include all of the systems listed above, with the code class and system boundaries clearly marked.
The drawings should include all of the components present at the time of submittal and a legend of t.je FAID symbols.
B.
Identification of the applicable ASME Code Editicn and Addenda C.
The period for which the program is applicable.
t D.
, Identify the component code class.
l
E.
For Pump testing:
Identify
~
1.
Each pump rce.oired to be tested (nare and nurbe'r) 2.
The test parcreters to be r.easured 3.
The test frequency F.
For valve testir.g:
Identify 1.
Ecch valve in ASME Section XI Categories A & B that will be exercised every three months during normal plant operation (ir.dicate whether partial or full stroke exercise, and for power cperated valves list the limiting value for strche tim 2.)
2.
Each valve in ASME Section XI Category A that will be leak,..
tested during refueling cutages (Indicate the leak test procedure ycu intend to use) 3.
Each valve in ASME Section XI Categories C, D and E that will be tested, the type of test and the test frequency.
For check valves, identify those thtt will be exercisad every 3 raanths and those that will only be exercised during cold shutdo.;n or refueling outages.
II.
Additional Information That Will Be Helpful in Speeding Up the Review Process A.'
Include the valve location coordinates or other appropriate locaticn infctmation which will expedite our locating the valves on the PSIDs.
B.
Provide P&ID drawings that are large and clear enough to be read easily.
C.
Identify valves that are provided with an interlock to other components and a brief description of that function.
t l
Relief Recuests from Section XI Pecaircr.ents The larcest area of concern for the I:R: staff, in the review of an l
inservice valve and pump testing program, is i n evaluating the basis for i
justifying r lief frem Sectica XI Recuirements.
It has been our expdrience e
./
that many requests for relief, submitted in these programs, do not pmvide adequate descriptive and detailed technical infomatien.
This cxplicit informationisnecessarytoprovidereasonableassurancethattheburden imposed on the licensee in complying with the code requirements is not justified by the increased level of safety cbtained.
Relief requests which are submitted with a justification such as
" Impractical", " Inaccessible", or any other categorical basis, will recuire edditional information, as illustrated in the enclosed examples, to allcw our staff to make an evaluation of that relief request.
The intentica of this guidance is to illustrate the content and extent of infomation required by the liRC staff., in the request for relief, to cake a proper evaluation dnd adequately document the basis for that relief in Our safety evcTuatiCn report.
The li?C staff feels that by receiving this infer.,ation in the program subm;ttal, subsequent requests for additional infor=?. tion and delays in completing our review can be consid:rably reduced or eli*ated.
l I.
Inforr. tion Recuired for NRC review of F.elief Recurs _ts A.
Identify ccmp nent for which relief is requested:
1.
fiame and number as given in FSAR 2.
Function 3.
ASME Section III Code Class 4.
For valve testing, also specify the ASME Section XI valve category as defined in IWV-2000 C.
Specifically identify the AS!*.E Code requirement that has been detemined to be impractical for each comp 0ntnt.
C.
Provide infomation to support the determination that the requirement in (B) is impractical; i.e., state and explain the basis'ior requesting relief.
D.
Specify the inservice testing that will be perfor ed in lieu of the ASME Code Section XI requirements.
E.
provide the schedule for implementation of the f.cocedure(s) in (D).
~
-5 II.
Examples to Illustrate Several Poss'ible Areas Where Relief May Be
}
Granted and the Extent and Content of Information Necessary to Make
~
An Evaluation A.
Accessibility:
Tne regulctica specifically grants relief from the code requirement because of insufficient access pro-visicas.
However, a detailed discussica of actual physical arrangement of the ccmponent in questica to illustrate the insufficiency of space for conducting the required test is necessary.
I Discuss in detail the physical arrangement of the component
_in qu'estien to deconstrate that there is not sufficient space. ___,
"to perform the code required inservice testing.
Unat" alt'ernative surveillcnce means which will provide an' acceptable level of safety have you censidered and why are these neans not fcasible?
B.
Environmental Conditions (e.g., High radiation level, High temperature, High hemidity, etc.)
Although it is pruder.: to naintain occupation radiation exposure for* inspection personnel as low as practicable, the request for relief from the code requirements cannot be granted solely en the basis of high radiation levels alone.
A balanced judgment
~
between the hardships and compensating increase in the level of safety should be carefully established.
If the health and safety of the public dictates the necessity of inservice testing, alternative reans or even decontamination of the plant if necessary should be provided or developed.
Pro' vide additicnal information regardinc the radiation levels at the required test location.
What alternative testing techniques which will provide an acceptable level of assurance of the integrity of the compenent in question have you considered and why are these techniques determined to be impractical?
1
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E.
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C.
Instrum2ntation is not originally provided P
Provide informtion to justify that compliance with the code requirements wauid result in undue burden or hardships withcut a compensating increase in the level of plant safety. k'ha t alternative testing methods wilich will previde an acceptable level of safety have you considered and why are these mthods detemined to be impractical?
D.
Valve Cycling During Plant Operction Could Put the Plant in an Unsafe Conditica The licensee shculd explain in detail why exercising t'ests -
during plant opsration could jeopardize the plant safety.
E.
Valve Testing at Cold Shutdo.:n or Refuelino Intervals :n lieu of the 3 Month Kequired Interval The licensee shculd explain in detail tihy each valve cannot be exercised durinn normal operat, ion.
Also, for the valves where a refueling interval is indicated, explain in detail why each i
valve cannot be exercised during cold shutdown intervals.
- III. Acceptance Criteria for Relief R?cuest The Licensee must s6ccessfully demonstrate that:
1.
Compliance with the code requirements would result in hardships or unusual difficulties without a compensating increase in the level of safety and concompliance will provide an acceptable level of quality and safety, or 2.
Proposed alternatives to the code requirements or portions thereof will provide an acceptable level of quality and safety.
Standard Porgat A standard format, for the valve portion of the pump and valve testiaq program and relief requests, is included as an attachment to this Guidance.
The NRC staff believes that this standard format will reduce the tim-spent by b6th the staff in ciur review and by the licensee in their preparation
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of the' pump and valve testing program and submittals. The standard fon:zt includes examples of relief requests which are intended to illustrate the application of the standard format and are not necessarily a specific plat relief request.
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a.'
ATTkCHP.ENT STANDARD TOPS T VALVE INSEP.VICE TESTING PP.02T.A'1 SU3 'ITTAL O
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9 of / 2.
0 0'
'S i
gib
,g$ g 9
REIGRKS O
I'1 S-
,5 (Not to be used for relief basis)
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G
.h 8.J #
5 E
R @t.
as i2 8-e r3 Valve C' [
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Category c>
g
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42 42 TTTFT U U
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a:
,e 3
i D-14 X
4 GA M
,6 DE fA C
DI l I
3 C-15 X
16 CK SA CV X
CS t*
3 C-15 X
16 CK SA CV 3
E-14, X
3 REL SA CV 3
D-I l X
X 4
GL 11 C
Q X
3 B-11 X
B-11 X
I 3
A-10 X,
2 B-10 X'
2 D-14 X
10 GA < M0 C
Q LT X
i MT 30 sec.
e e
8 O
e 1-
1 Q-Exercise valve (full stroke) for operability every (3) months LT -
Valves are leak tested per.Secticn XI Article IllV-3420 HT
' Stroke time teasurements are taken and cc.mpared to the stroke tim liraiting value per Section XI Article IW 3410 CV -
Exercise check valves to the position required to fulfill their 4..
fun: tion every (3) months
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SRV-Scfety and relief valves are tested per Section XI Article It!V-3510
)
DT - Test category D valves par Section XI Article IW-3600 i
ET -
Verify and record valve position before operations are parformed and n.-
3,';
after operations tre ccmpleted, and verify that valve is locked or g(.
sealed.
{.
CS - Exercise valve for operability every cold shutdown
.g P.R -
Exercise valve for operability every reactor refueling
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Relief Recuest Basis _
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Auxiliary Coolant System, Component Cooling System:
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Valve:
717 Category:
C-Class:
3 Function:
Prevent backflow from the reactor coolant pump cooling coils In;ractical test requirement: Exercise valve for cperability every three months Basis for relief:
To test thi's valve would require interruption of cooling water to the reactor coolant punps notor cooling coils.
This action could result in danage i
ta the reactor coolant pumps and thus place the I
4plant in an unsafe mode of operation.
Alt ernative This valve will be exercised for operability
~
Testing:
doring cold shutdowns.
2.
Valve:
834 Cat:; cry:
B-E Class:
3 i
Function:
Isolate the primary water from the component cooling surge tank during plant opertion.
It is no'rmally in the closed position, but routine 2
op'eration of this valve will occur during refueling i
and cold shutdowns.
Impractical Test Exercise valve (full stroke) for. operability
~
Requirement:
every three (3) months.
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Basis for Relief:
This valve is not required to change position during plant operation to acccaplish its safety function.
Exercising this valve will increase the possibility of surge tank line contamination.
Alternate Verify and record valve position before and Testing:
after each valve operation.
3.
Valve:
7443 c
Category:
A Class:
2 Function:
Isolate the residual heat exchangers from the cold leg R.C.S. backflow and accumulator backflow.
Test Requirements:
Seat leakage test Basis for This valve is located in a high radiation field Relief:
(2000 mr/hr) which would make'the required seat leakage test hazardous to test personnel.
We ir.tcr.d to ceat leak test tiro other valves (875B and 876B) which are in series with this valve and will also prevent backflow.
We feel that by complying with the seat leakage requirements we will not achieve a compensatory increase in the level of safety.
Alternative No alternative seat leak testing is proposed.
Testing:
9 4
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_ em N 9 5 G W%
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l 311.1 Section 2.10201 of the FSAR indicat'es that phovisions have been
['
been made with the Boston and !%ine Railroad to control traffic on the railroad right of way passing through the site.
- Also, it is indicated that special access control is not required with respect to the Exeter and Hampton Electric Company power transmission line which will be buried underground along the eastment within the exclusion area. With respect to either of j
the above two easements, it does not appear that the applicants have established adequate control of potential activities within 4
the exclusion boundaries.
For example, provisions to control traffic on the railroad richt of way do not necessarily cover the possible need to control the access of railroad employees 4
for the purpose of inspection, maintenance, or repair of the track.
Similarly, in the case of the power transnission line casement, it is not clear how the applicant is to exert its authority over the access of Exeter and Hampton Electric Corpany employees and their potential subcontractors for the purpose of inspection or repair of the transmission line.
The 3
applicants should provide sufficient information which would verify that they have adequate authority over activities of the type described above.
i ii.
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Section 6.5.1 460.1(a) Based upon infonnation presented in FSAR Section 6.5.1, the Seabrook Station does not meet Acceptance Cri teria ll.2.c of SRP 6.5.1, Rev.1.
The containment enclosure energency air cleaning system flow rate is not indicated in the control room nor is it alarned.
It is the staf f's position that the flow rate should be indicated in the control room and that alarns at high and low flows occur at the main control board (MCB).
For the fuel storage building emergency air cleaning system, Section 6.5.1.5.b indicates that an alarn will occur in the MCB on a high dif ferential pressure signal but the section does not state i
whether there is a p signal in the MCB or whether p is recorded in
.the MCB.
In addition, the section only indicates that there is flow instrunentation and not the particular type of flow instrunentation
~i 1,
other than Section 9.4.2 which states that a,n alarm in MCB will occur i
under of f-normal flow condi tions.
It is our position that the MCB I
include an indication of p and flow rate and that both parameters he recorded.
460.2 Table 6.5-1 of the FSAR indicates that heaters are not required for the containment enclosure emergency air cleaning system.
No justi-4 fication is provided as to why they are not required nor how an r
ef ficiency for methyl radiciodine of 991 was claimed without such heaters or hunidi ty control devices.
It is the staf f's position that credit for 85; renoval of methyl radioiodine will be given wi thout some form of humidity control.
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r 460.3 Table 6.5-1 and 6.5-2 has indicated that the noisture separator will act as a prefilter in the containnent enclosure emergency air clean-ing system filter systems.
Provide justification for the exception that no prefilter is required as called for by Acceptance Criteria ll.2.a of SRP 6.5.1, Rev. 1.
Include in your justification a dis-i cussion of particulate renoval ef ficiency in the moisture separatnr at times of low relative humidity or entrained noisture and an analysis of tiEPA filter particulate loading without a prefilter, i
t Since this system is used as a backup to the 112 purge, it should contain a prefilter or a medium ef ficiency filter.
460.4 Regulatory Position C.3.n of Regulatory Guide 1.52, Rev. 2, corres-ponds to Acceptance Criteria 11.4.1 of SRP 6.5.1, Rev.1.
The applicant's response to this regulatory position, as presented in Tables 6.5-1 and 6.5.2, is confusing.
Indicate whether the two ESF filter systems meet Acceptance Criteria 11.4.1.
L 460.5 Verify that the filter bank for the ESF systems are 4 inches in depth.
i 460.6 figure 9.4-2 shows one filter train of the containment enclosure r
i emergency air cleanino system with a roughing filter and the other train with a moisture separator.
Which train is correct?
C Section 9.3.5 460.7 What is the design flow rate of the Recovery Evaporation Reboiler Pump which is missing from Table 9.3-8?
f f
460.8 On pg. 9.3-66 of the FSAR, it is stated that "... dikes are provided around tanks to contain any spills".
In addition to the boron waste storage tanks, identify other tanks to which this statement refers.
I i
5 460.9 Uhat type of bed is the recovery demineralizer; Cation, Anion, Mixed?
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Section 10.4.2 460.10 Figure 10.4-1 of the FSAR does not show where the liquid from the 4
mechanical vacuum pumps goes and whether it is treated as a radio-f active liquid.
Please indicate how this liquid is collected, how it is processed, if at all, and whether it is discharged.
I i
460.11 The main condenser evacuation system has not been designated as a j
c non-nuclear safety (NNS) systen in Table 3.2-2 of the FSAR, nor in i
Section 10.4.2.
The components of this syst,em should be designed l-to your NHS quality group to meet the requirements of Acceptance
[
Cri teria 11.2 of SRP 10.4.2.
t Section 10.4.3 460.12 The turbine gland sealing system has not been designated as a NNS system in Table 3.2-2 of the FSAR, nor in Section 10.4.3.
The com-ponents of this system should be designed to your NNS quality group to meet the Acceptance Cri teria of SRP 10.4.3.
i 4
Section 11.2 460.13 There is a concern on the part of the staff as to the capability of th,e SGB system to treat the SGB if primary to secondary leaks are
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e
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.p..
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e
occurring in SGs of both units.
Our experience indicates that it is not unusual to find SGB rates of 30 gpm/SG.
It does not appear that the system has the capability to treat this flow rate even if the leak is occurring in the SG of only one of the units since the evaporators are limited to 25 gpm per evaporator and there are only 3 evaporators and since additional capability could not always be
)
guaranteed from the liquid waste evaporator.
Although the SGB system has demineralizers for treatment of SGB during periods of no prinary to secondary leak, the discharge fron the demineralizers can only be discharged to the condenser, and not to the circulating water system.
Our concern is that if the evaporators do not have the capability to handle the amount of SGB, then discharges to the waste test tank discharge header can only occur without treatment of the SGB. Explain the capability of the SGB system to treat th'e bicwdown under condi-tions where one of the SGB cvaporators is inoperative, primary to i
secondary leaks are occurring in the SG's of both units, and the basis for limiting SGB to a maximum of 75 gpm under these conditions.
460.14 From Figure 10.4-7, Sheet 1, it appears that distillate from the dis-tillate cooler will be discharged to both the waste test tank and to the main condenser.
Should valve SB-V208 be shown as normally closed j
in this figure?
460.15 Table 9.3-3 indicates that the leakage rate in the fuel storage l
building is 20 gpm.
Is this correct or should it be 20 gpd?
~
460.16 Provide justification for taking exception to Regulatory Guide 1.143 by designing the chemical drain tank and the chemical waste treatment tanks to Standard PS-15-69.
460.17 Frovide for the following conponents the volune or flow rate, as appropriate, to the piece of equipnent and the number per unit:
a) chemical drain tank b) chenical drain transfer pump c) chemical drain treatment punp d) chemical drain treatment tank
]
e) reactor coolant drain tank (RCDT) f)
RCDT pump g)
RCDT heat exchanger 460.18 Which demineralizers are considered regenerable and where does the regenerant solution go't Section 11.3 460.19 Figure 11.3-1 does not appear to show a flow path from the H com-2 pressors to the ll surge tank.
Please provide a drawing which 2
indicates this flow path.
460.20 Will the iodine guard hed of the RGWs be tested to the requirements of Regulatory Guide 1.140?
460.21 Discuss the impact upon the operation of the RGWS if the purge gas condenser is inoperable.
i 460.22 Figure 11.3-1, Sheet 1, appears to show flow from the regenerative compressor of the RGPS going to the H2 gas compressor but not to the in-line electrical heater.
Please clarify the flow path for this portion of the RGUS.
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460.23 Which control panel monitors the discharge flow from the PAB filtered exhaust system?
460.24 Figure 9.4-4, Sheet 1, does not include parameters to be nonitored and alarmed in the filte,r train.
Please include them.
460.25 Sections 9.4.12.2.b.3 and 9.4.12.2.c.3 indicate that the exhaust from the administration and service building (RCA) is directed
~
through an absolute filter and a radiation monitor. A drawing c
showing the location of the filter and the nonitor was not found in Section 9.4,11.3, or 11.5 of the FSAR.
Please indicate where a figure contain.ing such information may be found in the FSAR or provide one if one is not in the FSAR.
What parancters of the filter system are monitored and alarmed? This monitor should be included in the discussion presented in Sect, ion 11.5.
460.26 Does a high pressure in the exhaust from the PAB normal cleanup ex-haust. fans resuit in an alarm on the 11CB?
460.27 Section 11.3.2.2 states that some cubicles of the RGWS will.be con-
~
tinuously monitored for H and that in the event of high H concen-2 2
tration:
a) the af fected components of the process stream wil,1 be isolated and/or the affected component purged with N 25 b) the affected cubicle will be ventilated to reduce the H con-centration; and 2
i c) unnecessary personnel will be evacuated from the area.
l a
f p
1.
A
4 It appears that the ventilation to reduce the H c ncentration could 2
result in the addition of air in the ventilation systems in the anbient carbon delay bed and the hydrogen surge tank area, thus re-sulting in a potentially explosive mixture.
Another potential source of 0 could be the air conditioning units.
Provide an analysis to 2
show that tig addition of air in these cubicles of the RGUS would not result in a deflagration or an explosion.
O
,, 460 28 Our review of RGWS P&lDs concludes that there are no rupture discs in the system.
Please verify this.
If there are nipture discs then liquid seals sho.uld be provided downstream of the discs and the design of the system should include measures to prevent the prenanent loss l
of the liquid seals in the event of an explosion.
~
460.29j Based upon our review of the RGWS it appears that the critical location in the system for monitoring 0 is the hydrogenated vent 2
header which is the source of input to the RGMS.
Therefore, it would appear that one of the 0 analyzers should be analyzing the 2
influent upstream of the gas chiller.
The other analyzer should be sequentially monitoring the inputs from such streams as the reactor coolant drain tank, PDT, letdown degassifiers, CVCT, etc.
The rational for placing the 0 monitor af ter the dryer should be 2
e, presented.
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'5 460.30 The SER, at the construction permit stage, indicated that the RGWS was to be designed to withstand a H expl os ion.
The FSAR does not 2
indicate that system is designed to withstand such an explosion.
If this is a design change from the CP stage, provide justification.
Section 11.4 460.31 Table 11.4-3 of the FSAR provides the volume of solid wastes antici-c pa ted from thd Seabrook Station. The volume of evaporator bottoms and other liquid wastes appears to be extremely low based upon the anticipated flow rates to the BRS, SGB, and liquid waste evaporators, and when compared to the solid radwaste shipped fron operating PWRs.
.The volume of compactable waste is also low.
Please provide the bases for your volumes of evaporator bottoms and other liquid waste, compactable waste, and non-compactable waste.
460.32 Does the solid waste system neet the provisions of Branch Technical Position (BTP) ETSB 11-3?
460.33 Provide a description of the solidification system / process to be used at Seabrook showing its conformance with the acceptance criteria of SRP 11.4.
Section 11.5 460.34 In Section 11.5.2.3 of the FSAR, reference is made to the vent stack nonitor yet no monitor is mentioned in Section 11.5.
Please clarify whether there is a plant vent nonitor.
The vent stack is the final
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release point for various ef fluent streams, including the PAB and the fuel storage building.
It is the staf f's position that there must be a continuous radiation monitor located in the plant vent.
D 460.35 Acceptance Criteria II.C.1.a states that the gaseous and liquid pro-cess streams or ef fluent release points should be nonitored and sampled according to Tables 1 and 2 of SRP 11.5.
Informa tion provided in Section 11.5 of the FSAR indicates that the Seabrook Station does not meet this criteria in the following areas:
a)
Plant vent does not contain a continuous radiation nonitor for noble gas ef fluents. (see question above)
.b)
Containnent purge lines do not contain a process nonitor 1
nor the capability to isolate the purge line on a high radiation noni tor.
(Note: Area noni tors are not an ef fec-tive means for meeting 10 CFR Part 20 unrestricted area airborne concentration limits.)
c)
The fuel storage building does not contain a process monitor f rom it exhaust to the plant vent.
d)
The turbine gland steam condenser exhaust is discharged to the atmosphere unmoni tored.
c)
The turbine building sumps are to release their contents on a batch basis with only a sample taken and analyzed prior to release.
Since there is no means to isolate the sunp and since the release is not monitored, a monitor is required for turbine building ef fluent along with an automatic control feature to isolate the discharge on a high radiation si gnal.
. f)
A gross radioactivity monitor is required for the service water ef fluent line.
I g)
The capability to obtain a grab sanple in the stream from the following sources has not been provided:
(1) containment purge t
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-en s-e n
1:
(2) PAB ventilation system (3) fuel storage building (4) waste procesing building area handling radwaste (5) turbine gland steam condenser 1
(6) evaporator vent system (i.e., distillate coolers) i (7) SG flash tank distillate cooler (8) pressurizer and BRS vent systems (9) component cooling water system Commit to the installation of the above process and effluent monitors and the sampling of the above sources.
460.36 Does the design of the process and ef fluent monitoring systems meet the guidelines of Appendix 11.5-A of SRP 11.5, Regulatory Guide 4.15 (Position C), Regulatory Guide 1.97 (Position C and Table 2)?
460.37 Does the process and ef fluent nonitoring system have the capability to replace or decontaminate on-line nonitors without opening the pro-cess system or losing the capability to isolate the ef fluent stream?
460.38 Review of the FSAR provides no information with respect to the col-lection of radioactive wastes in the turbine building and their disposal. On page 11.2-7 of the FSAR it is indicated that the turbine building sump will collect radioactive wastes in this building.
Table 11.5-3 of the FSAR implies that a sample will be taken of this sump and that the sump' will be released on a batch basis.
From
^
Table 11.5-2 of the FSAR it can be inferred that there is no nonitor on the discharge from the sump.
\\
N
Provide the appropr'iate PAIDs which show the various flow inputs to the turbine building sump.
It is the staf f's position that unless a turbine building sump can be i.;olated such that no flow is allowed into the sump af ter a sample is taken, then this discharge point requires a radiation monitor and in addition an autonatic control feature which isolates the discharge on a high radiation signal.
i 460.39 It would appear that radiation nonitor RE-6503 has been labeled in-correctly in Table 11.5-1 hased upon figure 11.3-1 sheet 2.
The title would be discharge from carbon delay beds.
Please verify this.
i 460.40 How does.one tell fro.: the Seabrook F&Ius wnetner a monitor alarms
. locally or in the control room or in both locations? This also applies l
~ he location of indications for pressures, flow rates, radiation t
levels, etc.
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i 480.5 According to Section 6.2.1.2.6.1, the eight blo'cks that make up (6.2.1.2) the neutron shield are designed to rotate away from the reactor vessel due to the pressure rise in the reactor cavity from a t
LOCA.
Since this is a vent fios path that is not irmediately available at the time of pipe rupture, provide the following l
1 information.
(1)
Verification that the vent area and resistance to flow as a function of time af ter the break is based on a dynamic analysis of the subcompartment pressure response to the i
pipe rupture; (2)
Experimental data to support the validity of the dynamic analysis; and (3)
Analysis of the effects of any missiles that may be gener-ated by the rotation of the eight neutron shield blocks.
480.6 Describe the design provisior.s to ensure that the closure of (6.2.4) the containment mini-purge system isolation valves will not be prevented by debris which may become entrained in escaping con-tainment atmosphere.
To prevent debris from entering the mini-purge lines, debris screens-with the following characteristics should be installed:
r (1)
The debris screen should be seismic Category I design and installed about one pipe diareter away from the inner side of the inboard isolation valve.
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( 2)
The piping between the debris screen and the isolation valve should also be seismic Category I design.
(3)
The debris screen should be designed to withstand the LOCA dif ferential pressure.
(4)
The debris screen should be designed similar to that shown in the attached Figures 1 and 2.
480.7 Section 6.2.1.2.c.3.b of the FSAR states that a nodalization (6.2.1.2) sensitivity study on the pressurization analysis for the reac-tor cavity will be presented later.
Discuss your plans for pro-viding this information.
480.8 The results of your subcompartment analysis are incomplete.
(6.2.1.2)
We will require the following information in order to complete our review:
(a)
Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node.
Discuss the basis for establishing the dif feren-tial pressure on structures and components.
(b) For the compartment structural design pressure evaluation, provide the peak calculated differential pressure and time of peak pressure for each node.
Discuss whether the design differential pressure is uniformly applied to the compartment structure or whether it is spatially varied.
If the design differential pressure varies depending on the I
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proximity of the pipe break location, discuss how the vent areas and flow coefficients were determined to as-sure that regions removed from the break location are conservatively designed.
(c)
Provide a schematic drawing showing the compartment nodalization for the determination of maximum structural loads, 'and for the component supports evaluation.
Pro-tide sufficiently detailed plan and section drawings for several views, including principal dimensions, showing the arrangement of the compartment structure, major com-ponents, piping, and other major obstructions and vent areas to permit verification of the subcompartment nodali-zation and vent locations.
(d )
Provide the peak and transient loading on the major compo-nents used to establish the adequacy of the supports design.
This should include the load forcing functions (e.g., f (t),
X f (t), f (t)) and transient moments (e.g., M (t), M (t),
y z
x y
M (t)) as resolved about a specific, identified coordinate z
system.
Provide the projected area used to calculate these loads and identify the location of the area projections on plan and section drawings in the selected coordinate system.
This information should be presented in such a manner that confirmatory evaluations of the loads and moments can be made.
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480.9 The mass and energy release analysis for postulated LOCAs in FSAR
( 6.2.1)
Section 6.2.1.3 is based on a Westinghouse report, NS-TMA-2075,
" Westinghouse LOCA Mass and Energy Release Model for Containment Design." This report is currently under staff ' review and pre-sently has not been accepted.
Provide a comparison for the worst case LOCA of containment pressure and temperature response using the mass and energy release data based on NS-TMA-2075 and on the previously at-e cepted methodology on WCAP-8312A.
480.10 Verify that the Containment Recirculation Sump Screens are (6.2.2)
Seismic Category I and protected from the effects of pipe breaks.
480.11 Verify that the Containment Enclosure Emergency Cleanup System (6.2.3.3) is supplied by the Emergency Power Supply.
480.12 Describe the provisions for leak rate testing of the secondary (6.2.3) containment bypass leakage.
Include in your description any test necessary to measure the secondary containment bypass leakage.
480.13 Verify that all openings in the secondary containment (such as (6.2.3) personnel doors and equipment hatches) are equipped with alarms and position indicators that annunciate in the control room.
480.14 Either provide the containment isolation system information (6.2.4) identified as "later" in Table 6.2-83, or provide a schedule for submittal of this information.
. c.
i 480.15 Verify that the Hydrogen Recombiners are seismic Category I, (6.2.5) powered f rom Class IE electrical buses and desi,gned to func-tion in a post-accident environment.
480.16 Provide the maximum allowable leakage rate and the inleakage (6.2.6) limits for the secondary containment.
480.17 According to Table 6.2-83, the Main Feedsater Isolation Valves (6.2.4) remain open af ter an accident.
According to Section 6.2.1.1 the Main Feedwater Isolation Valves are isolated after an at-cident.
Correct this discrepancy in your FSAR.
480.18 Provide the parameters sensed for the containment ventilation (6.2.4) isolation signal.
480.19 Closed engineered safety feature systems outside containment (6.2.6)
(e.g., the emergency core cooling system (ECCS), containment spray system) will become extensions of the containment boundary following a LOCA. Discuss the capability for leak testing, the closed systems located outside containment and including the leakage in the off-site dose calculations.
480.20 Identify any branch lines that are located between the contain-(6.2.4) ment and an outside isolation valve.
Identify the General De-sign Criteria used to satisfy the isolation requirements for these branch lines.
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480.21 With regard to the leak rate testing of containment isolation (6.2.6) valves in the purge / vent systems, it is our position that ac-tive purge / vent systems (i.e., systems operated during plant operating Modes 1 through 4) be leak tested at least once every three months and passive purge system (i.e., systems adminis-tratively controlled closed during plant operating Modes 1 through 4) be leak tested at least once every six months.
Please indicate your approach to complying with this position.
480.22 Provide a description of the instrumentation used to monitor containment pressure, containment water level and containment s
hydrogen concentration.
In your description of containnent in-strumentation provide the information requested in Section II.F.1 of NUREG-0737.
480.23 Provide a discussion of Containment Isolation Dependability as requested in Section II.E.4.2 of NUREG-0737.
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