ML20236Q542

Requests Addl Info Re Proposed Inservice Testing Program for safety-related Pumps & Valves,Described in Util . Discussion of Request for Addl Info at Proposed Dec 1987 or Jan 1988 Util/Nrc Inservice Testing Conference Recommended

ML20236Q542
Person / Time
Site:	Calvert Cliffs
Issue date:	11/13/1987
From:	Mcneil S Office of Nuclear Reactor Regulation
To:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
References
TAC-64976, TAC-64977, NUDOCS 8711200117
Download: ML20236Q542 (17)

Text

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t November 13, 1987 i

Docket Nos. 50-317 DISTRIBUTION and 50-318

• Docketg11e:* SVarga NRCPDR BBoger ,

local PDR EJordan PDI-1 Rdg. - J Partl ow - )

. . SMcNeil ACRS(10).

Mr. J. A. Tiernan LTripp OGC ' 'i Vice President-Nuclear Energy j Baltimore Gas and Electric Company '

P.O. Box 1475 Baltimore, Maryland 21203

SUBJECT:

REQUEST _FOR. ADDITIONAL INFORMATION - PROPOSED PUMP AND VALVE INSERVICE TESTING PROGRAM (TACS 64976 & 64977) .

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We have determined that additional information is necessary to facilitate our ]

review of your proposed inservice test (IST) program for safety-related. pumps- i and valves as provided in your February 26, 1987 submittal. The request for I additional.information (RAI) is enclosed. Branch Technical Position RSB 5-1, l

" Design Requirements of the Residual Heat Removal System," is enclosed also, j to facilitate your response to Section I.B of the RAI. ';

Due to the scope of the questions presented in this RAI, we propose that it would be of benefit for this RAI to be discussed in a BG&E/NRC IST conference to be held in December of 1987 or January 1988. At this conference, preliminary draft responses to these questions should be made available for discussion. A formal response date for the RAI would be determined after the conference.

Please inform me of a proposed IST conference date by November 30, 1987. If you do not desire to hold such a conference, please provide the information as requested by January 15, 1988.

This request for information affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

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i Sincerely, I

~

Scott Alexander McNeil, Project Manager ,

Project Directorate I-1 i Division of Reactor Projects, I/II l

Enclosure:

As stated i cc: See next page PDI-1 g PDI-I kI-1 l l CVogan SMcNe RCapra  !

l  !(/g/87 10/13/87 10/[3/87 8711200117 317 l PDR ADOCK % PDR ,

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t . -Mr. J. A. Tiernan I

Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc:

Mr. John M. Gott, President Regional Administrator, Region I Calvert County Board of U.S. Nuclear Regulatory Commission Commissioners Office of Executive Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue D. A. Brune, Esq. King of Prussia, Pennsylvania 19406 General Counsel Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203 Jay E. Silberg, Esq.

Shaw, Pittman, Potts and Trowbridge 1800 M Street, NW j

Washington, DC 20036 4 Mr. M. E. Bowman, General Supervisor Technica1' Services Engineering Calvert Cliffs Nuclear Power Plant ,

MD Rts 2 & 4, P. O. Box 1535 .

Lusby, Maryland 20657-0073 Resident Inspector c/o U.S. Nuclear Regulatory Commission P. O. Box 437 Lusby, Maryland 20657-0073 Bechtel Power Corporation ATTN: Mr. D. E. Stewart Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Combustion Engineering Inc.

ATTN: Mr. W. R. Horlacher, III Project Manager P. O. Box 500 1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Department of Natural Resources Energy Administration, Power Plant Siting Program ATTN: Mr. T. Magette Tawes State Office Building Annapolis, Maryland 21204 I

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s REQUEST FOR ADDI110NAL INFORMATION '

AND PROPOSED MTETING AGENDA QUESTIONS AND COMMENTS CONCERNING PUMP AND VALVE INSERVICE TESTING PROGRAM BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET N05. 50-31/ AND 50-318 f s

I. VALVE TESTING PROGRAM A. General Ouestions and Comments j

1. Where the requirements of Section XI are more stringent than those identified in the Calvert Cliffs-Technical Specifications, the Section XI )

requirements must be adhered to or a request for relief from the i applicable Code requirement (s) must be submitted. . The statement in Section 1.1, page 1, of the Calvert Cliffs IST program implies that any Technical Specification requirements less conservative than the Section XI requirements will take precedence. ]

2. The NRC has concluded that the applicable leak test procedures and 1

requirements Appendix for containment isolation valves are determined by 10CFR50, J. Relief 1

from paragraphs IWV-3421 through 3425 for containment isolation valves presents no safety problem since the intent of IWV-3421 through 3425 is met by Appendix J requirements'. However, the licensee shall comply with Paragraphs IWV-3426 and 3427. See Section 3.1, page 3, of the Calvert Cliffs IST program discussion of " Category A Valves".

3. Provide a listing of all valves that are Appendix J, type C, leak rate tested which are not included in the IST program and categorized A or A/C.

4 What criteria is utilized for assigning limiting values of full-stroke '

times for power operated valves in the Calvert Cliffs IST program (see Section 3.9, on page 6 of the IST program)?

5 Amplify the term " minor maintenance" in Section 3.11, on page 6 of the IST program. This statement appears to conflict with Section IWV-3200 in that the valve may need to be tested to demonstrate operability prior to return to service.

6 The Code permits valves to be exercised during cold shutdowns where it is not practical to exercise them during plant operation. These valves must be specifically identified by the licensee and are full-stroke exercised during cold shutdowns. The NRC requires that the licensee provide a technical justification for each valve that cannot be exercised quarterly during power operations that clearly explains the difficulties or hazards encountered during that testing. The NRC staff will then verify that it is not practical to exercise those valves and that the testing should be performed during cold shutdowns. The cold shutdown justifications in the Calvert Cliffs IST program need to include more detailed information.

7. Review the safety-related function of any pumps or valves in the control room ventilation system to determine if they should be included in the IST program and tested in accordance with the requirements of Section XI.

Provide justification for those not included and tested per Section XI.

8. The NRC staff's position is that emergency diesel generators perform a safety-related function and therefore, the appropriate valves in the emergency diesel generator air start system should be included in the IST program and tested in accordance with the Code requirements.

9. Identify the valves which are full-stroke exercised on a cold shutdown  ;

frequency and not partial-stroke exercised quarterly during power l operation as required by Section XI, subsections IWV-3412 and 3522.

10. Review the safety-related function of the spent fuel cooling system to  ;

determined if it should be included in the IST program with the applicable '

system components tested in accordance with the requirements of Section XI.  ;

Provide justification for those not included and tested per Section XI.

11. Solenoid operated valves are not exempt from the stroke time measurement )

l requirements of Section XI. Their stroke times must be measured and corrective '

action taken if these times exceed the full-stroke time limiting values.

The NRC staff will consider granting relief from the trending requirements of Section XI (Paragraph IWV-3417 (a)) for these rapid acting valves.

However, in requesting this relief the licensee must assign a maximum i limiting stroke time of 2 seconds to these valves and perform corrective .

action as required by IWV-3417 (b) if the measured stroke times exceed the l 2 second limit. See valve relief reouest number A-1.

12. What alternate testing has been considered for verifying that the remote l position indication accurately reflects the actual valve position for the '

valves affected by valve relief request number A-2?

13. When flow through a check valve is used to indicate a full-stroke exercise of the valve disk, the NRC staff's position is that verification of the maximum flow rate through the valve as identified in any of ti,e plant's safety analyses, would be an adequate demonstration of the full-stroke requirement. Any flow rate less than this will be considered

! partial-stroke exercising unless it can be shown, by some means such as measurement of the differential pressure across the valve, that the check valve's disk position at the lower flow rate would permit maximum required flow through the valve.

B. Main Steam System If credit is taken for the operability of the atmospheric steam dumps 1(2)-CV-3938 and 3939, do they satisfy the requirements of Reactor Systems Branch Position RSB 5-17 (See enclosed Branch Technical Position RSB 5-1.)

Should these valves be included in the IST program and tested in accordance with the Code requirements?

C. Service Water Cooling System

1. How are valves 1(2)-SRW-314, 315, and 316 verified to.be full-stroke exercised quarterly?

2. How are valves 1-SRW-321, and 322 and 2-SRW-321 verified to be full-stroke exercised quarterly?

D. Circulating Saltwater Cooling System

1. How are valves 1(2)-SW-103,107, and 111 verified to be full-stroke exercised open quarterly?

2. Review the safety-related function of valves 1(2)-CV-5174 and 5175 (P&ID-No. O'4-49/450 Coords. H-1) to determine if they should be included in the IST program and tested to the Code requirements. Provide justification if not included and tested per Code requirements.

3. Review the safety-related function of valves 1(2)-CV-5150, 5152, and 5153 (P&ID No. OM-49/450 Coords. E-7, G-7, and-F-12) to determine if they should j be included in the IST program and tested to the Code requirements. i Provide justification if not included and tested per Code requirements.

E. Component-Cooling-System

1. How are valves 1(2)-CC-115,120, and 125 verified to be full-stroke l exercised open quarterly? l

2. Review the safety-related function of valves 1(2)-CV-3823 and 3825 (P&ID' No. OM-51/452 Sh. 2, Coords. A-7 and C-7) to determine if they should be included in the IST program and tested to the Code requirements.- Provide justification if not included and tested per code requirements.

F. Containment Ventilation System Provide a more detailed technical justification for not. full-stroke exercising valves 1(2)-CV-1410, 1411, 1412, and 1413 on a cold shutdown frequency.

G. Reactor Coolant and Water Process Sample System ,

Provide a more detailed technical justification for not full-stroke exercising valves 1(2)-SV-6529 quarterly.

4 H. Nitrogen Generating-Blanketino System Provide P&ID No. OM-68 for our review.

_ _ - - - ___ _ _ - _ .- - __ _. ]

., I. Plant Heatino System j 4

Valve 1-MOV-6579 is indicated to be normally open on P&ID No. OM-71.yet j its identified safety-related position is- closed. If this' valve must' change' position to perform its. safety function, it is not passive and must.be included in.the IST program.and tested in accordance with' Code requirements.

4 J. Reactor Coolant System Provide a more detailed technical justification for not full-stroke I exercising valves 1(2)-ERV-402 and 404. quarterly (see relief request No.

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RC-7). 1j K. Chemical and Volume' Control-System.

1. How are valves 1(2)-CVC-217 and 222' verified to full-stroke open quarterly? ,j

2. How are valves 1(2)-CVC-228 and 235 verified'to full-stroke open during ,

cold shutdowns? Provide'a more detailed technical justification for not ,I full-stroke exercising these valves quarterly (see cold shutdown:

justification).

3. Provide a more detailed technical justification for.not full-stroke exercising valves 1(2)-MOV-501 quarterly (see cold shutdown justification).

4 Review the safety-related function of valves; 1(2)-CVC-251 (P&ID No. 0M-73 Sh. 1, Coords. C-5) to determine.if they should be included.in the IST program and tested 'to the Code requirements. Provide justification if not included and tested per Code requirements.

5. How are valves 1(2)-CVC-186 verified to full-stroke' open : quarterly during .

1 power operations?

6 If valves 1(2)-CV-110P and 1100 (P&ID No. OM-73, Sh. 3, Coords. B-3) are Appendix J, type C leak-rate tested, theylshould be included in the IST program and tested to the Code requirements.

7. If either valve 1(2)-CVC-103 or- 105 is. normally in the open: position and.

J its. safety-related function (as a CIV) is in the closed position, it is an active valve and it should be full-stroke exercised ouarterly in accordance.with the Code requirements. Provide justification if not included and tested per Code requirements.

8. Review the safety-related function of valves 1(2)-MOV-504'(P&ID No. M-73, l Sh.1, Coords. F-3) to determine if they should be included in the IST program.and tested in accordance with the Code requirements. . Provide' justification if not included and tested'per Code requirements, l

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9. Review the safety-related function of valves 1(2)-CVC-257 (P&ID No. M-73, Sh.1, Coords. F-3) to determine if they should be included in the IST program and tested in accordance with the Code requirements. Provide justification if not included and tested per Code requirements.

L. Safety Injection & Containment Spray System

1. Could full-stroke exercising valves 1(2)-SI-113, 123, 133, and 143 during cold shutdowns result in a low temperature overpressurization of the reactor coolant system?

2. Valves 1(2)-CVC-401 and 410 appear to perform a safety-related function in the closed position, if so, they should be exercised to the closed position quarterly as required by the Code.

3. How are valves 1(2)-SI-422, 424, and 426 verified to full-stroke open quarterly?

4. How are valves 1(2)-SI-448 and 451 verified to full-stroke open quarterly? ,

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5. What are the consequences of failure of either valve 1(2)-MOV-659 or 660 in the closed position during quarterly testing? Would this render an entire safety system unavailable to perform its safety function?

6. How are valves 1(2)-SI-4146 and 4147 verified to full-stroke open during cold shutdowns?

7. What are the consequences of failure of either valve 1(2)-MOV-4142 or 4143 in the closed position during quarterly testing?

8. Review the safety-related function of valves 1(2)-MOV-654 and 656 (P&ID No. OM-74, Sh. 1, Coords. F-5 and D-5) to determine if they should be included in the IST program and tested to the Code requirements. Provide justification if not included and tested per Code requirements.

9. What is the proposed test frequency for disassembly and inspection of valves 1(2)-SI-215, 225, 235, and 245? (see valve relief request no.

SI-3)

10. What is the proposed test frequency for disassembly and inspection of valves 1(2)-SI-217, 227, 237, and 247? There is an apparent typographical error in the basis for relief (see valve relief request no. SI-5).

11. Provide a more detailed technical justification for not leak rate testing valves 1(2)-SI-217, 227, 237, and 247 in accordance with the Code requirements.

12. Provide a more detailed technical justification for not full-stroke exercising valves 1(2)-CV-306 quarterly in accordance with the Code requirements.

13. Provide the stroke time limit for valves 1(2)-CV-306, 14 Provide a more detailed technical justification for not full-stroke exercising valves 1(2)-MOV-651 and 652 quarterly in accordance with the ,

Code requirements. Do these valves perform a pressure boundary I' isolation function? Provide the stroke time limits for these valves.

15. Provide a more detailed technical justification for not full-stroke exercising valves 1(2)-MOV-651 and 652 quarterly in accordance with Code requirements. Are these valves verified to full-stroke during cold shutdown testing?

16. What is the proposed test frequency for disassembly and inspection of valves 1(2)-SI-316, 326, 330, and 340? (see valve relief request no.

SI-1).

17 How are valves 1(2)-SI-334 and 344 verified to full-stroke during quarterly testing?

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18. Provide the stroke time limits for valves 1(2)-CV-657 and 1(2)-MOV-658.

19. Is valve position indication verified once every two years for valves 1(2)-MOV-4144 and 4145 in accordance with the Code requirements?

20. Provide a more detailed technical justification for not full-stroke exercising valves 1(2)-S1-4148 and 4149 quarterly in accordance with the Code requirements. What is the proposed test frequency for disassembly and inspection of these valves?

M. Reactor Coolant Waste Processing System Provide the P&ID that shows valves 1(2)-ES-142 and 143.

N. Gas Analyzing System Are the following valves ever opened during power operation? Provide a more detailed technical justification for not exercising these valves quarterly during power operations.

1(2)-SV-6507A-F 1(2)-SV-6507G 1(2)-SV-6531 1(2)-SV-6540A-F 1(2)-SV-6540G I

-x__.-_-_-__--__---_____________._______.

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0. Auxiliary Feedwater System i

1. Do valves 1(2)-MS-103 and 106 perform a safety-related function in the closed position as well as the open-position?

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How ore valves 1(2)-MS-108 and 110 verified to full-stroke during I ouarterly testing? i i

3. How are valves 1(2)-MS-114 and 128 verified to full-stroke during  ;

quarterly testing? l 4 Do valves 1(2)-MS-129 and 130 perform a safety-related function in the closed position?

5. How are valves 1(2)-MS-201 and 202 verified to full-stroke during quarterly testing l 4

6 Review the safety-related function of valves 1(2)-CA-337 (P&ID No. M-800, Sh. 1, Coords. B-12) to determine if they should be included in the IST program and tested in accordance with the Code requirements. Provide i l

. justification if not included and tested per Code requirements.

II. PUMP TESTING PROGRAM

1. The NRC staff's position is that the emergency diesel generators perform a safety-related function and therefore, the emergency diesel generator fuel oil transfer pumps shculd be included in the IST program and tested in accordance with the Code requirements.

2. How is flow rate measured during ouarterly testing of the high pressure \

safety injection pumps 11, 12, 13, 21, 22, and 23? I

3. Pump relief request number 2 is not reouired since there is no lubricant level or pressure to observe.

4 J How is flow rate measured during quarterly testing of the low pressure safety injection pumps 11, 12, 21, and 22?

5.

How is flow rate measured during quarterly testing of the containment spray pumps 11, 12, 21, and 227 I

6. )

Provide a more detailed technical justification for not measuring inlet i pressure for the salt water pumps 11, 12, 13, 21, 22, and 23 during quarterly pump testing in accordance with the Code requirements. How is differential pressure measured for these pumps?

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7 How is flow rate measured during quarterly testing of the salt water pumps 11, 12, 13, 21, 22, and 237 1

8. Are vibration measurements being taken at all required locations on salt j water pumps 11,12,13,. 21, 22,. and 237 {

9. How is flow rate measured during quarterly testing of the service water i pumps 11, 12, 13. 21, 22, and 237-

10. How is flow rate measured during quarterly testing of the component .l cooling water pumps 11, 12, 13, 21, 22, and 237 '

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4 BRANCH TECHNICAL' POSITION RSB 5-1 DESIGN RE@IREMENT5 0F THE RESIDUAL HEAT REMOVAL SYSTEM I BACKGROUND h

GDC 19 states that, "A control room'shall be provided from which actions can.

be taken to operate the nuclear power unit'under normal conditions. . ." ,

Normal operating conditions including the shutting down of a reactor; therefore, I since the residual heat removal (RHR) system is one of several systems involved l

in the normal shutdown of all reactors, this system must be operable from the j control room.

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.GDC 34 states that " Suitable' redundance. . .shall be provided to assure that:

for onsite electrical power system operation (assuming offsite power is not i avMable) and for offsite electrical power system operation (assuming onsite i pc cr is not available), the system safety function can be accomplished,  !

assuming a single failure." '

In most current plant designs the RHR system has a: lower design pressure than the reactor coolant system RCS), is located outside of containment and'is j.

part of the emergency core c(ooling system (ECCS). However, it is possible for  !'

the RHR system to have different design characteristics. For example, the RHR system might have the same design pressure as.the RCS, or be located inside.of  ;

containment. Plants which may have RHR systems that deviate from current designs will be reviewed on a case-by-case basis. The functional, isolation, ,

pressure relief, pump protection, and test requirements for the RHR system are.  :

included in this position.

BRANCH POSITION A. Functional Requirements i

The system (s) which can be used to take the reactor from normal operating i conditions to cold shutdown

• shall satisfy the functional requirements'>1isted.  ;

below.

1. The design shall be such that the reactor can be taken from normal. 1 operating conditions to cold shutdown.using only safety grade systems.

These systems shall satisfy General Design Criteria 1 through 5.  ;

2. The system (s) shall have suitable redundancy in components and

• features, and suitable interconnections, leak detection, and isolation.

capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not >

available) the system function can be accomplished assuming a single failure. o

" Processes involved in cooldown are heat. removal, depressurization, flow' circulation, and reactivity control. The cold shutdown condition, as  ;

described in the Standard Technical Specifications, refers to a sub-  !

critical reactor with a reactor coolant temperature no greater than 200 F for a PWR and 212 F for a BWR.

l 5.4.7-12 Rev. 2 July 1981

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3. The system (s) shall be capable of being operated from the control )

. room with either only onsite or only offsite power available. In j demonstrating that the system can perform its function assuming a '

( single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.

4. The system (s) shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

B. RHR System Isolation Requirements-The RHR system shall satisfy the isolation requirements listed below.

1. The following shall be provided in the suction side of the RHR- )

system to isolate it from the RCS.

(a) Isolation shall be provided by at least two power-operated valves in series. The valve positions shall be indicated in the control room.

(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below 3 the RHR system design pressure. Failure of a power supply l shall not cause any valve to change position.

(c) The valves shall have independent diverse interlocks to protect against one or both valves being o>en during an RCS increase

( above the design pressure of the RiR system.

2. One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:

l (a) The valves, position indicators, and interlocks described in item 1(a)thru1(c)above,  ;

(b) One or more check valves'in series with a normally c16 sed  !

power-operated valve. The power-operated valve position shall  ;

be indicated in the control room. If the RHR system discharge i line is used for an ECCS function, the power-operated valve is -

to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure. ,

(c) Three check valves in series, or I (d) Two check valves in series, provided that there are design provisions to permit periodic testing of the check valves for ,

leak tightness and the testing is performed at least' annually. '

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5.4.7-13 Rev. 2 - July 1981

I C. Pressure Relief Requirements The RHR system shall satisfy the pressure relief requirements listed below.

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1. To protect the RHR system against accidental overpressurization when i it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a PWR with no steam bubble in the pres-surizer, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be considered in selection of the design bases.

2. Fluid discharged through the RHR system pressure relief valves must be collected and contained such that a stuck open relief valve will not:

(a) Result in flooding of any safety-related equipment.

(b) Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.

(c) Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment.

3. If interlocks are provided to automatically close the isolation (

valves when the RCS pressure exceeds the RHR system design pressure, adequate relief capacity shall be provided during the time period while the valves are closing.

l D. Pump Protection Requirements l

l The design and operating procedures of any RHR system shall have provisions to prevent damage to the RHR system due to overheating, cavitation or loss of adequate pump suction fluid.

E. Test Requirements The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in the RHR mode. Testability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22.

The preoperational and initial startup test program shall be in conformance with Regulatory Guide 1.68. The programs for PWRs shall include tests with supporting analysis to (a) confirm that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve.such mixing, and (b) confirm that the cooldown under natural circulation conditions can be achieved within the ifmits specified in the emergency operating procedures.

Comparison with performance of previously tested plants of similar design may be substituted for these tests.

5.4.7-14 Rev. 2 - July 1981

. F. Operational Procedures The operational procedures for bringing the 31 ant from normal operatin power

( to cold shutdown shall be in conformance wit 1 Regulatory Guide 1.33. or  !

pressurized water reactors, the operational procedures shall include specific '

procedures and information required for cooldown under natural circulation conditions.

G. Auxiliary Feedwater Supply The seismic Category I water supply for the auxiliary feedwater system for a 3 PWR shall have sufficient inventory to permit operation at hot shutdown for at ]

least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by cooldown to the conditions permitting operation of '

the RHR system. The inventory needed for cooldown shall be based on the longest cooldown time needed with either'only onsite or only offsite power ,

available with an assumed single failure. ,

i H. Implementation j rposes of implementing the requirements for plant heat removal Forthebtyforcompliancewitkthisposition,plantsaredividedintothe capabili I following three classes:

Class 1 -

Full compliance with this position for all plants (custom or standard) for which CP or 30A applications are docketed on or i after January 1, 1978. See Table 1 for possible solutions for I full compliance. l

( Class 2 -

Partial implementation of this position for all plants (custom or standard) for which CP or PDA applications are docketed before January 1, 1978, and for which an OL issuance is expected on or after January 1,1979. See Table 1 for recommended implementation for Class 2 plants. ,

1 Class 3 -

The extent to which the implementation guidance in Table 1 will  !

be backfitted for all operating reactors and all other plants  ;

(custom or standard) for which issuance of the OL is ex  !

before January 1, 1979, will be based on the combined I&pected E and DDR review of related plant features for operating reactors.

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5.4.7-15 Rev. 2 - July 1981

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