ML19318B862

Discusses Proposed Implementation of Revised Cooling Sys Requirements & Risk Assessment Re Overall Plant Heat Removal Capability

ML19318B862
Person / Time
Issue date:	01/19/1978
From:	Mattson R, Stello V Office of Nuclear Reactor Regulation
To:	Case E Office of Nuclear Reactor Regulation
Shared Package
ML19318B856	List: ML19318B855 ML19318B862
References
REF-GTECI-A-31, REF-GTECI-DC, RTR-REGGD-01.139, RTR-REGGD-1.139, TASK-A-31, TASK-EM-801-4, TASK-OR, TASK-OS NUDOCS 8006300201
Download: ML19318B862 (8)

Text

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EHCLOSURE 1

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'" ""%,i UMTED STATES NUCLEAR REGULATORY CCMMISSION

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wAsmHGTON, D. C. 20555

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JAN 191978

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MEMORANDUM FOR:

E. G. Case, Acting Director, Office of Nuclear Reactor Regu1ation FROM:

R. J. Mattson, Director, Division of Systems Safety Victor Stello, Jr., Director, Division of Operating Reactors

SUBJECT:

REVISED IMPLEMENTATION SCHEDULE OF TASK A-31, RHR H

SHUT 00WN REQUIREMENTS REFERENCE 1:

Memorandum from R. J. Mattson to E G. Case dated 11/8/77 re "Reques-t for.RRRC Consideration - Proposed Revision to Standard Review Plan 5.4.7, Residual Heat Removal System" l1 Introduction re.:

At Meeting Number 69 of the Regulatory Requirements Review Committee on.

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November 29, 1977, DSS reccmmended that the revised SRP 5.4.7 and Branch Technical Position RSB 5-1 be backfitted to all standard plants and all PWR plants that had received construction permits since November 1,1975 -

(see Ref. 1). Following the discussion of this proposed implementation

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schedule the RRRC requested that more definitive plans for implementation,.

including criteria for case-by-case review of older plants, be prepared for consideration by the RRRC. This memorandum rovides the requested implementation plan for plants under the cognizance of both OSS and 00R.

It also includes a brief discussion of risk assessments with respect to requirements for overall plant heat removal capability in response to questions raised at Meeting Number 69.

Procosed Imolementation of Revised Cooling Systems Recuirements The Engineering Methodology Standards Branch of the Division of Engineering Standards has prepared a draft guide on plant heat removal capability require-ments.

It is estimated that the guide will be issued for a 60-day public ccm=ent period by January 31, 1978. Following review of public comments, a revised guide will be prepared and sent to the ACRS, RRRC, and the NRC staff for review.-

The EMSB estimates that this additional review and preparation of the final modifications to the guide will require an additional six months. Hence, the estimated date for issuance of the effective guide is September 1978.

This issuance date was considered in the development of the implementation plan described below.

Contacts: Chuck Graves, NRR 49-27591 Brad Hardin, 00R 49-28050 O

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JAN 13 E3 E. G. Case

-2' There are a 'nummer' of differences in r'equirements between the existing SRP 5.4.7 and the proposed BTP RSB 5-1.

These differences and the resulting impacts are summarized in Table 1.

Table 2 provides a summary of specific components or systems affected by the requirements. Table 2 also describes various methods for full and partial compliance with RSB 5-1.

It should be noted that with regard to the RHR design features, all operating reactors and most reactors currently under OL review have a single drop line with two isolation valves connecting the RCS to RHRS and -

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thus are. susceptible to single failures..

For purposes of implementing the requirements for plant heat removal capabi,lity, plants are divided into the following three classes:

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] Class 1 - All plants (custom or standard) for which CP or PDA applications are docketed on or after January 1,1978.,

i Class 2 - All plants (custen. or standard) for which CP or PDA.

applications are docketed before January 1,1978 and for which an CL issuance is expected on er after January 1,.1979 '

C1' ass.3 - All operating reactors and all oth' r plants (custom

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e or standard) for which issuance of the CL is expected

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before January 1,1979.

The following implementation schedule is recommended:

(A) Revised SRP 5.4.7 will be used by DSS in the review of C1' ass 1 plants for CP or PDA until the proposed guide (which will be based on the revised SRP 5.4.7) becemes effective. It is anticipated that the guide would then be an RRRC Category 1

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( B) The guidance for partial implementation of revised SRP 5.4.7 presented in Table 2 will be used by 055 in the review. of C1 ass 2 pl ants.

( C) For Class 3 plants, the extent to which the implementation guidance in Table 2 will be backfitted will be based on the combined I&E and 00R review of related plant features for operating reactors. The format of Table 2 serves as a basis for documenting the degree of ccmpliance and for identifying possible alternate means for satisfying the intent of the proposed guide (or the revised SRP).

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A more specific list of review tasks is being drafted by I&E and 00R.

In all cases involving backfitting, the procedures described in 10 CFR 50.109 will be followed.

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. JAN 1 s 197a E. G. Case Risk Assessment Recardino Overall Plant Heat Removal Caoability The importance to safety of having reliable systems to transfer fission product decay heat to the environment while at or near nomal operating temperatures is indicated by the results presented in WASH-1400 for tr:nsient events. The ability of a typical FWR plant and a typical BWR plant to remove decay heat following a transient that caused a reactor The evaluation trip was evaluated in WASH-1400 on a probabilistic basis.

included both those events in which the reactor protect. ion system failed

( ATWS) and events in which the reactor protection system functioned as

. In the latter case, system and ?quipment failures that result designed.

in the inability to remove fission product decay heat ' result in a higher probability of core melt than that predicted for a large LOCA for both For these types of events, it was considered acceptable PWRs and BWRs.

to remain at or near nomal operating temperature and pressure for an

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The relatively 'igh probability of core melt due to the inability to remove fission product decay heat following anticipated transients, even with scram, is indicative that a significant safety benefit may be gained by The particular equipment and systems

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upgrading some equipment and systems.

ar? those needed to remain at hot standby conditions for extended periods l

of time or those necessary to cool and depressurize the reactor. coolant system so that the residual heat removal system can be operated,

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' reactor in a hot standby condition for an indefinite period, there have been and will be events that require eventual cooldown to pemit either long-tenn cooling with the RHR system or going to cold shutdown for inspection and repairs. Examples of such events are steam generator tube M

rupture, extended loss of offsite power, loss of reactor coolant pump seals, failure of steam generator relief valves to reclose, and water hammer irt the feedwa.ter lines. The principal objective of RSB 5-1 is to ensure that if such a shutdown i's required, procedures for operating qualified systems are avail-able to do it safely and in an orderly manner. The branch position is also based on the consideration that closed-loop long-tem cooling of the reactor at the low pressure and temoerature conditions associated with operation of the RHR system is preferred and inherently provides additional systems which can be used in preference to long-tem cooling at high temperatures and pres-In a pWR, long-tem cooling at hot standby conditions is dependent sures.

In on the water inventory available to the auxiliary feedwater system.

most designs there is-a limited volume of high-quality water available.

JAN 1 3 578 E. G. Case.

Hence, there is a long-tenn possibility of using water of poorer quality since the high-quality feedwater will be soon spent by steam release to the atmosphere via the steam generator pressure relief or dump valves. This could lead to premature degradation of the steam generator. t.o ng-term cooling capability of a BWR at hot standby conditions is dependent on the inventory of the condensate storage tank availabre in the RCIC system and ultimately on the inventory of, and heat removal capability from, the suppression pool. Relief is through.the safety / relief valves, and operating experience shows that the probability of a valve failing to close is relatively high. This could lead to a more rapid than desirable cooldown rate.

No quantitative assessment has been made of the reduction in the risk that would result frcm the improvements in plant cooling capability associated with implementation of the revised SRP 5.4.7.

As noted above,

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in WA5H-1400 it was considered acceptable to remain at or near normal operating temperature for extended periods of time. Therefore, the effect of a failure resulting (a) in the i.nability to cool down to the cut-in point of the RHR system or (b) in loss of the RHR cooling function was considered small and not evaluated.

Since the changes included in the revised SRP 5.4.7' are directed primarily at providing additional protection in situations where shutdown to cold conditions is required, continued plant licensing and operation is i

considered ccceptable pending consideration.of the need for such changes.

y Roge a son, pirector Division o Systems Safety

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Victor Stel', Jr.4 Director Division of Operating Reactors

Enclosures:

As Stated cc:

R. Minogue H. Thornburg e

N.Moseley(45,.es)

R. Boyd S. Varga T. Murley H. Denton

0. Ross T. Novak S. Israel C. Graves

3. Hardin S. Weiss v

TABLE 1.

Differences:Setween Requirements of BTP RSB 5-1 and Existing Standard Review Plan 5.4.7 Ossign Requirements Change from Existing SRP 5.4.7 of BTP RSB 5-1 and Comment on Imoact I. Functional Requirements for Taking to Cold Shutdown New reouirements. Affects systems

a. Capability Using Only Safety involved in taking reactor down to cut-in V

Grade Systems point of RHRS. No change for RHRS I'9"

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'b. Capability with Either Onsite existing SRP to RHRS with respect to themselves, single failure involving the requirement and w in e Fa ur.

f r parallel lines and independent power L

(Limited Action Outside CR to supplies (or ecu1 valent) for RHR isola-MeetSF) tioa valves has only recently been I

c. Reasonable Time for Cooldown applied to CP reviews. Also possible Assuming Most Limiting SF large. impact for CVCS, s'.eam dump

!.h and Only Offsite or Only *

. valves, CST and primary system de-f

.Onsite Power pressurization to meet a and b. New SRP gives 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> as reasonable time II. RWR Isolation (Type.3) under c. No impact from c. expected..

.a. Suction Side--At least Two Power Operated Valves with Existing SRP does not classify RHRS h

Po'sition Indication in CR and according to types 1, 2, and 3 used i

Interlocks to Prevent RHRS in BTP RSB 5-1. Type 3 classifica-Overpressurization tion appl,ies to most reactors.' Changes-

b. Discharge Side--Same as Suction from existing SRP for Types 1, 2, and Side or Three Check Valves 3 are in direction of increased flex-or Two Check Valves with ibility. No change in impact with Periodic Leak Testing or at respect to existing SRP. However, Least One Check Valve with protection of high-low pressure iso-

• Nonnally Closed Power Operated lation function from effects of Valve having CR Position single failure may result in need Indication.

If ECCS Function, for extra independent power supplies Open Valve by SIS When RCS and/or valves if corrective manual 0

Pressure Below RHRS Design action inside containment is not J

Pressure allowed (see discussion under U

O requirement I in Table 1 & 2).

III. RHR Pressure Relief

a. Provide Relief Capacity to New words with respect to meetis ASME Meet ASME Code, Considering Code and considering most limiting Most Limiting Transient When transient. 1.ittle impact expected.

"y RHR Connected to RCS

b. Collect and Contain Relief New requirement.

Impact 5

Discharge considered minor.

IV. Pump Protection

a. Protect Against Overheating, Words new, but no impact. Standard v

fon, or Loss of Pump design practice.

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. Design Requirements Change from Existing SRP 5.4.7 of BTP RSB 5-1 and Comment en Imcact V. Test Requirement New words in Section 5.4.7, but is

a. Design to Permit Line Testinn already in existing SRP in Table 7-1.

When Operating in RHR Mode. '

Meet IEEE 338 and R.G. 1.22 No impact,

b. Meet R.G. 1.68.

For PWRs,

.New requirement for tests and analyses for cooldown under natural circulation.

Test Plus Analysis for Cooldown Under Natural Circulation to Should be minor impact.

Confirm Adequate Mixing and Cooldown Within Limits Specified in EGP.

VI. Operational Procedure

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a. Meet R.G. 1.33.

For PWRs, New requirement. Minor impact.

Include Specific Procedures and Information for Cooldown Under Natural Circulation

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V8I. Auxiliary Feedwater Supply

a. Seismic Category I Supply for New requirement. Impact may depend

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Auxiliary FW for at least Four primarily on time for cooldown to Hours at Hot Shutdown Plus cut-in point of RHRS under natural Cooldown to RHR Cut-in Based circulation RCS conditions. Sizing on Longest Time for Only Onsite Seismic Category I aux'liari or Only Offsite Power and Assumed feedwater supply tank may vary Single Failure depending on how uniformly the RCS cools down with natural circulation.

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Pasibla Solutions for full Ccaplitnte uith BIP R58 5-1.tnd kecompended Inviemistition for Cisss 2 Plants 1Act[ 2.

Design kcquirements Process and/5ystem Possible Solution"for Necomenended Implementation for.

of Bip 858 5-1 or Conquenent/

_ full Compliance Class ? Plants (see Note 1)

II. Rilk isolation 10684 System Comply with one of allowable Compliance Requised. (Plants arrangemesints glVen.

nunm@lly paret (, e requirerAnt n

unoer calsting SNP 5,4.7.)

lit, kitR Pressure Relief

b. Collect and contain relief 2Wt Systcr Determine piping, etc., needed to meet Compliance will not be required.

dist.har9e requirensent and provide in design.

If it is shown 1141 adequate l

alternate methods of disposing of discharge are avellable.

V. Test Requirement le. Hect it.G. l.68, for PWNs test Run tests and confirming analysis to Compliance Required.

plus an.nlysis for cooldown under meet requirement, natural circulation to confirma adequate mixing and couldown within limits specified in E0P.

VI. Operational Procedure 4 Hect it.G.1.33. For INRs.

Develop procedures and infonmation f rom Compliance Required.

luclude specific procedures tests and analysis.

and infoDuatlofi for CooIdowls under natural circulation.-

Vll. Auxiliary leedwater Suinply Emerijency feedwater Supply 4 Seismic rategory i supply for Isom tests and asialysis obtain conser-Complianice will e.ot be required.

annillary fW for at least four vative estimate of auxiliary IW supply if it is shown that an adequale hours at hnt shutdown plus cool-to meet requirement and provide alternate seismic Category I duwn to Ri!R tut-in l>asid on Seismic Category I supply.

source is available, longest tis.e for only onsite or only of f site luneer and assumed single failure.

' Note 1: 1he implementation for Class ?

plants does not result in a major impact while providing additional capability to 90 to cold shulduwn.

The major inquct results froen the requiremient fur safety-grade steam i

dimqs valves.

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ENCLOSURE 2 PLANTS FOR hHICH ISSUAt!CE OF OL EXPECTED BEFORE 1/1/79*--CLASS 3 PLANT EXPECTED OL ISSUANCE DATE North Anna 1 01/78 Three Mile Island 2 01/78 Cook 2 01/78 Hatch 2

/78 Arkansas 2 06/78 North Anna 2 07/78 is, iy.

Diablo Canyon 1 09/78 Diablo Canyon 2 09/78 Zimmer 1 10/78 Sequoyah 2 11/78 i

Watts Bar 1 12/78 g.

• Based on December 1977 Licensing Projections s

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