ML17054A549

Summary of 840224 Meeting W/Util Re Administrative Matters Concerning Closing Out Draft SER Open Items

ML17054A549
Person / Time
Site:	Nine Mile Point
Issue date:	03/05/1984
From:	Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8403210213
Download: ML17054A549 (64)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 5, 1984 Docket No.:

50-410 APPLICANT:

FACILITY:

SUBJECT:

Niagara Mohawk Power Corporation (NMPC)

Nine Mile Point, Unit 2 SUMmARY OF MEETING WITH NMPC TO DISCUSS ADMINISTRATIVE MATTERS CONCERNING CLOSING OUT OPEN ITEMS ON NINE MILE POINT, UNIT 2 On February 24, 1984, the NRC staff met with representatives from NMPC to discuss administrative matters concerning closing out open items on Nine Mile Point, Unit 2 (NMP-2).

During the meeting, NMPC was given a list of open items which have been identified by the staff and are expected to be included in the Draft SER for NMP-2.

This list is included as Attachment 1.

NHPC outlined the steps to be taken by NMPC and their consultants in closing out open items.

NMPC provided a schedule for responding outstanding questions from the NRC staff (Attachment 2).

NMPC was requested to review the open items on the list, and the Draft SER when it is issued, to group the open items for discussions at meetings with the NRC staff.

These meetings are to be held to close out the open issues.

In order to facilitate the closing out of open issues NMPC was requested to develop a faster system for submitting formal responses to open issues.

The questions to be used as a basis for the Mechanical Engineering Branch (HEB)

SER review were provided to NHPC (Attachment 3).

This meeting 'was tentatively scheduled for March 27-29, 1984, but will be postponed until early June.

The next Instrumentation and Controls meeting was scheduled for April 4-5, 1984.

gualification testing of Diesel Generators was also discussed.

Additional details of differences between NMP-2 and Susquehanna and Zion Diesel Generators was requested.

r 84032i02i3 840305 PDR ADQCK 05000410 PDR

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Attendants at the meeting were as follows:

A.

N.

M.

J.

E.

Zallnick, NMPC Rademacher, NMPC

Haughey, NRC Lazevnick, NRC*

Tomlinson, NRC*

• diesel generator discussion only.

~keaL oigeog +g Attachments:

As stated Mary F. Haughey, Project Manager Licensing Branch No.

2 Division of Licensing cc w/ attachments:

See next page MFHagghey:pt ASchyencer 3/~/84 3/Q/84

1

Nine Mile Point 2

Mr. Gerald K. Rhode Senior Vice President Niagara Mohawk Power Corporation 300 Erie Boulevard West

Syracuse, New York 13202 CC:

Mr. Troy B. Conner, Jr.,

Esq.

Conner 8 Wetterhahn Suite 1050 1747 Pennsylvania

Avenue, N.W.

Washington, D. C.

20006 Mr. Richard Goldsmith Syracuse University College of Law E. I. White Hal 1 Campus

Syracuse, New York

~ 13210 Mr. Jay Dunkleberger, Director Technol ogi ca 1 Devel opm nt Programs New York State Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Ezra I. Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Resident Inspector Nine Mile Point Nuclear Power Station P. 0. Box 99

Lycoming, New York 13093 Mr. John W. Keib, Esq.

Niagara Mohawk Power Corporation 300 Erie Boulevard West

Syracuse, New York

. 13202 Jay M. Gutierrez, Esq.

U. S. Nuclear Regulatory Conmission Region I 631 Park Avenue King of Prussia, Pennsylvania 19406

I)

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complete its review of these items before staff will discuss the resolution of each report.

These items are listed below and of this report as indicated.

ATTACHN/gT, l~,

the operating license is issued.

The of these items in a supplement to this are discussed fur.her in the sections issue OSER Section

( 1) nearest population center 2.1 (2) long-term diffusion estimates 2.3.5 (3) seismic design of revetment ditch and flood protection berms 2.4.10 (4) protect i on against PMP 2.4.2.2 (5) protection of the main stack from wave forces from PNWS 2.4.10 (6) adequacy of the ultimate heat sink 2.4.11.2 (7) ground water level 2.4.12.2 (8) analysis of postulated rupture of a liquid radwaste tank 2.4.13 (9) recalculation of the changing stresses at the site, assuming shallower burial depths than in the original calculations 2.5.1, 2.5.2

( 10) an evaluation of the significance of the decoupled regional stress regimes in the Paleozoic and basement rocks measured in the site region 2'.1

( 11) assessment of seismic or aseismic origin of sedimentary structures 2.5.1, 2.5.2 02/21/84 Ni4IP-2 DRAFT SER

Issue DSER Section CaPS (12) monitoring program of the cooling tower fault designed to ascertain the strain or displacement rate on the faul t 2.5.1, 2.5.2 (13) magnitude of the fault movement for all seismic Category I structures in the power block 2.5.4.5.1 (14) excavation, backfill and geological mapping data of the main stack 2.5.4 '.3 (15) liquid faction potential analysis for the Category I

electrical duct bank and manhole 2.5.4.7

( 16) update of slope inclinometer and rock extensometer data 2.5.4.10

( 17) dynamic stability of the slopes of the revetment ditch 2.5.6.2.3, 2.5.6.2.4 (18)

PMP flood protection berm 2.5 '.3

( 19) turbine maintenance 3.5.1.3 (20) adequacy of tornado missile protection for diesel generator exhaust outside air intakes for HVAC systems safety-related buried piping 3.5.2 (21) effects of postulated pipe breaks 3.6.1 (22) stress and cumulative usage factor limits and inspection requirements for piping inside the break exclusion zone 3.6.2 (23) postulation of moderate energy cracks inside containment 3.6.2 and of high-energy cracks 02/21/84 1-15 NNP"2 DRAFT SER

Issue OSER Section (24) postulation of pipe ruptures 3.6.2 (25) feedwater isolation check valves 3.6.2 (26) design of pipe rupture restraints 3.6.2 (27) vertical floor flexibility in the seismic analysis 3.7.2 (28) results of the concrete containment ultimate capacity analysi s 3.8.1 (29) containment response to SRV/pool dynamic loads 3.8.1 (30) deviations from the applicable provisions of ASME Section III, Division 2 3.8.1 (31) deviations from the applicable requirements of ACI 349 as amended by RG 1.142 3.8.3, 3.8.4, 3.8.5 (32)

SRV/pool dynamic loads on containment interior structure 3.8.3 (33) consideration of upward seismic load effects in the foundation stability analysis of the screenwell building 3.8.5 (34) structural audit action items 3.8.6 (35) systems and locations to be monitored during the pre-operational testing program 3.9.2.1 (36) acceptance criteria for observed or measured vibration levels 3.9.2.1 02/21/84 1-16 NMP-2 DRAFT SER

Issue L"..

OSER Section (37) inclusion of all essential safety-related instrument lines in the vibration monitoring program 3.9.2.1 (38) seismic design of HVAC systems 3 '.2.2 (39) seismic methods used for the analysis of the safety-related piping in pipe tunnels 3.9.2.2 (40) documentation of analysis for combined lo'ads (LOCA and SSE) 3.9.2.4 (41) methodology of combining loads 3.9.3.1 (42) clarification of the BWR Mark II hydrodynamic loads 3.9.3.1 (43) assurance that downcomers will not develop fatigue cracks 3.9.3. 1 (44) design of piping and supports in the wetwell area.

3.9.3.1 (45) design of SRVs and attached discharge piping 3.9.3.2 (46) design and construction of ASME Class 1," 2 and 3

component supports 3.9.3.3 (47) stress categories and limits for core support structures and the applicable codes used for evaluation of the faulted condition 3.9.5 (48) response to IE Bulletin 80-07 3.9.5 (49) leak rate testing of isolation valves

3. 9'. 6 (50) preservice and inservice tes.ing of pumps and valves 3.9.6 02/21/84 1-17 NMP-2 DRAFT SER

Issue DSER Section (51) seismic and dynamic. equipment qualification program (52) pump and valve operability assurance 3.10 3.10 (53) dependability of containment isolation (purge valves) 3.10 (54) performance testing of relief and safety valves (II.D.1) 3.10 (55) qualification of accumulators on automatic depressurization system valves

( II.K.3.28) 3.10 (56) long-term operability of deep draft pumps 3.10 (57) environmental qualification of equipment 3.11 (58) irradiation fuel surveillance program 4.2 (59)

LPNS ( loose parts monitoring system) 4.4.6 and (Table 4.4.0)

(60) inadequate core cooling detection system (II.F.2) 4.4.7 (61) pipe break in the BWR scram system (62) lead factors in surveillance capsules 5.3.1 (63)

P-T (pressure-temperature) curves 5.3.2, 5.3.3 (64) ratio of neutron flux density of specimens in the surveillance capsule to peak neutron flux density at RPY 5.3.3 (65) reactor coolant pressure boundary inservice inspec ion and testing 5.2.4 02/21/84 1-18 NAP-2 DRAFT SER

Issue DSER Section (66) fracture prevention of containment pressure boundary 6.2.7 (67) control room habitability 6.4 (68) exceptions and deviations to RG 1.52, Rev.

2 6.5.1.5 (69) fission product control systems (70) inservice inspection of Class 2 and 3 components 6.5.3 6.6 (71) spent fuel storage pool materials surveillance (72) spent fuel storage pool materials surveillance (73) spent fuel pool design (74) light load handling.,system 9.1.4 (75) heavy loads 9.1.5 (76) failure of nonseismic buried pipe near safety-related buried pipe (77) backup nitrogen supply system (78) periodic air quality testing (79) flooding by rupture of nonseismic Category D piping,

vessels, or tanks or by failure of a backflow preven-tion device in the drainage system 9.3.3 (80) postaccident sampling (Ii.B.3) 9.3.2 02/21/84 1-19 sVMP-2 DRAFT SER

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Issue DSER Section (81) drainage of leakage water away from safety-related components or systems 9.3.3 (82) design capability of the CB HVAC system (410. 41 and 410. 42) 9.4.1 (83) protection against hydrogen accumulation in the battery rooms 9'.1 (84) outdoor temperatures assumed for sizing of the CB HVAC 9.4. 1 (85) spent fuel pool area ventilation system 9.4.2 (86) tornado mi ssle protection for diesel generator building louvers 9.4.4 (87) diesel generator building HVAC system conformance to GDC 4

9.4.4 (88) protection of essential electrical components from failure due to the accumulation of dust and particulate material 9.4.4 (89) potential systems interaction 9.5.1.II.B (90) administrative controls 9.5.1.III (91) fire brigade and fire brigade training 9.5.1.IV i

(92) qualification of fire doors

9. 5. 1.V. A (93) floor drains 9.5. 1. V.A 02/21/84 1-20 NMP-2 DRAFT SER

Issue OSER Section (94) safe shutdown capability 9.5.1.V.B (95) alternate shutdown capability 9.5.1.V

~ B (96) emergency lighting (97) installation of fire detectors 9.5.1.V.G

9. 5. 1. VI.A (98) qualification of the electric fire pump 9.5. 1. VI.A (99) valve supervision

9. 5. 1. VI.B (100) quality group classification information on the design of the turbine gland sealing system 10.4.3.5 (101) protection of safety-related systems from flooding~

~g from a postulated failure of a circulating water expansion joint or line failure as a result of an SSE 10.4.5 (102) parameters used for calculating liquid and gaseous source terms 11.1.2

( 103) assessment of the capability of liquid and gaseous radwaste systems for keeping the levels of radioactivity in effluents ALARA (104) assessment of charcoal absorber tank failure for

'10 CFR 100 dose guidelines 11.2.1, 11 '

2 11.3.1 11.3.1

( 105) process control program for the solid radwaste system 11.4.1, 11.4.2 (106) compliance program to meet 10 CFR 61 11.4.2 02/21/84 1-21 iVYip-2 DRAFT.;SER

1

Issue I

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OSER Section (107) high-range noble gas monitor (II.F.1) 11.5 (108) airborne radioactivity levels (471.1) 12.2 (109) airborne radionuclide concentration in liquid radwaste handling area (471. 3) 12.2 (110) conformance to RGs 1.8, 8.8, and 8.10 (471.4) 12.0 (111) dose rate criteria (II.B.2) (471.9) 12.3.2 (112) projected doses to individuals and dose rate maps (471.16) 12.3.2 (113) whole-body dose calculations (471.17) 12.3.2 (114) postaccident access and shield design review (471.19) 12.3.2 (115) crud buildup (471.19) 12.3.2 (116) postaccident vital area monitors (471.19) 12.3.2 (117) compliance with TMI II.B.2, shielding 12.3.2 (118) inhalation exposure (471.12) 12.4.2.2 (119) estimate of N-16 dose contribution (471.13) 12.4.2.2, (120) estimate of doses outside of plant structures (471. 14) 12.4.2.2 (121) dose assessment (471.11) 12.4 02/21/84 1-22 iVMP 2 ORAFT SER

Issue OSER Section (122) separation of health physics and chemistry functions (471.21) 12.5.1 (123) qualifications of Superintendent, Chemistry and Radiation Management (471.21)

(124) qualifications of temporary RPMs and commitment to ANSI 3.1 12.5.1 (125) training of health physics technicians 12.5.1 (126)

ANSI 18. 1 qualified health physics technician 12.5.1 (127) initial training program 13.2.1.1 (128) requalification training program 13.2.1.2 (129) immediate upgrading of reactor operator and senior reactor operator training and qualifications 13.2.1.4 (130) administration of training programs 13.2.1.4 (131)

STA training program 13.2.2 (132) emergency planning 13.3 (133) commitment to Section 5.3 of ANSI/ANS 3.2 13.5.1.1 (134) evaluation and development of procedures for tran-sients (I.C.1) 13.5.2C (135) upgraded emergency operating procedures 13.5.2 02/21/84 NMP-2 GRAFT SFR

Issue DSER Section (136)

NSSS vendor review of procedures (I.C.7) 13.5.2 (137) pilot monitoring of selected emergency procedures for NTOLs (I.C.8) 13.5.2 (138)

ATMS procedures 13.5.2 (139) licensed operator training program 13.2.1.1 (140) simulator training 13.2.1.1 (141) requalification (142) loss-of-air-supply tests (640.06) 14.2.7 (143) single-failure-proof cranes (NUREG-0612) and heavy load testing (NUREG-0554) (640.07) 14.2 (144) periodic testing of diesel generators (RG 1.108)

(640.08) 14.2.7 (145) applicability of RG 1

~ 140 to radwaste building exhaust (640.09) 14.2.7 (146) preoperational test abstracts (640.10, 640.13, 640.15, 640.16, 640.17, 640.19, 640.20, 640.21) 14.2.12 (147) protection of control room operators against accidental chlorine release (640. 18) 14.2.12 (148) startup test abstracts (640.23, 640.24, 640.26, 640.27, 640.29)

.14.2.12 02/21/84 1-24 NHP-2 DRAFT SER

~S

, ',i!!, 'I Issue DSER Section (149) incorporation of specific testing identified into test abstracts (640.34) 14.2.12

( 150) preoperational tests to be conducted after fuel load and tests to be exempted from prior notification (640.35) 14.2.12 (151) fuel handling accident 15.

(152) loss-of-coolant accident 15.

(153) leakage integrity from systems outside containment (III.D.F 1) 15.9.5 (154)

DCDR Summary Report (I.D.1)

18. 0 (155)

SPDS safety analysis and implementation plan (I.D.2) 18.0

( 156) Technical Specifications 16.0 (157) safeguards 1.9 Confirma or Issues At this point in the review there are some items that have essentially been resolved to the staff's satisfaction, but for which certain confirmatory information has not yet been provided by the applicant.

In these instances, the applicant has committed to provide the confirmatory information in the near future.

If staff review of the information provided for an item does not confirm preliminary conclusions, that item will be treated as open and the staff will report on-its resolution in a supplement to this report.

02/21/84 1-25 NiYP-2 DRAFT SER

v

Issue Section (1) sharing of fuel handling and fuel storage (2) long-term diffusion estimates 2.3.5 (3) fuel rod fracturing 4.2 (4)

Gadolinia thermal conductivity equation (incorporation in GESSAR II calculations) 4.2 (5) fire protection training 13.2.2 (6) full reactor isolation test abstract (640.28)

(7) test description for confirmatory in-plant tests of SRVs (640.30) 14.2.12

14. 2. 12 (8) loss of turbine generator and offsite power (640.32) 14.2.12

1. 10 License Condition Items There are certain issues for which a license condition may be desirable to ensure that staff requirements are met during plant operation.

The license condition may be in the form of a condition in the body of the operating licenses or a limiting condition for operation in the Technical Specifications appended to the license.

Item Section

( 1) activation of the rack bar heating system when lake temperature drops (TS) 2.4.7 "To be incorporated in the Technical Specifications appended to ne license.

02/21/84 1-26 HNP-2 DRAFT SER

Issue l~-j".-'] i Section (2)

Limiting conditions for operation, 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 (TS) 3.9.6 (3) fuel rod internal pressure criterion 4.2 (4) stability analysis prior to operation beyond Cycle 1

4.4.4 (5) crud deposition (TS) 4.4.5 (6) single loop operation (TS) 4.9 (7) natural circulation (TS) 4.9 (8) operation and surveillance of LPMS (TS) 4.9 (9)

MSIY leakage 6.7 (10) operation while RHR is in pool cooling mode (TS) 9.1.3 (11) Following the first refueling outage the applicant shall have made commitments acceptable to the staff regarding the guidelines of Se'ction

5. 1.2 through 5. 1.6 of NUREG-0612 (Phase II -

9 month responses to the NRC generic letter dated Oecember 22, 1980) 9.1.5 1.11 Unresolved Safet Issues Section 210 of the Energy Reorganization Act'f 1974, as

amended, reads as follows:

02/21/84 1-27 NMP-2 ORAFT SER

0

STATUS OF RESPOIISES

'IO BRANCH TECIINICAL VEST IOIIS TECIIIIICAL UESTlotf AREAS ATTACNflEIIT 2 P lplNG 210 STRUCTURE/220 1tlSTRUMENT RADIOLOGIC SE ISHlC-230 EQUIPHEttT AND

470, 471 GEOLOGY-231 QUAL.

COIITROLS*

EFFLUENT HYDROLOGY-240 270 QA 421 ER-OLS 460 GEOTECII-241 271 260 OTIIER CONTAltlHEtlT 100,250,251 FIRE POIIER REACTOR PIIYSICS REACTOR AUXILIARY 252,281,311 PROTECT lON SYSTEH CORE PERFORM SYSTEM SYSTEH 450,451,620 200 430 400 491 492 440 110 630 640 330 TOTALS I OF QUESTIOtlS RECEIVED RESPONSES COMPLETED I OF OUTSTAIIDltfG RESPONSES 46 51 46 46 30 16 103 96 8

50 29 99 1

4 19 13 51 33 118 66

'4 12 49 51 43 42 106 85 21 733 582 151 SCIIEDULE FOR RESPONSE COMPLETION NUMBER OF RESPONSES 1904 4IAIIUARY IIAIICII

.IUNI.

E I' EMIIFlf ffECU IIIElf 1

3 10 16 32 29 20 1985 ftARCII JUIIE-DECEMBER Responses will be discussed in meetings with the comntsston scheduled to co3n33ence tn February.

ATTACHMENT 3 MECHANICAL ENGINEERING BRANCH NINE MILE 2 SER QUESTIONS SECTION 3.6.2 210.17 Branch Technical Position MEB 3-1 requires that certain stress and cumulative usage factor limits be met in the break exclusion zone.

The criteria contained in the FSAR are not in compliance with these limits.

Provide justification for the criteria used.

In particu-lar, address those cases for Equation (10) exceeding 2.4 S

and the cumulative usage factor exceeding 0.1 for Class 1 piping.

210.18 Have occasional loads been considered in the evaluation of the sum of Equations (9) and (10) when comparing to the limits for, Class 2

piping in the break exclusion area?

210. 19 Provide assurance that 100% volumetric inservice examination of all pipe welds in the break exclusion area will be conducted during each inspection interval as defined in IlIA-2400, ASME Code,Section XI.

210.20 Breaks in non-nuclear high energy piping not seismically'analyzed (nor qualified) should be postulated at those locations which produce the greatest effect on an essential component or structure irrespective of the fact that the high stress or fitting criteria might not require a break to be postulated.

Provide assurance that the above criteria have been met.

210.21 What criteria are used for postulating moderate energy leakage cracks inside containment?

210.22 Discuss how high energy leakage cracks are considered?-

210.23 Discuss how pipe whip and jet impingement effects were determined for those postulated breaks in high energy piping that is not restrained.

l 210.24 Provide assurance that the tip deflection of a restrained whipping pipe does not adversely affect nearby safety-related components from performing their safety-related function.

210.25 Describe in more detail the design procedures and methodologies used in the jet impingement 'analyses.

Specifically, address

1) the jet loads and jet configurations used for circumferential and longitud-inal breaks, 2) how targets are determined, and 3) the acceptance criteria used to evaluate the effects on safety-related components and structures.

210.26 Provide the criteria used in the design of pipe rupture restraints including the auxiliary steel used to support the pipe rupture restraint; Provide.assurance.

that "the. pipe -rupture restraint and-.

supporting structure cannot fail during a seismic event.

210.27 Provide the design criteria for pipe rupture restraints that also support piping.

210.28 In order to assure the pipe break criteria has been properly implemented, the Standard

. Review Plan requires the

- review of sketches showing the postulated rupture locations and summaries of 44 the data developed to select postulated break locations including, for each point, the calculated stress intensity, the calculated cumulative usage factor, and the calcul.ated primary plus secondary stress range.

The vast majority of this information in the FSAR is either preliminary or incomplete.

Please provide a schedule for the completion of Tables 3.6A-2 through 3.6A-60 and Figures 3.6A-12 through 3.6A-39.

210.29 On page 3.6A-28 an amplification factor of between 1.0 and 1.1 to

account for pipe rebound is discussed.

Provide justification for the use of an amplification factor of less than 1.1.

210.30 Provide justification for shape factors of, less than unity as discussed on page 3.6A-35 of the FSAR.

210.31 Appendix 3C is very incomplete.

Provide a schedule for its completion.

210.32 Provide a list of all instances where full break area opening times in excess of one millisecond were used.

See page 3.6B-8.

210.33 Provide justification for using a thrust coefficient of less than 1.26 for saturated-steam-and-2-;Q

-for subcooled, water. as...discussed.

on page 3.6B-9.

210.34 Provide a list of all instances where mechanistic approaches were used to reduce break areas as discussed on page 3.6B-6.

210.35 Provide the basis for assuring that the feedwater isolation check valves can perform their function following a postulated break of, the feedwater line outside containment.

210.36 Discuss the types of protection used to mitigate the effects of jet impingement on safety-related components and structures.

SECTION 3.9.2 210.37 Provide the acceptance criteria to be used in determining if the vibration loads observed or measured during the pre-operational

4 testing are acceptable.

Specifically, address how the vibration amplitudes will be related to a stress load and what stress levels will be used for both steady-state and transient vibration.

It is the staff's position that all essential safety-related instru-mentation 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 instru-mentation lines that will be included in the initial test vibration monitoring. program.-

The essential instrumentation lines to be inspected should include (but are not limited to) the following:

a.

Reactor pressure vessel level indicator instrumentation lines (Used for monitoring both steam and water levels).

b.

Main steam instrumentation lines for monftoring main steam flow (used to actuate main steam isolation valves during high steam flow).

c.

Reactor core isolation cooling (RCIC) instrumentation lines on the RCIC steam line outside containment (used to monitor high steam flow and actuate isolation).

d.

Control rod drive lines inside containment (not normally pressurized but required for scram).

Due to a long history of problems dealing with 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. maintenance records for snubbers be documented as follows:

Pre-service Examination 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 to initial system pre-operational

testing, and should as a minimum verify the 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

'eaking from the snubber system.

6.

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

~I

During pre-operational

testing, snubber thermal movements 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.

b.

For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.

c.

Verify the snubber swing clearance at specified heatup and cooldown intervals.

Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.

The above described operability program for snubbers should be included and documented by the pre-service 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.

210.39

. Please provide a statement as to the compliance with NUREG-0619, "BMR Feedwater Nozzles and Control Rod Drive Return Line Nozzle Cracking".

210.40 Provide the basis used for the design of piping supports and anchors which separate seismically designed piping and non-seismic Category I piping.

Include in your discussion the loads and load combina-tions used and how the local pipe wall stresses are considered.

210.41 Describe the design considerations given to assure that an

adequate number of modes have been used in the dy'namic piping anlayses performed for:

1) seismic loadings, 2)

SRV loadings, r

3)

LOCA loadings, and

)

4) hydraulic transients (e.g.

steam and water-hammer.

210.42 Explain how in the design process the reinforcement thickness of branch connections are determined for both internal pressure and mechanical loads and incorporated into the fabricated piping.

Provide assurance that all branch connections decoupled from the main run piping on the piping analytical model are designed and fabricated to the required reinforcement area.'10.43 The staff finds insufficient information describing the design of safety-related HVAC ductwork and supports.

Provide the design basis used for qualifying, the HVAC ductwork and support structural integrity.

SECTION 3.9.3 210.44 Provide the basis for assuring that ASME Code Class 1, 2, and 3

piping systems are capable of performing their safety function under all plant conditions.

Describe the methodology used to assure the

s

~

functional capability of essential piping system when service limits C or D are specified.

provide a discussion of the design considerations used for sa'fety and relief valve 'oads and piping reactions.

Include in your discussion the basis for assuring that the valve end loads and the support arrangement for the affected piping are acceptable.

Describe those short-term and long-term actions being taken to preclude the occurrence of cracking in jet pump hold down beams as described in IE Bulletin 80-07.

Describe briefly the design considerations given to the piping stress analyses-for-the-mainsteam-piping-and-attached" safety *relief valve discahrge piping for the alternate shutdown cooling mode.

Specifically address the capability of the spring hangers to accomodate the additional weight of water during this mode.

Provide assurance (or a commitment) that the design of all safety-related mechanical components and their supports can withstand the effects of safety-relief valve discharge laods as defined in NUREG-0802, "Safety/Relief Valve quencher Loads Evaluation'or BWR Mark II and III Containments."

Provide assurance (or a commitment) that the design of all safety-related mechanical components and their supports can withstand the effects of loss-of-coolant accident loads as defined in NUREG-0808, "Mark II Containment Program Load Evaluation and Acceptance Criteria."

Provide the basis for assuring that a fatigue crack will not occur

1) in the safety relief valve discharge piping in the suppression pool wetwell airspace and 2) in the suppression pool downcomers.

The staff requests that ASME Code Class 1 piping fatigue evaluation

6

be performed and should include all cyclic loadings due to normal operation,

seismic, SRV discharge, LOCA chugging, and condensation oscillation loads (as appropriate).

Porvide for the staff review, typical examples of the amplified building response spectra used in the design of piping systems and including the following loadings:

a) seismic OBE b) seismic SSE c)

SRV loads d) LOCA-related loads Briefly, describe the attenuation of the hydrodynamic loads in the C

pl ant and. describe-to what"extent saf tey-rel ated-components-arc--'

designed to these loadings in the various areas in the plant.

Oescribe the design considerations given to the piping in the suppression pool wetwell with respect to stability of the piping and its supports during a

LOCA pool swell event.

Using the guidance of NUREG-0609, provide the methodology used and the results of the annulus pressurization (AP) analysis (asymmetric LOCA loads) for the reactor system and affected components including the following:

1.

reactor pressure vessel and supports, 2.

core supports and other reactor internals, 3.

control rod drives, 4.

ECCS piping attached to the reactor coolant system, 5.

primary coolant piping, and 6.

piping supports for affected piping systems.

N The results of the above analysis should specifical'ly address the effects of the combined loadings due to annulus pressurization and an SSE.

210.55 The staff review finds insufficient information regarding the design of component supports.

Per SRP Section 3.9.3, our review includes an assessment of design and structural integrity of the supports.

The review addresses three types of supports:

(1) plate and shell, (3) linear, and (3) component standard types.

For each of the above three types of supports, provide the following information (as applicable) for our review:

(a)

Describe (for typical support details) which part of the support is designed and constructed as component supports and which part is designed and constructed as building steel (NF vs AISC jurisdictional boundaries).

(b)

Provide the complete basis used for the design and construction of both the component support and the building steel up to the building structure.

Include the applicable codes and standards used in the design, procurement, installation, examination, and inspection.

(c)

Provide the loads, load combinations and stress limits used for the component support up to the building structure.

(d)

Provide the deformation limits used for the component support.

(e)

Describe the buckling cri'teria used for the design of component support.

210.56 The staff's review of your component support design finds that additional information is required regarding the design basis used for bolts.

(a)

Describe the allowable stress limits used for bolts in equip-ment anchorage, component supports, and flanged connections.

(b)

Provide a discussion of the design methods used for expansion anchor bolts used in component supports.

(c)

Identify where in the plant high strength bolts have been used.

210.57 It is the staff's position that for the design of component

supports, stresses produced by seismic anchor point motion of piping and the thermal expansion of piping should be categorized as primary stresses.

Confirm that Nine Mile Point 2 meets this criteria.

210.58 Provide a discussion of the use of stiff pipe clamps as addressed in IE Information Notice 83-80.-.-"= "

210.59 Provide assurance that any snubbers used as a vibration arrestor has properly considered the cyclic loadings which might cause fatigue failure.

210.60 Yalve discs are considered part of the pressure boundary and as such should have allowable -stress--limits.

Provide these limits for our review.

210.61 Provide the stress categories and limits for core support/structures and include the applicable codes used for evaluation of the faulted condition.;

SECTION 3.9.6 210.62 There are several safety systems connected to the reactor coolant

pressure boundary that have design pressure below the rated reactor coolant 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 testing to prevent exceeding the design pressure of the low pressure systems.

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 corrective 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.

Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and for systems rated at less than 50K of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.

The testing interval should average to be approximately one year.

Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.

The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute (GPM)

0 for each valve to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.

Significant increases over this limiting value would be an indica-tion of valve degradation from one test to another.

The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.

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 vs'olation-valves included-in-your-- '"--"

testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isola-tion valves.

Also discuss in detail how your leak testing program will conform to the above staff position.

Provide a schedule for completion of your program for inservice testing of pumps and valves including any request relief from ASME Section XI requirements.

V

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14arch 5, 1984 MEETING

SUMMARY

DISTRIBUTION:

~Document Control (50=410) j

(-LPDR NRC PDR NSIC PRC LBk'2 File

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EHylton MHaughey Region I Bordenick, OELD JLazevnick (w/o attachments)

ETomlinson (w/o attachments)

DTerao (w/o attachments)

HBrammer (w/o attachments)

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