ML19270F211

Forwards List of Category I-IV Matters Applicable to Preliminary Design Approval for SWESSAR/RESAR-41 Application.Requests Design Assessment for Each Matter Identified in Lists,If Extension of Assessment Is Sought

ML19270F211
Person / Time
Site:	05000495
Issue date:	01/24/1979
From:	William Kennedy SOUTHWEST ECONOMETRICS, INC.
To:	Boyd R Office of Nuclear Reactor Regulation
References
NUDOCS 7902050032
Download: ML19270F211 (41)

Text

. ,

/

[pA arc 9,% UNITED STATES

.. .,.- NUCLEAR REGULATORY CCMMisSION i

1, b( h j) i WASHINGTCN, D. C. 20555

%, L /j! JAN C , mg Docket No. STN 50 495 Mr. W. J. L. Kennedy, Vice President Stone & Webster Engineering Corporation P. O. Box 2325 Boston, Massacnusetts 02107

SUBJECT:

EXTENSION REVIEW MATTERS FOR PRELIMINARY DESIGN APPROVALS The Commission's August 22, 1973 policy statement on standardization includes a provision which allcws any Preliminary Design Approval (PDA) tht.t had been previously issued for a three-year term to be extended for two additional years. This provision applies to PDA-4 issued for the SWESSAR/RESAR-41 application.

As set forth in the policy statement, each application for a PDA extension will be subject to an assessment of the design with respect to the Category I, II, III, and IV matters approved for implementation since the regulatory recuirements cutoff date for the PDA in question. A tabulation of each Category I, II, III, and IV matter approved since January 17,1975, the regulatory requirements cutoff date for SWESSAR/RESAR 41, is provided in Enclosures A, B, C, and D, respectively, to this letter.

Should you desire PDA-4 to be extended for two additional years, we reouest that you provide an assessment of the SWESSAR/RESAR 41 design against each Categor. I, II, III, and IV matter icentified in the enclosures whicn is applicable te the SWESSAR/RESAR-41 design. Upcn receipt of your responses, the staff will eview them as folicws:

1) The staff will review your responses to determine wnether they are complete. If the staff determines that your responses are complete, we will administratively extend PDA-4 for two additional years subject to later staff acceptance of your proposed resolution for the acclicable Category II, III, and IV matters id%:ified in the enclosures.

79020500 9

2)

If the SWESSAR/RES, R-41 design is as soon as we are inte referenc ermed ailed pplican reference the SWESSAR/RESAR-41 d -

will require additional informationIt in .

is an detailed extension review. order to r detailed extension review guidelines: wil ey are dete onducted accor a) have clearly cellneatedetermir the conforms o which tt no changes to these to the desimatters.

It is anticicated of Category I matters.gn resulting from the st b) to wnicn the cesign conformsC alternative to these matters., or provides define t an ac design is no,t in substantial confoFor thos rmance with t:

demonstrate why conformanc ecessary.

you of the staff review may result in addit

~

c) nal rec whicn the design conforms alternatives are provided.

determin-ers or whe tive has been provided, will be required. s or nostaff-accect oved revisions d)

Catecory IV Matters - Category IV m t a ters are the:

deems to have sufficient r utes to safe e Direc being addressed during tne pDA exten will be treated ide

'a r -i s on revi. .

extension purposes.ntically to the Category II mat Ycur response to each matter identifi appendix.

as an amendment to the SWESSAR/RES Safety Analysis Report at this timeChanges

. ody of the Sk

-3 JAN 2 4 1979 The SWESSAR/RESAR-41 Safety Analysis Recort appendix addressing the extension review matters shculd be filed on the PDA docket.

We suggest that your submittal be filed as soon as practical. At the latest, however, your submittal should be filed by Aaril 1,1979 in order to provide adequate time for staff action prior to the expiration date of PDA-4.

If a Final Design Aporoval (FDA) application for the SWESSAR/RESAR-41 design is tendered, the PDA extension matters may also be filed and reviewed on the FDA docket. Review on the FDA docket is acceptable if (1) the extension review '. tatters can be resolved on a schedule consistent with the review ichedule established or contemplated for any construction permit application referencing the SWESSAR/RESAP-41 PDA, and (2) the construction permit applicant (s) agrees with that course of action.

The Commission has carefully considered the question of fees for PDA extension reviews. While committed to strong encouragement of the standardization program, the Commission also is responsible to a legislative mandate requiring that costs of review activities be recovered. Accordingly, the Commission has directed that we charge fees for PDA extension reviews in the following manner.

Based on the staff's completeness review of your assessment of each acplicable PDA extension review matter, we would conditionally extend the PDA if an acceptably complete assessment has been provided. There would be no fee associated with this extension, but the extension would be conditional in the sense that staff design approvals would be based on satisfactory resolution of the various issues to be addressed in later safety reviews.

Upon formal notification by a utility-applicant that it intended to reference your PDA during the extended term, that is, after the initial three-year period of validity, the staff would then review the assess-ment package. Such a review would be scheduled for completion prior to the tendering of the utility application. The cost of a PDA extension technical review conducted outside the context of the review of a Final Design Acproval application would be handled in a similar manner to the PDA approval fees which are due at the time of tendering of the application. That is, you will be charged the cost of the PDA extension review, on the basis of twenty-oercent of the cost as each application for the first five units involving the extended PDA is filed by a utility or utilities. However, if you file an application for an FDA for this Gesign prior to completion of the PDA extension review, the staff will, at your request, include the cost for the PDA extension review as part of the FDA review cost. Thus, in such instances, cost recovery for the PDA extension would be in the context of the FDA fee.

JAN ' 7379 4_

If you require any clarification of the matters discussed in this letter, please contact Thomas H. Cox, the staff's assigned licensing project manager.

.. Sic.~ rely,

• • gA

'Reser 5.' Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation

Enclosures:

A. Category I Matters B. Category II Matters C. Category III Matters D. Category IV Matters E. Description of Other Positions Identified as NRR Category IV Matters in Enclosure 0

ENCLOSURE A CATEGORY I MATTERS APPROVED BY RRRC --..

EFFECTIVE ~~

DATE DOCUMENT NO. REVISION TITLE

-' ~ '

. . . .gGE'7-~ "~'r'2

~

.,,,,,,3. -- - ' ~Controi of Cemeustible Gas Concentration in Containment Following a loss-of-Coolant Accident 9/1/78 RG 1.9 1 Selection, Design, and Qualification for D!asel-Generator Units Used as Onsite Electric Pcwer Systems at Nuclear Power Plants 1/9/76 RG 1.20 2 Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing 11/29/77 RG 1.28 1 Quality Assurance Program Requirements (Design and Construction) 6/20/78 RG 1.29 3 Seismic Design Classification 7/20/76 RG 1.31 2 Contrcl of Ferrite Content in Stainless Steel Weld Metal 1/14/77 RG 1,32 - 2 Criteria for Safety-Related Electric Pcwer Systems for Nuclear Power Plants

, 10/21/76 RG 1.33 1 Quality Assurance Program Requirements (Operation) 8/15/75 RG 1.35 '2 Inservice Inspection of Ungrouted Tendons in Pre-stressed Concrete Containment Structures (1) 5/77 RG 1.38 2 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage

_ and Handling of Items for Water-Cooled Nuclear Power Plants 7/12/77 RG 1.39 2 Housekeeping Requirements for Water-Cooled Nuclear Pcwer , Plants

. . .. .; . . . . l l.. ;::..

EFFECTIVE DATE DOCUMENT N0. REVISION TITLE (2) 11/29/77 RG 1.52 ' Design, Testing, and Maintenance for Engineered Safety Feature Atcosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled

... .. Nuclear Power Plants _ _. _- .. _<

(3) 3/22/77 RG 1.63 1 Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants 1/9/76 RG 1.64 2 Quality Assurance Requirements for the Design of Nuclear Power Plants 6/20/78 RG 1.68 2 Initial Test Programs for Water-Cooled Reactor Power Plants 9/26/75 RG 1.68.1 0 Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants 11/15/77 RG 1.72 1 Spray Pond Plastic Piping (1) 3/78 RG 1.84 12 Code Case Acceptability - ASME Section III Design and Fabrication (1) 3/78 RG 1.85 12 Code Case Acceptability - ASME Section III Materials 5/26/77 PG l.90 1 Inservice Inspection of Pre-stressed

- Concrete Containment Structures with Grouted Tendons 3/22/75 RG 1.92 1 Combining Modal Responses and Spatial Components in Seismic Response Analysis 2/6/76 RG 1.94 1 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel during the Construction Phase of Nuclear Power Plants 10/21/76 RG 1.95 1 Protection of Nuclear ?ower Plant Control Room Operators Against an Accidental Chlorine Release

EFFECTIVE DATE__ _ _ DOCUMENT NO. REVISION TITLE (4) 1/14/77 RG 1.99 1 Effects of Residual Elements on Predicted Radiatfon Damage to Reactor Vessel Materials

- 6/14/77'

~

RG 7.100 -

' Seismic Qua~lification of Electric ' : 7, Ecuipment for Nuclear Pcwer Plants (1) 10/76 RG 1.103 1 Post-Tensioned Pre-stressing Systems for Concrcte Reactor Vessels and Containments

'~

1/28/77 RG 1.106 1 Thermal Overicad Protection for Electric Motors on Mator-0peraced Valves 10/21/76 RG 1.107 1 Qualifications for Cement Grouting for Pre-stressing Tendons in Containment Structures (1) 5/77 RG 1.116 0-R Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Syste.as 9/27/77 RG 1.118 1 Periodic Testing of El;ctric Power and Protection Systems (5) 5/11/77 RG 1.120 1 Fire Protection Guidelines for Nuclear Pcwer Plants 11/15/77 RG 1.122 1 Development of Floor Design Response Spectra for 3eismic Design of Floor-Supported Equipment or Components (1) 7/77 RG 1.123 l Quality Assurance Requirements fer Control of Procurement of Items and Services for Nuclear Power Plants

~

ETFECTIVE DATE C0CUMENi NO. REVISION TITLE 1/14/77 RG 1.126 lb An Acceptable Model and Related Statistical Methods for the Analysis

.of Fuel Densification _ , ,

6/2d/73 RG 1.128 1 Installaticn Design and' Installetion ^

of Large Lead Storage Batteries for Nuclear Power Plants 2/18/77 RG 1.129 0 Maintenance, Testing and Replacement

.. of Large Lead Storage Batteries for Nuclear Power Plants (5) 5/26/77 RG 1.131 0 Qualification Tests of Electric Cables, Fielu Splices and Connections for Light Water Cooled Nuclear Power Plants 5/11/77 RG 1.132 0 Site Investigations for Foundations of Nuclear Power Plants 3/22/77 RG 1.134 0 Medical Certification and Monitoring of Parsonnel Requiring Operator Licenses 7/12/77 RG 1.135 0 Normal Water Level and Discharge at Nuclear Power Plants 8/31/77 RG 1.136 0 Material for Concrete Containments (7) 9/27/77 RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators 9/27/77 NUREG-0102 0 Interfaces for Standard Designs (SRP 1.8) 11/15/77 RG 1.133 0 Laboratory Investigation of Soils for Engineering Analysis and Design of Nuclear Power Plants 11/15/77 RG l.XXX 0 Permanent Dewatering Systems 11/29/77 RG 1.140 0 Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of LKR's 1/31/78 RG 1.la2 0 Safety-Related Concrete Structures 3/14/73 RG 3.19 0 Occupational Radiation Dose Assess-ment at L'AR's - Design Stage Man-Rem Estimates

EFFECTIVE .

DATE DOCUMENT NO. REVISION TITLE (8) 3/14/73 RSB 5-2 0 Reactor Coolant System Overpressure Protection

( 1) Indicates that the Category I assigned by RRRC for the previous revision o?'"~

tt.is document was retained. Review by the RRRC for reassignment of the category is not required for document revisior.s which do not result in an increase in requirements.

( 2) Revision 1 of this regulatory guide was assigned as a Cdtegory II matter effective January 9,1976. It is the intent of the 'RRC that revision 1 remain a Category II matter. However, revision 2 may be used in lieu of revision 1 if so desired by applicants.

( 3) Assigned as a Category II matter by the RRRC for those applications not previously reviewed to revision 0.

( 4) Category I for paragraph C.3 only. Paragraphs C.1, C.2, and C.4 are assigned by the RRRC as Category III matters.

( 5) In specifying Category I for this regulatory guide, the RRRC recognizes that the staff is utilizing Appendix A to BTP ASB 9.5-1 on operating reactors and all CP and CL applications now under review.

( 6) In specifying the Category I for this regulatory guide, the RRRC recognizes that the fire protection aspects are covered by Appendix A to BTP ASB 9.5-1 which is a Category II matter

( 7) Category I for all CF or PDA applications docketed after the implementation date shown in the published guide. Certain provisions of the guide are also assigned by the RRRC as Category II and Catecary III matters.

( 8) Category I for operating licenses issued prior to March 14, 1978. Assigned by the RRRC as a Catagory III matter for all other applications.

ENCLOSURE B CATEGORY II MATTERS APPROVED SY RRRC

~

EFFECTIVE DATE DOCUMENT REVISION TITLE

~ -

^ 11/12/75 RG T.27 72-

- Ultimate, Heat Sink for- fluclear Pcwer _-- ;

Plants 1/9/76 RG 1.52 1 Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmcsphere Cleanup System Air Filtra-

. tion and Adsorption Units of Light-Water-Cooled Nuclear Power Plants (1) 8/77 RG 1.59 2 Design Basis Floods for Nuc. ir Pcwer Plants (2) 3/22/77 RG 1.63 1 Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants (3) 5/16/78 RG 1.68.2 1 Initial Startup Test Program to Demon-strate Remote Shutdown Capability for Water-Cooled Nuclear Pcwer Plants 11/15/77 RG 1.91 1 Evaluation of Exp'osions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites 1/28/77 RG 1.97 1 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident 11/12/75 RG 1.102 1 Flood Protection for Nuclear Power Plants 9/15/76 RG 1.105 1 Instrument Setpoints 6/14/77 RG 1.108 1 Periodic Testing of Diesal Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants e

EFFECTIVE DATE DOCUMENT REVISION TITLE 3/22/77 RG 1.115 1 Prutection Against Low-Trajectory Turbine Missiles _

12/20/77 RG 1.117 1 Tornado Design Classification S/31/77 RG 1.124 1 Service Limits and Loading Comoinations for Class 1 Linear Type Component Support:

7/77 RG 1.130 0 .. Design Limits and Loading Combinations for Cliss 1 Plate- and Shell-Type Compo1ent Supports (4) 9/29/77 RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators S/18/76 RG 8.8 2 Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as low As is Reasonably Achievable (Nuclear PowerReactors) 8/18/76 BTP ASB Guidelines for Fire Protection for 9.5-1 Nuclear ?ower Plants Under Review and Const ruction 4/13/77 BTP Material Selection and Processing MTEB 5-7 Guidelines for B'WR Coolant Pressure Scundary Piping (5) 1/31/78 SRP 5.4.7 1 Residual Heat Removal System (6) 1/31/78 RG 1.141 0 Containment Isolation Provisions for Fluid Systems e

_3 (1) Indicates that the Category II assigned by RRRC for the previous revision of this document was retained. Review by the RRRC for reassignment of the category is not required for document revisions which do not result in an increase in requirements.

(2) Assigned as a Category I matter for those applications previously reviewed to revision 0.. Category JI for all other applications. ._. . _ _c (3) ~ Category II for operating reactors. Assigned by the RRRC as a Category III matter for all other applications.

(4) Category II for paragraph C.1 for all CP's or PDA's under review whose SER's have not been issued prior to the implementation date shown in the published guide. Paragraph C.9 for all operating reactors, CL applications, and CP and PDA applications under review whose SER's are completed prior to the implementation date shown in the published guide.

Certain provisions of this guide are also assigned by the RRRC as Category III matters.

(5) Category II for operating reactors and all other applications for which the issuance of the CL is expected prior to January 1,1979. Assigned by the RRRC as a Category III matter for all other applications.

(6) Category II for operating reactors and CL reviews. Assigned by the RRRC as a Category III matter for all other applications.

ENCLOSURE C CATEGORY III MATTERS APPRCVED BY RRRC i

EFFECTIVE DATE OCCUMENT REVIS!0 TITLE 5/16/78 RG 1.56 1 Maintenance of Water _ Purity in Boiling

. Water Reactors _. _

- ~

(1) 5/16/78 RG 1.68.2 l Initial Startup Test Program to Demon-strate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants (2) 1/14/77 RG 1.99 1 Effects of Residual Elements on Predicted

. Radiation Damage to Reactor Vessel Materials (3) 3/77 RG 1.101 1 Emergency Planning for Nuclear Power Plants 11/76 RG 1.114 1 Guidance on Being Operator at the Controls of a Nuclear Pcwer Plant 5/11/75 RG 1.121 0 Bases for Plugging Degraded PWR Steam Generator Tubes 11/29/77 RG 1.127 1 Inspection of Water-Control Structures Associated with Nuclear Power Plants (4) 9/27/77 RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators (5) 1/~i/78 -SRP 5.4.7 1 Residual Heat Removal System (6) '1/31/78 RG 1.141 0 Containment Isolation Provisions for Fluid Systems (7) 3/14/78 RSB 5-2 0 Reactor Coolant System Overpressuriza-tion Protection 4

(1) Assi ned by RRRC as a Category II matter for operating plants.

(2) .aragraph C.3 is a Category 7 matter.

(3) Indicates that the. Category III assigned by RRRC for the previous _ . _ _ _ _

revision of this document was retained . Review by the RRRC for -- --

reassignment of the category is not required for document revisions which do not result in an increase in requirements.

(4) Cdtegory III for paragraph C.2 for all CP and PDA applications under review whose SER's havc not been issued prior to the implementation date shown in the published gu' rte. Certain provisions of this guide are also assigned by the RRRC as Category II matters.

(5) Category III for CP or PDA applications docketed prior to January 1, 1978, and for which OL issuance is expected after January 1,1979, all Category II for all other applications.

(6) Assigned by RRRC as a Category II matter for operating reactors and OL applications.

(7) Assigned by RRRC as a Category I matter for OL's issued prior to March 14, 1978, and Category III for all other applications.

O

. ENCLOSURE D NRR CATEGORY IV MATTERS

~'

- , ; , '~ -

. A. ' RegulatoIy'iiuides' notl categcrized ,'- - - # -I'" "l , [ Z Issue Date Number Revision Title 12/75 1.13 1 Spent Fuel Storage Facility Cesign Basis 8/75 1.14 1 Reactor Coolant Pump Flywheel Integrity 1/75 1.75 l Physical Independence of Electric Systems 9/75 1.79 1 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 7/75 1.83 1 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes 2/76 1.104 0 Overhead Crane Handling Sys:2. f for

- Nuclear Power Plants d

e r

e

B. SRP Criterta Implementa- Apolicable tion Date Branch SRP Section Title

1. 11/24/75 MTEB 5.4.2.1 BTP HTEB-5-3, Monitoring of Secondary Side Water Chemistry in PWR Steam :T Generators

2. 11/24/75 CSB 6.2.1 BTP CSB-6-1, Minimum 6.2.lA Containment Pressure Model 6.2.18 for PWR ECCS Performance 6.2,1.2 Evaluation 6.2.1.3 6.2.1.4 6.2.1.5

3. 11/24/75 CSB 6.2.5 BTP CSB-6-2, Control of Combustible Gas Concentra-tions in Containment Following a Loss-of-Coolant Accident 4 11/2a '75 CSB 6.2.3 BTP CSB-6-3, Determination of Bypass Leakage Path in Dual Containment Plants

5. 11/24/75 CSB 6.2.4 BTP CSB-6-4, Containment Purging During Normal Plant Operations

6. 11/24/75 ASB 9.1.4 BTP ASB-9.1, Overhead Handling Systems for Nuclear Power Plarts

7. 11/24/75 ASB 10.4.9 BTP ASB-10.1, Design Geidelines for Auxiliary Feedwater System

~

Pump Drive and Power Supply Diversity for PWR's

8. 11/24/75 SEB 3.5.3 Procedures for Ccmposite Section Local Damage Predict'.on (SRP Section 3.5.3, par. II.l.C)

Implementa , Applicable tion Date Branch SRP Section lille

9. 11/24/75 SEB 3.7.1 Development ef Design Time History for Soil-Structure Interaction Analysis (SRP Section 3.7.1,' par. II.2)

10. 11/24/75 SEB 3.7.2 Procedures for Seismic System Analysis (SRP Section 3.7.2 par. II)

11. 11/24/75 SEB 3.7.3 Procedures for Seismic Sub-system Analysis (SRP Section 3.7.3, par. II)

12. 11/24/75 SEB 3.8.1 Design and Construction of Concrete Contaiirents) SRP Section 3.8.1, par. II)

13. 11/24/75 SEB 3.8.2 Design and Constiuction of Steel Containments (SRP Section 3.8.2, par. II)

14. 11/24/75 SEB 3.8.3 Structural Design Criteria for Category I Structures Inside Containment (SRP Section 3.8.3, par. II)

15. 11/24/75 SEB 3.8.4 Structural Design Critaris for Other Seismic Catagc.y I Str;ctur,e (SRP Section 3.J.4, par. II)

16. 11/24/75 SEB 3.8.5 Structural Design Criteria 'or Foundationt (SRP Section 3.8.5, par. II)

17. 11/24/75 SEB 3.7 Seismic Design Requirements for 11.2 Radwaste Sysems and Their Housing 11.3 Structures (SRP 3ection 11.2, BTP 11.4 ETSB 11-1, par. B.v)

4-Implementa- . Applicable tion Date Branch SRP Section Title

18. 11/24/75 SEB 3.3.2 Tornado Load Effect Combi-nations (SRP Section 3.3.2, par. II.2.d)

~

~

_ 19. 11/24[75 ~ SEB

'3.4.2 Dynamic Efects of Wave' Action .

(SRP Section 3.4.2, par. II)

20. 10/01/75 ASB 10.4.7 Water Hammer for Steam Generators with Preheaters (SRP Section 10.4.7 par. I.2.b)

21. 11/24/75 AB 4.4 Thermal-Hydraulic Stability (SRP Section 4.4, par. II.5)

22. 11/24/75 RSB 5.2.5 Intersystem Leakage Detection (SRP Section 5.2.5 par. II.4) and R.G.1.45

23. 11/24/75 RSB 3.2.2 Main Steam Isolation Valve Leakage Control System (SRP Section 10.3 par. III.3 and BTP RSB-3.2)

C. Other Positions Implementa- Applicable tion Date Branch SRP Section Title

1. 12/1/76 SEB 3.5.3 Ductility of Reinforced Concrete and Steel Suructural Elements Subjected to Impactive or Impulsive Loads

2. 8/01/76 SEB 3.7.1 Response Spectra in Vertical Direction

3. 4/01/76 SEB 3.8.1 SWR Mark III Ccntainment Pool 3.3.2 Dynamics

4. 9/01/76 SEB 3.8.4 Air Blast Loads

5. 10/01/76 SEB 3.5.3 Tornado Missile Impact

6. 6/01/77 RSB 6.3 Passive Failures During Long-Term Cooling Following LCCA

Implementa- '

Applicable tion Date Branch SRP Section Title

7. 9/01/77 RSB 6.3 Control Rocn Position Indica-tion of Manual (Handwheel) Valves in the ECCS

8. 4/01/77 RSB 15.1.3 -

Long-Term Recovery from Steamline -

Break: Operator Action 40 Prevent Overpressurization

9. 12/01/77 RSB 5.4.6 Pump Operability Requirements 5.4.7 6.3

10. 3/28/78 RSB 3.5.1 Gravity Missiles, Vessel Seal Ring Missiles Inside Containment

11. 1/01/77 AB 4.4 - Core Thermal-Hydraulic Analysis

12. 1/01/78 PSB 8.3 Degraded Grid Voltage Conditions

13. 6/01/76 CSB 6.2.1.2 Asymmetric Loads on Components

. Located Within Containment Sub-compartments 14, 9/01/77 CSB 6.2.6 Contair. ment Leak Testing Program

15. 1/01/77 CSB 6.2.l.4 Containment Response Due to Main Steam Line Break and Failure of MSLIV to Close

16. 11/01/77 ASB 3.6.1 Main Steam and Feedwater Pipe 3.6.2 Failures

17. 1/01/77 ASB 9.2.2 Design Requirements for Cooling Water to Reactor Coolant Pumas

18. 8/01/76 ASB 10.4.7 Design Guidelines for Water Hammer in Steam Generators with Top Feedring Design (BTP ASB-10.2)

19. 1/01/76 ICSB 3.11 Environmental Control Systems far Safety Related Equipment

~.

.r

_ ^ -

- T. . . _ ,. .

ENCLOSURE E DESCRIPTION OF OTHER POSITICNS IDENTIFIED AS NRR CATEGORY IV MATTE 5 IN ENCL.0SURE D Numbering scheme corresponds to that used in Item C of Enclosure 0; e.g., the first "Other Position" identified as a Category IV matter in Item C of Enclosure D is designated IV.C.1, etc.

e W

-g

ENCLOSURE E IV.C.1 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTS (3.5.3) SUBJECTED TO IMPACTIVE OR IMPULSIVE LOADS INTRODUCTION In the evaluation of overall respo.nse of reinforced concrete structural _

elements (e.g., missile barriers, columns, slabs, etc.) subjected to imoactive or impulsive loads, such as impacts due to missiles,' assumption of non-linear response (i.e., ductility ratios greater than unity) of the structural elements is generally acceptable provided that the safety functions of the structural elements ana those of safety-related systems and components supported or protected by the elements are maintained.

The following summarizes specific SEB interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements subjected to impactive and impulsive loads.

SPECIFIC POSITIONS

1. REINFORCED CONCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure controls design, the permissible ductility ratio ( u ) under impactive and impulsive laids should be taken as a = 0.05 for a_a. > .005 p

u = 10 for  :-a' < .005 where p and p'are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code.

1.2 If use of a ductility ratio greater than 10 (i.e. , u> 100) is required to demonstrate design adecuacy of structural elements against impactive or impulsive loads, e.g., missile impact, such a usage should be identified in the plant SAR.

Information justifying the use of this relatively high ductility value shall be provided for SE3 staff review.

1.3 For beam-columns, walls, and slabs carrying axial compression loads and subject to impulsive or impactive loads producing flexure, the permissible ductility ratio in flexure should be as follows:

(a) When compression controls the design, as defined by an interaction diagram,

~'

the permissible ductil_ity ratio , '

shall be 1.3.

(b) When the compression loads do not exceed 0.l fc ' Ag or one-third of that which would produce balanced conditions, which-ever is smaller, the permissible ductility ratio can be as given in Sectio <.1.1.

(c) The permissible dutility rat o shall vary linearly from 1.3 to that given in Section 1.1 for conditions between those specified in (a) and (b) . (See Fig 1.)

1.4 For structural elements resisting axial compressive impulsive or impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3.

1.5 For shear carried by concrete only u = 1.0 For shear carried by concrete and stirrups or bent bars u = 1.3 For shear carried entirely by stirrups a = 3.0 2.0 STRUCTURAL STEEL MEMBERS 2.1 For flexure compression and shear u = 10.0 2.2 For columns with slenderness ratio (1/r) equal to or less than 20 u = 1.3

where 1 = effective length of the member r = the least radius of gyration For columns with slenderness ratio greater than 20 9 = 1.0

' ~

2.3 For meinbers subjected to tension

-7 u = .5 $

where cu= uniform ultimate strain of the material

-cY = strain at yield of material IV.C.2 RESPONSE SPECTRA IN THE VERTICAL DIRECTION (3.7.1)

Subsequent to the issuance of Regulatory Guide 1.60, the report

" Statistical Studies of Vertical and Horizontal Earthquake Spectra" was issued in January 1976 by NRC as NUREG-0003. One of the important conclusions of this report is that the response spectrum for vertical motion can be taken as 2/3 the response spectrum for horizontal motion over the entire range of frequencies in the Western United States. According to Regulatory Guide 1.60, the vertical response spectrum is equal to the horizontal response spectrum between 3.5 cps and 33 cps. For the Western United States only, consistent with the latest available data in NUREG-0003, the option of taking the vertical design design response spectrum as 2/3 the horizontal response spectrum over the entire range of frequencies will be accepted.

For other locations, the vertical response spectrum will be the same as that given in Regulatory Guide 1.60.

IV.C.3 BWR MARK III CONTAINMENT POOL DYNAMICS (3.8.1 3.3.2)

1. POOL SWELL

a. Bubble pressure, bulk swell and froth swell loads, drag pressure. .cd other pool swell loads should be treated as abnormal pre;sure loads, Pa . Appropriate load combinations and load factors should be applied accordingly,

b. The pool swell loads and accident pressure may be ccmbined in accordance with their actual time histories of occurrence.

5

^

i i

oAS g

J :_

. i

/f Tt I

t t

R T

c g _

d

) o I F F $. .

- b o

j l 1

- I

_. A .b e -

R E

T

- )

k

- u a

-( c3

/. M o

, b o '

o e g s o

P o

_ R I

O n

P L

O .

1 4 ,

? '

G Tt 4

t N

c

-d

,e 3

$ eO y,,G I

F i -- E L

k T

\ '

Sda $Qn. V  ?.,n y j -

$b,Y .

)

S

} /

F \

l N

g )(r (b b o

g e

r P P ee hhl tl 3 a

'/ I sn i i s

4

2. SAFETY RELIEF VALVE (SRV) DISCHARGE

a. The SRV loads should be treated as live loads in all load combinations 1.5Pa where a load factor of 1.25 should be applied to the appropriate SRV loads,

b. A single active failure causing one SRV discharge must be considered in combination with the Design Basis Accident (DBA) .

~

c. ~ ippr~opriate muftiple SRV discharge should be consider'e'd in -' Z combination with the Small Break Accident (SBA) and Inter-mediate Break Accident (IBA).

d. Themal loads due to SRV discharge should be treated as TU for normal operation and T a for accident condi;....

e. The suppression pool liner should be designed in accordance with the ASME Boiler and Pressure Vessel Code, Division 1 Subsection NE to resist the SRV negative pressure, considering strength, buckling and low cycle fatigue.

IV.C.4 AIR BLAST LOADS (Pa, Ta, To as defined in ACI 359-740)

(3.8.4 )

The following interim position on air blast loadings on Nuclear Power Plant Structures should be used as guidance in evaluating analyses.

1. An equivalent static pressure may be used for structural analysis purposes. The equivalent static pressure should be obtained from the air blast reflected pressure or the overpressure by multiplying these pressures by a factor of two. Any proposed use of a dynamic load factor less than two should be treated on a case by case basis.

Whether the reflected pressure or the overpressure is to be used for individual structural elements depends on whether an incident blast wave could strike the surface of the element.

2. No load factor need be specified Mr the air blast loads, and the load combination should be:

U=D&L+B where, V is the strength capacity of a section D is dead load -

L is live load B is air blast load.

3. Elastic analysis for air blast is required for concrete structures of new plants. For steel structural elements, and also for rein-forced concrete elements in existing plants, some inelastic response may be permitted with appropriate limits on ductility ratios.

4. Air blast generated ground shock and air blast wind pressure may be ignored. Air blast generated missiles may be important in situations where explosions are postulated to occur in vessels which may fragment.

5. Overturning and sliding stability should be assessed by multiplying the structure's full projected area by the equivalent static pressure and assuming only the blast side of the structure is -

loaded. Justification for reducing the average equivalent static pressure on curved surfaces should be considered on a case by case basis.

6. Internal supporting structures should also be analyzed for the effects of air blast to determine their ability to carry loads applied directly to exterior panels and slabs. Moreover.in vented structures, interior structures may require analysis even if they do not support exterior structures.

7. The equivalent static pressure should be considered as potentially acting both inward and outward.

IV.C.5 TORNADO MISSILE PROTECTION (3.5.3)

As an interim measure, the minimum concrete wall and roof thickness for tornado missile protection (based on the acceotable tornado missile spectra identified in Section 3.5.1.4 of the Standard Review Plan) will be as follows:

Wall Thickness Roof Thickness Concrete Strength (psi) (inches) (inches) 3000 27 24 Region I 4000 24 21 5000 21 18 3000 24 21 Region II 4000 21 18 5000 19 16 3000 21 18 Region III 4000 18 16 5000 16 14 These thicknesses are for protection against local effects only. Designers must establish independently the thickness requirements for overall structura' response. Reinforcing steel should satisfy the provisions of Apcendix C, ACI 349 (that is , .2". minimum, E'4EF) . The regions are described in Regulatory Guide 1.76.

IV.C.6 PASSIVE ECCS FAILURES DURING LONG-TERM CCOLING FOLLCWING A LOCA (6.3)

Passive failures in the ECCS, having leak rates equal to or less than those from the sudden failure of a pump seal and which may occur during the long-term cooling period following a postulated LCCA; should be con-sidered. To mitigate the effects of such leaks, a leak detection system having design features and bases as described below should be included ~

in the plant design. -

The leak detection system should include detectors and alarms which would alert the operator of passive ECCS leaks in sufficient time so that appro-priate diagnostic and corrective actions may be taken on a timely basis.

The diagnostic and corrective actions would include the identification and isolation of the faulted ECCS line before the performance of more than one subsystem is degraded. The design bases of the leak detection system should include:

(1) Identification and justification of the maximum leak rate; (2) Maximum allowable time for operator action and justification therefor; (3) Demostration that the leak detection system is sensitive enough to initi. ate and alarm on a timely basis, i.e. , with sufficient lead time to allow the operator to identify and isolate the faulted line before the leak can create undesireable consequences such as flooding of re-dundant equipment. The minimum time to be considered is 30 minutes; (4) Demonstration that the leak detection system can identify the faulted ECCS train and that the leak can be isolated; and (5) Alarms that conform with the criteria specified for the control room alarms and a leak detection system that conforns with the require-ments of IEEE-279, except that the single fai? e criterion need not be imposed.

IV.C.7 CONTROL ROOM POSITION INDICATION OF MANUAL (HANDWHEEL) VALVES (6.3)

Regulatory Guide 1.47 specifies automatic position indication of each bypass or deliberately induced inoperable condition if the following three conditions are met:

(1) The byoass or inocerable condition affects a system that is designed to perform an autcmatic safety function.

(2) The bypass or inoperable condition can reasonably be expected to occur more frequently than once per year (3) The bypass or inoperabic condition is expected to occur when the system is normally required to operate.

Revision one of the Standard Review Plan in Section 6.3 req. ires -

conformance with Regulatory Guide 1.47 with the intent being that any manual (handwheel) valve which could jeopardize the operation of the ECCS, if inadvertently left in the wrong position, must have position indication in the control room. In the PDA extension reviews it is important to confirm that standard designs include this design feature. Most standard designs do but this matter was probably not specifically addressed in some of the first PDA reviews.

IV.C.8 LONG-TERM RECOVERY FROM STEAM LINE BREAK - OPERATOR ACTION TO (15.1.5) PREVENT OVERPRESSURIZATION (PWR)

A steam line break causes cooldown of the primary system, shrinkage of RCS inventory and depletion of pressurizer fluid. Subsequent to plant trip, ECCS actuation, and main steam system isolation, the RCS inven-tory increa.ses and expands, refilling the pressurizer. Without operator action, replenishment of RCS inventory by the ECCS and expansion at low temperature could repressurize the reactor to an unacceptable pressure-temperature region thereby compromising reactor vessel integrity. Anal-yses are required to show that following a main steam line break that (i) no additional fuel failures result from the accident, and (ii) the pressures fcilowing the initiation of the break will not compromise the integrity of the reactar coolant pressure boundary giving due considera-tion to the changes in coolant and material temperatures. The analyses should be based on the assumption that operator action will not be taken until ten minutes after initiation of the ECCS.

IV.C.9 PUMP OPERABILITY REOUIREMENTS (5.4.6 5.4.7 In some reviews, the staff has found reasonable doubt that some types of 6.3) engineered safety feature pumps would continue to perform their safety function in the long term following an accident. In such instances there has been followup, including pump redesign in some cases, to assure that long term performance could be met. The following kinds of infor-mation may be sought on a cas."by-case basis where such doubt arises.

a. Describe the tests performed to demonstrate that the pumps are capable of operating for extended periods under post-LOCA conditions, including the effects of debris. Discuss the damage to pumo seals caused by debris over an extended period of operation.

b. Provide detailed diagrams of all water cooled seals and ccmoo-nents in the pumps.

c. Provide a description of the composition of the pump shaft seals and the shafts. Provide an evaluation of loss of shaft s eal s. -

d. Discuss how debris and post-LOCA environmental conditions were factored into the specifications and design of the pump.

IV.C.10 GRAVITY MISSILES, VESSEL SEAL RING MISSILES INSIDE CONTAINMENT (3.5.1) Safety related systems should be protected against loss of function due to internal missiles from sources such as those associated with pressurized components and rotating equipment. Such sources would include but not be limited to retaining bolts, control rod drive assemblies, the vessel seal ring, valve bonnets , and valve stems. A description of the methods used to afford protection against such potential missiles, including the bases therefor, should be provided (e.g. , preferential orientation of the poten-tial missile sources, missile barriers, physical separation of redundant safety systems and components) . An analysis of the effects of such poten-tial miss'iles on safety related systems, including metastably supported equipment which could fall upon impingement, should also be provided.

-9_

IV.C.ll CORE THERMAL-HYDRAULIC ANALYSES (4.4)

In evaluating the thermal-hydraulic performance of the rcactor core.;he following additional areas should be addressed:

1. The effect of radial pressure gradients at the exit of open lattice cores.

2. The effect of radial pressure gradients in the upper plenum.

3. The effect of fuel rod bowing.

In addition,a commitment to perform tests to verify the transient analysis methods and codes is required.

IV.C .12 DEGRADED GRIO VOLTAGE CONDITIONS (8.3)

As a result of the Millstone Unit Number 2 low grid voltage occurrence, the staff has developed additional requirements concerning (a) sustained degraded voltage conditions at the offsite power scurce, and (b) inter-action of the offsite and onsite emergency power systems. These additional requirements are defined in the following staff position.

1. We require that a second level of voltage protection for the onsite power system be provided and that this second level of voltage pro-tection satisfy the following requirements:

a) The selection of voltage and time set points shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite system distribution levels; b) The voltage protection shall include coincidence logic to preclude spurious trips of the offsite power source;

c) The time delay selected shall be based on the following conditions:

(i) The allowable time delay, including margin, shall not exceed the maximum time delay that is assumed in the S/R accident analyses;

' ~~ ' '

(ii) The tiine delay ~ shall ~ minimize the effect of short ~Z duraticn disturoances from recucing the availacility of the offsite power source (s); and (iii) The allowable time duration of a degraded voltage condition at all distribution system levels shall not result in failure of safety. systems or components; (iv) The voltage sensors shall automatically initiate the disconnection of offsite power sources whenever the voltage set point and time delay limits have becn exceeded; (v) The voltage sensors shall be designed to satisfy the applicable requirements of IEEE Std. 279-1971 " Criteria for Drotection Systems for Nuclear Power Generating Stations"; and (vi) The Technical Specifications shall include limiting conditions for operation, surveillance requirements, trip set points with minimum and maximum limits, and allowable values for the second-level voltage protection sensors and associated time delay devicec.

2. We recuire that the system design automatically prevent load shedding of the emergency buses once the onsite sources are supplying pcwer to all sequenced loads on the emergency buses.

The design shall also include the capability of the load shedding feature to be automatically reinstated if the onsite source supply breakers are tripped. The automatic bypass and reinstatement feature shall be verified during the periodic testing identified in Item 3 of this position.

3. We require that the Technical Specifications include a test require-ment to demonstrate the full functional operability and independence of the onsite power sources at least once per 18 months during shut-down. The Technical Specifications shall include a recuirement for tests: (a) simulating loss of offsite power; (b) simulating loss of offsite power in conjunction with a safety injection actuation signal; and (c) simulating interruption and subsequent reconnection of onsite pcwer sources to their respective buses. *

4. The vo.itage levels at the safety-related buses should be optimized for the full load and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite power source by appropriate adjust-ment of the voltage tap settings of the intervening transformers.

We require that the adequacy of the design in this regard be ve.rified by actual measurement, and by correlation of measured

_ values with ana19 sis results. __.

IV.C.13 ASYMMETRIC LOADS ON COMPONENTS (6.2.1.2) LOCATED WITHIN CONTAINMENT SUBCCMPARTMENTS In the unlikely event of a ping rupture inside a major ccmconent sub-comoartment, the initial blowdown transient would lead to pressure loadings on both the structure and the enclosed component (s). The staff's generic Category A Task Action Plan A-2 is designed to develop generic resolutions for this matter. Our present schedule calls for completing A-2 for PWR's during the first quarter,1979. Dending completion of A-2, the staff is implementing the folicwing program:

1. For PWRs at the CP/PDA stage of review, the staff requires appli-cants to connit to address the safety issue as part of their appli-cation for an operating license.

2. For PWRs at the OL/FDA stage of review, the staff requires case-by-case analyses, including implementation of any indicated corrective measusres prior to the issuance of an operating license.

3. For BWRs, for which this issue is expected to be of lesser safety significance, the asymmetric loading conditions will be evaluated on a case-specific basis prior to the issuance of an operating license.

For those cases which analyses are required, we request the cerformance of a subccmpartment, multi-ncde pressure response analysis of the pressure transient resulting frcm postulated hot-leg and cold-leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity, pipe penetrations, and steam generator compartments. Provide similar analyses for the pressurizer surge and spray lines, and other high energy lines located in containment compartments that may be subject to pressurization. Show how the results of these analyses are used in the design of structures and ccmponent supports.

IV.C.14 CONTAINMEtlT LEAK TESTING PRCGRAM (6.2.6)

To avoid difficulties experienced in this area in recent OL reviews, the staff has increased its scope of inquiry at the CP/PDA stage of review. For this purpose, the following information with regard to the containment leak testing program should be supplied, Those systens that will rematn fluid filled for the Type A test a.

should be identified and justification given.

b. Show the design provisions that will pemit the personnel air-lock door seals and the entire air lock to be tested,

c. For each penetration,i.e., fluid system piping, instrument, electrical, and equipment and personnel access penerations, identify the Type B and/or Type C local leak testing that will be done.

d. Verify that containment penetrations fitted with expansion bellows will be tested at Pa. Identify any penetration fitted with expansion bellows that does not have the design capability for Type B testing and provide justification.

IV.C.15 CONTAINMENT RESPONSE DUE TO MAIN STEAM LINE (6.2.1.4) BREAK AND MSLIV FAILURE In recent CP and OL application reviews, the results of analyses for a postulated main steam line break accident (MSLB) for designs utilizing pressurized water reactors with conventional containments show that the peak calculated containment temperature can exceed for a short time period the environmental qualification temperature-time envelope for safety related instruments and ccmponents. This matter was also discussed in Issue No.1 of NUREG-0138 and Issue No. 25 of NUREG-0153. The signifiance of the matter is that it could result in a recuirement for requalifying safety-related equipment to higher time-temperature envelopes.

The staff's generic Category A Task Action Plans A-21 and A-24 are designed to develop generic resolutions for these matters. The presently scheduled completion dates for A 21 and A-24 (Short Term 0-t 4 n) are e rst cu m , , 1979 and ' u-th m m er, 1973, -esnectivelv,

.-una ng ::;.1 pie:::a a n u. au .1 c , 5:..: 1 r.w .n s; au w a1 e used as detailed below.

We have developed and are implementing a plan in which all applicants for construction oemits and operating licenses and those already issued con-struction permits must provide information to establish a conservative temperature-time envelope.

Therefore, describe and justify the analytical model used to conservatively detemine the maximum containment temperature and pressure for a spectrum of postulated main steam line breaks for various reactor power levels. Include the following in the discussion.

(1) Provide single active failure analyses which specifically identify those cafety grade systems and components relied upon to limit the ma'ss and energy release and containment pressure / ' '

~ temperature responce. The single failure analyses should

^ ~ ~

include, but not necessarily be limited to: main steam and connected systems isolation; feedwater auxiliary feedwater, and connected systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and auxiliary feedwater run-out control system; the loss of or availability.,0f offsite power; diesel failure when loss of offsite power is evaluated; and partial loss of containment cooling systems.

(2) Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

(3) Discuss and justify the heat transfer correlation (s) (e.g., Tacami, Uchida) used to calculate the heat transfer from the containment atmosphere to the passive heat sinks, and provide a plot of the heat transfer coefficient versus time for the most severe steam line break accident analyzed.

(4) Specify and justify the temperature used in the calculation of condensing heat transfer to the passive heat sinks; i.e.,

specify whether the saturation temperature corresponding to the partial pressure of vapor, or the atmosphere temperature (which may be superheated)was used.

(5) Discuss and justify the analytical model including the themodynamic equations used to account for the removal of the condensed mass from the containment atmosphere due to condensing heat transfer to the passive heat sinks; (6) Provide a table of the peak values of containment atmosphere temperature and pressure for the spectrum of break areas and power levels analyzed; (7) For the case which results in the maximum containment atmcsphere temperature, graphically show the containment atmoschere temcerature,

% ., --- ,4- 2 . 34., -

3, . ,, _ w -wg ,,

,4., Tent concrete

. .. . . .. .. . . . . . . ...p a . S "..'. a c a '. C u i a ad C o G ~. L 'l-ment atmosphere temperature response to the temperature profile used in the environmental qualification program for those safety related instruments and mechanical components needed to mitigate the consecuences of the assumed main steam line break and effect safe reactor shutdown;

(8) For the case which results in maximum containment atmosphere pressure, graphically show the containment pressure as a function of time; and (9) For the case which results in the maximum containment atmosphere pressure and temperature, provide the mass and energy release data in tabular form. .

In order to demonstrate that safety-related equipment has oeen adequately qualified as described above, provide the following information regard-ing its environmental qualification.

(1) Provide a comprehensive list of equipment required to be operational in the event of a main steamline break (MSLB) accident. The list should include, but not necessarily be limited to, the following safety related equipment:

(a) Electrical containment penetrations; (b) Pressure transmitters; (c) Containment isolation valves; (d) Electrical power cables; (e) Electrical instrumentation cable; and (f) Level transmitters.

Describe the qualification testing that was, or will be, done on this equipment.

Include a discussion of the test environment, namely, the temperature, pressure, moisture content, and chemical spray, as a function of time.

(2) It is our position that the thermal analysis of safety related equipment which may be exposed to the containment atmosphere following a main steam line break accident should be based on the following:

(a) A condensing heat transfer coefficient based on the recommendations in Branch Technical Position CSB 6-1,

" Minimum Containment Pressure Model for PWR ECCS Performance Evaluation,"should be used.

(b) A convective heat transfer coefficient should be used when the condensing heat flux is calculateu to be less than the convective heat flux. During the blowdown period it is appropriate to use a conservatively evaluated forced convection heat transfer correlation. For example,

Nu = C(Re)

Where Nu = Nusselt No.

Re = Reynolds No.

_ _._C. = erpirical constants dependent .on _ , _

geometry and Reynolds No.

Since the Reynolds number is dependent on velocity, it is necessary to evaluate the forced ficw currents which will be generated by the steam generaar blowdown. The CVTR experiments provide limited data in this regard. Convective currents of from 10 ft/sec to 30 ft/sec were measured locally. We recommend that the CVTR test results be extrapolated conservatively to obtain forced flow currents to determine the convective heat transfer coefficient during the blowdown period. After the blowdown has ceased or been reduced to a negligibly low value, a natural convection heat transfer correiation is acceptable.

(3) For each ccmponent where thermal analysis is done in conjunction with an environmental test at a temperature lower than the peak calculated temperature following a main steam 1 ne break accident compare the test thermal response of the compopant with the accident thermal analysis of the component. Provide the basis by which the component thermal response was developed from the environmental qualification test program. For instance, graphically show the thermocouple data and discuss the thermocouple locations, method of attachment, and performance characteristics, or provide a detailed discussion of the analytical model used to evaluate the component thermal response during the test. Th!3 ualuation should be performed for the potential points of failur u .ch as thin cross-sections and temperature sensitive parts wnere thermal stressing, temperature-related degradation, steam or chemical interaction at elevated temperatures, or other thermal effects could result in the failure of the component mechanically or electrically. If the component thermal response comparison results in the prediction of a more severe churmal transient for the accident conditions than for the qualification test, provide justification that the affected component will perform its intended function during a MSLB accident, or provide protection for the component whch would appropriately limit the thermal effects.

IV.C .16 ENVIRONMENTAL EFFECT OF PIPE FAILURES (3.6.1, 3.6.2) Identify the " break exclusion" regions of the main steam and feedwater lines. Compartments that contain break exclusion regions of main steam and feedwater lines and any safety related equipment in these compartments should be designed to with- -

stand the environmental effects (pressure, temperature, humidity and flooding) of a crack with a break area equal to the cross sectichal area of the treak excluded' pipe.

IV .C .17 DESIGN RECUIREMENTS FCR C00 LING WATER (5.4.1) TO REACTCR CCCLANT PUMPS Demonstrate that the reactor coolant system (RCS) pump seal injection flow will be automatically maintained for all transients and accidents or that enough time and information are available to permit gcorrective action by an operator.

We have established the following criteria for that portion of the component cooling water (CCW) system which interfaces with the reactor coolant pumps to supply cooling water to pumo seals and bearings during ndrmal operation, anticipated transients, and accidents.

1. A single active failure in the component cooling water system shall not result in fuel demage or a breach of the reactor coolant pressure boundary (RCPB) caused by an extended loss of cooling to one or more pumps. Single active failures include operator error, spuricas actuation of motor-operated valves, and loss of CCW pumps.

2. A pioe crack or other accident (unanticipated occurrence) shall not result in either a breach cf the RCPB or excessive fuel damage when an extended loss of cooling to two or more RC cumps occurs. A single active falure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Technical Position ASB 3-1.

In order to meet the criteria established above, an NSSS intar-face requirement should be imposed 01 the balance-of-plant CCW system that provides cooling water to the RC pump seals and notor and pump bearings, so that the system will meet the following con-ditions:

1. That pertion of the component cooling water (CCW) system which supplies cooling water to the reactor coolant pumps and motors may be designed to non-seismic Category I req. irements and Quality Group D if it can be demonstrated that the reactor coolant pumps will operate without component cooling water for at least 30 minutes without loss of function or the need for operator pro-tecti ~ve action. Irr addition, safety grade instrumentatiotr -- -- --

including alarms should be provided to detect the loss of -

component cooling water to the reactor coolant pumps and motors, and to notify the operator in the control rocm. The entire instrumentation system, including audible and visual alarms, should meet the requirements of IEEE Std 279-1971.

If it is not demonstrated that tNe reactor coolant pumps and motors will operate at least 30 minutes without loss of function or operator protective action, then the design of the CCW sys tem must meet the following requirements:

1. Safety grade instrumentation consistent with the criteria for the reactor protection system shall be provided to initiate automatic protection of the plant. For this case, the component cooling water supply to the seals and pump and motor bearings may be designed to non-seismic Ca tegory I require-ments and Quality Group D; or

2. The component cooling water supply to the pumps and motors shall be capable of withstanding a single active failure or a moderate energy line crack as defined in our Branch Technical Position APCSB 3-1 and be designed to seismic Category I, Quality Group D and ASME Section III, Class 3 requirements.

The reactor coolant (RC) pumps and motors are within the NSSS scope of design. Therefore, in order to demonstrate that an RC pump design can operate with loss of component cooling water for at least 30 minutes without loss of function or the need for operator action, the following must be provided:

1. A detailed description of the events following the loss of component cooling water to the RC pumps and an analysis demon-strating that no consequences important to safety may result from this event. Include a discussion of the effect that the Icss of cooling watar to the seal coolers nas on ne RC pump seals. Shcw that the loss of cooling water does not result in a LOCA due to seal failure.

2. A detail.ed analysis to shcw that loss of cooling water to the RC pumps and motors will not cause a loss of the flow coastdown characteristics or cause seizure of the pumps, assuming no administrative action is taken. The response should include a detailed description of the calculation procedure including:

a. The equations used. ~~

b. The parameters used in the equations, such as the design parameters for the motor bearings, motor, pump and any other equipment enter'ng into the calculations, and material property values for the oil and metal parts.

c. A discussion of the effects of possible variations in part dimensions and material properties, such as bearing clearance tolerances and misalignment.

d. A description of the cooling and lubricating sys' ems (with appropriate figures) associated with the RC pump and motor and their design criteria and standards.

e. Information to verify the applicability of the equations and material properties chosen for the analysis (i.e.,

references should be listed, and if empirical relations are used, provide a comparison of their range of appli-cation to the range used in the analysis).

Should an analysis be provided to demonstrate that loss of component cooling water to the RC pumps and motor assembly is '

acceptable, we will require certain modifications to the plant Technical Speci"ications and an RC pump test conducted under operating condtions and with component cooling water terminated for a specified period of time to verify the analysis.

IV.C.18 WATER HAMMER IN STEAM GENERATORS WITH TOP FEEDRING DESIGN (10.4.7)

Events such as damage to the feedwater system piping at Indian Point Unit No. 2, November 13, 1973, and at other plants, could originate as a consequence of uncovering of the feedwater sparger in the steam generator or uncovering of the steam generator feedwater inlet no::les. Subsequent events may in turn lead to the generation of a pressure wave that is propagated through the pipes and could result in unacceptable damage.

For CP/PDA and OL/FDA applications, provide the follow;ng for steam generators utilizing top feed:

1. Prevent or delay water draining from the feedring following a drop in steam generator water level by means such as J-Tubes;

2. Minimize the volume of feedwater piping external to the . steam _ _ _ _ _ .

- generator whch could pocket steam using the shortest possible- - - -

(less than seven feet) horizontal run of inlet piping to the steam generator feedring; and

3. Perform tests acceptable to the staff to verify that unacceptable feed-water hammer will not occur using the plant ocerating procedures for normal and emergency restoration of steam generator water level follcwing loss of normal feedwater and possible draining of the feedring. Provide the procedures for these tests for staff approval before conducting the tests.

Furthermore, we request that the following be provided:

a. Describe normal operating occurrences of transients that could cause the water level in the steam generator to drop below the sparger or nozzles to cause uncovering and allow steam to enter the sparger and feedwater piping.

b. Describe your criteria or show by isometric diagrams, the routing of the feedwater piping from the steam generators outwards to beyond the containment structure up to the auter isolation valve and restraint.

c. Describe any analysis on the piping system including any forcing functions that will be performed or the results

, of test programs to verify that ,either uncoverino of feedwater line, could not occur or that, if it did occur, unacceptable damage such as the experience at the Indian Point Unit No. 2 facility would not result with your design.

~

IV,C.19 E?lVIR0flMEllTAL C0tlTROL SYSTEMS FOR SAFETY RELATED ECilipMENT

( "" } Most plant areas that contain safety related equipment depend on the continuous operation of environmental control systems to maintain the environment in those areas within the range of environmental qualification of the safety related equipment installed in those areas. It appears that tnere are no requireme'nts for maintaining these environmental ~

control systems in operation while the plant is shutdown or in hot standby conditions. During periods when these environmental control systems are shutdown, the safety related equipment could be exposed to environmental conditions for which it has not been qualified. Therefore, the safety related equipment should be qualified to the extreme environmental conditions that could occur when the control equipment is shutdown or these environmental control systems should operate continuously to maintain the environmental conditions within the qualification limits of the safety related equipment. In the second case an environmental monitoring system that will alarm when the environmental conditions exceed those for which safety related equipment is qualified shall be provided. This environmental monitoring system shall (1) be of high quality, (2) be periodically tested and calibrated to verify its continued functioning, (3) be energized from continuous power sources, and (4) provide a continuous record of the environmental parameters during the time the environmental conditians exceed the normal 1imits.

m.

• O &

W

\